ML20114B963
| ML20114B963 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/08/1992 |
| From: | Sarah Turner HOLTEC INTERNATIONAL |
| To: | |
| Shared Package | |
| ML20114B957 | List: |
| References | |
| HI-91671, HI-91671-R03, HI-91671-R3, NUDOCS 9209010017 | |
| Download: ML20114B963 (28) | |
Text
_
,[
ATTAClit4Ef4T 1 5
CRITICALITY SAFETY EVALUATION OF 4
THREE MILE ISIAND FUEL STORAGE FACILITIES WITH FUEL OF 5% ENRICIDfENT
~
Preparo.1 for the GPU NUC2M CORPORATION by S';anley E. Turner, PhD, PE Holtec Project 10240 Holtec Report HI-91671 230 Normandy Circle 2060 Fairfax Ave.
Palm Harbor, FL 34683 Cherry Hill, NJ 08003 pb kDOC
[
P
=
l succ t."'tenanom to:co9-m-ino nur, 07 22 i n 1o no.011 5. c 3
1 I
i O
HOLTEC l
INTtaNATIONA KrVIEW AND CERT 2FICATICN LOG l
I i
CRITIC.ALITY 3ATETY ANALYS23 Of TIIREE MILE ISLAND DOCUMENT HAMEt TUrJ STORAGE FACILITIES W1118 rutL or 5% DrnIclo.ENT h0L77,C DOCUMENT 2.D. No.
- EI-91671 got,rge ;.RoJtcT NUMBER 8 10840 CUSTQh2R/ CLIENT OFU WUCLEAR CORPORATION 1
I u
REVISION BIACK QUALITY DROJECT I
ISSUE AUTnoR REVIti?'.~R AssURAWer MANACIR
(
j 4
NQ.
a DATE
& DATE I DATE
- f. DATE w-s.m. m xan s.E. runuza W G d e./w ;
ORIGINAL & ~ 4.Jw
'h;.c.w.as u x.ue
- s. E.h 5.s.Tuasee)
L
%.I,fn.) d k k n W %E FfRae, /in.J REV' 1 i
6.s.Tennsh ' w noru ELL h $,o{p
(
$.t.%RHCR i
~
@ r 14f. #,Afde '/s%[jk Y'-- a/a/r>
,,,, rgru g g l s. t. ma s.54 g,enteseW4 ft 7,(g, J
C_ v4w.s.ardrm pc 4,N,4
- )tust be Pro $ set Manager or his Desigr.ee.
NOTE:
Signatures and printed names are required in the review blocx.
'Jhis documant conforas to the requirements of the purchast g
spelfication and the r7plicable sections of the governing codes.
-r=.*__a T::::= :.
U 3
'd 6E821 FG/30/90
~. 1 N ' D 3170'i WOdd I
l TABLE OF CONTENTS
1.0 INTRODUCTION
AND
SUMMARY
1 2.0 CRITICALITY SAFETY ANALYSES 3
2.1 New Fuel Storage Vault 3
2.2 Pool A Storage Racks 5
2.2.1 Normal Storage of 5% Enrich &2 Fuel 5
2.2.2 Soluble Boron Concentrations for Accident Condition 7
2.3 Pool B Storage Ra.'ks 8
2.3.1 Fuel Burnup Requirements 8
2.3.2 Misplaced Fuel Assembly.
9 2.3.3 Dropped Fuel Assembly.
9 3.0 ANALYTICAL METHODOLOGY 11
+
4.0 REFERENCES
13 9
e
}
List of Tables Tablo 1
SUMMARY
OF CRITICALITY SAFETY ANALYSES NEW FUEL STORAGE VAULT 14 Tabla 2 DESIGN BASIS FUEL ASSEMBLY SPECIFICATIONS 15 Table 3 EUMMARY OF CRITICALITY SAFETY ANALYSES POOL A RACKS 16 List of Fleures Fig. 1 NEW FUEL VAULT CONFIGURATION 17 i
Fig. 2 REACTIVITY VARIATION WITH MODERATOR DENSITY 18 j
Fig. 3A CROSSECTION OF REGION 1 STORAGE CELL 19 IN POOL A Fig. 3D CROSSECTION OF REGION 2 STORAGE CELL 20 IN POOL A 2
Fig. 4 MINIMUM BURNUP REQUIREMENTS FOR FUEL IN
]
REGION 2 OF THE POOL A STORAGE RACKS 21 Fig. 5 CROSSECTION OF POOL B STORAGE LELLS 22 j
Fig.
