ML20062C706
ML20062C706 | |
Person / Time | |
---|---|
Site: | Fort Saint Vrain ![]() |
Issue date: | 10/30/1978 |
From: | Wessman G GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER |
To: | Gammill W Office of Nuclear Reactor Regulation |
References | |
NUDOCS 7811130200 | |
Download: ML20062C706 (56) | |
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sane [o$."cAtmonsa s2:3e (714) 455'J000 October 30, 1978 Mr. William Gammill '
t' Assistant Director for Advanced Reactors e Division of Project Management U. S. Nuclear Regulatory Commission Washington, D. C. 20555
Dear Mr. Gammill,
In accordance with my letter of October 25, 1978, we are enclos-Ing the first part of our response to your Request for Additional Information on H-451 graphite, dated September 19, 1978. Responses -
to the remaining questions wili be sent to NRC about November'24,1978.
If you have any questions regarding these responses, please do not hesitate to let us know, as we are most eager to expedite comple-tion of this review.
Sincerely,
, b!Zavtt:1L, G. L. Wessman, Director Plant Licensing Division GLW:mk h
a 7811130300'
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t After being duly sworn, the person known to me to be_G. L. Wessman of General Atomic Company, signed the within document this 2nd day of November 1978. ,
( WITNESS my hand and official seal.
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1 RESPONSE TO NRC OU_ESTICN 231.1 ON H-451 GRAPHITE i
QUESTION 231.1 The introduction to CLP-5588 indicates that the H-451 fuel and replaceable reflector elements "will be fabricated and assembled by General Atomic (GA) according to specifications and quality requirements comparable to those used for manufacture of the H-327 reference core." The imprecise wording of the quoted pro-tion of this statement requires considerable clarification for the following reasons:
/ (a) It is our understanding that the reference H-327 graphite fuel and replaceabic reflector elements were not " fabricated" by General Atomic but were, in fact, fabricated by an independent graphite manufacturing company, (only the machining and fuel loading of the as-fabricated graphite blocks were performed by General Atomic). We assume, ther efore, that the term, "fabrica-tion'," as used in the report, does not refer to the manufacture of the H-451 graphite. The fact that GA does not manufacture the H-451 graphite poses a problem that was raised previously on the LHTGR reviews.
In previous HTGR licensing actions (on the Summit, Fulton and GASSAR plants and also as part of generic HTGR graphite activ-ities) we have consistently stated our concern regarding the issue of graphite reproducibility. Graphite properties, in contrast to those of cost metallic alloys, are very heavily
. dependent on the type of precursor materials used in manufacture
( as well as the details of the fabrication process. This fact is partially acknowledged in the Mechanical Properties section (4.2.1) of the report, where it is stated (on p. 2-4) that the mechanical properties of the U-451 graphite elements will differ from those of H-327 elements in the reference core, primarily due to the use of a near-isotropic petroleu=-coke filler materici in place of the needle-coke filler used previously in H-327 graphite. Since the precursor material and fabrication processes ,
are so importantly related to the properties of the fabricated graphite elenents, please indicate how adequate assurance will be provided that (1) the H-451 graphite to be used in the Fort St. Vrain (FSV) reactor is the same as that used in the quali-fication test program for H-451 graphite (i.e., that the same precursor naterials and fabrication processes were used), and (2) the H-451 graphite to be used in future loadings, if manu-factured using different precursors or fabrication parameters or methods, will have the same thermal / irradiation performance and properties as the test-qualified grade. In responding to this question, please note that we _ understand that the state-of-the art. of nucle:.r e.raphite u.inufacture does not permit the l
prediction of therm.11/irr.Wiatien behavior f rom measurements of physical properties of unit radiated logs. Consequently, some other method of providir.g assurance of reproducibility must be provided.
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(b) It is our understanding that the fuel rods in the current H-327 graphite fuel elements were produced using an out-of- [
block forming and curing process. Are the fuel rods in the H-451 graphite elements produced in the same way? If not, discuss the potential effects of alternative fuel rod fabri- i cation processes on the performance of the graphite block, I fuel rods, and coated fuel particles. For example, if an in- !
block carbonization process is used in fabricating the fuel rods in the new H-451 elements, how does this process affect !
the potential for fuel particle coating failure due to block /
coating bonding and tearing? What effect will in-block car- t bonization have on the thermal response of the fuel (e.g. , will slumping of the fuel rod matrix material during in-block car-banization reduce the rod / block gap, thus improving heat con-duction)? Have any of these potential effects been accounted for in the accident analyses? l
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(c) There is no separate discussion of the replaccable reflector blocks provided in the GLP-5588 report. Are the reflector blocks [
i to be pure H-451 graphite or will they contain boronated graphite particles or other material? If any secondary caterial is to i be incorporated in the H-451 reflectors, please discuss its effects on properties and performance.
RESPONSE TO 231.la l The H-327 graphite logs for Fort St. Vrain were manufactured and the elements were machined by the Great Lakes Carbon Corporation (GLCC). ,
The fuel elements were loaded with fuel rods by General Atomic Co.
For FSV reload segments the H-451 logs will be manufact""-M from near-isotropic coke taken from the same batch (designated A3 in . - i and 2) as that used for the manufacture of preproduction lot 426 (Kel. 2).
Lot 426 is the preproduction lot used as a standard for production of the FSV reload elements and to be used in future H-451 production (Refs. 2 and 3). Preproduction lot 426 has been well characterized for properties .
and irradiation behavior (Refe. 1, 2, 3, 4), and the data obtained are used in the design data package for the H-451 fuel elements.
The H-451 purchase specification ensures reproducibility of coke structure as follows:
" Calcined coke samples...shall be fabricated into graphite specimens, which shall be tested for coefficient of thermal expansion (CTE) '
i ... The mean CTE values resulting from these tests shall fall within 20% of referc;ac valuc. catabliuhed on filler cokes used i i
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to manufacture the prototype, preproduction or production logs of grade H-451 that have been qualified by characterization and irra- ;
diation tests....and found to have satisfactory properties and i irradiation behavior".
! B The CTE test method to be employed is contained in RDT Standard E6-1T (Ref. 5). This standard has been developed to produce the degree of lot-to-lot reproducibility which is required to ensure consistent and l predictable properties and irradiation performance for specific graphite j j grades and to ensure traceability of the, graphite logs to production processes and raw materials that affect performance. Reproducibility of l other raw materials is ensured in the H-451 purchase specification as follows: l "The binder shall be a coal tar pitch of the same type used to manufacture prototype, preproduction or production logs of grade j H-451 and found to have satisfactory properties and irradiation i behavior". . -
- 4. f 1 "The impregnant shall consist of either a et t tar pitch or a petroleum pitch. The type used shall be certified to the pur-
! chaser." ,
The specification also requires that, " fabrication and processing i
- conditions will be the same and in the saec sequence as those used to j manufacture prototype preproduction or production logs of grade H-451 [
that have been qualified by characterfzation and irradiation tests and found to have satisfactory properties and irradiation behavior". [
J While we acknowledge at this time that irradiation behavior can- . l not be predicted or assured by correlations between unirradiated pro-porties and irradiation data, reasonable assurance of graphite repro-ducibility can be given by controlling the reproducibility of structure in the filler coke and standardizing the processing from preproduction
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(lot 426) to production lots (Ref. ?,). . -[
In the case of FSV reloads we need only he concerr.ed about within lot variation of coke quality since the reloads will be manu- j fact t. red from t he pre col e bat ch as preproduction lot 426, which was ;
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4 characterized and used as the medel for the production lots. For future production, when the current near-isotropic coke supply is exhausted, it will be necessary to produce a new batch. For reproducibility of the new batch of coke we will rely on the fabrication and processing requirements of the specification which are quoted above.
RESPONSE TO 231.lb It is not planned at this time to manufacture the fuel rods in the H-451 graphite fuel elements via the cure-in-place process. If the cure-in-place process is adopted in the future, a separate safety analysis report and request will be submitted.
RESPONSE TO 231.1c The replaceable reflector blocks will be made of pure H-451 graphite.
They were previously made of pure H-327 graphite.
REFERENCES
- 1. W. R. Johnson and G. B. Engle, " Properties of Unirradiated Fuel Element Graphites H-451 and.TS-1240", GA-A13752, January 31, 1976.
- 2. G. B. Engle,." Assessment of Grade H-451 Graphite for Replaceable Fuel and Reflector Elements in HTGR", GA-A14690, December 1977.
- 3. G. B. Engic and R. J. Price, " Strength Testing of Production Grade H-451 Graphite; Lots 472, 478 and 482", GA-A14269, March 1977.
- 4. R. J. Price and L. A. Beavan, " Final Report on Graphite Irradite tion Test OG-3", GA-A14211, January 1977.
- 5. "Near-Isotropic Petroleum-Coke Based Graphites for High Temper-ature Gas-Cooled Reactor Core Components", RDT Standard E6-1T, October, 1977.
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RESFONSE TO URC QUESTION 231.3 ON H-451 CRAPHITE QUESTION 231.3 (a) The report indicates that irradiation-induced dimensional changes in H-451 graphite in the axial dimension are about 50%
lower than for H-327 at peak temperatures and fluences and that axial stresses are, therefore, also lower. It is not clear, however, why axial dimensional changes should affect stress, as long as the changes are always negative; i.e., involving shrink-age and not expansion. Please show quantitatively the expected effect of lower shrinkage on axial stress. ,
(b) No discussion of the effects of differences in radial shrink-age on stress is provided in the report. Please show quantita-tively how these differences are accounted for in the design and accident analyses.
RESPONSE TO 231.3a Uniform dimensional changes of unrestrained elements, whether expan-sion or contraction, will not produce stress. Fuel element stresses are produced through the differential irradiation-induced shrinkages that occur because of the spatially varying temperature and fluence distributions.
