Research and Test Reactor Conversion...

Objectives and Constraints for Research Reactor

Conversion Design: Assessing Alternatives

24 June 2015 John G. Stevens, Ph.D. Manager of Research and Test Reactor Department, Nuclear Engineering Division Argonne National Laboratory International Reactor Conversion Technical Lead for M3

National Academy of Sciences – Status and Progress on Eliminating HEU Use in Fuel for Civilian Research and Test Reactors

USHPRR Reactor Conversion

Outline

Principle of Fuel Acceptability for Conversion

Objectives and Constraints of Multivariate Research Reactor Conversion

Alternatives Assessment Continuum

– Screening (“Rules of Thumb” and other Engineering Judgement)

– Scoping Studies

– Feasibility Studies

– Operational & Safety Analyses

– Redesigns to Address New or Altered Constraints

Down-Selection Process of UMo Monolithic for USHPRRs as Set of Reactors

ATR Conversion Analysis Status as Example

2

PRINCIPLE OF FUEL ACCEPTABILITY FOR CONVERSION

QUALIFIED Fuel Assembly

– Fuel assembly that has been successfully irradiation tested and is licensable from the point of view of fuel irradiation behavior

COMMERCIALLY AVAILABLE Fuel Assembly

– Fuel assembly that is available from a commercial manufacturer

SUITABLE Fuel Assembly

– Safety criteria are satisfied

– Fuel Service Lifetime comparable to current HEU fuel (e.g., Number of FA used per year is the same as or less than with HEU fuel)

– Performance of experiments is not significantly lower than with HEU fuel

To be ACCEPTABLE for LEU conversion of a specific reactor, a fuel assembly must be qualified, commercially available, and suitable for use in that reactor, then reactor operator & regulator must agree to ACCEPT fuel assembly for conversion

3

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The Multivariate Conversion Design Challenge Yacout compared EUHFR vs. USHPRR Operating Envelope

Note that BR2 and RHF designs for LEU assemblies are well-defined and vetted by collaboration between reactor operators and ANL. The

CEA and ANL have only recently begun sharing model detail for LEU designs of ORPHEE, and have not yet shared models of JHR.

Figure and footnote from Yacout, “MMM’s non-US fuel development efforts including HERACLES”, NAS Review 5th Meeting, 16 April 2015

• Fuel performance envelope plots indicate relative materials challenge of fuel use in the different reactors, and thus whether a fuel usable in one reactor should be able to meet the fuel integrity safety criteria in another reactor.

• But these envelopes do not describe reactivity, boiling margins, mission metrics, etc. • Reactor operation within fuel performance envelope is necessary, but not sufficient

Objectives & Constraints of Research Reactor Conversion

(1 of 4) Reactivity

– Reactivity clearly depends on much more than U-235 mass

• Parasitic absorption and self-shielding

• Moderator and Reflector efficiency

• Control system absorption efficiency and control rod/blade depletion

– Core lifetime for once-through or lifetime cores

– Cycle Lengths and Fuel Element Lifetime(s) for reasonable fuel utilization schedules for cores with fuel shuffling

– Critical states and associated power/flux shapes to preserve both safety margin and mission performance

Mission Performance Metrics

Safety Criteria, including but not limited to Fuel Performance Envelope

Minimization of Changes to Reactor System beyond fuel elements and core

Fabricability of Fuel Conforming to Specifications

Fuel Cycle Availability 5


(2 of 4) Reactivity


– High performance reactors have been tuned for near-optimal mission metrics

• Localized neutron and gamma fluxes at experiments, specific energy spectra

• Isotope production rates and resulting periodic throughput

– Power increase is one effective way to compensate for performance penalty, but impacts requirements for reactivity (e.g., shorter fuel lifetime at higher power) and safety margins

– Core management of fuel and reflectors can tune flux shapes and spectra, but impact reactivity and safety margins




