ML17215A306

Proposed Tech Spec Changes Re Reracking of Spent Fuel Pool. NSHC & SAR Encl

ML17215A306
Person / Time
Site:	Saint Lucie
Issue date:	03/13/1984
From:	FLORIDA POWER & LIGHT CO.
To:
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ML17215A305	List: ML17215A304 ML17215A306
References
NUDOCS 8403190228
Download: ML17215A306 (112)

Text

' DESIGN FE ATURES VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is l0,93 I + 275 cubic feet at a nominal Tavg of 572oF.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.I The meteorological tower shall be located as shown on Figure S.l -I.

5.6 FUEL STORAGE CRITICALITY 5.6. I The spent fuel storage racks are designed and shall be maintained with:

a. A keff equivalent to less than or equal to 0.95 when flooded with unborated water, which includes a conservative allowance of 0.024 hkeff for Total Uncertainty.

b. A nominal 8.96 inch center-to-center distance between fuel assemblies placed in the storage racks.

c. A boron concentration greater than or equal to I 720 ppm.

Region I can be used to store fuel which has a U-235 enrichment less than or equal to 4.5%. Region II can be used to store fuel which has achieved sufficient burnup such that storage in Region I is not required.

DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 56 feet.

CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than I I 88 fuel assemblies.  !

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5.7 COMPONENT CYCLIC OR TRANSIENT LIMITS 5.7. I The components identi fied in Table 5.7-I are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.

ST. LUCIE - UNIT 2 Amendment No.

8403190228 8403i3 PDR ADOCK 05000389 >

p PDR

No Si nificant Hazards Consideration Florida Power and Light Company (FPL) has determined that the proposed amendment involves no significant hazards considerations, focusing on the three standards set forth in IO CFR 50.92(c) as quoted below:

The Commission may make a final determination, pursuant to the procedures in 50.9I, that a proposed amendment to on operating license for a facility licensed under 50.2l(b) or 50.22 or for a testing facility involves no significant hazards considerations, if operation of the facility in accordance with the proposed amendment would not:

I. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3. Involve a significant reduction in a margin of safety.

FPL has determined that the activities associated with this amendment request do not meet any of the significant hazards consideration standards of IO CFR 50.92(c) and, accordingly, a no significant hazards consideration finding is justified. In support of this determination, the following background information is provided, followed by a discussion of each of the above three significant hazards consideration standards.

~Back round There is one spent fuel pool at St. Lucie Unit 2. The existing racks have 300 total storage cells, with a center-to-center spacing of l4 inches, which allows for the removal of one full core during that period of time when one-third of a core is stored in the fuel pool. With the 300 presently available storage cells, St. Lucie Unit 2

I would lose the full-core reserve storage capability after the second refueling, expected to be in l986. Therefore, to ensure that sufficient capacity continues to exist at St. Lucie Unit 2, FPL has contracted with Combustion Engineering (C-E) for new spent fuel storage racks whose design allows for more dense storage of spent fuel.

The new racks have an ultimate storage capacity of I I88 fuel assemblies, which will extend the full-core reserve storage capability until l998.

The new fuel storage racks will store fuel in two discrete regions of the spent fuel pool. Region I includes six modules having a total of 448 storage cells. Only one-half of these cells will be available for storage of fuel assemblies. The unused cells will be provided with cell blocking devices. The 224 available cells enable storage of fuel assemblies with Uranium-235 enrichments up to 4.5%, while maintaining the required subcriticality (Keff< 0.95).

Region II includes thirteen modules having a total of I I36 storage cells, of which 852 (75%) will be available for storage of fuel assemblies. The unused cells will act as neutron flux traps (to maintain the required subcriticality) and will be provided with cell blocking devices. Region II will be used to store fuel which has experienced sufficient burnup such that storage in Region I is not required.

The new fuel racks are fabricated from 304 stainless steel which is O.I35 inches thick.

Each cell is formed by welding along the intersecting seams which enables the assembled cells (module) to become a free-standing structure which is seismically qualified without depending on neighboring modules or fuel pool walls for support. The nominal center-to-center spacing of the cells within both Region I and II is 8.96 inches.

There is no spent fuel in the St. Lucie Unit 2 spent fuel pool at this time, nor is there expected to be any when the new spent fuel racks are installed. Therefore, no special administrative controls or procedures will be necessary to provide radiation

'protection, and the evaluation of a construction accident with respect to nuclear criticality or radioactivity'release is not necessary.

Evaluation The following evaluation demonstrates (with reference to the analysis contained in the attached Safety Analysis Report) that the proposed amendment involves no significant hazards considerations.

I. Involve a si nificant increase in the robabilit or conse uences of an accident revious I eva Iua ted.

The analysis of this proposed reracking has been accomplished using current accepted codes, standards, and NRC guidance as specified in Section 4.2 of the attached Safety Analysis Report (SAR).

In the course of the analysis, FPL identified the following potential accident scenarios:

(I) a spent fuel assembly drop in the spent fuel pool; (2) loss of spent fuel pool cooling system flow; (3) an extreme wind or seismic event; and (4) a spent fuel cask drop. The occurrence of these accidents is not affected by the racks themselves. Thus, the proposed reracking cannot increase the probability of these accidents. Furthermore, the spent fuel racks will be installed prior to storage of any spent fuel in the spent fuel pool. Therefore, there is no potential for an accident involving spent fuel during fuel rock installation.

Similarly, the analysis of the potential accidents, summarized below, has shown that there is no significant increase in the consequences of an accident previously analyzed.

The consequences of a spent fuel assembly drop have been evaluated with respect to nuclear criticality (Section 3.I of the SAR) and with respect to radioactivity release (Section 5.3 of the SAR). The presence of boron in the spent fuel. pool water ensures that the neutron multiplication factor (keff) remains less than the NRC acceptance criterion of 0.95 for all accident conditions. The consequences of a dropped fuel

'ssembly with respect to radioactivity'release are not affected'by. the new fuel rack.'esign itself. However, the analysis in 'Section 5.3 of this SAR included more conservative assumptions, relative to those in the previous FSAR analysis, to bracket future changes to fuel management. As a result, the predicted radioactivity releases are about 25% larger than those reported previously, but the increases are not "significant" 'ecause the results are only l% of NRC guidelines.

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Thus, the consequences of the spent fuel assembly drop accident would not be significantly increased from those previously evaluated.

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Total loss of spent fuel pool cooling flow has been evaluated and is reported in Section 3.2 of the attached SAR. As indicated in Section 3.2, more than 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> are available to restore cooling flow or to provide an alternate means for cooling before pool boiling results in a water level less than that which is needed to maintain acceptable radiation dose levels. Also, the analysis has shown that fuel cladding integrity is maintained.

Thus, the consequences of this type accident 'would not be significantly increased from the previously evaluated loss of cooling system flow accident.

The consequences of a seismic event have been evaluated and are summarized in Section 4;3 of the attached SAR. The new racks are to be designed and fabricated to meet the guidance of applicable portions of the NRC Regulatory Guides and codes and standards listed in Section 4.2 of the SAR. The maximum stresses within the fuel racks will be within the criteria specified in Section 4.4 of this SAR. Also, movement and deflection of the fuel rack modules does not result in contact with neighboring rack 'modules or the fuel pool-walls. The floor loading from the new racks filled with spent fuel assemblies does not exceed the structural capacity of the fuel handling building. As indicated in Section 4.I of the SAR, the fuel handling building walls, floors, and partitions are designed to withstand hurricane and tornado winds; the protection from these extreme winds is not affected by the new fuel rack design.

Thus, the consequences of an extreme wind or seismic event would not be significantly increased from previously evaluated events.

The consequences of a spent fuel cask drop outside the Fuel Handling Building has been evaluated and are reported in Section 5.3 of this SAR. As stated in Section 5.3, the spent fuel cask is prevented from dropping onto the spent fuel pool racks by design of the fuel handling building and overhead crane. Therefore, the consequences of the spent fuel cask drop are not affected by the new fuel rack design. The analysis in Section 5.3. includes more conservative assumptions, relative to those in the previous FSAR analysis, to bracket future changes to fuel management. As a result, the pre'dieted radioactivity release's'are about 25% la'rger than those reported pieviously, but the increases are not "significant" because the results are only 8% of NRC guidelines. Thus, the consequences of a cask drop accident would not be significantly increased from previously evaluated accident analysis.

It is concluded that the proposed amendment to rerack the spent fuel pools will not involve a significant increase in the probability or consequences of an 'accident previously evaluated.

2- Create the ossibilit of a new or different kind of accident from an accident reviousl evaluated.

FPL has evaluated the proposed rerackirig in accordance with the guidance of the NRC position paper entitled "Review and Acceptance of Spent Fuel Storage and Handling Applications", appropriate NRC Regulatory Guides, appropriate NRC Standard Review Plans, and appropriate Industry Codes and Standards as listed in Section 4.2 of the attached SAR. In addition, FPL has reviewed several previous NRC Safety Evaluation Reports for rerack applications similar to this proposal. As a result of this evaluation and these reviews, FPL finds that the proposed reracking does not, in any way, create the possibility of a new or different kind of accident from any accident previously evaluated for the. Spent Fuel Pool or Fuel Handling Building.

3., Involve a si nificant reduction in a mar in of safet .

The NRC Staff safety evaluation review process has established that the issue of margin of safety, when applied to a reracking modification, will need to address the fol towing areas:

Nuclear criticality considerations

2. Thermal-Hydraulic considerations

3. Mechanical, material and structura I considerations The established acceptance criterion for criticality is that the neutron multiplication factor (keff) in spent fuel pools shall be less than or equal to 0.95, including all

'uncertainties, under all conditions. 'This'margin of safety'has been adhered to in the criticality analysis methods for the new rack de'sign as discussed in Section 3.I of the attached SAR. That is, keff is always less than 0.95, including uncertainties at the 95/95 probability and confidence level.

In meeting the acceptance criteria for criticality in the spent fuel pool, the proposed amendment to rerack the spent fuel pools does not involve a significant reduction in the margin of safety for nuclear criticality.

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Conservative methods were used to calculate the increase in temperature of the water in the spent fuel pool and demonstrate maintenance of fuel cladding integrity. This evaluation uses the methods described in Section 3.2 of the SAR in demonstrating that the margins of safety are maintained. The proposed reracking allows an increase in the heat load in the spent fuel pool; the evaluation shows that the existing spent fuel cooling system, under normal conditions, will maintain the pool temperature below the design basis limit, assuming the maximum heat load in the pool. Since the design basis limit is met, there is not a significant reduction in the margin of safety. Also, the maximum fuel cladding temperature, assuming total loss of fuel pool cooling, would remain below 275 F, ensuring maintenance of fuel cladding integrity. Thus, there is no significant reduction in the margin of safety from a thermal-hydraulic or spent fuel cooling concern.

The main safety function of the spent fuel pool and the racks is to maintain the spent fuel assemblies in a safe configuration through'll environments and abnormal loadings, such as an earthquake, drop of a spent fuel assembly, or drop of any other object during routine spent fuel handling. The mechanical, material, and structural considerations of the proposed rerack are described in Section 4 of the attached SAR.

The analysis of Section 4 has shown that all criteria for fuel rack movement, stresses, floor loadings, etc., are met and that margins of safety are not significantly reduced.

ln summation, it has been shown that the proposed spent fuel storage facility modifications and proposed Technical Specifications do not:

Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3. Involve a significant reduction in a margin of safety.

