Merge devel into master, bringing master up to date

.github/ISSUE_TEMPLATE/bug_report.md

@@ -0,0 +1,31 @@
+---
+name: Bug report
+about: Report software bugs
+title: "[Bug]: "
+labels: Type:Bug
+assignees: ''
+
+---
+
+## Expected behavior
+A clear and concise description of what you expected to happen.
+
+## Describe the bug
+A clear and concise description of what the bug is.
+
+## To Reproduce
+Steps to reproduce the behavior:
+1. 
+2. 
+3. 
+4. 
+
+## Screenshots/output
+If applicable, add screenshots or program output to help explain your problem.
+
+## System Specifications:
+ - Version:
+ - Platform:
+ - Subsystem:
+
+## How can this issue be closed?

.github/ISSUE_TEMPLATE/feature_request.md

@@ -0,0 +1,16 @@
+---
+name: Feature request
+about: Suggest an idea for this project
+title: 'Feature: '
+labels: Status:1-New, Type:Feature
+assignees: ''
+
+---
+
+## Background and motivation
+
+## Description of idea
+
+## Implementation details
+
+## Potential snags

.github/dependabot.yml

@@ -0,0 +1,17 @@
+version: 2
+updates:
+  - package-ecosystem: "gitsubmodule"
+    directory: "/"
+    allow:
+      - dependency-name: "moose"
+    schedule:
+      interval: "monthly"
+    commit-message:
+      prefix: "auto_update_moose"
+    open-pull-requests-limit: 1 
+    labels:
+      - "Comp:Core"
+      - "Priority:1-Critical" 
+    reviewers:
+      - "smpark7"
+      - "nglaser3"

.github/pull_request_template.md

@@ -0,0 +1,38 @@
+## Summary of changes
+<!--- In one or more sentences, describe the PR you are submitting -->
+
+
+
+## Types of changes
+<!--- What types of changes does your code introduce? Put an `x` in all the boxes that apply: -->
+
+- [ ] Bug fix (non-breaking change which fixes an issue)
+- [ ] New feature (non-breaking change which adds functionality)
+- [ ] Breaking change (fix or feature that would cause existing functionality to change)
+
+## Required for Merging
+- [ ] I have read the **CONTRIBUTING** document.
+- [ ] My code follows the code style of this project.
+- [ ] My change requires a change to the documentation.
+- [ ] I have updated the documentation accordingly.
+- [ ] I have added tests to cover my changes.
+- [ ] All new and existing tests passed.
+  - [ ] CI tests pass
+  - [ ] Local tests pass (including Serpent2 integration tests)
+
+## Associated Issues and PRs
+<!--- Please note any issues or pull requests associated with this pull request -->
+
+- Issue: #
+
+
+## Associated Developers
+<!--- Please mention any developers who should be alerted of this PR -->
+
+- Dev: @
+
+
+## Checklist for Reviewers
+
+Reviewers should use [this link](https://arfc.github.io/manual/guides/pull_requests) to get to the 
+Review Checklist before they begin their review.

.gitmodules

@@ -4,3 +4,4 @@
 [submodule "moose"]
 	path = moose
 	url = ../../idaholab/moose
+	branch = devel

