ML111610136

License Amendment, Receipt and Storage of Low-Level Radioactive Waste

ML111610136
Person / Time
Site:	LaSalle
Issue date:	07/21/2011
From:	Billoch-Colon A T Plant Licensing Branch III
To:	Pacilio J M Exelon Nuclear
Billoch-Colon A
References
TAC ME3054, TAC ME3055
Download: ML111610136 (30)

Text

UNITED STATES .::.,,\-"'I;;, NUCLEAR REGULATORY COMMISSION

"" 01':; .... ... WASHINGTON, D.C. 20555*0001

<< 0 .... l: c/) i: July 21, 2011 0 "'-? .;..0 fJ Michael J. President and Chief Nuclear Exelon 4300 Winfield Warrenville, IL LASALLE COUNTY STATION, UNITS 1 AND 2 -ISSUANCE OF AMENDMENTS TO ALLOW RECEIPT AND STORAGE OF LOW-LEVEL RADIOACTIVE WASTE (TAC NOS. ME3054 AND ME3055)

Dear Mr. Pacilio:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 202 to Facility Operating License No. NPF-11 and Amendment No. 189 to Facility Operating License (FOL) No. NPF-18 for the LaSalle County Station (LSCS), Units 1 and 2, respectively.

The amendments are in response to an application submitted by Exelon Generating Company, LLC (Exelon) dated January 6,2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 100070298) as supplemented by letters dated August 20, October 14, and December 2,2010, and February 7,2011 (ADAMS Accession Nos. ML 102320599, ML 102880116, ML 103370375, and ML 110390445, respectively. ) The amendments change paragraph 2.B(5) of FOL Nos. NPF-11 and NFP18 to enable LSCS to possess and store byproduct material from Braidwood Station, Units 1 and 2, Byron Station, Unit Nos. 1 and 2, and Clinton Power Station, Unit 1 in the LSCS Interim Radwaste Storage Facility.

A copy of the Safety Evaluation is also enclosed.

The Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Sincerely, Araceli T. Billoch Colon, Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-373 and 50-374

Enclosures:

1. Amendment No. 202 to NPF-11 2. Amendment No. 189 to NPF-18 3. Safety Evaluation cc w/encls distribution via ListServ UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, DOCKET NO. LASALLE COUNTY STATION, UNIT AMENDMENT TO FACILITY OPERATING Amendment No. 202 License No. NPF-11 The U.S. Nuclear Regulatory Commission (the Commission) has found that: The application for amendment filed by the Exelon Generation Company, LLC (the licensee), dated January 6,2010, as supplemented by letters dated August 20, October 14, and December 2, 2010, and February 7, 2011, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission'S regulations set forth in 10 CFR Chapter I; The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; 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 set forth in 10 CFR Chapter I; 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 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. Accordingly, paragraph 2.8.(5) of the Facility Operating License No. NPF-11 is hereby amended to read as follows:

Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of LaSalle County Station, Units 1 and 2, and such byproduct materials as may be produced by the operation of Braidwood Station, Units 1 and 2, Byron Station, Unit Nos. 1 and 2, and Clinton Power Station, Unit 1. This license amendment is effective as of the date of its issuance and shall be implemented within 30 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION acob I. Zimmerman, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License Date of Issuance:

July 21, 2011 UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. 50-374 LASALLE COUNTY STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.189 License No. NPF-18 The U.S. Nuclear Regulatory Commission (the CommiSSion) has found that: The application for amendment filed by the Exelon Generation Company, LLC (the licensee), dated January 6,2010 as supplemented by letters dated August 20, October 14, and December 2,2010, and February 7,2011, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission'S regulations set forth in 10 CFR Chapter I; The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; 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 set forth in 10 CFR Chapter I; 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 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. Accordingly, paragraph 2.B.(5) of the Facility Operating License No. NPF-18 is hereby amended to read as follows:

Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of LaSalle County Station, Units 1 and 2, and such byproduct materials as may be produced by the operation of Braidwood Station, Units 1 and 2, Byron Station, Unit Nos. 1 and 2, and Clinton Power Station, Unit 1. This license amendment is effective as of the date of its issuance and shall be implemented within 30 days of the date of issuance.

fit:,i NUCLEAR RE9ULA TORY COMMISSION

()JMt# FvR Jacob I. Zimmerm ,Chief Plant Licensing Branch 111-2 Division of Operating Reactor licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License Date of Issuance:

July 21, 2011 ATTACHMENT TO LICENSE AMENDMENT NOS. 202 AND FACILITY OPERATING LICENSE NOS. NPF-11 AND DOCKET NOS. 50-373 AND Replace the following pages of the Facility Operating Licenses with the enclosed pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. Remove License NPF-11 License NPF-11 Page 3 Page 3 License NPF-18 License NPF-18 Page 3 Page 3

-License No. NPF-11 (4) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, 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) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of LaSalle County Station, Units 1 and 2, and such byproduct materials as may be produced by the operation of Braidwood Station, Units 1 and 2, Byron Station, Unit Nos. 1 and 2, and Clinton Power Station, Unit 1. C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth 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: Am. 198 09/16/10 (1) Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels not in excess of full power (3546 megawatts thermal).

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 201, arid the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. Am. 194 08/28/09 (3) DELETED Am. 194 08/28/09 (4) DELETED Am. 194 08/28/09 (5) DELETED Am. 194 08/28/09 (6) DELETED Am. 194 08/28/09 (7) DELETED Amendment No. 202

-3 License No. NPF-18 (5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70 possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of LaSalle County Station Units 1 and 2, and such byproduct materials as may be produced by the operation of Braidwood Station, Units 1 and 2, Byron Station, Unit Nos. 1 and 2, and Clinton Power Station, Unit 1. C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth 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 The licensee is authorized to operate the facility at reactor core power levels not in excess of full power (3546 megawatts thermal).

Items in Attachment 1 shall be completed as specified.

Attachment 1 is hereby incorporated into this license. (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 188, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. Am. 181 08/28/09 (3) DELETED Am. 181 08/28/09 (4) DELETED Am. 181 08/28/09 (5) DELETED Am. 181 08/28/09 (6) DELETED Am. 181 08/28/09 (7) DELETED Am. 181 08/28/09 (8) DELETED Am. 181 08/28/09 (9) DELETED Amendment No. 189 UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 202 TO FACILITY OPERATING LICENSE NO. NPF-11 AND AMENDMENT NO. 189 TO FACILITY OPERATING LICENSE NO. NPF-18 EXELON GENERATION COMPANY, LLC LASALLE COUNTY STATION. UNITS 1 AND 2 DOCKET NOS. 50-373 AND 50-374

1.0 INTRODUCTION

By letter to the U.S. Nuclear Regulatory Commission (NRC, the Commission) dated January 6, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession Number ML 100070298), as supplemented by letters dated August 20, October 14, and December 2, 2010, and February 7,2011 (ADAMS Accession Nos. ML 102320599, ML 102880116, ML 103370375, and ML 110390445, respectively), Exelon Generation Company, LLC (EGC, the licensee) submitted a license amendment request (LAR) to amend Facility Operating License (FOL) Nos. NPF-11 and NPF-18 for LaSalle County Station (LSCS), Units 1 and 2. The submittal proposed changes to license paragraph 2.B.(5) which would permit LSCS to possess and store byproduct material from Braidwood Station, Units 1 and 2 (Braidwood), Byron Station, Unit Nos. 1 and 2 (Byron) and Clinton Power Station, Unit 1 (CPS) in the LSCS Interim Radwaste Storage Facility (IRSF). Low-level radioactive waste (LLRW) is presently generated and stored at LSCS. The August 20, October 14, and December 2,2010, and February 7,2011, supplements contained clarifying information and did not change the NRC staff's initial proposed finding of no significant hazards consideration.

