ML20034B494

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Rev 0 to Final Rept Criticality Safety Analysis Millstone-2 New Fuel Storage Vault & Transfer Carriage W/5 % Enriched 14x14 Fuel Assemblies
ML20034B494
Person / Time
Site: Millstone Dominion icon.png
Issue date: 02/29/1988
From: Gerrald L
SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML20034B489 List:
References
ANF-88-028, ANF-88-028-R00, ANF-88-28, ANF-88-28-R, NUDOCS 9004270251
Download: ML20034B494 (30)
v • d • e


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FINAL REPORT CRITICALITY SAFETY ANALYSIS MILLSTONE-2 NEW FUEL STORAGE VAULT L

AND TRANSFER CARRIAGE f

WITH 5.0 PERCENT ENRICHED 14x14 FUEL ASSEMBLIES l

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i FINAL REPORT CRITICALITY SAFETY ANALYSIS-

't MILLSTONE 2 NEW FUEL STORAGE VAULT AND TRANSFER CARRIAGE-

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WITH 5.0 PERCENT ENRICHED 14X14 FUEL ASSEMBLIES-

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February 1988

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FINAL REPORT CRITICALITY SAFETY ANALYSIS MILLSTONE-2 NEW FUEL STORAGE VAULT AND TRANSFER CARRIAGE WITH S.0 PERCENT ENRICHED 14X14 FUEL ASSEMBLIES s

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Prepared by:

L. V. Gerfald, Criticality Safety Specialist Date Corporate Licensing Reviewed by:

J //P/fp J. E//Pieper, Cfitjtlility Safety Specialist Date Corp 5 rate Licensing Approved by:

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C. W. Malody, Mahager f

/Dr'e Corporate Licensing y

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ANF-88 028, Revision 0 FINAL REPORT CRITICALITY SAFETY ANALYSIS MILLSTONE-2 NEW FUEL STORAGE VAULT AND TRANSFER CARRIAGE WITH $.0 PERCENT ENRICHED 14X14 FUEL ASSEMBLIES e

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SUMMARY

The criticality safety of the new fuel storage vault and transfer carriage with S.O percent enriched 14x14 bundles is demonstrated in accordance with NUREG-0800 and ANSI /ANS-57.3-1983.

The subject system meets the applicable criticality safety criteria subject to i

the limits and controls given below.

1)

Fuel Design:

As specified in Section 2.0.

2)

System Design

As described in Section 3.0, 3)

If fuel assemblies are stored with plastic wrapping, the bottom of the wrapping shall be open to assure free drainage.

4)

Five hundred (500) ppm (minimum) dissolved Boron in water-during fuel l

movements in normally flooded systems unless four inch (minimum) edge-l edge bundle spacing (two bundles) is assured at all credible accident conditions.

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ANF 88 028, Revision 0 f

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.2.0 FUEL PARAMETERS j

The key bundle design parameters used in these calculations are listed. in i

Table 2.1.

The bundle is a 14x14 design with five guide tubes. Since the' guide tubes are l

much larger than the fuel rods, the bundle may be more easily visualized as a.

7x7 array of two cell types. Type 'F' is.a 2x2 fuel rod array and Type 'G' is a single guide tube.

Using this 7x7 description, the bundle ~ is composed as shown in Figure 2.1.

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Table 2.1 l

Bundle Parameten

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Parameter Desian Value Model Value Enrichment (wt% U-235) 5.0 (max.)

5.0 PelletDiameter(inch) 0.3700 0.3700 i

Pellet Density (%TD,)

94.0-95.0 Pellet Dish Volume (%)

1.0 0

j Active Fuel Length (1nch) 136.7 136.7 (min.).

j Clad 10/00(inch) 0.378/0.440 0.378/0.440 Rod Pitch (inch) 0.5800 0.5800 Gd/ Boron Content Variable None Fuel Rods Per Bundle 176 176 Guide Tube 10/00 (inch) 1.035/1.115 1.035/1.115 t

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5 ANF-88-028, Revision 0 3.0 STORAGE RACK GEOMETRY The coordinate system used is:

X:

East-West (East is +)

Y:

North-South (North is +)

Z:

Vertical (Up is +)

Multiple dimensions and array descriptions are listed in the order X, Y, Z.

The racks were modeled in accordance with Figure 9.8-1 of the FSAR.