Fuel Cycle Availability 6


(3 of 4) Reactivity



– Local power peaking low enough to assure margins to fuel damage

• Fuel performance envelope of fission rate, and fission density (burnup)

– Margins to boiling crises depend on details of flow channel geometries, power density, and larger system (pressure, flow, and inlet temperature ranges; relationship of element flow to bypass flows)

• Fuel swelling and Fluid-Structure-Interaction (FSI) alter flow geometry over time

– Transient and Accident safety margins depend on reactivity coefficients, fuel utilization (e.g., blister temperature decreases with fuel burnup), and larger system (hydraulics and potential release paths)

– Fuel handling safety depends on masses, geometries, materials



Fuel Cycle Availability

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(4 of 4) Reactivity




– Reflector modifications within or surrounding core effective, may perturb experiments

– Control system (rods/blades, timing, activation logic)

– Changes to experimental infrastructure (cold sources, neutron guides, in-core-devices)


– Reliability

– Cost

Fuel Cycle Availability

– Fresh: Mandate for M3 program is LEU (i.e., enrichment < 20%)

• NRC Domestic Licensing basis (10 CFR 50.64)

• Schumer Amendment of Chapter 11 of the Atomic Energy Act of 1954 (42 U.S.C. 2151 et seq.)

– Spent: Identification of Disposition Path part of selection of fuels to be deployed 8

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MITR, 22-27 Elements MURR, 8 Elements NBSR, 30 Elements ATR, 40 Elements HFIR, Single Element

(540 plates)

The Multivariate Conversion Design Variables

Fuel System: Uranium Density, Fuel Performance Envelope, Integral Burnable Absorber, Integral Moderator

Fuel Element Geometry: Uranium, Moderator, and Burnable Absorber distributions to adapt to self-shielding, but with impacts on experiments and safety margins

Reflector(s): Balance neutron economy with flux shape impacts on experiments and safety margins

Operating Power: Flux directly proportional to power, but margins inversely proportional

Fuel Utilization: Cycle Lengths and Fuel Shuffling Scheme designed to balance fuel consumption against experimental/isotope production and throughput

Changes beyond Fuel and Core: Control System, Experimental Infrastructure and/or Balance of Plant

Self-Shielding Illustrated by Plate Powers

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HEU CR @ 13" Power & Heat Flux Peaking Factors

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Peakin

g F

acto

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LEU CR @ 13" Power Peaking Factors

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Plate powers illustrate local thermal flux available to U

Note much wider spread in plate power when LEU replaced HEU without geometry change

High density LEU consumes thermal neutrons at outer plates

This also implies diminishing return of adding U mass in low-thermal-flux regions

Non-uniform fuel thickness allows design to exploit behavior

MURR HEU: All Meat Same Thickness MURR LEU : All Meat Same Thickness

LEU CR @ 13" Heat Flux Peaking Factors

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MURR LEU : Four Distinct Meat Thicknesses

Graphics from McKibben, J. C., et al., “Feasibility Analyses for HEU to

LEU Fuel Conversion of the University of Missouri Research Reactor (MURR),” MURR Technical Data Report No. 0125, September 2009..


(1 of 4)

Screening (“Rules of Thumb” and other Engineering Judgement)

– U-235 mass in LEU reactor must exceed U-235 mass in HEU reactor

– For fixed or constrained core and element geometries, many options can be excluded by simple consideration of total volumes, thus associated masses and power density

– Generally focused on reactivity, perhaps on estimated fuel operating envelope

Scoping Studies

– Neutronics models developed for representative core(s) to allow formal, if often simplified, comparison of alternatives (fuel, geometry, utilization)

– Experimental performance generally evaluated by proxy (thermal flux) or simplified metrics

– Safety margins generally treated by simplified analyses such as hot-stripe thermal hydraulics for characteristic flow channels

Feasibility Studies

Operational & Safety Analyses

Redesigns to Address New or Altered Constraints

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(2 of 4)


