-
Notifications
You must be signed in to change notification settings - Fork 10
/
ORNL-4622.txt
30321 lines (17604 loc) · 709 KB
/
ORNL-4622.txt
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
21
22
23
24
25
26
27
28
29
30
31
32
33
34
35
36
37
38
39
40
41
42
43
44
45
46
47
48
49
50
51
52
53
54
55
56
57
58
59
60
61
62
63
64
65
66
67
68
69
70
71
72
73
74
75
76
77
78
79
80
81
82
83
84
85
86
87
88
89
90
91
92
93
94
95
96
97
98
99
100
101
102
103
104
105
106
107
108
109
110
111
112
113
114
115
116
117
118
119
120
121
122
123
124
125
126
127
128
129
130
131
132
133
134
135
136
137
138
139
140
141
142
143
144
145
146
147
148
149
150
151
152
153
154
155
156
157
158
159
160
161
162
163
164
165
166
167
168
169
170
171
172
173
174
175
176
177
178
179
180
181
182
183
184
185
186
187
188
189
190
191
192
193
194
195
196
197
198
199
200
201
202
203
204
205
206
207
208
209
210
211
212
213
214
215
216
217
218
219
220
221
222
223
224
225
226
227
228
229
230
231
232
233
234
235
236
237
238
239
240
241
242
243
244
245
246
247
248
249
250
251
252
253
254
255
256
257
258
259
260
261
262
263
264
265
266
267
268
269
270
271
272
273
274
275
276
277
278
279
280
281
282
283
284
285
286
287
288
289
290
291
292
293
294
295
296
297
298
299
300
301
302
303
304
305
306
307
308
309
310
311
312
313
314
315
316
317
318
319
320
321
322
323
324
325
326
327
328
329
330
331
332
333
334
335
336
337
338
339
340
341
342
343
344
345
346
347
348
349
350
351
352
353
354
355
356
357
358
359
360
361
362
363
364
365
366
367
368
369
370
371
372
373
374
375
376
377
378
379
380
381
382
383
384
385
386
387
388
389
390
391
392
393
394
395
396
397
398
399
400
401
402
403
404
405
406
407
408
409
410
411
412
413
414
415
416
417
418
419
420
421
422
423
424
425
426
427
428
429
430
431
432
433
434
435
436
437
438
439
440
441
442
443
444
445
446
447
448
449
450
451
452
453
454
455
456
457
458
459
460
461
462
463
464
465
466
467
468
469
470
471
472
473
474
475
476
477
478
479
480
481
482
483
484
485
486
487
488
489
490
491
492
493
494
495
496
497
498
499
500
501
502
503
504
505
506
507
508
509
510
511
512
513
514
515
516
517
518
519
520
521
522
523
524
525
526
527
528
529
530
531
532
533
534
535
536
537
538
539
540
541
542
543
544
545
546
547
548
549
550
551
552
553
554
555
556
557
558
559
560
561
562
563
564
565
566
567
568
569
570
571
572
573
574
575
576
577
578
579
580
581
582
583
584
585
586
587
588
589
590
591
592
593
594
595
596
597
598
599
600
601
602
603
604
605
606
607
608
609
610
611
612
613
614
615
616
617
618
619
620
621
622
623
624
625
626
627
628
629
630
631
632
633
634
635
636
637
638
639
640
641
642
643
644
645
646
647
648
649
650
651
652
653
654
655
656
657
658
659
660
661
662
663
664
665
666
667
668
669
670
671
672
673
674
675
676
677
678
679
680
681
682
683
684
685
686
687
688
689
690
691
692
693
694
695
696
697
698
699
700
701
702
703
704
705
706
707
708
709
710
711
712
713
714
715
716
717
718
719
720
721
722
723
724
725
726
727
728
729
730
731
732
733
734
735
736
737
738
739
740
741
742
743
744
745
746
747
748
749
750
751
752
753
754
755
756
757
758
759
760
761
762
763
764
765
766
767
768
769
770
771
772
773
774
775
776
777
778
779
780
781
782
783
784
785
786
787
788
789
790
791
792
793
794
795
796
797
798
799
800
801
802
803
804
805
806
807
808
809
810
811
812
813
814
815
816
817
818
819
820
821
822
823
824
825
826
827
828
829
830
831
832
833
834
835
836
837
838
839
840
841
842
843
844
845
846
847
848
849
850
851
852
853
854
855
856
857
858
859
860
861
862
863
864
865
866
867
868
869
870
871
872
873
874
875
876
877
878
879
880
881
882
883
884
885
886
887
888
889
890
891
892
893
894
895
896
897
898
899
900
901
902
903
904
905
906
907
908
909
910
911
912
913
914
915
916
917
918
919
920
921
922
923
924
925
926
927
928
929
930
931
932
933
934
935
936
937
938
939
940
941
942
943
944
945
946
947
948
949
950
951
952
953
954
955
956
957
958
959
960
961
962
963
964
965
966
967
968
969
970
971
972
973
974
975
976
977
978
979
980
981
982
983
984
985
986
987
988
989
990
991
992
993
994
995
996
997
998
999
1000
ORNL-4622
R UC-BO Reactor Technology
MOLTEN-SALT REACTOR PROGRAM“ S s
SEM!ANNUAL PROGRESS REPORT
For Penod Endmg August 31 1970
OAK RIDGE NATIONAI. LABORATORY |
- ' operated by S Lo
umow C'ARBlDE CORPORATION -
S for the L T e
u s ATOMIC ENERGY commssnon
' pySTRIBUTIOR OF THIS DOCUMENT IS UNLOUITED -
e
e Prmted in the United ¢ States ot Amenca Avallable from L o
' S _National Technical Informatlon Service .~ - i
US Department ‘of Commerce Sprm'fneld V|rg|n|a 22151 S L
“Price: Printed Copv $3.00; Microfiche $065 Lo B R I i
- T This report was prepared as ‘an_ account of work sponsored bv the Umted Fs
5 vol. ) States Government: “Neither . the - United - States nor the United States Atomic | i
- Energv Commlssmn nor any of- the:r empioyees nor_any of their ‘contractors, - LT _ "-"Y;
p -subcontractors “or the:r emplovees makes any. warranty, express or implied, or | .. - ST e Ei'j’
| IR assumes any - legal lnabuletv or. respons:bmty for the accuracv. completeness or ',i e “ .
