3F0999-07, Application for Amend to License DPR-72,to Change ITS Proposed in LAR 239,rev 0,increasing Licensed Capacity for Spent Fuel Assembly Storage in SFP & Revise Configuration for Storage of Fresh Fuel.Proprietary Encl Withheld
| ML20212C068 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 09/16/1999 |
| From: | Holden J FLORIDA POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20138D882 | List: |
| References | |
| 3F0999-07, NUDOCS 9909210126 | |
| Download: ML20212C068 (19) | |
Text
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F12rida Power
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September 16,1999 3F0999-07 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
Subject:
License Amendment Request #239, Revision 0 Enhanced Spent Fuel Storage
Dear Sir:
Florida Power Corporation (FPC) hereby submits a request for an amendment to its Facility Operating License No. DPR-72 for Crystal River Unit 3 (CR-3) in accordance with 10 CFR 50.90. The changes to the CR-3 Improved Technical Specifications (ITS) proposed in the attached License Amendment Request (LAR) #239, Revision 0, increase the licensed capacity for spent fuel assembly storage in the CR-3 Spent Fuel Pool (SFP) and revise the configuration for storage of fresh fuel.
FPC requests approval of LAR #239 by October 1,2000. Delivery of the new spent fuel racks to CR-3 is scheduled to be completed in December 2000, with rack replacement to begin in January 2001.
During the rack replacement activity, an interim configuration will exist where fuel will be stored in both the existing racks and the replacement racks. To accommodate this interim configuration, FPC requests that the amendment be issued.dlowing both the existing ITS and the proposed ITS to be effective during the implememation period.
FPC requests that the amendment be issued with an
. implementation peris ending September 1, 2001.
The supporting technical information provided as Attachment D contains information considered by Westinghouse Electric Company, LLC to be proprietary.
The Westinghouse application for withholding proprietary information from public disclosure pursuant to 10 CFR 2.790 is provided as Attachment G. Accordingly, FPC requests that Attachment D be withheld from public disclosure. A non-proprietary version of the supporting technical information suitable for release to the public is provided as Attachment H.
New regulatory commitments are identified in Attachment I.
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- J PDR ADOCK 05000302 pO P
PDR CRYSTAL RIVER ENER1Y COMPLEX: 15780 W. Power Line Street
- Crystal River, Florida 34428-6708 e (352)795-6486 A Moride Progrees cornpeny
q U. S. Nucl=r Regul-tory Commi sion 3F0999-07 Page 2 of 3 If you have any questions regarding this submittal, please contact Mr. Sid Powell, Manager, Nuclear Licensing at (352) 563-4883.
Sincerely, E$tL J.d J. Holden Vice President and Site Director JJH/rer xc:
Regional Administrator, Region II NRR Project Manager Senior Resident Inspector Attachments:
A.
Description of Changes, Reason for Request, and Evaluation of Request, No Significant Hazards Consideration, Environmental Impact Evaluation B.
Proposed Improved Technical Specifications and Bases Changes - Strikeout and Shadowed Text C.
Proposed Improved Technical Specifications and Bases Changes - Revision Bar Format D.
Enhanced Spent Fuel Storage Project-Engineering Input (Proprietary)
E.
Criticality Safety Analysis of the Crystal River Unit 3 Pool A for Storage of 5% Enriched Mark B-11 Fuel in Checkerboard Arrangement with Water Holes F.
Criticality Safety Analysis of the Westinghouse Spent Fuel Storage Racks in Pool B of Crystal River Unit 3 G.
Westinghouse Application for Withholding Proprietary Information from Public Disclosure H.
Enhanced Spent Fuel Storage Project-Engineering Input (Non-Proprietary)
I.
List of Commitments
p 4 '
- U. S. Nucl:ar Regulstory Comrnission 3F0999-07.
. Page 3 of 3 STATE OF FLORIDA COUNTY OF CITRUS-John J. Holden. states.that he is the Vice President and Site Director for Florida Power Corporation; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information, and belief.
il John Jr Folden Vice President and Site Director Sworn to and subscribed before me this _ l N day of Sed.
1999,by John J. Holden.
S$itaTure of N6tary Public 8, Notary PuMic.Steleof Rodde Ll$A ANN MC8R10E State of Florida I
I Wy Comm. Esp, Det.25,1989
- Comm No. CC 505458 L/3M 14 # #
/M!d//dl (Print, type, or stamp Commissioned Name of Notary Public)
Personally Produced Known I
-OR-Identification
FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 ATTACHMENT A LICENSE AMENDMENT REQUEST #239, REVISION 0 ENHANCED SPENT FUEL STORAGE i
Description of Changes, Reason For Request, and Evaluation of Request No Significant Hazards Consideration Environmental Impact Evaluation I
i
U.S. Nuclear Regulatory Commission Attachment A 3F0999-07 Page 1 of15 FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NO. 50-302/ LICENSE NO. DPR-72 LICENSE AMENDMENT REQUEST (LAR) #239, REVISION 0 ENHANCED SPENT FUEL STORAGE LICENSE POCUMENT INVOLVED:
CR-3 Improved Technical Specifications (ITS)
PORTIONS: Sections 3.7.14, 3.7.15, 4.3.1, and 4.
