#include "openmc/physics.h"

#include "openmc/bank.h"

#include "openmc/bremsstrahlung.h"

#include "openmc/chain.h"

#include "openmc/constants.h"

#include "openmc/distribution_multi.h"

#include "openmc/eigenvalue.h"

#include "openmc/endf.h"

#include "openmc/error.h"

#include "openmc/ifp.h"

#include "openmc/material.h"

#include "openmc/math_functions.h"

#include "openmc/message_passing.h"

#include "openmc/ncrystal_interface.h"

#include "openmc/nuclide.h"

#include "openmc/photon.h"

#include "openmc/physics_common.h"

#include "openmc/random_dist.h"

#include "openmc/random_lcg.h"

#include "openmc/reaction.h"

#include "openmc/search.h"

#include "openmc/secondary_uncorrelated.h"

#include "openmc/settings.h"

#include "openmc/simulation.h"

#include "openmc/string_utils.h"

#include "openmc/tallies/tally.h"

#include "openmc/thermal.h"

#include "openmc/weight_windows.h"

#include <fmt/core.h>

#include "openmc/tensor.h"

#include <algorithm> // for max, min, max_element

#include <cmath> // for sqrt, exp, log, abs, copysign

namespace openmc {

//==============================================================================

// Non-member functions

//==============================================================================

void collision(Particle& p)

{

// Add to collision counter for particle

++(p.n_collision());

p.secondary_bank_index() = p.local_secondary_bank().size();

// Sample reaction for the material the particle is in

switch (p.type().pdg_number()) {

case PDG_NEUTRON:

sample_neutron_reaction(p);

break;

case PDG_PHOTON:

sample_photon_reaction(p);

break;

case PDG_ELECTRON:

sample_electron_reaction(p);

break;

case PDG_POSITRON:

sample_positron_reaction(p);

break;

default:

fatal_error("Unsupported particle PDG for collision sampling.");

}

if (settings::weight_windows_on) {

auto [ww_found, ww] = search_weight_window(p);

if (!ww_found && p.type() == ParticleType::neutron()) {

// if the weight window is not valid, apply russian roulette for neutrons

// (regardless of weight window collision checkpoint setting)

apply_russian_roulette(p);

} else if (settings::weight_window_checkpoint_collision) {

// if collision checkpointing is on, apply weight window

apply_weight_window(p, ww);

}

}

// Kill particle if energy falls below cutoff

int type = p.type().transport_index();

if (type != C_NONE && p.E() < settings::energy_cutoff[type]) {

p.wgt() = 0.0;

}

// Display information about collision

if (settings::verbosity >= 10 || p.trace()) {

std::string msg;

if (p.event() == TallyEvent::KILL) {

msg = fmt::format(" Killed. Energy = {} eV.", p.E());

} else if (p.type().is_neutron()) {

msg = fmt::format(" {} with {}. Energy = {} eV.",

reaction_name(p.event_mt()), data::nuclides[p.event_nuclide()]->name_,

p.E());

} else if (p.type().is_photon()) {

msg = fmt::format(" {} with {}. Energy = {} eV.",

reaction_name(p.event_mt()),

to_element(data::nuclides[p.event_nuclide()]->name_), p.E());

} else {

msg = fmt::format(" Disappeared. Energy = {} eV.", p.E());

}

write_message(msg, 1);

}

}

void sample_neutron_reaction(Particle& p)

{

// Sample a nuclide within the material

int i_nuclide = sample_nuclide(p);

// Save which nuclide particle had collision with

p.event_nuclide() = i_nuclide;

// Create fission bank sites. Note that while a fission reaction is sampled,

// it never actually "happens", i.e. the weight of the particle does not

// change when sampling fission sites. The following block handles all

// absorption (including fission)

const auto& nuc {data::nuclides[i_nuclide]};

if (nuc->fissionable_ && p.neutron_xs(i_nuclide).fission > 0.0) {

auto& rx = sample_fission(i_nuclide, p);

if (settings::run_mode == RunMode::EIGENVALUE) {

create_fission_sites(p, i_nuclide, rx);

} else if (settings::run_mode == RunMode::FIXED_SOURCE &&

settings::create_fission_neutrons) {

create_fission_sites(p, i_nuclide, rx);

// Make sure particle population doesn't grow out of control for

// subcritical multiplication problems.

