View source for RS-10-007, LaSalle Station, Units 1 and 2 - Irsf LAR Support Technical Report Supporting Engineering Change (EC) No. 375636 Type: Document Change Request (DCR) Subtype: Dsgn

{{Adams
| number = ML100070301
| issue date = 09/28/2009
| title = LaSalle Station, Units 1 and 2 - Irsf LAR Support Technical Report Supporting Engineering Change (EC) No. 375636 Type: Document Change Request (DCR) Subtype: Dsgn
| author name = Propst N
| author affiliation = URS Corp
| addressee name = 
| addressee affiliation = Exelon Nuclear, NRC/NRR
| docket = 05000373, 05000374
| license number = 
| contact person = 
| case reference number = 29487-NCS0097, FOIA/PA-2010-0209, RS-10-007
| document report number = 375636
| package number = ML100070297
| document type = Report, Technical
| page count = 173
}}

=Text=
{{#Wiki_filter:ATTACHMENT 3 LaSalle Station IRSF LAR Support Technical Report Supporting Engineering Change (EC) No. 375636  

LASALLE STATION UNITS 1 & 2 

Project Number 29487-NCS0097 

LASALLE STATION IRSF LAR SUPPORT TECHNICAL REPORT SUPPORTING ENGINEERING CHANGE (EC) NO. 375636  TYPE: DOCUMENT CHANGE REQUEST (DCR) SUBTYPE: DSGN PREPARED FOR 

EXELON NUCLEAR 

PREPARED BY 

URS - Washington Division 510 Carnegie Center Princeton, NJ 08540 

Revision: 0 Status: Final

Project Name: Client: STUDY REVISION PAGE TECHNICAL REPORT SUPPORTING ENGINEERING CHANGE DOCUMENT CHANGE REQUEST IRSF 50.59 LAR Support Exelon Nuclear-LaSalle County Nuclear Station Discipline:
Process Project Number: 29487-NCS0097 Latest Revision: 0 Revision Signatures EC Number 375636 l k,'it&#xa3;/o'l Prepared by Noani PropstDeChecked by Ed Taylor Date Approved by Don Gardner-SDE Approved by (if required)(title)Date Status Draft Final Rev.No.B o Date 8/10/09 9/28/09 Prepared By E.R.T.N.P.Pages All All Description of Changes 90%Draft For Use and Reference TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 1 of 67 TABLE OF CONTENTS  

==1.0 INTRODUCTION==
AND BACKGROUND.........................................................3 2.0 ENGINEERING CHANGE DESIGN CHANGE REQUEST OJECTIVES..5 3.0 DISCUSSION OF PRIOR IRSF 50.59 SAFETY ASSESSMENTS.................6 4.0 REGULATORY GUIDANCE.............................................................................7 4.1 R EGULATORY ISSUE 

==SUMMARY==
 (RIS) 2008-32, I NTERIM L OW L EVEL R ADIOACTIVE W ASTE S TORAGE AT REACTOR S ITES, DATED D EC. 30, 2008..........................................7 4.2 G ENERIC LETTER 81-38......................................................................................................7 4.3 NUREG 0800 SRP 11.4A, D ESIGN GUIDANCE FOR T EMPORARY S TORAGE OF L OW-L EVEL R ADIOACTIVE W ASTE, R EV. 3 DATED 2007..........................................................8 4.4 SECY-94-0198...................................................................................................................
8 4.5 IRSF A CCEPTANCE C RITERIA............................................................................................9 5.0 RADIOACTIVE WASTE CHARACTERISTICS (RECEIVING STATION AND FORWARDING STATION)................................................................................20 6.0 INITIAL DURATION OF STORAGE.............................................................21 7.0 COMMON TECHNICAL ASSESSMENTS....................................................22 7.1 D EWATERED R ESIN F LAMMABLE G AS G ENERATION ASSESSMENT................................22 7.2 HDPE (P OLY) H IGH I NTEGRITY C ONTAINERS (HIC) C ONTAINER I NTEGRITY ASSESSMENT....................................................................................................................23 7.3 W ASTE A CCEPTANCE C RITERIA (WAC)..........................................................................24 7.4 E FFLUENT R ELEASE M ONITORING ASSESSMENT.............................................................25 7.4.1 Purpose.................................................................................................................................25 7.4.2 Current Facility Configurations............................................................................................25 7.4.3 Waste Characteristics and Airborne Activity Monitoring Practicalities...............................26 7.4.4 NRC Regulatory Positions on Release (Effluent) Monitoring..............................................26 7.4.5 NUREG-0800 - 11.4 (SRP Appendix 11.4-A) Design Guidance for Temporary Storage of Low-Level Radioactive Waste, Reference 4.....................................................................................3 0 7.4.6 NRC Inspection Manual, Inspection Procedure 65051 (Reference 7), Subsection 03.01, paragraphs g and h, indicate the following:........................................................................................33 7.4.7 EPRI Guidelines for Operating an Interim Onsite Low Level Radioactive Waste Storage Facility - Revision 1, Reference 10,.................................................................................................35 7.4.8 IAEA.....................................................................................................................................37 7.4.9 Conclusions..........................................................................................................................37 7.4.10 References........................................................................................................................38 8.0 LASALLE IRSF TECHNICAL ASSESSMENT.............................................39 8.1 IRSF F ACILITY D ESCRIPTION...........................................................................................39 8.1.1 Physical Description.............................................................................................................39 8.2 S ITE D OSE C RITERIA V ALUES , L OCATION , AND B ASES...................................................51 8.2.1 Onsite Radiation Protection Considerations.........................................................................51 8.2.2 Offsite Radiation Protection Considerations........................................................................51 8.3 P HYSICAL S ECURITY P ROGRAM ASSESSMENT.................................................................52 8.4 D ESIGN BASIS E VENT ASSESSMENTS...............................................................................52 8.4.1 Fire Assessment....................................................................................................................52 TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 2 of 67  8.4.2 Tornado Assessment.............................................................................................................
54 8.4.3 Flood Assessment.................................................................................................................55 8.4.4 Seismic Assessment..............................................................................................................55 8.5 S HIELDING ASSESSMENT..................................................................................................56 8.5.1 Facility and Shielding Description.......................................................................................56 8.5.2 Regulatory Requirement for Which Shielding is Necessary.................................................56 8.5.3 Shielding Analysis Methodology..........................................................................................57 8.5.4 IRSF Loading Strategies.......................................................................................................5 7 8.5.5 Impacts of Accepting Clinton, Byron and Braidwood Waste at the LaSalle IRSF for shielding purposes...............................................................................................................................58 8.6 IRSF D ECONTAMINATION C APABILITY...........................................................................59 8.7 C ONTAINER D ROP ASSESSMENT......................................................................................60 8.8 IRSF CONTAINER R EPACKAGING CAPACITY...................................................................60 8.9 IRSF L IQUID I DENTIFICATION C APABILITY.....................................................................61 8.10 IRSF ANNUNCIATION TO CONTINUOUSLY M ONITORED A REA...................................61 8.11 UFSAR AND T ECHNICAL SPECIFICATIONS R EVIEW....................................................61 8.11.1 Technical Specifications and Facility Operating Licenses...............................................61 8.11.2 UFSAR Revisions............................................................................................................61 8.11.3 Other Licensing Basis Documents...................................................................................62

==9.0 CONCLUSION==
S.................................................................................................63 10.0 10 CFR 50.59 REVIEW-LASALLE..................................................................65 APPENDICES Appendix A - Discussion of Prior IRSF 50.59 Safety Assessments Appendix B- Dewatered Resin Flammable Gas Generation Assessment Appendix C- HDPE (Poly) High Integrity Containers (HIC) Container Integrity Assessment Appendix D - Waste Acceptance Criteria (WAC)
Appendix E- Fire Hazard Analysis Report for Interim Radwaste Storage Facility at the LaSalle County Nuclear Station TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 3 of 67  

==1.0 INTRODUCTION==
AND BACKGROUND Exelon Nuclear (Exelon) designed and constructed an Interim Radwaste Storage Facility (IRSF) in the mid-1980s at LaSalle County Nuclear Station (LaSalle). The facility was designed to store radioactive waste on an interim basis, i.e., at the time arbitrarily specified by NRC as up to five years. The primary reason for the IRSF was to offset lack of disposal in case existing disposal facilities, such as the Southeast Compacts Barnwell Disposal Facility in Barnwell South Carolina, ceased accepting radioactive waste from utilities not in the Southeast Compact. The Low-Level Radioactive Waste (LLRW)
Policy Amendments Act of 1985 (LLRWPAA) allows compacts with operating LLRW disposal sites to deny access to generators in States and compacts that have not developed 

disposal capacity on their own. 

The LaSalle IRSF was designed to specific NRC regulatory guidance documents at the time, primarily Generic Letter (GL) 81-38 entitled Storage of Low-Level Radioactive Wastes at Power Reactor Sites. Other regulatory guidance documents that were utilized for radiological dose criteria were 10 CFR 20 for dose outside the building shield walls, and, 40 CFR 190 for dose at the nearest public dose receptor. Waste containers were accepted for interim storage if they met NRC and DOT requirements, as well as burial site waste acceptance criteria requirements, for eventual shipment to the disposal facility. 

Approximately ninety per cent of the Radwaste projected to be stored in the LaSalle IRSF in that period of time was Class A, with the balance being Class B/C waste. Class A, B, and C low-level radioactive wastes are generally acceptable for near-surface disposal and are defined in 10CFR61. Section 10CFR61.55 lists the radioactivity concentration limits of specific radioactive materials allowed in each low-level waste class. Class A low-level radioactive waste contains the lowest radioactive concentration and constitutes the vast bulk of radioactive waste. Class B contains the next lowest radioactive concentration.
Class C waste has the highest radioactive concentration allowed to be disposed of in a low-level waste disposal facility. 

Most Class A waste meets categorization as Low Specific Activity (LSA) waste, such as condensate resin. The class B/C waste is typically high specific activity resins generally with isotopics being driven by high levels of Co-60, thereby resulting in relatively high container contact dose rates (15 or more R/hr). Class B/C Radwaste is typically generated at the Station by the Reactor Water Cleanup System (RWCS) or the Spent Fuel Cleanup System (SFCS) and is stored in High Density Polyethylene (HDPE) High Integrity Containers (HICs) in a dewatered bead resin waste form. The dewatered resins meet Generic Letter 81-38 requirements for Stabilized Waste. 

The LaSalle IRSF features 30-inch concrete shield walls on the peripheral wall system for direct transmission protection and a concrete roof that is 12 to 15 inches thick for 

skyshine protection purposes.  

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 4 of 67  On July 1, 2008 the Barnwell Disposal Facility in the Southeast Compact closed its doors to out-of-compact Radwaste, which precluded LaSalle from shipping Class B/C Radwaste to an outside disposal facility. Class A waste generated by LaSalle is still able to be disposed at the Envirocare of Utah LLRW Disposal Complex in Clive, Utah.
Thus the need for utilizing the LaSalle IRSF for storing Class B/C Radwaste for an extended period, perhaps life-of-plant or more, became apparent. The LaSalle IRSF new purpose is limited to storing Class B/C Radwaste only. Class A Radwaste may continue to be staged in this facility awaiting shipment to disposal. 

Additionally, other Midwest nuclear stations located in Illinois that did not build an IRSF heretofore will also need extended Radwaste storage. In early 2009, Exelon made a decision to store Radwaste from the Byron Nuclear Station (Byron), Braidwood Nuclear 

Station (Braidwood), and Clinton Nuclear Station (Clinton) at the LaSalle IRSF. As only Class B/C Radwaste will need to be shipped, the original volumetric capacity of the LaSalle IRSF is capable of handling the small number of additional expected shipments 

annually from the Exelon sister nuclear stations in Illinois. 

Regulatory guidance related to extended storage of Radwaste in IRSFs must be considered to determine whether current design and operations of the existing LaSalle IRSF is acceptable, including review of prior Safety Evaluations (50.59) Reports, for this new mission. Additionally, this Technical Report provides the technical bases for Exelons License Amendment Request (LAR) to be submitted to NRC to allow possession of Class B/C waste from the Byron, Braidwood, and Clinton Stations. 

Review of a previous applicable precedent regarding Radwaste shipment to the Tennessee Valley Authority Sequoyah Nuclear Station from their Watts Bar Nuclear Station was carried out prior to making the decision to ship Exelon sister station Radwaste to LaSalle. The NRC approved a License Amendment Request (LAR) dated December 17, 1999 regarding the shipment of Radwaste to the Sequoyah Station IRSF on July 18, 2000 (Amendment No. 257 to Facility Operating License No. DPR-77 and Amendment No. 248 to Facility Operating License No. DPR-79 for the Sequoyah Nuclear Plants, Units 1 and 3, respectively) from Watts Bar. 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 5 of 67 2.0 ENGINEERING CHANGE DESIGN CHANGE REQUEST OJECTIVES This Engineering Change Document Change Request (EC DCR) is prepared according to Exelon Procedure CC-AA-104 and referenced calculations prepared according to Exelon Procedure CC-AA-309. As updated NRC guidance applicable to extended storage of Radwaste at the LaSalle IRSF needed to be assessed against existing IRSF design and operational requirements, and, a larger number of Class B/C Radwaste containers is expected compared to original configurations, including additional Class B/C Radwaste containers from the sister nuclear units, a LaSalle EC DCR report was deemed necessary. 

This EC DCR report: 
: 1. Provides a review of the previous 50.59s against expected current requirements,  
: 2. Assesses original and current regulatory requirements,  
: 3. Characterizes expected Class B/C Radwaste from the shipping stations,  
: 4. Defines duration of storage,  
: 5. Assesses potential container flammable gas generation,  
: 6. Assesses container integrity for extended storage,  
: 7. Provides a Waste Acceptance Criteria for the Receiving Station 
: 8. Evaluates continuous effluent release monitoring,  
: 9. Considers EPRI Operational Guidelines,  
: 10. Provides a description of the existing IRSF,  
: 11. Evaluates design basis events,  
: 12. Documents that NRC physical security program requirements are met,  
: 13. Provides a container drop assessment,  
: 14. Updates shielding assessments,  
: 15. Evaluates Stations Container Decontamination Capability,  
: 16. Evaluates IRSF container repackaging capability,  
: 17. Evaluates liquid identification capability,  
: 18. Assesses IRSF annunciation to the Main Control Room, and  
: 19. Evaluates various operational considerations.  

A 50.59 Applicability Review Form is also included, as well as a review of necessary 

LaSalle Station UFSAR and Technical Specification changes. 

