1 Overview of ACR Reactor Physics Toolset Qualification Program

by Hank ChowSenior Reactor Physicist,

Reactor Core Physics Branch, AECL

Presented to US Nuclear Regulatory CommissionChalk River Laboratories

2004 July 27-29

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Presentation Objectives

• Provide an overview of the physics code qualification plan designed to demonstrate the adequacy of the physics analysis methodology for ACR applications

• Provide the basis and context for subsequent detailed topical presentations in these three days

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Presentation Outline

• IST Code validation methodology• Physics code suite validation process• Code assessment / development / enhancement• Validation Matrix document• Qualification Plan• Validation database • Data from other criticality facilities• Coverage of phenomena• Summary

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Physics Toolset

AECL standard reactor physics toolset:• WIMS-IST - lattice-cell code • DRAGON-IST - super-cell code • RFSP-IST – 2-group finite-core simulation codeSupplementary tools used: • MCNP – Monte Carlo neutron transport code• DONJON – multi-group diffusion-theory codePhysics codes used by regulators and consultants:HELIOS / PARCS / NESTLE / MCNP

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IST Codes Validation Methodology

• Technical Basis Document− safety concerns, accident scenarios vs governing phenomena − PKPIRT (Phenomena Key Parameters Importance Rank Table)

• Validation Matrix− Phenomena (PH) vs accidents, also relative importance of PH− PH vs measurement data

• Validation Plan• Validation Analysis Tasks• Validation Report

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Physics Code Suite Assessment And Validation Process

Code Validation

Manual

Validation Matrix

Document

DCD

Design Analysis

Safety Analysis

Technical Basis

Document

Benchmark –other data sources

Benchmark -MCNP

Benchmark –ACR Measurements

Benchmark –ZED-2 data

Code Qualification

Plan Document

Code Validation Analysis

Code Assessments

Code Developments

ZED-2 Experimental

Program

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Code Assessments

• Codes must be capable of modelling the particular nuclear characteristics of the ACR design− assessed over entire range of conditions in postulated

transients• Assessment reports prepared for WIMS, DRAGON and

RFSP• Need for code enhancements identified

− particularly for effects of large heterogeneity (such as in checkerboard voiding)

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Code / Model Development

• WIMS (details by D. Altiparmakov – day 2)− ACR specific data libraries− use MCNP results as benchmark− optimal spatial mesh discretization− improved resonance reaction calculations − enhanced geometry capabilities

• RFSP (details by H. Chow / B. Rouben – day 2)− accuracy of two-group representation− spatial mesh subdivision in the core model− WIMS grid-based method in kinetics calculations− micro-depletion method in static core simulations− discontinuity factors − multi-group finite-difference and nodal methods

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Validation Matrix Document

• ACR Physics VM prepared• Discipline specific; not code-specific• Phenomena identification and importance ranking• Validation database requirements specified

− sources of data− cover core transient response for the full range of anticipated

conditions

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Physics Phenomena and Codes Capturing the Phenomena

PH No. Reactor Physics Phenomenon Primary Code(s) PH1 Coolant-Density-Change Induced Reactivity WIMS-AECLPH2 Coolant-Temperature-Change Induced Reactivity WIMS-AECLPH3 Moderator-Density-Change Induced Reactivity WIMS-AECLPH4 Moderator-Temperature-Change Induced Reactivity WIMS-AECLPH5 Moderator-Poison-Concentration-Change Induced Reactivity WIMS-AECLPH6 Moderator-Purity-Change Induced Reactivity WIMS-AECLPH7 Fuel-Temperature-Change Induced Reactivity WIMS-AECLPH8 Fuel-Isotopic-Composition -Change Induced Reactivity WIMS-AECL/RFSP-ISTPH9 Refuelling-Induced Reactivity RFSP-ISTPH10 Fuel-String-Relocation Induced Reactivity RFSP-ISTPH11 Device-Movement Induced Reactivity DRAGON/RFSP-ISTPH12 Prompt/Delayed Neutron Kinetics RFSP-ISTPH13 Flux-Detector Response RFSP-ISTPH14 Flux and Power Distribution in Space and Time WIMS-AECL/RFSP-ISTPH15 Lattice-Geometry-Distortion Reactivity Effects WIMS-AECL/RFSP-ISTPH16 Coolant-Purity-Change Induced Reactivity WIMS-AECL

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Relevance of Physics PH in Accident Events

Small LOCA, Single Channel Accident

Secondary Coolant Failure

LOR

Moderator System

ID No.

PHENOMENA

Large LOCA

Small Out-of-

Core LOCA

PT

Failure

Off-Stag.

