ML070540522

License Amendment, Revised Section 4.3, Fuel Storage, of the Technical Specifications to Allow for Contingent Installation of a Temporary Spent Fuel Storage Rack

ML070540522
Person / Time
Site:	Monticello
Issue date:	03/09/2007
From:	Tam P NRC/NRR/ADRO/DORL/LPLIII-1
To:	Conway J Nuclear Management Co
Shared Package
ML070710005	List: ML070540522 ML070710375
References
TAC MD0302
Download: ML070540522 (22)

Text

March 9, 2007 Mr. John T. Conway Site Vice President Monticello Nuclear Generating Plant Nuclear Management Company, LLC 2807 West County Road 75 Monticello, MN 55362-9637

SUBJECT:

MONTICELLO NUCLEAR GENERATING PLANT - ISSUANCE OF AMENDMENT RE: CONTINGENT INSTALLATION OF A TEMPORARY SPENT FUEL STORAGE RACK (TAC NO. MD0302)

Dear Mr. Conway:

The Commission has issued the enclosed Amendment No. 150 to Renewed Facility Operating License No. DPR-22 for the Monticello Nuclear Generating Plant (MNGP). The amendment consists of changes to the Technical Specifications in response to your application dated March 7, 2006, as supplemented by letters dated May 30, September 7, December 15, 2006, and January 2, 2007.

The amendment revises Section 4.3, "Fuel Storage," of the MNGP Technical Specifications to allow for installation of an additional temporary 8 x 8 (64-cell) high-density spent fuel storage rack in the spent fuel pool to maintain full core off-load capability. We understand that such contingent installation is likely needed from January 2007 to approximately the fall of 2008, when the MNGP Independent Spent Fuel Storage Installation is expected to be operational.

A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

/RA/

Peter S. Tam, Senior Project Manager Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-263

Enclosures:

1. Amendment No. 150 to DPR-22

2. Safety Evaluation cc w/encls: See next page

Mr. John T. Conway March 9, 2007 Site Vice President Monticello Nuclear Generating Plant Nuclear Management Company, LLC 2807 West County Road 75 Monticello, MN 55362-9637

SUBJECT:

MONTICELLO NUCLEAR GENERATING PLANT - ISSUANCE OF AMENDMENT RE: CONTINGENT INSTALLATION OF A TEMPORARY SPENT FUEL STORAGE RACK (TAC NO. MD0302)

Dear Mr. Conway:

The Commission has issued the enclosed Amendment No. 150 to Renewed Facility Operating License No. DPR-22 for the Monticello Nuclear Generating Plant (MNGP). The amendment consists of changes to the Technical Specifications in response to your application dated March 7, 2006, as supplemented by letters dated May 30, September 7, December 15, 2006, and January 2, 2007.

The amendment revises Section 4.3, "Fuel Storage," of the MNGP Technical Specifications to allow for installation of an additional temporary 8 x 8 (64-cell) high-density spent fuel storage rack in the spent fuel pool to maintain full core off-load capability. We understand that such contingent installation is likely needed from January 2007 to approximately the fall of 2008, when the MNGP Independent Spent Fuel Storage Installation is expected to be operational.

A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

/RA/

Peter S. Tam, Senior Project Manager Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-263

Enclosures:

1. Amendment No. 150 to DPR-22

2. Safety Evaluation cc w/encls: See next page DISTRIBUTION PUBLIC LPL3-1 r/f RidsNrrDorlLpl3-1 RidsNrrPMPTam RidsNrrLATHarris RidsOGCRp RidsAcrsAcnwMailCenter RidsNrrDirsltsb G. Hill, OIS RidsRgn3MailCenter A. Stubbs W. Wang J. McGuire Y. Orechwa RidsNrrDorlDpr Package Accession No.: ML070710005 Amendment Accession No.: ML070540522 Tech. Spec. pages Accession No.: ML07010375 OFFICE NRR/LPL3-1/PM NRR/LPL3-1/LA NRR/AADB/BC NRR/SBPB/BC NRR/SNPB/BC NRR/EGCA/BC OGC NRR/LPL3-1/BC NAME PTam THarris MKotzalas* JSegala* FAkstulewicz* KManoly KWinsberg LRaghavan DATE 3/1/07 3/1/07 6/6/06* 1/29/07* 10/25/06* 3/9/07 3/7/07 03/9/07

• Safety evaluation transmitted by memo dated as shown. .

OFFICIAL RECORD COPY

NUCLEAR MANAGEMENT COMPANY, LLC DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 150 License No. DPR-22

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Nuclear Management Company, LLC (the licensee), dated March 7, 2006, as supplemented by letters dated May 30, September 7, December 15, 2006, and January 2, 2007, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.2 of Renewed Facility Operating License No. DPR-22 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 150, are hereby incorporated in the license. NMC shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

L. Raghavan, Chief Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: March 9, 2007

ATTACHMENT TO LICENSE AMENDMENT NO. 150 RENEWED FACILITY OPERATING LICENSE NO. DPR-22 DOCKET NO. 50-263 Replace the following page of Renewed Facility Operating License DPR-22 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

REMOVE INSERT 3 3 Replace the following page of Appendix A, Technical Specifications, with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

REMOVE INSERT 4.0-2 4.0-2

2. Pursuant to the Act and 10 CFR Part 70, NMC to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operations, as described in the Final Safety Analysis Report, as supplemented and amended, and the licensees filings dated August 16, 1974 (those portions dealing with handling of reactor fuel) and August 17, 1977 (those portions dealing with fuel assembly storage capacity);

3. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NMC to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required;

4. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NMC to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and

5. Pursuant to the Act and 10 CFR Parts 30 and 70, NMC to possess, but not separate, such byproduct and special nuclear material as may be produced by operation of the facility.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission, now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

1. Maximum Power Level NMC is authorized to operate the facility at steady state reactor core power levels not in excess of 1775 megawatts (thermal).

2. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 150, are hereby incorporated in the license. NMC shall l operate the facility in accordance with the Technical Specifications.

3. Physical Protection NMC shall implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Renewed License No. DPR-22 Amendment No. 1 thru 150

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 150 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-22 NUCLEAR MANAGEMENT COMPANY, LLC MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263

1.0 INTRODUCTION

By application dated March 7, 2006 (Accession No. ML060720212), as supplemented by letters dated, May 30 (Accession No. ML061520416), September 7 (Accession No. ML062560369),

December 15, 2006 (Accession No. ML063610073), January 2, 2007 (Accession No. ML070030075), and January 29, 2007 (Accession No. ML070290325), the Nuclear Management Company, LLC (the licensee), requested changes to the Technical Specifications (TSs) for the Monticello Nuclear Generating Plant (MNGP). The proposed amendment would revise Section 4.3, "Fuel Storage," of the MNGP TSs to allow for installation of an additional temporary 8 x 8 (64-cell) high-density spent fuel storage rack in the spent fuel pool to maintain full core off-load capability.

The licencee's supplements cited above provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register on April 3, 2006 (71 FR 16599).

2.0 REGULATORY EVALUATION

The licensee provided a list of applicable regulatory requirements in Section 6.2 of Enclosure 1, of its March 7, 2006 application. The licensee also stated that MNGP was constructed before the General Design Criteria (GDC) of Title 10 of the Code of Federal Regulations (10 CFR)

Part 50, and Standard Review Plan (SRP) were promulgated. However, the concepts, guidance, and criteria expressed by the GDC(s) that the licensee listed in Section 6.2 of the enclosure are generally applicable.

2.1 Regulatory Evaluation Regarding Criticality While MNGP was licensed prior to the issuance of the GDC of 10 CFR Part 50, it was designed and constructed to comply with Northern States Power Companys understanding of the intent of the Atomic Energy Commission (AEC) proposed 70 GDCs for Nuclear Power Plant Construction Permits issued by the AEC for public comment in July 1967. MNGP complies with AEC GDC 66, Prevention of Fuel Storage Criticality, that states:

Criticality in new and spent [fuel] storage shall be prevented by physical systems or processes. Such means as geometrically safe configurations shall be emphasized over procedural controls.

AEC GDC 66 is nearly identical to GDC 62 of 10 CFR Part 50 in wording and intent. The latter states:

Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.

Thus, conformance with either of these GDCs meets the regulatory intent of minimizing the potential for a criticality event.

The current Nuclear Regulatory Commission (NRC) regulatory requirements for maintaining subcriticality in spent fuel pools (SFPs) are given by 10 CFR 50.68, Criticality Accident Requirements, and 10 CFR 70.24, Criticality Accident Requirements. The former sets forth requirements to prevent a criticality event, while the latter set forth requirements for a monitoring system capable of detecting a criticality. MNGP does not have an exemption to the requirements of 10 CFR 70.24. However, Section 1.3.6, Plant Fuel Storage and Handling Systems, of the Updated Safety analysis Report (USAR), states that for the MNGP, the licensee has elected to comply with the criticality accident requirements of 10 CFR 50.68 in lieu of the 10 CFR 70.24 requirements for the handling and temporary storage of new fuel and non-fuel special nuclear material.

2.2 Regulatory Evaluation Regarding Radiological Doses The NRC staff reviewed the potential impact of the proposed changes on previously analyzed design-basis accident (DBA) radiological consequences, and the acceptability of the revised analysis results. The regulatory requirements for which the NRC staff based its acceptance are the accident dose criteria in 10 CFR Part 50.67, as supplemented in Regulatory Position 4.4 of Regulatory Guide 1.183 (RG 1.183), Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, and 10 CFR Part 50 Appendix A, GDC-19, Control Room. Additionally, the NRC staff used the guidance in Section 15.0.1, Radiological Consequence Analysis Using Alternative Source Terms, of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants in performing this review. The NRC staff also considered relevant information in the MNGP USAR.

The NRC staff also considered relevant information from the safety evaluation associated with Amendment No. 145, dated April 26, 2006, which describes the most recent NRC staff review of the licensees dose analysis of the fuel handling accident at MNGP applying an alternate source term. Additionally, the NRC staff referenced Amendment No. 34, dated April 14, 1978, which increased the capacity of MNGP SFP from 740 fuel assemblies to 2237 fuel assemblies.

2.3 Regulatory Evaluation Regarding Thermal-Hydraulic Issues GDC 61 specifies, in part, that fuel storage systems shall be designed with residual heat removal capability having reliability and testability that reflects the importance to safety of decay

heat removal, and with the capability to prevent significant reduction in fuel storage coolant inventory under accident conditions.

SRP Section 9.1.3 provides criteria for design and performance of the SFP cooling and cleanup (SFPCC) system. It provides the acceptance criteria for the SFPCC system and make-up water system. An acceptable design as stated in SRP 9.1.3 is one in which the integrated design is in accordance with the specified requirements in GDC 2, 4, 5, 44, 45, 61, 63, and 10 CFR Part 20. to Matrix 5 of Section 3.1 of NRC Review Standard RS-001, Revision 0, provides the NRC staff review guidance used to determine the adequacy of SFP cooling capability. It supersedes the guidance of paragraphs III.1.d. and III.1.h. of SRP Section 9.1.3. Section 3.1 of RS-001 states that the licensee demonstrates adequate SFP cooling capacity by either performing a bounding evaluation, or committing to a method of performing outage-specific evaluations. The analysis conditions to be assumed for bounding and cycle-specific analysis are given in Section 3.1.1 of RS-001. Section 3.2 of RS-001 provides guidance regarding requirements for adequate makeup supply.