2.0 REGULATORY EVALUATION

2.1 Background

Storage of LLRW on-site at LSCS is permitted per paragraph 2.B.(5) of FOLs Nos. NPF-11 and NPF-18 for Unit 1 and Unit 2, respectively.

Paragraph 2.B.(5) references Title 10 of the Code of Federal Regulations (10 CFR) Part 30, "Rules of general Applicability to Domestic Licensing of Byproduct Material" and 10 CFR Part 70, "Domestic Licensing of Special Nuclear Material." The portions of these regulations applicable to on-site storage of LLRW are incorporated in the LSCS Unit 1 and Unit 2 licenses granted pursuant to 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities." Therefore, the licensee has requested the proposed LAR in accordance with 10 CFR 50.90. "Application for amendment of license, construction permit, or early site permit." The proposed LAR involves on-site storage of LLRW that was generated at Braidwood, Byron, and CPS. Braidwood, Byron, and CPS are licensed to the same licensee ENCLOSURE (EGC) and the disposition under 10 CFR Part 50 of the LAR is consistent with NRC practices as described in Section 3.2 "Application of Legal Authority" of Attachment 1 "Evaluation of Proposed Change" of the licensee's submittal dated January 6,2010. The licensee's supplement dated August 20,2010, confirms that EGC has established financial accounting processes to ensure that all operational costs associated with transportation and storage of the LLRW generated at Braidwood, Byron, and CPS will be allocated to Braidwood, Byron, and CPS. In addition, EGC ensures that accounting liability equivalent to the expected disposal cost for the container holding LLRW will remain the responsibility of the station that generated the LLRW. Therefore, all financial liability requirements and eventual decommissioning costs associated with the LLRW containers generated in Braidwood, Byron and CPS will be allocated to Braidwood, Byron and CPS, respectively, even though the container is stored in the LSCS LLRW storage facility.

2.2 Purpose

of the Interim Radwaste Storage Facility (lRSF) On July 1, 2008, the EnergySolutions, LLC LLRW Disposal Facility in Barnwell, South Carolina, stopped accepting LLRW from non-compact LLRW generators, including LSCS, Byron, Braidwood, and CPS. Due to the cost of constructing new storage facilities, and the existing excess storage capacity at the LSCS IRSF, EGC has requested regulatory approval for LSCS to take possession of Class B and Class C LLRW from Braidwood, Byron, and CPS for storage in the existing LSCS IRSF. Class A, Band C waste are defined in 10 CFR 61.55, "Waste Classification." 10 CFR 61.55 lists the radioactivity concentration limits of specific radioactive materials allowed in each LLRW class for near-surface disposal.

Class A LLRW contains the lowest radioactive concentration, Class B LLRW contains the next higher radioactive concentration and Class C LLRW has the highest radioactive concentration allowed to be disposed of in a near-surface LLRW disposal facility.

The LAR describes that the LSCS IRSF was built in the mid-1980s and was originally designed to store 270 containers of Class A, B, and C waste on an interim basis in order to offset the postulated lack of permanent disposal capability.

Currently, Class A waste can continue to be disposed at the LLRW to the EnergySoJutions, LLC LLRW Disposal Complex in Clive, Utah, leaving the LaSalle IRSF with additional storage capacity that was originally intended for storage of Class A waste. Therefore, storing the additional Class BIC waste will not constrain the LaSalle LLRW storage capability.

The LAR proposes the anticipated maximum storage duration of 80 years. The IRSF is located inside the LSCS Security Protected Area in a Restricted Area. The LAR states that the IRSF was designed and constructed under NRC regulatory guidance in effect at the time, primarily Generic Letter (GL) 81-38, "Storage of Low-Level Radioactive Wastes at Power Reactor Sites." GL 81-38 was issued based on the NRC expectation that many nuclear power reactor licensees were taking or planning on taking steps to provide for additional onsite storage of LLRW generated onsite due to the anticipated reduction in waste disposal availability in the United States. The NRC staff provided guidance in GL 81-38 on acceptable methods of evaluating new, proposed onsite storage facilities using the provisions of 10 CFR 50.59, "Changes, Tests and Experiments." The submittal describes that there were two 10 CFR 50.59 evaluations performed; one in 1992 and the second in 1994. The submittal further states that the 1992 evaluation was focused on mostly Class A radioactive waste, with some Class BIC containers stored in the IRSF. Additionally, this 1992 document focused exclusively on waste solidified in concrete, bitumen, and/or DOW polymer binders. The 1994 evaluation considered that a new combustion source was introduced (i.e., high-density polyethylene containers or of dewatered resins). 2.3 Applicable Regulations Subsequent to the issuance of GL 81-38, the NRC issued NUREG 0800, "Standard Review Plan," (SRP), including Chapter 11, Section 11.4, "Solid Waste Management System," (SWMS), Revision 3, dated March 2007, (ADAMS Accession No. ML070710397).

Section 11.4 provides guidance that is applicable to a broad spectrum of activities associated with the design and management of LLRW systems, including design guidance from Branch Technical Position (BTP) 11-3, Revision 3, "Design Guidance for Solid Radioactive Waste Management Systems Installed in Light-Water-Cooled Nuclear Power Reactor Plants," (SRP dated March 2007, ADAMS Accession No. ML070730202).

BTP 11-3 provides guidance on SWMS design guidelines and operation, addressing process parameters, waste stabilization or dewatering, waste form properties, free liquid detection, quality assurance, waste storage, and portable solid waste systems. BTP 11-3 focuses primarily on wet and liquid wastes for the purpose of ensuring complete stabilization and dewatering.

Section 11.4 of the SRP also incorporates relevant criterion from 10 CFR 50 Appendix A, "General Design Criteria." The SRP criteria are not a substitute for NRC regulations and compliance with them is not required.

Rather, they are criteria that are acceptable to meet NRC regulations.

The NRC staff has used selected criteria from Section 11.4 as evaluation criteria to assess the acceptability of the proposed scope of the LAR which is limited to the storage of LLRW from Braidwood, Byron and CPS in the LSCS IRSF. Appendix D, "Waste Acceptance Criteria" of Attachment 3 of the licensee's January 6, 2010, submittal states that the facility's WAC requires that all radioactive waste shall be packaged and loaded in accordance with applicable U.S. Department of Transportation (DOT) regulations and NRC regulations as described in 10 CFR Part 71, "Packaging and Transportation of Radioactive Material." The NRC regulatory requirements related to LLRW packaging, container design, transportation of LLRW, or the duration of storage of LLRW, are not within the scope of the proposed Technical Specification changes in the licensee's amendment request, and are not addressed in this NRC safety evaluation.

The licensee in its supplement dated February 7, 2011, acknowledged this limited scope of the LAR and associated NRC review. The applicable regulations for protection of individuals receiving occupational exposure and members of the public are contained in 10 CFR Part 20 "Standards for Protection Against Radiation." As stated in Section 11.4 of the SRP, the facility design and operation should assure that radiological consequences of design basis events do not exceed a small fraction (10 percent) of 10 CFR Part 100, "Reactor Site Criteria," dose limits which translates to dose limits of 2.5 roentgen equivalent man (rem) whole body and 30 rem thyroid. The NRC staff notes that for accidents involving the dispersion of solid waste activity, the controlling dose would be from inhalation of the airborne activity and should consider all dose significant radionuclides.

The licensee assessed the controlling dose for these events by using the NRC staff accepted dose conversion factors for inhalation as documented in Federal Guidance Report No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion." The licensee evaluated the inhalation dose from all dose significant radionuclides and used an acceptance limit of 2.5 rem. Section 9.5.1.1, "Fire Protection Program," Revision 0 and Regulatory Guide (RG) 1.189, "Fire Protection for Nuclear Power Plants," Revision 2, of the SRP address the review of the SWMS and waste storage facilities given the use or presence of flammable materials.