Feeder Break

End-

Fitting Failure

Loss of Feed-water

Steam Line break

Fuel Handling Accidents

Fast

Slow

Loss of

Flow

Loss of Mod

Inventory

Loss of

Mod. Heat Sink

Limited Core Damage

Accidents

PH1 Coolant-Density-Change Induced Reactivity � � � � � � � � � � �

PH2 Coolant-Temperature-Change Induced Reactivity � � � � �

PH3 Moderator-Density-Change Induced Reactivity � � � � �

PH4 Moderator-Temperature-Change Induced Reactivity � � � � �

PH5 Moderator-Poison-Concentration-Change Induced Reactivity

� � � � � �

PH6 Moderator-Purity-Change Induced Reactivity � � �

PH7 Fuel-Temperature-Change Induced Reactivity � � � � � � � �

PH8 Fuel-Isotopic-Composition-Change Induced Reactivity � � � � � � � � � � � �

PH9 Refuelling-Induced Reactivity �

PH11 Device-Movement Induced Reactivity � � � � � � � � � � � � � �

PH12 Prompt/Delayed Neutron Kinetics � � � � � � �

PH13 Flux-Detector Response � � � � � � � � � � � PH14 Flux and Power

Distribution (Prompt/Decay Heat) in Space and Time

� � � � � � � � � � � � � �

PH15 Lattice-Geometry-Distortion Reactivity Effects

� � � �

PH17 Moderator Level Change Induced Reactivity � � �

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Sample Importance Ranking (In-Core LOCA Event)P h a s e P h e n o m e n o n S ig n if . C o m m e n t

In i t ia l c o n d i t io n s P H 1 4 (p o w e r d is t r ib u t io n )

S

P re - s h u td o w n p h a s e P H 0 4 (m o d e ra to r t e m p e ra tu r e r e a c t iv i ty )

P M o d te m p r i s e s d u e to m ix in g w i th c o o la n t

P H 0 5 (m o d e ra to r p o i s o n r e a c t iv i t y )

P M o d p o is o n c o n c e n tr a t io n i s d i lu te d b y m ix in g w i th c o o la n t

P H 0 6 (m o d e ra to r p u r i ty r e a c t iv i t y )

P M o d p u r i t y d e c re a s e s d u e to m ix in g w i th c o o la n t

P H 1 1 (d e v ic e s ) P R R S r e s p o n s e - z o n e d ra i n in g , M C A in s e r t io n

P H 1 4 (p o w e r d is t ) P C o re p o w e r d is t r ib u t io n d u e to R R S (M C A in s e r t io n )

P H 0 1 ( c o o l . d e n s i t y ) S V o id in g in a s in g le c h a n n e l P o s t - s h u td o w n p h a s e P H 0 4 (m o d te m p

r e a c t iv i t y ) P M o d te m p r i s e s d u e to m ix in g w i th

c o o la n t P H 0 5 (m o d p o i s o n

r e a c t iv i t y ) P M o d p o is o n c o n c e n tr a t io n i s d i lu te d b y

m ix in g w i th c o o la n t P H 0 6 (m o d p u r i t y

r e a c t iv i t y ) P M o d p u r i t y d e c re a s e s d u e to m ix in g

w i th c o o la n t P H 1 1 (d e v ic e s ) P S O R a n d L IS S w o r th P H 1 4 (d e c a y h e a t ) P P H 0 3 (m o d d e n s i t y

r e a c t iv i t y ) S M o d d e n s i t y w i l l d e c r e a s e o n b o i l in g

P H 0 7 ( fu e l te m p ) S F u e l te m p fe e d b a c k h a s e f f e c t o n S D S 1 d e p th

P H 0 8 (x e n o n / f u e l i s o to p ic )

S X e n o n e f f e c t s a n d e f f e c t o f fu e l i s o to p ic o n S D S 1 d e p th

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PH Validation Considerations• Separation of PHs

− simultaneous occurrence of PHs inherently related (e.g. moderator temperature and density)

− simultaneous occurrence in experimental set-up or accident transients (e.g. coolant and fuel temperature changes)

− a few PHs likely to occur at the same phase of an accident event(e.g. moderator heat-up and poison dilution in an in-core LOCA)

• Requirement: Biases and Uncertainties − for a frozen code version and data library− input data used to model the real physical situation− inaccuracies in basic nuclear data− code numeric, and approximations in solution method and model− within the bounds of measurement accuracy

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Code Qualification Plan

• Qualification Plan Document Rev. 1 (2004 Feb) issued• Periodically updated (e.g. next rev. will include plans to

address checkerboard void)• Code specific document• Integrate all activities related to qualification of the

three major physics codes − requirements from design and safety applications − conclusions from Code Assessment Reports− information from VM document− measurement database− schedule

