ML20197C997
| ML20197C997 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 11/17/1978 |
| From: | Wessman G EMVGA, GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER |
| To: | Gammill W Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7811210276 | |
| Download: ML20197C997 (21) | |
Text
so- 23o 7 failMSitETJSERfMMMMMCMHiMMD355LNWa5EtWC'JPA'UEATf3EgiENERAL ATOMIC 2EEELMWR2iFIALTJ2KNEM7&EIEWWlE"EE0Cil22M222WGIEntm CENE AL MIC COMPANY SAN DIEGO, CALIFORNIA 92138 (714)455 3000 November 17, 1978 Mr. William Gammill Assistant Director for Adva.:ced Reactors Division of Project Management U. S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Gammill,
In accordance with my letters of October 25 and October 30, 1978, we are enclosing 50 copies of the second part of our response to your Request for Additional Information on H-451 graphite, dated September 19, 1978.
This submittal completes our response to your request.
If you have any questions regarding these responses, please do not hesitate to let us know.
Sincerely, t
M1224%/
G. L. Wessman, Director Plant Licensing Division
/
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STATE OF CALIFORNIA ss.
COUNTY OF SAN DIEGO )
After being duly sworn, the person known to me to be G. L. Wessman of General Atomic Company, signed the within document this 17 day of November 1978.
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, a 1 RESPONSE TO NRC QUESTION 231.2 ON H-451 GRAPHITE QUESTION:
231.2 In the discussion of elastic modulus, the report indicates that the axial modulus of H-451 graphite is " typically" lower than that of H-327 graphite and that, therefore, the axial stresses in H-451 graphite elements are lower than those in H-327 graphite elements for a given strain. It is then stated that although the radial modulus of H-451 graphire is higher than that of H-327 its effect on H-451 stress compensated for by the higher radial strength of H-451.
(a) The inclusion of the work " typically" in the report state-ment on axial modulus implies that there are instances where H-327 has a lower axial modulus than.H-451. If this is so, is there some uncertainty associated with the axial modulus value? What is the source of this uncertainty? How is this factored into the design?
(b) Although the lower H-451 modulus results in a lower stress for a given strain, the reverse must also be true; that is, for a given stress, there will be a greater strain. How is this factored into the design for steady state and transient conditions?
(c) Please provide some quantification of the effect of radial modulus on stress. Show, by means of a " worst case" example, how the higher radial modulus of H-451 graphite is compensated by higher radial strength.
RESPONSE TO 231.2a Inspection of Tables 4.3 and 4.9 of GLP-5588 show that in the unirradiated condition, the values of the axial elastic modulus for the midlength and center position of the graphite log are 10.3 x 106 kPa (1.5 x 106 psi) and 7.93 x 106 kPa (1.15 x 106 psi) for H-327 and H-451 graphite, respectively. The changes in these elastic modull as a function of temperature and dose are shown in Figure 1. Although the H-451 modulus, generally increases at a faster rate than that of H-327 (see Figure 1), its initially lower value ensures that it " typically" will have a lower value at most temperatures and fluences.
The exception to this occurs at temperatures of 900 C and fast flux ex-posures greater than 6.5 x 10 21 n/cm . At these conditinns, the irradiated moduli of these two graphites are equal. The irradiated modulus is the pro-
-.. .- - ~ . , ~ . - . - . .,
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i duct of the unirradiated value given above and the increase due-to irradiation l taken from Figure.1. At.the peak FSV fluence of 8 x 10 21 n/cm and 900 C, the H-451 axial modulus is approximately 12% greater than that of H-327. Since only a small= fraction of the graphite in the core experiences these conditions, the phrase " typically lower than that of H-327 graphite" was used to describe this situation.
The standard deviation of the unirradiated H-451 modulus is less than 4%
of the mean'value. For irradiated graphite, 95%Lof the data are in the' range of -10% to.+30% of the design values. This uncertainty is the result of un-certainties in measurement and .to.a lesser extent, matertal variability. -As I shown in the response to Question-231.6, the impact of this uncertainty upon design margin is not large.
RESPONSE TO 231.2b The total graphite strain computed in the design stress analyses is a' summation of the creep strain, the thermal strain, the irradiation induced strain, and the elastic strain. The strain resulting from a given stress, '
the elastic strain, is accounted for 1r the determination of total strain.