6-MINIMUM BURNUP REQUIREMENTS FOR FUEL IN POOL B 23 i
11 i
I L
1.0 INTRODUCTION
and
SUMMARY
The present study was undertaken for the purpose of documenting the capability of the fuel storage facilities at Three Mile Island to safely accept fuel of 5% initial enrichment.
The three fuel storage facilities are the following:
New Fuel Storage Vault, designed to receive and store fresh fuel in the normally dry condition, Pool A with high density poisoned storage racks originally designed for 4.6% enriched fuel, and Pool B with unpoisoned racks originally qualified for 4.3%
enriched fuel.
Criticality safety analyses and accident evaluations for each of the three fuel storage facilities at Three Mile Island are presented in separate sections of this report.
I Calculations reported here confirm that the new fuel storage vault can safely receive and store fuel of a 5.00% i O.05%
enrichment within the limits of the USNRC guidelines.
To accomplish this, two rows of storage positions must be blanked off j
and remain unused for fuel in order to provide the additional neutron leakage required to maintain the necessary sub-criticality margin under postulated accident conditions (non-fuel components are acceptablo).
The stiorage racks in Pool A use Boral* absorber panels for reactivity control and were initially qualified for fuel of 4.6%
enrichment with a
substantial margin below the regulatory guidelines.
Subsequent data on the as-built Boron-10 loading in l
the Boral absorber panels indicated that there was additional margin that would enable the racks to accommodate fuel of higher enrichments.
Revised calculations of the racks with the as-built boron-10 loading confirm that the racks in Region 1 of Pool A can l
safely accommodate 5% enriched fuel with a maximum k of 0.947 g,
including all uncertainties. The interf ace between storage modules (including Region 1 to Region 2) includes an adequate water gap to preclude any increase in reactivity.
Pool B racks are unpoisoned and were previously qualified for l
unburned fuel of 4.3% enrichment using a stainless-steel and water
)
j flux trap between storage cells as a
means of controlling l
l reactivity.
By using some of the available margin in reactivity
]
below the 0.95 k, limit, the Pcol B racks wrtre shown to be capable g
of cafely accommodating fuel of 4.37% enrichment. Calculations for l
accident conditions of a single fuel assembly of 5% enrichment placed in a Pool D rack confirmed that the maximum reactivity would remain within the USNRC guideline and therefore is acceptable.
Calculations to determine the required discharge fuel burnup of 5%
enriched fuel showed that a burnup of only 2.48 MWD /KgU was adequate for unrestricted storage in the Pool B racks.
A fuel assembly of 5% enrichment dropped and lying across the top of a Pool B rack does not have a reactivity any higher than that of a single isolated assembly suspended in water.
i Both Pool A and Pool B use soluble boron as a means of
{
augmenting reactivity control. Since credit for the soluble poison is acceptable for accident conditions (under the double contingency calculatio's were also made to determine the minimum principle),
n soluble boron concentration required in Pool A to assure that the required suberitical reactivity margin is maintained under the accident conditions of mislocated fuel assemblies.
To protect against the postulated mis-location of a fresh fuel assembly in 1
Pool A will require a minimum of 150 ppm soluble boron.
In Pool B, with fuel of 4. 37% enrichment, credit for soluble boron is not I
necessary for postulated accident convitions.
A concentration of 200 ppm,
- however, would enable the Pool B racks to safely accommodate fresh fuel of 5% enrichment.
.~
2.0 CRITICALITY SAFETY ANALYSES 2.3 New Fuel Etorage Vault The new fuel storage vault is intended for the receipt and storage of fresh fuel under normally dry conditions where the reactivity is very low.
To assure the criticality safety under accident conditions and to conform to the requirements of General Design Critorion 62, " Prevention of Criticality in Fuel Storage and Handling",
two separate criteria must be satisfied as defined in NUREG-0800, Standard Review Plan 9.1.1, "New Fuel Storage".
These critoria are as follows:
When fully loaded with fuel of the highest anticipated reactivity and flooded with clean unborated water, the maximum reactivity, including uncertainties, shall not exceed a k,,, of 0.95.
With fuel of the highest anticipated reactivity in place and assuming the optimum hypothetical low density moderation, (i.e., fog or foam), the maximum reactivity shall not exceed a k,,, of 0.98.
The new fuel storage vault normally provides a6 x 11 cell array of storage locations arranged on a 21.125 inch lattice spacing. Preliminary c'alculations showed that it is necessary to blank-off, and keep empty, two rows of storage locations as indicated in Fig.
1.
With this rostriction, the new fuel storage l
vault can accommodate a maximum of 5% enriched fuel within the two guidelines identified above.
