ML18043A345
| ML18043A345 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 12/18/1978 |
| From: | Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Ziemann D Office of Nuclear Reactor Regulation |
| References | |
| TASK-09-01, TASK-9-1, TASK-RR NUDOCS 7812260206 | |
| Download: ML18043A345 (33) | |
Text
consumers Power company.
General Offices: 212 Wesi Michigan Avenue, Jackson, Michigan' 49201
- Area Code 517 788-0550 December 18, 1978 Director, Nuclear Reactor Regulation Att:
Mr Dennis L Ziemann, Chief Operating Reactors Branch No 2 US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT - PROPOSED TECH SP~C CHANGE REQUEST - FUEL STORAGE Attached are three (3) original and. thirty-seven (37) conformed copies of a request for a change to the Palisades Technical Specifications.
This.change is necessary to allow the storage of 3. 27% enriched fuel in the new and spent fuel racks.
It is requested that this change be issued prior to February 1, 1979.
The requested changes involve a single issue.
Therefore, a check in the amount of $4,ooo is attached pursuant to-*10 CFR-*170.22 - Class III change.
David P Hoffman Assistant Nuclear Licensing Administrator CC:
JGKeppler, USNRC REGULATORY DOCr\\ET FILE COPY 781.2260~
~' -
.t' CONSUMERS POWER COMPANY Docket 50-255 Request for Change to the Technical Specifications License DPR-20 For the reasons hereinafter set forth, it is requested that the Technical Specifications contained in Provisional Operating License DPR-20, Docket 50-255, issued to Consumers Power Company on October 16, 1972 for the Palisades Plant be changed as described in Section I below:
I.
Changes A.
Change Technical Specification 5.4.l.a to read:
"a.
Unirradiated fuel may be stored in the new fuel storage rack which is designed to ensure an effective multiplication factor of less than 0.95 under the worst credible conditions for fuel enriched to 3.30 weight percent U-235."
B.
Delete the word "also" from Technical Specification 5.4.l.b.
C.
Change Technical Specification 5.4.l.c to read:
"c.
New fuel enriched to 3.27 weight percent U-235 may be stored in the poisoned high capacity racks which are designed to en-sure an effective multiplication factor of less than 0.95 when flooded with unborated water."
D.
Change Technical Specification 5.4.2.e to read:
"e.
The fuel placed in the spent fuel pool and stored in the poisoned high capacity storage racks shall not contain more than 40.12 grams of U-235 per axial centimeter of active fuel assembly sub-ject to a maximum assembly average loading of 3.27 weight percent U-235.
The fuel placed in the spent fuel pool and stored in the unpoisoned lower capacity racks shall not contain more than 38.3 grams of U-235 per axial centimeter of active fuel assembly, sub-ject to a maximum assembly average loading of 3.05 weight percent U-235."
II. Discussion The purpose of this Technical Specifications change is to allow storage of 3.27% U-235 enriched fuel in the new and spent fuel racks.
The technical bases for these changes are the attached NUS report entitled "Criticality Analysis for 3.27 W/O Enriched Fuel Palisades High Density Fuel Rack" and "Palisades Site New Fuel Storage Array Criticality Safety Reanalysis."
In addition to this, the new fuel storage racks have been modified such that Technical Specification 5.4.l.a no longer has'significance.
l
-'---- - III.
Conclusion f
Based on the foregoing, both the Palisades Plant Review Committee and the Safety and Audit Review Board have reviewed these changes and find*
them acceptable.
CONSUMERS POWER CO~IPANY By Sworn and subscribed to before me this 18th day of December 1978.
r:nda: R Thayer~Nary Public Jackson County, Michigan My commission expires July 9, 1979.
2
5.3 5.3.2 NUCLEAR STEAM SUPPLY SYSTEM (NSSS) (Contd)
Reactor Core and Control
- a.
The reactor core shall approximate a right circular cylinder with an equivalent diameter of about 136 inches and an active height of about 132 inches.
- b.
The reactor core shall consist of approximately 43,000 Zircaloy-4 clad fuel rods containing slightly enriched uranium in the form of sintered uo2 pellets.
The fuel rods shall be grouped into 204 assemblies.
A core plug or plugs may be used to replace one or more fuel assemblies subject to the analysis of the resulting power distribution.
- c.
The fully loaded core shall contain approximately 211,000 pounds uo2 and approximately 56,000 pounds of Zircaloy-4.
Poison may be placed in the fuel bundles for long-term reactivity control.
- d.
The core excess reactivity shall be controlled by a.combination of boric acid chemical shim, cruciform control rods, and mechanically fixed boron rods where required.
Forty-five control rods shall be distributed throughout the core as shown in Figure 3-5 of the FSAR.
