PRODUCTION OF NP237A N D PU238 IN T H E R M A L POWER REACTORS D .
R . V O N D Y , J. A.
Onk Rid?? Yniionni Laboratory. O ak
L A N E , A N D A. T. G R E S K Y
RidqP. I mn.
The production of Np217and Pu23xwas calculated for present-generation reactors. Results indicate that Np2" production now averages about 3 kg. per year per 1000 Mw(th) (at 0.8 plant factor); it was estimated ~ cross section i s about three that 100 kg. per year would b e produced in 1975. Because the P u * ? capture times that of Np217,high efficiency for conversion of N p 2 " to Pu'ih requires removal of the product after relative short exposures. The maximum conversion of Np2j' to PuZJhin a single exposure i s about 20y0, but b y recycling the Np, over-all conversion of 50 to
60y0can b e realized.
~ b use . of Puzn8as a heat soiirce for space power applications temperature and fast-to-thermal flux ratio. T h e high fast Tis of interest because of its very desirable nuclear decay flux level is favorable because of resonance captures, especially properties relative to more conventional radioisotopes. hsn@ in L-23fi. .4 high thermal temperature is favorable because a t a emits 3.4 m.e.v. alpha particles a n d has a half life of 89.0years. given flux level the cT23fi resonance capture rate remains constant? \vhereas Lvith the lower thermal cross sections, the UZ3j 11 is Lvorth nearly as much per unit \jri loss of
Recycler
0 144
n 164
0 0 0 0
259 351 425 484
0 287 0 380 0 449
0 0 0 0
511
0 0 0 0 0
568 578 622 0 641
0 618
0 62' 0 634 0 639 n 643 0 646 0 648 0 649 0 656 0 542
0 6 56
0 668
0 678 0 686 0 692
0 0 0 0
500 519 568 590 606
69' 701 720 570
each mafrrial each prorrssinq !imp.
Puzs'IProduction from Np23i
To convert Xp*37to Pu23eefficiently it is necessar). to remove the product after relatively short exposures. This is because the PufSrcapture cross section is about three times as high as the YTpZ3jcapture cross section causing significant loss of this product.
Production and Worth of P U * and ~ ~ N p Z 3 jin Representative Fuel Cycles (Discounted at 87&) .vp23? .
__~ iVorth" ~~
-
Equilibrium cycleh Firrt cycled
0 118 0 218 0 303 0 3-3 0 436 0 488 0 5'32 0 0 ' 5 0 602 0 629 0 652 0 6'2 0 689 0 '03 0 715 0 '25 0 734 0 780 582
These are first-cycle production rates. Recycle of the L-Z36 in spent fuel would increase the Kp*37 production per Mw installed capacity. Enrichment of spent fuel by addition of UZ3j rather than by re-enrichment in a diffusion plant \vould increase Npz3' production.
1,7235
First cycle?
0.25
given in the Appendices to the report: the future total production of s ~ \\'as * estimated. ~ ~T h e results are given in Table
~~~~~~~~~~~
Cycle
0.20
VI.
Detailed Calculation of Production
NpZ3'Production in
m
n
0. I 5
cvclr
I:\ixniirr
Comparison of NpZai Production in Specific Reactor Types
T h e d a t a developed in these studies permit a n estimation of the SpZ3:production in any given reactor, I n each case. it is only necessary to know: the initial V 3 5 enrichment. the UZ3: depletion per cycle. the ratio of fast-to-thermal neutron flux in the fuel. and the thermal neutron temperature. Using these d a t a , the production rate can be estimated from Figure 2 and Table I . Production rates Ivere rhus calculated for a number of specific reactor designs and the results are summarized in Table IV,
C'llmulalrd i V i c ~/lrlPI ; illill, Friiitiori (J/ Oirqirial .Vp237
r;/ f l i/IOl/ niirrr -up , I / . v pPPI C ' ) i I P
Power Leuel, Mut
Enrichmmt,
86 84 1735 1735
2 6 3 1 92
5'
CwIr
ioadinq.
4. 1 0122 x 105 0 6098 X 105 44 22 x 105 3 . 4 0 X 105
Fract ion drplrtion. qffrrtiw
0 273 0 469 0 724
Produrti on.
e..
~~
~
~
~~
,fuel cyclr. mills li. w h r r at $ 7 0 O / q .
108 240 38,000 44.000
~~~
Pu2@ ~-
~
in
0 047 0 090 0'8
~~~
~
"07th
~~~
in
fuel Product ion.
4.
cycIP;
ni ills lkzuhre at S500/,q.
3 0c 10 0c
14 000e
0 144
... 0 32 10.000 0.38 Volup of n;b23' may diffpr f r o m that used hPrP. Produrtton f r o m '28R not coniidrrrd: rrnctor i s a propored H and M' moritimr s / m m ypnprtitor horminq a metal-to-water ratio t o 7 and t o bP o p r a t r d at 260" C. thrrmnl t r m p r n t u r p : 0..3l.dpower conwrsion. 7 year f u e l cyclr. .\-ot r ~ c o r r r a b l r orcine t o contominotion h y hiqhrr plutonium isotoprs. Brttis-KAPL drsiqn ~ r r d - h i a r i i ~HzO rt moderatrd 1 m5 burner. ./.h2"2conre?trr -lnrqr p o r r ~ rrpactor: 0.288 porc P T e ,Ipproximate/y .5Qc; of thP total plulonium production. conwrsion. 80% plant f a c t o r . 8-war j u r l cvrle. f RP!tis-RAPL desi,qn serd-blanket i1,O modprated t p r v r l p thorium ronvrrtPr~lar,qepouerrractor. A coarse estimnfr was niadr her? of Pquiiibrium recyclr? 2-yFal f u e l cyclc. . .