Regulatory Issues Summary (RIS) 2008-32, "Interim Low Level Radioactive Waste Storage at Reactor Sites," clarifies the NRC's position regarding LLRW storage at reactor sites. RIS 32 described that the positions on LLRW storage remain consistent with what was described in the Commission Paper (SECY)94-138, "Review of Existing Guidance Concerning the Extended Storage of Low Level Radioactive Waste." SECY-94-138, discusses that the NRC staff has eliminated any implication of a 5-year storage period limitation, and that the use of high-integrity containers (HICs) is an acceptable method for long-term storage of LLRW. SECY 94-198, consolidates all previous NRC staff guidance into a single document; addresses different aspects of managing LLRW after access to a licensed disposal facility is denied; and clarifies the meaning of the 5-year "limit" for onsite storage. SECY 94-198 also removes the statement that power reactor licensees need a 10 CFR Part 30 license to store low-level waste (LLW) beyond five years and states that licensees need only place waste into storage in a form suitable for disposal when they have assurance that the future disposal site will accept the waste in that form. 3.0 TECHNICAL EVALUATION

3.1 Proposed

FOL Changes Presently, LSCS FOL Nos. NPF-11 and NPF-18, paragraph 2.B.(5), state the following:

Pursuant to the Act and 10 CFR Parts 30,40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of LaSalle County Station, Units 1 and 2. The proposed paragraph 2.B.(5) states: Pursuant to the Act and 10 CFR Parts 30,40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of LaSalle County Station, Units 1 and 2, and such byproduct materials as may be produced by the operation of Braidwood Station, Units 1 and 2, Byron Station, Units Nos. 1 and 2, and Clinton Power Station, Unit 1. 3.2 NRC Evaluation 3.2.1 SRP Section 11.4 Applicable Acceptance Criteria The NRC staff has determined that the acceptance criteria listed below, contained in SRP Section 11.4, is applicable to the proposed LAR.

-3.2.1.1 SWMS Classification SRP Section 11.4 Acceptance Criteria No.1, states: The SWMS design parameters are based on expected radionuclide distributions and concentrations consistent with reactor operating experience for similar designs, as evaluated under SRP Section 11.1. The licensee confirmed that the LLRW classifications system for disposal (Le., Class A, B, C LLRW), is based on the radionuclide concentrations in accordance with 10 CFR 51.55 classification criteria.

The licensee analyzed an isotopic mix of resin and determined that the largest contributor to the dose rate on a container was Cobalt-50 (Co-50). Additionally, the submittal describes radiation shielding dose assessments based on Co-50 such as to provide a bounding, conservative dose assessment due to the hjgher gamma energies of Co-50 compared to other radionuclides present in the isotopic mix. The NRC staff concludes that the SWMS design parameters are based on expected radionuclide distributions and concentrations consistent with the reactor operating experience for similar designs as they relate to wastes produced during normal operation and anticipated operational occurrences.

Therefore, the NRC staff concludes that the LLRW radionuclide distributions and concentrations are acceptable and in accordance with SRP Criteria No.1. 3.2.1.2 Stabilization and/or Dewatering Methods SRP Section 11.4 Acceptance Criteria No.3, states: All liquid and wet wastes will be stabilized in accordance with a [Process Control Program] PCP before offsite shipment, or provisions will be made to verify the absence of free liquid in each container and procedures to reprocess containers in which free liquid is detected in accordance with the requirements of Branch Technical Position (BTP) 11-3. SRP Section 11.4 Acceptance Criteria No.4, states: Other forms of wet wastes will be stabilized or dewatered (subject to the licensed disposal facility's waste acceptance criteria) in accordance with a PCP, or provisions will be made to verify the absence of free liquid in each container and procedures to reprocess containers in which excess water is detected in accordance with the requirements of BTP 11-3. The licensee's submittal describes that the liquid and wet wastes are stabilized in accordance with the respective plant's PCP by de-watering the LLRW in the high density polyethylene (HOPE) containers.

It further indicates that the LLRW resin is de-watered to the 10 CFR 51 regulatory criteria for liquid to not exceed 1 percent of the waste volume as part of the facility's WAC. Other forms of wet wastes were discussed and will be stabilized in the same manner as above. All liquid and wet wastes are expected to be stabilized in accordance with the PCPs before offsite shipment, or provisions will be made to verify the absence of free liquid in each container; and procedures to reprocess containers in which free liquid is detected are developed in accordance with the requirements of BTP 11-3. The NRC staff finds that the licensee has a program for ensuring that liquid and wet wastes are appropriately stabilized to render the waste immobile and thereby mitigate the consequences of potential ruptures to shipping containers during transport.

Therefore the NRC staff finds the stabilization and/or dewatering methods proposed are acceptable.

3.2.1.3 Operations and Maintenance Procedures SRP Section 11.4 Acceptance Criteria No.9, states: The SWMS contains provisions to reduce leakage and facilitate operations and maintenance in accordance with the provisions of Regulatory Guide 1.143 and BTP 11-3, as they relate to wastes produced during normal operation and anticipated operational occurrences.

The submittal describes the licensee's container inspection procedures used to identify potential container degradation and leakage. The site has IRSF sumps installed for collection of potential liquid leakage, with procedures for emptying the sump if necessary.

Section 4.2, "Important Processes and Cabling" of Appendix E, "Fire Hazard Analysis Report for Interim Radwaste Storage Facility at LSCS" of Attachment 3, "LaSalle Station IRSF LAR Support Technical Report Supporting Engineering Change No. 375636" of the licensee's January 6, 2010, submittal indicates that during handling operations, the building ventilation system is turned off and the truck bay doors closed thereby reducing the impact of any anticipated operational occurrence such as a container drop and an inadvertent effluent release. Site procedures provide actions to take upon identification of an abnormal operation; such as continuous air monitor alarm, sump alarm, response to sump overflow, crane failure, container drop accident, fire, or upon identification of a degraded container.

LSCS has a 10 CFR Part 50 license that requires the licensee to have a radiological protection program sufficient to manage operational occurrences such as contamination from leaking containers, and provisions for radiological surveys, decontamination, as low as reasonably achievable (ALARA), and re-packaging of LLRW stored in degraded containers and reprocessing in the radwaste building.

The SWMS contains provisions to reduce leakage and facilitate operations and maintenance as they relate to wastes produced during normal operation and anticipated operational occurrences.

The NRC staff concludes that the licensee has adequate provisions to reduce leakage and facilitate operations and maintenance in that LSCS container inspection procedures provide an adequate method of identification of container degradation, the IRSF has building sumps to collect potential leakage, and a radiological protection program and procedures are in place to respond to and manage operational occurrences.

Based on the above, the NRC staff finds these measures are acceptable.

3.2.1.4 Mixed Wastes SRP Section 11.4 Acceptance Criteria No. 12, states: Mixed wastes (characterized by the presence of hazardous chemicals and radioactive materials) will be processed and disposed in accordance with 10 CFR 20.2007 ["Compliance with environmental and health protection regulations"], as it relates to compliance with other applicable Federal, State, and local regulations governing any other toxic or hazardous properties of radioactive wastes. Per Appendix D of Attachment 3 of the licensee's January 6, 2010, submittal, the facility's WAC Condition 26 prohibits the receipt of mixed waste at the LSCS IRSF. Therefore, the NRC staff concludes that mixed waste will not be stored in the IRSF. 3.2.1.5 Gaseous Effluent Releases SRP Section 11.4 Acceptance Criteria No. 13, states: All effluent releases (gaseous and liquid) associated with the operation (normal and anticipated operational occurrences) of the SWMS will comply with 10 CFR Part 20 and Regulatory Guide 1.143, as they relate to the definition of the boundary of the SWMS beginning at the interface from plant systems, including multiunit stations, to the points of controlled liquid and gaseous effluent discharges to the environment or designated onsite storage locations, as defined in the PCP and [Off site Dose Calculation Manual] ODCM. As discussed above, the NRC staff reviewed RG 1.143 and determined that it does not apply to storage facilities for LLRW. With regard to gaseous effluent releases, Section 7.4.6 of Attachment 3 to the licensee's January 6,2010, submittal, states that the LSCS storage facility will continue to be utilized for dry-waste storage-only and no dry-waste segregation, compaction, processing or repackaging will be performed in the facility.