Four of these control rods may consist of part-length absorbers.
5.3.3 Emergency Core Cooling System An emergency core cooling system shall be installed cons.isting of various subsystems each with internal redundancy.
These subsystems
- shall include four safety injection tanks, three high-pressure and two low-pressure safety injection pumps, a safety injection and refueling water storage tank, and interconnecting piping as shown in Section 6 of the FSAR.
5.4 FUEL STORAGE 5.4.l New Fuel Storage
- a.
Unirradiated fuel may be stored in the new fuel storage rack which is designed to ensure an effective multiplication factor of less than 0.95 under the worst credible conditions for fuel enriched to 3.30 weight percent U-235.
5-3 Amendment
,r
--5-.4 FUEL STORAGE (Contd)
- b.
New fuel may be stored in shipping containers.
- c.
New fuel enriched to 3.27 weight percent U-235 may be stored in the poi-soned high capacity racks which are designed to ensure an effective multi-plication factor of less than 0.95 when flooded with unborated water.
- d.
The new fuel storage racks are designed as a Class I structure.
5.4.2 Spent Fuel Storage
- a.
Irradiated fuel bundles will be stored, prior to off-site shipment in the stainless steel-lined spent fuel pool.
- b.
The spent fuel racks are designed to maintain fuel in a geometry which insures an effective multiplication factor of 0.95 or less with new fuel flooded with unborated water.
- c.
The spent fuel pool water boron concentration shall be verified at least once monthly to be equal to or greater than 1720 ppm.
- d.
The spent fuel racks are designed as a Class I structure.
- e.
The fuel placed in the spent fuel pool and stored in the poisoned high capacity storage racks shall not contain more than 40.12 grams of U-235 per axial centimeter of active fuel assembly subject to a maximum as-sembly average loading of 3.27 weight percent U-235.
The fuel placed in the spent fuel pool and stored in the unpoisoned lower capacity racks shall not contain more than 38.3 grams of U-235 per axial centimeter of active fuel assembly, subject to a maximum assembly average loading of 3.05 weight percent*U-235.
- f.
Spent fuel shipping casks shall not be moved in the fuel storage building until such time as the NRC has reviewed and approved the spent fuel cask drop evaluation.
- g.
Fuel stored in the higher capacity storage racks as described in t.he SER supporting Amendment No.* 28, shall have decayed for a minimum of 12 months if the storage racks are not supported by similarly designed, adjacent rac~s and the spent fuel pool wall or the cask anti~tipping device.ll)
References (l)Until needed for fuel storage, two A-type racks in the northeast corner of the spent fuel pool will be removed and replaced with the cask anti-tipping device to provide necessary seismic restraint.
FSAR, Appendix A.
FSAR, Appendix B.
5-4 Amendment
~
....... 1.........
PALISADES SITE NEW FUEL STORAGE ARRAY CRITICALITY SAFETY REANALYSIS
- Drafted by :-=---=--=-~-=~-7,
,'---.. _~_'l-1_1 __
/._..,_., r_R-1--~ef._,.. _1,r __
C. 0.
Bro~m Prepared & Accepted by]l...- il\\_.::.~ __ :(__,\\---u-.( :/'-:--' n /J Manager, licensing & Compliance EiJ{ON ft\\JUCLEAR COMPANY, Inc..
- 9 INTRODUCTION The design enrichment of Palisades reloaJ l3atch II \\-Jas i11c1*eased in order to provide capability for extended,burnup beyond the present design criteria governing the specification of reloads E and G.
A reanalysis of the new fuel storage pool facilities at Palisades has been performed, to confirm that criticality criteria will still be*
met at the higher enrichment, as proposed in Task 2.6 of Palisades Batch H Extended Burnup Redesign and License Evaluation Project.*
The results of the rritirlllit_v rriln;'ll_v~ic; nrr hrirHJ r*f*JiPrtf'il 1'1'iP1' le*
the completion of the project *for the convenience* of Consumers Power Company.
The criticality reanalysis is independent of the remaining mechanical, reactor safety limit, and cost analysis required to com-plete the project.
At present the Palisades new fuel storage array may store Exxon Batch "E" or similar new fuel assemblies having a maximum average enrichment of 3.2 wei~ht % 235u.
This report summarizes the results of a reanalysis of the storage array for the storage of 3.27 weight %
235u (ave.) Exxon Batch "H" fuel*assemblies \\'lith the burnable poison rods removed.
SUMMARY
The following criticality safety analysis of the new fuel storage array, containing steel box beams in alternate storage locations as described in XN-309, demonstrates the array to be adequately subcritical for fuel assemblies at 3.3 weight% 235u.
