ML20056E541

Safety Evaluation Supporting Amend 155 to License DPR-40

ML20056E541
Person / Time
Site:	Fort Calhoun
Issue date:	08/12/1993
From:	Office of Nuclear Reactor Regulation
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ML20056E539	List: ML20056E538 ML20056E540 ML20056E541
References
NUDOCS 9308240237
Download: ML20056E541 (19)

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UNITED STATES

[ )k; i j NUCLEAR REGULATORY COMMISSION WASMNGTON. D.C. 20555&M SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.155 TO FACILITY OPERATING LICENSE NO. DPR-40 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION. UNIT NO. 1 DOCKET NO. 50-285

1.0 INTRODUCTION

By letter dated December 7, 1992 (Reference 1), Omaha Public Power District (OPPD) requested an amendment to the Technical Specifications (TS) for the Fort Calhoun Station, Unit No.1 (FCS). The amendment would change TS 2.8, 3.2 (Table 3-5), 4.4.2, 5.10.3, and Figure 2-10, and is intended to permit expansion of the spent fuel pool (SFP) storage capacity from 729 assemblies to 1083 assemblies and to allow higher enrichment fuel to be stored in the SFP.

The licensee intends to accomplish the proposed expansion by replacing the existing racks with higher-storage-density racks consisting of 11 fru-standing rack modules in a discrete-zone, two-region storage system.

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The licensee has determined that the SFP, in its present configuration, will retain a full-core reserve discharge capacity through the end of Fuel Cycle 15 in 1995.

The licensee estimates that implementation of the proposed changes will defer the date for loss of full-core storage capability to the year 2007.

The licensee concluded that reracking of the SFP was the best available option to obtain the additional storage capacity for spent fuel assemblies necessary for continued operation of the facility.

The TS changes that OPPD has proposed include (1) revising TS 2.8 to increase the allowable fuel enrichment of assemblies stored in the SFP and to place additional restrictions on SFP boron concentration when unirradiated fuel is stored in the SFP, (2) revising TS 3.2 to incorporate changes to the test material surveillance program, (3) revising TS 4.4 to incorporate the design features of the new SFP racks, and (4) making editorial changes to TS 3.2 and 5.10.3.

The licensee's safety evaluation for the proposed SFP storage capacity expansion was detailed in the licensing report, " Licensing Report for Spent Fuel Storage Capacity Expansion, Fort Calhoun Nuclear Station," dated November 1992 (hereafter referred to as "the licensing report"). This report was prepared by Holtec International and was submitted as Attachment C to the licensee's letter dated December 7, 1992. The licensee clarified some aspects of its submittal during a technical meeting between the staff, OPPD, and the licensee's contractor, Holtec International, on necember 15, 1992.

Requests i

9308240237 930812 PDR ADOCK 05000295 P

PDR

for additional information (RAI) were issued by the staff on February 4, March 19, April 2, and April 8, 1993. OPPD responded to the RAls on March 19, April 28, and May 14, 1993.

In addition, an audit meeting was held at US Tool & Die (USTD), the manufacturer for OPPD, on April 20, 1993, and the staff issued a trip report (Reference 2) on April 30, 1993.

The March 19, April 28, and May 14, 1993, letters provided clarifying t

information that did not change the initial proposed no significant hazards consideration determination.

2.0 EVALUATION 2.1 Civil Enaineerina and Geosciences 2.1.1 High Density Racks The high-density spent fuel storage racks are seismic Category I equipment, and are required to remain functional during and after a safe shutdown earthquake (SSE). OPPD used a computer program, DYNARACK, for dynamic analysis to demonstrate the structural adequacy of the FCS spent fuel rack design under earthquake loading conditions. The proposed spent fuel storage racks are free-standing and self-sepporting equipment and are r.ot attached to the floor of the storage pool. A nonlinear dynamic model consisting of inertial mass elements, spring elements, gap elements, and friction elements as defined in the program was used to simulate three-dimensional dynamic behavior of the rack and the stored fuel assemblies, including frictional and hydrodynamic effects. The program calculated nodal forces and displacements at the nodes and then obtained the detailed stress field in the rack elements from the calculated nodal forces.

Two seismic analyses were performed: the 3-D single-rack model analysis and the 3-D whole pool multi-rack analysis. The main purpose of the whole pool multi-rack analysis was to investigate the fluid-structure interaction effects between racks and pool walls as well as those among the racks. These seismic analyses were performed utilizing the direct integration time-history method.

Four sets of seismic time histories were calculated from the plant response spectra as described in the FCS Final Safety Analysis Report (FSAR)

(Reference 3), each set consisting of three statically independent time histories for two horizontal and the vertical directions. The average calculated response spectra generated from these time histories envelop the FCS design response spectra.

In the 3-D single-rack model analysis, three rack geometries were considered for the calculation of stresses and displacements:

(1) 8 ft x 10 ft, (2) 8 ft x 11 ft, and (3) 10 ft x 12 ft.

Each rack was considered fully i

loaded, partially loaded, and almost empty with two different coefficients of 1

friction between the rack and the pool floor (g=0.2 and 0.8) to identify the worst-case response for rack movement and for rack member stresses and strains.

. Each of the three racks was subjected to the controlling loading condition (dead load + thermal load + SSE). The calculated stresses in tension, compression, bending, combined flexure and compression, and combined flexure and tension were compared with corresponding allowable stresses specified in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (1986 edition),Section III, Subsection NF. Tables 6.7.3-6.7.26 of Reference 1 present the stress factors for various rack geometries, friction and loading configurations. The stress factor is defined as the ratio of the calculated stress to the allowable stress of the ASME Boiler and Pressure Vessel Code,Section III, Subsection NF. The limiting value of each stress factor is 2.0 for load combinations corresponding to Level D service limits.