Appendix D of Attachment 3, WAC Condition 33, states that the LSCS storage facility shall only receive aqueous liquids and other applicable waste forms which have been solidified or otherwise stabilized.

Attachment 3 of the January 6,2010, submittal indicates that the EnergySolutions 8-120 HDPE containers provide a ventilation path with high efficiency particulate air (HEPA) filtration to prevent container over-pressurization and filter potential effluent releases.

The submittal additionally indicates that the HDPE storage container has an engineered vent path with an HEPA filter. The HEPA filter is designed to remove potential particulate radioactivity effluent that could be generated when the container breathes, such as from container flexing during load lifting or from gas generation.

Truck bay doors will be closed during handling operations and there will be no processing or treatment of LLRW in the IRSF. Area radiation monitoring instrumentation will be used as well as radiation survey monitoring for direct radiation, airborne radioactivity will be performed by LSCS staff and radiological controls established such as posting of radiation areas and high radiation areas. The licensee concluded that the handling and storage of LLRW is unlikely to result in an effluent release. All effluent releases (gaseous and liquid) associated with the operation (normal and anticipated operational occurrences) of the SWMS are expected to comply with 10 CFR Part 20 and RG 1.143, as they relate to the definition of the boundary of the SWMS beginning at the interface from plant systems, including multi-unit stations, to the points of controlled liquid and gaseous effluent discharges to the environment or designated onsite storage locations, as defined in the PCP and in the ODCM. The NRC staff review found that gaseous effluent releases are unlikely from the IRSF. This conclusion is based on restrictions that ensure that the physical form of LLRW is in solid form; the HOPE storage container vent path has an installed HEPA filter that will remove potential particulate radioactivity; the doors are closed and ventilation shut off during handling operations; and there will be no LLRW processing, treatment, or repackaging activities in the IRSF. Based on the adequacy of the licensee's container inspection program to identify potentially degraded containers prior to leakage, and the installed sumps to collect potential container leakage, the NRC staff finds the licensee's treatment of gaseous effluent releases is acceptable.

3.2.2 SRP Section 11.4, Appendix A, Generally Applicable Guidance The NRC staff has determined that the generally accepted guidance listed below, contained in SRP Section 11.4, Appendix A, is applicable to the proposed LAR. 3.2.2.1 Dose Rates SRP Section 11.4, Appendix A, Generally Applicable Guidance 111.1, states: The quantity of radioactive material allowed and the shielding configurations will be dictated by the dose rate criteria for both the site boundary and unrestricted areas or site. The 40 CFR Part 190 limits will restrict the annual dose from direct radiation and effluent releases from all sources of uranium fuel cycle, and 10 CFR 20.1302 ["Compliance with dose limits for individual member of the public"] limits the exposure rates in unrestricted areas. Offsite doses from onsite storage must be sufficiently low to account for other uranium fuel cycle sources (e.g., an additional dose of less than or equal to 0.01 mSv (1 mrem) per year is not likely to cause the 40 CFR Part 190 limits, as implemented under 10 CFR 20.1301 (e) to be exceeded.

Onsite dose limits associated with temporary storage will be controlled per 10 CFR Part 20, including the ALARA principle of 10 CFR 20.1101. The submittal states that the maximum dose rate on an individual container is limited to 380 radiation absorbed dose (R) per hour (hr). The submittal continues to indirectly establish a limit on the total quantity of radioactive material allowable in order to meet the 10 CFR 20 dose rate criteria of 2 millirem (mrem) in an hour at the site boundary/unrestricted areas and the SRP Appendix 11.4-A acceptance criteria of 1 mrem per year for a real member of the public in the unrestricted area. For the determination of the indirect limit on total radioactivity and the source term evaluation, a Co-60 source term is derived based on container loading and placement restrictions whereby containers on the periphery of the IRSF are limited to a 25 Rlhr contact, and the average contact dose rate limit for upwardly exposed containers (Le., the upper layer of containers) is 50 Rlhr. This strategy places the higher dose rate containers in the center area of the IRSF, and lower radiation level containers in the peripheral areas.

In determining the source term, the licensee analyzed an isotopic mix of resin and determined that the largest contributor to the dose rate on a container was Co-50. The licensee concludes that the use of the radioisotope Co-50 in dose assessments is a conservative assumption since Co-50 has higher energy gamma photons than most other radionuclides.

Further, the licensee stated that the use of Co-50 is conservative since concrete shielding is less effective for higher energy photons, and therefore dose calculations based on Co-50 will produce a higher dose than the dose from radioisotopes with lower energy photons. The derived Co-50 source term was derived using the Microshield, Version 5.05 computer program to calculate an equivalent amount of Co-50 necessary to produce the limiting 25 Rlhr (-40 Curies) and 50 Rlhr (-80 Curies) container dose rates. The licensee's derived Co-50 source term assumed no credit for dose reduction from radiological decay in the Co-50 source term, in spite of Co-60 having a relatively short 5-year half-life compared to the 80-year storage period that will result in declining dose rates on each container throughout the storage period. Using this source term, the licensee performed a direct radiation dose analyses using modeling and shielding assumptions of a full-capacity inventory IRSF as described in Attachment 2 to the letter dated February 7,2011. The Monte Carlo (MCNP) computer code was used to perform the dose assessment.

The MCNP code tracks random photon emissions that travel through the shield walls and roof and determines the photon fluence at various receptor locations of interest.

The photon f1uence is then converted to effective dose equivalent using conversion factors obtained from American National Standards Institute/American Nuclear Society-6.1.1 dated 1991, "Neutron and Gamma-Ray Fluence-To-Dose Factors." For onsite occupational dose protection purposes, the licensee concludes that for storage periods with a fulllRSF, the dose rates are less than 1 mrem/hr outside the IRSF, less than 1 mrem/hr in the IRSF control room, and are likely to be greater than 5 mrem/hr in the truck bay requiring posting as a Radiation Area. During container handling operations, dose rates in the truck bay are likely to exceed 100 mrem/hr and will likely require posting and radiological controls as a high Radiation Area. For off-site member of the public dose assessment, the licensee's assessment is that with a completely fuIlIRSF, the maximum potential annual dose to a real off-site individual is 0.3 mrem, and that the dose in the unrestricted area will not exceed 2 mrem in an hour. The NRC staff concludes that the calculational methods using Microshield and MCNP techniques, combined with a conservative source term of the Co-60 radioisotope, and a container loading strategy are adequate methods to provide a conservative estimate of the potential off-site doses to members of the public and on-site occupationally exposed individuals.

For offsite members of the public, the NRC staff concludes that the licensee's calculation of the maximum potential annual dose of 0.3 mrem to a real off-site individual has been adequately assessed.

The potential maximum public dose is less than the acceptance criteria of 1 mrem per year and is a small fraction of the U.S. Environmental Protection Agency (EPA) 40 CFR 190 dose limit of 25 mrem/yr for all radiation sources in the uranium fuel cycle. This dose is well below the 10 CFR 20 dose limits of 100 mrem/yr for members of the public, well below the EPA 40 CFR 190 dose limit for members of the public of 25 mrem/yr, and meets the 10 CFR 20.1101(b) criteria for ALARA. Therefore, the NRC staff finds that the licensee's methods described should ensure that the dose rates both on and off site are acceptable.