Comparison of Figures 4-1 and 4-Sa (CLP-5588) shows that H-451 not o'nly has a lower irradiation-induced axial dimensional change but also has less differential strain between any two sets of temperature contours. It is these lower differential strains that lead to lower irradiation-induced stresses in the H-451 graphite fuel elements.
RESP _0NSE TO 231.lb Because of radial strain turn-around at higher fluences (see Figures 4-2 and 4-Sb GLP-5588), the irradiation-induced etrain behavior in the radial orientation is more complex than that in the axial orientation.
For this reason, no general comparison between the stresses caused by radial irradiation strain in H-327 and H-451 graphites was included in the text of Section 2.4.1.2. The eficers ei these r.: dial nhrinkares are, however, included in the analyses and the results are shown in Tabic 2-1 of GLP-5588.
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6 RESP 0:;SE TO 1:RC QUESTIO!1 231.7 ON H-451 GRAPilITE 231.7 Ou est ion The computer programs FESIC and SAFE GRAPIIITE, which are reported to have been used to calculate the operating and shutdown strain and stress distributions, have not undergone a licensing review which included model verification. Please provide some oiscussion, therefore, of veri- i fication efforts made on these codes, along with pertinent data and key - '
charts. .
Response
The computer programs FESIC and SAFE GRAPHIT were used in the analysis and design of the FSV initial core, and they were accepted by the AEC in its review of the FSAR. They have been checked numerous times by hand calcula- l tions and comparison with other codes. There is no formal documentation of this verification. The following coeparisons were made during the code develop- <
ment:
SAFE GRAPiiIT - 2-D clastic analysis, by comparison with its predecessor, SAFE CRACK, which has been accepted by NRC for PCRV analysis
- creep analysis of graphite, by simulating uniaxial creep and relaxation tests l 1
- simple beam and uniaxial models by comparison with hand calculations ,
FESIC - comparison of axial stresses with those computed by SAFE GRAPillT. The agreement was excellent, as shown below. -
t MAXI! RIM AXIAL STRESSES TIME 'FESIC SAFE-CRAFI Initial Startup 111.1, psi 124.1' psi .
62.1 days . 42.1 psi .44.1 psi 124.2 days -
120.5 psi 121.9 psi 186.'3 days 158.0 psi 153.2 psi ,
248.4 days 143.3 p:,i' 131.i psi 310.5 days ,
194.3 psi 195.1 psi 310.5 days 262.5 psi *262.4 psi' (shutdown) s It is noted th.it the FF. SIC and SAFB GR.\l' lilt codes are completely independent and that they use different visco-clastic formulations.
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- checking of element material behavior such as thermal '
strains, creep strains, irradiation strains, etc., by
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comparison with hand calculations for a typical time step
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- checking of initial clastic web stresses by means of simpic beam theory formulas .
There is no existing experimental verification of these codes because adequate techniques have not been developed to measure internal irradiation induced stresses. The assumption is made that measured differential i material properties, when employed in a conventional finite element analysis program, will provide a successful simulation of integral structural behavior for complex geometries.
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i RESPONSE TO MPC QUESTION 211.9 GN H-451 CRAPHITE i
QUESTION ,
t 231.9 In the comparison of stress and strength values to obtain de sa margins, the report text (line 7, page 2-6) states that mean ,
strength values were used whereas the footnote to Table 2-1 !
defines design margin in terms of a minitum strength. Please clarify this' apparent contradiction and compare the design mar-gin values that would be obtained using mean strength values versus minimum values. ;
i RESPONSE ;
Both the GLP 5588 text (line 7, page 2-6) and Table 2-1 define design strength and design margin in the same terms. Because the strength of graphite varies as a function of intra-element position, the design strength is conservatively chosen from that area which exhibits the low-est strength. This lowest (minimum) strength location is in the midlength ,
center position of the element. The mean strength of the graphite at this position is evaluated over the log population. The resulting value is used for design purposes and is listed in Tables 4-2 and '4-8 (GLP-5588).
Thus, the design margin is defined as the mean strength value measured at the minimum (lowest) strength location in the element (mi,dlength center position) minus the stress at the area of concern. This definition is used both in the report text and in Tabic 2-1, as illustrated by the iden-tical design margins quoted in the text and in Tabic 2-1.
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. i RESPONSE TO NRC QUESTIONS 231.10 AND 231.11 ON H-451 GRAPH 1TE QUESTIONS 231.10 The statement on page 3-3 of the report, viz. that the reaction dependence on fractional graphite burnoff is not particularly sensitive to the change in graphite type, is not strongly sup-ported by the , data shown in Figures 6-1 and 6-2 of report reference 3-2. Neither is the claim disproven. All that is shown is a rather limited data base covering a narrow range of burnoffs for H-451 graphite (in Figure 6-2) and a single (high) moisture concentration and helium flow rate. How closely do k these conditions match expected or measured moisture in leakage conditions in FSV? To what extent would the mode of. oxidation, e.g., whether reaction rate or diffusion controlled, surface or in-depth oxidation, be affected by the coolant conditions? If the reaction rate of H-451 graphite were significantly greater or lesser than that of H-327 (for example, double or half the rate at a given te=perature), how much effect would this have on the predicted accident consequences listed in report Table 3-17 231.11 (a) As stated in the report, the rate of hydrolysis of exposed fuel kernels is dependent upon local fuel temperatures and steam
! concentration. The local steam concentration (adjacent to the fuel rod) will depend upon the concentration in the coolant channel and the rate of reaction of steam with the block graphite.
Thus, the lower the steam-graphite reaction rate, the greater the amount of hydrolysis of exposed fuel particles. Please pro-vide a quantative measure of the sensitivity of the fuel hydrolysis rate and associated release of noble gas fission products to the ,
steam-graphite reaction rate. For example, if the steam-graphite reaction rate were half the rate assumed in GLP-5588, how much would this affect the hydrolysis and noble gas release?
(b) If the fuel rods in the H-451 blocks are fabricated by in-block carbonization, discuss the potential effect of this fabrica- .
tion method on coating failures caused by coating-graphite bonding and tearing (see Q 231.15) and the subsequent potential for fuel hydrolysis due to the ruptured coatings.
RESPONSE TO 231._10 AND 231.11a ,
The statement in the H-451 SAR that "the reaction dependence on frac-tional graphite burnoff is not particularly sensitive to the change in graphite type" is not based on the data given in Figures 6-1 and 6-2 of j l
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10 SAR Ref. 3-2. It is based on data taken from the attached Figure 1, taken from Ref.1 below, where seven different nuclear grade graphites are shown to exhibit this insensitivity.
The experimental conditions given in Figures 6-1 and 6-2 of GLP-5588 Ref. 3-2 are for accelerated laboratory tests and represent a transient accident condition in the Fort St. Vrain reactor. The results presented in Figures 6-1 and 6-2 exhibit similar chemical reactivity over the tem- -
perature of interest. For comparative purposes, the experimental chemical reactivities evidenced in Figurc 6-1 for H-327 were transposed onto Figure 6-2 (see Figure 2). The standard deviation range for the data points shown are included in the figure for both graphite types. It is clear that
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within the uncertainty range of the data, the H-327 reaction rates and H-451 reaction rates are approximately equal over the temperatures of i in*erest, as stated in GLP-5588.
In view of their similarity in chemical reactivity, predicting a dif-forent temperature dependent change in mode of oxidation of H-327 vs. H-451 would be very difficult. However, assuming H-327 has a slightly lower rate of oxidation than H-451, the latter might be expected to approach in-pore diffusion and laminar sub-layer diffusion rate limiting modes at a slightly lower temperature than H-327.
' Neverthelcss, a parameteric calculation of the kind requested in both of these questions has been performed as documented in Ref. 2. The reaction rate of the steam with the fuel element graphite was arbitrarily increased and decreased by a factor of ten to establish the effect on noble gas re-Icase during fuel hydrolysis, the amount of core graphite oxidized, and the PCRV pressure rise in a simulated reactor steam ingress accident condi- i I
tion (large steam leak, correct plant response, peak PCRV pressure marginally below relief setpoint). These calculations were obtained using the OXIDE-3 I
computer code and data base documented in Ref. 1. Specifically:
(1) The method of calculation for hydrolysis was essentially the same .
as in the FSV FSAR Section 14.5.2.2. Contrast Figures 14.5-9, 14.5-10, and 14.5-11 of the FSAR with Figures 3-13, 3-14, and 3-15 of Ref. 1. Also, the resulting time-dependent fission product relcare has the same qualitative behavior (contrast ISV FSAR rigute 14.5-12 with Figure 5-12 of Eci. 2). In
11 particular, the OXIDE-3 code methods assume an all carbide fuel core (Table 3-3 of Ref. 1).
(2) The graphite type was assumed to be H-327 in consideration of the steam-graphite reaction rate. However, it is noted that the FSV FSAR used a somewhat conservatively higher overall reaction rate in accounting for the reaction rate dependence on prior burnoff, catalyst effects, steam pressure dependence, etc.
(3) The calculations were performed for GA Reference 2000 and 3000 MW(t) HTCR plant designs of 1974. However, the results were essentially the same for both size plants and are considered to apply also to a FSV size plant.
Results of the parametric calculations show that a factor of ten increase in reaction rate decreases the noble gas release due to hydrolysis only by 15%. A factor of ten decrease it reaction rate increases the noble gas release by about 20%. This is a relatively small effect. The total amount of graphite oxidized throughout the transient varies by about a factor of two for a tenfold change in reaction rate. This is less than the theoretical square root dependence for steady state diffusion conditions, since the time for temperatures to cool down during the transient is too short for steady state H2 O concentration profiles to be built up. The peak PCRV pressure was found to vary by less than one psi for a tenfold change in reaction rate.