The SFP cooling system is described in Chapter 10 of the MNGP USAR. Section 10.2.2 provides design basis information for the SFP cooling and demineralizer system, including the SFP peak fuel pool temperatures for both normal (planned) and abnormal (emergency) refueling scenarios. The system description along with the applicable design basis information included in Chapter 10 provides the criteria needed to evaluate the impact that the increased SFP heat load has on the ability of the SFP system to comply with the plant design basis and GDC 61. In meeting the GDC, the licensee should demonstrate that sufficient spent fuel pool cooling capacity and make-up sources are available during refueling, and time is available prior to pool boiling to supply makeup water following a loss of forced cooling.

2.4 Regulatory Evaluation Regarding Crane and Heavy Loads NUREG-0612, Control of Heavy Loads, provides guidelines and recommendations to assure safe handling of heavy loads by prohibiting, to the extent practicable, heavy load travel over stored spent fuel assemblies, fuel in reactor core, safety-related equipment, and equipment needed for decay heat removal.

NUREG-0612 and NUREG-0554, Single Failure Proof Cranes for Nuclear Power Plants, provided the basis for review of the licensee-proposed handling of heavy loads during the SFP expansion. NUREG-0612 endorses a defense-in depth approach for handling of heavy loads near spent fuel and safe shutdown systems. General guidelines for overhead handling systems that are used to handle heavy loads in the area of the reactor vessel and SFP are given in Section 5.1.1 of NUREG-0612. They are as follows: (1) definition of safe load paths; (2) development of procedures for load handling operations; (3) training and qualification of crane operators in accordance with Chapter 2-3 of American National Standards Institute (ANSI)

B30.2 -1976; (4) use of special lifting devices that meet guidelines in ANSI N14.6 -1978; (5) installation and use of non-custom lifting devices in accordance with ANSI B30.9 - 1971; (6) inspection, testing, and maintenance of cranes in accordance with Chapter 2-2 of ANSI B30.2 - 1976; and (7) design of crane in accordance with Chapter 2-1 of ANSI B30.2- 1976 and CMMA-70.

NUREG-0554 identifies features of design, fabrication, installation, inspection, testing, and operations of single-failure-proof overhead crane handling systems used for handling heavy

loads. It recommends special single-failure-proof features in the crane hoisting machinery and reeving. Single-failure-proof means the cranes hoisting and breaking systems will safely retain the load following any single failure in them.

2.5 Regulatory Evaluation Regarding Structural Issues SRP Section 3.7.1, Seismic Design Parameters, provides guidance for defining seismic analysis input data, in accordance with the acceptance criteria in GDC 2 and 10 CFR Part 100, Appendix A. Specific criteria are set forth in RG 1.60, Design Response Spectra for Seismic Design of Nuclear Power Plants, RG 1.61, Damping Values for Seismic Design of Nuclear Power Plants, and RG 1.92, Combining Model Responses and Spatial Components in Seismic Response Analysis. Design of category I structures such as spent fuel pools is specified in SRP Section 3.8.4, Other Seismic Category I Structures.

3.0 TECHNICAL EVALUATION

The licensee proposed to revise the MNGP licensing basis to allow for installation of an additional temporary 8 x 8 (64 cell) high-density spent fuel storage rack, manufactured by the Programmed and Remote Systems Corporation (PaR), in the SFP to maintain full core off-load capability. Approval of the license amendment will permit the licensee to temporarily increase the current SFP capacity from 2237 to 2301 fuel assemblies.

Prior to the spring 2007 refueling, the licensee began staging replacement fuel assemblies.

The staging of the new assemblies created a shortfall of storage location, and without the addition of new storage space, the SFP will not contain a sufficient amount of storage locations to accommodate a full core off-load if needed. The MNGP Independent Spent Fuel Storage installation (ISFSI) is expected to be operational by fall 2008, and movement, of spent fuel from the SFP to the ISFSI cask storage location, after the ISFSI is operational, will ensure that long-term full core off-load capability is maintained. The licensee stated that contingent installation of the temporary spent fuel rack is required from January 2007, to approximately the fall of 2008, and that prior to transfer of spent fuel assemblies to the ISFSI (assuming the new rack is installed), the new rack will be emptied, and removed from the SFP, and subsequently an application for amendment submitted to return the SFP capacity to the present limit of 2237 fuel assemblies.

3.1 Thermal-Hydraulic Issues 3.1.1 SFP Cooling The normal fuel pool cooling and cleanup (FPCC) system provides the primary means for forced cooling of the SFP. The Residual Heat Removal (RHR) system in the fuel pool cooling mode is also capable of providing forced cooling of the SFP. If the SFP heat load is greater than the heat removal capacity of the FPCC system heat exchangers, one subsystem of the safety-related RHR system is designed, and procedures are in place, to direct RHR operation in the fuel pool cooling mode to either supplement the FPCC system, or provide cooling to the SFP. Operation of one RHR subsystem in the fuel pool cooling mode is capable of providing sufficient SFP cooling for all heat loads up to and including an emergency full core offload (following installation of the PaR 64 cell fuel storage rack)

The FPCC, which is the normal spent fuel pool cooling system, consist of two parallel pumps and heat exchangers with associated piping instrumentation and controls, thus providing the

SFP with two redundant cooling subsystems. Heat is rejected from the FPCC to the Reactor Building Closed Cooling Water (RBCCW) system, which, in turn, is rejected to the Service Water system. The FPCC has a design capacity of 5.7 x 106 BTU/hr.

The RHR system is capable of being aligned to cool the SFP. The major equipment of the RHR system consist of two RHR heat exchangers, four main injection RHR pumps, and associated piping, valves and instrumentation. Each RHR pump is designed to deliver at least 4,000 gpm . The RHR system is cooled by the Residual Heat Removal Service Water (RHRSW) system. Using a process fluid maximum temperature of 140 EF and a maximum cooling water temperature of 90 EF, the maximum cooling capability of the RHR system in the fuel pool cooling mode is approximately 26.4 MBTU/hr.

