plutonium recycling in thermal power reactors vol.i

IAEA-143 (VOL.

I)

PLUTONIUM RECYCLING IN THERMAL POWER REACTORS

PROCEEDINGS OF A PANEL ON PLUTONIUM RECYCLING IN THERMAL POWER REACTORS SPONSORED BY THE INTERNATIONAL ATOMIC ENERGY AGENCY AND HELD IN VIENNA 21-25 JUNE 1971

A TECHNICAL REPORT PUBLISHED BY THE INTERNATIONAL ATOMIC ENERGY AGENCY, VIENNA, 1972

The IAEA does not maintain stocks of reports in this series. However, microfiche copies of these reports con be obtained from INIS Microfiche Clearinghouse International Atomic Energy Agency Komtner Ring 11

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ABSTRACT

A Panel on "Pu Recycling in Thermal Power Beactors" was convened by the Agency in Vienna from 21 to 25 June 1971* This was the ihird such panel held by the Agency on Pu Utilization in thermal power reactors, with the first two held in 1964 and 1968. Increased interest in Pu recycle was evidenced by the increased participation and the number of papers submitted. Attendees from twelve countries and three international organizations totaled 50» submitting 53 papers compared with 23 attendees and 19 papers in 1968.

The objective of the Panel as stated by the Agency was to "review the present status of major developments in the recycling of Pu in industrial thermal power reactors." Emphasis was placed on: (a) (b) (o) (d) (e)

Beviewing national programs on Pu recycling, Demonstration programs, Technology and performance of Pu Fuels, Physics of Pu recycle, and Economic aspects of Pu recycling.

After presentation of the papers by the participants, and discussion of each of these by the Panel, two working groups were formed, each led by a chairman, for the purpose of preparing the summary, conclusions and recommendations to the Agency. The working groups encompassed all subjects covered by the Panel and were divided into the following areas: - Economic studies and national and demonstration programs - Physios and fuels technology.

The reports of these two working groups were reviewed, modified and accepted by the Panel and are presented in tme next section of this document.

In addition to the spécifie inforaation presented in the aforementioned working group reports, the following general observations, conclusions and recommendations are givens 1. Increased activities and interest in Pu recycle was particularly illustrated by the number of demonstration programs initiated since the 1968 Panel. These are generally associated with water reactors, although some work is being conducted in graphite moderated reactors. 2*

Information presented did not refute the finding of the 1968 Panel that Pu recycle technological feasibility is established for water reactors. The Panel recognized that additional work is required for optimum recycling, but the cost of such developments may not be justified because of ultimate benefits to be derived or length at time before the Pu is used in fast reactors.

3. Because of the large experience with pellet fuels and the lack of substantial benefits to be gained from proven alternate fuel types, the pellet type fuel is the type that will be used initially for thermal recycle in water reactors. 4« Pu may have a higher value in thermal reactors other than light «rater reactors, but insufficient data were presented for firm conclu-

sions to be drawn. As in 1968 the Panel agrees that on present evidence it is likely that Pu will ultimately be used in fast reactors.

The extent to which Pu is recycled in thermal reactors will be determined by the timing of fast reactor introduction. 5. Much important and valuable data were presented and key conclusions were drawn by the Panel. It is believed that these would be extremely beneficial to the nuclear industry. Therefore, it is recommended that the proceedings of the Panel be published and given wide distribution by the Agency. 6. The Agency should continue to serve as a point of information exchange through sponsorship of meetings and publications. 7. Another panel meeting should be held around the end of 1973 when a substantial amount of information is likely to become available (l) from the various Demonstration Programs, (2) on the timing of introduction of fast reactors, (3) on fuel technology, etc.

C O N T E S T S

OF

V O L. I

X•

II. III.

Meeting Arrangements ...................................................2 Panel Survey on the Physios of Pu Recycling and Fuel Technology .............................................»»..»...3

17.

Panel Survey on Pu Recycling Economics, Demonstration and National Programmes ..................................9 "aneJ «oncxusions •»••«««•««•««•««•••«•••••••••«•••••••»..*«•••••«••••««lp Recommendations of the Panel to the IAEA ...............................16

"• VI.

PL/447-1

"Plutonium Kecyole in Can au-Type Eeavy Water Reactors11 by 1C.F. Duret, Canada ••••••••••.••.•...••..........•..................19

PL/447-2

"Thermal Reactor Plutonium Fuel Irradiation Programs in the United States" by F.O. Dawson, USA ...............................33

PL/447-3

"Results from USAEC Plutonium Utilisation Frogran Conducted by Battelle-Korthwest" by F.Q. DawsoB, USA .......................45

PL/447-4

"General Description of the Belgian Programme OB Plutonium Recycling in Thermal Reactors" by F. Bairiot, Beugium ..............91

PL/447-5

"Measurements in Cold Critical Assemblies Performed at Ifol, Belgium", by F. de Waegh, Belgium ..................................99

PL/447-6

"Evolution of FueJ Under Irradiation - Belgian Experience", by H. Bairiot, Belgium ...............................................133

PL/447—7

"Progress on Fabrication Techniques of Plutonium Fuel" - Part I: "Ceramic Pelletized Fuel Pins"; Part lit "Coated Particles", by J.J. Euet, E. \Tanden-Bemden, E. Bairiot, Belgium •.........•..••••.157

PL/447-8

"Security Rulos and Quality Control", by E. Vanden Bemden, Belgium .169

PL/447-9

"Belgian Demonstration Programme on Plutonium Recycle" Part I: "Plutonium Recycling in the BR 3 Reactor"; Part lit "Loading of Two Plutonium Assemblies in the Dodewaard Reactor"; Part Hit "Recycle Concept for the SENA Reactor", by H. Bairiot, F. de Waegh, Belgium

PI/447-10

««•••«•«•«««••••••«««««««••««««*««»»»»•»•««••••«««••••««•»

"Dégradation de la valeur du Pu dans les Réacteurs thermiques à eau légère, au cours de recyclages successifs" (Degradation of the Pu value in LWR during successive recycle), by M. Bregeon, Prance ••••••••••••«•••«•«•••••*•*•••«••««••«•*««»«••••••••«•••••••«

PL/447-11

"Etude économétrique du recyclage du Pu dans les réacteurs & eau légère" (Econometric Study of Pu Becycle in LWR), by

PL/447-12

"Perspectives du recyclage Pu er France" (Perspectives of Pu

PL/447-13

"Evaluation of "j;7Pu Alpha Value Toy Using Standard Resonance

PL/447-14

"Critical Experiments on Two Regional Water Lattice Core", by

PL/447-15

"Evaluations on Performances of Low Plutonium-enriched Riel s

PL/447-16

"Experiences in the Plutonium Fuel Development Laboratory on Scrap Recovery and Vaste Disposal", by I. Ito, Japan •••••••••••••375

PL/447-17

"Experiences on Pellets Fabrication for Thermal Reactor Use", by T. Iluto, Japan •••••••••••••••••••••••••••••••••••••••••••••••«401

PL/447-18

"Sol-gel Processing of Mixed Oxide for Thermal Reactor Use",

PL/447-19

"Safety Considerations on Equipments Design in the Ne* Plutonium

PL/447-20

"Experiences on Quai siy Control and KUF Origination During

PL/447-?!

