19076324204

Committed 04 Nov 2025 04:51PM UTC coverage: 82.008% (-3.1%) from 85.155%

Build # 19076324204

Build Type

Pull #3252

github

Committed by

web-flow

Commit Message

Merge b4a815914 into bd76fc056

Pull Request Pull Request #3252: Adding vtkhdf option to write vtk data

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77.84

/src/physics_mg.cpp

#include "openmc/physics_mg.h"

#include <stdexcept>

#include "xtensor/xarray.hpp"
#include <fmt/core.h>

#include "openmc/bank.h"
#include "openmc/constants.h"
#include "openmc/eigenvalue.h"
#include "openmc/error.h"
#include "openmc/material.h"
#include "openmc/math_functions.h"
#include "openmc/message_passing.h"
#include "openmc/mgxs_interface.h"
#include "openmc/particle.h"
#include "openmc/physics_common.h"
#include "openmc/random_lcg.h"
#include "openmc/settings.h"
#include "openmc/simulation.h"
#include "openmc/tallies/tally.h"
#include "openmc/weight_windows.h"

namespace openmc {

void collision_mg(Particle& p)
{
  // Add to the collision counter for the particle
  p.n_collision()++;

  // Sample the reaction type
  sample_reaction(p);

  if (settings::weight_window_checkpoint_collision)
    apply_weight_windows(p);

  // Display information about collision
  if ((settings::verbosity >= 10) || p.trace()) {
    write_message(fmt::format("    Energy Group = {}", p.g()), 1);
  }
}

void sample_reaction(Particle& p)
{
  // 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)

  if (model::materials[p.material()]->fissionable()) {
    if (settings::run_mode == RunMode::EIGENVALUE ||
        (settings::run_mode == RunMode::FIXED_SOURCE &&
          settings::create_fission_neutrons)) {
      create_fission_sites(p);
    }
  }

  // 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.macro_xs().absorption > 0.) {
    absorption(p);
  }
  if (!p.alive())
    return;

  // Sample a scattering event to determine the energy of the exiting neutron
  scatter(p);

  // Play Russian roulette if survival biasing is turned on
  if (settings::survival_biasing) {
    // if survival normalization is applicable, use normalized weight cutoff and
    // normalized weight survive
    if (settings::survival_normalization) {
      if (p.wgt() < settings::weight_cutoff * p.wgt_born()) {
        russian_roulette(p, settings::weight_survive * p.wgt_born());
      }
    } else if (p.wgt() < settings::weight_cutoff) {
      russian_roulette(p, settings::weight_survive);
    }
  }
}

void scatter(Particle& p)
{
  data::mg.macro_xs_[p.material()].sample_scatter(p.g_last(), p.g(), p.mu(),
    p.wgt(), p.current_seed(), p.mg_xs_cache().t, p.mg_xs_cache().a);

  // Rotate the angle
  p.u() = rotate_angle(p.u(), p.mu(), nullptr, p.current_seed());

  // Update energy value for downstream compatability (in tallying)
  p.E() = data::mg.energy_bin_avg_[p.g()];

  // Set event component
  p.event() = TallyEvent::SCATTER;
}

void create_fission_sites(Particle& p)
{
  // 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.macro_xs().nu_fission /
                p.macro_xs().total;

  // Sample the number of neutrons produced
  int nu = static_cast<int>(nu_t);
  if (prn(p.current_seed()) <= (nu_t - int(nu_t))) {
    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;

    // Sample the cosine of the angle, assuming fission neutrons are emitted
    // isotropically
    double mu = 2. * prn(p.current_seed()) - 1.;

    // Sample the azimuthal angle uniformly in [0, 2.pi)
    double phi = 2. * PI * prn(p.current_seed());
    site.u.x = mu;
    site.u.y = std::sqrt(1. - mu * mu) * std::cos(phi);
    site.u.z = std::sqrt(1. - mu * mu) * std::sin(phi);

    // Sample secondary energy distribution for the fission reaction
    int dg;
    int gout;
    data::mg.macro_xs_[p.material()].sample_fission_energy(
      p.g(), dg, gout, p.current_seed(), p.mg_xs_cache().t, p.mg_xs_cache().a);

    // Store the energy and delayed groups on the fission bank
    site.E = gout;

    // We add 1 to the delayed_group bc in MG, -1 is prompt, but in the rest
    // of the code, 0 is prompt.
    site.delayed_group = dg + 1;

