21458205742

Committed 28 Jan 2026 10:40PM UTC coverage: 81.974% (-0.02%) from 81.994%

Build # 21458205742

Build Type

Pull #3751

github

Committed by

web-flow

Commit Message

Merge a14249281 into 008d58460

Pull Request Pull Request #3751: Resolve conflict with weight windows and global russian roulette

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78.65

/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_windows_on) {
    auto ww = search_weight_window(p);
    if (ww.is_valid()) {
      if (settings::weight_window_checkpoint_collision)
        apply_weight_window(p, ww);
    } else {
      apply_russian_roulette(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 there are no weight windows
  if (!settings::weight_windows_on)
    apply_russian_roulette(p);
}

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_windows_on) {	1,783,060,477✔
35	auto ww = search_weight_window(p);	3,057,087✔
36	if (ww.is_valid()) {	3,057,087!
37	if (settings::weight_window_checkpoint_collision)	3,057,087!
38	apply_weight_window(p, ww);	3,057,087✔
39	} else {
NEW 40	apply_russian_roulette(p);	×
41	}
42	}
43
44	// Display information about collision
45	if ((settings::verbosity >= 10) \|\| p.trace()) {	1,783,060,477!
46	write_message(fmt::format(" Energy Group = {}", p.g()), 1);	×
47	}
48	}	1,783,060,477✔
49
50	void sample_reaction(Particle& p)	1,783,060,477✔
51	{
52	// Create fission bank sites. Note that while a fission reaction is sampled,
53	// it never actually "happens", i.e. the weight of the particle does not
54	// change when sampling fission sites. The following block handles all
55	// absorption (including fission)
56
57	if (model::materials[p.material()]->fissionable()) {	1,783,060,477✔
58	if (settings::run_mode == RunMode::EIGENVALUE \|\|	1,777,796,823!
59	(settings::run_mode == RunMode::FIXED_SOURCE &&	×
60	settings::create_fission_neutrons)) {
61	create_fission_sites(p);	1,777,796,823✔
62	}
63	}
64
65	// If survival biasing is being used, the following subroutine adjusts the
66	// weight of the particle. Otherwise, it checks to see if absorption occurs.
67	if (p.macro_xs().absorption > 0.) {	1,783,060,477!
68	absorption(p);	1,783,060,477✔
69	}
70	if (!p.alive())	1,783,060,477✔
71	return;	113,644,773✔
72
73	// Sample a scattering event to determine the energy of the exiting neutron
74	scatter(p);	1,669,415,704✔
75
76	// Play russian roulette if there are no weight windows
77	if (!settings::weight_windows_on)	1,669,415,704✔
78	apply_russian_roulette(p);	1,668,643,207✔
79	}
80
81	void scatter(Particle& p)	1,669,415,704✔
82	{
83	data::mg.macro_xs_[p.material()].sample_scatter(p.g_last(), p.g(), p.mu(),	2,147,483,647✔
84	p.wgt(), p.current_seed(), p.mg_xs_cache().t, p.mg_xs_cache().a);	1,669,415,704✔
85
86	// Rotate the angle
87	p.u() = rotate_angle(p.u(), p.mu(), nullptr, p.current_seed());	1,669,415,704✔
88
89	// Update energy value for downstream compatability (in tallying)
90	p.E() = data::mg.energy_bin_avg_[p.g()];	1,669,415,704✔
91
92	// Set event component
93	p.event() = TallyEvent::SCATTER;	1,669,415,704✔
94	}	1,669,415,704✔
95
96	void create_fission_sites(Particle& p)	1,777,796,823✔
97	{
98	// If uniform fission source weighting is turned on, we increase or decrease
99	// the expected number of fission sites produced
100	double weight = settings::ufs_on ? ufs_get_weight(p) : 1.0;	1,777,796,823!
101
102	// Determine the expected number of neutrons produced
103	double nu_t = p.wgt() / simulation::keff * weight * p.macro_xs().nu_fission /	1,777,796,823✔
104	p.macro_xs().total;	1,777,796,823✔
105
106	// Sample the number of neutrons produced
107	int nu = static_cast<int>(nu_t);	1,777,796,823✔
108	if (prn(p.current_seed()) <= (nu_t - int(nu_t))) {	1,777,796,823✔
109	nu++;	111,087,570✔
110	}
111
112	// If no neutrons were produced then don't continue
113	if (nu == 0)	1,777,796,823✔
114	return;	1,666,708,615✔
115
116	// Initialize the counter of delayed neutrons encountered for each delayed
117	// group.
118	double nu_d[MAX_DELAYED_GROUPS] = {0.};	111,088,208✔
119
120	// Clear out particle's nu fission bank
121	p.nu_bank().clear();	111,088,208✔
122
123	p.fission() = true;	111,088,208✔
124
125	// Determine whether to place fission sites into the shared fission bank
126	// or the secondary particle bank.
127	bool use_fission_bank = (settings::run_mode == RunMode::EIGENVALUE);	111,088,208✔
128
129	// Counter for the number of fission sites successfully stored to the shared
130	// fission bank or the secondary particle bank
131	int n_sites_stored;
132
133	for (n_sites_stored = 0; n_sites_stored < nu; n_sites_stored++) {	222,178,143✔
134	// Initialize fission site object with particle data
135	SourceSite site;	111,089,935✔
136	site.r = p.r();	111,089,935✔