doc/content/bib/references.bib

@@ -0,0 +1,36 @@
+
+@article{placzek_milnes_1947,
+	title = {Milne's {Problem} in {Transport} {Theory}},
+	volume = {72},
+	url = {https://link.aps.org/doi/10.1103/PhysRev.72.550},
+	doi = {10.1103/PhysRev.72.550},
+	abstract = {A modified derivation of the Wiener-Hopf solution of Milne's problem is given in a form suitable for application to problems in the theory of neutron diffusion.},
+	number = {7},
+	urldate = {2024-05-31},
+	journal = {Physical Review},
+	author = {Placzek, G. and Seidel, W.},
+	month = oct,
+	year = {1947},
+	note = {Publisher: American Physical Society},
+	pages = {550--555},
+	file = {APS Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\IJVWPMDS\\PhysRev.72.html:text/html;Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\TVUWZQ6J\\Placzek and Seidel - 1947 - Milne's Problem in Transport Theory.pdf:application/pdf},
+}
+
+@article{rulko_variational_1995,
+	title = {Variational {P1} {Approximations} of {General}-{Geometry} {Multigroup} {Transport} {Problems}},
+	volume = {121},
+	issn = {0029-5639},
+	url = {https://doi.org/10.13182/NSE121-393},
+	doi = {10.13182/NSE121-393},
+	abstract = {A variational approximation is developed for general-geometry multigroup transport problems with arbitrary anisotropic scattering. The variational principle is based on a functional that approximates a reaction rate in a subdomain of the system. In principle, approximations that result from this functional “optimally”determine such reaction rates. The functional contains an arbitrary parameter α and requires the approximate solutions of a forward and an adjoint transport problem. If the basis functions for the forward and adjoint solutions are chosen to be linear functions of the angular variable Ω, the functional yields the familiar multigroup P1 equations for all values of α. However, the boundary conditions that result from the functional depend on α. In particular, for problems with vacuum boundaries, one obtains the conventional mixed boundary condition, but with an extrapolation distance that depends continuously on α. The choice α = 0 yields a generalization of boundary conditions derived earlier by Federighi and Pomraning for a more limited class of problems. The choice α = 1 yields a generalization of boundary conditions derived previously by Davis for mono-energetic problems. Other boundary conditions are obtained by choosing different values of α. We discuss this indeterminancy of a in conjunction with numerical experiments.},
+	number = {3},
+	urldate = {2024-05-31},
+	journal = {Nuclear Science and Engineering},
+	author = {Rulko, Robert P. and Tomašević, Djordje and Larsen, Edward W.},
+	month = dec,
+	year = {1995},
+	note = {Publisher: Taylor \& Francis
+\_eprint: https://doi.org/10.13182/NSE121-393},
+	pages = {393--404},
+	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\8NHDX7DM\\Rulko et al. - 1995 - Variational P1 Approximations of General-Geometry .pdf:application/pdf},
+}

doc/content/getting_started/theory.md

@@ -16,7 +16,46 @@ the MOOSE finite element framework, enabling highly flexible and scalable reacto
 
 ## Multigroup Neutron Diffusion
 
-More information to be added in future.
+The neutron diffusion equation is an approximation to the Boltzmann transport equation, and is derived by taking the zeroth and first moment with respect to $\hat\Omega$, the direction of neutron travel. 
+A full derivation of the diffusion equation will not be presented here for conciseness. Importantly, when taking the first moment of the transport equation the angular flux is assumed to be linearly anisotropic. 
+The approximations reduce the phase space through the elimination of angular dependence, but the resulting equation also has reduced fidelity when compared to the transport equation, particularly in regions where the neutron flux has strong angular dependence. 
+A non-exhaustive list of regions in which the neutron flux strongly depends on angle is near material interfaces between materials with highly dissimilar neutronic properties, within strong absorbers, and within near-void regions.
+As a deterministic method, the neutron diffusion method also requires discretization of the continuous energy dependence into energy groups consisting of non-overlapping, finite energy ranges across the entire energy spectrum. 
+This energy discretization creates a system of equations referred to as the multigroup neutron diffusion equations: 
+
+!equation
+\frac{1}{v_g} \frac{\partial\phi_g}{\partial t}-\nabla \cdot D_g \nabla \phi_g +\Sigma^R_{g} \phi_g =\sum^G_{g \neq g'} {\Sigma^s_{g'\rightarrow g} \phi_{g'}}+ \chi^p_{g} \sum^G_{g'=1} {\left(1- \beta \right) \nu_{g'} \Sigma^f_{g'}\phi_{g'} }+\chi^d_{g} \sum^I_i {\lambda_i C_i}
+
+The delayed neutron precursor distributions are governed by:
+
+!equation
+\frac{\partial C_i}{\partial t}=\sum^G_{g'=1}{\beta_i \nu \Sigma^f_{g'}\phi_{g'}}-\lambda_i C_i-\vec{u} \cdot \nabla C_i
+
+Notably, there are two production terms of neutrons, the prompt fission source and delayed neutron precursor decay source. 
+The first term describes the neutrons immediately born from fission, and the second term describes the neutrons born from the radioactive decay of neutron-emitting radionuclides, commonly called delayed neutron precursors. 
+The multigroup neutron diffusion equations are generally impossible to solve analytically for realistic problems and are therefore typically solved with numerical methods. 
+Moltres utilizes the [Finite Element Method](https://mooseframework.inl.gov/help/faq/what_is_fem.html) (FEM) capabilities provided by the MOOSE framework.  
+In FEM, the spatial domain is discretized into finite mesh elements and the time dependence is modeled through discrete time steps and time integration methods. 
+
+!table id=diff_table caption= Terms in multi-group neutron diffusion equation and their associated kernels
+|Term in Multi-Group Diffusion Equation | Associated Kernel |Definition of Term|
+| - | - | - |
+| [!eq](\frac{1}{v_g} \frac{\partial\phi_g}{\partial t}) | [NtTimeDerivative](NtTimeDerivative.md) | Time rate of change of energy group g |
+| [!eq](- \nabla \cdot D_g \nabla \phi_g) | [GroupDiffusion](GroupDiffusion.md) | Streaming term of energy group g |
+| [!eq](\Sigma^R_{g} \phi_g) | [SigmaR](SigmaR.md) | Removal from energy group g |
+| [!eq](\sum^G_{g \neq g'} {\Sigma^s_{g'\rightarrow g} \phi_{g'}}) | [InScatter](InScatter.md) | In-scattering into energy group g | 
+| [!eq](\chi^p_{g} \sum^G_{g'=1} {\left[1- \beta \right] \nu_{g'} \Sigma^f_{g'}\phi_{g'} } ) | [CoupledFissionKernel](CoupledFissionKernel.md) | Prompt fission neutron source |
+| [!eq](\chi^d_{g} \sum^I_i {\lambda_i C_i}) | [DelayedNeutronSource](DelayedNeutronSource.md) | Delayed fission neutron source |
+
+!table id=prec_table caption= Terms in delayed neutron precursor equation and their associated kernels
+| Term in Delayed Precursor Equation | Associated Kernel(s) | Definition of Term |
+| - | - | - |
+| [!eq](\frac{\partial C_i}{\partial t}) | [ScalarTransportTimeDerivative](ScalarTransportTimeDerivative.md) | Time rate of change of precursor population | 
+| [!eq](\sum^G_{g'=1}{\beta_i \nu \Sigma^f_{g'}\phi_{g'}}) | [PrecursorSource](PrecursorSource.md) | Production of precursor from fission 
+| [!eq](-\lambda_i C_i) | [PrecursorDecay](PrecursorDecay.md) | Loss of precusors due to radioactive decay | 
+| [!eq](-\vec{u} \cdot \nabla C_i) | [DGCoupledAdvection](DGCoupledAdvection.md), [DGFunctionConvection](DGFunctionConvection.md), [DGConvection](https://mooseframework.inl.gov/source/dgkernels/DGConvection.html) | Advection of the precursors |
+
+For the advective term in the delayed neutron precursor equation, $-\vec{u}\cdot \nabla C_i$, there are three kernels. DGCoupledAdvection is used when the velocity is a variable in the simulation, DGFunctionConvection is used when the velocity is a function, and DGConvection is for when the velocity is constant.
 