T e ;usetulness ‘of - any information, - apparatus, product .or process disclosed, or_ I '
- rrepresents that its use would not :nfnnge prlvately owned nghts. L e e
A
-
ORNL-4622
UC-80 — Reactor Technology
Contract No. W-7405-eng-26
MOLTEN-SALT REACTOR PROGRAM
SEMIANNUAL PROGRESS REPORT
FOR PERIOD ENDING AUGUST 31, 1970
M. W. Rosenthal, Program Director
R. B. Briggs, Associate Director
P. N. Haubenreich, Associate Director
LEGAL NOTICE
This report was prepared as an account of work
sponsored by the United States Government. Neither
the United States nor the United States Atomic Energy
Commission, nor any of their employees, nor any of
their contractors, subcontractors, or their employees,
makes any warranty, express or implied, or assumes any |
legal liability or responsibility for the accuracy, com-
pleteness or usefulness of any information, apparatus,
. { product or process disclosed, or represents that its use
‘ would not infringe privately owned rights,
JANUARY 1971
OAK RIDGE NATIONAL LABORATORY
Oak Ridge, Tennessee
operated by
UNION CARBIDE CORPORATION
for the
U.S. ATOMIC ENERGY COMMISSION
FISTRIBUTION OF TEIS DOCUMENT IS UNLIXITED
.,'
This report is one of a series of periodic reports in which we describe the progress of the program. Other reports
issued in this series are listed below.
ORNL-2474 Period Ending January 31, 1958
ORNL-2626 Period Ending October 31, 1958
ORNL-2684 Period Ending January 31, 1959
ORNL-2723 Period Ending April 30, 1959
ORNL-2799 Period Ending July 31, 1959
ORNL-2890 Period Ending October 31, 1959
ORNL-2973 Periods Ending January 31 and April 30, 1960
ORNL-3014 Period Ending July 31, 1960
ORNL-3122 Period Ending February 28, 1961
ORNL-3215 Period Ending August 31, 1961
ORNL-3282 Period Ending February 28, 1962
ORNL-3369 Period Ending August 31, 1962
ORNL-3419 Period Ending January 31, 1963
ORNL-3529 Period Ending July 31, 1963
ORNL-3626 Period Ending January 31, 1964
ORNL-3708 Period Ending July 31, 1964 5.
ORNL-3812 Period Ending February 28, 1965 %
ORNL-3872 Period Ending August 31, 1965 -
ORNL-3936 Period Ending February 28, 1966 -
ORNL4037 Period Ending August 31, 1966
ORNL4119 Period Ending February 28, 1967
ORNL4191 Period Ending August 31, 1967
ORNL4254 Period Ending February 29, 1968
ORNL4344 Period Ending August 31, 1968
ORNL4396 Period Ending February 28, 1969
ORNL4449 Period Ending August 31, 1969
ORNL4548 Period Ending February 28, 1970
» F,. (o
w
Contents
INTRODUCTION ..ttt ettt et ettt e s ettt et ee et e
SUMM A RY .o e e e e e e
PART 1. MOLTEN-SALT REACTOR EXPERIMENT
1. MSRE OPERATIONS .. ... ittt ittt ittt ettt ttaette e s tee et iaeeiaennn, 1
1.1 Chronological ACCOUNL . ... ...ttt ittt it ittt e ittt ettt et et te e ea e 1
1.2 Operations Analysis .. ....uuitnntie ittt ittt it e et e 2
1.2.1 Gamma Spectrometry ... . ..ouuttn it tnarcee ettt e et e 2
1.2.2 Noble-Metal Migration . ...........c.iurininininiit ittt et eeeaaennn. 2
1.2.3 Radiation Heating .. ........ .0ttt ittt it ettt iinennns 4
=\ 1.2.4 Fluorine Evolution form FrozenSalt ........... .. ... . .. ... ... ... ......... 4
-y
* 2. COMPONENT DEVELOPMENT . ........oueuse e TR 6
2.1 Freeze-Flange Thermal Cycle Test ... ... ... .. ceiiiienn i it ieteeeeenan 6
2.2 PUMIDS ittt e e e e e e 6
2201 Mark 2Fuel Pump ... o i e e e 6
2.2.2 OilPump Endurance Test ... ... ... oottt it et et 6
PART 2. MSBR DESIGN AND DEVELOPMENT
3 DESIGN o e e e et e 8
3.1 Single-Fluid 1000-MW(e) MSBR Design Study Report ...... ... 8
3.2 Molten-Salt Demonstration Reactor Design Study ..................... et 8
320 Introduction . ...ttt ittt et et e ettt 8
3.2.2 General DesCription ... ...ttt ettt ettt it e, 9
