| | i i ' . w E -

O A K R I D G E N A T I O N A L L A B O R A T O R Y

o p e r a t e d by

U N I O N CARBIDE CORPORATION fo r the

U.S.ATOMIC ENERGY C O M M I S S I O N

ORNL- TM- 2994

LU

Q

Q LU >

LU

o LU

THE DISTRIBUTION OF FISSION PRODUCT TRITIUM IN A

ZIRCALOY-CLAD UO BLANKET ROD FROM PWR-1

J.H, Goode and C M . Cox

NOTICE Th is document contains information of a prel iminary nature

end was prepared pr imar i ly for internal use at tfie Oak Ridge Nat ional

Laboratory. It is sub|ect to rev is ion or correct ion and therefore does

not represent a f ina l report.

nv THIS DOCUMENT IS UNUNITED

LEGAL NOTICE

This report was prepared as an account of Government sponsored work. Neither the United States,

nor the Commission, nor ony person acting on behalf of the Commission

A Makes any warronty or representation, expressed or implied, with respect to the accuracy,

completeness, or usefulness of the information contained in this report, or that the use of

ony information, apparatus, method, or process disclosed in this report moy not infringe

privately owned rights, or

B. Assumes ony l iab i l i t ies with respect to the use of, or for damages resulting from the use of

ony information, apparatus, method, or process disclosed in this report.

As used in the above, "person acting on behalf of the Commission" inci udes any employee or

contractor of the Commission, or employee of such contractor, to the extent that such employee

or controctor of the Commi ssion, or employee of such contractor prepares, di s semi notes, or

provides access to, any information pursuant to his employment or contract with the Commission,

or his employment with such contractor

DISCLAIMER

This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency Thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

DISCLAIMER Portions of this document may be illegible in electronic image products. Images are produced from the best available original document.

INTRALABORATORY CORRESPONDENCE OAK RIDGE NATIONAL LABORATORY

June 19, 1970

To: Recipients of ORNL-TM-2994

Report N o . : ORNL-TM-2994 Classif ication: Unclassified

Author(s): J , H. Goode and C. M . Cox

Subject: The Distribution of Fission Product Tritium in a Z i rca loy-Clad

UO Blanket Rod from PWR-1

Request compliance wi th indicated act ion:

p. 22, third l ine from top should read:

6.84 X 10 moles/cm x 1.37 x 10 g/mole = 9.371 x 10 g Cs/cm

p. 23, second line from bottom should read

7.239 X 10^ rt Apr i l 1, 1970 7.239 X 10^ mCi x 2.22 x 10^ dpm/mCi = 1.61 x 10^^ dpm tri t ium in rod on

p. 23, bottom line should read:

11 3 1.60 X 10 dpm Ho recovered oa AO/ r I I i. J • U ynj—" A ^ = 99.4% recovery of calculated y ie ld 1.61 X 10 dpm H^ formed

^ ' * * ' N , T . Bray, Supervisor Laboratory Records Department Technical Information Division

/wm

INTRALABORATORY CORRESPONDENCE OAK RIDGE NATIONAL LABORATORY

June 19, 1970

To: Recipients of ORNL-TM-2994

Report N o . : ORNL-TM-2994 Classif ication: Unclassified

Author(s): J . H. Goode and C. M . Cox

Subject: The Distribution of Fission Product Tritium in a Z i rca loy-Clad

UO Blanket Rod from PWR-1.

Request compliance wi th indicated act ion:

p. 22, third line from top should read:

6.84 X 10 moles/cm x 1.37 x 10 g/mole = 9.371 x 10 g Cs/cm

p. 23, second line from bottom should read

7.239 X l o ' rt Apr i l 1, 1970

1 9 11 7.239 X 10 mCi x 2.22 x 10 dpm/mCi = 1.61 x 10 dpm trit ium in rod on

p. 23, bottom line should read:

11 3 1.60 X 10 dpm Ho recovered nn AO/ r I I . J • U

Y^— A * = 99.4% recovery of calculated y ie ld 1.61 X 10 dpm H« formed

^ ' ^ N . T . Bray, Supervisor Laboratory Records Department Technical Information Division

/wm

ORNL-TM-2994

Contract No. W-7405-eng-26

CHEMICAL TECHNOLOGY DIVISION

METALS AND CERAMICS DIVISION

THE DISTRIBUTION OF FISSION PRODUCT TRITIUM IN A ZIRCALOY-CLAD UO2 BLANKET ROD FROM PWR-1

J. H. Goode Chemical Technology Division

C. M. Cox Metals and Ceramics Division

L E G A L N O T I C E This report was prepared as an account of Goveroment sponsored work Neither the United Stal«s, nor the Commission, nor any person acting on behalf of the Commlealon

A Makes any warranty or representation, expressed or implied, with respect to the accu­racy, completeness, or usefulness of the Information contained In this report, or that the use of any Information, aFqwratus, method, or process disclosed In this report may not Infringe privately owned rights, or

B Assumes any liabilities with respect to the use of, or for damages resulting from the use of any Information, ai^Mratus method, or process disclosed in this report

As used in the above, "person acting on behalf of the Commission" includes any em­ployee or contractor of tlte Commission, or employee of such contractor, to the extent that such employee or contractor of the Commission, or employee of such contractor prepares disseminates, or provides access tp, -any information pursuant to hia employment or contract with the Commission, or his ejaplbyui^tt j t^^syc^ gwptt-acfOr^

i'M^i,Sjix^[ _ J^l siAtiY . ^ "? -i^tZ

JUNE 1970

OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee

operated by UNION CARBIDE CORPORATION

for the U. S. ATOMIC ENERGY COMMISSION

1

THE DISTRIBUTION OF FISSION PRODUCT TRITIUM IN A ZIRCALOY-CLAD UO2 BLANKET ROD FROM PWR-1

J. H. Goode and C M. Cox

ABSTRACT

The given Z i rca loy-2-c iad UO2 blanket rod had been irradiated to 3 .41% FIMA at a time-averaged linear heat rate of 4.84 kw/ f t , with a clad temperature of 259°C, for the ful l l i fe of the first core of the Shippingport Pressurized Water Reactor. It was destructively examined to determine the distribution of fission product t r i t ium. There was apparently no loss of tr i t ium from the rod; we found 7% of the y ie ld of trit ium in the Zircaloy metal cladding (apparently as the trit ide) and 93% in the UO2 fuel .