Scoping Studies

Feasibility Studies

– Neutronics models formalized for representative core(s), experiments, and representative fuel utilization schemes to allow formal comparison of alternatives (full set of design variables)

– Experimental performance evaluated at a level of detail appropriate for differentiation of alternatives and determination of adequacy of proposed designs

– Safety margins generally treated with detail for steady-state thermal hydraulics including impacts of tolerances and uncertainties. Reactivity coefficients and most-likely-limiting transient generally evaluated, but not full suite of transients and accidents

– Significant Model Verification and Validation for reactor specific case via HEU operating data and other experiments (e.g., Croft-and-Waters tests for boiling margin)

– Key output of Feasibility Study is one or few proposed design(s) of suitable LEU element



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(3 of 4)


Scoping Studies

Feasibility Studies


– For the fuel element proposed by Feasibility Study

– Neutronics models extended as necessary for representative core(s), experiments, and representative fuel utilization schemes to allow formal comparison of alternatives (full set of design variables). Transition schemes from HEU to LEU operations evaluated.

– Experimental performance evaluations refined to allow optimization

– Full Safety Basis Developed including impacts of tolerances and uncertainties and full suite of transients and accidents. Per NUREG-1537, all SAR sections impacted by fuel change are evaluated.

– Significant Model Verification and Validation for reactor specific case continues

– Output of suite of studies and documents is an approved license for LEU use


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(4 of 4)


Scoping Studies

Feasibility Studies



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Redesign to adapt to

new or altered

constraints

Screening

Scoping

Feasibility Study

Operation & Safety Analyses

USHPRR Down-selection to Monolithic Path

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• Scoping and feasibility studies were performed by USHPRR Working Group (reactors and Argonne) between 2005 and 2009 in parallel with fuel system development

• Per Meyer February 2015 NAS Review presentation: “First ‘real’ data on monolithic fuel acquired in 2006”

• At May 2007 USHPRR Working Group meeting, all five reactor organizations declared their preference for the monolithic fuel form based on multi-variate considerations (reactivity, ability to tune power peaking, mission performance)

Image of slide from facilitated discussion of “Overview of Conversion Feasibility Study Open Issues and Timelines” by Stevens, 23 May 2007

Screening: Reactivity “Rule of Thumb” U Density

Requirements for USHPRR from Matos’ 1996 Paper

Estimated LEU density= HEU density * (HEU enrichment/LEU enrichment) * (Factor for Parasitic Absorption and Self-Shielding)

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Table from Matos, “LEU Conversion Status of U.S. Research Reactors September 1996”, 1996 International RERTR Meeting, Seoul, Korea, 6-11 October 1996.

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Scoping: MCNP Reactivity for MURR Use of UMo

Image of slide from “University of Missouri Research Reactor” presentation at first USHPRR Working Group Meeting, February 2006. Arrows and highlight rectangles added by Stevens herein. Also see McKibben, Kutikkad, and Foyto, “Current Status of the Missouri University Research Reactor HEU to LEU Conversion Feasibility Study,” 2006 International RERTR Meeting, Cape Town, Republic of South Africa, Oct. 29-Nov. 2, 2006.

• MCNP model was used for direct material replacements, showing that dispersion fuels had no excess reactivity once Xe and Sm burned in (scoping did not calculate fission product for each case, but rather used a representative value)

• Enrichments above 20% were identified as necessary for monolithic fuels to match HEU reactivity unless geometry changes were made to the fuel

Scoping: Early MITR Studies for Experimental

Performance with UMo

“A number of options for an LEU fueled MIT reactor have been studied, including promising designs using mixed H2O/D2O moderator or using an inner beryllium reflector. The mixed moderator design, although promising from a flux point of view, has limitations with reactivity and operational concerns. The inner beryllium reflector design appears to be more feasible and can, under some circumstances, deliver fluxes needed to experimental facilities. Issues with power peaking, both near the reflector and in a possible fueled annulus, will need to be solved to assure adequate cooling in all areas.