if (p.local_secondary_bank().size() >= settings::max_secondaries &&

!settings::use_shared_secondary_bank) {

fatal_error(

"The secondary particle bank appears to be growing without "

"bound. You are likely running a subcritical multiplication problem "

"with k-effective close to or greater than one.");

}

}

p.event_mt() = rx.mt_;

}

// Create secondary photons

if (settings::photon_transport) {

sample_secondary_photons(p, i_nuclide);

}

// If survival biasing is being used, the following subroutine adjusts the

// weight of the particle. Otherwise, it checks to see if absorption occurs

if (p.neutron_xs(i_nuclide).absorption > 0.0) {

absorption(p, i_nuclide);

}

if (!p.alive())

return;

// Sample a scattering reaction and determine the secondary energy of the

// exiting neutron

const auto& ncrystal_mat = model::materials[p.material()]->ncrystal_mat();

if (ncrystal_mat && p.E() < NCRYSTAL_MAX_ENERGY) {

ncrystal_mat.scatter(p);

} else {

scatter(p, i_nuclide);

}

// Advance URR seed stream 'N' times after energy changes

if (p.E() != p.E_last()) {

advance_prn_seed(data::nuclides.size(), &p.seeds(STREAM_URR_PTABLE));

}

// Play russian roulette if there are no weight windows

if (!settings::weight_windows_on)

apply_russian_roulette(p);

}

void create_fission_sites(Particle& p, int i_nuclide, const Reaction& rx)

{

// If uniform fission source weighting is turned on, we increase or decrease

// the expected number of fission sites produced

double weight = settings::ufs_on ? ufs_get_weight(p) : 1.0;

// Determine the expected number of neutrons produced

double nu_t = p.wgt() / simulation::keff * weight *

p.neutron_xs(i_nuclide).nu_fission /

p.neutron_xs(i_nuclide).total;

// Sample the number of neutrons produced

int nu = static_cast<int>(nu_t);

if (prn(p.current_seed()) <= (nu_t - nu))

++nu;

// If no neutrons were produced then don't continue

if (nu == 0)

return;

// Initialize the counter of delayed neutrons encountered for each delayed

// group.

double nu_d[MAX_DELAYED_GROUPS] = {0.};

// Clear out particle's nu fission bank

p.nu_bank().clear();

p.fission() = true;

// Determine whether to place fission sites into the shared fission bank

// or the secondary particle bank.

bool use_fission_bank = (settings::run_mode == RunMode::EIGENVALUE);

// Counter for the number of fission sites successfully stored to the shared

// fission bank or the secondary particle bank

int n_sites_stored;

for (n_sites_stored = 0; n_sites_stored < nu; n_sites_stored++) {

// Initialize fission site object with particle data

SourceSite site;

site.r = p.r();

site.particle = ParticleType::neutron();

site.time = p.time();

site.wgt = 1. / weight;

site.surf_id = 0;

// Sample delayed group and angle/energy for fission reaction

sample_fission_neutron(i_nuclide, rx, &site, p);

// Reject site if it exceeds time cutoff

if (site.delayed_group > 0) {

double t_cutoff = settings::time_cutoff[site.particle.transport_index()];

if (site.time > t_cutoff) {

continue;

}

}

// Set parent and progeny IDs

site.parent_id = p.current_work();

site.progeny_id = p.n_progeny()++;

// Store fission site in bank

if (use_fission_bank) {

int64_t idx = simulation::fission_bank.thread_safe_append(site);

if (idx == -1) {

warning(

"The shared fission bank is full. Additional fission sites created "

"in this generation will not be banked. Results may be "

"non-deterministic.");

// Decrement number of particle progeny as storage was unsuccessful.

