ML17216A616

Rev 1 to St Lucie Unit 1:New & Spent Fuel Storage Criticality Safety Evaluation for Natural U Axial Blanket Fuel.

ML17216A616
Person / Time
Site:	Saint Lucie
Issue date:	02/28/1986
From:	Brown O, Malody C, Patten T SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17216A614	List: ML17216A613 ML17216A615 ML17216A616
References
XN-NF-83-36, XN-NF-83-36-R01, XN-NF-83-36-R1, NUDOCS 8607150141
Download: ML17216A616 (29)

Text

XN-NF-83-36, Rev. 1 Issue Date: 2/28/86 ST. LUCIE UNIT 1 NEW AND SPENT FUEL STORAGE CRITICALITY SAFETY EVALUATION FOR NATURAL URANIIN AXIAL BLANKET FUEL Prepared by

0. C. Brown February l986 E ON NUCLEAR COMPANY, INC.

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XN-NF-83-36, Rev. 1 ST. LUCIE UNIT 1 NEW AND SPENT FUEL STORAGE CRITICALITY SAFETY EVALUATION FOR NATURAL URANIUM AXIAL BLANKET FUEL Prepared by:

rown, ngsneer ae BWR Neutronics Approved by:

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0 I XN-NF-83-36, Rev. 1 ST. LUCIE UNIT 1 NEW AND SPENT FUEL STORAGE CRITICALITY SAFETY EVALUATION FOR NATURAL URANIUM AXIAL BLANKET FUEL TABLE OF CONTENTS Section Page

1.0 INTRODUCTION

.............................................. 1 2.0

SUMMARY

OF RESULTS ........................................ 2 3.0 SPENT FUEL STORAGE POOL ANALYSIS .......................... 3 3.1 Design- Base Fuel Assembly Description ..................... 3 3.2 .Storage Array Description ................................. 3 3.3 Storage Array Reactivity .................................. 3 3.4 C oncluslons ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ si ~ o ~ oo ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5 4.0 NEW FUEL STORAGE ROOM ANALYSIS ............................ 6 4.1 Design Base Fuel Assembly Description ..................... 6 4.2 Storage Array Description ................................. 6 4.3 Storage Array Reactivity .................................. 6 4.4 C OnC 1 us 1 OllS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ 7 5.0 FUEL INSPECTION ELEVATOR, UPENDER, AND TRANSFER TUBE ...... 8 6.0 CALCULATIONAL METHODS ..................................... 9 7.0 COMPUTER MODEL REVIEW AND VALIDATION ..............;....... 10 8.0 R EFERENCES ................................................. 11 APPENDIX 1 - SECOND-PARTY REVIEW DOCUMENTATION

XN-NF-83-36, Rev. 1 ST. LUCIE UNIT 1 NEW AND SPENT FUEL STORAGE CRITICALITY SAFETY EVALUATION FOR NATURAL URANIUM AXIAL BLANKET FUEL LIST OF TABLES Table Page St. Lucie Unit 1 Nominal Fuel Assembly Parameters ......... 12 Spent Fuel Storage Pool Reactivity Sensitivity Calculation R esults ...........'........................;............... 13 New Fuel Storage Room Reactivity Calculation Results ...... 14

) ~ l XN-NF-83-36, Rev. 1 ST. LUCIE UNIT 1 NEW AND SPENT FUEL STORAGE CRITICALITY SAFETY EVALUATION FOR NATURAL URANIUM AXIAL BLANKET FUEL

1.0 INTRODUCTION

This report summarizes the results of three criticality safety analyses performed for the handling and storage of new and spent fuel at the St. Lucie Unit 1 Nuclear Generating Station. Specifically, the analyses addressed the following areas:

1) Spent Fuel Pool (Section 3.0)

2) New Fuel Storage Racks (Section 4.0)

3) Fuel Inspection Elevation, Upender, and Fuel Transfer Tube (Section 5.0)

An Exxon Nuclear Company (ENC) fuel assembly design which includes natural uranium axial blankets on the assembly ends and a central fuel region enriched to a maximum of 4.0 wtX U-235 was assumed for the analyses.

Detailed descriptions of the handling and storage areas is provided in the criticality safety evaluation report sumnarized in Reference 1.

\

XN-NF-83-36, Rev. 1 2.0

SUMMARY

OF RESULTS Calculations performed for the handling and storage of 4.0 wtX U-235 enriched fuel assemblies external to the St. Lucie Unit 1 reactor core indicate that the applicable criticality safety criteria are met.