VOL.
3
NO. 4
OCTOBER 1964
295
Table VIII.
1 - m = fraction processing 107s each cycle Q( m ) = total conversion (infinite c\ d e s )
Optimum Conversion of NpZ3'to P U ? ~ as % Dependent on Processing Loss Ojtiniuni
Fraction Procpssin? Losr of Each .tfnfprial Each Cjclr
Eracizonnl Dfpletivn of .Ipz3i Each Exposure
(0) 0 00423 0 0192 0 0495 0 1016
0 05 0 10 0 15 0 20
(0)
n y Q -1
m =
~
YO(.
Optimum
Q( m ) =
Con7 ersion
1 0 0 0 0
0 854 715 584 460
-
]
_
_
- 1)
ml 1 - mx
where
Results of calculations for a value of a of 3 are given in Table VI11 Losses owing to capture in Xpz38have been neglected Calculations Lvere made of cumulative conversion ivith successive exposures as dependent o n the consumption of Np237 a t each exposure. Results are sho\vn in Table YII. I t ivould likely be possible to add unconverted material to fresh feed to realize a large number of exposures. Processing losses Lvhich were not considered could significantly reduce recovery as indicated by the effect on infinite cycle recovery. If processing losses are large. there is a n optimum exposure for batch operation. Considrr that a n equal fractional loss of each material occurs in each stage of processing (each cycle). Equations of interest are as follo~vs: Let 1 - ,Y = optimum fraction Xp237consumption each exposure n = ratio of neutron absorption rate in PuZ3*per atom to the capture rate in NpZ3'
Acknowledgment
Contributions bvere made to this rtudy by P R . Karten,
L L Bennett, and E . H. Gift. O R N L . Literature Cited (1) Vondy. D. R., "Development of a General Method of Explicit
Solution to the Nuclide Chain Equations for Digital Machine Calculations," 1'. S. A t . E n ~ r g yComm. Rept. ORNL-TM-361 and .\ddendum. 1962. RECEIVED for review January 15, 1964 ACCEPTED A u g u s t 5. 1964
Division of Nuclear Chemistry and Technology, 146th Merting ACS. Denver, Colo., January 1964. This work was done under USAEC contract W-7405-eng-26 at Gak Ridge National Laboratory. operated by Union Carbide Corp. Nuclear Division.
NEPTUNIUM RECOVERY AND PURIFICATION A T HANFORD R . E.
ISAACSON AND B. F. JUDSON
Hanford .4tomic Produrts Operation. Genrral Electric Co
,
Richland. H'nsh
Neptunium i s routinely recovered from irradiated fuel elements at Hanford's two main separations plants. Initial development tests were started in the Purex plant in 1958, then in the Redox plant in 1959, and recently culminated in the installation of new production systems in both plants for improved recoveries. Both recovery flowsheets employ solvent extraction techniques based on the relative extractability of neptunium(V1). The neptunium i s coextracted with uranium and plutonium in the plant's first extraction cycles and then partiiioned and decontaminated in separate neptunium cycles. Excellent decontamination from fission products i s achieved without interfering with mainline uranium and plutonium production. Recovered neptunium i s purified b y anion exchange and shipped offsite for subsequent irradiation to plutonium-238. Over-all separation factors of uranium and fission products from neptunium are greater than 10' and 1 O"', respectively.
have recently been installed in H a n ford's Purex and Redox plants to permit continuous production-scale recovery of neptunium-237. Neptunium is formed in Hanford's nuclear reactors as a by-product of plutonium-239 production and is recovered and purified in the chemical processing plants for subsequent conversion to plutonium-238. 'The production reactor and chemical processing complex a t Hanford is operated by the General Electric C o . under prime contract to the U . S. Atomic Energy Commission. The Purex and Rcdox plants employ solvent extraction systems to separate and decontaminate plutonium from irradiated ELV PROCESS S Y S T E M S
296
l & E C PROCESS D E S I G N A N D DEVELOPMENT
uranium for subsequent use in iveapons components. The irradiated uranium is recovered and decontaminated from fission products for reuse in the V235fuel and lveapons systems. Recently. the Purex plant has also been employed for recovery of useful fission products-including strontium-90. cesium-1 37. cerium-1 44, promethium-rare earths. and technetium-99. Recovery of the transuranic neptunium-237 \vas started a t Purex in 1958 and at Redox in 1959 using main-line process equipment with special campaign operations for decontamination of accumulated material ( 7 . 3, 6). The use of the main plant system permitted rapid response to the .4EC:'s early program needs for plutonium-238 as a heat sourcp i n space