"Experiences and Future Considerations on External Radiation

PL/447-2?

"an Economical Consideration of Fuel Recycling for Light Water

PL/447-23

"Calculation of the Capacities of Plutonium Recycle in Light Water Reactor Systems", by H. Oi, et.al., Japan ••••••••••••••••••471

PL/447-24

"Pu Recycling Demonstration Program in the 12IKAKA Ko. 1 Reactor",

PL/447-25

"A Reference Design of PuO.-UO. Assembly for JPDR-II Core'1, by

PL/447-26

"Local Power Distribution in Pu-loaded BWR Core", by Y. Fukai

PL/447-27

"Characteristics of Pu02-UU2 Fuel Assemblies in Commercial Boiling Water Reactors" by S. Yamada et.al., Japan •••••••••••••••515

? tQ

CONTENTS

PL-447-28

OF

VOL. E l

"Heavy Aater B«aotor Project in Japan Especially on Plutonium Utilization", by Shiro Shima and Sandamu Sawai, Japan •••••••••••••533

PL/447-29

"Expérience from the IFA/159/160 Irradiation in the ttBWR", by Y. Nak amura, et.al., Japan •••••••••••••••••••••••••••••

.553

PL/447-30

"PNC Irradiation Program in the Saxton Core III1*, by Y. Nakamura, et.al., Japan • •••••.••••••••••••••••••••••••«

.575

PL/447-31

"Review on Plutonium Recycling Experiments in Japan", by Y. Nakamura, Japan ••••••••••••••••••••••••••••••••••••••••••••••••707

PL/447-32

"The VAK-Kahl Plutonium Recycle Demonstration Program" by 5. Knoglinger et.al», PHG ••••••••••••••••••••••••••••••••••••••••••593

PL/447-33

"Parametric Study of Pu Recycling in Obrigheim Power Station" by C. Steinert, 11. Dillig, FRO ••••••••••••••••••••••••«•••••••••••611

PL/447-34

"Revue of CNBN Reactor Physics Work on Pu Recycle in LWR'i',1 by F« Pistella, Italy ••••••••••••••••••••••••••••••••••••••••••••••••625

PL/447-35

'Verification of Plutonium Build-up Calculation Methods Based on Post-irradiation Measurements on a Uranium Assembly of the Qarigliano BWB", by A. Ariemma et.al., Italy •••••••••••••••••649

PL/447-36

"Irradiation Experiment» at CNEN for the Thermal Recycle of Plutonium Fuel", by A. Calza Bini et.al, Italy •••••••••••••••••••••669

PL/447-37 PL/44/-38 PL/447-39 PL/447-40

PL/447-41 PL/447-42 PL/447-43

"Progress in the Preparation of Plutonium Packed Particle Fuel at CNEN and AGIP-SucJeare for Thermal Utilization", by G* Cogliati et.nl., Italy ••••••••••••••••••••••••••••••••••••••••••721 "Progress in Pelletizatlon of Plutonium Thermal Fuel at CNBN*' by V. DiStefano et.al., Italy •••«•••••••••••••••••••••••••••••••••747

"Fabrication and Quality Control Experience at CNEN to Prepare Plutonium Fuel Rods", by G. Fossati et.al*, Italy ••••••••••••••••••767 "Thermal, Hydraulic and Mechanical Experimental and Design Activities at CNBN for the Thermal Recycle of Plutonium Fuel", G.P. Cali, et.al., Italy ••••••••••••••••••••••••••••••••••••••••••o05

"Main Results of ENEL Garigliano 3WR Plutonium Demonstration

Program up to 1971", by A. Ariemma et.al., Italy ••••••••••••••••••o851 "Review of the Italian Programs to Reoyole Plutonium in Thermal Reactors", Italy (presented by Dr. Schileo) ••••••••••••••••••••••••885 "The Present Status of the Thermal Cross-sections of the Main

Plutonium-isotopes and of their Fission Products*1, by fl.D. Lemmel,

PL/447-44

"Thermal Reactor Physics Measurements Using Pu Bearing Fuels" by G.fi. Kinchin, et.al. U.K. ........................................951

PL/447-45

"Plutonium Recycle in Heavy Water Reactors", by 3.P. Rastogi, India.,973

PL/447-46

"Review of Plutonium -Technology Work in India", by U.P. Rastogl .....983

PL/447-47

"Recyclage au Plutonium dans les Reacteurs à Eau Légère" by F. Lafontaine, France - EURATOM .......................................989

PL/447-48

"Review of National Program", by Mr. ífachácek, C3SH ..................Iol3

PL/447-49

"Pu Recycling Economical Studies" by Mr. Balriot, Belgium ............1017

PL/447-50

"Review of National Program" by Mr. Osteriundh, Sweden...............1023

H./447-51

"Review of National Program", by Mr. Duret, Canada ...................1027

PL/447-52

"Activities on Pu Recycling in the Fed. Hep. of Germany" by W. Etzel, Ch. Breest, France ...........••..•.«.....•.....•.•.••••.1031

PL/447-53

"Physique du Recyclage Pu" by Mr. Naudet, France ...............••.•••1035

Annexes:

A) Formally Proposed Topics for Panel Discussions ••••••«•»«•»••••••••*••••••••s :'vG!¿ranR in ixt* 'jr.te*d "¿avas, will cover major irradiation programs being conducted in the United States which will demonstrate the utilization of plutonium in current power reactors. It has become clear with the U.S. nuclear Industry that recycling of plutonium in thermal power reactors will be required by 1975. After that date there will be a rapid rise in available plutonium. Storage of this plutonium for later use in fast reactors is not required nor economically justified. Therefore» the U.S. Industry is taking the necessary steps required to be ready to use the plutonium when it becomes available. 2.0 PLUTONIUM FUEL CHARACTERISTICS

