ML20066E127
| ML20066E127 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 11/01/1982 |
| From: | ARKANSAS POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20066E110 | List: |
| References | |
| NUDOCS 8211120437 | |
| Download: ML20066E127 (99) | |
Text
{{#Wiki_filter:i j ARKANSAS POWER & LIGHT COMPANY ARKANSAS NUCLEAR ONE
- UNITS 1 & 2 SPENT FUEL POOL RERACK LICENSING SUBMITTAL
?
November 1, 1982 Pr i 8211120437 821105 . DRADOCK05000g
+
1 1 Y TABLE OF CONTENTS s
1.0 INTRODUCTION
2.0 GENERAL DESCRIPTION 2.1 Region 1 Design 2.2 Region 2 Design 3.0 DESIGN BASES ' 3.1 Design Loads 3.2 Specified Loads and Definitions 3.3 Applicable Codes, Standards, and Regulations 4.0 NUCLEAR CONSIDERATIONS 4.1 Introduction 4.2 Criticality Analysis for Region 1 4.3 Criticality Analysis for Region 2 4.4 Acceptance Criterion for Criticality 5.0 THERMAL-HYDRAULIC CONSIDERATIONS < 5.1 Spent Fuel Pool Cooling System Design Bases 5.2 Local Fuel Bundle Thermal-Hydraulic Assum tions and Considerations q, 5.3 Description of Analytical Method and Types of Calculations Performed l 5.4 Acceptance Criteria 5.5 Rack / Pool Thermal-Hydraulics 6.0 STRUCTURAL AND SEISMIC CONSIDERATIONS 6.1 Assumptions 6.2 Seismic Analysis 4 k J
]
i j TABLE OF CONTENTS (continued)
/
6.3 Fuel Rack Structural Analysis 6.4 Fuel Handling Crane Uplift Analysis 6.5 Fuel Bundle / Module Impact Evaluation 1
, 6.6 Acceptance Criteria 6.7 Pool Structural Analysis 7.0 MATERIAL CONSIDERATIONS q
8.0 INSTALLATION CONSIDERATIONS i 3 8.1 Rack Modules Handling Conside.ations
,, 8.2 Radiation Protection Considerations 9.0 R/.DI0 LOGICAL CONSIDERATIONS 9.1 GeneralDedcriptionofRadiologicalAspects
- / 9.2 Description ci Radiological Aspects of Increased j j, Spent Fuel Storage 10.0 ACCIDENT EVALUATIONS 10.1 Cask Drop Accident Analysis-i 10.2 Fuel Assembly Drop Accident Analysis 10.3 Loss of Spent Fuel Pool Cooling 11.0 COST AND BENEFIT ASSESSMENT 11.1 Alternatives 11.2 Commitments of Material Resources
12.0 REFERENCES
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1.0 INTRODUCTION
This report is submitted in support of Arkansas Power and Light Company's application to amend the Arkansas Nuclear One operating licenses for reracking modifications to the Arkansas Nuclear One Unit-1 (ANO-1) and Unit-2 (ANO-2) spent fuel storage pools. This report is in accordance with the guidelines of the Nuclear ; Regulatory Commission's " Standard Review and Acceptance Plan for Spent Fuel Storage and Handling Applications." These proposed modifications will increase spent fuel storage capacities by replacing the present spent fuel storage racks with storage racks as described in Section 2.0 of this report. Reracking the ANO spent fuel pools will increase the ANO-1 pool storage capacity from 589 spaces to 968 spaces and the ANO-2 pool storage capacity from 485 spaces to 988 spaces. Arkansas Power and Light Company is responsible for the modification of the spent fuel storage pools. Westinghouse Electric Corp. is retained to design, analyze, fabricate, and provide technical assistance during the installation of the spent fuel racks. Structural Dynamics Technology, Inc. is retained to perform the spent fuel pool structural analysis and provide technical assistance. 1
2.0 GENERAL DESCRIPTION The spent fuel pool and its associated systems are described in the ANO-1 FSAR Section 9.6 and in Section 9.1.3 of the ANO-2 FSAR. Figures 9-11 and 9.1-1 are reproductions of FSAR figures which show the general arrangement of the ANO-1 & 2 pools and the associated fuel handling equipment. This arrangement is not changed as a result of this modification. The spent fuel storage pool rack arrangement for ANO-1 is shown in Figure 2-1 and the arrangement for AN0-2 is shown in Figure 2-2. Fuel storage is divided into two regions within each pool. Region 1 (approximately 200 assemblies), is of high density fuel assembly spacing obtained by utilizing a neutron absorbing material and is reserved for core off loading. Region 2 is also of a high density fuel assembly spacing providing normal storage for spent fuel assemblies. Region 1 is designed to accommodate non-irradiated fully enriched fuel. Region 2 is designed to accomodate irradiated fuel that has sustained approximately 80 percent of the design burnup. Placement of fuel in Region 2 is determined by burnup calculations and controlled administrative 1y by Arkansas Power & Light. Fuel which does not meet this criterion may be placed in Region 2 in a checker board fashion. In these cases, vacant spaces surrounding the assembly being inspected will be controlled administratively to prevent inadvertent assembly insertion. No physical barrier is necessary between the two regions. The 2 , l
I racks meet the requirements of the NRC Position for Review & ' Acceptance of Spent Fuel Storage and Handling Applications, I dated April 14, 1978, and modified January 18, 1979, with the exception that, for Region 2 storage, credit is taken for fuel burnup based on the proposed Revision 2 of USNRC Regulatory Guide 1.13.
; 2.1 REGION 1 DESIGN The Region 1 storage racks are composed of individual storage cells made of stainless steel. These racks utilize a neutron absorbing
- material, Boraflex, which is attached to each cell. The cells within a module are interconnected by grid assemblies to form an integral structure as shown in Figure 2-3. Each rack module is provided with leveling pads which contact the spent fuel pool floor and are remotely adjustable from above through the cells at installation. The modules are neither anchored to the floor nor braced to the pool walls.
l The fuel rack assembly consists of three major sections which are s the leveling pad assembly, the lower and upper grid assembly, and the cell assembly. Figure 2-4 illustrates these sections. The major components of the leveling pad assembly are the leveling pad and the leveling pad screw. The top of the support plate is l 3
welded to the base plate. The leveling pad assemblies transmit the loads to the pool floor and provide a sliding contact. The leveling pad screw permits the leveling adjustment of the rack. The lower grid attaches the cell assembly to the base plate. The lower grid consists of box-beam members and the base plate. The bottom of the cell assembly is welded to the lower grid. The upper grid consists of the box-beam members. The upper part of the cell assembly is welded to the upper grid. The upper and lower grid assemblies maintain the center-line to center-line spacing between the cells and provide the structural connections between the cells to form a fuel rack assembly. The major components of the cell assembly are the fuel assembly cell, the Boraflex (neutron absorbing) material, and the wrapper. The wrapper is attached to the outside of the cell by spot welding the entire length of the wrapper. The wrapper covers the Boraflex material and also provides for venting cf the Boraflex to the pool environment. Depending on the criticality requirements, some cells have a Boraflex wrapper on all four sides, some on three sides, and some on two sides. 4
2.2 REGION 2 DESIGN The Region 2 storage racks consist of stainless steel cells assembled in a checkerboard pattern, producing a honeycomb type structure. Each cell has attached to its outer wall a stainless steel wrapper plate creating a pocket opened at the top and bottom. This is referred to as a " spacer pocket" design and is shown in Figure 2-5. lhe spacer pockets are designed to accept poison inserts if future need arises. This design is also provided with leveling pads which contact the spent fuel pool floor and are remotely adjustable from above through the cells at installation. The modules are neither anchored to the floor nor braced to the pool walls. The fuel rack assembly consists of two major sections which are the base support assembly and the cell assembly. Figure 2-6 illustrates these sections. The major components of the base support assembly are the leveling pad, the leveling pad screw, and the support plate. The top of the support plate is welded to the fuel rack base plate. The leveling pads transmit the loads to the pool floor and provide a sliding contact. The leveling pad screw permits the leveling adjustment of the rack. 5
The components of the cell assembly are the cell enclosure and the wrapper plates. I The wrapper plates are attached to the outside of the cell by spot welding along the entire length of the wrapper. The wrapper forms the pocket and establishes the size of the noncell locations. Rack module data for both units are described in Table 2-1. 6
-- . ._ . . - - . - . _ - - - . . . - - . - . . . . . .~ - _ _ - -. ._ - - -
l ,' TABLE 2-1 Unit 1 Unit 2 Region 1 Region 2 Region 1 Region 2 Number of Storage 220 748 234 754 Locations Number of Rack 2 (10x11) 2 (11x12) 2 (9x9) 4 (9x10) Arrays 4 (11x11) 1 (8x9) 2 (8x10) 2 (9x9) l 1 (8x9) Center-to-Center 10.65 10.65 9.8 9.8 Spacing (Inches) Cell I.D. (Inches) 8.97 8.97 8.58 8.58 Type of Fuel B&W 15x15 B&W 15x15 C.E. 16x16 C.E. 16x16 Rack Assembly (10x11) (11x12) (9x9) (9x10) 4 Dimensions (Inches) 106.7x117.4x169.0 118.2x128.8x169.0 89.2x89.2x188.875 89.2x99.0x188.875 j (11x11) (8x9) (8x10) 118.2x118.2x169.0 79.4x89.2x188.875 79.4x99.0x188.875 i (9x9) j 89.2x89.2x188.875 l ~ (8x9) 79.4x89.2x188.875 Dry Weights (lbs) 27,500 (10x11) 19,500 (11x12) 20,300 (9x9) 16,000 (9x10) 18,000 (11x11) 18,000 (8x9) 14,200 (8x10) 14,400 (9x9) 13,000 (8x9) i i l 7 l . . .