The vault was modeled as a room 32'-9"x16'x11'4.7".

The vault was reflected with 30 cm of concrete at all six faces.

l This close-fitted reflection at the top and bottom is conservative relative to l

the actual bundle positioning and the vault height.

All 76 storage locations were' filled with the subject fuel design.

There are eight 2x4 modules and three lx4 modules of storage cells.

l The bundles are spaced 20.5 inches center-center within ' modules.

The edge-edge module module spacing modeled is 31 inches (X) and 28 inches (Y).

All materials of construction were neglected in the model.

All neutron absorptions occur in the fuel, the moderator, or the reflector.

This is a conservative model.

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ANF 88 028, Revision 0 4.0 CALCULATION METHODS All computer codes and cross sections are part of the SCALE (1) system.

The neutron multiplication factors, keff or k-inf, were calculated using t

XSDRNPM, a one-dimensional discrete ordinates transport code, or KENO-Va, a three-dimensional Monte Carlo code.

Sixteen group (Hansen-Roach) cross sections were used with resonance correc-tions by BONAMI/NITAWL.

l All codes and cross sections have been extensively benchmarked against i

critical experiment data.

4.1 METHODS VERIFICATION The SCALE codes and cross sections have been extensively benchmarked against data from critical experiments.

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Supplemental benchmarking was performed immediately before the calculations reported here. The experiments selected are described in' References 2 and 3.

The results are listed in Table 4.1.

The average and standard deviation are 1.00265 and 0.00490, respectively.

The 95 percent upper limit (UL) on the KEN 0 k rf is calculat.ed by pooling the e

KEN 0 variance and the bias variance.

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7 ANF-88-028, Revision 0 Table 4.1 Benchmark Calculation Results from KENO Va 16 Group Cross-Sections Calculated k gg Cnne No.

e Reference 2 Experiments

.2378 1.00395 0.00376 2384 1.00037 0.00306 2388 0.99886 0.00341 2420 1.00038 0.00367 2396 0.99443 0.00360 l

2402 1.00694 0.00283 l

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10 1.00181 0.00412 11 0.99786 0.00413 12 0.99885 0.00487 31-1.00442 0.00421

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8 ANF 88 028, Revision 0 For example, the 95 percent UL keff for Case 2378 is calculated below, keff (95% UL) = 1.00395 - 0.00265 + 1,66

  • sqrt(3.76E-3**2 + 4.90E-3**2)

= 1.00130 + 0.01025 - 1.01155 The 1.66 multiplier is the one-sided Student t (5%) with about 80 degrees of freedom.

For reference, the bias corrected results are reported in Table 4.2.

The 95% upper limit (UL), which is the parameter used in judging accept-i ability, exceeds 1.0 in every case after bias correction.

The average 95% UL is 1.0102.

Therefore, the results remain conservative.

All results in this report have agl been bias corrected, unless otherwise stated. Therefore, these results would tend to be conservative'by about 0.0027.

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ANF-88-028, Revision 0 Table 4.2 i

Bias-Corrected Benchmark Results (95kUL)

Case No.

keff 2378-1.00130 i 0.00376 1.01155 2384 0.997716 1 0.00306 1.00731 i

2388 0.996206 0.00341 1.00612 2420 0.997726 0.00367 1.00789

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t 2396 0.991776 1 0.00360 1.00187-2402 1.00429 0.00283 1.01368 5

2411 1.00958 0.00286-1.01900 2407 1.00382 1 0.00332 1.01364

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0.998266 0.00487 1.00973 10 0.999156 0.00412 1.00978 11 0.995206 1 0.00413~

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31 1.00177 1 0.00421 1.01249:

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8 10 ANF-88 028, Revision 0 5.0 CALCVLATION RESQL15 The racks were conservatively modeled with uniform interspersed moderation and with concrete reflection.

The input for a typical KENO run _ is attach for reference.

The KENO Va results are listed in Table 5.1.

The fully flooded result is actually that for a single bundle surrounded by 30 I

cm of water.

As will be shown in Section 6.2, flooded bundles on 20.5 inches centers are effectively isolated (i.e., the bundle-bundle interactions are negligible.

The peak reactivity occurs at full flooding.

The 95% upper limit on the peak k rf is:

e k rf (95% UL) = 0.9045 - 0.00265 + 1,66

  • SQRT(0.0038**2+0.0049**2) = 0.9121 e

i Therefore, the system meets the 0.95 limit on keff.