For the configuration shown in Fig. 1, calculations were made with the 27-group NITAWL-KENO-Sa code package for both the usual B&W f '.' a l 00:0 tlics er.d U.=
newer Westinghouse 15 x 15 fuel i
assemblies at 5% U-235 enrichment.
In these calculations, it was found that the B&W fuel gave the higher reactivity for the low-a
. ~,, +, - - -.
,v.--
-..-r
- ~. -. -
\\
1 l
1
)
l density "optilnum" moderation condition while the Westinghouse fuel gave the higher reactivity under the fully flooded accident 4
condition.
]
)
Results of the calculations for various water densities are i
shown in Fig.
2, indicating that the limiting reactivity co 41 tion
)
for low-density optimum moderation occurs at a
hypo' ical moderator density of 9%.
With all 54 locations filled wii Si enriched fuel (two blanked-off rows), the maximum reactivit3 l
including uncertainties, was calculated to be 0.953 which is well below the regulatory limit of 0.98.
For the fully flooded accident condition, the maximum k, g
including uncertainties, calculated for fuel of 5% enrichment, was O.949 which is.Within the regulatory limit and therefore acceptable.
Table 1 summarizes the criticality calculations for both accident conditions.
Based upon these calculations (and the data in Fig. 2) and the assunption of two blanked-off rows of storage cells, it is concluded that the new fuel vault at Three Mile Island will safely accommodate 5% enriched fuel assemblies of either the B&W or Westinghouse design with assurance that the maximum reactivities conform to the criteria of SRP 9.1.1.
l 1 l 1
1 2.2 Pool A Storage Racks l
2.2.1 Normal storage of 5% Enriched Fuel l
{
The spent fuel storage racks in Pool A,
pictorially j
)
described in Figs. 3A and 3B, were initially designed (" for fuel of 4.6% initial enrichment bt t with a substantial margin below the reactivity limit.
Subs.
ntly, when as-built B-10 loading data I
became availabic for th.
aral panels, the storage racks were re-l evaluated.
At the sate came, the design basis temperature was j
changed to 4* C (temperature of maximum water density)) to assure l
that the calculations were made at the temperature corresponding to j
the highest possible reactivity.*
i i
i calculations were also made for Westinghouse-design fuel of 5% initial enrichment in both Regions 1 and 2.
Initial j
calculations showed that the Westinghouse fuel gave a slightly higher reactivity in the storage racks and, therefore, subsegaent l
evaluations in Regior. I were based on the Westinghouse type fuel.
l However, in Region 2, the B&W fuel design showed a slightly higher reactivity than the Westinghouse fuel design, as indicated by the following results:
Calculated Reactivity (not including uncertainties)
B&W Fuel Westinchouse Fuel 2% fuel 0 4.92 MWD /KgU O.9245 0.9212 5% fue'l at 42 MWD /KgU O.9062 0.8997 Therefore, the B&W fuel design remains ~ controlling in Region 2.
4 Fuel design specifications, as used in the calculations, are shown in Table 2.
See INPO Significant Event Report dated 24 October 1990, relating to " Reactor Coolant System Temperature Below Analyzed Limit for Extended Time Period"
1 i
Using the as-built B-10 loading of 0.0222 i O.0016 g/cm2 in Region 1 (rather than the original design loading of 0.0211 i i
O.0011 g/cm2 ), the racks were re-evelaated at 4'C for both B&W anc.
l Westinghouse fuel of 5% enrichment.
Results of the criticality I
calet.lations are summarized in Table 3.
The maximum k,,, was calculated to be 0.947 for 5% Westinghouse fuel in Region 1,
including $1.1 uncertainties (95%/95%).
These calculations assumed
]
the same uncertainties as in the previous evaluation, except that the actual tolerance on B-10 loading was used.
Thorofore, it is i
concluded that the Regicn 1 storage racka can safely accommodate fuel of 5% enrichment within the limits of the USNRC guidelines, i
since the original design calculations for Region 2 had j
already extended the analysis to include 5% enriched fuel, little change resulted from the revised calculations.
The principal changes resulted from the reduction in design temperature to 4 *C, the use of as-built B-10 loading in the Boral poison panels, and i
the use of the KENO-Sa code (rather than KENO IV), which permitted I
a more accurate geometrical description in the check calculations.
I The revised calculations resulted in a negligible change to the curve of enrichment-dependent burnups required for storage in Region 2 shown in Fig.
4, but did result in a small reduction in the reference k,,,,
The maximum calculated k,,, in Region 2 is 0.933 including all uncertainties as indicated in Table 3, compared to
{
the previous value of' O.939.
It may be noted that there is a substantial margin below the regulatory limit of 0.95 for k,,,.
