ORNL-TM-7207 - the Molten Salt Energy Technologies Web Site

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1 e INTRODUCTION Molten-salt reactors’ (PISRS) have been under study and development in the United States since about 1947, with most of the work since 1956 directed toward high-performance breeders for power production in the Th-233U fuel cycle. terminated in September 1976 in response to guidance2 provided by the Energy Research and Development Administration (now Department of Energy) (ERDA/DOE) in March 1976. A brief study of alternative MSRs3 which em- phasized their antiproliferation attributes was carried out in late 1376. This study concluded that MSRs without denatured fuel probably would not be sufficiently prs%iferation-resistant for unrestricted worldwide deploy- ment. Subsequently, a more extensive study was undertaken at Oak Ridge National Laboratory (ORNL) to identify and characterize denatured molten- salt reactor (B8fSR) concepts for possible application in antiproliferation situations, This work began as part of the effort initiated by ERDA in response to a nuclear policy statement by President Ford on October 28, 1976;~ it was continued under the Nonproliferation ~lternative systems ~ssessment program (NASAP) $5 which was established in response to the Nuclear Power Policy Statement by President Carter on April 7, 1977,6 and TIE National ~nergy ~lan.’ The most recent development effort in this area was The DMSR is only one of a large number of reactors and associated fuel cycles selected for study under NASAP. However, it is also a member of a smaller subgroup that would operate primarily QII the ~ h-2”~ fuel cycle. Molten-salt X-~~C~QTS, in general, are particularly well suited to this fuel cycle because the fluid fuel and the associated core design tend to enhance neutron economy, which is particularly important for effective res~urce utilization. In addition, the ability of the ~ ~~lten fuel to re- tain plutonium (produced from neutron captures in the 238~ denaturant) in a relatively inaccessible form appears to contribute to the proliferation resistance of the system. The MSR concept also offers the possibility of system operation within a sealed containment from which IIQ fissile mate- ria% is removed and to which only denatured fuel or fertile material is added during the life of the plant. This combination of properties sug- gests the possibility of a fuel. cycle with a %ow overall cost and signifi- cant resistance to pr~liferati~n. e
3 The primary purpose of this study was to identify and characterize one or more BXSR concepts with antiproliferation attributes at least equivalent to those of a "conventional" light-water reactor (EWR) oper- ating on a once-through fuel cycle. The systens were also required to show an improvement over the LWR in terms of fissile and fertile resource utilfzation. Considerable effort was devoted to characterizing features of the concept(s) that would be expected to affect the assessment of their basic technological feasibility. These features included the estimated costs and time schedule for developing and deploying the reactors and their anticipated safety and environmental features. Although the older MSR studies were directed toward a high-perfor- manse breeder [and a reference molten-salt breeder reactor (MSBR) design8 was developed], the basic concept is adaptable to a broad range of fuel cycles, Aside from the breeder, these fuel cycles range from a plutonium burner ~ Q 2P 3 3 ~ production, through a DMSR with break-even breeding ana complex on-site f ission-product processing 99 to a denatured system with a 30-year fuel cycle that is once-through with respect to fission-product cleanup and fissile-material recycle. Of these, the Past one currently appears to offer the most advantages for development as a proliferation- resistant power SQUPC~. Consequently, this report is concentrated on a conceptual DMSR with a %@year fuel cycle and no special chemical processing for fission-product removal; other alternatives are considered only briefly- Section 2 contains a general description of the BMSR concept, with emphasis on those features that would be the same for all BMSR fuel cycles. Section 3 presents a more detailed treatment of the reference- concept DMSR covering the neutronic and thermal-hydraulic characteristics of the reactor core, fuel-salt chemistry, reactor materials, plant safety considerations, and system-specific environmental considerations. A general treatment of the antipsoliferation attributes of the concept is also included. The next section (Sect. 4) addresses potential alternatives to the reference concept and their perceived advantages and disadvantages. Section 5 addresses the commercialization considerations for DMSRs, in- cluding the perceived status, needs, and potential research, development,
Page 1: i *w . . I
Page 5 and 6: P iii CONTENTS ....................
Page 7: CONCEPTUAL DESIGN CHARACTERISTICS O
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Page 21 and 22: 15 Initial scoping studies showed t
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Page 25 and 26: :.:..y ,. . . w, E9 The code calcul
Page 27 and 28: h 3. Additional 238Y is added as re
Page 29 and 30: ...z,y .. . c .A, Table 9. Actinide
Page 31 and 32: ';x.w' .... Relative power distribu
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Page 35 and 36: 29 Table 16. Effect of neutron spec
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Page 39 and 40: .. . . . . . . . . . 'a# 0.1 0.2 0.
Page 41 and 42: 35 (Doppler effect), (2) changing t
Page 43 and 44: where Ci = relative delayed-neutron
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Page 55 and 56: 49 Variation of fuel composition wi
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;: c.... x. . w ..i 53 Table 26, St
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would be expected to have an equili
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.... . . . . :i ,.*,&s 57 Corrosion
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. ;;..,..I . . . . . . "LW 0 9 m -
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of the fissioned atom. ea ~arly ass
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63 n e results of a program of solu
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65 Table 27. Sources and rates Sf p
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operation of the PISRE. 67 analyses
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UP3/UF4 ratio would be possible. DM
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$ .p9 3.3.3.1 Preparation of initia
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73 Contamination of fuel by seconda
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*w- through the walls of the primar
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77 4 will apparently be necessary (
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such contamination. For a demonstra
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8% 3.4 Reactor Materials Although s
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: ....., . .’ -. ..WY used in the
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..... i.. ....., . .Yd 85 QRPIL-DWG
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... w . . .y and to salt containing
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With the different ~Bjectives of ns
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t.;$- ... quite different from thos
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93 during noma1 operation. Minor am
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............. . . .. . .=w as SSm,
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. . . :.w . . . . , 4. ALTERNATIVE
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.. . ... . . . . . . .. . . "W 3. T
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.,.. w .." by removing some uranium
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, . . . . . . , -c%w probably still
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and the ~ dvent (or carrier) salt a
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BO 7 TPle fuel makeup requirements
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5. CQPIPIERCIWLPZATION CONSIDERATIO
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11% At the close of MSRE operation,
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12YS 75 22”” 955 30” 1””
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115
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aa 7 some components in common with
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Ae eo un t No * Table 3%. Capital c
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- ,.,.,.,., p .. ..... . G is based
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123 Table 34. Summary of annual non
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...... . . . . ., x.:w* 4 i.. ... _
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E2 7 There are no detailed estimate
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129 particularly if such activities
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. .. . . . . . . . ;.:w b 133 the i
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_... ..&/ ; .; with reactor scram m
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137 includlng $900 million for the
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139 The authors want to acknowledge
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.:. _.i .. ..... -.=* 45. A. L. Mat
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...... &;.& 3 14 5 74. W, R. Grimes
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p 103. 6. Re Hudson 11, CQNCEPT-5 U
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Account No. -0- Three- digit digit
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.... ..
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156 77. He 6. MacPherson, Institute
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