Worst case reactivities calculated for each area are as follows:

Calculated Limiting Criteria Area keff (95K CL) keff (9N CL) Reference Spent Fuel Pool 0.918 0.95 New Fuel Storage Room 0.925 0.98 Fuel Elevator, Upender, 0.924 0.95 Transfer Tube

3t XN-NF-83-36, Rev. I 3.0 SPENT FUEL STORAGE POOL ANALYSIS The high capacity (HI-CAP') spent fuel storage racks were designed to accomnodate 728 fuel assemblies and are generally defined in Attachment A of Reference I.

3.1 Desi n Base Fuel Assembly Descri tion The St. Lucie Unit 1 spent fuel storage racks are currently designed and licensed to accept fuel assemblies enriched to 3.7 wtX U-235. Table I summarizes fuel assembly parameters for this fuel design, as well as parame-ters for the Exxon Nuclear fuel design. Also listed in Table I are fuel assembly k calculation results obtained using the CCELL(4) computer code.

(CCELL is a pin cell calculation code used to cell-average resonance corrected cross section data for input into other codes, i.e., KENO, etc.)

These results indicate a +0.015 bk change in fuel assembly reactivity between the two fuel types.

3.2 Stora e Array Descri tion (S ent Fuel Pool Analysis)

The spent fuel storage array design and dimensions are described in detail in Reference 1. The array consists of square stainless steel cans having an inside dimension of 8.4835 inches. The cans have a nominal wall thickness of 0.25 inches and-are fixed in a square-pitched array on nominal 12.53-inch centers.

3.3 Storage Array Reactivity (k ) (Spent Fuel Pool Analysis)

For the nominal storage array geometry defined in Section 3.2, keff was calculated for the Exxon Nuclear fuel enriched to 4.0 wtX, U-235 using the KENO IV(5) Monte Carlo computer code. (Cross section data from the XSDRN 123-energy group library were prepared for input into KENO IV using the

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XN-NF-83-36, Rev. I, NITAWL(6) and XSDRNPM(6) codes. A detailed description of the calculation method is given in Section 6.0.) For a pool water temperature of 68oF, a keff of 0.874 + 0.005 was calculated assuming an effectively infinite array.

From results summarized in Reference I, a 8< of +0.034 has been established between the nominal and worst case (minimum offset) storage array conditions.

Applying this hk to the KENO-calculated nominal keff value results in a worst case k ff of 0.918(*) at the 955 confidence level.

For postulated accidents such as the dropping of a fuel assembly on top of the racks or having an assembly by accident achieve any other abnormal location in the pool, credit may be taken for realistic initial pool condi-tions(2). These conditions include taking credit for soluble boron (> 1720 ppm) present in the pool water. At about 1700 ppm, boron storage array reactivity (keff ) is reduced aPProximately 20% hk. Hence, Postulated accidents defined above are of little concern from a criticality safety standpoint.

To better understand the reactivity sensitivity of the storage arrangement to stainless steel can wall thickness and fuel cell center-to-center spacing, several calculations were performed using the one-dimensional transport code XSDRNPM. A cyl indricized-equivalent nominal storage cell was modeled with a reflective cell boundary to appr oximate an infinite array. Results of these calculations are summarized in Table II. For a stainless steel can wal 1 thickness decrease of 0.05 inches, storage array reactivity increases about

+0.005 hk. For a storage cell center-to-center spacing decrease from 12.4375 inches to 12.0000 inches, results indicate a reactivity increase of about

+0.020 &.

• keff(MC) = 0.874 + (2)(0.005) + 0.034 = 0.918 (95K confidence level).

XN-NF-83-36, Rev. 1 3.4 Conclusions (S ent Fuel Pool Anal sis)

The results of this analysis demonstrate that Exxon Nuclear design fuel defined in Table I and enriched to 4.0 wtX U-235 can be stored in the St.

Lucie Vnit 1 storage pool and continue to meet NRC criticality safety criteria, i.e., the neutron multiplication factor in the pool shall be ( 0.95 under worst credible conditions. The adequacy of the calculational methods used in this analysis is discussed in Section 7.0.

