Sodium-cooled Fast Reactor Design Principles and Approach II

Transcription

1 Sodium-cooled Fast Reactor Design Principles and Approach II Seminar at KAIST April 24, 2013 YOON IL CHANG Visiting Professor, KAIST World Class University Program Supported by NRF Grant Funded by MEST

2 Outline Coolant Properties Intermediate Loop Reactor Configuration Options Sodium Purification Reactor Outlet Temperature Actinide Burning Sodium Void Reactivity 2

3 Fast Reactor Coolant Properties Sodium Lead 56Na- 44K 22Na- 78K 44.5Pb- 55.5Bi Helium Density, g/cc Viscosity, c.poise Melting point, o C Boiling point, o C Spec. heat, J/gK Th. cond, W/mK

4 Sodium Fire Liquid sodium reacts readily with air and oxidation reaction can occur in a runaway manner leading to sodium fire. Sodium burning is accompanied by production of dense sodium oxide smoke. The heat itself is much less than that of conventional hydrocarbon fires. Sodium flame height is also an order of magnitude lower, allowing a close approach for fire fighting. 4

5 Design Solutions Sodium in the primary system is blanketed with inert gas and is maintained in double containment. Reactor vessel sodium free surface is covered with inert gas, and the gap between reactor vessel and guard vessel as well as the gap between sodium pipe and guard pipe around it are also filled with inert gas. A variety of techniques are available for sodium leak detection. The principal technique relies on the detection of sodium aerosols in the annulus gap, produced by oxygen and water impurities in inert gas atmosphere. In the secondary sodium system, a gap between the pipe and its insulation enable leak detection, but no preventive measures are deployed. 5

6 Sodium-Water Reaction Sodium reacts exothermically with water, producing caustic sodium hydroxide and hydrogen gas. Therefore, it is highly desirable to avoid potential contact of sodium with water/steam in steam generator design. One conservative approach is double-wall tube steam generator: EBR-II SG operated without a single tube leak for the entire 30-year life. 6

7 Design Solutions A secondary sodium loop is always added to isolate primary sodium system from BOP. Steam generator tube leaks will not impact the primary system. Steam generator relief and dump system is also employed to relieve pressures in the secondary sodium system in the event of a major steam/water leak into sodium. Rupture disc is located in the SG and sodium-water reactor products are collected in a separator tank. 7

8 Summary of Sodium Leak Experience Primary sodium is maintained in double containment, blanketed with inert argon gas. There has been no leaks in the primary system. Sodium leaks in the secondary loop piping are easily detected and the resulting smokes or fires are extinguished without undue difficulty or consequence. Few reactors were plagued with leaks in steam generators during early years of operations. Most leaks were at tube-to-tubesheet welds, a significant fraction due to manufacturing (welding) defects. Some were due to material defects and lack of post-weld heat treatment. 8

9 Intermediate Heat Transport System Consists of intermediate heat exchanger and piping Not a major cost penalty Isolates the primary heat transport system from sodiumwater reaction and sodium fire In the absence of intermediate heat transport system, the steam generator becomes a safety-grade system, a challenging task. 9

10 Schematic of Intermediate Sodium Loop POWER SECONDARY SODIUM S T E A M G E N E R A T O R STEAM TURBINE GENERATOR STEAM P U M P HOT SODIUM IHX CORE 10

11 Reactor Configuration Options Pool-type: Reactor core, IHX, and primary pumps are all housed in a single reactor vessel. Loop-type: Only reactor core is housed in the reactor vessel, and IHX and primary pumps are housed in separate vessels, connected to the reactor vessel through piping. Hybrid-type: Combines the features of both pool and loop types. 11

12 Pool-type Configuration

13 Loop-type Configuration

14 Early experimental reactors have been mostly loop-type Loop EBR-I BR-5/10 DFR Fermi-1 Rapsodie BOR-60 KNK-II Joyo FFTF FBTR Pool EBR-II CEFR 14

15 Both types have been built for demonstration and near-commercial projects. Loop BN-350 SNR-300 CRBR Monju Pool Phenix PFR BN-600 SuperPhenix PFBR BN

16 Historical Background Early small experimental reactors were straightforward to build as loop-type. Pool-type was introduced with EBR-II in Initially, it was logical to layout the primary system components interconnected with pipings. However, it became increasingly clear it would be difficult to design a piped system with valves, etc. to achieve assurance of a high level of leak tightness of the sodium coolant. The EBR-II designers decided to place all primary system components and pipings inside a single reactor vessel. Also a guard vessel was added for added assurance. 16

17 Historical Background (cont d) The success of EBR-II led to the prototype plants, Phenix in France and PFR in U.K. to be based on pool-type. Although BN-350 was loop-type, the follow-on BN-600 as well as BN- 800 in Russia were based on pool-type. SuperPhenix (1240 MWe) was also pool-type. Argonne planned on the pool-type in EBR-III, but AEC changed the direction to loop-type with FFTF and CRBR. Japan followed the U.S. lead for loop-type Monju. Argonne re-introduced pool-type in early 80s and recent U.S. designs, PRISM and SAFR are pool-type. 17