FUEL ASSEMBLY DESCRIPTION
. The Exxon nilt<:h "II" ftJP.l ilS')r~rnbly desirJn i<;
d1~pir:tr:d HI ri<jtff(! l.
As indicated, this 15xl5 lattice arrangement includes a single instru-ment tube, eight guide tubes for poison rods, and eight guide bars.
The remaining positions within the assembly are occupied by 208 uo2 fuel rods.
- Letter to W. A. Walker from R. K. Robinson dated April 21, 1978, Palisades Batch H Extended Burnup Redesign and License Evah1ation (RJE:l02:7R).
e e
~
FIGURE 1 PALISADES (EXXON BATCH 11H 11
) FUEL ASSEMBLY 0 0 O-lzJb 6 0 0 O-r!1-0 0 00
. "88818 g~g-gg*gg 8[88 g
- ,.,k,,,.., no 1~,, "...., "),,..) :""" 1*
~} '--' '-" '- I._.
I 100 0 0 000000 O<]C L'tl
'.Ctiocooooooooceoo 1010 0 0 0 0 0 0 000 0 0 c,o 0,000000000000010 10:00000 000 00 oo (_)IQ t o~*o o.* o o o o o o o o C) e ul 01:
00000000000001~
j o_oooooooooorQQIO
,o 06100 0 000~0 o,o 00
- o oolPoo o ooC>o 0.000 iO 00 O~OOOOQ~QO 00
!c___
---~~~
0 I
=
Ins trumcnt TubP 0
= Poison Rod Guide Tube m *= Guide Bar 0
= Fuel Rod 235 Inner envelope - 3.43 wt.%
U 2 '>5 Outer envelope - 2.90 wt.%
~ U
\\
~
The fuel assembly specifications and the lattice cell parameters assumed in this evaluation are given in Table I.
The bundle-averaged cell parameters were calculated by including the zi~conium associated with the instrument tube, guide tubes, and the guide bars in the zirconium clad of each fuel rod.
Water associated wJth each guide bar, instrument and guide tube was included by increasing the unit cell dimension (rod lattice pitch). Such*assumptions permit a conser-vative estimation of the effect on reactivity of the extra zirconium and water within the fuel assembly.
CALCULATIONAL RESULTS Values of k were calculated using the CCELL(l) computer code O>
for 3.3 wt.% 235u enriched uo2 rods as described in Table I.
Results for both the nominal and bundle-averaged lattice cell parameters are give*n in Table II alon9 with results 1>reviously reported i11 XN-J09.
Comparison of the bundle-averaged k values for the two enrichments O>
shows a +0.013 6k change per l.16 g/cm increase in fuel assembly 235u loading (axi:l).
The maximum keff of the storage array as discussed in XN-309 occurs when the array is fully flooded and at the 95~ confidence level is calculated to be 0.928 (abnormal or worst case condition).
To esti~ate the maximum keff of the storage array with the higher enriched Exxon Batch "H" fuel assemblies, the reactivity bias calculated between
'fuel assemblies is adde~ to the maximum array keff value.
Estimation of Storage Array kef~ (max.)
0.928 max. previous array keff
+0.013 enrichment bias, Ak
().)
0.941 max. estimated new array ke~f
~
~-
.9 TABLE I PALI SADES (EXXON BATCH "H")
.. FUEL ASSEMBLY PARAMETERS Lattice Pitch, in.
Clad OD, in.
Clad Material Clad Thickness, in.
uo2 Pellet Diameter, ini Pellet Density, (%TD)
Ave. Enrichment, (wt.~ U-235)
. Active Fuel Rods Rod Array Effective Lattice Dimensions, in.
Poison Rod Guide Tubes (Zr-4)
Instrument Tube (Zr-4)
Guide Bars (Zr-4)
- Approximate dimensions.
NOMINAL LA TTJ CE r.n I P/\\fV\\MrHl~S 0.550 0.417 Zr-4 0.030 0.350 93.2 3.3 (specified) 208 1.5 x 15 o.25 x 8.25 0.416 11 0D x 0.013" wall 0.415"0D x 0.029" wall 0.398 11 x 0.450 11 13UNDLE-/\\VERAGED CTI I PJ\\fMMf rn:*,
0.569 0.422 Zr-4.
0.032 0.350 93.2
. 3.3 208 15 x 15 8.25 x 8.25 N/A N/A N/A
TABLE I I..
INFINITE MEDIA MULTIPLICATION FACTORS PALISADES (EXXON GATCH "E" AND wH") FUEL ASSEM8LIES Axial 235u Lattice Cell Parameters
- Loading, Nominal 8undle-Averaaed Case Enrichment g/cm k
ilk k
ilk a>--
ClO--
a>--
ClO--
(Base) 3.2 Gatch *"E" 37.20 l.401
- 1. 402 2
3.3 Batch "H" 38.36 1.409
+0.008 1. 415
+0.013 NOTE:
In comparing the 8atch "E" and "H" infinite multiplication constants, it should be noted that in all cases no credit is taken for burnable poison.
