Bojan Petrovic

Pioneering role of IRIS in the resurgence of Small Modular Reactors

Abstract This paper presents an overview of the first 10 years of the IRIS project, summarizing its main technical achievements and evaluating its impact on the resurgence of small modular reactors (SMRs). SMRs have been recurrently studied in the past, from early days of nuclear power, but have never gained sufficient traction to reach commercialization. This situation persisted also in the 1990s; the focus was on large reactors based on the presumed common wisdom of this being the only way to make the nuclear power plants competitive. IRIS is one of several small reactor concepts that originated in the late 1990s. However, the specific role and significance of IRIS is that it systematically pursued resolving technology gaps, addressing safety, licensing, and deployment issues and performing credible economics analyses, which ultimately made it possible—together with other SMR projects—to cross the “skepticism threshold” and led the making of a convincing case—domestically and internationally—for the role and viability of smaller reactors. Technologically, IRIS is associated with a number of novel design features that it either introduced or pursued more systematically than its predecessors and ultimately brought them to a new technical level. Some of these are discussed in this paper, such as the IRIS Safety-by-Design, security by design, the innovative thermodynamic coupling of its vessel and containment, systematic probabilistic risk assessment-guided design, approach to seismic design, approach to reduce the emergency planning zone to the site boundary, active involvement of academia, and so on. Many individuals and organizations contributed to that work, too many to list individually, and this paper attempts to pay tribute at least to their collective work.

The SPES3 Experimental Facility Design for the IRIS Reactor Simulation

IRIS is an advanced integral pressurized water reactor, developed by an international consortium led by Westinghouse. The licensing process requires the execution of integral and separate effect tests on a properly scaled reactor simulator for reactor concept, safety system verification, and code assessment. Within the framework of an Italian R&D program on Nuclear Fission, managed by ENEA and supported by the Ministry of Economic Development, the SPES3 facility is under design and will be built and operated at SIET laboratories. SPES3 simulates the primary, secondary, and containment systems of IRIS with 1 : 100 volume scale, full elevation, and prototypical thermal-hydraulic conditions. The simulation of the facility with the RELAP5 code and the execution of the tests will provide a reliable tool for data extrapolation and safety analyses of the final IRIS design. This paper summarises the main design steps of the SPES3 integral test facility, underlying choices and phases that lead to the final design.

IRIS safety-by-design? and its implication to lessen emergency planning requirements

International Reactor Innovative and Secure (IRIS) is an integral configuration pressurised light water reactor that has been in development since late 1999 by an international consortium. Its design and safety characteristics have been amply reported. In this paper the safety-by-design? IRIS philosophy is reviewed to show how the projected safety performance (most accidents either eliminated or inherently mitigated, Core Damage Frequency (CDF) due to internal events of the order of 10−8 events/year) exceeds the current norm of nuclear reactors. The IRIS project plans to use this enhanced safety response to explore the possibility of lessening, or even eliminating, the off-site emergency planning requirement. A review is given of previous attempts to attain this relaxation of licensing regulations and of current goals for advanced reactors. Finally, the proposed methodology is outlined. It consists of a combined deterministic and probabilistic approach, including a review of the defence in-depth, and a risk informed analysis of a wide spectrum of accidents, rather than an evaluation limited to a few design-based accidents.