The results show that the stress factor varies from 0.01 (minimum) to 0.32 (maximum). Most stress factors are below 0.30, indicating that the induced stresses in the rack due to the postulated loading conditions are very small when they are compared to the allowable stresses of the ASME Code.

Tables 6.7.3-6.7.26 of Reference I also show the calculated horizontal displacements at the top and baseplate levels of the rack. The displacements at the baseplate level and at the top level are about 0.08 inch and 0.25 inch, respectively.

These computed horizontal rack displacements show that rack-to-rack impacts and rack-to-wall impacts would not occur during a SSE event.

OPPD also calculated the weld stresses of the rack under the SSE loading condition. Three weld locations were considered: (1) baseplate-to-rack, (2) baseplate-to-pedestal, and (3) cell-to-cell connections. Table 6.7.27 of Reference I shows the ratio of the calculated weld stress to the allowable stress specified in ASME Code Section III, Subsection NF.

The calculated ratios are in the range of 0.16 to 0.53 indicating that the weld connection design of the rack is adequate and acceptable.

Ir. the 3-D whole pool multi-rack analysis, all 11 racks were modeled together with the pool structure. All racks were considered fully loaded with one coefficient of friction (p=0.5) between the racks and the pool floor and were subjected to a loading condition of (dead load + thermal load + SSE).

The results of the multi-rack analysis indicate that the calculated stresses of a rack are higher than those obtained from the corresponding single-rack analysis. However, all calculated stresses of the multi-rack analysis are smaller than the allowable stress of the ASME Boiler and Pressure Vessel Code,Section III, Subsection NF, and are acceptable. The results of the multi-rack analysis also show that there is no rack-to-wall impact expected between a rack and its adjacent pool wall.

However, the multi-rack analysis indicates that there are rack-to-rack impacts. OPPD demonstrated that all rack-to-rack impacts are above or below the active fuel region of the racks. The impact loads are confined within and redistributed to the hardened impact protection i

system (3/16 inch bumper bars added to rack top corners), so that all stresses in the racks are smaller than the allowable stresses of the ASME Code. Thus, the analysis results are acceptable.

Although OPPD relied upon the results of relatively small computed rackstresses, strains and displacements as well as no potential for overturning under SSE conditions to demonstrate the structural design adequacy of the spent fuel rack, the staff made an independent assessment of the safety margin against overturning of a rack in order to supplement the findings obtained from the OPF0's DYNARACK analysis. The assessment is based on the principle of energy conservation, whereby the kinetic energy resulting from the maximum velocity of the rack induced by a SSE loading is equated to the potential energy that is needed to raise the rack to a position where the center of gravity of the rack is about to move beyond the line connecting the two supporting legs of the rack. A conservative factor of safety is defined as the ratio of the potential energy needed to raise the rack to the point of tipping over to the kinetic energy imparted to the rack by the SSE. The hydrodynamic effect is not considered in the staff analysis in order to reach a conservative conclusion.

The staff chose a rack geometry of 8 ft x 10 ft for overturning analysis since this geometry has the narrowest width among the three racks OPPD used and is identified as the most critical geometry based on the results of the OPPD's structural analyses. The factor of safety of about 1.3 against overturning was calculated when a fully loaded rack was considered.

This calculated factor of safety is larger than 1.1 factor, provided in the SRP Section 3.8.5, and indicates that the overturning of the rack would not occur under a SSE loading condition.

Based on (1) the OPPD's comprehensive parametric study (e.g., varying coefficients of friction, different geometries and fuel loading conditions of the rack), (2) large factor of safety of the induced stresses and strains of the rack when they are compared to the corresponding allowables provided in the ASME Boiler and Pressure Vessel Code,Section III, Subsection NF, (3) OPPD's overall structural integrity conclusions supported by both single-and multi-rack analyses, (4) the staff's independent assessment based on simplistic but conservative assumptions, and (5) the reliable fabrication methods and quality assurance / quality control (QA/QC) procedures adopted by OPPD and its vendors while manufacturing the rack structures, the staff concludes that the rack modules will perform their safety function and maintain their structural integrity under postulated loading conditions, and, therefore, are acceptable.

However, it is quite likely that the racks will move during or after seismic events. Therefore, OPPD is required to institute a surveillance program that inspects and maintains the originally installed rack gaps after occurrence of an earthquake equivalent to or larger than an operating-basis earthquake (OBE), if any occurs.

2.1.2 Spent Fuel Storage Pool The spent fuel pool structure is a reinforced concrete structure and is designed as a seismic Category I structure. The dimension of the FCS pool t

5-structure is approximately 21-feet wide and 33-feet long with a 12-feet thick reinforced concrete slab. The internal surface of the pool structure is lined with stainless steel to ensure water tight integrity.

The pool structure was analyzed by using the finite element computer program, ANSYS, to demonstrate the adequacy of the pool structure under fully loaded high density fuel racks with all storage locations occupied by fuel assemblies. The fully loaded pool structure was subjected to the load combinations specified in Standard Review Plan (SRP) Section 3.8.4 including the thermal loads. OPPD identified the six critical locations of the fuel pool slab and wall sections adjoining the pool slab based on the dynamic analysis.

Tables 8.5.2 and 8.5.3 of Reference I show the factors of safety for bending moments and shear forces, respectively, at the critical locations of the pool structure. The f actors of safety vary from 1.57 to 2.48 for bending moments and from 1.03 to 2.86 for shear forces at different critical locations.

The pool structural analysis was completed with adequate conservatism (e.g., use of consolidated fuel mass, use of a lower bound of concrete compressive strength, etc.) (Reference 4), and the factors of safety are acceptable.

