An Overview of Radiolysis Studies for the Molten Salt

Reactor Experiment (MSRE) Remediation Project

Global 2001 Conference

Paris, France

A. S. Icenhour, D. F. Williams, L. D. Trowbridge, L. M. Toth, G. D. Del Cul

Oak Ridge National Laboratory

Oak Ridge, Tennessee, USA

865-576-5315, icenhouras@ornl.gov

The MSRE was operated from 1965 to 1969 as a test-bed for molten salt reactor technology

• Fuel consisted of LiF-BeF2-ZrF4-UF4

• Initially used 235U, final fuel charge 233U (~ 220 ppm 232U)• Reactor shut down in December 1969—fuel salt drained into

two tanks where it solidified and has remained for more than 30 years

• In 1994 it was discovered that 233UF6 had migrated throughout the reactor piping systems, leading to several very dangerous conditions

• A remediation program was initiated to address the hazards and to remove and store the 233U

Simplified overview of the MSRE systemO R N L D W G 9 9C -12 R

0 10 20

S C A LE (ft)

D R A IN TA N K C E L L

R E A C T O R C E L L

R E A C T O R V E S S E L

O F F -G A S L IN E

VA LV E 5 6 1

H IG H B AY

V E N TH O U S E

H E PAF ILT E R S

M O N ITO R E DS TA C K

C H A R C O A L B EDC EL L

A U X ILIA RYC H A R C O A L

B ED S

C O O L A N T S A LTD R A IN TA N K

F U E L D R A INTA N K S

F U ELP R O C ES S IN G

C E LL

R E A C T IV E G A SR E M O VA L S Y S T EM

S A M P LE REN R IC H E R B O X

Characteristics of 233U

• Relatively high alpha activity (~5000 times 235U, ~0.2 times 239Pu)

• 233U contains an impurity isotope, 232U (concentrations typically in the range 10-1500 ppm)

• A daughter of 232U (208Tl, 2.6 MeV gamma) results in high gamma radiation fields

2 3 25 3 1

2 2 85 4 0

2 2 45 6 8

2 2 0

6 2 9

2 1 6

6 7 8

0 5 02 1 2 2 1 2 2 1 2

8 7 8 3 6 %

2 0 8

6 8 9 1 9 1 3 6 6 5 5 6

0 1 5

2 9 86 0 5

1 0 64

Uy

Th Ra Rn

Po

Po B i Pb

Pb

M eV

M eV

. . . .

.

..

. . .

.

.

.

.

y d s

s

n s m

h

M eV M eV

M eV

M eV

M eV

6 4 %

2 .1 7 M eV

0 .1 M eV

2 0 8

= 2 .6 M eV m

T l

3 0 5.

During the remediation project, a number of needed radiolysis studies were identified

Material Studied Reason

MSRE fuel salt Determine cause of UF6 production

Uranium-laden charcoal Evaluate potential for production ofexplosive CxF compounds

2NaF UF6 Determine amount of pressurizationthat could be experienced in storage

Uranium oxides containingresidual fluorides and water

Evaluate radiolytic effects for long-term storage of these materials

Gamma radiation sources

ORNL 60Co irradiator~105 rad/h

High Flux Isotope Reactorspent nuclear fuel element

~107– 108 rad/h

G age

Valve

Scupper bracket

Tubing

Sta in less stee llifting fram e

Irrad ia tionvessel

SNF e lem enton a cadm ium

lined post

Pool floor

Pool wa ll

Scupper

Pool w ater leve l

OR NL DW G 99C-487

HFIR SNF irradiation

configuration

Typical gas-yield curve for radiolysis experimentsNot all characteristics may be seen

D ose (W -h /g so lid )

G-value = 2.68 x slope

Plateau

Inductionperiod

=# molecules produced

O RNL DW G 99C-6870

100 eV deposited in solid

Gas

yie

ldm

mol

pro

duct

g so

lid

MSRE Fuel Salt

• LiF-BeF2-ZrF4-238UF4 (64.5, 30.3, 5.0, 0.13 mol%) irradiated in HFIR SNF elements