If the iron / steel structural members of the rack had been modeled, the keff would have been lower than that reported here.

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~11 ANF-88-028, Revision 0 Table 5.1 idew Fuel Racks:

5.0% Enriched Fuel Interspersed Moderation Effects KENO-Va Results Interspersed-Water Density (Vo1%)

keff 1

0.7095'i 0.0047 2.5

.0.7139 1 0.0035.

5 0.8566 0.0032 7.5

.0.8158 0.0034 100 0.9045 0.0038 8

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C 12 ANF-88-028, Revision 0 6.0 SENSITIVITY STUDIES The key parameters controlling reactivity are:

1)

Fuel enrichment:

The enrichment is fixed at the maximum. credible value (5.0%).

i 2)

Moderation:

Data on interspersed moderation effects within the new fuel storage are in Section 5.0.

Other moderation effects are covered in Section 6.1.

3)

Bundle-Bundle Spacing:

Spacing ef.fects due to dimensional tolerances, eccentric positioning and those due to bundle handling accidents are covered in Section 6.2.

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6.1 Moderation Effects (Full Floodina)

The nominal bundle design (176 fuel rods and 5 guide tubes) is composed as follows:

The average water / fuel volume ratio (Vw/Vf) is 2.03.

If the entire 14x14 array was fuel rods, the Vw/Vf would be 1.71.

Since reactivity may be' changed if fuel rods are removed from the bundle, generic bundle designs with Vw/Vf ratios in the range 2.0-4.0 were evaluated.

The generic bundles were modeled with the nominal pellet and clad dimensions but with a pitch selected to yield the desired Vw/Vf.

This is a conservative model since the zircaloy of the guide tubes is not in the model.. Reference data for models with all zircaloy included are also provided. The calculation sequence was as follows:

1)

Self-shielded 16 group cross sections were generated using BONAMI/NITAWL.

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~13 ANF-88-028, Revision 0 Table 6.1 Fuel Zone Composition Nominal Design Values Material.

Volume %

28.70' UO2 Pellet-Clad Gap 1.25 Clad (zirc)-

11.66-Moderation (water) 58.'39 Total-100.00-4 e

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Cell-weighted cross sections were generated using an XSDRNPM model of an infinite rod array.

b 3)

These cell-weighted cross sections were used to simulate a 8.12"x8.12"x infinite bundle in KENO-Va or XSDRNPM models of bundles in an infinite array or a single bundle with full water reflection.

Generic bundle characteristics are listed in Table 6.2.

Listed in Table 6.3 are XSDRNPM results for generic rods / bundles.

The results include the k-inf for an infinite rod lattice (cell-weighting run) and the keff for a single bundle with full water reflection (FWR).

The bundle was modeled as a 12.0834 cm radius cylinder (infinite length) surrounded by a 30 cm of water.

l The Table 6.3 results indicate a peak keff (conservative model) near 0.934 l

assuming that fuel rods are withdrawn in the optimum sequenca.

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Table 6.2 Generic Bundle Characteristics Clad OD Pitch Total Removed l

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Fuel Rods Fuel Rods Guide Tube Zr in Model 2.0 1.13148 1.54689 177.8

-1.8 2.03 1.13162 1.55465 176.0 0 (nominal) 2.5 1.1335 1.65629 155.1 20.9 3.0 1.1355 1.75889 137.5 38.5 3.5 1.13752 1.85585 123.5 52.5 i

4.0 1.13952 1.94796 112.1 63.9 i

No Guide Tube Zr in Model 2.0 1.1176 1.53894 177.8

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2.03 1.1176 1.54667 176.0 0 (Nominal) 2.50 1.1176 1.64778 155.1 20.9 3.0 1.1176 1.74987 137.5 38.5 3.5 1.1176 1.84631 123.5 52.5 4.0 1.1176 1.93797 112.1 63.9 1

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ANF-88-028, Revision 0-Table 6.3 Fuel Rod Removal Effects (Generic Rods / Bundles)

XSDRNPM Results (Infinite Length Rods / Bundles)

Rod FWR Bundle k rr___

yw/Vf k-inf e

Guide Tube Zr in Model 2.0 1.5086 0.9062 2.5 1.5252 0.9219 3.0 1.5298 0.9286 3.5 1.5279 0.9296 4.0 1.5210 0.9267 No Guide Tube Zr in Model 2.0 1.5102-0.9103 2.03 1.5119 0.9117 2.5 1.5268 0.9262 3.0 1.5314 0.9329 3.5 1.5295 0.9340 W

4.0 1.5226 0.9311 L

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17 ANF-88-028, Revision 0 Safety of a~ single bundle is assured with any credible' number of fuel rods removed (or added).