3.3 DESCRIPTION
OF CHANGES:
A.
Section 3.7.14, " Spent Fuel Pool Boron Concentration',
Required Action A.2.2, currently requires verifying by administrative means that a spent fuel pool verification has been performed in Pool A and in Pool B, Region 2, since the last movement of fuel assemblies in the spent fuel pool. This Required Action is revised to delete reference to
" Region 2" as part of Storage Pool B boron concentration verification since all of the new racks that will be placed into Pool B will be of the same design.
The fuel storage racks currently in Pool B are a combination of " Region 1" design and
" Region 2" design. The new Pool B racks are solely " Region 2" design. Because the racks in Pool B will be of the same design after the rack replacement Pool B will no longer have different designated regions.
B.
Section 3.7.15, " Spent Fuel Assembly Storage" 1.
LCO 3.7.15. This Limiting Condition for Operation (LCO) requires that the combination of initial enrichment and burnup of each fuel assembly stored in Pools A and B be within the acceptable region of Figures 3.7.15-1, 3.7.15-2, 3.7.15-3, or stored in accordance with the Final Safety Analysis Report (FSAR). This LCO is revised to delete reference to Figure 3.7.15-3 as a conforming change with eliminating the regions and their designation as a part of Pool B.
Also, the reference to storage in accordance with the FSAR is deleted.
2.
Surveillance Requirement (SR) 3.7.15.1.
This SR requires verification by administrative means that the initial enrichment and burnup of a fuel assembly is in accordance with Figures 3.7.15-1, 3.7.15-2, 3.7.15-3, or stored in accordance with the FSAR. This SR is revised to delete reference to Figure 3.7.15-3 as a conforming change with eliminating the regions and their designation as a part of Pool B. Also, the reference to storage in accordance with the FSAR is deleted.
U.S. Nuclear Regulatory Commission Attachment A 3F0999-07 Page 2 of15 3.
Figure 3.'7.15-1, "Burnup versus Enrichment Curve for Spent Fuel Storage Pool A." This Figure specifies the acceptable area of Burnup and Enrichment values for storage of a fuel assembly in Pool A.
An existing note is revised to clarify the area above the curve as the acceptable area of Burnup and Enrichment for unrestricted storage. A note is added to clarify that a fuel assembly of any combination of enrichment and burnup can be loaded in a one-out-of-two checkerboard configuration with empty cells. No change is made to the plot of paired Burnup and Enrichment values that derme Figure 3.7.15-1.
4.
Figure 3.7.15-2, "Burnup versus Enrichment Curve for Spent Fuel Storage Pool B, Region 1."
This Figure specifies the acceptable area of Burnup and Enrichment values for storage of a fuel assembly in the existing Region 1 of Pool B. The existing plot of paired Burnup and Enrichment values that define Figure 3.7.15-2 is replaced with two plots of new values for Pool B. Reference to Region 1 of Pool B is deleted. A note has been added to each area of enrichment versus burnup defined in the new figure to identify acceptable fuel assembly storage.
5.
Figure 3.7.15-3, "Burnup versus Enrichment Curve for Spent Fuel Storage Pool B, Region 2."
This Figure specifies the acceptable area of Burnup and Enrichment values for storage of a fuel assembly in the existing Region 2 of Pool B.
The figure is deleted in its entirety as a conforming change with i
eliminating Region 2 of Pool B.
C.
Section 4.3.1, " Criticality",
1.
Subsection 4.3.1.1, item c.
This item specifies the nominal center-to-center distance between fuel assemblies in Region 1 of Pool B.
The dimension of center-to-center distance is being changed from 10.6 inches to 9.11 inches. In addition, reference to " Region 1" of Pool B is deleted, thereby making the dimension applicable to all of Pool B.
2.
Subsection 4.3.1.1, item d.
This item specifies the nominal center-to-center distance between fuel assemblies in Region 2 of Pool B. This item is being deleted in its entirety based on all of Pool B being the same design.
3.
Subsection 4.3.1.1, item e.
This item specifies the nominal center-to-center distance between fuel assemblies placed in Pool A.
Item "e" is being renumbered to "d" as a result of the above deletion. No change is being made to the center-to-center distance.
U.S. Nuclear Regulatory Commission Attachment A 3F0999-07 Page 3 of 14 D.
Section 4.3.3, " Capacity" This section specifies the maximum storage capacity of the spent fuel storage pool.