It is concluded from these parametric results that a change in fuel element graphite reaction rate on the order of two would have almost a negligible effect on predicted noble gas release due to fuel hydrolysis (probably within the accuracy of the calculation). The amount of graphite .
oxidized might change by about 20% but this would produce an imperceptible change in peak primary coolant pressure (e.g. conclusions regarding PCRV safety relief valve lifting are unchanged from the FSAR).
l RESPONSE TO 230.11b It is not planned at this time to manufacture the fuel rods in the H-451 graphite elements via the cure-in-place process. If the cure-in-place pro-
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cess is adopted in the future, a separate safety analysis report and request will be submitted.
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REFERENCES
- 1. M. B. Peroomian, A. k'. Barse11, and J. C. Sacger, "0XIDE-3: A t Computer Code for Analysis of HTGR Steam or Air Ingress Accidents", l General Atomic Report CA-A12493 (LTR-7), January 15, 1974. ,
' 2. A. W. Barse11 and M.* B. Peroomian, " Consequences of Water Ingress into the HTGR Primary Coolant", General Atomic Report GA-A13171, [
April 15, 1975. !
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I 3
L*J Temperature dependence of steam-
- - .s.^. --- N.
%+, me* 8 :
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graphite reaction rate data .
=- lt -
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6 - . . ..: 112 0 in helium with a helium ficw S..._
(-
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15 e
RESPONSE TO NRC QUESTION 231.12 ON H-451 CRAPHITE !
QUESTION: f l 231.12 P'rovide data to support the claim (on report page 3-4) that l the fission product retention characteristics of H-451 graphite ,
I are similar to H-327. ,
f L
RESPONSE
The retention characteristics of graphite are discussed here in terms of sorptivity and diffusivity. l
- 1. Sorptivity on Graphite !
1.1 Sorptivity of Strontium Most of the data for sorption of strontium on graphite are for H-327 !
- h graphite. There are fewer data for sorption on H-451 graphita. These j data indicate no significant difference in sorptivity of strontium on l r
H-327 and H-451 graphites. !
The data on sorption of strontium on H-327 graphite are presented in I I
Tables 1 through 3. These data are the results of three experiments I
(Sr-1, Sr-5, and Sr-6) performed at General Atomic using the mass spectrometric - Knudsen cell method. (For a description of the method . I see Refs. 1 through 3.)
The data in Tables 1 through 3 apply only to the Freundlich region. 'All the values of in P in Tables 1 through 3 were derived from a fit to the experimental data for each concentration, C. The experimental data were in the forn in P = a-b/T. The standard deviatfor -
P is 5% or 1csc.
The values of in P given in the three tab]cs are within the temperature
16 :
and concentration ranges of the measurements. ;
All of the data in Tables 1 through 3 have been combined to obtain by the Icast-squares method (Ref. 5) the general' fit function: l in P F(Pa) = (19.38 Shfff) + (-0.324 + O88) y in C (mmol/kg). (1) f There are no GA data for strontium sorption on graphite in the H1..rian regime. The treatment of, sorption in the Hentian regime descrfaud in [
Ref. 6 is. retained; thus, the transition concentration Cgis given by in C = ~2.12 in the Henrian isotherm expression:
t In PH(Pa) = (19.38 - ) + (-1.324 + , ) in C + in C (mmol/kg) . (2) t The data for sorption of strontium on H-451 (Ref. 7) are listed in Table 4 and are compared in Figure 1 with the least-squares fit to the sorption data on H-327 according to Eq. 1. There is reasonable agreement ,
between the sorption isotherms for H-327 and H-451 graphites. The H-451 data had some associated experimental problems at low concentrations (Ref. 7),
which account for the major deviations apparent in Figure 1 for values of f
4' in C (mmol/Kg) less than -0.5.
i 1.2 Sorptivity of Cesium Most of the data for sorption of cesium on graphite are for H-327 graph-
- ite. There are fewer data for sorption on H-451 graphite. These data indicate no significant difference in sorptivity of cesium on H-327 and H-451 graphites.
A series of experiments was performed with H-327 graphite. The data, previoncly unpuMished, are given in Table 5: these data vere obtained using the mass-spectrometer-Knudson ec11 method (Refs.1 through 3).
l 1
-o l
17 All the data in Tabic 5 have been combined to obtain, by the least-squares method (P.cf. 5), the general fit functions:
In P (Pa) = (19. 747 - ) + (-2.077 + 6 ) In C (mmol/kg) (3) p for the Freundlich region and in PH(Pa) + (19.747 - 0 R ) + (-3.077 + 6f10) in C t+ in C (mmol/kg) (4) i for the Henrian region with in C g = 0.545 - 7.775 A0 ' T(K) (5)
The sorptivity data for cesium on H-451 graphite are given in Table 6 i
(see also Ref. 4). The least-squates fit to these data yield the follow-ing fit functions:
In Pp(Pa) = (24.00 - 30) + (-1.561 + 6123) In C (mmol/kg) (6) in the Freundlich region and l In PH(Pa) = (24.00 - 30) + (-2.561 + 6123) in C + in C (mmol/kg) (7)
( t i
in the Henrian region with r
~3 in C g~ = 2.035 - 1.786*10 T(K) (8)
The data for sorptivity of cesium on H-327 graphite and on H-451 graphite ,
are compared in Figure 2 in terms of the general fit functions. This comparison indicates there is no significant difference in cesium sorp-tivity on H-327 and H-451 graphite.
i
- 2. Diffusivity in Graphite i
2.1 Diffusivity of Strontium e
There are no data for the diffusion of strontium in H-451 graphite.
l There are, however, data for diffusion in several other types of graphite.
l '
These data are nhown in Tabic 7 and in rigure 3. There appears to be no dependence on the type of graphite. This conclusion has also been reached
18 by Rowland (Ref. 23). Therefore, it is concluded that the diffusion coefficient of strontium in H-451 does not differ significantly from that in H-327 graphite.
2.2 Diffusivity of Cesium Permention coefficients for cesium in unirradiated H-451 and H-327 graphite have been measured out-of-pile by Mysels and Burkett (Refs. 24,25) and by Chandra and Norman (Ref. 26), res'pectively. The permention coefficient, n, is defined by C2-C1
( J=w L (9) where J is the steady state flux, C , the cesium concentration in the 2
graphite at the graphite-source interface, C1 , the cesium concentration in the graphite at the graphite-sink interface and L is the thickness of the graphite. The experimental apparatus used by Mysels and Birkett and Chandra and Norman were similar, and their results are comparable.
The permeation cocf ficients neasured by Mysels and Birkett over the temperature range 873 to 1288 K are compared in Figure 4 with the per-meation coefficient ceasured by Chandra and Norman at 1273 K. The data in Figure 4 show that the out-of-pile permeation coefficient for cesium in unirradiated H-451 graphite is about ten times larger than that in unirradiated H-327 graphite.
The fact that the permeation coefficient of cesium in unirradiated H-451 graphite differs from that in unirradiated H-327 graphite, while the -
' ' ~ "~
diffusivities of strontium in various graphite types do not differ is attributed to the different modes of transport of the two elements in graphite. The bulk of the data on strontium transport in graphite
19 indicate that diffusion can be described by Fick's Law (Ref. 6).
In the case of cesium, however, transport in graphite is known to be a more complex phenomenon (Ref. 26).
All the cesium permeation experiments discussed above were conducted out-of-pile. It is probable that in-pile, the differences in transport be-havior of cesium on H-451 and H-327 graphites will be significantly reduced as a result of the irradiation induced structural changes in the graphites. The effect of an irradiation field on transport in graphite has been shoun (Ref. 27) for in-pile experiments with H-451 graphite.
Experiments in which the same experimental apparatus has been used both in-pile (Ref. 27) and out-of-pile (Ref. 24, 25), demonstrate that the permeation coefficient for cesium in H-451 graphite is significantly reduced in-pile.
The permeation coefficient of cesium in H-451 graphite in-pile has been
-14 2 found to be 3x10 m /s at 1123 K and a fast neutron fluence of 4.6x 10 n/m (Ref. 27). This value is shown in Figure 4 along with data on ,
the permeation coefficients of cesium in unirradiated H-451 and H-327 graphites. It is clear in the figure that the permeation coefficient 25 at 1123 K is reduced by a factor of 100 at a fast fluence of 4.6x10 n/m2 . As noted *n Appendix A of.the FFAR, the_ median fast fluence in , ,
Fort St. Vrain is 4.8x10 n/m .
l Also shmm in Figure 4 .i , t he fit t o t he d.w.i of licya n t , et.al., (Ref. 10) which was given in Appendix A of the FSAR and used to determine the
, . . . . . .-_ ~ . -- .- . - - -
20 ;
cesium source terms in Section 3.7 of the FSAR. It can be seen that the in-pile permeation coefficient of cesium in H-451 graphite is substantially {
- less than that used in the FSAR for II-327 graphite.
Thus, although the permeation coef ficient of cesium in unirradiated graphite is dif f erent for H-451 vs. H-327 graphite, in-pile experiments indicate that t cesium permention in H-451 graphite is less than that assumed in FSAR analy-t l ses for H-327 graphite, ,
e i
h 4
r i
J L
e y
b t
l ,
t 6
l
21
References:
- 1. "HTGR Ease Program Quarterly Progress Report for the Period Ending February 28, 1967," USAEC Report GA-7801, General Atomic, Division of General Dynamics, April 20, 1967.
- 2. "HTGR Base Program Quarterly Progress Report for the Period Ending May 30, 1969," USAEC Report GA-9372, Gulf General Atomic, June 27, 1969.
- 3. "HTGR Base Program Quarterly Progress Report for the Period Ending August 31, 1970," USAEC Report GA-10288, Gulf General Atomic, September 30, 1970. .
- 4. "HTCR Esse Program Quarterly Progress Report for the Period Ending February 28, 1974," USAEC Report GA-Al2916, General Atomic Company, March 29, 1974.
- 5. Doring, D.A., "Funfit N-Space Function Fitting,"'GAMD-8630, Gulf General Atomic, May 1, 1968.
USAEC Report GA-A13168 (GA-LTR-16), General Atomic Company, December 6, 1974.
