ML19289C545
| ML19289C545 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 01/10/1979 |
| From: | Speis T Office of Nuclear Reactor Regulation |
| To: | Wessman G GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER |
| References | |
| NUDOCS 7901170309 | |
| Download: ML19289C545 (4) | |
Text
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PTan JAN 101979 ARB R/F RIreland Docket File RMattson
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NRR R/F
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4Crutchfield Mr. George '4essman, Director LPOR ACRS(16)
Plant Licensing Division GKuzmyc::
Vendor File General Atomic Company RSchambercer JBucnanan, NSIC RSoyd TAbernathy, TIC Cear Mr. Wessman:
HGearin HDenton EGCase
SUBJECT:
REQUEST FCR ADDITIONAL INFORMATION ON H J51 GRAPHITE
- !e have reviewed your rescanses of October 30, 1973 and Noveccer 17, 1975 to our cuestions pertaining to our review of crace H-451 grachite. In order to complete this review we find that additional infor ation will be necessary which is muested in the enclosure. Based on cur discussiens with you on January 10, 1979 we anticipate tnat our safety evaluaticn will be issued by April 1,1979.
The findings of our safety evaluation, if favorable will be centinoent on successful irradiation perfomance of the H 451 test 21eeents in the Fort St. Vrain reacter. Establishment of successful irradiation cer#cr-ance will recuire cost irradiation exmination in accordance with cur colicy on fuel surveillance transmitted to Mr. J. K. Fuller of the Public Service Cocoany of Colorado on January 3,1979. In this recard, a concit::ent to sucoly post irradiation examination infcmation en the eicht H-451 test fuel eleeents scheduled for insertion with the first Fcrt St. Vrain reload core would be recuired prior to NRC final accroval for a full reicad of fuel using H-451 graphite.
If ycu have any cuestiens, clease let us know.
Sinceraly.
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% P. sm Themis P. Speis Chief Advanced Reactors 3 ranch Division of Droject Manacement
Enclosure:
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ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION Review of Grade H-451 Graphite, General Atomic R2 port, GLP-5588 130.2 In your response to Q130.la, you indicated that unirradiated H-451 graphite does not experience fatigue damage under reversed stress cycling unless stressed above a homologous stress limit of 0.63 in the axial direction or 0.74 in the radial orientation.
These homologous stress limits are understood to be established on the basis of 50 percent survival.
Indicate your rationale for not using limits for 99 percent survival.
If the limits for 99 percent survival are used, will your conclusion be still the same?
231.17 The response to ist round cuestion 231.3b indicated that, because of the high degree of complexity of strain behavior in the radial orientation (as compared with axial strain behavior), no general comparison of stresses caused by irradiation strain in H-327 and H-451 graphites has been provided. Yet the effects of these radial shrinkages are said to have been included in the analyses. No twi th-standing the assarted comolexity in radial strain behavior, some linkage must be provided, at least in the form of an exemolar calculation, between the radial strains and the stress provided in Taole 2-1 of GLP-5588.
231.18 As a point of clarification only, it should be noted that code verification usually implies a comoarison of code credictions with experimental data, not a comparison with results from another code.
Thus, the comparison of axial stresses computed by the SAFE GRAPHIT code with stresses computed by the FESIC code or with hand calculations (response to Q231.7), while interesting, does not, in itself, constitute true verification.
231.19 (a) Although there are no data for the diffusion of strontium in H-451 graphite, it is stated in the report that, based on the data shown in Table 7 and Figure 3, "there appears to be no decendence on the type of graphite." Please indicate which of the referenced data are for near-isotropic graonites.
If little or no strontium diffusion data exist for near-isotrooic needle-coke graphites, discuss the applicability of the data to H-451 graphite.
(b) The strontium diffusion data in Figure 3 of the report have a fairly wide scatter (1 to 2 orders of magnitude) in the lower range of temperatures whereas at high temperatures the scatter is smaller.
Please indicate the range of expected operating temoeratures in the Fort St. Vrain H-451 blocks.
231.20 (a) The report indicates that, whereas strontium diffusion in graphite can be described by Fick's Law, cesium transport is a more comolex phenomenon. Yet cesium transport is described in the recort in terms of a permeation coefficient, defined by the eauation J =, (C -C )/L, (report equation 9), which is a form of Fick's Law.
1 2 Please discuss this apparent contradiction; that is, discuss now a
e
. process that is acknowledged to be more complex, can be described by Fick's Law, when Fick's Law is known to be generally applicable only to very simple diffusion cases.
(b) There appears to be only one datum (see report Figure 4) indicating the effect of irradiation on cesium permeation coefficients for H-451 graphite and none for H-327.
It is di fficult, therefore, to accept the conclusion that "the permeation coefficient for Cs in H-451 graphite is significantly reduced in-pile." Please discuss the effect of using the permeating coefficients for unirradiated graphite as interim values until more data are obtained on cesium diffusion in H-451 in-pile.
Please indicate whether such data are to be obtained on the 8 test elements currently scheduled for the first reload in Ft. St. Vrain.
231.21 The response to 1st-round question 231.16' indicates (Table 1) that the creep tests on H-451 g~rachite will not be completed until 1983, whereas H-451 graphite reload fuel elements could be placed in Fort St. Vrain as soon as late 1980 or early 1981. Moreover, much of the current creep datt base on near-isotropic graphite appears to have been generated on Gilsocarbon-based, near-isotropic graphite rather than petroleum coke-based, near-isotropic graohites.
In view of (a) the current limitations on the data base on H-451 graphite and (b) the uncertainty regarding the applicability of Gilsocarbon graphite creep data, please indicate how the test data yet to be developed would lead the expected irradiation exposures of reload elements. Also show how the planned post-irradiation examination program on the 8 test elements and fu'.re reload elements will provide dimensional change data that would e used to verify the predicted changes that are based, in part, on expected creep behavior.