ML19347F452
| ML19347F452 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 06/12/1981 |
| From: | Charles Brown, Hansen L, Nilson R SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML18046A641 | List: |
| References | |
| XN-NF-542, NUDOCS 8105190292 | |
| Download: ML19347F452 (20) | |
Text
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O' Issue Date: 06/12/80 XN-NF-542 PALISADES NUCLEAR GENERATING STATION SPENT FUEL STORAGE POOL CRITICALITY SAFETY REANALYSIS Prepared by:
0 0 #80 C. 0. Brown, Licensing Engineer Criticality Safety Concurred by: Ib 4//4/[d L. E. Ha'nsen, Senior Specialist Criticality Safety and Security N
Approved by:
R.' Nilson, Manager Corporate Licensing and Compliance May 1980 ERON NUCLEAR COMPANY,Inc.
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XN-NF-542 TABLE OF CONTENTS PALISADES NUCLEAR GENERATING STATION SPENT FUEL STORAGE P00L CRITICALITY SAFETY REANALYSIS Page No.
INT (000CTION.......................
j
-1 S,lMMARY.........................
2 FLEL ASSEMBLY DESCRIPTION 2
SPENT FUEL STORAGE POOL DESCRIPTION 3
CALCULATIONAL METHODS..................
RESULTS OF PREVIOUS STORAGE POOL CRITICALITY SAFETY 4
EVALUATIONS.......................
RESULTS OF NEW STORAGE POOL k CALCULATIONS........ eff 4
6 STORAGE P0OL ACCIDENT CONDITIONS.............
7 CONCLUSIONS.......................
14 REFERENCES........................
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XN-NF-542 LIST OF TABLES PALISADES NUCLEAR GENERATING STATION SPENT FUEL STORAGE POOL.
CRITICALITY SAFETY REANALYSIS Table No.
Page No.
I PALISADES (EXXON NUCLEAR BATCHES H AND I)
FUEL ASSEf1BLY PARAMETERS...........
8 II CALCULATIONAL RESULTS OF THE PALISADES SPENT FUEL STORAGE P0OL CRITICALITY SAFETY REANALYSIS BY P. 500NG (NUS) IN JULY 1978...................
9 Ii!
PALISADES NOMINAL MAIN POOL STORAGE ARRAY k
VERSUS ENRICHMENT CALCULATIONS.....
10 eff
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XN-NF-542 LIST OF FIGURES PALISADES NUCLEAR GENERATING ~ STATION SPENT FUEL STORAGE POOL CRITICALITY SAFETY REANALYSIS Figure No.
Page No.
1 PALISADES (EXXON NUCLEAR BATCHES H AND I)
FUEL ASSEMBLY DESIGNS.............
11 2
MAIN POOL STORAGE CELL-NOMINAL.GE0 METRY ARRANGEf1ENT..................
12 3
PALI.ADES MAIN P00L STORAGE ARRAY WORST CASE k,ff VERSUS ENRICHf1ENT..........
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t
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XN-NF-542 PALISADES NUCLEAR GENERATING STATION SPENT FUEL STORAGE P0OL CRITICALITY SAFETY REANALYSIS INTRODUCTION In November 1976 Consumers Power Co. submitted to the USNRC a request to install high density spent fuel storage racks at the Palisades Nuclear Generating Station. The subsequent safety analysis (I) showed the pool to be adequately subcritical for fuel assemblies 235 enriched to 3.05 wt. %
0.
With the trend in fuel management toward higher burnup, average reload fuel assembly enrichments have increased to as high as 3.267 w/o, (Exxon Nuclear Batch H). Thus, in July 1978 a reanalysis (2) of the high density storage racks was per-formed by NUS Corp. for the average Batch H reload fuel enrichment of 3.267 w/o.
The intent of this reanalysis is to define a maximum average 2S fuel assembly enrichment (i.e. maximum axial Uloading),which will continue to meet applicable criticality safety criteria.