Young's Modulus of H-451 in the axial direction varies between 1.15 x 106 and 3.1 x 106 psi with irradiation. Using a mean graphite strength of 1970 psi, the contribution of elastic strain will be approximately an order of mag-nitude less than the irradiation induced strain shown in Figures 4-1 and 4-2.
RESPONSE TO 231.2c The results of the stress analyses for normal operating and shutdown conditions have been presented in Table 2-1 of.GLP-5588. The effects of radial moduli are.shown for both H-327 and H-451 graphite. Although the higher H-451 modulus results in higher H-451 stress, the higher radial strength of H-451 provides a greater design margin; 5,757 kPa (835 psi) and 9,480 kPa (1,375 psi) ,
for H-327 and H-451, respectively.
As discussed in the response to Question 231.8, the off-normal case that creates the highest graphite stresses is the maximum worth' rod withdrawal accident. The highest stresses for this transient occur in unirradiated graphite at the beginning of life.
,, . 3 For the H-327 fuel elements, the maximum radial stress during this worst case accident is 1689 kPa (245 psi). This leaves a H-327 design cargin of ,
4792 kPa (695 psi).
The H-451 elements have higher radial stresses due to their higher radial '
modulus. The H-451 caximum radial stress is 2799 kPa (406 psi) with a design targin of 7956 kPa (1154 psi). Thus, the higher H-451 stress is compensated for by the higher radial strength in this worst off-normal case.
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Ficure 1 Chanac in clastic :"odalus zith neutron fluence at temperature 9
- Peak fast neutron fluence in FSV = 8x10'1 n/cm2 .
Median fast neutron fluence in FSV = 4,5x10 21 n/cm .
. .- 5 RESPONSE TO NRC QUESTION 231.4 ON H-451 GRAPHITE
-QUESTION 231.4 (a) Although the report makes the rather sweeping statement that irradiation-induced strains make the major contributions
.to the operating stresses within the graphite elements, whereas ,
thermal strains contribute strongly to the shutdown stresses, !
it is not self-evident why this is so. Please provide-some dis- !
cussion of this area.
(b) If the-thermal expansion of H-451 graphite is about 30%
higher than that.of H-327 in the axial direction, and if thermal !
strains contribute strongly to the shutdown stresses as claimed,- i then why-is the maximum calculated shutdown tensile stress in the axial' direction lower for H-451 than for H-327 (139 vs 237 psi, respectively)?
I RESPONSE TO 231.4 (a) j The initial secondary operating stresses in the unirradiated f fuel element are due solely to thermal expansion and are compressive in the hotter portions of the element and tensile in the cooler portions. During operation, irradiation causes the hotter graphite to shrink faster than the colder graphite. This irradiation-induced shrinkage can overcome the smaller j magnitude thermal expansion, and the colder portion of the graphite which was originally in tension goes into compression while the hotter portion goes into tension. In the portion of graphite now in tension, the thermal expansion serves to reduce the net shrinkage. Upon shutdown, the block returns to an isothermal condition, and the greatest decrease in temperature (and therefore the greatest decrease in thermal expansion) occurs in the previously hotter graphite.
Thus, at the time of maximum operating stress, thermal expansion is genecally both maximum and of opposite sign to the irradiation strains in this area of maximum tensile stress. During-shutdown, the removal of.
.the major portion of these thermal expansion strains (i.e. , the relative thermal contraction) produces a major change in stress state. Their re-moval allows the contribution from the irradiation strain to predominate so that an increase from the operating tensile stress is observed. '
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As an example, details of the: finite' element location giving the H-451-element maximum axial' shutdown stress'(958 kPa (139 ps'i)) in Table-2-1 GLP-5588 are presented below:
" thermal romroom\
temp. _ , thermal +
\ j
- irradiation
- irradiation . Stress *
. Initial: Thermal Operating. Strains +4.72 x 10-3 0.0 ~+4.72 x 10-3 -64 psi
~0perating Strains Before' Shutdown +4.08 x 10-3 -3.66 x 10-3 +0.42 x 10-3 49 psi Shutdown Strains (0.340 F)~ +4'.96 x 10-4 -3.66 x 10 -3 --3.16 x 10-3 139 psi i
.-RESPONSE TO 231.4 (b)
The thermal expansion'of H-451 graphite is larger than H-327 in both the axial and radial directions. Comparisons of Tables 4-4 and 4-10 GLP-5588 show that.the thermal expansion of H-451 graphite.is up to 50%
and 200% higher than that of H-327 graphite in the radial and axial direc-tions, respectively, for temperatures of interest. Therefore, the H-451 ;
thermal strain is a larger component of both the operating and shutdown total strains than in the case of H-327 graphite. However, it is the total strain 4 that determines the stress. Comparisons of Figures 4-1 and 4-4 (GLP-5588) show that the irradiation strain of H-327 can be up to a factor of two greater j than that of H-451. It is the contribution from this irradiation strain that produces the greatest total strain, and the'efore stress, in the H-327 l graphite. !