INTRODUCTION

Studies are in progress at ORNL to determine the fate of tr i t ium that is produced

In fuel elements by ternary fission of uranium and plutonium. Previous studies, wi th

stainless steel and Z i rca loy-2-c lad U O „ samples from LWR fuels and with stainless

steel-clad PuO^-UO^ samples of LMFBR fuels, indicated that tr i t ium diffused

through stainless steel cladding when the linear heat rating was higher than 5 kw/ f t .

The clad temperature was 105 to 350°C for the < 5 kw / f t ratings and 500 to 1000°C

for the higher ratings (10 to 25 kw/ f t ) . Thus, i t is uncertain whether the primary

variable affecting tr i t ium loss Is clad temperature or the linear heat rating. An

experimental Z i rca loy-2-c lad fuel rod retained about 50% of the trit ium at an

estimated heat rating of 5 kw / f t and a surface temperature of 125°C.

This report presents additional information on the fate of t r i t ium; the subject of

the study was an irradiated blanket rod from the first core of the Shippingport Pres­

surized Water Reactor (PWR). This 0.410-in.-diam by 10.26-in.- long Z i r ca loy -2 -

clad rod (Rod 10) contained 26 normal-enrichment UO^ pellets and was the north­

east corner rod of a total of 120 in Bundle 0320, the middle bundle in a vert ical

2

2-3 stack of 7 in Blanket Assembly K-10. The assembly was inserted into Blanket

Region 2 for the in i t ia l cr i t ica l experiments of PWR-1 (December 18, 1957), and was

removed at the end of l i fe of Core 1 on February 9, 1964, after 27,780 effective fu l l 4

power hours (EFPH) of operation. The rod was sent to ORNL by the Westinghouse

Expended Core Faci l i ty for use in fuel reprocessing studies.

This report represents the combined efforts of Individuals from several divisions

of the Oak Ridge National Laboratory. Members of Chemical Development Section B

of the Chemical Technology Division (J. H. Goode, V. C. A. Vaughen, and technicians

L. A. Byrd, G. D. Davis, and O. L, Kirkland) planned and carried out the experimental

work and the primary tr it ium separations. Operations Division and Metals and Ceramics

Division personnel (R. L. Lines, G. A. Moore) punctured the rod to release fission

gases and sectioned the fuel rod. Analyt ical Chemistry Division members (C. E. Lamb,

F. L. Layton, J. H. Moneyhun, J. E. Morton, W. L Mul l ins, H. A. Parker, J. R. Sites,

R. R. RIckard, and E. I. Wyatt) performed the chemical, radiochemical, and mass

spectrometric analyses. C. M. Cox, of the Fuels Evaluation Group of the Metals and

Ceramics Division, performed the reactor and physics calculations.

PROCEDURES

The intact fuel rod was weighed, measured to determine the diametric dimensions

at 0° and 90° in the mid-plane, punctured to release fission gases, and cut into 10

segments (Fig. 1). About 3.25% of the fuel and cladding was "consumed" by the

alumina cut-of f wheel. The segments were indiv idual ly canned and weighed to

determine the weight of material lost during cutting (Table 1).

We leached or dissolved weighed portions of the fuel , cladding, or both (Table 2)

in a closed system, using an argon-4% hydrogen purge gas to carry vapors or gases

from the dissolver to the sampling points (Fig. 2). In- l ine absolute filters prevented

entrained particles from entering the gas handling equipment. Hot copper oxide

(~ 700°C) was used to convert hydrogen and trit ium to water, and the water was

sorbed onto dry Linde Type 5A molecular sieve. The water was later desorbed at

ORNL DWG 70-6598

SEGMENT ; NUMBER

TOP OF ROD

' 1 8 10 / / ? ^ / / / / J / J / ^ y / / / / / / / / / / / / / / / / J ^ ? J J J

V ^ / / / ' / / / / } ^ J ' / / - / / / / V / / / / / / ^ ^ - ^ y ^ ^ ' / / / /

_L 0.4088 in.

to 0.4100 In.

9/16"\ 1-1/8" I 1-1/8" I 1-1/8" I 1-1/8" I 1-1/8" I 1-1/8" | 1-1/8" | 1-1/8" |9/J6'

(0.26

T CO

PUNCTURE IN SPACE

Fig. 1. Sectioning Diagram, Rod 10, Bundle 0320, Blanket Assembly K-10, PWR Core 1.

4

Table 1. Sectioning of PWR-1 Blanket Rod (Rod 10, Bundle 0320, Assy. K-10)

Starting Weight Cut Weight Cutting Loss

197.9 g 191.4g

6.5 g

Segment Number

1

2

3

4

5

6

7

8

9

10

Total

HRLEL Net Weight^

7.2195

20.8563

23.4652

22.1377

23.3708

23.4252

22.9705

19.8419

21.5345

6.5389

191.3605

Our Net Weightb

7.24

20.84

23.49

22.22

23.37®

23.44

22.93

19.85

21.51

6.57

191.46

Crushed Weight^

d n.a.

20.93

23.54

22.14

n.a.

n.a. '>

n.a.

n.a.

n.a. J

n.a.

Cladding Weight

7.21

> 11.33

3.78®

^ 14.64

6.52

Fuel Weight

55.28

19.59®

73.09

Analyt ical balance at ORNL High Radiation Level Examination Laboratory after cutt ing.

Ohaus "D ia l -o -Gram" triple beam balance in hot cel l at Building 4507.

Segments crushed to separate cladding from fuel .

Not applicable.

Estimated from other weight ratios.