2.1 Introduction The technical feasibility of utilizing plutonium in thermal reactors has been successfully demonstrated through the success of numerous experimental programs (Table I). The irradiation program in the Plutonium Recycle Test Reactor (PRTR) was characterized by testing advanced mixed-oxide fuel concepts to moderate burnups under specific power conditions higher than those currently employed in commercial power reactors'- ' . Experiments are being conducted in the Saxton reactor to evaluate the performance of mixedoxide fuel at linear power and burnup levels consistent with modern PWR technology1 This paper is based on work performed under the U.S. Atomic Energy Commission Contract AT(45-1)-1830 45

The success of these and other experimental programs 1s attested to by the universal acceptance of plutonium utilization in light water reactors. These efforts have been culminated by the Initiation of demonstration programs In commercial power reactors (Table II) with the rapid full-scale Implementation of plutonium utilization 1n commercial plants anticipated'- ' • •". Table I

MAJOR DOMESTIC EXPERIMENTAL UO^-PuO, IRRADIATIONS

Reactor PRTR PRTR PRTR (High

No. of Rods

Fabrication Method

Clad

»»1860 «v 700 ^1730

Z1rc. Z1rc. Z1rc.

1296 638 -v 250 4

Zlrc. Z1rc/SS Z1rc. Z1rc.

Power Density) EBWR Saxton-Core II

Swaged VI pac Pellet-VipacSwaged Vipac Pellet-Vipac Pellet Pellet

Peak Burnup, MUD/MTM

Peak Rod Power, KU/Ft.

Peak Heat Flux 2, BTU/Hr-Ft

%17,000 •-18,500 "-13,000

•v/17 «-16 x21.5

390,000 370,000 500,000

* 3,000 --29,000 '.38,000* --13,000

8 16 -.19 '-10

200,000 530,000 630,000 230,000

Saxton-Core III Dresden- I "December 1970 - Goal Peak Burnup -«50,000 MUd/MTM on Rods from Saxton Core II Program. Table II MAJOR DOMESTIC PLUTONIUM DEMONSTRATION PROGRAMS

Reactor Type BUR

PUR

Reactor

No. of Units

Program Objective

Big Rock Point

16 Assemblies 32 Rods

Fuel Concept Evaluation

Big Rock Point

3 Assemblies 204 Rods

Performance Demonstration

Dresden I

11 Assemblies 99 Rods

Performance Demonstration

Vermont-Yankee

4 Assemblies 48 Rods

Performance Demonstration

4 Assemblies 720 Rods

Performance Demonstration

San Onofre

2.2 Plutonium Fuel Types Plutonium recycle fuel evaluation studies have Included a variety of ri ?! fuel types and fabrication processes1 ' J. Pellet fuel, because of a vast amount of satisfactory U02 experience, Is the reference design for plutonium utilization in commercial reactors. Packed particle fuel (vipac) 1s considered to be the most promising alternate fuel fornr . Under the operating conditions presently employed or projected for commercial power reactors in the near future, neither pellet nor packed particle mixed-oxide fuels appear roí to be confronted with Inherent performance limitations.1- J 2.3 Irradiation Behavior - Effects of Plutonla Additions to U09 Although the small PuO?. additions to UO?. for thermal reactor fuel has an Insignificant effect on most irradiation performance characteristics, there are Important considerations related to performance that are unique with plutonium enriched fuels.

2.3.1 Plutonium Uniformity

Sintered mixed-oxide fuel material produced from powder prepared by the coprecipitation process Is essentially uniform (U, Pu)0? solid solution rgi with small localized PuO? nonuniformityL J. However, mechanical mixing of UO? and Pu02 powders prior to pressing and sintering, which Is the most commonly used method of preparing mixed-oxide fuels, results in the formation of non-uniform solid solution with localized regions of high PuO? concentration'- •*. Mechanically mixed and heterogeneous 1 y enriched mixed-oxide 1s also used for packed-particle Uniform solid solution forms rapidly in regions of mechanically mixed mixed-oxide fuel operating above columnar grain growth temperatures KI700°C) by a vaporization-condensation process «whereas homogenization occurs more slowly In the equiaxed grain growth regions (1400 - 1700°C) by solid state diffusion. Therefore, localized regions of high plutonium concentration persist In a significant volume of the fuel In a rod operating under normal conditions. The high fission density associated with these localized regions could affect such fuel performance characteristics as swelling and fission gas release although there 1s no definitive supportive evidence.

47

2,3.2

Fuel Clad Mechanical Interaction

Core-clad mechanical Interaction phenomena, which are more severe in pellet fuels than In particle fuels'- % are sensitive to fuel type and design and should not be affected by Pu02 additions to U02 except as the Pu02 additions may affect swelling behavior. Likewise, fuel cladding reaction in 7.ircaloy-clad fuels'- * " is probably related to stolchiometry and oxygen partial pressure in the rod and not plutonium concentration.

Postirradiation length measurements and diameter profiles were obtained on vioac fuel rods and rods containing hot-press and cold-presssinter oellet fuel irradiated in PRTR to compare the dimensional behavior FQ-I of the three fuel types1 J. The vipac rods contained crushed, pneumatically impacted UO>«-PuOp fuel vlbratlonally compacted to a smear density of nominally 86' TO. The hot-press and the cold-press-sinter pellet density was approximately 93% TD and the rods wore assembled with a nominal 0.3 mm diametral gap. Observations related to the rod dimensional data are summarized as follows: '

Cladding ovality and length changes, which seem to be related, are more pronounced in the cold-press-sinter pellet rods than in the hot-press DO!lot «wd vipac rods. The fuel 1n the vipac rods próvidos cladding support and thereby reduces the amount of ovality. Considerably less ovality occurred in the hot-press pellet rods than in the rods which contained cold-Dress-sinter pellets although both were assent»led with a nominal 0.3 mm diametral gap. It is suspected that the hotpress pellet fuel, because of Its high stoichiometry (0/M • --2.15) and reduced thermal conductivity, operated at higher temperatures than the cold-press-sinter fuel. The increased expansion resulted in closure of the fuel-clad gap and prevented cladding collapse and distortion.

*

Although the Zircaloy cladding may be calculated to be freestanding, ovality results from creep which occurs early in the irradiation.

48

*

The creep rate may be enhanced during neutron Irradiation.

