ML20053A626
| ML20053A626 | |
| Person / Time | |
|---|---|
| Site: | 07002949 |
| Issue date: | 04/23/1982 |
| From: | Andognini G ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| To: | Ketzlach N NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| Shared Package | |
| ML20053A624 | List: |
| References | |
| NUDOCS 8205270027 | |
| Download: ML20053A626 (15) | |
Text
.
NOS82-326
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April 23, 1982 c, u n t,.m.., v. n.
E L f f Y stsC opt st a f s ONS Director of Nuclear Materi:11 Safety and Safeguards U.
S. Nuclear Regulatory Commission Wash in gton,
D. C.
20555 Attn:
Mr. Norman Ketzlach Subj:
Special Nuclear Material License Response to NRC Questions for Palo Verde Nuclear Generating Station Docket No. 70-2949 (SNM 1887)
Attached please find Combustion Engineering's responses to NRC ques-tions 10 and 11 as requested in NRC Ictter to G. C. Andognini dated January 8, 1932.
Also attached are amended pages to our submittal dated February 17, 1982, in support of the Arizona Public Service C
Company application for authorization to receive, store, inspect and transport Special Nuclear Material.
The criticality analysis per-formed by Combustion Engineering on the dry storage of new fuel in the spent fuel pool will be transmitted to you upon our receipt of this information.
Should you have any questions or require additional information, l
please contact Mr. Steven R. Frost at (602) 271-3348.
l l
Sincerely, fWMO CCA/.1Rp/jc
(
Attachments l
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Director, Region V, USNRC NRC I:esident Inspector - PVNGS NRC Project Manager - E.
1.i c i t ra
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l 82052706 M
- Attachment NOS82-326
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Page 5 amended April 23, 1982 3.
Page 20-21, Section 1.2.1 c.
Describe the maximum number of fuel assemblics in the inspection area (e.g., maximum number of assemblies at an inspection s tation, minimum edge-to-edge distance between individual as-semblies and between the inspection area and the fuel storage area).
PVNGS Response No more than one new fuel shipping container (two fuel assemblies) will be in the new fuel inspection station area (one inspection
[
station per unit) at any time and no more than one fuel assembly will be out of a shipping container at a time in the new fuel inspection area. Procedures will require main-taining at least 6 inches edge-to-edge separation between fuel as-semblies (the edge to edge dintance provided in the C-E shipping container).
The minimum possible edge-to-edge distance between a fuel assembly in the new fuel inspection area and one in the new fuel storage area is 48 inches.
m
- Attachment NOS82-326
(
Page 17 amended April 23, 1982 9.
Page 25, Section 2.2.1 Confirm the array in the Fuel Building shipping container laydown a.
area is no larger than that in a single shipment or specify the criteria for larger storage arrays. If larger arrays are to be used, provide justirication for the larger storage arrays.
b.
Identify the shipping container used (e.g., NRC Certificate of Compliance No.) and specify the numSer of fuel assemblies per container.
PVNGS Response a.
See Response to Item 3a and 3b.
b.
The shipping containers to be used under NRC Certificate of Compliance No. 6078, Revision No. 7, are Combustion Engineering Models Nos. 927Al and 927C1 shipping packages.
The steel fuel bundle shipping containers consist of a strongback and fuel bundle clamping assembly, shock mounted to a steel outer C
container. A minimum 1/4" thick, 6" x 6" x 8" high steel sepa-rators are bolted between fuel bundles. The Model No. 927Al container is approximately 43" in diameter by 189" long with an approximate gross weight of 6,200 lbs.
The Model No. 927C1 container is approximately 43" in diameter by 216" long with an approximate gross weight of 7,000 lbs.
Each shipping container carries two (2) fuel bundles, and the fuel is usually shipped in eight (8) containers per shipload.
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Attachment PVNGS FSAR 1;0S82-326
'Apri,1 23, 1982 (c
9.
AUXILIARY SYSTEMS 9.1 FUEL STORAGE AND HANDLING 9.1.1 NEW FUEL STORAGE 9.1.1.1 Design Bases The following design bases are imposed on the storage of new fuel:
A.
Accidental criticality shall be prevented for the most reactive arrangement of new fuel stored, with optimum moderation, by assuring that K is less eff than 0.98.
This design basis shall be met under any normal or accident conditions.
B.
The requirements of Regulatory Guide 1.13 shall be met.
C.