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Validation Database

• ZED-2 Measurements− using existing fuel in ACR-like lattice and conditions

• 28-el natural uranium buckling measurement, flux map and substitution

• 37-el and 43-el LVRF substitution and fine flux measurements − SEU fuel lattices and ACR test bundles

• NRU Dy irradiation data• Data from other criticality facilities

− examples: DCA, ECO measurements• “Surrogate” data generated using MCNP• Commissioning physics test data (supplementary)

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Coverage of PHs by ZED-2 Measurements• Coolant-void reactivity (fresh and “irradiated” fuel) • Combined coolant- and fuel-temperature reactivity (fresh and

“irradiated" fuel)• Combined moderator-temperature and -density reactivity, cooled and

voided lattice (fresh fuel)• Fine flux in cooled and voided lattices (fresh fuel)• Moderator-poison reactivity for cooled and voided lattices (fresh fuel) • ZED-2 global flux shape with heterogeneous fuel compositions• Absorber devices reactivity effect and flux perturbations• Contributions from the delayed photo-neutrons to the total delayed

fraction

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Dy Modelling and Dy burnout

• Modelling of Dy burnout via WIMS burnup calculations (Part of PH08 fuel composition change induced reactivity)

• Existing ZED-2 measurement data (fresh fuel)− 37-element bundle, 7-rod substitution measurements− 43-element bundle, 7-rod substitution measurements

• Existing data for Dy depletion− PIE of fuel elements with Dy irradiated in NRU (La content, Dy

content and fractions of Dy isotopes)• Details to be described in presentation by R. Jones

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Inter-code Comparisons against MCNP

• An integral component of the validation program -complementary to comparisons to measurement data

• Cover the ranges of parameters overlapping with available measurement data - verify how MCNP performs vs measurements

• Extend to cover power reactor conditions and postulated accident conditions where no measurement data exist

• Can include cross-PH effects• Full-core flux/power distribution data

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Validation of WIMS/DRAGON/RFSP Full Core Calculations – Test Facility Measurements and Inter-Code Comparisons

• ZED-2 measurements (and data from other test facilities)− full core modeling of ZED-2 by WIMS/RFSP− lattice properties of background lattice and test fuel bundles generated by

WIMS− measured critical moderator height used in the model− δδδδk of the “before” and “after” core state represents code-error in

simulation of the perturbation• MCNP vs DRAGON and MCNP vs RFSP

− 2-D super-cell models, 2x2, 3x3 array of channels – checkerboard void − 2-D slice of reactor core models - comparison of radial flux shape changes

upon coolant voiding− 3-D full core model

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Validation of PHs by Measurements from ZED-2, DCA, ECO, NRU and Inter-Code comparisons

Phenomenon SEU fuel ZED-2 measurements+

Existing data from Other Facilities Code-to-code validation

PH01 Coolant Density ZED-2 DCA, ECO MCNP (Cell) PH02 Coolant Temp ZED-2 DCA MCNP (Cell) PH03, PH04 Mod Density/Temp

ZED-2 DCA MCNP (Cell)

PH05 Mod Poison ZED-2 DCA MCNP (Cell) PH06 Mod Purity ZED-2 MCNP (Cell) PH07 Fuel Temp ZED-2 DCA MCNP (Cell) PH08 Fuel Isotopic ZED-2 NRU MCNP (Cell) PH09 Refuelling MCNP (Core) PH10 (NA) PH11 Device Movement ZED-2 MCNP (Cell, Super-cell)PH12 Prompt/Delayed N ZED-2 PH13 Flux Detector ZED-2 PH14 Flux/Power Dist. ZED-2 DCA, ECO MCNP (Cell, Core) PH15 Lattice Geometry MCNP (Cell, Core) PH16 (NA)

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Data from Other heavy-water moderated and light-water cooled facilities

• DCA, FUGEN, Japan• ECO, CIRENE Project, Italy• DIMPLE, SGHWR, Winfrith, UK• Savannah River Laboratory

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DCA (Japan) Data• Heavy water in Al tank (1 cm

thick, 3 m dia. 3.5 m height)• Light water cooled• 28-element cluster (2 m long) • Enriched UO2 fuel and (Pu,U)O2

fuel• Square lattice pitch 22.5 cm

(121 channels) or 25 cm (97 channels)

• Ambient temperature• Measurements: criticality, CVR,

reaction rates, flux profiles

Calandria Tube

Pressure Tube

Fuel Cluster

Heavy Water Level

Al Spacer50(W)¡ ¿100(L)30Pieces50(W)¡ ¿130(L)24Pieces

Absorber Sandwich20 Al12.7 B4C(35w/o)-C

9.7 Al =1.5

Al Void Tank20 (Al)

80 (SUS)

(Al)20

20(Al)

20(Al) 36(SUS) 36(SUS)

Dum

p Valve

1.10

0 Concrete

50

30 (Al)