3.2.3 Buildingsand Containment . ............c.iiiiuriinerrnnnnerrnneneennnnennnnn. 9
324 ReaCtOr ..\ttt ittt ittt et e e e e e 16
3.2.5 Primary Heat Exchangers ...................... e 18
3.2.6 Primary Drain Tanks ... ........couiiiiiiiiii ittt it iiniteennnnnnnnens 18
3.2.7 Fuel Salt Storage Tank ........ ..ottt 21
3.28 Discard Salt Storage Tanks . ... ...ttt ittt it e e 22
T S 3.2 Steam System . . . i e et it et 24
PQ“ 3.3 - Afterheat Temperatures in Empty MSBR Heat Exchangers . ..............0.ccoiiunninnean... 24
= 3.4 Industrial Study of 100-MW(e) Molten-Salt Breeder Reactor .....................ccuouun... 25
iii
iv
4, REACTOR PHY SICS ...ttt ittt et iaeaeeetoatetraeeansnsarssesseenasneonens 26
4.1 Physics Analysis of MSBR ....... ..ot e 26
4.1.1 Fixed-Moderator Molten-Salt Reactor ............ ... ... i, 26
42 MSRExperimental Physics . . ... ..ottt it i ittt it 31
" 4.2.1 233U Capture-to-Absorption Ratio in the Fuel of the MSRE ... ...................... 31
5. SYSTEMS AND COMPONENTS DEVELOPMENT .. ... ... . ittt ieieinnenns 35
5.1 GaseousFissionProduct Removal ......... ... . i ittt iiiiinennens, 35
5.1.1 GaS SeParator . ..ottt ettt 35
5.1.2 Bubble Generator . ... ..o ittt it i it e e et 38
5.1.3 WaterTestLoop .. ................... e e ettt e et 39
5.2 Molten-Salt Steam Generator ... ....c.cittet ittt ae ittt ietee ettt 39
5.2.1 Steam Generator Industrial Program ............. ... it iiiiiiiiriniiinnnens 39
5.2.2 Steam Generator Tube Test Stand (STTS) ... ....ccoviiniiiiiiiiiiinnnns v... 40
5.2.3 Molten-Salt Steam Generator Technology Loop(SGTL) .........ccciveiiinnnnnnn. 40
5.2.4 Development Basis for Steam Generators Using Molten Salt as the Heat Source .......... 40
5.3 Sodium Fluoroborate Test LOOp .. ....... ..ottt it iiinennarannnnns 41
5.3.1 Water Addition Test ... ... .. it i e et ieiea e, 41
5.3.2 Conclusions ............covnunnnn. P 44
54 MSBR PUMDS . ...t i e ettt e e 44
54.1 MSBE Salt Pump Procurement . .. ... ..ottt ittt ieiinnninneennnnns 44
542 MSBESaltPump TestStand . .......... ... ittt 45
SA3 ALPHA PUmp .. .oi ittt ittt ittt e e et e iesearenearoneaaenanaeans 45
55 RemoteWelding . ....... ... ... ... .. i, et et 45
5.5.1 Operational Prototype Equipment .. ... ... ... .. . it iienenannn 45
55.2 CuttingDevelopment ... .......oiuiuiririin i inerennenenoeneonsenneanennnn 45
5.5.3 Welding Development . ..o .. ... it iit it ittt iaeenrceeeevanenncacanaaenanan 47
554 External Interest . .........ouuiiuiiintniineenentieeeateararnaeaaaneaaa, 49
5.5.5 Design Studies ... ... .. i i i i i i e et e e e 49
6. MSBR INSTRUMENTATION AND CONTROLS . ... ...ttt it ieeeinenaaannnenns 51
7. HEAT AND MASS TRANSFER AND THERMOPHYSICAL PROPERTIES ........................ 53
7.1 Heat Transfer ... ... .. it ittt ettt ettt tteaetaaaananeaaaan, 53
7.2 Thermophysical PIOPETtIES - . . . ..o ottt ettt e et e ettt ettt eeeaieeeenns 54
7.3 Mass Transfer to Circulating Bubbles . ... ... .. ... .. .. .. . i i iiiiiiannnnn 57
PART 3. CHEMISTRY
8. FISSIONPRODUCT BEHAVIOR .. ... iiiiiiiittiineeteeernnseneerneneoacooacananenaenas 60
8.1 Noble Metal Fission Product BEHAVIOT .. .o e ottt e et e e e iee e e e et 60
8.2 Short-Term Fission Product Deposition Tests .......... oottt iniirtiecvannnnneonnns 66
8.3 Fission Product Deposition on the Fifth Set of Graphite and Hastelloy N Samples
fromthe MSRE COTE .....vvtiiit it tiiiietieiieeennaeenereananoonranasanaannnns 68
8.3.1 Effect of Surface Roughness ........................: e et eaeeaaeeaaaan, 68
8.3.2 Effect of Flow ConditionsonDeposition......... ...ttt iirieennannn. 68
8.3.3 Comparison of Deposition of Noble Metal Fission Products on Graphite A
and Hastelloy N ... ottt e e i ittt e tetnesaanennenns 69
C
o
.
t) 1 ar
[
$)
9.