The storage racks and facilities shall be qualified as Seismic Category I.
D.
Storage shall be provided for at least one-third i
core of new fuel.
9.1.1.2 Facilities Description The rack assemblies are made up of individual racks similar to those shown in figure 9.1-1.
A minimum edge-to-edge spacing between fuel assemblies, as required by section 9.1.1.3.1, is maintained between assemblies in adjacent rows.
These spacings are the minimum values after allowances are made for rack fab-l rication tolerances and the predicted deflections resulting from postulated accident conditions,. discussed in sec-tion 9.1.1.3.1.
The specific location af the new fuel racks in the fuel build-ing is shown in figures 1.2-6, 1.2-7, 1.2-12, and 9.1-2.
9.1-1
PVNGS FSAR FUEL STORAGE AND HANDLING r
The stainless steel construction of the storage racks is compatible with water and zirconium clac fuel.
The top structure of the racks is designed such that there is no opening between adjacent fuel cavities that is as large as the cross-section of the fuel bundle.
In addition, the outer structure of the racks precludes the inadvertent placement of a bundle against the rack closer than the prescribed edge-to-edge spacing.
9.1.1.3 Safety Evaluation The new fuel storage rack design and location, discussed in section 9.1.1.2, ensures that the design bases of sec-tion 9.1.1.1 are met.
The capability of PVNGS new fuel storage is described below.
9.1.1.3.1 Criticality Safety
/'
(
The following postulated accidents were considered in the design of the new fuel storage racks.
A.
Flooding; complete immersion of the entire storage array in pure, unborated, room temperature water.
B.
Envelopment of the entire array in a uniform density aqueous foam or mist of optimum de:rity that maximizes the reactivity of the finite array as described in 8
item D of the criticality safety assumptions.
It is postulated that these conditions could be present as a result of fire fighting.
C.
A fuel assembly dropped from a height of 2 feet onto the rack which then falls horizontally across the top of the rack.
D.
Tensile load on the rack of 5000 pounds (limited by adjustment o' the motor stall torque or load limiting device of the crane used to load fuel into the racks.)
(
Amendment 8 9.1-2 March 1982
'NOS82-326 Aprri 23, 1982 FUEL STORAGE AND HANDLING Although the above accident conditions have been postulated, the fuel handling equipment, new fuel racks, and the building arrangement are designed to minimize the possibility of these accidents or the effects resulting from these accidents by:
A.
Providing positive hoist travel limits and ir.terlocks.
to ensure proper equipment operation and sequencing.
B.
Limiting the crane loads when installing fuel into or removing fuel from the fuel rack.
C.
Designing the new fuel racks for SSE conditions and dropped fuel bundle conditions.
D.
Maintaining K less than 0.95 in the event eff the fuel area becomes flooded.
E.
Designing the new fuel handling crane to preclude the new fuel handling crane, or any part thereof, from falling into the new fuel handling area.
The following assumptions are made in evaluating criticality safety.
A.
Under postulated conditions of complete flooding by unborated room temperature water, the storage array is treated as an infinite array of assemblies having an infinite fueled length.
B.
Under postulated conditions of envelopment by aqueous j
foam or mist, a range of foam or mist densities is I
examined to ensure that the maximum reactivity of the array is established.
The foam or mist is assumed to be pure water.
C.
For the analyses presented here, the poisoning effects of rack structure have been neglected.
Prior calcula-tions have shown this to be a conservative assumption, where the degree of conservatism depends on the exact rack structure design.
It is also assumed that no supplemental fixed poisons are utilized in the storage array.
March 1982 9.1-3 Amendment 8
PVNGS FSAR FUEL STORAGE AND HANDLING D.
Two concrete storage cavities are utilized for new fuel storage.
Each cavity is approximately 8 feet 8
by 23 feet and contains 45 fuel assemblies.
Three racks (figure 9.1-1) are installed in each cavity forming a 3-by 15-foot array of fuel assemblies.
The array is assumed to be surrounded on all six faces by a 2-foot thick, close-fitting reflector of concrete.
This assumption is conservative since the concrete walls are several inches away from the outer rows of fuel assemblies, the floor is several inches below the bottom of the active fuel, and the materials above the active fuel provide a substantially poorer reflector than the assumed thick concrete reflectors.
Calculations indicate that the assumption of concrete reflectors is conservative relative to the assumption of thick water reflectors.
r E.