// This step is needed so that the sum of all progeny is equal to the

// size of the shared fission bank.

p.n_progeny()--;

// Break out of loop as no more sites can be added to fission bank

break;

}

// Iterated Fission Probability (IFP) method

if (settings::ifp_on()) {

ifp(p, idx);

}

} else {

site.wgt_born = p.wgt_born();

site.wgt_ww_born = p.wgt_ww_born();

site.n_split = p.n_split();

p.local_secondary_bank().push_back(site);

p.n_secondaries()++;

}

// Increment the number of neutrons born delayed

if (site.delayed_group > 0) {

nu_d[site.delayed_group - 1]++;

}

// Write fission particles to nuBank

NuBank& nu_bank_entry = p.nu_bank().emplace_back();

nu_bank_entry.wgt = site.wgt;

nu_bank_entry.E = site.E;

nu_bank_entry.delayed_group = site.delayed_group;

}

// If shared fission bank was full, and no fissions could be added,

// set the particle fission flag to false.

if (n_sites_stored == 0) {

p.fission() = false;

return;

}

// Set nu to the number of fission sites successfully stored. If the fission

// bank was not found to be full then these values are already equivalent.

nu = n_sites_stored;

// Store the total weight banked for analog fission tallies

p.n_bank() = nu;

p.wgt_bank() = nu / weight;

for (size_t d = 0; d < MAX_DELAYED_GROUPS; d++) {

p.n_delayed_bank(d) = nu_d[d];

}

}

void sample_photon_reaction(Particle& p)

{

// Kill photon if below energy cutoff -- an extra check is made here because

// photons with energy below the cutoff may have been produced by neutrons

// reactions or atomic relaxation

int photon = ParticleType::photon().transport_index();

if (p.E() < settings::energy_cutoff[photon]) {

p.E() = 0.0;

p.wgt() = 0.0;

return;

}

// Sample element within material

int i_element = sample_element(p);

const auto& micro {p.photon_xs(i_element)};

const auto& element {*data::elements[i_element]};

// Calculate photon energy over electron rest mass equivalent

double alpha = p.E() / MASS_ELECTRON_EV;

// For tallying purposes, this routine might be called directly. In that

// case, we need to sample a reaction via the cutoff variable

double prob = 0.0;

double cutoff = prn(p.current_seed()) * micro.total;

// Coherent (Rayleigh) scattering

prob += micro.coherent;

if (prob > cutoff) {

p.mu() = element.rayleigh_scatter(alpha, p.current_seed());

p.u() = rotate_angle(p.u(), p.mu(), nullptr, p.current_seed());

p.event() = TallyEvent::SCATTER;

p.event_mt() = COHERENT;

return;

}

// Incoherent (Compton) scattering

prob += micro.incoherent;

if (prob > cutoff) {

double alpha_out;

int i_shell;

element.compton_scatter(

alpha, true, &alpha_out, &p.mu(), &i_shell, p.current_seed());

// Determine binding energy of shell. The binding energy is 0.0 if

// doppler broadening is not used.

double e_b;

if (i_shell == -1) {

e_b = 0.0;

} else {

e_b = element.binding_energy_[i_shell];

}

// Create Compton electron

double phi = uniform_distribution(0., 2.0 * PI, p.current_seed());

double E_electron = (alpha - alpha_out) * MASS_ELECTRON_EV - e_b;

int electron = ParticleType::electron().transport_index();

if (E_electron >= settings::energy_cutoff[electron]) {

double mu_electron = (alpha - alpha_out * p.mu()) /

std::sqrt(alpha * alpha + alpha_out * alpha_out -

2.0 * alpha * alpha_out * p.mu());

Direction u = rotate_angle(p.u(), mu_electron, &phi, p.current_seed());

process_charged_secondary(p, u, E_electron, ParticleType::electron());

}

// Allow electrons to fill orbital and produce Auger electrons and

// fluorescent photons. Since Compton subshell data does not match atomic

// relaxation data, use the mapping between the data to find the subshell

if (settings::atomic_relaxation && element.has_atomic_relaxation_ &&

i_shell >= 0 && element.subshell_map_[i_shell] >= 0) {

element.atomic_relaxation(element.subshell_map_[i_shell], p);

}

phi += PI;

p.E() = alpha_out * MASS_ELECTRON_EV;

p.u() = rotate_angle(p.u(), p.mu(), &phi, p.current_seed());

p.event() = TallyEvent::SCATTER;

p.event_mt() = INCOHERENT;

return;

}

// Photoelectric effect

double prob_after = prob + micro.photoelectric;

if (prob_after > cutoff) {

// Get grid index, interpolation factor, and bounding subshell

// cross sections

int i_grid = micro.index_grid;

double f = micro.interp_factor;

tensor::View<const double> xs_lower = element.cross_sections_.slice(i_grid);

tensor::View<const double> xs_upper =

element.cross_sections_.slice(i_grid + 1);

for (int i_shell = 0; i_shell < element.shells_.size(); ++i_shell) {

const auto& shell {element.shells_[i_shell]};