In view of the calculated factors of safety, the staff concludes that the OPPD pool structural analysis demonstrates the adequacy and integrity of the pool structure under full fuel loading, thermal loading, and SSE loading conditions. Thus, the storage fuel pool design is acceptable.

2.1.3 Fuel Handling Accident The following three accident cases were evaluated by 0 PPD:

(1) drop of a fuel i

assembly with handling tools weighing 2780 lbs, which enters an empty cell and impacts the baseplate, (2) drop of a fuel assembly with handling tools weighing 2780 lbs, which impacts the top of a rack, and (3) drop of the spent fuel pool gate (11,522 lbs) onto top of a fuel rack.

3 The analysis results of Accident Case 1 above show that the load transmitted to the liner through the structure is properly distributed through the bearing

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pads located near the fuel handling area and that the liner is not damaged by impact. The analysis results of Accident Drop Case 2 shows that damage will be restricted to a depth of between 4.4" and 7.3" below the top of the rack, which is above the active fuel region.

The analysis results of Accident Drop Case 3 show that damage to the rack is confined to the region above the active fuel area with a depth of 1.17" below the top of the rack. The load transmitted to the liner through the structure is bounded by the load induced from the seismic loading conditions. The staff l

reviewed the OPPD analysis results submitted and concurs with the findings.

i Based on the review and evaluation of the OPPD's submittal, the staff audit at the vendor's manufacturing facilities, and additional information and analysis l

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provided by OPPD, it is concluded that the OPPD's structural analysis and design of the spent fuel rack modules and the spent fuel pool structure are adequate to withstand the effects of the required loads. The analysis and design are in compliance with current licensing basis set forth in the FSAR and applicable provisions of_ the SRP, and, therefore, are acceptable provided that OPPD commits to implement a surveillance program that inspects and maintains the originally installed rack gaps after occurrence of an earthquake equivalent to or larger than an OBE.

2.2 Materials and Chemical Enoineerino 2.2.1 Structural Materials The licensee has hired an outside contractor to perform a safety analysis for the proposed license amendment (LIC-92-0340A). The contractor has selected the following structural and welding materials for use in the proposed storage rack modification:

ASME SA240-304 for all sheet metal stock - this is a Type 304 stainless steel which conforms to ASME standard SA240 ASME SA240-304 for the internally threaded support legs ASME SA564-630 for the externally threaded support spindle - this is a precipitation-hardened stainless steel, which has been heat-treated to 1100*F Weld material - Type R308L stainless steel conforming to ASME Specification SFA-5.9 t

ASME SA240-304 stainless steel is a common austenitic alloy which has been used in nuclear applications. ASME Specification SA-240 requires that Type 304 stainless steel have a minimum yield strength of 30 ksi and a minimum ultimate tensile strength of 75 ksi. The choice of Type 304 stainless steel for the sheet metal stock and the internally threaded support legs is reasonable. The high nickel content stabilizes the austenitic phase of the steel at room temperature.

The high chromium content imparts reasonable corrosion resistance to oxidizing effects of most electrolytes at low concentrations. The steel is, however, susceptible to corrosion in acidic solutions (pH < 7.0) containing chloride or fluoride anions. These anions can lead to pitting or stress corrosion cracking of the material. The corrosion i

effects by chloride or fluoride anions are not as pronounced in basic media (pH > 7.0).

l The licensee has opted to use a SA564-630 steel, heat-treated at Il00*F, for the externally threaded support spindle. Type 630 (also called 17-4 precipitation-hardened (PH) steel) is a martensitic, PH stainless steel.

l Martensitic, PH, stainless steels have increased strength, without suffering considerable loss of ductility.

The corrosion resistance, however, is not

i quite as high as that of austenitic stainless steels. Heat treatments and aging of martensitic, PH stainless steels above 1025'F increase their i

resistance to stress corrosion cracking.

The procured specimens have been manufactured to conform to ASME Section.II Specification SA-564. SA-564 requires that a 630 steel heat-treated at 1100*F have a minimum yield strength of 115 ksi, a minimum tensile strength of 140 ksi, a minimum elongation (2-in. specimen) of 14%, and a minimum charpy L

impact energy of 25 ft-lb. This steel has been selected primarily for its high strength. This steel will also reduce the chance of galling with the Type 304 internally threaded support legs.

It should be noted that control of water impurities in nuclear plant spent fuel pool water is typically provided by the spent fuel pool demineralizers in i

the spent fuel cooling system. The demineralizers fun'ction to keep the chemistry of the spent fuel pool water approximately the same as that of the

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reactor coolant system, minimizing the probability of abnormal chemistry i

incursions during refueling operations when the two systems link together.

However, control of spent fuel pool chemistry also serves to reduce corrosion i

4 effects by keeping the concentrations of water impurities at low levels.

Therefore, stress corrosion cracking or pitting, induced by residual chloride or fluoride ions in the fuel pool, should not be a problem with either the SA240-304 stainless or SA564-630 precipitation-hardened stainless steels.

I The Type R308L weld filler metal has been selected for its similar composition i

to the Type 304 structural material. This material conforms to ASME Code

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Section II Specification SFA-5.9, and is suitable in gas metal arc welding and i

gas tungsten arc welding applications of both the SA240-304 and SA564-630 steels.

l 2.2.2 Poison Material l

t Boral - patented material produced by AAR Brooks and Perkins Boral is a cermet composite material made of Type 1100 aluminum and boron j

elements. The composite panel consists of two outer Type 1100 aluminum sheets which clad a sintered plate of boron carbide in a Type 1100 aluminum matrix.