Therefore,,the curve in Fig. 4 represents the acceptable burnup domain for fuel of various initial enrichments with a conservative estimat0 of the maximum k,,,.
The fit to the burnup-limit data in Fig. 4 is the following:
Minimum Acceptable Burnup (MWD /KgU) in Region 2 of. Pool A
= -29.846 + 20.908 E - 2.046 E2+ 0.1441 E8 for Enrichments (E)-of 2% to 5%
v - - _ _
l p
i The data in Fig. 4 extrapolate to a zero-burnup intercept, calculated to be 1.75% enrichment.
The intermediate burnup limits 1
i between 1.75% and 2% enrichment can be read from the dotted line or 3
j may be calculated from the polynomial fit equation.
Although the polynomial fit to the curve was developed over the range from 2%E to 5%E, it may still be used below 2% enrichment with a small but conservative over-prediction (i.e.,
produces a
zero-burnup l
intercept of 1.67%E) j 2.2.2 Soluble Boron Concentrations for Accident Conditions s
]
The accident condition in which a fresh fuel assembly of j
the maximum allowab e reactivity is inadvertently loaded into an otherwise filled Region 2 rack has also been evaluated.
Conser-I vative calculations (NITAWL-KENO Sa) show that this accident l
conditior. cauld result in exceeding the regulatory limit on k,,, in the absence of soluble boron, although criticality would not be achieved (maximum k,,, of 0.973, including uncertainties). Normally about 2000 ppm of boron are present and the calculations show that, with the 2000 ppm soluble boron present, the maximum k,,, was 0.755.
Calculations were also made to determine the soluble boron required to protect against the misplaced assembly accident.
For this accident condition, calculations showed that only 150 ppm are required to maintain the k,,,
at 0.95 (including uncertainties).
Credit for the presence of soluble boron is allowable for accident conditions under the double contingency principle (April 14, 1978 f
USNRC letter).
l 1
i 3
[
l l,-
4 3
2.3 P_gpl B St2rgae R3cks The Pool B storage racks, pictorially described in Fig. 5 i
are unpoisoned racks previously qualified for fuel of 4.3%
l enrichment based upon CASMO-2E calculations.
Re-evaluation with l
CASMO-3 confirmed the previous calculated maximum k,,,.
- However, f
corresponding KENO-Sa calculations (benchmarked against cases with large water gaps) indicate that CASMO-3 overpredicts reactivity for a
l configurations with water-gaps comparable to those in the Pocl B i
racks ( 4.131 inches). Nevertheless, the calculations reported here l
were corrected to the CASMO-3 result to assure conservatism.
5 With axial leakage included in a three-dimensional KENO-Sa i
calculation, the corrected (conservative) maximum k,,, was O.946, including uncertainties.
Increasing the allowabla enrichment to 4.37%, the maximum reactivity would increase to 0.949, confirming that the Pool B racks can safely accommodate 4.37% enriched fuel.
1 If fully loaded with 5% enriched fuel, the reactivity in Pool B would be 0.970.
An estimated soluble boron concentration of 200 ppm B would be adequate to allow the Pool B racks to safely accept 5% fuel within USNRC guidelines.
Calculations have been made to dete: mine the discharge burnups
(
required for fuel of e'nrichments up to 5% to be safely stored in the Pool B racks.
These calculations and the evaluation of i
potential accident conditions are described in the following l
paragraphs.
l 2.3.1 Fuel Burnup Requirements Burnup calculations were made for fuel of 4.5%, 4.75%'and 5%
initial enrichment and the restart option in CASMO-3 used to determine the reactivities in the storage rack.
The rer, tart calculations were then interpolated to obtain the burnup required to yield the same reactivity as the reference calculation (4.37%
unburned fuel).
As an allow.nce for uncertainty in depletion
J
?
l 4
calculations, the required burnups determined above were increased l
by 5%.
Results of these calculations are shown in Fig. 6 and indicate that a minimum burnup of 2.48 MWD /KgU was adequate for 5%
enriched fuel co be acceptable for storage in the pool B racks.
j The data in Fig. 6 have been fitted to a polynomial exprettsion as j
follows:
l
}
Hinimum Acceptable Burnup (MWD /KgU) in Pcol B l
= 703.497 - 448.751*E
- 94.5773*E2-6.57357*E3 For Enrichments, E, between 4.37% and 5.0%
1 i
i i
2.3.2 Misplaced Fuel Assembly Although credit for the soluble baron present in the pool 4
water is allowed under accident conditions, calculations were made for a fresh fuel assembly of 5% enrichment accidentally installed in a Pool B cell, with surrounding cells filled with fuel of the maximum pemissible reactivity.