Currently, this section specifies that the capacity is limited to no more than 1,357 fuel assemblies and six failed fuel containers. The value of the storage capacity is being j
changed from 1,357 to 1,474. No change is being made to the provision for storage of six failed fuel containers.
REASON FOR REQUEST:
Increased Spent Fuel Storage Capacity The current licensed CR-3 spent fuel storage capacity, specified in ITS Section 4.3.3, is 1,357 assemblies. There are 177 fuel assemblies in the CR-3 reactor core.
The currently available storage capacity for spent fuel at CR-3, allowing for the required Full Core Reserve (FCR), is projected to be exceeded in the year 2013. The CR-3 operating license has an expiration date of December 3,2016. Thus, the current capacity of the spent fuel pools is not adequate to allow all spent fuel discharged from the reactor to be stored onsite for the remainder of the CR-3 operating license.
The replacement racks will provide an additional 117 locations for storage of fuel assemblies.
As a result, the new racks will increase the total storage capacity of the spent fuel pools from the current 1,357 to 1,474 fuel assemblies.
The increased storage capacity will provide adequate spent fuel storage capacity for the remainder of the CR-3 operating license.
This increase in the licensed capacity of spent fuel storage to 1,474 assemblies is reflected in the proposed revised ITS Section 4.3.3.
Degradation of the Neutron Absorbing Material in the Current Pool B Racks The existing fuel racks in both Region 1 and Region 2 of the CR-3 Spent Fuel (SF) Pool B contain the neutron absorbing material Boraflex*. Boraflex* is a polymer-based compound of boron carbide. As discussed in NRC Generic Letter 96-04, Boraflex* degrades under long term exposure to the gamma radiation from the spent fuel.
FPC analyzed the remaining Boraflex* to determine if it would be adequate to maintain the existing racks subcritical as required by the CR-3 ITS. This analysis concluded that, based on the current degradation rate, the Boraflex* was adequate to maintain the racks subcritical. The analysis methodology and results were submitted for NRC review by License Amendment Request (LAR) #245, Revision 0, (FPC letter 3F1098-15 dated October 30,1998). The NRC 1
U.S. Nuclear Regulatory Commission Attachment A 3F0999-07.
Page 4 of15 approved this methodology by Amendment No.175 to Facility Operating License No. DPR-72 for CR-3, dated April 27, 1999.
The safety evaluation accompanying Amendment No.175 states that, assuming the worst case projected weight loss, the Boraflex* in the Pool B racks will be acceptable until the year 2019.
Upon NRC approval of LAR #239, FPC will replace the existing racks containing Boraflex*
with new racks containing Boral*. Assuming NRC approval of this LAR by the requested date, and based on delivery of the racks in December 2000, the rack replacement is expected to be completed no later than fall 2001. Thus, the Boraflex* is adequate to maintain the spent fuel racks in the pool subcritical as required by ITS until the existien racks are replaced.
The degrading Boraflex* also results in elevated levels of silica in the spent fuel pool water.
These elevated levels of silica are a problem with respect to proper water chemistry. The silica is removed by the SFP demineralizers in the SFP cooling and cleanup system. Spent resins must be disposed of as radioactive waste. The continued degradation of the Boraflex* is expected to result in further increases in silica levels in the SFP water. Increased use of the fuel pool cleanup system is needed to remove this increased silica content. This would cause an increased volume of spent resins which must be disposed of as radioactive waste. The increased generation of radioactive waste would result in higher personnel dose. FPC has used reverse-osmosis to reduce the silica level in the spent fuel pool water.
Revised Loading of Pool A Pool A is currently licensed for storage of irradiated fuel having acceptable combinations of enrichment and burnup as specified in ITS Figure 3.7.15-1. FPC has determined that spent fuel storage is optimized if Pool A is used for temporary storage of fresh fuel and irradiated fuel with some remaining useful life before being placed in the reactor, and if Pool B is used primarily for storing spent fuel with no remaining useful life.
Holtec International performed a criticality analysis of the CR-3 Pool A with fuel of the maximum licensed 5.0 weight percent initial enrichment. This annlysis identified that fresh fuel, and irradiated fuel with burnup less than a certain value, must be loaded in a one-out-of-two checkerboard configuration with empty cells. LAR #239 proposes changes to the ITS to allow for such checkerboard loading of fresh fuel in Pool A. The acceptable combinations of initial enrichmem and accumulated burnup for fuel assemblies which are to be loaded in this checkerboard configuration as well as those which can be loaded anywhere in Pool A (i.e.,
unrestricted) are specified in revised ITS Figure 3.7.15-1.
I
1 U.S. Nuclear Regulatory Commission Attachment A 3F0999-07 Page 5 of 14 EVALUATION OF CHANGES Current Configuration-Pool A is currently licensed for storage of fuel assemblies having enrichment-burnup combinations specified in ITS Figure 3.7.15-1. Pool A fuel racks have boron as the neutron poison in the form of boron carbide (B4C) in the rack walls.