8.3.4 Concentration Profiles of Fission Productsin Graphite .............................
8.3.5 Electron Microscope Examination of Particles in the Gas Phase Above Fuel Salt
Mthe MSRE ... . . i i i i i it e i et e
8.4 A Possible Origin of Smokes and Mists Emitted by MSREFuel .............................
8.5 Synthesis of Niboium Fluorides . ...........o ittt ittt et e eeeeenn
8.6 Molybdenum and Niobium Fluoride SolutionsinMoltenLi;BeF, ..........................
8.7 MassSpectroscopyofNiobiumFluorides...............................; ..............
8.8 Raman Spectra of Crystalline MoF; and )
PROPERTIES OF THE ALKALI FLUOROBORATES . .. ... ... . . i i,
9.1 Solubility of BF; Gasin Fluoride Melts . . .. ... o ittt et e e,
9.2 Studies of Hydrogen Evolution and Tritium Exchange in Fluoroborate Coolant Salt . ............
9.3 Reaction of Water with the Sodium Fluoride—Sodium Tetrafluoroborate Eutectic ..............
9.4 Raman Spectra of Molten NaBF,; and NaF-NaBF, t0606°C .. ...........uur e,
9.5 Electronic Polarizability of the Tetrafluoroborate Ion and Various Gaseous Halide
Molecules as Compared with the Free (Gaseous) HalideIons ..............................
9.5.1 The Polarizability of the Tetrafluoroboratelon ................. S
9.5.2 Test of a Recent Theoretical Set of [on Polarizabilities .............................
9.5.3 Effect of Interaction Between Halide Ions on Polarizability ... .......................
10. PHYSICAL CHEMISTRY OF MOLTEN SALTS . ... i i i e i e e
¥ i
A"
7’
<7
)
10.1 Liquidus Temperature of the Salt Mixture LiF-BeF, -ThF, (70-15-15mole %) .................
10.2 Equilibrium Phase Relationships in the System LiF-BeF,CeF3 ... ......ooovione ...
10.3 Phase Relationships in the System CeFy-ThF, ... ... ... o . i i
104 The Instability of PuOF and the Solubility of Pu(IIl) in MSBR Solvent Salt ...................
10.5 Oxide Chenustry of Protactinium in MSBR Fuel Solvent Salt ..............................
10.6 Entropnes of Molten Salts . .. ... ... i it i e e e e
10.7 All-Metal Conductance Cell for Use inMolten Salts . . ... . ... .. e
10.8 Glass Transition Temperatures in NaF-BeFy MiXtures .. .......ououeeereenenreneneenannnn..
10.9 Correlations of Electrical Conductivities in Molten LiF-BeF,-ThF,; Mixtures ..................
10.10 Coordination Effects on U(IV) Spectrain FluorideMelts ...................ccoovvinnn...
11. CHEMISTRY OF MOLTEN-SALT lfiEACTOR FUEL TECHNOLOGY. ceveine i .k ..... .
12.
11.1 Remova! of Fluoride from Molten LiCl .. ............. e S .....
11.2 Distribution of Sodium and Potass:um in the Metal Transfer Process ét650°C ................
11.3 Removal of Lanthanides from Lithium Chlorides on Zeolites .......... e
11.4 Zirconium Platinide as a Removal Agent for Rare Earths . .. .. - e . Dt e
DEVELOPMENT AND EVALUATION OF AN ALYTICAL METHODS FOR
MOLTEN-SALTREACTORS ........ciiiiieninnnnnnns e [
12.1 - Spectral Studies of MSRE Fuel Samp]es....‘. e e,
12.2 Spectral Studies of Actinide Ions in Molten Fluoride Salts . . ..................... e
12.3 Reference Electrode Studiesin MoltenFluorides .. .......... ... ... ... i L.
13.
14.
15.
