Behaviour of small-sized BWR fuel under reactivity initiated accident conditions in comparison with standard BWR fuel

Nuclear Engineering and Design 143 (1993) 285-294 North-Holland

285

Behaviour of small-sized BWR fuel under reactivity initiated accident conditions in comparison with standard BWR fuel Kazuaki Yanagisawa

Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Ibaraki-ken 319-11, Japan Received 2 December 1991, revised version 8 September 1992

This report describes the results of reactivity initiated accident (RIA) experiments performed on small-sized BWR fuel for which the outer diameter of 11 mm is smaller than 12.3 mm of a conventional BWR. Evaluation of the fuel behaviour was based on in-core instrumentation data and data obtained from post-pulse irradiation examination. The principal matter of consideration was to determine the fuel failure threshold and the failure mechanism. The results obtained were: (1) The failure threshold of the small-sized BWR fuels was not less than 260 cal/g fuel (220 cal/g fuel in enthalpy) and no measurable differences in failure thresholds were observed between the small-sized BWR, the conventional BWR and the NSRR standard fuels. The failure mechanisms of the small-sized BWR fuels were either fuel melting/brittle fracture or cladding rupture depending upon the fuels whether lined by Zr or not. (2) No measurable difference in the threshold for mechanical energy release were found between the small-sized BWR fuels and the conventional BWR (and NSRR standard) fuels.

1. Introduction

2. Experiment

To reduce the mechanical and thermal load of fuel rods arising either from pellet-cladding mechanical interaction or from rod internal pressure built up by released fission product (FP) gas, the rod outer diameter (O.D.) of BWR fuel has been reduced from 14.3 mm (7 × 7 type) to 12.3 mm (conventional 8 x 8 type) [1]. To further increase fuel performance margins at extended high burn-up stages, the fuel rod with an outer diameter reduced to 11.0 mm was developed and used in commercial BWR's widely, especially in Europe [2-5]. In the present work, we denoted this type of BWR fuel as the small-sized BWR fuel. From the safety point of view, the reliability of the fuels under R I A conditions is studied mainly in order to increase data base in the Licensing Guideline for RIA, Japan [6]. Obtained results were compared with Nuclear Safety Research Reactor (NSRR) standard data which has been used as design limit of R I A [7] and with data obtained from conventional BWR fuels [8,9]. All experiments are made with zero burn-up (fresh fuel) at the N S R R of the Japan Atomic Energy Research Institute (JAERI).

2.1. Test fuel rods Test fuel rods shown in Fig. 1, were fabricated by the Nuclear Fuel Industries, Ltd. The approximate U-enrichment of the U O 2 pellets contained in the fuel rods was 3.9 w / o for the ends of the columns and 10 w / o for the active columns, respectively. All pellets had 96.4% of the theoretical density (T.D.). The cladding was fully annealed Zircaloy-2. Table 1 summarizes the characteristics of the fuel rods. The small-sized BWR fuel rods had three different parameters in the fabrication: the Zr l i n e r / p r e s surization (0.65 MPa), the Zr liner/non-pressurization and non-liner/pressurization, respectively. The thickness of Zr liner was about 10% of that of the cladding. The rod internal pressure (prepressurization by pure helium gas) was changed from rod to rod with a magnitude between 0.1 MPa and 0.65 MPa. The initial plenum volume was about 2.4 cm 3. As reference rods, 8 x 8 BWR fuels consisting of Zr l i n e r / p r e s s u r i z e d ( < 0.6 MPa), Zr l i n e r / n o n pressurized and non-liner/pressurized were used, for

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K. Yanagisawu / Behal,iour of small-sized BWR fuel

286

°'umn ,! 135

/ RODNo L ( Zeroorientation) (~ Top end plug @ Plenum spring @ Push rod @ Movement marker @ Disk (Hf)

(&) @ @ @ @

I -1

q RODBOTTOM Measurementsarein millimetres

Stackmeasurement core @ End pellet (39w/o enrich.) Zircaloy-2cladding UO2 pellet (10w/o enrich) @ Pt/Pt-13*/,Rh thermocouples Spacer (~) Cladding Rod internal pressure extensometer sensor

Fig. 1. Drawing of test fuel rod. which the characteristics were described in detail elsew h e r e [8-10].