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- Calculation of graphite stress requires knowledge of the stra'in distribu-tion in the total web and cannot be done from the hmited strain infc -
tion presented in this table. Stress values are given only to demc .cate relative increases in stress at shutdown.
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7 For the specific example cited in Question 231.4 (b), the strains for H-451 have been given previously-in the response to Question 231.4 (a). For the same finite element location and conditions, the H-327 strains are:
thermal [romroom temp.
thermal +
\ / irradiation irradiation Stress
- Initial Thermal -3 Operating Strains +1.76 x 10 0.0 +1.76 x 10-3 -56 psi Operating Strains Before Shutdown +1.76 x 10-3 -4.63 x 10-3 -2.87 x 10-3 153 psi Shutdown Strains
(@ 340 F) +9.11 x 10-5 -4.63 x 10 -3 -4,54 x 10 -3 237 psi i
Determined from differential strains across the fuel element web.
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,- .. 3 l RESPONSE'TO'NRC QUESTION 231.5 ON H-451 GRAP'HITE ~
OVESTION: ,
231.5 (a) Clarification is.needed with respect to the relationship of ;
the H-451 fuel! elements ~to the current oscillation problem in .
the FSV core. Provide
- 1) the' anticipated loading schedule for H-451 fuel elements;
- 2) in relation to this schedule, the point at,'or.before which, corrective measures to prevent oscillations will have'been completed.
(b) Unless there is a firm commitment that the corrective measures.
will' positively stop any potential fluctuations over the full' :
power. range, and assuming that inadvertent fluctuations.may still ,
be possible, for the transition . core' and the ~ full H-451 core provide- !
- 1. an assessment of the potential for component damage, owing to gap flow induced. motion, taking into account the differ-ences in irradiation induced shrinkage and thermal expan-sion of H-327 and H-451 graphites and the corresponding changes in gap size, and
- 2. an assessment of the effects on the magnitude of temperature fluctuations and other fluctuations characteristics that may occur with the substitution of H-451 for H-327 graphite.
RESPONSE TO 231.5 (a)
The earliest time at which H-451 graphite fuel elements (other than test elements) could be loaded into the Fort St. Vrain reactor' core is late in
~1980 or early in 1981. Public Service Company of Colorado and General Atomic Company are engaged in a major effort to determine the cause of the fluctuations
~
and to institute corrective measures. However, until the results of further testing have been analyzed, .it' is not possible to define a date' by which' corrective measure will have been completed.
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1 RESPONSE TO 231.5 (b.1)
Determination of fuel element loads during fluctuations has been explained in detail in the Core Fluctuation Investigation Status and Safety I
.. .. 9 ,
I l
Evaluation-Report transmitted via letter P-7E137 from J. K. Fuller (PSC)- 1 WilliamGammill'(NRC), August 11, 1978.
Because the maximum' radial shrinkages of the'H-327 and H-451 fuel elements are' calculated-to. differ by less than 0.1% (pg. 2-7, GLP-5588), the characteristics of the fluctuations should be unaffected by the change in graphite type.