This EC DCR also supports a potential License Amendment Request (LAR) for LaSalle since the LaSalle Docket License will likely need revision to accept Radwaste from the other sister stations located in Illinois. 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 6 of 67 3.0 DISCUSSION OF PRIOR IRSF 50.59 SAFETY ASSESSMENTS Appendix A discusses LaSalle IRSF 50.59 assessments performed in 1992 and 1994 and presents observations and recommendations related to current extended storage efforts as well as potential future shipments of Radwaste to LaSalle IRSF shipped from Byron, Braidwood, and Clinton Stations. This evaluation has helped guide the current technical assessment activities and 50.59 review efforts for addressing extended storage and shipments of Class B/C Radwaste to LaSalles IRSF. 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 7 of 67 4.0 REGULATORY GUIDANCE 4.1 Regulatory Issue Summary (RIS) 2008-32, "Interim Low Level Radioactive Waste Storage at Reactor Sites", dated Dec. 30, 2008 This document was prepared by NRC to clarify the current NRC staff position regarding the long-term, interim storage of low level radioactive waste at facilities licensed under Title 10, Part 50, and to provide an acknowledgement of the proposed NEI/EPRI Guidelines for Operating an Interim On-Site Low-Level Waste Storage Facility, published in April 2008. The RIS delineates the evolution of NRC regulatory requirements, and, based on these predecessor documents, the RIS establishes a current NRC position for long-term on-site LLRW storage. The RIS summarizes prior NRC guidance documents that specifically address interim storage of LLRW on reactor sites. The pertinent guidance discussed herein, are: Generic Letter (GL) 81-38; SECY-94-198; and NUREG-0800 SRP 11.4A. The RIS also indicates that the EPRI Operating Guidelines for Interim LLRW Storage Facilities report is generally consistent with the RIS and other NRC guidance such as NUREG-0800.
Therefore, where applicable, this document is referred to in this Technical Report.
4.2 Generic Letter 81-38 NRC developed its existing guidance for the long-term storage of LLRW in the 1980's and early 1990's. The agency issued Generic Letter 81-38 "Storage of Low-Level Radioactive Wastes at Power Reactor Sites," which provides much of the substantive guidance related to the long-term storage of LLRW generated at nuclear power plants, on November 10, 1981. 

GL 81-38 was developed basically as a result of radwaste disposal reduction in the United States. In the 1970's there were six nuclear waste disposal sites located in 1)
Sheffield, Illinois; 2) Maxey Flats, Kentucky; 3) West Valley, New York; 4) Hanford, Washington; 5) Beatty, Nevada; and 6) Barnwell, South Carolina. Three of the sites were forced to close before 1980. In 1978, the facility in Sheffield, Illinois, was closed due to reaching maximum capacity. The Maxey Flats, Kentucky site was closed in 1977. The site in West Valley, New York site was closed in 1975. The Beatty, Nevada, site was closed twice in 1979 due to mishandling of the waste during transport and currently only accepts intra-compact waste. The Hanford, Washington, site was temporarily closed in 1979 as well and the Governor of South Carolina reduced the waste accepted at the 

Barnwell site in response to the Hanford closure. 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 8 of 67  GL 81-38, which is still in force except for its five year storage limitation, is considered the original staff guidance on the storage of low level radioactive waste and its requirements are addressed in this document. 

4.3 NUREG 0800 SRP 11.4A, "Design Guidance for Temporary Storage of Low-Level Radioactive Waste", Rev. 3 dated 2007 This document provides specific NRC guidance to licensees for increasing on-site Radwaste storage capacity. It is part of the NUREG-0800 Standard Review Plan, Revision 3, for the Review of Safety Analysis Reports for Nuclear Power Plants, Chapter 11.4, Solid Waste Management System. The LSCS-UFSAR Section 11.4.1.2 indicates that the IRSF was designed in accordance with the version of this appendix current at that time. The Generally Applicable Guidance, and the guidance for Stabilized Radioactive Waste Storage are considered applicable to the IRSF. Standard Review Plan 11.4 is for Solid Waste Management Systems in the short-term, and specifically refers to 11.4-A for long-term storage in IRSFs, as follows: 

Appendix 11.4-A provides guidance for applicants when considering onsite low-level radioactive waste storage capabilities for periods that may last several years but are significantly less than the life of the plant.
4.4 SECY-94-0198 Nuclear Regulatory Commission (NRC) Paper SECY-94-198 entitled Review of Existing Guidance Concerning the Extended Storage of Low Level Radioactive Waste was issued to the NRC Commissioners on August 1, 1994 by the Executive Director for Operations at NRC. The objective of this SECY document was to provide the Commission with results of a review of NRC guidance concerning on-site storage of LLRW and to inform the Commission of needed changes.  

Two key statements made by the NRC staff in the SECY are: 
: 1. The staffs primary intention in prescribing the five-year limit in 1981 (via GL 81-38) was to encourage the development of new (disposal) facilities. In retrospect, it has not been an important factor in achieving this objective. In the 

revised guidance (which was drafted to replace GL 81-38 and was appended to the SECY letter) the staff eliminated any language that implies that a five year term is a limit, beyond which storage would not be allowed, or which imposes any special review requirements. 
: 2. The SECY also stated that the staff continues to believe that placing waste into storage in a form suitable for disposal is desirable, but only if there is sufficient assurance that the waste form will be ultimately acceptable for disposal. Because that assurance is in many cases not available, the staff proposes to revise the TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 9 of 67  guidance to allow generators to have additional flexibility when waste form requirements for individual facilities have not been established. Use of HICs were confirmed by the staff as an acceptable method for long term storage, as long as radiolytic gas generation and the potential for embrittlement of 

polyethylene containers was addressed for extended storage pursuant to 

established NRC guidance. 

The SECY letter itself provide insight into the NRCs thinking at the time, but since the enclosure (Generic Letter on Extended Storage of Low-Level Radioactive Waste) was never issued as a final GL, no official NRC guidance resulted from SECY-94-198. The SECY was however referenced in Regulatory Issue Summary 2008-32  for guidance purposes, but the RIS 2008-32 provided specific relief from the five-year storage limit and referenced other guidance documents as appropriate for extended storage, including NUREG 0800, SRP 11.4-A issued in 2007. The RIS also provided specific guidance for 

extended storage not previously promulgated by NRC. 

4.5 IRSF Acceptance Criteria Primary NRC regulatory guidance for interim storage of radioactive waste was reviewed to identify criteria that need to be met for continued operation of the existing IRSF facility for an extended storage function and for receiving Class B/C Radwaste either from the LaSalle station of from sister stations in Illinois. 

Key NRC guidance documents evaluated were: 
: 1. Generic Letter 81-38 entitled Storage of Low-Level Radioactive Wastes at Power Reactor Sites,  
: 2. NUREG 0800 SRP 11.4-A entitled,  Design Guidance for Temporary Storage of Low-Level radioactive Waste and,  
: 3. Regulatory Issue Summary (RIS) 2008-32 entitled, Interim Low Level Radioactive Waste Storage at Reactor Sites.
A thorough review of these guidance documents resulted in a tabulation of LaSalle IRSF Regulatory Acceptance Criteria (Acceptance Criteria) that define the minimum essential design and operational requirements for the LaSalle IRSF (see Table 4.5-1). Many of the Acceptance Criteria are bounded by previous 50.59 assessment findings, but a number require new evaluation and analysis to confirm the criteria are adequately met. This EC DCR Report fully documents the acceptability of the existing IRSF for meeting the Acceptance Criteria and forms a licensing basis for applying for an LAR for the LaSalle 

station.
TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 10 of 67 Table 4.5-1 LASALLE IRSF REGULATORY ACCEPTANCE CRITERIA Applicable Regulatory

SourceGuidance/Regulatory Wording Acceptance Criteria and/or Expected Compliance Basis Licensing Basis Status ITEM 1:  IRSF Radiation Dose and Monitoring Considerations Standard Review Plan (SRP) 11.4-A - 

Generally Applicable Guidance 1 Generic Letter (GL) 81-38 

Generally Applicable GuidanceNRC Inspection Procedure65051 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 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. 
: 1. The parameters that will be used to assure that 40 CFR 190 dose criteria are met will be a combination of container contact dose rates, container placement, existing and supplementary shielding, and decay credit where applicable. 
: 2. Offsite nearest resident 40 CFR 190 dose rates will be maintained less that 0.01 mSv (1 mrem) per year recommendation, with 24/7 occupancy assumed. 
: 3. Site boundary or onsite locations where some occupancy by a member of the public could be postulated will consider a conservative occupancy and may consider actual other fuel cycle dose commitments in establishing dose acceptance criteria. Per the Midwest ODCMs, a value of 2.5 mR/yr is currently used as a limit for contained sources such as the IRSFs. 
: 4. Radiation levels in the vicinity of these radwaste storage facilities will be controlled by combinations of permanent and temporary shielding, container placement, and remotely controlled operations. Acceptance criteria are selected to minimize occupational dose commitments. Administrative RP controls will also be an important factor during waste handling and will be proceduralized to maintain occupational doses ALARA. For members of the public, design and controls are such that only a small fraction of 40 CFR 190 limits are used by these 

facilities. 
: 5. Exelons established programs for radiation protection, surveys, monitoring, labeling, and reports and record retention, in conformance with 10CFR20, will be used for these radwaste storage facilities. 
: 6. The historically and currently applied design limits for onsite locations outside of the LaSalle IRSF [inside the protected area] is about 1 mrem/hr.   
: 7. Separate limits remain to be established in the internal IRSF location such as the truck bay, during both typical storage conditions and during waste handling. These will be determined on an ALARA basis. Generally, New Dose Analyses Required because of waste characteristic changes.
The guidance is in keeping with the existing design of these 

facilities. No changes are planned in dose acceptance criteria for location outside of the LaSalle IRSF unless justified on an ALARA basis. Calculated doses must continue to meet regulatory limits.
Analyses will use existing dose assessments as either reference material or justified as bounding, where applicable.   

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 11 of 67 Table 4.5-1 LASALLE IRSF REGULATORY ACCEPTANCE CRITERIA Applicable Regulatory

SourceGuidance/Regulatory Wording Acceptance Criteria and/or Expected Compliance Basis Licensing Basis Status SRP 11.4-A - Gen Applicable Guidance 6A SECY-94-198 GL 81-38 Wet Radioactive

Waste Storage 

Section (c) 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. 
: 1. Radioactive isotopes in the radwaste stored in the LaSalle IRSF would generally be in a solid form, or dewatered resins contained with minimal dispersal mechanisms even in the event of a container leak.   
: 2. Failure mechanisms include: 
: a. Container degradation and leakage 
: b. Events, such as Dropped Container or Fire 
: 3. No outside storage is associated with the LaSalle IRSF.  
: 4. Container decontamination to the identified standards will be part of the Waste Acceptance Criteria for storage in these facilities.
New Evaluations Required because of differences in waste form, packaging and storage arrangement. Additionally, new evaluations will speak to regulatory guidance issued since the original 50.59 safety evaluations. Longer-term storage issues 

need to be addressed.   

Historically, accidentanalyses addressed container drops. PBAPS was for a bounding RWCU liner drop with radioactivity comparable to a current 100 R/hr contact liner.
Drop conservatively treated as an unconfined dry powder, with airborne fraction per NUREG-1320.
New evaluation required to determine whether this needs to continue to be a design basis event, whether any update is required, and whether the same approach could be used for the LSCS IRSF, where only a drop of a concrete solidified container 

has been performed. Add explicit defenses on no Fire design basis release, or perform a related accident analysis.
Regulatory

Information

Summary (RIS) 2008-32 GL 81-38 Wet Radioactive

Waste Storage 

Section (c) 5 When evaluating interim long-term on-site LLRW storage, Part 50 licensees must consider the applicability of the general design criteria listed in Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50, specifically Criteria 61, 63 and 64.
Criterion 61, Fuel Storage and Handling and Radioactivity Control, specifies that fuel storage and handling, radioactive waste and other systems that may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions.
Criterion 63, Monitoring Fuel and Waste Storage, states that appropriate systems shall be provided in fuel storage, radioactive waste systems, and associated handling areas to (1) detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions. Criterion 64, Monitoring Radioactivity Releases, specifies that there must be a method for monitoring the level of radioactivity in effluent release pathways and to the plant environs. 
: 1. 50.59 reviews and supporting technical assessments with respect to extended storage will be designed to assure adequate safety under normal and postulated accident conditions, including 

external events. 
: 2. The instrumentation systems to detect excessive radiation levels and initiate appropriate actions are a combination of fixed radiation and airborne activity monitors, supplemented by surveys. 
: 3. In general, no airborne radioactivity releases are anticipated from these facilities. The waste is in closed, filtered vent containers, with the radioactivity non-volatile. The combination of continuous air monitors, and effluent monitors when the HVAC system is in service, provides adequate monitoring. As stated in Section 7.4.4.1, the CAMs are a plus, but not required.
New Evaluations and Justification Required because of differences in waste form, packaging and storage arrangement. Some additional discussion on external events may be beneficial to record. Additional justification may be required with respect to continued adequacy of existing design features. SRP 11.4-A -
Gen Applicable Guidance 6G 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. Airborne activity is expected to be minimal for these closed and filtered-vented containers. However, design features are provided to monitor airborne particulate both in the truck bay (CAMs) and in facility HVAC exhausts. These aspects will be confirmed and detailed.
New Evaluations and Justification Required because of differences in waste form, packaging and storage arrangement, plus storage duration. Additional justification may be required with respect to continued adequacy of 

existing design features. 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 12 of 67 Table 4.5-1 LASALLE IRSF REGULATORY ACCEPTANCE CRITERIA Applicable Regulatory

SourceGuidance/Regulatory Wording Acceptance Criteria and/or Expected Compliance Basis Licensing Basis Status ITEM 2:  IRSF Design Considerations SRP 11.4-A Generally Applicable

Guidance 6B GL 81-38 Generally Applicable

Guidance Section (d) 2 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 water ingress is from external sources and remains uncontaminated by plant-generated radioactivity. The provision for liquid drainage collection will be described. The LaSalle IRSF has storage area drains, with leakage routed to a closed-end sump in the truck bay. Radioactive liquid would only be expected in the event of container failure. Sampling is expected to be by hand, with analysis done in station laboratory facilities. Collected liquid will be directed to Radwaste, if contaminated, or sent to a storm drain system if uncontaminated. These liquid collection systems comply with the ALARA principal. These aspects will be confirmed and detailed. Existing Design and Justification is adequate. SRP 11.4-A - Gen Applicable 

Guidance 5 The facility should include design features, in accordance with 10 CFR 20.1406, that would minimize, to the extent practicable, contamination of the waste facility and environment; facilitate eventual decommissioning; and minimize, to the extent practicable, the generation of extraneous radioactive waste. This requirement applies to storage facilities used to process and store liquid, wet, dry solid, and stabilized wastes. 
: 1. All IRSF operations must comply with RegGuide 8.8, latest revision. 
: 2. The IRSF must contain features for container decontamination. 
: 3. The IRSF must follow guidance found in RegGuide 4.21 Minimization of Contamination and Radioactive Waste Generation - Life Cycle Planning 
: 4. Waste destined for the LaSalle IRSF is decontaminated as required for offsite shipment as radioactive waste. 
: 5. If possible, the utilization of storage area epoxy coating protection suitable for decontamination will be verified. Existing Design and Justification is generally adequate. 

