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Photoneutron reaction kinematics and error of commonly used approximations:

An exact closed-form classical equation and a transcendental relativistic equation for the energy-angle relationship of the neutrons produced from the photoneutron reaction are derived in the laboratory coordinate system. Additional formulas are derived to describe the restriction on scattering angle in the double-valued photoneutron energy regime for incident photon energies near the kinematic threshold energy of the target nucleus. The full range of photoneutron reaction independent variables (target nucleus mass number, neutron separation energy, and photon energy) encountered in practical photoneutron production applications are explored to give the applicable range of classical kinematics and other approximations in the calculation of the photoneutron energy-angle relationship. A six-way comparison study of classical, relativistic, and approximation equations presented in this work and those implemented in radiation transport codes is performed. The well-known and widely-used approximations presented without derivation by Wattenberg and Hanson in the late 1940s are only applicable to energy spread of neutrons produced from gamma-ray photoneutron sources, the intended application of the approximations. The Wattenberg and Hanson approximations are inappropriate for general purpose radiation transport simulation and give unphysical negative photoneutron energies for incident photon energies near the kinematic threshold. Recent versions of the MCNP code appear to use erroneous equations related to inelastic neutron scattering to describe photoneutron kinematics resulting in the overestimation of photoneutron energy. Simulated photoneutron spectra are significantly hardened and cannot be used in contemporary nuclear science applications such as characterization of accelerator-driven photoneutron sources. The Monte-Carlo code developer and user communities need to exhaustively review, correct, document, verify, and validate photoneutron physics implementations in the codes that are in common use.


Operator Action–Induced Two-Phase Flow Condition Resulting in Performance Degradation of Interfacing Passive System

This study investigates the degradation of the heat transfer performance of a closed-circuit intermediate natural circulation heat transport loop used as a passive safety system in a nuclear power plant (NPP). The degradation arises from the strong thermal-hydraulic (TH) coupling of the loop operating characteristics, saturation temperature and pressure, and natural circulation flow rate, which determine the heat rejection rate to the TH boundary conditions imposed on the hot side of the loop by the transitory state of the primary reactor coolant system (RCS) of the NPP. Several operator actions related to a feed-and-bleed emergency operating procedure (F&B) are postulated, and system TH code simulations are performed to demonstrate how the F&B can induce two-phase flow conditions in the RCS. Natural circulation two-phase flow regimes in the RCS hot leg can significantly reduce the heat transfer to the circulating working fluid of the interfacing heat transport loop over long periods, sometimes lasting over 24 h, of passive system mission time. A transient performance indicator for the passive system mission is introduced for use in the passive reliability assessment and quantitative comparison of transient simulations. The need to consider human factors in the design and operation of NPP passive safety systems is stressed.


Optimal working fluid charge and degradation thresholds for a closed-circuit intermediate natural circulation heat transport loop

The study derives the optimal working fluid charge of a closed-circuit intermediate natural circulation heat transport loop based on the passive residual heat removal system (PRHRS) of SMART (System integrated Modular Advanced ReacTor). The study finds that efficient natural circulation flow and heat transfer regimes can be maintained in a narrow band of working fluid charge mass owing to the small secondary side volume of the once-through helical-coil steam generator design of SMART. Overcharging condition is a severe passive system degradation mode that quickly leads to system failure. Overcharging is characterized by sustained two-phase flow instabilities, mainly liquid slugging in the hot leg of the PRHRS, and an increased operating pressure of the PRHRS coupled with a compensating increase in the operating pressure of the reactor coolant system (RCS). Increased pressures maintain the required temperature differential across the steam generator tubes and heat sink heat exchanger to achieve a prescribed quasi-steady state heat removal rate capacity. Undercharging condition is a benign degradation mode relative to overcharging and is characterized by a slow upward drift of the RCS pressure and temperature of the liquid water entering the primary side of the steam generator. Only the undercharged case of a completely dry steam generator secondary side leads to PRHRS failure-to-start via vapor lock.


