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Advanced CANDU Reactor (ACR-700)

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Advanced CANDU Reactor (ACR-700) - g 4~~~~~~...,, Z ,,A,I "%,,, ira&-n ZED- 2 Measurements using Full-Core I CANFL EX 0.95% SEU (Scheduled in 2004) hi- Flux-Map Set Up For Square Lattice i * Flux-maps and buckling measurements in hexagonal lattice ILE '.# - existing ZED-2 aluminum PT/CT assemblies with 3 coolants - H20, D20 and air coolant, at 3 or more lattice spacings * Flux-maps and buckling measurements in square-lattice - repeat of above using square lattice - copper activation foils will be positioned across the lattice and into the heavy water-reflector to measure the thermal flux peak in reflector on voiding * Substitution measurements - 0.95% SEU into various ref. lattices (validation of substitution method) - 7-rod substitutions inACR-type PT/CT assemblies with H20,D 20 and air coolant into ref lattice of 0.95% SEU (inexisting ZED-2 aluminum PT/CT) * CANFLEX LVRF (1%SEU, Dy with NU incentral element) . MOX (simulating irradiated ACR fuel) * Temperature Reactivity Effects of Fuel, Coolant, and Moderator for various configurations 1 I L I I If II fI - I 4 41 .4 E 4 0I PS 21 P 22 '7 1. ,n_- I - - - I------ -- 7- - I .'., '.',' _A '': I I , _ 11;11 '' , : .. ' - I Ea~I 1. ,7; ': :'_- - I -i .- ' ' ;iA. 7; 1 hWL~~~Z - -FW Summary 4 * Coolant Void Reactivity in CANDU lattice is caused by spatial and spectral changes of neutron flux upon voiding * The physics design of ACR manipulates these changes to achieve a slightly negative coolant void reactivity with H20 coolant, by: - Using a tighter lattice pitch and a larger gap between pressure tube and eAs AECL calandria tube than the existing designs TECHNOLOGIES INC. - Adding burnable poison (Dy) in the central fuel pin - Using SEU fuel * Confirmation of Negative CVR by: - Comparisons of AECL's physics codes with international physics codes - Verification of negative CVR in ZED-2 reactor experiments at CRL /7 23 PI24 r I.C_ _ _ I 1-_ ( _ [ __ -1 I (~~~~. L..t I [ ______-_I_______I___-I( I I [t - : -F---_ < f - F . r I - F _[ ~~ -- T- F--___F~~~~~~~~7-r _f____K-.-F---- __F- _('__ Outline ACR Fuel * Introduction to CANDU fuel * ACR fuel design Peter G. Boczar * Experience relevant to ACR fuel Director, Reactor Core Technology Division - CANFLEX CRS Subcommittee on Future Plant Designs Washington D.C. - extended burnup experience January 13, 2004 - low void reactivity fuel A AECL * ACR fuel qualification TECHNOLOGIES INC. eli 777 Characteristics of CANDU Fuel CANDU 6 37-element Fuel --- - ad-A. ....... 1.N5.............. <;pT=;ctt, * Small, simple, light-weight - 20" length, 4"dia, 50 lb / bundle - CANFLEX has only 8 components I * Inexpensive - low fuel cycle costs (dollars/unit energy) * Efficient - good use of uranium * Excellent performance - -2 million bundles fabricated; - 2 clad defects per million elements - on-power defect detection, location and removal * Easy to manufacture and localize - CANDU fuel ismanufactured domestically in 7countries - CANDU (and its fuel) licensed in many different regulatory jurisdictions O., U0 2 Pellets Clad, CANLUB, Endcaps, Endplates . Clad * U02, high density (for dimensional stability) - thin, collapsible (-0.016") * Chamfers and end-dishes (reduce inter-pellet stresses on clad, - excellent heat transfer to coolant volume for fission gas) - low neutron absorption, Zr-4 * CANLUB - graphite coating applied to inside of A;^ .X\ clad provides protection against .t A_ power ramp failures * Endcaps - seal the fuel element - thin to reduce neutron absorption, good heat transfer - profiled to interact with fuel channel " 'I """ -' i VT \1 T,\ ,, and fuel handling components 'O r . Endplates mc 11 A:;_ - thin to minimize neutron absorption - flexible to accommodate fuel element differential expansion 1W5 - strong and ductile to provide structural support and element separation ,,t I M Spacers, Bearing Pads CHF-Enhancing Buttons (CANFLEX) * Appendages are attached on the 1/4 and 3/4 bundle planes * Inter-element spacers - provide element separation at the bundle midplane * Bearing pads - provide element-to- , pressure tube separation Pt 7 4 I I I I 11 I I I I ~II I I I I 1 I I I I II I I I I 1 - I F 1 1- A.' ACR Fuel Design CANFLEX Geometry * Evolutionary extension of current fuel * ACR fuel based on CANFLEX Mk IV - extensive experience base on underlying geometry technologies - 43 elements, 2 element sizes * Based on 3 underlying technologies - greater "subdivision" reduces ratings and - CANFLEX geometry facilitates achievement of higher burnup - lowvoidreactivityfuel - "buttons" increase CHF - extended burnup - qualified for NU fuel * Key design features - higher bearing pads further improve CHF - 2.1% U235 in outer 42 elements compared to Mk IV - 7.5% Dy in nat. U02 in central element - 21 MWd/kg burnup PA10 44k ; %, CANFLEX Geometry Evolution of CANDU Fuel Al. Other Design Features Smmary of CANFLEX NU Qualification * Design requirements documented in Design Requirements, * Optimized pellet design Design Verification Plan - in smaller elements (highest ratings) * Tests and analysis confirmed that