Spent Fuel Management and Storage in Korea

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1 2010 Int l Seminar on Spent Fuel Storage, nov.15-17, Tokyo, Japan Spent Fuel Management and Storage in Korea Jongwon CHOI

2 Contents 1 Spent Fuel Management NPP & SF Status Policy Storage R&D Activities for Dry Storage Storage System Spent Fuel Integrity

3 Nuclear Power Plant in Korea 2 Site In Operation Units (MWe), As of Oct Under Construction Total Kori 4 (3,137) 4 (4,800) 8 (7,937) Wolsong 4 (2,779) 2 (2,000) 6 (4,779) Yonggwang 6 (5,900) - 6 (5,900) Ulchin 6 (5,900) 2 (2,400) 8 (8,300) Total 20 (17,716) 8 (9,200) 28 (24,516) Yonggwang (#1,2,3,4,5,6) Seoul Daejeon Ulchin (#1,2,3,4,5,6) Wolsong (#1,2,3,4) Shin-Wolsong (#1,2) Kori (#1,2,3,4) Shin-Kori (#1,2,3,4) ~10 NPPs will be added in Hydro1.4% Fossil Fuel 59.6% Nuclear 39.0%

4 Spent Fuel Amount 3 Amount (1000 MTU) SF Amount 7,960 11,000 PWR CANDU TOTAL 19, Year Annual arising : 700 tu/yr CANDU ~100 tu/yr 4 units ~400 tu/yr PWR ~20 tu/yr 16 units ~320 tu/yr NPP Site Capacity (ton) Accu. (ton) Expected Saturation (year) Kori 2,190 1, Yonggwang 2,670 1, Ulchin 2,350 1, Wolsong 9,440 5, Sum 16,650 10,872 -

5 National Policy for RadWaste rd AEC s Decision (Dec. 2004) Final repository for LILW should be constructed by Delayed until 2012 for Safety Reason Spent Fuel Management - All spent fuel will be stored at plant sites until Future national policy for SF management will be decided through public participation taking into consideration of national and international trends on policy and technology development.

6 RadWaste Management Act 5 This Act has been effective since January 1 st, Key contents are: - Establish. of a new Org., responsible body for radwaste management Korea Radioactive Waste Management Corporation (KRMC) - Establish. of the Radioactive Waste Management Fund which will be paid by the Radwaste generator. KRMC is responsible for managing the RadWaste Fund

7 Organization related with RadWaste Management 6 6

Overall Plan for Radwaste Management 7 SFP (4) Dry

Point (to be determined) Power Reactors PHWRs (4)

Interim Storage Deep Geological Repository Post

LILW Interim Storage under Construction Research

8 Overall Plan for Radwaste Management 7 SFP (4) Dry Storage Interim Storage (to be determined) End Point (to be determined) Power Reactors PHWRs (4) PWRs (16) SFP (16) ANFC? Interim Storage Deep Geological Repository Post Irradiation Examination PIEF Facility Hot Cell LILW Interim Storage under Construction Research Reactors HANARO KRR 1 and 2 to be determined transported to the United States (1998)

Spent Fuel Storage at NPP * As of Dec. 2009 (Unit : MtU) Plant Site Capacity Yr of Saturation Accum.

Total 13,246 10,083 To secure the on-site storage capacity PWR s storgae facility has been expanded by

CANDU SFs have been transferred to the dry storage facility (concrete silo) since 1991 Kori Unit 1 :

9 Spent Fuel Storage at NPP * As of Dec (Unit : MtU) Plant Site Capacity Yr of Saturation Accum. Kori 2, ,685 Yonggwang 2, ,623 Ulchin 2, ,294 Wolsong 5,980(9,440) ,481 Total 13,246 10,083 To secure the on-site storage capacity PWR s storgae facility has been expanded by re-racking and transshipment. CANDU SFs have been transferred to the dry storage facility (concrete silo) since 1991 Kori Unit 1 : Transshipment to Units 3&4 Unit 3&4 : Addition of high density racks Ul chin Unit 1&2 : Full re-racking of AR pool Young kwang Unit 1 : Transshipment to Units 3&4 Wol song Unit 1 : 300 Concrete silos for 162,000 bundles

10 Expansion of Storage Capacity 9 PHWR(CANDU) SF PWR SF High Density Reracking 1 2 Concrete Silo MACSTOR/KN-400 Transshipment between NPPs

11 Concrete Silo for CANDU SF Storage 10 Concrete Silo System Capacity: 540 Bundle (60 Bundle/Basket x 9 Basket) Out Diameter: 3.07 m Height: 6.52 m Total 300 Silos (~3,200 MtU) installed from 1990

MACSTOR/KN-400 for CANDU SF Storage 11 Air

site (2010) Economy : reduce of Area by 2.

