# Reaction rate tallies failed if \`reaction\_rate\_mode\` is flux in the operator kwargs

**URL:** <https://openmc.discourse.group/t/reaction-rate-tallies-failed-if-reaction-rate-mode-is-flux-in-the-operator-kwargs/5373>\
**Category:** User Support\
**Created:** [March 27, 2025, 10:49am UTC](https://openmc.discourse.group/t/reaction-rate-tallies-failed-if-reaction-rate-mode-is-flux-in-the-operator-kwargs/5373 "2025-03-27T10:49:49Z")\
**Posts on this page:** 1\
**Page:** 1

<div class="post-metadata">

**Author:** ![cjwiguna](https://yyz2.discourse-cdn.com/free1/user_avatar/openmc.discourse.group/cjwiguna/32/3728_2.png) [@cjwiguna](https://openmc.discourse.group/u/cjwiguna)\
**Post date:** [March 27, 2025, 10:49am UTC](https://openmc.discourse.group/t/reaction-rate-tallies-failed-if-reaction-rate-mode-is-flux-in-the-operator-kwargs/5373/1 "2025-03-27T10:49:49Z")

</div>

Greetings!

So this is my tallies:

```auto
# Reactions and productions tally in fuel
rxFuelTally = mc.Tally(name="rxFuelTally")
rxFuelTally.filters = [mc.MaterialFilter([mox])]
rxFuelTally.nuclides = sorted(['Th232', 'Cm242', 'Pu238', 'Am241', 'Pa231', 'Pu239', 'U234', 'Pu240', 'Pu242', 'U235', 'U232', 'Np235', 'Np236', 'Np237', 'Cm243', 'Np239', 'Np238', 'Am242', 'Pu241', 'Am243', 'U233'])
rxFuelTally.scores = ['total', 'absorption', 'nu-fission', 'fission',
                      '(n,gamma)', '(n,nc)', '(n,2n)', '(n,3n)', 
                      '(n,a)', '(n,d)', '(n,t)']

```

If i run depletion like this:

```auto
model.deplete(
    timesteps=time_steps,
    method=burnup_algo,
    operator_kwargs={
        'chain_file': chainLoc,
        'normalization_mode': "fission-q", # change to fission-q to compute CR (140325)
        'fission_yield_mode': 'average', # add for variation, 110325
        'reduce_chain' : True,
        'reduce_chain_level' : 5,
        # 'reaction_rate_mode' : "direct"
        'reaction_rate_mode' : "flux",
        'reaction_rate_opts' : {
            'energies' : np.logspace(-5, 7 + np.log10(2), 3000, base=10),
            'reactions' : ['fission', '(n,gamma)', '(n,2n)', '(n,3n)', '(n,4n)', '(n,a)', '(n,p)'],
            'nuclides' : sorted(['Th232', 'Cm242', 'Pu238', 'Am241', 'Pa231', 'Pu239', 'U234', 'Pu240', 'Pu242', 'U235', 'U232', 'Np235', 'Np236', 'Np237', 'Cm243', 'Np239', 'Np238', 'Am242', 'Pu241', 'Am243', 'U233']),
        }
    },
    directory=saveDir,
    power_density=powerDensity, timestep_units='d'
)

```

Many nuclides aren’t scored in tallies (right), and those nuclides are scored when `reaction_rate_mode = "direct"` (left):

 ![image](https://global.discourse-cdn.com/free1/uploads/openmc/original/2X/6/65a6c5de3e9c3eeceb5690dcbdaecd7680a43e9e.png)

I even add a miniscule amount of those isotopes like this:

```auto
for nucinc in sorted(['Th232', 'Cm242', 'Pu238', 'Am241', 'Pa231', 'Pu239', 'U234', 'Pu240', 'Pu242', 'U235', 'U232', 'Np235', 'Np236', 'Np237', 'Cm243', 'Np239', 'Np238', 'Am242', 'Pu241', 'Am243', 'U233']):
    fuel.add_nuclide(nucinc, 1e-24)

```

But it still fails to tallied those isotopes unless if i use ‘direct’ mode. Is this behavior expected?

Thanks!