137	site.particle = ParticleType::neutron;	111,089,935✔
138	site.time = p.time();	111,089,935✔
139	site.wgt = 1. / weight;	111,089,935✔
140
141	// Sample the cosine of the angle, assuming fission neutrons are emitted
142	// isotropically
143	double mu = 2. * prn(p.current_seed()) - 1.;	111,089,935✔
144
145	// Sample the azimuthal angle uniformly in [0, 2.pi)
146	double phi = 2. * PI * prn(p.current_seed());	111,089,935✔
147	site.u.x = mu;	111,089,935✔
148	site.u.y = std::sqrt(1. - mu * mu) * std::cos(phi);	111,089,935✔
149	site.u.z = std::sqrt(1. - mu * mu) * std::sin(phi);	111,089,935✔
150
151	// Sample secondary energy distribution for the fission reaction
152	int dg;
153	int gout;
154	data::mg.macro_xs_[p.material()].sample_fission_energy(	333,269,805✔
155	p.g(), dg, gout, p.current_seed(), p.mg_xs_cache().t, p.mg_xs_cache().a);	222,179,870✔
156
157	// Store the energy and delayed groups on the fission bank
158	site.E = gout;	111,089,935✔
159
160	// We add 1 to the delayed_group bc in MG, -1 is prompt, but in the rest
161	// of the code, 0 is prompt.
162	site.delayed_group = dg + 1;	111,089,935✔
163
164	// If delayed product production, sample time of emission
165	if (dg != -1) {	111,089,935✔
166	auto& macro_xs = data::mg.macro_xs_[p.material()];	1,507✔
167	double decay_rate =
168	macro_xs.get_xs(MgxsType::DECAY_RATE, 0, nullptr, nullptr, &dg, 0, 0);	1,507✔
169	site.time -= std::log(prn(p.current_seed())) / decay_rate;	1,507✔
170
171	// Reject site if it exceeds time cutoff
172	double t_cutoff = settings::time_cutoff[static_cast<int>(site.particle)];	1,507✔
173	if (site.time > t_cutoff) {	1,507✔
174	continue;	737✔
175	}
176	}
177
178	// Set parent and progeny ID
179	site.parent_id = p.id();	111,089,198✔
180	site.progeny_id = p.n_progeny()++;	111,089,198✔
181
182	// Store fission site in bank
183	if (use_fission_bank) {	111,089,198!
184	int64_t idx = simulation::fission_bank.thread_safe_append(site);	111,089,198✔
185	if (idx == -1) {	111,089,198!
186	warning(	×
187	"The shared fission bank is full. Additional fission sites created "
188	"in this generation will not be banked. Results may be "
189	"non-deterministic.");
190
191	// Decrement number of particle progeny as storage was unsuccessful.
192	// This step is needed so that the sum of all progeny is equal to the
193	// size of the shared fission bank.
194	p.n_progeny()--;	×
195
196	// Break out of loop as no more sites can be added to fission bank
197	break;	×
198	}
199	} else {
200	p.secondary_bank().push_back(site);	×
201	}
202
203	// Set the delayed group on the particle as well
204	p.delayed_group() = dg + 1;	111,089,198✔
205
206	// Increment the number of neutrons born delayed
207	if (p.delayed_group() > 0) {	111,089,198✔
208	nu_d[dg]++;	770✔
209	}
210
211	// Write fission particles to nuBank
212	NuBank& nu_bank_entry = p.nu_bank().emplace_back();	111,089,198✔
213	nu_bank_entry.wgt = site.wgt;	111,089,198✔
214	nu_bank_entry.E = site.E;	111,089,198✔
215	nu_bank_entry.delayed_group = site.delayed_group;	111,089,198✔
216	}
217
218	// If shared fission bank was full, and no fissions could be added,
219	// set the particle fission flag to false.
220	if (n_sites_stored == 0) {	111,088,208!
221	p.fission() = false;	×
222	return;	×
223	}
224
225	// Set nu to the number of fission sites successfully stored. If the fission
226	// bank was not found to be full then these values are already equivalent.
227	nu = n_sites_stored;	111,088,208✔
228
229	// Store the total weight banked for analog fission tallies
230	p.n_bank() = nu;	111,088,208✔
231	p.wgt_bank() = nu / weight;	111,088,208✔
232	for (size_t d = 0; d < MAX_DELAYED_GROUPS; d++) {	999,793,872✔
233	p.n_delayed_bank(d) = nu_d[d];	888,705,664✔
234	}
235	}
236
237	void absorption(Particle& p)	1,783,060,477✔
238	{
239	if (settings::survival_biasing) {	1,783,060,477✔
240	// Determine weight absorbed in survival biasing
241	double wgt_absorb = p.wgt() * p.macro_xs().absorption / p.macro_xs().total;	4,219,259✔
242
243	// Adjust weight of particle by the probability of absorption
244	p.wgt() -= wgt_absorb;	4,219,259✔
245
246	// Score implicit absorpion estimate of keff
247	p.keff_tally_absorption() +=	4,219,259✔
248	wgt_absorb * p.macro_xs().nu_fission / p.macro_xs().absorption;	4,219,259✔
249	} else {
250	if (p.macro_xs().absorption > prn(p.current_seed()) * p.macro_xs().total) {	1,778,841,218✔
251	p.keff_tally_absorption() +=	227,289,546✔
252	p.wgt() * p.macro_xs().nu_fission / p.macro_xs().absorption;	113,644,773✔
253	p.wgt() = 0.0;	113,644,773✔
254	p.event() = TallyEvent::ABSORB;	113,644,773✔
255	}
256	}
257	}	1,783,060,477✔
258
259	} // namespace openmc

openmc-dev / openmc / 21458205742

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