vi
12.4 Kinetic Studies on the Ni/Ni(II) Couple in Molten LiF-BeF,-ZxF, .. ..... ... ... .. .. ..., . 115
12.5 Analytical Studies of NaBF, CoolantSalt ................... P 116
12.6 1In-Line Chemical ANalyses .. .........ovueeirruennrunnnansnssserseeeeeenannaennonnnns 118
PART 4. 'MATERIALS DEVELOPMENT
EXAMINATION OF MSRECOMPONENTS . ... ..ottt e eeeeaeeraaiaaeaas 119
13.1 Examination of Tubing from the MSRE Radiator ............. ... ... il e 120
13.2 Examination of Thermocouple Wells from the MSRE CoolantCircuit . ....................... 126
GRAPHITE STUDIES .. . ...\t tttitniiaeeeaaseeeaneeeeeeeeinnnninens e, 134
14.1 Procurement of New Gradesof Graphite ........ ... it 134
14.2 General Physical Property Measurements . .........coiiuiiiiiiiiiiiniiniiiirnneannnnnens 135
14.3 Graphite Fabrication: Chemistry ........... .. .ot e aareea 135
144 Measurement of Thermal Conductivity of Graphite ......... et ee et eeaeintea e 141
145 X Ray Studies ........cctiritinniiiiirateenananesesacaeransnosnnanasssssasanans 141
14,6 Reduction of Graphite Permeability by Pyrolytic CarbonSealing ........................... 142
14,7 HFIR Irradiation Program e e e e e e 145
HASTELL DY N . ittt ittt teteaeaeaoasaeneansarurosnseeonasonsensasesaneananan 149
15.1 Influence of Titanium on the Strengthening of an Ni-Mo-Cr Alloy . ...... e eseaee e 149
15.2 Effect of Composition on the Postirradiation Mechanical Properties
of Modified Hastelloy N .. ............... e e et e, 153
15.3 Comparison of Laboratory and Commercial Heats of Modified Hastelloy N ................... 156
154 Effect of Prior Aging on Irradiation Damage at 760°C ... .........utururneereneneeennnn. 159
15.5 Weldability of Commercial Alloys .............. oot 160
15.6 Mechanical Properties of Unirradiated Commercial Modified Alloys ......................... 161
15.7 - Electron Microscope Studies ..........iiuiiiiiiiiiii ittt e 164
158 Comrosion Studies . .. .. .. iiiiit ittt tnneee e aeaar ettt 165
15.8.1 Results of Thermal Convection Loop Tests .. ......iitinttnieinrinrnnenneenannns 166
15.8.2 Fertile-Fissile Salt . ... .. ... i i i it i i iie i iiireeanannnnn 168
1583 Blanket Salt . ..... ...ttt ittt it i eaeiasaans 168
1584 Coolant Salt ...... ...ttt ittt ittt ia et areeianennnsannns . 168
1585 Summarny ....... ittt ittt ittt [ 171
159 Forced-Convection Loop, MSR-FCL-1 ... ..o uiiiiniiii ittt it iierennannncnnnns 173
1591 Operations ..........oiriiioiiiiininnnneernonscnncesarssanssnna et 173
15.9.2 Cold Finger Corrosion Product Trap ......... ..ottt 174
15.9.3 Metallurgical Analysis ................... e 175
15.9.4 Forced-Convection Loop MSR-FCL-2 .......... et r e eeereeeeeeieaeae e 176
15.10 Corrosion of Hastelloy NinSteam . ............cooviiiiiiiiiiann.. e 178
15.11 Evaluation of Duplex Tubing for Use in Steam Generators . ................cooceouueeenenn 181
ik
™)
"/
‘\4‘/\“
£ o i"'/ 4
-/
vii
16. SUPPORT FOR CHEMICALPROCESSING ...............ciiivinn.. P 184
16.1 Construction of a Molybdenum Reductive Extraction TestStand ........................... 184
16.2 Fabrication Development of Molybdenum Components ............ ... it iuiinenennnnn. 184
16.3 Weldingof Molybdenum .............. ittt it ia it iaaaanann 185
164 Development of Bismuth-Resistant Filler Metals for Brazing Molybdenum .................... 189
16.5 Compatibility of Structural Materials withBismuth . ............. ... ... .. .. ot 189
16.6 Chemically Vapor Deposited Coatings . ........ ...ttt iiiiiniiinnnnennnas 195
16.6.1 TungstenCoatings ......... ...t tniunmiii ittt iiearanaanaannn 195
16.6.2 Molybdenum Coatings .. ....... ...t iiiitiittinernnernerneeanenneeannenans 197
16.7 Molybdenum Deposition fromMoFg . ... ... i 197
17. FLOWSHEET ANALY SIS .. i it i ittt it ittt ittt ca i easinenans 199
17.1 Protactinium Isolation Using Fluorination—Reductive Extraction ..................... ... ... 200
17.2 Rare-Earth Removal Using the Metal Transfer Process . ........... ... oo, 201
17.3 Removal of Uranium from Fuel Salt by Oxide Precipitation ................. ... ... ....... 202
18. MEASUREMENT OF DISTRIBUTION COEFFICIENTS IN MOLTEN-SALT-METAL SYSTEMS ....... 204
18.1 Metal Transfer ProcessStudies .. ... ... . i ittt iiienennn 204
18.2 Solubility of Neodymium in Li-Bi-ThSolutions . .. ... ... ...t innnns 207
18.3 Solubility of LaOClin Molten LiCl. . . ... ..ottt ittt it e ie e ieeceeeeeaanannns 208
18.4 Oxide Precipitation Studies. . ... ... ottt i e e 208
PART 5. MOLTEN-SALT PROCESSING AND PREPARATION
19. ENGINEERING DEVELOPMENT OF PROCESS OPERATIONS . ... ... ... .. .. i, 211
19.1 Reductive Extraction Engineering Studies .......................... P 211
19.2 Design of a Processing Materials Test Stand and the Molybdenum Reductive
Extraction Equipment . ........ ... . e e 212
19.3 Contactor Development: Pressure Drop, Holdup, and Flooding in Packed Columns ............. 213
19.4 Contactor Development: Axial Mixing inPackedColumns ..................... ... ... ..., 215
19.5 Axial Mixing in Open Bubble COIUMNS . . ..\ evvtveeeeneennneaeeeeaaaannn. e 216
19.6 Demonstration of the Metal Transfer Process for Rare-Earth Removal . . ........ e e 217
19.7 Development of a Frozen-Wall Fluorinator .......................... fheeer i 219
19.8 Electrolytic Cell Developmeht ............... et e et 220
20. CONTINUOUS SALT PURIFICATION SYSTEM ... ... ..ttt i iiiieinenn, e 224
"
o
Introduction
The objective of the Molten-Salt Reactor Program is
the development of nuclear reactors which use fluid
fuels that are solutions of fissile and fertile materials in
suitable carrier salts. The program is an outgrowth of
the effort begun over 20 years ago in the Aircraft
Nuclear Propulsion program to make a molten-salt
reactor power plant for aircraft. A molten-salt reactor —
the Aircraft Reactor Experiment — was operated at
- ORNL in 1954 as part of the ANP program.