In the Batch "E" fuel assembly, burnable poison rods are_ replaced by fuel rods (see XN-309) and for the 13atch "H" fuel ar,sembly the p~ison rods are
- removed.
These assumed conditions tend to maximize the fuel assembly k values.
Q)
*-:-~****"--* -----* ---. --~~~~-----
,......,.,.,.~,,,.~--..===*.f'C"r."*"""-**5-~*-*:!'*~,**H....
fl'* I
~
Hence, the maximum estimated*keff of the new fuel storage array under the worst credible conditions for fuel enriched to 3.3 wt.% 235u is 0.941.
CONCLUSIONS It has been demonstrated above that the Palisades new fuel storage array as described fo XN-309 may store Exxon Batch "H" fuel assemblies at a maximum enrichment of 3.3 wt.% 235u and continue to meet criticality safety requirements.
Namely, under postulated abnormal or \\~orst credible c6nditions as outlined in XN-309, the effective m11ltiplic~tinn ~nn~t~nt of the storl\\~fe tln*;iy 1*1ill re111ai11 bel01v the li111iting value of ll.~l~l.
The keff of the ne1'/ array is estimated to be 0.941, which is well within the established limit.
.,~
REff'ERENCES
- l.
W.W. Porath, "CCELL Users Guide, 11 BNW/JN-86, Battelle Pacific Northi,ies t. Laboratories, February 1972.
CRITICALI'IY ANALYSIS FOR 3.27 w/o ENRICHED FUEL PALISADES HIGH DENSITY FUEL RACK by P. Soong' July 1978
- I.
- 1. 0 INPUT INFORMATION Table 1 presents the specificution of high enrichment fuel to he used !n the Palisades reactor. The enrichment distribution in a fuel bundle is shown in
\\
F!gure 1.
The rack drawings on wh!Ch the criticality analysis was based are as follows:
5097 M 2000'Re\\lision S 509.7 M 2001 Revision 5, 3. sheets 5097 M 2004 Revision 3 5097 M 2005 Revision 3 5097 M 20Q7*Revision 4.
5097 M 2008 Revision 4 Figure 2 shows the poison cans to be installed in the main pool, while Figure 3 shows the poison cans to be installed in the tilt pool.
I
~*
y
~
Table 1.
Fuel Data MECJIMl I Cf..l o::s I G::
r:;.P..~~*~ETEP.S PJ\\LJ SADES FUEL FUEL ROD
. Fuel f*~ teri al Fuel E:irich.-::ent, w/o Pellet Diameter, (in:)
Dish Volu~2 Per Pellet (Total %)
- Pellet Length, (in.)
Pellet Density, (g/cc)
Pellet Density, (3 TD)
Clad l*~terial Clad ID, (in.)
Clad 00, (in.)
Clad Thickness, Nominal, (in.)
Diametral Gap, Cold Nomi.!lal, (in.)
Active.Lef}gth,. (in.)
Total Rod length, (in.)
Fill Gas and Pressure Fuel Weights:
uo2, kgs/rod U-235, grc~s/rod
?:u1:1ber of Active Fuel Rods per UOz S~ndle PALISADES Fl!El (c E..)
uo2 Sintered Pellets.
2.73 (Region 0) 0.359 2.0~
0.600 10.25
.93.5 Zircaloy-4 0.3655 0.4135 0.024 0.0065 132.0 140.5 He, above at~ospheric 2.13 51.3 Exxc:~ i:UCLEft.R FUEL (e.:;....;.. o;.)
uo2 Sintered Pellets 2.822 (Lead Assewblies~*
0.350 2.0~
- o. 274' 10.30 94.0.
Zircaloy:-4 0.3580 0.4150 0.0285 0.0030 131.8 139.452 He at 250 psia 2.0965 55.63/44.35/34.56 216 Fue 1 P.od Array, Square 216 lS*x 15 15 x* 15
. "'-f\\*:er*29~ enrir:h:.:ent, ccns1sts of: 10 pins ct 1.87 \\*:/o, t;S (3 2.~0 \\*.*/o er.::
l :, :, (; 3. C: 1 *,: / c.
I FUEL P.CD *~:~ntinued)
. __ _:__fuel Rod r itch, (in.)
Spacers
. Type
~-*,.
Number Pc::- ;;ssembly flumber*w1t~in Active Fuel Guide. Bar :~ teri a 1 Guide Bar :*'.?xir.iurn Dimensions (in.)
Guide ~ar L~~gth (in.)