Therefore, FPL has determined that the proposed amendment involves no significant hazards considerations.

SPENT FUEL POOL RERACK SAFETY ANALYSIS REPORT

TAB"E OF CONTENTS

~Pa e

1. 0 INTRODUCTION

1. 1 License Amendment Requested 1.2 Current Status 1-2 1.3 Interfaces with Other Organizations 1-2 1.4 Summary of Report 1-2 1.5 Conclusions 1-3 2.0

SUMMARY

OF SPENT FUEL RACK DESIGN 2-1 3.0 NUCLEAR AND THERMAL-HYDRAULIC CONSIDERATIONS 3-1

3. 1 Neutron Multiplication Factor 3-1 3.1.1 Normal Storage 3-1

3. 1.2 Postulated Accidents 3-2 3.1. 3 Calculation Methods 3-2 3.1.4 Fuel Rack Modification 3-4

3. 1.5 Acceptance Criterion for Criticality 3-4 Calculations for the Spent Fuel Pool (Bulk)

'-4 3.2 Decay Heat 3.2. 1 Design Bases 3-4 3.2.2 System Description 3-6 3.2.3 Safety Evaluation 3-6 3.3 Potential Fuel and Rack Handling Accidents 3-10 3.4 Technical Specifications 3-10 3.5 References 3-11 4.0 MECHANICAL, MATERIAL, AND STRUCTURAL CONSIDERATIONS 4-1 4.1 Description of Structure 4-1

4. 1. 1 Description of Fuel Handling Building 4-1

4. 1.2 Description of Spent Fuel Racks 4-1 4.2 Applicable Codes, Standards, and Specifications 4'-4 4.3 Seismic and Impact Loads 4-6 4.4 Loads and Load Combinations 4-7 4.5 Design and Analysis Procedures 4-7

4. 5; 1 Methodology Summa'ry 4-7 .

4.5. 2 Computer Code Descri pti ons 4-9 4.6 Structural Acceptance Criteria 4-11 4.7 Materials, equality Control, and Special Construction Techniques 4-12 4.7. 1 Materials 4-12 4.7.2 equality Control 4-12 4.7.3 Construction Techniques 4-13

TABLE OF CONTENTS

~Pa e 4.8 Testing and In-Service Surveillance 4-13 4.9, References 4-13 5.0 COST/BENEFIT ASSESSMENT AND ENVIRONMENTAL IMPACT 5-1

5. 1 Cost/Benefit Assessment 5-1 5.1. 1 Need for Increased Storage Capacity 5-1

5. 1.2 Construction Costs 5-2

5. 1.3 Consideration of Alternatives 5-2 5.1.4 Resources Committed 5-3

5. 1.5 Thermal Impact on the Environment 5-3 5.2 Radiological Evaluation 5-4 5.2. 1 Solid Radwaste 5-4 5.2.2 Gaseous Radwaste 5-4 5.2.3 Personnel Exposure 5-5 5.2.4 Rack Disposal 5-7 5.3 Accident Evaluation 5-7 5.3. 1 Spent Fuel Handling Accidents 5-7 5.3.2 Acceptability 5-13 5.3.3 Fuel Decay 5-13 5.3.4 Loads Over Spent Fuel 5-14 5.3.5 Conclusions 5-14

LIST OF TABLES Table ~Pa e 5-1 Estimated Spent Fuel Pool Capacity Requirements 5-16 5-2 Replacement Cost for St. Lucie Unit 2 5-17 5-3 Gamma Isotopic Analysis for St. Lucie Unit 1 5-4 Spent Fuel Cask Drop Accident Assumptions 5-5 Spent Fuel Cask Drop Accident Releases 5-22 5-6 Spent Fuel Cask Drop Offsite Doses 5-23 5-7 Fuel Handling Accident Assumptions 5-24 5-8 Fuel Handling Accident Releases 5-26 5-9 f Fuel Handl i ng Acci dent Of si te Doses 5-27

LIST OF FIGURES

~Fi ere Title ~Pa e 3-1 Spent Fuel Rack Module for Region I 3-12 3-2 Spent Fuel Rack Module for Region II 3-13 3-3 St. Lucie Unit 2 Spent Fuel Storage Rack 3-14 Criticality vs. Burnup for Fuel Assemblies in Region II 3-4 Spent Fuel Storage Rack Initial Enrichment 3-15 vs. Burnup for Fuel Assemblies in Region II 4-1 Typical Spent Fuel Storage Rack Module 4-15.

4-2 Typical Spent Fuel Rack Module L-Insert 4-16 4-3 Spent Fuel Storage Module Installation 4-17 4 4 Spent fuel Storage Module 4-18 4-5 L- In se rts '-19 4-6 SAP IV Computer Code Membrane Stresses 4-20 and Bending Moments

INTRODUCTION LICENSE AMENDMENT REQUESTED Florida Power 5 Light (FPL) has contracted for the purchase of new spent fuel storage racks to be placed into the spent fuel pool of St. Lucie Unit 2. These new racks increase the amount of spent fuel that can be stored in the existing spent fuel pool. This Safety Analysis Report supports this request for a license amendment to the St. Lucie Unit 2 Facility Operating License NPF-16 for the technical specification revisions required as a result of installation and use of the new spent fuel storage racks.

CURRENT STATUS There is one spent fuel pool at St. Lucie Unit 2. The existing racks in the pool have 300 total storage cells, with a center-to-center spacing of 14 inches, which allows for the removal of one full core during that period of time when one-third of a core is stored in the fuel pool. Per the FSAR, Section 9. 1, additional storage racks of the same design may be installed to bring total capacity, to 675 spent fuel assemblies (approximately three full cores). With the 300 presently available storage cells, St. Lucie Unit 2 would lose the full-core reserve storage capability after the second refueling, expected to be in 1986; with additional racks, up to the maximum of 675, the capability would be lost in 1992.

Therefore, to ensure that sufficient capacity continues to exist at St. Lucie Unit 2 to store discharged 'fuel assemblies, FPL has decided to remove the present racks and has contracted with Combus-tion Engineering (C-E) for new spent fuel storage racks whose design (MAX CAP') allows for more dense storage of spent fuel. The new racks have a usable storage capacity of 1076 cells, extending the full-core-reserve storage capability until 1998.

Further references within this report to the "spent fuel racks" refer to the new design. The details of how the spent fuel rack

'esign meets the design requirements is provided in this report.

INTERFACES WITH OTHER ORGANIZATIONS FP8L has overall responsibility for this modification. C-E is responsible for the design and fabrication of the new spent fuel storage racks and engineering assistance in reviewing the spent fuel pool cooling system. Ebasco Services, Inc. is responsible for reviewing building structural analysis and accident evaluation.

SUMMARY

OF REPORT This report follows the guidance of the NRC position paper entitled "Review and Acceptance of Spent Fuel Storage and Handling Applica-tions", April 1978, as amended by NRC letter dated January 18, 1979.

Section 2.0 presents a summary of the spent fuel rack design.

Sections 3.0 through 5.0 of this report are consistent with the section/subsection content'f the above NRC position paper, Sections III through V.

This report contains the nuclear, thermal-hydraulic, mechanical, material, structural, and radiological design criteria for the fuel racks. The nuclear and thermal-hydraulic aspects of this report (Section 3.0) address the neutron multiplication factor, considering normal. storage and handling of spent fuel as well as postulated accidents, with respect to criticality and the abi 1'ity of the spent fuel pool cooling system to maintain sufficient cooling.

Mechanical, material, and structural aspects (Section 4.0) involve the capability of the fuel assemblies, storage racks, and spent fuel pool system to withstand effects of natural phenomena and other service loading conditions.

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The environmental aspects of the report (Section 5.0) concern the thermal and radiological release from the facility under normal and accident conditions. This section also addresses the occupational radiation exposures, generation of radioactive waste, need for expansion, commitment of material and non-material resources, and a cost-benefit assessment.

CONCLUSIONS On the basis of the evaluations and information presented in this report, plus operating experience with high density fuel storage at St. Lucie Unit 1, FP&L concludes that the proposed modification of St. Lucie Unit 2 spent fuel storage facilities provides safe spent fuel storage, and that the modification is consistent with the facility design and operating criteria as provided in the FSAR and operating license.

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SUMMARY

OF SPENT FUEL RACK DESIGN The spent fuel racks are designed to store the 14 x 14 design fuel from St. Lucie Unit 1 and the 16 x 16 fuel from St. Lucie Unit 2.

These racks have an initial capacity to store a total of 1076 fuel assemblies in two regions of the spent fuel pool. By expansion of Region II into Region I, the spent fuel racks have an ultimate storage capacity of 1188 fuel assemblies.

Region I initially contains four 7xl1 modules and two 7x10 modules, i.e., six modules with 448 storage cells. Only one-half of these cells will be available for storage of fuel assemblies. The unused cells will be provided with blocking devices. All cells contain "L" inserts which are stainless steel neutron absorbers (see Figure 4-2). The 224 available cells enable storage of fuel assemblies

.with Uranium-235 enrichments up to 4.5l while maintaining the required subcriticality (k ff < 0.95).

Region II initially contains one 8x10 module and twelve 8x11 mod-ules, i.e., thirteen modules with 1136 storage cells, of 'which 852 (75Ã) are available for storage of fuel assemblies. The unused cells are neutron flux traps (to maintain the required subcritical-ity) and are provided with cell blocking devices. Region II is used to store fuel which has experienced sufficient burnup such that storage in Region I is not required.

The spent fuel racks are fabricated from 304 stainless steel which is 0.135 inches thick. Each cell is formed by welding along the intersecting seams which enables the assembled cells to become a free-standing module which is seismically qualified without depend-ing on neighboring modules or fuel pool walls for support:

center-to-center spacing of the cells within both Region I The'ominal and II is 8.96 inches.

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NUCLEAR AND THERMAL-HYDRAULIC CONSIDERATIONS NEUTRON MULTIPLICATION FACTOR The following subsections describe the conditions in the spent fuel pool which are assumed in calculating the effective neutron multi-plication factor (Keff ff) the analysis methodology, and the analysis

'esults.

Normal Stora e

a. The analysis considers the most limiting storage condition. In Region I the racks are designed to store 4.5 wt X U-235 in a checkerboard fashion (Figure 3-1) with cell blocking devices in every other storage location.

In Region II the fuel is stored in 3 out of 4 locations with cell blocking devices in I out of 4 locations (Figure 3-2). In the criticality analysis for Region II, credit was taken for reactivity depletion in the spent fuel (consistent with Regula-tory Guide 1. 13, "Spent Fuel Storage Facility Design Hasis",

Draft R2).

b. The moderator is assumed to be pure water at a temperature applicable to the design basis condition which yields the highest reactivity.

c. The Region I and Region II arrays were assumed to be infinite in lateral extent and infinite in length.

d. 'echanical uncertainties (manufacturing tolerances, uncertain'ty of assembly position in storage racks, material tolerances,,

-etc.) are treated by performing sensitivity studies for the various uncertainties and applying an uncertainty in the K ff value.

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e. No control element assemblies (CEAs) or non-contained burnable poisons are assumed to be present.