2.2. Fuel assemblies and irradiation capsule Two types of fuel assembly were used. T h e first, shown in Fig. 2, was a triplet configuration d e s i g n e d principally for d e t e r m i n i n g rod failure threshold. T h r e e rods were located 120 ° a p a r t on a

pitch circle having a radius of 18 mm. This minimized radial power distortion a n d avoided t h e r m a l interaction a m o n g the rods. T h e m i d p o i n t of the active colu m n of the test rods was located at the N S R R core midplane. As shown in Fig. 1, each rod h a d i n s t r u m e n tations. A t h e r m o c o u p l e to m e a s u r e the t e m p e r a t u r e at the o u t e r surface of the cladding was positioned at the m i d p o i n t of the active column of each rod. A cladding e x t e n s o m e t e r to m e a s u r e the rod axial move-

Table 1 Summary of fuel rod physical parameters Fuel rod

H7-H10, H14

Fuel O/U Density ( M g / m 3) Outer diameter (mm) Enrichment (w/o of 235U) End form Grain size (/zm) Cladding Zr liner Cladding O.D. b (mm) I.D. c (ram) Diametral gap (mm) Prepressurization (MPa) Active fuel column (m)

Sintered and ground UO 2 2.013 2.013 10.56 10.56 9.38 9.38 10.0, 3.9 10.0, 3.9 Dish + chamfer 8 8 Fully annealed Zircaloy-2 Yes a Yes 11.00 11.00 9.55 9.55 0.17 0.17 0.65, pure He 0.10 0.135 0.135 0.123 (10%) d + 0.012 (3.9%) d 2.4 2.4

Plenum volume (m/) Note: a b c d

10% of cladding wall thickness, O.D.: outer diameter, I.D.: inner diameter, enrichment.

H11-H13

H15-H18 2.006 10.56 9.50 10.0, 3.9 8 No 11.00 9.67 0.17 0.65 (I.135 2.4

K. Yanagisawa / Behauiourof small-sized BWR fuel ment, a movement marker to estimate fuel maximum expansion, and a pressure sensor to monitor the change of rod internal pressure, were utilized. The second type of assembly shown in Fig. 2, was a single rod configuration designed principally to investigate fuel fragmentation and the mechanical energy generation. The rod was located in the center of the irradiation capsule. The axial center of the test fuel was set corresponding to the middle of the NSRR core. Adding to the instruments mentioned above, pressure sensors were provided at both ends of the capsule to measure the pressure pulse generated upon fuel fragmentation. A water level sensor (floating buoy) was also installed to measure the movement of the water column ejected upwards by the steam explosion. The floating buoy has a magnet coil inside and travels along the guide stem according to the water column movements. As the span a l to a unit coil is 6 mm, the water column velocity V is given by V= Al/zlt = 6 mm/At, where At is the half-wave period of the output signal from buoy/magnet movement. Details on determination of the water column velocity and generated mechanical energy were described elsewhere [7,9].

-- ~

287

All fuel rods were immersed in stagnant water at room temperature (about 20°C) and atmospheric pressure inside a sealed irradiation capsule.

2.3. Pulse history The half-width of the power of the NSRR pulse irradiation has a minimum of about 4.4 ms at a maximum integral power of 110 MW. s. The value of this width varies from 4.4 to 20 ms depending on the magnitude of inserted reactivity. The effect of pulse width variation in this experiments is, however, negligible since the pulse-width is far below the thermal time constant of the fuel rods. A typical history of power pulse in NSRR is described in detail elsewhere [9]. The integral value of the reactor power P (MW- s) was used to estimate the deposited energy Eg(cal/g fuel) in each test rod. Hence, Eg =KgP, where the power conversion ratio Kg(cal/g fuel per MW" s), is the ratio of fuel rod power to reactor power. This ratio was determined through fuel burn-up analysis [11] taking the radial and axial power skew into consideration. The ratio to convert the deposited energy to radially

200

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Fig. 2. Irradiation capsule for triplet rod configuration (left) and for single rod configuration (right), where the wire net used to protect the capsule inner surface from the fragmented hotter UO 2 fuel is not shown.

258 219

0.701 0.831

D e p o s i t e d e n e r g y ( c a l / g fuel) F u e l e n t h a l p y ( c a l / g fuel)

Fill gas p r e s s u r e : P0 ( M P a ) a R e a c h e d P e a k P r e s s u r e : Pm ( M P a )

. . . .