XN-NF-83-36, Rev. 1 4.0 NEW FUEL STORAGE ROOM ANALYSIS The new fuel storage array was analyzed in 1979(1) for CE fuel assemblies enriched to 3.70 wtX U-235. Results of this analysis indicated the maximum storage array reactivity (keff) to be about G.92 for all degrees of uniformly interspersed moderation.

4.1 Desi n Base Fuel Assembly Description The Exxon Nuclear fuel assembly defined in Section 3.1 and Table I was used in reactivity calculations for the new fuel storage array.

4.2 Stora e Array Descri tion (New Fuel Stora e Room)

The St. Lucie Unit 1 new fuel storage room consists of a 10x10 fuel assembly array arrangement with the two middle rows {running north and south) removed.

The room has concrete walls around the entire array and cells are spaced on 21-inch centers. The fuel is normally stored in the dry condition.

Additional details concerning the arrangement can be found in Attachment B of Reference 1.

4.3 Storage Array Reactivity (k ) (New Fuel Stora e)

Since the new fuel is stored dry, reactivity calculations were performed for varying degrees of moderation in the event the array became moderated. For the case of full flooding, the array would remain subcritical due to neutron isolation between fuel assemblies. For the case of uniform moderation interspersed within and between fuel assemblies in the storage array, i.e.,

aqueous foam, etc., 123 energy group KENO-IV calculations were performed for water densities in the array ranging from 15% water (by volume) to 2.5X.

Results of these calculations are summarized in Table III and indicate the maximum keff of the array to be 0.925 at the 95K confidence level.

7 XN-NF-83-36, Rev. 1 4.4 Conclusions (New Fuel Stora e)

The results summarized in Section 4.3 and Table III'emonstrate that the maximum keff or the St. Lucie Unit 1 new fuel storage room will not exceed the limiting criterion of 0.98 for all degrees of uniform moderation in the array. Specifically, the maximum reactivity occurs for a moderator void fraction between 0.90 and 0.95 and is estimated to be about 0.925 at the 95K confidence level.

XN-NF-83-36, Rev. 1 5.0 FUEL INSPECTION ELEVATOR, UPENDER AND TRANSFER TUBE For the fuel inspection elevator, upender and transfer tube, keff calcula-tions were performed for three worst case situations in the previous evalua-tion, see Attachment C of Reference 1. The case which produced the highest reactivity involved the fuel inspection elevator. For this case, it was assumed that one fuel assembly was in the elevator and one additional assembly was located four (4) inches edge-to-edge from the elevator assembly.

Assuming this condition, a KENO calculation was performed for the Exxon Nuclear fuel assembly defined in Table I. This case was run with 18 energy group cross section data prepared using CCELL and gave a resulting keff of 0.914 + 0.005 or 0.924 at the 95K, confidence level. Hence, it is concluded that 'the fuel inspection elevator, upender and transfer tube will all meet the keff < 0 g5 criterion for the 4.0 wtX U-235 enriched fue'l.

0 I' I

e 9 XN-NF-83-36, Rev. I 6.0 CALCULATIONAL METHODS The KENO-IV Monte Carlo code(5) was used to calculate the reactivities of the storage arrays and fuel elevator. Multigroup cross section data from the XSDRN 123 group library were generated for input into KENO-IV using the NITAWL(6) and XSDRNPM(6) codes. Specifically, the NITAWL code was utilized to obtain cross section data adjusted to account for resonance self-shielding by the Nordheim Integral Method. The XSDRNPM code, a discrete ordinates one-dimensional transport theory code, was then used to prepare spatially cell-weighted cross section data representative of the fuel assembly for input into KENO-IV. Cross section data for the fuel elevator keff calcula-tion were prepared using the CCELL code. CCELL is a combination of the HRG(8) and THERMOS(9) codes, and produces multigroup cross section data treated in a similar fashion to the NITAWL/XSDRNPM methodology.

10 XN-NF-83-36, Rev. 1 7.0 COMPUTER MODEL REVIEW AND VALIDATION The calculated worst case keff (0.918) for the spent fuel storage array is about A less than the comparable value for 3.7 wtX fuel (0.947) reported in Reference 1. This is apparently due to a high degree of conservatism in the Reference 1 calculational model. The computer code models used for the calculations in this report have been benchmarked against experimental data and adequately reproduce the critical values. (On the average, the critical value is reproduced to within one standard deviation.) Detailed results of these benchmark calculations are given in Reference 7. In addition, the results and conclusions of this criticality safety evaluation have been second-party reviewed by an individual knowledgeable in the area of critical-ity safety. The findings of this review are given in Appendix I.