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i , J i f 3.0 DESIGN BASES 1 j The function of the. spent fuel storage racks is to provide for i storage of spent fuel assemblies in a flooded pool, while maintaining a coolable geomotry, preventing criticality, and I protecting the fuel assemdlies from excess mechanical or thermal loadings. 1 i i 3.1 DESIGN LOADS 1 The dry weights for each type of specific module are listed in J , Table 2-1. 1 3.2 SPECIFIED LOADS AND DEFINITIONS 1 i
- The following are load combinations specified for the modified 4
racks: 4 i r Elastic Analysis Acceptance Limits , (1) D + L Normal Limits of NF 3231.la (2) D + L + E Normal Limits of NF 3231.la { (3) D + L t T g Lesser of 2 Sy or S Stress Range u l (4) D + L + T g +E Lesser of 2 Sy or S Stress Range u { -(5) D + L + T a +E Lesser of 2 Sy or S uStress Range (6) D + L + Ta + E' Faulted Condition Limits of NF 3231.lc i l 17 L \ l-l
Definitfor.s: 0 - Dead loads or their related internal moments a.1d forces including any permanent equipment and hydrostatic loads. L - Live loads or their related internal ~ moments and forces including any movable equipment loads. E - Operational Basis Earthquake Loads E' - Design Basis Earthquake Loads T g Operating Thermal Loads T a Accident Thermal Loads 3.3 APPLICABLE CODES, STANDARDS, AND REGULATIONS The modified spent fuel storage racks are designed by Westinghouse Electric Corp. with the applicable provisions of the following codes, standards, and regulations:
"NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" dated April 14, 1978 and revised January 18, 1979, with the exception that credit is taken for burnup for all storage locations except for those reserved for full core discharge (in Region 1 of both Units). -
18
3.3.1 NRC Regulatory Guides R.G. 1.13 Proposed Revision 2 - Spent Fuel Storage Facility Design Basis R.G. 1.29 Seismic Design Classifications R.G. 1.44 Control of the Use of Sensitized Stainless Steel R.G. 1.60 Design Response Spe..ra for Seismic Design of Nuclear Power Plants R.G. 1.61 Damping Values for Seismic Design of Nuclear Power Plants R.G. 1.92 Combining Nodal Responses and Spatial Components in Seismic Response Analysis R.G. 1.124 Service Limits and Loading Combinations for Class I Linear-Type Component Supports 3.3.2 NRC Standard Review Plans SRP 3.7 Seismic Design SRP 3.8.4 Other Category I Structures 19
SRP 9.1.2 Spent Fuel Storage SRP 9.1.3 Spent Fuel Pool Cooling and Cleanup System 3.3.3 Industry Codes and Standards American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NF. American National Standards Institute, N210-1976, " Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations." American National Standards Institute, N16.1-1975, " Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors." l l 20
4.0 NUCLEAR CONSIDERATIONS
4.1 INTRODUCTION
The spacer pocket spent fuel rack design is a high density rack design in which explicit credit is taken for tne reduction in reactivity due to fuel burnup. The spent fuel rack design described herein employs two separate and different arrays which will be considered as two separate spent fuel racks. The smaller array, referred to as Region 1, is designed on the basis of the currently accepted NRC guidance on spent fuel rack design.E13 The larger array, Region 2, is designed to take into consideration the changes in fuel and fission product inventory resulting from depletion in the reactor core. 4.2 CRITICALITY ANALYSIS FOR REGION 1 1 4.2.1 Neutron Multiplication Factor Criticality of fuel assemblies in the spent fuel storage rack is prevented by the design of the rack which limits fuel assembly interaction. This is done by fixing the minimum separation between assemblies and inserting neutron poison between assemblies. 21
The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95 percent probability at a 95 percent confidence level that the effective multiplication factor (K,ff) of the fuel assembly array will be less than 0.95 as recommended in ANSI N210-1976 and in Reference 1. The following are the conditions that are assumed in meeting this design basis. 4.2.2 Normal Storage
- a. The fuel assembly contains the highest enrichment authorized without any control rods or any noncontained burnable poison and is at its most reactive point in life.
The following assembly parameters were modeled: Unit 1 Unit 2 Number of Fuel Rods per Assembly 208 236 Rod Zirc-4 Clad 0.D. (inch) 0.430 0.382 Clad Thickness (inch) 0.0265 0.025 Fuel Pellet 0.D. (inch) 0.3686 0.325 Fuel Pellet Density (% TD) 95 95 1 Fuel Pellet Dishing (%) 2.03 3.37 Rod Pitch (inch) 0.568 0.506 22
Number Zirc-4 Guide Tubes 17 5 Guide Tube 0.D. (inch) 0.530 0.980 Guide Tube Thickness (inch) 0.016 0.04 f The assembly is conservatively modeled with water replacing the assembly grid volume and no U-234 or U-236 in the fuel pellet. No U-235 burnup is assumed.
- b. The storage cell nominal geometry is shown on Figure 4-3.
- c. The moderator is pure water at the temperature within the design limits of the pool which yields the largest reactivity. A conservative value of 1.0 gm/cm3 is used for the density of water. No dissolved boron is included in the water.
- d. The array is either infinite in lateral extent or is surrounded by a conservatively chosen reflector, whichever is appropriate for the analytical model. The nominal case calculation is infinite in lateral and axial extent.
Poison plates are not necessary on the periphery of the rack module except for the sides of the module adjacent to another rack module. Calculations for the racks with the poison removed on sides adjacent to the pool wall indicate a less reactive configuration than the nominal case of an infinite rack. Therefore, the nominal case of an infinite array of poison cells is a conservative assumption. I 23 l 1
- e. Mechanical uncertainties and biases due to mechanical tolerances during construction are treated by either using
" worst case" conditions or by performing sensitivity studies to obtain the appropriate values. The items included in the analysis are:
poison pocket thickness
- stainless steel thickness
- can ID
- center-to-center spacing
- can bowing The calculational method for treatment of uncertainty and bias is discussed in Section 4.2.4 of this report.
- f. Credit is taken for the neutron absorption in full length structural materials and in solid materials added specifically for neutron absorption. The minimum poison loading is assumed in the poison plates and B4 C particle self shielding is included as a bias in the reactivity calculation.
4.2.3 Postulated Accidents Most accident conditions will not result in an increase in K eff of the rack. Examples are the loss of cooling systems (reactivity decreases with decreasing water density) and 24
dropping a fuel assembly on top of the rack (the rack structure pertinent for criticality is not deformed and the assembly has more than eight inches of water separating it from the rest of the stored assemblies which precludes interaction). However, accidents can be postulated which would increase reactivity such as a theoretical example of an inadvertent drop of an assembly between the outside periphery of the rack and the pool wall. Therefore, for accident conditions, the double contingency principle of ANS N16.1-1976 is applied. This states that it shall require two unlikely, independent, concurrent events to produce a criticality accident. Thus, for accident conditions, the presence of soluble boron in the storage pool water can be assumed as a realistic initial condition. The presence of approximately 1600 ppm boron in the pool water will decrease reactivity by approximately 30% AK. In perspective, this is more negative reactivity than is present ; in the poison plates (25% AK) so K f r the rack would be eff less than 0.95 even if the poison plates were not present. Thus, K eff 50.95 can be easily met for postulated accidents, since any reactivity increase will be much less than the negative worth of the dissolved boron. i l l 25
For fuel storage applications, water is usually present. However, accidental criticality when fuel assemblies are stored in the dry condition is also accounted for. For this case, possible sources of moderation, such as those that could arise during fire fighting operations, are included in the analysis. The " optimum moderation" accident is not a problem in poisoned fuel storage racks. The presence of poison plates removes the conditions necessary for " optimum moderation" so that K eff continually decreases as moderator density decreases from 1.0 gm/cm3 to 0.0 gm/cm3 in poison rack designs. 4.2.4 Criticality Analytical Method The calculation method and cross-section values are verified by comparison with critical experiment data for assemblies similar to those for which the racks are designed. This benchmarking data is sufficiently diverse to establish that the method bias and uncertainty will apply to rack conditions which include strong neutron absorbers, large water gaps, and low moderator densities. l The design method which ensures the criticality safety of fuel assemblies in the spent fuel storage rack uses the AMPX system of codes [2,33 for cross-section generation and KENO IV E43 for reactivity determination. l l 26 l
,t
/
The 218 energy group cross-section library [23 that is the common starting point for all cross-sections used for the benchmarks and the storage rack is generated from ENDF/B-IV data. The NITAWL program [33 includes, in this library, the self-shielded resonance cross-sections that are appropriate for each particular geometry. The Nordheim Integral Treatment is used. Energy and spatial weighting of cross-sections is
- performed by the XSDRNPM program E33 which is a one-dimensional Sn transport theory code. These multigroup cross-section sets are then used as input to KEN 0 IVE 43 which is a-three-dimensional Monte Carlo theory program designed for reactivity calculations.
A set of 27 critical experiments has been analyzed using the above method to demonstrate its applicability to criticality analysis and to establish the method bias and variability. The I experiments range from water moderated, oxide fuel arrays , separated by various materials (Boral, steel, water)~that simulate LWR fuel shipping and storage cc.aitions,E5'83 to dry, hardened spectrum uranium metal cylinder arrays with various interspersed materialsE73(Plexiglas,steelandair)that demonstrate the v5de range of applicability of the method. The rest't o ao descriptive facts about each of the 27 benchmark critical experiments are given in Table 4-1. The average K,ff of the benchmarks is 0.9998 which demonstrates that there is no bias associated with the method. The standard 27
t deviation of the K eff values is 0.0057 AK. The 95/95 one-sided s tolerance limit factor for 27 values is 2.26. Thus, there is a 95 percent probability with a 95 percent confidence level that
\
the uncertainty in reactivity, due to the method, is not t greater than 0.013 AK. The total uncertainty (TV) to be~added to a criticality calculation is: I ' TV = [(ks)2 method + (ks)2 nominal 3 where (ks) method is 0.013 as discussed above and (ks) nominal is the statistical uncertainty associated with the particular KENO calculation being used. The most important effect on reactivity of the mechanical tolerances is the possible reduct; in the water gap between the poison plates of adjacent storage cells. For ANO-1 and 2, the worst combination of mechanical tolerances (i.e., sheet metal thickness, cell I.D. maximum, rack grid assembly, and cell bowing) will result in a reduction of the water gap between adjacent cells. For a single can it is found that reactivity does not increase significantly because the increase in reactivity due to the water gap reduction on one side of the can is offset by the decrease in reactivity due to the increased water gap on the opposite side of this can. The analysis, for the effect of mechanical tolerances, however, assumes a worst case of a rack composed of an array of groups of four cans where the water gap
\
28 5 i
between the four cans is reduced. The reactivity increase of this configuration is found and is included as a bias term in calculating the K,ff of the rack. Some mechanical tolerances are not included in the analysis because worst case assumptions are used in the nominal case analysis. An example of this is eccentric assembly position. Calculations were performed which show that the most reactive condition is the assembly centered in the can which is assumed in the nominal case. Another example is the reduced width of the poison plates. No bias is included here since the nominal KEN 0 case models the reduced width explicitly. The final result of the uncertainty analysis is that the criticality design criteria are met when the calculated effective multiplication factor, plus the total uncertainty (TV) and any biases, is less than 0.95. These methods conform with ANSI N18.2-1973, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants", Section 5.7, Fuel' Handling System; ANSI N210-1976,
" Design Objectives for LWR Spent Fuel Storage Facilities at Nuclear Power Stations", Section 5.1.12; ANSI N16.9-1975,
" Validation of Calculational Methods for Nuclear Criticality Safety"; NRC Standard Review Plan, Section 9.1.2, " Spent Fuel Storage"; and the NRC Guidance, "NRC Josition for Review and Acceptance of Spent Fuel Storage and Landling Applications".