This calculations used a three dimensional HENO-Sa model with a correction f actor derived from CASMO-3 calculations.
For the accident condition of a fresh fuel assembly of 5% enrichment loaded into i Pool B rack otherwise j
filled with 4.3% enriched fuel, the maximum calculated k,,, value j
was 0.9495, which is still within the regulatory limit.
With Pool B filled wi'th 5% enriched fuel, the maximum k,,, would be 0.970 in the absence of soluble boron.
The soluble boron, usually worth about 0.20 Sk, would be more than adequate to compensate for any fuel mislocation accident.
2.3.3 Dropped Fuel Assembly The accident of a dropped fuel assembly of 5% enrichment and assuming the assembly came to rest on top of a filled rack module, would be more th;.a adequately accommodated by the soluble boron i,
i present in the pool water.
Nevertheless, the consequences of this accident have been evaluated in the absence of soluble boron.
1 In its final assumed position on top of the rack, the dropped ussembly is separated from the fuel in the racks by more than 12
)
inches.
A separation distance of 1? inches has for many years been
]
accepted as adequate to preclude any neutronic coupling, and this has recently been re-confirmed by critical experiments.
Thus, The reactivity of the system would be the same as that of the isolated assembly lying on top of the racks.
i The reactivity of a single fuel assembly suspended in clean water has a maximum reactivity of 0.9631rrespective of the storage rack.
Calculations of the system of a rack with a dropped fuel assembly on top (using KENO-Sa) gave essentially the same reactivi-ty as an isolated assembly of 5% enrichment.
Thus, the system
}
reactivity is determined by the single dropped assembly of 5%
j enrichment, irrespective of the rack loading or configuration. This confirms that the system with a dropped sssembly will have no l
higher reactivity than the reactivity of the single isolated l
assembly in water.
p i
6 l
[
I j r
.m,
...m.
3.0 ANALYTICAL METl!GDOIDGY In the fuel rack analyses, the primary criticality analyses of the Mgh densit.y spen 1 fuel storage racks were performed with a two-dimensional multi-group transport theory technique, using the CASMO-3(2) computer code and a Monte Carlo technique utilizing the NITAWL-KEHn-Sa computer package'3)
NITAWL was used with the 27-group SCALE
- cross-section library") and the Nordheim integral treatment for U-238 resonance shielding ef fects.
Benchmark calculations, presented in Appendix A, indicate a bias of 0.0000 with an uncertainty of i O.0024 for CASMO-3 and 0.0113 1 0.0017 (9S%/95%)m for NITAWL-KENO-Sa. Since the SCALE crossection library as used by NITAWL has scattering matrices only at 20' C and 277
- C, the NITAWL-KENO-Sa calculations were used only at 20' C and temperature corrections determined by normalizing to CASMO-3 cell calculations at the temperature of highest reactivity (70'C).
CASMO-3 has been beachmarked against critical experiments with water gaps up to 2.576 inches.
However, in a parametric study, it i
was found that there is an error in the CASMO-3 calculations depending on the magnitude of the vator gap and the number of mesh-points used.
At the 4.131 inch water gap in the Pool B storage racks, the error in the CASMO-3 calculation was found to be about 0.01 ok.
Despite this error, the CASMO-3 results were used to assure that the fincl evaluated k values were conservative.
g CASMO-3 was also used bath for burnup calculations and, f
where burnup limits were invo'ved, the calculated limits were i
increased by 5% as an allowance for uncertainty in the depletion k
r calculations.
In
- addition, the small reactivity increments
]f jl 5
associated with manufacturing tolerances obtained in the previous evaluation were assumed-to remain applicable.
Previous calcula-tions have confirmed a continuous reduction in reactivity with
" SCALE" is an acronym for Etandardized Computer Analysis for Licensing Evaluation, a standard cross-section set developed by ORNL for the USNRC.
I t
4
)i storage time (af ter Xe decay) due primarily to Pu-241 decay and Am-241 growth.
I j
In the geometric model used in the calculaticns, each fuel rod and its cladding were described explicitly in both the CASMO-3 and KENO-Sa models.
Reflecting boundary conditions (zero neutron current) were used in the radial direction which has tae effect of creating an infinite array of storage en11s. in X-Y l
directions.
In the KENO-Sa model, the actual fuel assembly length was used in the axial direction, assuming a thick (30 cm) water reflector.
i a
Monte Carlo (KENO-Sa) calculations inherently include a statistical uncertainty due to the rand:m nature of neutron tracking.