* There was no evidence of general or local clad swelling 1n any of the rods. * There was no evidence of ridging at pellet Interfaces In either the hot-press or cold-press-sinter pellet fuel rods. 2.3.3

Thermal Conductivity and Restructuring Kinetics

Small Pu02 additions to UOZ do not have a significant effect on the thermal conductivity of UO^ ••. Hence, fuel restructuring kinetics, which are affected by temperature and temperature gradient In the fuel, should not be significantly different for U0>: and U02-PuO¿> fuels operating under comparable conditions. Restructuring phenomena are more dependent on fuel type. Structural evidence of certain operating conditions in the fuel can be quickly eradicated by time-temperature-dependent "¡ass transport roí and diffusion phenomena during subsequent reduced power operation1- J. 2.3.4

Fission Product Migration

The transport processes which affect fission product migration in oxide fuels during irradiation are not affected by small PuO; additions in UOo . Fission products such as zirconium, niobium, cerium, and prasedymium, whose oxides are thennodynamically more stable than U0k and PuOj, do not redistribute significantly (Figure 1). Conversely, fission product cesium, ruthenium, and rhodium, whose oxides are thermodynan'icdlly less stable than U02 and PuO¿, migrate significantly during irradiation (Figure 1). ?.3.5

PIuton1um Rodi stribution

A significant change in radial plutonium concentration can occur ranidly in mixed-oxide fuel rods during irradiation'- •". For mixed-oxide fuel operating above columnar grain growth temperatures (>1700:>C), radial plutonium redistribution is coincident with the fuel restructuring phenomena and occurs as a result of the preferential evaporation of

49

DISTANCE (ROMTHERMAL CENTER

FIGURE 1. RADIAL FISSION PRODUCT DISTRIBUTION IN A U07-2 wt.% Pu02 FUEL ROD AFTER IRRADIATION IN PRTR FOR 230 HOURS. THE DISTRIBUTION OF FISSION PRODUCTS MAS DETERMINED BY MICROGAMMA SCANNING.

50

gross amounts of uranium oxide from the central hot region of the fuel to the cooler peripheral regions near the cladding"- ". Thermal diffusion can also cause plutonium enrichment 1n the high temperature region of the fuel^. Electron microprobe scanning analyses of a vlbrationally compacted UO--2 wt.SJ Pu02 fuel specimen irradiated for only 230 hours in PRTR

show that the average plutonium concentration in the region of the fuel that operated at columnar grain growth temperatures and above, i.e., >1700°C, is about 10% higher than the nominal as-fabricated concentration (Figure 2). The plutonium concentration in the once-molten fuel region is about 18% higher than the nominal value. Plutonium redistribution effects in mixed-oxide fuels have been shown to be stoichlometry dependent*2.3.6

Fission Gas Release

Fission gas release from oxide fuel 1s not significantly affected by adding plutonium to U02 or bumup level and is most sensitive to fuel temperature*- . Fission gas release fractions for pellet and particle fuels are comparable when evaluated on the basis of volumetric average fuel rod temperature. The higher surface area of particle fuels should result in a higher release by recoil-knockout; however, the contribution of this release mechanism to the total is relatively small under normal operating conditions. Measurements made on mixed-oxide rods during the irradiation show that the release of sorted gases and moisture from the fuel does not contribute C191 to internal pressure1- J. The released gases apparently react rapidly with Zircaloy cladding and/or the lower temperature fuel. Internal gas pressure resulting from the release of fission gases Increases in a

stepwise manner during reactor shutdown-startup power cycles with no significant changes occurring during steady power operation. This behavior, which is the same for UO, and UOz-PuO., fuels, is consistent with the r?ni diffusion-trapping release mode1- J and the postulated containment of released fission gases in the central cavity of the rod until the fuel 1s cooled and cracking occurs'- ••.

.

CJ1 Is»

í\ ............ ..,

FIGURE 2.

RADIAL PLUTONIUM œNCENTRATIQN IN AN IRRADIATED !)02-2 wt.S Pu02 FUEL SPECIMEN AS DETERMINE!) UY LlhCTRON f-'ÏCROPRODE SCANNING. FUFL ROD WAS IRRADIATED FOR 230 HOURS IN PRTR.

2.3.7

Defect Behavior

Presently, there is no defect experience with mixed-oxide fuels in commercial power reactors. Comparative experiments to investigate the performance of defective and intentionally defected Zircaloy-clad mixedoxide fuels were conducted in the Plutonium Recycle Test Reactor (PRTR)"- •" and Engineering Test Reactor (ETR)*- •*. A defect test was performed on a mixed-oxide fuel rod in the ETR pressurized loop facility. The pellet-containing rod was nominally 14.4 mm OD clad in 0.76 mm thick Zircaloy-4 cladding with a 68-cm active fuel length. The UO?-4 wt.2 PuO? pellets were assembled in the cladding with a 0.25 mm diametral gap. Peak burnups of approximately 3000 MWd/MTM were achieved. The system pressure was 2000 psi at an average coolant temperature of 255°C. Tne rod was intentionally defected by drilling a 0.51 im diameter hole through the cladding in the peak power region after the rod was irradiated to a peak burnuo of about 1200 MWd/MTM. The defected rod was subsequently irradiated for 17.3 effective full power days at a linear heat rating of about 13.5 kW/ft which produced maximum fuel temperatures of approximately 1650°C. Rotation of the element increased the heat rating at the defect to about 20 kW/ft with associated maximum fuel temperatures of 2400°C for an additional 14.1 effective full power days. Gross gamma activity levels monitored in the loop piping system increased in proportion with the power but remained well within acceptable limits. There were no Indications of particulate fuel release and there were no activity bursts associated with changes in reactor power level, some of which were severe, that characterize the defect behavior of particle fuel.

The gross gamma scan of a companion nondefected rod shows a smooth fission product activity distribution that is related to the axial power profile. The gamma scan of the defected rod shows two localized variations In the fission product activity curve at and above the peak power position. The perturbations, which are possibly the result of axial fuel relocation

53

or slumping, do not appear to be related to the position of the defect. Pre- and postirradiation diameter profile measurements on the defected rod and companion nondefected rods show that the cladding has plastically deformed and become oval shaped during Irradiation 1n the pressurized loop. The maximum ovallty Increase in the defected rod was 152 um. Cladding collapse on the defected rod must have occurred during the relatively short predefect irradiation of 35 effective full power days since no collapse type deformation 1s expected to occur after defecting because of pressure equalization. There was no evidence of clad ridging at pellet Interfaces 1n any of the rods. These results show that clad deformation, even in cladding that 1s calculated to be freestanding, commences during the early stages of irradiation under these conditions.