12 MACSTOR/KN-400 for CANDU SF Storage 11 Air Inlet Air Outlet Weather Cover Shield Plug CANDU Fuel Fuel Basket Storage Cylinder High-dry Storage Facility 7 modules at Wolsong site (2010) Economy : reduce of Area by 2.7 times compared to concrete silo Cooling: Passive Natural Cooling

Design Parameters of MACSTOR/KN-400 12 Storage System Lifetime 50 years Temp.

13 Design Parameters of MACSTOR/KN Storage System Lifetime 50 years Temp. Limit in operation 66 24,000 bundles KN-400 system System: 40 cylinders Cylinder : 10 baskets Basket : 60 bundles 5,000 Bundles generation in a PWHR / year 24,000 bundles = 1.2 x All SF generation in 4 PHWR in a year = 44.4 x Silo dry storage Dimension Structure Material Thickness 22 (L) x 12.5 (W) x 7.5 (H) m Reinforced Concrete Side : 0.98 m Top : 1.08 m Cooling Time Average Burnup Average Heat Flux Initial U mass PHWR Spent Fuel Bundle Max. Temp in dry storage Minimum 6 years 7,800 MWd/MtU 6.08 Watt 19.2 kgu / Bundle 168

14 SF transshipment between NPPs 13 Transshipment Year No. of Spent fuels Transfer Cask Kori 1 Kori KSC-4 Kori 1 Kori KSC-4, KN-12 Kori 2 Kori KN-12 Kori 2 Kori KSC-4, KN-12 Kori 4 Kori KN-12 Total Total 928 assembles moved to neighboring units between

15 Contents 14 Spent Fuel Management NPP & SF Status Policy Storage R&D Activities for Dry Storage Storage System Spent Fuel Integrity

16 Backgrounds of SF Dry Storage 15 ISFSF should be in commission by 2016 Some prerequisites for dry storage (to be considered) - Integral spent fuel history & properties database : new fuel design, increased burnup due to improved op. tech. - Thermal cycling limitation due to to in-site transshipment for expansion of the on-site storage capacity - Technical criteria for safe dry storage system such as the long-term integrity of SF and storage facility material

17 Storage System Develop. Plan 16 Stage 1 Core Design Tech. Stage 2 System Design Stage 3 Performance Test, V&V Conceptual Design Evaluation Product System Design Transportation & PSA System Manufac. and Performance Test Storage System Develop. Burnup Credit Transfer Cask Design System Operation Demon. Facility Installation Serious Accident Evaluation Radiation Monitoring System Devel. Core Material Development New Storage System SF Integrity Degradation Model Unit/Integral Test of SF Model V&V Integral Integrity Evaluation Project Schedule Start Licensing Installation

ID NDR Burnup Discharege BU Discharege BU Fuel type

Cooling Date Enrichment Fuel Cladding Coolant Chem.

Date Thermal-Hydro Discharge Date Temp: Room Coolant :

18 Fuel Environ. Change 17 Manufacture Burn Wet Storage Dry Storage Fuel ID NDR Burnup Discharege BU Discharege BU Fuel type Discharge BU Damage Info. Cladding Info. Initial Enrich. Power History Cooling Info. Cooling Date Fuel Dimension Damage Info. Cooling Date Enrichment Fuel Cladding Coolant Chem. Transfer History Fuel Pellet Radiation Level Loading Date Thermal-Hydro Discharge Date Temp: Room Coolant : 150 atm Coolant : ~1.5 atm Air : ~1 atm Clad Pressure: ~320 C ~40 C ~20 C ~20 atm Cladding: ~450 C Clad: ~150atm Clad: ~150 atm Pellet: ~800 C ~100 C ~400 C

19 Fuel Supply History in Korea 18 Plant Year Kori-1 (WH14) KOFA OFA Kori-2 (WH16) KOFA STD ACE7 Kori-3/4 (WH17) KOFA V5H RFA ACE7 YGN-1/2 (WH17) KOFA V5H RFA ACE7 UCN-1/2 (WH17) KOFA V5H RFA ACE7 S-UCN-1/2(APR ) Guardian PLUS7 YGN-3/4 (OPR) OPR Guardian PLUS7 YGN-5/6 (OPR) OPR Guardian PLUS7 UCN-3/4 (OPR) OPR Guardian PLUS7 UCN-5/6 (OPR) Guardian PLUS7 S-KR-1/2 (OPR) Guardian PLUS7 S-KR-3/4 (APR) PLUS7 STD CANDU Wolsong(CANDU) S- Wolsong (OPR ) Guardian PLUS7 KOFA : Korean FA, OFA : W s Optimized FA, STD : W s Standard FA V5H : Vantage5H FA, RFA : Robust FA, NGF : Next Generation FA OPR : OPR FA, Guardian : OPR FA with Debris Filtering Grid, PLUS7 : Advanced OPR FA, STD CANDU : Standard CANDU Fuel Bundle

