ML20211P882

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Redacted/non-proprietary Rev 6 to Transtor PWR Criticality Analysis
ML20211P882
Person / Time
Site: 07109268
Issue date: 08/09/1999
From:
SIERRA NUCLEAR, INC.
To:
Shared Package
ML20138D213 List:
References
BNFL1.10.06.14, BNFL1.10.06.14-R06, NUDOCS 9909140125
Download: ML20211P882 (31)
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Text

SNC NO.:

BNFL01.10.06.14 CLIENT NO.:

BNFL-01 REVISION NO.:

6 j

1 DESIGN CALCULATIONS Transtor PWR Criticality Analysis Prepared by SIERRA NUCLEAR CORPORATION for l

BNFL I

l Approved by: 43 Ar DFs Date:

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Approved by:

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TranStor" PWR Criticality Analysis SNC NO.: BNFL1.10.06.14 REVISION CONTROL SHEET FOR DESIGN DOCUMENTS i

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- 1. Introduction / P urpo se........................................................................................................... I 2.Results.................................................................................................................................2

3. Design Input and Assumptions............................................................................................ 3 3.1. Design Inp ut......................................................................................................

3.2. Assu mpt io n s...............................................................................................

4. M odel / M ethodology.................................................................................-......... 3 6 4.1. General Model Description.................................................................................. 3 6 4.2. Detennination of Most Reactive TranStor PWR System Configuration................. 36 4.3 Preliminary Maximum Enrichment Determination for 24 Assembly Loading Con figuration..................................................................................................................... 3 4.4. Pellet Diameter Studies............................................................................................. 3 9 4.5. Quantification of 24 Assembly Loading Configuration Suberitical Margins............ 40 4.6. Maximum Enrichment Determination for 20 Assembly leading Configuration....... 40 4.7. Trojan-Specific End-Drop Hypothetical Accident Condition (HAC)........................ 40 4.8. Computer Codes........................................................................................................ 4 1 4.9. Upper Subcritical Limit (USL)................................................................................... 41
5. References......................................................................................................................42
6. Calculations................................................................................................

6.1. KENO Va Model Description..................................................................................... 43 6.1.1. Fuel Assembly Construct................................................................................. 4 8 6.1.2. Right Corner L-Spacer Construct...................................................................... 4 8 6.1.3. Left Comer L-Spacer Construct........................................................................ 4 8 6.1.4.0.75 inch Bar Spacer Construct........................................................................... 49 6.1.5. Main Cross Spacer Construct............................................................................... 49

- 6.1.6. Poison Sheet Construct........................................................................................ 5 0 6.1.7. Central Fuel Sleeves and Contents Construct...................................................... 51 6.1.8. Debris Canister Content Construct................................................................... 51 6.1.9. Right and Left Comer Sleeves and Contents Construct...................................... 51 6.1.10. Fuel Sleeve and Small Flux Trap Combination Constmets............................... 52 6.1.11. Global Configuration Construct......................................................................... 52 6.1.12. Material Specifications...................................................................................... 5 3 6.1.13. KENO Va Control Specifications...................................................................... 54 6.2. Detennination of Most Reactive TranStor" PWR System Configuration................. 54 6.2.1. Assembly Shifling Evaluation............................................................................. 55 6.2.2. Canistered versus Intact Fuel Evaluation............................................................. 61 i

6.2.3. BORAL Attachment Pin-Hole Modeling Evaluation.......................................... 61 6.2.4. Most Reactive Intemal Water Moderator Density Evaluation.........................,. 64

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6.2.5. Cask Center-tohter Pitch Evaluation........................................................: 66

'6.2.6. Most Reactive Extemal Water Moderator Density Evaluation........................... 70 i

6.2.7. Ouide Tube Modeling Evaluation........................................................................ 73 6.2.8. Clad Modeling Evaluation................................................................................... 76 6.2.9. Confirmation of Most Reactive Configuration.................................................... 80 6.2.10. Most Reactive TranStor" PWR Configuration Summary................................ 81 F ^. Bl#L Revie6en Peupered Does Checked Does Page W d j

i sdycetinnSd* PWR Crnicality Ans1pis 6

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SNC Sierra Nuclear Corporation 6.3. Preliminary Maximum Allowable Enrichment Detennination for 24 Assembly Lo adin g Con fi guration..............................................................................