Although it is possible to design cores with higher reactivity, greater ex-core thermal flux, and greater in-core facility fast flux, no design has yet been identified to accomplish all three simultaneously. It is evident that sufficient operational flexibility must be designed into an LEU MIT reactor so that flux enhancement options for immediate experimental needs can be easily implemented. It only by safely meeting the needs of the scientific community that conversion of all research reactors to LEU can be fully accepted.”

18 Quote from Newton, Pilat, and Kazimi, “Flux Enhancement Options for an LEU-Fueled MIT Reactor”, 2004 International RERTR Meeting, Vienna, Austria, 7-12 November 2004. Highlights added by Stevens herein

Scoping: HFIR Comparison of U7Mo Dispersion and

U10Mo Monolithic Conversion Potentials

19 Quote from Primm et al, “Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2007,” ORNL/TM-2007/45, November 2007. Highlights added by Stevens herein

Feasibility Study: Initial ATR Conversion Studies

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ATR HEU fuel assembly – 19 curved plates, each with 20 mil thick fuel meat zone – 45 degree arc – Integral burnable absorber used to control power peaking within

inner 4 and outer 4 plates of HEU – Distinct UAlx loading also used to control HEU plate power distribution

(~1.1 gU/cc in 1,2,18,19; ~1.3 gU/cc in 3,4,16,17; ~1.6 gU/cc in 5-15) – 48 inch fuel length (others <24”) – No grid flexibility (i.e., always 40 elements loaded)

Studies culminating in 2009 Feasibility Determination considered LEU with:

– U7Mo Dispersion and U10Mo Monolithic

– Power Peaking controlled by:

• Distinct dispersion loading between differing plates (as in HEU)

• Burnable absorber in outer plates (as in HEU), both B4C and Cd evaluated (but in abstract, i.e., without fabrication technique identified)

• Distinct enrichments in different plates

• Distinct foil thicknesses in different plates (as for MURR LEU, based upon concept of distinct UAlx densities in ATR HEU)

– Several performance metrics evaluated, but not for representative cores

2009 Selected LEU design: U10Mo monolithic with distinct foil thicknesses and burnable absorber in outer plates

Redesign to Address New or Altered Constraints:

ATR Pre-Conceptual Design of Base Fuel Conversion

As Described in Pope, Dehart, Nigg, Jamison, and Morrell, “Fuel Element Design and Analysis for Advanced Test Reactor Conversion to Low Enriched Uranium Fuel,” RRFM 2014, Ljubljana, Slovenia, 30 March - 3 April 2014

Original ATR Feasibility Study design not considered acceptable in 2012 due to:

– Gas production by integral burnable absorber, which would reduce blister threshold

– No fabrication scheme considered viable for extra layer of absorber in plate

Emphasis on Experimental Performance Requirements for Representative Cores

– NR 1. Operational cycle length of 56 days at 120 MW

– NR 2. Fast to thermal neutron flux ratio within 5% of current values in a pressurized water loop test

– NR 3. Greater than 4.8 × 1014 fissions per second per gram 235U in a specimen with l gram 235U per linear inch in a standard in pile tube (SE or SW) operating at 60 MW

– NR 4. 3/1 lobe power split with south corner lobes operating at three times the lobe power of the northern lobes.

– NR 5. Gamma to neutron flux ratio within +10%/ 0% of current values.

“Nine concepts for an LEU fuel design were proposed with multiple variations of each”

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Quote from Pope, Dehart, Nigg, Jamison, and Morrell, “Fuel Element Design and Analysis for Advanced Test Reactor Conversion to Low Enriched Uranium Fuel,” RRFM 2014, Ljubljana, Slovenia, 30 March - 3 April 2014. Brackets and emphasis added by Stevens herein.