3.1.2 Fuel Pool Heat Up Analysis The proposed increase in the storage capacity of the SFP will result in an increase in the number of spent fuel assemblies stored in the SFP for the bounding analysis cases where the refueling offload results in all the available storage spaces filled. The increase in the number of fuel rods stored in the fuel pool will result in an increase in the total decay heat associated with the stored fuel and a corresponding increase in the FPCC cooling loop heat load. The licensee performed thermal-hydraulic analyses to evaluate the effect of the increased storage capacity on the FPCC heat loads and the corresponding FPCC water temperature. Discharge scenarios were considered for both partial and full-core discharges to the SFP.

3.1.2.1 Partial Core Offload (Normal Refueling)

The temporary PaR 8 x 8 high-density spent fuel rack will be installed just prior to Refueling Outage 23 (RFO-23) or during Cycle 24 if needed to perform a full-core offload. The licensee analyzed a scenario that conservatively bounds the actual number of irradiated fuel assemblies that are expected to be stored in the SFP with the addition of the temporary PaR 8 x8 high-density storage rack installed. The licensees evaluation included as part of the overall heat load, the base decay heat load from the fuel that has been discharged prior to Cycle 23, based on the actual fuel assembly burnup and operating power. The decay heat for the current SFP configuration and the decay heat of the future discharges were combined to provide a total decay heat load on the SFP cooling system. Since operating cycle length increased from 18 to 24 months beginning with Cycle 23, the assumed discharge for the partial core offload beginning with Cycle 23 increased from 141 to 152 assemblies.

The FPCC system consists of two redundant cooling subsystems. Under normal operation only, one FPCC subsystem is required to be in operation. The loss of flow from one FPCC pump, results in the greatest reduction in system heat removal. The licensee performed a SFP cooling analysis and found that for the maximum normal cooling configuration, operation of only one of the two circulating pumps is required to maintain FPCC temperature to less than or equal to 140E F when subjected to a maximum normal heat load of 5.60 MBtu/hr.

In the December 15, 2006, supplement, the licensee indicated that in the event of insufficient cooling capability by the non-safety related FPCC system, one subsystem of the safety-related RHR system can be placed in service in the fuel pool cooling mode to either supplement or replace the FPCC system for cooling of the SFP. The licensee also stated that operation in this mode is sufficient to provide SFP cooling, for all heat loads up to and including an emergency full-core offload (i.e., with the PaR 64-cell rack included), and that in the event offsite power is unavailable, power to safety-related RHR pumps meets single failure criteria.

In evaluating the maximum SFP bulk water temperature, the licensees analysis was based on a cooling configuration that credits operation of only one of the two FPCC cooling pump and two FPCC heat exchangers. While the maximum decay heat load increased slightly due to additional fuel storage in the pool, the licensee found that it did not add significantly to the cooling requirements of the existing configuration, and that the present heat removal system would have adequate capacity to continue to maintain the SFP temperature within the current 140E F temperature limit, provided the fuel discharge is not completed prior to 216 hours9 days <br />1.286 weeks <br />0.296 months <br /> post-shutdown.

The licensee performed a cycle-specific offload analysis to confirm the applicability of the bounding analysis assumptions, and will use administrative controls (e.g. controlling the number of assemblies transferred over a time period) to ensure that the number of fuel assemblies moved to the SFP does not result in SFP heat loads greater than 5.60 MBtu/hr.

The peak SFP bulk temperature for this case (SFP normal heat load) continues to satisfy the licensing-basis SFP temperature of 140E F, for normal planned refueling offload given in Section 10.2.2 of the MNGP USAR. Accordingly, the NRC staff finds the analysis results acceptable.

3.1.2.2 Full Core Offload The licensee identified two full-core offload scenarios for which evaluations were performed.

The first was a normal (non-emergency) full-core offload, and the second, an emergency full-core offload after 150 hours6.25 days <br />0.893 weeks <br />0.205 months <br />. The licensees thermal-hydraulic analysis shows that the heat loads anticipated for the SFP can be adequately cooled by either the normal fuel pool cooling system or by the RHR system in the fuel pool cooling mode connected to the SFP. The licensee provided the following results for two scenarios:

5.55 MBtu/hour for normal discharge, 216 hours9 days <br />1.286 weeks <br />0.296 months <br /> after shutdown (FPCC system) 24.71 MBtu/hour for emergency full-core offload, 150 hours6.25 days <br />0.893 weeks <br />0.205 months <br /> after shutdown (RHR system).

Thus, the USAR Section 10.2.2.3 heat load of 5.60 MBtu/hr will be met at 216 hours9 days <br />1.286 weeks <br />0.296 months <br /> after shutdown. Similarly, the projected decay heat due to an emergency full-core offload is less than the worst-case heat removal capability of the RHR system of approximately 26.4 MBtu/hr with 90o F Mississippi River water.

Thermal-hydraulic analysis also shows that with the temporary PaR 8X8 high-density spent fuel storage rack installed, the SFP peak bulk temperature remains within the current licensing basis maximum of 140o F for all offload scenarios.

The installation of the temporary PaR 8 X 8 high-density spent fuel storage rack, therefore, does not add significant cooling requirements to the existing configuration. The present heat removal systems have adequate capacity to maintain the SFP temperature within the current temperature limits specified for MNGP in Section 10.2.2.1 of the MNGP USAR, which states that during refueling outages, full core offloads are allowed because heatloads are explicitly calculated and compared to cooling capabilities prior to any fuel movement that would increase the spent fuel pool heat load. Since the full core offload would occur during a planned refueling, and therefore is a planned offload, the NRC staff requested additional information on the analysis performed by the licensee for this scenario. Specifically, the NRC staff wanted to

know how the analysis was performed and what equipment was credited. In its December 15, 2006, response, the licensee stated that the cooling analyses for the non-emergency full-core offload case conservatively assumed the FPCC system was not operational and that the entire coolant inventory was supplied by one RHR subsystem. The maximum heat load for this case was calculated to be 24.0 MBtu/hr compared to the maximum SFP cooling capability for this configuration of 26.4 MBtu/hr. In summary, the licensees analyses showed that the heat removal capability of one RHR subsystem is sufficient to remove the maximum anticipated heat load for this case.