. 7. "IkTGRFuelsandCoreDevelopmentProgramQuarterlyProgressReportfor
.the Period Ending November 30, 1974," USAEC Report GA-A13253, General Atomic Company, January 31, 1975.
- 8. "HTGR Chemistry Quarterly Progress Report for the Period October-December 1961," General. Dynamics / General Atomic Division, January 11, r 1962.
- 9. "ESADA Chemistry Quarterly Report, July-September, 1963," General Dynamics / General Atomic Division, October 28, 1963.
- 10. " TARGET A Program for a 1000-MW(e) High Temperature Gas-Cooled Reactor,
- USAEC Report CA-5104, Cencral Dynamics, General Atomic Division, February 29, 1964. .
- 11. Bromley, J., A. R. Paddon, and K. Moul, "The Diffusion of Cesium, Strontium and Bariun Through Porous Graphites," AERE Raport-R3471, Atomic Energy Research Establishment, Harwell, 1962.
- 12. Doyle, L. B., Report N.A.A.-SR-225, North American Aviation, Inc. 1952.
- 13. Young, C.T., "High Temperature Diffusion of Individu?.1 Fission Products ~
~ ~ '
Frem Small Uranium Ir.pregnated Graphite 5amples Under Deutcon Bombard-ment," Report N.A.A.-SR-247, North American Aviation, Inc. , Iby 28, 1953.
. - _ - - . . = --- . -
. . i
- 22
- 14. Large, N.R., Report AERE C/M-318, Atomic Energy Research Establishment, Harwell, Berkshire, 1957.
- 15. Sandalls, F.J., and M.R. Walford, " Laboratory Determinations of Strontium Diffusion Coefficiente, in Graphite," Report AERE R-6911, Atomic. Energy i
Research Establishment, Harwell, Berkshire,1972.
- 16. C. F. Wallroth, N.L. Baldwin, C. B. Scott and L. R. Zumwalt, "Postirradia- [
tion Examination of Peach Bottom Fuel Test Element FTE-3", USAEC Report j GA-A13004, General Atomic, August 15, 1974. ;
- 17. Faircloth, R.L., F.C.W. Pummery, and B.A. Rolls, " Diffusion of Barium, I Strontium and Cerium in Various Grades of Reactor Graphite," Thermo- ,
dynamics: Proceedingb of the Sym_posium on Thermodynamics with Emphasis on Nuclear Materials and Atomic Transport in Solids II 133-152, Inter-national Atomic Energy Agency, Vienna, 1966.
- 18. Bryant, E.A., et al., " Rates and Mechanisms of the Loss of Fission Products from Uranium-Graphite Fuel Materials," Nucl. Sci. Eng. 15, 288-295 (1963).
~
- 19. de Nordwall, H.J. , et al. , " Post Irradiation Radiochemical Analysis of Pluto Loop A Charge 16," Report AERE-R-5405, Atomic Energy Research Establishment, Harwell, Berkshire, June 1967.
- 20. "HTGR Base Program Quarterly Progress Report for the Period Ending February 28, 1969," USAEC Report GA-9227, Gulf General Atomic, i March 28, 1969.
- 21. "HTGR Base Program Quarterly Progress Report for the Period Ending "
i November 30, 1969," USAEC Report GA-9815, Gulf General Atomic, December 30, 1969, 22.. Absassin, J.J., R.J. Blanchard, and J. Carriot, " Determination Des Coef-ficients De Diffusion Du Strontium Dans Le Graphite P JHAN, 3 En Function De'La Temperature,", Compte Reader DMG N Dr 11/74, C.E.A. - D.M.E.C.N, Department De Metallurgie de Grenoble, 5 Mars 1974. ,
- 23. P. R. Rowland, " Mechanisms of Fission Product Migration in Nuclear j Graphite", DP-902, AEE Winfrith, October, 1974: paper presented at Fourth London International Carbon and Graphite Conference, 23-27 ,
September 1974.
- 24. "HTGR Fuels and Core Development Program Quarterly Progress Report [
for The Period Ending May 31, 1975", USERDA Report GA-A13444, General ;
Atomic Co., June 30, 1975.
- 25. "HTGR Fuels and Core Development Program Quarterly . Progress Report for . . 5 The' Period Ending August 31, 1976", USERDA Report GA-A14046, General :
Atomic Co., September 24, 1976.
F
' .- . . 23
- 26. D. Chandra and J.11. Norman, " Diffusion of Cesium Through Graphite",
J. Nucl. Mater. Q, 293 (1976).
- 27. Illt;R Fuels and Core Development Program Quarterly Progress Report for The Period Ending February 28, 1977", USERDA Report GA-A14298, General Atomic Co., March, 1977.
e l
t i . .
i Data for Strontium Sorption on H-327 Graphite (Experiment Sr-1) J Table 1. ,
i i '
conc C' I conc C(*) -
la P (Pa)
(g Sr/kg carbon) (mmol Sr/kg carbon) In C 800 C 900 C 1000 C 1100 C 1200 C 1300 C 1400 C 1500 C 4.34 49.32 3.898 -4.65 -2.75 -1.15 3.66 41.59 3.728 -4.96 -3.01 -1.37 .025 3.09 35.11 3.559 -5.68 -3.64 -1.92 .450 2.72 30.91 3.431 -6.41 -4.26 -2.45 .903 2.48 28.18 3.339 -6.66 -4.45 -2.58 .990 2.19 24.89 3.214 -6.66 -4.47 -2.62- -1.04 1.806 20.52 3.022 -6.9S -4.77 -2.91 -1.31 1.654 18.80 2.934 -7.75 -5.47 -3.54 -1.90 -
1.558 17.70 2.874 -5.84 -3.85 -2.15 1.475 16.76 2.819 -6.16 -4.12 -2.38 1.320 . 15.00 2.708 -6.39 -4.36 -2.62 -1.12 1.191 13.53 2.605 -6.77 -4.73 -3.00 -1.49 0.941 10.69 2.370 -7.32 -5.22 -3.43 -1.88 .53 0.740 8.41 2.129 -8.27 -6.08 -4.22 -2.61 -1.21 0.641 7.28 ^1.986 -6.49 -4.60 -2.97 -1.54 0.533 6.06 1.801 -7.19 -5.22 -3.52 -2.04 0.495 5.62 1.727 -7.66 -5.64 -3.89 -2.36 0.464 5.27 1.663 -7.91 -5.87 -4.11 -2.57 0.414 4.70 1.549 -8.28 -6.22 -4.43 -2.88 0.360 4.09 1.409 -8.62 -6.54 -4.74 -3.18 -1.80 0.309 3.51 1.256 -8.97 -6.88 -5.08 -3.50 -2.11 0.222 2.52 0.925 -7.56 -5.67 -4.02 -2.56 0.1936 2.20 0.788 -7.97 -6.02 -4.33 -2.84 0.1629 1.85 0.616 -8.06 -6.14 -4.47 -2.99 0.1178 1.34 0.292 -8.84 -6.84 -5.09 -3.55 -2.19 0.1035 1.18 0.162 -9.13 -7.12 -5.37 -3.83 -2.46
- 0.0825 0.937 -0.0645 -9.93 -7.83 -6.00 -4.38 -2.95 0.0756 0.859 -0.152 -10.23 -8.11 -6.26 -4.63 -3.18 0.0704 0.800 -0.223 -8.45 -6.60 -4.97 -3.53 0.0604 0.686 -0.376 -8.91 -7.03 -5.37 -3.90 '
0.0556 0.632 -0.459 -9.12 -7.26 -5.62 -4.16
' s, .
Table 1 (Continued) cone C' conc da) in P (Pa)
(g Sr/kg carbon) (r.o1 Sr/kg carbon) in C 800 C 900 C 1000 C 1100Cl1200C 1300 C 1400 C 1500 C 0.0I.91 0.558 -0.583 -9.64 -7.69 -5.97 -4.46 0.0447 0.508 -0.677 -9.90 -7.95 -6.22 -4.68 0.0381 0.433 -0.837 -8.34 '
-6.56 -4.99 0.385 -0.954 -8.61 -6.84 -5.26 0.0339 0.0301 0.342 -1.073 -8.93 -7.12 -5.51 0.'0261 0.297 -1.215 -9.28 -7.46 -5.86 0.265 -1.329 -9.55 -7.70 -6.07 0.0233 0.244 -1. 4 0') -9.71 -7.83 -6.16 0.0215 l
(*)C = 1000 C'/88.
<~ -~ .
~.
t ..
Table 2. Data for Sorption of Strontium on H-327 Graphite (Experiment Sr-5) cone C' conc C(a) In P(Pa)
(e=ol Sr/kg carbon) 1000 C I1100C 1200 C 1300 C 1400 C 1500 C 1600 C (g Sr/kg carbon) in C 0.434 4.93 1.596 -7.44 -4.89 -2.69 0.354 4.02 -
1.392 -8.10 -5.68 0.308 3.50 1.253 -8.77 -6.32 -4.20 0.284 3.23 1.172 -6.46 -4.46 0.256 2.91 1.068 -7.29 -5.14 -3.25 0.226 2.57 0.943 -7.40 -5.49 0.210 2.39 0.870 -7.66 -5.70 -3.99.
0.192 2.18 0.780 -8.02 -6.00 -4.23 -
0.166 1.89 0.635 -8.43 -6.39 -4.60 0.139 1.58 0.457 -8.82 -6.78 -5.00 '
O.114 1.30 0.259 -9.15 -7.19 -5.47 -3.96 0.0488 -8.05 -6.12 -4.43.