Some additional discussion may be required with LSCS IRSF truck bay activities requiring personnel entry given anticipated higher dose rates.
GL 81-38 General Information

SectionFacility design and operation should assure that radiological consequences of design basis events (fire, tornado, seismic event, flood) should not exceed a small fraction (10%) of 10 CFR Part 100, i.e., no more than a few rem whole body dose. 
: 1. Fire suppression systems will be considered, if evaluation determines an unacceptable fire hazard. 
: 2. IRSF must be designed to local building code tornado and seismic standards or be retrofitted to meet those standards. 
: 3. IRSF must be located in a non-flood event area or be retrofitted to mitigate flooding effects.
Additional analyses and justifications likely required with respect to design basis event selection and response. See, for instance, NRC document Risk-Informed Decision-Making for Nuclear Material and Waste Applications - Draft for Trial Use" MAY 11, 2005 For seismic, flood, and tornado events prior 50.59 assessments are expected to be effectively bounding.

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 13 of 67 Table 4.5-1 LASALLE IRSF REGULATORY ACCEPTANCE CRITERIA Applicable Regulatory

SourceGuidance/Regulatory Wording Acceptance Criteria and/or Expected Compliance Basis Licensing Basis Status ITEM 2:  IRSF Design Considerations SRP 11.4-A - Gen Applicable 

Guidance 3 GL 81-38 Generally Applicable

Guidance Section (c) If possible, the preferred location of the additional storage facility is inside the plants 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). 
: 1. The LaSalle IRSF is located in the facilitys protected area.  
: 2. Regardless of storage location a physical security program will be in-place. 
: 3. Features related to physical security are addressed separately and are considered as Safeguards Information under Exelons control The storage facility physically protected and provided with restricted access for radiation protection purposes. The Applicability Review Form (LS-AA-104-1001) identified potential programs that could be affected by the proposed changes including the Security Program and Radiation Protection Program. Completion of program reviews in accordance with the 50.59 procedure and the Resource Manual will be coordinated with responsible site organizations to determine potential program impacts.SRP 11.4-A -
Gen Applicable 

Guidance 4 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). 
: 1. Design basis accidents evaluation will include dropped containers 
: 2. The storage building design, crane control features, and related procedures by transfer or transport container vendors will allow a system that prevents accidental dropping of containers or dropping of equipment onto waste containers.
: 3. Appropriate Station procedures, such as those below, will be in-place that controls storage container crane movements.
LOP-WX-33%2C e v 001%2C ABNOR M LOP-WX-32%2C e v 010%2C IRSF G.Existing Design and Justification is adequate. SRP 11.4-A - Stabilized Waste 

Storage 4A 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. 
: 1. Evaluations to confirm that the LaSalle IRSF design is sufficiently robust such that seismic events will not adversely affect waste containment will be made. 
: 2. These facilities inherently provide curbing and drainage control function, and any spilled dewatered resins would be expected to be contained within the building. Existing Design and Justification is adequate. Expected response of existing designs to external events is being prepared to demonstrate no significant hazard.
TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 14 of 67 Table 4.5-1 LASALLE IRSF REGULATORY ACCEPTANCE CRITERIA Applicable Regulatory

SourceGuidance/Regulatory Wording Acceptance Criteria and/or Expected Compliance Basis Licensing Basis Status ITEM 2:  IRSF Design Considerations SRP 11.4-A - Stabilized Waste 

Storage 4D 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. 
: 1. Design basis events will be assessed and evaluated against appropriate regulatory criteria. 
: 2. A fire hazard assessment applicable to expected combustible material loading in the truck and storage bays will be performed. 
: 3. Personnel training and administrative procedures to ensure both control of radioactive materials and minimum personnel exposures will be in-place.
Existing Design is generally adequate.
Particular attention will be applied to HDPE combustibility issues.
GL 81-38 Introduction In general, it is preferable to store radioactive material in solid form. 
: 1. Liquid wastes are excluded from IRSF storage. 
: 2. Solid and wet wastes (dewatered resins) are authorized for IRSF storage. This waste must meet the definition of stabilized radioactive waste storage guidance from SRP 11.4-A.
Existing Design and Justification is adequate. 

SRP 11.4-A -
Stabilized Waste 

Storage 1 Stabilized radwaste for storage purposes is defined as waste that meets stabilized waste criteria for licensed facilities. For purposes of this document, resins or filter sludge dewatered to the above criteria are defined under this waste classification/criteria. Only stabilized radwaste is anticipated to be stored in the LaSalle IRSF as dewatered resins. Existing Design and Justification is adequate. 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 15 of 67 Table 4.5-1 LASALLE IRSF REGULATORY ACCEPTANCE CRITERIA Applicable Regulatory

SourceGuidance/Regulatory Wording Acceptance Criteria and/or Expected Compliance Basis Licensing Basis Status ITEM 3:  Waste Form and Packaging Considerations SRP 11.4-A - Gen Applicable 

Guidance 2 GL 81-38 Position III(b)
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 should be based on data that demonstrate minimal corrosion from the anticipated internal and external environment for a period well in excess of the planned storage duration. Container integrity after the period of storage should be sufficient to allow handling during transportation and disposal without container breach.
For proposed increases in storage capacity for more than five years (long-term), the application and review procedures will be pursuant to 10 CFR 30 with consideration of container integrity and retrievability, volume reduction, influence on state planning for disposal, and implications for de facto onsite disposal. 
: 1. Stored containers must have a design life in excess of credible storage periods to prevent container integrity degradation that would affect suitability for transport for disposal. High Integrity Containers are assumed to meet this criterion in that they are designed to maintain integrity in a disposal site for no less than 300 years. Poly-HICs required specific container integrity evaluation for extended storage. 
: 2. Stored containers placement and retrieval must be ALARA. 
: 3. Stored container contents must be prepared in accordance with a Station specific Process Control Program. 4. Consideration will be given to vendor information, material selection, coatings, etc. The protection provided by the storage facilities will serve to minimize effects of external weather condition 
: 5. 10 CFR 30.51 (Transfer of by Product Material), 30.53 (Test), and other 10 CFR 30 requirements, when applicable will be followed  
: 6. IRSF containers must be corrosion resistant and be compatible with the waste form. This may be achieved by material selection or suitable coatings. 
: 7. Containers for IRSF storage must be identified on an Approved Container List. The Approved Container List will identify the smallest practicable group of containers that been found to have certification of compliance, specification, or testing that ensures compliance with 10 CFR and 49 CFR transportation of radioactive materials requirements.
Existing Design and Justification, along with the approaches discussed to the left, is adequate.
EXELONs selection of 8-120 Poly-HICs needs to be evaluated against these acceptance criteria. SRP 11.4-A - Stabilized Waste Storage 4B If liquids exist in a corrosive form, licensees should implement proven measures to protect the container (i.e., special liners or coatings) and/or neutralize the excess liquids. If deemed appropriate and necessary, highly non-corrosive materials (e.g., stainless steel) should be used. Potential corrosion between the solid waste forms and the container should also be considered. In the case of dewatered resins, highly corrosive acids and bases can be generated, which will significantly reduce the longevity of the container. The PCP should implement steps to assure the above does not occur; provisions should be made to govern container material selection and precoating to ensure that container breach does not occur during temporary storage periods. 
: 1. Due to long-term storage, consideration will be given to the need for special liners or coatings or materials to prevent chemical attack. These aspects will be confirmed and detailed. 
: 2. Process Control Procedures, such as the following, will address package pH control, hazardous chemicals, incompatible materials, and other waste management issues; they will be verified and 

detailed.
RW-AA-10 "Radwaste Process D Existing Design and Justification, along with the approaches discussed to the left, is adequate. May also be incorporated into any license application with 

respect to trans-shipment of waste between stations.
TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 16 of 67 Table 4.5-1 LASALLE IRSF REGULATORY ACCEPTANCE CRITERIA Applicable Regulatory

SourceGuidance/Regulatory Wording Acceptance Criteria and/or Expected Compliance Basis Licensing Basis Status ITEM 3:  Waste Form and Packaging Considerations SRP 11.4-A - Gen Applicable 

Guidance 2 GL 81-38 Generally Applicable

Guidance Section III(b)
Gas generation from organic materials in waste containers can also lead to container breach and potentially flammable/explosive conditions. To minimize the number of potential problems, licensees should evaluate the waste form gas generation rates from radiolysis, biodegradation, or chemical reaction with respect to container breach and the creation of flammable or explosive conditions. Unless storage containers are equipped with special vent designs that allow depressurization and do not permit the migration of radioactive materials, resins highly loaded with radioactive material, such as BWR reactor water cleanup system resins, should not be stored for longer than approximately 1 year. 
: 1. All LaSalle IRSF stored containers will feature special vent designs, which allow depressurization and do not permit the migration of radioactive materials. 
: 2. Gas generation will be predicted and all containers will use filtered vents to prevent pressurization.  
: 3. Adequate venting of shield bells will also be considered, unless otherwise justified 
: 4. With the expected flammable gas generation rate an accumulation of flammable concentrations is conceivable in the containers or the shield bell. A bounding case for flammable gas release will be developed and is expected to be acceptable with adequate venting and building ventilation. Within the waste containers there are no ignition sources and a relatively damp environment. 
: 5. An accumulation of flammable gas and a static discharge, for instance during bell removal, is considered unlikely, but an evaluation of its significant may be required. Existing Design and Justification, along with the approaches discussed to the left , is adequate. SRP 11.4-A - Stabilized Waste 

Storage 4C GL 81-38 Solidified Radioactive

Waste Storage 

Section (d) 3 Provision should be made for additional reprocessing or repackaging due to container failure and/or, as required for final transporting and burial as per DOT and burial site criteria. Contamination isolation and decontamination capabilities should be developed. When significant handling and personnel exposure can be anticipated, ALARA methodology should be incorporated as per Regulatory Guides 8.8 and 8.10. 
: 1. A Waste Acceptance Criteria (WAC) document will be developed for the LaSalle IRSF. 
: 2. All waste destined for LaSalle IRSF storage must comply with the WAC. 
: 3. Provisions for container retrieval and removal to a designated reprocessing/repackaging area must be provided for stored waste, including ALARA considerations. 
: 4. The specifics of reprocessing/repackaging options for container failure and/or transportation/disposal compliance assessment after extended storage will be evaluated and best approaches identified. Storage containers will be assessed and dewatering verification will be evaluated on a case-by-case basis. 
: 5. The LaSalle IRSF provides inherent contamination control. Decontamination of storage areas could be required upon facility decommissioning, and suitable epoxy coatings are used to facilitate this effort (an attempt to verify their use in the storage area will be made). Container contamination will be controlled to regulatory limits at the time of preparation, and is not expected to increase unless container leaks should occur. Provisions for resurveying and decontamination, if necessary, will be described. 
: 6. ALARA evaluations are based on dose acceptance criteria inside and immediately outside of the building. Task specific handling and dose control process ALARA may need generation.
Additional discussion on reprocessing and repackaging is needed, including any justification of features not considered necessary. Clear Waste Acceptance Criteria are considered necessary with respect to trans-shipped waste. 

Existing design bases and justification related to contamination control appear adequate. 

Some more explicit ALARA evaluation may be required for the LSCS IRSF with respect to truck bay related 

activities. SRP 11.4-A -
Stabilized Waste 

Storage 2 Any storage plans should address container protection and any reprocessing requirements for eventual shipment and burial. The goal is to provide container protection that would assure that no reprocessing is required for eventual shipment and burial. The specific details of final examinations before shipping and contingencies for reprocessing will be verified and detailed. Additional discussion on reprocessing and repackaging is needed, including any justification of features not considered necessary.
GL 81-38 Wet Radioactive

Waste Storage Section (c) and SRP 11.4- A - Guidance is specific designed for waste that is in a liquid or slurry form. Dewatered waste to be stored in the IRSF is prepared in accordance with SRP 11.4-A - Section V Stabilized Waste Storage with free standing liquid limited in accordance with disposal site criteria. EXELON will not be storing wet waste in the IRSF per its Waste Acceptance Criteria guidelines. Thus free standing liquid will meet expected burial ground requirements. Over time due to the hygroscopic nature of resins, free standing liquids may not meet expected burial ground requirements and could require dewatering. Such a limited increase in free standing liquid will not have a significant impact on container design or effluent potential under normal, upset or accident conditions. Existing Design and Justification, along with the approaches discussed to the left , is adequate.
TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 17 of 67 Table 4.5-1 LASALLE IRSF REGULATORY ACCEPTANCE CRITERIA Applicable Regulatory

SourceGuidance/Regulatory Wording Acceptance Criteria and/or Expected Compliance Basis Licensing Basis Status ITEM 3:  Waste Form and Packaging Considerations Section IV Wet Radioactive 

Waste Storage GL 81-38 Generally Applicable Guidance Section (d) 5 Total curie limits should be established based on the design of the storage area and the safety measures provided. 

To Be Determined; however we expect to argue that this parameter is not a meaningful control. Controls will generally be based on container contact dose rates, container placement, and storage capacity. Existing Design and Justification, along with the approaches discussed to the left , is adequate.
Table 4.5-1 LASALLE IRSF REGULATORY ACCEPTANCE CRITERIA Applicable Regulatory

SourceGuidance/Regulatory Wording Acceptance Criteria and/or Expected Compliance Basis Licensing Basis Status ITEM 4:  IRSF Container Inspection Considerations SRP 11.4-A - Gen Applicable 

Guidance 2 GL 81-38 Generally Applicable

Guidance Section III(b) Licensees should implement a program providing for at least periodic (quarterly) visual inspections of container integrity (e.g., swelling, corrosion products, leaks, or breach). Inspections can be accomplished by the use of television monitors; by walkthroughs if storage facility layout, shielding, and container storage array permit; or by selecting waste containers that are representative of the types of waste and containers stored in the facility and placing them in a location specifically designed for inspection purposes. All inspection procedures developed should minimize occupational exposure. The use of high-integrity containers (300-year lifetime design) would permit an inspection program of reduced scope. 
: 1. The IRSF will utilize TV monitors for remote inspection or be capable of retrofitting to include this feature. 
: 2. IRSF inspections will be performed in accordance with the ALARA principal. 
: 3. The ability to visually inspect containers is provided by the radwaste storage facilities, but the logistics, identified concerns, and related inspection frequencies will need to be evaluated.  
: 4. Exelon Procedures RW-AA-102 (Radwaste Storage Facility/DAW Waste Container Inspections, 104 (Radwaste Storage Facility/Waste Container Inspections), 105 (Guidelines for Operating an Interim on Site Low Level Radioactive Waste Storage Facility) will be utilized for this purpose. 