5

Table 2. Analyt ical Scheme for PWR Fuel Rod

Analyzed*^ Segment Portion of Leach Number Segment Treatment No. Solution Of f -Gas M o l . Sieve

1, 10 Al l Leached, 12M_ L-1 X X X HNO3

2 , 3 , 4 Cladding Leached, 12 M L-2 X X X HNO3

2, 3, 4 Cladding^ Dissolved, L-5, X X X Zirflex*^ L-7,

L-8

2 , 3 , 4 Fuel Leached, 12 M_ L-3 X X X HNO3

6, 7, Al I Leached, 12 M L-4 X X X 8, 9 HNO3

6,7, b 8, 9 Cladding Dissolved, L-6 X X X

Z i r f lex^

5 A l l Archive Sample _ _ _

3 85 Analyzed for uranium, plutonium, Hof Kr, gamma-emitting fission products as appl icable.

Random 1 to 2-g samples of sidewall cladding.

' ' 6 M N H . F ~ 1 M N H , N O „ . — 4 — 4 3

ORNL DWG 70-6597

ABSOLUTE HOT FILTERS CuO

(700°C)

fl

M CONDENSER

PURGE Ar-4% H,

DISSOLVER

MOLECULAR SIEVE TRAPS WET TEST

METER

OFF-GAS HOLDER

SAMPLE FOR

SAMPLE FOR

^H^ U, Pu, F.P.'s

Fig. 2. Sample Points for Heavy Elements, Tritium, and Fission Products in PWR Blanket Rod.

7

500-600°C, under vacuum, and col lected in a cold trap at l iquid nitrogen 1/5 .

temperatures, using established techniques. ' The ice was dissolved in a standard

POPOP-PPO scint i l lat ion l iquid for tr i t ium counting in a Packard TrI-Carb sc in -3

t i l la t ion spectrometer. Large (approximately 300 cm ) gas samples were taken from

the gas holder, mixed with hydrogen carrier, and oxidized with hot copper oxide;

the resulting water was concentrated by sorption onto molecular sieve. This water 3

was also desorbed and counted for t r i t ium. Small (5 cm ) aliquots of gas were 85

analyzed for Kr In a gamma spectrometer, as was the molecular sieve prior to

desorption of the water.

The end plugs, cladding segments from which the fuel had been mechanically

removed, the broken fuel fragments, and complete segments were leached and/or

dissolved in boi l ing 12 M HNO^- Aliquots of the solutions were purif ied by

dist i l la t ion, and the trit ium content of the dist i l late was determined by counting

in a l iquid scint i l lat ion spectrometer. Selected segments of the Z i rca loy-2 cladding

were dissolved in 6 M N H . F - - 1 M N H - N O q (Zir f lex solution) to release tr i t ium

and hydrogen contained within the metal; both the solutions and the off-gases were

analyzed for t r i t ium. Finally, portions of the cladding that had been leached in

12 M H N O „ were heated to release contained gases; these gases were analyzed by

mass spectrometry.

Material balances, based on actual weights and analyses of solutions for uranium,

plutonium, and fission products, were made and compared with the calculated quantities.

RESULTS

Z i rca loy-2 and UO^ Material Balance

The overall material balance Indicated that a 3.25% weight loss (approx. 1.0 g

of Z i rca loy-2 and 5,5 g UO«) occurred during sectioning of the rod wi th a cut-of f

saw (Table 1). We accounted for 99.4% of the UO^ and Zircaloy tubing from

Segments "^2, 3, 4, 6, 7, 8, and 9 (Appendix A) and 99.6% of the fuel and cladding

for the entire rod (Appendix B).

8

Burnup

The ni t r ic acid leach solutions, containing fuel and fission products, were 137

analyzed for uranium and Cs (Table 3).

137 Table 3. Uranium and Cs in Rod 10, Bundle 0320

Leach No.

1

2

3

4

Volume (ml)

98

100

250

250

Total

Uranium

mg/ml

0.022

0.639

190.0

259.0

mg

2.2

63.9

47,500.0

64,750.0

112,316.1

dpm/ml

7.86 X 10^

1.68 X 10^

3.93 X 10'°

5.21 X 10'°

137-Cs

dpm

7.86 X 10^

1.68 X 1o'°

9.825 X 10^^

13.025 X l o ' ^

22.87 X l o ' ^

137 We then calculated the amount of Cs in the entire rod:

5 1.123 X 10 mg U ^n-7 A-T I I/^ • I I O O A ^-g^^j 2 — = 127.47 g UO2 in L -1 , - 2 , - 3 , - 4 .

137 152.06 g U O , ,3 ,37 2.287 X 10 dpm Cs x , , ^ . ^ — H T S ^ = 2.728 x 10 dpm Cs in rod

• 127.47 g UOp

(See Appendix C).

137 . The fuel burnup, based on the Cs act iv i ty , was calculated. The fuel p in,

3 containing 0.357-in.-dIam solid UO« pellets (10.08 g/cm ) operated through four

seed loadings, as summarized in Table 4. The burnup calculat ion assumed the

ideal ized operating history shown in Fig. 3, used the fission yields in Table 5, and

used spectrum-averaged neutron cross-sections that were determined by a previously 8 20 3

described technique. The fuel burnup was calculated as 7.67 x 10 fissions/cm

or 3 .41% FIMA. The corresponding fuel isotopic and trit ium concentrations, along

wi th average heat rates at various stages of the fuel pin l i fet ime, are given in Table 6.

9

4 Table 4. Operating History of Shippingport Core 1

Seed 1 Seed 2 Seed 3 Seed 4

Date of ful l power startup

Date of shutdown

Equivalent ful l power days

12-23^57

11-2-59

241.9

5-7-60

8-16-61

329.2

10-24-61

11-26-62

305.4

1-20-63

2-9-64

281.0

Table 5. Fission Yields for Themial Neutrons

Fissionable Isotope

235u

236u

238u

239 Np

239p,

240p,

241 Pu

''W 0.062

(0.062)°

(0.062)

(0,065)

0.0656

(0.0656)

0.0646

atoms/fission

Tritium

0.00013

(0.00013)

(0.00013)

(0.0002)

0.00023

(0.00023)

(0.00023)

Values in parentheses are rough estimates, but gross errors should not signif icantly affect the results, as Indicated in Table 8.