The results of the defect experiments conducted to date show that (1) the defect performance of U02-Pu02 fuels 1s essentially the same, although there 1s evidence that plutonium redistribution is enhanced in defected mixed-oxide fuels possibly as a result of a change in stoichlometry,

i.e., increased O/M"- ••, (2) defect behavior, I.e., activity release, is most sensitive to fuel type, fuel temperature, and power history'- ', and (3) the defect performance of both pellet and particle oxide fuels 1s excellent under normal operating conditions. It appears that the defect behavior of ceramic fuels is more sensitive to surface heat flux than to fuel type'- * •'. Consequently, as more severe operating conditions are imposed on the fuel, defect behavior could become a performance limitation.

2.4 Transient Behavior

The behavior of oxide fuels when subjected to accidental high energy, short duration, power excursions is an important safety consideration In both thermal and fast reactors. Experiments were conducted to Investigate and compare the transient behavior of the different pi utoni urn-enriched fuel types'- * •* and to Investigate the possible effect of large Pu02 particles or agglomerates in mixed-oxide fuels'- ^.

54

The results of comparative transient experiments performed with unirradiated pellet and particle fuel pins show that[8'25^: • The cladding failure threshold energy of approximately 270 cal/gm fuel 1s essentially the same for both pellet and particle fuels. •

Both pellet and particle fuels fall by clad melting; however, because of higher fuel temperatures in particle fuels at comparable energy depositions, more fuel is expelled from the pin (Figure 3).

•

The extent of Ztrealoy-water reaction 1s comparable for both pellet and particle fuels and Increases with increasing total energy deposition.

Transient tests were conducted on unirradiated oxide pelletcontaining thermal recycle fuel pins to Investigate the possible effects of large single 550 \aa diameter PuO, particles, which could possibly be F261 present 1n the fabricated fuel, on the transient behavior1. The results are compared with the results of similar tests conducted on pins containing enriched U02 and U02-PuO? pellet fuel which did not contain abnormally large PuO? particles. These tests demonstrate that below the energy deposition range of 200 to 213 cal/gm of fuel the energy generated in a single 550 urn diameter Pu02 microsphere is readily absorbed by the surrounding U02 matrix and no external effects occur. Above this energy range, large Pu02 particles near the surface of oxide pellets (0 to 800 inn) can penetrate the cladding (Figure 4).

The cladding perforation threshold for thermal

recycle pellet fuel containing 550 »m diameter Pu02 agglomerates is in the energy deposition range of 200 to 213 cal/gm of fuel. Perforation occurs by localized melting (Figure 5} caused by the expulsion of PuO? particles located near the pellet surface through the cladding. Fuel loss to the water is limited to only the PuO? particle. Other tests have shown that the prompt cladding failure threshold due to melting for unirradiated pins containing enriched U02 and U02-PuO? pellet fuel is in

55

O» O»

3. ÎH; APPFARAtCF CF /.IRCAl.;lV-CLA!; V.'PAC AND :'ïl...F: -CONTAINING UD RAKSIb.\T IRRADIAIICN A; ?7£ cal/gin UC?.

!»S AFTER

FIGURE 4. POSTIRRADIATION APPEARANCE OF :A Z1RCALOY-CLAD PIN SlSJtCTED TO A TRANSIENT ENERGY DEPOSITION 0F 213 CAL/S.M 0F a EL (CLADDING HAS PENETRATED BY Ti» PjOo MICROSPHERES LOCATED NEAR THE SURFACF OF U0? PELLETS.)

en oo

20Oft m

FIGURE 5. ZIRCALOY CLADDINS PENETRATION CAUSED BY THE EXPULSION OF A PuU2 MICROSPHERE LOCATED NEAR THE SURFACE OF THE U02 FUEL DURING TRANSIENT IRRADIATION.

the range of 225 to 274 cal/gm of fuel. Therefore, the presence of large 550 urn diameter particles In the fuel has slightly reduced the threshold energy deposition required for cladding perforation and the subsequent release of fission products. However, there were no Indications of the effects of prompt fuel dispersal caused by the expulsion of the Pu02 particles Into the surrounding water when tested at these energy levels. Because the presence of single 550 urn diameter Pu02 particles 1n mixed-oxide fuels does not appear to significantly affect the cladding failure threshold energy from that of mixed-oxide fuels with the normal PuO? particle size and distribution, product specifications which limit the maximum Pu02 particle size below 550 ;.-m diameter do not appear to be warranted from the standpoint of transient fuel performance considerations. Specifications should be developed which limit the maximum PuO? particle size. These experiments have shown that such a limit Is greater than 550 .;m diameter. 3.0

PLUTONIUM FUEL FABRICATION

3.1 Fuel Fabrication Synopsis

Before the advent of the U.S. Atomic Energy Commission sponsored Plutonium Utilization Program at Pacific Northwest Laboratory in 1956, there was essentially no experience in producing piutoniurn-containing fuel elements. Largely through the success of this program, fabrication methods and techniques were developed which permit production of mixedoxide fuels by either the palletizing or the packed particle processes. The success of the program 1s attested to by the Increased number of programs devoted to the commercial implementation of plutonium utilization

In thermal reactors and the general acceptance of the need for and technical feasibility of plutonium utilization.

Solid mixed-oxide fuel pellets clad In Zircaloy 1s considered to be the current reference fuel design for the commercial utilization of Tal plutonium In water cooled thermal reactors'* J. This choice 1s a natural result of the vast amount of satisfactory experience accumulated with

59

UO; fuels using this basic design. Mixed-oxide pellet fuels can be prepared by sintering either coprecipltated (U, Pu)02 powder or a mechanical mixture of U02 and PuO5

4.2 Calculation of Multiplication Factors A large number of "clean" plutonium fueled, H?0 moderated lattice experiments were reported at the 1968 IAEA panel meeting'- ^. Also, described was a set of analysis methods and how the results from the use of these methods compared with the lattice experiments. Since that time, changes In some of the cross section data were made (See Table VI). Also, Improvements In the basic theory were developed as follows: Slowing Down a. Include Intermediate resonance formulation 1n HRG b. Allocate resonance cross section over more than one fine group 1n HRG. c. c.

UpscatteHng correction for H 1n H20 1n HRG. Dancoff correction using Carlvik's method rather than ANL-5800 tables.

Thermal Izatl on

a. Current calculation of diffusion coefficient 1n BATTELLE-REVISED-THERMOS

The most significant changes In the cross section data were In the resonance Integrals for 23SU and r-9Pu. The low value of the Integral for ;i:î?;U Is arrived at by simply adding UP all Individually resolved resonances developed from resonance parameters. The modifications in the slowing down calculations also had an appreciable effect on the calculated kgff's» particularly for the tight lattices where resonance capture 1s high.