18 Pool-type Design Characteristics All primary system components are contained in a single reactor vessel, leading to a compact primary system, a small footprint containment building, and hence a lower capital cost. All radioactive sodium is contained in a single reactor vessel and hence a low probability for release. Primary system boundary is a simple geometry with no vessel wall penetrations. Can tolerate minor leakage of primary heat transport system within the reactor vessel. In the event of reactor vessel leakage, the free surface continues to cover all components so that pumped flow can continue, namely designed to handle faulted condition. 18

19 Pool-type Design Characteristics (cont d) Thermal inertia and heat transport arrangements make the primary system more tolerant of transients and component failures. Reactor vessel can be maintained at a nearly isothermal cold sodium temperature and hence, combined with a lower fluence level, leads to longevity. Can accommodate passive decay heat removal systems based on natural circulation. Low pressure drop between the core and IHX allows placement of primary pumps in the cold pool. The reactor vessel is maintained at a lower pressure than the loop-type. 19

20 Sodium Purification System Solubility of oxygen and hydrogen in sodium increases with temperature. In a cold trap, as the temperature is reduced, the impurities precipitate as sodium oxide or sodium hydride. In EBR-II the original sodium was used for over 30 years without any makeup. 20

21 Experience of EBR-II Primarty Cold Trap Steady- State Cold Trap 120 o C Oxygen 1.2 ppm 0.7 ppm Hydrogen 80 ppb 55 ppb Primary source: moisture introduced during refueling Tritium is also removed along with hydrogen Cold trap efficiency improves with time as impurity crystals accumulate, providing more surface area for further crystallization. Then the effectiveness gradually lessens as it becomes choked with precipitates. Typically operated for 7 years. 21

22 Impurity Levels Before and After Cold Trapping 22

23 Cesium is Removed in Cesium Trap 23

24 Thermal Efficiency in Superheated Steam Cycle 24

25 Worldwide SFR Temperature Experience Reactor Outlet Temperature EBR-I 316 EBR-II 463 Fermi FFTF 503 CRBR 535 DFR 350 PFR 560 Phenix 560 SuperPhenix 545 KNK-II 550 SNR BN BN Joyo 500 Monju

26 U.S. Trend in SFR Core Outlet Temperature Following the cancellation of CRBR Project (535 o C), there were intensive debates on optimum reactor core outlet temperatures. Extensive tradeoff studies were carried out in the 1980s, which resulted in an industry consensus that an optimum outlet temperature is in the range of 950 o F or 510 o C. A higher outlet temperature results in an increased thermal efficiency, improving the economics. On the other hand, a higher temperature induces many cost penalties due to design changes to cope with reduced safety margins and structural design margins. 26

27 Typical Tradeoff Study Result 27

28 U.S. SFR Designs Since 1980s Outlet T Net Th. Efficy. o C MWth MWe % LSPB SAFR PRISM Mod A PRISM Mod B S-PRISM SMFR ~ ABTR ~ ABR ~

29 Breeder, Burner, or Break-even? Breeding is possible only with external blankets: Axial blanket is extension of fuel pin below and above active core. Radial blanket around the core utilizes bigger diameter pins for higher fuel volume fractions. If axial and radial blankets are replaced with reflectors, then the reactor becomes a burner. If only sufficient amount of blanket, a break-even or selfsufficient cycle can be achieved. 29

30 Fuel for Initial Fast Reactors Today there are excess plutonium inventory from LWR reprocessing, which can be utilized as startup fuel for initial fast reactors. LWR produces about 300 kgtru/gwe-yr. The TRU inventory requirement for a fast reactor including 2-3 reloads is of the order of tons/gwe. Conventional burner (w/o blankets) will require annual makeup of kgtru/gwe-yr depending on the reactor size. The startup of fast reactors is dictated by the availability of TRU (Pu) or the reprocessing capacity. The initial fast reactors will utilize the available Pu stockpiles, and hence a burner is a natural choice, gradually transitioning to a breeder. Hence, burning rate itself is not important. 30

31 Specific Fissile Inventory vs. Reactor Size 31

32 Relative Radiological Toxicity Effective lifetime of nuclear waste can be reduced from ~300,000 to ~300 years Transuranic Elements (Actinides) 10 1 Natural Uranium Ore Current Waste SFR Waste Fission Products ,000 10, ,000 1,000,000 Years 32

33 Actinide Burning Once actinides are removed from the waste streams destined for the repository, the recovered products have to be burned (or transmuted) to achieve benefits of a shorter waste lifetime. LWR thermal spectrum is not effective in burning actinides. Only fast reactors can effectively burn actinides, at the same time generating energy. 33

34 Transmutation Probabilities (in %) Isotope Thermal Fast Np Pu PU Pu Pu Pu Am Am-242m Am Cm Cm Cm

35 Evolution of Actinides in Thermal Spectrum (Pu recycle is typically limited to a single pass and cannot transmute minor actinides) 35