 ## Heat Transfer and Fluid Flow
 

doc/content/source/actions/NtAction.md

@@ -1,20 +1,42 @@
 # NtAction
 
-!alert construction title=Undocumented Action Class
-The NtAction has not been documented. The content listed below should be used as a starting point for
-documenting the class, which includes the typical automatic documentation associated with an Action;
-however, what is contained is ultimately determined by what is necessary to make the documentation
-clear for users.
+!alert note
+This action sets up ONLY the variables, kernels, and BCs for neutron diffusion,
+NOT the variables, kernels, and BCs for precursor tracking.
+To generate these, see [PrecursorAction](PrecursorAction.md).
 
 !syntax description /Nt/NtAction
 
 ## Overview
 
-!! Replace these lines with information regarding the NtAction action.
+```NtAction``` greatly reduces input file length and complexity by automatically
+setting up variables, kernels, and BCs for the neutron diffusion equations.
+When including only the required input parameters, this action constructs the
+neutron group variables and their associated kernels:
+[SigmaR](SigmaR.md), [GroupDiffusion](GroupDiffusion.md),
+[InScatter](InScatter.md), and [CoupledFissionKernel](CoupledFissionKernel.md).
+With ```account_delayed``` set to ```True```,
+[DelayedNeutronSource](DelayedNeutronSource.md) is also constructed,
+otherwise this kernel is not constructed and the simulation does not consider
+the delayed neutron precursor effects on the neutron flux distributions.
+
+For more information regarding the use of ```NtAction``` please refer to the
+tutorials located [here](tutorials.md), specifically the +Multiphysics Reactor
+Simulations+ section.
 