Boral is typically manufactured in two thicknesses in accordance with AAR t

Brooks and Perkins Specification BPS-9000-01, " Item Specification for Boral, a

'l Neutron Shielding Material." Manufacture of Boral panels of.other thicknesses i

is done in accordance with the guidelines of BPS-9000-02.

BPS-9000-01 and -

9000-02 fall within the confines of the manufacturer's QA program for nuclear-

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grade materials.

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. The licensee intends to install the Boral sheets by freely inserting them between the 304 stainless steel walls of the rack assemblies and the 304 stainless steel sheaths.which are to be welded to the wall. The licensee has stated in its presentation to NRR on December 15, 1992, that the offset j

between-the rack assembly wall and the sheath cover is 0.08010.004 in. This f

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value applies to both the Region I racks and the Region II racks. The licensee stated in its safety evaluation that Boral is being procured by special order according to the following dimensions, parameters, and tolerances :

Boral sheet width 7.25 i 0.13 in.

Boral sheet length 128 in.

Boral sheet thickness 0.075f0.004 in.

Boral B' Areal Density 0.0151 g/cm2 (0.0140 g/cm min.)

2 These parameters conform to those stated in the manufacturer's specifications and material information documents. The length and the width of the sheet material falls within the tolerances imposed by the length and width imposed by the Type 304 stainless steel sheath; however, the sheet thickness (0.07510.004 in.) is not statistically different from the offset between the i

rack wall and the sheath (0.08010.004 in.).

It is evident that the insertion of the Boral panels will create a tight fit. A number of independent studies by industry organizations and by NRC contractors have shown that the reaction of Boral with water or moisture may generate hydrogen gas. This is evident from the following reaction between water and aluminum (Type 1100 is this case):

2Al(s) + (3+x)H 0(1)=== Al 0.xH 0(s) + 3H (9) 2 23 2

2 Production of hydrogen may result in deformation of the rack cells by imparting additional stresses on the walls.

Information Notice 83-29, " Fuel Binding Caused by fuel Bundle Deformation," was issued to alert the industry to this concern. The licensee has stated in its presentation to NRR on December 15, 1992, that holes will be drilled in the sheath areas to create a vent path for any hydrogen which may be produced by a water-aluminum reaction.

The licensee has also created a Boral surveillance program to characterize the performance of the Boral panels during the remaining lifetime of the plant.

This program is in accordance with the NRC letter of April 14, 1978, to all i

nuclear power licensees, which stated the staff's position that " Methods for verification of long-term material stability and mechanical integrity of special poison materials utilized for neutron absorption should include actual tests."

The licensee's Boral Surveillance Program calls for placing 10 Boral test coupons, suspended on a mounting jacket, in a designated cell that is surrounded by four spent fuel assemblies.

The coupons will be taken from the same lot as that used for the manufacture of Boral in the construction of the racks. The mounting jacket will be constructed from the same Type 304 stainless steel alloy as is used in the construction of the fuel pool racks.

Two additional, unjacketed coupons will be preserved as control specimens.

The surveillance program calls for removing and testing one coupon from the mounting jacket at the following intervals after the installation of the racks:

1 yr, 2 yr, 4 yr, 7 yr,10 yr,15 yr, 20 yr, 25 yr, 30 yr, and 35 yr.

During each of the first six cycles, the spent fuel assemblies surrounding the coupon mounting jacket will be replaced with freshly discharged spent fuel assemblies. This is done to ensure that the Boral coupons are exposed to a slightly higher radiation exposure than that which would be experienced by the Boral panels in the racks.

From the seventh refueling cycle on, the surrounding spent fuel assemblies will remain in place.

Each coupon, upon its removal from the mounting jacket, will be analyzed according to the following tests:

visual observation and photography neutron attenuation dimensional measurements (length, width, and thickness) weight and specific gravity Of these tests the neutron attenuation and the dimensional measurements are the most important since they are used to determine whether the coupons are exhibiting any signs of boron loss or structural deformation, respectively.

The gravimetric analyses will be performed to augment the results of the neutron attenuation studies if boron loss is indicated.

The OPPD's license amendment request submittal indicates that material selection for the Fort Calhoun spent fuel rerack modification has been satisfactorily thought out.

Except for the internally threaded support spindle, the rack is to be constructed mostly from a Type 304 stainless steel fabricated according to an approved ASME Section Il specification. The internally threaded support spindle will be constructed from an ASME Section II approved precipitation-hardened stainless steel. Boral is an acceptable poison material; however, since the Boral may generate hydrogen when in contact with water or moisture, care must be taken to provide a sufficient path to allow potential hydrogen generation to vent from the sheath area. After reviewing the licensee's submittal, the staff concludes that the licensee's selection of structural, welding and poison materials meets current industry and regulatory standards for use in the construction of the new rack modules.

2.3 Plant Systems 2.3.1 Control of Heavy Loads During the period from June 1993 to July 1994, the time period proposed for the reracking operation at Fort Calhoun, 574 out of the 729 cell locations will be occupied with spent fuel. The licensee has committed to develop a rack change-out and fuel reshuffle scheme to minimize the potential of damage to the stored fuel assemblies during the handling evolutions associated with the reracking effort. All work in the SFP area will be controlled and performed in accordance with specific written procedures and administrative controls to preclude the movement of a rack over any stored fuel.

i 1 In the safety evaluation issued pursuant to Amendment No. 57 to Facility Operating License DPR-40 for Fort Calhoun Station, Unit No. 1_ dated March 25, 1981, the staff found that there is adequate assurance that a single failure of the hoisting or braking mechanism of the auxiliary building crane will not result in a loss of the capability of the crane to retain a critical load.