Cooldown procedure success criteria map for the full break size spectrum of SBLOCA

A success criteria map defining successful implementation of a cooldown procedure to mitigate small-break loss of coolant accident (SBLOCA) with failed high pressure safety injection system for the OPR1000 is generated for the full break size spectrum from 0.5 in to 2.5 in diameter breaks. The success criteria map is a multi-dimensional response surface that gives the expected peak clad temperature (PCT) as a function of break size and operator action time to start the 55.6 degrees C/hr reactor coolant system cooldown and depressurization involving secondary side steam dump through manual manipulation of the atmospheric dump valves (ADV). The PCT response surface is constructed through a regression analysis using a Gaussian process model (GPM) coupled to an adaptive sampling procedure that greatly reduces the computational cost and analysis time by limiting the number of required best estimate simulations of the SBLOCA with the MARS code to approximately 100. Break flow rates and times to core heat up each span several orders of magnitude when considering the full break size spectrum so an innovative solution of transforming the break size and ADV actuation time input variables using exponential and logarithm functions was implemented such that the regression could be performed on the scaled and transformed input space.


Pressure drop-flow rate curves for single-phase steam in Combustion Engineering type steam generator U-tubes during severe accidents

Characteristic pressure drop-flow rate curves are generated for all row numbers of the OPR1000 steam generators (SGs), representative of Combustion Engineering (CE) type SGs featuring square bend U-tubes. The pressure drop-flow rate curves are applicable to severe accident natural circulations of single-phase superheated steam during high pressure station blackout sequences with failed auxiliary feedwater and dry secondary side which are closely related to the thermally induced steam generator tube rupture event. The pressure drop-flow rate curves which determine the recirculation rate through the SG tubes are dependent on the tube bundle geometry and hydraulic diameter of the tubes. The larger CE type SGs have greater variation of tube length and height as a function of row number with forward flow of steam favored in the longer and taller high row number tubes and reverse flow favored in the short low row number tubes. Friction loss, natural convection heat transfer coefficients, and temperature differentials from the primary to secondary side are dominant parameters affecting the recirculation rate. The need for correlation development for natural convection heat transfer coefficients for external flow over tube bundles currently not modeled in system codes is discussed.


Implicit Treatment of Technical Specification and Thermal Hydraulic Parameter Uncertainties in Gaussian Process Model to Estimate Safety Margin

The Gaussian process model (GPM) is a flexible surrogate model that can be used for nonparametric regression for multivariate problems. A unique feature of the GPM is that a prediction variance is automatically provided with the regression function. In this paper, we estimate the safety margin of a nuclear power plant by performing regression on the output of best-estimate simulations of a large-break loss-of-coolant accident with sampling of safety system configuration, sequence timing, technical specifications, and thermal hydraulic parameter uncertainties. The key aspect of our approach is that the GPM regression is only performed on the dominant input variables, the safety injection flow rate and the delay time for AC powered pumps to start representing sequence timing uncertainty, providing a predictive model for the peak clad temperature during a reflood phase. Other uncertainties are interpreted as contributors to the measurement noise of the code output and are implicitly treated in the GPM in the noise variance term, providing local uncertainty bounds for the peak clad temperature. We discuss the applicability of the foregoing method to reduce the use of conservative assumptions in best estimate plus uncertainty (BEPU) and Level 1 probabilistic safety assessment (PSA) success criteria definitions while dealing with a large number of uncertainties.


Photoneutron production in heavy water reactor fuel lattice from accelerator-driven bremsstrahlung

The coupling of low-energy electron linear accelerators (eLINACs) to a large heavy water reactor is proposed to create an accelerator-driven photoneutron source (ADS). Photoneutron yields of 1012 pn/s per kW of beam power can be achieved in the ADS-CANDU concept where the wide fuel channel spacing of heavy water reactors represents a near-optimal geometry for conversion of accelerator-driven bremsstrahlung into photoneutrons in the heavy water moderator with minimal parasitic attenuation of photons in the fuel channels. Twenty MeV electron beam energy is most efficient at producing photoneutrons despite having the largest fuel channel shielding effect (28% attenuation compared to an infinite heavy water medium). The majority of photoneutrons are produced during source-photon first flights, so the spatial distribution and emission spectra of the ADS in the secondary (γ,n) converter are correlated to the doubly differential angle and energy distribution of the bremsstrahlung emitted from the (e-,γ) converter. Compton-scattered photons and tertiary bremsstrahlung originating from the electron–positron pair and recoil electron secondary particles are important contributors to the photoneutron yield of higher energy eLINACs systems. Photonuclear data for the deuterium photoneutron reaction cross section and secondary electron transport and tertiary bremsstrahlung production physics implemented in Monte-Carlo radiation transport codes can dominate ADS simulation results and are the same magnitude as the physical phenomena such as the fuel channel shielding effect.