Pergamon

PII: S0956-053X(96)00017-7

Waste Management, Vol. 15, No. 8, pp. 593-598, 1995 Copyright © 1996 Elsevier Science Ltd Printed in the USA. All rights reserved

0956-053X/95 $9.50 + 0.00

ORIGINAL CONTRIBUTION

TRITIUM INVENTORY PREDICTION IN A CANDU PLANT

M. J. Song, S. H. Son and C. H. Jang Korea Electric Power Research Institute. Daejon, Korea

ABSTRACT. The flow of tritium in a CANDU nuclear power plant was modeled to predict tritium activity build-up. Predictions were generally in good agreement with field measurements for the period 1983-1994. Fractional contributions of coolant and moderator systems to the environmental tritium release were calculated by least square analysis using field data from the Wolsong plant. From the analysis, it was found that: (1) about 94% of tritiated heavy water loss came from the coolant system; (2) however, about 64% of environmental tritium release came from the moderator system. Predictions of environmental tritium release were also in good agreement with field data from a few other CANDU plants. The model was used to calculate future tritium build-up and environmental tritium release at Wolsong site, Korea, where one unit is operating and three more units are under construction. The model predicts the tritium inventory at Wolsong site to increase steadily until it reaches the maximum of 66.3 MCi in the year 2026. The model also predicts the tritium release rate to reach a maximum of 79 KCi/yr in the year 2012. To reduce the tritium inventory at Wolsong site, construction of a tritium removal facility (TRF) is under consideration. The maximum needed TRF capacity of 8.7 MCi/yr was calculated to maintain tritium concentration effectively in CANDU reactors. Copyright © 1996 Elsevier Science Ltd

1. INTRODUCTION

In a pressurized heavy water reactor, tritium is gen- erated by several different mechanisms, such as ternary fission within the fuel elements and neutron absorption of deuterium. Because heavy water is used as both coolant and moderator, a considerable amount of tritium is produced by thermal neutron absorption by deuterium. Tritium generated by this mechanism generally remains in the systems as tritiated heavy water (TDO). Tritium decays into helium, emitting a /3-radiation with a half-life of 12.3 years. The energy level of the/3-radiation is so low that it causes negligible external radiation hazard. However, tritiated water or water vapor, if inhaled, can cause internal radiation damage.

Wolsong unit-1 is a CANDU-type pressurized heavy water reactor with a capacity of 680 MWe. It has been operating since 1983. Specific activities of tritium in coolant and moderator have increased ever since. As of the end of 1994, the specific activities of tritium in coolant and moderator of Wolsong unit-1 were 1.56 and 38.5 Ci/kg • D20, respectively, l Such tritium levels correspond to a tritium inventory of 10.1 MCi for Wolsong unit-1. During operation and maintenance of the plant, heavy water leaks from the systems, some

RECEIVED 13 NOVEMBER 1995; ACCEPTED 5 MARCH 1996.

of which is recovered and the remainder is released into the environment. Tritium in tritiated heavy water is also released together with the lost heavy water. By 1994, a total of 100 KCi of tritium has been released from the plant.'

Currently, three more CANDUs are under con- struction at the site. As the new units start operation, the total tritium inventory at the site will rapidly increase and so does the tritium release into the environment. Accurate predictions of tritium build-up for and release from Wolsong site are needed to assess environmental impacts and provide valuable data for establishing tritium management plans.

In this study, the behavior of tritium in the Wolsong nuclear power plant was modeled to calculate the tri- tium build-up in the systems and release rate into the environment. The results of the modeling were verified with the actual measurements. Additionally the future tritium inventory for the plant was projected based on the model. The effectiveness of a Tritium Removal Facility (TRF) in reducing tritium inventory will be also discussed.

593

2. TRITIUM INVENTORY IN A CANDU

2.1 Generation and Accumulation Tritium is generated within the fraction of heavy water which is under neutron flux. Tritium can also be

594 M . J . S O N G ET AL

added to the systems through make-up systems. During the operation of plants, heavy water containing tritium leaks from the systems, most of which is recovered and returned to the systems. The remaining heavy water is released into the environment, together with tritium. Tritium is also removed from the systems by radioactive decay. Figure 1 shows a schematic of the flow of tritium in the coolant and moderator in a C A N D U heavy water reactor.