29
4.2.5 Rack Modification For normal operation and using the method described in the above sections, the K,ff for the rack is determined in the following manner: K,ff = Knominal + Bmech + Bmethod + Opart
+ [(ksnominal) + (ksmethod) where:
Knominal = n minal case KEN 0 K,ff Bmech = K,7f bias to account for the fact that mechanical tolerances can result in water gaps between poison plates less than nominal. Bmethod = method bias determined from benchmark critical comparisons. t B part
= bias to account for poison particle self-shielding.
ksnominal = 95/95 uncertainty in the nominal case KEN 0 K,ff. l ks method
= 95/95 uncertainty in the method bias, l
l l 30 l - _ _ _ _, ._
-. . . , - _ = - -. .- .- - = -. . _ . - . _ _
t Substituting calculated values, K,7f is shown to be.less than 0.95. Since K,ff is less-than 0.95 including uncertainties at a 95/95 probability / confidence level, the acceptance criterion for criticality is met. 4.3 CRITICALITY ANALYSIS FOR REGION 2 +
- 4.3.1 Analytical Methods The methods used in the analysis of Region 2 include NITAWL,-
l
- XSDRNPM and KENO-IV for basic reactivity determination, along with LEOPARD [83, CINDER E 93 and TURTLE E 103 for. reactivity equivalencing. LEOPARD and CINDER are used to calculate the isotopic compositions and cross-sections of the fuel as a
! function of burnup history and subsequent decay time. These cross-sections are then input to the TURTLE code, which is used ' to determine the reactivity equivalence (in the Region 2 rack) l of assemblies with different initial enrichments and burnups. j The reactivity equivalencing is extended back to an unirradiated assembly, which is then analyzed using NITAWL, XSDRNPM and KEH0-IV in a fashion similar to the analysis for Region 1. l l The accuracy of the burnup dependent isotopics is given in Table 4-2. These measurements were taken from the Saxton Core II 113 E and the LEOPARD predictions show excellent agreement. The agreement between measurement and prediction 31
not only verifies the accuracy of the isotopic predictions, it also verifies the accuracy of the cross-sections of the actinides and therefore indirectly the reactivity worth. In order to account for uncertainties in the prediction of the actinide number densities, an uncertainty of 5 percent of the worth of the actinides (.009 = .05 x .18) will be applied to the fincl rack multiplication factor. The accuracy of the reactivity calculations is shown in Table 4-3, giving the results of 101 critical experiments analyzed with LEOPAR0[123, When comparing the reactivities of spent and unirradiated fuel assemblies, an uncertainty arises in the ability to predict the depletion of the actinides in concert with the accumulation of fission products. Due to a lack of clean experimental information on the reactivity of spent fuel, the uncertainty must be inferred from comparisons of actual and calculated reactor reactivity lifetimes using LEOPARD and TURTLE. Lifetime calculations are typically within 2 percent of predictions. An extremely conservative estimate of the uncertainty due to depletion can be made by taking 5 percent of l the change in reactivity due to burnup. As an example, take a l B&W 15x15 assembly with a 3.10 w/o initial enrichment at 21000 MWD /MTV. The total reactivity change from fresh to spent is .188 AK. The corresponding uncertainty in the Region 2 multiplication factor is then .0094 AK (= .05x.188 AK). 32
1 In order to guarantee the applicability of these calculations ' for long-term storage, fission product decay after discharge was taken into account using CINDER. The fission products were permitted to decay for 30 years after discharge, and the time at which the cell reactivity peaked was chosen for the design 4 basis. The maximum reactivity occurs at approximately 100 hours after shutdown (primarily due to the decay of Xe135),at which point it begins to decrease, continuing throughout the 30 year time span. 4.3.2 Reactivity Equivalency s One of the basic principles behind the spacer pocket design is the concept of reactivity equivalencing. In this concept, a constant rack km contour is constructed in enrichment-burnup space using LEOPARD and TURTLE. The intersection point at zero burnup is then calibrated using KENO-IV. Figure 4-2 shows the constant km contour based on a high enrichment endpoint of 4.10 l w/o and 36000 MWD /MTU. The advantage of this approach is that l
- LEOPARD and TURTLE are used only to calculate relative reactivities as a function of burnup (a calculational ability
- qualified by many years of reactor design experience), while l the actual rack reactivity determination is performed by the more powerful Monte Carlo method.
i i 33 i i
The principal motivation behind reactivity equivalencing is the relationship between assembly km and rack km as a function of initial enrichment. If a constant assembly km contour is constructed in enrichment /burnup space, the rack km increases as the enrichment increases. If the rack is designed to contain assemblies with high initial enrichments, a substantial amount of usable margin at lower enrichments would be lost by using the assembly km contour rather than the rack km contour. Reactivity equivalencing eliminates this unnecessary conservatism and permits more flexible storage capability at lower burnups. 4.3.3 Reactivity Determination The final K eff f r Region 2 is determined using the same methods described in Section 4.2 for Region 1. The actual conditions for this determination are defined by the zero burnup intercept point in Figure 4-2. In this instance, the intercept point is at 1.4 w/o U235 The design model for Region 2 will therefore be an unirradiated assembly at 1.4 w/o initial enrichment. The K eff f r Region 2 is also determined assuming a checker board storage configuration, i.e., alternate spacer pocket occupation, with unirradiated assemblies at 4.1 w/o enrichment. 34
Adjacent vacant spaces between stored assemblies in this case will be controlled administratively to prevent assembly introduction. 4.3.4 Postulated Accidents Most accident conditions will not result in an increase in K,77 of the rack. Examples are the loss of cooling systems (reactivity decreases with decreasing water density) and dropping a fuel assembly on top of the rack (the rack structure pertinent for criticality is not deformed and the assembly has more than eight inches of water separating it from the active fuel in the rest of the rack which precludes interaction). However, accidents can be postulated which would increase reactivity such as inadvertent drop of an assembly between the outside periphery of the rack and the pool wall. Therefore, for accident conditions, the double contingency principle of ANS N16.1-1975 is applied. This states that it is unnecessary i I to assume two unlikely, independent, concurrent events to ensure protection against a' criticality accident. Thus, for accident conditions, the presence of soluble boron in the storage pool water can be assumed as a realistic initial condition since its absence would be a second unlikely event. l 35
The presence of the approximately 1600 ppm boron in the pool water will decrease reactivity by approximately 30% AK. Thus K,77 5 0.95 can be easily met for postulated accidents, since any reactivity increase wi;I t,c :;;uch less than the negative worth of the dissolved boron. For fuel storage applications, water is usually present. However accidental criticality when fuel assemblies are stored in the dry condition is also accounted for. For this case, possible sources of moderation, such as those that could arise during fire fighting operations, are included in the analysis. This " optimum moderation" accident is not a problem in fuel storage racks because possible water densities are too low (5 0.01 gm/cm3) to yield K gf values higher than for full density water and the rack design prevents the preferential reduction of water density between the cells of a rack (e.g., boiling between cells). 4.3.5 Manufacturing Biases The construction tolerances for the spacer pocket racks allow for the nominal center-to-center spacing to be randomly reduced for individual cells. This change will result in an increase in K eff which will be treated conservatively as a bias. The 36
effect of the tolerances on pocket height and material thicknesses also result in an increase in K eff which will be treated conservatively as a bias. Another center-to-center spacing reduction can be caused by the asymmetric assembly position within the storage cell. The inside dimensions of a nominal storage cell are such that if a fuel assembly is loaded into the corner of the cell, the assembly centerline will be displaced from the cell centerline. This means that adjacent asymmetric fuel assemblies would have their center-to-center distance reduced from the nominal. Analysis shows this reduction may increase reactivity. This will also be treated as a bias although the asymmetric positioning of assemblies within storage cells will in reality be random. The final K,ff for Region 2 is constructed according to the following formula:
+
K,ff = K nom +Bmeth + Bmech + B asym + [(ksmeth) + (ksnom) r (kspu) + (ksbu) where: K is the eigenvalue from KEN 0 for the nominal storage l nom l l configuration, l 37
B meth is the bias in the method, B mech is the bias induced by material thickness variations and mechanical tolerances, B asym is the bias induced by the potential placement of the assemblies asymmetrically in the can, ks meth is the method uncertainty (95/95), ks nom is the uncertainty (95/95) on the nominal eigenvalue, ks pu is the uncertainty on the plutonium reactivity, and ks bu is the uncertainty on reactivity as a function of irradiation. While it may be argued that ks bu and ks pu are not independent and should not be combined statistically, it should be considered that the reactivity of fuel as a function of burnup dependsimplicitlyonthedroductionrateofplutonium. The two uncertainties are so closely related that accounting for
- them twice is a conservative form of double acounting.
38
, The final K,ff for Region 2 from this construction is less than 0.95, including all uncertainties at a 95/95 probability / confidence level. Therefore, the acceptance criterion for criticality is met.