Instrument:~ion Tube Material Instrumen~=tion Tube Dimef)s i ans (in.)
~
Instrume~t~:ion Tu~e End Fittin; :*:aterial Spacer O:its".de Dimension ( ~ n.)
Fuel Asse:"".:!y Pitch with Cruci".'-:*:7.1 Control Blade {in.)
Ftiel Asse:7'.:~... Pitch Without Cr~:~form Control B1~=e (in.)
Fuel Data (continued)
PALISADES FUEL 0.550
- leaf Spring,
.. Z_i rca l oy-4
- 8
\\
/
8 Zircaloy-4 0.44 x* 0.44 140.652 Zircaloy-4 0~4135 OD x 0.3655 ID Zircaloy-4 8.181
- 8. 61.4 8.355 Maximum As::;::-:bly En*:elope 8.336 l~ngth Bet~~2n Tie Plates, 190.652 (in.)
Total Assc::-::iy Length, less Lo~cr :~e Plate 147.646 Aligning r~~, {in.)
..3*
e
- EXXOH
~UCLEAR FUEL 0.550 Zi rca 1 oy-;-4 with Inconel 718 Springs 10 9
2'ircaloy-4
. 0.450 x 0.398 140.752 Zirca1oy-4 0.415 OD x 0.3580 ID Z{rc;~loy-4 8.195.
8.614 8.355 s:324s 140.652 147.432
Fuel Enrichment Distributit notes:
- 1.
- 2.
64 type 6 rods @ 2.90 w/o u235 (lo~ enrich~ent) 144 tJ~e 5 rods @ 3.43 w/o u235 (high enric~~ent)
The pellet densities are 93.0 ! l 1/2~ T.D.
- 3.
- 4.
The guide tubes (8) are non-fuel locations to accorrnocate a poison (re~ovaole) cluster.
The instn;.ment tube (1) is for an i*
a L ncore e... ec... or.
d~sign is.identical to relcacs E, F, and G.
- 6.
- Clad thic'kness:
O. 72 ~
Ef£. Rod Ler.;th:
3348 r:.~
Rod 0.D:
10.54 rr:::i He ~ill gas press~re: 20 a~~
Othe:-wise., mech.anicall~t tr:~
- lG UF\\R E i\\t1\\N F'KEE P.t>.T H i I
"/
A A
I\\ ~
PT I
T'YP I i--'--- 4 D 2 8
---t"->t
~s. 5 Co+/-->O<a SG.Uf\\~E lNS\\DE AI.
ALL CROSS SE.Cl IONS 5
n 5 Cc.+/-. o OCO SG-uAR. ~
Figure 2.
Poison Can, 10.25 x 10.25 inch Spacing
. 5r-r'JO"'
t-- I I
B--~(GR.l)
- *:. SC f\\ Lf:
I,, *- _,-, "'
r-
J
.~
j~ c:.
I I
9r-TY f..>.
I..
I 7.189 J---1;
. I TYP
~- 3.595 r>-!
i
~---* -
!I
- 1 Iii j:.06 SQUARE INSIDE AT
__,.~
r.s--~
ALL CROSS. SECT/Oi~S
~------- 10.00 !.OG. SQUARE --------\\::'(
Figure 3
- Poison Can, 10.69 x 11.25 inch Spacing r--'I C' " "
t-~ -
.--~ I (:., R L..__.J S.:ALE :C.'-:::.1'- 0" I
p
. E!
4
4
- 2. 0.
GENERAL METHOD OF ANALYSIS The preliminary design caJcula tlons were performed us in; PDO-7 (1) wi~h
... (2}
(3}
NUMICE-2 cross sections. The latter ls NUS' versicn of LEOPARD.
These are common tools used in L V'/R lattice.design and have been extem-slvely test~d for accuracy. The LEOPAP.D code has been validated against (4) a large number of critical experiments.
The KENO (S) code in conjunction with 123-group. AMPX (G) averaged cross sections was employed for the final design calculations. KENO is a multi-group Monte Carlo criticality program. It traces the life history of a large number of neutrons as they travel from one point to another in a system.
Infonnation based on collision of these neutrons with the materials in the system is used to detennine the criticality condition of the system.
AMPX ls a modular code system for generating multi-group cross sections in a format suitable for KENO input. NITAvVL, one of the.subroutines in AMPX, performs the Nordheim integral treatment for a medium which contains resonance *absorbers. The calculation in the resolved resonance region, plus a rational approximation in the unresolved, prepares fine group cross sections. These are merged with smooth cross sections to form a complete fine group set for each resonance absorber.
KENO has also been benchmarked. ORNL (7) performed calculations, using KENO and 123-group cross sections, against a number of critical experi.-
ments which involve arrays of clusters of 0. 762 cm diameter fuel rods of 5% enriched uranium.