ORNL DWG 70-6596

cr

UJ

UJ

> _J UI cr

r 23 DECM957 9FEa i964

1 1.0 r-

0

1 1 1 1 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400

TIME (days)

Fig. 3. Idealized Power History for PWR Core 1.

11

Table 6. Summary of Calculated Operating Conditions and Isotopic Concentrations for Shippingport Fuel Rod 0320-10°

Exposure Period

I "i." I Seed ., , Initial ••• — April I,

Startup 1 2 3 4 1970

Average heat rate, kw/ft 0,37 5.07 4.61 4,59 4,58

Concentrations, atoms/cm UO2X 10-24

235u 1.7-4'^ 4.28-5 6.57-6 1.15-6 2.33-7 2,33-7

U 0 1,88-5 2,35-5 2.36-5 2.31-5 2.31-5

U 0 1.68-8 2.15-8 2.16-8 2.12-8 ~ 0 238

U 2,426-2 2.41-2 2.39-2 2.37-2 2.35-2 2.35-2 U 0 1,61-8 1,60-8 1.58-8 1.57-8 ~ 0

239 Np 0 2.32-6 2.29-6 2.27-6 2.26-6 ~ 0

239

Pu 0 6.29-5 6.87-5 6.84-5 6.78-5 7.01-5

^'^Pu 0 2.11-5 4,23-5 4.84-5 5,00-5 5.00-5

^'^'Pu 0 4.10-6 1.18-5 1.45-5 1.52-5 1.10-5

l^^Cs 0 1.16-5 2.51-5 3.70-5 4,75-5 4.12-5

^H 0 3.02-8 7.19-8 1,10-7 1.42-7 1.00-7

FPP= 0 1.85-4 3.99-4 5.91-4 7,67-4 7.67-4 % FIMA 0 0.82 1.77 2.63 3.41 3.41

°Based on ' 3 7 Q activity = 1.7965 x 10^' dpm/g UO2 on April 1, 1970, and a UO2 density of 10.18 c/cm3.

^1.7-4= 1.7 x 10"^ etc.

FPP= total fission product pairs.

12

For convenience, the sources of the tr i t ium existing on Apr i l 1, 1970, are summarized

in Table 7.

Table 7. Calculated Sources of Tritium Present on Apri l 1, 1970

Fissionable Isotope Fraction of Total Tritium

235. U 0.100

236u 3 x 1 0 - 5

^^^U 0.017

Np 9 X 10

^^^Pu 0.723

240p o 9 i n - 4 Pu 2.2 X lU

241 Pu 0.160

The burnups in the mirror image bundle (Bundle 0202, Blanket Assembly J-11) 90 T 2

were first estimated to average 6.97 x 10 fissions/cm , although a later document 20 3 9

reported the average burnup as 5.7 x 10 fissions/cm . The other rods in the

mirror image Bundle 0202 had burnups ranging from 4.9 to 7.5 x 10 fissions/cm . 239 235

Stachew reported that the ratio of fissions of Pu to fissions of U in some of

the high-burnup blanket rods at the end of the third seed (21,035 EFPH) was sl ightly 3

greater than 3 :1 ; our calculations indicated a ratio of 2.2:1 at the end of the third

seed l i fe and 3.0:1 at the end of the fourth for rod 10 in Bundle 0320. We expect

that these calculations provide a good estimate although firm data are not available

on other rods examined after the fourth seed.

Fission Product Concentrations

There appears to be only slight differences between the top and bottom halves

of the rod, as Indicated by radiochemical analyses (Table 8).

13

Table 8. Radiochemical Analysis of Top and Bottom Halves of PWR Blanket Rod

Fission Product

Pu a

3H

'5^Cs

" " C e

'="£>.

Top Half of Rod

3.06 X 10 cpm/mg U

1.12 X lO^dpm/mg U

1.94 X 10 dpm/mg U

2.07 X 10^ dpm/mg U

1.88 X 10^ dpm/mg U

5.42 X 10^ dpm/mg U

Bottom Half of Rod

3.17x 10 cpm/mg U

1.08 X 10 dpm/mg U

1.87 X lO'^ dpm/mg U

2.01 X 10^ dpm/mg U

1.61 X 10^ dpm/mg U

5.44 X 10^ dpm/mg U

Linear Heat Rating

Lynam examined rods from a similar blanket assembly (F-5) at the end of the

second refueling (Seed 3), He reported no microstructural changes in the UO« from

a rod calculated to have experienced peak heat ratings, during operation of Core 1, 2 11

of 333,000 Btu/hr- f t , or about 10,3 kw / f t . The l i fet ime average heat rating 2 10

for the given bundle was reported to be 155,000 Btu/hr- f t , or about 5 kw / f t .

Rubin and Lynam calculated that the peak center temperature of the pellets in the 12

high-rated rod was 1440°C and that the pellet surface temperature was 400°C.

Westinghouse's best estimate of the surface temperature of the cladding was 531 °F

(277°C), wi th an average of 520°F (271 °C),

9f 11 The rod we examined was from a bundle having sl ightly lower temperatures.

Based on a total operation of 1157.5 equivalent fu l l power days and an energy

release of 200 Mev/f ission, the time averaged linear heat rate of rod 10 is 4.84 kw/ f t ,

corresponding to a cladding surface heat f lux of 154,000 Btu/hr- f t . By comparison

wi th Table 6, this is seen to be representative of the heat rate throughout the pin

l i fet ime; however, short periods of higher power can be expected due to seed changes

and control rod positioning. Using the time-averaged heat rate, the fuel center

temperature was calculated as 660°C. This calculat ion was made with the PROFIL

14

13 code assuming a cladding surface temperature of 259°C and using the Asamoto

14 . et a l . correlation for UO^ thermal conductivi ty. The fuel surface temperature

was calculated as 350°C.