-3.2.2.2 Interim Radwaste Storage Facility (IRSF) Location SRP Section 11.4, Appendix A, Generally Applicable Guidance 111.3, states: If possible, the preferred location of the additional storage facility is inside the plant's protected area. If adequate space in the protected area is not available, the licensee should place the storage facility on the plant site and establish both a physical security program (fence, locked and alarmed gates and doors, and periodic patrols) and a restricted area for radiation protection purposes.

The facility should not be in a location that requires transportation of the waste over public roads unless no other feasible alternatives exist. Licensees must conduct any transportation over public roads in accordance with the NRC and DOT regulations (10 CFR Part 71 and 49 CFR Parts 171-180).

The licensee states that IRSF is the same facility currently used by LSCS and is located within the plant's security protected area and within the radiological Restricted Area. The NRC staff reviewed the proposed location to ensure that the requirements regarding ALARA and applicable dose limits are appropriately applied for the proposed storage location.

Given that the location of the IRSF has not changed and is inside the plant's security protected area and a radiological Restricted Area, the NRC staff finds that the location of the IRSF continues to be acceptable.

As the LLRW will be received from other facilities, the NRC staff notes that any transportation over public roads must be conducted in accordance with the NRC and DOT regulations (10 CFR Part 71 and 49 CFR Parts 171-180).

3.2.2.3 Container Drop Accidents SRP Section 11.4, Appendix A, Generally Applicable Guidance 111.4, states: Licensees should implement operational safety features to prevent the accidental dropping of containers from cranes and forklifts or the puncturing of containers from forklifts during the movement and transportation of radioactive waste containers.

Personnel should receive training in the proper operation of such equipment and instruction on the use of methods to securely hold containers on such equipment (e.g., tie-downs, gates, cages). The NRC reviewed the submittal to determine whether the licensee had operational safety features to prevent the accidental dropping of containers from cranes and forklifts or the puncturing of containers from forklifts during the movement and transportation of radioactive waste containers.

A review was conducted to determine if personnel would receive training in the proper operation of such equipment and instruction on the use of methods to securely hold containers on such equipment (e.g., tie-downs, gates, cages). Section 8.7 "Container Drop Assessment" of Attachment 3 of the January 6, 2010, submittal, describes the measures in place to prevent the accidental dropping of containers.

These measures include crane safety features such as verification that the crane grapple is fully engaged when lifting containers, use of crane limit switches, crane surveillance testing and maintenance; operator training; procedural controls; closed circuit television (CCTV) monitoring, dry-runs, training and pre-job briefings.

-11 Section 4.2 "Important Processes, Equipment, and Cabling" of Appendix E "Fire Analysis Report for IRSF LSCS", of Attachment 3 of the January 6,2010, submittal also states that a container drop is considered to be highly unlikely due to the design of the container lifting equipment.

The licensee notes that the LSCS LLRW storage facility cranes are equipped with a winch system, which permits them to be manually retracted over the truck bay without electrical power or functioning drive motors. The licensee's response to the NRC request for additional information, included in the supplement dated October 14, 2010, states that the LSCS truck bay is controlled as a radiological control area and that personnel access in the truck bay is controlled in accordance with EGC procedures RP-AA-1008, "Unescorted Access to and Conduct in Radiologically Controlled Areas," and RP-AA-403, "Administration ofthe Radiation Work Permit Program".

The NRC staff concludes that the licensee has provided acceptable operational safety features, procedures, and training to prevent accidental dropping or container puncture during LLRW handling operations.

Therefore, the NRC staff finds that the above guidance has been satisfied.

3.2.2.4 Effluent Release Pathways SRP Section 11.4, Appendix A, Generally Applicable Guidance 1I1.6.A, states: Licensees shall monitor potential release pathways of all radionuclides present in the stabilized waste form as described in Appendix A to 10 CFR Part 50. Surveillance programs shall incorporate adequate methods for detecting failure of container integrity and measuring releases to the environment.

For outside storage, licensees shall conduct periodic direct radiation and surface contamination monitoring to ensure that levels are below limits specified in 10 CFR 20.1301 and 10 CFR 20.1302,10 CFR Part 71, and Subpart I (Class 7) of 49 CFR Part 173. All containers should be decontaminated to these or lower levels before storage. With regard to gaseous effluent releases, Section 7.4.6 "Effluent Release Monitoring Asessment:

NRC Inspection Manual, Inspection Procedure 65051, Observation Discussion" of Attachment 3 to the licensee's January 6, 2010. submittal, states that the LSCS IRSF will continue to be utilized for dry-waste storage-only and no dry-waste segregation, compaction, processing or repackaging will be performed in the facility.

Appendix D, Attachment 3, WAC Condition 33, states that the LSCS IRSF shall only receive aqueous liquids and other applicable waste forms which have been solidified or otherwise stabilized.

In addition, Section 8.7 of Attachment 3, to the licensee's January 6, 2010, submittal, states that a container drop is considered a very unlikely event given crane features, operator training, and procedural controls.

Appendix E, Section 4.2 of Attachment 3 of the submittal also states that a container drop is considered to be highly unlikely due to the design of the container lifting equipment.

The licensee notes that the LSCS IRSF cranes are equipped with a winch system which permits them to be manually retracted over the truck bay without electrical power or functioning drive motors. The licensee acknowledges in Section 7.4.3 Waste Characteristics and Airborne Activity Monitoring Practicalities" of

-Attachment 3 to the licensee's January 6, 2010, submittal, that should a postulated container drop occur, this could result in some initial airborne particulate if a container is breached.

However, this activity is a manned operation, so such an accident would be immediately detected.

Section 7.4.5, "NUREG-0800

-11.4 (SRP Appendix 11.4-1) "Design for Temporary Storage of Low-Level Radioactive Waste" of Attachment 3 to the licensee's January 6, 2010, submittal notes that periodic container inspection is used to detect any container degradation and that LSCS off-site doses are monitored according with the ODCM. Table 4.5-1, "LaSalle IRSF Regulatory Acceptance Criteria," page 11 of 67, in Attachment 3 of the licensee's January 6, 2010, submittal states that outside storage is not permitted at the LSCS IRSF storage facility.

The NRC staff concludes that in accordance with 10 CFR 20.1302 requirements for radiation surveys, the licensee has procedures for performing surveys upon receipt of the LLRW, and for performing radiological surveillance during and after handling and storage activities, including on-site and off-site environmental monitoring.

The results of the licensee surveys will be used to determine and demonstrate long term compliance with public dose limits in 10 CFR 20.130 "Dose limits for individual members of the public." For the outside storage of LLRW, the NRC staff reviewed the submittal to assess whether adequate provisions were in place to handle severe environmental conditions, up to and including the design-basis event for the waste storage facility.

Table 4.5-1 of Attachment 3 to the January 6, 2010, submittal indicates that "[n]o outside storage is associated with the LaSalle IRSF." Therefore, it is the NRC staff's understanding that storage of all Classes of material under review in this application will be in the IRSF. In terms of the contamination of waste in the facility and the environment, 10 CFR 20.1406 "Minimization of contamination," is applicable to applicants for plants licensed after August 30, 1997, combined licenses, and standard design certifications.

As LSCS was licensed before 1997, this section does not apply to the LSCS. Nevertheless, the licensee described operating practices and procedures that meet these criteria.

The licensee describes that the use of the IRSF is restricted to storage activities, and will not perform any type of waste processing or repackaging that could generate contamination in the IRSF. Additionally, the licensee provided a description of the radiological protection program and radiation survey procedures to survey incoming containers to provide early identification of contamination, and inspection procedures to identify degraded containers.