0.0924 1.05 0.0679 O.772 -0.259 -8.75 -6.76 -5.02 0.0562 O.639 -0.448 -9.28 -7.19 -5.35 0.517 -0.660 -7.92 -5.94 -4.18 0.0455 -5.04 0.0310 0.352 -1.043 -8.68 -6.75 0.285 -1.254 ' -9.17 -7.17 -5.3S 0.0251 -6.15 0.195 -1.632 -10.01 -7.96 0.0172 -8.64 -6.77 0.0116 0.132 -2.026 0.103 -2.269 -9.08 -7.20 0.00910 -9.97 -7.96 0.00612 0.0695 -2.666 -6.62
-2.900 -8.19 0.00484 0.0550 -8.68 -7.04 0.00398 0.0452 -3.096 -7.41
-3.441 -9.35 0.00282 0.0320 C = 1000 C'/88.
_s
n .n -
Table 3. Data for Strontium Sorption on H-327 Craphite (Experiment Sr-6)
Cone C' Cone c (a) In P(Pa)
(g Sr/'/g carbon) (mmol Sr/kg carbon) In C 1000*C 1100*C ! 1200*C l 1300*C 1400*C l 1500'C 0.342 3.89 1.357 -6.44 -4.14 0.284 3.23 1.172 -7.43 -5.13 0.248 2.82 1.036 -8.26 -5.78 0.219 2.49 0.912 -8.76 -6.42 1 0.202 2.30 0.831 -9.15 -6.86 -4.88 .
0.180 2.05 0.716 -9.82 -7.47 -5.44 0.170 1.93 0.658 -7.64 -5.66 -3.94 0.148 1.68 0.520 -8.27 -6.30 -4.57
-6.92 -5.14
~
0.134 1.52 0.421 -8.97 0.110 1.25 0.223 -7.35 -5.60 -4.07 0.0948 1.08 0.0744 -7.92 -6.09 -4.48 0.0712 0.809 -0.212 -8.62 -6.77 -5.15 0.0597 0.678 -0.388 -9.33 -7.36 -5.63 -4.09 0.0463 0.526 -0.642 -8.02 -6.29 -4.75 0.0352 0.400 -0.916 -8.92 -7.07 -5.44 0.0244 0.277 -1.283 -9.91 -7.96 -6.24 0.0186 0.211 -
-1.554 -8.61 -6.82 0.0150 0.170 -1.769 -9.03 -7.27 0.0128 0.145 -1.928 -9.43 -7.62 0.0116 0.132 -2.026 -9.72 -7.84 0.00983 0.112 -2.192 -10.01 -8.20 (a).C = ,1000 C'/88.
- 1 e .
5.
. . . _ _ .- h _. _
i .
' o r -
Table 4. Data for Strontium Sorption on H-451 Graphite (Experiment Sr-8)
In T (Pa)
Cone C' Cone C(a)
(g Sr/kr. carbon) (e=ol Sr/kg carbon) In C 800 C 900 C - 1000 C 1100 C 1200 C 1300 C 1400*C 1500 C 1600 C e
8.824 100.3 4.608 -4.39 -2.10 .'
7.833 89.1 4.489 -3.34 -1.36 0.311 5.271 59.9 4.093 -3.92 -1.89 -0.178 3.753 42.7 3.754 -5.44 -3.21 -1.34 2.790 31.7 3.456 -5.89 -3.66 -1.78 -0.179 2.083 23.7 3.164 -4.86 -2.89 -1.22 -
1.612 18.3 2.908 -5.74 -3.75 -2.05 1.170 13.3 2.587 -7.16 -5.03 -3.22 1 0.9369 10.6 2.365 -6.32 -4.36 -2.66 0.7462 8.48 2.138 -7.06 -5.07 -3.35 0.6185 7.03 1.950 -7.72 -5.68 -3.91 -
0.4558 5.18 1.645 -6.36 -4.57 -3.00 0.3510 4.10 1.412 -6.56 -4.82 -3.31 0.2132 2.42 0.885 -7.46 -5.68 -4.12 -2.75 0.1359 1.54 0.435 -6.81 -5.20 -3.7S 0.1000 1.14 0.128 -7.S6 -6.23 -4.80 0.0786 0.893 -0.113 -9.12 -7.35 -5.79 -4.41 0.0598 0.680 -0.386 -8.09 -6.49 -5.08 0.0546 0.620 -0.477 -9.18 -7.55 -6.11 0.0455 0.517 -0.660 -8.06 -6.53 -5.16 0.0399 0.453 -0.791 -8.59 -7.00 -5.57 0.0252 0.286 -1.250 -9.17 -7.56 -6.13 0.168 -1.783 -9.67 -8.06 -6.63 0.0148 I*I C - 1000 C'/(88). .
Table S. Data for Cesium Sorption on Unitradiateu H-327 Craphite Conc. C' Conc. C(*} y p (pg g Cs (kgcarbon [kgcarbon/mol InCs C } 600*C 700*C 800*C 900*C 1000*C 1100*C 1200*C 1300*C i
Run Cs-2 2.315 7.91 2.068 -3.43 2.17 7.42 2.004 -3.77 1.935 6.61 1.889 -4.56 -1.85 1.75 -
5.98 1.788 -5.44 -2.66 .
1.67 5.71 1.742 -5.83 -3.05 1.54 5.26 1.660 -6.42 -3.63 .
1.485 , 5.08 1.625 -6.77 -3.93 1.37 4.68 1.543 -7.24 -4.31 ,
1.33 4.54 1.513 -7.45 -4.60 -2.27 1.19 ,
4.07 1.404 -8.40 -5.31 -2.80 1.145 i 3.91 1.364 -8.45 -5.49 -3.09 0.979; 3.35 1.209 -6.44 -3.87 0.897' 3.07 1.122 -7.00 -4.43 -2.32 0.7675' 2.62 0.963 -7.83 -5.21 -3.05 0.660 2.26 0.815 -5.85 -3.63 -1.74 .
0.496 1.70 0.531 -7.06 -4.69 -2.69 0.387 ' 1.32 0.278 -7.97 -5.56 -3.53 -1.77 0.256 .
i 0.875 -0.134 -6.72 -4.61 -2.78 0.1575 0.538 -0.620 -7.76 -5.57 -3.69 -2.09 0.0539 0.184 -1.693 -9.38 -7.13 -5.19 -3.54 0.0167 4 0.0571 -2.863 -6.88 -5.16 -3.66
- Table 5 (Continued) ~ }'
Cone. C' Cone. C("} in P (Pa)
- g Cs j (kg carbon /
[nnelCsj
\kg carbon / In C( } 600*C 700*C 800*C 900*C 1000'C 1100'C 12'00*C 1300*C Run Cs-3 0.899 3.07 1.122 -8,57 -5.66 -3.29 0.8625 2.95 1.082 -9.02 -6.02 -3.59 I
0.7835 2.68 0.986 -6.67 -4.06 0.7485 2.56 0.940 -7.01 -4.40
~
0.6705 2.29 0.829 -7.54 -4.99 '
~
0.644 2.20 0.788 -7.76 -5.25 0.5745 1.96 0.673 -8.40 -5.77 0.558 1.91 0.647 -8.56 -5.93 -3.76 0.468 1.60 0.470 -6.49 -4.24 0.4325 1.48 0.392 -9.61 -6.92 -4.71 -2.83
~
0.360 1.23 0.207 -7.54 -5.22 0.3285 1.12 0.113 -7.88 -5.56 -3.60 0.2545 0.870 -0.139 -8.62 -6.18 -4.13
'.2255 O '
O.771 -0.260 -9.01 -6.55 -4.47 .
0.177 0.605 ,-0.503 -9.55 -7.11 -5.04 0.1405 0.480 -0.734 -7.56 -5.47 -3.68 -
0.1045 0.357 -1.030 s -8.29 -6.14 -4.28 0.0584 0.200 -1.609 -7.16 -5.25 -3.63 0.0395 0.135 -2.002 -7.77 -5.93 -4.37 .
0.01985 0.0678 -2.691 -7.23 -5.54 o
m , ., .
~
i l '
s
. Table 5 (Continued) .
Cone. C' Conc. C(*} in (Pa) e g Cs j /mmol'Cs j in C( 700*C 800*C 900*C 1000*C 1100*C 1200*C 1300*C .
kkgcarbon/ \kg carbon / 600*C Run Cs-4 .
1.044 3.57 1.273 -5.17 -2.41 1.015 3.47 1.244 -6.02 -3.30 .
0.9955 3.40 1.224 -6.53 -3.79 .
0.9715 3.32 1.200 -7.16 -4.29 .
0.9505 3.25 1.179 -7.54 -4.71 0.9265 3.17 1.154 -8.15 -5.22 -2.82 .
0.907 3.10 1.131 -8.64 -5.64 -3.20 0.885 3.02 1.105 -9.22 -6.09 -3.54 -
0.8595 2.94 1.078 -6.46 -3.74 0.836 2.85 1.047 -6.79 -4.06 0.795 2.72 1.001 -7.03 -4.27 .
0.7525 2.57 0.944 -7.55 -4.62 0.735 2.51 0.920 -7.76 -4.99 -2.70 ,
0.681 2.33 0.846 -8.10 -5.22 -2.86 .
0.6635 2.27 0.829 -5.44 -3.07 ,.
0.6315 2.16 0.770 -8,59 -5.76 -3.43 0.557 1.90 0.642 -9.09 -6.18 -3.78 .
^
, , , , , - - ,n-- -
Table 5 (Continued)
Cone. C' Cone. C(a)
/ g Cs j [ c: tol Cs j in P (Pa)
\kg carbon / \kg carbon / In C( ) 600*C 700*C 800*C 900*C 1000*C 1100*C 1200*C 1300*C
, Run Cs-4 (Continued) 0.520 1.78 0.577 -9.73 -6.65 -4.13 0.485 1.66 0.507 -9.93 -6.89 -4.39 .
0.4715 .