In this safety evaluation, the staff also determined that the crane structure is designed and qualified to support the critical load during the Design i

Earthquake and the Maximum Credible Earthquake. The licensee has committed to employ the main hoist of the auxiliar; building crane, which is rated at 75 tons, in the reracking process.

In the licensing report prepared for the l

licensee by Holtec International, the main hoist of the auxiliary building crane is described as being qualified as single-failure-proof in accordance with the criteria of NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants," July 1980. The auxiliary hoist is not single-failure-proof.

The licensee also stated that a single-failure-proof lifting rig will be employed in the reracking operation.

Special lifting devices will comply with ANSI N14.6-1978, and non-custom lifting devices will be installed and used in accordance with ANSI B30.9-1971.

In addition, the licensee committed to conduct an inspection of the crane to be used in the reracking process before reracking operations begin.

Heavy loads associated with the reracking operation will not be carried in the

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SFP area until all fuel in the pool has decayed for a minimum of 2 months.

This provides for sufficient time for decay of gaseous radionuclides in the l

fuel such that an assumed accidental release of gases from damage to all stored fuel assemblies would result in a potential off-site dose of less _than i

10% of 10 CFR Part 100 limits.

The licensee has evaluated safe load paths for j

the reracking in accordance with the criteria of NUREG-0612 and has found them to be acceptable. The licensee also committed to administratively impose restrictions on the handling of all racks in the process of being transported.

I In addition, the licensee committed to establish a training program for crew members involved in the reracking process.

The staff has previously determined that the design of the crane to be used in the reracking operation is single-failure-proof. The licensee has committed j

to employ in the reracking process a single-failure-proof handling system 1

designed to meet the criteria of Section 5.1.6 of NUREG-0612. Cranes and

-associated lifting devices which conform to the criteria of NUREG-0612 for j

single-failure-proof handling systems satisfy the guidance of Regulatory l

Guide 1.13 and Section 9.1.5 of the SRP, NUREG-0800, and the-requirements of General Design Criteria 4 and 61 of Appendix A to 10 CFR Part 50 with regard to the design of heavy load handling systems. Therefore, the staff finds that the licensee has committed to employ an acceptable heavy load handling system in the reracking process.

The licensee has also committed to conduct operator training, perform a crane

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inspection, utilize safe load paths, and develop procedures for the reracking operation. These commitments comply with the criteria of Section 5.1.1 of i

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NUREG-0612.

In addition, the licensee plans to impose other administrative restrictions on the handling of the rack modules. These plans and commitments are consistent with the defense-in-depth approach of NUREG-0612 and the guidance of Section 9.1.5 of the SRP, and are acceptable.

Although the provision of a single-failure-proof handling system substantially reduces the probability of a heavy load drop event, the licensee committed not to carry heavy loads associated with the reracking operation in the SFP area until all fuel in the pool has decayed for a minimum of two months. This commitment reduces the consequences of a potential heavy load drop event and is acceptable.

2.3.2 Spent Fuel Pool Thermal-Hydraulics The licensing report states that the decay heat load calculation for the SFP was performed in accordance with the provisions of Branch Technical Position ASB 9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling," Rev. 2, July 1981.

In order to evaluate the total potential decay heat load, an inventory of 1159 fuel assemblies accumulated through the end of Cycle 27 was assumed to be present in the SFP. The heat load from the i

base inventory of 1159 fuel assemblies was calculated based on an assumed period of full power operation of approximately 4.3 years for each assembly.

For conservatism the quantity of fuel assemblies assumed to be in the SFP for the thermal-hydraulic analysis was selected to be greater than the actual capacity of the SFP.

t The licensee's analysis assumes that,15 months after the projected refueling for Cycle 27, the full-core of 133 fuel assemblies is discharged to the SFP.

The core is assumed to be transferred to the SFP at a rate of three assemblies per hour following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> decay in the reactor vessel. Of these assemblies, 45 are assumed to have 4.3 years of operation, 44 are assumed to have 2.9 year: of operation, and the last 44 are assumed to have 1.4 years of operation. The 88 fuel assemblies with the lower burnup are reloaded into the core prior to the end of a 56-day outage along with 45 new fuel assemblies.

After 30 days of full power operation and 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br /> decay in the reactor vessel, these 133 fuel assemblies are discharged to the SFP at a rate of three assemblies per hour.

Because of the time varying fuel assembly inventory described above, a transient calculation was performed to evaluate bulk pool temperature.

Convective heat transfer and evaporative cooling from the pool surface, and heat removal through operating SFP hea; exchangers were credited in the analysis. The heat removal rate through the single operating SFP heat exchanger was calculated based on a temperature effectiveness factor obtained by rating the heat exchanger on a proprietary thermal-hydraulic computer code.

In obtaining the temperature effectiveness value, the heat exchanger was assumed to be fouled to the design maximum extent, and five percent of the heat exchanger tubes are assumed to be plugged.

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The maximum SFP bulk temperature was calculated to be 135*F at 122 hours0.00141 days <br />0.0339 hours <br />2.017196e-4 weeks <br />4.6421e-5 months <br /> following the reactor shutdown for the first full-core offload.

The heat load removed by the SFP cooling system at this time was determined to be 20.72 million Btu /hr.

The maximum bulk temperature is below the 140*F design temperature limit for the SFP prescribed in the Fort Calhoun Station Updated Safety Analysis Report (USAR). The licensing report also evaluated the transient response of the SFP following a loss of all forced cooling.

The loss of cooling was assumed to occur coincident with the maximum bulk temperature.

The response was evaluated assuming no makeup water addition.