Pool B is divided into two distinct geometric areas or regions, identified as " Region 1" and
" Region 2.".
There are eight storage racks in Pool B, four of " Region 1" design and four of
" Region 2" design. The Region I racks are used for temporary staging of fresh fuel prior to being loaded into the core and for storage of spent fuel assemblies. These racks are licensed for storage of fresh, unirradiated fuel in a checkerboard configuration with fuel assemblies having combinations of initial enrichment and burnup as shown in ITS Figure 3.7.15-2. The Region 2 racks are licensed for storage of spent fuel assemblies having combinations of initial enrichment and burnup values tLat comply with current ITS Figure 3.7.15-3. Pool B fuel racks have boron as the neutron poison in the form of Boraflex*, a polymer-based compound of boron carbide, in the rack walls.
Overview of the Rack Replacement Process The following is a summary of the anticipated process of removing the existing racks from Pool B and installing the new racks. The process includes removing the fuel from the existing racks, removing those racks, decontaminating the existing racks onsite and shipping them off-site for ultimate disposal, and installing the new racks in Pool B.
The configuration of the racks in Pool B is shown in Figure 1-1 of Attachment D. The spent fuel pools will remain flooded during the replacement of the fuel storage racks in accordance with existing ITS requirements. The plan for removing the existing racks and installing the replacement racks will prevent their movement over fuel assemblies unless the missile shields
. are in place, in accordance with FPC procedures and practices that implement NUREG-0612,
" Control of Heavy Loads at Nuclear Plants."
The fuel in the four racks in the south end of Pool B will be moved to the four racks in the north end of Pool B or to Pool A. Each of these four racks will be removed from the pool one at a time. After these four racks are removed, debris in the pool will be removed.
The racks will be decontaminated to the extent possible to minimize personnel exposure during 1
the rack replacement. One' method of minimizing exposure is by hydrolazing the racks underwater prior to removing them from the pool. This underwater hydrolazing is expected to reduce the background radiation dose rate below that which would occur if the rack was l
1
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U.S. Nuclear Regulatory Commission Attachment A 3F0999-07 Page 6 of15 exposed to air without this hydrolazing. The racks will be lowered into the decontamination pit adjacent to Pool B for rinsing. After the rinsing is complete the racks will be wrapped in plastic sheeting to minimize airborne contamination. The decontamination process will be completed on one rack at a time and that rack will be set aside on the spent fuel pool floor before the next rack is removed from the pool.
After the four racks have been decontaminated to the extent possible, they will remain on the spent fuel pool floor inside the Auxiliary Building until transported to an off-site facility for final disposal. The old racks will be removed and the new racks brought in through the Auxiliary Building floor hatch.
The first four new racks will be placed in the fuel pool ano luel will be transferred to them from the remaining four existing racks in Pool B. These remaining racks must be moved to the East End of Pool B before they can be removed in order to avoid lifting the racks over the spent fuel in Pool B. These racks will be decontaminated to the extent possible, as were the first four racks. Just as with the first four racks, these remaining racks will be stored inside the Auxiliary Building until they are transported to an off-site facility for final disposal.
A more detailed discussion of the rack replacement process is provided in Section 1.4.2.2 of Attachment D.
Interim Rack Configurations and Implementation of Revised ITS Existing ITS Figures 3.7.15-2 and 3.7.15-3 specify the acceptable values of initial enrichment and burnup for fuel assemblies to be stored in the existing racks of Pool B, Regions I and 2, respectively. The proposed ITS Figure 3.7.15-2 specifies the values of initial enrichment and burnup for fuel assemblies to be stored in the new Pool B racks. During the process of removing the existing racks and installing the new racks, there will be an interim period during which fuel will be stored in both the existing and the new racks. As a result, both the existing and the proposed ITS sections, figures, and associated ITS Bases will be applicable during this interim period. To accommodate this situation administrative controls will be implemented to maintain both the current and proposed versions of ITS and ITS Bases in the controlled copies of the CR-3 ITS during the rack replacement process. When all existing racks have been replaced with the new racks, the current ITS and associated ITS Bases will be removed.
Pool B Replacement Racks The eight existing rack modules in Pool B are being replaced with eight new rack modules designed by Westinghouse Electric Company, LLC. All of these new racks are of the same design. As a result, Pool B will no longer have different regions. With the new racks, Pool B will be used primarily for the storage of spent fuel. The neutron poison to be used in the racks is boron in the form of Boral*, L aoron carbide and aluminum-composite sandwich. A detailed I
U.S. Nuclear Regulatory Commission Attachment A 3F0999-07 Page 7 of 15 description of the replacement racks, including appropriate figures, is provided in Section 1.3 of Attachment D.