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XN-NF-83-36, Rev. 1

8.0 REFERENCES

Letter (with attachments) from R. E. Uhrig (FPL) to V; Stello (USNRC),

"St. Lucie Unit 1 Docket No. 50-335 Proposed Amendment to Facilities Operating License DPR-67," dated October 4, 1979.

2. Letter to all Power Reactor Operators from B. K. Grimes (NRC), "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14, 1978.

3. American Nuclear Society Standard, ANS-N18.2.

4. W. W. Porath, "CCELL Users Guide," BNW/JN-86, Pacific Northwest Labora-tories, February 1972.

5. L. M. Petrie and N. F. Cross, "KENO IV: An Improved Monte Carlo Criticality Program," ORNL-4938, Oak Ridge National Laboratory, November 1975.

6. N. M. Green, et. al., "AMPX - A Modular Code System for Generating Coupled Multigr'oup Neutron-Ganma Libraries from ENDF/B, " ORNL-TM-3706, Oak Ridge National Laboratory, March 1976.

7. C. 0. Brown, "Criticality Safety Benchmark Calculations for Low-Enriched Uranium Metal and Uranium Oxide Rod-Water Lattices," XN-NF-499, Exxon Nuclear Company, Inc., April 1979.

8. J. L. Carter, Jr., "HRG-3: A Code-for- Calculating the Slowing-Down Spectrum in the P1 or B1 Approximation," BNWL-1432, Battelle-Pacific Northwest Laboratories, June 1970.

9. D. R. Skeen and L. J. Page, "THERMOS/BATTELLE: The Battelle Version of the THERMOS Code," BNWL-516, Battelle-Pacific Northwest Laboratories, September 1967.

12 XN-NF-83-36, Rev. 1

.Table I St. Lucie Unit 1 Nominal Fuel Assembly Parameters Type: CE 14xl4 Lattice Pitch: 0.580 inch Clad O.D.: 0.440 inch Clad Material: Zr-4 Active Fuel Rods: 176 No. of Control Rod Guide Tubes: 5 Guide Tube Material: Zr-4 Guide Tube O.D.: 1.115 inch Guide Tube Tk: 0.040 inch Eff. Array Dimension: 8.12 inch x 8.12 inch Active Fuel Length: 136.7 inch CE 14x14

'\WWWWWW ENC 14xl4 Enrichment, wtX 3.7 4.0 (assumed)

Pel]et OD, inch 0.3765 0.3700 Stacked Fuel Density, g/cc 10.054 10.199 Clad Tk, inch 0.028 0.031 Moderator-to-Fuel Volume Ratio(>) 1.934 2.034 Axial U-235 Loading, g/cm '41.45 43.91(2) k (CCELL) 1.447 1.462 This ratio takes into account the zirconium tubing and water associated with the instrument and control rod locations within the fuel assembly.

This value corresponds to an axial U-235 loading at 4.0 wtX and does not include the axial blankets of natural uranium.

13 XN-NF-83-36, Rev. 1 Table II St. Lucie Unit 1 Spent Fuel Storage Pool Reactivity Sensitivity Calculation Results (NITAML/XSDRNPH)

Nominal Storage Case Oescription Array 6k~ Comments 1 (base) Enrichment - 4.0 wtX Nominal Arrangement Temperature - 6BoF Mall Thickness - 0.25" C-C Spacing - 12.4375" 2a Mall Thickness - 0.20" + 0.0052 2b Mall Thickness - 0.10" + 0.0249 3a C-C Spacing - 12.0000" + 0.0199 Min. Offset Condition 3b C-C Spacing - 11.5625" + 0.0462

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14 XN-NF-83-36, Rev. 1

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Table III St. Lucie Unit 1 New Fuel Storage Room Reactivity Calculation Results Fuel Assembly: ENC 14x14 (4.0 wtX U-235)

Number of Assemblies in Room: 80 Array Geometry: Per Ebasco Services Dwg No. 8770-6-832, Rev. 3 Calculation Model: KENO IV with 123 group cross sections prepared using NITAWL/XSDRNPH Case Vol% Hater in Array keff + ~ Comnents 2.5 0.825 + 0.006 5.0 0.905 + 0.005(1) Max. Calculated keff 10.0 0.898 + 0.006 15.0 0.811 + 0.006 Peak keff estimated to be about 0.925 (at the 95K confidence level) and to occur between 2.5 and 5.0 vol% water in the storage array.