The rack is assumed to be filled to capacity with fuel
(, _
assemblies.
F.
No burnable poison shims or other supplemental neutron poisons (e.g., CEAs) are assumed to be present in the fuel assemblies.
Criticality safety margins are maintained by:
A.
Limiting the size of the array to 90 assemblies.
B.
Defining an overall array configuration as shown in figure 9.1-1.
1 C.
Providing adequate mecnanical separation of fuel assemblies in the array, even under postulated accident conditions.
The mechanical separation provided is discussed in sec-tion 9.1.1.2.
In evaluating criticality safety, neutron cross section data i
for representative fuel rod cells, and material between and k-Amendment 8 9.1-4 March 1982
Attachment NOS82-326 4
April 23, 1982
---j 8 S9"i l/18 (TYP),lNSIDE DIMENSION OF THE BOX g
c BOXES MADE FROM 0.109" (MIN) STAIN LESS STEEL SHEET STOCK 00000 u
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9" MIN (TYP) 22" MIN (TYM WWWWW W
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NEW FUEL STORAGE RACK (150" ACTIVE FUEL LENGTil)
FicJure 9.1-1 March l'382 Amendment 8
Attaciunent
- NOS82-326 j
April.-23, 1982 1
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.1.
Question 10 - Page 25-26, Section 2.2.2 j
i
~
l ProNde the results of the' nuclear criticality safety an'alysis for the.' fuel l
assembly arrays in both. the new an'd the spent fuel pool. storage areas.
}
gguer The result of the. nuclear criticality analysis for the spent fue'l pool with 2
the fuel stored in a checkerboard array.is Keff = 0.8893.
This analysis was done using the following input.
i.
. ~
4.3'wt' % U-235 (4.0 wt % U-235 is i
Fuel enrichment the license limit) 1
!!onolith 304 sta'inless steel 0.120 inches j
wall thickness.
0.1'75 inches 1
"L" insert 304 stainless steel thickn~ess center-to-center. monolith 9.515 inches.
The result of the nuclear criticality analysis using KEi;0 for the new. fuel l
racks is.Keff = 0.7767 i.0070.
This' analysis was done using the following Input.
.4
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~
4.3 wt % U-235 (4.0 wt % U-235 is
. (
Fuel enrichment licenselimit)
' Stainless steel can.
6 4
l Thickness' around each fuel assembly 0.10 in,ches' 3 Mist density (H O) 2 0.0625 gm/cm i
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,. April'23,,1982 1
2.
Question 11 - Page 26, Section 2.2.3 Provide a-demonstration that shows the effect of all degrees.of moderation on'Keff of the arrays.
Describe the method (s) of analysis used and their validation.
' Answer i
The only rack ~which is currently a'nalyzed 'for mist
- conditions is' the new
' N
. fuel rack.
The results of the KEf;0 analysis for the new fuel rack for 4.3
-(
wt % U-235 (4.0 wt' % U-235 is license limit) without the 0.10 inch stain-
- less' steel can around the fuel assemblies are as follows:
f ll0 density,gms/cm1 gerf 3
l
'0f080 i
0.9144 t,0067
~
s i
O.0625
, 0.9277 t.0080 l
'.0.045,
0.8932..0070 i
A single ' case was run ' for the d.0625 gins /cm3 water density including the 0.10 inch stain 1.ess steel c~ n' surrounding the, fuel assembly and. resulted in a
a Kef f. of 0.7767 i.0070. -
Since this is.well below the. design limi.t of Kef f ' = 0.98 further variations of - the mist. density. were not deemed necessary.
/
i The method' of analysis and its validation is enclosed.
Since no critilcal
' ('.
. experiments are available for. mist conditions it is C-E's policy to' provide
(
large: margins between the calculated Keff of the rack and the flRC design limit 'of Xeff = 0.98.
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Attachment Nos82-326
. April 23, 1982 Qualification of Analytical Methods Used in Spent Fuel Storage Rack Analyses
(
I.
purpose The purpose of this paper is to provide qualification of the calculationa'l model and evaluation of calculational uncertainties and/or biasfactors lL used in analyzing spent fuel storage racks, especially the HI-CAP racks employing steel boxes and super HI-CAPS containing boron carbide poison.
This is based on the analysis of a variety of reactor and laboratory experi-4 ments.