85 Kr Content of Rod 10, Bundle 0320

The rod was punctured in a high vacuum apparatus above the pellet column by 85 3

laser and the released gas was analyzed for Kr, H, and mass distribution by 15

mass spectrometry. The total volume of released gas was 1.232 cc. Moss analysis

showed the fo l lowing composition:

'5'xe 7% ""? He

H^O

N , +

^2

CO

7%

43

4

5

1

A

CO^

Xe

Kr

1%

< 1

37

2

Ce 23

136. ' ^ ^ X e 26

Xe 44

9 85 The puncture gas contained a total of 1.85 x 10 dpm of Kr, representing a release

12 of 0.16% of the 1.16 X 10 dpm found in the rod (Table 9). By comparison, Lynam

reported fission gas releases of 0.26 to 0.93% in the blanket rods. We actual ly 85

accounted for 88% of the calculated y ie ld of Kr for Rod 10, Bundle 0320 (Appendix

C), but recovered sl ightly more than the calculated y ie ld for the mirror image rod.

We did not count the molecular sieve traps used during the leaches of the end plugs 85

and the unfueled cladding; however, based on the ratio of dpm Kr per mg U in the 85

rod, the quantity of Kr would be insignif icant.

Tritium in Cladding

Samples of leached cladding were heated to 1400°C in a high vacuum apparatus

to release gases contained wi th in the metal. A known "spike" of deuterium in neon

was added to determine the degree of recovery of sample after mass spectrometer

analysis for Hy D_, and T_. The average analyses were: '

15

85 Table 9. Release of Kr From Rod 10

Sample

L-1 Off-gas

L-2 Off-gas

L-3 Off-gas

L-3 Mo l . Sieve

L-4 Off-gas

L-4 M o l . Sieve

L-5 Off-gas

L-6 Off-gas

Puncture Gas

Volume (ml)

4,600

4,900

16,500

-

17,000

-

9,500

6,500

1.232

«=K: dpm/m1

58.5 X 1

1.34 X

2.18x

-

2.73 X

-

3.59 X

5.40 X

1.50 X

io3

10=

10^

10^

10^

10^

lo'

r Total (

S3.91 X

6.57 X

3.60 X

2.23 X

4.65 X

1.24 X

3.41 X

3.51 X

1.85 X

8.34 X

dpm

10^

,0=

lo"

lo'

lo"

lo'

lo'

10«

lo'

io"

b % Total Released

<0.01

0.06

31.03

0.19

40.09

0.11

0.29

0.03

0.16

71.96'=

See Table 2 for identi f icat ion of leaches (L-).

L

Remaining fraction in undissolved Segment ^5 and cutting losses. The total quantity of Kr was calculated to be:

Q-iA i n ' l ^ 85^ 15Z06 g UO2 , , . i n l 2 . 85^ . , 8 . 3 4 x 1 0 dpm Kr x . - - , . - , ^ , .-.^ = 1.16 x 10 dpm Kr in rod, 127.47 g UO2

16

Cladding Segment 7. "Z Numbers ppm ppm

2 , 3 , 4 <0 .02 64-130

6, 7, 8, 9 < 0.02 9

These analyses for segments 2, 3, and 4 generally confirm those of Westinghouse for

the hydrogen content of the cladding; their analyses ranged from 52 to 93 ppm after 4

Seed 4. The trit ium content was too low to be measured by mass spectrometry.

We dissolved four samples of Z i rca loy-2 cladding in Zi r f lex Reagent, 6 M

NH,F—1 M N H . N O o * Three of these samples had been leached with 12 M

HNO« and one had not been leached. Two samples (L-5 and L-6) were dissolved

in the hot cel l using the usual apparatus shown in Fig. 2, and two (L-7 and L-8)

were removed from the hot cell and dissolved in the laboratory. In the laboratory,

we used two gas bubbler-scrubbers containing 1 J ^ H«SO. to remove ammonia from

the gas stream, and a dry- ice-cooled cold trap was used to remove the water vapor.

The scrubbers were inserted between the reflux condenser and the copper oxide unit.

A l l solutions and gases were analyzed for trit ium after each dissolution (Table 10),

Table 10. Tritium Content of Z i rca loy-2 Cladding

Sample H_, dpm x 10 3|_ Leach from Weight, 2

No. Segment No. g Solution Scrubbers M o l . Sieve Gas dpm/g Z r - 2

0.02 4.81 X 10^

0.01 5.48 X 10^

<0.01 2.75 X 10^

<0.01 3.49 X 10^

Leached with ni t r ic acid.

Not leached with ni t r ic ac id.

L-5°

L-6''

L-7°

L-8°

2,3,4

2,3,4

2,3,4

2,3,4

1.35

2.10

1.46

1.02

0.32

0.68

0.36

0.33

none

none

0.01

<0.01

6.15

11.00

3.64

3.23

17

A negl igible amount of tr i t ium was found in the sulfuric acid scrubbers and,

since 90-95% was found as water on the molecular sieve, the results suggest that

the tr i t ium was probably in the metal as zirconium tr i t ide. Metal lographic exam­

ination of Zircaloy cladding at Westinghouse showed that the hydrogen was present 4

as hydride platelets distributed near the outer (cooler) portion of the cladding.

The tr i t ide-hydr ide apparently reacted with the water in the Zi r f lex reagent,

releasing T „ -H_ , rather than being evolved as N H ^ or NTo» Greater than 80%

of the y ie ld of ammonia from the Zircaloy was found in the scrubber. Another

possibility is that the tr i t ium may have been held as a gas in the Zircaloy lat t ice;

however, the metallographic evidence of hydride platelets suggests that this is not

the case. Addit ional laboratory work to answer this question is indicated.

o Assuming an average trit ium content of 3,68 x 10 dpm/g (L-5, L-7, L-8) for

the Zircaloy, the 29,75 g of sidewall cladding would contain 1.13 x 10 dpm of

t r i t ium. (The two end plugs, on the basis of total tr i t ium per unit area of interior

surface, would contain only about 2 x 1 0 dpm.) This quantity of tr i t ium is

equivalent to about 17 parts per b i l l ion by weight, thus confirming the mass spectro­

meter analyses.

Tritium in the UO^

Leach solutions 1 through 4 were analyzed for t r i t ium, other fission products,

and uranium and plutonium (Table 11), The dissolved fuel , as a whole, contained

1.10 X 10 dpm/mg U. Thus, we recovered 1.237 x 10 dpm of H„ from 127.49 g 11

of UO^ In L-1 through L-4, or a total of 1.483 x 10 dpm from al l of the fuel in

the rod.