RW-AA-102 "Radwaste Storage F Radwaste Storage Facility/Waste Contai RW-AA-105 "Guidelines for Opera Existing Design and Justification, along with the approaches discussed to the left , is adequate. SRP 11.4-A - Stabilized 

Radioactive

Waste Storage 3 GL 81-38 Section V (c) 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, tornadoes) unless otherwise justified. 
: 1. To the extent practical, only containers certified for radwaste transportation and disposal, and designed for an extended life without containment degradation (such as approved high integrity 

containers) will be used. 
: 2. Accident mitigation design and control procedures, generally and specific to the IRSF, address postulated waste handling and external events based on the safety analyses for these facilities. Existing Design and Justification, along with the approaches discussed to the left , is adequate. 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 18 of 67 Applicable Regulatory

SourceGuidance/Regulatory Wording Acceptance Criteria and/or Expected Compliance Basis Licensing Basis Status ITEM 5:  IRSF Inventory and Records Management Considerations GL 81-38 Applicable Guidance Section (d) 6SRP 11.4-A - Gen Applicable Guidance 6F Inventory records of waste types, contents, dates of storage, shipment, etc., should be maintained. 

Licensees should maintain inventory records by waste types, waste 

contents, radionuclides and radioactive material, dates of storage, shipment, and other relevant data. These types of data will be collected and records maintained. This will include typical manifest data forms and their required data. 

The IRSF system identifying the location, waste form and type, packaging, storage emplacement date, generating station, etc. must be operational. 

Existing Design and Justification, along with the approaches discussed to the left, is adequate. Standard shipping manifest data would be expected to meet this requirement.   

This data may also be used, by procedure, for decay in place 

credit. SRP 11.4-A -
Gen Applicable 

Guidance 6E 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). 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. Existing Design and Justification, along with the approaches discussed to the left , is adequate.
TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 19 of 67  

TABLE 4.5 - 1 Embedded Documents  (Printed for Convenience) 

Level of Use Reference 1 of 8 LOP-WX-33 Revision 1 March 19, 2008 ABNORMAL IRSF OPERATION PROCEDURE A. PURPOSE The purpose of this procedure is to provide generic instruction to maintain the Interim Radwaste Storage Facility (IRSF), and its systems, and the IRSF crane in a safe condition should any abnormal or unforeseen events occur.
B. REFERENCES 1. LOP-WX-32, IRSF General Use Procedure. 
: 2. LOP-WX-34, Guidance for the Placement of Containers in the Interim Radwaste Storage Facility.
C. PREREQUISITES 1. None. 

D. PRECAUTIONS 1. DO NOT carry loads over personnel. 
: 2. IF electrical power fails, THEN all breakers should be opened and all crane controls placed into the OFF position. 3. The load MUST NOT exceed the crane capacity (20 tons). 
: 4. During lifting operations, the hoist must be centered directly over the load. Side or end pulling is prohibited. 5. OPERATE crane controls in a smooth manner. DO NOT swing or jerk loads. IF the load swings, THEN STOP motion until the load has stopped swinging. 6. EXERCISE appropriate Radiation Protection precautions to prevent or mitigate potential contamination events. 7. Personnel in the IRSF should exercise good radiological ALARA practices to maintain their personnel dose. E. LIMITATIONS AND ACTIONS 1. Bridge and trolley motion are independent and can be performed simultaneously.
Level of Use Reference 2 of 8 LOP-WX-33 Revision 1 March 19, 2008  2. During crane operation, the focus of all TV cameras can be adjusted using the appropriate focus switch located on the respective camera control panel. 3. Bridge mounted TV cameras provided for area surveillance, and should be used as much as possible to ensure safe operations. 4. The bridge and trolley movements are restricted by limit switches to prevent inadvertent movement of a load into any of the building's four walls or shield wall. The BRIDGE & TROLLEY LIMIT SWITCH BYPASS is a key-lock 

switch (ON/OFF) located on the upper right hand portion of the Operator Console. When the switch is in the ON position, the interlocks associated with 

the bridge and trolley are de-activated , thus allowing the bridge and trolley to travel past their interlocked positions to their end travel points. EXERCISE 

caution when the BYPASS switch is in the ON position. The (Bridge, Trolley) coordinates of the four corners of the crane's working range are presented below.
The bridge and trolley coordinates can be viewed through cameras 4 and 5, respectively. Coordinates with Coordinates with      BYPASS SWITCH "ON" BYPASS SWITCH "OFF"   a. Northwest (0.0, 0.0) (3.0, 1.0)   b. Northeast (0.0, 53.0) (2.0, 52.5) 
: c. Southwest (115.5, 0.0) (113.5, 1.0) 
: d. Southeast (115.5, 53.0) (113.0, 52.5) 
: 5. The bridge, trolley, hoist, and speed controls are interlocked to prevent inadvertent movement of a load into the shield wall. The interlocks function as 

follows:   a. Hoist - Speed:  The bridge or trolley will not operate in fast speed while the hoist is out of the "nested position". The 

LOAD NESTED light on the Operator Console will 

be de-energized. b. Hoist - Bridge: 
: 1) Truck Bay - When the crane is in the truck bay and the hoist is lowered (i.e., the LOAD ABOVE SHIELD WALL light is de-energized), the bridge will not move South when near the shield wall. The South limit is Level of Use Reference 3 of 8 LOP-WX-33 Revision 1 March 19, 2008 approximately 18.25 to 19.75 when referring to the Bridge Grid coordinate as viewed on Camera 4. 2) Storage Area - When the crane is in the storage area and the hoist is lowered (i.e., the LOAD ABOVE SHIELD WALL light is de-energized), the bridge will not move North when near the shield wall. The North limit is approximately 33.25 when referring to the Bridge Grid coordinate as viewed from Camera 4. c. Hoist - Trolley: When the crane is near the shield wall (i.e., the bridge interlock has been activated) and the hoist is lowered (i.e., the LOAD ABOVE SHIELD WALL light is de-energized), the trolley cannot move out 

of the center slot position in the shield wall (i.e.,
East-West movement is prevented when the 

TROLLEY ON CENTER SLOT light is energized). 6. IF IRSF Fire Detectors alarm the IRSF Ventilation will automatically shutdown. 

==F. PROCEDURE==

CAUTIONIf a concern with possible airborne contamination exists or any containers will be moved, SECURE IRSF Ventilation in accordance with LOP-WX-32, IRSF General Use Procedure.
NOTEAny number of abnormal events/anomalies are possible, therefore contingency actions for general categories should 

exist to assist the crane operator in handling the event and maintaining or restoring the IRSF to its normal condition. In 

all cases, NOTIFY the Radwaste S pecialist or designee of the abnormal event, and if necessary the Radiation Protection Shift Supervisor. 1. Area Radiation Monitor Alarm. a. At any time during the movement of containers in or out of the storage area spurious alarms may occur. The Radiation Protection Technician 

should verify conditions.
Level of Use Reference 4 of 8 LOP-WX-33 Revision 1 March 19, 2008   b. Once the container has been moved and the higher dose rate condition no longer exists, the RPT should reset the ARM alarm. c. If the alarm will not reset, and the alarm is not due to equipment malfunction, the RPT should assess the source of the radiation problem. d. If a dose problem does exist the Radiation Protection Shift Supervisor and the Radwaste S pecialist shall be notified. Also non essential personnel should be evacuated from the IRSF. e. If any container movements are required the Radwaste S pecialist must be informed and concur with any movement of containers.  (Container movements shall be in accordance with References 1+2)  2. Continuous Air Monitor Alarm. 
: a. During the movement of containers in or out of the storage area the background radiation levels may cause spurious alarms. The RPT in 

attendance should verify this condition. b. After the container move is complete and background radiation levels are no longer a concern the alarm shall be reset. c. If after a period of time the alarm will not clear, a confirmatory air sample should be taken to ensure no airborne situation exists. d. If a problem does exist, the Radiation Protection Shift Supervisor, Shift Manager , and the Radwaste S pecialistshall be informed of the situation. e. All non essential personnel shall exit the IRSF, and actions should be established to recover the IRSF. 3. IRSF Sump Overflow. 
: a. The Radiation Protection Shift Supervisor and the Radwaste S pecialist shall be notified of the condition. b. If storage area sump (curbed sump) is in overflowing condition refer to LOA-0PL01J-M203 and follow actions listed. c. For the truck bay sump the following should be completed. 
: 1) Radiation Protection Technician should verify contamination levels.
Level of Use Reference 5 of 8 LOP-WX-33 Revision 1 March 19, 2008    2) Arrangements should be made to empty the sump. Since this sump is not equipped with pumps the water will have to be barreled or placed into similar container. 3) Once removed from the sump Chemistry should be contacted to perform isotopic analysis on the water to determine proper disposal of the material.
NOTE- Curbed sump in the storage area sump which its only input is from floor drains (4) in the storage area. - Non-curbed sump in the truck bay sump and its only input is from floor drains (2) located in the truck bay 

area. 4) If it is determined that it is possible that a container could be the source, the Radwaste Specialist or designee shall concur with any further plans to move or remove any containers within the storage area. d. A determination should be made as to the cause of the input/overflow (there are normally no inputs to these sumps). 4. IRSF Crane failure. 
: a. In the event that the IRSF crane fails to operate with a container attached to the crane, the Radwaste Specialist shall be notified of the condition. b. The Radiation Protection Shift Supervisor shall be notified to assess radiological conditions. c. Due to the vast number of situations that could occur, Radiation Protection, Radwaste, and applicable maintenance departments should formulate an action plan to be taken prior to any attempt to recover the 

container.
Level of Use Reference 6 of 8 LOP-WX-33 Revision 1 March 19, 2008 NOTEIf the bridge needs to be moved back to the north wall there is 

a hand crank located in the truck bay of the IRSF which will 

pull the bridge back to the north wall. d. As dictated by dose rates, personnel should establish boundaries outside the IRSF, if necessary, and also evacuate the IRSF. 5. Container Drop Accident. 
: a. Radwaste Specialist and the Radiation Protection Shift Supervisor shall be notified of situation and the location of the container. b. As dictated by radiological conditions, personnel should leave the IRSF affected areas, and establish boundaries outside the building. c. Container integrity should be checked with the cameras on the crane to determine the urgency of the situation. d. Once the situation is stable Radwaste, Radiation Protection and any additional parties should concur on any further actions due to the exposure 

potential, and possible radiological consequences for incorrect actions. 6. Unstable Stack of Containers in the Storage Area. 
: a. This has the potential to be a very significant problem due to the location of the containers and the dose rates involved with the correction of a stack 

that has toppled. b. IF a stacking concern was noticed during container placement, the container being set into place SHALL be returned to its previous stable location. No further movements shall occur without the concurrence of 

the Radwaste Coordinator or designee. c. IF the stacking concern was noticed while retrieving a container. 
: 1) CONTACT Radwaste S pecialist. 2) If the crane is supporting the container then, refer to LOP-WX-34, Guidance for the Placement of Containers in the Interim Radwaste Storage Facility, for placement in another location.
NOTE Level of Use Reference 7 of 8 LOP-WX-33 Revision 1 March 19, 2008 Consideration should be made to question was that container being used for shielding?    3) IF the concern was noticed prior to the crane supporting the container then STOP all crane movement until concurrence is received from the Radwaste S pecialist. 4) PRIOR to the movement of the containers to new locations LOP-WX-34, Guidance for the Placement of Containers in the Interim Radwaste Storage Facility, MUST be referred to for proper 

location selection. 5) Containers after being unstacked should be inspected to determine cause of stacking problem prior to placing container back into 

storage, the containers integrity and stackability shall be verified 

and corrected if need be. 7. FIRE in the IRSF. 
: a. CALL Emergency Number 2211 and notify of fire, location, conditions, etc. b. SECURE IRSF ventilation system. 
: c. Support Fire fighting efforts as needed. 
: d. CONTACT Radiation Protection Shift Supervisor and the Radwaste S pecialist. 8. Degraded Container. a. If leaking or damaged container is noticed the Radwaste S pecialist and the Radiation Protection Shift Supervisor SHALL be notified of conditions. b. Prior to the movement of any containers the Radwaste Specialist ordesignee and Radiation Protection shall agree on the methodology to be used. c. When the container is retrieved caution should be used to prevent spreading contamination. d. If the container was the top container, thought should be given to the fact the container may have been being used for shielding with respect to Level of Use Reference 8 of 8 LOP-WX-33 Revision 1 March 19, 2008 skyshine. If container is not going to be placed back into the same 

location another container should be used for shielding. 9. RESTORE the IRSF as follows: 
: a. VERIFY crane is parked in the truck bay. 
: b. PLACE IRSF ventilation system back into operation. 
: c. IF posted, THEN SECURE Fire Watch in the truck bay. d. CHECK that Continuous Air Monitor is still on and no alarm condition exists. e. IF there was any change in container locations, additions or deletions to the storage area. ENSURE Radiation Protection has verified the dose rate at the IRSF exterior walls to be <1 mR/hr as described in Reference 2, Attachment C.
G. CHECKLISTS 1. None.
H. TECHNICAL SPECIFICATION REFERENCES 1. None.
Level of Use Reference 1 of 32 LaSalleStation UNIT COMMON OPERATING PROCEDURE IRSF GENERAL USE PROCEDURE LOP-WX-32 Revision 10 March 18, 2008 HINK CT EVIEW S T A R TOPProcedure Responsibility/Review/Approval RequirementsResponsible Department Head:SOS Minimum Review Type:TR Required Cross-Discipline Review(s):N/A Approval Position Required:SOS Specific Requirements: 
: 1. None.
Level of Use Reference 2 of 32 LOP-WX-32 Revision 10 March 18, 2008 TABLE OF CONTENTS A. PURPOSE
............................................................................................................................2 B. PREREQUISITES
...............................................................................................................2 C. PRECAUTIONS
..................................................................................................................2 D. LIMITATIONS
....................................................................................................................2

==E. PROCEDURE==

......................................................................................................................2E.1 Lifting, Weighing, Surveying or Transporting a Container.....................................2 E.2 Movement of Containers in the Storage Area.........................................................2 E.3 Retrieving a Container from the Storage Area........................................................2 F. REVIEW AND SIGNOFF
...................................................................................................2 G. REFERENCES
....................................................................................................................2 ATTACHMENTSA IRSF Crane Controls And Indication.......................................................................2 B IRSF Ventilation System Operation........................................................................2 C IRSF Crane Startup Checkout..................................................................................2 FIGURE1 IRSF Layout.............................................................................................................2 Level of Use Reference 3 of 32 LOP-WX-32 Revision 10 March 18, 2008 IRSF GENERAL USE PROCEDURE A. PURPOSE A.1  Objective   The purpose of this procedure is to provide direction for the normal operation of the Interim Radwaste Storage Facility (IRSF).
B. PREREQUISITES NOTE Specific Prerequisites are addressed in each of the scenarios covered by this procedure. The prerequisites listed below are administrative controls that must be adhered to during IRSF 

operation. B.1  When transferring a container to the IRSF, ensure that:  B.1.1  All transfers occurring totally within the Owner Controlled Area are regulated by the site radiation safety procedures. B.1.2  The cask/process shield transported on-site will be secured to the conveyance. B.2  The Radiation area boundary (as defined by 10CFR20) outside the IRSF must be controlled whenever a container with radioactive contents is moved by the crane within the truck bay. B.3  A Fire Watch must be posted whenever a truck cab is in the truck bay.  (An Operating EA can perform this function as part of operations within 

the IRSF.)  B.4  Truck bay diesel exhaust fan must be OFF at all times.
B.5  All IRSF ingress/egress (exterior building doors) must be closed, except for personnel access. The Ventilation system must be OFF when handling 

waste packages. B.6  If required, a shield liner (less than 15 R/hr on contact) should be available for placement on top of another liner, for any sky shine shielding 

purposes.
Level of Use Reference 4 of 32 LOP-WX-32 Revision 10 March 18, 2008  B.7  IRSF Ventilation system must be turned off prior to opening the cask/process shield. Ventilation shall not be turned on until all container movement/placement has ceased and the Continuous Air Monitor (CAM) in the Truck bay/Mezzanine area is operating and not alarming. B.8  Radiation Protection shall be cognizant of any radioactive container moves in the IRSF. B.9  All personnel MUST attend a pre-job briefing which must include descriptions and discussion of all planned crane moves applicable 