'T 4.4 ACCEPTANCE CRITERION FOR CRITICALITY l The neutron multiplication factor in spent fuel pools shall be less than or equal to 0.95, including all uncertainties, under all l conditions. Generally, the acceptance criterion for postulated accident conditions can be K,ff 5 0.98 because of the' accuracy of the methods used coupled with the low probability of occurrence. For instance, in ANSI N210-1976, the acceptance criterion for the " optimum moderation" condition is K,ff 5 0.98. However, for storage pools j which contain dissolved boron, the use of realistic initial conditions ensures that K,ff (( 0.95 for postulated accidents as i discussed in Sections 4.2.3 and 4.3.4. Thus, for simplicity, the ! acceptance criterion for all conditions will be K,ff 5 0.95. i l 1 39
TABLE 4-1 E BENCHMARK CRITICAL EXPERIMENT 4'S'83 General Enrichment Separating Characterizing Description w/o U235 Reflector Material Separation (cm) K,ff
- 1. UO 2.35 water water 11.92 1.004 i .004
- 2. 2 r" d lattice " " "
8.39 0.093 i .004
- 3. 6.39 1.005 i .004
- 4. " "
4.46 0.994 i .004 S. Stainless steel 10.44 1.005 i .004
- 6. 11.47 0.99'. t .004
- 7. 7.76 0.c _ i .004
- 8. 7.42 1 :s4 i .004
- 9. boral 6.34 1. s t .004
- 10. 9.03 0. t .004 11.
5.05 1.001 i .004
- 12. 4.29 water 10.64 0.999 i .005 13.
Stainless steel 9.76 0.999 i .005 14. 8.08 0.998 i .006
- 15. .
boral 6.72 0.998 i .005
- 16. U metal cylinders 93.2 bare air 15.43 0.998 i .003 i 17. paraffin air 23.84 1.006 i .005 18.
-bare air 19.97 1.005 i .003
- 19. paraffin air 36.47 1.001 i .004 20.
bare air 13.74 1.005 i .003
- 21. paraffin air 23.48 1.005 i .004 l
- 22. bare plexiglas 15.74 1.010 i .003 23.
paraffin plexiglas 24.43 1.006 i .004
, 24. bare plexiglas 21.74 0.999 i .003
] 25. paraffin plexiglas 27.94 0.994 i .005 l 26. bare steel 14.74 1.000 i .003
- 27. bare plexiglas steel 16.67 0.996 i .003 40
TABLE 4-2 SAXTON CORE II IS0 TOPICS R0D MY, AXIAL ZONE 6 20 Precision LEOPARD Atom Ratio Measured (a) (%) _ _ _ _ Calculation
~ -
U-234/U 4.65 x 10 5 129 4.60 x 10 s U-235/U 5.74 x 103 10.9 5.73 x 10 3 U-236/U 3.55 x 10 4 15.6 3.74 x 10 4 U-238/U 0.99386 10.01 0.99385
~ -
Pu-238/Pu 1.32 x 10 3 2.3 1.222 x 10 3 Pu-239/Pu 0.73971 10.03 0.74497 Pu-240/Pu 0.19302 0. 2 0.19102 Pu-241/Pu 6.014 x 10 2 10.3 5.47 x 10 2
~ -
Pu-242/Pu 5.81 x 10 3 0.9 5.38 x 10 3 Pu/U(b) 5.938 x 10 2 0.7 5.970 x 10 2
- ~
Np-237/U-238 1.14 x 10 4 115 0.86 x 10 4 l Am-241/Pu-239 1.23 x 10 2 115 1.08 x 10 2 Cm-242/Pu-239 1.05 x 10 4 110 1.11 x 10 4
- ~
Cm-244/Pu-239 1.09 x 10 4 120 0.98 x 10 4 (a) Reported in reference 11 (b) Weight ratio 41
TABLE 4-3 BENCHMARK CRITICAL EXPERIMENTS LEOPARD COMPARISONS 4 Description of Number of LEOPARD k,ff Using Experiments (a) Experiments Experimental Bucklings UO2 Al clad 14 1.0012 SS clad 19 0.9963 Borated H20 7 0.9989 Subtotal 40 0.9985 .; U-Metal
- Al clad 41 0.9995 l Unclad 20 0.9990 Subtotal 61 0.9993 l
Total 101 0.9990 t f I l (a) Reported in reference 12 L 42
O 4 9 h
+t g #N' \@
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\ \ '
C s (x \ \ N N \ \ N \ \ \i
- u. \ s
\
4
.N x
\ s s s , s s s N
mP s 9 43
FIGURE 4-2 MIf1IllCM BURNUP VS. INITIAL ENRICHMENT FOR REGION 2 STORAGE 36 -=- =- n- - - - - - - - - - - - - - - - - - - -
= =- = - t =:- u,==n=.====;=========...._....._=r====f=---
= - - - - - - - - - - -
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. . . . . .. . =--'.=:r=.iiH:-5_8.::'
?i=~ G = n=# -:i=E =r-C.== m==;==9====tM=.= := #=5~2N=M.= ~~
....- ._-J2......._ -
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= = n .7 _
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,4 , _
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=- -
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- s. = = = =- r= --_: -- -- z= ==-:_ ._ -. i _ . . _ - . . _ . . _ . . . _ _ _ _
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20 ' a,_
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?._=-_r+.===
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. = = ==
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. .. . . .. .. a ( . .L.. .
O _ _ . 1.0 2.0 3.0 4.0 Initial Assembly Average Enrichment (w/o U-235 ) 44 - o
L Figure 4-3 Region 1 Cell I.ayout 6
'! t 4
~E 4 .*[f. 'k,
(~-T! ----,# 3 - , _~~__3 _ , _ - ,__g, u, [ c.11 Inr.1osurn , k x m,ca n e,
'i L. Can ^
s _ . ._.._...__.-~ I.D. i
- h. {
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,
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5.0 THERMAL-HYDRAULIC CONSIDERATIONS The ANO spent fuel pool cooling and purification systems are designed to maintain water quality and clarity and to remove the decay heat from the stored fuel. 5.1 SPENT FUEL POOL COOLING SYSTEM DESIGN BASES The ANO-1 cooling system was originally designed to maintain the spent fuel pool water at 150 F with a heat load based upon removing the decay heat generated from 1-1/3 cores. One batch irradiated for 930 days and cooled for 100 days, one batch irradiated for 720 days and cooled for 150 hours, one batch irradiated for 410 days and cooled for 150 hours, and one batch irradiated for 100 days and cooled for 150 hours. The ANO-2 cooling system was originally designed to maintain the pool temperature at 150 F with 2-3/4 cores in the pool assuming that one full core which contains three batches irradiated for one, two, and three years is placed in the pool seven days after reactor shutdown along with five batches of fuel from previous refuelings. f 5.1.1 Heat Loads and Pool Temperatures i The ANO-1 and 2 pool cooling systems have the capability of maintaining the spent fuel pool at or below 150 F with a heat load from the typical combinations of spent fuel listed in Tables 5-1 and 5-2 that have been cooled for 168 hours or more. 46
This analysis was performed using the guidelines of APCSB 9-2,
" Residual Decay Energy for Light Water Reactors for Long Term Cooling."
5.2 LOCAL FUEL BUNDLE THERMAL-HYDRAULIC ASSUMPTIONS AND CONSIDERATIONS A local fuel bundle thermal-hydraulic analysis is performed to determine the maximum fuel clad temperatures which may occur as a result of using the spent fuel racks in the Arkansas spent fuel pools. Key assumptions used in the analysis are:
- The nominal water level is 23 feet above the top of the fuel storage racks.
- The maximum fuel assembly spent fuel pool decay heat output is i 1.36 x 105 8tu/Hr.
l
- The maximum temperature of the water at the inlet to the storage cells is 150 F when' the cooling system is operational.
- Under postulated accident conditions, when no pool cooling systems are operational, the maximum temperature at the inlet to the cells is assumed to be equal to the saturation temperature at atmospheric pressure or 212 F.
47
5.3 DESCRIPTION
OF ANALYTICAL METHOD AND TYPES OF CALCULATIONS PERFORMED A natural circulation calculation is employed to determine the thermal-hydraulic conditions within the spent fuel storage cells. The model used assumes that all downflow occurs in the peripheral gap between the pools walls and the outermost storage cells and all lateral flow occurs in the space between the bottom of the racks and bottom of the pool. The effect of flow area blockage in the region is conservatively accounted, and a multi-channel formulation is used to determine the variation in axial flow velocities through the various storage cells. The hydraulic resistance of the storage cells and the fuel assemblies is conservatively modeled by applying large uncertainty factors to loss coefficients obtained from various sources. Where necessary, the effect of Reynolds Number on the hydraulic resistance is considered, and the variation in momentum and elevation head pressure drops with fluid density is also determined. The solution is obtained by iteratively solving the conservation equations (mass, momentum, and energy) for the natural circulation loops. The flow velocities and' fluid temperatures that are obtained are then used to determine the fuel cladding temperatures. An elevation view of a typical model is sketched in Figure 5-1 where the flow paths are indicated by arrows. Each cell shown actually l corresponds to a row of cells that are located at the same distance from the pool walls. This is more clearly shown in a plan view, Figure 5-2. 48 l
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As shown, the lateral flow area underneath the storage cells decreases as the distance from the wall increases. This counteracts the decrease in the total lateral flow that occurs because of flow that branches up and flows into the cells. This is significant because the lateral flow velocity affects both the lateral pressure drop underneath the cells and the turning losses that are experienced as the flow branches up into the cells. These effects are considered in the natural circulation analysis. The most recently discharged or " hottest" fuel ascemblies are assumed to be located in various rows during different calculat.ons in order to ensure that they may be placed anywhere within the pool without violating safety limits. In order to simplify the calculations, each row of the model must be composed of storage cells having a uniform decay heat level. This decay heat level may or may n,ot correspond to a specific batch of fuel, but the model is constructed so that the total heat input is correct. The " hottest" fuel assemblies are all assumed to be placed in a given row of the model in order to ensure that conservatively accurate results are obtained for those assemblies. In fact, the most conservative analysis that can be performed is to assume that all assemblies in the pool (or rows in the model) have the same decay heat rate. This maximizes the total natural circulation flow rate which leads to conservatively large pressure drops in the downcomer and lateral flow regions which reduces the driving pressure drop across the limiting storage location. 51
Since the natural circulation velocity strongly affects the temperature rise of the water and heat transfer coefficient within a storage cell,
.the hydraulic resistance experienced by the flow is a significant parameter in the evaluation. In order to minimize the resistance, the design of the inlet region of the racks has been chosen sur.1 as to maximize this flow area. Each storage cell has either one large flow opening or multiple smaller openings. The use of large and/or multiple flow holes virtually eliminates the possibility that all flow into the inlet of a given cell can be blocked by debris or other foreign material that may get into the pool.