Tritium in Puncture Gas

85 After determination of the Kr content, we analyzed 0.546 cc of the puncture

4 gas for t r i t ium, and found 7.53 x 10 dpm. The total volume, 1.232 cc, therefore

5 contained 1.70 x 10 dpm, or about 0.001% of the trit ium in the rod.

18

Table 11. Tritium Content of UO« Fuel

Leach No.

1

2

3

4

Vol., ml

98

100

250

250

3 H„ in Solution,

dpm E

2.10 X 10^

5.16 X 10^

5.32x10^°

6.98 X l o ' °

Other ^H^°

dpm E

0.37 X 10^

0.02 X 10^

0.02 X 10^°

0.03 X l o ' °

u. Emg

2.2

63.9

4.75 X

6.47 X

10^

10^

^ H , dpm/mg U

1.12x 10*

8.11 X 10^

1.12x 10^

1.08 X 10^

Molecular sieve traps, off-gases, etc.

Overal l Tritium Distribution

Tritium distribution in the complete rod was:

Zi rcaloy cladding: 1.13 x 10 dpm = 7.08%

U O , fuel pellets: 1.483 x l o ' ^ dpm = 92.92% ^ 5

Puncture gas: 1.70 x 10 dpm = < 0 . 0 1 %

Total 1.596 X l o ' ' dpm = 100%

We recovered 99.4% of the theoretical y ie ld of trit ium (Appendix C), indicating

that there was no signif icant loss of trit ium from this blanket rod, which had operated

at a time-averaged linear heat rating of 4.84 kw / f t and a clad temperature of 259°C.

About 93% was in the UO« fuel and 7% In the Z i rca loy-2 cladding.

REFERENCES

1. J. H. Goode and V. C. A . Vaughen, ORNL Experiments on the Behavior of Tritium

During Head-End Processing of Irradiated Reactor Fuels, ORNL-TM-2793

(February 1970).

2. H. M. Roth, USAEC, Letter to J. A. Swartout, Deputy Director, Oak Ridge

National Laboratory (April 1964).

19

J. C, Stachew, Isotopic Analysis of Natural UO9 Fuel Irradiated to 22,000

Mwd/MTU: Theory vs Experiment, WAPD-TM-766 (August 1968).

A, J, Fiorel l l and C, Santore, ed., PWR Core 1 Component Exomination Program

Summary Report, WAPD-320 (October 1969).

J. H. Goode, Hot Cel l Evaluation of the Release of Tritium and Krypton-85

During Processing of ThO^-UO^ Fuels, ORNL-3956 (April 1966).

235 M. E. Meek and B, F. Rider, Summary of Fission Product Yields for U, oqo nog oA] ^ ^ ° U , Pu, and Pu, APED-5398, Class I (March 1, 1968),

Dudey, et a l „ Chem, Eng. Highlights, May 1967 to Apr i l 1968, ANL-7450.

Fuels and Materials Development Program, Quarterly Progress Report for

Period Ending March 3, 1969, ORNL-4420 (August 1969), pp. 25-28.

R. L. Henke, ed,. Summary Design Report for Melbg Test in PWR Core 2

Blanket, WAPD-TM-483 (March 1966).

M. Peterson, Bettis Atomic Power Laboratory, Personal Communication.

L. R, Lynam, Metal lurgical Examination of PWR Core 1 Blanket Fuel Rods

at the End of the Third Seed Life, WAPD-TM-433 (August 1964),

B, Rubin and L, R, Lynam, "Examination of Fuel and Cladding in PWR-1

Blanket Rods," Nuc l . AppI, 2, 499-506 (1966).

C, M, Cox and F. J. Homan, PROFIL - A One-Dimensional FORTRAN IV

Program for Computing Steady-State Temperature Distributions in Cyl indrical

Ceramic Fuels, ORNL-TM-2443 (March 1969).

R. R. Asamoto, F. Anselln, and A. E, Conti , The Effect of Density on the Thermal

Conductivity of Uranium Dioxide, GEAP-5493 (Apri l 1968),

G. A. Moore, ORNL, personal communication,

J, R, Sites, ORNL Analyt ical Chemistry Division, personal communication.

C. E. Lamb and F. L. Layton, ORNL Analyt ical Chemistry Division, personal

communication.

20

APPENDIX A

Experimental Material Balance (Rod 10, Bundle 0320)

Starting wt. rod

Our cut wt.

Wt. end plugs

Wt. fueled tubing

Archive sample {^5}

Fuel + clad for expts.

197.90 g

191.46 g

-13.73 g

177.73 g

-23.37 g

154.36 g

(1.034 factor to original rod)

(3.78 g tubing, 19.59 g UO2)

(Seg. *Z 3, 4, 6, 7, 8, 9)

Seg. #2, 3, 4 E

Leached clad

UO2 wt.

UO^ by anal.

UO^ Difference

66.61 g

-17.33 g

55.28 g

-53.99 g

1.29 g (-2.3%)

Seg. #6, 7, 8, 9 E

Leached clad

UO2 wt.

UO^ by anal.

UO« difference

87.73 g

-14.64 g

73.09 g

-73.48 g

0.89 g (+ 1.2%)

Total U O ^ found (by anal.)

Total cladding (by wt.)

Fuel + clad found

Fuel + clad input

Difference

Recovery

127.47 g

25.97 g

153.44 g

154.36 g

0.92 g (-0.60%)

99.40%

21

APPENDIX B

Experimental Material Balance

UO2 in Seg. *2, 3, 4 53.99 g

UO2 in Seg. #5 19.59 g

UO2 in Seg. #6, 7, 8, 9 73.48 g

Total UO2 147.06 g x 1.034 cutt ing loss = 152.06 g

Z r -2 in end plugs 13.73 g

Z r -2 in Seg. ^^Z 3, 4 11.33g

Z r -2 in Seg. #5 3.78 g

Z r - 2 in Seg. #6, 7, 8, 9 14.64 g

Total Z r -2 43.48 g x 1.034 cutting loss = 44.96 g

152.06 g UO2 + 4 4 9 6 g Z r -2 = 197.02 g total found.