6.4. Pellet Diameter S tudies......................................................................

6.5. Quantification of 24 Assembly Loading Configuration Subcritical Margins............. 95 6.6. Maximum Enrichment Determination for 20 Assembly Loading Configuration....... 96 6.7. Trojan-Specific End Drop Hypothetical Accident Condition (HAC)......................101 6.7.1. Description of the Trojan-Specific Shipping End-Drop HAC Configuration...101

...... 103 6.7.2. Description of Most Reactive Trojan Fuel...............

..... 103 6.7.3. Trojan-Specific End-Drop HAC Results and Conclusions....

6. 8. U S L Applicability.........................................................................

6.9. Electro nic Files..........................................................

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NW Date RevwHm

r. 6 red Date p,g,y,,g Chent/Propect: BNFL Suyect. Transtor'" PWR Criticahty Analysis

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KDW 0842m 101i OtMM iv Calculation Number: BNFLI.10 0614

4 SNC Sierra Nuclear Corporation

1. Introduction / Purpose The purpose of this calculation is to evaluate the criticality of the SNC TranStor PWR system. The TranStor* PWR basket is designed to hold a maximum of 24 PWR assemblies during transport and storage. The TranStor" PWR system is designed not to exceed a kwr f 0.95 under worst case cask conditions when the basket is loaded with n

maximum allowed enrichment fuel assemblies. This calculation determines the maximum P_ R basket.

W allowed fuel enrichment in the 20 and 24 assembly TranStor Four of the fuel locations in the basket have been designed to hold canistered partial or damaged fuel assemblies, or canistered fuel debris in larger than normal sleeves. A canistered assembly may contain fuel material having higher reactivity than the Tanaining assemblies in the basket. The four canistered fuel locations are assumed to contain a maximum reactivity load as calculated in Reference 5.5. As discussed in Section 4 of this document, the maximum reactivity load calculated in Reference 5.5 will bound any possible configuration inside the canisters, including partial fuel assemblies, damaged fuel assemblies, and fuel debris. The analyses also bound any configurations that may occur if fuel pellets exit a damaged rod had move freely within the canisters.

The CSAS25 control sequence of the SCALE 4.3 modular. code system (Refs. 5.1 and 5.9) was used to perform the calculations documen.ted herein. SCALE 4.3 is a standardized nuclear analysis software system developed by Oak Ridge National Laboratory for the U.S. Nuclear Regulatory Commission.

'Ihe following changes were made between this revision of the document and the previous revision:

^

all criticality calculations were re-performed to include all component dimensional e.

tolerances all free space that exists between MSB intemal components was removed such that e

the tightest packing of fuel sleeves was analyzed

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an evaluation was added to determine the effect of poison sheet attachment pin holes e

on kwr the scope of the criticality design basis was clearly identified in tenns of the assembly e

specifications that are bounded.

This revision completely supercedes all previous revisions of this document.

This revised calculation file is written in accordance with QA procedures Q4P 3.0, Rev.

7, and QAP 3.2, Rev. 5. Revised QA procedures were in effect by the date of felease of this document. However, as documented in Memorandum 99-241 (froni Ken Wright /

Ram Srinivasan to Howard Wong / Ed Fuller), a request to have this document written to the previous QA procedures was granted.

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2. Results The analyses presented in this calculation file provide the basis for establishing assembly -

loading acceptance criteria that ensure the criticality safety of the TranStor" PWR system during all phases of operation (loading / unloading, transfer, storage, and shipping).

The assembly loading acceptance criteria are based on analyses of variations in assembly parameters that may effect criticality performance. The assembly loading acceptance criteria for the 20 and 24 assembly basket loading options are presented in Table 3.2-2.

Table 2-1 presents the maximum allowable enrichment results for each of the PWR fuel assembly classes in the 20 and 24 assembly basket loading configurations. The maximum allowable enrichtnent results apply to all assembly parameter ranges for each assembly class as presented in Table 3.2-2 and Figure 3.2-1. The maximum allowable enrichment results are bar ed on analyses of the system configuration that.is most reactive.