1.61 0.476 -10.15 -7.07 -4.55 -
0.419 1.43 0.358 -7.37 -4.87 -2.76 0.3755 1.28 0.247 -7.57 -5.10 -3,00 '
. 0.333 1.14 0.131 -8.01 -5.57 -3.50 0.3125 1.07 0.068 -8.38 -5.86 -3.72 1
0.288 0.984 -0.016 -8.72 -6.17 -4.01 -
0.258 O.882 -0.126 -8.90 -6.42 -4.31 0.2295 0.784 -0.243 -9.27 -6.78 -4.68 -2.87 O.172 O.588 -0.531 -7.28 -5.17 -3.35 0.1525 , 0.521 -0.652 -7.68 -5.52 -3.65 0.138 0.472 -0.751 -8.11 -5.96 -4.10 ,
0.108 0.369 -0.997 -6.34 -4.47 -2.88 0.08765 0.300 -1.204 -6.92 ~5.06 -3.47 0.05865 0.200 -1.609 -5.80 -4.19
, 0.04605 0.157 -1.852 -6.65 -4.99
-u -
A
t ,, .,
. l Table 5 (Continued) - -
Cone. C' Conc. C(*)
[ g Cs j [mmolCsj in P (Pa)
\kg carbon / \kg carbon / In C(b) 600*C 700*C 800*C 900*C 1000*C 1100*C 1200*C 1300*C Run Cs-6 1.0595 3.62 1.286 -7.12 -4.39 -2.16
- 0.779 2.66 0.978 -9.90 -6.92 -4.47 0.692 2.37 0.863 -10.44 -7.45 -5.03 -3.02 0.5005 1.71 0.536 -9.12 -6.47 -4.29 -2.45 .
0.2355 0.805 -0.217 -11.92 -9.01 -6.63 -4.62 -2.88 0.1014 0.347 -1.058 -11.15 -8.60 -6.45 -4.59 -3.01 .
Run Cs-8(c) 2.097 7.17 1.970 -3 81 2.026 6.92 1.934 -4.08 -1.90 .
1.5695 5.36 1.679 -3.33 -1.32 1.182 4.04 1.396 -4.76 -2.64 -0.90 0.8395 2.87 1.054 -6.33 -4.18 -2.40 -
0.t875 1.67 0.513 -8.69 -6.16 -4.08 -2.32 0.2885 0.986 -0.014 -7.71 -5.43 -3.49' -1.82 .
0.175 0.598 -0.514 -8.78 -6.43 -4.44 -2.73 -1.27 ,
0.08695 0.297 -1.214 -9.89 -7.50 -5.47 '3.72
-2.24 0.03525 0.120 -2.120 -11.16 -8.71 -6.64 -4.85 -3.33 -1.99 0.0158 0.0540 -2.919 -9.89 -7.79 -5.98 -4.44 -3.09 ,
0.0081 0.0277 -3.586 -11.21 -9.13 -7.34 -5.81 -4.48 0.00705 0.0241 -3.726 -12.61 -10.47 -8.63 -7.06 -5.69 0.0068 0.0232 -3.764 -11.94 -10.15 -8.63 .
Azure a suuntAnueuf ,
Conc. C' Cone. C(*} .
A* i ~ ' *) -
g Cs
) (kgcarbon/=molCs) 800'C 900'c 1000*C 1100'c 1200*C 1300*C l,
- kg carbon / \ In C(b) 600*C 700*C Run Cs-9(c) 0.96225 3.29 1.191 -8.63 -5.73 0.9212 3.15 1.147 -9.60 -6.43 -3.84 .
0.8466 2.89 1.061 -10.26 -7.12 -4.55 -
0.7027 2.40 0.875 -8.07 -5.50 -3.54 0.5410 1.85 0.615 -9.07 -6.51 -4.41 -2.63 0.38595 1.32 0.278 -10.23 -7.61 -5.44 -3.61 0.2492 0.852 -0.160 -8.90 -6.60 -4.66 .
0.14275 0.488 -0.717 -10.16 -7.73 -5.68 -3.90
-6.60 -4.71 0.0768 0.262 -1.339 -11.37 -8.78 -
O.02635' O.0901 -2.407 -12.94 -10.23 -7.93 -5.95 0.00945 0.0323 -3.433 -11.12 -8.86 -6.90 0.0044 0.0150 -4.200 -12.16 -9.95 -8.04 -6.42
-4.825 -13.11 -10.84 -8.88 -7. 21 )
O.00235 0.00803 .
I Run Cs-11 0.4216 1.44 0.365 -8.45 -5.72 0.29615 0.010 -9.85 -7.17 -4.95
)01
. 0.21685 '0.741 -0.300 -8.26 -6.00 .
0.166.9 0.570 -0.562 -9.11 -6.87 0.13105 0.448 -0.803 -9.89 -7.53
. 0.10235 0.350 -1.050 -10.86 -8.34 -6.20 ,
0.07165 0.245 -1.406 -9.05 -6.88 0.0453 0.155 -1.864 -10.10 -7.87 -5.95 -
)C = 1000 C'/(133)(2.2); this equation provides the correction for particle size and for conversion of concentration units to mmol/kg carbon.
(b)1n represents the natural logarithm. w I (c) Data for lowest three concentration values are in the pressure decline region
'__________ a __.
Table 6. Data for Cesium Sorption on H-451 Graphite (Run Cs -39) .
Co- C' t 1 ,p (p,)
700'C 800*C 900*C 1000'C 1100*C N iaTbon kkg r on in C(b) 600*C 7.953 2.074 -5.49 -2.79 -0.61 2.327 5.663 1.734 -7.50 -4.53 -2.11 1.657 '
4.901 1.389 -8.61 -5.44 -2.87 1.434 '
1.188 4.060 1.401 -6.44 -3.76 -1.54 1.024 3.500 1.253 -7.10 -4.30 -2.00 2.678 0.985 -8.08 -5.18 -2.80 0.7836
- 2.215
- 0.795 -6.13 -3.69 0.6481 1.488 0.397 -7.14 -4.65 -2.54 0.4355 '
0.045 -8.27 -5.64 -3.42 0.3062 1.046 0.7129 -0.338 -9.32 -6.66 -4.40 0.2086
-0.856 7 -7.58 -5.24 -3.21 0.1243 0.4248
-1.268 -7.88 -5.62 -3.66 0.0823 0.2813
(*)C = 1000 C'/(133)(2.2); this equation provides the correction for particle size and for conversion of concentration units to etnol/kg carbon.
(b)1n re. presents the natural logarithm.
Table 7 . Diffusion Coefficients for Strontium in Graphite ..
Temp R f*
2 Type of Graphite (*K) D (cm /sec) Comments No.
8 EL}'-85 i=pregnated (H) 1273 1 x 10 {
HLM-85 base stock 1273 1x10]# Frofiles CL-I-C6 1273 4 x 10 Saturati n concentration
-6 Dii-3C 1273 8 x 10 -6 1-10 v= ole /g GL-I-S10 1273 3 x 10
~0 Speer 711 TS 1273 %3 x 10 Profiles, initial loading 9 10 pmole/g Speer 711 TS or HLM-85
~
10 1273 1 x 10 Profiles .
1360 11 Type A British reactor grade 3.5 x 10_6 Profile analysis - approximately 1070 1.4 x 10_7 confirmed by permention -
770 2.7 x 10 measurements L ding 300 paole/g 1380
-5 EY-9 1.5 x 10
-6 1180 7.9 x 10
-6 1035 2.5 x 10 -6 1030 2.7 x .10-6 970 1.9 x 10
-6 880 1.1 x 10 HX-10 1380 1.3 x 10_5 6 -
1180 7.7 x 10-
. 770 9.6 x 10 .
~
HX-12 1370 6.9x10j .
1180 -
5.7 x 10_7
~
990 7.1 x 10 .
990 9.6x10[7 830 4.4 x 10 .
- -- ,n-
-s ., ,
Table 7 (Continued) _ , _ ,
Temp Ref.
2 Type of Graphite (*K) D (cm /sec) Comments No.
14 A'JF 1973 1.5x10[ Calculated from Refs. 12,13 1873 3 x 10- -
1773 4 x 10_5 9 1073 3.8 x 10 A.G.~ 1860 Profiles 15 9
1690 3.0x10[h 4.0 x 10 C < 0.7 p= ole /g ..
1610 1.5 x 10,73 1480 1.1 x 10 ,
A.G.L. 9 ,
1100 2.0x10[6 C = 20 pmole/g 1333 1.8 x 10 Ear 67 . 1860 6.9x10[6 C < 0.7 puole/g 1750 2.1 x 10_6 7 i 1655 7.7 x 10 -
i 1580 1.9 x 10_7 ,
1485 3.7 x 10 '
I "I .. : 16" ;
1590 1.1 x 10-
-7 Bri:ish Aches 6n pitchcoke 1575 1.6 x 10 Spec pure i 1610 3.0 x 10-
-f2 Dr:3cn fuel. tube (HX30) 1150 5 x 10 j Profiles 15 1220 5 x 10 e
t
Table 7 (Continued)
Temp 2
R*f* '
Type of Graphite (*K) D (cm /sec') Co=ments No.
H327 graphite (FTEs) " bulk" 1055 4.6x10[ Profiles 16 1250 S.2 x 10_9 C = 0.2 to 1.4 pmole/g -
1250 1.1 x 10_
1113 8.1 x 10_j);
1113 6.3 x 10 ;10
. 1241 2.9 x 10- -
1241 2.5 x 10_10 1230 4.1 x 10_9 9
. 1230 1.6 x 10 , ,
CL*-163 doubic furfuryl 1026 3.7 x 10,~ Profiles 17 impregnated graphite 1026 5.1 x 10_8 7 Leading >600 pmole/g 1121 2.4 x 10 1196 4.4 x 10-Self-made admixed and graphi- 1823 2.8x10[5 Release rate of recoils from 18 ti:cd (UO2 + C + binder).or 2198 6.6 x 10 U 1: pregnated AUC graphite 2473 2.0 x 10
-4 imprggnatdcylinders.