The calculated minimum time from the loss of pool cooling to the onset of bulk boiling conditions was determined to be 9.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> following the loss of forced cooling, and the maximum rate of inventory loss due to boiling was calculated to be 33.4 gpm.

e To verify that cladding integrity is not threatened, a model was developed to calculate the maximum local cladding temperature. The model was used to determine the location of minimum flow in an idealized, axially symmetric arrangement of fuel assemblies. The calculation assumed that the fuel assembly located in the minimum flow region is the most thermally limited. As an additional conservatism, the fuel assembly cladding was assumed to have a crud deposit which covered the entire surface.

For both unblocked and 50%

blocked flow conditions, the calculation indicated no incidence of nucleate boiling and no potential for fuel cladding damage.

Section 9.1.3 of the SRP provides guidance in evaluating the heat load imposed on the SFP cooling system. The guidance specifies evaluation of the following two scenarios:

a single active failure assumed coincident with a SFP inventory consisting of one refueling offload after 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> decay, one refueling offload after one year decay, and one refueling offload after 400 days decay; and a full-core offload after 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> decay, one refueling offload after 36 days decay, and one refueling offload after 400 days decay with no assumed equipment failures. The staff determined that the maximum heat removal rate calculated for the licensee's analysis conservatively bounds the decay heat rate calculated for these scenarios.

The maximum SFP bulk temperature determined from the licensee's analysis, which assumes a full-core offload with a single train of SFP cooling in operation, was calculated to be 135*F. This temperature is below the 140*F design temperature specified in the USAR and below the temperature criteria specified in SRP Section 9.1.3 for normal and abnormal discharges to the SFP.

Therefore, the guidance of Section 9.1.3 of the SRP is met with regard to providing adequate cooling for the postulated spent fuel inventory under normal abnormal operating conditions following the rerack.

i The licensee calculated a minimum time of 9.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> to reach bulk boiling conditions in the SFP following a loss of all forced cooling.

Based on this determination, the staff concludes that adequate time is available to provide makeup water to the SFP prior to the onset of bulk boiling.

Therefore, the i

. staff finds that the guidance of Section 9.1.3 of the SRP is met with regard to provision of makeup water.

For the potential fuel inventory following the proposed reracking of the SFP, the cooling and makeup water supply to the SFP is adequate to meet the intent of applicable guidance contained in Section 9.1.3 of the SRP. Therefore, the staff finds the proposed reracking acceptable with regard to potential SFP thermal-hydraulic concerns.

Licensee calculations of local fuel cladding temperature provide additional assurance that SFP cooling is adequate to protect cladding integrity following the proposed reracking.

The staff has reviewed the licensee's submittals for the proposed SFP reracking with regard to control of heavy loads and thermal-hydraulic concerns.

The staff determined that the licensee's commitment to comply with the criteria of NUREG-0612 with regard to control of heavy loads during the reracking operation is acceptable. The licensee's analysis demonstrated the adequacy of SFP cooling and makeup water systems in supporting the potential increased decay heat load permitted by the reracking process. The staff found this analysis to be acceptable in addressing potential SFP thermal-hydraulic concerns.

2.4 Radiation Protection 2.4.1 Occupational Exposure Control The licensee estimated in its December 7, 1992, application that total occupational exposure for planned reracking activities would be between 5 and 10 person-rem with no diving operations.

This overall estimate is based on individual dose estimates for each series of anticipated activities to be performed during the reracking operation. These activities include removing, washing (hydrolasing), and decontaminating empty racks, removing underwater appurtenances, installing new racks, and preparing old racks for shipment.

The licensee has indicated that the removal of underwater appurtenances shall be performed using remote handling tools to the greatest extent possible.

If diving operations are required, dose rates to divers would be minimized by placing fuel at a distance from the areas where the divers would be working.

Careful monitoring and adherence to procedures should assure that the radiation dose to the divers is maintained as low as is reasonably achievable (ALARA).

Further, if divers are used, the licensee has committed to the guidance provided in Appendix A to draft Regulatory Guide DG-8006 (" Control of Access to High and Very High Radiation Areas in Nuclear Power Plants").

The licensee noted that detailed procedures prepared with consideration of ALARA principles will be utilized.

In addition, Omaha Public Power District noted that continuous air samplers would be utilized where a potential for significant airborne activity exists and that personnel would wear protective clothing and, as appropriate, respiratory protective equipment.

Further, work activities are to be governed by Radiation Work Permits (RWP) specifying appropriate radiation protection measures.

In addition to the routine use of self-reading dosimeters and thermoluminescence dosimeters, extremity badges and alarming dosimeters will be utilized as appropriate, i

The licensee further stated that work activities, personnel traffic, and equipment movement will be monitored and controlled to minimize contamination and maintain personnel exposures ALARA. Therefore, based on our review of the licensee's application and information provided by the licensee by letter dated March 19, 1993, the staff finds the proposed radiation protection aspects of the spent fuel pool rerack acceptable.

2.4.2 Solid Radioactive Waste The licensee states in its application that the existing spent fuel storage racks will be ultimately shipped to a licensed disposal site as low-level radioactive waste. However, the racks may be required to be stored temporarily on site due to the uncertainty of operation of the Central States Compact low level radioactive waste facility. The licensee states that it has the capacity to properly store spent fuel storage racks on site on a temporary basis.

In addition, the licensee notes that, while a small amount of additional (spent) resins may be generated by the pool cleanup system on a one-time basis, a significant increase in the volume of solid radioactive waste is not expected to result from the increased storage capacity.

Based on our review, the staff finds that the licensee's plans for handling and disposing of solid radioactive waste generated in connection with the planned reracking operation meet regulatory requirements and are, therefore, acceptable.