Characteristics of the most reactive system configuration include the following:

zero bumup credit zero boron credit in the spent fuel pool all UO2 assumed to have 96% theoretical density e

all fuel pellet dishes and chamfers filled with UO2 fuel assemblies shifted into most reactive position within their respective sleeves all component dimensional tolerances included to maximize system reactivity (i.e.,

e worst case flux trap dimensions, worst case poison sheet dimensions, closest packing of fuel sleeves) fuel assembly pellet-to-clad gap filled with normal density water optimal (i.e., maximum reactivity) interior and exterior moderation credit only taken for 75% of B-10 atom density in poison sheets most reactive canistered fuel matrix e

three-dimensional (3-D) infinite array of transfer casks containing MSBs radial neutron shield removed from transfer cask (maximum cask-to-cask neutron a

interaction in x-y plane of 3-D cask array) no neutron attenuation material other than extemal moderator above or below fu e

sleeves (maximum cask-to-cask neutron interaction in z-direction of 3-D cask array).

All of the calculated kerr values, based upon the maximum allowable enrichment for each assembly class specification as presented in Section 6, were less than their corresponding upper subesitical limit (USL) values. As described in Section 4.9, the USL value

+sents the maximum allowable kerrvalue after accounting for bias, uncertainty, and administrative margin.

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Sierra Nuclear Corporation Table 2-1. Maximum Allowable Enrichment Results for the 24 and 20 Assembly Loading Configurations of the TranStor PWR System 24 Assembly Loading 20 Assembly Loading n Euration ConGguration t

Assembly Class Maximum Allowable Maximum Allowable Enrichment (wt%)

Enrichment (wt%)

W 14x14 4.90 5.00 W 15x15 4.40 4.60 W 15x15 ANF 4.30 4.60 W l7x17 4.40 4.70 W 17x17 OFA 4.30 4.60 W 17x17 ANF 4.29 4.58 B&W 15x15 4.40 4.80 4.70 B&W Mark BW 4.39 CE 14x14 4.80 5.00 CE 14x14 St. Lucie 5.00 5.00 CE 15x15 A 4.80 5.00 CE 15x15 B 4.80 5.00 CE 'l5x15 C 4.60 4.80 CE 16x16 4.90 5.00 CE 15x16 5.00 5.00 W 14x14 SS 5.00 5.00 W l5x15 SS 5.00 5.00 W 17x18 5.00 5.00

' The assembly class specification are presented in Table 3.2-2.

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3. Design Input and Assumptions I

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r SNC Sierra Nuclear Corporation Table 3.2-2. TranStor* PWR System Criticality Design Basis Assembly Class Specifications Assembly Class W 14x14 W 15x15 _

W 15x15 ANF W 17x17 2 0A000 2 OA2%

2 OA200 2 0.3700 Diameter ( ches) 2 0.0225 2 0.0240 2 0.0240 2 0.0220 in

)

2 0.3444 2 0.3640 2 0.3560 2 0.3190 eHe ame er s 147.0 s 147.0 s 147.0 s 147.0 hi s) 2 0.0236 2 0.0258 2 0.0258 m 0.0236 E*hh

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Rod Pitch (inches) 0.5560 0.5630 0.5630 0.4960 See Figure See Figure See Figure See Figure Lattice Layout 3.2-1 3.2-2 3.2-2 3.2-3 t

Assembly Class W 17x17 OFA W 17x17 ANF W 17x18 W 14x14 SS i

Clad Outer 2 0.3600 2 0.3500 2 0.3400 2 0.4200 D.iameter(inches) 2 0.0220 2 0.0220 2 0.0120 2 0.0160 "f;,,ess 2 0.3080 2 0.3000 2 0.2865 2 0.3820

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s 147.0 s 147.0 s 147.0 s 125.0

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Rod Pitch (inches) 0.4960 0.4960 0.4220 0.5560 See Figure See Figure See Figure See Figure Lattice Layout 3.2-3 3.2-3 3.2-4 3.2-5 Assembly Class arameter W 15x15 SS B&W 15x15 l B&W Mark BW I CE 14x14 2 0.4200 2 0.4200 2 0.3750 2 0.4300 Di eter( ches) 2 0.0160 2 0.0260 2 0.0230 2 0.0250 e

ss 2 0.3820 2 0.3670 2 0.3230 2 0.3680 (in s 140.0 s 125.0 s 147.0 s 147.0 h inch s) 2 0.0155 2 0.0304 2 0.0258 2 0.1217 u j) e c8 P j

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Revision Prepad Dew Checked Date Pop 22 d Cint/ Project BNFL OWM IA e

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m SNC Sierra Nuclear Corporation Rod Pitch (inches) 0.5630 0.5680 0.5020 0.5800 See Figure See Figure See Figure See Figure Lattice Layout -

3.2-2 3.2-6 3.2-3 3.2-7 Assembly Class Parameter CE 14x 4 St.