3.1.2.3 Effects of SFP Boiling In the unlikely event there is a complete loss of cooling, the SFP bulk water temperature will begin to rise and will eventually reach the boiling temperature.

The licensee has performed analyses to demonstrate minimum time-to-boil and the maximum boil-off rate. The licensee indicated a minimum time-to-boil of 10.3 hours0.125 days <br />0.0179 weeks <br />0.00411 months <br />, and a maximum boil-off rate of 43 gpm. The NRC staff noticed that the time-to-boil and maximum boil-off rate that the licensee quoted was based on an initial pool temperature of 120E F, and a heat load of 20.0 MBtu/hr. Since the pool temperature limit is 140E F, the NRC staff requested the corresponding information for the case where a 140E F initial pool temperature is assumed. In its December 15, 2006, response the license indicated that the time-to-boil for this case to be 8.3 hours0.125 days <br />0.0179 weeks <br />0.00411 months <br />, and the corresponding boil-off rate to be 53 gpm.

In summary, the licensee has adequate time to align and supply sufficient water from a variety of sources to the SFP prior to the time-to-boil. Ample makeup can be provided by various means, e.g., through use of the filter/demineralizer backup connections (100 gpm), and RHRSW system connection (1000 gpm). Based on the above, the NRC staff finds the time-to-boil analysis acceptable.

3.2 Handling of Heavy Loads NUREG-0612 provides recommendations and guidelines to assure safe handling of heavy loads in proximity to or over safe shutdown equipment or irradiated fuel in the spent fuel area.

The guidelines are meant to ensure that either (1) the potential for a load drop is extremely small, or (2) the consequences of a postulated accidental load drop do not result in the violation of radiological or criticality limits, or compromise safe shutdown.

The licensee stated that the main hoist of the reactor building crane will be used for handling operations involving installation/removal of the temporary high-density fuel storage rack. The main hoist of the reactor building overhead crane has a rated capacity of 85-tons and is single-failure proof. The maximum lift weight during installation/removal of the high-density fuel rack module is approximately 12,500 lbs. Therefore, the weight being handle during the installation of the rack is well within the capacity of the reactor building crane.

NUREG-0612, Section 5.1.1 recommends that (1) safe load paths be developed, (2) procedure be developed for load handling, (3) crane operators be trained and qualified, (4) special lifting devises used meet criteria of ANSI N14.6, (5) slings used are in compliance with the requirements specified in NUREG-0612, (6) the reactor building crane be inspected, tested, and maintained in accordance with the requirements of NUREG-0612, and (7) the crane be designed to meet the applicable criteria and guidelines of Chapter 2-1 of ANSI B30.2-1976.

The licensee discussed how the safe handling of heavy loads by the reactor building crane will

be ensured by following the defense-in-depth approach guidelines of NUREG-0612. The licensee stated the following:

(1) Safe load paths for the Reactor Building overhead crane are procedurally defined and reviewed by the Plant Operations Review Committee (PORC). Movement of the temporary PaR high-density spent fuel storage rack module will conform to the specified safe load path requirements. The temporary spent fuel storage rack module will not, during installation, be suspended over any portion of the SFP containing spent fuel assemblies.

(2) Although existing MNGP procedures covering the handling of heavy loads are adequate for handling of the temporary PaR 8 x 8 high-density spent fuel storage rack module, these procedures will be augmented as necessary to emphasize this specific task as part of the licensees modification process to be used for this temporary installation.

These procedures will be comprehensive with respect to load handling, exclusion areas, equipment required, inspection and acceptance criteria before load movement, and steps/sequence to be followed during load movement, as well as defining safe load paths and special precautions.

(3) Crane operators are trained and qualified for the tasks they perform, and perform their duties in accordance with procedures that are in compliance with ANSI B30.2-1976.

Any special training required will be identified and implemented as part of the NMC modification process.

(4) The PaR fuel storage rack lifting rig is similar to the rigs used in the initial SFP rack installation and subsequent re-racking in 1977 for high-density fuel storage racks at MNGP. Any special lifting devices necessary for this installation will meet the criteria of ANSI N14.6.

(5) Slings used in conjunction with the reactor building overhead crane comply with the requirements of ANSI B30.9-1971. Slings have a minimum safety factor of 5. Slings are not derated for dynamic loading since these loads are a small percentage of the overall static load and can be disregarded.

(6) Inspection, testing, and maintenance of the reactor building overhead crane complies with ANSI B30.2-1976.

(7) The reactor building overhead crane was manufactured prior to issuance of CMAA-70 and ANSI B30.2, and was designed to EOCI 61, "Specification for Electric Overhead Traveling Cranes," a precursor to CMAA-70. The NRC has previously concluded that since the Reactor Building overhead crane met the criteria for a single-failure proof crane (i.e., met the applicable provisions of draft Regulatory Guide 1.104), and therefore, it had met the applicable NUREG-0612 guidelines.

The approach above described by the licensee fully satisfies the criteria of Section 5.1.1 of NUREG-0612 and is acceptable. Also since the Reactor Building crane to be used is single-failure-proof, the NRC staff finds the use of the crane for rack installation to be acceptable.