Tables VII, VIII, IX and X show comparisons between previous and current calculated k 's for U02, Pu-Al and U02-PuO? lattice experiments'- •". Generally the modified methods give k f . results more nearly In line with oxperlments than the previously reported results. The lower 238U resonance Integral results 1n considerable Improvement 1n the correlation for the U0? Yankee experiments, particularly for the tight lattice cases. The Pu-Al data show much better correlation. However, the k.^'s of the Al AK 2 wt.2 Pu experiments are still over-predicted by about 3% -TT. Because

Table VI CHANGES IN BASIC DATA

Quantity

Measured 26.5 ± 0.3

1.

Age In H,0

2.

Resonance Integrals» I. (107 to 0.5 eV)

a.

?3S

U I-c I»f a

b.

238

c.

239

U !..

a

d.

2 Pu Ua

26.5

25.5

140 t 8 280 t 11 0.50 t 0.02

148 286 0.517

140 280 0.500

280 *, 12

282

269

215

177 312 0.567

-

Pu I.c I«f

f.

Integral Value Calculated from Master Library Previous Current

310 i 20 -

331 0.650

8370 i 380

8424

8467

164 514

164

557i 33» 524 i16 1280 + 60

517

0.319

0.317

1144

1112

Table VII FOR UO, (YANKEE) CRITICÁIS

Moderator/Fuel (Volume)_ 1.048 1.405 1.853 3.357 4.078 4.987

Mean

Calculated Keff Previous Current 0.9731 0.9786 0.9828 0.9886 0.9888 0.9888 0.9835

1.0035 1.0017 1.0002 0.9989 0.9980 0.9969 0.9999

67

AKgff

Current/Previous (10-3) +30.4 +23.1 +17.4 +10.3 + 9.2 + 8.1 +16.4

Table VIII CALCULATED OF KeiT FOR —Al-Pu CRITICALS r- —— —————— — —VALUES ——————— ————— ——— ——

Lattice Spacing

0.75 0.80 0.85 0.90 0.95

Mean Lattice Spacing (in.) 0.85 1.05 1.30

Al-1 .8 wt.% Pu

Previous .0323 .0274 .0240 .0223 .0187 1 .0249

Al-2 wt.% Pu

Previous 1 .0439

Current 1.0265 1 .0237

1.0074 1.0050

1 .0477 1.0428 1.0522 1 .0375

1.0079

1 .0448

1.0228

Current 1 .0088 1.0101 1 .0081

, M f Al-5 wt.% Pu Boron in H,0 (WDDm) Previous 1.0232 1.0234 1.0125

0 0 0

Mean 1.05

K«tf

0 48 101 164 229 285

Mean

1.0202

3 .0222 1.0215

Current 1.0090 1.0140 1.0125

1.0197 1 .0234 .0206 .0249 .0222 .0199 .0193

1 .0118 1.0140 1.0032

1.0217

1.0080

1.0100

i .0070 1.0058 1.0080

Table IX CALCULATED VALUES OF Kpff FOR EBWR AND SAXTON FUELED CftlTICALS EBWR (U02-1.5 wt.2 Pu02)

Lattice Spacing (In.) 0.55 0.60 0.71 0.80 0.90 0.93

Kc

Previous "

1.0041

0.9925 0.9914 0.9965

0.9994 0.9996 0.9983

0.9967 0.9998

0.999? 1 .0012

0.9958

1.000.1

0.9980

Mean

Current

SAXTON (U02-6.6 wt.% Pu02)

Lattice Spacing (1n.)

0.52 0.56 0.735 0.792 1.04

Mean

l^ff

Previous

Current

1.0115 1.0094 1.0119 1.0130 1.0266

1.0094 1.0081 0.9996

1.0145

J .0079

1.0041 1.0184

(1.0053 without 1.04 in. lattice)

69

Table X FOR UO,-PuO, FUELED CRITICÁIS

Lattice Spacing

UOo-2 wt.Z PuOo * ff 16% Z'fOpu

8* ?l tOPu Previous Current 0.9916 1 .0022 0.9974 1 .0073

0.9933 0.9993 0.9930

1.0090 1.386 1.0074 1.611 0.9920 Mean 1.0010

1.0043 1 .0029

0.80 0.93 1.05 1.143 1.32

0.0012

0.9991 0.9990

Previous

Current

1.0027 1 .0000 1 .0076 1.0077 1.0043 -

1.0018

1.0045

1.0029

0.9972 1 .0048 1 .0067 1 .0039 -

UO,-4 wt.% PuO,

0.80 0.85 0.93 1.05 1.143 1.386 1.60 1.70

Mean

0.9C69 1..0007 0.9969 1.0038 1.0008 0.9976 0.9905 0.9867

0.9955

70

24%

Previous 0.9957 1.0001 1 .0044 1.0080 1.0064 1 .0035 1.0030

2*0pu

Current 0.9975 0.9990 0.9996 1 .0051 1.0051 1.0034 1 .0016

of almost universal difficulty In predicting these experiments, the experimental data have been re-examined and the fuel Is being analyzed, but so far no discrepancies have been uncovered.

The calculated Individual and mean values of kgff for almost all of the UO?-PuO? data are within ± 1%. As stated previously, these correlations are with the low (269b vs. 282b) resonance Integral for ?38U. There Is no Integral data to justify the low value. Also, other correlations performed by Batte!le on Yankee reactor burnup data are Improved If the higher resonance Integral value of 282b 1s used'. This 1s seen in Figure 6. The only difference 1n the two calculations Is the ?3!JU resonance Integral. The "1969 Cross Section Evaluation" curve Is for 269b and the "Hybrid Cross Section" curve 1s for 282b. It 1s seen that the ?39Pu buildup at high exposures tracks much better when I», of 282b 1s used. Several Infinite medium multiplication factor measurements (lu) have been made in the Physical Constants Test Reactor which has the advantage of eliminating leakage and core-reflector soectrum problems associated with small critical experiments'. Table XI presents data on two such lattices. In another series of PCTR experiments, the change in kœ with PuO? particle size was determined and is shown in Figure 7. The latter is of Interest since mechanically blended UO?-PuO^ results in a powder containing finite particles of PuO? in U0?. It is seen from Table XI that for the 0.9 wt.*. Pu02 experiment, the calculated k», of 1.105 compares quite favorably with the experimental value of 1.104, while the difference in the k^ values for the 2 wt.* PuO? experiment 1s almost 3%. It was determined that compensating differences In experimental and calculated "• " and "f" values of approximately 1.55. yielded the good agreement for the

0.9 wt.% PuO? lattice. Although there was some compensation between "-.," and "f" for the 2 wt.% PuO? experiment, it was insufficient to result in agreement on k.,.'s. The calculated "f" was 2% low while the experimentally determined "nf" was approximately 3» low.