B~ varying the spacing between clusters and by 7
varying the number of rods in a cluster, they stt.:dled theJeffect of diiierent self-shielding appro:-::imations for lT-238 resonance in the multi-group sets
\\
of cross sections. For the ten critical experimen:s all with water reflec-tors, the average calculated keff was 0.9914 :'.: 0.0020. Based on this Information, a bias factor of.Ck = + 0. 0086 is established to cover defi-ciency in calculational methods and microscopic cross section data in the master library.
NUS completed several benchmark problems including one of the Yankee it.1 l
t (9) d... w t". h "t"
.1.
{9}
er ca expsnmen s an
~v10 es ing ouse en ica experiments, us1ng the KENO Monte Carlo method d,escrib~d above. Our results indicate that the above bias factor is conservative.
KENO is a versatile code; however, it is often inefficient in detecting small reactivity changes in sensitivity studies.* In such cases, PDQ was utilized even in the final design phase and results were normalized to the base KENO case *
-*- ~--
3.0 (l)
I I I I
.e MAJOR ASSUMPTIONS
_.::Jl fuels are assumed to be fresh and unirradia ted.
(2)
- -:om!nal water temperature is taken as 68°F, but the final result is 0
adjusted for the maximum water density at 40 F.
(3)
Xo reactivity credit is given to the soluble boron in the spent fuel storage pool, exce'pt in a~cident cases.
(4)
~~o reactivity credit is given to burnable poison and control rods (if any) in the fuel assemblies.
(SJ
- 'he rack ls infinite ln size in all three dimensions.
(6)
No credit is taken for neutron absorption by structural materials other than those associated with the stainless steel cans and the enveloped B 4 C plates...
Note U:at because of Assumption (S), there L; no difference in physics interp:-etatlons of ~ of the storage lattice cells and keff of the entire rack syster:.:. These two terms are used interchangeably in this report.
,_I
- J I
.9 e
I 4.0 I
RESULTS The procedure of the analysis was as follows:
Determination of the rack k= at nominal conditions Determination of calculational uncertainties Determination of the sensitivity of k=> to design variables of interest for licensing input*
e Evaluation of the effect of pool conditions and mechanical design variables on k=
The res*ults are presented in this order in the following sections.
4.1 Nominal Rack Analvsis The storage cell in the rack is depicted schematically in Figure 4_. Its principal features are:
Fuel Assembly Poison Can inside Dimensiqn Can wall thickness B4C Slot thickness B4C Plate thickness Storage Cell Spacing
/0 See Table 1 and Figure 1 8
- 5 6" (main pool) 9
- 00" (tilt pool)
= 0.125"
= 0. 250"
=0.210" I 0
- 2 5 11 x 10. 2 5 11 (main pool) 10.69" x il.25" (tilt pool)
'"rj IQ c:.
fll
,fl,
(/)
..+
0..,
OI IQ m
~
..+
..+
0
(/)
ro
'O 0
OI 0
m
- i 10 r\\.>
~~
I>
'/
- f.
i.* ',*.:.,._.;L...,....,,..,..,.
- r,-.
,"JI',
- t.
-=.:.
-lITT,.
..(\\
I I I
~
---~~ :=- ___ -. - -~ __ ff_
-~
-1 1----1---_., ______ ---~*-.* -- --***-- --
.,__~v~*--~~**---"~~~-~1--~-1----
t~
- ~*--*------- -*-
-.t\\
- -l'
- l.,
-t I
- --*--**--- -------- *---*. -----.. ***-- -***I 1*
~
.:)
1-----11-----+*----~---- *---**----* ----*
1..)
-fl
- t' I, \\
I
~
..... ------------* --**-** ---** *--****---~;_:*.~. *-... _..
-~*--=---half ~pacing
- °
,I
.r
- 0 The nominal water temperature was chosen as 68 F ar:= the enrichment.
3.43 and 2.90 w/o U-235, fresh and unirradiated.
The rack was analyzed with NUMICE-2/PDQ-7 and KENO-IV. The k= at nominal conditions is as follows:
Nominal Storage Can Lattice Cell Spacing IO/OD PDO k=
KENO k=
Original 10.25" x 10.25" 8
- 5 6 ~ /9
- 5 6 "
.8490
.. 8517::.0043 Analysis 10.25" x 11.25" 9.0/10.0
.8580
.8629::.0045 This Work 10.25 x 10.25 8.56/9.56
.8569
.8693::.0042
- 10. 69 x 11. 2 5 9.0/10.0
.8397
- 8 51 6:::
- o o 3 4 The original analysis introduced a 6.k= of +.015 for modeling deficiency and
+. 0053 for Zr bar penalty. The latter is not required in this work, because Zr bars were discretely represented in the KENO geometry. The nominal k=
now becomes:
Original Analvsis This Work Main pool Tilt pool Main pool Tilt pool rack i;ack rack rack KENO cell k=
- 8517,
.8629
.8693
.8516 KENO modeling
.0150
.0150
.0150
.0150 Zr bar penalty
.0053
.0053 Not required
.8720
.8832
.8843
.8666 Taking into account a ~km of
- 0040 for 8 4c particle self-shielding effect, the KENO kc:cfor the storage lattice cell becomes 0.8883.