3.1.2 Postulated Accidents The double contingency principle of ANSI N16. 1-1975 states that it is not necessary to assume concurrently two unlikely independent events to ensure protection against a criticality accident. This contingency principle is applied for the following postulated accidents: 1) dropping of a fuel element on top of the racks,

2) dropping of other objects into the spent fuel pool, 3) deforma-tion and relative position of racks due to tornado or earthquake, and 4) loss of one spent fuel pool cooling pump. The drop of the spent fuel cask onto the spent fuel racks was not considered in the criticality analysis since it is physically impossible to tip the cask into the pool (see Section 5.3. 1). The technical specifica-tions require that the boron concentration in the spent fuel pool be greater than or equal to 1720 ppm to ensure that k ff remains less than or equal to 0.95 for these accidents.

3.1.3 Calculation Methods A calculation bias has been determined from the comparison between the calculations and experiments and a calculation methodology has been determined such that K ff will be eff less than the calculated value 95 percent of the time with a 95 percent confidence level.

The total uncertainty value to be applied to the value of K ff for eff the storage racks is obtained from the expression:

N Total Uncertainty = Calculation Bias + [ E ak.] = 0.024 i=1 where hk.1 are the values of all other uncertainties, including mechanical, neutronic, and thermal variations.

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CEPAK The CEPAK lattice program is employed to calculate the basic broad group cross section data for the fuel assembly, spent fuel rack structure, and water. This program is a synthesis of a number of computer codes, many of which were developed at other laboratories, e.g., FORM, THERMOS, and CINDER (References 3-1, 3-2, and 3-3).

These codes are interlinked in a consistent way with inputs from an extensive library of differential cross section data.

NUTEST NUTEST is a two-dimensional integral transport code which employs the collision probability technique to compute sub-region dependent reaction rates in an explicit geometric representation of the fuel rods and associated structure of a fuel assembly. This code is used to calcul-ate the flux advantage factors which are applied as correc-tion factors to the basic broad group cross sections computed by the CEPAK lattice program to account for heterogeneous lattice effects not represented in either the multigroup spectrum or homogenized cell spatial calculation, e.g., heterogeneous fast fission effect in fuel pellets.

DOT-2W The spatial flux solution and multiplication factor for an infinite array of individual or clusters of fuel storage cells are computed with the two dimensional, discrete ordinates transport code, DOT-2W (Reference 3-4). The major features of the method used in this .code are:

a) 'nergy dependence is considered using the multigroup treatment.

b) The derivative terms and spat'ial dependence are approximated using a finite difference technique.

c) Dependence upon the direction variables is treated using the discrete ordinates method.

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d) The scattering integral is evaluated using a discrete ordinates quadrature in combination with a Legendre expansion of the scattering kernel to approximate anistropic scattering.

3. 1.4 Fuel Rack Modification The present spent fuel racks for St. Lucie Unit 2 are being complete-ly replaced as described in Section 1.2. The design of the new spent fuel racks is described in Section 4.1.

3. 1.5 Acce tance Criterion for Criticalit The acceptance criterion for the neutron multiplication factor (K ff) is that it be less than or equal to 0.95, including uncer-tainties, under all postulated conditions. For Region I the result-ing neutron multiplication factor (k ff) is 0.942 including all uncertainties and calculational biases. For Region II reactivity depletion is a function of the percentage of burnup achieved, not of the initial enrichment. The resulting reactivity, including uncer-tainties and calculational biases, as a function of burnup for several initial enrichments is shown in Figure 3-3. The minimum allowable burnup for a given initial enrichment is that correspond-ing to keff 0.95. When these minimum burnup values are adjusted upwards to account for axially-dependent burnup distribution.and are :

plotted as a function of initial enrichment, regions of acceptable and unacceptable burnup are identified (see Figure 3-4).

3.2 DECAY HEAT CALCULATIONS FOR THE SPENT FUEL POOL (BULK) 3,2,1 ~0 The Fuel Pool System provides continuous 'cooling for spent fuel assemblies stored in the fuel pool. This permits storage of spent fuel assemblies in the fuel pool from the time the fuel is unloaded from the reactor vessel until it is shipped offsite.

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The Fuel Pool Cooling System removes the decay heat from one-third of a core batch, which is assumed to have undergone irradiation for three cycles, placed in the spent fuel pool five days after shutdown and eleven previous refueling batches. With one fuel pool pump operating and one fuel pool heat exchanger in service and with a component cooling water temperature of 100'F, the maximum spent fuel pool water temperature does not exceed 131'F.

The Fuel Pool Cooling System also removes the decay heat produced in the fuel from a full core off-load placed in the spent fuel pool seven days after reactor shutdown, in addition to the decay heat from eleven previous batches of one-third core each. With two fuel pool pumps and one fuel pool heat exchanger operating, and with a component cooling water temperature of 100'F, the maximum spent fuel pool water temperature does not exceed 148'F. Evaporative cooling

.effects are neglected, for conservatism, in determining this maximum temperature. If Region II were expanded into Region I to reach the ultimate capacity of 1188 fuel assemblies, the total decay heat load would be less than that for the full core off-load case described above since such an expansion would preclude the high decay heat typical of the full core off-loaded into Region I.

V The Fuel Pool Cooling System also includes purification equipment designed to remove soluble and insoluble foreign matter from the fuel pool water and dust from the fuel pool surface. This maintains the fuel pool water purity'nd clarity, permitting visual observa-tion of. underwater operations.

The minimum design limit depth of water over the spent fuel to maintain the radiation dose levels to less than 2.5 mrem/hr is 9 feet'. The 'technical specifications require a m'inimum depth of 23 feet of water over the stored spent fuel assemblies; this is suffi-cient to meet the 2.5 mrem/hr dose level during fuel movement, to limit the maximum continuous radiation dose levels in working areas to much less than 2.5 mrem/hr during normal storage, and to ensure 3-5

that the offsite dose consequences during an accident are accept-able, Make-up to the fuel pool is from the refueling water tank, which is maintained at a technical specification limit of greater than or equal to 1720 ppm boron.

3.2.2 S stem Descri tion A description of the Fuel Handling and Storage Systems is provided in Section 9. 1 of the FSAR: The PSI diagram of the Fuel Pool System is shown on Figures 9. 1-6 and 9. 1-7a of the FSAR. The system process flow data are shown in Table 9. 1-5 of the FSAR. Radiation monitoring for the spent fuel pool area and Fuel Handling Building stack is discussed in Section 12.3.4 of the FSAR. The Fuel Pool Cooling System components and piping are guality Group C, seismic Category I and are described completely in FSAR Section 9. 1.3.

3.2.3 Safet Evaluation The calculations for the amount of thermal energy that may have to be removed by the spent fuel pool cooling system are made in accord-ance with Branch Technical Position ASB 9-2, "Residual Decay Energy for Light-Water Reactors for Long-Term Cooling", which is part of the Standard Review Plan (NUREG-0800).

3.2.3. 1 Normal Maximum Fuel Pool Cooling With Single Failure With one-third of a core batch, which is assumed to have undergone irradiation for three cycles, placed in the spent fuel pool five days after 'reactor shutdown and eleven previous refueling batches, the heat load is less than 15.3 x 10 Btu/hr. Under these condi-tions, with one'uel pool pump operating (single active failure of second pump) and one fuel pool heat exchanger in service, the spent fuel pool temperature does not exceed 131'F.

3-6

3.2.3.2 Abnormal Maximum Fuel Pool Cooling For a full core unloading, it is assumed that one full core is placed in the fuel pool seven days after reactor shutdown, along with eleven previous refueling batches of one-third core each. The resultant heat load from one full core and eleven refueling batches is 30.3 x 10 6 Btu/hr, the maximum heat load in the fuel pool. Under these conditions, with both the fuel pool pumps in service, the maximum fuel pool water temprature is 148'F. Pursuant to the guidance in SRP 9. 1.3 (NUREG-0800), there is no spent fuel pool water bulk boiling, and a cooling system single failure need not be, considered for this condition. This temperature is less than the value (150'F), referred to as T in Section 4.4.

3.2.3.3 Accident Maximum Fuel Pool Cooling Notwithstanding the guidance in SRP 9. 1.3, the fuel pool temperature for the full core off-load case described above is evaluated with one fuel pool pump inoperable. The resulting fuel pool equilibrium temperature is 161'F. Again, there is no spent fuel pool water bulk boiling.

3.2.3.4 Overall Performance Evaluation The fuel pool is provided with a seismic Category I Fuel, Pool Cooling System which maintains the water temperatures within accept-able limits. Two seismic Category I fuel pool pumps and two seismic Category I fuel pool heat exchangers are available before, during, and after a postulated Safe Shutdown Earthquake (SSE), and thereby provide adequate fuel pool cooling for the normal, abnormal and acciden't maximum heat loads described above. As noted previously,.

SRP 9. 1.3 does not suggest single active failure considerations even for the abnormal maximum fuel pool cooling analyses.

3-7

The purification loop normally runs intermittently during fuel pool operation to maintain the fuel pool water purity and clarity. The purification system can be operated with either the fuel pool ion exchanger or fuel pool filter bypassed. Local sample points are provided to permit analysis of fuel pool ion exchanger and fuel pool filter efficiencies.

For the abnormal conditions where pool temperature is above 140'F, there will be no detrimental effects to fuel movements, cooling system operation, fuel and fuel assemblies, or pool structures. The fuel pool ion exchanger will be manually isolated before cooling water temperature reaches 140'F. All fuel pool cooling components are designed for at least 200'F. Instrumentation is available locally with a control room alarm to monitor pool water temperature and enable shutdown of the purification system (an intermittent operation) so as to prevent ion exchanger resin damage. Addition-ally, any concerns related to increased fuel pool bulk temperature, are not expected to affect the environmental conditions to which the refueling, equipment operators will be exposed (e.g., air tempera-ture, humidity of airborne activity).

All connections to the fuel pool are made so as to preclude the possibility of siphon draining of the fuel pool. Any leakage from the fuel pool cooling system is detected by reduction in the fuel pool inventory. Makeup to the fuel pool is from the refueling water tank.

Unacceptable levels of radioactivity to maintenance personnel from the spent fuel pool are not anticipated while a heat exchanger is undergoing repairs. The fuel pool heat exchangers are enclosed in a

~ v separate room from the fuel pool. This design feature assures that maintenance personnel are not subjected to unacceptable levels of radioactivity. Ouring such repair, radiation levels in the fuel pool area are monitored continuously and access to this area is regulated accordingly; 3-8

3.2.3.5 Total Loss of Fuel Pool Cooling Although SRPs 9. 1.2 and 9. 1.3 (NUREG-0800) do not suggest analysis of the total loss of fuel pool cooling, this scenario was evaluated to ensure that ample time is available to restore cooling or to provide makeup water to the spent fuel pool. Multiple sources (seismic and non-seismic) of makeup water exist as discussed in Subsection 9. 1.3.3. 1 of the FSAR.

Under the assumption that all fuel pool cooling is lost and that makeup water is not supplied until after the pool reaches boiling, the decay heat in the spent fuel will cause boiling in a time period dependent upon the decay heat load.

Utilizing the maximum fuel pool decay heat load (Section 3.2.3.2),

the fuel pool water inventory would take 3.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> to boil, with a subsequent boi 1-off rate of 62.8 gpm. At this boil-off rate, 1.4 days would pass before the water level- dropped to the level (nine

'feet above the fuel) required to maintain acceptable radiation dose levels (see Section 3.2. 1). This is more than enough time to provide indication and alarms to the operators and to provide necessary repairs or supply makeup water to the spent fuel pool (see FSAR Section 9.1.3.3.1).