71

!•

¿OOIAC

200

• x-

* FXPERIMENTAl POINT

»...*&*:•

— LEAST SQUARE FIT OF EXPERIMENTAL DATA

zaoo

X

jj

..**

144

:*U ••*""

•tii x"

'..-'"

14.00

!

aë ^ *

1 *

-^íl*

''

~

-*Wr^ -•* ST™ •

'

i

*

s

1 OO*

>

'*

•-

;«f

/

a 20 "

^

T

x s, *

..y.

^"

^

:

...s- -*•» '

•/

''

'/

000

î 36

0. SSï g -

i aie '•-

/4

Sj

"* P §

2.V.

/"

•SÍJf"

ia03 j. ^

/

-r^' *T

-'

*

/

lijt'"" -'24.

' ,'

:

!.«

*'* -

%'

i

V, '

fi

a os

/

- ¡132 [

y"'"^ /"" ^^ ,+ '"'•* (8S!

2!

/fc H/Pu (3eff ) V'eff/V ; Atom Ratio Cale, sec" Meas., sec"1

810

34

33 t 3

567 686

74 73

71 ± 6 76 + 3 128 ± 8 133 ± 8

293 293

125 137

6

'°Pu) UO?-2 wtX PuO?(8% 'Pu) UO?-2 wtS PuO>(24£ ¿J'-Pu) UO?-2 wtX PuO^(8# ;!'-°Pu)

243

142

7

2.35 wtS enr. 1%

330**

4 5

2li

160

151 ± 3 (135 ± 1)** 153 t 8

*"~De~scr1bed in" detail 1n Tables XII and XIII. ** H/*3°U atom ratio in this case *** Measured by a different method (Reference 46). T471 J

4.4 Calculations of Power Distributions'-

An analysis has been made of twelve experiments conducted as part of the joint USAEC-CNEN experimental program'- •". This study involves six configurations for each of two fuel types, 2.35 wt% enriched U0? and 2 wU PuO?-U02 (8%?:'Pii) rods. The six configurations were (1) a regular uniform loading of rods; (2) the same loading but with a water hole in the center (I.e., the central fuel rod was

79

removed); (3) water slab (a row of fuel rods removed); (4) water cross; (5) a 7x7 rod array surrounded by water slots; (6) a similar 9x9 rod array. In these twelve experiments* spatial power distributions were measured by gamma-scanning selected fuel rods. The main Interest was 1n the rod-to-rod power distribution, especially the effects of water slots on the power peaking. The series of experiments, ranging from the simplest (regular) array to one simulating a 7 x 7 o r 9 x 9 bundle, provides a systematic test for the evaluation of calculatlonal methods. In the analysis of H?0-moderated and reflected experiments (especially clean critical experiments) one generally assumes the reactor 1s composed of two regions: core and reflector. Few-group cross sections are calculated, assuming that an Infinite medium spectrum applies 1n each region. The few-group cross sections are then used 1n a diffusion theory calculation of T491 In an earlier study1- J, 1t was pointed out that such a tworegion, Infinite medium model generally does not predict the power distribution well, although it :nay yield a satisfactory value for k eff* In 9eneral» tnis method shows a pronounced trend, such that If the power distribution Is normalized at the center of the core, the power near the core-reflector Interface Is consistently underestimated. A simple modification, which resulted In considerably imoroved correlations, was reported in Reference [49]. This simple modification consisted in introducing an extra reflector region (one lattice unit thick, adjacent to the core) which 1s represented by cross sections averaged over a spectrum characteristic of the core. The multi group transport theory codes HRG and Battelle-RevisedTHERMOS were used to generate four-group cross sections for core and reflector regions. These cross sections were used In the twodimensional diffusion theory code 2DB in an x-y calculation of power distributions and kfiff

BO

Four mesh points per cell were used 1n the 2DB calculations. This mesh description was carried out two lattice units Into the reflector; then the mesh points were more widely spaced. An axial buckling of 8.9 nr" was used consistently.

In our current analysis we have compared three variations of our calculational model: Model 1. The usual two-region, infinite-medium model (described in previous section); Model 2. A simple modification consisting of an additional reflector region whose cross sections are obtained from cell calculations performed for the core (described In the previous section). In this model all water gaps also contained these modified reflector cross sections. Model 3. A more detailed representation of the differences in spectrum in successive rows of fuel and water. This was accomplished with THERMOS calculations in slab geometry, with appropriate homogenized regions of core, reflector and gaps; editing was done over the proper spatial points to obtain average cross sections for each "row" of fuel and water. For analysis of the U0?. loadings, five sets of core cross sections were used to represent fuel rods in various locations, and four sets of cross sections were used to represent water. For analysis of the UO?-PuO? loadings, three sets of core cross sections and three sets of water cross sections were used. Power distributions calculated using the three models described above were compared with measured distributions. The trend that is so evident in the regular lattices when using Model 1 (I.e., calculated power consistently underestimated near core-reflector boundary) was significantly reduced when the modified models were used. The simple modification used in Model 2 gives a better correlation than the more refined modification used in Model 3. This 1s evident especially In the case of the mixed oxide loading.

81

The goodness of each correlation Is represented by a standard devia tion, a, defined

£

(ï -«,)*/« - 1)

where N Is the number of rods measured,

• Pmeas meas

and

for

N S - £ 1

The definition of ? Implies an "effective" normalization such that the average fractional deviation, 5, is zero. This definition thus

makes c Independent of the particular choice of normalization, and provides us a meaningful, consistent measure for purposes of comparing methods. The u and buildup of ^'Am were calculated using the BATTELLE-REVISEDTHERMOS^41^ code, and are shown 1n Table XVIII. The results show that calculations in which the effects of ?*iPu decay are Ignored result In

multiplication factors that are too large. Some specific results of exchanging 1.

2lll

Pu for

2lfl

Am are:

A significant reduction 1n "r,f".

2. The change 1s mainly in "ñ" since "f" changes very little. 3.

The buildup of 2!iiAm has a larger reactivity effect than the depletion of ¿aiPu.

The reason an exchange of ?h¿Pu for ? : * J Am changes the "-f" value is clear when one compares the thermal cross sections of the two isotopes. The capture cross sections for the two isotopes are of the same order of magnitude but the fission cross section of ?-"";Pu 1s approximately 300 times that of 2'4lAm. This explains also why "~." changes significantly, whereas "f" does not. The value of "ñ",

reflects the change in total fissions in the fuel rod, which is significant, whereas the value of "f" reflects the change in total absorptions in the rod, which is less significant. The values of "p" and "f are fluxand volume-weighted averages over the specified regions.