To minimize the statistical uncertainty of-the KENO-calculeted reactivity, a minimum of 250,000 neutron histories in j
500 generations of 500 neutrons each, were accumulated in each calculation.
l l
l l --.
4.0 REFERENCES
1.
Holtec international, "Licens.ing Report for Pool A
Reracking, Three Mile Island Unit I,
Report HI-89407, Nov. 28, 1989.
2.
A.
- Ahlin, M.
- Edenius, H.
- Haggblom, "CASMO A Fuel Assembly Burnup Program," AE-RF-76-4158, Studsvik report (proprietary).
A. Ahlin and M. Edenius, "CASMO - A Fast Transport Theory Depletion Code for LWR Analysis," ALLS, Iransactiona, Vol.
26, p. 604, 1977.
M. Edenius et al., "CASMO Benchmark Report," Studsvik/ RF-78-6293, Aktiebolaget Atomenergi, March 1978.
"CASMO-3 A Fuel Assembly Burnup Program, Users Manual",
Studsvik/NFA-87/7, Studsvik Energ!technik AB, November 1986 M. Edenius and A. Ahlin, "CASMO-3 : New Features, Benchmar-king, and Advanced Applications"', Nuclear Science and Enaineerina, 100, 342-351, (3988) 3.
R.M. Westf all, et. al., "NITAWL-S : Scale System Module for Performing Resonance Shielding and Working Library Production" in SCALE:
A Modular Code System for P3.II.52rmina Standardized Cououter Analyses for Idgensina Evaluation., NUREG/CR-02OO, 1979.
L.M. Petrie and N.F. Landers," KENO Va. An Inproved Monte Carlo Criticality Program with Supergrouping" in Scale: A Modular Code System for oerformino Standardized Comouter Analyses for Licensino Evaluation, NUREG/CR-02OO, 1979.
4.
R.M. Westfall et al.,
" SCALE:
A Modular Cods System for performing Standardized Computer Analyses for Licensing Evaluation," NUREG/CR-02OO, 1979.
5.
M.G. Natrella, Experimental Statistics National Bureau of Standards, Handbook 91, August 1963.
l 1
Table 1
SUMMARY
OF CRITICALITY SAFETY ANALYSES NEW FUEL STORAGE VAULT Optimum Flooded Moderation Type of Fuel Westinghouse D&W Moderator Density 100%
9%
Fuel Enrichment, wt% U-235 5.00 5.00 Reference k,,
(KENO-Sa) 0.9341 0.9371 U
Calculational bias, 6k )
0.0113 0.0113 Uncertainties Bias i O.0017 i O.0017 KENO statistics (95%/95%)
i O.0012 i O.0042 Lattice Spacing i O.0010 i O.0010 Fuel enrichment i O.0014 i O.0014 Fuel density 0.0019 i O.0019 Statistical combination i O.0033 i O.0050 of uncertainties (3)
Total O.9454 i O.0033 0.9484 i O.0050 Maximum neactivity (k.)
0.9487 0.9534 d)
Appendix A (2)
Square ~. root of sum of squares.
- l
_ _ _ _ _ - _ _ ~
i Table 2 DESIGN DASIS FUEL ASSEMBLY SPECIFICATIONS B&W Westinghouse FUEL ROD DATA Outside diameter, in.
O.430 0.422 Cladding thickness, in.
0.0265 0.0243 Cladding inside diameter, in.
O.377 0.5734 Cladding material Zr-4 Pellet density, % T.D.
95.0 Stack density, g UO,/cc 10.225 10.29 i O.21 Pellet diameter, in.
O.369 0.3659 Maximum enrichment, wt % U-235 5.00 i O.05 FUEL ASSEMBLY DATA Fuel rod array 15x15 Number of fuel rods 208 Fuel rod pitch, in.
0.568 Number of control rod guide and 17 inctrument thimbles Thimble O.D., in. (nominal) 0.530 Thimble I.D., in. (nominal) 0.498 1 _
__-__-__--________-_A
i 86 i
Table 3
SUMMARY
OF CRITICALITY SAFETY ANALYSES POOL A RACKS l
Region 1 Region 2 Design Basis burnup at O
42 MWD /KgU 5.% initial enrichment Temperature for analysis 4'C (39'F) 4'C (39'F) l Type Fuel Westinghouse B&W i
l Reference k, (CASMO-3) 0.9403 0.9097 U
Calculational bias, bk )
0.0000 0.0000 Uncertainties i
Bias i O.0024")
i O.00240)
B-10 loading 0.0021(2) i O.0026(2)
Boral width O.0007(3) i O.0009(3)
Inner box dimension i O.0012(3) i O.0016(3) l Water gap thickness i O.0036(3)
NA SS thickness i O.0008(3) i O.0001(3)
Fuel enrichment i O.0026(3) i O.0026(3)
Fuel density i O.0025(3) i O.0025(3)
Eccentric position i O.0025(3)
Negative (3)
Statistical combination i O.0067 i O.0051 l
of uncertainties")
l l
Allowance for Burnup Uncertainty NA
+ 0.0210 i
Total 0.9403 i O.0067 0.9307 i O.0051 Maximum Reactivity (k )
O.9470 0.9358 03 Appendix A (2)
As-Built Boral Panel Loading (3)
Rorack Licensing Report,.1I-89407, September 1990
")
Square root of sum of squares.