For the normal decay heat load (Section 3.2.3.1), the time to reach boiling conditions is 12.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and the subsequent boil-off rate is 31.7 gpm. At this boil-off rate, 2.8 days would pass before the water level dropped to the level required to maintain acceptable radiation dose levels. This is more than enough time to restore fuel pool cooling and provide makeup water.

for both the normal and maximum decay heat loads analyzed assuming a total loss of fuel pool cooling, the maximum cladding temperature was conservatively predicted to be 275'F, thus ensuring fuel clad-ding integrity.

3-9

POTENTIAL FUEL AND RACK HANDLING ACCIDENTS The spent fuel racks will be installed prior to storage of any spent fuel in. the spent fuel pool. Therefore, there is no potential for an accident involving spent fuel during fuel rack installation.

3.4 TECHNICAL SPECIFICATIONS The proposed changes to the technical specifications are summarized as follows:

1. Specification 5.6. 1a. is changed to indicate the total uncer-tainty value applied to the value of Keff ff for the spent fuel storage racks.

.2. Specification 5.6. lb. is changed to show the nominal center-to-center distance for the spent fuel storage racks.

3. Specification 5.6. 1c. is added to specify the required boron concentration for the spent fuel pool.

4. Specification 5.6. 1 is changed to delete the statement regard-ing dry storage of new fuel for the first core. This statement is no longer applicable.

5. Specification 5.6. 1 is changed to add a statement defining the enrichment/burnup requirements for storage of fuel in each region of the fuel pool.

6. Specification 5.6.3 is changed to show the capacity of the spent 'fuel storage r'acks.

'I 3-10

REFERENCES 3-1 FORM - A Fourier Transform Fast Spectrum Code for the IBM-7090, McGoff,- D. J., NAA-SR-Memor 5766, September 1960.

3-2 THERMOS - A thermalization Transport Theory Code for Reactor Lattice Calculations, Honeck, H., BNL-5816, July 1961.

3-3 CINDER - A One Point Depletion and Fission Product Program, England, T. R., WAPD-TM-334, Revised June 1964.

3 4 R. G. Soltesz, et. al., "Users Manual for DOT-2W Discrete Ordinates Transport Computer Code," WANL-TME-1982, December 1969.

3-11

FUEL ASSEMBlY CELL BLOCKING DEVICE "L" INSERT Figure R% SYSTEMS COMBUSTION ENGINEERING. INC.

SPENT FUEL RACK MODULE FOR REGION I 3-12

FUEL ASSEMBLY CELL BLOCKING DEVICE Figure KQ svsTEMs COMBUSTION ENGINEERING. INC SPENT FUEL RACK MODULE FOR REGION II 3-2 3-13

1.0 CO 0.9 ENRICHMENTS I

0 4.5 W/0 m

~nC Cg)Z 0.8 4.0 W/0 ITl I 3.5 W/0

~ n~ 3.0'/0 0.7 2.5 W/0 g) C 8 C m z zco 0.6 m~8 A

OO 0 Q Z Ill ~

Il D 0.5 n 10000 20000 30000 40000 50000 BUR NUP, IVIWD/MTU

D I- 30,000 A

ALLOWABLEBURNUP FOR REGION II U BURNUP ~ (11/00) x (ENRICHMENT W/0) cL'0,000 13,900 0'-

BURNUP REQUIRING STORAGE IN REGION I

~ 10000 1.5 2.0 2.5 . 3.0 3$ 4.0 4.5 5.0 INITIALU-235 ENRICHMENT, W/0 "

Figure SPENT FUEL STORAGE RACK CL% SYSTEMS INITIALENRICHNIENT vs BURNUP FOR FUEL ASSEMBLIES IN REGION II COMBUSTION ENGINEERING INC 3-15

4.0 MECHANICAL, MATERIAL, AND STRUCTURAL CONSIDERATIONS

4.1 DESCRIPTION

OF STRUCTURE 4.1.1 Descri tion of the Fuel Handlin Buildin A description of the Spent Fuel Storage Pool is provided in Section 9.1.2 of the FSAR. The spent fuel storage pool is located in, and an integral part of, the Fuel Handl ing Building. The general arrangement of the Fuel Handling Building showing the location of the spent fuel storage facilities is given on Figures 1.2-16 and 1.2-17 of the FSAR and the Fuel Handling Building design is describ-ed in Section 3.8.4. 1 of the FSAR. The Fuel Handling Building exterior walls, floors and interior partitions provide radiation shielding to plant personnel and protect the equipment from the effects of adverse atmospheric conditions including hurricane and tornado winds, temperature, external missiles and corrosive environ-ment. The design loading conditions and allowable stresses for the Fuel Handling Building are described in subsection 3.8.4.3 of the FSAR. The dynamic analysis of the Fuel Handling Building is des-cribed in Section 3.7 of the FSAR. The Spent Fuel Pool walls and floors are lined with type 304 stainless steel. The seismic cate-gory I spent fuel pool is being analyzed considering the spent fuel racks and additional fuel assemblies, to ensure that the design meets the criteria specified in FSAR Section 3.8.4.3.

4.1.2 Descri tion of S ent Fuel Racks 4.1.2.1 Design and Fabrication of Spent Fuel Racks The, spent fuel storage racks are fabricated with 304 stainless steel having a maximum carbon content of 0.0655. The racks are monolithic honeycomb structures with square fuel storage locations as shown in Figure 4-1. Each storage location is formed by welding stainless steel sections along the intersecting seams, permitting the assemb-4-1

led cavities to become the load bearing structure, as well as framing the storage cell enclosures. Each module is free standing, and seismically qualified without mechanical dependence on neighbor-ing modules or pool walls. This feature enables remote installation (or removal if required for pool maintenance) with minimal effort.

Reinforcing plates at the upper peripheral edges provide the requir-ed strength for handling.

Stainless steel bars, which are inserted horizontally through the rectangular slots in the lower region of the module, support the fuel assemblies. These support bars, when welded in place, support an entire row of fuel assemblies. Semicircular passages at the bottom of every cell wall allow cooling water to flow to all cells.

The size of the openings precludes blockage by any crud accumula-tions.

Loading of the fuel racks is facilitated via a movable lead-in funnel assembly containing four lead-in devices. The openings of the funnel assembly are symmetrical and the assembly sits on top of the rack module.

The module wall thickness is 0. 135 inch 304 stainless steel. The L-inserts are 0. 188 inches thick and are shown, along with the cell blocks, in Figure 4-'. As indicated in Figure 4-3, L-inserts are used only in Region I and cell blocks are used in both Regions I and II. The cell blocks for Region II are removable and are similar to those for Region I shown in Figure 4-5. The nominal pitch of the spent fuel racks is uniform throughout the nineteen modules to be contained in the spent fuel pool. This pitch is 8.96 inches center-to-center in both horizontal directions. Additional details are shown in Figures 4-4 and'4-5.

Region I is located within 6 modules and comprises a total of 448 cavities.. Region I is the high-enrichment, core off-load region.

The'fuel assemblies are to be stored in every other location in a checkerboard configuration (see Figure 3-1). The checkerboard arrangement makes 505 of the Region I ultimate storage capacity 4-2

I initially available for storage of fuel with high fissile concentra-tions. The unused. cavities are fitted with cell blocking devices to prevent inadvertent insertion of fuel into these locations. Region I may be expanded into Region II, if required, to store additional fuel which has not reached the required burnup. This would be accomplished by the addition of L-inserts and cell blocking devices into Region II.

Region I is designed for a total of 224 usable cavities for enrich-ments up to and including 4.5 w/o U-235. The cavities in Region I contain an L-insert (Figure 3-1). The L-shaped stainless inserts lock into the storage cavity using a spring locking mechanism on the upper end (Figure 4-2). This locking mechanism snaps into one of the holes in the four surrounding cell walls. These L-shaped 304 stainless inserts are neutron absorbers.

Region II consists of a total of 1136 cavities. klithin Region II, fuel assemblies are stored in 75% of the total cavities (see Figure 3-2) for an initial available storage capacity of 852 cavities.

Cell blocking devices'are used to preclude placement of fuel assemb-lies into every fourth cavity, which remains empty and provides a flux trap for reactivity control.'igure 4-3 shows the installation of Region I and Region II modules.

4. 1.2.2 Support of Spent Fuel Racks The spent fuel racks have been designed for direct bearing onto the spent fuel pool floor. A 10" support plate under the peripheral cells provides the bearing surface for the racks. Fuel rack module leveling is accomplished by placing 10" square stainless steel shims between the support plates and the fuel pool liner.

4.1.2.3 Fuel Handling The design of the spent fuel racks will not affect the conclusions of the fuel handling accidents presented in the FSAR (Section 4-3

15.7.4) -and summarized by the NRC in the Safety Evaluation Report (NUREG-0843). That is, the radiological doses for the postulated fuel cask and fuel assembly drop accidents are well within the 10 CFR 100 criteria.

The fuel handling accidents are presented in Section 5.3 of this report. In these analyses the maximum radial power peaking factor experienced by a spent fuel assembly was assumed to be 2.0 in order to bracket future fuel management changes, whereas the analysis in the FSAR assumed 1.65. Therefore, there are increases in the doses for the fuel handling accidents. These increases are not related to the spent fuel rack design and do not affect the conclusions of the analysis since the doses are still well within NRC guidelines (see Sections 5.3.1.1 and 5.3.1.3).

4.2 APPLICABLE CODES, STANDARDS, AND SPECIFICATIONS The spent fuel racks are designed in accordance with the following:

1. Code of Federal Regulations 10CFR Part 50:

a) Appendix A "General Design Criteria for Nuclear Plants,"

Criteria 2, 3, 4, 5, 61, 62, 63.

b) Appendix 8 "guality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants."

2. ASME Boiler and Pressure Vessel Code Section III, Subsection NF, "Nuclear Power Plant Components."

3; 'merican'Society for Testing Material's Documents:

a) ASTM - A240 - Specification for Corrosion Resisting Chromium Nickel Steel Plate, Sheet & Strip for Fusion-Welded Unfired Pressure Vessels.

4-4

b) ASTM - A276 - Specification for Stainless and Meat Resist-ing Bars and Shapes

4. American National Standards Institute:

a) ANSI - N210, Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations, 1976.

b) ANSI - N16.1, Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors, 1975.

5. United States Nuclear Regulatory Comnission:

a) Standard Review Plan, Section 9. 1.2, Rev. 2 "Spent Fuel Storage."

b) Regulatory Guide 1.13, Rev. 2 Draft, "Spent Fuel Storage Facility Design Basis."

c) Regulatory Guide 1.26, Rev. 3 "guality Group Classifica-tion and Standards for Water, Steam and Radioactive Waste Containing Components of Nuclear Power Plants."

d) Regulatory Guide 1.29, Rev. 3 "Seismic Design Classifica-tion."

e) Regulatory Guide 1.31, Rev. 2 "Control of Stainless Steel Welding" as modified by Bran'ch Technical Position MTEB-51, "Interim .Position on Regulatory Guide 1.31, "Control of Stainless Steel Welding."

f) Regulatory Guide 1. 122, "Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components, Rev. 1, February 1978.

4-5

g)'egulatory Guide 1.70, "Standard Format and Content of Safety Analysis Report for Nuclear Power Plants, Rev. 3."

h) NRC Guidance "Review and Acceptance of Spent Fuel Storage and Handling Applications" April 1978, and Modifications dated January 18, 1979.