    // If delayed product production, sample time of emission
    if (dg != -1) {
      auto& macro_xs = data::mg.macro_xs_[p.material()];
      double decay_rate =
        macro_xs.get_xs(MgxsType::DECAY_RATE, 0, nullptr, nullptr, &dg, 0, 0);
      site.time -= std::log(prn(p.current_seed())) / decay_rate;

      // Reject site if it exceeds time cutoff
      double t_cutoff = settings::time_cutoff[static_cast<int>(site.particle)];
      if (site.time > t_cutoff) {
        continue;
      }
    }

    // Set parent and progeny ID
    site.parent_id = p.id();
    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;
      }
    } else {
      p.secondary_bank().push_back(site);
    }

    // Set the delayed group on the particle as well
    p.delayed_group() = dg + 1;

    // Increment the number of neutrons born delayed
    if (p.delayed_group() > 0) {
      nu_d[dg]++;
    }

    // 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 absorption(Particle& p)
{
  if (settings::survival_biasing) {
    // Determine weight absorbed in survival biasing
    double wgt_absorb = p.wgt() * p.macro_xs().absorption / p.macro_xs().total;

    // Adjust weight of particle by the probability of absorption
    p.wgt() -= wgt_absorb;

    // Score implicit absorpion estimate of keff
    p.keff_tally_absorption() +=
      wgt_absorb * p.macro_xs().nu_fission / p.macro_xs().absorption;
  } else {
    if (p.macro_xs().absorption > prn(p.current_seed()) * p.macro_xs().total) {
      p.keff_tally_absorption() +=
        p.wgt() * p.macro_xs().nu_fission / p.macro_xs().absorption;
      p.wgt() = 0.0;
      p.event() = TallyEvent::ABSORB;
    }
  }
}