To determine the fuel element loads during fluctuations for.H-451 graphite, the H-327 element loads given in the . fluctuation SER were increased ,
by the square root of the ratio of the Young's moduli. The H-451 graphite element loads were calculated for various impact orientations to be:
Flat face 2600 lb f
. Top and bottom edge 1360 lb f Side edge 500 lb f Dowel loads 205 lb f These loads. represent a 24% increase over the fluctuation loads calculated with H-327 elements and result from the higher modulus of H-451 graphite. Because H-451 graphite has a 66% greater radial strength than H-327, its use reducas the likelihood of damage due to fluctuation induced loadings. .l RESPONSE TO 231.5 (b.2) ;
Given the extent of current understanding of the core fluctuations, ;
it is not believed that the substitution of H-451 graphite will have any measurable j effect on the characteristics of core fluctuations. l 1
. 10 l l
1 RESPONSE TO KRC QUESTION 231.6 ON H-451 GRAPHITE QUESTION:
231.6 The graphite mechanical properties that are identified in the report as the ones that determine the element stresses, stress-strength
=argins, and element deformations c ' e (1) modulus of elasticity, (2) tensile strength, (3) creep properties, (4) irradiation-induced -
dimensional changes, (5) thermal expansion, and (6) thermal con-ductivity. Please illustrate the relative importance of these properties by tabulating the relative effect (as fraction of the total) that each of these properties has on element stresses, stress-strength =argins, and ele =ent deformation (even if these values can only be estimated), along with a meervra of the uncer-tainty for each nu=ber. Tables should ha provided for both normal and of f-normal conditions. Please previde a short description also of the types of parametric studies thac were cade in establishing the magnitude of these effects. Indicate where further work is on-going or planned (including general schedule), and where further information on future work is reported.
RESPONS E:
The effect of caterial property variations on stress, design margin, and dimensional change is quite complex. Because the effect of any specific varia-tion will be dependent on the interaction with the various other material pro-perties, the magnitude of the resulting change will dif fer both as a function of the stress state at any specific point in the fuel element and as a function of time. Thus, there are nearly an infinite variety of answere to this question.
For these reasons, Question 231.6 has been answered in terms of the effect of caterial property changes at the time of maxicum stress, for both operating and l shutdown conditions, and at the time of maximum dimensional change.
l This method does not necessarily present the greatest percent change in the resulting stress, design margin, or dimensional change, but does identify the magnitude of change at the time of mini =um design targin. For example, the
! caximum operating stress occurs at zero days and results only from initial startup thermal stresses. At this time there are no creep effects, and any change in creep parameters, therefore, has zero effect on stress and stress i
. a.- - . .. ~ - - . . . . . . e- .. .. .- - . . __ . . . .
.. . 11 margin. Towards-the end of life of the element, a +20% variation in the steady state creep coefficient can result in +38%-increase in operating stress. However, the operating stress at this time is.sufficiently low that even this 38% increase
. produces a. stress of only one quarter the stress during initial startup.
The accompanying tables, Tables 1 and 2, demonstrate the effect-that varia-tions in the measured material properties have upon the maximum stresses, stress margins, and dimensional changes for aormal and off-normal conditions, respec-tively. These parameter studies were made using the FESIC computer code to analyze the H-451 fuel element with the maximum stress while varying the material properties one at a time. Table 2 is representative of the worst case off-normal condition, a maximum worth rod withdrawal accident. This case is discussed in detail in the response to Question 231.8. Since this worst case produces the greatest stress in an.unirradiated element due to.the lack of alleviatinh
~ '
irradiation shrinkage, creep and 1rradiation dimensional changes are not con-sidered in this studylof parameter variation.
General Atordic's efforts are currently concentrated on developing new -
stress codes with' improved efficiency and capability. All efforts to deter-mine the effects of variations in the material property parameters will be directed toward these new codes. No further work is planned with FESIC of SAFE GRAFIT.
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TABl.E 1 11/16/78 Relative Ef fect of Material Property Vaa sations on Fuel Element Stresses Stress-St rength Hargins, and Element Dimensional Changes (for normal operating and shutdown conditions)*
ANALYTICAL ASSUMPTIONS AND RESULTS
% Change That Results % ChanEe That Results
% Change in Maximum Operating Stress ** In Maximum Shutdown Stress Assumed in and Miniatun Design Margin and Minimum Design Margin % Change Approximate Variation Material Design Design in Maximum Haterial Property in Heasured Data Property Stress Margin Stress Margin Dimensional Change
- 1. FLxhalus of a 5 5% mean +5% +5% -0.2% +3.5% -0.25% 0%
Elasticity
- 2. Tensile o dev.
5 11% mean (axial) +20% 0% +21% 0% +21% 0%
S"""'h o dev.