ALARA practices, and cask handling safety practices. B.10  Assembly and disassembly of casks shall be in accordance with approved vendor procedures. B.11  The requirement to weigh or survey a container shall be determined by the Radwaste Specialist or designee. B.12  Prior to retrieving a container from the storage area for loading into a shipping cask, verify with the Radwaste Specialist , or designee, that the shipping cask configuration will accommodate the container. B.13  Grapple installation of testing shall be in accordance with approved vendor procedures.
C. PRECAUTIONS C.1  DO NOT carry loads over personnel.
C.2  The load MUST NOT exceed the crane capacity (20 tons).
C.3  During lifting operations, the hoist must be centered directly over the load.
Side or end pulling is prohibited. C.4  ENSURE loads are properly and safely on the lifting device.
C.5  Crane controls shall be operated to their full motion, due to the possibility of crane damage if controls are held at a mid position.  (Referring to 

Bridge, trolley and hoist joy sticks.)  C.6  EXERCISE appropriate radiation protection precautions to prevent or mitigate potential contamination events. C.7  OPERATE the crane controls in a smooth manner. DO NOT swing or jerk loads. If the load swings, STOP motion until the load has stopped swinging.
Level of Use Reference 5 of 32 LOP-WX-32 Revision 10 March 18, 2008  C.8  Grapple MUST be parallel with crane rails before nesting load. C.9  When disconnecting grapple from the crane ensure grapple power cable is disconnected prior to lifting crane block. C.10 Both of the following methods shall be used to control crane movements at all times:  A spotter/signalman, who can see the load and crane path, is in constant communication with the crane operator using a dedicated source of communication. The crane operator has sufficient cameras to ensure the crane movement will NOT result in hitting any equipment or shield walls in the area.
D. LIMITATIONS D.1  Bridge and trolley motion are independent and can be performed simultaneously. D.2  During crane operation, the focus of all TV cameras can be adjusted using the appropriate focus switch located on the respective camera control 

panel. D.3  Bridge mounted TV cameras are provided for area surveillance, and should be used when possible to ensure safe operation. D.4  The bridge and trolley movements are restricted by limit switches to prevent inadvertent movement of a load into any of the building's four 

walls or shield wall. The (Bridge, Trolley) coordinates of the four corners of the crane's working range are presented below. The bridge and trolley coordinates can be viewed through cameras 4 and 5, respectively.  (See 

Figure 1). Coordinate Limit with BYPASS SWITCH"ON" Coordinate Limit with BYPASS SWITCH "OFF" Northwest (0.0, 0.0) (3.0, 1.0) 

Northeast (0.0, 53.0) (2.0, 52.5) 

Southwest (115.5, 0.0) (113.5, 1.0) 

Southeast (115.5, 53.0) (113.0, 52.5)
Level of Use Reference 6 of 32 LOP-WX-32 Revision 10 March 18, 2008  D.5  The bridge, trolley, hoist, and speed controls are interlocked to prevent inadvertent movement of a load into the shield wall. The interlocks 

function as follows:  D.5.1  Bridge:
D.5.1.1  Truck Bay - When the crane is in the truck bay and the trolley is not on the center slot (i.e. TROLLEY ON CENTER SLOT light is de-energized) the bridge will not move south near the shield wall. The south limit is approximately 18.25 to 

19.75 when referring to the bridge coordinates as viewed on camera 4. D.5.1.2  Storage - When the crane is in the storage area and Area  the trolley is not on the center slot (i.e. TROLLEY ON CENTER SLOT light is de-energized) the bridge will not move north near the shield wall. The north limit is at approximately 33.25 when referring to the bridge coordinates as viewed in camera 4. D.5.2  Trolley: If the bridge is interlocked out as described in 5.1.1 or 5.1.2 above, trolley motion east or west is prohibited. D.5.3  Hoist - Speed: The bridge or trolley will not operate in fast speed while the hoist is out of the nested position. The LOAD NESTED light will be 

de-energized. D.5.4  Hoist - Bridge:
CAUTIONThe "LOAD ABOVE SHIELD WALL" light was for a design container and grapple proposed to be used in the IRSF, this 

should NOT BE used for current containers being stored. The LOAD NESTED LIGHT should be energized when passing 

containers through to the storage area. D.5.4.1  Truck Bay - When the crane is in the truck bay and the trolley is on center slot (i.e. TROLLEY ON CENTER SLOT light is 

energized), if you were to lower the load and clear the "LOAD ABOVE SHIELD WALL" light, the bridge will not move south near the shield wall. The south limit is approximately 18.25 to 19.75 when referring to the bridge coordinates as viewed in   camera 4.
Level of Use Reference 7 of 32 LOP-WX-32 Revision 10 March 18, 2008  D.5.4.2  Storage - When the crane is in the storage area and    Area  the trolley is on center slot (i.e. TROLLEY ON CENTER SLOT light is energized), if you were to lower the load and clear the "LOAD ABOVE SHIELD WALL" light, the bridge will not move north near the shield wall. The north limit is approximately 33.25 when referring to the bridge coordinates as viewed in camera 4. D.6  The BRIDGE & TROLLEY LIMIT SWITCH BYPASS is a key switch located on the upper right hand portion of the Operator Console. D.6.1  When the switch is in the ON position, the interlocks associated with the bridge and trolley are de-activated , thus allowing the bridge and trolley to travel past their interlocked positions to their end travel points. D.6.2  EXERCISE caution when the BYPASS switch is in the ON position. With the limits bypassed and a container on the crane it will be possible to hit the shield wall, extreme caution should be exercised. D.7  Smearable contamination on waste containers must not exceed 50,000 dpm/100 cm 2 per the Safety Evaluation (reference 6). Containers must be decontaminated to less than 50,000 dpm/100 cm 2 prior to storage in the IRSF.
NOTES[1] Section E contains individuals scenarios for different evolutions with regard to movement of containers in 

the IRSF. Each scenario contains specific prerequisites to be performed in addition to the administrative prerequisites in Section B of this procedure. [2] The crane operator should be familiar with the crane and its controls, and the IRSF Ventilation system by reading Attachments A and B of this procedure. D.8  Stackable HICs MUST be stacked within 1/2" of center, in relation to a HIC below it.
Level of Use Reference 8 of 32 LOP-WX-32 Revision 10 March 18, 2008  D.9  Continuous airborne monitors are provided in the following two locations:
IRSF Truck Bay and the Ventilation System - on the mezzanine level near 

the exhaust louver. D.9.1  If the IRSF Truck Bay CAM specified in Step D.9 is NOT operating, a portable CAM providing gross radioactive airborne contamination indication and alarm capability will be placed in the IRSF Truck Bay.
The portable CAM is required to be operating when activities with the potential to change radiological conditions are being performed in the 

IRSF. D.10Area radiation monitors are provided in the following two locations: IRSF Truck Bay and IRSF Control Room. D.10.1If any of the Area Radiation Monitors in Step D.10 are not operating, then an RP qualified individual will be required to be present when activities are being performed in the IRSF with the potential to change radiological 

conditions. The RP qualified individual shall have a radiation dose rate monitoring device and shall have positive control over the activities within 

the area.
Level of Use Reference 9 of 32 LOP-WX-32 Revision 10 March 18, 2008 

==E. PROCEDURE==
E.1  Lifting, Weighing, Surveying, or Transporting a Container E.1.1  VERIFY the following prerequisites are complete:  E.1.1.1  The main power disconnect for the crane on the IRSF Crane Starter Panel is CLOSED. E.1.1.2  IRSF Ventilation secured (per Attachment B).
E.1.2  PERFORM and DOCUMENT in IRSF log book, proper operation of the IRSF crane.  (If it has not already been previously performed that shift.)  E.1.3  If required, DISASSEMBLE the cask. E.1.4  If required, INSTALL a weighing device on the crane. E.1.5  ATTACH the crane with the weighing device to the container, as applicable. E.1.6  LIFT the container to obtain weight, as applicable.
E.1.7  If required, PERFORM a radiological survey of the bottom of the container. E.1.8  RECORD the weight of the container and the radiological survey results , as applicable. E.1.9  If applicable, RETURN the container to the vehicle to remove the weighing device. E.1.10  If applicable, REMOVE weighing device. E.1.11  If required INSTALL the grapple. E.1.12  As applicable, TEST the grapple.
E.1.13  ATTACH the grapple to the container, as applicable.
E.1.14  VERIFY the grapple is fully engaged, as applicable.
E.1.15  LIFT the container, using the crane with the grapple, as applicable.
E.1.16  If required, POSITION the container to enable a radiological survey. E.1.17  As applicable, PEFORM a radiological survey.
E.1.18  If the container is to be stored in the IRSF, REFER to Step E.2.
Level of Use Reference 10 of 32 LOP-WX-32 Revision 10 March 18, 2008  E.1.19  If the container is not to be stored in the IRSF, RETURN the container to a vehicle. E.1.20  As applicable, REASSEMBLE the cask.
E.1.21  RESTORE the IRSF as follows:
E.1.21.1 VERIFY the crane is parked in the truck bay.
E.1.21.2 PLACE the IRSF ventilation system back into operation (per Attachment B). E.1.21.3 SECURE the Fire Watch that was posted in the truck bay.  (as applicable).
E.1.21.4 OPEN the main power disconnect for the crane on the IRSF Crane Starter Panel. E.1.22  Transporting a Container to the Storage Area
. E.1.22.1 VERIFY the following prerequisites are complete:
E.1.22.1.1  The main power disconnect for the crane on the IRSF Crane Starter Panel is CLOSED. E.1.22.1.2  The IRSF Ventilation is secured (per Attachment B).
E.1.22.2 PERFORM and DOCUMENT in IRSF log book, proper operation of the IRSF crane.  (If it has not already been previously performed that shift.)  E.1.22.3 OBTAIN LOP-WX-34 Attachment B, Container Placement and Movement Sequence from the Radwaste Coordinator or designee. E.1.32.4 VERIFY that the appropriate Storage Location and Level is vacant.
NOTEPreviously stored containers may need to be moved to place container per E.3. E.1.22.5 As applicable, REMOVE the crane hook, INSTALL the appropriate container attachment device on the crane block, and test, as applicable. E.1.22.6 From the IRSF Crane Operator Console and using the truck bay camera, POSITION the attachment device over the container. E.1.22.7 ATTACH the lifting device to the container, and VERIFY the lifting device is fully engaged, as applicable.
Level of Use Reference 11 of 32 LOP-WX-32 Revision 10 March 18, 2008 NOTEThe crane operator must be aware of Height Level indication 

which corresponds to the point where the container touches bottom. Refer to Attachment A for a description of the Height 

Level Indicator. E.1.22.8 The Container contamination level shall be checked (if not previously done) and logged in the IRSF logbook:  E.1.22.8.1  If
 < 10,000 dpm/100cm 2 proceed with storage of the container. E.1.22.8.2  If
 > 10,000 dpm/100cm 2 but < 50,000 dpm/100 cm 2 the Radiation Protection Shift Supervisor and Radwaste Specialist shall be notified. Any further operation with the respective container will be with their concurrence. E.1.22.8.3  If
 > 50,000 dpm/100cm 2 , REPLACE cask lid and contact the Radiation Protection Shift Supervisor and Radwaste Specialist
.The container must be deconned prior to placing it in storage.
NOTES[1] Refer to Figure 1 for IRSF configuration and bridge/trolley coordinates. [2] MONITOR Height indication when operating the hoist. E.1.22.9 From the IRSF Crane Operator Console and using the appropriate cameras, PERFORM the following:  E.1.22.9.1  VERIFY the BRIDGE & TROLLEY LIMIT SWITCH BYPASS is OFF. E.1.22.9.2  VERIFY the GRAB BYPASS switch is OFF.  (The red GRAB BYPASS light will be de-energized.)
Level of Use Reference 12 of 32 LOP-WX-32 Revision 10 March 18, 2008  E.1.22.9.3  LIFT and TRANSPORT the container through the notch in the IRSF wall as follows:  E.1.22.9.3.1  Using the Hoist Control Joystick, RAISE the container to the nested position. The LOAD NESTED, LOAD ABOVE SHIELD WALL and LOAD ON HOOK lights will illuminate. The SLACK 