• Pressure monitoring and gas sampling used to quantify radiolytic yields and gas composition

• Measured G(F2) = 0.012–0.02 molecules F2/100 eV, example reactions:LiF + Li· + F·

F· + F· F2

• Induction period observed (i.e., no gas was produced) and maximum damage limit (~2%) demonstrated at high total dose

• Periodic heating of the salt, combined with the presence of F2 produced the volatile UF6 at MSRE: UF4 + F2 UF6

Gamma radiolysis of MSRE surrogate fuel salt (LiF-BeF2-ZrF4-238UF4) in HFIR SNF elements

2NaF·UF6 complex

• 233U sorbed on NaF pellets forming 2NaF·UF6

• Containers of 2NaF·UF6 referred to as “NaF traps”• Two of the traps have been instrumented with pressure transducers• Pressure rise observed for these traps

– UF6 + (,,) UF5 + 1/2F2

– uranium may be reduced further

• Estimated G-values– G(F2) ~ 0.44 molecules F2/100 eV

– G,(F2) ~ 0.02 molecules F2/100 eV (based on work with other fluoride salts, i.e., LiF, BeF2)

Pressure monitoring data for two NaF traps

20

30

40

50

60

70

80

90

100 deg F

0

50

100

150

200

250

psia

10/28/95 07/24/98 04/19/01

Trap 6 Trap 1 Temp. Predict 1 Predict 6

Pressure in NaF/UF6 traps in Bldg 3019, ORNL

Radiolysis of NH3-treated C2.6F

• Fluorinated charcoal in the charcoal bed at MSRE treated with NH3

to remove explosive potential and forming NH4F

• Radiolysis of NH4F can lead to reformation of CxF

• 60Co irradiation resulted in G(gas) ~ 0.24 molecules/100 eV• Gas analyses of the treated charcoal bed revealed products

consistent with the reported results for the decomposition of NH4F (i.e., production of H2 and N2). The material in the charcoal bed was irradiated by both alpha and gamma radiation.

• Results of these analyses used to establish the frequency of retreatment of the charcoal bed with NH3

Gamma radiolysis of UO2F2·xH2O

• UO2F2·xH2O is an intermediate compound formed in the conversion of UF6 to U3O8

• UO2F2·xH2O irradiated in both the 60Co source and in HFIR SNF elements (x varied from 0.4 to 1.7)

• Color change, CO2/O2 production, and U(IV) increase provide evidence that gamma radiation releases oxygen from the UO2F2 matrix: UO2F2 + UOF2 + O

• Oxygen radicals react with other oxygen radicals or carbon impurities, producing O2 and CO2

• G(O2) = 0.007–0.03 molecules O2/100 eV• A steady state was reached in HFIR SNF elements providing a measure of the

damage limit to UO2F2•xH2O • Damage measured by amount of U(IV): 7–9 wt% at 108 rad/h

HFIR SNF element radiolysis of UO2F2 • 0.4 H2O (loaded in helium, dose rate 107–108 rad/h)

10

12

14

16

18

20

22

0 20 40 60 80 100 120 140 160

Dose (W-h/g)

Pre

ssu

re (

Psi

g)

0

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0.08

Gas

yie

ld (

mm

ol/g

)

G~0.01 molecules gas/100 eV

Inserted in new SNF element

Inserted in new SNF element

Before After

Summary of results for gamma radiolysis experiments for uranium oxides with sorbed water

• Samples of U3O8 and UO3·xH2O with water contents up to 10 wt% were irradiated

• In general, little or no pressure increase observed• Sample pressure decreased for some experiments• Little hydrogen produced• Oxygen depleted for many samples

Follow-on alpha radiolysis experiments on uranium oxides with water and fluoride

impurities are being conducted

• Samples are doped with 244Cm

• Samples are designed for long-term pressure monitoring and for periodic gas sampling

Summary

• A number of radiolysis experiments have been conducted in support of the MSRE remediation program

• Results from these studies were used to establish criteria for the intermediate and long-term storage of materials removed from the MSRE