“The two ELF [Enhanced LEU Fuel] fuel element designs, ELF Mk 1A and ELF Mk 1B, have been identified as leading candidates for conceptual design for an LEU element for ATR. This is the result of a traditional engineering design process involving many design variations. The field was narrowed based on reactor physics evaluations and thermal-hydraulics (TH) analysis against Naval Reactor performance requirements and nuclear safety criteria. From the initial pool of candidates, three were selected for a detailed trade study, ISBA-Cd [Integral Side-plate Burnable Absorber], ISBA-B, and ELF. All three of these use [U10Mo monolithic] fuel meat thickness variation to achieve radial power flattening. From the neutronic and TH analyses performed in the trade study, no significant discriminators were found causing any of these three concepts to be discarded. Ultimately, the fabricability and fuels performance questions associated with elements containing burnable absorbers lead to the adoption of ELF, a fuel element containing no poisons.”

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From Pope, Dehart, Nigg, Jamison, and Morrell, “Fuel Element Design and Analysis for Advanced Test Reactor Conversion to Low Enriched Uranium Fuel,” RRFM 2014, Ljubljana, Slovenia, 30 March - 3 April 2014

23

USHPRR: Fuel Challenges and Recent Progress

Reactor LEU Fuel Needed

LEU U-Mo Monolithic Fuel Element

Design Status

MITR

Base Monolithic

HEU: 15 finned plates; complex fuel loading

LEU: 19 plates ≥12 mil cladding; no fins

Completing Accident Analyses

Unfueled region leads to high

local burnup; Thin 8.5 mil fuel

PSAR submitted to NRC NBSR

Fewer thinner plates with thinner clad in LEU with respect to HEU

Drafting PSAR MURR

ATR/ ATRC

Fuel foil length 48 inch thickness 8 – 18 mil

Pre-conceptual design completed

HFIR Complex

Monolithic

Contoured LEU fuel: Radial/axial shape to be optimized

Absorbers: Boron in side plates; others TBD

Complex fuel options under consideration

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USHPRR & EUHFR Alternatives Assessment Tracking

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Ranking of each fuel system for each reactor is a sensitive matter since

combination of evaluations by M3, the reactor operator, and other parties are all

considered as part of negotiations for collaborative paths toward conversion.

U-Mo

Monolithic

Less than 20%

HERACLES U-

Mo

Dispersion-

Coated,

Less than 20%

KOREAN U-

Mo

Dispersion-

Other,

Less than 20%

U3Si2-4.8g/cc,

at 27%

U3Si2-4.8g/cc,

Less than 20%TRIGA - 30/20

U3Si2-6.5g/cc,

~20%TRIGA - 45/20

Nothing

(i.e., remain

HEU)

Fuel Previously Qualified or Fuel Under QualificationPotential or

Unplanned Qualification

1 ExistsScale

Uncertain

Steps

Uncertain

Potential

Showstopper

Unknown or

Rejected

2 Yes Probable Uncertain Improbable No

3 ExistsBeing

CommercializedIdentified Unknown Showstopper

4Good

(Evaluated)

Expected Good

(need Evals)

Uncertain

(Ambiguous)

Expected Poor

(need Evals)Showstopper

5Good

(Evaluated)

Expected Good

(need Evals)

Uncertain

(Ambiguous)

Expected Poor

(need Evals)No

6 Yes No

7 NNSA Supports Uncertain Uncertain Uncertain No

Will this fuel support HEU minimization?

Is an HEU supply readily available to support operation until fuel is

available?

A path to qualify and manufacture the fuel

(for required conditions and scale)

Is the projected facility lifetime complementary to the possible fuel

availability schedule?

A disposition path

Reasonable loss of performance for identified mission?

Will this fuel support operation with reasonable cycle length and fuel

consumption?

