13710627373

Committed 07 Mar 2025 12:00AM UTC coverage: 85.127%. First build

Build # 13710627373

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Commit Message

Add option for survival biasing source normalization (#3070)

Co-authored-by: Patrick Shriwise <pshriwise@gmail.com>
Co-authored-by: Paul Romano <paul.k.romano@gmail.com>

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89.36

/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.wgt = 1. / weight;
    site.parent_id = p.id();
    site.progeny_id = p.n_progeny()++;

    // 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;

    // 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)	1,785,735,523✔
27	{
28	// Add to the collision counter for the particle
29	p.n_collision()++;	1,785,735,523✔
30
31	// Sample the reaction type
32	sample_reaction(p);	1,785,735,523✔
33
34	if (settings::weight_window_checkpoint_collision)	1,785,735,523✔
35	apply_weight_windows(p);	1,785,735,523✔
36
37	// Display information about collision
38	if ((settings::verbosity >= 10) \|\| p.trace()) {	1,785,735,523✔
39	write_message(fmt::format(" Energy Group = {}", p.g()), 1);	×
40	}
41	}	1,785,735,523✔
42
43	void sample_reaction(Particle& p)	1,785,735,523✔
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()) {	1,785,735,523✔
51	if (settings::run_mode == RunMode::EIGENVALUE \|\|	1,776,459,168✔
52	(settings::run_mode == RunMode::FIXED_SOURCE &&	×
53	settings::create_fission_neutrons)) {
54	create_fission_sites(p);	1,776,459,168✔
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.) {	1,785,735,523✔
61	absorption(p);	1,785,735,523✔
62	}
63	if (!p.alive())	1,785,735,523✔
64	return;	116,621,527✔
65
66	// Sample a scattering event to determine the energy of the exiting neutron
67	scatter(p);	1,669,113,996✔
68
69	// Play Russian roulette if survival biasing is turned on
70	if (settings::survival_biasing) {	1,669,113,996✔
71	// if survival normalization is applicable, use normalized weight cutoff and
72	// normalized weight survive
73	if (settings::survival_normalization) {	4,213,044✔
NEW 74	if (p.wgt() < settings::weight_cutoff * p.wgt_born()) {	×
NEW 75	russian_roulette(p, settings::weight_survive * p.wgt_born());	×
76	}
77	} else if (p.wgt() < settings::weight_cutoff) {	4,213,044✔
78	russian_roulette(p, settings::weight_survive);	142,197✔
79	}
80	}
81	}
82
83	void scatter(Particle& p)	1,669,113,996✔
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);	1,669,113,996✔
87
88	// Rotate the angle
89	p.u() = rotate_angle(p.u(), p.mu(), nullptr, p.current_seed());	1,669,113,996✔
90
91	// Update energy value for downstream compatability (in tallying)
92	p.E() = data::mg.energy_bin_avg_[p.g()];	1,669,113,996✔
93
94	// Set event component
95	p.event() = TallyEvent::SCATTER;	1,669,113,996✔
96	}	1,669,113,996✔
97
98	void create_fission_sites(Particle& p)	1,776,459,168✔
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;	1,776,459,168✔
103
104	// Determine the expected number of neutrons produced
105	double nu_t = p.wgt() / simulation::keff * weight * p.macro_xs().nu_fission /	1,776,459,168✔
106	p.macro_xs().total;	1,776,459,168✔
107
108	// Sample the number of neutrons produced
109	int nu = static_cast<int>(nu_t);	1,776,459,168✔
110	if (prn(p.current_seed()) <= (nu_t - int(nu_t))) {	1,776,459,168✔
111	nu++;	111,052,953✔
112	}
113
114	// If no neutrons were produced then don't continue
115	if (nu == 0)	1,776,459,168✔
116	return;	1,665,405,489✔
117
118	// Initialize the counter of delayed neutrons encountered for each delayed
119	// group.
120	double nu_d[MAX_DELAYED_GROUPS] = {0.};	111,053,679✔
121
122	// Clear out particle's nu fission bank
123	p.nu_bank().clear();	111,053,679✔
124
125	p.fission() = true;	111,053,679✔
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);	111,053,679✔
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++) {	222,108,832✔
136	// Initialize fission site object with particle data
137	SourceSite site;	111,055,153✔
138	site.r = p.r();	111,055,153✔
139	site.particle = ParticleType::neutron;	111,055,153✔