SUMMARY
The criticality safety reanalysis of the Palisades spent fuel storage pool, as described in Reference 1, and this report demon-strates the pool to be adequately subcritical, i.e. keff 0.95 at 235 the 95% CL(confidence level), for fuel assembly axial U loadings up to and including 44.11 g/cm (3.80 w/o). The worst case k f
eff.-..-.--
XN-NF-542 the racks is estimated to be 0.946 at the 95% CL for fuel assemblies enriched to 3.80 w/o.
FUEL ASSEMBLY DESCRIPTION The Exxon Nuclear Batch H and I fuel assembly designs are depicted in Figure 1.
As indicated, the 15x15 lattice arrangement includes a single instrument tube, eight guide bars, and eight locations for removable poison rods, gadolinia-bearing fuel pins or water rods.
Current plans call for a Batch I reload of 68 fuel assemblies. Of these fuel assemblies 48 will contain eight water rods, twelve will contain eight gadolinia-bearing fuel pins and eight will provide eight poison rod loca tions.
The fuel assembly parameters assumed in this evaluation are given in Table I.
From this information bundle-averaged cell parameters were
' calculated by including the zirconium associated with the instrument tube, guide tubes and the guide bars in the zirconium clad of each fuel rod. Water associated with each guide bar, instrument and guide tube was included by increasing the unit cell dimension (rod lattice pitch).
For producing cell-averaged cross section data the above assumptions permit a conservative estimation of the effect on reactivity of the extra zirconium and water within the fuel assembly by maintaining the correct assembly water-to-fuel volume ratio.
1 SPENT FUEL STORAGE P0OL DESCRIPTION
(
The Palisades spent fuel storage pool consists of racks comprising the gmain pool and tilt pit pool.
The main pool storage racks have 8.56"
XN-flF-542 square ID storage cells and a 10.25" center-to-center spacing, see Figure 2.
In the tilt pit pool the rack is designed to store control rods as well as fuel assemblies. This rack has a 9.0" storage cell ID and cells are located on 10.69"xil.25" centers. As demonstrated in Reference 2 the reactivity of the main pool storage racks is s2.0% ak/k higher than the tilt pit pool rack.
Hence. the main pool rack design is analyzed with respect to criticality safety as the limiting case.
The neutron absorber plate is composed of B C bonded in a carbon 4
matrix. The plate is 0.21" thick and 8.26" wide with a minimum 10B 2
loading of 0.0959 g/cm. As shown in Figure 2 each plate is centered width-wise in each storage cell wall.
CALCULATIONAL METHODS The KEf10 IV Monte Carlo code (3) was utilized to calculate the reactivity (keff) f the Palisades spent fuel storage pool. fiul ti-group cross section data from the XSDRN 123 group data library were generated for input into KEN 0 IV using the NITAWL( ) and XSDRNPMI#)
codes. $pecifically, the NITAWL code was utilized to obtain cross section data adjusted to account for resonance self-shielding using the fiordheim Integral fiethod. The XSDRNPfi code, a discrete ordinates one-dimensional transport theory code, was then used to prepare spatially cell-weighted cross section data representative of the fuel assembly for imput into KEN 0 IV...
l XN-NF-542 RESULTS OF PREVIOUS STORAGE P0OL CRITICALITY SAFETY EVALUATIONS In order to show the storage pool to be adequately subcritical for Exxon Nuclear Batch H fuel enriched to 3.267 w/o, a reanalysis (2) of the pool was performed by P. Soong of NUS in July 1978.
In this analysis storage array k,ff values under nominal conditions were calculated using the " KEN 0 code in conjunction with 123-group AMPX averaged cross sections".(2) The PDQ-7 code with NUMICE (NUS version of LEOPARD) cross sections was then used to calculate the reactivity changes resulting from variations in storage rack conditions. The reactivity for " worst case" conditions was then calculated by summing the KEN 0 calculated k,ff for nominal conditions and the Ak values calculated using the PDQ-7 code. Table II sumarizes eff the results of these calculations for both the main pool and the tilt pit pool.