For a given plant, tritium inventory is a function of power level, capacity factor, and release rate. To calcu- late tritium inventory, the following were assumed:

• Coolant and moderator are completely separated. • Complete mixing of total heavy water inventory

within the systems. • Make-up flow rates are equal to release flow rates,

or the heavy water inventories in the systems are constant.

• Recovered coolant is returned to the coolant system.

• Recovered moderator is returned to the moderator system.

• Plant average capacity factor is 84.4%.

From Fig. 1, a mass balance for tritium was set

: heavy W ater inventory in syStem

~ n r y ~ ~lel..,ntO besvy water leakage Imvironment

[wavy water make-up to

N *:T cor~entratzon in ~ Y~O: a l~rp~2^" / / make-up heavy water X Y~/o~/s~,~ " /

~ N: T c~centration in system - ' --W-

pecay [ _ _ . ~ ~.: Decay constanl

FIGURE 1. A schematic of the flow of heavy water and tritium in the coolant and moderator systems in a C A N D U plant.

d ( M N ) _ 4 ) t r N o r n a + F N ° - h M N - L N (1) dt

The first term on the right-hand side is the tritium generation rate, the second term is tritium from make- up flow, the third term is natural decay rate, and the last term is tritium release rate. The definition of each parameter is summarized in Table 1. Because total heavy water inventory is constant, equation (1) can be rewritten as follows

d____NN + (A + ML----)N = 4)or ND ~ - a + --F-F N° (2) dt M

The general solution of equation (2) is

N = e A~' [ + ~ o ' ( - - ~ N ° + S) + C] (3)

where S = th or No ~ - a

L X ~ = X + ~ - -

A¢ is an effective decay constant. Because tritium concentration is zero at the start of the operation, then

1 (FN0 + S) C - A~

Finally, the tritium concentration (T atoms/kg - D20) can be expressed as follows

(FN° + S) N - [1 - e-ae ' ] (4)

Ae

Assuming tritium-free make-up heavy water, specific activity of tritium (Bq/kg - D20) can be expressed as follows

TABLE 1 Definitions and Values of the Plant Parameters Used in the Activity Analysis 2

Coolant Moderator

a Plant capacity factor 0.844 0.844 L Heavy water loss rate (kg/s) 1.18 x 10-4 7.93 × 10 -6 F Make-up flow rate (kg/s)* 1.18 x 10 4 7.93 × 10 -6 N o Tri t ium in make-up flow (kg-t) * 0 0 M Total heavy water inventory (kg) 1.32 x 105 2.57 X 105 m Heavy water under neutron flux (kg) 6.03 × 103 1.902 x 105 N D Number of deuterium per kg • D20 (kg-l) *+ 6.01 × 1025 6.01 × 1025 A Trit ium decay constant (s -1) 1.78 x 10 -9 1.78 X 10 -9 A e Effective tritium decay constant ( s 1) 2.674 x 10 -9 1.811 x 10 9 o" Neutron absorption cross-section (cm 2) 3.64 × 10 -28§ 4.19 x 10 2811 ~b Thermal neutron flux ( c m 2 s t) 1.235 × 1014 2.30 x 10 TM

* Equal to loss rate. t Assumed to be zero. :~ 2 D-Atoms/D20* Na/20 g - D20* 10 -3 g/kg* 0.9975 kg - D20/kg = 6.01 × 1025 D/kg.

-4 , 0 5, 0 5 28 2 § %(5.1 x 10 barn) (~r/4)' (i93.6/563)' = 3.26 x 10 c m . II %(5.1 x 10-4 barn)*(zr/4)°'5*(293.6/341) °'5 = 4.19 × 10 28 cm 2.

T R I T I U M INVENTORY P R E D I C T I O N 595

A = A N = A S [ l_eAr , ] Ae

The maximum specific activity of tritium is

A A m Am - ~-e S : ~ ~ o" No -77-~ a M

2.2 Specific Activity of Tritium in Coolant The fraction of the coolant under neutron flux is about 0.046. The specific activity in coolant, in Curie per kg • D20, can be calculated from equation (6) using the values shown in Table 1; t is measured in years.

Acoolan t = 1.873 [1 - e -0"0843t] (7)

It takes 27 years for the specific activity to reach 90% of the equilibrium value.