- 4. 2 Calculational Uncertainties Two types of uncertainties were considered:
- t.
,J Benchmarks KENO statistics An uncertainty of 0. 0086~.k:> is assigned to the KENO results to account for the KENO benchmark against critical experi:::wnts.
A Akm uncertainty equal to 2cr is assigned to'* the KENO k::;) for the results to be at 95% confidence level. In this analysis I the 2J is 0. 0084 6_k::;).
Therefore, the km of the reference rack including calcula tional uncertainties is as follows:
KENO (nominal rack}:
Benchmark:
Statistics:
km (Nominal; 95% G. L.):
4.3 Sensitivity Analvsis 0.0086 0.0084 0.0170
.0.8883
- 0. 9053 (rounded to 0. 905)
The following km sensitivity calculations were performed for the rack with PDQ-7:
km vs. Pool Water Temperature k= vs. Enrichment km vs. Boron Loading e
k= vs. Storage Cell Spacing These calculations assume the rack at nominal conditions (base case) except for one parameter which is varied around the nominal value to determine its
/3
i!.
effect en ka:.. The results of these sensitivity analyses are presented in Table 2.
4.4 Pool, Materials and Mechanical Tolerances Analvses r
These analyses were performed to determine 6.k= penalties for the war~~
possible off-nominal *situations regarding storage. pool water temperature, material properties and mechanic.;al tolerances. The following worst situations were considered:
Most Reactive Water Temperature Materials Enrichment at nominal plus 2% of the nominal value.
Mechanical Design Tolerances Eccentric fuel loading in the storage can One of every 100 poison plates missing Variation of poison can dimensions Spacing \\.~ria tion Non-straightness of poison cans Variation of B4C loading and thickness Based on the sensitivity analyses, the most reactive iNater temperature is 40°F and the Ak= penalty corresponding to the temperature change from 68 to 40°F is +. 002 8. In normal storage conditions, the pool water temperature is approximately 140°r. The actual~ km will be negative at the operating temperature, however, no credit is taken for this negative reactivity change.
Based on the sensitivity analyses, the penalty is +. 003 7 6k::o for a possible enrichment variation and +.0006 for possible change of uo2 density.
./q
\\*It *-*
e i
I I I TABLE 2 kc:o SENSITIVITY RESULTS Ak=
- a.
k= vs. Temperature 0
68
- F (base case) 0 40
+.0028 100
-.0039 212
-.0376
- b.
k= vs
- Enrichment
- 3. 2 67 (base case) 0
-3.332
+.0037.
3.201
-
- 0037 k= vs. Boron Loading 50% (base case) 0 45
+.0035 40
+.0074 k= vs. Storage Cell Spacing 10.25" x 10.25 11 (base case) 0 10.31x10.31
-.0076 10.19 x 10.19
+.0081 10.13 x 10.13
+.0165 l0.07xl0.07
+.0253 I.:;;-
ii
-~
.1
\\ !
l
-V I
- I The ec~entric fuel loading situation assumes that four fuel assemblies are I
clustered to a common corner and touc!1 the poison cans; this effectively represents reduction of the *spacing on one side and increase of the spacing on the other. This situation was analyzed with PDQ-7 at nominal conditions.
The net ~ff ect is -
- 0019 ~kca for which no credit is taken.
The poison cans will be inspected during fabrication to assure that no B 4c plate is missing. However, should one B 4c plate be missing out of 100 plates in a 5 x 5 storage cell zone, there will be a O.OCH8 Akca increase as calculated with PDQ-7.
The Ak= penalty due to the worst mechanical tolerances and variation of.
m.aterial composition was determined based on PDQ-7 results. An itemized list is given as follows:
Enrichment Variation Fuel Rod B4C Slab Width B 4c Slab Thickness and B 4 C Loading.
Variation of Spacing Can Dimensions B 4c Slot Thickness B 4c Slab Missing (1/100)
Bow and Twist Can Wall Thickness
~ot mean square sum Akca
.0037
.0006
.0008
.0038
.0081
.0097
.0045
.0018
- . 0165
.0061
.0228 In view of the random occurrence of these changes, it is more appropriate to take the root mean square sum than the arithmetic sum of all ti_k=> values to represent the combined net penalty. Thus, the net penalty for mechanical tolerances is equal to 0. 0228.