// Check threshold of reaction

if (xs_lower(i_shell) == 0)

continue;

// Evaluation subshell photoionization cross section

prob += std::exp(

xs_lower(i_shell) + f * (xs_upper(i_shell) - xs_lower(i_shell)));

if (prob > cutoff) {

// Determine binding energy based on whether atomic relaxation data is

// present (if not, use value from Compton profile data)

double binding_energy = element.has_atomic_relaxation_

? shell.binding_energy

: element.binding_energy_[i_shell];

// Determine energy of secondary electron

double E_electron = p.E() - binding_energy;

// Sample mu using non-relativistic Sauter distribution.

// See Eqns 3.19 and 3.20 in "Implementing a photon physics

// model in Serpent 2" by Toni Kaltiaisenaho

double mu;

while (true) {

double r = prn(p.current_seed());

if (4.0 * (1.0 - r) * r >= prn(p.current_seed())) {

double rel_vel =

std::sqrt(E_electron * (E_electron + 2.0 * MASS_ELECTRON_EV)) /

(E_electron + MASS_ELECTRON_EV);

mu =

(2.0 * r + rel_vel - 1.0) / (2.0 * rel_vel * r - rel_vel + 1.0);

break;

}

}

double phi = uniform_distribution(0., 2.0 * PI, p.current_seed());

Direction u;

u.x = mu;

u.y = std::sqrt(1.0 - mu * mu) * std::cos(phi);

u.z = std::sqrt(1.0 - mu * mu) * std::sin(phi);

// Process secondary electron at the photon collision site.

process_charged_secondary(p, u, E_electron, ParticleType::electron());

// Allow electrons to fill orbital and produce auger electrons

// and fluorescent photons

if (settings::atomic_relaxation) {

element.atomic_relaxation(i_shell, p);

}

p.event() = TallyEvent::ABSORB;

p.event_mt() = 533 + shell.index_subshell;

p.wgt() = 0.0;

p.E() = 0.0;

return;

}

}

}

prob = prob_after;

// Pair production

prob += micro.pair_production;

if (prob > cutoff) {

double E_electron, E_positron;

double mu_electron, mu_positron;

element.pair_production(alpha, &E_electron, &E_positron, &mu_electron,

&mu_positron, p.current_seed());

// Process secondary electron at the photon collision site.

Direction u = rotate_angle(p.u(), mu_electron, nullptr, p.current_seed());

process_charged_secondary(p, u, E_electron, ParticleType::electron());

// Process secondary positron at the photon collision site.

u = rotate_angle(p.u(), mu_positron, nullptr, p.current_seed());

process_charged_secondary(p, u, E_positron, ParticleType::positron());

p.event() = TallyEvent::ABSORB;

p.event_mt() = PAIR_PROD;

p.wgt() = 0.0;

p.E() = 0.0;

}

}

void process_charged_secondary(

Particle& p, Direction u, double E, ParticleType type)

{

int idx = type.transport_index();

if (idx == C_NONE || E < settings::energy_cutoff[idx])

return;

if (settings::electron_treatment == ElectronTreatment::TTB) {

thick_target_bremsstrahlung(p, type, u, E);

}

if (type == ParticleType::positron()) {

Direction photon_u = isotropic_direction(p.current_seed());

p.create_secondary(

p.wgt(), photon_u, MASS_ELECTRON_EV, ParticleType::photon());

p.create_secondary(

p.wgt(), -photon_u, MASS_ELECTRON_EV, ParticleType::photon());

// The annihilation photons are now emitted during the parent photon

// collision. Offset the pair-production Q value in the energy balance so

// heating matches the prior explicit positron slowing-down sequence.

p.bank_second_E() -= 2 * MASS_ELECTRON_EV;

}

}

void sample_electron_reaction(Particle& p)

{

// TODO: create reaction types

if (settings::electron_treatment == ElectronTreatment::TTB) {

thick_target_bremsstrahlung(p);

}

p.E() = 0.0;

p.wgt() = 0.0;

p.event() = TallyEvent::ABSORB;

}

void sample_positron_reaction(Particle& p)