CE 15x15 A CE 15x15 B CE 15x15 C 2 0.4400 2 0.4135 2 0.4135 2 0.4135 Diameter ( ches)

Clad Eckness 2 0.0260 2 0.0225 2 0.0225 2 0.0225

  • 't 2 0.3795 2 0.3500 2 0.3500 2 0.3500

(;n s 147.0 s 135.0 s 135.0 s 135.0

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2 0.0703 20.0 20.0 20.0 Rod Pitch (inches) 0.5680 0.5500 0.5500 0.5500 See Figure See Figure See Figure See Figure g;

g 3.2-7 3.2-8 3.2-9 3.2-10 Assembly Class arameter CE 16x16 CE 15x16 Clad Outer 2 0.3800 2 0.3600 D.iameter(inches)

Clad Thickness 2 0.0230 2 0.0220 (inches)

PeHe ameter 2 0.3230 2 0.3100 (inches)

Active Fuel s 153.0 s 95.0 Lanoth(inches)

PerWije 2 0.1122 20.0 Tube (inches )

Rod Pitch (inches) 0.5060 0.4720 Lattice Layout See Figure 3.2-11 See Figure 3.212 i

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, _BNFL ftswision suycet, henstcn'" PWR Crisahty Anahd5 6

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s e D e e e e e e e e D e e e s e e O c e s e D e e e s e e e e e e e e e e e e e e e e O c e D e e O c e O c e e e e e e e e e e e e e e o e e e e e e e e e eeeee e

Fuel Rod h

GuideTube J

Water Hole or Instmment Tube Figure.3.2-1.W 14x14 Assembly Class Lattice Layout Note: It is assumed that displacing any moderator within an assembly (i.e., within a water hole or guide tube) with a non-fuel material will cause her to decrease. It is also assumed that replacing any fuel rod within an assembly with a non-fuel rod or integral bumable poison rod will cause kert o decrease.

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SNC Sierra Nuclear Corporation s e e e e e e e e e e e e e e e e e e e e e e e e e e e e s e e D e e Oc e e D e e D e e e s e'e e e e c e s s e e e s s e e e O c e s s e Oc e s e e e Oc e e e e e e e e D e e s e e e e e e e e e e e e e e e s ec e s e e s e Os e e e s e e e e e e e e e e e e e s e c e s s e e e e e e D e e e e e e D e c e e e D e e c e' 0 e 0 0 e 0 0 0 0 0 0 0 0 0 e e 0 0 0 0 0 0 0 0 0 0 0 0 0 0 e e e e e e e e e e e e e s e e s e e e e e e e e eeese e

Fuel Rod h

GuideTube

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Water Hole or Instrument Tube Figure 3.2-2. W 15x15, W 15x15 ANF, and W 15x15 SS Assembly Class Lattice Layout Note: It is assumed that displacing any moderator within an assembly (i.e., within a water hole or guide tube) with a non-fuel material will cause kert to decrease. It is also assumed that replacing any fuel rod within an assembly with a non-fuel rod or integral burnable poison rod will cause ketr to decrease.

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Fuel Rod h

GuideTube Waterliole or Mtrument Tube Figure 3.2-3. W 17x17, W 17x17 OFA, W 17x17 ANF and B&W Mark BW Assembly Class Lattice Layout Note: It is assumed that displacing any moderator within an assembly (i.e., within a water hole or guide tube) with a non-fuel material will cause ken to decrease. It is also assum that replacing any fuel rod within an assembly with a non-fuel rod or integral burnable poison rod will cause ken to decrease.

