Comparison of MPACT and Shift for MAGNOX Reactors

Brian J. Ade (ORNL) August, 2019 SCALE User Group Meeting

ORNL is managed by UT-Battelle, LLC for the US Department of Energy Reactor Venture Background

• This work is funded by NNSA/NA-22 (Nonproliferation R&D), typically referred to as the “Reactor Venture” • Goal is to advance the current state of the art of modeling and simulation of MAGNOX gas-cooled reactors and heavy-water moderated reactors • Activities include: – Developing MPACT to simulate neutronics for these reactors – Integrating AGREE (University of Michigan) with MPACT for thermal feedback – Performing verification of the MPACT using Monte Carlo – Performing validation using fuel samples from Calder Hall and associated operating data

2 Current Nuclear M&S

• New modeling and simulation (M&S) methods are needed for national security applications

• Enables simulations of the true operating conditions in the reactor without major assumptions

• Enables characterization of the true isotopic distribution in the reactor of interest

• Current models for these reactors are typically small and only include neutronics – Simple 1-D calculations (ORIGEN) using pre-generated libraries • Does not account for spatially-varying effects, no thermal feedback – Slightly more complex 2-D calculations without thermal feedback • Does not account for some spatially-varying effects, no thermal feedback – Full-core 3-D Monte Carlo (KENO, MCNP) • Spatially varying effects can be captured, but do not generally include thermal feedback • Number of depletion regions limited by memory (needed spatial detail is impossible) • VERY expensive to run (weeks to months simulation time)

3 Reactor Complexity

• Reactors are inherently complex • Neutron energy (spectrum) and burnup variation – Power varies throughout the reactor throughout the reactor leads to a 3-D distribution of leading to a 3-D distribution of burnup isotopics – Because the spectrum varies, two regions that • MAGNOX reactors are complex for a have the same burnup may have different variety of reasons isotopics – 5-6 fuel slugs are stacked atop one another, rather than a single pin • Very large variations in isotopics observed both axially and radially in the core, and radially within the – Graphite changes density/shape as a pin function of irradiation – Graphite temperate has a significant impact on the neutron spectrum leading to variation in the nuclides in the fuel

• Heavy water moderated reactors (HWRs) are complex for other reasons – Moderator level used for reactor control – May be refueled while on-line

4 Previous M&S in the NA-22 Reactor Venture

• For MAGNOX reactors – 1-2 axial regions per fuel slug and a single radial region → 1000-2000 depletion “regions” modeled – No thermal modeling of the temperatures in the reactor and not feedback between thermal and neutronics – Requires weeks to months of time to obtain a solution

• Simplifications required to run the model (limited depletion regions) lead to deficiencies in the prediction of the 3D distribution of isotopes

5 Full-Core Monte Carlo Depletion Difficulties

• For Monte Carlo, statistical uncertainty (�) is related to the Original calculations: number of particles (histories) simulated 95% of regions have an uncertainty of less than 1% To reduce the uncertainty by a � � ∝ factor of 2, we must increase the ��� number of particles by a factor of 4!

� � � If the number of depletion regions is ∝ = doubled, we would also need to � � ��� �×��� double the number of particles to maintain the original uncertainty 10x regions, same number of particles: • Goal calculations would include >10 axial regions per fuel slug 30% of regions have an and >10 radial regions to capture radial depletion uncertainty of less than 1% – 1000 → hundred of thousands up to millions depletion regions – Would require and increase in the number of particles by 100 maintain same level of uncertainty (100M to 10B) – 2-week simulation turns into a 4-year simulation

6 MAGNOX Moderator Temperature Effects

• Higher moderator temperatures lead to faster (harder) neutron Cs-137 spectra and pushes flux into key resonances of Pu-239 and U-235 Eu-154 • Shifts in the neutron spectrum lead to changes in the isotopes that cause fission (U-235 and Pu-239), leading to different fission products

• Changes production of certain fission products due to the differences in U-235 and Pu-239 fission yield distributions

Eu-154

Pu Quality Cs-137

7 Status and Plan

• Phase 1: Square-pitched gas-cooled reactors (MAGNOX-type) Can currently model neutronics, but without thermal feedback – Performance testing: Can the software simulate the large, detailed models we want? – Accuracy/verification testing: Does the software provide accurate solutions for the problems of interest? – Code development: Extraction of isotopics, integration of thermal solver, other issues identified in verification testing. – Validation: Comparisons with UK measured reactor operating data and isotopic data

• Phase 2: Hexagonal-pitched heavy-water reactors (NRX-type) Can currently model thermally, but limited neutronic capabilities – Plan for HWRs similar: Performance testing, verification testing, code development, validation – Main code development need is hexagonal geometry capability in MPACT and controlling reactivity through moderator level.