Based on review of the licensees submitted information concerning the handling of heavy loads associated with this amendment request, the NRC staff finds that the licensee has provided adequate assurance that its planned actions for the handling of heavy loads for the installation

of the storage racks are consistent with the defense-in-depth approach to safety described in NUREG-0612. The NRC staff believes that the use of the reactor building crane will enable the licensee to maintain safety during the handling of heavy loads associated with the installation of the PaR 8 x 8 high-density spent fuel storage rack module.

3.3 Seismic and Structural Design Issues The licensee provided most of the safety analysis information from a 1977 PaR report for the structural design and analyses for other high-density spend fuel storage racks, specifically 8x10, 8x11, 9x12, and 10x11 modules which the licensee indicated bounded the 8x8 module design. The structural safety and seismic response analyses for those racks were performed in 1977. The input response spectrum used in seismic response analyses was an artificial time history response spectrum that envelops the Iowa site response spectrum for the Duane Arnold Energy Center (DAEC) who procured the racks. NMC has arranged with DAEC to use an 8x8 rack that was not installed in the DAEC SFP to be installed at the MNGP in the event of a full core off-load. The NRC approved the use of these PaR racks at DAEC in 1978 (DAEC Amendment No. 45, July 7, 1978). Although some related regulations have changed since then, considering that there is no significant safety concern that requires reevaluation of the approved fuel racks, the proposed new 8x8 fuel rack module configuration will be acceptable if all design parameters and loading (static, dynamic, seismic and other loads, as well as the loading conditions) responses of the 8x8 rack are bounded by that of approved fuel rack module configurations, or within the acceptable criteria.

3.3.1 Structure of the 8x8 Spent Fuel Rack The structure of the 8x8 spent fuel rack is similar to that of the approved spent fuel racks. The licensee stated that (Section 3.0 of Enclosure 1, September 7, 2006, letter) it is constructed of bolted anodized aluminum with a boral neutron absorber in an aluminum matrix core clad with 1100 series aluminum at alternating cell locations. The module consists of an 8x8 array of tubes. The rack is approximately 4.5 feet-square by 14 feet high. Nominal fuel element center-to-center spacing is 6.625 inches. The previously analyzed fuel rack modules were constructed by the same manufacturer, PaR, with the same material and design. All fuel racks have the same height of 14 feet, but the center-to-center spacing is 7.0 inches for 8x12 and 9x12 models, and 6.625 inches for 8x11 and 10x11 models - the same as the 8x8 model.

Based on the information provided by the licensee, the structure of the proposed temporary spent fuel rack is similar to that of previously approved spent fuel rack models at DAEC.

3.3.2 Dynamic Properties of the 8x8 Spent Fuel Rack In its January 29, 2007, letter, the licensee provided the calculation of the natural frequencies of the 8x8 spent fuel rack to determine its dynamic properties as a seismic response qualification check. In Enclosure 3 of its March 7, 2006, application, the licensee stated that two models were used in the fuel rack seismic analysis in 1977 - a detailed SAP IV finite element model and a simplified ANSYS model. The simplified model was first verified by comparing its calculation results with that of the detailed model for the 10x10 rack module. Since only small differences were observed, and because of limitations of computer capability at that time, the simplified ANSYS model was then used to analyze all other rack designs.

The new calculation for the 8x8 fuel rack design was conducted by using the same simplified ANSYS model and applying the same procedure as that used in the 1977 analysis except: (1)

a recent version of the SAP software, SAP 2000, was used; (2) the 1977 simplified model defined that the bottom element of the rack was at the same elevation at the bottom casting, while the actual model used has to assign a 1-inch rigid body connection between them because the SAP 2000 software does not allow two nodes to be assigned to the same location; and (3) the model nodes were equally spaced so that the weight was more uniformly distributed.

The new analysis first calculated natural frequencies for the 10x10 design to check the accuracy of the model used in SAP 2000. After obtaining good agreement between the SAP 2000 simplified model output and the detailed SAP IV model output, analyses were carried out for the 10x11 and 8x11 designs to further confirm the calculation accuracy, and finally the 8x8 design. The comparison of the analysis results was summarized in the licensees January 29, 2007, submittal. The analysis results showed that:

! The fundamental horizontal and vertical frequencies of the simplified and the detailed models differ by no more than 10%.

! The fundamental horizontal frequency is 9 Hz and the vertical frequency is 23 Hz for the 8x8 fuel rack design, which are higher than the lowest fundamental frequencies for all previously analyzed fuel rack designs: 8 Hz for horizontal frequency and 13.2 Hz for vertical frequency.

The NRC staff checked the input parameters for the fuel rack designs analyzed and found them to be reasonable and in accordance with the current engineering practice.

3.3.3 Seismic Response of the 8x8 Fuel Rack Design Although no seismic analysis was performed for the 8x8 spent fuel rack configuration, and the response spectrum of the MNGP site is different from that of the Iowa site, the licensee stated that the original qualification report (System Report on the High-Density Rack Design, to the Reference 5.1) provides assurance of the seismic suitability of the 8x8 fuel rack configuration (January 29, 2007, letter). This claim was mainly based on:

! As shown in Figures AA and BB of Enclosure 2 of the licensees September 7, 2006, letter, the artificial time history response spectrum used in the 1977 seismic analyses envelopes the MNGP site response spectrum for frequencies higher than 2.5 Hz for horizontal component, and higher than 5.0 Hz for vertical component.

! The calculated 8x8 fuel rack natural frequencies are 9.0 Hz in the horizontal direction and 23 Hz in the vertical direction - well within the response spectra envelope.

After reviewing the 1977 seismic analysis results, the NRC staff finds that the licensees conclusion is acceptable based on the following considerations:

(1) The 1977 seismic analyses for the spent fuel rack designs listed above were accepted by the NRC in accordance with the regulatory requirements at that time. Since no backfit is required for the analyses, the results are still acceptable.