<10- p= ole /g Loading 0
Pluto loop fuel tubes 1173 D calculated from steep part 19 1273 2x10[9 1 x 10 . f c ncentration profile.
-8 1393 2 x 10 Loading %0.1 pnole/g 0
, 1333- 1 x 10 1673 2 x 10
-8 1743 2 x 10 .
4 0
v a
I Table 7 (Continued) _ . , _ . . . -
Temp Type of Graphite 2 Ref.
(*K) D (cm /sec) Comments No.
H327 Graphite
_ Permeation (in helium) 20,21 1248 C %4 pmole/g 4.7 x 10_7 7 . ...
1248 C %3.4 prole /g 1273 4.3 2.4 xx 10_9-10 C s0.08 umole/g P rnenti n (in v cuum) 1273 C 10.2 pmole/g 1273 31 xx 10_10 10 ;; C %0.045 pmole/g Pc:hiney P JHAN 1373 9 Profile analysis 3
1423 4.15x10[3 -
22 .
2.56 y 10 , C s 0.01 paole/g 1463 6.34 x 10_~
1573 1673 3.16 x 10_7 9.29 x 10
-6 1773 3.17 x 10 -6 1833 5.70 x 10 Pcic.incy P JEAN -9
- 3 1373 3.5 x 10 -
Act v ty change in source 22
. 1423 5.2 x 10_8 C s 0.01 puole/g 3
1463 1573 8.2 x 10_7 5.4 x 10 1673 1.2 x 10 6
- 1773 1.7 x 10 8.1 x 10 -6 1883 d
s .
e 8
40
. . i 1000 C A o -
O r e
H 321 3 -
O B00'C
'I A 1000*c -
a O 1200'C r H 451 O 1400 c 12000C V 1600 C e 14000C 8000C O D 3 -
1600 C .
Q O a - 0 o O
_ O o g -5 -
A m V
- O O
6 V
A V
1 -
g O
O ^
4 -
0 0
9 -
O D .
O
-10 -
.jg i i l I _ l I I l 3 2 .I O 1 2 3 4 5 l In C (mmot/kg) l Fig. 1. Comparison of Strontium Data for 11-451 Graphite With Iso-thern2 for l1-327 Graphtte l
e 41 Fig. 2: Compar1non of Sorption Isotherms for Cesf um on if-327 and II-451 Gyaphite 10 -
~ /
/
- - H-451 /
10 -- H-327 ,'
- - - - - - - - 95 % confidence
~
limit for 11-451 data /
, /
/ /
s
/ /
/
10 -
./ /
- ,/ /
/
I
~ s* / /
/
' /
. ,- / ,
1
/
T 1300 C' / ,
f ,
/
- / j
,s' // j
/
~
' /
/
10 -
/ /
- # /
5 '
,' j ,/
s' / ,'
E
~2 -#' ,- /
/
10 -
f
/
- /
l
/
,s' r
/ f' /
\
s' <
,' /
o
-3 / ,/ /
10 -
/
' ' s' /
/ /
o .-
/
900 ,s C' -
-4
~
- /,"
10 -
,' ,/ ,
s' / ,
,s' ,/ ,
/ -
10~ ~~
- /
i 10
-6 P'l ll 1 1 II! I I I Il l t t i
~
~ ~I 30-10 10 1 10 C(mmol/kg)
lt2 ,
' -3 I-
- , 1, Ia I. I, -
D -
.s -
. 10 -
10
-5 ,
- O -
( -
+ .
i _
6 _.
t 10 -
+
i -
O -
4 c. _ -
- i, -
e "s 10-1 + .
+
. A6 -
( 10'
-~
T
'l SELECTED REFERENCE $ -
(.. -
x .
t ,
a 3 - O -
10 - 1 -
REr. -
0 15 x .-
x -
0 18 iD
-10 r ^ 19 0 D x d
- I 15 -
~
~
D 20,21 D x 16 :
_ _ h
.ii +
to -
22 -
~
~
d d x !
I(
.s r /
10
.n !. I I I 4 5 6 7 8 9 10 .
10 /T (*K)
Fig. 3. Strontium dif fusion coef ficients in graphite .
L
43 Fig. 4: Comparison of permention Coefficients in H-327 and H-451 Graphite
-10
?q?^$5N e.g. . .
.g , ,. e i' i. e i*.
.~~ T -
4, \~ T- . . . . .i;; iisi l iit-si : iii
!!I' \lI
'I!!! l lll. 1 !lll !ll,'
4 10-11
! N~i l '^\l l' ll' ll
~'
n =r 0
4 __ t _ :.- - a=_Es\ -
_a .=.u---
. ,l i,
.l; g ['l ~'~}- siti .I- * :: leta t ill !!!!
t, .
- Illi l i ! o.-Q'j 4).
\!Ill. ~, I!!; illi Illi ,
u i jlI-I i -
w '
l ll !I 8 10-12 1 ! 1 i\l .
u __ ----\-__ - - ~ ~ -
g .._q .( _\
o - -
t- m y .i . , . .
.y iT 3
g . .-
t, i;..
g i i!1 .I: t it i il ,\i'. i'!i iiI ,
B . .
l Illl llll illi ' !. I l\i l !!il li
$ ii{ l Ii Il81 I!!
- \ i i i j 10-13
!!3 l !! I \ 2 -
. _:_ 3--
~
%, _ m
. . .s i ... i i .. 6 A. .
( l :il ! i : _M +i i i*.i-
_ ,.__, [ 1 1' I!Il lllQl!.Iilll l Sh, '!I _. h!!l 10-14 ll ._,
l
!\
l 1
!!l i !.
5 6 7 8 9 10 11 12 13 4
. 10 /T(K) .
Open synbols are dita of Mysels & Birkett (Refs. 24, 25) on unitradia ted H-451 out-of-pile.
Solid Symbol is datum of Chandra & Norman (Ref. 26) on unirradiated H-327 out-of-pile.
Italf-filled symbol is datum of Mysels and Birkett. . .(Ref. . 27) on H-451 in-pile.
Q Average of eight values at 1133 K.
O Average of t enty-one values at 1143 K.
Broken line is the fit to the data of Bryant, et.al., (Ref. 18) r,1ven in the FS\' FSAR and used to determine cesium source terms in the PSAR.
44 e
RESPONSE TO NRC OUESTION 231.14 DN H-451 GRAPHITE QUESTION 231.14 Reference the data sources which provided the basis for the equa-tion for irradiation strain (Eq. 4-2 in GLP-5588). If these data were obtained on near-isotropic graphites other than H-451, discuss the applicability of the data to the prediction of H-451 irradiation strain.
( RESPONSE The principal sources of data for the equation for irradiation-induced strain (Eq. 4-2 and Table 4-1 of GLP-5588) were measurements on H-451 graph-ite irradiated in capsules OG-1 and OG-2 (Ref.1). Secondary data sources were high fluence points obtained on H-429 graphite irradiated in the same capsules (Ref. 2) and low temperature data points from UKAEA irradiation on a near-isotropic prototype grade designated PI (Ref. 3). Grade H-429 was a prototype for H-451 which employed the same raw materials. Grade PI was an extruded petroleum coke based graphite whose thermal ex-5 pansion coefficients (3.5 x 10-6 oC-1 parallel to extrusion and 4.1 x 10-6 oC-1 perpendicular to extrusion, measured between 200C and 1200C)
I (,
j are similar to those of H-451. More recent data on H-451 graphite irradiated to full FSV exposure (Ref. 4) confirm the applicability of Eq. 4-2.
REFERENCES
- 1. R. J. Price and L. A. Beavan, " Final Report on Graphite Irradiation
- Test OG-2", ERDA Report GA-A13556, October 1975.
t
- 2. W. R. Johnson and G. B. Engle, " Properties of Unirradiated Fuel "'ement Graphites H-451 and TS-1240", GA-A13752, January 31, 1976. !
- 3. R. J. Price, " Property Changes in Near-Isotropic Graphite Irradiated at 3000 to 6000C: A Literature Survey" ERDA Report GA-A13478, June 1975.
- 4. R. J. Price and L. A. Beavan, " Final Report on Graphite Irradiation Test OG-3", GA-A14211, January 1977. "
y . . _ , _ _ 3 - _ ,
I s ' . . *j - .
45 RESP 0MSE TO NRC QUESTION 231.16 ON 11-451 GRAPillTE
_UESTION Q
231.16 (a) The creep equation provided in GLP-5588, Eq. 4-3 appears ,
to have been derived originally for Gilsocarbon-based graphites, whereas H-451 graphite is a petroleum coke-based graphite; (the transient creep term in Eq. 4-3 is identical to that for Gils-carbon isotrop,1c graphite creep, as presented in Gulf General Atomic Report GA-B12331 (November 1, 1971)). Please discuss the rationale for extending the use of a creep equation for Gilsocarbon-based isotropic graphites to a petroleum coke-based
( graphite such as H-451.
(b) What tests have been performed, are in progress, or are planned to verify the applicability of Gilsocarbon-based graphite creep data to the H-451 graphite? For ' tests in progress or plan- '
ned, compare schedules to the predicted exposure schedule for the first H-451 elements to be irradiated in FSV.
(c) Equation 4-3 is said to be valid for both tensile and com-pressive stresces at temperatures between 5000 to 12000C, fast fluences up to 1026 n/m2, and creep strains up to 2.5 x 10-2 in tension and 2 x 10-2 in compression. Please list the actual temperatures, fluences, stresses, and strains to which H-451 graphite has been tested along with the maximum values expected in the FSV reactor. Indicate how compressive creep data are to '
be applied to the prediction of tensile creep and vice versa.
s RESP 0NSE TO 231.16a ,
The equation for irradiation induced creep strain in H-451 graphite ,
(Eq. 4-3 of GLP-5588) was derived from restrained shrinkage experirants conducted at RCN, Petten on near-isotropic petroleum coke-based graphites, ,
including H-451 (Ref. 1). The data for the secondary creep coefficient are shown in Figure 1.
