armi.physics.neutronics.crossSectionGroupManager module

Cross section group manager handles burnup-dependent properties of microscopic cross sections.

Blocks are specified to be in a certain cross section type and burnup group. Together, these form the cross section group. By advancing blocks by their burnup into different groups, we capture some of the physical effects related to depletion.

XS types are typically single capital letters like A BU groups are also capital letters. A XS group of AB is in XS type A and burnup group B.

This module groups the blocks according to their XS groups and can determine which block is to be deemed representative of an entire set of blocks in a particular xs group. Then the representative block is sent to a lattice physics kernel for actual physics calculations.

Generally, the cross section manager is a attribute of the lattice physics code interface (e.g. the mc2Interface.

Examples

csm = CrossSectionGroupManager() csm._setBuGroupBounds(cs[‘buGroups’]) csm._addXsGroupsFromBlocks(blockList) csm.createRepresentativeBlocks() representativeBlockList = csm.representativeBlocks.values() blockThatRepresentsBA = csm.representativeBlocks[‘BA’]

The class diagram is provided in xsgm-class-diagram

Class inheritance diagram for crossSectionGroupManager.

armi.physics.neutronics.crossSectionGroupManager.describeInterfaces(cs)

Function for exposing interface(s) to other code

armi.physics.neutronics.crossSectionGroupManager.getXSTypeNumberFromLabel(xsTypeLabel: str) → int

Convert a XSID label (e.g. ‘AA’) to an integer.

Useful for visualizing XS type in XTVIEW.

2-digit labels are supported when there is only one burnup group.

armi.physics.neutronics.crossSectionGroupManager.getXSTypeLabelFromNumber(xsTypeNumber: int) → int

Convert a XSID label (e.g. 65) to an XS label (e.g. ‘A’).

Useful for visualizing XS type in XTVIEW.

2-digit labels are supported when there is only one burnup group.

class armi.physics.neutronics.crossSectionGroupManager.BlockCollection(allNuclidesInProblem, validBlockTypes=None)

Bases: list

Controls which blocks are representative of a particular cross section type/BU group

This is a list with special methods.

createRepresentativeBlock()

Generate a block that best represents all blocks in group.

calcAvgNuclideTemperatures()

Calculate the average nuclide temperatures in this collection based on the blocks in the collection.

If a nuclide is in multiple components, that’s taken into consideration.

\[T = \frac{\sum{n_i v_i T_i}}{\sum{n_i v_i}}\]

where \(n_i\) is a number density, \(v_i\) is a volume, and \(T_i\) is a temperature.

See also

terrapower.physics.neutronics.mc2.mc2Writers.Mc2Writer._getAllNuclideDensities

uses these values

getWeight(block)

Get value of weighting function for this block

getCandidateBlocks()

Get blocks in this collection that are the valid representative type.

Often, peripheral non-fissile blocks (reflectors, control, shields) need cross sections but cannot produce them alone. You can approximate their cross sections by placing them in certain cross section groups. However, we do not want these blocks to be included in the spectrum calculations that produce cross sections. Therefore the subset of valid representative blocks are used to compute compositions, temperatures, etc.

Tip

The proper way to treat non-fuel blocks is to apply a leakage spectrum from fuel onto them.

class armi.physics.neutronics.crossSectionGroupManager.MedianBlockCollection(allNuclidesInProblem, validBlockTypes=None)

Bases: armi.physics.neutronics.crossSectionGroupManager.BlockCollection

Returns the median burnup block. This is a simple and often accurate approximation.

class armi.physics.neutronics.crossSectionGroupManager.AverageBlockCollection(allNuclidesInProblem, validBlockTypes=None)

Bases: armi.physics.neutronics.crossSectionGroupManager.BlockCollection

Block collection that builds a new block based on others in collection

Averages number densities, fission product yields, and fission gas removal fractions.

armi.physics.neutronics.crossSectionGroupManager.getBlockNuclideTemperatureAvgTerms(block, allNucNames)

Compute terms (numerator, denominator) of average for this block.

This volume-weights the densities by component volume fraction.

It’s important to count zero-density nuclides (i.e. ones like AM242 that are expected to build up) as trace values at the proper component temperatures.

class armi.physics.neutronics.crossSectionGroupManager.SlabComponentsAverageBlockCollection(allNuclidesInProblem, validBlockTypes=None)

Bases: armi.physics.neutronics.crossSectionGroupManager.BlockCollection

Creates a representative 1D slab block.