RESULTS OF NEW STORAGE P0OL k CALCULATIONS eff In order to demonstrate the criticality safety of the storage pool for higher fuel assembly enrichments, additional k calculations eff were performed using the KEN 0 IV code.
Using a geometric model of the nominal main pool storage arrangement as described in Reference 2 and depicted in Figure 2, k was calculated for Exxon Nuclear Batch H eff fuel (See Table I) enriched to 3.267 w/o. The resulting k value of eff 0.875 +.005 represents an infinite array and is within two standard deviations of the KEN 0 k,ff reported by P. Soong in Reference 2 and shown in Table II of this report. Hence, based on this result it is concluded that the calculational model used for this calcu-lation gives conservative results for the main storage pool..._ _
XN-NF-542 The above duplication calculation for Bat'ch H fuel represented a fuel assembly design of 208 active fuel rods. To determine the effect on S
array k f a higher fuel assembly axial U loading based on the eff number of fuel rods, the above case was rerun with 216 fuel rods.
For this case k,ff was calculated to be 0.365 I.005.
This result indicates a slightly greater reactivity worth of the water holes in the 208 fuel rod arrangement relative to having those positions filled with eight additional fuel rods. Since the fuel assembly design with fewer active rods gives a higher array keff, all subsequent calculations assume 208 fuel pins.
Having established a representative calculational model, additional reactivity calculations were performed for increased fuel assembly enrichments in the nominal main pool storage arrangement. These results are summarized in Table III and are shown graphically in Figure 3.
Also shown in Table III are final worst case k values estimated eff at the 955 confider.ce level for the storage pool based on fuel 23S assembly axial U loading. These values were calculated from the following expression:
keff (WC) = keff(f1 m) + 2c + W + T o If(fiom) = the nominal storage array k k*
eff l
- where, one standard deviation W = Root-mean square of the worst case tolerance variations, see Table II.
T = Maximized moderator temperature variation, see Table II.
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XN-NF-542 No reactivity penalty was assessed for either 8 C particle self-4 shielding or calculational bias defined by benchmark calculations.
In the KEN 0 IV calculational model B C was conservatively modeled 4
to account for the effects of self-shielding (i.e. the boron was lumped in the center of the neutron absorber plate). With regard to calculational bias Exxon Nuclear Co.. has had extensive experience using the above computer code calculational model and benchmark calcula-tions(5) show no significant bias using these methods as described.
STORAGE POOL ACCIDENT CONDITIONS Since the storage pool will under normal operating conditions contain s2,000 ppm soluble boron (I) in the water, the actual k eff of the storage array, based on the described pool conditions, will be s20% lower (2) than calculated. Thus, for the cccident conditions of a fuel assembly lying either across the racks or up against the outside of the racks, the storage array reactivity will remain well below the limiting value of 0.95.
In the event of a single failure in the storage pool cooling system, based on the assumed conditions presented in Section 6 of Reference 1, the bulk pool temperature would not exceed 118 F for a 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> normal off load.
For this condition the maximum surface temperature of a fuel rod is less than 230*F providing greater than 9*F margin to local boiling.(I) Both initial (2200 MW ) and stretch (2650 MW )
t t
power cores and their applicable design peaking factors have been considered in establishing limiting thermal conditions.(I) From -
XN-NF-542 the standpoint of neutronics, any localized fuel rod surface boiling inside the fuel assembly will have a negative effect on array re-activity (i.e. k,ff of the array will decrease).
CONCLUSIONS This analysis conservatively demonstrates the reactivity of the 235 Palisades spent fuel storage pool for fuel assembly axial U
loadings of 1 44.11 g/cm (3.80 w/o fnr Batch I fuel with 208 active fuel rods) to be less than 0.95 under existing assumptions of worst credible storage array conditions described in this report.
- Hence, at the 95% confidence level the k f the storage pool will be eff 1 0.946.
The analytical efforts, the results of which are presented in Table III, were reviewed by a second party knowledgeable in the performance of criticality safety evaluations...-.