In order to determine the impact of a partial blockage on the thermal-hydraulic conditions in the cells, an analysis is also performed for various assumed blockages. The analyses that have been described only address the flow through the storage cells. As noted in the discussion of criteria, it is also required that the flow and temperatures in the axial gap between adjacent storage locations be evaluated. In order to preclude the possibility of i stagnant conditions in these gaps, flow relief areas are provided at the bottom of the cell assemblies. This flow area also ensures that air or steam cannot be trapped in the rack structure. The thermal-hydraulic conditions in the gap region are evaluated by using a parallel path thermal-hydraulic model of the gap and cell under consideration. This analysis considers the gamma heat generation in the cell enclosure and cell wrapper in addition to the decay heat input. Using the cell flow velocity and driving pressure differential obtained from the previously 52
described pool analyses, the flow velocity in the gap and the axial temperature distributions of the coolant and structure are determined.
'The radial temperature distributions through the various components are also considered.
5.4 ACCEPTANCE CRITERIA The criteria used to determine the acceptability of the design from a thermal-hydraulic viewpoint are summarized as follows: 5.4.1 The design must allow adequate cooling by natural circulation and by flow provided by the spent fuel pool cooling system. The coolant should remain subcooled at all points within the pool when the cooling system is operational. When the cooling system is postulated to be inoperable, adequate cooling implies that the temperature of the fuel cladding should be sufficiently low that no structural failures would occur and that no safety concerns would exist. 5.4.2 For normal operations, the maximum pool temperature shall not exceed 150 F. For conservdtism, the temperatures of the storage racks and the stored fuel are evaluated assuming that ! the temperature of the water at the inlet to the storage cells is 150 F during normal operation. 53
5.4.3 The rack design must not allow trapped air or steam and direct ganna heating of the storage cell walls and the water between adjacent storage locations must be considered. 5.5 RACK / POOL THERMAL-HYDRAULICS 5.5.1 Design Bases The Spent Fuel Pools and Pool Cooling Systems are designed to keep the pool water temperature below 120 F for normal refueling operations and 150 F for full core discharge situations. Under normal refueling conditions the fuel is discharged over a four day period after at least three days cooling inside the reactor vessel. The full core discharge is expected to take four days also with three days cooling in the reactor vessel prior to moving any fuel. The heat released from the fuel stored in the pool is determined in accordance with Branch Technical Position APCSB 9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling." Tables 5-1 and 5-2 show the expected loading in the pools. In the event mixed oxide fuel'becomes available, the heat load in the pool will be slightly higher. The increase is apparent only in fuel which has decayed for a relatively long period of time and contributes little additional heat load to the pool. 54
t I , 5.5.2 System Description The Spent Fuel Cooling Systems are described in the ANO FSAR Section 9.4.2 for ANO-1 and Section 9.1.3 for ANO-2. 5.5.3 Design Evaluation During normal operation, the Spent Fuel Cooling System serves to maintain the pool water at temperatures below 120 F. Under normal conditions, the pool temperature is maintained at approximately 120 F as stated in the AN0 FSAR Section 9.6.2.4.3.5 for Unit 1 and Section 9.1.3.2 for Unit 2 by recirculating spent fuel cooling water from the spent fuel pool through pumps and cooler (s) and back into the pool. For maximum normal conditions, full core offload, the pool , temperature is maintained at approximately 150 F. An analysis of pool responses to loss of forced cooling is presented in Section 10.3 of this dccument. The heat loads shown in Tables 5-1 and 5-2 represent the largest h' eat loads expected in the spent fuel pools. The calculations are based upon 18-month cycles for both units. 55
TABLE 5-1 ANO-UNIT 1 SPENT FUEL OPERATING & COOLING TIMES To (EFPD) Ts (Days) Dis. Date # Assemb. Heat Load (108 Btu /Hrl Batch 1 500 6310 1/77 60 0.03 Batch 2 772 5914 2/78 56 0.05 Batch 3 1101 5489 4/79 61 0.08
- , Batch 4 891 4849 1/81 60 0.06 Batch 5 1074 4150 12/82 68 0.09 Batch 6 1171 3512 9/84 64 0.10 Batch 7 1198 2965 3/86 68 0.12 Batch 8 1099 2417 9/87 72 0.12 Datch 9 1068 1870 3/89 68 0.13 Batch 10 1068 1322 9/90 64 0.16 Batch 11 1068 775 3/92 64 0.30 Batch 12 1068 227 9, 33 64 1.19 Batch 13 862(1) 7 4/94 64 9.19 Batch 14 506 7 4/94 64 8.87 Batch 15 150 7 4/94 64 7.53 l t To = Operation Time Ts = Cooling Time (1) Full Core Discharge after 150 Days Operation into Cycle 15 56
TABLE 5-2 ANO-UNIT 2 SPENT FUEL OPERATING & COOLING TIMES To (EFPD) Ts (Days) Dis. Date # Assemb. Heat Load (108 Btu /Hr) Batch 1 325 6067 3/81 61 .0.03 Batch 2 612 5520 9/82 60 0.05 Batch 3 873 5155 9/83 56 0.07 Batch 4 892 4607 3/85 60 0.08 Batch 5 961 4060 9/86 52 0.08 Batch 6 1056 3512 3/88 68 0.11 Batch 7 1068 2965 9/89 68 0.12 Batch 8 1068 2417 3/91 68 0.12 Batch 9 1068 1870 9/92 68 0.14 Batch 10 1068 1322 3/94 68 0.18 Batch 11 1068 775 9/95 68 0.34 Batch 12 1068 227 3/97 68 1.39 Batch 13 852(1) 7 8/97 68 10.70 Batch 14 506 7 8/97 68 10.33 Batch 15 150 7 8/97 68 8.76 t To = Operation Time Ts = Cooling Time (1) Full Core Discharge after 150 Days Operation into Cycle 15 57
6.0 STRUCTURAL AND SEISMIC CONSIDERATIONS 6.1 ASSUMPTIONS The purpose of the seismic and stress analysis is to analyze the proposed modified spent fuel module under various loading conditions. The racks are evaluated for both operating basis earthquake (0BE) and safe shutdown earthquake (SSE) conditions and meet Seismic Category I requirements. A detailed stress analysis is performed to verify the acceptability of the critical load components and paths under normal and faulted conditions. The racks rest freely on the pool floor and are evaluated to ensure that under various loading conditions they do not impact each other, nor do they impact the pool walls. 6.2 SEIMIC ANALYSIS The dynamic response of the fuel rack assembly during a seismic event is the co;.4ition which produces the governing loads and stresses on the strut ure. The dynamic response, internal stresses, and loads are obtained from a siismic analysis which is performed in two phases. The first phase is a time history analysis on a , simplified nonlinear finite element model. The second phase is a response spectrum analysis of a detail rack assembly finite element model. The damping values used in the seismic analysis are two percent damping for OBE and four percent damping for SSE as specified in NRC Regulatory Guide 1.61. 58
.. - . . - _ ~ .. . _ . - - .