The worst tolerance rack k f; becomes 0.9309 (or rounded to O. 931).
e !
... -/~
'!I' J
~"*~*
4.5 Fuel Handling Accident Analvsi s The soluble boron which is usually present i!1 a P\\VR spent fuel pool w!ll cause a substantial reduction in pool reactivity. *Based on ci.l::- ?ast expe-rience in similar rack design work, the worth of soluble !Joron is estimated to be as follows:
Soluble boron ppm 1900 4000 Estimated
!\\ kco
-0.24
-0.40 In the presence of soluble boron in the pool, the reactivity of the pool will be so low that none of the fuel handling accidents as postulated in the original analysis file G-RA-06 will threaten criticality safety *of the rack_
- system *
- 4. 6 Sumrnarv of Results The rack system is so designed that it will remain subcritica1 and will not exceed the keff limit of 0. 95 under any storage conditions. Table 3 sum-marizes the results presented in *Section 4.1 through 4. 5 above. The maximum rack keff is 0.931, at 95% confidence level, for the worse possi-ble conditions. Note that the quoted maximum rack keff value represents, in fact, k::o of an infinite array of storage cells with infinitely long active fuels *
/7
'... *',......... "*.~**-*****"*~...... -*~... ***.... ***~.......,~******-.. t;.--.......................,,........ _
..., __ ~
~* -
i.*~
TADLE 3 OF CRITICALITY ANALYSIS RESULTS
(, *
. al File G--RA-06 Thls Work i
.1 pool rack kco Tilt p9ol ruck kco Most rea.ctlve ruck 1-.*..
- st Worst
- worst rnces
- Accident Tolerances Accident Tolcr.unces Accident 720 0~8787 0.8832 0.8899
.8043
.* 0090
.0084 e,
.0040
.0040
.0086
.0086
.0216
.0210
.9053 B2 0.8999 0.9048 0.9115 E
- 3
-0
- )63 0.9030
.0031 0.9079 0.9146
.0020
.9001 c:
Ql 1J "O
l.D 0
.0036
.0037 I
~
.. 0083
.000(;
I
.0006
.OOOA
(.'.)
~ 9.
. 0030
.0084
. 0081 c:
.0107
.0097 0
.0043
- 0045
....... u CJ
.0018
.0018 Cl)
.0165 OJ OJ
.0061 Cl}
.0170
.0220 J
199 0.9266 0.9249*
0.9316
. 9~,~i
- ., )'.'jf*
~ " 'V I 5.0
- r i
I I I REFE?.ENCES
- l.
Cadwell, W. R., "-PDQ-7 Reference Manua_l." Bettis Atomic
- Power Laborato~, Vv.APD-TM'.""678 (January 1967).
- 2. - Kirn, Y. S., "NUMICE-2: A Spectrum Dependent Non-Spatial Cell Depl.eti'on Code for CDC-6600." NUS Corporation, NUS-894, Rev. 2 (March 1976).
- 3.
Barry, R. F., "LEOPARD - A Spectrum Dependent Non-Spatial Depletion Code for the IBM 70*94." \\Vesting house Electric Corporation I vVCAP.:..32 69-2 6 (September 1963).
- 4.
- l.
Strav:bridge, L. E., "Calculation of Lattice Parameters and Criticality for Uranium Water Moderated Lattices." \\Vestinghouse Electric Corporation, \\VCAP-32 69-35 (September 1963).
- 5.
Petrie, L. M. and Cross, N. F., "KENO-IV: An Improved Monte 6
- Carlo Criticality Program." Qak Ridge National Laboratory, ORl'.JL-4938 (November 1975)*.
Greene, N. M., et al., "AMPX: A Modular Code System for Generating Coupled Multigroup Neutron - Gamma Libraries from ENDF/B." Oak Ridge National Laboratory, ORNL/TM-3706 (March 1976).
- 7.
Petrie, L. M. and McCarty, P. G., "Validation of Monte C3rlo Calculations of Shipping Cask Systems." ORNL, CONF-731101-14 (1973).
- s.
Davison, P. W. et al., "Yankee Critical Experiments -
Measurements on Lattices of Stainless Steel Clad Slightly Enriched Uranium Dioxide Fuel Rods in Light \\"later,"
YAEC-94 (April *1959) *
- 9.
Learner, R.*D., et al., "Critical Experimen~s Performed \\vith Clustered and Uniform Arrays of Rodded Absorbers," \\Vesting house Electric Corporation, WCAP-3269-39 (~ovember 1965).
17'