l 0
4
. 4
- m+
>g* ':9Xlm;0>02'as;:0!jij;:$:,:<!E:.v):.,.:s u sssp
- .!:is: sl%:.S< 'i:i23... i.:. p. ;...' ','F.v.
- m+" " ' "sg i+i
~
!s jij.::!O:?;;;4!! s 5 fT. THICK C0NCRETE
' 0$s9 '-
- 8A 20:u; 08:0R 26 M$i8!
.. t.s:p.:
.s a
y 0.4 c...
v.ss xs+
- s$0;0 39.6" 100 CM v.v
+w u.:.
p 4.............,
s20
...y 820
.::0.s::.,
s.,
n
.v
- . 4.:.
21.125" LATTICE
.'sv.
k
.<sv
- 89
p..y.:.
...o.
,,,;g:gi.:
...s.
.w,v..v...,:.
- M:pti:i i l
'M2;
..::pj!.is EM
' M Q
M Mi
- igiisjiji:,
x 3sv.81 o
- s. v
- .2.:.
- i:.Ijl5h,y 4 Nf 63IN.
f..z.,a....:
.+.s.~
- .f.p[*....my 160 CW
.u.
isij;p;g.s:
,e.sy y m:
.N.v,.v.
ys
,. A. s..,...
+v.v q a'
.. s,
.+.m..,
, sss.sv.
.:9: y -
... :>s.v.
v.s s
,.(s$+
%u. +s * *
.. sy
- v. s Oss:
's2 :
OR:
i i
IMM MEMQ@lj l ',,,
v>s
..s
.v.v.
- h. v. :..,
. :.vaM.
v...
- u.:.
s..
Sss:i mms,
Ri$!
- .s..aa:,
.v.
- %p
.. v.v
- a. +.a,.
v.v 8.g: <.. sis
+>O.v::-
s x :s v
- g. :g;e.:
so:e sa. v m.0:m :
- .:a..
O. 9.. g:a p.:su g
...........................s a
16" 40.64 CM.
.sv
.*1;@+i' v
3-
?
=
?
?
r Fig. 1 NEW FUEL VAULT CONFIGURATION
' C>
J f
4 3
1.00 t
y 0.9s _
g
- g g
.c I
d I
~
0.90 o
-l l
H m
l 0.85
/
g a
~
0.80 f
b 0.75
/
a D
[
~
D a.
q h
0.70-n u
0.65 O.,60 s
.7 e,
o
- i 4
s a r s,
10
1 MODERATOR DENSITY, gm/cc Fto. 2 REACTIVITY VARIATION WITH MODERATOR DENSITY 1 I
I I
- s.
J 7.500"
- 0.063" BORAL
.L_._._s_..
IN 0.088" i 0.003" GAP L
1 j--
i i
- ~
d G
0.081"* 0.004" THK 0.0211 g B-10/en.2 s
I q
q l
L.:h
~
- )
- .w:
[ b TUEL OD 0.3686 IN.
i i
l
! Oh DENSITY 10.225 g/cc i
j NUWBER 208 (15 X 15) 1 i OOG i
ctAo io 0.377 iN.
j
- OOOA n00 00 0.430 iN.
.l SS SHEATHING 0.024" i O.003" iOOOOQ PITCH 0.568 IN.
i i
! OOSOOk Taiustts
! + c'87 "
- 5" 00 0.530 iN'.i l
! 000000' ID 0.498IN
? 0000000 NO. 17
! OOOOOOOOb
- 004000OOOk i
WATER GAP
! OOOOOOOOOOb 1.697
- 0.06" E OOOOOOOOOOOk
! O000000000004 i
t i
<i --nnnt s
9.00 1 0.06.. BOX l.0". nns, i
l-49909xrrrrr@00%J-I l
5%_.. _ __
4/j N.075"
- 0.005" 0
s, 4
SS BOX 11.07 i 0.06" D
._r----0-------------------L---z'4--
i i
NOT TO SCALE I Fig. 3A CROSSECTION Of REGION 1 STORAGE CELL IN POOL A 1 l,
\\.