{

// TODO: create reaction types

if (settings::electron_treatment == ElectronTreatment::TTB) {

thick_target_bremsstrahlung(p);

}

// Sample angle isotropically

Direction u = isotropic_direction(p.current_seed());

// Create annihilation photon pair traveling in opposite directions

p.create_secondary(p.wgt(), u, MASS_ELECTRON_EV, ParticleType::photon());

p.create_secondary(p.wgt(), -u, MASS_ELECTRON_EV, ParticleType::photon());

p.E() = 0.0;

p.wgt() = 0.0;

p.event() = TallyEvent::ABSORB;

}

int sample_nuclide(Particle& p)

{

// Sample cumulative distribution function

double cutoff = prn(p.current_seed()) * p.macro_xs().total;

// Get pointers to nuclide/density arrays

const auto& mat {model::materials[p.material()]};

int n = mat->nuclide_.size();

double prob = 0.0;

for (int i = 0; i < n; ++i) {

// Get atom density

int i_nuclide = mat->nuclide_[i];

double atom_density = mat->atom_density(i, p.density_mult());

// Increment probability to compare to cutoff

prob += atom_density * p.neutron_xs(i_nuclide).total;

if (prob >= cutoff)

return i_nuclide;

}

// If we reach here, no nuclide was sampled

p.write_restart();

throw std::runtime_error {"Did not sample any nuclide during collision."};

}

int sample_element(Particle& p)

{

// Sample cumulative distribution function

double cutoff = prn(p.current_seed()) * p.macro_xs().total;

// Get pointers to elements, densities

const auto& mat {model::materials[p.material()]};

double prob = 0.0;

for (int i = 0; i < mat->element_.size(); ++i) {

// Find atom density

int i_element = mat->element_[i];

double atom_density = mat->atom_density(i, p.density_mult());

// Determine microscopic cross section

double sigma = atom_density * p.photon_xs(i_element).total;

// Increment probability to compare to cutoff

prob += sigma;

if (prob > cutoff) {

// Save which nuclide particle had collision with for tally purpose

p.event_nuclide() = mat->nuclide_[i];

return i_element;

}

}

// If we made it here, no element was sampled

p.write_restart();

fatal_error("Did not sample any element during collision.");

}

Reaction& sample_fission(int i_nuclide, Particle& p)

{

// Get pointer to nuclide

const auto& nuc {data::nuclides[i_nuclide]};

// If we're in the URR, by default use the first fission reaction. We also

// default to the first reaction if we know that there are no partial fission

// reactions

if (p.neutron_xs(i_nuclide).use_ptable || !nuc->has_partial_fission_) {

return *nuc->fission_rx_[0];

}

// Check to see if we are in a windowed multipole range. WMP only supports

// the first fission reaction.

if (nuc->multipole_) {

if (p.E() >= nuc->multipole_->E_min_ && p.E() <= nuc->multipole_->E_max_) {

return *nuc->fission_rx_[0];

}

}

// Get grid index and interpolation factor and sample fission cdf

const auto& micro = p.neutron_xs(i_nuclide);

double cutoff = prn(p.current_seed()) * p.neutron_xs(i_nuclide).fission;

double prob = 0.0;

// Loop through each partial fission reaction type

for (auto& rx : nuc->fission_rx_) {

// add to cumulative probability

prob += rx->xs(micro);

// Create fission bank sites if fission occurs

if (prob > cutoff)

return *rx;

}

// If we reached here, no reaction was sampled

throw std::runtime_error {

"No fission reaction was sampled for " + nuc->name_};

}

void sample_photon_product(

int i_nuclide, Particle& p, int* i_rx, int* i_product)

{

// Get grid index and interpolation factor and sample photon production cdf

const auto& micro = p.neutron_xs(i_nuclide);

double cutoff = prn(p.current_seed()) * micro.photon_prod;

double prob = 0.0;

// Loop through each reaction type

const auto& nuc {data::nuclides[i_nuclide]};

for (int i = 0; i < nuc->reactions_.size(); ++i) {

// Evaluate neutron cross section

const auto& rx = nuc->reactions_[i];

double xs = rx->xs(micro);