36 Goal of Actinide (TRU) Burning Remove actinides from the waste stream destined for a repository. Then, actinides cannot be just stored and they need to be burned at SFRs. But how fast you burn is not important, since actinides are valuable resources in the long term. In other words, burning actinides is not the principal reason for SFRs. 36

37 Fallacies of Actinide Burning Repository not necessary if actinides eliminated? First proposed by Charles Bowman at LANL Role of accelerator for transmutation? Spallation neutron source is expensive, so augmented with subcritical reactor: neutron multiplication = 1/(1-k). Higher actinide transmutation rate in the absence of U-238, but with k around 0.98, the design is similar to critical reactor in terms of neutronics and thermal-hydraulic constraints. Realities of partitioning and transmutation (P&T) Omega project aimed to partition all elements of fission products and transmute or use, but abandoned after 20 years of R&D. Fast reactors optimized for maximum burning? Proposed to compete with accelerator-driven transmutation LWR/SFR ratio, kg/gwe-yr, etc. are not proper figure of merit. 37

38 LWR Spent Fuel Radioactivity Normalized to EPA Cumulative Release Limits Radio-nuclide Activities at 10 years Activities at 1,000 years Activities at 10,000 years Sr-90 60, Cs , I Tc Other F.P. 1, Actinides 76,000 19,000 4,000 38

39 Assumptions for a Simple System analysis Excluding CANDU reactors, the PWR capacity was about 15 GWe in Assume 1 GWe addition per year through 2030, and additional 5 GWe by 2040, reaching a steady-state capacity of 40 GWe. As PWRs reach their lifetimes, SFRs are introduced as their replacements. Assume SFR requires 10 metric tons of TRU inventory (initial core inventory plus 2-3 reloads), with various conversion ratio: CR = 1.0 self suffiicient CR = 0.8 makeup requirement of 200 kgactinide/gwe-yr CR = 0.6 makeup requirement of 400 kgactinide/gwe-yr 39

40 Capacity, GWe Nuclear Capacity Assumption Total Capacity LWR Potential SFR Capacity Year 40

41 Cumulative Actinides, T Cumulative Actinide Inventory LWR Only w/o SFR CR = 0.6 SFR CR = 1.0 CR = 0.8 LWR w/ SFR Year 41

42 Summary on Actinide Burning The SFR core design should be optimized for normal performance characteristics to improve safety and economics. The initial SFRs can operate as burners, with blankets replaced by reflectors. The conversion ratio, whether 0.8 or 0.6, is not important. What s important is that a large inventory will be tied up in the SFR fuel cycle generating energy. (10 tons vs. ~0.1 tons variation) To achieve a lower CR, the TRU enrichment has to be increased, which implies detrimental effects on fuel performance, as well as safety and economics. 42

43 Sodium Void Reactivity If sodium is voided, by boiling or gas entrainment, the reactivity effect is positive in general. The positive sodium void reactivity has been a concern with MOX-fueled SFRs: Unprotected loss-of-flow event leads to sodium boiling, introducing positive reactivity, which can lead to promptcritical transient overpower and core meltdown classical hypothetical core disruptive accident (HCDA). MOX fuel elements typically fail at the mid-plane. A massive failures might release the plenum gas bubbles. However, such events are benign in metal-fueled SFRs: Unprotected loss-of-flow event does not lead to sodium boiling: HCDA initiators are prevented. Fuel element failures, if any, occur at the top of fuel column. 43

44 Important Safety Coefficients Power reactivity coefficient should remain negative at all operating power range. Coolant temperature reactivity coefficient should remain negative at all operating power range. If both reactivity coefficients remain negative, then whether one sub-component, such as sodium density or void coefficient, is positive or negative is not important; of course, if sodium voiding can be avoided. 44

45 Magnitude of Sodium Void Reactivity In small reactors, especially with enriched uranium fueling, the void reactivity is very small or negative. BN-600 with enriched uranium fueling has a slightly negative void reactivity. Following Chernobyl accident, a negative void reactivity was required by the Russian regulation, but it will have to change with MOX fueling and BN-800. In large SFRs with Pu-fueling, the sodium void reactivity is in the range of $

46 Physics of Sodium Void Reactivity Sodium void reactivity consists of leakage and spectral components. In general core designers try to spoil the geometry to increase the leakage component. However, in reality the spectral component dominate the leakage component. Therefore, the design options to reduce the void reactivity are very limited. The best approach is to eliminate the potential of sodium boiling: Emphasis on accident prevention by exploiting inherent safety as demonstrated in EBR-II. 46

47 Inherent Passive Safety Demonstrated in EBR-II The landmark tests conducted on EBR-II in April 1986 provide an integral demonstration of the superior passive safety characteristics of sodium cooled fast reactors, if properly designed. Two major accident initiators were simulated: Loss-of-flow without scram from full power; Loss-of-heat-sink without scram from full power. 47

48 Pool-type Sodium-cooled Fast Reactor POWER SECONDARY SODIUM S T E A M G E N E R A T O R STEAM TURBINE GENERATOR STEAM P U M P HOT SODIUM IHX CORE 48

49 Loss-of-Flow w/o Scram Sequence 49

50 Loss-of-Flow without Scram Test in EBR-II 50