The simplified nonlinear finite element model is used to determine the fuel rack response for the structural characteristics of a submerged rack assembly. The nonlinearities of the fuel rack assembly which are accounted for in the model are in the gap between the fuel cell and the fuel assembly, the boundary conditions of the fuel rack support locations, and energy losses at the support locations. The WECK computer program E23'143 is used to determine the nonlinear time history response of the fuel assembly / fuel rack system. The fuel assembly to cell impact loads, and overall rack response are obtained from the nonlinear time history results. The detail model is a three-dimensional finite element representation of a rack assembly consisting of discrete three-dimensional beams interconnected at a finite number of model points. The results of the nonlinear time history model are incorporated in the detail model. Since the detail model does not account for the nonlinear effect of a fuel rack assembly, the internal loads and stresses for the rack assembly obtained from this model are corrected by load correction factors. The load correci.o.7 l l factor is derived from the nonlinear model results and is applied to the components in the stress analysis. The responses of the model from accelerations in three directions are combined by the Square Root Sum of the Squares method in the stress analysis. The loads in ! two major components (support pad assembly and fuel cell) are examined, and the maximum loaded section of each of these components 59
is found. These maximum loads from the detail model are corrected by the nonlinear load correction factors and used in the stress analysis to obtain the stresses within the rack assembly. 8 6.3 FUEL RACK STRUCTURAL ANALYSIS J The stress analysis for the racks is performed using the load combinations specified in the "NRC Position For Review and Acceptance of Spent FuelAStorage and Handling Applications," as described in Section 3.2. J The thermal loads due to rack expansion relative to the pool floor are negligible since the support pads are not structurally - restrained in the lateral direction. The major seismic loads are produced by the operational basis earthquake (0BE) and safe shutdown earthquake (SSE) events. It is noted from the seismic analysis that the magnitude of stresses vary considerably from one geometrical location to the other in the model. Consequently, the maximum loaded cell assembly, grid assembly, and the leveling pad assembly are analyzed. Such an analysis envelopes the other areas of the rack assembly. Because of structural symmetry of the cell assembly about the x and y axes, the x and y direction horizontal seismic events produce i identical loads. Consequently, the margins of safety for the 60 9
multi-direction (x and y directions simultaneously) seismic event is computed by multiplying the undirectional loads by J2. The loads described in the seismic analysis section are adjusted by load modification factors obtained from the nonlinear analysis. The computed stresses are below the allowable stresses as required by the NRC Position Paper. 6.4 FUEL HANDLING CRANE UPLIFT ANALYSIS A fuel handling crane uplift analysis is performed to demonstrate that the rack can withstand the maximum 3000 pound uplift load of the fuel handling crane without violating the criticality acceptance criteria. In this analysis the uplift load is assumed to be applied to a fuel cell. Resulting stresses are within acceptable limits, and there is no change in rack geometry of a magnitude which causes f the criticality acceptance criteria to be violated. 6.5 FUEL BUNDLE / MODULE IMPACT EVALUATION i l An analysis is performed to eval'uate the effect of an isppact load due to fuel assembly and fuel storage cell interaction during a seismic event. The fuel rack system consists of an array of cells which form the fuel rack structure and fuel assemblies. The fuel rack system is located in the spent fuel pool and is submerged in water. 61
- _ . . . _ . __ __ ._. _ - ~ _ _ _ _ _ _ , _ - , _ . . _ _ _ _ . _ _ _ . . _ . _ _
Since the fuel assembly is stored within the cell, the gap between the fuel assembly grid and cell changes (i.e., opens und closes) during a seismic event. From the equation of motion for such a system it is evident that the fuel rack system is nonlinear. This condition necessitates the performance of c transient dynamic analysis. The mathematical features of the nonlinear fuel rack model facilitate the determination of the fuel assembly / cell interaction and hydrodynamic mass (fluid mass) effects on the fuel rack response during seismic excitation. The effect of fuel assembly and fuel storage cell impact force on the rigid body displacements is obtained from the nonlinear analysis. The analysis is conducted with a minimum coefficient of friction of 0.2, and it is shown that the rigid body displacement is minimal. Thus, impact between adjacent rack modulcs or between a rack module and the pool wall is precluded. The fuel assembly and fuel storage cell impact forces obtained from the nonlinear analysis are used'to evaluate the effects on the fuel rack structure and fuel assembly structure. These loads are within the allowable limits of the fuel rack module materials and fuel assembly materials. Therefore, there is no damage to the fuel assembly or fuel rack module due to impact loads. 62
6.6 ACCEPTANCE CRITERIA The fuel racks are analyzed for the normal and faulted load combinations of Section 3.2 in accordance with the "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications." The major normal and upset conditions loads are produced by the operational basis earthquakes (OBE). The thermal stresses due to rack expansion relative to the pool floor are negligible since the support pads are not structurally restrained in the lateral direction. The faulted condition loads are produced by _the safe shutdown earthquakes (SSE) and a postulated fuel assembly drop accident. The stresses are below the allowable stresses as required by the ASME B&PV Code, Section III, Subsection NF. In summary, the results of the seismic and structural analyses show that the ANO Units 1 and 2 spent fuel storage racks meet all the structural acceptance criteria adequately. f 63
6.7 POOL STRUCTURAL ANALYSIS The existing structures were analyzt:d for the modified fuel rack loads using the STARDYNE finite element computer program. The finite element models consisted of the pool walls and floor, foundation walls, cask laydown area, and fuel transfer canal area with the lowest elevation of 335'-0" assumed as a fixed boundary. Figures 6-1, 6-2 and 6-3 show details of the models prepared. The pool walls and floor slabs are modeled utilizing two layers of three dimensional solid " brick" elements with three degrees of freedom per node. In order to permit recovery of cube surface stresses, extremely thin quadrilateral membrane elements coincident with the nodes forming the surface of the solid elements were utilized. Recovery of these stresses along with cube centroidal stresses permitted the calculation of the required resultant forces and moments for an American Concrete Institute Code evaluation. The foundation walls were modeled using only one layer of brick elements through the thickness. The pool liner was not modeled since it is not acceptable to count I on its stiffness contribution td the overall pool capacity. l However, stress evaluation of this component was conducted using results obtained from the computer analysis. In order to correctly represent boundary conditions, modeling of the floor diaphrams and shear walls attached to the spent fuel pool was ( accomplished utilizing STARDYNE matrix elements. 64
6.7.1 Loads and Loading Combinations Table 6-1 lists the individual load components to which the spent fuel pool facilities were subjected. In all cases, except for fuel rack loads, the loads are consistent with the original loads used in the analyses performed and documented in the FSAR for each unit. Table 6-3 describes the final load combinations investigated for both normal and accident conditions. These combinations are based on the strength design method. 6.7.2 Seismic Loading Original plant response spectra and damping values were used to analyze the spent fuel facilities. Since only one horizontal and one vertical spectra were originally generated, the horizontal spectra was assumed to act in two horizontal directions. NUREG 0800 states that the three directions of the earthquake must be combined in an SRSS fashion; however, a strict SRSS procedure does'not maintain the signs of the force components which are of obvious importance for concrete structures. Therefore, the requirements of NUREG-0800 were met by formation of resultant earthquakes based on the three directions of earthquake applied simultaneously. 65 e - -
--.a
Components of seismic loads include vertical earthquake effects, which are factored dead weight, plus vertical rack forces and the horizontal effects, which are pool wall inertia forces, pool hydrodynamic forces, building seismic forces, and fuel rack horizontal forces. Table 6-2 shows the resulting factors used to produce a 0.lg effective OBE earthquake simulation. 6.7.3 Structural Acceptance Criteria The resulting forces and moments for the controlling load combinations were compared against the ultimate strength of the concrete sections. This comparison was initially carried out without thermal relaxation. All load combinations not involving thermal loads must pass this initial criteria check. Any load combination or element involving thermal loads, which meets the criteria without consideration of thermal relaxation, was eliminated from further processing. The load combinations failing to qualify on this first ' pass were selectively investigated by relieving the thermal moment as the section cracks. This procedure for post processing ensured that the contr671ing load combinations were addressed in the most efficient manner possible. i 66
J < 6.7.4 Results Using the above methods, models, loads, loading combination, r and acceptance criteria, the existing spent fuel facilities were determined to safely support the loads generated by the new fuel racks. i 1 1 s k 1
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TABLE 6-1 ARKASAS NUCLEAR ONE SPENT FUEL STORAGE FACILITY STRUCTURAL EVALUATION INDIVIDUAL LOAD CASE DESCRIPTION TABLE Load Case No. Notation Description i Dc Dead weight of the concrete. 2 H Hydrostatic pressure due to water in the pool. , 3 F Accident flood lood. 4 To(I) Normal operating thermal food. 5 To(I) Accident thermal load. 6 E,,(2) 1 ood generated by east-west 19 earthquake.
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7 E ns 8 D fr Fuel rock dead weight load. 9 FR y Reaction food of fuel rocks during Ig vertical earthquake. 10 FR ew Reaction lood of fuel rocks during Ig east-west earthquake. ll FR ns Reaction lood of fuel rocks during Ig north-south earthquake. 12 C Vertical crone reaction forces. NOTE: ,(1 ) includes effects of thermal momen't on the east foundation wall due to 280 thermal gradient.
. (2) Includes effect of pool hydrodynamic lood, pool wall, horizontal inertial ,
forces and building seismic response. g 7 ; seuense m-0 '" 71
APL-02-007 July 10 1982 s TABLE 6-2 ARi<ANSAS NUCLEAR ONE
. SPENT FUEL STORACE FACILIT/ STRUCTURAL EVALUATION
SUMMARY
OF COMPOSITE LOADS COMPOSITE - LOADS LOAD FACTORS DESCRIPTION De H F To To E,e E ns D fr FRy FR ew FR ns CIII D 1.0 1.0 - - - - - - - - - - Dead Load L - - - - - - - 1.0 - - - 1.0 Live Load _y ? To - - - 1.0 - - - - - - - Operating Thermal Load
" ~
To 1.0 Accident Thermal Load L - - - - - - - - - - - Ei 0.067 0.067 :- - - 0.1 0.1 - 0.067 0.1 0.1 - Loads Generated by' E2 0.067 0.067 - - - 0.1 -0.1 - 0.067 0.1 -0.1 - .19 Operating E3 -0.067 -0.067 - - - 0.1 0.1 - - 0.1 0.1 - Basis Earthquake E4 -0.067 -0.067 - - - 0.1 -0.1 - - 0.1 -0.1 - F. - - 1.0 -- - - - - - - - - Flood Load 1 . (1) Crone load will be considered only if it is determined to contribute to live food in a conservative manner. u x 2 m $ 1 {NR I,fk
lJ' - APL-02-007 July 1,1982 TABLE 6-3 ARKANSAS NUCLEAR ONE j SPENT FUEL STORACE FACILITY STRUCTURAL EVALUATION LOAD COMBli4ATION
SUMMARY
TABLE No Lood Combination Reference (I) I 1.4D + l .7L + 1.9E Load Case 2 2 .75 (1.4D + l .7L + 1.7T,) Load Case 4 3 .75 (1.4D + 1.7L + l.7To+ 1.9E) Load Case 5 Load Cose a 4 D+L+To + E' 5 D+L+T+F o Load Case b 6 D+L+TA Load Cose c 7 D + L + Tc+ 1.25E' Load C'ose d NOTES: (I) Reference (2) Section 3.8.4. / (2) E' = Represents a food, generated by .20g safe shutdown earthquake (SSE). E' = 2.0 E s e e A .1-- 4 [ 73
7.0 MATERIAL CONSIDERATIONS Construction materials conform to the requirements of ASME B and PV Code, Section III, Subsection NF. All the materials used in the construction are compatible with the storage pool environment and do not contaminate the fuel assemblies or the pool water. The racks are constructed from Type 304 stainless steel. The neutron absorbing material, Boraflex, used in the spent fuel rack construction is manufactured by Brand Industrial Services, Inc., and fabricated to safety related nuclear criteria of 10CFR50, Appendix B. Boraflex is a silicone based polymer containing fine particles of boron carbide in a homogenous, stable matrix. Boraflex has undergone extensive testing to study the effects of gamma irradiation in various environments and to verify its structural integrity and suitability as a neutron absorbing material.E153 Tests were performed at the University of Michigan exposing Boraflex to 1.03 x 1011 rads gamma radiation with a substantial concurrent neutron flux in borated water. These tests indicate that Boraflex maintains' its neutron attenuation capabilities before and after being subjected to an environment of borated water and 1.03 x 1011 rads gamma radiation.E183 Long term borated water soak tests at high temperatures were also conducted.E173 It was shown that Boraflex withstands a borated water immersion of 240 F for 260 days without visible distortion or 74
softening. Boraflex maintains its functional performance characteristics and shows no evidence of swelling or loss of ability to maintain a uniform distribution of boron carbide. During irradiation, a certain amount of gas may be generated. A conservative evaluation of the effect of gas generation on the spent fuel pool building atmosphere indicates that the maximum gas generation would be less than 0.01 percent of the total room volume. Additionally, the majority of gas generation is nitrogen, oxygen and CO2 - The actual tests verify that Boraflex maintains a long-term material stability and mechanical integrity, and can be safely utilized as a poison material for neutron absorption in spent fuel storage racks. t 75
8.0 INSTALLATION CONSIDERATIONS Installation of the proposed spent fuel storage racks is scheduled I .to begin March, 1983. All residing spent fuel assemblies will remain in the pools, while the spent fuel rack modules are replaced. This will be accomplished by shuffling the spent fuel a semblies to one area of the pool and replacing the empty modules with the new modules, then moving the spent fuel assemblies into the new rack modules and replacing the remaining empty rack modules. 8.1 RACK MODULE ASSEMBLY HANDLING CONSIDERATIONS i 2 The overhead cranes in the auxiliary building at ANO will be used 3 for removing the existing rack modules and lowering the new modules into the pool. No loads exceeding 2000 pounds will be allowed over the stored fuel assemblies at any time. All ANO fuel handling
- specifications and procedures will be observed at all times.