2.3 Specific Activity of Tritium in Moderator The specific activity in the moderator can be calculated using the values shown in Table 1. The fraction of the moderator under neutron flux is 0.739 and is much larger than that of coolant. Therefore, the specific activity of tritium in the moderator system increases faster than that in the coolant system. The equilibrium specific activity is also much larger. Using the same method used for the coolant, the specific activity in the moderator can be calculated as follows

A m o d e r a t o r = 96.09 [1 - e 0"0571t] (8)

2.4 Comparison of Predicted Values and Field Measurements The operation records of Wolsong unit-1 are shown in Table 2. For the analysis, it was assumed that the

(5) plant was operating continuously with a capacity factor of 84.4% with heavy water loss rate of 3970 kg/year. At the end of year i, the specific activity of tritium (Ag) can be expressed as follows

(6) A, = N~ A

= S i A-~---(1-e-'%.iAti)+Ai leae. i ±ti (9) I~e, i

where Si = 6 or N D ~ - ai

Zi

At i-- 1 yea r=3 .154E7s

Ni, Sg, L~ and a~ represent tritium concentration at the end of the year i, tritium generation rate, heavy water loss rate, and capacity factor during the year i. Annual capacity factors and total heavy water loss rates of unit- 1 are shown in Table 2. In order to calcu- late A~.i, heavy water loss rates (Li) from coolant and moderator are required separately. Heavy water loss rate from each system was calculated using tritium release rate and measured specific activity of each system. The results are shown in Table 3.

The calculated and measured specific activities of tritium in the coolant and moderator are shown in Fig. 2. Predicted values are generally in good agreement with measured ones up to the seventh year. There were considerable discrepancies between calculated and measured activities afterwards. The inadvertent injection of the small amount of the recovered moderator into

TABLE 2 Operation Records of Wolsong Unit-1 Showing Plant Capacity Factors, Specific Activities of Tritium in the Coolant and Moderator, Heavy

Water Loss Rate and Tritium Release Rate into the Environment I

Year Capacity Acool* Amod t Heavy water loss Tritium release (Ci/year) factor (%) (Ci/kg) (Ci/kg) (kg • D20/year)

Vapor Liquid Total Vapor Liquid Total

1983 61.9 0.10 3.51 2950 1220 4170 1306.3 92.8 1399.1 1984 66.8 0.24 6.22 4317 1077 5394 270118 177.6 2879.3 1985 94.4 0.30 11.71 3718 494 4212 2424.7 324.4 2749.2 1986 79.7 0.50 16.22 4714 480 5194 6528.8 994.9 7523.8 1987 92.9 0.54 18.02 2957 659 3616 8467.0 1372.5 9839.5 1988 79.4 0.68 21.27 3446 638 4084 8082.0 2014.2 10,096.1 1989 91.0 0.85 27.93 2477 386 2863 6097.8 1633.6 7730.6 1990 85.9 1.20 33.60 3071 367 3438 6231.7 1397.7 7629.4 1991 91.1 1.31 34.14 2976 369 3345 6946.8 2518.9 9465.7 1992 85.9 1.39 35.40 3107 245 3352 10,499.4 1136.4 11,635.8 1993 100.8 9953.5 1252.0 11,205.5 1994 82.6 1.56 38.50 3682 327 4009 13,062.0 4874.8 17,936.8

Avg. 84.4 3401 569 3970

Total 82,302 17,790 100,091

*Specific activity of coolant. tSpecific activity of moderator.

596 M . J . S O N G ET AL.

the coolant system is suspected as the probable cause of a sudden increase in activity in 1990. There is some improvement of accuracy in estimating the tritium activities in the coolant and moderator systems by using annual capacity factors and heavy water loss rates instead of life-time averaged values. However, the differences between the two methods are not signi- ficant. Therefore, the simpler method using average plant capacity factor and heavy water loss rate can be used to estimate future tritium build-up in Wolsong unit-1 and subsequent units.