The licensee goes on to state that the IRSF has sumps to contain any liquid leakage that may occur. The NRC staff concludes that potential gaseous effluent releases are unlikely from the LSCS LLRW storage facility.

This conclusion is based on the restrictions and licensee controls described above, including on-site and off-site environmental monitoring, and the requirement that the physical form of LLRW is in solid form. Therefore, the NRC staff finds that effluent monitoring is not necessary for the LSCS LLRW storage facility, and that the licensee has procedures for container inspection that will be adequate to detect potential container failure. 3.2.2.5 Collection of Liquid Drainage SRP Section 11.4, Appendix A, Generally Applicable Guidance 1I1.6.B, states:

-13 Licensees should incorporate provisions for collecting liquid drainage, including provisions for sampling all collected liquids. Routing of the collected liquids should be to radwaste systems if contamination is detected or to normal discharge pathways if the water ingress is from external sources and remains uncontaminated by plant-generated radioactivity.

Section 8.9, "IRSF Liquid Identification Capability" of Attachment 3 to the January 6, 2010, submittal indicates that separate sump systems (hot and cold) are provided.

The hot sump is located in the truck bay which allows hot liquid drainage from the storage bay and decontamination area. Its high-level alarm provides annunciation in the IRSF. The instrumentation on the sump has liquid level indication and provisions for sample collection.

The licensee indicated that sump levels are part of the routine surveillance program and sampling procedures, and sampling equipment is already in place. The NRC staff reviewed the submittal to determine whether provisions for collecting liquid drainage, including provisions for sampling all collected liquids were included.

The NRC staff found that the existing provisions for the collection and sampling of liquid drainage remain sufficient.

Therefore, the NRC staff concludes that the licensee has provided leak collection equipment and instrumentation that is adequate to collect and sample potential liquid leakage. 3.2.2.6 External Weathering and External Corrosion SRP Section 11.4, Appendix A, Generally Applicable Guidance 111.6.0, states: Licensees should assure container integrity against corrosion from the external environment, including external weather protection where necessary and practical.

Storage containers should be raised off storage pads where water accumulation can be expected to cause external corrosion and possible degradation of container integrity.

The NRC staff reviewed the submittal to determine whether the selected containers were protected against corrosion from the external environment, including external weather protection where necessary and practical.

Item 3 of Table 4.5-1 of Attachment 3 of the licensee's January 6, 2010, submittal, states that the protection provided by the storage facilities will serve to minimize effects of external weather condition.

Section 8.1.1.6, "Heating and Ventilation System" of Attachment 3 also states that the LSCS I RSF contains two patterns of ventilation and heating system for cooling and heating. In addition, Appendix B states that the ventilation damper lineup will depend on outside weather conditions and that there are two modes of operation; summer and winter. The submittal describes that LLRW will be stored inside a climate-controlled building with heating and air-conditioning that provides protection against external weathering.

The NRC staff concludes that storage is inside a climate-controlled building that provides suitable protection from external weather factors. The NRC staff also reviewed the South Carolina Department of Health and Environmental Control's certificate of compliance (COC) for the EnergySolutions 8-120 HOPE containers.

Compatibility of the container materials with the waste forms and with environmental conditions external to the containers is necessary to prevent significant container corrosion.

Container

-selection was reviewed to determine whether data demonstrated that minimal corrosion from the anticipated internal and external environment for a period would be well in excess of the planned storage duration.

Additionally, container integrity was reviewed to ensure that after the period of storage the container should be sufficient to allow handling during transportation and disposal without container breach. The NRC staff found that the COC provides that the HDPE containers are suitable containers for disposal of Class A, B, and C waste, and are approved for containment of resins, solid waste, filters, solidified resins, and sludges. In addition, SECY-94-198 acknowledges that HICs are acceptable by today's standards for storage in that waste can be easily removed and placed into a form meeting new (future) waste disposal acceptance criteria.

The licensee's container inspection program was reviewed for indications of container swelling, corrosion products and signs of breaching.

The NRC staff review was to ensure that the license had a program which provided for at least periodiC (quarterly) visual inspections of container integrity (e.g., swelling, corrosion products, leaks, or breach). The licensee stated that periodic visual inspections and surveillance will be performed of the storage bay with CCTV monitors and with direct visual inspections, including sump level monitoring.

Item 4.1.4 of Section 4.0 "Main Body" of EGC procedure RW-AA-105, Revision 2, "Guidelines For Operating an Interim On Site low level Radioactive Waste Storage Facility" (included in Attachment 3 to the licensee's January 6, 2010, submittal), provides guidance that storage containers should be raised off storage pads, where water accumulation can be expected to cause external corrosion and possible degradation of container integrity.

The NRC staff concludes that the HDPE containers are compatible with the waste form, and the proposed inspections are adequate to identify potential container degradation.

The NRC staff notes that this review did not include long-term container integrity.

3.2.2.7 Material Inventory limits SRP Section 11.4, Appendix A, Generally Applicable Guidance 1I1.6.E, states: licensees should establish total radioactive material inventory limits (in becquerels and curies), based on the design of the storage area, dose limits for members of the public, and safety features or measures being provided (e.g., radiation monitoring).

Per Table 4.5-1, page 18 of 67, in Attachment 3 of the licensee's January 6,2010, submittal, the licensee indirectly established a total radioactive material inventory as described in the guidance above. The licensee stated that: Container contact dose rates, isotopic considerations, container placement, timing, and supplementary shielding are the controlling parameters that are used to assure that dose limits are met. In this context, a simple total radioactivity inventory limit has no practical value. The NRC staff reviewed the submittal to determine whether radioactive material inventory limits (in becquerels and curies) had been established, based on the design of the storage area, dose limits for members of the public, and safety features or measures being provided (e.g., radiation

-15 monitoring).

The licensee indirectly established a total radioactive material inventory as described above. The NRC staff concluded that the use of a dose based criteria and a container loading and placement restrictions are an acceptable alternative to establishing a specific total radioactive material inventory limit since dose based criteria and container loading and placement restrictions provide an equivalent method of ensuring that potential radiation doses are limited to within regulatory limits. 3.2.2.8 Maintenance of Inventory Records SRP Section 11.4, Appendix A, Generally Applicable Guidance 1I1.6.F, states: Licensees should maintain inventory records by waste types, waste contents, radionuclides and radioactive material, dates of storage, shipment, and other relevant data. The NRC staff reviewed the submittal to determine whether the license maintains inventory records by waste types, waste contents, radionuclides and radioactive material, dates of storage, shipment, and other relevant data. Attachment 3 to the January 6, 2010, submittal describes procedures for inspection, survey, and recordkeeping for incoming LLRW containers, and routine radiation surveys during the storage period. The results of the licensee surveys will be used to determine and demonstrate term compliance with public dose limits in 10 CFR 20.1301. The NRC staff reviewed the descriptions to ensure that administrative controls are appropriate to limit the exposure rates in unrestricted areas. The NRC staff concludes that the procedures described are adequate to perform surveys upon receipt of the LLRW, and radiological surveillance during and after handling and storage activities, including on-site and off-site environmental monitoring.

3.2.2.9 Ventilation Exhaust and Airborne Monitoring SRP Section 11.4, Appendix A, Generally Applicable Guidance 1I1.6.G, states: The facility design should incorporate provisions for a ventilation exhaust system (for storage areas) and an airborne radioactivity monitoring system (building exhaust vents) where there is a potential for airborne radioactivity to be generated or to accumulate.