5 19% mean (radial) -20% 0% -21% 0% -21% 0%
- 3. Creep (steady upper and lower 90% +20% 0% 0% -5% +0.3% 0% -
state creep confidence on the -20% 0% 0% +7% -0.5% 0%
coefficient) mean is s 122%
4, Irradiation o s 15% mean +15% 0% 0% +5% -0.3% +15%
induced -15% 0% 0% -5% +0.3% -15%
Dimensional Change
- 5. Thermal o s 5% mean +5% +5% -0.2% +3.5% -0.25% -0.25%
Expansion -5% -5% +0.2% -3.5% +0.25% +0.2%
(max. dimensional i change is negative)
- 6. Thermal o 5 10% mean -10% +11% -0.5% +11% -0.8% +0.4%
Conductivity
- The effects of parameter variations are analyzed at the time of aluimum design margin.
- Nxis.um operating st ress occurs during startup (unirradiated graphite) due to theratal strains. Therefore, creep and irradiation induced strain are zero.
H b)
. . _ _ + _ _ . . . .
TABLE 2 11/16/78 Relative Effect of Material Property Variations of Fuel Element Stresses, Stress-Strength Margins, and Element De f orma tions (for worst case off-normal condition - maximum worth rod withdrawal for unirradiated elements)
ANALYTICAI. ASSUMPTIONS AND RESUI.TS
% Change
% Change % Change in in Maximum Assumed in Accident Conditions
- Dimensional Approximate Material Design Change Material Property Variation in Measured Data Property Stress Margin (at time = 0)
- 1. Modulus Elast.i city dm 5 5% mean +5% +5% -1.2% 0%
- 2. Tn:nsile Strength a < +20% 0% +25% 0%
dev. - 11% mean (axial) dev. s 9 mean (r dial) -20% 0% -25% 0%
- 3. Creep Upper and lower 90% confidence
- creep is 0 for initial life (time = 0) on the mean is 5 122%
- 4. Irradiation Induced a s 15% mean
- irradiation induced dimensic- kunge Dimensional Change is O for initial life (time = n
- 5. Thermal Expansion a < 5% mean +5% +5% -1.2% +5%
-5% -5% +1.2% -SI
- 6. Thermal Conductivity o
-< 10% mean -10% +11% -2.6% 0.5%
dev.
ORWA is analyzed at the beginning of .cl element life for determination of maximum stresses.
14 RESPONSE TO NRC QUESTION 231.8 ON H-451 GRAPHITE QUESTION:
231.8 The report indicates that " initial core operating conditions" were used in calculating maximum stresses and deformations for the H-327 and H-451 graphite elements. Please explain what is meant by the phrase, " initial core operating condition". Are these " worst case" conditions ? If they cover only normal operation and not transient or accident conditions, please include analytical results for the worst case off-normal condition. Please show what effect, if any.,
is had from assuming a mixed H-327/H-451 core in place of a complete H-451 core.
RESPONSE
The graphite fuel element stresses and deformations expected during normal reactor operation and presented in Table 2-1 of GLP-5588 were calculated using power distributions and fast flux values appropriate to segments 1 through 6.
These six segments constitute the initial core. The projected operating histor-les used in these analyses cover the first six cycles of operation (Segment 6 remains in the core through these six cycles) and assume full power normal operation. These " initial core operating conditions" were used for comparison of H-327 and H-451 graphite element stresses under similar conditions.
The off-normal case that produces the maximum graphite stress is the maxi-mum worth rod withdrawal accident. This accident produces the greatest temper-l ature differences across the web between fuel and coolant holes and, therefore, l
results in the greatest thermally induced stress. As stated in the FSV FSAR Sec tion 14. 2. 2. 6, the worst conditions for the reference design accident (with-drawal of 0.012 Ak rod worth at end of equilibrium cycle) assume the following accident sequence.
l l 1. Malfunction of rod withdrawal prohibit (FSV FSAR Section 14.2.2.1).
- 2. Malfunction of automatic scram at 140% power.
l 3. Lack of manual operator action.
- 4. For the maximum worth rod at end of cycle, 0.012 Ak, the scram signal on reheat steam temperature of 1075 F occurs 105 seconds after accident initiation.
l
15 These conditions were assumed for the worst case off-normal graphite stress analysis.
During the rod withdrawal accident, the local power density rises, and the temperature gradients across the graphite webs increase. At the time of the peak gradient, the temperature differential will be up to six times the normal operating value.