// if cross section is zero for this reaction, skip it

if (xs == 0.0)

continue;

for (int j = 0; j < rx->products_.size(); ++j) {

if (rx->products_[j].particle_.is_photon()) {

// For fission, artificially increase the photon yield to account

// for delayed photons

double f = 1.0;

if (settings::delayed_photon_scaling) {

if (is_fission(rx->mt_)) {

if (nuc->prompt_photons_ && nuc->delayed_photons_) {

double energy_prompt = (*nuc->prompt_photons_)(p.E());

double energy_delayed = (*nuc->delayed_photons_)(p.E());

f = (energy_prompt + energy_delayed) / (energy_prompt);

}

}

}

// add to cumulative probability

prob += f * (*rx->products_[j].yield_)(p.E()) * xs;

*i_rx = i;

*i_product = j;

if (prob > cutoff)

return;

}

}

}

}

void absorption(Particle& p, int i_nuclide)

{

if (settings::survival_biasing) {

// Determine weight absorbed in survival biasing

const double wgt_absorb = p.wgt() * p.neutron_xs(i_nuclide).absorption /

p.neutron_xs(i_nuclide).total;

// Adjust weight of particle by probability of absorption

p.wgt() -= wgt_absorb;

// Score implicit absorption estimate of keff

if (settings::run_mode == RunMode::EIGENVALUE) {

p.keff_tally_absorption() += wgt_absorb *

p.neutron_xs(i_nuclide).nu_fission /

p.neutron_xs(i_nuclide).absorption;

}

} else {

// See if disappearance reaction happens

if (p.neutron_xs(i_nuclide).absorption >

prn(p.current_seed()) * p.neutron_xs(i_nuclide).total) {

// Score absorption estimate of keff

if (settings::run_mode == RunMode::EIGENVALUE) {

p.keff_tally_absorption() += p.wgt() *

p.neutron_xs(i_nuclide).nu_fission /

p.neutron_xs(i_nuclide).absorption;

}

p.wgt() = 0.0;

p.event() = TallyEvent::ABSORB;

if (!p.fission()) {

p.event_mt() = N_DISAPPEAR;

}

}

}

}

void scatter(Particle& p, int i_nuclide)

{

// copy incoming direction

Direction u_old {p.u()};

// Get pointer to nuclide and grid index/interpolation factor

const auto& nuc {data::nuclides[i_nuclide]};

const auto& micro {p.neutron_xs(i_nuclide)};

int i_temp = micro.index_temp;

// For tallying purposes, this routine might be called directly. In that

// case, we need to sample a reaction via the cutoff variable

double cutoff = prn(p.current_seed()) * (micro.total - micro.absorption);

bool sampled = false;

// Calculate elastic cross section if it wasn't precalculated

if (micro.elastic == CACHE_INVALID) {

nuc->calculate_elastic_xs(p);

}

double prob = micro.elastic - micro.thermal;

if (prob > cutoff) {

// =======================================================================

// NON-S(A,B) ELASTIC SCATTERING

// Determine temperature

double kT = nuc->multipole_ ? p.sqrtkT() * p.sqrtkT() : nuc->kTs_[i_temp];

// Perform collision physics for elastic scattering

elastic_scatter(i_nuclide, *nuc->reactions_[0], kT, p);

p.event_mt() = ELASTIC;

sampled = true;

}

prob = micro.elastic;

if (prob > cutoff && !sampled) {

// =======================================================================

// S(A,B) SCATTERING

sab_scatter(i_nuclide, micro.index_sab, p);

p.event_mt() = ELASTIC;

sampled = true;

}

if (!sampled) {

// =======================================================================

// INELASTIC SCATTERING

int n = nuc->index_inelastic_scatter_.size();

int i = 0;

for (int j = 0; j < n && prob < cutoff; ++j) {

i = nuc->index_inelastic_scatter_[j];

// add to cumulative probability

prob += nuc->reactions_[i]->xs(micro);

}

// Perform collision physics for inelastic scattering

const auto& rx {nuc->reactions_[i]};

inelastic_scatter(*nuc, *rx, p);

p.event_mt() = rx->mt_;

}

// Set event component

p.event() = TallyEvent::SCATTER;

// Sample new outgoing angle for isotropic-in-lab scattering

const auto& mat {model::materials[p.material()]};

if (!mat->p0_.empty()) {

int i_nuc_mat = mat->mat_nuclide_index_[i_nuclide];

if (mat->p0_[i_nuc_mat]) {

// Sample isotropic-in-lab outgoing direction

p.u() = isotropic_direction(p.current_seed());

p.mu() = u_old.dot(p.u());