3. TRITIUM RELEASE FROM WOLSONG UNIT-I

3.1 Leakage, Recovery and Loss of Heavy Water In a CANDU, heavy water leaks from coolant and moderator systems during operation. About 70-80% of leaked heavy water is recovered and the remainder is released into the environment and lost. As shown in Table 3, the heavy water vapor loss originates mostly from the coolant system. The heavy water liquid loss occurs mostly during plant maintenance. During the first few years of operation, a large amount of liquid

TABLE 3 Calculated Heavy Water Loss Rates from the Coolant and

Moderator Systems of Wolsong Unit-1

Year Coolant loss rate (kg. D20/year)

Moderator loss rate (kg • D20/year )

Vapor Liquid Sum Vapor Liquid Sum

1983 2270.4 1201.3 3 4 7 1 . 7 679.6 18.7 698.3 1984 3897.8 1078.2 4976.0 418.0 - - 418.0 1985 3554.6 472.0 4026.6 163.4 22.0 185.4 1986 4371.7 420.8 4792.5 342.3 59.2 401.5 1987 2539.6 597.0 3136 .5 417.4 62.0 479.5 1988 3131.8 552.6 3684 .5 314.2 85.4 399.5 1989 2300.7 329.9 2630.5 176.3 56.1 232.5 1990 2967.3 332.7 3300.0 103.7 34.3 138.0 1991 2877.5 306.0 3183.5 98.5 63.0 161.5 1992 2918.3 220.9 3139.2 188.7 24.1 212.8 1993 1994 3475.6 206.1 3 6 8 1 . 6 206.4 120.9 327.4

A n~aared

2,5 L " ' opacilyratl0r&rclt~t.~ . 50 cak~l~t~l uxang average i ca~'ily I~lor & lelta.~ tale 1 i

2.0 ............... [ .................. l ...................................... ] ......... i ................................. 4@

g , . , [ - . ............... ! .................. ............... -

1.o .................... . ... zo o g. • : • ipillulell IL..,i(Z ,o

0.5 1~- ..... ( clllp~iy flOr & l i m e i cil~lmed mini averilt

0 2 6 8 10 12 14 operatin 8 year

F I G U R E 2 . C o m p a r i s o n o f c a l c u l a t e d t r i t i u m a c t i v i t i e s w i t h a c t u a l

measurements (1983-1994) for Wolsong unit-1.

was lost due to frequent maintenance. After a few years of operation (when the plant becomes stabilized), annual loss of heavy water liquid became stabilized. For Wolsong unit-l, the annual heavy water leak rate was about 17.7 tons, of which about 13.7 tons were recovered and about 4 tons were eventually lost. I About 85% of the heavy water loss was in vapor form.

Annual tritium release rates are also shown in Table 2. Though the annual heavy water loss rate becomes more or less constant after a few years of operation, the annual tritium release rate fluctuates considerably. About 82% of the released tritium was in vapor form. After 1992, the annual tritium release rate exceeded 10,000 Ci/year. It is conceivable that the tritium release rate will steadily increase as the specific activities of tritium in the coolant and moderator sys- tems increase continuously. For future tritium release rate prediction, it is necessary to find the fractional contribution of the coolant and moderator systems to environmental tritium release.

The fraction of tritium release from each source can be estimated using heavy water and tritium release rates. For vapor- or liquid-phase tritium release, the following mass balance equation can be set

R = Xc L Ac + (1 - X¢) L A m (10)

o r 0 = ) ( c ( A c - A m ) + ( A m - - - ~ - - )

where R and L are annual tritium release rates and heavy water loss rates in either vapor or liquid form. X~ is a fraction of tritium release in either vapor or liquid phase from the coolant system. Ac and A,, are the specific activities of tritium in coolant and moderator, respectively. The following least square analysis was used to find optimum Xc

d u R ._ ~= X~(Ac.~ - Am,i) + (Am,r- --')L~ = 0 (11)

dX~

N

Xo(A<.,- Am.,) (A.,.,- ) Xc = '=' (12) N

i ~ (Ac,~ - AmjY

where subscript i represents operating year. The above equation can be used for the estimation of

the fractional contributions of coolant and moderator to tritium release in both vapor and liquid forms. The results of the analyses are shown in Table 4. From the data for Wolsong unit-l, about 95% of vapor and about 86% of liquid heavy water loss come from the coolant system. The contribution of the moderator system to total heavy water loss was far less than that of the coolant system. This was expected because the moderator system is operating at much lower pressure and temperature than the coolant system. Nevertheless, the contribution of the moderator system

TRITIUM INVENTORY PREDICTION 597

to tritium release was 63.8%, which was larger than that of the coolant system due to the high tritium concentration of the moderator system. About 61% of vapor-phase and 78% of liquid-phase tritium were released from the modera tor system.