4.3 SEISMIC AND IMPACT LOADS Maximum loads transmitted to the floor by the spent fuel racks at St. Lucie Unit 2 are given below.

Floor-Rack Interface Loads KIPS North-South East-West Vertical Down Dead Weight 0 59.7 Live 0 OBENS 104.1 0 102.9 EW 28.6 40.4 OBEy 0 0 8.8 SSENS 98.3 0 97.2 SSEEW 42.1 59.5 SSEV 0 0 15.5 The seismic analysis of the spent fuel rack inc'ludes an assessment of the maximum sliding and tipping that can be expected. The racks are installed with a nominal gap of 2"'between modules and are a minimum of 12 inches from the pool walls. The analysis has shown that the maximum motion of the racks, including tipping, sliding and thermal expansion is less than the gap between adjacent modules, therefore, no contact is predicted.

4-6

0' LOADS AND LOAD COMBINATIONS The loads and load combinations used in the structural analysis of the spent fuel racks are listed below and are consistent with NRC guidance in "Review an Acceptance of Spent Fuel Storage and Handling Applications" (Reference 4-1).

Load Combination (Elastic Anal sis) Acce tance Limit D+ L limits of NF 3231.1a D+ L+

0+L+To E

E'ormal Normal limits of NF 3231.1a Lesser of 2Sy or Su stress range D+ L+ To+ E Lesser of 2Sy or Su stress range D+ L+ Ta+ E Lesser of 2Sy or Su stress range D + L + Ta + Faulted .Condition Limits of NF 3231. 1c The abbreviations in the table above are those used in Section 3.8.4 of the Standard Review Plan where each term is defined except for Ta which is defined as the highest temperature associated with the postulated abnormal design conditions.

Consistent with NRC guidance (Reference 4-1), the provisions of NF 3231. 1 shall be amended by paragraphs c.2, 3, and 4 of the Regula-tory Guide 1. 124 entitled "Design Limits and Load Combinations for Class 1 Linear-Type Components Supports."

4.5 DESIGN AND ANALYSIS PROCEDURES 4.5.1 Methodo1 o Summa r The spent fuel storage racks are designed to withstand forces generated during normal operation, an Operating Basis Earthquake, or a Safe Shutdown Earthquake. Lateral and vertical seismic loads 4-7

along with fluid forces are considered to be acting simultaneously on the fuel racks. The racks are designed to assure rack structural integrity while at the same time keeping the fuel in a subcritical state.

Linear response spectrum methods are use'd for the vertical direc-tion. The lateral seismic responses of the spent fuel storage racks are determined using a non-linear time history analysis. Non-linear time history analyses are p'erformed for the lateral directions primarily because of fuel impacting. The effects of impacting structures significantly influence the stresses in both the storage structure and the fuel and, because they are non-linear in nature, can only be accounted for by performing more complex non-linear time history analyses.

The seismic input used for these analyses consists of the vertical response spectrum and the lateral acceleration time histories corresponding to the pool floor elevation at St. Lucie Unit 2. The analyses are performed in accordance with Reg. Guide 1. 122, Revision 1, February 1978.

The first step in the analytical procedure is to determine the dynamic characteristics of the fuel storage racks. This is. done by developing a three-dimensional finite element model of the structure and solving for the natural frequencies and mode shapes in air. The finite element code used in the study is SAP IV (see Section 4.5; 2).

The resulting dynamic characteristics are then incorporated into a non-linear representation of the entire system, which includes the fuel and the storage racks. The.CfSHOCK computer code (see Section 4.5;2), is used to determine the non-linear time history response of the system. The effects of impacting between the fuel and the storage rack are represented in the CESHOCK model. Because of the 4-8

close proximity of the structures, hydrodynamic coupling effects between the fuel, the storage rack and the pool are also included in the model. -(See Reference 4-2 for additional information.)

The racks are analyzed using a finite element model in the SAP IV code and the loads from Section 4.3. SAP IV output consists of membrane stresses and bending moments for each element as shown in Figure 4-6. When dealing with this type of element, the results are given per unit length; therefore, the stress caused by the moment will be arrived at by the following expressions:

oB

= MC/I for a unit length strip, I = lt C = t/2 C/I = 6/t W 0= 6M/t This approach applies to Mx, M , and Mx thus, the total stress in any one direction will be:

2 t memb The beams used at the bottom of each cavity opening to support the stored fuel assembly are included in the model as, lumped masses with no structural rigidities. The beams are then analyzed for a clearer understanding of the existing stress situation. Again the stress caused by the moment will be arrived at by the expression a = MC/I and the shear stress by the expression cr = P/A.

Com uter Code Descri tions The computer codes used in these analyses are described in the following subsections.

4-9

SAP IV SAP IV is a general structural analysis program for static and dynamic analysis of linear elastic complex structures. The finite element displacement method is used to solve for the disp'lacements and compute the stresses of each element of the structure. The structure can be composed of unlimited number of three-dimensional truss, beam, plate, shell, solid, plane strain-plane stress, thick shell, spring, axisymmetric elements. The program can treat thermal and various forms of mechanical loading as well as internal element loadings. Oynamic analysis options consist of eigenvalue solutions yielding frequencies and mode shapes, response history by mode superposition, response history by direction integration, and response spectrum analysis.'arthquake type of loading as well as time varying pressure can be treated. The output consists of displacements at each nodal point as well as internal member forces for each element.

The program being used at C-E is essentially equivalent to the version verified, documented, and released by the University of California (Reference 4-3).

CESHOCK The CESHOCK computer code performs transient, dynamic analyses of non-linear elastic systems. These systems can be either axial models having one degree-of-freedom per node or lateral ones having one rotational and one translational degree of freedom per node.

The response of a system is determined by numerically integrating (using a Runge-Kutta-Gill technique) its equations of motion.

Excitation can take the form of either initial conditions or time histories of applied accelerations, velocities displacements or 4-10

forces. The non-linearities can consist of gaps, friction, hyster-esis or non-linear springs. Hydrodynamic action can also be model-ed, with both on-diagonal (added mass) and off-diagonal (coupling) terms being considered.

The program automatically searches the response time histories and prints out the maximum and minimum values of all nodal accelera-tions, and member loads and can generate an optional output tape containing the complete response histories.

"CESHOCK" is an extensively modified, proprietary version of the "SHOCK" computer code developed by V. K. Gabrielson and R. T. Reese of Sandia Laboratories (Reference 4-4). It differs from the orig-inal in the areas of damping, coefficient of restitution, friction, hydrodynamic effects, hysteresis, input of time histories, output options, allowable problem size and the manner of inputting stiff-ness elements. CESHOCK has been verified by demonstration that its solutions are substantially identical to those obtained by hand calculations or from accepted analytical results via an independent computer code (References 4-4 and 4-5).

4.6 STRUCTURAL ACCEPTANCE CRITERIA The allowable stress limits for normal and faulted conditions as defined in the ASME Code,Section III, Subsection NF are:

Normal. Operating Conditions: Primary Membrane (P ) = 16,500 psi Primary Bending (Pb) = 16,500 psi Shear (~ ) = 11,000 psi Faulted Conditions: P = 33,000 psi P = 33,000 psi b

= 22,000 psi s

4-11

For similar spent fuel storage racks supplied by Combustion Engineer-ing for another utility, the maximum bending stress under the faulted condition was 29, 199 psi as compared to the allowable limits given above. The maximum stress for the St. Lucie Unit 2 storage racks will also be within the above allowable limits.

4.7 MATERIALS, QUALITY CONTROL, AND SPECIAL CONSTRUCTION TECHNIQUES 4.7.1 Material s The spent fuel storage racks are fabricated from 304 stainless steel with a maximum carbon content of 0.065% and conform to ASTM specifi-cation A240. The stainless steel for the storage cell walls is

0. 135 inches thick and for the L inserts the stainless steel is

0. 188 inches thick.

As stated in Section 9. 1 of the FSAR, the fuel racks being replaced are also made of 304 stainless steel and the fuel pool chemistry control (see Section 3.2.2) is the same as that described in the FSAR. Therefore, the compatibility of the fuel rack material with the pool water should be the same as that suamarized in the FSAR safety evaluation report (NUREG-0843).

4.7.2 ualit Control Florida Power 8 Light's and Combustion Engineering's Quality Assur-ance Programs ensure that all manufacturing and installation activ-ities conform to acceptable quality requirements throughout all areas of performance. The pertinent requirements of 10CFR50, Appendix B, and Combustion Engineering quality assurance report CENPD-210-A, Rev. 3 and Specification 00000-WQC-5.2 will be follow-ed. In addition', Florida Power 8 Light's Topical QA Report FPL-NQA-4-12

100A (approved by the NRC) describes guality Assurance requirements with which the design, procurement, and fabrication of the new fuel storage racks will comply.

4.7.3 Construction Techni ues Therei is no spent fuel in the St. Lucie Unit 2 spent fuel pool at this time, nor is there expected to be any when these spent fuel racks are installed. Therefore, no special administrative controls or procedures will be necessary to provide radiation protection.

Standard construction techniques and procedures will be utilized during installation to ensure worker safety and compliance with guidelines from the manufacturer.

4.8 TESTING AND IN-SERVICE SURVEILLANCE Since the new fuel racks do not include boron poison plate inserts, there is no requirement for special testing or surveillance.

4.9 REFERENCES

4-1 NRC Guidance "Review and 'Acceptance of Spent Fuel Storage and Handling Applications", dated April 1978 and modified January 19, 1979.

4-2 Longo, R., and Bailey, D. F., "Seismic Analysis of Spent Fuel Racks" ANS paper TS-7308 presented at the ANS Topical Meeting on Options for Spent Fuel Storage at Savannah, Georgia, Sep-tember 26-29, 1982.

'-3 'Bathe, K. J., Wilson, E. L., and Peterson, F. E., "SAP IV - A Analysis Program for Static and Dynamic Response of 'tructural 4-13

Linear Systems.", Report No. EERC,73-11, Earthquake Engineering Research Center, University of California - Berkeley, June 1973.

4-4 SCL-DR-65-34, "SHOCK - A Computer Code for Solving Lumped Mass Dynamic Systems", V. K. Fabrielson, January, 1966.

4-5 Topical Report on Dynamic Analysis of Reactor Vessel Internals Under Loss-of-Coolant Accident Conditions with Application of Analysis to CE 800 Mwe Class Reactors," Combustion En ineerin ,

Inc., Report CENPD-42, August 1972 (Proprietary).

4-14

L-INSERT L-INSERT LOCKING HOLE FUEL ASSEMBLY SUPPORT PLATE SLOT FLOW PASSAGES Figure K% SYSTEMS.

COMBUSTION ENGINEERING. INC.

TYPICAL SPENT FUEL STORAGE RACK MODULE 4-1 4-15

Figure TYPICAL SPENT FUEL RACK MODULE 4-2 SYSTEMS L-INSERT COM BUSriON ENGiNEERING. INC.