}

}

}

void elastic_scatter(int i_nuclide, const Reaction& rx, double kT, Particle& p)

{

// get pointer to nuclide

const auto& nuc {data::nuclides[i_nuclide]};

double vel = std::sqrt(p.E());

double awr = nuc->awr_;

// Neutron velocity in LAB

Direction v_n = vel * p.u();

// Sample velocity of target nucleus

Direction v_t {};

if (!p.neutron_xs(i_nuclide).use_ptable) {

v_t = sample_target_velocity(*nuc, p.E(), p.u(), v_n,

p.neutron_xs(i_nuclide).elastic, kT, p.current_seed());

}

// Velocity of center-of-mass

Direction v_cm = (v_n + awr * v_t) / (awr + 1.0);

// Transform to CM frame

v_n -= v_cm;

// Find speed of neutron in CM

vel = v_n.norm();

// Sample scattering angle, checking if angle distribution is present (assume

// isotropic otherwise)

double mu_cm;

auto& d = rx.products_[0].distribution_[0];

auto d_ = dynamic_cast<UncorrelatedAngleEnergy*>(d.get());

if (!d_->angle().empty()) {

mu_cm = d_->angle().sample(p.E(), p.current_seed());

} else {

mu_cm = uniform_distribution(-1., 1., p.current_seed());

}

// Determine direction cosines in CM

Direction u_cm = v_n / vel;

// Rotate neutron velocity vector to new angle -- note that the speed of the

// neutron in CM does not change in elastic scattering. However, the speed

// will change when we convert back to LAB

v_n = vel * rotate_angle(u_cm, mu_cm, nullptr, p.current_seed());

// Transform back to LAB frame

v_n += v_cm;

p.E() = v_n.dot(v_n);

vel = std::sqrt(p.E());

// compute cosine of scattering angle in LAB frame by taking dot product of

// neutron's pre- and post-collision angle

p.mu() = p.u().dot(v_n) / vel;

// Set energy and direction of particle in LAB frame

p.u() = v_n / vel;

// Because of floating-point roundoff, it may be possible for mu_lab to be

// outside of the range [-1,1). In these cases, we just set mu_lab to exactly

// -1 or 1

if (std::abs(p.mu()) > 1.0)

p.mu() = std::copysign(1.0, p.mu());

}

void sab_scatter(int i_nuclide, int i_sab, Particle& p)

{

// Determine temperature index

const auto& micro {p.neutron_xs(i_nuclide)};

int i_temp = micro.index_temp_sab;

// Sample energy and angle

double E_out;

data::thermal_scatt[i_sab]->data_[i_temp].sample(

micro, p.E(), &E_out, &p.mu(), p.current_seed());

// Set energy to outgoing, change direction of particle

p.E() = E_out;

p.u() = rotate_angle(p.u(), p.mu(), nullptr, p.current_seed());

}

Direction sample_target_velocity(const Nuclide& nuc, double E, Direction u,

Direction v_neut, double xs_eff, double kT, uint64_t* seed)

{

// check if nuclide is a resonant scatterer

ResScatMethod sampling_method;

if (nuc.resonant_) {

// sampling method to use

sampling_method = settings::res_scat_method;

// upper resonance scattering energy bound (target is at rest above this E)

if (E > settings::res_scat_energy_max) {

return {};

// lower resonance scattering energy bound (should be no resonances below)

} else if (E < settings::res_scat_energy_min) {

sampling_method = ResScatMethod::cxs;

}

// otherwise, use free gas model

} else {

if (E >= settings::free_gas_threshold * kT && nuc.awr_ > 1.0) {

return {};

} else {

sampling_method = ResScatMethod::cxs;

}

}

// use appropriate target velocity sampling method

switch (sampling_method) {

case ResScatMethod::cxs:

// sample target velocity with the constant cross section (cxs) approx.

return sample_cxs_target_velocity(nuc.awr_, E, u, kT, seed);

case ResScatMethod::dbrc:

case ResScatMethod::rvs: {

double E_red = std::sqrt(nuc.awr_ * E / kT);

double E_low = std::pow(std::max(0.0, E_red - 4.0), 2) * kT / nuc.awr_;

double E_up = (E_red + 4.0) * (E_red + 4.0) * kT / nuc.awr_;