The revised ITS Figure 3.7.15-2 proposed in this amendment request specifies the acceptable vQes of initial enrichment and burnup for the spent fuel assemblies in Pool B.
t Fuel and Rack Handling FPC-qualified operators will make the fuel movements required for the rack replacement project. FSAR Section 9.6.2.7 provides the controls for performing these fuel movements.
The plan for fuel movements within Pool B will consider the spent fuel to be discharged from the core during the Cycle 11 refueling outage, scheduled for fall 1999, and placed into Pool B.
FPC procedures and practices that implement the criteria of NUREG-0612 will be followed for I
conducting all load-handling operations. No heavy loads will be moved over spent fuel in the fuel pools without the missile shields being installed. The equipment, controls, and processes
{
currently planned to be used for removing the existing Pool B racks and installing the replacement racks are described in greater detail in Section 1.4 of Attachment D.
Fuel Criticality Analyses Separate criticality analyses of fuel storage in Spent Fuel Pools A and B were performed by Holtec International and are provided as Attachments E and F, respectively. Both analyses I
were performed using fuel of the maximum licensed 5.0 weight percent initial enrichment.
For normal conditions, no soluble boron in the pool water was assumed.
For accident j
conditions, soluble boron was assumed consistent with the " double-contingency" principle.
The analyses for both pools were performed using the principle of reactivity equivalencing.
This principle allows for decreased reactivity based on fuel burnup.
Pool A The Pool A criticality analysis is presented in Holtec Repc,rt H1-992285, dated August 1999, provided as Attachment E of this amendment request.
This analysis was performed to evaluate the acceptability of loading fresh 5.0 weight percent fuel in a one-out-of-two checkerboard configuration in Pool A.
The report addresses both normal storage and accident storage conditions for this checkerboard loading. The analysis demonstrates the acceptability of storing fresh fuel in a one-out-of-two checkerboard configuration with cells filled with water. The maximum reactivity of the Pool A racks with fresh fuel of 5.0 weight percent initial enrichment stored in this checkerboard configuration was determined to be 0.8466.
This checkerboard configuration in Pool A is the proposed method for fresh fuel storage. The enrichment-burnup limits for fuel storage in Pool A are specified in revised ITS Figure 3.7.15-1.
U.S. Nuclear Regulatory Commission Attachment A 3F0999-07 Pnge 8 of 15 Pool B The Pool B criticality analysis is presented in Holtec Report HI-992128, dated May 1999, provided as Attachment F of this amendment request. This analysis evaluated the criticality of fuel stored in the replacement storage rack modules.
The analysis determined the minimum burnup value for fuel assemblies of various initial enrichments in unrestricted storage and in peripheral cell storage (i.e., those cells adjacent to the fuel pool walls). The burnup value for assemblies stored in the periphery is less restrictive as a result of neutron leakage in that region.
Both sets of these minimum burnup values are shown in Figure 1-1 of the report. The revised enrichment-burnup limits for unrestricted storage and for peripheral cell storage are specified in revised ITS Figure 3.7.15-2.
The results of the analysis confirm that the storage rack modules can sathy accommodate fuel with initial enrichments of up to 5.0 weight percent, with assuranci that the maximum reactivity, including calculation and manufacturing uncertainties, will be less than 0.95 km, with 95% probability at the 95% confidence level, provided the fuel conforms to the enrichment-burnup limits defm' ed in Figure 1-1 in the report.
Thermal-Hydraulic Analysis FPC determined the expected fuel pool heat loading from the proposed increased capacity of fuel pool storage by performing calculations. The maximum heat load generated in the pools is based on off-loading the full reactor core.
This maximum Spent Fuel pool heat load, in conjunction with limiting service water temperature and heat exchanger fouling, was used to determine Spent Fuel pool cooling requirements. The available Spent Fuel pool cooling methods while the core is fully off-loaded are:
one Decay Heat Removal train aligned to cool the Spent Fuel pools; e
two Spent Fuel Cooling Trains running in parallel; or e
one Spent Fuel Cooling pump aligned through two Spent Fuel Heat Exchangers.
e The assumptions, methodology, and results of these calculations are described in detail in Section 2.0 of Attachment D.
The thermal-hydraulic analysis was performed using the 1994 version of the ANSI /ANS-5.1 standard. This is a change from the 1979 version of the model that was part of the CR-3 licensing basis. The 1994 version has multipliers that are more conservative than those in the 1979 version. Use of the 1994 version was implemented pursuant to 10 CFR 50.59.
l 1
U.S. Nuclear Regulntory Commission Attachment A 3F0999-07 Page 9 of 15 Fuel Handling Accidents Three different orientations were analyzed for the drop of a fuel assembly. The assembly assumed was a standard Babcock and Wilcox (B&W) fuel assembly plus the handling tool, weighing 2,750 pounds. The three different orientations considered were:
(A) the fuel assembly drops vertically onto the top of a rack; (B) the fuel assembly drops in an inclined position onto the top of a rack; and (C) the fuel assembly drops through an empty cell location in the rack without impacting the rack.