Notes

• Ignores lumped fission products since there is no foreseeable need for burn calculations in 1D slab geometry since it is used for low power neutronic validation.

• Checks for consistent component dimensions for all blocks in a group and then creates a new block.

• Iterates through components of all blocks and calculates component average number densities. This calculation takes the first component of each block, averages the number densities, and applies this to the number density to the representative block.

class armi.physics.neutronics.crossSectionGroupManager.FluxWeightedAverageBlockCollection(*args, **kwargs)

Bases: armi.physics.neutronics.crossSectionGroupManager.AverageBlockCollection

Flux-weighted AverageBlockCollection

class armi.physics.neutronics.crossSectionGroupManager.CrossSectionGroupManager(r, cs)

Bases: armi.interfaces.Interface

Looks at the reactor and updates burnup group information based on current burnup.

Contains a BlockCollection for each cross section group.

Notes

The representative blocks created in the CrossSectionGroupManager are ordered alphabetically by key.

Construct an interface.

The r and cs arguments are required, but may be None, where appropriate for the specific Interface implementation.

Parameters

• r (Reactor) – A reactor to attach to

• cs (Settings) – Settings object to use

Raises

RuntimeError – Interfaces derived from Interface must define their name

name: Optional[str] = 'xsGroups'

The name of the interface. This is undefined for the base class, and must be overridden by any concrete class that extends this one.

interactBOL()

Called at the Beginning-of-Life of a run, before any cycles start.

interactBOC(cycle=None)

Update representative blocks and block burnup groups.

Notes

The block list each each block collection cannot be emptied since it is used to derive nuclide temperatures.

interactEOC(cycle=None)

EOC interaction.

Clear out big dictionary of all blocks to avoid memory issues and out-of-date representers.

interactCoupled(iteration)

Update XS groups on each physics coupling iteration to get latest temperatures.

clearRepresentativeBlocks()

Clear the representative blocks.

xsTypeIsPregenerated(xsID)

Return True if the xs id is pre-generated.

createRepresentativeBlocks()

Get a representative block from each cross section ID managed here.

createRepresentativeBlocksUsingExistingBlocks(blockList, originalRepresentativeBlocks)

Create a new set of representative blocks using provided blocks.

This uses an input list of blocks and creates new representative blocks for these blocks based on the compositions and temperatures of their original representative blocks.

Notes

This is required for computing Doppler, Voided-Doppler, Temperature, and Voided-Temperature reactivity coefficients, where the composition of the representative block must remain the same, but only the temperatures within the representative blocks are to be modified.

Parameters

• blockList (list) – A list of blocks defined within the core

• originalRepresentativeBlocks (list) – A list of unperturbed representative blocks that the new representative blocks are formed from

Returns

• blockCollectionByXsGroup (dict) – Mapping between XS IDs and the new block collections

• modifiedReprBlocks (dict) – Mapping between XS IDs and the new representative blocks

Raises

ValueError – If passed list arguments are empty

getNextAvailableXsTypes(howMany=1, excludedXSTypes=None)

Return the next however many available xs types.

Parameters

• howMany (int, optional) – The number of requested xs types

• excludedXSTypes (list, optional) – A list of cross section types to exclude from using

Raises

ValueError – If there are no available XS types to be allocated

makeCrossSectionGroups()

Make cross section groups for all blocks in reactor and unrepresented blocks from blueprints.

disableBuGroupUpdates()

Turn off updating bu groups based on burnup

Useful during reactivity coefficient calculations to be consistent with ref. run.

See also

enableBuGroupUpdates

enableBuGroupUpdates()

Turn on updating bu groups based on burnup

See also

disableBuGroupUpdates

getNucTemperature(xsID, nucName)

Return the temperature (in C) of the nuclide in the group with specified xsID.

Notes

Returns None if the xsID or nucName are not in the average nuclide temperature dictionary self.avgNucTemperatures

updateNuclideTemperatures(blockCollectionByXsGroup=None)

Recompute nuclide temperatures for the block collections within the core.

Parameters

blockCollectionByXsGroup (dict, optional) – Mapping between the XS IDs in the core and the block collections. Note that providing this as an arugment will only update the average temperatures of these XS IDs/block collections and will result in other XS ID average temperatures not included to be discarded.

Notes

This method does not update any properties of the representative blocks. Temperatures are obtained from the BlockCollection class rather than the representative block.

armi.physics.neutronics.crossSectionGroupManager.blockCollectionFactory(xsSettings, allNuclidesInProblem)

Build a block collection based on user settings and input.