J
'e i
1 l
i j
7.500"
- 0.063" BORAL IN 0.106"*
0.003" GAP i
=
=
2 0.09 E"
- 0.004" ThK 0.026 g B-10/cm 3
g
.w.w.w
.u wz.w u u w xu u w M b FUEL 00 0.3686 iN.
k
" QQ DENSITY 10.225 g/cc i
NuueER 208 (15 x 15) 00b CLAD ID 0 sOOOk a00 00 0.'377 IN' 430iN.
E00004 PITCH 0.568 IN.
l 0.216"
- 0.006" 004OO i
000000 'TNiuetEs 00 0.498 iN.'
0.530iN s
10 s 0000000 NO. 17 00000000-
! 000000000k
! 0000000000b s 000000000004 s OOOOO9000000L l
6000( 9.00 : 0.0s~ sox to. )OOE j
i% _O L v v v v v v 0 0 0 0 0:,,, f.
OO
__ __ 4 u.
Jr N
S$EA HI NOT TO SCALE 0.075 5 80 LATTICE SPACING 9.216" i 0.06" Figi 3B CROSSECTION OF REGION 2 STORAGE CELL l
IN POOL A l
l l
t j _
i 5 l
4s.e
~
4e.e
~
O as.e 3
s ACCEPTAB_E BURNUP ae.e uviulN 25.e r
\\
2e.e p
15.e LINACCEPTABLE BURNLP DOMAIN 5
/
te.e G.e o
t !n e
a t t t t i I I f i t i e t i e a t i e e e e t t
- 9.s 2.e 2.s 3.e 3.s 4.e 4.s s.e t
INITIAL ENRICHMENT, WTs U-235 i
Fta. 4 MINIMUM BURNUP REQUIREMENTS FOR FUEL IN REGION 2 0F THE POOL A STORAGE RACKS 21 -
l t
i l
i
- 000005000500000 g i
l 000000000000000 a
i A
__m
~.
. _ __J i
i
._._.___._._._._._._._._._._._._._.1._._._._._
i I
l c
y y-
)OO '
- 000000000000000 OOC 20 0
! 000000000000000 i
j OOC 20 0.
9 000009000000000 l
l OOC 20 0
- 000000000000000 i
OOC 20 0 -
000000000000000 i
OOC DOO i
l 009009000000000 l
- 0 04
)OO i
- 000000000000000
!_ 4"
. s OOC 20 0 i
l 000000000000000 i
WAR $ CAP OOC 20 0 i
- 000000000000000 i
i OOC 4
DOO i O00000000000000
- 0 04 20 0 000000000000000
- OOC 20 0 i
s O00000000000000
! OOC 20 0 000000000000000 l OOC 20 0 i O00000000000000
! OOC 20 0 s 00000000000000
- OOC w_ _ - --
_0 l
_4 s
l*
Wsm I
i 0.187" SS BOX 9.12" 1.D.
j
._._.__._j___._____._____._.___._._._.___._._._._.q._._.___._.
i i
i l
7____________y i 000000000000000 i
000000000000000 l
i nnnnnannnannnnn e
i Fig. 5 CROSSECTIM 0F POOL B STORAGE CELLS l 5
6 i
e 2.5 3
1 3
3 2.0 N
s'
/
k 1.5
/
5 d
~
a.
tu j
1.0 1
UD O
0.5-0.0 ii d ii 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0 INITIAL ENRICHMENT, w tss U-235 Fto. 6 MINIMUM BURNUP REQUIREMENTS FOR STORAGE OF. FUEL IN POOL 8, -
i
t.'
ATTACHMENT 2 TABLE 1.
RADIOLOGICAL CONSEOUENCE FOR FHA & 8FCD2 2-HR 8.B.
DOBE (REM)
ACCIDENTS
_.. THYROID (*)
W.B.
DOSE (*)
FRA in FHB (1) 0.94 0.078 (1.28)
(0.26)
FHA in RB (2) 56.0 0.46 (62.3)
(1.31)
SFCDA (3) 2.61 0.0506 (3.38)
(0.051) 10CFR 100 LIMIT 300 25 llQIEEi
~
(1): FHA in FHB =
Fuel Handling Accident in the Fuel Handling Building (2): FHA in RB Fuel Handling Accident in the Reactor
=
Building Spent Fuel Cask Drop Accident (3): SFCDA
=
Numbers in parentheses are FSAR values.
(*): (xxx)
=
__.___m__---_.--__2