ML20114D687

Tritium Leach Test on H-327 Graphite

ML20114D687
Person / Time
Site:	Fort Saint Vrain
Issue date:	09/01/1992
From:	Stithem A PUBLIC SERVICE CO. OF COLORADO
To:
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ML20114D686	List: ML20114D685 ML20114D687
References
NUDOCS 9209090260
Download: ML20114D687 (9)

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. P-92273 TRITIUM LEACH TEST ON H-327 GRAPHITE Oc /l. .r.d-_

Arthur R. Stithem Project Engineering Public Service Compar.y of Colorado i

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' I. BACKGROUND The initial decommissioning activities for the Fort St. Vrain High Temperature Gas Cooled Reactor involve the removal of all defueling elements and accessible reflector blocks with the fuel handling machine, and subsequent flooding of the Prestressed Concrete Reactor Vessel (PCRV) cavity. Flooding of the PCRV cavity will provide biological shielding for workers associated with PCRV dismantlement activities. The large permanent graphite reflector blocks and side spacer graphite blocks which remain in the PCRV will be submerged in tha shield water ur.til removal. A portion of the tritium in these reflector blocks will leach into the shleid water during tne decommissioning process (Reference 1).

A study of British Magnox reactor graphite described the leach behavior of irradiated reactor graphite in domineralizea water at ambient temperature '

(Reference 2). The British study was used to estimate the amount of tritium ,

that would be released into the PCRV shield water during decommissioning.

During reactor operation there were four main sources of tritium production:

ternary fission, activation of He-3 in the helium primary coolant, boron activation, and activation of lithium impurities in fuel and reflector graphite.

The-Fort St. Vrain Activation Analysis (contained in Refere.1ce 1) estimated a total of approximately 100,000 curies of tritium in the permanent graphite reflector blocks. The tritium leach behavior described in the British Magnox study indicated that approximately 0.5% of the total tritium (=500 curies) would teach into the PCRV shield wates after 100 days.

l Public Service Company of olorado contracted with Babcock & Wilcox to l

perform a leach test on two a radiated replaceable graphite reflector blocks from the Fort St. -Vrain core in order to gain further insight into the behavior of tritium impregnated graphite in a water environment. The results of the B&W L test are compared with the British Magnox study in thio report.

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11. TEST BLOCKS The large permanent graphite re' lector blocks are composed of HLM

! graphite. This grade of graphite has a max;rnum lithium specification of 2 ppm. The high purity graphite used in fuel elements and replaceable graphite reflector blocks (grades H-327 and H-451) have a maximum lithium specification ofless than or equal to 0.1 ppm. During the reacter defueling process HLM graphite was completely inaccessibls by non-destructive means. Two grade H-327 replaceable graphite reflector blocks wera removed from Region 30 tor leach testing. Region 30 had not been refueled during the operational life of the reactor, and therefore received the maximum neutron flux exposure possible.

The in-core locations of the two test blocks are shown in Figures 1 and 2.

One small half length top reflector block was removed from column 3 in the reflector layer immediately above the fuel. This block was directly in the primary coolant f!ow path and has coolant holes (Figure 3). The other reflector block removed was a large full length solid side reflector taken from column 17 at approximately the mid-core level. This block was -

located between the active core and permanent side reflector graphite, and does not have coolant holes (Figu e 4).

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k Ill. GRAPHITE ANALYSIS Frior to teaching, each test block had graphite samples removed to determine initial tritium and lithium content. The small half len0th graphite reflector block was sampled in a single location aaproximately 1 inch from the bottom on the serial number side face,1 inch deep (Figure 3). The large solid graphite reflector block was sampled in two locations. T.co six inch long cores were ren:oved from the serial number side l ace, approxirnately six inches from the top and bottom (Figure 4), Each six loch coro sample was analyzed for tritium at 2 inct increments from the auter surface to the interior end. Tritium concentrations in the large block exhibited a wide degree of variability, but in general wure 3 to 4 orders of magnitude higher near the surface than in the deep interior.

H-327 GRAPHITE LITHlUM ANALYSIS Location Lijp_-gLgl large Block Upper Samp!s @ 6 inches < 0.1 Large Block Lower Sample @ 6 inches < 0.1 Small Block Sample @ 2 inches < 0.1 H 327 GRAPHITE TRITIUM ANALYSIS Samole DeD1b it-3Jy-Cilal 1

Large Block Upper Sample @ 0 inches 6.4 5 E.1 Large Block Upper Sample @ 2 inches 6.65 E-2 Large B Ock Upper Sample @ 4 inches 4.16E-1 Large Block Upper Sample (D 6 in:5es 5.54E-4 Large Block Lower Sample @ 0 inches 7.7 5 E-2 Large Block Lower Sample @ 2 inches 1.9 6 E-1

- Large Block Lower Sarnple @ 4 inches < 2,87E-5 Large Block '.ower Sample @ 6 inches < 1.94E-5 Small Block @ 1 inch 1.02E + 0 British large Small Graohi,tt Block Blocin Tritium Concentration (p Cuiies/Cm ) 1.18E + 1 3.07E 1

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. Tota! Tritium Content (p Cu ies) . 1,830 27,300* 64,600

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IV. LEACH TEST lhe sinall block was placed in a 55 gallon stainless steel tank and filled with 165 liters of. daionized water. The large block was placed in an 80 gallon stainicss steel tank and filled with 210 liters of deionized water. The water !n each tank was slowly circulated and remained at room temperature. Samples were taken from each tank twice each day for the first 7 days of the teach test, and then once por day for the following 23 days.

A comparison of the large and small blocks to the Bi;tish Magnox study samp!a is given below:

Sample Sample Surface / Leach Volume / Leachate pamole . Volume Jyttface Area Volume Surface Area , Volume British sample 88.4cc 134cm2 1.52cm4 2.2dem C.3 L Large Bbck a9,000cc e,460cm2 0.14cm 4 16.85cm 21o L small Block 36,200cc 27,874em2 0.77cm4 5.92cm 165 L Tritium from each block was released at a nearly constant rate throughout the first 30 days of the teach test (Figure 5). The smal. a:ock released more total tritium than the large block. A comparison of the H 327 graphite block leach

. fractions to the-British Magnox graphite leach fraction is given in Figure 6.

British Large Small Graphite Block Block

@30 Days @30 Dava @30 Days Total Tritium Released 8.0 p-ci 26.2 g-ci 28.8 gmi Release Fraction 0.44 % 0.096% 0.045 %

. V. CONCLUSION Data from the H 327. graphite teach test indicates that tritium is released from this grade of graphiro at a Tw, nearly constant rate.

Durin0 the first 30 days of this leach test, less than 0.1 % of the tritium content was released from each t, lock, significantly below the 0.44% indicated by the British Magnox study for a similar leach period.

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REFERENCES

1. Fort St. Vrain Proposed Demmmissioning Plan; submitted to the NRC in PSC letter, Warembourg to Weiss, dated April 17,1992 (P 92162).

2. 1.F. White, et.al., " Assessment of Management Modes for Graphite Reactor Decommissioning", CUR 9232 EN, Commission of the European Communities.

This document was submitted to the NRC by PSC letter, Brey to Weiss, dated October 10,1991 (P 91299

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