Starting wt. 197.90 g

Calc. starting wt. -197.02 g

Difference 0.88 g (-0.44%)

Recovery 99.56%

22

APPENDIX C

Calculated Fission Product Yields and Recoveries

Cs: From Table 6, there are 4 .12x 10 atoms Cs/cm present today.

19 3 4.12 X 10 atoms/cm , „> ir ."5 • 137_ / 3

= 6.84 X 10 moles Cs/cm . 23

6.023 X 10 atoms/mole

C O o 0 107 0

6.84 X 10 moles/cm x 1.37 x 10 g/mole = 9.731 x 10 g Cs/cm .

9.371 X 10"^ g /cm^ x 9.85 x lo"^ mCi /g = 9.230 x 10^ m C i / c m ^

9.230 X 10^ mCi/cm'^ x 2.22 x 10^ dpm/mCi = 2.05 x 10^^ dpm ' '^ '^Cs/cm^

1 3 12 3 13 137 1.494 X 10 cm X 2.05 x 10 dpm/cm = 3.06 x 10 dpm Cs in rod today.

2.73 X 10 dpm Cs recovered „ - „ , . . ^ p5= = oy% recovery or calculated y ie ld .

3.06 X 1 0 ^ dpm Cs y ie ld

Kr: Calculations indicated 75°/o total fissions due to Pu and 25% due to 235 35

U (p. 11, text). Y ie ld of Kr should therefore be:

(0.75)(1.215 X 1 0 " ' % yield) + (0.25)(3.049 x 10"^% yield) = 0.09112 + 0.07622

oc Kr y ie ld = 0.1673%.

From Table 6, 7.65 x 10 fissions/cm x 1.494 cm = 1.146 x 10

total fissions in rod.

22 3 19 1.146 X 10 fissions X 1.673 x 10 atoms/fission = 1.917 x 10 atoms

Kr formed.

23

1 .917x10 atoms Kr o i co i n " ^ i £5 . , =75 = 3.1b2x 10 moles Kr formed.

6.023 X 10 atoms/mole

3.182 X lO"^ moles x 8.5 x 10 g/mole = 2.705 x 10~ g Kr formed in rod.

2.705 X 10~ g X 4061 x 10 C i / g = 10.98 x lO" Ci Kr formed in rod.

19 19 8*5 1.098 C i x 2.22 X 10 dpm/Ci = 2.437 x 10 dpm Kr formed.

10 10 ft'^ 2.437 X 10 dpm x 0.54 decay factor = 1.316 x 10 dpm Kr today.

17 85 1.16 X 10 dpm Kr recovered QO-. , . i . • • i .

pr—-—j™ = 88% recovery ot calculated y ie ld . 1.32 X 10 dpm Kr formed

um: From Table 6, there were

17 3 1 3 18 1.00 X 10 atoms/cm x 1,494 x 10 cm = 1.494 x 10 atoms tr i t ium

present today.

18 1.494 X 10 atoms trit ium ^ . „ i n " ^ i ^ •..' • j

= 2.4o X lU moles tr i t ium m rod. 23

6.023 X 10 atoms/mole

2.48 X 10 moles x 3 g/mole = 7.44 x 10 g tr i t ium present.

7.44 x 10 g X 9.73 x 10 mCl /g = 7.239 x 10 mCi tr i t ium present.

7.239 X 10^ mCI x 2.22 x 10^ dpm/mCi = 1.61 x 10^° dpm tri t ium in rod

on Apri l 1, 1970.

10 3 1.60 x 10 dpm H2 recovered

yrr s = 99.4% recovery of calculated y ie ld . 1.61 X 10 dpm H„ formed

25

INTERNAL DISTRIBUTION

R. D. Ackley R. E. Adams C. M. Adamson, Jr, L. G. Alexander E. D. Arnold M. J. Bell R. L Bennett R. E. Blanco J. O. Blomeke E. S. Bomar W. D. Bond G. E. Boyd R. A. Bradley R. E. Brooksbank K. B. Brown F. R. Bruce W. D. Burch W. H. Carr W. L„ Carter G. 1. Gathers J. M. Chandler S. D. Clinton H. Eo Cochran

C. F. Coleman J. H. Coobs C. M . Cox R. L. Cox D. J, Crouse F. L Culler, Jr. J. E. Cunningham F. C Davis W. Davis, J r H. G. Duggan J. H. Evans D. E. Ferguson L. M. Ferris B. C. Finney R. B. Fitts J, D. Flynn M., H. Fontana E. J. Frederick J. H. Frye, Jr. W. Fulkerson

53. 5 4 55.

56-65. 66. 67. 68. 69. 70. 7 1 . 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

F, J. Furman H. W. Godbee H. E. Goel ler J. H. Goode A. T. Gresky W. R. Grimes Wo S. Groen ie r

P. A. Haas R. G. Haire R. L. Hamner W. O. Harms F. E. Harrington C, C Hows, Jr. M. R. Hi l l N. W. Hil l F. J. Homan D. E. Horner F. J. Hurst A. R. Irvine F. A. Kappelmann P, R. Kasten A. H. Kibbey F, G. Kitts B. B. Klima C. E. Lamb J. M. Leitnaker T. B. Lindemer C. S. Lisser R. E. Leuze M„ H. Lloyd J. T. Long A. L, Lotts R. S. Lowrie H. G. MacPherson R. E. MacPherson, D. W. Magnuson S. Mann E. J. Manthos C. D. Mart in, Jr. J. P. McBride, Jr. R. W. McClung K, H. McCorkle W. L McDuffee

26

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.

A. B. Meservey J. G. Moore W. L. Moore L. E. Morse C. A. Mossman J. P. Nichols E. L. Nicholson K. J. Notz C. H. Odom A. R. Olsen J. R. Parrott W. A. Pate P. Patriarca W. L. Pattison W. H. Pechin R. B. Pratt R. H. Rainey J. M. Robbins J. T. Roberts J. D. Rollins A. D. Ryon C. F. Sanders J. M. Schmitt J. L. Scott J. D. Sease L. B. Shappert J. W. Snider W. G. Stockdale W. C. T. Stoddart O . K« Tallent R. E. Thoma R. E. Toucey D. B. Trauger W. E. Unger V» C. A. Vaughen L N . Washburn D. C. Watkin C. D. Watson A. M . Weinberg J. R. Weir G. A. West M . E. Whatley W. M . Woods E. 1. Wroblewski R. G. Wymer

152. 153.