4-16

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4-17

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8.740

.188 8.740 SECTION A.A I

I I

I A 4 DETAILZ 164.7/8 46 WELDED CELL BLOCKING DEVICE "L"INSE RT MODIFIED "L"INSERT Figure SYSTEMS L-INSERTS 4-5 COMBUSTION ENGINEERING. INC 4-19

I

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~ L Figure POWER SAP IV COMPUTER CODE MEMBRANE STRESSES AND BENDING MOMENTS 4-6 COMBUSTION ENGINEERING. INC 4-20

COST/BENEFIT ASSESSMENT ANO ENVIRONMENTAL IMPACT 5.1 COST/BENEFIT ASSESSMENT 5.1.1 Need for Increased Stora e Ca acit A. FP&L currently has no contractual arrangements with any fuel reprocessing facilities.

FP&L executed three contracts with the Oepartment of Energy (OOE) on June 16, 1983 pursuant to the Nuclear Waste Policy Act of 1982, but the storage/disposal facilities will not be available for spent fuel. storage any earlier than 1998.

B. Table 5-1 includes a proposed refueling schedule for St. Lucie Unit 2, and the expected number of fuel assemblies that will be transferred into the spent fuel pool at each refueling until the total existing capacity is reached. At present the licens-ed capacity of Unit 2 is 675 storage cells. All calculations in the table for loss of full core reserve (FCR) are based on the number of licensed total cells in the pool.

C. The St. Lucie Unit 2 spent fuel pool contains no spent fuel assemblies at this 'time.

D. At present, there are no control rod assemblies or other components stored in the St. Lucie Unit 2 spent fuel pool.

E. Adoption of this proposed spent fuel storage expansion would not necessarily extend the time period that spent fuel assemb-lies would be stored on site. Spent fuel. could be sent offsi'te for final disposition 'under existing legislation, but the government facility is not expected to be available until 1998.

As matters now stand and until alternate storage facilities are available, spent fuel assemblies on site will remain there.

5-1

F. The estimate date when the spent fuel pool will be filled with the proposed increase in storage capacity is provided in Table 5-1.

5.1.2 Construction Costs Total construction cost associated with the proposed modification is 2.4 million dollars. This figure includes the cost of designing and fabricating the spent fuel racks; engineering costs for C-E, Ebasco and FPSL; and installation and support costs at the site.

5.1.3 Consideration of Alternatives A. There are no operational commercial reprocessing facilities available for FPL's needs, nor are there expected to be any in the foreseeable future.

B. At the present time, there are no existing available indepen-dent spent fuel storage facilities. While plans are being formulated by the DOE for construction of a spent fuel storage facility per the Nuclear Waste Policy Act of 1982, this facil-ity is not expected to be available to accept spent fuel any earlier than 1998.

C. At present, FPL has no license to transship fuel between facilities, nor are presently installed storage racks at Turkey Point Units 3 and 4 capable of (or licensed to) store fuel generated at St. Lucie Unit 2. St. Lucie Unit 1 will lose full core reserve capacity upon startup of cycle 8 spent fuel in 1986. Therefore, transfer of St. Lucie Unit, 2 spent fuel to

't. Lucie Unit 1 would only compound that. unit.'s storage problem'nd's not a viable option.

5-2

D. Estimates for costs of replacement power were calculated based on the last official rate of return. The assumption was made that the unit could be operated without maintaining full core reserve, thus cycle 10 in 1996 would be the last refueling possible with existing storage capacity. Table 5-2 indicates the average yearly fuel cost increases for St. Lucie Unit 2 after 3 years of reactor shutdown. Plant shutdown would place a heavy financial burden on Florida residents within FP8L's service area and can not be justified.

5.1.4 Resources Committed Reracking of the spent fuel pool will not result in any irreversible and irretrievable commitments of water, land and air resources. The land area now used for the spent fuel pool will be used more effi-ciently by safely increasing the density of fuel storage.

The materials used for new rack fabrication are discussed in Section 4.7. 1 of this report. The materials are not expected to signifi-cantly foreclose alternatives available with respect to any other licensing actions designed to improve the possible shortage of spent fuel storage capacity.

5. 1.5 Thermal Im act on the Environment The thermal analysis for the spent fuel pool cooling system is presented in Section 3.2 of this report. That analysis included the maximum spent fuel decay heat load for normal, abnormal, and acci-dent conditions. As a result of the reracking of St. Lucie Unit',

the maximum decay heat load for normal storage increased from 12.5 x 10 Btu/hr to 15.3 x 10 Btu/hr and for abnormal and accident 6

storage the heat load increased from 29.9 x 10 Btu/hr to 30.3 x 10 Btu/hr. This increased heat load results in an increase of 6'F in the maximum fuel pool water temperature for the "normal" case, and 1'F for the "abnormal" and "accident" cases. Since the evaporation rate is assumed to be zero, the increased decay heat load is also 5-3

the increased load on the cooling system and the increased heat rejected to the environment. The total plant heat load rejected to the environment is about 6.4 x 10 9 Btu/hr (see Section 5. 1.2.2. 1 of

~

the Environmental Report Operating License). Therefore, the per-centage increase in heat rejected to the environment is less than 0.05K for normal fuel storage and 0.01% for accident storage.

5.2 RADIOLOGICAL EVALUATION 5.2.1 Solid Radwaste No spent fuel is currently stored in the Unit 2 spent fuel pool.

St. Lucie Unit 2 began commercial operation in August 1983 and will not be refueled until at least October 1984. Therefore, no solid radioactive waste is currently generated by the spent fuel pool purification system, nor will any solid radwaste be generated during the actual reracking operation. S'ince the great majority of contam-ination collected by the purification system derives either from freshly unloaded fuel or the intermixing of spent fuel pool fluid with primary fluid during refueling, increasing the storage capacity has no significant effect on the quantity of waste collected. The effective change in increasing the storage capacity of the spent .

fuel pool is the retention of older fuel elements in the pool beyond the time when ".hey would have otherwise been shipped offsite for disposal or reprocessing.

Operati.ng plant experience with high density fuel storage (St. Lucie Unit I), has indicated that only 35 cubic feet of resins are generat-ed per year.

."' '5;2.2 Gaseous"Radwaste Krypton-85 would only be released to the pool water and subsequently to the refueling building atmosphere from leaking fuel assemblies.

For normal operating conditions, most of the krypton would come from 5-4

the most recently discharged batch of fuel. After the most recent batch has cooled in the pool for 12 months, the pressure buildup in a fuel pin which causes'the release of krypton will be small. Thus, any increase in Krypton-85 activity attributed to the increase in spent fuel pool storage capacity is expected to be small compared to the total quantity of all noble gases released.

Operating plant experience with high density fuel storage (St. Lucie Unit I), indicates that there were no measurable continuous releases of Krypton-85 over the past two years from the Fuel Handling Building ventilation system.

5.2.3 Personnel Ex osure A. Data is not available regarding recent gamma isotopic analysis of St. Lucie Unit 2 spent fuel pool water. In lieu, recent gamma isotopic analysis is provided in Table 5-3 for St. Lucie Unit I, which presently uses high density spent fuel storage racks. Fuel pool activities for St. Lucie Unit 2 are expected to be within the values presented in the FSAR in Section

'1.1.2.2 and Table 11.1-15.

B. External dose equivalent rates are not available for the St.

Lucie Unit 2 spent fuel pool; however, operating plant experi-ence (St. Lucie Unit 1) indicates radiation levels less than I MR/HR above the surface of the spent fuel pool.

C. Data is not available regarding airborne radionuclides for the St. Lucie Unit 2 spent fuel pool; however, operating plant-experience (St. Lucie Unit 1) indicates airborne levels are typically less than the minimum detectable activity of (GeLi Detector) used for gamma spectroscopy.

the'quipment D. Since operating experience has shown minimal airborne radio-activity, no increases in dose rate in the spent fuel pool area or at the site boundary are anticipated.

5-5

E. As stated in 5.2. 1, based upon operating plant experience with high density fuel storage racks, there will be no significant increase in the radioactive waste generated by the spent fuel pool purification system. This is because there is no signifi-cant increase in the radioactivity levels in the spent fuel pool water with high density fuel storage racks; therefore spent fuel pool cooling and purification filters need not be replaced more frequently. Thus the annual man-rem burden is not expected to increase due to the increased fuel storage.

F. Most of the crud associated with spent fuel storage is released soon after fuel is removed from the reactor. Once fuel is placed into the spent fuel storage racks, additional crud contribution is minimal.

The highest possible water level will be maintained in the spent fuel pool (> 23 feet above the fuel assemblies) to keep exposure as low as reasonably achievable. A fuel pool skimmer is used during purification operations to remove any floating pollutants, thereby further precluding the deposition of crud on the spent fuel pool walls.

G. Sections 5.2.3.e and 5.2.3.f indicate that operating experience with high density racks has shown no significant increase in radioactivity levels in spent fuel pool water or dose rates above the spent fuel pool. As stated in Sections 5.2.1 and 5.2.3.e, operating experience has shown no increase in the processing of solid radioactive waste or annual man-rem burden associated with it.

The St. Lucie Plant Radiation Protection program is described .

in Chapter 12 of the FSAR. Based on experience with the reracking of St. Lucie Unit 1, no modifications to the program

're anticipated as a result of the St. Lucie Unit 2 rerack.

5-6

5.I.4

~ ~ ~Rk Oi The ex>sting spent fuel storage racks have never been used to store spent fuel and consequently are not contaminated. The racks will be disposed through routine industrial means. The total weight of these racks is approximately 122,540 pounds. The racks consist of eight modules; four weighing approximately 13,140 pounds each and the other four approximately 17,520 pounds each.

5.3 ACCIDENT EVALUATION 5.3. 1 S ent Fuel Handlin Accidents 5.3.1.1 Cask Drop As discussed in FSAR Section 9. 1, the construction of the Fuel Handling Building, the design of the cask handling crane and the travel limit switch interlock circuitry are such that the spent fuel cask cannot traverse over the spent fuel. From the spent fuel cask storage, the crane moves the cask loaded with spent fuel assemblies out of the Fuel Handling Building to the decontamination area onto the shipping vehicle for offsite shipment. As discussed in FSAR Subsection 9. 1.4.3, although it is not likely, the potential drop of a spent fuel cask is about 43 ft just outside the Fuel Handling Building. It is conservatively postulated that a cask drop accident of 43 ft results in the damage of all the assemblies contained in the cas,k and the instantaneous release of activity to the atmosphere.

It is assumed that a spent fuel cask containing 10 irradiated fuel assemblies is in the pr'ocess of being moved with the cask suspended from the crane above the transport vehicle. Through some unspeci-fied failure, the cask becomes disengaged from the crane and falls 43 ft onto an unyielding surface. The activity from the damaged fuel bundles is postulated to be released to the environment, at ground level.

5-7

The analysis is based on Standard Review Plan 15.7.5. The models used for evaluation of the radiological doses are described in Appendix 158 of the FSAR.

The parameters and assumptions used are presented in Table 5-4. The data for a realistic analysis are also included in Table 5-4 for comparison purposes.

The cask drop accident analysis assumes that all the fuel is damaged and all of the gap activity is released at ground level to the environment. This consists of 10 percent of the noble gases, 30 percent of the Kr-85 and 10 percent of the radioactive iodine in the fuel rods at the time of the accident. The values assumed for individual fission product inventories included a radial peaking factor of 2.0 and a radioactive decay of 90 days.

The activity release to the environment is presented in Table 5-5.

The maximum offsite dose produced is no more than 24.0 rem for the two-hour inhalation thyroid dose at the exclusion area boundary.