The maximum stresses frame rod withdrawal accident would occur only during the initial life of the graphite fuel element. If the element has undergone more than approximately 50 days of irradiation, the hotter portions of the fuel element web will have gone into tension because the hotter portions shrink more than the cooler portions under 1rradiation. During the transient, the hotter portion of the web has a greater thermal expansion; therefore, the transient would temporarily reduce the value of the peak tensile stress.
The stress analysis conducted for the rod withdrawal accident conserva-tively assumes the combination of an unirradiated element with the maximum worth rod at end of cycle, 0.012 ok, and a scram signal at reheat steam tem-perature of 1075 F, 105 seconds after accident initiation.
During this accident, H-451 fuel elements would experience maximum thermal stresses of 2875 kPa (417 psi) in the axial direction and 2799 kPa (406 psi) in the radial orientation. The design margins for the axial and radial directions are 10707 kPa (1553 psi) and 7956 kPa (1154 psi) for the axial and radial directions, respectively. The H-451 fuel elements retain considerable design margin even in this worst case sf*uation. The correspond-ing design margins for H-327 graphite are 7100 kPa (1030 psi) in the axial direction and 4792 kPa (695 psi) in the radial direction.
The transition to H-451 graphite fuel Alements will take place on a region-by-region basis . It is not planned to mix H-451 and H-327 fuel elements in the same layer within a region during the transition. As stated in Section 2.4.3 of GLP-5588 and further shown in the dimensional change curves (Figures 4-1 and 4-4), H-451 graphite will experience less axial dimensional change than H-327. Therefore, the maximum change in the height of the fuel columns will be reduced, and the relative shrinkages between regions will remain
I i
16 I within the original design envelope. The maxi =u::1 radial shrinkage of the H-451 fuel elements is only 0.1% greater than the H-327 elements, so there will be no significant change in the maxi =um radial gaps in the core. As f urther discussed in the response to Question 231.13, no interaction will occur between dowels and sockets between the vertically adj acent H-451 replace-able relectors and H-327 permanent reflectors.
' ' l 17 ,
RESPONSE TO NRC QUESTION 231.13 ON H-451 GRAPHITE QUESTION:
231.13 (a) In the discussion of the permanent loss of forced cooling event
[ design basis accident No. 1 (DBA-1)l, the report text (p. 3-5) implies that there is little concern about the maximum thermal con-pressive load in the fuel blocks and reflectors because these loads are small compared to the graphite axial (primary, or dead-weight) compressive stress. This seems to contradict the statement in GASSAR (p. 4. 2-18B) that the secondary (thermal) stresses are by far the largest components of the total stress in the elements.
The apparent contradiction may stem from a confusion over terminology or working. Please clarify.
(b) Since the thermal expansivity of H-451 graphite is greater than that of H-327, substitution of H-451 fuel and reflector blocks would appear to increase the thermal stresses in the dowel pins and other points of contact, particularly where dissimilar types of graphite are in contact (for example, where the bottom reflectors meet the core support. blocks). Please discuss the potential for increased thermal stresses of this nature, expecially those that might be gen-erated during a DBA and subsequent shutdown.
RESPONSE TO 231.13a The apparent contradiction between the statements concerning the DBA-1 event on p. 3-5 of GLP-5588 and the statements in GASSAR arises from an error in the eighth line in the last paragraph on page 3-5. The corrected sentence should read: "These loads are small compared with the graphite axial strength."
! Comparing the maximum primary compressive stress (140 kPa, or 20 psi) in the fuel elements at the bottom of the core to the peak axial shutdown stresses given in Table 2-1 of GLP-5588 shows that the secondary stresses are indeed the largest components of the stress in the element. This agrees with the statement in GASSAR (p. 4.2-18B) .
RESPONSE TO 231.13b The dowel / socket system in FSV fuel and reflector elements is dimensioned so that proper engagement of the dowel in the socket is assured even if an irradiated element having the maximum shrinkage is placed on top of or under-neath an unirradiated element. To confirm the validity of this statement, a':
core locations where H-327 elements could contact H-451 elements, a calcula-
18-l tion of the minimum dowel / socket clearance was performed for the case where a fresh, unirradiated H-451' element sits atop an irradiated H-327 element. The radial shrinkage of the H-327 element was taken as 0.8%, the value given in FSAR Section 3.4.2.1.2, since this is conservative with respect to the maximum calculated value of 0.56% presented in Section 2.4.3 of GLP-5588. The differ-
'ence between the thermal ' strains of H-327 and H-451 at 1600 C (a temperature
-3 well above that expected during a DBA) of 2.2x10 1nch/ inch was also used in the calculations as well as the worst combination of dimensional' tolerances.