NF -

C a p s u l e p r e s s u r e : at b o t t o m ( M P a ) at t o p ( M P a ) M a x i m u m w a t e r c o l u m n velocity ( m / s ) Mechanical energy conversion ratio (%)

F a i l u r e ( F ) / N o failure ( N F ) T i m e to f a i l u r e by p r e s s u r e (s)

.

. .

.

NF

0.16

.

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.

4.6 2 . 4 + 1.3

-

1588

0.145 0.194

258 219

529-1 Y 0.1

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.

.

0.16

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1.05 - 0.60

-

0.647 0.749

275 233

Y 0.65

H8

12.5 4.4-+3.0

'~

23.5 12.27

1.79 0.36

1246

0.690 0.804

258 219

N 0.65

H15

a value at 20°C, w h e r e p l e n u m v o l u m e at this c o n d i t i o n w a s 2.4 m l , b incipient c r a c k i n g o c c u r r e d , d a m a g e d m u c h b u t did n o t fail, T/C failed, r e f e r e n c e only.

2.3

8.0 5.3+2.2

M a x i m u m fuel c o l u m n m o v e m e n t ( % )

Peak (%) Average (%)

R e s i d u a l d i a m e t r a l strain

Peak (%) Residual (%)

Axial strain

1.05 0.30

~ Y 0.65

P e a k c l a d d i n g s u r f a c e t e m p e r a t u r e (°C)

H7

F u e l rod No.

Experimental parameter (1) Z r liner (Y : Yes, N : N o ) (2) Fill gas p r e s s u r e ( M P a )

F 5.5

0.46

6.5 3.0-+1.5

-

1612

0.142 0.201

275 233

529-2 Y 0.1

H12

F 3.4

-+ 5.6 2.47

23.4 12.5

2.11 0.57

0.326 0.432

275 233

N 0.65

H16

3.1

F

0.36

6.9 3.8--+1.4

IC b

0.56

NF ~

-

24.9 12.5 _+6.5 0.98

1.92 0.49

1467

0.568 0.698

268 227

N 0.65

HI7

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1231

0.199 0.351

268 227

529-3 Y 0.1

H13

1.00

1764

0.316 0.620

268 227

Y 0.65

H9

S u m m a r y of fuel b e h a v i o u r o f the small-sized B W R b a s e d on i n - c o r e m e a s u r e m e n t s a n d p o s t - p u l s e i r r a d i a t i o n e x a m i n a t i o n

Table 2

H14

HI8

F

0.299 0.523 3.51 0.014

-

-

(1258) d

0.646 (2.41) d

370 314

F

0.295 0.163 6.93 0.048

-

-

(1094) d

0.570 (0.674) d

415 353

F

0.364 0.102 6.32 0.041

-

-

(1943) a

0.638 (1.408) d

407 347

' ~---- 5 2 9 - 4 ------*~---- 529-5 ----0 ~---- 529-6 -----* Y Y N 0.65 0.65 0.65

H10

p,.

K. Yanagisawa / Behauiour of small-sized BWR fuel averaged fuel enthalpy within the fuel pellet has been described elsewhere [12]. It should be noted that the fuel enthalpy was used frequently in the Licensing Guideline for R I A in Japan. The following discussion will present the values of both the deposited energy and the corresponding radially averaged fuel enthalpy. The axial profile was relatively fiat, although locally sharp peaks with about 12% higher than the average were revealed at both the bottom and top ends of the fuel.

3.1. Failure threshold Table 2 summarizes the 6 experiments conducted, 3 experiments with a triplet rod configuration and 3 with a single rod configuration. Fuel behaviour derived from in-core measurements and post-pulse irradiation examination (PIE) is included. Failure thresholds of the fuel rods as a function of energy deposition are shown in Fig. 3. In the figure, data from conventional 8 × 8 BWR referential rods [7-9] were included for comparison. As shown, the small-sized BWR fuels tested

ENERGY DEPOSITION(k Jig. fuel) 10

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ENERGY DEPOSITION(cal/g.fuel)

Fig. 3 Failure/no failure thresholds of the small-sized BWR fuels, conventional 8× 8 BWR fuels [8-10], and NSRR standard fuels [7].

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3. Results and discussion

LEGEND:

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LLI ~) IJ_

0.6

,

,

08.

,

0.0

1.0

PRESSURE DIFFERENCE BETWEEN ROD INTERNAL AND COOLANT(MPa) Fig. 4. Deposited enthalpy of fuel rod vs. difference between internal and external rod pressure, where both conventional LWR fuel design criteria and R I A standard [6] are also

shown.