} // namespace openmc

1	#include "openmc/physics_mg.h"
2
3	#include <stdexcept>
4
5	#include "xtensor/xarray.hpp"
6	#include <fmt/core.h>
7
8	#include "openmc/bank.h"
9	#include "openmc/constants.h"
10	#include "openmc/eigenvalue.h"
11	#include "openmc/error.h"
12	#include "openmc/material.h"
13	#include "openmc/math_functions.h"
14	#include "openmc/message_passing.h"
15	#include "openmc/mgxs_interface.h"
16	#include "openmc/particle.h"
17	#include "openmc/physics_common.h"
18	#include "openmc/random_lcg.h"
19	#include "openmc/settings.h"
20	#include "openmc/simulation.h"
21	#include "openmc/tallies/tally.h"
22	#include "openmc/weight_windows.h"
23
24	namespace openmc {
25
26	void collision_mg(Particle& p)	1,783,060,477✔
27	{
28	// Add to the collision counter for the particle
29	p.n_collision()++;	1,783,060,477✔
30
31	// Sample the reaction type
32	sample_reaction(p);	1,783,060,477✔
33
34	if (settings::weight_window_checkpoint_collision)	1,783,060,477!
35	apply_weight_windows(p);	1,783,060,477✔
36
37	// Display information about collision
38	if ((settings::verbosity >= 10) \|\| p.trace()) {	1,783,060,477!
39	write_message(fmt::format(" Energy Group = {}", p.g()), 1);	×
40	}
41	}	1,783,060,477✔
42
43	void sample_reaction(Particle& p)	1,783,060,477✔
44	{
45	// Create fission bank sites. Note that while a fission reaction is sampled,
46	// it never actually "happens", i.e. the weight of the particle does not
47	// change when sampling fission sites. The following block handles all
48	// absorption (including fission)
49
50	if (model::materials[p.material()]->fissionable()) {	1,783,060,477✔
51	if (settings::run_mode == RunMode::EIGENVALUE \|\|	1,777,796,823!
52	(settings::run_mode == RunMode::FIXED_SOURCE &&	×
53	settings::create_fission_neutrons)) {
54	create_fission_sites(p);	1,777,796,823✔
55	}
56	}
57
58	// If survival biasing is being used, the following subroutine adjusts the
59	// weight of the particle. Otherwise, it checks to see if absorption occurs.
60	if (p.macro_xs().absorption > 0.) {	1,783,060,477!
61	absorption(p);	1,783,060,477✔
62	}
63	if (!p.alive())	1,783,060,477✔
64	return;	113,644,773✔
65
66	// Sample a scattering event to determine the energy of the exiting neutron
67	scatter(p);	1,669,415,704✔
68
69	// Play Russian roulette if survival biasing is turned on
70	if (settings::survival_biasing) {	1,669,415,704✔
71	// if survival normalization is applicable, use normalized weight cutoff and
72	// normalized weight survive
73	if (settings::survival_normalization) {	4,219,259!
74	if (p.wgt() < settings::weight_cutoff * p.wgt_born()) {	×
75	russian_roulette(p, settings::weight_survive * p.wgt_born());	×
76	}
77	} else if (p.wgt() < settings::weight_cutoff) {	4,219,259✔
78	russian_roulette(p, settings::weight_survive);	142,340✔
79	}
80	}
81	}
82
83	void scatter(Particle& p)	1,669,415,704✔
84	{
85	data::mg.macro_xs_[p.material()].sample_scatter(p.g_last(), p.g(), p.mu(),	2,147,483,647✔
86	p.wgt(), p.current_seed(), p.mg_xs_cache().t, p.mg_xs_cache().a);	1,669,415,704✔
87
88	// Rotate the angle
89	p.u() = rotate_angle(p.u(), p.mu(), nullptr, p.current_seed());	1,669,415,704✔
90
91	// Update energy value for downstream compatability (in tallying)
92	p.E() = data::mg.energy_bin_avg_[p.g()];	1,669,415,704✔
93
94	// Set event component
95	p.event() = TallyEvent::SCATTER;	1,669,415,704✔
96	}	1,669,415,704✔
97
98	void create_fission_sites(Particle& p)	1,777,796,823✔
99	{
100	// If uniform fission source weighting is turned on, we increase or decrease
101	// the expected number of fission sites produced
102	double weight = settings::ufs_on ? ufs_get_weight(p) : 1.0;	1,777,796,823!
103
104	// Determine the expected number of neutrons produced
105	double nu_t = p.wgt() / simulation::keff * weight * p.macro_xs().nu_fission /	1,777,796,823✔
106	p.macro_xs().total;	1,777,796,823✔
107
108	// Sample the number of neutrons produced
109	int nu = static_cast<int>(nu_t);	1,777,796,823✔
110	if (prn(p.current_seed()) <= (nu_t - int(nu_t))) {	1,777,796,823✔
111	nu++;	111,087,570✔
112	}
113
114	// If no neutrons were produced then don't continue
115	if (nu == 0)	1,777,796,823✔
116	return;	1,666,708,615✔
117
118	// Initialize the counter of delayed neutrons encountered for each delayed
119	// group.
120	double nu_d[MAX_DELAYED_GROUPS] = {0.};	111,088,208✔
121
122	// Clear out particle's nu fission bank
123	p.nu_bank().clear();	111,088,208✔
124
125	p.fission() = true;	111,088,208✔
126
127	// Determine whether to place fission sites into the shared fission bank
128	// or the secondary particle bank.
129	bool use_fission_bank = (settings::run_mode == RunMode::EIGENVALUE);	111,088,208✔
130
131	// Counter for the number of fission sites successfully stored to the shared
132	// fission bank or the secondary particle bank
133	int n_sites_stored;
134
135	for (n_sites_stored = 0; n_sites_stored < nu; n_sites_stored++) {	222,178,143✔