The two-hour and entire event doses for both thyroid and whole body are shown in Table 5-6. Results are well within the acceptance criteria of 10 CFR 100 and NRC guidelines in Section 15.7.5 of the NRC Standard Review Plan (NUREG-0800). These doses indicate that results are lower than NRC guidelines by at least a factor of three and that those guidelines would be met even for a radial power peaking factor of 6.2, as compared to the value of 2.'0 assumed in this analysis.'.3.

1. 2 Overhead Crane No cran'e capable of carrying heavy loads can move in the area of the spent fuel pool. (The analysis of light 1'oads is summarized in Section 5.3.4 of this report.) Protection against dropping the spent fuel cask into the spent fuel storage pool is provided by the basic layout of the Fuel Handling Building (refer to FSAR Figures 1.2-16 and 1.2-17). The cask handling arrangement (roof opening vs.

5-8

pool location) makes it impossible to pass the cask over the spent fuel pool. The cask is assigned a separate storage pool adjacent to the spent fuel pool. The cask decontamination and washdown areas are located outside the Fuel Handling Building. The plan at Eleva-tion 96.83 feet shows the location and size of the roof opening through which the fuel cask crane hoist ropes can pass. The hori-zontal movement of the ropes, and therefore of the crane hook, is limited by the roof opening. The plan at elevation 19.50 feet shows the location and dimensions of the spent fuel pool. A comparison of the two plans shows that the crane hook is prevented from approach-ing the spent fuel pool by the limits of the roof opening.

Additional protection is afforded by the trolley bumpers and a set of limit switches working together with bridge and trolley brakes to prevent movement of the hook into the restricted area as shown on Partial Plan A of FSAR Figure 1.2-16. The primary set of limit switches and bridge and trolley brakes is backed up by an indepen-dent secondary set designed to perform the same function. Under these conditions the hook movement within the building is limited to a narrow corridor sufficient to bring the cask into the building and place it in the center of the cask storage pool. There are no safety related components located under the travel path of the spent fuel cask.

A cask dropped onto either of the separating walls between the spent fuel pool and cask storage pool, after dropping onto and tipping over the exterior wall, falls back into the cask storage pool. The design of the cask yoke will prevent the cask from dropping in other than vertical orientation. Please refer to FSAR Figures 9. 1-21 and 22 which illustrate the concept of the double yoke.

It is apparent that two failure modes exist:.

1. One side of the double yoke fails and the non-redundant main hook does not. The consequence of this is that, since the cask 5-9

will still be supported on three sides, the cask center of gravity will not drop and no significant pendulum motion can be initiated.

2. One side of the double yoke fails, followed by failure of the non-redundant main hook. The cask is expected to fall straight down.

In summary, the double yoke arrangement provides adequate protection against the initiation of pendulum motion in the event of a yoke failure. There is no conceivable failure mode which would cause the cask to fall over the separating wall into the spent fuel pool. A set of redundant upper limit switches prevents the crane from lifting the cask above Elevation 62.5 ft.

5.3.1.3 Fuel Assembly Drop The possibility of a fuel handling accident is remote because of the many interlocks and administrative controls and physical limitations imposed on the fuel handling operations (refer to FSAR Subsection 9.1.4). All refueling operations are conducted in accordance with prescribed procedures under direct surveillance of a supervisor technically trained in nuclear safety and fuel handling.

Notwithstanding the above, the fuel handling accident is assumed to occur as a consequence of a failure of the fuel assembly lifting mechanism resulting in the dropping of a raised fuel assembly onto the spent fuel pool.

The earliest anticipated time at which a spent fuel assembly could be handled is three days after shutdown. For this evaluation, dropping of a fuel assembly is assumed to occur, breaching the cladding and releasing the volatile fission products from the gas gap of the fuel pins. In addition to the area radiation monitor located in the spent fuel cask area, portable radiation monitors 5-10

capable of emitting audible alarms are located in this area during fuel handling operations. Doors in the Fuel Handling Building are closed to maintain controlled leakage characteristics in the spent fuel pool region during refueling operations involving irradiated fuel-. Should a fuel assembly be dropped in the fuel transfer canal or in the spent fuel pool and release radioactivity above a pres-cribed level, the airborne radiation monitors sound an alarm, alerting personnel to the problem. Airborne radiation monitors in the Fuel Handling Building isolate the normal Fuel Handling Building Ventilation System and automatically initiate the filtration systems (refer to Subsection 6.2.3 of the FSAR).

The models used for evaluation of the radiological doses are des-cribed in Appendix 15B of the FSAR.

, Assumptions and parameters used in evhluating the fuel handling accident are consistent with Regulatory Guides 1. 13 "Spent Fuel Storage Facility Design Basis" December 1975 (Rl) and 1.25 "Assump-tions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for BWR and PWRs" March 1972 (RO) recommendations, as shown in Table 5-7. The radioactive inventory of the 236 fuel rods was obtained by multiplying the activity of the most radioactive fuel rod 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after shutdown by a factor of 236. The calculational methods and assumptions described in Regulatory Guide 1.25 (RO) apply since:

a) the values for maximum fuel rod pressurization, b) peak linear power density for the highest power assembly discharged, c) maximum centerline operating fuel temperature for the assembly in item b) and d) average burnup for the peak assembly in item b) above 'bove, are less than the corresponding values in Regulatory Guide 1.25 (RO)..

The assumptions and parameters acceptable for a realistic analysis are also indicated in Table 5-7 for comparison purposes.

5-11

The radioactivity released to the environment is presented in Table 5-8. The maximum offsite dose produced is no more than 3.0 rem two-hour inhalation thyroid dose at the exclusion area boundary.

The two-hour and entire event doses for both thyroid and whole body are shown in Table 5-9. Results are well within the acceptance criteria of 10 CFR 100 and NRC guidelines in Section 15.7.4 of the NRC Standard Review Plan (NUREG-0800). These doses indicate that results are less than NRC guidelines by a factor of 25 and that those guidelines would be met easily even for a radial power peaking factor greater than that assumed in this analysis.

5.3. 1.4 Spent Fuel Handling Machine The spent fuel handling machine, as shown on FSAR Figure 9. 1-14, is a traveling bridge and trolley that rides on rails over the spent fuel pool, fuel transfer pool and cask loading pit. Motors on the bridge and trolley position the machine over the spent fuel assembly storage racks, the new fuel elevator, the upending machine, and fuel shipping cask. An overhead crane is used to transfer new fuel from the new fuel storage racks to the new fuel elevator. The spent fuel handling machine hoist is provided with a long handling tool which engages the fuel assembly to'e moved. Once the fuel assembly is grappled, the cable and hoist winch raise the fuel assembly.. The machine then transports the fuel assembly from the upending machine to the spent fuel storage racks (spent fuel), from the new fuel elevator to the upending machine (new fuel), or from the spent fuel storage racks to the fuel shipping cask.

The controls for the spent fuel handling machine are mounted on a console located on the spent fuel handling machine trolley. Coor-dinate location of the 'bridge is indicated by synchros at the console, and the trolley position is indicated by a pointer and .

target system.

5-12

During withdrawal or insertion of a fuel assembly, the load on the hoist cable is monitored to ensure that movement is not being restricted. Overload and underload setpoints are provided to interrupt hoist operation at preassigned levels of cable load, thereby protecting fuel assemblies during hoisting operations.

Positive locking is provided between the grappling device and the fuel assembly to prevent inadvertent uncoupling. The drives for both the bridge and the trolley provide close control for accurate positioning, and brakes are proved to maintain the position once achieved. In addition, interlocks are installed so that movement of the spent fuel handling machine is not possible'when the hoist is withdrawing or inserting an assembly.

Manually-operated handwheels are provided for bridge, trolley, and winch motions in the event of a power loss.

5.3.2

~ ~ ~Abel i ~

As described above in Section 5.3.1.2, it is not possible to drop the spent fuei cask onto the spent fuel racks. For the drop of the cask outside the fuel handling building, the conservative analysis summarized in Section 5.3. l. 1 above demonstrated that resulting radiological doses are well within acceptance criteria.

In evaluating the drop of a fuel assembly from the spent fuel handling machine into the spent fuel pool (Section 5.3.1.3), it was conservatively assumed that all fuel rods in the dropped fuel assembly were damaged. The analysis showed that the radiological doses were also well within acceptance criteria.

~

P 5-13

f 5.L3

~ ~ ~F1 D The present technical specification on fuel decay time prior to removal of irradiated fuel from the reactor vessel and the storage capacity of the spent fuel racks ensure that the total decay heat load from the spent fuel pool is less than the capacity of the fuel pool cooling heat exchanger. The fuel pool cooling analysis is described in Section 3.2.

Also, the technical specification on fuel decay time prior to removal of fuel from the reactor vessel ensures that the assumptions on radioactivity content of the fuel in the fuel assembly and cask drop accident analyses (Sections 5.3. 1. 1 and 5.3.1.3) are met.

5.3.4 Loads Over S ent Fuel As described in Section 9. 1 of the FSAR, there are no light loads (objects less than or equal to the weight of a fuel assembly) which could be carried over spent fuel in either the containment building or in the spent fuel pool which would result in fuel damage more severe than that assumed in the accident analysis (i.e., damage of all fuel rods in one fuel assembly).

As described in Section 5.3. 1.2 above, the designs of the fuel handling building and the overhead crane ensure that heavy loads cannot be carried over the spent fuel pool.

5.3.5 Conclusions Since the spent fuel cask cannot fall onto the spent fuel racks, the higher capacity of the spent fuel rack design does not affect, by itself, results of the fuel cask drop accident described above (Section 5.3. 1. 1). Also, the spent fuel rack design does not affect 5-14

the fuel assembly drop accident analysis because the design of the fuel racks precludes damage to more than one fuel assembly. There-fore, the fuel assembly drop accident is bounded by the drop of a single fuel assembly, as described in Section 5.3.1.3.

Since there will be only a negligible change in radiological condi-tions in the fuel handling building as a result of the higher capacity of the new fuel racks, no change in FP&L's radiation protection program is anticipated. Also, since the environmental consequences of the fuel handling accidents described in Sections 5.3. 1. 1 and 5.3. 1.3 are well within NRC guidelines, there will be no change to or impact on the previous determinations of the Final Environmental Statement. Therefore, the new fuel rack design and the proposed license amendments will not significantly affect the quality of the human environment and issuance of a 10 CFR 51.5(c)(1) negative declaration is appropriate.