3.2 Tritium Release Rate Comparison Trit ium release rates of vapor and liquid phases can be calculated using equation (10) and Table 4. Total tritium release rate was calculated as follows by summing tritium release rates of both phases.

R -- 3720A c + 250A m

= 6968(1 - e 0"0843t) + 24023(1 - e ~'°571') (13)

Tritium release rates in vapor and liquid phases from Wolsong unit-1 were compared in Fig. 3. Calculated release rates are generally in good agreement with the field measurements for 12 years. After 3 years of opera- tion, the annual tritium release rate became more than 1000 Ci/year. Actual release rates of Wolsong unit-1 became lower than the predicted values after 7 years.

Annual tritium release rates of a few operating C A N D U 6 plants 2 are compared in Fig. 4. The four plants in the figure have similar heavy water mainte- nance records, with about 4 tons of heavy water loss annually. The calculated tritium release rates agree well with Wolsong unit-1 and other plants. Thus, the calculated curve can be used as a representative of C A N D U 6 plants.

TABLE 4 Fractional Contribution of Coolant and Moderator to Total Heavy

Water Loss and Tritium Release

D20 loss D20 loss rate Tritium release fraction (kg/year) fraction

Vapor Liquid Vapor Liquid Vapor Liquid Total

Coolant 0.950 0.860 3231 489 0.387 0.218 0.362 Moderator 0.050 0.140 170 80 0.613 0.782 0.638

4. P R E D I C T I O N O F T R I T I U M BUILD-UP AND RELEASE AT W O L S O N G SITE

At Wolsong site one CANDU6 unit is in operation and three units are under construction. These units will start operation in 1997, 1998, and 1999. Thus, tritium inventory and release rate at site will increase as the operating plant-year increases. Using the results of pre- vious analyses, tritium inventory build-up can be predicted with fairly good precision. The annual tritium production rate largely depends on the plant capacity factor. Worldwide C A N D U 6 plants show a wide range of plant capacity factors from 71.6 to 91.4%, as of the end of 1994. 3 The lifetime averaged capacity factor of four operating CANDU6 plants is about 80%. Thus for the prediction, plant capacity factors of 75, 80, 85 and 90% were used.

The results of the analyses are shown in Fig. 5. Tritium inventory at Wolsong site increases steadily as more units are added. The inventory reaches a maximum of 66.3 MCi in the year 2026 for an 80% capacity factor, then decreases afterwards as some plants will be closed down. A change of 5% in capacity factor resulted in a 4.2 MCi difference in maximum tritium inventory. Trit ium release rates were also predicted using the results of the previous analyses and shown in Fig. 6. It reaches a maximum of 79 KCi/year in

2 . 0 1 # . . . . . I ' I ' i

=~ 1.010 4

S 10 3 ~ L 2 . , a f ~ . ~ . . ~ . . L . ~ o J t Gen t iny -2 5.0 f / ~ f f / i - i I o gmbalse

Z ~ ! i I • w,~ms-t

0.010 ° -- ' - - - • ' ' •

10 12 14 Operating year

FIGURE 4. Comparison of tritium release rate in CANDU6 plants. Calculated release rate is also shown. Annual heavy water loss rate of 3970 kg/year and plant capacity factor of 84.4% were used for the calculation.

calculated total release • measured ~ t a l release

2.0 lO 4 . . . . . . . . . calculated vapor phase release . . . . V " - measured vapor phase release . . . .

calculated liquid phase release [] measured liquid phase release ] [

1~,o , .................. f ................... ! ................... ~ .................... t ................... : ~ ! .......

-~ 1~o~ ................. i ................... i ~ ~ i ..... ...,.,.7,'L...i .......

0.010 0

o 2 6 8 10 12 14 year

FIGURE 3. Comparison of calculated and measured tritium release for Wolsong unit-1 during the l~riod 1983-1994. Annual heavy water loss rate of 3970 kg/year and plant capacity factor of 84.4% were used for the calculation.

I ~ 90% capacity factorl | o 85% capacity factor 1

8,0107 --'v'-] - - - . & ~ 8 0 % c a p a c i t y f a c t o r l l . . . . I . . . . -.....[ ,~. 75% capacity r a c e r I i - ~

: i i ~ 6.olo 7 : ~ i

!