i 8.2 RADIATION PROTECTION CONSIDERATIONS
- A complete review of all removal and installation procedures will be performed by the ANO Health Physics and ALARA committee as outlined in Section 9.0. All radiation protection measures will be followed l
at all times to insure minimum occupational exposures. Based upon AN0's experience with similar projects, a collective dose of 16 person-rem is estimated in replacing the ANO racks in Unit-1 and ! Unit-2. i 76 l
-. -. - - - - . , . . , . - . - . - , . - , . - . . - . . . - - - . _ . _ - - . . _ , , , - ~ . . , - .
9.0 RADIOLOGICAL CONSIDERATIONS 9.1 GENERAL DESCRIPTION OF RADIOLOGICAL ASPECTS The radiological safety of the project is the responsibility of the ANO Health Physics Superintendent. He is assisted by his staff with support from the Corporate Health Physicist. The gamma radiation levels in the spent fuel pool area are monitored continuously by the area radiation monitoring systems. Additionally, radiation and contamination surveys are conducted in the spent fuel pool area on a weekly basis. In areas where a potential exists for significant airborne radioactivity, breathing zone air samples are taken and appropriate respiratory protective equipment is worn. Personnel working in radiologically controlled areas are required to wear protective clothing as specified by the applicable Radiaticn Work Permit (RWP). Contamination control measures are used to prevent the spread of contamination and to protect personnel from internal exposure from radioactive material. Risk level zones are established and posted. Movement of material, equipment, and personnel from a high risk zone to a lower risk zone requires specific contamination control measures to minimize the spread of I contamination across boundaries. Personnel monitoring devices are worn by all personnel working in the radiologically controlled area. The minimum requirement is a self-reading dosimeter and a thermoluminescent dosimeter (TLD). 77 l
4 Additional monitoring of the underwater divers will be by multiple whole body TLD's and extremity TLD's. 9.1.1 ALARA Considerations The work to be performed will be reviewed by the ALARA committee, which reviews all work in radiologically controlled areas when the estimated collective dose for the job exceeds 1 person-rem. The committee will review the radiological survey data, job scope, steps to be taken to avoid unnecessary exposure, temporary shielding, and changes which may occur during the work which may require additional radiological safety practices. 9.1.2 Underwater Radiation Surveys Underwater radiation surveys will be performed in all areas where divers must work or have the need for access to the work l area. An underwater radiation monitoring instrument and/or thermoluminescent dosimeters will be used, when applicable, to perform dose rate measurements in the pool. 9.1.3 Diving Operations Prior to diving operations, the spent fuel assemblies stored in the pool will be rearranged in such an array as to provide the lowest practicable dose rates to divers while minimizing the 78' l ~
. -- r.- . .-_ ^ - - _ _ , .
amount of rearrangement of the spent fuel. These two criteria should lessen the effects of radiation from both direct radiation from the fuel and from contaminated particles in the water that may be stirred up by fuel movement. Underwater work and access areas will be established for divers to ensure that exposures received are maintained ALARA. Health Physics Technicians will provide continuous coverage when divers are in the water. Their duties will be to provide health physics support, minimize personnel exposure, and enforce good radiation work practices and compliance with RWP requirements. The Health Physics Technicians and the diving supervisor will be in direct communication with the divers and will continually observe the divers while they are in the pool and warn them if they approach high radiation / exclusion zones. The divers will not be permitted to exceed dose rates of 1 rem / hour whole body. Divers will wear protective clothing items inside their rubber diving suits to protect them from contamination when they remove the diving suits and exit the controlled area. TLD's will be worn inside th' diving suits on the head, chest, back, legs, and extremities. Self reading dosimeters (SRD) will be sealed in plastic bags and worn inside the diving suits. The SRD's will be read and recorded after each dive. The TLD's will be read daily and a tabulation of each divers cumulative whole body and extremity doses will be prepared and reviewed by the diving supervisor and the cognizant Health Physics 79
~ . . - - _.- . . -
Supervisor. This information will be used to maintain exposures within administrative limits and to allocate the exposure evenly among the divers. 9.1.4 Spent Fuel Pool Decontamination and Clean-up The Spent Fuel Pool Purification Systems provide purification, clarification, and decontamination of pool water by recirculation through filters and a demineralizer. During the re-rack project the pool water will be sampled weekly to determine the concentrations of radionuclides in the pool. A portable filtered water vacuum system will be used as necessary to remove loosely deposited contamination from the fuel rack surfaces, pool floor, and walls near diver working areas to reduce the radiation exposures to the divers. 9.1. 4.1 Solid Waste Disposal The solid waste generated from the project will be in the form of demineralizer resins, filters, compactable trash, and non-compactable waste. These wastes will be disposed through , normal plant methods for shipment to a licensed waste burial facility. ! c 80
9.1.4.2 Decontamination of Removed Racks The rack sections will be rinsed with a low pressure spray of demineralized water or spent fuel pool water as they are removed from the pool. If further decontamination is necessary the racks will be removed to an area with adequate containment to allow the use of hydrolasing equipment. Personnel performing decontamination on the racks will wear appropriate protective clothing and respiratory protection equipment. After decontamination, the racks will be moved to another area to be packed for disposal. The decontamination operations are expected to remove significant quantities of loose contamination while causing a relatively low exposure to the decontamination workers, but will reduce subsequent personnel exposures to workers packaging, handling, and shipping the rack sections. J 9.1.5 Disposition of Old Racks Four options are available for the old racks, as follows:
- 1. Burial without volume reduction.
1
- 2. Burial with volume reduction.
- 3. Decontaminated to releasable criteria of Regulatory Guide 1.86 and disposal.
81
- 4. Transfer to a licensed facility for decontamination and/or disposal.
i An evaluation will be made for disposal of the racks based upon the contamination levels when removed, person rem exposures estimated for decontamination, person-rem exposures estimated for volume reduction, and cost / benefit factors for each of the options. Ultimate disposal will be one of the four options listed after the evaluations have been performed. 9.1. 6 Current Radiological Data From Spent Fuel Pools f fhe most recent radiological survey data from the spent fuel pool area indicates the following parameters: DOSE RATES: Unit I Fuel Transfer Bridge 4 mR/hr Unit I Side of Pool 1 mR/hr Unit II Fuel iransfer Bridge 4 mR/hr Unit II Side of Pool 1 mR/hr AIRBORNE RADI0 ACTIVITY: Gross Particulate < 1.0 E-9 uCi/mL Gross Iodine < 1.0 E-9 uCi/mL Isotopic analysis not performed on air samples less than 1.0 E-9 uCi/mL. 82
Spent Fuel Pool Water. (July 1, 1982) Isotope uCi/mL Cs-137 2.37 E-2 Cs-134 9.91 E-3 Co-60 8.73 E-4 Co-58 2.09 E-4 Zr-95 6.71 E-6 Sb-125 6.12 E-5 l TOTAL 3.42 E-2 I
9.2 DESCRIPTION
OF RADIOLOGICAL ASPECTS OF INCREASED SPENT FUEL STORAGE As of July 1, 1982, 244 assemblies are stored in AN0-1 and 60 assemblies are stored in ANO-2 spent fuel pools, respectively. The I current radiological data relating to the spent fuel pools are shown in Section 9.1.6.
- Currently solid wastes are collected from the.'spent fuel pools and concentrated in two areas: (1) The spent fuel pool ion exchangers which remove ionic material from the pool water and (2) The l particulate filters which remove particulates larger than 5 microns.
The ANO-1 ion exchanger resin volume istappr,oximately 20 ft3 and the AN0-2 resin volume is approximately 32 ft3 (useful). Both are currently changed about once per year. Both AN0-1 and ANO-2 particulate filter systems contain two filters which are changed l l approximately once per year. 83 l 3
9.2.1 Solid Waste Generation
\,
The ANO-1 spent fuel pool demineralizer (ion exchanger) resin volume of approximately 20 ft3 represents <10% of the total ANO-1 annual solid waste in the form of spent resins. ANO-2 spent fuel pool ion exchanger volume is approximately 32 ft: (useful) and represents about 10% of the total AN0-2 annual solid waste in the form of spent resin. These resins are designed to be changed based on an increase in differential pressure rather than on the lack of ability to remove radioactive ionic material. Neither the frequency of fuel addition nor the annual amount of fuel to be added to the spent fuel pool will be changed due to the rack modifications. Therefore, the annual amounts of contaminants added to the pools are_not expected to increase significantly. Since the
', s resins are designed to be changed annually, and the annual amount of contaminants is not expected to increase, no appreciable increase in solid waste in the form of spent fuel
. pool ion exchanger resins is expected.