CABLE light will de-energize. E.1.22.9.3.2  Using the Trolley Control Joystick, MOVE the trolley west until aligned with the notch centerline as viewed in camera No. 5. E.1.22.9.3.3  Using the Bridge Control Joystick, MOVE the bridge south. As the Bridge approaches the shield wall the TROLLEY ON CENTER light will illuminate, which at this point prevents a trolley movement while in the notch. E.1.22.9.3.4  Using the Bridge Control Joystick, MOVE the bridge south through the notch. After passing through the notch, (the Trolley 

on Center Slot Light will de-energize) the bridge, trolley and hoist 

controls will be used to deliver the container to the designated grid 

location.  (Subject to the interlocks described in Section D of this 

procedure.)
NOTE If placing a container in the storage area with Bridge coordinate 31.5' or  115.5', it may be necessary to use the Bridge/Trolley Bypass Switch. If bypass switch is used, extreme caution should be exercised in placement of container. E.1.23  From the IRSF Crane Operator Console and using the appropriate cameras, PERFORM the following:  E.1.23.1 Using the Bridge and Trolley Control Joysticks, PROCEED to the appropriate grid location. E.1.23.2 While monitoring the Height Level indication and using the hoist control joystick, carefully LOWER the container to the designated grid location and level. The SLACK CABLE light will illuminate when the load is placed into position. The LOAD ABOVE SHIELD WALL Light will 

de-energize. E.1.23.3 RELEASE the container. E.1.23.3.1  VERIFY the Load on Hook light de-energizes.
Level of Use Reference 13 of 32 LOP-WX-32 Revision 10 March 18, 2008  E.1.24  While monitoring Height Controller indication, RAISE the hoist to the nested position. The LOAD NESTED and LOAD ABOVE SHIELD WALL lights will illuminate and the SLACK CABLE light will 

de-energize. E.1.25  DOCUMENT container placement as follows:
E.1.25.1 In the IRSF log book, LOG container location, container number, stored by, date stored, and comments (if any). E.1.25.2 MOVE the tag for the container to the appropriate Tag Board location in the IRSF Crane Operator Control Room corresponding to the grid 

coordinate. E.1.26  If additional container moves are required continue per E.2.5 thru E.2.14. E.1.27  Using the bridge and trolley controls as was performed to move the hoist to the designated location, RETURN the hoist to the truck bay through the 

notch as follows. E.1.27.1 Using the Bridge and Trolley Control Joysticks, RETURN the hoist to the approximate vicinity of the notch. E.1.27.2 Using the Bridge and Trolley Control Joysticks, MOVE the bridge and trolley as necessary to obtain the TROLLEY ON CENTER light. E.1.27.3 Using the Bridge Control Joystick, MOVE the bridge north through the notch. After passing through the notch, POSITION the crane to install the 

cask lid. E.1.28  RESTORE the IRSF as follows:
E.1.28.1 VERIFY the crane is parked in the IRSF truck bay.
E.1.28.2 PLACE the IRSF Ventilation System back into operation (per Attachment B). E.1.28.3 SECURE Fire Watch if used.
E.1.28.4 OPEN the main power disconnect for crane on the IRSF Crane Starter Panel. E.1.28.5 NOTIFY Radiation Protection to verify and log in the IRSF logbook that the dose rate is less than 1 mrem/hr at the outside surface of the IRSF 

walls. If dose rates exceed 1 mrem/hr NOTIFY the Radwaste S pecialist.
Level of Use Reference 14 of 32 LOP-WX-32 Revision 10 March 18, 2008 NOTES[1] Refer to Figure 1 for IRSF configuration and bridge/trolley coordinates. [2] MONITOR Height indication when operating the hoist. E.2  Movement of Containers in the Storage Area E.2.1  VERIFY the following prerequisites are complete:  E.2.1.1  The main power disconnect for the crane on the IRSF Crane Starter Panel is closed. E.2.1.2  The IRSF Ventilation secured.  (Attachment B)
E.2.2  PERFORM and DOCUMENT in IRSF log book, proper operation of the IRSF crane.  (If it has not already been previously completed that shift). E.2.3  PERFORM LOP-WX-34 concurrently with steps E.2.3 through E.2.12.
E.2.4  From the IRSF Crane Operator Console and using the appropriate cameras, PERFORM the following:  E.2.4.1  VERIFY the BRIDGE & TROLLEY LIMIT SWITCH BYPASS is OFF.
E.2.4.2  If not already in place, ATTACH lifting device to crane block, and test as applicable. E.2.4.3  VERIFY the GRAB BYPASS switch is OFF.  (The red GRAB BYPASS light will de-energize)  E.2.4.4  Using the Bridge and Trolley Control Joysticks, PROCEED to the designated grid location as follows:  E.2.4.4.1 Using the Hoist Control Joystick, RAISE the hoist to the nested position. The LOAD NESTED, LOAD ABOVE SHIELD WALL and LOAD ON HOOK lights will illuminate. The SLACK CABLE light will de-energize. E.2.4.4.2 Using the Trolley Control Joystick, MOVE the trolley west until aligned with the notch centerline as viewed in camera No. 5. E.2.4.4.3 Using the Bridge Control Joystick, MOVE the bridge south. As the Bridge approaches the shield wall the TROLLEY ON CENTER light will illuminate, which at this point prevents a trolley movement while in the 

notch.
Level of Use Reference 15 of 32 LOP-WX-32 Revision 10 March 18, 2008  E.2.4.4.4 Using the Bridge Control Joystick, MOVE the bridge south through the notch. After passing through the notch, (the Trolley on Center Slot light 

will de-energize) the bridge, trolley, and hoist controls will be used to move the center of the designated grid location (subject to the interlocks 

described in Section D of this procedure).
NOTE If placing a container in the storage area with Bridge coordinate 31.5' or 115.5', it may be necessary to use the Bridge/Trolley Bypass Switch. If bypass switch is used extreme caution should be exercised in placement of container. E.2.5  VERIFY that the appropriate Storage Location and Level is vacant. E.2.6  VERIFY the container number to be moved is correct as indicated on Attachment B of LOP-WX-34, Guidance for the Placement of Containers in the Interim Radwaste Storage Facility. NOTIFY the Radwaste Specialist or designee if any container is not in the expected position. E.2.7  While monitoring the Height Level Indicator indication and using the hoist control joystick, carefully LOWER the hoist to the specified container. The LOAD NESTED and LOAD ABOVE SHIELD WALL lights will de-energize. The SLACK CABLE light will illuminate. E.2.8  ATTACH the attachment device to the load, and VERIFY the attachment device is fully engaged. E.2.9  RAISE the container to the nested position. The LOAD NESTED, LOAD ABOVE SHIELD WALL and LOAD ON HOOK lights will illuminate.  

The SLACK CABLE light will de-energize. E.2.10  Using the Bridge and Trolley Control Joysticks, PROCEED to the new designated grid location. E.2.11  While monitoring the Height Level Indicator indication and using the hoist control joystick, carefully LOWER the container to the designated grid location and level. The SLACK CABLE light will illuminate when the load is placed into position. The LOAD ABOVE SHIELD WALL 

light will de-energize. E.2.12  RELEASE the container.
Level of Use Reference 16 of 32 LOP-WX-32 Revision 10 March 18, 2008  E.2.13  While monitoring the Height Controller indication, RAISE the hoist to the nested position. The LOAD NESTED and LOAD ABOVE SHIELD WALL lights will illuminate and the SLACK CABLE light will 

de-energize. E.2.14  DOCUMENT container placement as follows:
E.2.14.1 In the IRSF Logbook, LOG container number, location and level, previous location, date, moved by, and comments (if any). E.2.14.2 MOVE the tag for the container to the appropriate Tag Board location in the IRSF Crane Operator Control Room corresponding to the new grid 

coordinates. E.2.15  REPEAT Steps E.2.5 through E.2.14 until all required container movements are complete. E.2.16  Using the Bridge and Trolley Control Joysticks, RETURN the hoist to the truck bay through the notch as follows:  E.2.16.1 Using the Bridge and Trolley Control Joysticks, RETURN the hoist to the approximate vicinity of the notch. E.2.16.2 Using the Bridge and Trolley Control Joysticks, MOVE the bridge and trolley as necessary to obtain the TROLLEY ON CENTER light. E.2.16.3 Using the Trolley camera, VERIFY the reference mark on the wall as being the notch centerline as the container passes through the notch. E.2.16.4 Using the Bridge Control Joystick, MOVE the bridge north through the notch. E.2.17  If no further container movements are required, the lifting device may be removed. E.2.18  RESTORE the IRSF as follows:
E.2.18.1 VERIFY the crane is parked in the truck bay.
E.2.18.2 PLACE the IRSF ventilation system back into operation (per Attachment B). E.2.18.3 SECURE the Fire Watch posted in the truck bay.  (as applicable).
E.2.18.4 OPEN the main power disconnect for the crane on the IRSF Crane Starter Panel.
Level of Use Reference 17 of 32 LOP-WX-32 Revision 10 March 18, 2008  E.2.18.5 NOTIFY Radiation Protection to verify and log in the IRSF log book that the dose rate is less than 1 mR/hr at the outside surface of the IRSF wall.  

If the dose rate exceeds 1 mR/hr NOTIFY the Radwaste S pecialist.
Level of Use Reference 18 of 32 LOP-WX-32 Revision 10 March 18, 2008 NOTES[1] Refer to Figure 1 for IRSF configuration and bridge/trolley coordinates. [2] MONITOR Height indication when operating the hoist. E.3  Retrieving a Container from the Storage Area E.3.1  VERIFY the following prerequisites are complete:  E.3.1.1  The main power disconnect for the crane on the IRSF Crane Starter Panel is CLOSED. E.3.1.2  The IRSF Ventilation secured (per Attachment B).
E.3.2  PERFORM and DOCUMENT in IRSF log book, proper operation of the IRSF crane.  (If it has not already been previously completed that shift.)  E.3.3  PERFORM LOP-WX-34 concurrently with steps E.3.3 through E.3.13.
E.3.4  DISASSEMBLE the cask, as required.
E.3.5  If the cask has not been previously smeared clean, have the Radiation Protection Technician smear inside of the cask.
NOTE If tamper-proof seal on the secondary lid is broken, the secondary lid MUST be removed, inspected, replaced and seal 

installed. E.3.6  INSTALL the appropriate container attachment device on the crane block. E.3.6.1  If applicable, PERFORM required testing. E.3.7  From the IRSF Crane Operator Console and using the appropriate cameras, PERFORM the following:  E.3.7.1  VERIFY the BRIDGE & TROLLEY LIMIT SWITCH BYPASS is OFF.
E.3.7.2  VERIFY the GRAB BYPASS switch is OFF.  (The red GRAB BYPASS light will de-energize)
Level of Use Reference 19 of 32 LOP-WX-32 Revision 10 March 18, 2008  E.3.8  MOVE the hoist and attachment device to the appropriate grid location as follows:  E.3.8.1  Using the Hoist Control Joystick, RAISE the hoist to the nested position. The LOAD NESTED, LOAD ABOVE SHIELD WALL and LOAD ON HOOK lights will illuminate. The SLACK CABLE light will de-energize. E.3.8.2  Using the Trolley Control Joystick, MOVE the trolley west until aligned with the notch centerline as viewed in camera No. 5. E.3.8.3  Using the Bridge Control Joystick, MOVE the bridge south. As the Bridge approaches the shield wall the TROLLEY ON CENTER light will illuminate, which at this point prevents trolley movement while in the 

notch. E.3.8.4  Using the Bridge Control Joystick, MOVE the bridge south through the notch. After passing through the notch, (the Trolley on Center Slot light 

will de-energize) the bridge, trolley, and hoist controls will be used to move the center of the designated grid location.  (Subject to the interlocks 

described in Section D of this procedure.)
NOTE If retrieving a container from the storage area with Bridge coordinate 31.5' or  115.5', it may be necessary to use the Bridge/Trolley Bypass Switch. E.3.8.5  Using the Bridge and Trolley Control Joysticks, PROCEED to the appropriate grid location and level. E.3.9  RETRIEVE the container as follows:
E.3.9.1  While monitoring the Height Level Indicator indication and using the hoist control joystick, carefully LOWER the container Lifting Device to 

the designated grid location and level. The SLACK CABLE light will illuminate when the load is placed. The LOAD ABOVE SHIELD WALL 

Light will de-energize. E.3.9.2  ATTACH the attachment device to the container and VERIFY that the attachment device is fully engaged. E.3.9.3  Using the Hoist Control Joystick, RAISE the hoist to the nested position. The LOAD NESTED, LOAD ABOVE SHIELD WALL and LOAD ON HOOK lights will illuminate. The SLACK CABLE light will de-energize.
Level of Use Reference 20 of 32 LOP-WX-32 Revision 10 March 18, 2008  E.3.10  Using the Bridge and Trolley Controls, RETURN the container to the truck bay as follows:  E.3.10.1 Using the Bridge and Trolley Control Joysticks, RETURN the hoist to the approximate vicinity of the notch. E.3.10.2 Using the Bridge and Trolley Control Joysticks, MOVE the bridge and trolley as necessary to obtain the TROLLEY ON CENTER light. E.3.10.3 Using the Trolley camera, VERIFY proper alignment with the notch centerline as the container passes through the notch. E.3.10.4 Using the Bridge Control Joystick, MOVE the bridge north through the notch. After passing though the notch, MOVE the crane to the handling 

area. E.3.11  LOWER the container into the shipping cask.
E.3.11.1 When fully lowered, RELEASE the container.
E.3.12  VERIFY the container lifting slings are secured at the top of the container, as required. E.3.13  RE-ASSEMBLE the cask, as required.
E.3.14  DOCUMENT container removal from the IRSF as follows:
E.3.14.1 In the IRSF Logbook, LOG container removal from the IRSF.
E.3.14.2 REMOVE the tag for the container from the appropriate Tag Board location in the IRSF Crane Operator Control Room corresponding to the 

grid location. E.3.15  RESTORE the IRSF as follows:
E.3.15.1 VERIFY the crane is parked in the truck bay.
E.3.15.2 PLACE the IRSF ventilation system back into operation (per Attachment B). E.3.15.3 SECURE the Fire Watch already posted in the truck bay (as applicable).
E.3.15.4 OPEN the main power disconnect for the crane on the IRSF Crane Starter Panel. E.3.15.5 NOTIFY Radiation Protection to verify that the dose rate is less than 1 mR/hr at the outside surface of the IRSF wall. If the dose rate exceeds 1 mR/hr NOTIFY the Radwaste S pecialist.
Level of Use Reference 21 of 32 LOP-WX-32 Revision 10 March 18, 2008 F. REVIEW AND SIGNOFF F.1 None G. REFERENCES G.1  Vendor Manual:
G.1.1  Whiting Corporation; Parts List, Operating and Maintenance Instructions for the Whiting 20 Ton Crane (Serial No. 11965) Vendor Manual No. 235. G.2  ANSI B30.2.0 - 1976; Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder). G.3  29CFR1910.179; Labor - Occupational Safety and Health Administration - Overhead and Gantry Cranes. G.4  LOP-WX-34, Guidance for the Placement of Containers in the Interim Radwaste Storage Facility. G.5  10CFR20.
G.6  50.59 Safety Evaluation, "Final Report-Low level waste storage - LaSalle Interim Radwaste Storage Facility."  G.7  RP-AA-601, Surveying Radioactive Material Shipments Level of Use Reference 22 of 32 LOP-WX-32 Revision 10 March 18, 2008 ATTACHMENT A IRSF CRANE CONTROLS AND INDICATION The bridge, trolley and hoist are controlled by joystick controls. Each joystick control is "stepped" to its 100% position in five increments from its neutral position. Each step corresponds to a predetermined resistance which is placed into the applicable motor windings. Each step is automatically timed at two 

seconds.The hoist joystick control will also normally be operated in the 100% position when raising and lowering the hook. Caution should be used whenever controls are not operated at full motion due to possible damage to resistors. When the load is being lowered to its placement point, the hoist control will be operated through the stepped positions to provide very fine speed control. For this particular control, the stepped increments act as very effective brakes for the hoist. Operation of the joystick controls as described above will provide good control of the load and cause less wear and tear on the crane motors.
DESCRIPTION OF CONTROL OR INDICATIONBRIDGE CRANE LIGHTS    On/Off switch for the general area lighting mounted on the crane. BRIDGE POSITION LIGHT    On/Off switch for the light which illuminates the Bridge grid markings. TROLLEY POSITION LIGHT    On/Off switch for the light which illuminates the Trolley grid markings. BRIDGE AND TROLLEY BYPASS   This is a key switch, when on bypasses ALL interlocks associated with the movement of the Bridge or Trolley. Refer to the description in Limitations and Actions section E of this 

procedure.HEIGHT LEVEL INDICATION   Height level indication is a digital read out in inches located on the Operator Console. This read out is the indication of the number of 