(hereinafter called the tested fuels) did not fail below a failure limit of the NSRR standard fuels (260 c a l / g fuel) and that of the conventional 8 × 8 BWR fuels. It is important to point out that the value 260 c a l / g fuel (or 220 c a l / g fuel in enthalpy) has been used for Japanese R I A licensing limit established through past NSRR experiments [8,13]. As shown in Fig. 4, fuel enthalpy in each test is plotted as a function of the difference between internal and external rod pressure. Previous NSRR experimental data for fuel failure and acceptable design criteria of LWR fuel are also included [8,13]. The tested fuels did not fail below a failure limit of the NSRR standard fuels and that of the conventional 8 × 8 BWR fuels. The threshold enthalpy for failure of it was estimated to be 227 c a l / g fuel, indicating that the acceptable design criteria for the LWR fuel were also valid. 3.2. Failure mechanism Figure 5 shows metallo/ceramography of three different kinds of the tested fuels failed at 275 c a l / g fuel. Transversal sections shown were cut from locations close to failure. The type Zr liner/pressurized fuel, as shown in the upper part of the photograph, was defected by m e l t /

290

K. YanagLsawa / Bettat iour qf small-sized BWR /~e/

As-pou~nea

BOTTOM

~,o.J

TOP

x,~,,

Fig. 5. Micrograph of Zr lined/pressurized fuel (top), Zr lined/non-pressurized fuel (middle) and non-lined/pressurized fuel (bottom); all specimens were cut close to the failure locations after pulse irradiation of 275 cal/g fuel,

291

K. Yanagisawa / Behaviour of small-sized BWR fuel brittle mechanism. That is, a cladding was defected in brittle manner accompanying with local melt of cladding. Zr liner around the location was melted. There was an interaction between fuel and cladding within 11% of cladding wall thickness. No interaction occurred at other locations. The extent of ballooning was 13% in maximum. External and internal oxide scale thickness were 18% and 0.6%, respectively, of the initial wall thickness. The type Zr liner/non-pressurized fuel, as shown in the middle of the photograph, was defected by melt/brittle mechanism, too. As to this case, oxidation was significant. External and internal oxidation was 17% and 24%. The extent of ballooning was 7% in maximum. The type non-liner/pressurized fuel, as shown in the lower part of the photograph, was defected by ballooning followed by the rupture of cladding. A significant ballooning of 26% in maximum, resulted in a lot of wrinkles at the cladding outer surface. Hence many external notches having maximum depth by 0.053 mm (8% of cladding wall thickness) were caused by this deformation. External oxidation of 4% wall thickness was observed only. Observed failure modes in these tests were typical for the tested fuel parameters. 3.3. Mechanical energy release Figure 6 shows the result of water column movement and capsule pressure transient measurements in the tested Zr liner/pressurized fuel at 415 cal/g fuel. The jumping movement of the water column and cap-

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LEGEND: O NSRR STANDARD FUEL(I~.7 ) A 8x8 BWR TYPE FUEL(I~et8-9 l • THIS EXPERMENT:Srna s ~ BWR

W ~1.5

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350

'

|

. . . .

400

I

450

. . . .

]

. . . .

500

550

ENERGY DEPOSITION (callg.fuet) Fig. 7. Capsule peak pressure generated by fuel fragmentation.

sule pressure increase were initiated about 0.07 s after the pulse peak. The water column was driven to a height of 144 mm after which it fell back to the initial position. The maximum water column velocity was about 6.93 m / s . The pressure measured at the capsule bottom was 0.295 MPa. The nuclear to mechanical energy conversion ratio for this case is about 0.048%. The capsule bottom maximum pressure measurements are summarized in Fig. 7. The data from NSRR standard fuels [7] and conventional 8 × 8 BWR fuels [8,9] are also shown. This shows no significant difference in the threshold of mechanical energy release between the tested fuels and the others. The mechanical energy generation measurements are summarized in Fig. 8. In past NSRR experiments [7], conversion ratios up to the energy level of 600 cal/g fuel have ranged 0.004-1%. The mechanical energy generation from the tested fuels (0.014-0.048%) are within the range of NSRR standard fuels.

I.U 4

b.l ..J ~ 2

It. C) I-Z LLI

(3_ < U

hi 0

~r

TIME(s) Fig. 6. Variation of capsule pressure at bottom and at top, and water column velocity of Zr lined/pressurized fuel at 415 cal/g fuel.