8 DEFENSE NUCLEAR NONPROLIFERATION R&D

MAGNOX Basics Fuel Elements to be Measured • MAGNOX = “magnesium non-oxidizing” – Uranium metal fuel “slugs” clad in MAGNOX alloy Fuel Load Channel 5-Slug ELEMENT (MWD/TE) Date Channels • MAGNOX cores are constructed using graphite(DDMMMYY) blocks with holes (“channels”) for fuel and08 -control28-Q-1 rods561

08-28-Q-3 1005 • CO2 flows through the graphite holes and around the fuel; Control fins on the fuel element enhance heat transfer. 08-28-Q-6 341 Channel 06-30-B-1 436.9 16JUL02 • Originally built with the primary purpose of plutonium Measurement production and secondary purpose06-30-B-3 of electricity782.5 16JUL02 generation Location – X 06-30-B-6 266.9 16JUL02

• Calder Hall 08-22-K-1 268.9 03JUL02

– 182 MWt, 46 MWe 08-22-K-3 480.4 03JUL02

– 1696 fuel channels, three sizes08-22-K-6 163.6 03JUL02 – Core: 9.45 m diameter, 6.4 m 03-26-B-1 31.5 tall – Fuel: 1.02 m (40”) length 03-26-B-3 56.8 2.92 cm (1.15”) diameter03-26-B-6 21.3 – Contains a number of 5- and 6- high fuel channels 2 OFFICIAL

9 Calder Hall ”Charge Pan” Layout

• Calder Hall contains three different cooling channels sizes – Larger Channels in the center of the core for increase coolant flow

• Control rods are placed in the interstitial space between the fuel elements

• Operates with control rods mostly Pitch = 20.3 cm inserted

• Control rod and fuel layout are R non-symmetric, so true full-core M Rc models will be required Zone A: Rc,A = 5.28 cm Rf Zone B: Rc,B = 5.02 cm Zone C: Rc,C = 4.58 cm Rf = 1.46 cm RM = 2.04 cm

10 Calder Hall Modeling Complexities

• As-designed, MPACT uses fuel unit cells as the basic building blocks of a model – Works very well for typical PWRs because control rods are place in fuel locations. In MAGNOX reactors, the control rods are between fuel pins

• MPACT already has the capability to model some MAGNOX reactors.

• For Calder Hall, special modeling techniques are needed due to changing fuel channel diameter

• Modifications to MPACT to allow modeling varying coolant channel sizes and control rods between fuel pins implemented, and preliminary testing has been completed – New ”cyls” capability will allow multiple fuel pins per cell, allowing specification of control elements between pins

Current method fails because Current method: break the Current method works well for the channel is too large for the unit cells into sub-regions zone C that has smaller channels smaller unit cells

Zone C Zone C Zone A 11 Full-Core Model Testing

• A number of successively larger models were simulated in order to test scaling to full-core models Fuel pin view – Largest model is a ¼-core Calder Hall “simplified” model • Approximated zone A coolant channel, all fuel channels loaded with 6 slugs. – 2544 fuel slugs (6 slugs per fuel channel, 424 total fuel channels) – 10 radial regions and 11 axial regions per fuel slug 2544*10*11 = 279840 depletion regions – 13 depletion steps to 1875 MWd/MTHM

– ~3.0 hours of runtime on 384 total cores Thermal flux distribution • Pin powers, burnup, etc, easily extracted from the Core layout simulation output using VeraView

• Sub-pin isotopic results being generated, but writing the data to a common file results in a memory bottleneck – SOLVED

12 Increasing Axial Fidelity

• Research question: What level of axial fidelity is needed to obtain accurate Full-length fuel simulations? approximation, Actual fuel slugs, Actual fuel slugs, no axial reflector with axial reflector • Underlying goal is to create an accurate no axial reflector representation of the isotopic distribution in the reactor