2.0 lo ~ : i i i i

o.oloo ~'i"

1980 1990 2000 2010 2020 2030 year

FIGURE 5. Estimated tritium inventory at Wolsong site with respect to various capacity factors.

5 9 8 M . J . S O N G ET AL.

the year 2012 for 80% capacity factor. A 5% change in capacity factor resulted in about 5 KCi/year difference in maximum tritium release rate. After 2012, when unit 1 will be decommissioned, tritium will be released from three units only, so the tritium release rate from the site will decrease.

To reduce the tritium inventory at Wolsong site, construction of a tritium removal facility (TRF) is under consideration. The removed tritium will be immobilized in metal beds as metal hydride. Therefore, tritium inventory and tritium release rate with and without TRF operation were compared. The target tritium concentrations in coolant and moderator sys- tems with TRF in operation were set as 0.5 and 10 Ci/kg • D20, as Ontario Hydro did. 4"5 The following operation mode was assumed.

1. TRF starts full operation from the beginning of the year 2006.

2. Achieve target concentration after 7 years of operation.

+ 9 0 % capacity factor| 85% capacity factor I

1.0105 ~ 80% capacity factorl ,

-----n.---75% capacity factor I ]

6.010 4

4.0 lo 4

• r" 2 .010 4 l -

0 . 010 0

1980 1990 2000 2010 2020 2030 year

FIGURE 6. Annual tritium release rate prediction at Wolsong site with respect to various capacity factors. Tritium release from the decommissioned unit was assumed to be zero.

Inventory w/o TRF I Release rate w/o TRF I I : I : ' o ' . . . . . ate wITRF I 8.0 lO ~' 8.0 l o 4

. . . . . . . . I ' ' ' ' I ' ' ' ! . . . .

,.olo, ,.01o

i ,.olo, 3.01o,

"E 2.010 ~ 2.0x~ ~

1.0 107 1.0 104 '~

0.010 0 0.010 0

1980 1990 2000 2010 2020 2030 year

FIGURE 7. Effects of TRF operation on tritium inventory and release rate. Average plant capacity factor was assumed to be 80%.

3. Tritium levels will be maintained at the target activities afterward.

A maximum tritium removal capacity of 8.7 MCi/year was calculated for 80% capacity factor to meet the above operation mode. The effects of TRF on the tritium inventory and release rate were shown in Fig. 7. There were dramatic decreases in tritium inventory and release rate with TRF in operation.

5. CONCLUSIONS

Tritium concentrations in the coolant and moderator systems of Wolsong unit-1 were calculated using a simple model and compared with the plant operating record. Tritium release rates were also calculated from the model and compared with the plant record. The model predictions were generally in good agreement with the measured record. From this study the fol- lowing conclusions were drawn.

• About 64% of tritium release from the Wolsong unit 1 came from the moderator system.

• Tritium inventory at Wolsong site reaches a maximum of 66.3 MCi (with plant capacity factor of 80%) in the year 2026 and decreases afterwards. With the introduction of TRF in the year 2006, tritium inventory will be reduced dramatically.

• Assuming average plant capacity factor of 80%, a maximum TRF capacity of 8.7 MCi/year will be needed to meet the following operation mode: - The target tritium concentrations in coolant

and moderator systems are 0.5 and 10 Ci/kg • D20, respectively.

- Achieve target concentration after 7 years of operation.

- Tritium levels will be maintained at the target activities afterward.

R E F E R E N C E S

1. Annual Book o f Wolsong unit-l, KEPCO, Seoul, Korea (1983-1994).

2. Song, M. J. et al. Tritium reduction of Wolsong nuclear power p l a n t s - final report. KEPCO, Seoul, Korea (1993).

3. Howles, L. Load factors: 1994 annual review. Nucl. Eng. Int. A p r i l : 26 (1995).

4. Wong, K. Y. et al. Canadian Tritium Experience. Canadian Fusion Fuels Technology Project, Ontario Hydro (1984).

5. Sood, S. K., Quelch, J. and David, R. B. Fusion technology experience at Ontario Hydro's Darlington tritium removal facility and heavy water upgraders. Fusion Eng. Design 12:365 (1990).

Open for discussion until 19 December 1996