Our major goal now is to achieve a thermal breeder
reactor that will produce power at low cost while
simultaneously conserving and extending the nation’s
fuel resources. Fuel for this type of reactor would be
233yUF, dissolved in a salt that is a mixture of LiF and
BeF,, but it could be started up with >?>5U or
plutonium. The fertile material would be ThF, dis-
" solved in the same salt or in a separate blanket salt of
similar composition. The technology being developed
for the breeder is also applicable to high-performance
converter reactors.
A major program activity until recently was the
operation of the Molten-Salt Reactor Experiment. This
reactor was built to test the types of fuels and materials
that would be used in thermal breeder and converter
reactors and to provide experience with the operation
and maintenance of a molten-salt reactor. The MSRE
operated at 1200°F and produced 7.3 MW of heat. The
initial fuel contained 0.9 mole % UF,, 5% ZrF,;, 29%
BeF,, and 65% "LiF, a mixture which has a melting
point of 840°F. The uranium was about 33% 235U.
The fuel circulated through a reactor vessel and an
external pump and heat exchange system. All this
equipment was constructed of Hastelloy N, a nickel-
molybdenum-iron-chromium alloy with exceptional re-
sistance to corrosion by molten fluorides and with high
strength at high temperature. The reactor core con-
tained an assembly of graphite moderator bars that
were in direct contact with the fuel.
Heat produced in the reactor was transferred to a
‘coolant salt in the primary heat exchanger, and the
coolant salt was pumped through a radiator to dissipate
the heat to the atmosphere.
ix
Design of the MSRE started in the summer of 1960,
and fabrication of equipment began early in 1962. The
reactor was taken critical on June 1, 1965. Operation at
low power begin in January 1966, full power was
reached in May, and sustained power operation was
begun in December. In September 1967, a run was
begun which continued for six months, until terminated
on schedule in March 1968.
Completion of this six-month run brought to a close
the first phase of MSRE operation, in which the
objective was to demonstrate on a small scale the
attractive features and technical feasibility of these
systems for civilian power reactors. We believe this
objective has been achieved and that the MSRE has
shown that molten-fluoride reactors can be operated at
temperatures above 1200°F without corrosive attack on
either the metal or graphite parts of the system, that
the fuel is completely stable, that reactor equipment
can operate satisfactorily at these conditions, that
xenon can be removed rapidly from molten salts, and
that, when necessary, the radioactive equipment can be
repaired or replaced.
The second phase of MSRE operation began in
August 1968, when a small facility in the MSRE
building was used to remove the original uranium
charge from the fuel salt by treatment with gaseous F,.
In six days of fluorination, 221 kg of uranium was
removed from the molten salt and loaded onto ab-
sorbers filled with sodium fluoride pellets. The decon-
tamination and recovery of the uranium were very
good. _
After the fuel was processed, a charge of 223U was
added to the orginal carrier salt, and in October 1968
the MSRE became the world’s first reactor to operate
on 233 U. The nuclear characteristics of the MSRE with
the 222U were close to the predictions, and, as
expected, the reactor was quite stable.
In September 1969, small amounts of PuF; were
added to the fuel to obtain some experience with
plutonium in a molten-salt reactor. The MSRE was shut
down permanently- December 12, 1969, so that the
funds supporting its operation could be used elsewhere
in the research and development program. An in-
spection of parts of the reactor is under way, and
samples of materials are being removed for examina-
tion.
Most of the Molten-Salt Reactor Program is now
being devoted to the technology needed for future
molten-salt reactors. Conceptual design studies are
being made of breeder reactors, and the program
includes work on materials, on the chemistry of fuel
and coolant salts, on processing methods, and on the
development of components and systems. ,
Because of limitations on the chemical processing
methods available at the time, until three years ago
most of our work on breeder reactors was aimed at
two-fluid systems in which graphite tubes would be
used to separate uranium-bearing fuel salts from tho-
rium-bearing fertile salts. In late 1967, however, a.
one-fluid breeder became feasible because of the de-
velopment of a process that uses liquid bismuth to
isolate protactinium and remove rare earths from a salt
that also contains thorium. Our studies showed that a
one-fluid breeder based on the new process can have
fuel utilization characteristics approaching those of our
two-fluid designs. Since the graphite serves only as
moderator, the one-fluid reactor is more nearly a
scaleup of the MSRE. These advantages caused us to
change the emphasis of our program from the two-fluid
to the one-fluid breeder; most of our design and
development effort is now directed to the one-fluid
system.
