12776996362

Committed 14 Jan 2025 09:49PM UTC coverage: 84.938% (+0.2%) from 84.729%

Build # 12776996362

Build Type

Pull #3133

github

Committed by

web-flow

Commit Message

Merge 0495246d9 into 549cc0973

Pull Request Pull Request #3133: Kinetics parameters using Iterated Fission Probability

Run Details

318 of 330 new or added lines in 10 files covered. (96.36%)

1658 existing lines in 66 files now uncovered.

50402 of 59340 relevant lines covered (84.94%)

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91.3

/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()++;

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

  // Sample the reaction type
  sample_reaction(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 (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
    p.nu_bank().emplace_back();
    NuBank* nu_bank_entry = &p.nu_bank().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)	1,939,604,940✔
27	{
28	// Add to the collision counter for the particle
29	p.n_collision()++;	1,939,604,940✔
30
31	if (settings::weight_window_checkpoint_collision)	1,939,604,940✔
32	apply_weight_windows(p);	1,939,604,940✔
33
34	// Sample the reaction type
35	sample_reaction(p);	1,939,604,940✔
36
37	// Display information about collision
38	if ((settings::verbosity >= 10) \|\| p.trace()) {	1,939,604,940✔
UNCOV 39	write_message(fmt::format(" Energy Group = {}", p.g()), 1);	×
40	}
41	}	1,939,604,940✔
42
43	void sample_reaction(Particle& p)	1,939,604,940✔
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()) {	1,939,604,940✔
51	if (settings::run_mode == RunMode::EIGENVALUE \|\|	1,937,955,456✔
UNCOV 52	(settings::run_mode == RunMode::FIXED_SOURCE &&	×
53	settings::create_fission_neutrons)) {
54	create_fission_sites(p);	1,937,955,456✔
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.) {	1,939,604,940✔
61	absorption(p);	1,939,604,940✔
62	}
63	if (!p.alive())	1,939,604,940✔
64	return;	120,885,540✔
65
66	// Sample a scattering event to determine the energy of the exiting neutron
67	scatter(p);	1,818,719,400✔
68
69	// Play Russian roulette if survival biasing is turned on
70	if (settings::survival_biasing) {	1,818,719,400✔
71	if (p.wgt() < settings::weight_cutoff) {	4,596,048✔
72	russian_roulette(p, settings::weight_survive);	155,124✔
73	}
74	}
75	}
76
77	void scatter(Particle& p)	1,818,719,400✔
78	{
79	data::mg.macro_xs_[p.material()].sample_scatter(p.g_last(), p.g(), p.mu(),	2,147,483,647✔
80	p.wgt(), p.current_seed(), p.mg_xs_cache().t, p.mg_xs_cache().a);	1,818,719,400✔
81
82	// Rotate the angle
83	p.u() = rotate_angle(p.u(), p.mu(), nullptr, p.current_seed());	1,818,719,400✔
84
85	// Update energy value for downstream compatability (in tallying)
86	p.E() = data::mg.energy_bin_avg_[p.g()];	1,818,719,400✔
87
88	// Set event component
89	p.event() = TallyEvent::SCATTER;	1,818,719,400✔
90	}	1,818,719,400✔
91
92	void create_fission_sites(Particle& p)	1,937,955,456✔
93	{
94	// If uniform fission source weighting is turned on, we increase or decrease
95	// the expected number of fission sites produced
96	double weight = settings::ufs_on ? ufs_get_weight(p) : 1.0;	1,937,955,456✔
97
98	// Determine the expected number of neutrons produced
99	double nu_t = p.wgt() / simulation::keff * weight * p.macro_xs().nu_fission /	1,937,955,456✔
100	p.macro_xs().total;	1,937,955,456✔
101
102	// Sample the number of neutrons produced
103	int nu = static_cast<int>(nu_t);	1,937,955,456✔
104	if (prn(p.current_seed()) <= (nu_t - int(nu_t))) {	1,937,955,456✔
105	nu++;	121,148,676✔
106	}
107
108	// If no neutrons were produced then don't continue
109	if (nu == 0)	1,937,955,456✔
110	return;	1,816,805,988✔
111
112	// Initialize the counter of delayed neutrons encountered for each delayed
113	// group.
114	double nu_d[MAX_DELAYED_GROUPS] = {0.};	121,149,468✔
115
116	// Clear out particle's nu fission bank
117	p.nu_bank().clear();	121,149,468✔
118
119	p.fission() = true;	121,149,468✔
120
121	// Determine whether to place fission sites into the shared fission bank
122	// or the secondary particle bank.
123	bool use_fission_bank = (settings::run_mode == RunMode::EIGENVALUE);	121,149,468✔
124
125	// Counter for the number of fission sites successfully stored to the shared
126	// fission bank or the secondary particle bank
127	int n_sites_stored;
128
129	for (n_sites_stored = 0; n_sites_stored < nu; n_sites_stored++) {	242,300,544✔