CANFLEX met all requirements . larger chamfers, deeper dishes, shorter pellets - strength * more internal void for accommodating fission gas release - impact and cross-flow * reduces inter-pellet clad strain - fueling machine compatibility, endurance - sliding wear * Slightly thicker clad - fuel performance (NRU irradiations) - to accommodate higher coolant pressures and temperatures - CHF thermal hydraulic * Demonstration Irradiation (Dl) in Point Lepreau 1998 to 2000 - 2 channels, 24 bundles - irradiation of 24 bundles currently taking place inWolsong 1 * Design qualification program documented in Fuel Design Manual * Ready for commercial implementation in CANDU 6 reactor Pg 13 I'M14 tjwl experience with Low Void Reactivity Fuel Overview of LVRF Testing * Dy 0 -U0 pellet fabrication * Reactor physics ACR fuel is variant of LVRF 2 3 2 0 - measurement of thermal properties - ZED-2 measurements Generic testing done for - corrosion behavior of U02 * void reactivity - 37-element LVRF (NU burnup, with negative void reactivity inCANDU 6) * Bundle fabrication . fine structure - CANFLEX LVRF (3x NU burnup, with negative void reactivity in CANDU 6) * Irradiation testing in NRU & PIE - WIMS validation - Dy-doped demountable elements * Thermalhydraulics with Dy levels of 1to 15% - measurements .1 - prototype bundles - modeling I*, U * Safety experiments - interactions with Zircaloy - grain-boundary inventory * CANFLEX LVRF currently being qualified for Bruce Power implementation - enrichment, Dy content tailored to meet station needs 37-element LVRF CANFLEX LVRF PR15 - synergistic with ACR fuel qualification Pg 6 L I I I I -I L I I I II II I III I I I I ) __ I- IW I __ I I I I I II~~~~~~~~~I_ I I I I I I I I Extended Burnup Irradiation Experience NRU Fuel String * Power reactor experience - >230 37-element bundles achieved burnups > 17 MWd/kg in Bruce A * Research reactor experience - >24 bundle and element irradiations in NRU > 17 MWdIkg . 15 irradiations with burnups greater than 21 MWd/kg - 10 of 24 irradiations also experienced power ramps - several irradiations ongoing * Qualified irradiated fuel databases - 28-element, 37-element and CANFLEX * Good confidence in ACR fuel performance based on our experience - ACR power envelope is below the high power envelope for which we have experience - ACR fuel pellet design is optimized for extended burnup, based on our experience base and assessments Pg1 PpIP km", ",IAM , . |: A At_ e P~~~~ll !w R Power Envelope vs. Bruce A Experience i-VI ACR Fuel Qualification - - Bruce A Experience - - - Dy Doped Centre Elements * Will ensure ACR has full thermal integrity, structural SEU Outer Elements integrity, and compatibility with interfacing systems E 7n -- SEU Inner Elements 5 { * Comprehensive, integrated set of in-reactor tests, out- ,, 60 V, reactor tests, and analyses I.- 0 50 * Qualified computer facilities, codes, and staff L 40 * US fuel consultants providing guidance X 30 i 20 E 10 a, Cw 0 5 10 15 20 25 30 Element Burnup (MWd/kgU) PgR19 PR20 I -- - - p~:ir;- Approach Summary * Systematically evaluate impact of all significant operating * ACR fuel builds on an extensive experience base and damage mechanisms, - CANFLEX geometry individually and in combination - low void reactivity fuel * Confirm consequences are within acceptable limits via - enriched fuel (extended burnup performance) combination of * ACR fuel qualification will be facilitated through recent AECL - in-reactor tests experience in fuel qualification - out-reactor tests - CANFLEX Mk IVfuel with natural uranium - analyses, and - current qualification of CANFLEX-LVRF for Bruce Power - engineering judgment * ACR fuel qualification will entail out-reactor tests, in-reactor tests, and analyses * Envelope all permitted operational and design * Numerous background papers on CANDU fuel have been sent to US configurations NRC * Ensure sufficient margins exist that account for burnup, * ACR fuel report, summarizing ACR fuel design, experience base, fuel peak element rating, coolant temperature and flow rate design requirements, and qualification plan will be sent to US NRC shortly I-22 I i I * .,.',,..' ,- . ,. az'aeA> Aq"AECL TECHNOLOGIES INC. Pe23 C 1'L~ ___ [l --- ----- -_[ __,_f _ I __ L.~~~~~~ i.~~~~~~~~~~ Ii I I I r I - , -.. C.. I * r----r--- F ~-- - [U- -V-- F 7,-I i-I - F- - t-r -F -W- Ir- Fr-- rd X F-1 r I 1-- 1, 1-w-77777777 I, orical Perspective of AECL's PRA Projects * AECL brings many insights of their long PRA ACR PRA Methodology experience to the ACR PRA: - SDMs - 1978-1983: CANDU 6 and Ontario Hydro's NPPs - CANDU 600 Probabilistic Safety Study - March 1988 Raj Jaitly - Wolsong 213/4 PRA - March 1995 Manager, PSA and Safety Design - KEPRI- Wolsong 2/3/4 Level 2 PRA Review - 1997 ACRS Subcommittee on Future Plant Designs - Qinshan CANDU Unit 1and 2 PRA - May 2001 10% Washington D.C. - Generic Level 2 PRA for internal and external events 2002 A'' .ft~to~'-V;, January 13, 2004 - Pickering A Return to Service PRA Review - 1999 - Lepreau Refurbishment Project Level 2 PRA - ongoing A"AECL - Preliminary PRAs for CANDU 3 and CANDU 9 (1994,1997) TECHNOLOGIES INC.
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