• Simplest assumption is that fuel is not stacked and neutrons simply leak from the system axially – Results in a very typical axial power and burnup shape that is well-predicted by basic reactor theory End plug • Simulating actual fuel slugs with end locations caps, results in a much more complex axial power shape – Additional thermal neutrons in the ends of the fuel due to surrounding graphite

• Adding as-designed axial reflectors to the model further changes the axial power shape

13 2D-slice and 3D Full-Core Convergence Issues

• Initially encountered convergence issues using when CMFD for 2D core slice calculations – CMFD solver would eventually crash

• Turning CMFD off resolved the issue, however the calculations required a significant number of outer iterations

• 2D slices required:

– 4700 iterations for 1.0E-6/1.0E-5 (keff/flux), 2.0 hours of CPU time on 54 processors

– 13000 iterations for 1.0E-6/1.0E-6 (keff/flux), 5.5 hours of CPU time on 54 processors • Solution: – CMFD relaxation – MPACT option: cmfd_dhat_relaxation = 0.5 • Relaxation on the CMFD correction factor – Same simulations converge and require 10 minutes on 54 processors – Equivalent eigenvalue and power distribution compared to the cases without CMFD

14 Radial Reflector Sensitivity on Power Distribution

• Currently difficult to model true core barrel shape in MPACT – Possible using new “cyls” method, but this would require significant effort to generate the model – Currently approximated using a “blocked” boundary – graphite blocks are included or excluded depending on their distance from the core center • Question: – What impact this boundary treatment have on results? – Without major modification, what is the best choice of reflector layout? • Solution – Generate a number of different MPACT 2D model that include increasing amount of radial reflector square blocks, compare the power distribution to the Shift reference solution

15 Radial Reflector Sensitivity Results “20” Layout Fuel Reflector Shaded area represents • Radial pin power boundary of distribution quite “true” core + sensitive to the reflector radial reflector

“0” Layout “9” Layout Reflector Reflector

-6.0% 6.2% Diff 2.1% Diff Diff

16 2D Core Slice Results • Excellent agreement between MPACT and Shift pin power distributions 1.7 • Large eigenvalue differences (51-group library) Shift keff: 1.0493 MPACT keff: 1.0332 (1610 pcm) • Eigenvalue differences under investigation 1.0 – Modeling error? – XS group structure? – Other XS issue? – Temperature treatment?

0.3

17 3D BOL Pin Power Distribution

XY-66 2.3

XY-56

XY-55

XY-45

XY-28 XY-55 XY-56 XY-66 1.0

XY-5

0.0 XY-5 XY-28 XY-45 18 3D BOL Pin Power Distribution vs. Shift

XY-66 5.4

XY-56

XY-55

XY-45 2.0

XY-28 XY-55 XY-56 XY-66 0.0

-2.0 XY-5

-5.4 XY-5 XY-28 XY-45 19 3D BOL Pin Power Distribution Summary

• keff and pin power RMS summary 52G 47G keff,diff 1400 pcm 260 pcm Radial: 0.69 % 1.00% Axial: 1.04 % 1.41% 3D: 1.32 % 1.86%

• Axial ends produce 2.0 largest bias from Shift

• Radially, channels that have 5 instead of 6 fuel elements 1.5 produce larger biases

• Overall, very positive initial power 1.0 distribution results 0.8

20 Initial Depletion Calculations

• No control rod insertion or movement

• No thermal feedback

• ~250,000 depletion regions

• 13 statepoint calculations

• ~13 hours on 384 processors

VeraVeiw Depletion Animation 21 Near-Term Activities and Upcoming Challenges

• Addition of control rods and their movement vs irradiation • Modeling the actual cycle – Initial burnup distribution in a “jump-in” type calculation – True control rod layout and movement (non-symmetric) – Modeling true graphite density, thermal conductivity, etc. • Development of a XS library for gas-cooled reactors • Integration of AGREE thermal solver and benchmarking using operating data • Comparison to measured data (temperatures, fuel isotopics when they become available) • Heavy-water moderated reactor activities

22 Questions?

“There's no sense in being precise when you don't even know what you're talking about.”

―John von Neumann

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