For (A) and (B) it was determined that the deformation of the fuel rack will be localized at the top of the rack. The plastic deformation is between 1.9 and 6.0 inches. It was concluded that the functional capability of the fuel racks is maintained during these fuel drop scenarios. For (C), the bearing load on the steel liner was determined to be 506 kips.
This compares favorably with the allowable bearing load of 688 kips. The punching load on the concrete was 506 kips. This compares favorably with the allowable load of 2,322 kips for punching shear.
Additional details regarding the assumptions, methodology, and results are provided in Section 3.3 of Attachment D.
Seismic Analyses of the Pool B Replacement Racks I
The storage racks are required to be Seismic Class I equipment and must remain functional during and after a Safe Shutdown Earthquake (SSE). The racks are free standing and the fuel is free to move inside the cell. Therefore, a non-linear seismic analysis was performed. The ANSYS code, a general-purpose computer code for finite element analysis, was used for all rack structural and seismic analyses.
One set of synthetic time histories, consistent with the CR-3 Auxiliary Building seismic design requirements, was created for use in the non-linear se'ismic analysis of the racks. Three models were used in the evaluation of the fuel racks. These were (1) Static Model of the Fuel Rack Structure; (2) Single Rack Dynamic Model; and (3) Whole Pool Multiple Rack Dynamic Model. Damping, fluid interactions, and friction coefficient were modeled in the analysis.
The seismic analyses performed employed non-linear dynamic analyses. Sliding and uplifting behavior was included in the model. These analyses demonstrated that there is no sliding even when using the conservatively low value of 0.2 for the coefficient of friction. The seismic time history analyses demonstrated that there was no uplift of a fully loaded rack. The cliding and tilting motion is less than the clearances, and there is no impact between adjacent rack modules or between a rack and the pool walls.
l
U.S. Nuclear Regulatory Commission Attachment A 3F0999-07 Page 10 of 15 The fully loaded rack configuration has been analyzed and shown to experience no sliding or uplift. However, it was recognized that rack uplift and tip-over might be more severe for a partially loaded fuel rack. As a result three different fuel-loading configurations for the racks were analyzed. These were: (1) empty rack; (2) quarter-filled rack; and (3) half-filled rack.
No overturning or sliding occurred in these loading configurations.
All of the rack configurations evaluated meet appropriate factors of safety against tilting.
Additional details regarding the seismic analyses models, assumptions, process, and results are provided in Section 3.4 of Attachment D.
Materials and Quality Control The new fuel racks have been designed and will be constructed with consideration given to mechanical and material qualifications, neutron poison, fuel handling qualifications, fuel dimensional compatibility, and accident conditions. The rack designs, material selection and fabrication process will comply with the applicable ASTM standards for service in the nuclear and the boric acid environments. The governing quality assurance requirements for fabrication of the racks meet the quality assurance and quality control requirements of 10 CFR 50 Appendix B. All materials used in the construction of the racks are compatible with the spent fuel pool environment. Surfaces that make contact with the fuel assemblies are made from austenitic stainless steel. Other materials are corrosion resistant and will not adversely impact the stored fuel assemblies.
The new racks will be made of annealed austenitic 304 L Series stainless steel, with stainless steel welds. These racks have been in use for 30 years in the nuclear industry with no significant deterioration of the stainless steel.
The new racks will contain baron carbide as the neutron poison in the form of Boral*. Boral*
is a metallic composite of a hot rolled (sintered) aluminum matrix containing baron carbide sandwiched between and bonded to Type 1100 alloy aluminum. The boron carbide is a stable chemical compound. Boral* has a long history of use in spent fuel, aols where it has maintained its neutron attenuation capability. Because its operating history has demonstrated that Boral* is chemically stable, FPC does not plan to conduct a prograra of periodic surveillances of Boral* coupons.
Additional details regarding the material of rack construction, the Boral* neutron absorbing material, and quality assurance provisions are provided in Section 3.6 of Attachment D.
U.S. Nuclear Regulatory Commission Attachment A 3FJ999-07 Page 11 of15 NO SIGNIFICANT HAZARDS CONSIDERATION FPC has evaluated the proposed License Amendment Request (LAR) for the spent fuel rack replacement project against the criteria of 10 CFR 50.92(c) to determine if any significant hazards consMeration is involved.
FPC has concluded that this proposed LAR does not involve a significant hazards consideration. The following is a discussion of how each of the 10 CFR 50.92(c) criteria are satisfied.