¥
(rh
o
L)
v
N N
| Summary
PART 1. MOLTEN-SALT REACTOR EXPERIMENT
1. MSRE Operations
The MSRE remained shut down, with the salt frozen
in the tanks and the primary systems intact, awaiting
the scheduled examinations. Procedures were readied,
and tools were designed and procured. Equipment to
become surplus was identified. Specimens of radiator
tubes and thermocouple wells in the coolant salt line
were cut out for examination.
Continuing analysis included the gamma spectrometry
data and other evidence of noble-metal fission product
behavior. Comparison of observed deposition with a
model based on conventional mass transfer theory
showed reasonable agreement, suggesting that the noble
metals quickly migrate, as extremely small particles, to
solid or gas interfaces.
Examination of data from previous experiments
indicates that fluorine evolution from the frozen fuel
can be easily controlled.
2. Component Development
Component development for the MSRE was con-
cluded with the termination of endurance tests of the
Mark 2 fuel pump (at 16,680 hr of salt circulation) and
a lubricating oil pump (at 58,900 hr).
PART 2. MSBR DESIGN AND DEVELOPMENT
3. Design
‘A comprehensive report on the conceptual design
study of a 1 OOO-MW(e) molten-salt bréeder reactor
power station is near completion, and copies of the
final draft have been cn'culated at ORNL for comment
and review.
‘Since the first molten-salt power reactor could be a
demonstration unit in the 100-to-300-MW(e) size range,
a conceptual design study has been initiated to investi-
x1
gate the features of such a plant of 300-MW(e) size. The
plant being studied is a converter that uses periodic
discard to rid the fuel salt of fission product poisons
rather than awaiting development of continuous chem-
ical processing systems needed for breeders.
The basic arrangement of the demonstration unit is
much the same as that of the 1000-MW(e) MSBR. It is
of the single-fluid type, with ? LiF-BeF, -ThF 4-UF, fuel
salt that flows upward through an unclad graphite-
moderated and -reflected core. The heat is transferred
in primary heat exchangers to a circulated sodium
fluoroborate coolant salt, which transports the energy
to supercritical-pressure steam generators and to steam
reheaters. The steam turbine power cycle has a 1000°F
throttle temperature, reheat to 1000°F, a regenerative
feedwater heating system, and special provisions to
attain high-temperature feedwater. The overall plant
thermal efficiency is about 44%.
The Hastelloy N reactor vessel is about 26 ft in
diameter X 32 ft high with a wall thickness of about 2
in. The reactor core design is very similar to that used
successfully in the MSRE, consisting of 4 X 4 in.
vertical graphite elements. Flow passages are sized to
provide the desired salt-to-graphite ratios in the various
regions of the core and an essentially uniform salt
temperature rise through all regions. The salt enters at
about 1050°F and leaves at 1300°F. The reactor has a
conversion ratio above 0.8. The relatively low power
density of about 10 W/ cm?® permits the graphite to last
the lifetime of the plant and ehmmates the need for a
removable head on the reactor vessel.
The primary heat exchangers are of the U-shell type
with '%-in.-OD U-tubes. The maintenance arrangement
is to cut the heads from the shell, operating from the
side . through plugged openings in the cell wall, to
expose the tube sheets for location and plugging of
faulty tubes.
During normal operation the fuel salt circulating
pumps overflow to a drain tank, and fission product
gases extracted from the salt circulating system pass
through the drain tank for holdup and decay. The drain
tank is cooled by a natural circulation NaK system
which rejects heat to an elevated pool of water. Double
barriers, with radiant heat transfer across gas-filled
annuli, are used in both the drain tank and the water
pool to assure isolation of the NaK from the fuel sa]t
and the water,
A fuel salt storage tank is provided for the fuel salt in
event repairs are needed on the primary drain tank. This
xii
storage tank can also be used for the fluorination
process to recover the uranium from a spent fuel salt
charge and for addition of UF¢ and H, to new carrier
salt for reconstituting a fresh charge. Storage facilities
for spent carrier salt are provided in the reactor
building. It is estimated that the salt charge would have
to be replaced about three tlmes during the 30-year
lifetime of the plant.
Temperatures producéd by the heat from decay of
deposited fission products were calculated for primary
heat exchangers of sizes from 94 to 281 MW, and the
results were extrapolated to the 563-MW size of the
reference design. The calculations assumed that the
exchangers were drained of primary and secondary salts
and that the heat was radiated to surroundings at
1000°F. The maximum temperatures were found to be
in the range of 1700 to 2100°F for a 141-MW
MSBE-size exchanger, but extrapolated results indicated
that the temperatures in a 563-MW MSBE exchanger
would be unacceptably high,
Preparations were nearly completed for issuance of a
request for proposals for an industrial study of a
1000-MW(e) MSBR.