Conclusions

Reactor Conversion is a Multivariate Nonlinear Design Challenge

Alternative Designs Assessed Across Continuum of Depth

The M3 Reactor Conversion Program does revisit constraints and alternatives regularly

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Redesign to adapt to

new or altered

constraints

Screening

Scoping

Feasibility Study

Operation & Safety Analyses

Thank you for your attention

27

Acknowledgements

Many thanks to the team efforts of staff at NBSR, MURR, MIT, BNL, ORNL, INL, SCK, ILL, CEA, Areva-CERCA, TUM, and ANL

Supporting Material

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Specific Discussion of Challenges for

Conversion of the USHPRR

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MITR

MITR fuel assembly

– Rhomboid

– Finned Clad on each fuel plate surface to increase heat transfer

Flexible Number

– Fuel Elements and

– Dummy/Experiment Assemblies

Number of plates increased from 15 HEU to 19 LEU to preserve cycle length & improve thermal margin

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Power Increase 6 MW HEU to 7 MW LEU to preserve operational flexibility

Mitigations will not require significant system modification

Pre-Mitigated Penalty 5-10%

No Post-Mitigated Penalty

MURR Very Compact Core Design

– Core Volume 33 liters – Fuel Meat 4.3 liters

MURR fuel assembly – 24 curved plates – 45 degree arc – No grid flexibility

Weekly refueling for > 90% capacity factor for > 20 years

Weekly cycle and initial control blade position key to efficient isotope production

LEU fuel assembly:

– 23 plates (for moderation)

– Plate thicknesses reduced

– Variable fuel meat thickness for power peaking control

– Thinner clad for better moderation (fuel utilization)

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Power Increase 10 MW HEU to 12 MW LEU to preserve Production Rates

NBSR

NBSR fuel assembly

– Unfueled region at core axial centerline provides flux peak for experiments

– 34 slightly curved plates per assembly

No grid flexibility, Compact LEU core design would not preserve sufficient range and flexibility of experiments

Pre-Mitigated Penalty ~10%


– Thermal loss to be overcome by improved instruments

– Cold losses completely overcome by upgrade of cold source – potential for small gains in cold neutron performance

Key Challenge will be timing i.e., planning and execution have little margin for problems or adjustments

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ATR

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ATR fuel assembly – 19 curved plates – 45 degree arc – No grid flexibility – Integral burnable absorber used to

control power peaking within inner and outer plates of HEU

– 48 Inch plate length (others <24”)

5 Lobes of reactor operated at distinct powers, so operational flexibility a key

Detailed performance penalties still being determined, but preliminary estimates are 5-10%

Burnable Absorber was key challenge, but was designed out FY13 by selecting distinct foil thickness in plates. FY13 Conceptual Design now in Preliminary Design rigor.

HFIR

HFIR fuel assembly

– Involute plates maintain constant water gap between concentric cylinders

– Graded fuel meat within plates to control power peaking at radial edges of fuel

– Burnable Absorber in inner element “filler” of HEU to further control radial peaking

Single-Use Core

Original Power Rating 100 MW

– was reduced to 85 MW due to vessel pressure concern coupled with 60s safety basis

– Plan to return to 100 MW with LEU via modern safety basis

Pre-Mitigated Penalty 15% BOC


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HFIR Key Challenges

Onset of Nucleate Boiling (ONB) at Core Exit is the active thermal constraint

Definitive Complex Fuel – Radial grading of fuel must be maintained for LEU

– Burnable absorber apparently still necessary in inner element

– Axial grading of last several cm at exit apparently necessary to avoid ONB

– Key current design efforts aim to:

1. Get rid of axial “toe” in earlier LEU design to keep grading one dimensional

2. Get absorber out of plate and into structure, like BR2 has done

35

Power Increase from 85 MW HEU to 100 MW LEU presents additional challenges

– Computational Fluid Dynamics (CFD) Safety Basis will be required to show that sufficient thermal margin exists at current system pressure

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