136	// Initialize fission site object with particle data
137	SourceSite site;	111,089,935✔
138	site.r = p.r();	111,089,935✔
139	site.particle = ParticleType::neutron;	111,089,935✔
140	site.time = p.time();	111,089,935✔
141	site.wgt = 1. / weight;	111,089,935✔
142
143	// Sample the cosine of the angle, assuming fission neutrons are emitted
144	// isotropically
145	double mu = 2. * prn(p.current_seed()) - 1.;	111,089,935✔
146
147	// Sample the azimuthal angle uniformly in [0, 2.pi)
148	double phi = 2. * PI * prn(p.current_seed());	111,089,935✔
149	site.u.x = mu;	111,089,935✔
150	site.u.y = std::sqrt(1. - mu * mu) * std::cos(phi);	111,089,935✔
151	site.u.z = std::sqrt(1. - mu * mu) * std::sin(phi);	111,089,935✔
152
153	// Sample secondary energy distribution for the fission reaction
154	int dg;
155	int gout;
156	data::mg.macro_xs_[p.material()].sample_fission_energy(	333,269,805✔
157	p.g(), dg, gout, p.current_seed(), p.mg_xs_cache().t, p.mg_xs_cache().a);	222,179,870✔
158
159	// Store the energy and delayed groups on the fission bank
160	site.E = gout;	111,089,935✔
161
162	// We add 1 to the delayed_group bc in MG, -1 is prompt, but in the rest
163	// of the code, 0 is prompt.
164	site.delayed_group = dg + 1;	111,089,935✔
165
166	// If delayed product production, sample time of emission
167	if (dg != -1) {	111,089,935✔
168	auto& macro_xs = data::mg.macro_xs_[p.material()];	1,507✔
169	double decay_rate =
170	macro_xs.get_xs(MgxsType::DECAY_RATE, 0, nullptr, nullptr, &dg, 0, 0);	1,507✔
171	site.time -= std::log(prn(p.current_seed())) / decay_rate;	1,507✔
172
173	// Reject site if it exceeds time cutoff
174	double t_cutoff = settings::time_cutoff[static_cast<int>(site.particle)];	1,507✔
175	if (site.time > t_cutoff) {	1,507✔
176	continue;	737✔
177	}
178	}
179
180	// Set parent and progeny ID
181	site.parent_id = p.id();	111,089,198✔
182	site.progeny_id = p.n_progeny()++;	111,089,198✔
183
184	// Store fission site in bank
185	if (use_fission_bank) {	111,089,198!
186	int64_t idx = simulation::fission_bank.thread_safe_append(site);	111,089,198✔
187	if (idx == -1) {	111,089,198!
UNCOV 188	warning(	×
189	"The shared fission bank is full. Additional fission sites created "
190	"in this generation will not be banked. Results may be "
191	"non-deterministic.");
192
193	// Decrement number of particle progeny as storage was unsuccessful.
194	// This step is needed so that the sum of all progeny is equal to the
195	// size of the shared fission bank.
UNCOV 196	p.n_progeny()--;	×
197
198	// Break out of loop as no more sites can be added to fission bank
UNCOV 199	break;	×
200	}
201	} else {
UNCOV 202	p.secondary_bank().push_back(site);	×
203	}
204
205	// Set the delayed group on the particle as well
206	p.delayed_group() = dg + 1;	111,089,198✔
207
208	// Increment the number of neutrons born delayed
209	if (p.delayed_group() > 0) {	111,089,198✔
210	nu_d[dg]++;	770✔
211	}
212
213	// Write fission particles to nuBank
214	NuBank& nu_bank_entry = p.nu_bank().emplace_back();	111,089,198✔
215	nu_bank_entry.wgt = site.wgt;	111,089,198✔
216	nu_bank_entry.E = site.E;	111,089,198✔
217	nu_bank_entry.delayed_group = site.delayed_group;	111,089,198✔
218	}
219
220	// If shared fission bank was full, and no fissions could be added,
221	// set the particle fission flag to false.
222	if (n_sites_stored == 0) {	111,088,208!
UNCOV 223	p.fission() = false;	×
UNCOV 224	return;	×
225	}
226
227	// Set nu to the number of fission sites successfully stored. If the fission
228	// bank was not found to be full then these values are already equivalent.
229	nu = n_sites_stored;	111,088,208✔
230
231	// Store the total weight banked for analog fission tallies
232	p.n_bank() = nu;	111,088,208✔
233	p.wgt_bank() = nu / weight;	111,088,208✔
234	for (size_t d = 0; d < MAX_DELAYED_GROUPS; d++) {	999,793,872✔
235	p.n_delayed_bank(d) = nu_d[d];	888,705,664✔
236	}
237	}
238
239	void absorption(Particle& p)	1,783,060,477✔
240	{
241	if (settings::survival_biasing) {	1,783,060,477✔
242	// Determine weight absorbed in survival biasing
243	double wgt_absorb = p.wgt() * p.macro_xs().absorption / p.macro_xs().total;	4,219,259✔
244
245	// Adjust weight of particle by the probability of absorption
246	p.wgt() -= wgt_absorb;	4,219,259✔
247
248	// Score implicit absorpion estimate of keff
249	p.keff_tally_absorption() +=	4,219,259✔
250	wgt_absorb * p.macro_xs().nu_fission / p.macro_xs().absorption;	4,219,259✔
251	} else {
252	if (p.macro_xs().absorption > prn(p.current_seed()) * p.macro_xs().total) {	1,778,841,218✔
253	p.keff_tally_absorption() +=	227,289,546✔
254	p.wgt() * p.macro_xs().nu_fission / p.macro_xs().absorption;	113,644,773✔
255	p.wgt() = 0.0;	113,644,773✔
256	p.event() = TallyEvent::ABSORB;	113,644,773✔
257	}
258	}
259	}	1,783,060,477✔
260
261	} // namespace openmc

openmc-dev / openmc / 19076324204

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