The methods of cross section generation are essentially those of C-E's physics design procedures modi fied appropriately for use in four-t l
group transport, discrete ordinate method. criticality calculations,. and Monte Carlo codes.
i II. Calculational Uncertainty and Bias The rcsults of the analysis of a series of 'UO2 critical experiments are summarized in Table I.
Table I includes the mean and standard deviation. -
Although the spatial solution for the flux distribution was obtained by use of. a di f fusion theory code such as PDQ-7, transport corrections fo'r the i
reflector and heterogeneous lattice effects were employed.
These calcula-tions suppot use of the di f ferential cross-section data base. and ~. broad group crnss section generation codes.
(
~
Since fuel storage arrays do involve the spacing of the fuel 'assemb' lies at larger separation distances than in typical PWR reactor lattices, the pre-dictive capability of the calculational model was tested -on the following experiments.
In these -analyses done for this memo, the spatial flux.
solution was obtained directly with the transport code, ANISN. ' To assess the accuracy of the calculational model in predicting the multiplication factor of fuel assemblies having a separation distance sufficiently large.
so as to be isolated, analyses were carried out for a group of subcritical exponential experiments on clusters of 3.0 w/o U0 fuel pins clad with 2
type 304 S.S. and moderated by Hp (page 165 of _ Reference 7).
The cluster 6
sizes analyzed vary from 181 to 301 fuel rods so as to encompass the range -
of sizes typical of current PWR fuel assemblies.
The multiplication factors for the lattices analyzed using axial bucklings r duced from the e
reported relaxation lengths are tabulated below.
No. of Fuel Rods Keff 181 0.9966 211 1.0011 235 0.9966
(
265 0.9988 301 0.9984 i
w
Attachment NOS82-326
, April.23, 1982 These results indicate that the calculational model predicts the multipli-cation factor for small clusters of fuel rods in a water environment to a.
(
high degree of accuracy, i.e. a bias of
.0017.
To ascertain whether the. calculational mode can predict the reactivity characteristics of thick stainless steel plates and boron poisoned plates an analysis was made of PNW experimental (Reference 8) critical separations of 2.35 w/o U-235 U0 subcritical clusters.
The results using the Monte 2
Carlo code KEMO IV are shown in Table II.
Method of Calculation
. The calculation methods for these experimental comparisons which are also useo to determine reactivity for fuel rack
- storage, fuel shipping containers plus other fuel configurations found in fuel manufacturing areas are based on CEpAK cross sections.
Using an apprepriate. buckling value and taking proper account of resonance absorption, three fast groups ar.e col-lapsed from 55 fine energy mesh groups' in FORM and the one ther.aal group is collapsed from 29 thermal energy groups in. THERMOS.
In addition, each component such as water gap, or poison plate has its thermal cross section determined by a
slab THERMOS calculation employing the proper fuel environment.
FORM and THERMOS are sub-programs of CEPAK.
For one dimensional analyses such as BNL exponential experiments the discrete ordinates code ANISN (Reference 9) is used.
For two' dimensional analyses 00T-2W (Reference 10) is used.
For three dimensional analyses (such as the critical separation experiments) KEN 0.IV (Reference 11) is C
'used.
Results The above analyses indicate a mean error between predicted and measured multiplication factors of +.00135 and a calculational uncertainty of.00714 at the 95/95 confidence level for the complete series of UO2 experiments.
Thus, using CEPAK cross sections we conclude the following -
~
Total Number of Results 41 Mean Value (17) 1.00138 Standard Deviation = 6 0.00337 6 Multiplier for 9/95 confidence 2.118 95/95 Confidence Level Uncertainty 0.00714 Bias (4 -1.0)
+.00138 Uncertainty Minus Blas
.00575 8
=
pe
. -. - ~
attaemaent NOS82-326
' April 23, 1982 It will be noted that the seven no boron steel cases have a bias of 0.00207 (i.e. the calculated value is.00207 grea'ter than the critical Keff value of unity) which is greater than the mean bias.
The three boral cases have a bias of -0.00435 with unity particle sel f-shiel ding factor for the 8 C.
Because of the size and distribution of the boron carbide particle ~s 4
the boron allows more transmission than an equivalent homogeneous boron carbide mixture.
ficutron transmission experiments conducted by the University of 14ichigan for Brooks & Perkins Inc. (Reference 12) are consis-tent with using a 0.9 self-shielding factor in the third of four CEPAX neutron group and a 0.75 self-shiciding factor in the thermal group.