50(SUS)

63 (Al)

Upper Grid Plate45 (Al)40

0

20(Al)

10(Al)

Al Support for Upper Grid Plate50(th)¡ ¿355(W)¡ ¿Total

Al Grid Plate45(Al)

D2O

ρ

60

Upper gridPlate with Al Support

VacantAltubes

Bare Fuel Part aboveD2O level

Core

Lower grid plate AbsorberAl SpacerSteel base

D2O

2.00

0

3.10

0

3. 005

Calandria Tube

Pressure Tube

Fuel Cluster

Heavy Water Level

Al Spacer50(W)¡ ¿100(L)30Pieces50(W)¡ ¿130(L)24Pieces

Absorber Sandwich20 Al12.7 B4C(35w/o)-C

9.7 Al =1.5

Al Void Tank20 (Al)

80 (SUS)

(Al)20

20(Al)

20(Al) 36(SUS) 36(SUS)

Dum

p Valve

1.10

0 Concrete

50

30 (Al)

50(SUS)

63 (Al)

Upper Grid Plate45 (Al)40

0

20(Al)

10(Al)

Al Support for Upper Grid Plate50(th)¡ ¿355(W)¡ ¿Total

Al Grid Plate45(Al)

D2O

ρ

60

Upper gridPlate with Al Support

VacantAltubes

Bare Fuel Part aboveD2O level

Core

Lower grid plate AbsorberAl SpacerSteel base

D2O

2.00

0

3.10

0

3. 005

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Configuration of DCA core with 25 PuO2-UO2 test fuel assemblies

250

250

PLUTONIUM FUEL ASSEMBLYURANIUM FUEL ASSEMBLY

3005

Fuel and coolant channe l

Tota l Number of Fue l Channe l : 97

250

250

250

250

PLUTONIUM FUEL ASSEMBLYURANIUM FUEL ASSEMBLY

3005

Fuel and coolant channe l

Tota l Number of Fue l Channe l : 97

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Cross-sectional View of DCA 28-rod 1.2 wt% UO2 Fuel Assembly

15.0616.68

120.8136.5

Modera tor (D2O)Calandria Tube (Al)

Air GapPressure Tube (Al)

Clad (Zry-2)

Fuel Pe lle t (P UO2-UO2)Coolant

2.02.0

26.25 60.0

95.15

14.7

15.0616.68

120.8136.5

Modera tor (D2O)Calandria Tube (Al)

Air GapPressure Tube (Al)

Clad (Zry-2)

Fuel Pe lle t (P UO2-UO2)Coolant

2.02.0

26.25 60.0

95.15

14.7

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DCA UO2 Bundle and (P,U)O2 Test Bundles

Fuel Type 0.54 wt% PuO2-UO2

0.84 wt% PuO2-UO2

1.2 wt% UO2

Fuel pellet Density (g/cm3) Diameter (mm) Enrichment (wt%)

10.17 14.69

5SPu*

10.17 14.72

8SPu**

10.36 14.80

1.203 Composition (wt%) U-235 U-238 Pu-238 Pu-239 Pu-240 Pu-241 Pu-242 O

0.6214

86.782 0.000102 0.4304 0.04115 0.004359 0.000303

12.12

0.6194

86.503 0.000145 0.6849 0.06584 0.006960 0.000510

12.12

1.057

86.793

12.15 Fuel pin (mm) Clad material Clad inner diameter Clad outer diameter Gap material

Zircaloy-2

15.06 16.68

He gas

Zircaloy-2

15.06 16.68

He gas

Al

15.03 16.73

Air *5SPu: PuO2 weight fraction in PuO2-UO2 = 0.542 wt%

**8SPu: PuO2 weight fraction in PuO2-UO2 = 0.862 wt%

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ECO Data

Ref: Misure di Buckling con Reticli al Torio nel Reacttore ECO, F. Amoroso et al RT/F(75)2

• Reactor at ISPRA (Italy)• Vertical core, side and bottom graphite reflector• 89 fuel rods• 19 and 37-element bundles• SEU (2.35%) and (Th, U)O2 fuel• Coolant: Air, H2O, D2O and H2O-D2O • Lattice pitches: 18.8, 22.3, 25.5 and 28.1 cm• Measurement of critical heights and bucklings

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Summary• Physics code qualification program considers:

− phenomenon-based framework of the AECL standard code validation methodology

− assessment results of the code applicability and capabilities with respect to ACR nuclear design

− developments required to bridge the gaps and eliminate code deficiencies• Key documents (VM, Qualification Plan) prepared• Requirements on validation data

− ZED-2 experimental program− data from other facilities− inter-code comparisons

• All physics PH are addressed• Program is underway – execution schedule linked to ZED-2

experimental program

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