XN-NF-542 7ABLE I Palisades (Exxon Nuclear Batches H and I)
Fuel Assembly Parameters Nominal Lattice Pitch, in.
0.550 Clad OD, in.
0.417 Clad Material Zr-4 Clad Thickness, in.
0.028 UO Pellet Diameter, in.
0.350 7
Pellet Density, % pT 94 + 1.5 Percent Dish 1.0 No. Active Fuel Rods 208 (Batch H) 208, 216 (Batch I)
Ave. Enrichment 3.267 (Batch H - 208 active rods)'
3.260 (Batch I - 208 active rods) 3.232 (Batch I - 216 active rods)
Rod Array 15x15 Eff. Array Dimensions, in.
8.25 x 8.25 No. Guide Bars (Solid Zr) 8 No. Guide Tubes 8
GT OD, in.
0.416 GT TK, in.
0.011 No. Instrument tubes 1
0.415 IT TK, in.
O.329
.a.
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XN-NF-542 TABLE II Calculational Results of the Palisades SpentFuef2}toragePoolCriticalitySafety by P. Soong (NUS) in July 1978 Reanalysis Batch H Fuel (3.267 w/o)
Nominal Storage lattice Case Description Cell k=
PD0-7 KENO 1
Main Pool 0.8569 0.8693*+.0042 2
Tilt Pit Pool 0.8397 0.851610034 Worst Case Parametrics, ak 3
Enrichment Variation
.0037 4
UO., Density Variation
.0006 5
B C Slab Width
.0008 6
B C thickness and loading
.0038 7
V riation in Spacing
.0081 8
Storage Can Dim. Variation
.0097 9
B C Slot thickness
.0045 10 B$CSlabMissing(1)
.0018 11 Bdw and Twist
.0165 12 Storage Can thickness
.0061 TOTAL
.0228* (Root-mean square sum)
Other, ok 13 Temperature Variation
.0028*
14 Two Standard Deviations
.0084*
15 Particle Self Shielding (B C)
.0040*
4 16 Benchmark Bias
.0086*
Total Sum of (*) Values, k=
17 Maximum Credible Worst Case 0.9309 These Values summed to establish maximum " worst case" storage pool keff*
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XN-NF-542 TABLE-III Palisades Nominal Main Pool Storage Array k,ff Versus Enrichment Calculations Fuel Assembly Parameters - See Table I Storage Array Description - Main Pool (Nominal)
See also Figure 2 Nominal 235 KENO IV Enrichment, Axial 0
(123 group)
Worst Case
- kgff Case w/o Loading, g/cm k,ff + o at 95% CL 1
3.267 37.92 0.875 +.005 0.9106 2
3.50 40.63 0.892 7.005 0.9276 3
3.70 42.95 0.906 7.004 0.9396 4
3.80 44.11 0.9460 (est.)
5 3.90 45.27 0.918 +.004 0 '516 For Cases 1, 2, 3, and 5 worst case values are calculated ar follows:
k,ff(WC) = k,ff(Nom) + 20 + W + T where W = 0.0228 (Root-Mean square sum of tolerance variations, see Table II).
T = 0.0028 (Maximized moderator temperature variations).
Case 4 worst case k value taken from graph, see Figure 3.
gff..