2.4.3 Design-Basis Accidents In its application, the licensee evaluated the possible consequences of postulated accidents, included means for their avoidance in the design and operation of the facility, and provided means for mitigation of.their 1

consequences should they occur. The licensee has evaluated the effect of the changes on the calculated consequences of a spectrum of postulated design-basis accidents (DBAs) and concludes that the effect of the proposed TS changes is small and that the calculated consequences are within regulatory requirements and staff guideline dose values. Since the licensee proposes to utilize higher enrichment fuel, the staff reevaluated the fuel-handling accident for Fort Calhoun to consider the effects of increased enrichment and burnups.

In its evaluation for Fort Calhoun, issued on August 14, 1991, the staff conservatively estimated offsite doses due to radionuclides released to the

. atmosphere from a fuel handling accident. The staff concluded that the plant mitigative features would reduce the doses for this DBA to below the doses specified in NUREG-0800.

Since the applicant intends to utilize higher enrichment fuel the staff reanalyzed the fuel-handling DBA for this case. According to NUREG/CR-5009 (February 1989), increasing fuel enrichment to 5.0 weight percent U-235 with a maximum burnup of 60,000 MWD /T increases the doses for a fuel-handling accident by a factor of 1.2.

The licensee proposes to limit enrichment to 4.2 weight percent U-235. Therefore, the 1.2 factor increase in dose, displayed in Table 1 below, bounds the dose consequences of the licensee's proposal.

In Table 1, the new and old DBA doses are presented and compared to the guidelines doses in NUREG-0800 (established based on 10 CFR Part 100).

Table 1 Radiological Consequences of Fuel Handling Design Basis Accident (rem)

Exclusion Area low Population Zone Thyroid Thyroid Staff r

Evaluation August, 1991 53 19 Bounding Estimates for Higher Enrichment Fuel Burnup*

63.6 22.8 Regulatory Requirement (NUREG-0800 Chapter 15.7.4) 75 75

• Factor of 1.2 greater than original estimate for iodine.

The staff concludes that the only potential increased doses resulting from the fuel handling accident with increased enrichment are the thyroid doses; these doses remain well within the dose limits set forth in NUREG-0800 and are, therefore, acceptable.

2.5 Reactor Systems Two separate storage regions are provided in the spent fuel pool with independent criteria defining the highest potential reactivity in each of the two regions. Region 1 is designed to accommodate new fuel with a maximum enrichment of 4.20 weight percent (w/o) U-235 or spent fuel regardless of its discharge burnup.

Region 2 is designed to accommodate fuel of maximum initial enrichments up to 4.20 w/o which has accumulated. minimum irradiation levels

within the acceptable burnup domain depicted in Figure 2-10.

Region 2 may also accommodate any fuel assembly of maximum enrichment of 4.20 w/o which is mechanically coupled with a full-length control element assembly (CEA) with the provisions discussed later in this safety evaluathn.

The analysis of the reactivity effects of fuel storage in Region I and 2 was performed with the two-dimensional transport theory code, CASM0-3.

Independent verification calculations were made with the KENO-Sa Monte Carlo computer code using the 27-group SCALE cross-section library. Since the KENO-Sa code package does not have burnup capability, depletion analyses and the determination of small reactivity increments due to manufacturing tolerances were made with CASMO-3. These codes are widely used for the analysis of fuel rack reactivity and have been benchmarked against results from numerous critical experiments. These experiments simulate the Fort Calhoun spent fuel racks as realistically as possible with respect to t

parameters important to reactivity such as enrichment, assembly spacing, and t

absorber thickness. These two independent methods of analysis (KENO-Sa and CASMO-3) showed good agreement both with experiment and with each other. The intercomparison between different analytical methods is an acceptable technique for validating calculational methods for nuclear criticality safety.

To minimize the statistical uncertainty of the KENO-Sa calculations, t

1,250,000 neutron histories in 2,500 generations of 500 neutrons each were accumulated in each calculation.

Experience has shown that this number of histories is sufficient to assure convergence of KENO-Sa reactivity calculations. The staff concludes that the analysis methods used are acceptable and capable of predicting the reactivity of the Fort Calhoun storage racks with a high degree of confidence.

The criticality analyses were performed with several assumptions which tend to maximize the rack reactivity. These include:

(1) Unborated pool water at the temperature yielding the highest reactivity (4 C) over the expected range of water temperatures.

(2) Assumption of infinite array of storage cells in all directions (except for the assessment of peripheral effects and certain abnormal conditions where neutron leakage is inherent).

(3) Neutron absorption effect of structural material is neglected.

In addition, the design-basis fuel assembly was a Westinghouse 14x14 array of fuel rods with 20 rods replaced by 5 control rod guide tubes. Since Fort Calhoun also contains Combustion Engineering (CE) and Advanced Nuclear Fuels i

(ANF, now Siemens Nuclear Power Corporation) fuel designs, calculations were made for each of these fuel types and the Westinghouse fuel was determined to be the most reactive.

The staff concludes that appropriately conservative assumptions were made.

, For the nominal storage cell design, uncertainties due to boron loading tolerances, boral width tolerances, tolerances in cell lattice spacing, stainless steel thickness tolerances, and fuel enrichment and density tolerances were accounted for. These uncertainties were appropriately determined at least at the 95 percent probability, 95 percent confidence (95/95 probability / confidence) level.

In addition, a calculational bias and uncertainty were determined from benchmark calculations as well as an allowance for uncertainty in depletion calculations and the effect of the axial distribution in burnup. When combined with all known uncertainties, the final maximum calculated reactivity resulted in a k-eff of 0.928 (Region 1) and 0.935 (Region 2). This result meets the staff's criterion of k-eff no greater than 0.95 including all uncertainties at the 95/95 probabilii.y/

confidence level and is, therefore, acceptable.