5-15

7 (83K17)/ds--:56 TABLE 5-1 ESTIHATED SPENT FUEL POOL CAPACITY REQUIREHENTS ST. LUCIE UNIT 2 Approx. Cycle Total No. Assemblies Spaces Required Total No. Spaces Excess Augmented Cycle Startup in Pool from all for Full Core Needed During Storage Storage No. Date Previous C cles Reserve This C cle Available Re uired Existing New Existing New Capacity(1) Capacity(2) Capacity Capacity 217 675 1076 08/01/83 217 675 1076 2 12/06/84 80 217 297 378 779 3 05/15/86 164 217 381 294 695 4 12/01/87 240 217 457 218 619 5 05/01/89. 328 217 130 531 6 'I2/01/90 404 217 621 455 05/01/92 492* 217 709 367 34 12/01/93' 568 217 785 291 110 05/01/95 656 217 873 203 198 10 12/01/96 732ee 217 949 127 274 11 05/01/98'2 820 217 1037 39 362 12/01/99 896+ 217 1113 438 53 13 05/01/2001 984 217 1201 526 . 141 14 12/01/02 1060++ 217 1277 602 217 15 05/01/04 1148 217 1365 690 305 (1) Licensed Capacity At Present ~ 675 Cells (2) New Rack Installation = 1584 Total/1076 Initially Usable Cells

• FCR Lost at 458 Stored Assemblies with Existing Capacity ee Last Refueling Possible with Existing Capacity (Assumes No FCR Requirements)

+ FCR Lost at 859 Stored Assemblies with New Capacity

++ Last Refueling Possible with New Capacity Racks (Assumes No FCR Requirements)

TABLE 5-2 REPLACEMENT COST FOR ST. LUCIE UNIT 2 Base Case Energy Gen. Outage Case Re lacement Cost Prod. Cost By SL 2 Prod. Cost

$ 000 000 MWH $ O00) ($ 000 ~$ /MWH 1998 7,295,560 3,812 7,970,697 675,137 177.11 1999 8,422,404 4,818 9,343,839 921,435 191. 25 2000 9,069,199 4,883 10,117,841 1,049,642 214.96 NOTES; SL 2 out 5/1/98 through 12/31/2000 Based on SYP long term base case (7/22/83) 5-17

TABLE 5-3 GAMMA ISOTOPIC ANALYSIS FOR ST. LUCIE UNIT 1 PEAK WIDTH = 0.00 FWHM. CONFIDENCE LEVEL = 4.66 1-SIGMA NUCLIDE HALF-LIFE UCI / ml ERROR I ERROR X MPC AR-41 1.83H 5.726E-07 NUCLIDE NOT DETECTED KR-85 10.72Y 5.062E-04 NUCLIDE NOT DETECTED KR-85M 4.48H 2. 170E-06 NUCLIDE NOT DETECTED KR-87 76.30M 6.082E-06 NUCL IDE NOT DETECTED KR-88 2.84H 7.847E-06 NUCLIDE NOT DETECTED XE-131M 11. 90D 7.595E-05 NUCLIDE NOT DETECTED XE-133 5. 24D 5.922E-06 NUCLIDE NOT DETECTED XE-133M 2. 19D 1.676E-05 NUCL IDE NOT DETECTED XE-135 9.08H 2.027E-06 NUCLIDE NOT DETECTED XE-135M 15. 65M 1.856E-05 NUCLIDE NOT DETECTED XE-138 14. 17M 4. 921E-05 NUCLIDE NOT DETECTED I-130 12. 36H 1.807E-06 NUCLIDE NOT DETECTED I-131 8.04D 2.322E-06 NUCLIDE NOT DETECTED I-132 2.30H 2.078E-06 NUCLIDE NOT DETECTED I-133 20.80H 2.406E-06 NUCLIDE NOT DETECTED I-134 52.60M 2. 187E-06 NUCLIDE NOT DETECTED I-135 6.61H 2.376E-06 NUCLIDE NOT DETECTED NA-24 15.00H 3.758E-07 NUCLIDE NOT DETECTED CR-51 27.70D 1.873E-05 NUCLIDE NOT DETECTED MN-54 312.50D 1.493E-05 1. 347E-06 9.02 0.00 MN-56 2.58H 1.400E-06 NUCLIDE NOT DETECTED CO-57 270.90D 7.839E-06 1. 491E-06 19.02 0.00 CO-'58 70.80D < 1.660E-03 6.072E-06 0.37 0.00 FE-59 44.60D 1.965E-06 NUCLIDE NOT DETECTED CO-60 5.27Y 3.744E-04 3.124E-06 0.83 0.00 ZN-65 244.10D 2. 181E-06 NUCLIDE NOT DETECTED NI-65 2.52H 2.008E-06 NUCLIDE NOT DETECTED 5-18

TABLE 5-3 (Cont'd)

GAMMA ISOTOPIC ANALYSIS FOR ST. LUCIE UNIT 1 PEAK WIDTH = 0.00 FWHM. CONFIDENCE LEVEL = 4.66 1-SIGMA NUCLIDE HALF-LIFE UCI / 1111 ERROR g ERROR '5 MPC BR-82 35.30H 1.770E-06 NUCLIDE NOT DETECTED AG-110 252.20D 1.782E-06 NUCLIDE NOT DETECTED SN-113 115.00D 3.903E-06 NUCLIDE NOT DETECTED SB-124 60.20D 1.098E-05 9.856E-07 8.'8 0.00 W-187 23.90H 5.660E-06 NUCLIDE NOT DETECTED NP-239 2. 35D 7.261E-06 NUCLIDE NOT DETECTED RB-88 17.80M 5.229E-05 NUCLIDE NOT DETECTED ZR-95 63.98D 2.560E-06 NUCLIDE NOT DETECTED NB-95 35.15D 1.468E-06 NUCLIDE NOT DETECTED ZR-97 16. 90H 3.723E-05 NUCLIDE NOT DETECTED NB-97 72.10M 3.356E-06 NUCLIDE NOT DETECTED MO-99 66.02H 1.105E-05 NUCLIDE NOT DETECTED TC-99M 6.02H 1.800E-06 NUCLIDE NOT DETECTED RU-103 39.35D 2. 391E-06 . NUCLIDE NOT DETECTED SB-125 2.77Y < 3.757E-05 5.973E-06 15.90 0.00 CS-134 2.06Y 9.525E-04 5.162E-06 0.54 0.00 CS-136 13. 10D 1.471E-06 NUCLIDE NOT DETECTED CS-137 30. 17Y 1.398E-03 5.860E-06 0.42 0.00 BA-140 12.79D 1.047E-05 NUCLIDE NOT DETECTED PR-144 17. 28M 6.151E-04 NUCLIDE NOT DETECTED LA-140 40.22H 3.028E-07 NUCLIDE NOT DETECTED CE-141 32.50D 3.037E-06 NUCLIDE NOT DETECTED CE-144 284.30D < 1.382E NUCLIDE NOT DETECTED AC-228 5.75Y 1.633E-05 NUCLIDE NOT DETECTED RA-224 5.75Y < 4.398E-05 NUCLIDE NOT DETECTED PB-212 5.75Y 3.832E-06 NUCLIDE NOT DETECTED 5-19

TABLE 5-3 (Cont'd)

GAMMA ISOTOPIC ANALYSIS FOR ST. LUCIE UNIT 1 PEAK WIDTH = 0.00 FWHM. CONFIDENCE LEVEL = 4.66 1-S IGMA NUCLIDE HALF-LIFE UCI / ml ERROR 5 ERROR TL-208 5. 75Y < 2.382E-06 NUCLIDE NOT DETECTED TH-234 1000.00Y < 1.943E-06 NUCLIDE NOT DETECTED RA-226 1000.00Y < 5.161E-05 NUCLIDE NOT DETECTED PB-214 1000.00Y < 5.089E-06 NUCLIDE NOT DETECTED BI-214 1000.00Y < 4.885E-06 NUCLIDE NOT DETECTED K-40 1000.00Y < 3.860E-06 NUCLIDE NOT DETECTED TOTAL ACTIVITY = 5.028E-03 UCI/UNIT; TOTAL X MPC =

5-20

TABLE 5-4 SPENT FUEL CASK DROP ACCIDENT ASSUMPTIONS Design Realistic Basis Basis Assum tions Assum tions

l. Data and assumptions used to estimate radioactive sources from postulated accident:

Power level 2700 Mwt 2700 Mwt b) Burnup (any time during core life) NA NA c) Spent fuel damaged 100% 0. 1%

d) Number of assemblies in core 217 217

2. Data and assumptions used to estimate activity release a) Assumed spent fuel decay

. time prior to accident 90 days 90 days b) Radial peaking factor 2.0 1.65 c) Fuel rod release fractions: 10K noble gas 10K noble gas (all of the gas gap 30% Kr-85 305 Kr-85 activity is released) 10% iodine 105 iodine d) Number of assemblies in a cask 10 10

3. Dispersion Data a) EZ/LPZ distance (kilometers) 1. 56/1. 61 1.56/1.61 b) Assumed release level Ground liround c) X/Q's (Sec/m )

EAB (0-2 hr) 1.6 E-04 2.6 E-05 LPZ (0-8 hr 7.1 E-05 7.4 E-06 H

5-2T

TABLE 5-5 SPENT FUEL CASK DROP ACCIDENT RELEASES Activity Release to Environment (Curies) For the Design Basis Analysis

~jsoto e ~Act ivt t I-131 2.91 E+02 I-132 I-133 I-134 I-135 Kr-85m W Kr-85 1.89 E+04 Kr-87 Kr-88 Xe-131m 2.36 E+01 Xe-133 Xe-135 Xe-138 5-22

TABLE 5-6 SPENT FUEL CASK DROP OFFSITE DOSES Two Hour Exclusion Area Entire Event Low Boundar Dose Po ulation Zone Dose Thyroid 2.4 x 10 rem 1.1 x 10 rem Whole Body 6.1 x 10 rem 2.7 x 10 rem 5-23

TABLE 5-7 FUEL HANDLING ACCIDENT ASSUMPTIONS Design asis Rea istic Parameter Assum tions Assum tions

1. Data and assumptions used to estimate radioactive sources from postulated accidents:

a) Power level (MWT) 2700 b) Radial peaking factor 2.0 1.55 c) Burnup 4.5 full-power 4.5 full-power years at 80K years at 80%

plant factor plant factor d) Decay time (hr) 72 72 e) Number of failed rods 236 16 f)

Iodine'700 Fraction of fission product gases contained in the gap region of the fuel rods (percent)

Kr-85 30 30 Other Noble Gases 10 10 10 10

2. Data and assumptions used to estimate activity released:

a) Gap activity released to pool (percent) 100 100 b) Minimum water depth above damaged rods, (ft) 23 23 c) Pool decontamination factor for noble gases

. d)'ool decontamination factor for iodine Inorganic 133 Organic 1 Overall 100 500 5-24

TABLE 5-7 (Cont'd)

FUEL HANDLING ACCIDENT ASSUMPTIONS Design asis ea 1 stl c Parameter Assum tions Assum tions e) Iodine chemical form released to fuel. handling Inorganic iodine (percent) 75 75 Organic iodine (percent) 25 25 f) Filter Efficiency Iodine inorganic (percent) 95 95 Iodine organic (percent) 50 95 Iodine particulate 99 99

3. Dispersion Data:

a) EAB/LPZ distance (kilometers) 1. 56/1. 61 1.56/1.61 b) Assumed release level Ground Ground c) X/Q's (sec/m )

EAB 1.6 x 10 5

2.6 x 10-6 LPZ 7.1 x 10 7.4 x 10

4. Dose Data a) Method of dose calculation Appendix 15B, Section 15B.1 b) Dose conversion factors Appendix 15B, Table 158-1.

5-25

TABLE 5-8 FUEL HANDLING ACCIDENT RELEASES Activity Released to the Environment (Curies) For the Design Basis Analysis ISOTOPE 236 RODS I-131 3.4(+1)

I-133 8.5 0)

Xe-131m 3.7(+2)

Xe-133 9.3 +4 Xe-135 1.4 +2)

Kr-85 1.9(+3) 5-26

TABLE 5-9 FUEL HANDLING ACCIDENT OFFSITE DOSES Two Hour Exclusion Area Entire Event Low Boundar Dose Po ulation Zone Dose Thyroid 3.0 x 10 rem 1.3 x 10 rem Whole Body 1.1 x 10 rem 4.6 x 10 rem 5-27