// find lower and upper energy bound indices

// lower index

int i_E_low;

if (E_low < nuc.energy_0K_.front()) {

i_E_low = 0;

} else if (E_low > nuc.energy_0K_.back()) {

i_E_low = nuc.energy_0K_.size() - 2;

} else {

i_E_low =

lower_bound_index(nuc.energy_0K_.begin(), nuc.energy_0K_.end(), E_low);

}

// upper index

int i_E_up;

if (E_up < nuc.energy_0K_.front()) {

i_E_up = 0;

} else if (E_up > nuc.energy_0K_.back()) {

i_E_up = nuc.energy_0K_.size() - 2;

} else {

i_E_up =

lower_bound_index(nuc.energy_0K_.begin(), nuc.energy_0K_.end(), E_up);

}

if (i_E_up == i_E_low) {

// Handle degenerate case -- if the upper/lower bounds occur for the same

// index, then using cxs is probably a good approximation

return sample_cxs_target_velocity(nuc.awr_, E, u, kT, seed);

}

if (sampling_method == ResScatMethod::dbrc) {

// interpolate xs since we're not exactly at the energy indices

double xs_low = nuc.elastic_0K_[i_E_low];

double m = (nuc.elastic_0K_[i_E_low + 1] - xs_low) /

(nuc.energy_0K_[i_E_low + 1] - nuc.energy_0K_[i_E_low]);

xs_low += m * (E_low - nuc.energy_0K_[i_E_low]);

double xs_up = nuc.elastic_0K_[i_E_up];

m = (nuc.elastic_0K_[i_E_up + 1] - xs_up) /

(nuc.energy_0K_[i_E_up + 1] - nuc.energy_0K_[i_E_up]);

xs_up += m * (E_up - nuc.energy_0K_[i_E_up]);

// get max 0K xs value over range of practical relative energies

double xs_max = *std::max_element(

&nuc.elastic_0K_[i_E_low + 1], &nuc.elastic_0K_[i_E_up + 1]);

xs_max = std::max({xs_low, xs_max, xs_up});

while (true) {

double E_rel;

Direction v_target;

while (true) {

// sample target velocity with the constant cross section (cxs)

// approx.

v_target = sample_cxs_target_velocity(nuc.awr_, E, u, kT, seed);

Direction v_rel = v_neut - v_target;

E_rel = v_rel.dot(v_rel);

if (E_rel < E_up)

break;

}

// perform Doppler broadening rejection correction (dbrc)

double xs_0K = nuc.elastic_xs_0K(E_rel);

double R = xs_0K / xs_max;

if (prn(seed) < R)

return v_target;

}

} else if (sampling_method == ResScatMethod::rvs) {

// interpolate xs CDF since we're not exactly at the energy indices

// cdf value at lower bound attainable energy

double cdf_low = 0.0;

if (E_low > nuc.energy_0K_.front()) {

double m = (nuc.xs_cdf_[i_E_low + 1] - nuc.xs_cdf_[i_E_low]) /

(nuc.energy_0K_[i_E_low + 1] - nuc.energy_0K_[i_E_low]);

cdf_low = nuc.xs_cdf_[i_E_low] + m * (E_low - nuc.energy_0K_[i_E_low]);

}

// cdf value at upper bound attainable energy

double m = (nuc.xs_cdf_[i_E_up + 1] - nuc.xs_cdf_[i_E_up]) /

(nuc.energy_0K_[i_E_up + 1] - nuc.energy_0K_[i_E_up]);

double cdf_up = nuc.xs_cdf_[i_E_up] + m * (E_up - nuc.energy_0K_[i_E_up]);

while (true) {

// directly sample Maxwellian

double E_t = -kT * std::log(prn(seed));

// sample a relative energy using the xs cdf

double cdf_rel = cdf_low + prn(seed) * (cdf_up - cdf_low);

int i_E_rel = lower_bound_index(nuc.xs_cdf_.begin() + i_E_low,

nuc.xs_cdf_.begin() + i_E_up + 2, cdf_rel);

double E_rel = nuc.energy_0K_[i_E_low + i_E_rel];

double m = (nuc.xs_cdf_[i_E_low + i_E_rel + 1] -