The KENO result for an explicit model of a single bundle (176 fuel rods, 5 guide tubes,136.7" long) flooded and reflected with full density water is 0.9045 0.0038.

A single flooded bundle was also modeled using the conservative homogeneous (cell-weighted) cross sections in KENO-Va; Result:

0.9054 i 0.0035.

s Therefore, the cross sections (heterogeneous-homogeneous) and the codes (KENO-XSDRNPM) agree very well.

6.2 Bundle Soacina Effects An infinite array of generic bundles with the nominal Vw/Vf (2.03) were modeled with XSDRNPM.

The system modeled was fully flooded, The guide tube Zr was not in the model.

The infinite length bundles were spaced as indicated in Table 6.4.

An infinite array of bundles is acceptable with all center-center. spacing greater than about 15 inches.

No credible combination of dimension tolerances and eccentric positioning could result in spacings approaching 15" or less.

Two closely-placed bundles -(flooded, full water reflection) were also modeled to determine the effect of fuel handling accidents.

The results are in Table 6.5.

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i 18 ANF-88-028, Revision 0 Table 6.4 Bundle Spacing Effects Infinite x Infinite Bundle Array Fully Flooded XSDRNPM Results Center Center Spacing Edge-Edge Spacing (inch)

(inch) k-inf-20.5-12.38 0.9128 20 11.88 0.9131 18 9.88 0.9158 16 7.88 0.9244 14 5.88 0.9525 12 3.88 1.0481 10 1.88-1.3218 -

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19-ANF-88-028, Revision 0 Table 6.5-Two Closely Placed Bundles l

Flooded, Full. Water Reflection Zero Boron l

KENO-Va Results (Explicit Model) l Edge-edge Spacing k rf-l (inch) e 0

1.0379 1 0.0053 4

0.9270 0.0052 i

8 0.9041 0.0049-12 0.9002 0.0054 l

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4 20 ANF-88 028, Revision 0 Criticality can result. if two bundles are brought together in a flooded system.

Since a minimum spacing between bundles has been specified and since flooding of the vault is an independent and very unlikely occurrence, no single accident condition in the new fuel vault can result in criticality.

a The four-inch minimum edge-edge spacing is met by two bundles on the transfer carriage.

If an in-transit bundle could be brought closer than four inches to a bundle on the carriage, then 500 ppm (minimum) dissolved boron will be required.

For fuel handling in normally flooded systems, a minimum dissolved boron content is specified to assure safety if-bundles are accidentally brought together.

The effect of boron on the keff of two edge-edge bundles is shown in Table 6.6.

The reflector water (30 cm thick) also contained the indicated boron content.

The specified 500 ppm (minimum) will assure criticality safety at any single credible accident condition during fuel handling.

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21 ANF-88-028, Revision 0 Table 6.6-Dissolved Boron Effects Two Edge-Edge Bundles Flooded, Full Water Reflection KENO-Va Results (Explicit Model)

PPM Boron keff 0

1.0409 r 0.0045 500 0.9371 1 0.0037 1000 0.8739 0.0030 1500 0.8247 0.0038

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22 ANF-88-028, Revision 0 7.0 KENO INPUT LISTING 9

The new fuel vault model, including ' compositions and geometry, is in.the listing below.

The water density was changed for _ other runs reported in Section 6.0.