Restrained shrinkage tests do not give independent values.for tran-
- _. , _ _ _ sient creep; the transient creep term derived earlier for Gilsocarbon- L based graphites (Ref. 2) was therefore retained. This is justified be- !
cause the transient creep strain depends primarily on Young's modulus of the graphite, and the Young's modulus taken for Gilsocarbon-based graphite f (1.1 x 106 p:.,1) is e inil.ir to that of 11-45] (CI.I'-55Sd, Table 4-3) .
P A
/*.*,* .
! 46 J-RESPONSE TO 231.16b Irradiation creep tests completed, in progress, or currently planned for H-451 graphite are summarized in Table 1. The table includes scheduled completion dates. The earliest time at which H-451 graphite fuel elements (other than test elements) could be placed in the Fort St. Vrain reactor ,
1 core is late in 1980 or early in 1981.
RESPONSE TO 231.16e i
The actuni temperatures, fluences, stresses and strains,to which
( H-451 graphite has been creep-tested are given in Table 1. Corresponding values expected for FSV fuel elements are:
Operating temperature rang 6 (oC): 420-1210 Maximum fluences (x1021 n/cm2): 8.0 Maximum stress (MPa): 1.3 Maximum creep strain (%): 0.52 Irradiation creep constants for graphite are the same in tension and compression. During analysis of the data, values from both types of test are combined and the lumped values are used for the prediction of creep strains in both tension and compression.
REFERENCES
- 1. H. J. Veringa and R. Blackstone, Carbon, Jj, 279 (1976)
- 2. R. J. Price, " Review of Irradiation-Induced Creep in Graphite Under HTGR Conditions", ERDA Report CA-A12332, November 1972.
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Figure 1 Steady state irradiation creep constant for near-isotropic petrolcun coke nr.qhites as a function' of irradiation temperaturc
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IRRADIATION CREEP EXPERIMENTS ON H-451 GRAPHITE Maxi- Max.
Tempera- Maximum mum Creep I ture Fluence (n/cm2 ) Stress Strain Completion Type of Test (OC) (E > 0.18 MeV, HTGR) (MPa) (7.) Date Laboratory RCN, Pctten Restrained shrinkage 870-1115 3.7.x 10 21 10.8 +1.66 1975 21 +0.41 JRC, Petten Controlled tensile load 835 3.3 x 10 -
6.0 1977 21 JRC, Petten Controlled tensile load 840 7.4 x 10 6.0 W1. 0 1980 21 20.7 -2.7 1977 ORNL Controlled compressive load 900 1.2 x 10 21 20.7 1982 ORNL Controlled compressive load 600 s8 x 10 ---
21 20.7 1983 ORNL Controlled compressive load 1250 $8 x 10 ---
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49 RESPONSE TO NRC QUESTION 130.1 ON H-451 GRAPHITE QU ESTION 130.1 In Section 2.4, Graphite Structural Analysis, the following informa-tion should be provided:
(a) The manner in which fatigue behavior of the H-451 Graphite is considered.
(b) A description of the method of analysis used to determine the seismic stresses in the graphite fuel and reflector elements.
( (c) The loads and load combinations which give rise to the tensile stresses shown in Table 2-1.
RESPONSE TO 130.la It has been shown that unirradiated H-451 graphite does not experience fatigue damage under reversea stress cycling unless stressed above a homo-logous stress limit of 0.63 in the axial direction or 0.74 in the radial orientation (Ref. 1). Homologous stress is defined as the maximum applied fatigue stress divided by the tensile strength. Irradiation increases both the strength of the specimens and the homologous stress limits for
( fatigue endurance.
Both secondary (Table 2-1) and primary loadings are well below these homologous stress limits, and during normal operation, no fatigue inducing mechanism exists. Therefore, no fatigue analysis is necessary for the H-451 graphite elements.
RESPONSE TO 130.lb The seismic analysis of the FSV H-327 reactor internais was conducted as a dynamic analysis, treating the core as a monolithic beam. This analysis gave responses in terms of g-lohds at vhriobs lochtions' ins ~ide the PCRV. ~The"
~ '
excitation spectrum developed for the core internals system was a relative displacement spectrum utilizing the response levels of the PCRV obtained from the PCRV dynamic analysis. hinun lateral carthquake loads were deter-l mined. Having established the maximun response, in terms of g-loads, a l
l
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50 stress analysis was conducted. The maximum g-load was applied laterally as a body force to all the internal components, which were then analyzed using static analysis techniques.
The maximum dowel force was calculated at the reflector element inter-face immediately below the keyed top reflector plenum element within each column. The dowel shear force was calculated using one half the mass of the column based on the assumption that the column behaves as a rigid beam with end support, and the entire load is borne by one dowel with no interface friction. The maximum dowel shear load for the safe shutdown
- carthquake was deternined to be 400 lbs.
Testing of full scale structures under simulated seismic loading was also conducted to substantiate the strength of the fuel elements.
Compression testing and impact fatigue tests were conducted on FSV H-327 fuel elament bodies, yielding one cycle strength greater than 110,000 lbf (Ref. 2). This compares with a 0.26g design seiscic loading of approxi-mately 1480 lbs (determined from a 0.26g lateral loading by 20 fuel elements across the core diameter for the 0.lg horizontal ground acceleration safe shutdown earthquake).
Dowel strength tests were conducted to evaluate the shear strength of g
the fuel element dowel / socket system. As stated in Appendix A.l.5.2 of the FSAR, a single H-327 dowel / socket system withstood a minimum shear force of over 950 lbs before fa'ilure occurred. The mean failure load in these tests was 1148 lbs.
H-451 fuel elements would be expected to experience increased aeismic loading due to their increased radial modulus. The loading is estimated ,
to increase as the square root of the ratio of the moduli, or by a factor of 1.3. 'Since the radial strength of H-451 fuel elements is greater than that of H-327 elements by more than a factor of 1.6, the increase in load is more than accounted for by the corresponding increase in H-451 strength.
l l RESPONSE TO 130.lc Table 2-1 of GLP-55SS is a compilation of the raaximum secondary stresses caused by thereal and irradiation induced dimensional change for
+
N
- 51
-l operating and shutdown conditions. This table does not include combinations of primary loading stresses.
The major tensile primary stress will be from a seismic event. As shown in the answer to question 130.lb, the load will be small (about 1480 lbs.
lateral load for the safe shutdown earthquake). Based on test results, this load is estimated to produce less than 7.0 psi surface compressive stress and less than 13.0 psi tensile strees in the element interior. '
i REFERENCES
- 1. R. J. Price, " Cyclic Fatigue of Near Isotropic Graphite: Influence ;
of Stress Cycle and Neutron Irradiation", CA-A14588, November 1977.
- 2. L. Sevier, "LHTGR Graphite Fuel Element Seismic Strength", GA-A13920, April 30,1976, p. 3 and pp. 42-46.
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'i 4' 52 RESPONSE TO NRC QUESTION 400.1 ON H-451 CRAPHITE QUESTION 400.1 Provide a discussion of H-451's response to the design basis earthquake in comparison to that for H-327. The discussion ,
should parallel other safety analyses given in GLP-5588 and demonstrate that the structural damage potential for H-451 is less than or equal to that for H-327.
b
RESPONSE
(,
As stated in Section 2.4.3 of the H-451 SAR, the difference in maximum radial shrinkages of H-327 and H-451 is less than 0.1%. Therefore, the overall gap distribution within the core will not change significantly, and H-451 dimensional change is not expected to affect the core seismic behavior.
H-451 fuel elements would be expected to see increased seismic loading vs. H-327 due to their higher radial modulus. Using simple spring-mass relations, this loading can be determined to increase as the square root of the ratio of. the moduli. Thus, using the modulus data in Tables 4-3 and 4-9, H-451 fuci element impact force is expected to be 30% greater r' than tha't of the H-327 elements. Because H-451 fuel elements have more than 60% higher radial strength, the structural damage potential for H-451 is less than that for H-327 fuel elements.
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53 RESPONSE TO NRC OUESTION 400.2 ON H-451 GRAPHITE QUESTION: ,
400.2 Confirm your plan to irradiate test elements of H-451 graphite in the Fort St. Vrain reactor and later perform post irradiation ,
examinations on these elements. Describe how these examinations i will be used to verify that the irradiated mechanical properties of H-451 remain superior or equal to rhe properties of H-327.
Also, will the examinations permit verification of the irradiated mechanical properties of H-451 used in the structural analysis.
(' i
RESPONSE
Public Service Company of Colorado (PSC) has submitted to NRC a J request for a technical specification change to permit installation of eight fuel test elements into the FSV reactor core at the first refuel-ing. The safety evaluation for these test elements submitted to NRC by PSC is contained in the General Atomic Report, " Safety Analysis Report for Fort St. Vrain Reload 1 Test Elements FTE-1 through FTE-8", GLP-5494, June 30, 1977, including Amendments 1 and 2. As discussed in that report, ,
the structural graphite of all eight FTEs is type H-451.
/ Amendment 2 of the safety analysis report describes the proposed i Department of Energy (DOE) funded post-irradiation examination (PIE) work-scope for the test elements. Amendment 2 was submitted to NRC by PSC via correspondence P-78102, June 20, 1978. As noted in that transmittal, any changes in DOE funding would require changes in the described PIE program. As presently conceived, the PIE program would include non- ,
j destructive and destructive strain, stress, and strength examination of [
. I the H-451 structural graphite in three of the test elements. However, the principal data on the' effects of irr*diation on the properties of H-451 graphite will be derived from the OG series of irradiation capsules !
in the Oak Ridge Research Reactor.
The test elements described in CLP-5494 are currently on-site at FSV. They will be placed in the reactor core at the first refueling if ,
the following conditions are net:
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- 54
- a. NRC approves the technical specification change requested by PSC.
i
- b. Insertion of the test elements continues to be agreeable to PSC.
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