These self-shiciding factors which are used in designing' boron containing fuel racks make the bias for these boral cases +0.00008.
References:
- 1. T. C. Engelder, et al, " Spectral Shift' Control Reactor, Basic. Phisics Program", B&W-1273, tiovember,1963.
- 2. R. 11. C1 ark, et al, " Physics Verification Program Final Report", B&W-
- ^
3647-3, liarch,1967.
- 3. P.
W. Davison, et al, " Yankee Critical Experiments"., Y'AEC-94, April,
1959.
~
- 4. W. J. Eich and W. P. Rocacik, " Reactivity and ticutron Flux ~ Studies in 14ulti-Region Loaded Cores", WCAP-1443, 1961.
(
- 5. F.
J.
Fayers, et al, "An Evaluation of Some Uncertainties in the Comparison Between Theory and Experiments for Regular. Light Water
- Lattices, Brit. tiuc. En. Soc. J., 6, April,1967.
- 6. J. R. Brown, et al, " Kinetic and Buckling licasurements on tattice's of Sli ghtly Enriched Uranium and UO R ds in Light Water", WAPD-176, 2
1958.
- 7. G.
A.
- Price,
" Uranium Water Lattice Compilation Part I,
Bril' j
Exponential Assemblies", Bril-50035 (T-449), December,1966.
- 8. S.
R.
Bierman,'E.
D.
Clayton and B.
li. Durst, " Critical Separation Between Subcritical Clusters of 2.35 w/o U-235 Enriched U0 R ds in 2
llater with Fixed ticutron Poisons". Ptil-2438, October,1977.
- 9. !!ard W.
- Engle, Jr.,
"A Users llanual for Afil sfi, A One Dimensional Discrete Ordinates Transport Code With Anisotropic Scattering K-1693",
liarch 30, 1967.
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. Attachment
. NOS82-326..
, April'23, 1982 TABLE I'
(.
'Results, of Analysis of Critical U02 Systems
~
No)
_attice 2
8 tot Keff*
-1
-B&W (1)
I
.88-2 1.00121
.II'
,172-2.
1.00534 2
~.
3
-X-
.79-2
.99838 4.
XIII
. 701-2 1.00419
~
~
11.00550 JS XX
~.202-2
~-
6'
' B'8W '( 2) 1
. 861-2 1.00269 7
2
'420-2 1.00443
~
~
8 yankee (3)'
1
.408-2' 1.00088 2
.53.1-2 1.00115
~
9 10 3
'.633-2 1.00136
^
11
'-Ya'nkee (4)',
4
.688-2
'1.00244 i
! ('
12 '
'Winfrith (5)
R1-20
.6'60-2 l'.00214 13 R1-80
.626.99942 R3
-.510-2 1.00422 14 i
]
J 5
l 15 Bettis.(6) 1
.326-2 1.00053 l
16' 2
.355-2 1.00046
~
17'
'3
.342-2 1.00106 1.00208 l
Average
.+.00206 j
- Using calculated radial bucklings and measured axial bucklings.
i 9
e
Attachment NOS82-326
. April 23, 1982 TABLE II Calculated Keff Values For Separation Experiments Monte Carlo Expt #
Type Poison Plate Keff (STD Deviation)
~
15 None 1.00227
,.00534 04 None 0.99912
.00540 49 Hone 1.00221
.00473-18 None 1.00813
.00489 21 None 0.99589
.00461 28 304 S Steel 0.0 w/o Boron 1.00393
.00308 05*
304 S Steel 0.0 w/o Boron 1.00329
.00303 29 304 S Steel 0.0 w/o Boron,
1.00271
.00302 27 304 S Steel 0.0 w/o Boron 1.00418
.00273 26 304 S Steel 0.0 w/o Boron 0.99811'
.00279 34 304 S Steel 0.0 w/o Boron 0.99793
.00297 35 304 S Steel 0,0 w/o Boron 1.00436
.00290
,e 32 304 S Steel 1.05 w/o Boron 0.99970
.00524 33 304 S Steel 1.05 w/o Baron 1.01173
.00491-(
38 304 S Steel 1.62 w/o Boron 1.00289
.00512 39 304 S Steel 1.62 w/o. Boron 1.00208
.00506 20 Boral 0.99585
.00301 16 Boral 1.00020
.00288 17 Boral 0.99519
.00286 Mean K rr Value 1.00157 e
1 Standard deviation
.00419 5
G m
=
,,