 ## Example Input File Syntax
 
-!! Describe and include an example of how to use the NtAction action.
+An example input file without the ```NtAction```, showing only the portion
+affected by the action:
+
+!listing tutorial/eigenvalue/nts.i
+	start=Variables end=Materials remove=Precursors
+
+
+And then the same information created with the ```NtAction```:
+
+!listing tutorial/eigenvalue/nts-action.i
+	block=Nt
 
 !syntax description /Nt/NtAction
 

doc/content/source/bcs/VacuumConcBC.md

@@ -4,12 +4,25 @@
 
 ## Overview
 
-This object adds the $\frac{\phi}{4}-\frac{D_g}{2}\hat{n}\cdot\nabla\phi_g = 0$ vacuum boundary
+This object adds the $\phi_g+\alpha D_g\hat{n}\cdot\nabla\phi_g = 0$ vacuum boundary
 condition of the multigroup neutron diffusion equations. The weak form after applying integration
 by parts to the neutron diffusion term ([GroupDiffusion](/GroupDiffusion.md)) is:
 
 !equation
-\int_{\partial V}\psi D_g\nabla\phi_g\cdot\hat{n}\ dS = \int_{\partial V}\psi\frac{\phi_g}{2}\ dS
+-\int_{\partial V}\psi D_g\nabla\phi_g\cdot\hat{n}\ dS =
+\int_{\partial V}\psi\frac{\phi_g}{\alpha}\ dS
+
+The value of $\alpha$ varies depending on the BC type selected using the `vacuum_bc_type`
+parameter. The available options are:
+
+1. Marshak: $\alpha=2$
+2. Mark: $\alpha=\sqrt{3}$
+3. Milne: $\alpha=3\times 0.710446$
+
+The Marshak and Mark BCs are derived from vacuum boundary condition approximations using
+$P_1$ and $S_2$ transport methods, respectively. The Milne BC is derived from the exact analytical
+solution of the Milne problem [!citep](placzek_milnes_1947) and has been proven to be accurate for
+a wide range of diffusive problems [!citep](rulko_variational_1995).
 
 ## Example Input File Syntax
 

include/bcs/VacuumConcBC.h

@@ -4,10 +4,13 @@
 #include "ScalarTransportBase.h"
 
 /**
- * Implements a simple VacuumConc BC for neutron diffusion on the boundary.
- * VacuumConc BC is defined as \f$ D\frac{du}{dn}+\frac{u}{2} = 0\f$, where u is neutron flux.
- * Hence, \f$ D\frac{du}{dn}=-\frac{u}{2} \f$ and \f$ -\frac{u}{2} \f$ is substituted into
- * the Neumann BC term produced from integrating the diffusion operator by parts.
+ * Implements a simple VacuumConc BC for neutron diffusion on vacuum boundaries.
+ * VacuumConc BC is defined as \f$ D\frac{du}{dn}+\frac{u}{\alpha} = 0\f$, where u is neutron flux.
+ * Hence, \f$ D\frac{du}{dn}=-\frac{u}{\alpha} \f$ and \f$ -\frac{u}{\alpha} \f$ is substituted
+ * into the Neumann BC term produced from integrating the diffusion operator by parts.
+ *
+ * The three types of vacuum BCs available are Marshak ($\alpha=2$), Mark ($\alpha=\sqrt{3}$), and
+ * Milne ($\alpha=3\times 0.710446$). VacuumConcBC defaults to Marshak if `bc_type` is not set.
  */
 class VacuumConcBC : public IntegratedBC, public ScalarTransportBase
 {
@@ -22,9 +25,20 @@ class VacuumConcBC : public IntegratedBC, public ScalarTransportBase
 
 protected:
   virtual Real computeQpResidual() override;
-
   virtual Real computeQpJacobian() override;
 
-  /// Ratio of u to du/dn
+  enum BC_TYPE
+  {
+    MARSHAK,
+    MARK,
+    MILNE
+  };
+
+  // Ratio of u to du/dn
   Real _alpha;
+
+  // Milne vacuum boundary extrapolation coefficient
+  // Derived from the exact analytical solution to the Milne problem. See MooseDocs-based
+  // documentation for more information.
+  Real _milne_extrapolation_coefficient = 3 * 0.710446;
 };

include/kernels/INSBoussinesqBodyForce.h

@@ -19,7 +19,7 @@ class INSBoussinesqBodyForce : public Kernel
 protected:
   virtual Real computeQpResidual() override;
   virtual Real computeQpJacobian() override;
-  virtual Real computeQpOffDiagJacobian(unsigned jvar);
+  virtual Real computeQpOffDiagJacobian(unsigned jvar) override;
 
   // Parameters
   const VariableValue & _T;