Effects of SN method numerics on pressure vessel neutron fluence calculations

An accurate prediction of the reactor pressure vessel (PV) fast neutron fluence (E > 1.0 MeV or E > 0.1 MeV) is necessary to ensure PV integrity over the design lifetime. The discrete ordinates method (S N method) is the method of choice to treat such problems, and the DORT S N code is widely used as a standard tool for PV fluence calculations. The S N numerics and the corresponding DORT numerical options and features offer alternative choices that increase flexibility but also impact results. The effects of S N numerics based on PVfluence calculations for two pressurized water reactors are examined. The differencing schemes [linear, zero-weighted (ZW), and θ-weighted (TW)] and their interactions with spatial and angular discretization are also examined. The linear and TW (θ = 0.9) schemes introduce unphysical flux oscillations that for certain groups and positions may exceed 10%. The ZW scheme produces smooth results ; however, its results differ from the other two schemes. A good compromise for PV fluence calculations is a TW scheme with a small θ value (i.e., θ = 0.3), which reduces the uncertainty to ∼3%. Angular discretization and spatial mesh size employed in typical calculations introduce another ∼3 and ∼2% uncertainty, respectively. The analysis further shows that the fixup is not necessary for the negative scattering source. The pointwise convergence criterion is also not a critical issue in the fast energy range because of a relatively fast convergence rate. Similarly, acceleration parameters impact mainly the execution time and only marginally the results. The root-mean-square combined uncertainty for standard PV fluence calculations due to the options analyzed is ∼5%.

Fuel Cycle Analysis of the SABR Subcritical Transmutation Reactor Concept

Abstract A fuel cycle analysis was performed for the SABR transmutation reactor concept, using the ERANOS fast reactor physics code. SABR is a sodium-cooled, transuranic (TRU)-Zr-fueled, subcritical fast reactor driven by a tokamak fusion neutron source. Three different four-batch reprocessing fuel cycles, in which all the TRUs from spent nuclear fuel discharged from light water reactors are fissioned to >90% (by recycling four times), was examined. The total fuel residence time in the reactor was limited in these three cycles by a radiation damage limit (100, 200, or 300 displacements per atom) to the cladding material. In the fourth cycle the fuel residence time was determined by trying to achieve 90% burnup in a once-through cycle without reprocessing.

Pulsed neutron interrogation for detection of concealed special nuclear materials

A new neutron interrogation technique for detection of concealed Special Nuclear Material (SNM) is described. This technique is a combination of timing techniques from pulsed prompt gamma neutron activation analysis with silicon carbide (SiC) semiconductor fast neutron detector technology. SiC detectors are a new class of radiation detectors that are ultra-fast and capable of processing high count rates. SiC detectors can operate during and within nanoseconds of the end of an intense neutron pulse, providing the ability to detect the prompt neutron emissions from fission events produced by the neutrons in concealed SNM on a much faster pulsing time scale than has been achieved by other techniques. Neutron-induced fission neutrons in 235U have been observed in the time intervals between pulses of 14-MeV neutrons from a deuterium-tritium electronic neutron generator. Initial measurements have emphasized the detection of SNM using thermal-neutron induced fission. Neutron pulsing and time-sequenced neutron counts were carried out on a hundreds of microseconds time scale, enabling the observation of prompt fission neutrons induced by the die-away of thermal neutrons following the 14-MeV pulse. A discussion of pulsed prompt-neutron measurements and of SiC detectors as well as initial measurement results will be presented.

Neutron fluence rate measurements in a PGNAA 208-liter drum assay system using silicon carbide detectors

Abstract Pulsed prompt gamma neutron activation analysis (PGNAA) is being implemented for the nondestructive assay (NDA) of mercury, cadmium and lead in containers of radioactive waste. A PGNAA prototype system capable of assaying 208-liter (55-gallon) drums has already been built and demonstrated. As part of the evaluation of this system, the thermal neutron fluence rate distribution in a drum containing a combustible waste surrogate was measured during PGNAA runs using a silicon carbide neutron detector. The fast charge-collection time of this detector type enabled the investigation of the neutron kinetics at various locations within the matrix during and between pulses of the system’s 14-MeV neutron source. As expected, the response of a SiC detector equipped with a lithium-6 fluoride layer is dominated by thermal neutron-induced events between pulses. The measurement results showed that the thermal neutron fluence rate is relatively uniform over a radial depth of several centimeters in the matrix region that contributes a significant fraction of the prompt gamma radiation incident on the system’s photon detector.