The calculated maximum reactivity in Region 2 includes a burnup-dependent allowance for uncertainty in depletion calculations and, as mentioned above, provides an additional margin below the limiting k-eff criterion of no greater than 0.95.

Although not included in the criticality analyses, subsequent decay of Pu-241 with long-term storage results in a significant decrease in reactivity. This will provide an increasing subcriticality margin and further compensate for any uncertainty in the depletion calculations.

The attached Figure 2-10 shows that Region 2 can safely accommodate fuel of various initial enrichments up to 4.20 w/o and discharge burnups, provided the combination falls within the acceptable domain illustrated by the solid line.

This reactivity equivalencing method is the standard one used for storage rack reactivity evaluations and is acceptable.

The licensee has previously received credit for the reactivity decrease associated with the placement of used CEAs into fuel assemblies which do not meet the burnup requirements for storage in Region 2.

Calculations for this configuration assumed that the CEA was depleted to 75% of its initial boron-10 loading.

Since depletion to only 85% is estimated for used CEAs, additional conservatism is added with the additional depletion assumption.

In addition, after installation, a clip is attached to tie the CEA and fuel assembly together.

The grapple on the fuel-handling machine will be unable to remove the clip, thus preventing subsequent inadvertent CEA removal.

Based on the conservative depletion assumption and the mechanical latching of the CEAs, the staff finds that continued credit for CEA insertion is acceptable.

Because of the high neutron leakage from the boundary cells which face the pool walls, additional calculations were performed to determine the burnup required for safe storage in these cells. KENO-Sa calculations for these cells, assuming fuel of 4.20 w/o initial enrichment irradiated to 27,000 MWD /MTU with the remainder of the racks filled with fuel of 32,000 MWD /MTU burnup, resulted in a reactivity less than that for Region 2 filled with fuel of 4.20 w/o enrichment irradiated to 32,000 MWD /MTU.

Therefore, these boundary cells can safely accommodate fuel of the enrichment-burnup combination shown by the lower curve in Figure 2-10.

i Most abnormal storage conditions will not result in an increase in the k-eff of the racks. However, it is possible to postulate events, such as the inadvertent misloading of an assembly with a burnup and enrichment combination outside of the acceptable area in Figure 2-10 or dropping an assembly between the pool wall and the fuel racks, which could lead to an increase in reactivity. However, for such events, credit may be taken for the presence of boron in the pool water since the staff does not require the assumption of two unlikely, independent, concurrent events to ensure protection against a criticality accident (Double Contingency Principle). The Double Contingency Principle may be applied since Fort Calhoun TS 3.2, Table 3-4, Item 5 requires the sper,t fuel pool boron concentration to be determined once per 31 days.

In addition, plant test procedures implement the requirements of Table 3-4, Item 5, and establish an acceptable minimum limit. The reduction in k-eff caused by the boron more than offsets the reactivity addition caused by credible accidents.

In fact, the licensee has confirmed that a minimum boron concentration of only 80 ppm boron would be adequate to assure tha; the limiting k-eff of 0.95 is not exceeded.

The following Technical Specification changes have been proposed as a result of the requested spent fuel pool reracking. The staff finds these changes acceptable as well as the associated Bases changes.

t (1) TS 2.8 (10) has been modified to reflect the increase in allowable maximum enrichment to 4.2 w/o U-235.

(2) TS 2.8 (11) has been modified to reflect allowable storage in the peripheral cells of Region 2 for assemblies noted in Figure 2-10.

(3) Figure 2-10 has been modified to reflect the new criticality analyses evaluated and approved above.

1 (4) TS 4.4.2 has been modified to reflect the fact that both Region 1 and 2 cells are surrounded by Boral.

Based on the review described above, the staff finds that the criticality aspects of the proposed modifications to the Fort Calhoun spent fuel pool storage racks are acceptable and meet the requiremants of General Design Criterion 62 for the prevention of criticality in fuel storage and handling.

2.6 Administrative Chanaes TS 3.2, Item 10 on Table 3-5, is being revised to correct the spelling of "fuseable" to " fusible."

l TS 5.10.3 is being revised to correct references.

The references to TS 2.8(12) and 4.8.4 are incorrect and are being corrected to TS 2.8 and 4.4, respectively.

It is also proposed to replace the word " bundle" with

" assembly" to be consistent throughout the TS.

i i

j

. o These changes are administrative in nature; and therefore, the staff finds them acceptable.

3.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Nebraska State official l

was notified of the proposed issuance of the amendment. The State official had no comments.

4.0 ENVIRONMENTAL CONSIDERATTON Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact has been prepared and published in the Federal Reaister on August 5, 1993 (58 FR 41811). Accordingly, based upon the environmental assessment, the Commission has determined that the issuance of this amendment will not have a significant effect on the quality of the human environment.

5.0 CONCLUSION

i The Commission has concluded, based on the considerations discussed above, l

that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common i

defense and security or to the health and safety of the public.

6.0 REFERENCES

1.

OPPD, letter to US NRC, " Application for Amendment of Operating License,"

December 7, 1992.

i 2.

US NRC, trip report, " Trip Report: Meeting with Omaha Public Power District on the Structural Aspects of the Proposed High Density Spent Fuel Racks," April 30, 1993.

t 3.

Fort Calhoun Station, Unit 1, Final Safety Analysis Report.

s 4.

OPPD, letter to US NRC, " Request for Additional Information Concerning the Fort Calhoun Station Spent Fuel Pool Rerack," May 14, 1993.

l 1

Principal Contributors:

Y. S. Kim J. Medoff S. Jones J. Minns D. Carter L. Kopp Date: August 12, 1993 l

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