84

TABLE XVIII THE EFFECT OF PLUTONIUM-241 DECAY ON CALCULATED VALUES OF "ñ" and "f"

Atom Ratio H/Pu 123.5

Parameter ñ f ñf

152.8 G9

n f ñf

en

203.3

288.7

Reduced ^Pu Density by 10Ï Value of % Decrease

Parameter

1.7869 0.94724 1.6926

0.10 0.02 0.11

1.7791 0.94738 1 .6855

0.53

1.7951

1.7932 0.93*52 1.6758

1.7858 0.93470

0.52

0.93478 1.6780

0.11 0.03 0.13 0.10 0.03 0.13

1.7939 0.91254

0.49

1 .6370

0.49

0.11 0.05 0.16

1.8023

0.46

0.87522 1.5774

0.02

1.8070

1.8027

1.8009

0.91258

0.91234

ñf

1 .6451

1 .6430

r\

1.8107

1.8087

f

0.87537 1.5850

0.87493

1.8132

f

1.8150 0.84403

ñf

1.5319

r,

from Original

Added Corresponding Amount of '-lflAm Value of % Decrease Parameter From Original

1.7886 0.94741 1 .6945

n f

ñf

413.9

Cal cul a ted Parameters Using Original Densities

1.5825

1 .6692

0.00 0.53

0.00 0.52

0.00

0.48

0.84359

0.10 0.05

0.84385

0.44 0.02

1.5296

0.15

1.5248

6.46

REFERENCES

1. N. 0. Freshley, F. E. Panlsko, and R. E. Skavdahl, Th* on High Temperature Bthaoiof of VQyX&t 9ml* in JWTC.JUNE PKS*l% AINE Syoposlui S Nuclear Fuels, Delavan, Wisconsin, October 2. N. 0. Freshley and S. toldsrtth, Gyrating jfe&wtoM» tíitít Plut,ynim Puelo in iww, AINE SyMposlu» on Plutonio» Futís Technology, Scottsdale, AH zona, Octobtr 19(7. 3. Seutton Tîutoniœ Pvogvm - S**lannual P*ogv*o* Report fa* tí» Btfiod texM&m» «2» M8, «CAP-3385-18, Nay 1969.

4. j. R. ToMonto, J. S. Tulenko, and A. Veras, A l*xcn*üpat¿on Vfoapm of llutaniv* SaaifoU in a Soiling UKfcn» ^«krtor, Trant. AM. Nucí. Soc.»

Vol. 12 (1), 3une 1969.

5. XKl-WwttngHaut* Plvtoxtkm ¡toayot* Deuonetofoiion Pvogvctn PvogreM fa* 1** V*rtod Kndtng April. 13?0, «CAP-4167-1, Nay 1970. 6. Plutonitt* y*tli*aUon in Soiling Voto* Soaotot* - Pbtae U - 1669 Report, NEDC-12111, July 1970.

7. N. D. Freshley and T. B. Burley, '/¿taKtonott» Ornatod Conxtío Vuelo, WWL-SA-2412, February 1969.

8. N. 0. Freshley, / CoHrxtsioor. of PelUt and Tipao Ouefoof WM.-SA-3466, July 1970.

9. C. S. CaléMll and K. H. Puechl , ^uícn¿«*-;ro*í«* Díoxtd* Votado* Hollttt MmtJboitaKt, AINE Sy^x>s1UM on Plutonfun Fuels Technology, Scottsdale, Arizona, October 1967. 10. Pltaxjním ffUliaaMon frogmm fvtuntoal /.otiPiïl** CHtseterly Itepovt Mmrt, April, Me* 1970, BÜHL-1442, Juna 1970. p. 3.18-3.24.

11. H. J. Bailey and N. 0. Freshley, Txvadtatten teope**i*a of Mgh

ftat* toufmaUoally .Hapaotod WZ-PIUH Pu*l*> BMC-356, April 1967.

12. H. Balrlot and A. Lhost, Irradiation of Msttotstm Fi»U in m-9, AINE SyMOslm on Plutonlm Fuels Technology, Scottsdale, Arizona, October 1967. 13. 6. KjaarhaUi rhaUi and and E. Rol s tad, Jh-Coro &,»% of tí» Mtotontotit fntefcellc llcn beuxum Vml and Ctaââtog» Nuclear Applications and ogy, Vol. 7 (347), October 1969. Technology 14. Vothnioat AottoiUto 8*po*t - ASC Seaosov VevolojXMnlf ma ?*ahnology f*ogxm» - July» Auguu, Sept***» 2970, BNUL-1 522-1 , October 1970.

88

IS, Raatton Plutoniva Progron - Smt^enmoH Trogr*t» Kmart far Uta P*vtod «fimo 90» 2979» «CAP-3385-20, Octobtr 1969.

16. R. 1. Glbby, 7h* Kffaot, of Pluttwtm Ganta** on «*« Thermal Conductivity

of (tf» JWCi Solid Solutions. J. Nucí. Mataríais, Vol. 38, 163-177,

February 1971.

17. J. K. Bahl.and M. 0. Freshley, Pfefettfe* mA Motion Product in Uiaxd-Cxido Pnalo during Irradiation, Trans. A». Nucí. Soc., Vol. 13 (2), 1970.

18. N. Bober, C. Sari and 6. Schumchtr, Plutonic* xegration in a

'¡radiio* in mm&QxM* total ¿feat during Trradiation, Nuclear Applications and

Ttchnology. Vol. 9, 233 (1970). 20. R. N. Carrol 1 and 0. Slsmn, Pi**ion-Ga» Bklum* ¿teeing Hotticnixg in .^2, Nucí. AppK, 2, 142 (1966).

21 . H. j. F. Notley and J. R. NacEwan, Steyuia* Stiltm* of Kwton Goo >X2 **&» Nuclear Applications and Technology, Vol. 2, 477 (1966). 22. M. D. Freshley and F. E. Panlsko, rha ireaHation tofaarto* of , BNML-366, March 1967. 23. M. D. Freshley, R. 6. Vheeler, 0. N. Batch and 6. N. Hesson, íüri?íK0d fallar* cf » trotteur* 'jtâ* ma Defeated ¡fuel Sod in fffff?., Trans. A». Nucí. Soc., Vol. 9 (2) 1966. 24. 3. H. Locke, :