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XN-NF-542 FIGURE 1 I **
PALISADES (EXXON NUCLEAR BATCHES W AND H) FUEL ASSEMBLY DESIGNS LLLLGLLLLLGLLLL LLLHHHHHHHHHLLL LLLHHPHHHPHHLLL LHHHHHHHHHHHHHL GHHHHHHHHHHHHHG Exxon Nuclear Batch H L H P H H H H H H it H H P h L LHHHHHHHHHHHHHL L - 2.90 w/o Fuel LHHHHHHIHHHHHHL LHHHHHHHHHHHHHL H = 3.43 w/o Fuel LHPHHHHHHHHHPHL GHHHHHHHHHHHHHG G = Guide Bar LHHHHHHHHHHHHHL LLLHHPHHHPHHLLI I = Instrument Tube LLLHHHHHHHHHLLL LLLLGLLLLLGLLLL P = Poison Rod Location LMMMGMMMMMGMMML fiMMHHHHHHHHHMMM Exxon Nuclear Batch I MMMHHPHHHPHHMMM MHHHHHHHHHHHHHM L = 2.52 w/o Fuel GHHHHHHHHHHHHHG MHPHHHHHHHHHPHM fi = 2.90 w/o Fuel MHHHHHHHHHHHHHM i
MHHHHHHIHHHHHHM H = 3.43 w/o Fuel MHHHHHHHHHHHHHM MHPHHHHHHHHHPHM G = Guide Bar GHHHHHHHHHHHHHG MHHHHHHHHHHHHHM I = Instrument Tube MMMHHPHHHPHHMMM MMMHHHHHHHHHMMM P = 2.52 w/o Fuel LMMMGMMMMMGMMML with 4.0 w/o Gadolinia*
or, Water Hole; or, Poison Rod Location Gadolinia-bearing pin locations may vary slightly from P locatior.s I
as shown.
- Correction of typegraphical error. L
XN-NF-542 Poison Plate:
B C in Carbon Matrix Width-8.26" Thickness - 0.21" 10 2
B Ldg. - 0.0959 g/cm (min.)
I
_y____________y-I i
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Inside E
I Poison Plate Water l
F]"
Slot (0.25")
J, Gap lr 4
(0.155")
I j
l Outside Water Gao (0.69")
Fuel Assembly l
4 l
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3.25" x 8.25" Stainless Steel j
)0.125" thick
][
g
{
i l
f i
Stainless Steel I
l u
--w 21 (0.25" thick)
______________9_
l I
Storage Cell:
10.25" Center-to-center Can ID-8.56" Can 00-9.56" Inside Water Gap-0.155" Outside Water Gap-0.69" Note: With the poison plate in place a gap of 0.04" (total) is allowed in the poison plate slot.
Figure 2 Palisades !!ain Pool Storage Cell Noninal Arrangement -.
X*I 'iF-5e2 0.96 (Plotted values reoresent k,ff 1 2e l/
0.95 0.946 0.94 -
1
~~
Array k,ff 0.93 --
0.92 --
D
~~
0.91 --
I i
4; l
0.90 --
3.0 3.1 3.3 3.5 3.7 3.9 4.0 235 wt %
U Figure 3 Palisades Horst Case itain Pool k,ff Versus Enrichment.._.. _
XN-NF-542 REFERENCES 1.
" Spent Fuel Pool Modification Description and Safety Analysis,"
Consumers Power Company, Palisades Nuclear Generating Station, Docket No. 50-255 (Nov.1976).
2.
- Soong, P., " Criticality Analysis for 3.27 w/o Enriched Fuel Palisades High Density Fuel Rack," NUS Corp. (July 1978).
3.
L.M. Petrie and N.F. Cross, " KEN 0 IV: An Improved tionte Carlo Criticality Program," ORNL-4938, Oak Ridge National Laboratory (November 1975).
4.
N.M. Greene, et al., "AMPX - A Modular Code System for Generating Coupled Multigroups Neutron-Gamma Libraries from EtlDF/B, "0RNL-TM-3706, Oak Ridge National Laboratory (March 1976).
5.
C.O. Crown, " Criticality Safety Benchmark Calculations for Low-Enriched Uranium Metal and Uranium 0xide Rod-Water Lattices", XN-NF-499 Exxon Nuclear Co.
(April 1979).
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XN-NF-542 ISSUE DATE: 06/12/8 PALISADES NUCLEAR GENERATING STATION i
SPENT FUEL STORAGE POOL CRITICALITY SAFETY REANALYSIS 3
Distribution C.O. Brown (2) i L.E. Hansen R. Nilson H. G. Shaw I
Consumers Power.o.
(10)/HG Shaw Document Control (5) r
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