4. Surface temperature of cladding From past NSRR experiments it is known that the cladding temperature is strongly associated with the fuel failure mechanism, especially when the failure occurred by cladding embrittlement due to oxidation a n d / o r local wall thinning at high temperatures. In Fig. 9, peak cladding surface temperatures of the tested fuels at axial mid location are plotted against energy deposition. In addition, data obtained from NSRR standard fuels [7] and those obtained from

K. Yanagisawa / Behat'iour of small-sized BWR [~a'l

292

PEAK FUEL ENTHALPY(cal/g.luel) 100 200 300 4O0 500 I

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2000

Z

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THIS EXPERIMENT

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LEGEND: NSRR STANDARD:~t(7)

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IFGEND: 8x8 BWR :Ref.(8 - 10 )

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I I |I~sVERAGED TEMPERATUREOF NSRR O FUEL ROD: Ref.( 7 )

o.0~

=

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ENERGY

oO I

I I I I 200 400 6OO ENERGY D E P O S I T I O N (cal/g.fuel) Fig. 8. Conversion ratio from nuclear to mechanical energy generation as a function of energy deposition, where N S R R data are also shown [7].

conventional 8 × 8 BWR fuels [8-10] are shown. It is clear that there is no significant difference in temperatures between the tested and conventional 8 × 8 BWR fuels. Temperatures of the tested fuels were comparable with those of NSRR standard fuels, too.

,

I

200

300

DEPOSITION

(cal/g.fuel)

Fig. 9. Measured cladding peak temperatures as a function of energy deposition, where data from conventional 8 x 8 B W R

fuels [8-10] and from N S R R standard fuels [7] are shown.

tively long and ranged 1-6 s, while, for PWR fuels pressurized up to 3.3 MPa, it was relatively short and ranged < 1 s. Hence, the time-to-fuel failure is dependent on the initial pressurization level. Please note that

0.2

I

5. Fuel rod pressure Figure 10 shows time dependent changes of the pressure at a deposition energy of 275 c a l / g fuel. The pressure increased abruptly among the tested fuels just after irradiation starts. The maximum increase of rod pressure at pulse irradiation ranged 0.05-0.3 MPa. The pressures then decreased slightly due to ballooning. At last, the pressure of the tested fuels decreased to the atmospheric level because of failure. In the figure, the time-to-fuel failure is indicated by arrows. The measurement was, as shown in Fig. 11, summarized together with data from fuels with different prepressurization [10,14]. For BWR fuels pressurized up to 0.6 MPa, the time-to-fuel failure was rela-

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TIME(s) Fig. 10. Pressure increase of the tested fuel rods vs. time, where the internal pressure value of each rod was normalized to zero at the beginning of pulse irradiation.

K. Yanagisawa / Behaviour of smafi-sized BWR fuel , "~ 2 8 0 L

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Ref(lO)

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ENERGY DEPOSITION (kJlg.fuel) •~l0.8 0.9 1.0 1.1 1.2

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NSRRFaitureTWe~d . . . . . . . . . . . . . . . . . -I ILeg~d: mess.~'_oe~al I M: Melt, M~=MetURupturel -I I R=Rupture I 1 I i I , l 2.0 3.0 4.0 5.0 6.0 TIME TO FAILURE(s

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TIME TO FAILURE(s)

I--

AZr tinerlunpressunzedl ~

Z UJ

0 Non4iner/wessuTized I ~

@:

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Fig. 11. Energy deposition vs. time to fuel failure for BWR fuels [10] (top) and for PWR fuels [14] (bottom).

-,0.

the time-to-failure of the tested fuels which failed by the melt/brittle mechanism was in the range of 3-6 s.

rr ")." "" " .:. [i'-,

6. C l a d d i n g d i m e n s i o n a l c h a n g e

In Fig. 12, the radial permanent strain of the tested fuels is shown. In addition, data obtained from NSRR standard fuels [7] and from conventional 8 x 8 and 7 x 7 BWR fuels [8-10] are also given. Obviously, the radial permanent strain in the tested n o n - l i n e r / pressurized fuel was similar to that of n o n - l i n e r / pressurized 7 × 7 BWR fuel. It is worthy of mentioning that the wrinkle deformation combined with cladding external notches took place in both cases. The extent of radial permanent strain in the present Zr lined fuels was similar to those of NSRR standard and Zr lined 8 X 8 BWR fuels. In NSRR non-pressurized standard fuel, as shown in Fig. 13, the cladding axial permanent strain and maximum fuel column movement increased linearly with increasing energy deposition. The cladding axial permanent strain in the tested fuels was not so much due to a marked influence of ballooning.