Section 7.4.5.3 of Attachment 3 to the January 6,2010, submittal, describes the IRSF ventilation system with heating and air conditioning system for climate control and a rooftop exhaust vent. The licensee indicates that the ventilation system is turned off and truck bay doors closed during container handling, such that a container drop event should not result in a direct effluent release. Continuous airborne radioactivity monitors are provided in the IRSF truck bay and in the IRSF ventilation system. The NRC staff reviewed the facility design to determine whether a ventilation exhaust system (for storage areas) and an airborne radioactivity monitoring system (building exhaust vents) existed where there is a potential for airborne radioactivity to be generated or to accumulate.

-The NRC staff review concludes that the existing LSCS IRSF ventilation exhaust system and airborne radioactivity monitoring remain acceptable.

3.2.2.10 Container Protection During Storage and Provisions for Reprocessing SRP Section 11.4, Appendix A, Generally Applicable Guidance V.2, states: Any storage plans should address container protection and any reprocessing requirements for eventual shipment and burial. As discussed previously container protection will be provided by storing containers inside a climate-controlled building, by use of container handling procedures, and container inspection procedures for identification of potential container degradation.

Section 8.8, "IRSF Container Repacking Capacity" of Attachment 3 indicates that any needed reprocessing or re-packaging for degraded containers will not be performed in the IRSF, and instead, would be accomplished by moving the container to the plant's radwaste building.

The NRC staff reviewed the submittal to determine whether the storage plans address container protection and any reprocessing requirements for eventual shipment and burial. The NRC staff concludes that the licensee has provided adequate container protection measures and the criteria processes for reprocessing or repackaging for degraded containers by moving such containers to the radwaste building have been met and are acceptable.

3.2.2.11 Uncontrolled Releases During Storage SRP Section 11.4, Appendix A, Generally Applicable Guidance V.3, states: Casks, tanks, and liners containing stabilized radioactive waste should be designed with good engineering judgment to preclude or reduce the probability of uncontrolled releases of radioactive materials during handling, transportation, or storage. Licensees must evaluate the accident mitigation and control procedures and their ability to protect the facility from design basis events (e.g., fire, flooding, and tornadoes) unless otherwise justified.

As discussed above, the South Carolina Department of Health and Environmental Control issued a COC for the EnergySolutions 8-120 HOPE containers.

The COC provides that the HOPE containers are suitable containers for disposal of Class A, B, and C LLRW, specifically including resins, solid waste, filters, solidified resins, and sludges. The HOPE containers are subject to a routine inspection program which should therefore reduce the probability of uncontrolled releases during storage. The NRC staff reviewed COC for the EnergySolutions 8-120 HOPE containers to determine whether the casks, tanks, and liners containing stabilized radioactive waste are designed with good engineering judgment to preclude or reduce the probability of uncontrolled releases of radioactive materials during handling, transportation, or storage. Section 8.4 "Design Basis Event Accident" of Attachment 3 to the licensee's January 6,2010, submittal, describes the design basis event assessments related to the LSCS IRSF storage facility.

The licensee performed assessments of design basis events including fires, tornadoes, floods, seismic activity, and container drops events.

-The storage facility has been analyzed for the LSCS design basis flood and the construction of the building is adequate to withstand tornados and resultant missiles.

In addition, the LSCS IRSF is designed to adequately limit the impacts of an operating basis earthquake.

The NRC staff reviewed Appendix E of Attachment 3 to the January 6, 2010, submittal for the fire hazards analysis (FHA) and the licensee evaluation of potential hydrogen gas generation provided in Appendix B of the submittal.

The NRC staff review focused on whether procedures were adequately developed for early detection, prevention, and mitigation of accidents (e.g., fires). Storage areas and facility designs incorporated appropriate features and capabilities for handling accidents and provide safeguard systems. The licensee's submittal identified in situ combustible materials used in plant systems, structures, and components and specified suitable fire protection.

The licensee's hydrogen gas evaluation (Appendix B of Attachment

3) concludes that the concentration of hydrogen will not exceed 50 percent of the lower explosive limit for hydrogen and thus meets the guidance provided in Section 4.1.8 of RG 1.189. The installed electrical cable meets the flame spread test requirements of the Institute of Electrical and Electronics Engineers 383 and the fire protection systems were designed in accordance with the appropriate National Fire Protection Association standards.

Therefore, the NRC staff finds that the licensee's FHA meets the guidance provided in RG 1.189, Section 6.2.3 and in SRP 11A-A for a waste storage facility and is acceptable.

The NRC staff concludes that the licensee has evaluated the LSCS IRSF for the effects of design basis events. These evaluations confirm that the facility is designed to protect stored LLRW from the effects of the postulated events and a container drop accident.

Based on the FHA and hydrogen gas evaluation, the NRC staff finds that adequate prevention/mitigations measures are sufficient given the presence of flammable materials.

Also, based on a satisfactory inspection program and accident mitigation and control procedures, the NRC staff finds that the casks, tanks, and liners containing stabilized radioactive waste are sufficiently designed to preclude or reduce the probability of uncontrolled releases during handling, transportation, or storage. 3.2.2.12 Material Control and Accountability SRP Section 11A, Appendix A, Generally Applicable Guidance VA.A, states: All stabilized radwaste should be located in restricted areas where effective material control and accountability can be maintained.

While structures are not required to meet seismic criteria, licensees should employ good engineering judgment to ensure that radioactive materials are contained safely, such as by the use of curbs and drains to contain spills of dewatered resins or sludge. Attachment 3 of the January 6, 2010, submittal, describes that the IRSF is located within the plant's security protected area and within the plant's Restricted Area. The submittal addresses in Appendix D of Attachment 3, the record keeping requirements to meet material accountability and radioactive material controls.

Additionally, the submittal describes the selection of containers, use of radiation monitoring techniques, container handling procedures, equipment to properly handle and inspect containers, and sumps used to contain potential leakage of liquid from the containers.

The NRC staff reviewed the submittal to ensure that all stabilized radwaste would be located in Restricted Areas where effective material control and accountability can be maintained.

Additionally the NRC staff reviewed whether measures were in place to ensure that spills of dewatered resins or sludge are appropriately contained.

As discussed previously, the IRSF is located in a Restricted Area, the WAC requires material control and accountability, and that good engineering features are provided to contain spills. Therefore the NRC staff concludes that adequate material accountability and control measures are present. 3.2.2.13 Contamination Isolation and Decontamination SRP Section 11.4, Appendix A, Generally Applicable Guidance V.4. C, states: There should be provisions for additional reprocessing or repackaging in the event of container failure and/or as required by DOT regulations and license disposal facility criteria for final transportation and disposal.

Licensees should develop contamination isolation and decontamination capabilities.

When significant handling and personnel exposure can be anticipated, licensees should incorporate ALARA methodology in accordance with Regulatory Guides 8.8 ["Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be as Low as is Reasonably Achievable, June 1978"] and 8.10 ["Operating Philosophy For Maintaining Occupational Radiation Exposures as Low as is Reasonably Achievable, May 1977"]. Section 8.6, "IRSF Decontamination Capability" of Attachment 3 to the January 6, 2010, submittal, discusses methods of reprocessing or repackaging in the event of container failure or loss of container integrity (swelling, breaching, cracking, etc.). In the event of container failure, LSCS has the capability to transfer containers to the plant's radwaste building and use eXisting in-plant radioactive waste processing equipment to reprocess (e.g., de-water) or repackage the LLRW into a new container.

Radiation survey methods have been developed for detecting and controlling contamination, including use of sumps to contain liquid spills. In addition, ALARA methods are in place to minimize occupational dose by the use of CCTV cameras and shielding techniques are employed when performing operations in the LSCS IRFS. The NRC staff reviewed the submittal to address whether provisions for additional reprocessing or repackaging in the event of container failure were available.

Additionally, the NRC staff reviewed whether ALARA methods were incorporated when significant handling and personnel exposure can be anticipated.