inches left above the floor. An indication of "0.0" means the hoist is full down.  (i.e. If the indication read "55" this would indicate that 

there are 55 inches of downward travel left with 

the hoist.) POWER ON LIGHT (RED)    Indicates the Main Power Disconnect at the IRSF Crane Starter Panel is Closed, and that 

power is available to the crane and its components.
Level of Use Reference 23 of 32 LOP-WX-32 Revision 10 March 18, 2008 ATTACHMENT A (Continued)
IRSF CRANE CONTROLS AND INDICATION POWER ON/OFF PUSHBUTTONS   Powers up or off the crane and its associated controls/indications.LOAD NESTED LIGHT (BLUE)   Indicates that the Hoist is in the full up position.  (cradled)SLACK CABLE LIGHT (AMBER)   Indicates that all tension has been removed from the hoist cables (i.e. the crane hoist is resting on something). LOAD ON HOOK LIGHT (RED)   Indicates that tension has been applied to the cable of the hoist. FAST/SLOW CONTROL SWITCH   Provides selection of operating the crane functions in either fast or slow speeds, this includes bridge, trolley and hoist motion. If selected to the slow all functions operate in 

slow.BLOCK CW/CCW PUSHBUTTONS   These provide the ability to rotate the attachment connection point (i.e. The point where the hook, or other lifting devices are connected to the 

block) in either the clockwise or counter 

clockwise directions. GRAB OR RELEASE PUSHBUTTONS  Provide for the remote lifting of containers when the drum or liner lifting devices are attached to 

the crane. GRAB OPEN/CLOSED LIGHTS (WHITE)  Provide the indication for the operations of the grab devices, indicating that the grab is either fully open or fully closed. GRAB BYPASS SWITCH    Normally, if the Grab Jaws are NOT in either a FULLY ENGAGED or FULLY DISENGAGED position, that prevents movement of the Hoist either UP or DOWN. This switch bypasses that interlock to allow movement of the Hoist. CONTROL CIRCUIT GROUND FAULT  Indicates that there is a continuous ground in the control circuits of the crane. This will cause a loss of power to the crane to prevent damage to the crane controls or cameras.
Level of Use Reference 24 of 32 LOP-WX-32 Revision 10 March 18, 2008 ATTACHMENT A (Continued)
IRSF CRANE CONTROLS AND INDICATION POWER LOSS HOIST     Indicates a specific problem with the Hoist. Normally the Hoist Control breaker (labeled 

HCB) should be checked within the IRSF Crane 

Starter Panel for indication of fault, and reset if 

possible. If problem persists inform supervisor, and Electrical Maintenance Department. POWER LOSS BRIDGE    Indicates a specific problem with the Bridge. Normally the Bridge Control Breaker (labeled 

BCB) should be checked within the IRSF Crane 

Starter Panel for indication of fault, and reset if 

possible. If problem persists inform supervisor, and Electrical Maintenance Department. POWER LOSS TROLLEY    Indicates a specific problem with the Trolley. Normally the Trolley Control breaker (labeled 

TCB) should be checked within the IRSF Crane 

Starter Panel for indication of fault, and reset if 

possible. If problem persists inform supervisor, and Electrical Maintenance Department. POWER LOSS GRAB     Indicates a specific problem with the Grab controls. Normally the Grab Control Breaker (labeled GCB) should be checked within the 

IRSF Crane Starter Panel for indication of fault, and reset if possible. If problem persists inform supervisor, and Electrical Maintenance Department.
Level of Use Reference 25 of 32 LOP-WX-32 Revision 10 March 18, 2008 ATTACHMENT A (Continued)
IRSF CRANE CONTROLS AND INDICATION CAUTION The "LOAD ABOVE SHIELD WALL" light was for a design container and grapple proposed to be used in the IRSF, this should 

NOT BE used for current containers being stored. The LOAD NESTED LIGHT should be energized when passing containers 

through to the storage area.
LOAD ABOVE SHIELD WALL LIGHT (GREEN)      This light was intended to provide an indication that a container would be high enough to clear 

through the notch and pass into the storage area.
Light energizes with approximately 6" of 

upward travel left. In other words the container 

would clear the lowest portion of the notch (shield wall), and could safely pass through.  

This was for a design container and grapple 

proposed to be used in the IRSF, this should 

NOT BE used for current containers being 

stored. The LOAD NESTED LIGHT should be 

energized when passing containers through to 

the storage area.
TROLLEY ON CENTER SLOT LIGHT (GREEN)      This light indicates that the Trolley is centered within the shield wall notch and can safely pass 

through the notch with a container. This light is activated by limit switches which are mounted 

on the west wall of the IRSF building. As the Bridge is moved south near the shield wall, the light will not illuminate until approximately 18 

ft as indicated on the Bridge grid coordinates.

Also if the Bridge is located in the storage area and traveling north to the truck bay this light will illuminate at approximately 33.5 ft as indicated on the Bri dge grid coordinates. If the "Trolley on Center Slot" li ght does not light and the Bridge stops, the trolley was not properly centered (east/west).
Level of Use Reference 26 of 32 LOP-WX-32 Revision 10 March 18, 2008 ATTACHMENT B IRSF VENTILATION SYSTEM OPERATION NOTES[1] During times when the IRSF is solely being used for storage of radioactive waste, i.e., no container movement in progress, the normal IRSF ventilation system is in operation.
Ventilation damper lineup will depend on outside weather conditions. There are two modes of operation; summer and 

winter.[2] The IRSF is equipped with a Continuous Air Monitoring system that alerts Operator's to the presence of abnormal airborne contamination conditions. [3] The Exhaust Fan EF-1 located in the Truck Bay of the IRSF has been permanently disabled due to Radiation Protection concerns with an unmonitored release path. [4] The IRSF Ventilation System is interlocked to TRIP when the Fire Detection System is initiated. A. IRSF Ventilation system startup CAUTIONALL portions of the ventilation system MUST BE secured when 

required per applicable steps. The IRSF Ventilation MUST not be 

restarted while cask/process shield are open or if container movements are in progress. 1. Spring/Summer Mode (Storage Area/Truck Bay)   a. At local Heater Panel MS-HV1, VERIFY CLOSED breaker HV-1, IRSF Supply Fan Breaker. b. PLACE MS-HV1-CS Control Switch in AUTO. 
: c. At Panel TCP-1, PLACE HV-1 Selector Switch in FAN. 
: d. The following will occur:
Level of Use Reference 27 of 32 LOP-WX-32 Revision 10 March 18, 2008 ATTACHMENT B (Continued)
IRSF VENTILATION SYSTEM OPERATION NOTEThere is no remote indication for the following steps. Verification should be performed by OBSERVING the component functions in the following manner. 1) HV-1, IRSF Supply Fan STARTS. 2) Damper MOD-1, Outside Air Intake Damper OPENS. 
: e. IRSF temperatures can now be controlled by regulating the Outside Air Mixing Unit Regulating Damper as follows:
NOTEMOD-2, IRSF Mixing Damper controls IRSF temperatures. The 100% position allows maximum outside air volume to enter the system to provide maximum cooling. The 0% position allows minimum outside air volume to enter the system to provide minimum cooling. This damper can be repositioned at any time 

during operation to provide desired cooling. 1) At Panel TCP-1, POSITION MOD-2, Mixing Damper Control Switch to provide desired cooling. 2. Winter Mode (Storage Area/Truck Bay) 
: a. At local Heater Panel MS-HV1, VERIFY CLOSED breaker HV-1, IRSF Supply Fan Breaker. b. PLACE MS-HV1-CS Control Switch in AUTO. 
: c. At Panel TCP-1, PLACE HV-1 Selector Switch in HEAT.
Level of Use Reference 28 of 32 LOP-WX-32 Revision 10 March 18, 2008 ATTACHMENT B (Continued)
IRSF VENTILATION SYSTEM OPERATION   d. The following will occur:
NOTEThere is no remote indication for the following steps. Verification should be performed by OBSERVING the component functions in the following manner. 1) HV-1, IRSF Supply Fan STARTS. 2) Damper MOD-1, Outside Air Intake Damper OPENS. 
: 3) MOD-2, Outside Air Mixing Unit Regulating Damper moves to a minimum position.
NOTES UH-1, UH-2, and UH-3, Internal Heaters are now enabled and can be controlled from thermostats in the IRSF Truck bay. The IRSF Control Room Ventilation System is interlocked to TRIP when the Fire Detection System is initiated. 3. IRSF Control Room System Startup. a. At local thermostat in the IRSF Control Room, VERIFY the Cooling Thermostat is set at 78&deg;F. b. At local thermostat in the IRSF Control Room, VERIFY the Heating Thermostat is set at 70&deg;F.
Level of Use Reference 29 of 32 LOP-WX-32 Revision 10 March 18, 2008 ATTACHMENT B (Continued)
IRSF VENTILATION SYSTEM OPERATION NOTEIn automatic AC-1, IRSF Control Room Air Conditioning Unit and RTC-1, IRSF Condenser will cycle ON and OFF to maintain preset temperature for cooling purposes and DH-1, Duct Heater will cycle ON and OFF to maintain preset temperature for Heating purposes. c. At Panel TCP-2, PLACE the IRSF Control Room Air Ventilation System in AUTO.NOTEThe IRSF Equipment Room Ventilation System is interlocked to TRIP when the Fire Detection System is initiated. 4. Equipment Room Ventilation System Startup. a. At Panel TCP-3, VERIFY System Selection UH-4, UH-5, UH-6, IRSF Equipment Room Heaters are in Auto. b. VERIFY the Master Switch for EF-2 is in AUTO. 
: c. At Panel TCP-3, PLACE the System Selection - Exhaust Fan EF-2 Switch in AUTO. d. Motor operated Damper for EF-2 will open (located in the south/east corner). 
: e. EF-2, IRSF Equipment Room Exhaust Fan will now cycle ON and OFF to maintain a temperature of 80&deg;F. f. UH-4, UH-5, UH-6, IRSF Equipment Room Heaters will now cycle ON and OFF to maintain at local thermostat settings.
Level of Use Reference 30 of 32 LOP-WX-32 Revision 10 March 18, 2008 ATTACHMENT B (Continued)
IRSF VENTILATION SYSTEM OPERATION B. IRSF Ventilation Shutdown.
NOTE The IRSF Ventilation MUST be shutdown if a cask/process shield is open or if cask movements are in progress. 1. Storage Bay/Truck Bay   a. At Panel TCP-1, PLACE HV-1 Selector Switch in OFF. 
: b. The following will occur:
NOTEThere is no remote indication for the following steps. Verification should be performed by OBSERVING the component functions in the following manner. 1) HV-1, IRSF Supply Fan STOPS. 2) Damper MOD-1, Outside Air Intake Damper CLOSES. 
: 3) IRSF Exhaust Fan STOPS. 
: c. At Panel TCP-1, PLACE HV-1 Selector Switch in OFF. 
: 2. System Shutdown (Control Room). 
: a. At Panel TCP-2, PLACE the IRSF Control Room Air Ventilation System in OFF. 3. System Shutdown (Equipment Room) 
: a. At Panel TCP-3, PLACE the System Selection - Exhaust Fan EF-2 Switch in OFF.
Level of Use Reference 31 of 32 LOP-WX-32 Revision 10 March 18, 2008 ATTACHMENT C IRSF CRANE STARTUP CHECKOUT  1. ENERGIZE the Crane Control Panel by depressing the green power ON push button. The red power ON light should illuminate and the red lights for POWER LOSS HOIST, POWER LOSS BRIDGE, POWER LOSS TROLLEY, and POWER LOSS GRAB should 

de-energize. 2. ENERGIZE the crane bridge lights by placing the BRIDGE CRANE LIGHTS switch and the BRIDGE POSITION LIGHT switch to ON. 3. ENERGIZE the trolley lights by placing the TROLLEY LIGHTS switch to ON.
NOTEThere are six cameras which can be used to view the truck bay. Cameras 1, 2, 3, and 6 can be maneuvered by using the camera 

controls on the upper half of the Operators Control Console.
Cameras 4 and 5 are fixed in position to view the grid coordinates. 4. SELECT a camera for each screen by depressing the white push button beneath each screen.NOTES[1] If the GRAB is not used, PLACE the GRAB BYPASS switch to ON and VERIFY the red GRAB BYPASS light illuminates. [2] If the GRAB is used, PLACE the GRAB BYPASS switch to OFF. 5. VERIFY that the hoist controls operate in the SLOW and FAST speeds.
NOTEAllow block to come to a full stop before reversing rotation. 6. VERIFY load block operation (rotation) by operating the block clockwise and counter-clockwise.
Level of Use Reference 32 of 32 LOP-WX-32 Revision 10 March 18, 2008 FIGURE 1 IRSF LAYOUT RW-AA-10Revision 5 Page 1 of 12 RADWASTE PROCESS DESCRIPTION
: 1. PURPOSE 1.1. The Radwaste Process describes the activities associated with program improvement, and processing radioactive waste in a safe, reliable and cost-effective manner while complying with licensing and station requirements. 
: 2. TERMS AND DEFINITIONS