Capacity Factor in Korea 19 1984 ~ 1994 1998 ~ 2007

20 Capacity Factor in Korea ~ ~ 2007 Source: Ministry of Knowledge and Economy 2008 Nuclear Power Note

21 Discharge Burnup Increase 20 Discharge Burnup Increase as longer NPP operating cycle 235 U enrichment period 3.2 wt.% 1980 s 4.0~4.2 wt.% 1990 s 4.2~4.5 wt.% 2000 s(early) 4.5 wt.% 2000 s(current) Current average discharge burnup : 45 GWd/MTU 4.5 wt.% enriched spent fuels from 2010 with average 55GWd/MTU burnup Number of Assemblies % < U wt.% 86.4% < U wt.% Initial 235 U Enrichment (w/o) Production Amount [tu] Maximum Burnup Average Burnup Production Amount Time [year] Burnup of Spent Nuclear Fuel [GWd/tU]

Spent Fuel DB (1/2) SNF DB : planed tracing system on ID basis Mining

(AR) Interim Storage Final Disposal Uranium Provenance Fuel ID Fuel Type

Fuel Loading Date Fuel Discharge Date Discharge Burnup Fuel Inventory Fuel

Type (Wet / Dry) Fuel Discharge Date Storage Cask Information Fuel

Disposal Date Storage Cask Information Transport Procedure Fuel Inventory

22 Spent Fuel DB (1/2) SNF DB : planed tracing system on ID basis Mining Milling Conversion Enrichment Fuel Fabrication Reactor Operation Storage (AR) Interim Storage Final Disposal Uranium Provenance Fuel ID Fuel Type Cask Type Fuel Storage Facility Fuel Storage Location Initial Enrichment Fuel Loading Date Fuel Discharge Date Discharge Burnup Fuel Inventory Fuel Activity Defect Fuel Check Heat Generation End Cycle Enrichment Storage Type (Wet / Dry) Fuel Discharge Date Storage Cask Information Fuel Inventory Fuel Activity Heat Generation Fuel Undertaking Date Fuel Disposal Date Storage Cask Information Transport Procedure Fuel Inventory Fuel Activity Fuel Disposal Date Storage Cask Information Transport Procedure Fuel Inventory Fuel Activity

23 Spent Fuel DB (2/2) SNF DB Module Structure Fuel Storage Information Storage Facilitty Cask Serial Number Cask Type Storage Date Defect Fuel Information Defect Fuel Rod Occurrence Date Management Facility Extracted Ass. No. Storage Facility Storage Location Specific Matters FA Information Manage Facility Total Assembly Number Occurrence Cause Minimum Burnup Period Final Burnup Period Initial Core Loading Date Final Core Discharge Date Ass. Arrangement Type Fuel Type Reactor Equipment Fuel Rod Information Management Facility Separation Factor for Assembly/Fuel Rod User ID Cause Information Acquisition Date Extracted Ass. No. Fuel Defect Check Radiation Information Initial Uranium Enrichment Fuel Material Mass in Assembly Discharge Burnup Total Mass of Assembly Volume of Assembly Heat Generation Rate Evaluation Date Heat Generation Rate Total Radioactivity Evaluation Method Total Radioactivity Total Radioactivity Evaluation Date Remaining Uranium Inventory Defect Fuel Rate Nuclide Activity Detection Method Nuclide Activity Detection Date Nuclide Activity Nuclide Code Radiation Intensity H-3 C-14 Co-60 Sm-151 Cs-135 Cs-137 U-235 U-238 Np-237 Pu-238 Pu-239 Pu-240

24 Tech. Roadmap for SF Integrity study 23 Objective Technology development for the dry-stored SF integrity evaluation Test Procedure Prototype Code Code Verification Patent Commercialziation Properties Analysis Prototype Model Material Degradation Model Spent Fuel Degradation Model Hot Test Senario Test Facility PIE Test Test Facility design & construction Hot test Long-term behavior test Hot test database Model verification Integral Code Develop Comprehensive Evaluation Technology Application - Design Requirement and improvement(s) - Basic data for license (2016~) Phase 1 ( ) Phase 2 ( ) Phase 3 ( ) Technology Application