4, Reactor Physics
We have continued our investigation of MSR core
configurations having sufficiently low power density
that the graphite moderator should not require replace-
ment over the life of the plant (i.e., 30 years) and have
considered a power level of 300 MW(e) as well as 1000
MW(e). Fuel-cycle analyses have been performed for
such cores over a range of thorium concentrations from
10 to 18 mole %. With continuous processing, as for the
reference single-fluid MSBR, we find breeding ratios
comparable with that of the reference breeder (1.06)
but somewhat higher specific fuel inventories, associ-
ated with the lower power density. Consequently, the
fuel yields and conservation coefficients are lower than
for the reference breeder. Fuel-cycle cost is about 0.7
mill/kWhr(e) for the 1000-MW(e) reactor and about 1.1
to 1.2 mills/kWhi(e) for the 300-MW(e) reactor. Opti-
mum thorium concentrations, when optimized in terms
of the fuel conservation coefficient, are a little higher
than for the reference breeder, that is, 15 to 17 mole %,
as compared with 12%.
Calculations have been performed for a series of batch
fuel cycles, in which the fuel salt, including thorium
and plutonium, is assumed to be discarded after several
years of operation, with only the uranium carried over
to the subsequent cycle. These reactors, of course, have
conversion ratios substantially below unity. For ex-
ample, a 300-MW(e) reactor with 30-year graphite life,
plutonium feed, and an 8-year batch processing cycle |
would have a lifetime-average conversion ratio of about
0.84 and a fuel-cycle cost of 0.8 to 0.9 mill/kWhr(e).
The same reactor could become a breeder by addition
of appropriate chemical processing equipment.
The measurement of the capture-to-fission cross-
section ratio, @, for 233U in the MSRE circulating fuel
salt has been successfully completed. The measured
value is 0.1233 * 0.0039. The corresponding calculated
value using the cross sections and computational meth-
ods used in predicting MSBR performance is 0.1226.
The uncertainty of +3,2% in the measured value, if
applicable to the calculated value for the MSBR,
corresponds to an uncertainty in breeding ratlo of
approximately +0.008.
S. Systems and Cbmponents Development
The gas removal efficiency of the centrifugal gas
separator while operating with circulating water con-
taining various concentrations of entrained air bubbles
was found to be affected most by an instability of the
vortex in front of the recovery hub. Since the initial
separation of the gas bubbles to the central vortex
appears to be virtually 100%, the gas takeoff port is
being redesigned in an attempt to improve the vortex
stability. The bubble generator designs previously tested
have shown a tendency for the liquid flow to oscillate
along the trailing edge of the generator and for the gas
flow to distribute itself unevenly among the gas feed
holes. Testing has shown a design resembling a jet pump
to have some promise, and testing is continuing. A
water test loop which will permit testing full-sized
MSBE bubble generators and gas separators has been
designed and is being fabricated.
The program for obtaining design studies of molten-
salt-heated steam generators from industrial firms has
been rewritten and now consists of four tasks: The first
two will result in conceptual designs of steam genera-
tors for the ORNL 1000-MW(e) reference design and
for an alternate steam cycle to be suggested by the
industrial firm. The third task will show how the
i
o
"y oy
w, Ny
designer would propose to scale one of the two
conceptual designs for use with a molten-salt reactor of
about 150 MW(t). The fourth task will describe a
research and development program necessary to assure
the adequacy of the design of the Task IIl steam
generator. Further work on the facility for testing
steam generator tubes and tube arrangements of up to 3
MW(t) capacity has been suspended. Instead, we are
proceeding to prepare a conceptual design of a loop
with a capacity in the range of 50 to 150 kW which will
be used to study steady-state operation of sections of
molten-salt-heated steam generators. We are preparing a
report which will describe the status of the molten-salt
steam generator technology, show which elements of
the LMFBR program will be of use in the molten-salt
technology program, and define the areas which need
further experimental study.
A test was run in the sodium fluoroborate circulation
loop to determine if a small quantity of water injected
into the salt in the pump bowl could be detected by
monitoring the off-gas stream for a change in con-
taminant level. An on-line water detector failed to show
a positive response when the water was injected. A
temporary increase in collection rate was noted on two
cold traps, but less than 50% of the injected water was
accounted for by the total weight of material collected.
The data did not yield any clear-cut evidence as to the
fate of the missing water. The water injection test
concluded the current phase of the fluoroborate circula-
tion program in the PKP loop, and the loop was shut
down and drained on April 13, 1970. Since the initial
operation in March 1968, 11,567 hr of circulating have
xiii
6. MSBR Instrumentation and Controls
Further studies were conducted of the part-load
steady-state behavior of the reference MSBR plant.
Cases studied involving both constant and variable
secondary salt flow rate with load produced un-
acceptably low temperatures at the reactor inlet and/or
in the secondary salt cold leg. The addition of a variable
bypass of the secondary salt around the primary heat
exchanger allowed the plant to be operated with
constant steam and reactor inlet temperatures and with
an essentially constant secondary salt cold leg tempera-
ture for loads between 20 and 100% of design load. The
maximum deviation of the cold leg temperature from
its design value of 850°F was about 8°F. Allowing the
steam temperature to increase with decreasing load
permitted the plant to operate with constant reactor
inlet and secondary salt cold leg temperatures over the
same power range. The maximum increase in steam
temperature was to approximately 1110°F, occurririg at
50% load.
7. Heat and Mass Transfer and Thermophysical
Properties
Heat Transfer. — A new test section is being installed
in the inert-gas-pressurized flow system which will
“enable direct determination of the effect on heat