The NRC staff concludes that the ability to transfer degraded containers to the plant's radwaste building for reprocessing and repackaging are sufficient in event of container failure; appropriate contamination control measures exist, and ALARA techniques are provided.

Therefore, the NRC staff finds these measures and methods in the event of container failure acceptable.

3.2.2.14 Mitigation of Accidents SRP Section 11.4, Appendix A, Generally Applicable Guidance V.4.D, states:

Licensees should develop and implement procedures for early detection, prevention, and mitigation of accidents (e.g., fires). Storage areas and facility designs should incorporate good engineering features and capabilities for handling accidents and provide safeguard systems, such as fire detectors and suppression systems (e.g., smoke detectors and sprinklers).

If water sprinkler systems are used, floors should be sloped to drain into local floor sumps or curbed to prevent water runoff to uncontrolled areas. Licensees should establish personnel training and administrative procedures to ensure both control of radioactive materials and minimum personnel exposures.

Fire suppression devices may not be necessary if combustible materials in the area are minimal. The waste that will be stored in the LSCS IRSF is Class B/C LLRW which is similar to the types of LLRW currently licensed and stored in the LSCS IRSF. As discussed in Section 1 of Attachment 4 to the January 6, 2010, submittal, the licensee considered the bounding design basis events were a waste handling accident, involving a container drop, and a postulated fire in the storage bay. The design bases analyses provided evaluates postulated fire, tornado, flood, seismic, and container drop events. The licensee determined that the original LSCS IRSF 10 CFR 50.59 evaluations for tornado, flood, and seismic bound the proposed LLRW storage. The licensee used conservative assumptions to maximize the calculated dose from the container drop accident and the container fire event. In both evaluations, the licensee assumed that all affected containers have a contact dose rate of 380 rem/hr. This is the maximum allowable contact dose rate for HICs stored in the IRSF, based on the requirements of the licensee's ODCM. No water sprinkler systems are used, therefore consideration of runoff of fire sprinkler water to uncontrolled areas is not needed. For the fire event, the licensee assumed that six HICs are involved.

This is the maximum number of HICs allowed to be grouped together based on procedures to isolate combustible materials.

The licensee assumed that the containers are grouped in three stacks, two containers high. The licensee stated that normal stacking practice would place lower dose rate containers on top of higher dose rate containers to reduce the dose from air scattered radiation, however to maximize the calculated accident dose the licensee conservatively assumed that each HIC has the maximum allowable contact dose rate. The licensee chose the most restrictive airborne release fraction for resin fires from the measured values listed in the U.S. Department of Energy handbook, "Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities," dated December 1994. The licensee then conservatively rounded up this value by approximately 20 percent. The licensee evaluated the container drop accident using the release fraction based on a dry powder spill scenario from NUREG-1320, "Nuclear Fuel Cycle Facility Accident Analysis Handbook," dated May 1988. The licensee asserts that for the worst-case de-watered resin, which is expected to be the consistency of wet sand, as well as for the larger bead resin, released particulate activity would be less likely to become airborne than for a dry powder. In addition, the referenced dry-powder spill scenario includes an airborne contribution resulting from shear stress as the powder falls through the air, which would not be applicable for a container drop. In addition, the licensee conservatively rounded up the release fraction for dry powder by apprOXimately 20 percent. The NRC staff agrees that the use of a release fraction based on a dry-powder spill is conservative and therefore acceptable for use in the dose consequence analysis.

-The licensee assumed for each event scenario that all the airborne activity is instantly released to the environment with no credit for particle settling within the IRSF or in transit to the exclusion area boundary.

The licensee noted that during container handling the IRSF ventilation system is isolated to maximize particulate settling should any releases occur. In addition, EGC procedure LOP-WX-33, Revision 1, "AbnormallRSF Operation Procedure," March 2008, directs that the IRSF ventilation system be secured during any fire event. The NRC staff notes that the assumption of an instantaneous release to the environment with no credit for settling is conservative.

LSCS IRSF Internal Design Basis Events Event Calculated Dose Limit Container Drop 0.389 rem 2.5 rem Container Fire 2.334 rem 2.5 rem Based on the assumptions above and a review of the licensee's calculations, the NRC staff concludes that the licensee's IRSF design basis dose assessment used conservative assumptions to conclude that the accident dose will remain below the acceptance criteria stated in the SRP and is therefore acceptable.

3.2.2.15 Ventilation Exhaust System and Storage Area Monitoring SRP Section 11.4, Appendix A. Generally Applicable Guidance VA-E, states: The facility design should incorporate provisions for a ventilation exhaust system (for storage areas) and an airborne radioactivity monitoring system (building exhaust vents) where there is a potential for airborne radioactivity to be generated or to accumulate.

Attachment 3 of the January 6, 2010, submittal, describes the ventilation exhaust system for the affected areas and operation of monitoring and exhaust systems. The NRC staff confirmed that a ventilation exhaust system for storage areas and an airborne radioactivity monitoring system were present where there is a potential for airborne radioactivity to be generated or to accumulate.

Therefore, the NRC staff finds that adequate ventilation and monitoring are present for storage areas where airborne radioactivity may be generated or accumulate.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Illinois State official was notified on July 15, 2011, of the proposed issuance of the amendments.

The State official had no comments.

5.0 ENVIRONMENTAL

CONSIDERATION Pursuant to 10 CFR Sections 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact was published in the Federal Register on February 17, 2011 (76 FR 9379). Accordingly, based upon the environmental assessment, the Commission has determined that issuance of this amendment will not have a significant effect on the quality of the human environment.

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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 amendments will not be inimical to the common defense and security or to the health and safety of the public. Principal Contributors:

Stephen Garry, NRR John Parillo, NRR Phil Qualls, NRR Date of issuance:

July 21,2011 July 21, 2011 Mr. Michael J. Pacilio President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555 LASALLE COUNTY STATION, UNITS 1 AND 2 -ISSUANCE OF AMENDMENTS TO ALLOW RECEIPT AND STORAGE OF LOW-LEVEL RADIOACTIVE WASTE (TAC NOS. ME3054 AND ME3055)

Dear Mr. Pacilio:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 202 to Facility Operating License No. NPF-11 and Amendment No. 189 to Facility Operating License (FOL) No. NPF-18 for the LaSalle County Station (LSCS), Units 1 and 2, respectively.

The amendments are in response to an application submitted by Exelon Generating Company, LLC (Exelon) dated January 6,2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 100070298) as supplemented by letters dated August 20, October 14, and December 2,2010, and February 7,2011 (ADAMS Accession Nos. ML 102320599, ML 102880116, ML 103370375, and ML 110390445, respectively. ) The amendments change paragraph 2.B(5) of FOL Nos. NPF-11 and NFP18 to enable LSCS to possess and store byproduct material from Braidwood Station, Units 1 and 2, Byron Station, Unit Nos. 1 and 2, and Clinton Power Station, Unit 1 in the LSCS Interim Radwaste Storage Facility.

A copy of the Safety Evaluation is also enclosed.

The Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Sincerely, IRA! Araceli T. Billoch Colon, Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-373 and 50-374

Enclosures:

1. Amendment No. 202 to NPF-11 2. Amendment No. 189 to NPF-18 3. Safety Evaluation cc w/encls distribution via ListServ DISTRIBUTION:

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• SE Memo date OFFICE LPL3-21PM LPL3-2/LA IHPBIBC AFPB/BC AADBIBC* OGC LPL3-2/BC NAME ABillochCol6n SRohrer UShoop AKlein TTate RHarper JZimmerman(M Mahoney for) DATE 7/18/11 7/18/11 7/14/11 7/14/11 2/14/11 7/14/11 7121/11 OFFICIAL RECORD COPY