Similarly, the particulate filters are designed to be changed annually, based on differential pressure. As in the case of the' ion exchanger resins, no significant change in the i frequency of replacement due to the modified racks is expected. s l (
'i 84
The number of ANO-1 storage locations was increased in 1976, and no significant increase in radioactive solid waste generation in the form of resins or filters resulted. 9.2.2 Airborne Activity ' The design of ANO-1 does not permit the measurement of radioactive gases released from individual ventilation systems, but data are available for overall plant releases. As shown in the AN0-1 FSAR Section 9.6,2.4.2, the data related to the 233 percent increase in fuel storage capacity in 1976 do not indicate an appreciable ircrease in the Kr-85 release rate. However, as a method of estimation, the FSAR analysis assumed a 233 percent increase in the annual Kr-85 release to predict a maximum release rate. The results indicate that even with this conservative treatment, the increase of total plant Kr-85 annual release would be less than 1.13 percent. i Since spent fuel storage activities are to remain basically unchanged, there are no data to predict nor reason to expect results different from thos'e realized from the earlier modifications, i.e., no appreciable increase in release rate in , either unit. l l l 85
9.2.3 Personnel Exposure The radionuclide concentrations in the spent fuel pools as presented in the applicable FSARs are based on one percent failed fuel and a crud burst model. Exposures based on this are presented in ANO-1 FSAR, Section 9.6.2.3. No significant increase in personnel doses was made due to the increase in
~
storage capacity. As stated in Section 9.6.2.4.2 of the ANO-1 FSAR, " Fuel being transferred is the controlling contributor to the basic dose rates, not the stored fuel." l In regard to personnel exposure received during filter and resin changes, based on experience it is estimated that the annual exposure is <0.2 man-rem / unit. This annual exposure is not expected to increase, since the frequency of change is not expected to be altered. t 5 t 86
10.0 ACCIDENT EVALUATIONS The following analyses are related to postulated accidents identified in the "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications". 10.1 CASK DROP ACCIDENT ANALYSIS Arkansas Nuclear One Unit 1 and Unit 2 administrative procedures prevent the spent fuel cask from being moved over the spent fuel pools; hence, the cask drop accident is not credible. 10.2 FUEL ASSEMBLY DROP ACCIDENT ANALYSIS A fuel assembly drop accident analysis is also performed to ensure that, in the unlikely event of dropping a fuel assembly, accidental deformation to the rack will not cause the criticality acceptance criteria to be violated, and that the spent fuel pool liner will not i be perforated. 10.3 LOSS OF SPENT FUEL P0OL COOLING Under postulated accident conditions where all non-Category 1 spent fuel pool makeup systems become incperative, alternate Seismic Class 1 methods for makeup to the spent fuel pool water are 87 1 -
'i i
available in each unit. Section 9.6.1.3 of the ANO-1 FSAR describes
- .these methods for Unit 1, and Section 9.1.3.3.1 of the ANO-2 FSAR details the methods for Unit 2.
l I ] Although it is highly unlikely that a complete loss of cooling capabilitycouldoccur,-tNemodifiedstorageconditionsare' analyzed
- to this condition.
i Basis: A
- a. No pool cooling implies that temperature of water at inlet to i 1 j spent fuel racks is 212 F, which corresponds to the saturation temperature at the pool surface.
- b. The nominal water level above the top of the racks is l
maintained.- . c. Maximum. fuel loading cases of Table 5-1 for Unit 1 and Table 5-2 for Unit 2 are assumed.
- d. The assemblies that are evaluated are initially placed into the pool at seven days after shutdown.
i l e. The peak rods are assumed to have 60% greater heat output than average rods. i i 5 88 y
,,, , , , , , . .- -- . ~ . , . - , r- -- -
t
- f. All storage cells are filled and all downflow occurs in the peripheral gaps.
The criterion for this. condition is that adequate cooling must exist to maintain the temperature of the f'el cladding sufficiently low to prevent structural failures and safety concerns. A J Pe i 89 '
11.0 COST AND BENEFIT ASSESSMENT Arkansas Nuclear One's present spent fuel rack capacities are 589 spaces for Unit-1 and 485 spaces for Unit-2. As Figure 11-1 and 11-2 indicate, ANO-1 and AN0-2 will lose full core discharge capacities in 1986 with these conditions. Currently, ANO does not have contracts or options for the use of any "away from reactor storage" or reprocessing facilities, nor do we expect to have these options available within the next 10 years. 11.1 ALTERNATIVES Without the increased storage capacity provided by reracking the spent fuel pools, Arkansas Power and Light would be forced to shutdown Unit-1 and Unit-2 in 1989 due to the inability to refuel the units. This would cause a significant increase in electrical power cost to the citizens served by Arkansas Power & Light due to the cost of replacement power. l 11.2 COMMITMENT OF MATERIAL RESOURCES t The proposed replacement rack modules will be fabricated from type 304 stainless steel material with these modules in Region 1 i utilizing the neutron absorbing material, Boraflex. A total amount i i 90
. , . .- - -. .. . _ _ _ -. . - - . . .. .. _ . . = .
5 4-of approximately 347,000 pounds of stainless steel and approximately 9,600 pounds of Boraflex is expected to be used. The use of these i materials will have no significant affect on the availability of l these material resources. i s i l l i l } } 4 i i 91
. , , _ _ , , _ . _ _ , , _ _ _ , _ . , - . ~ . . _ , _ . . _ _ . . . _ , . . _ . _ . . . _ _ , . - , _ . . . _ . . . . . _ . . _ _ _ - . . -
Figure 11-1 SPENT FUEL STORAGE AT ANO - UNIT 1- , REFUELING REFUELING DISCHARGED TOTAL SPACES TIME NUMBER ASSEMBLIES IN POOL REMAINING Sprfr- 1977 1 56 56 533 Spring 1978 2 56 112 477 Spring 1979 3 64 176 413 l Spring 1981 4 68 244 345 Fall 1982 5 72 316 273 Fall 1984 6 68 384 205 *584 l Spring 1986 (1) 7 64 448 141 *520 ! Fall 1987 8 64 512 77 *456 Spring 1989 9 64 576 13 *392 Fall 1990 10 64 640 *328 Spring 1992 11 64 704 *264 Fall-1993 .12 64 768 *200 Spring 1995 (2) 13 64 832 *136 Fall 1996 14 64 896
- 72
! Spring 1998 15 64 960
- 8 l
NOTE: * (Reracked) l (1) Full core discharge capability is lost with the present capacity of l 589 spaces. (Core = 177 Assemblies) (2) Full core discharge capability is lost assuming a reracked capacity of 968 spaces. 92
Figure 11-2 SPENT FUEL STORAGE AT ANO - UNIT 2 REFUELING REFUELING DISCHARGED TOTAL SPACES TIME NUMBER ASSEMBLIES IN P0OL REMAINING i Spring 1981 1 60 60 425 Fall 1982 2 52 112 373 Fall 1983 3 68 180 305 *808 Spring 1985 4 68 248 237 *740 Fall 1986 (1) 5 68 316 169 *672 Spring 1988 6 68 384 101 *604 Fall 1989 7 68 452 33 *536 Spring 1991 8 68 520 *468 Fall 1992 9 68 588 *400 Spring 1994 10 68 656 *332 Fall 1995 11 68 724 *264 Spring 1997 12 68 792 *196 Fall 1998 (2) 13 68 860 *128 Spring 2000 14 68 928
- 60 i
NOTE: * (Reracked) r (1) Full core discharge capability is lost with the present capacity of 485 spaces. (Core = 177 Assemblies) (2) Full core discharge capability is lost assuming a reracked capacity of 988 spaces. l 93
12.0 REFERENCES
- 1. Nuclear Regulatory Commission, Letter to All Power Reactor Licensees, from B.K. Grimes, April 14, 1978, "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications."
- 2. W. E. Ford III, et al, "A 218-Group Neutron Cross-Section Library in the AMPX Master Interface Format for Criticality Safety Studies," ORNL/CSD/TM-4 (July 1976).
- 3. N. M. Greene, et al, "AMPX: A Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B," 0RNL/TM-3706 (March 1976).
- 4. L. M. Petrie and N. F. Cross, " KEN 0 IV--An Improved Monte Carlo Criticality Program," ORNL-4938 (November 1975).
- 5. S. R. Bierman, et al, " Critical Separation Between Subcritical Clusters of 2.35 wt % U 235 Enriched U0 Rods in Water with 2
Fixed Neutron Poisons," Battelle Pacific Northwest Laboraties PNL-2438 (October 1977).
- 6. S. R. Bierman, et al, " Critical Separation Between Subcritical Clusters of 4.29 wt % U 235 Enriched UO 2 Rods in Water with Fixed Neutron Poisons," Battelle Pacific Northwest Laboratories PNL-2615 (March 1978).
94
- 7. J. T. Thomas, " Critical Three-Dimensional Arrays of U (93.2)--Metal Cylinders," Nuclear Science and Engieering, Volume 52, pages 350-359 (1973).
- 8. Barry, R. P. , " LEOPARD - A Spectrum Dependent Non-Spatial Depletion Code for the IBM-7094," WCAP-3269-26, September 1963.
- 9. England, T. R., " CINDER - A One-Point Depletion and Fission Product Program," WAPD-TM-334, August 1962.
- 10. Altomare, S. and Barry, R. F., "The TURTLE 24.0 Diffusion Depletion Code," WCAP-7758-A, January 1975.
- 11. Nodvik, R. J., "Saxton Core II Fuel Performance Evaluation,"
WCAP-3385-56, Part II, " Evaluation of Mass Spectrometric and Radio-hemical Analysis of Irradiated Saxton Plutonium Fuel," j July 1970.
- 12. Strawbridge, L. E. and Barry, R. F. , " Criticality Calculation 1
for Uniform Water-Moderated Lattices," Nuclear Science and Engineering 23. 58 (1965). '
- 13. WECAN " Documentation of Selected Westinghouse Structural Analyses Computer Codes", WCAP-8252.
- 14. WECAN " Benchmark Problem Solution Employed for Verification of the WECAN Computer Program", WCAP-8929.
95
- 15. J.S. Anderson, "Boraflex Neutron Shielding Material -- Product Performance Data," Brand Industries, Inc., Report 748-30-1, (August 1979).
- 16. J.S. Anderson, " Irradiation Study of Boraflex Neutron Shielding Materials," Brand Industries, Inc., Report 748-10-1,(July 1979).
- 17. J.S. Anderson, "A Final Report on the Effects of High Temperature Borated Water Exposure on BISCO Boraflex Neutron Absorbing Materials," Brand Industries, Inc., Report 748-2-1, (August 1978).
p l l f i ll 96 l
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