Under these assumptions the minimum possible dowel / socket clearance was cal-culated to be 0.02 inches, a value that is considered ample given the conser-vatism of the ' computations. Therefore, the use of dissimilar types of graphite in the core will not-increase the potential for contact or interference stresses.
It should also be noted that the bottom reflector elements, which are ;
made of H-327 graphite and rest upon the core support blocks, are designed for the life of the plant. There are currently no plans or. proposals to change these transition elements to H-451 graphite.
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RESPONSE TO NRC QUESTION 231.15 ON H-451 GRAPHITE l QUESTION:
231.15 Many of the graphite mechanical properties reported in Section 4.1 of GLP-5588 appear to be for unirradiated graphite at room tempera '
ture. Because the strength of graphite tends to increase with tem-perature and irradiation to a maximum before declining, we assume that these particular property values were chosen with the intention of being conservative, but this is not so stated in the report.
Please discuss why only unirradiated, room temperature values are provided for ultimate tensile strength, ultimate compressive strength, elastic modulus and Poisson's ratio, and discuss briefly the ef fects of irradiation and temperature on these properties. If these pro-perties have even been measured ac a function of irradiation and temperature, please reference the data.
RESP 0NSE Both the modulus of elasticity and the strength of graphite are increased by fast neutron irradiation. The changes in the modulus of elasticity as a function of fluence and temperature have been presented in Figure 1 of the response to question 231.2. It can be seen that increases of 170% are possible under extreme conditions. However,' analysis has shown that under'the most extreme conditions of temperature and fluence experienced by FSV fuel elements, the use of unirradiated properties provides a conservative calculation of the design margin because at the low stresses calculated for the fuel elements in FSV, the increase in graphite strength with irradiation compensates for the increase in the modulus of elasticity. Since the use of unirradiated pro-perties is conservative and permits some simplification in the computer codes, unirradiated, room temperature values of elastic modulus, ultimate strengths, and Poisson's ratio are used in the fuel element structural analysis.
The effects of irradiation and tamperature on the modulus of elasticity have been presented in the response to Question 231.2 (see Figure 1). These curves were obtained from data in References 1 and 3. The tensile and com-pressive strengths of unirradiated and irradiated graphite, S o and S i , respec-tively, are related to the corresponding elastic moduli (Eo and E i ) by the equation:
1
. _. ~ _ _. _
l a 6 20 Si IE t n So (Eo j i
where n = 0.40 for H-327 in the radial orientation i n = 0.67 for H-327 in the axial orientation n = 0.64 for H-451 in the axial or radial orientation These values of the exponent n have been obtained from correlations of experi-mental data (Refs. 2, 3).
The effect of irradiation on Poisson's ratio has been measured in several experiments. These measurements have been discussed and su==arized in Section 3.4 of Reference 4. It was concluded that those experiments showed no systematic effect of irradiation on Poisson's ratio. Recent measurements of Poisson's ratio for H-451 graphite irradiated at 900 C (Ref. 5) have shown that Poisson's ratio is actually constant for creep strains less than 0.6%. Since the maximum creep strain for the H-451 elements in the FS.V core does not exceed 0.52%, it is appropriate to use a constant value of Poisson's ratio at all temperatures and fluences considered in the stress calculations.
REFERENCES
- 1. Everett, M. R., et. al., " Irradiation Perfermance and Selection of Graphites for HTGRs," Dragon Repcrt DP-877, April 1974
- 2. Beavan, L. A., " Irradiation Strain in H-327 Graphite," General Atomic Report GA-Cl2509, April 1973.
- 3. Price, R. J., and L. A. Beavan, " Final Report on Graphite Irradiation Test OG-3," General Atomic Report GA-A14211, January 1977.
4 Price, R. J., " Mechanical Properties of Graphite for High-Temperature Gas-Cooled Reactors: A Review," General Atomic Report GA-A13524, September 1975.
- 5. Kennedy, C. R., W. H. Cook, and W. P. Eacherly, "Results of Irradiation Creep Testing Graphite at 900 C." 13th Biennial Conference on Carbon, Irvine, CA, July 1977.
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