MILLSTONE 14x14, 5.0% ENR, NEW FUEL RACKS, 5% WATER READ PARAMETERS TME=290.0 GEN =103 NPG=500 LIB 41 TBA=2.0 FLX YES FDN=YES XSl=NO NUB-YES PWT=YES END PARAMETERS READ MIXT SCT-1 MIX = 1 FUEL PELLET, 5.0% ENR 92235 1.175834E-03 92238 2.205852E-02 8016 4.646871E-02 MIX = 2 40302 4.251812E-02 MIX = 3 5% WATER 8016 1.6690-3 1001 3.3380-3 MIX = 4 CONCRETE 8016 4.607448E-02 1001 1.374186E-02 13027 1.745493E-03 20040 1.520656E-03 26000.3.472435E-04 14028 1.662057E-02 11023 1.747307E-03 END MIXT READ GE0 METRY

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23 ANF-88-028, Revision 0 UNIT 1 FUEL ROD

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-CYLINDER 1 1 0.4699 2P173.609 CYLINDER -0 1 0.48006 2P173.609-CYLINDER 2 1 0.5588'2P173.609 CUB 0ID 3 1 4PO.7366 2P173.609 UNIT 2 GUIDE TUBE, 1.035"ID/1.115"0D COM=' UNIT 2 IS GUIDE TUBE' CYLI 3 1 1.31445 2P173.609 CYLI 2 1 1.41605 2P173.609 CUB 0 3 1 4Pl.4732 2P173.609 UNIT 3 COM=" 2X2 ARRAY OF FUEL RODS ARRAY 1 2R-1.4732 -173.609 UNIT 4 COM=" 'THIS IS THE BUNDLE ARRAY 2 2R-10.3124 -173.609 ADD MODERATION-FOR 20.5" CENTERS CUB 0 3 1 4P26.035 2P173.609 UNIT 5 COM=" THIS IS A IX4 ARRAY OF CELLS l

ARRAY 3 -26.035 -104.14 -173.609 UNIT 6 COM=" MODERATION FOR 6 FEET C-C BETWEEN 2X4 MODULES E-E SPACING = 72" -41" = 31" CUB 0 3 1 2P39.37 2P104.14 2P173.609 l

UNIT 7 COM=" MODERATION FOR 61.5 INCH C-C SPACING (IX4 - 2X4) l E-E SPACING = 61.5-20.5 - 10.25 - 30.75" CUB 0 3 1 2P39.0525 2P104.14 2P173.609

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24 ANF-88-028, Revision 0 l

VNIT 8 i

COM=" RACKS AT NORTH END OF ROOM

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l ARRAY 4 -456.565 -104.14 -173.609 ADD MODERATION TO WALL.AT EAST-WEST AND TO ROOM CENTER AT N-S l

CUB 0 3 1 2P499.ll 104.14 -139.7 2Pl73.609 t

UNIT 9 COM=" RACKS AT SOUTH END OF ROOM

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ARRAY 5 -326.39 -104.14 -173.609 ADD MODERATION TO WALL AT EAST-WEST AND TO ROOM CENTER AT N-S CUB 0 3 1 2P499.11 139.7 104.14 2P173.609 GLOBAL UNIT 10 ARRAY 6 3R0.0 ADD 30 CM CONCRETE REFLECTION REPL 4 2 6R5.0 6 END GE0 METRY READ ARRAY ARA =1 NUX-2 NUY-2 NUZ-1 FILL F1 END FILL ARA =2 NUX-7 NUY=7 NUZ-1 LOOP I

3 171 171 111 2 264 264 111 2 441 441 111 END LOOP ARA =3 NUX-1 NUY=4 NUZ-1 FILL F4 END FILL ARA-4 NUX-15 NUY-1 NUZ-1 FILL 5 7 5 5 6 5 5 6 5 5 6 5 5 7 5 END FILL ARA =5 NUX-13 NUY=1 NUZ-1 FILL 5 5 6 5 5 6 5 5 6 5 5 7 5 END FILL ARA-6 NUX-1 NUY-2 NUZ-1 FILL 9 8 END FILL END ARRAY s

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r, 25 ANF-88-028, Revision 0 READ BOUNDS ALL-VACVVM END BOUNDS READ START NST-1 END START READ BIAS10-301 2 7 END B.lAS END DATA

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e 26 ANF-88-028, Revision 0 I

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8.0 REFERENCES

1.

" SCALE:

A Modylar Code System for Performina Standardized Computer Analyses for Licensina Evaluation," NUREG/CR-0200.

2.

M.N. Baldwin, et.al., " Critical Exoeriments Succortino Close Proxim Water Storaae of Power Reactor Fuel", BAW-1484-7, July 1979.

3.

S.R. Bierman, B.M. Durst, and E.D. Clayton, " Critical Seoaration' Between Suberitical Clusters of 4.31% Enriched UO2 Rods in Water with Fixed.

Neutron Poisons," NUREG/CR-0073, May 1978.

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