(1)
Involve a sigmficant increase in the probability or consequences of an accident previously evaluated.
The LAR proposes to increase the onsite storage capacity of spent fuel and to revise the fresh fuel-loading configuration.
The licensee is replacing the existing spent fuel storage racks with new storage racks with a different neutron absorbing material. The licensee has reanalyzed the criticality of the revised storage configuration for fresh fuel.
The replacement storage racks and the revised fuel storage configuration do not affect any structure, system or component, nor process related to the operation of CR-3. As a result, the proposed LAR will not change the probability or consequences of any accidents related to operation previously evaluated. Thus, only those accidents that are related to movement and storage of fuel assemblies could be potentially affected by the proposed LAR.
Fuel handling accidents (FHA) are analyzed in Section 14.2.2.3 of the CR-3 Final Safety Analysis Report (FSAR). These include a FHA inside the Reactor Building (RB) and a FHA outside the RB. The LAR involves storage of fuel assemblies, which is an activity conducted outside the RB only. Therefore, only the FHA outside the RB is potentially affected.
The FHA outside the RB is postulated as the dropping of a fuel assembly into the spent fuel storage pool that results in damage to a fuel assembly and the release of the gaseous fission products. The current FHA assumes all 208 fuel pins in the dropped assembly are damaged. The results of that analysis demonstrate that the applicable 10 CFR 100.11 dose acceptance criteria are satisfied. Thus, the consequences of a FHA are not increased by the installation of the high-density racks. The high-density racks only increase the storage capacity and do not change the frequency or method for handling fuel assemblies. Thus, the probability of a FHA is not increased.
The increased spent fuel storage capacity will result in a negligible increase in the heat input to the spent fuel pool and'its cooling system. The limiting heat load is from the combined impact of stored fuel and a full core off-load. The full core off-load accounts for approximately 90% of that heat load.
The increase in stored fuel capacity, numerically less than 10%, is comprised of fuel that has been stored the longest
U.S. Nuclear Regulatory Commission Attachment A 3F0999-07 Page 12 of15 resulting in less decay heat. Thus, the impact of the increased spent fuel storage capacity on the total heat load is less than 1 %.
The increased fuel pool capacity and the revised fuel loading configuration do not increase the probability of a full core off-load.
The FSAR specifies the normal upper limit of the fuel pool cooling system as 160 F.
Administrative controls regarding when fuel movements from the reactor to the fuel pool can be completr.d are implemented to assure this upper limit is not exceeded.
Because neither the probability nor the consequences of a FIIA are increased, and because there is not any significant additional heat input to the spent fuel pools, it is concluded that the LAR does not involve a significant increase in the probability or consequences of an accident previously evaluated.
(2)
Create the possibility of a new or different kind of accidentfrom any accident previously evaluated?
Onsite storage of spent fuel assemblies in the spent fuel pools is a normal activity that CR-3 has been designed and licensed for. As part of assuring that this normal activity can be performed without endangering public health and safety, the ability of CR-3 to safely accommodate different possible accidents in the spent fuel pools such as dropping a fuel assembly or the misloading of a fuel assembly have been analyzed.
The increased spent fuel pool storage capacity proposed by the LAR does not change the methods of fuel movement or fuel storage. Thus, the proposed LAR does not create any new or different kind of accident from those previously evaluated.
The process of replacing the storage racks will involve removing the existing racks from the pool and installing new racks. These movements of the storage racks will be performed with the racks empty of all fuel. Even empty, these racks are of such weight as to be considered heavy loads. Movement of these empty racks create the potential for a heavy load drop. Movement of these empty racks will be restricted such that they will not be moved over any spent fuel stored in the spent fuel pools without the missile shields installed over the spent fuel pools. This will eliminate the potential for a rack to impact stored fuel if it were dropped.
Because only activities currently performed at CR-3 are affected, i.e., the same types of activities will be performed with the increased onsite fuel assembly storage capacity and revised configuration for fresh fuel storage, the LAR does not create the possibility of any new or different kind of accident from any previously evaluated.
)
U.S. Nuclear Regulatory Comrnission Attachment A 3F0999-07 Pnge 13 of 15 (3)
Involve a significant reduction in a margin of safety?
The CR-3 Improved Technical Specifications (ITS) specifies required margin to criticality (suberiticality margins) for the spent fuel storage racks when fully loaded with spent fuel. This margin is having the effective neutron multiplication factor, Ln, 1
of the spent fuel storage racks maintained less than or equal to 0.95 when flooded with unborated water. The LAR proposes no change to this margin.
The new racks have been analyzed to demonstrate that this required margin is satisfied when fully loaded with fuel enriched to the maximum enrichment allowed by the CR-3 license. Maintaining this margin is assured by remaining within the limits on initial enrichment and fuel burnup that are specified in the ITS.