154-155. 156.

157-166. 167. 168. 169.

O. O. Yarbro H. C. Young Central Research Library Document Reference Section Laboratory Records Laboratory Records, RC ORNL Patent Of f ice Nuclear Safety Information Cente

27

EXTERNAL DISTRIBUTION

AEC, Washington 20545 170. 171. 172. 173. 174 175. 176. 177. 178.

179-182. 183. 184

185-188. 189. 190. 191. 192. 193.

W. G. Belter, RDT P. R. Clark, Division of Nuclear Reactors J. Crawford, RDT G. W. Cunningham, RDT A. Van Echo, RDT J. S. Gr i f fo, Div. of Space Nuclear Systems N . Haberman, RDT P. A. Halpine, RDT J. A. McBride, Div. of Materials Licensing W. H. McVey, RDT W. A. Nelson, RDT H. Schneider, RDT J. M . Simmons, RDT E. E. Sinclair, RDT A. N . Tardlff, RDT W. R. Voigt, Jr., Div. of Production C. E. Weber, RDT M. J. Whitman, RDT

AEC - A N U USAEC, 9800 S. Cass Avenue, Argonne, I l l inois 60439 194. M . E. Jackson, RDT Senior Site Representative

AEC - Canoga Park Area Of f ice , USAEC, P. O. Box 591, Canoga Park, Cal i fornia 91305 195. C. W. Richards

AEC - Gul f General Atomic, P. O . Box 2325, San Diego, Cal i f . 92112 196. R. H. Ball, RDT Senior Site Representative

AEC - IDO, P. O. Box 2108, Idaho Falls, Idaho 83401 197. K. K. Kennedy, Idaho Operations Of f i ce

AEC - ORNL RDT Site Of f ice 198. D. F. Cope, RDT Senior Site Representative 199. R. L. Philippone

AEC - ORO, Oak Ridge, Tennessee 37830 200-214 DTIE

215. Laboratory and University Division

•

28

AEC - Richland Operations Of f i ce , USAEC, P. O. Box 550, Richland, Washington 99352

216. W. Devine, Jr., Director, Production Reactor Division 217. P. H. Hoisted, RDT Senior Site Representative, PNL 218. J. Morabito, FFTF Project Manager, RDT 219. J. Shirley, FFTF Project Administrator, RL

A l l i ed Chemical Nuclear Products, Inc., Gen. Chem. Div., Box 35, Florham Park, N. J. 07932

220. R. I. Newman 221. E. M. Shank

ANL, 9700 South Cass Avenue, Argonne, I l l inois 60439 222. A. Amorosi, LMFBR Program Of f ice 223. Leslie Burris, Jr., Associate Director of Chemical Engineering Div. 224. S. Lawroski 225. W. J. Mecham 226. M . Nevi t t , Director, Metal lurgy Division 227. W. B. Seefeldt 228. C. E. Stevenson 229. R. C. Vogel 230. D, S. Webster

ATCOR, Inc., 5 Westchester Plaza, Elmsford, New York 10523 231. Lois Seulowitz, Technical Librarian

At lant ic Richfield Hanford Co., At tent ion: Document Control, P. O. Box 250, Richland, Washington 99352

232. W. P. McCue 233. G. C. Oberg 234 M i l t Szulinski 235. R. E. Tomlinson

Atomics International, Div. of North American Rockwell, Inc., P. O. Box 309, Canoga Park, Cal i fornia 91304

236. R. Aron stein 237-240. Simcha Golan

241. C. A. Tri l l ing

Babcock & Wilcox, Research and Development Div., P. O. Box 1260, Lynchburg, Virginia 24505

242. E. M. Croft 243. E. W. Dewell

29

Battelle Pacific Northwest Laboratory, P. O. Box 999, Richland, Washington 99352 244 E. R. Astley 245. D. C. Boyd 246. Ro E. Burns 247. E. D. Clayton 248. Director, PNL 249. E. D. Grazzini 250. D. E. de Ha las 251. A. M . Piatt 252. R. C. Smith

BNL, Upton, New York 11973 253. E. E. Winche

Combustion Engineering, Inc., Nuclear Power Dept., P. O. Box 500, Windsor, Connecticut 06095

254-255. W. P. Stoker

E. I. du Pont de Nemours & Co., Savannah River Laboratory, Aiken, S. C. 29801 256. L. H. Meyer 257. D. A. Orth 258. E. B. Sheldon

General Electric Company, 283 Brokaw Rd., Santa Clara, Cal i f . 95050 259. R. G. Barned 260. A. So Gibson 261. W. H. Reas

Gul f General Atomic, P. O. Box 608, San Diego, Cali fornia 92112 262. Stanley Donelson 263. A. J. Good John

264-265 R. H. Graham 266. A. R. Matheson 267. R. E. Norman 268. J. N . Siltanen 269. K. B. Steyer

Idaho Nuclear Corporation, P. O. Box 1845, Idaho Falls, Idaho 83401 270. J. A. Buck ham 271. W. J. Maeck

Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 272. Manson Benedict

National Lead Company of Ohio, Cincinnat i , Ohio 273. S. Marshall

30

National Lead, New Projects Dept., 11 Broadway, New York, N. Y. 10000 274 S. H. Brown

Nuclear Fuel Services, Inc., Wheaton, Maryland 20902 275. W. H. Lewis, Vice President

Nuclear Materials and Equipment Corp., Apol lo, Pennsylvania 15613 276. C . S . C a l d w e l l 277. W. J. Ross 278. M . Selman

United Nuclear Corp., Grasslands Rd., Elmsford, New York 10523 279. A. Strasser

Westinghouse Electric Corp., P. O. Box 158, Madison, Pa. 15663 280. E. F. Beckett 281. W. C. Ritz 282. J, D. Sutherland