diff --git a/docs/source/capi/index.rst b/docs/source/capi/index.rst index eb6493e0e..222f81c4b 100644 --- a/docs/source/capi/index.rst +++ b/docs/source/capi/index.rst @@ -355,7 +355,7 @@ Functions Get density of a material. :param int32_t index: Index in the materials array - :param double* denity: Pointer to a density + :param double* density: Pointer to a density :return: Return status (negative if an error occurs) :rtype: int diff --git a/docs/source/devguide/workflow.rst b/docs/source/devguide/workflow.rst index 57f2b1c68..3bad04cae 100644 --- a/docs/source/devguide/workflow.rst +++ b/docs/source/devguide/workflow.rst @@ -4,7 +4,7 @@ Development Workflow ==================== -Anyone wishing to make contributions to OpenMC should be fully acquianted and +Anyone wishing to make contributions to OpenMC should be fully acquainted and comfortable working with git_ and GitHub_. We assume here that you have git installed on your system, have a GitHub account, and have setup SSH keys to be able to create/push to repositories on GitHub. @@ -81,7 +81,7 @@ features and bug fixes. The general steps for contributing are as follows: openmc-dev/openmc as the target. At a minimum, you should describe what the changes you've made are and why - you are making them. If the changes are related to an oustanding issue, make + you are making them. If the changes are related to an outstanding issue, make sure it is cross-referenced. 5. A committer will review your pull request based on the criteria diff --git a/docs/source/io_formats/geometry.rst b/docs/source/io_formats/geometry.rst index ff2f6eb74..ac48e48d2 100644 --- a/docs/source/io_formats/geometry.rst +++ b/docs/source/io_formats/geometry.rst @@ -48,7 +48,7 @@ Each ```` element can have the following attributes or sub-elements: :periodic_surface_id: If a periodic boundary condition is applied, this attribute identifies the - ``id`` of the corresponding periodic sufrace. + ``id`` of the corresponding periodic surface. The following quadratic surfaces can be modeled: diff --git a/docs/source/io_formats/statepoint.rst b/docs/source/io_formats/statepoint.rst index 591dc2777..f96d7181d 100644 --- a/docs/source/io_formats/statepoint.rst +++ b/docs/source/io_formats/statepoint.rst @@ -149,7 +149,7 @@ All values are given in seconds and are measured on the master process. finalization. - **transport** (*double*) -- Time spent transporting particles. - **inactive batches** (*double*) -- Time spent in the inactive - batches (including non-transport activities like communcating + batches (including non-transport activities like communicating sites). - **active batches** (*double*) -- Time spent in the active batches (including non-transport activities like communicating sites). diff --git a/docs/source/methods/cmfd.rst b/docs/source/methods/cmfd.rst index 344a4cc1a..66d545ac1 100644 --- a/docs/source/methods/cmfd.rst +++ b/docs/source/methods/cmfd.rst @@ -266,11 +266,11 @@ and eq. :eq:`eq_cell_bound` can be written in this generic form, The parameter :math:`\widetilde{D}_{l,m,n}^{u,g}` represents the linear coupling term between current and flux. These current relationships can be -sustituted into eq. :eq:`eq_neut_bal` to produce a linear system of multigroup +substituted into eq. :eq:`eq_neut_bal` to produce a linear system of multigroup diffusion equations for each spatial cell and energy group. However, a solution to these equations is not consistent with a higher order transport solution unless equivalence factors are present. This is because both the diffusion -approximation, governed by Fick's Law, and spatial trunction error will produce +approximation, governed by Fick's Law, and spatial truncation error will produce differences. Therefore, a nonlinear parameter, :math:`\widehat{D}_{l,m,n}^{u,g}`, is added to eqs. :eq:`eq_cell_cell` and :eq:`eq_cell_bound`. These equations are, respectively, diff --git a/docs/source/methods/cross_sections.rst b/docs/source/methods/cross_sections.rst index c6d1f44f4..2210dfa45 100644 --- a/docs/source/methods/cross_sections.rst +++ b/docs/source/methods/cross_sections.rst @@ -9,7 +9,7 @@ Continuous-Energy Data ---------------------- In OpenMC, the data governing the interaction of neutrons with various nuclei -for continous-energy problems are represented using an HDF5 format that can be +for continuous-energy problems are represented using an HDF5 format that can be produced by converting files in the ACE format, which is used by MCNP_ and Serpent_. ACE-format data can be generated with the NJOY_ nuclear data processing system, which converts raw `ENDF/B data`_ into linearly-interpolable diff --git a/docs/source/methods/energy_deposition.rst b/docs/source/methods/energy_deposition.rst index fbcccc26b..4675aee3f 100644 --- a/docs/source/methods/energy_deposition.rst +++ b/docs/source/methods/energy_deposition.rst @@ -7,7 +7,7 @@ Heating and Energy Deposition As particles traverse a problem, some portion of their energy is deposited at collision sites. This energy is deposited when charged particles, including electrons and recoil nuclei, undergo electromagnetic interactions with -surrounding electons and ions. The information describing how much energy +surrounding electrons and ions. The information describing how much energy is deposited for a specific reaction is referred to as "heating numbers" and can be computed using a program like NJOY with the ``heatr`` module. diff --git a/docs/source/methods/introduction.rst b/docs/source/methods/introduction.rst index 924d4858d..eacdd70f5 100644 --- a/docs/source/methods/introduction.rst +++ b/docs/source/methods/introduction.rst @@ -45,7 +45,7 @@ following steps: - Initialize the pseudorandom number generator. - - Read the contiuous-energy or multi-group cross section data specified in + - Read the continuous-energy or multi-group cross section data specified in the problem. - If using a special energy grid treatment such as a union energy grid or diff --git a/docs/source/methods/neutron_physics.rst b/docs/source/methods/neutron_physics.rst index 7c1490647..e9061eea1 100644 --- a/docs/source/methods/neutron_physics.rst +++ b/docs/source/methods/neutron_physics.rst @@ -182,7 +182,7 @@ Inelastic Scattering -------------------- Note that the multi-group mode makes no distinction between elastic or -inelastic scattering reactions. The spceific multi-group scattering +inelastic scattering reactions. The specific multi-group scattering implementation is discussed in the :ref:`multi-group-scatter` section. The major algorithms for inelastic scattering were described in previous diff --git a/docs/source/methods/parallelization.rst b/docs/source/methods/parallelization.rst index 29748807a..d9fc15c9f 100644 --- a/docs/source/methods/parallelization.rst +++ b/docs/source/methods/parallelization.rst @@ -293,7 +293,7 @@ Cost of Nearest Neighbor Algorithm ---------------------------------- With the communication cost of the traditional fission bank algorithm -quantified, we now proceed to discuss the communicatin cost of the proposed +quantified, we now proceed to discuss the communication cost of the proposed algorithm. Comparing the cost of communication of this algorithm with the traditional algorithm is not trivial due to fact that the cost will be a function of how many fission sites are sampled on each node. If each node @@ -398,7 +398,7 @@ equation :eq:`k-to-source`, we can relate the stochastic eigenvalue to the integral of the noise component of the source distribution as .. math:: - :label: noise-integeral + :label: noise-integral N\hat{k} = Nk + \sqrt{N} \int \hat{\epsilon}(\mathbf{r}) \: d\mathbf{r}.