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XN-NF-83-36, Rev. I APPENDIX I SECOND-PARTY REVIEM DOCUMENTATION

HJ$ 84 UGLEULR coMPANY,Oc.

'i ~ ~ ~ XN-NF-83-36 Rev. 1 internal Correspondence Distribution Date: May 3, 1983 To: C. O. Brown From: J. E. Pieper

Subject:

CRITICALITY SAFETY ANALYSIS POR ST. LUCIE UNIT 1 NEW AND SPENT FUEL STORAGE, XN>>NF-83-36 Ref: A. S. Jameson (Canbustion Engineering) to R. W. Winnard (Florida Power t Light), September 18, 1979, "Bases for Updating the St. Lucie-1 Technical Specification 5.3.1".

I have completed a reviev of the calculations and the analysis approach used in the study of the St. Lucie Unit 1 new, spent fuel storage and fuel inspection elevator, upender and transfer tube. I believe that in conjunction vith the referenced letter, your analysis adequately deaan-strates the criticality safety of Exxon Nuclear 4.0 vt.% U fuel design when used in this equipment.

The cross-section library used for the analyses vere traced through the HITAWL, XSDRNPM, NITAWL code stream. The KENO and CCELL runs vere also revieved.

The folloving s~cific runs vere revieved:

Job Name Date of Run Time of Run CCELL XCEL 940 25/03/83 12 55.19 NITAWL NZTAWGE 28/03/83 13 34 40 XSDRNPM XSDRNMP 28/03/83 13.57.27 NZTAWL NZT?lW81, 28/03/83 15.26.40 ZENDZV KSTLUGQ 28/03/83 19.20.06 XSDRNPM XSDRNRU 05/04/83 20.42.17 XSDRNPM XSDRHRX 05/04/83 20 24.10 XSDRNPM XSDRNRO 05/04/83 20.46.15

'SDRNPM XSDRNRZ 05/04/83 20.46.27 XSDRNPM XSDRN7U 04/04/83 19 29.26 XSDRNPM XSDRNDR 31/03/83 13.20.36 NZTAWL NIT?LNH3 31/03/83 12.57.20 CCELL XCE19LG 14/04/83 16.04.53 XENO-2 KEN?2OZ 19/04/83 19.54.07 CCELL XCEL9LF 08/21/83 12.58.58 CCELL XCEL9FU 05/04/83 11.56.45 CCELL XCEL9LC 05/04/83 12 26.17 CCELL XCE19LD 05/04/83 12.26.18 NZTAWL NZTAWST 08/04/83 13.24.21 NITAWL NZTAWSV 07/04/83 17.25.33

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XN-NF-83-36 Rev. 1 C. O. arche May 3, 1983 Job Name Date of Run Time of Run NIThWL NIThWUO 06/04/83 15.46.14 NITAHL NITKWSU .07/04/83 17.25.32 XSDRNPM XSDRNV1 08/04/83 15.46.14 XSDRNPM XSDRN9R 07/04/83 17.44.21 XSDRNPM XSDRNXB 06/04/83 15.58.31 XSDRNPM XSDRN95 07/04/83 17.44-35 KENO-IV XSTLU2D 08/04/83 17.59.02 KENO-IV XSTLUG3 07/04/83 22.15.34 XENO-IV XSTLUSP 07/04/83 20.11.34 KENO-IV KSTLUHL 07/04/83 22.15.36 The following items ~e included in the review of the codes:

XENO: Options Corz'ect Mixtures Geometry Adequate Histories

.

eff Sections (Data or Lihrary tape)

Cross CCELL Geometry Atom Densities Verified Resonance Treatment Options X

NITASL: Input Tape Output Tape ZD Runners Resonance Parameters A>here used)

XSDRNPM: Input Tape Output Tape Gecxnetzy Options KQctures Convergence clc

XN-NF-83-36, Rev. 1 Issue Date: 2/28/86 ST. LUCIE UNIT 1 NEM AND SPENT FUEL STORAGE CRITICALITY SAFETY EVALUATION FOR NATURAL URANIIN AXIAL BLANKET FUEL 1

.DISTRIBUTION

0. C. Brown

'L. D. Gerrald T. J. Helbling (5)

J. S. Holm T. M. Patten G. N. Mard Document Control (2)

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