Calculating the Second Eigenpair in Criticality Calculations Using the Monte Carlo Method with Source Points Pairing as an Efficient Net-Weight (Cancellation) Algorithm

Abstract The Monte Carlo method is widely used to compute the fundamental eigenfunction and eigenvalue for nuclear systems. However, the standard power iteration method computes only the fundamental eigenmode, while it would be beneficial to also compute the higher eigenfunctions and eigenvalues to support the reactor transient analysis, stability analysis, and assessments of nuclear safety, as well as to enable certain source convergence acceleration techniques. Modifications to the power method have been developed that in principle can accomplish this goal, but they typically lead to unphysical positive and negative particles requiring a procedure to compute the net-weight deposition. In this paper, we present a new mechanism that enables the Monte Carlo procedure, with certain modifications, to compute the second eigenfunction and eigenvalue for one-dimensional (1-D) problems. The method could in principle be extended to higher harmonics and general geometries. The results from numerical examples, including a 1-D, two-group, multiregion example, are consistent with reference results. Moreover, the extra computational cost of this method is of the same order of magnitude as the conventional Monte Carlo simulations. This method can be applied solely to solve for the high eigenmodes, or implemented as a part of a net-weight computation mechanism when negative particles are present in the modified power iteration method.

Neutronic Evaluation of a Liquid Salt–Cooled Reactor Assembly

Abstract This study presents a thorough parametric neutronic analysis of a plate-based tristructual isotropic (TRISO) fuel particle bearing liquid salt–cooled reactor assembly. The analyses presented investigated the effects of altering fuel enrichment, packing fraction, plate region thicknesses, assembly structure thicknesses, assembly size, numbers of plates per assembly, use of burnable poison materials, replacement of assembly and plate carbon material with silicon carbide, and use of uranium nitride fuel kernels. The effects or trends observed included reactivity behavior, discharge burnup, cycle length, and other key design parameters such as moderator temperature coefficients, coolant density coefficients, control blade worth, and impacts upon power peaking (i.e., power and flux distributions). This study is based upon two-dimensional lattice physics calculations involving the SERPENT 2 code and by using the nonlinear reactivity model as a reasonable tool for predicting discharge burnup. The reported results show that the system’s reactivity can be significantly altered by varying these design parameters, thus providing a starting point for future design optimization studies, and it is understood that future studies will need to be expanded to equilibrium full core analysis for more complete and accurate design and safety assessments, which is also a work in progress.

PRELIMINARY EVALUATION OF THE SCALE MAVRIC CODE AND FW-CADIS METHOD FOR EFFICIENT SCOPING OF THE RADIATION FIELD THROUGHOUT THE IRIS CONTAINMENT

Abstract The integral configuration of the International Reactor Innovative and Secure (IRIS) with its relatively thick downcomer region within the reactor vessel and compact spherical steel containment offers potential for a significant dose reduction but also presents challenges for the related deep-penetration shielding analyses due to the large spatial domain. It is necessary to determine the radiation field throughout the 25-m-diam spherical containment and into the adjoining auxiliary building. The shielding analysis is being performed using the “traditional” deterministic SN and Monte Carlo approaches and codes (TORT and MCNP, respectively). In the preliminary, scoping phase, the radiation field is sought “everywhere” throughout the power plant to identify any possible shielding issues. This is very challenging for typical Monte Carlo variance-reduction methods, which are devised and may work very well to provide results in a limited region or for individual “detectors” rather than everywhere. However, the recently developed FW-CADIS method, implemented within the MAVRIC sequence of the SCALE code system, aims to address this problem. It uses forward and adjoint deterministic SN calculations to generate effective biasing parameters for Monte Carlo simulations throughout the problem. Previous studies have confirmed its potential for obtaining Monte Carlo solutions with acceptable statistics over large spatial domains. The objective of this work was to evaluate the applicability of FW-CADIS/MAVRIC to efficiently perform the required shielding analysis of IRIS. For that purpose, a representative model was prepared, retaining the main problem characteristics, i.e., a large spatial domain (>10 m in each dimension) and significant attenuation (more than 12 orders of magnitude), but geometrically rather simplified at this stage of evaluation. The obtained preliminary results indicate that the FW-CADIS method implemented through the MAVRIC sequence in SCALE will enable determination of the radiation field throughout the large spatial domain of the IRIS nuclear power plant.