These limits must be complied with before the fuel can be stored in the spent fuel pool. The LAR proposes revised limits on fuel burnup (no change to fuel enrichment is proposed) to ensure that the existing suberiticality margins are not reduced.
The current CR-3 licensing basis, as reflected by the Final Safety Analysis Report (FSAR), allows the use of administrative controls, e.g., curves of initial fuel assembly enrichment versus burnup, as a means of preventing criticality in the spent fuel pools.
The use of these curves would be continued under this proposed amendment. The changes to these curves proposed by this LAR consist of revising the values of burnup and adding notes to restrict loading of certain fuel assemblies to specific configurations.
These curves have been included in the CR-3 operating license and their use implemented by site procedures since initial issue of the license. From this previous use CR-3 personnel are familiar with the practice of using administrative controls as curves of fuel assembly enrichment versus burnup for placing fuel assemblies in the spent fuel pool in order to prevent criticality.
A mis-loaded fuel assembly was analyzed. The analysis demonstrated that misloading of one assembly does not result in exceeding the criticality margin regulatory limit of Ln = 0.95. This analysis assumed no neutron poison, i.e. soluble boron, in the spent fuel pool water. This is a conservatism since the license requires a minimum of 1925 ppm boron. (Typically the fuel pool water contains approximately 2000 ppm boron.)
ENVIRONMENTAL IMPACT EVALUATION FPC reviewed the CR-3 Final Environmental Statement (FES), dated May 1973, and considered whether its assumptions and conclusions remain valid with the increased spent fuel storage. As explained in Appendix D, Section 6.0, Radiological Evaluation, no significant increase in the annual generation of solid and gaseous radioactive waste are expected as a result of the increased spent fuel capacity. Also, no significant increases in the radioactivity levels in the spent fuel pool water nor in the frequency of changing the spent fuel pool
l U.S. Nuclear Regulatory Commission Attachment A 3F0999-07 Page 14 of 15 demineralizer resin and filter media are expected. The FES concluded that routine operation of CR-3 is expected to contribute a negligibly small incremental dose to that which area residents would receive from natural background. Based on the above, FPC expects this conclusion of the FES to remain valid.
FPC evaluated the non-radiological impacts on the environment of increasing the number of spent fuel assemblies stored onsite. This evaluation considered numerous aspects including, but not limited to, generation or modification of non-radioactive liquid, gaseous and solid effluents, potential for spills of various liquids from tanks, (e.g., transformer oils, fuel, chemicals, hazardous wastes). This evaluation concluded that these effluents and activities would not be impacted by the increased capacity.
The evaluation also considered the heat load to the environment due to this increased capacity.
The heat load in the spent fuel pool is primarily from those assemblies most recently discharged from the reactor core. Generally, the additional fuel assemblies to be stored in the pools as a result of the increased capacity will have been discharged from the reactor a considerable time ago. As a result, the decay heat of these assemblies will be low and the heat added to the pool from them is a small fraction of the heat from the assemblies most recently discharged from the reactor core. No change to the number of assemblies to be transferred to the spent fuel pool from a core reload nor to the frequency of such transfers are proposed.
Thus, any additional heat load to the environment from this increased capacity is not significant. In addition, the heat load from the spent fuel pool is insignificant when compared to the normal heat discharged to the environment as a result of secondary steam condensation during routine plant power operation.
No changes to the heat load from normal power operation are proposed. This evaluation concluded that no significant environmental impacts result from this capacity increase.
The NRC Staff document "OT Position for Review and Acceptance of Spent Fuel Storage and IIandling Applications" dated April 14, 1978, indicates that an environmental impact statement or environmental impact appraisal shall be prepared in which five specific factors should be applied, balanced, and weighed. FPC has addressed the following five factors in determining the environmentsl impact of LAR #239:
(1) the specific needs that require increased storage capacity; (2) the total construction [ costs]
associated with the proposed modification; (3) alternatives to increasing spent fuel storage capacity; (4) whether commitment of material resources would significantly foreclose alternatives for other licensing actions; and (5) whether or not there will be any significant increase in the amount of heat released to the environment. A detailed discussion of how these factors are satisfied is presented in Attachment D.
FPC reviewed the " Final Generic Environmental Impact Statement on Handling and Storage of Spent Light Water Power Reactor Fuel", (FGEIS) dated August 1979 (NUREG-0575), to ensure that its assumptions and conditions are consistent with this proposed license amendment
U.S. Nuclear Regulatory Commission Attachment A 3F0999-07 Page 15 of15 request (LAR #239).. A finding of the FGEIS was that the alternative of increasing the capacity of individual spent fuel storage pools is environmentally acceptable. This review performed by FPC concluded that LAR #239 is consistent with the basis of the FGEIS. Based on that review, FPC concludes that the above FGEIS finding is applicable to LAR #239.
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