(2) The structural dynamic analysis showed that the natural frequencies of the 8x8 fuel rack design are bounded by that of previously analyzed fuel rack designs. Even if the results differ by 50%, the analysis results using the simplified model are still within the

frequency ranges on which the time history response spectrum used in the previous seismic analyses envelops the MNGP site response spectrum. Therefore, it is reasonable to conclude that the seismic response of the 8x8 fuel rack is bounded by the previous seismic analysis results.

(3) In the 1977 seismic analysis, the maximum seismic response was obtained by using the square root sum of the squares method to combine 3 orthogonal seismic response components, per guidance of RG 1.92.

3.3.4 Other Structural Safety Related Evaluations The designed SFP floor loading capacity is 2.7 ksf (Section 3.0 (4) of Enclosure 1 of September 7, 2006, letter), and the existing fully loaded MNGP SFP floor loading is 2.1 ksf with 2209 fuel storage cells. If the proposed 8x8 fuel rack is installed, the fuel storage cells will increase to 2273 (64 cells more), equivalent to about 3 percent of loading increase (i.e., the floor loading will reach 2.163 ksf, assuming that the fuel racks are uniformly installed in the SFP). The total SFP floor loading will still be smaller than the design SFP floor capacity (2.7 ksf) with an acceptable factor of safety (FS) of 1.25.

The 1977 seismic analysis showed that the maximum horizontal displacement (sliding) is 1.05 inches without significant rocking or liftoff for the fuel rack designs analyzed. In the letter of September 7, 2006, the licensee stated that (i) the minimum clearance distance from the proposed 8x8 fuel rack to the nearest adjacent object in the SFP at MNGP is 6 inches; and (ii) the coefficient of friction used in the 1977 analysis was 0.2 and the tests conducted at that time provided coefficient of friction values from 0.23 to 0.29. Although the 8x8 fuel rack is smaller than the 8x10 fuel rack (the smallest fuel rack configuration used in the 1977 analysis), the safety margin of the maximum displacement under seismic loading condition is acceptable considering that the minimum clearance distance is about 5 times of the calculated maximum displacement, and conservative input parameters were used in the previous analyses.

3.3.5 Summary of Seismic and Structural Evaluation Based on review of the licensees submitted information, the NRC staff finds that: (1) the structural and dynamic properties of the PaR 8x8 high-density fuel storage rack design are bounded by those of the accepted similar fuel rack designs; (2) the time history response spectrum used in seismic response analyses conducted in 1977 for the accepted fuel storage rack designs envelopes the MNGP site response spectrum over a wide frequency range, and the natural frequencies of the 8x8 fuel rack are well within this range; and (3) the factor of safety for fuel rack pool floor loading capacity and maximum rack displacement for the 8x8 fuel rack are acceptable. On the basis of its findings, the staff concludes that the installation of a temporary PaR 8x8 high-density spent fuel storage rack at MNGP is acceptable from the perspective of structural safety.

3.4 Radiological Assessment The NRC staff reviewed the regulatory and technical analyses, as related to the radiological consequences of DBAs, performed by the licensee in support of the proposed license amendment. The NRC staff reviewed the assumptions, inputs, and methods used by the licensee to assess these impacts.

To support the proposed increase of spent fuel storage capacity, the licensee evaluated the impact of the contingent installation of the temporary storage racks in the SFP on the current DBA dose analyses as discussed in the MNGP USAR. The DBA that is potentially affected by the proposed change to the spent fuel storage capacity is the fuel-handling accident (FHA). By Amendment No. 145, dated April 26, 2006, the NRC staff approved implementation of an alternative source term for the FHA, per the provisions of 10 CFR 50.67. In support of that amendment request, the licensee determined that the radiological consequences of the FHA would be within the offsite and control room dose acceptance criteria specified in SRP 15.0.1 and GDC-19, and within the dose criteria given in 10 CFR 50.67.

With respect to the postulated FHA if the temporary spent fuel storage rack is installed, the licensee did not propose changes to the procedures and equipment used to move fuel; therefore, there would be no change to the projected fuel damage due to a postulated FHA.

Additionally, the licensee did not propose changes to fuel burnup, decay time, or pool water level; therefore, the postulated radiological source term would remain the same. Because adding additional spent fuel storage does not increase the amount of fuel assumed to be damaged in a FHA, and because there are no changes to the source term, the NRC staff finds that the current licensing-basis FHA dose analysis remains applicable after the expansion of the spent fuel storage capacity. The NRC staff finds that the current licensing basis DBA dose analyses remain bounding for the proposed changes to TSs Section 4.3.

The licensee has evaluated the impact of the proposed expanded spent fuel storage on the consequences of a loss of SFP cooling. The licensee determined that in the unlikely event of a total loss of SFP cooling, there would be sufficient time for the operators to provide alternate means of cooling (see Section 3.1 above) to preclude fuel damage and radioactive material release. Therefore, there are no projected radiological consequences due to loss of SFP cooling.

Based on the preceding discussion, the NRC staff finds that the current DBA dose analyses remain bounding for the contingent temporary installation of expanded spent fuel storage capacity in the SFP. The NRC staff finds with reasonable assurance that the licensees estimates of the exclusion area boundary, low-population zone, and control room doses will continue to comply with the applicable regulatory criteria identified in Section 2.0 above.

Therefore, the proposed temporary expansion of the onsite spent fuel storage within the SFP is acceptable with regard to the radiological consequences of postulated DBAs.

The licensee also analyzed the installation and removal of the temporary PaR 8X8 high-density spent fuel storage rack and concluded that these activities have a minimal impact on occupational radiation exposure to operating personnel, and that they will not create significant radiological waste. The NRC staff has no reason to dispute the licensees analyses results on occupational exposure and nuclear waste generation.