4. C o n c l u s i o n s

The conclusions reached in the present experiments are summarized as follows: (1) The failure thresholds for the small-sized BWR fuel rods were not less than 268 c a l / g fuel (227 c a l / g fuel in enthalpy). Below 260 cal/g fuel (220 cai/g fuel in enthalpy) used as NSRR experimental

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ENERGY DEPOSITION (cal/g.fuel) Fig. 12. Radial permanent strain of the fuel rods vs. energy deposition, where conventional BWR fuels [8-10] and NSRR standard fuels [7] are included.

data line for fuel failure, no fuel failure occurred in the tested fuel rods. The failure mechanism in Zr lined fuel was fuel melting/brittle fracture of the 3.0

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ENERGY DEPOSITION (cai/gfuel) Fig. t3. Axial permanent strain vs. energy deposition; data

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K. Yana~cisawa / Behat'iour qf small-sized BWR liwl

cladding while that in non-lined fuel was ballooning followed by rupture. No significant differences in the threshold for mechanical energy release were found between the tested fuels and the N S R R standard (and the conventional 8 × 8 BWR) fuels. Above 268 c a l / g fuel, the Zr liner tended to melt due to the interaction between the lined cladding and the hot UO 2 fuel. At 275 c a l / g fuel, interaction occurred in the Zr lined/pressurized fuel within 11% of the cladding wall. Around failure location, external and internal oxidation occurred in the tested fuel failed by a fuel melt/brittle failure mechanism. However, external oxidation took place only in the tested fuel, failed by rupture. It was revealed that the monitoring of the rod internal pressure could detect fuel failure. The time to fuel failure in the tested fuels ranged 3-6 s. In general, the time to fuel failure of BWR fuels pressurized up to 0.6 MPa was longer than that of PWR fuels pressurized up to 3.3 MPa.

Acknowledgments Appreciation is addressed to Messrs. T. Sato, K. Takeuchi, K. Tsuda, and M. Goto, Nuclear Fuel Industries, Ltd., Japan for the fabrication and characterization of the fuel rods. The technical support obtained from Mr. O. Horiki, Head of the NSRR Operation Section, JAERI, during the course of the N S R R exper-

iment and the efforts of Mr. C. Dianshan, Associate professor, Chinese Institute of Atomic Energy, who made the data evaluation for the mechanical energy release from fragmented fuel while at JAERI, arc gratefully acknowledged, Thanks arc also due to Dr. Fujishiro, Head of Reactivity Accident Lab. ,IAERI for his fruitful discussions and comments on this work.

References [l] K. Yanagisawa, JAERI-M 90-120 (1990), in Japanese. [2] D. Bender, O. Bender and D. Urban, Kerntechnik 50(4) (1987) 222. [3] T. Watanabe and K. Sato, Proc. Sixth Pacific Basin Nuclear Conf., Beijing, 454 (1987). [4] G. Lill, H. Knaab and D. Urban, proc. ANS-ENS Int. Topical Mtg. on LWR Fuel Performance, Vol. l. 74 (1991). [5] T. Howe, S. Djurle and G. Lysell, ibid, vol. 2, 828 (1991). [6] Nucl. Safety Comm., Evaluation Guideline for Reactivity Initiated Events in Light Water Power Reactors (in Japanese, 1984). [7] T. Fujishiro, T. Inabe and M. Sobajima, Proc. Annual CNA/CNS Conf. Canada (1988). [8] T. Hosbi, K. Iwata, T. Yoshimura and M. Ishikawa, JAERI-M 8836 (1980). [9] K. Yanagisawa, T. Fujishiro, A. Negrini and F. Franco, J. Nucl. Sci. Technol. 27(1) (1990) 56. [10] K. Yanagisawa, JAERI-M 92-021 (1992), in Japanese. [11] ASTM-E 321-79 (1985). [12] N. Ohnishi, T. Inabe, J. Nucl. Sci. Technol. 19(7) (1982) 528. [13] K. Yanagisawa, Nucl. Sci. Technol. 28(5) (1991) 459.