140	site.wgt = 1. / weight;	111,055,153✔
141	site.parent_id = p.id();	111,055,153✔
142	site.progeny_id = p.n_progeny()++;	111,055,153✔
143
144	// Sample the cosine of the angle, assuming fission neutrons are emitted
145	// isotropically
146	double mu = 2. * prn(p.current_seed()) - 1.;	111,055,153✔
147
148	// Sample the azimuthal angle uniformly in [0, 2.pi)
149	double phi = 2. * PI * prn(p.current_seed());	111,055,153✔
150	site.u.x = mu;	111,055,153✔
151	site.u.y = std::sqrt(1. - mu * mu) * std::cos(phi);	111,055,153✔
152	site.u.z = std::sqrt(1. - mu * mu) * std::sin(phi);	111,055,153✔
153
154	// Sample secondary energy distribution for the fission reaction
155	int dg;
156	int gout;
157	data::mg.macro_xs_[p.material()].sample_fission_energy(	333,165,459✔
158	p.g(), dg, gout, p.current_seed(), p.mg_xs_cache().t, p.mg_xs_cache().a);	222,110,306✔
159
160	// Store the energy and delayed groups on the fission bank
161	site.E = gout;	111,055,153✔
162
163	// We add 1 to the delayed_group bc in MG, -1 is prompt, but in the rest
164	// of the code, 0 is prompt.
165	site.delayed_group = dg + 1;	111,055,153✔
166
167	// Store fission site in bank
168	if (use_fission_bank) {	111,055,153✔
169	int64_t idx = simulation::fission_bank.thread_safe_append(site);	111,055,153✔
170	if (idx == -1) {	111,055,153✔
171	warning(	×
172	"The shared fission bank is full. Additional fission sites created "
173	"in this generation will not be banked. Results may be "
174	"non-deterministic.");
175
176	// Decrement number of particle progeny as storage was unsuccessful.
177	// This step is needed so that the sum of all progeny is equal to the
178	// size of the shared fission bank.
179	p.n_progeny()--;	×
180
181	// Break out of loop as no more sites can be added to fission bank
182	break;	×
183	}
184	} else {
185	p.secondary_bank().push_back(site);	×
186	}
187
188	// Set the delayed group on the particle as well
189	p.delayed_group() = dg + 1;	111,055,153✔
190
191	// Increment the number of neutrons born delayed
192	if (p.delayed_group() > 0) {	111,055,153✔
193	nu_d[dg]++;	1,540✔
194	}
195
196	// Write fission particles to nuBank
197	NuBank& nu_bank_entry = p.nu_bank().emplace_back();	111,055,153✔
198	nu_bank_entry.wgt = site.wgt;	111,055,153✔
199	nu_bank_entry.E = site.E;	111,055,153✔
200	nu_bank_entry.delayed_group = site.delayed_group;	111,055,153✔
201	}
202
203	// If shared fission bank was full, and no fissions could be added,
204	// set the particle fission flag to false.
205	if (n_sites_stored == 0) {	111,053,679✔
206	p.fission() = false;	×
207	return;	×
208	}
209
210	// Set nu to the number of fission sites successfully stored. If the fission
211	// bank was not found to be full then these values are already equivalent.
212	nu = n_sites_stored;	111,053,679✔
213
214	// Store the total weight banked for analog fission tallies
215	p.n_bank() = nu;	111,053,679✔
216	p.wgt_bank() = nu / weight;	111,053,679✔
217	for (size_t d = 0; d < MAX_DELAYED_GROUPS; d++) {	999,483,111✔
218	p.n_delayed_bank(d) = nu_d[d];	888,429,432✔
219	}
220	}
221
222	void absorption(Particle& p)	1,785,735,523✔
223	{
224	if (settings::survival_biasing) {	1,785,735,523✔
225	// Determine weight absorbed in survival biasing
226	double wgt_absorb = p.wgt() * p.macro_xs().absorption / p.macro_xs().total;	4,213,044✔
227
228	// Adjust weight of particle by the probability of absorption
229	p.wgt() -= wgt_absorb;	4,213,044✔
230
231	// Score implicit absorpion estimate of keff
232	p.keff_tally_absorption() +=	4,213,044✔
233	wgt_absorb * p.macro_xs().nu_fission / p.macro_xs().absorption;	4,213,044✔
234	} else {
235	if (p.macro_xs().absorption > prn(p.current_seed()) * p.macro_xs().total) {	1,781,522,479✔
236	p.keff_tally_absorption() +=	233,243,054✔
237	p.wgt() * p.macro_xs().nu_fission / p.macro_xs().absorption;	116,621,527✔
238	p.wgt() = 0.0;	116,621,527✔
239	p.event() = TallyEvent::ABSORB;	116,621,527✔
240	}
241	}
242	}	1,785,735,523✔
243
244	} // namespace openmc

openmc-dev / openmc / 13710627373

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