3.5 Criticality Safety Analysis The licensee stated that it has evaluated a Holtec Report, "Criticality Safety Evaluation of the Spent Fuel Storage Racks in the Duane Arnold Energy Center for Maximum Enrichment Capability," dated August 1997, and Nuclear Associates International Corporation Report NAI 77-86, Revision 0, "Summary Report Nuclear Criticality Analysis for the Spent Fuel Racks of the Duane Arnold Nuclear Power Plant," dated November 1977. The licensee determined that these reports are applicable for the fuel designs in use at the MNGP. The Holtec criticality

analysis provides the primary basis for the determination that the installation and use of PaR 8X8 high-density fuel storage rack is acceptable from the criticality standpoint.

The Design Features section of the MNGP TSs (Specification 4.3.1.1.a) specifies that the fuel racks in the SFP may only hold fuel assemblies with a maximum k4 less than or equal to 1.33 in the core configuration at cold conditions. Specification 4.3.1.1.b requires that for this condition with high-density fuel racks fully flooded with unborated water, the keff be less than or equal to 0.95.

The Holtec analysis confirms that the PaR high-density fuel storage rack is capable of safely accepting the fuel assemblies in the MNGP inventory, with enrichments up to 4.95 weight percent U-235, assuming a conventional loading of gadolinia. The criticality analysis demonstrates that for GE 8X8, GE 9X9 and GE 10X10 fuel, the keff of the PaR high-density SFP fuel storage racks does not exceed 0.95 provided the k4 of the fuel assemblies contained in the rack does not exceed 1.39 in the standard core configuration.

Thus, since the MNGP reload design requirement by TS limits the k4 of stored fuel to be less than or equal to 1.33 in the core configuration at cold conditions, the PaR fuel storage rack will maintain a keff less than or equal to 0.95 fully loaded with fuel that meets the MNGP reload design requirement.

The Holtec criticality analysis for the PaR 8X8 high-density fuel storage racks was performed with the NITAWL-KENO5a code package, a three-dimensional Monte Carlo code package with the 238-group SCALE cross-section library, and the CASMO-3 computer code, a two-dimensional multi-group transport theory code. The NITAWL-KENO5a code package has been extensively benchmarked against critical experiments under conditions that reflect the variables for fuel storage in the SFP.

The following conservative assumptions were made in the criticality analysis:

! The fuel racks were assumed to contain the most reactive fuel for the case being analyzed, without any control rods or burnable poison, except gadolinia, as appropriate.

! The fuel assemblies were conservatively evaluated for uniform average enrichment, i.e.,

the distribution in enrichments normally used in boiling-water reactor fuel was represented by an average value.

! The moderator was assumed to be pure, unborated water at a temperature corresponding to the highest reactivity (4o C).

! The criticality safety analyses were based upon the assumption of an infinite array of storage cells in the radial direction, i.e., no credit was taken for radial neutron leakage.

! Neutron absorption in minor structural members were neglected, i.e., the spacer grids were assumed to be replaced by water.

! In the CASMO-3 model, the flow channel was homogenized with the immediately surrounding water.

In addition, the following manufacturing effects were taken into account in the reactivity calculations of the PaR racks:

! Tolerance in B-10 loading in the Boral

! Tolerance in the lattice spacing

! Tolerance in the fuel enrichment

! Tolerance in the UO2 density

! Fuel eccentricity

! Flow channel removal

! Tolerance in clad thickness

! Tolerance in water rod thickness.

Holtec analyses of the reactivity consequences of the following accident/abnormal conditions are either bounded by the normal condition or lead to negligible reactivity effects:

Temperature and void reactivity effects Eccentric position of fuel assemblies Dropped fuel assembly Seismic event Fuel mis-location event.

The NRC staff completed its evaluation of the criticality analysis delineated above. Based on this evaluation, the NRC staff finds the proposed amendment acceptable from the perspective of this criticality analysis.

3.6 Summary of NRC Staffs Technical Evaluation As set forth above in Sections 3.1 through 3.5, the NRC staff evaluated the issues of thermo-hydraulic consideration, handling of heavy load over the SFP, structural and seismic consideration, radiological consequences due to a fuel-handling accident, and criticality. Based on such evaluation, the NRC staff finds the proposed amendment acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Minnesota State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is

no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (71 FR 16599). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, 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 defense and security or to the health and safety of the public.

Principal Contributors: W. J. Wang A. Stubbs Y. Orechwa J. McGuire Date: March 9, 2007

Monticello Nuclear Generating Plant cc:

Jonathan Rogoff, Esquire Commissioner Vice President, Counsel & Secretary Minnesota Department of Commerce Nuclear Management Company, LLC 85 7th Place East, Suite 500 700 First Street St. Paul, MN 55101-2198 Hudson, WI 54016 Manager - Environmental Protection Division U.S. Nuclear Regulatory Commission Minnesota Attorney Generals Office Resident Inspector's Office 445 Minnesota St., Suite 900 2807 W. County Road 75 St. Paul, MN 55101-2127 Monticello, MN 55362 Michael B. Sellman Manager, Nuclear Safety Assessment President and Chief Executive Officer Monticello Nuclear Generating Plant Nuclear Management Company, LLC Nuclear Management Company, LLC 700 First Street 2807 West County Road 75 Hudson, MI 54016 Monticello, MN 55362-9637 Nuclear Asset Manager Commissioner Xcel Energy, Inc.

Minnesota Pollution Control Agency 414 Nicollet Mall, R.S. 8 520 Lafayette Road Minneapolis, MN 55401 St. Paul, MN 55155-4194 Regional Administrator, Region III U.S. Nuclear Regulatory Commission Suite 210 2443 Warrenville Road Lisle, IL 60532-4351 Commissioner Minnesota Department of Health 717 Delaware Street, S. E.

Minneapolis, MN 55440 Douglas M. Gruber, Auditor/Treasurer Wright County Government Center 10 NW Second Street Buffalo, MN 55313 November 2005