130	// Initialize fission site object with particle data
131	SourceSite site;	121,151,076✔
132	site.r = p.r();	121,151,076✔
133	site.particle = ParticleType::neutron;	121,151,076✔
134	site.wgt = 1. / weight;	121,151,076✔
135	site.parent_id = p.id();	121,151,076✔
136	site.progeny_id = p.n_progeny()++;	121,151,076✔
137
138	// Sample the cosine of the angle, assuming fission neutrons are emitted
139	// isotropically
140	double mu = 2. * prn(p.current_seed()) - 1.;	121,151,076✔
141
142	// Sample the azimuthal angle uniformly in [0, 2.pi)
143	double phi = 2. * PI * prn(p.current_seed());	121,151,076✔
144	site.u.x = mu;	121,151,076✔
145	site.u.y = std::sqrt(1. - mu * mu) * std::cos(phi);	121,151,076✔
146	site.u.z = std::sqrt(1. - mu * mu) * std::sin(phi);	121,151,076✔
147
148	// Sample secondary energy distribution for the fission reaction
149	int dg;
150	int gout;
151	data::mg.macro_xs_[p.material()].sample_fission_energy(	363,453,228✔
152	p.g(), dg, gout, p.current_seed(), p.mg_xs_cache().t, p.mg_xs_cache().a);	242,302,152✔
153
154	// Store the energy and delayed groups on the fission bank
155	site.E = gout;	121,151,076✔
156
157	// We add 1 to the delayed_group bc in MG, -1 is prompt, but in the rest
158	// of the code, 0 is prompt.
159	site.delayed_group = dg + 1;	121,151,076✔
160
161	// Store fission site in bank
162	if (use_fission_bank) {	121,151,076✔
163	int64_t idx = simulation::fission_bank.thread_safe_append(site);	121,151,076✔
164	if (idx == -1) {	121,151,076✔
UNCOV 165	warning(	×
166	"The shared fission bank is full. Additional fission sites created "
167	"in this generation will not be banked. Results may be "
168	"non-deterministic.");
169
170	// Decrement number of particle progeny as storage was unsuccessful.
171	// This step is needed so that the sum of all progeny is equal to the
172	// size of the shared fission bank.
UNCOV 173	p.n_progeny()--;	×
174
175	// Break out of loop as no more sites can be added to fission bank
UNCOV 176	break;	×
177	}
178	} else {
UNCOV 179	p.secondary_bank().push_back(site);	×
180	}
181
182	// Set the delayed group on the particle as well
183	p.delayed_group() = dg + 1;	121,151,076✔
184
185	// Increment the number of neutrons born delayed
186	if (p.delayed_group() > 0) {	121,151,076✔
187	nu_d[dg]++;	1,680✔
188	}
189
190	// Write fission particles to nuBank
191	p.nu_bank().emplace_back();	121,151,076✔
192	NuBank* nu_bank_entry = &p.nu_bank().back();	121,151,076✔
193	nu_bank_entry->wgt = site.wgt;	121,151,076✔
194	nu_bank_entry->E = site.E;	121,151,076✔
195	nu_bank_entry->delayed_group = site.delayed_group;	121,151,076✔
196	}
197
198	// If shared fission bank was full, and no fissions could be added,
199	// set the particle fission flag to false.
200	if (n_sites_stored == 0) {	121,149,468✔
UNCOV 201	p.fission() = false;	×
UNCOV 202	return;	×
203	}
204
205	// Set nu to the number of fission sites successfully stored. If the fission
206	// bank was not found to be full then these values are already equivalent.
207	nu = n_sites_stored;	121,149,468✔
208
209	// Store the total weight banked for analog fission tallies
210	p.n_bank() = nu;	121,149,468✔
211	p.wgt_bank() = nu / weight;	121,149,468✔
212	for (size_t d = 0; d < MAX_DELAYED_GROUPS; d++) {	1,090,345,212✔
213	p.n_delayed_bank(d) = nu_d[d];	969,195,744✔
214	}
215	}
216
217	void absorption(Particle& p)	1,939,604,940✔
218	{
219	if (settings::survival_biasing) {	1,939,604,940✔
220	// Determine weight absorbed in survival biasing
221	double wgt_absorb = p.wgt() * p.macro_xs().absorption / p.macro_xs().total;	4,596,048✔
222
223	// Adjust weight of particle by the probability of absorption
224	p.wgt() -= wgt_absorb;	4,596,048✔
225
226	// Score implicit absorpion estimate of keff
227	p.keff_tally_absorption() +=	4,596,048✔
228	wgt_absorb * p.macro_xs().nu_fission / p.macro_xs().absorption;	4,596,048✔
229	} else {
230	if (p.macro_xs().absorption > prn(p.current_seed()) * p.macro_xs().total) {	1,935,008,892✔
231	p.keff_tally_absorption() +=	241,771,080✔
232	p.wgt() * p.macro_xs().nu_fission / p.macro_xs().absorption;	120,885,540✔
233	p.wgt() = 0.0;	120,885,540✔
234	p.event() = TallyEvent::ABSORB;	120,885,540✔
235	}
236	}
237	}	1,939,604,940✔
238
239	} // namespace openmc

openmc-dev / openmc / 12776996362

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