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SNC Sierra Nuclear Corporation iOO999 9 9 9 9 9 9 9 9 9 9 0 0 4 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 0 0 0 9 9 9 9 9 9 9 9 9 9 0 0 9 9 9 9 9 9 9 9 9 0 0 0 0 0 0 0 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 99 9 O O O O O O O O 0 9 9 9 9 9 9 9 9 9 9 9 9 0 0 0 0 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 0 0 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 G 9 9 9 9 9 0 9 9 9 9 9 9 9 9 9 9 9 9 9 0 0 0 6 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 0 0 0 9 0 0 9 9 9 9 9 99999990 p

FuelRod Water Hole or Instrument Tube Figure 3.2-4. W 17x18 Assembly Class Lattice Layout Note: It is assumed that displacing any moderator within an assembly (i.e., within a water hole or guide tube) with a non-fuel material will cause hn to decrease. It is also assumed that replacing any fuel rod within an assembly with a non-fuel nod or integral burnable poison rod will cause hn to decrease.

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_e Fuel Rod h

GuideTube Figure 3.2-5. W 14x14 SS Assembly Class Lattice Layout j

l Note: It is assumed that displacing any moderator within an assembly (i.e., within a water hole or guide tube) with a non-fuel material will cause ha to decrease. It is also assumed that replacing any fuel rod within an assembly with a non-fuel rod or integral bumable poison rod will cause he to decrease.

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GuideTube Water Hole or Instmment Tube Figure 3.2-6. B&W 15x15 Assembly Class Lattice Layout Note: It is assumed that displacing any moderator within an assembly (i.e., within a water hole or guide tube) with a non-fuel material will cause kerr to decrease. It is also assumed that replacing any fuel rod within an assembly with a non-fuel rod or integral bumable poison rod will cause kerr to decrease.

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SNC Sierra Nuclear Corporation seoss e e e e e e e e s e s e e e e e e e e ee e e s e s e e e e s s e s e e e e e e e s e s e ee e e e e e e e e e e e e e e e e e e e e e e e e I

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Fuel Rod Guide Tube Figure 3.2-7. CE 14x14, CE 14x14 St. Lucie Assembly Class Lattice Layout Note: It is assumed that displacing any moderator within an assembly (i.e., within a water hole or guide tube) with a non-fuel material will cause ken to decrease. It is also assume that replacing any fuel rod within an assembly with a non-fuel rod or integral bumable poison rod will cause ken to decrease.

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Fuel Rod Water Hole or Instrument Tube Figure 3.2-8. CE 15x15 A Assembly Class Lattice Layout Note: It is assumed that displacing any moderator within an assembly (i.e., within a water hole or guide tube) with a non-fuel material will cause ken to decrease. It is also assumed that replacing any fuel rod within an assembly with a non-fuel rod or integral bumable poison rod will cause ken to decrease.

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Fuel Rod Water Hole or Instrument Tube Figure 3.2-9. CE 15x15 B Assembly Class Lattice Layout Note: It is assumed that displacing any moderator within an assembly (i.e., within a water hole or guide tube) with a non-fuel material will cause kerr to decrease. It is also assume that replacing any fuel rod within an assembly with a non-fuel rod or integral bumable poison rod will cause kerr to decrease.

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Fuel Rod Water Hole or Instrument Tube Figure 3'.2-10. CE 15x15 C Assembly Class Lattice Layout Note: It is assumed that displacing any moderator within an assembly (i.e., within a water hole or guide tube) with a non-fuel material will cause kerr to decrease. It is also assumed that replacing any fuel rod within an assembly with a non-fuel rod or integral bumable poison rod will cause kert o decrease.

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Fuel Rod Guide Tube Figure 3.2-11. CE 16x16 Assembly Class Lattice Layout Note: It is assumed that displacing any moderator within an assembly (i.e., within a water hole or guide tube) with a non-fuel material will cause brr to decrease. It is also assumed that replacing any fuel rod within an assembly with a non-fuel rod or integral burnable poison rod will cause brr to decrease.

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Water Hole or Instrument Tube Figure 3.2-12. CE 15x16 Assembly Class Lattice Layout Note: It is assumed that displacing any moderator within an assembly (i.e., within a water hole or guide tube) with a non-fuel material will cause karto decrease. It is also assumed that replacing any fuel rod within an assembly with a non-fuel rod or integral bumable poison rod will cause ker to decrease, i

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