25 Objects of the Study 24 Spent Fuel Structural Components Assembly Fuel and Clad Metal Cask Components Concrete Facility

26 Ongoing Project (1/2) 25 To develop reference degradation mechanism model under SF characteristics and storage condition Target Developing prototype model Evaluating SF structural deformation Analyzing previous SF test data Developing integrity evaluation code system Product Patent/report/article for degradation models Integrity evaluation code design and model PIRT (Phenomena Identification and Ranking Table)

27 Ongoing Project (2/2) 26 To develop hot-test scenario of 2 nd Phase experiments Target Characterization test scenario of SF Unit test scenario of SF Integrated test scenario of SF Product Conceptual design of unit test experiments Patent application for conceptual design Scenarios for Char./Unit/Integrated test

28 Major Degradation Mechanisms 27 Clad Degradation Creep Rupture Hydrogen re-orientation Delayed Hydride Cracking Oxidation Stress Corrosion Cracking Diffusion Controlled Cavity Growth Pellet Degradation Oxidation Fragmentation

29 Characteriz. Test of SF in PIEF 28 FA Pool Test Rod Hot Cell Test Specimen Hot Cell Test Transportation Undertaking Unloading Dismantling Cutting Spent Fuel Integrity Degradation Damage Creep He-reorientation Storing Pool Test NDT DT Macro Texture Visual Inspection Dimension Measuring Burnup Measuring Visual Inspection Oxidation Measuring Dimension Measuring ECT FGR/Internal Pressure Micro Texture Gamma Scaning Density Measuring Burnup&H Analysis SEM / EPMA

30 Characteriz. Test of SF in PIEF 29 Own Assembly Rod Design type Array type Fuel Name STD KOFA OFA STD KOFA STD 16ACE7 OFA KOFA V5H RFA 17ACE7 KSNP Guardian PLUS7 Owned Assembly Enrich (U-235 wt %) 2.1~ Burnup (GWd/tU) 17~ Cladding Zry-4 Zry-4 Imp. Zry-4 Owned Rods Enrich (U-235 wt %) ~ ~ ~ Burnup (GWd/tU) 32~42 7~12 40~56 34~54 1.8~2 48~55 55~58 Cladding Westinghouse type 14X14 16X16 17X17 16X16 Zry-4 Low Tin Zry-4 Adv. Zry-4 Zirlo OPR Zirlo Zry-4 Zirlo Zirlo Sum Charac terizati on Test NDT DT Visusal Test Dimension Measure ECT γ-spectroscopy Rod Internal Pressure/FGR Measure Ceramography/Metallography Chem. Analy(Burnup,Hydroge) Pellet/Clad Material Properties

31 Conclusions 30 Long-term SF management program will be established through public engagement. For mid-term policy of SF management, interim SF storage facility will be in operation by 2016 Since June 1 st in 2009, we have started R&Ds on SF Integrity Study. Integrity Evaluation Code and Indigenous test data will be expected.

32 31

Implementation Schedule for 1 st Phase 2009 2010 2011 6 8 10 12 2 4 6 8 10 12 2 4 Degradation model development of Storage System Material structural fabric list pre-exist degradation mechanism

33 Implementation Schedule for 1 st Phase Degradation model development of Storage System Material structural fabric list pre-exist degradation mechanism analysis grouping structural fabrics technologies for each storage type domestic/international meeting reference model reference model evaluation degradation causes and behaviour analysis reports International Workshop prototype model develop reports, articles pre-exist degradation model analysis reference model selection reference model evaluation prototype model develop Degradation model Development of SF foreign/domestic integrity data compilation & analysis degradation causes and behaviour analysis integrity evaluation code requirements analysis model/code connection scheme develop expert advice domestic/international meeting reports International Workshop model code, reports, ariticles Development of Experimental Plan for the 2 nd Phase previous spent fuel test database analysis spent fuel selection for pre-test test item list for spent fuel integrity properties spent fuel properties test senario degradation unit mechanism test item list dry storage performance verification technology develop unit mechanism test procedure unit test/full test conceptual design of experimental equipments relation procedure check between model and test results expert advice domestic/international meeting reports International Workshop reports

The ROK s Nuclear Energy Development and Spent Fuel Management Plans and Options

The ROK s Nuclear Energy Development and Spent Fuel Management Plans and Options Jungmin KANG Visiting Professor, KAIST Presentation given at 2012 Resilience and Security of Spent Fuel in East Asia WORKING