minor typo fix

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,

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

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.

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

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

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}.