===2.1. Benchmarking===
A process of contacting other utilities or industry experts to obtain information for improving performance.
2.2. Burial Site: A licensed burial facility established per 10CFR Part 61 for the burial of Low Level Radioactive Waste (LLRW).
2.3. De-watering: The process of removing water from a burial or shipping container for the purposes of meeting vendor processing, burial site criteria, and/or storage 

requirements.
2.4. LLRW: Low Level Radioactive Waste.
2.5. Low Activity Waste or Low Specific Activity Waste (LSA): Class 7 Radioactive Material with a limited specific activity (activity/weight) which satisfies the 

descriptions in 49CFR 173.403.
2.6. Radwaste Personnel: Management personnel assigned to activities involving radwaste planning, packaging and shipping, waste minimization, waste processing, and contract management.
2.7. Re-useable materials:
The process of removing re-usable materials from the waste stream and directing the materials back into the plant for re-use.
2.8. Self Assessment:
A process of performing a self critical assessment of the radwaste program for the sole purpose of ensuring the business plan initiatives, EPRI/NRC/INPO, etc. criteria are being met.
2.9. Source Reduction: Policies and practices that help prevent the generation of waste at the source.
2.10. Storage: A process for the temporary holding of waste for future burial, treatment or decay. 2.11. Volume Reduction:
A process where waste is treated by the utility or vendor for the sole purpose of reducing its final burial or storage volume.
RW-AA-10Revision 5 Page 2 of 12 2.12. Waste Minimization Program:
A process developed for the purpose of reducing the amount of waste generation at the source. 
: 3. RESPONSIBILITIES 3.1. Radwaste Personnel
: Are responsible for governance and oversight of radwaste processes in order to control waste generation at the source, process radwaste using the most efficient and effective technology, and providing program improvements.  
: 4. MAIN BODY The following flowcharts contain the process activities needed to control Radwaste in a safe, reliable and cost effective manner while complying with design bases, licensing, and applicable regulatory requirements.
The written instructions in the section following the flowchart provide more detail for each activity in the flowchart.
RW-AA-10Revision 5 Page 3 of 12 FIGURE 1 Program Improvement Flowchart Page 1 of 1 Business need:
Program improvement Yes No Business need met:
Program improvement realized Are program improvements required?
4.1.9.Develop program performance standards or initiatives 4.1.3. Perform Corporate/Site self-assessments 4.1.2. Perform industry benchmarking, implement EPRI/NRC/INPO recommendations (when warranted) 4.1.1. Develop standard processes/performance indicators 4.1.4. Evaluate monthly performance indicators 4.1.5. Trend, analyze and report results 4.1.6. Evaluate self
 - assessment requirements 4.1.7.Develop and perform a follow-up self-assessment based upon developed performance standards 4.1.8.Business need:
Program improvement Yes No Business need met:
Program improvement realized Are program improvements required?
4.1.9.Develop program performance standards or initiatives 4.1.3. Perform Corporate/Site self-assessments 4.1.2. Perform industry benchmarking, implement EPRI/NRC/INPO recommendations (when warranted) 4.1.1. Develop standard processes/performance indicators 4.1.4. Evaluate monthly performance indicators 4.1.5. Trend, analyze and report results 4.1.6. Evaluate self
 - assessment requirements 4.1.7.Develop and perform a follow-up self-assessment based upon developed performance standards 4.1.8.
RW-AA-10Revision 5 Page 4 of 12 FIGURE 2 Minimize Liquid Inputs into RW and Process Page 1 of 1 Business need:minimize liquid inputs toRW and processLiquid re -cycled forplant re - use in storagetanksDischarge liquids toriver/lakeRecycleRiver/lakeDetermine dischargecriteria: to river/lake orrecycle for plant use?
4.2.4.Liquid RW processingthrough filters/demins/concentration, etc.(RW inputs) 4.2.3.Liquid RW storage on-sitewaste collection 4.2.2.Source reduction (equip.,floor,chemdrains)(RW inputs) 4.2.1.Business need:
minimize liquid inputs to RW and process Liquid   re-cycled forplant re - use in storage tanks Discharge liquids to river/lake Recycle River/lake Determine discharge criteria: to river/lake or recycle for plant use?
4.2.4.Liquid RW processing through filters/
demins / concentration, etc.
 (RW inputs) 4.2.3.Liquid RW storage on
 -site waste collection 4.2.2.Source reduction (equip., floor ,chem drains) (RW inputs) 4.2.1.
RW-AA-10Revision 5 Page 5 of 12 FIGURE 3 Minimize Liquid Waste Flowchart (Resin, Sludges, Filter Media)
Page 1 of 1 Business need: minimize liquidwaste (resin, sludges, filter media,etc.)Spent -resin/filter media andsludge (CU, FP, RW, Cond.System) slurried for wastecollection4.3.1.Place waste into wasteprocessing tanks forprocessing4.3.2.Radioactive wastedirect to burialOnsite Storage DAWBuilding / IRSFDAW -Landfill -Metals -Scrap YardDe-water/solidifywasteVendor4.3.3.RW processingsegregation, vol.reduction, thermaltreatment,compactionVendor4.3.5.Non-RadwasteOn-site StorageVendor/Ownerpackages wasteVendor4.3.4.RW or non-RWdisposal?Vendor4.3.6.Burial or On-sitestorage?Vendor4.3.7.RadwasteBurialBusiness need: minimize liquidwaste (resin, sludges, filter media,etc.)Spent -resin/filter media andsludge (CU, FP, RW, Cond.System) slurried for wastecollection4.3.1.Place waste into wasteprocessing tanks forprocessing4.3.2.Radioactive wastedirect to burialOnsite Storage DAWBuilding / IRSFDAW -Landfill -Metals -Scrap YardDe-water/solidifywasteVendor4.3.3.RW processingsegregation, vol.reduction, thermaltreatment,compactionVendor4.3.5.Non-RadwasteOn-site StorageVendor/Ownerpackages wasteVendor4.3.4.RW or non-RWdisposal?Vendor4.3.6.Burial or On-sitestorage?Vendor4.3.7.RadwasteBurial RW-AA-10Revision 5 Page 6 of 12 FIGURE 4 Processing DAW/Metals (Vendor Processing)
Page 1 of 1 Business need:Minimization, processing ofDAW/MetalsReturn material for plant use Burial site for LLW waste On - site storage LLW wasteFree release to land fillVendor volume reduction, free release, sorting, compaction, thermal treatment, and shipping process Vendo r 4.4.6.Yes NoWaste for vendorprocessing or storage?
Utility 4.4.4.Storage VendorprocessingFree release, burial, or utility storage?
Utilty/vendor 4.4.8.BurialFree releaseUtility storageUtility waste or vendorpacking and transport 4.4.5. Re - usable materials?
Utility 4.4.3.On - site waste collection 4.4.2. Utility waste minimization program 4.4.1. Business need:Minimization, processing ofDAW/MetalsReturn material for plant use Burial site for LLRW On - site storage LLRWFree release to land fillVendor volume reduction, free release, sorting, compaction, thermal treatment, and shipping process Vendo r 4.4.6.Yes NoWaste for vendorprocessing or storage?
Utility 4.4.4.Storage VendorprocessingFree release, burial, or utility storage?
Utilty/vendor 4.4.7.BurialFree releaseUtility storageUtility waste or vendorpacking and transport 4.4.5. Re - usable materials?
Utility 4.4.3.On - site waste collection 4.4.2. Utility waste minimization program 4.4.1.
RW-AA-10Revision 5 Page 7 of 12 4.1. Program Improvement Perform industry benchmarking, implement EPRI/NRC/INPO recommendations (when warranted) 4.1.1. 4.1.1. The Exelon benchmarking involves a process of understanding our own radioactive waste processing 

systems, determining what to benchmark and preparing data collection, identifying the first quartile utilities (INPO 

Performance Criteria), Analyzing the benchmarking 

information, and developing a plan for program 

improvement. 1. Other sources of information which are valuable for developing improvement plans are EPRI, NRC Inspection Reports, INPO Assessments.
Perform Corporate/Site self-assessments 4.1.2. 4.1.2. Obtaining an understanding of the data from Step 4.1.1. perform corporate/site self-assessment. 1. Develop a Corporate or Site Self-Assessment Plan per approved standard procedures. 2. The purpose of self-assessments is to determine and provide the mechanism to correct any 

performance issues.
4.1.3. Evaluate the information and develop site specific performance standards or initiatives.
An example of an initiative is developing a project plan to install pre-filters upstream from a deminerializer and the 

steps necessary to complete the project.
Develop standard processes/performance indicators 4.1.4. 4.1.4. Develop performance indicators based upon establishing goals (when applicable) to trend performance on a 

monthly, quarterly, or semi-annual basis. Typical performance indicators include, DAW and Metal pounds generated, Burial Volume, Filter Media, Resin generation, Liquid In leakage, and processing costs.
Evaluate monthly performance indicators 4.1.5. 4.1.5. If the performance indicators are outside the established goals, then go to Step 4.1.6, Trend Analyze, and Report 

Results. Develop programperformance standards orinitiatives 4.1.3.Develop programperformance standards orinitiatives 4.1.3.
RW-AA-10Revision 5 Page 8 of 12 Trend, analyze and report results 4.1.6. 4.1.6. Using the data, evaluate and trend the applicable results and report these results to management.
Evlauate self-assessment requirements 4.1.7. 4.1.7. Determine if another self-assessment is required to identify the causal factors associated with poor 

performance.
Develop and perform a follow-up self-assessment based upon developed perfromance standards 4.1.8. 4.1.8. Obtaining understanding of the data indicating poor performance develop a follow up self assessment using 

the approved self assessment process.
Are program improvements required?
4.1.9. 4.1.9. If program improvements are required, then GO to Step 4.1.3. Otherwise business need met. 

RW-AA-10Revision 5 Page 9 of 12 4.2. Minimize Liquid Inputs into RW and Process Source reduction (equip., floor, chem drains)(RW inputs) 4.2.1. 4.2.1. In leakage source reduction is controlled by managing the leakage rate going into Radwaste Systems. This is 

accomplished by establishing leakage rate goals and tracking/trending program performance based upon the 

number of gallons processed from the Equipment, Floor Drain and Equipment Drain Systems against the station 

goals. Out of variance trending conditions are monitored first by identifying or locating all sources of in leakage into radwaste system. Major in leakage sources can be 

attributed by comparing the delta between the normal system design in leakage from abnormal equipment leaks.
Potential sources for in leakage can be attributed to excessive equipment packing leaks, valve leakage, sample lines being flushed excessively, ground water in leakage, draining non-radioactive contaminated systems 

into floor drains, and etc.
Out of variance leaks are prioritize based on quantity and quality  (i.e. determining which leaks would have the most detrimental effect upon the radwaste processing media),
the top priority leaks are brought to the managers attention so corrective actions can be initiated.
Liquid RW storage on-site waste collection 4.2.2. 4.2.2. Various collection tanks are used to provide temporary storage of Floor, Equipment, and Chemical drain in 

leakage prior to processing these liquids for plant 

discharge or re-use.
Liquid RW processing through filters/demins/
concentration, etc.(RW inputs) 4.2.3. 4.2.3. The normal process for processing water is filtration, ion-exchange media (deminerializers), evaporation, concentration, neutralization, and reverse osmosis to 

improve the water quality.
RW-AA-10Revision 5 Page 10 of 12 Determine discharge criteria: to river/lake or recycle for plant use?
4.2.4. 4.2.4. This is an analytical process where the processed liquids are sampled, and a determination is made to send the 

liquids to either the onsite storage tanks for plant re-use or 

discharged to the river or lake. 1. Some criteria to consider are:
Alarm set points,  Auto isolation,  Resin Cost,  Environmental Impact,  Plant Water Quality,  And, accident monitoring systems Once the determination is made the business need is met with either the liquid re-cycled for plant re-use or discharged to river/lake. 4.3. Minimization of Liquid Waste (Resin, Sludges, Filter Media, and etc.)
Spent - resin/filter media and sludge (CU, FP, RW, Cond.
System) slurried for waste collection 4.3.1. 4.3.1. Processing media is collected from various plant systems (Fuel Pool, Condensate, Clean-up, and etc. by sluicing the media to dedicated Radwaste Tanks (Spent Resin Tank(s), and Sludge Tanks. 1. The media is collected until a time it is determined to process the waste for either de-watering or 

solidification.
Place waste into waste processing tanks for processing 4.3.2. 4.3.2. Dedicated tanks used to temporarily store spent processing media.
De-water/solidify waste Vendor 4.3.3. 4.3.3. The spent processing media is transported to the on-site vendor processing skid(s) where the waste is processed 

into containers.
De-watered waste is processed to meet either burial site criteria or vendor off-site processing 

criteria.
Solidified waste is processed to meet burial site criteria.
RW-AA-10Revision 5 Page 11 of 12 Vendor/Owner packages waste Vendor 4.3.4. 4.3.4. After de-watering or solidification processing has been completed, the waste containers are seal and packed into either Nuclear Regulatory Commission or Department of 

Transportation containers meeting 49CFR Shipping 

Requirements.
RW processing segregation, vol.
reduction, thermal treatment, compaction Vendo r 4.3.5. 4.3.5. Some waste streams such as sludges, resin, and some filter media is volume reduced by thermal treatment at an off-site processing facility.
RW or non-RW disposal?Vendor 4.3.6. 4.3.6. If the waste meets the vendors free release criteria the waste is segregated for disposal at an approved land fill.
Burial or On-site storage?Vendor 4.3.7. 4.3.7. Waste stream not meeting the vendors free release criteria are sorted and packaged for disposal at an 

approved burial site or is sent back to the utility for 

temporary storage. 4.4. Processing DAW/Metals(Vendor Processing)
Utility waste minimization program 4.4.1. 4.4.1. DAW waste generation is minimized through an approved standardized procedure.
On-site waste collection 4.4.2. 4.4.2. DAW and Metals are collected and segregated on-site.
Re-usable materials?
Utility 4.4.3. 4.4.3. A cursory check is made at the collection point for any materials that can be recycled. 1. Typical materials are:
Recyclable and washable materials,  Tools,  Scaffolding materials,  Partially used solvent cans.
RW-AA-10Revision 5 Page 12 of 12 Waste for vendor processing or storage?
Utility 4.4.4. 4.4.4. A determination is made to either direct the waste for volume reduction or placed in temporary storage on-site for future processing.
Utility waste packaging and transport 4.4.5. 4.4.5. DAW and Metal wastes are package in containers meeting The Department of Transportation 49CFR Shipping Requirements.
Vendor volume reduction, free release, sorting, compaction, thermal treatment, packaging and shipping process Vendor 4.4.6.4.4.6. DAW waste is processed by a vendor through one or more of the following processing methods:
Compaction Sort and survey for free release Thermal treatment 1. Metal waste is processed by a vendor through one or more of the following processing methods:
Decontamination Metal melt Free release, burial, or utility storage?
Utilty/vendor 4.4.7. 4.4.7. After packaging the waste is either sent to local landfill, an approved burial site, or to the utility for temporary storage. 
: 5. DOCUMENTATION - None 6. REFERENCES - None 7. ATTACHMENTS - None RW-AA-102Revision 3 Page 1 of 4 Level 2 - Reference Use RADWASTE STORAGE FACILITY/ DAW WASTE CONTAINER INSPECTIONS
: 1. PURPOSE 1.1 Provide guidance on periodic (normally quarterly) inspections of waste containers that are placed into the DAW long-term storage. 
: 2. TERMS AND DEFINITIONS 2.1. Storage Facility: A building, facility, or location on-site in which waste containers are placed into long-term storage. A Storage Facility does not include outside areas used for the temporary storage (staging) of waste containers.
These include; DAW Facility - Radwaste storage facility at Braidwood, Byron, Dresden, LaSalle, and Quad Cities.
Peach Bottom DAW Area within Low-Level Waste Storage Facility (LLRWSF)
TMI Interim Solid Waste Staging Facility (ISWSF)
Oyster Creek DAW Area within Low-Level Waste Storage Facility (LLRWSF) 2.2. Waste Container: A box, drum, liner, or other package holding radioactive waste or mixed waste.
2.3. DAW: Dry Active Waste.
2.4. Mixed Waste: A hazardous waste (as defined by 40 CFR 261.3), which is also radioactively contaminated. 
: 3. RESPONSIBILITIES 3.1. Chemistry/Radwaste Specialists and Radiation Protection Department are responsible for the operation of site DAW Storage Facilities. 3.2. The Radiation Protection Department is responsible for surveys and postings of the Storage Facility. 
: 4. MAIN BODY 4.1. All waste containers transferred between buildings are regulated by site radiation protection procedures. 4.2. Storage Facilities will not be used to process, package or decontaminate radioactive material unless such activity is allowed by Storage Facility safety evaluation.
RW-AA-102 Revision 3 Page 2 of 4 4.3. Containers and Contents 4.3.1. Waste containers stored in the Storage Facility will be stored in accordance with site procedures. Site storage procedures are written in accordance with individual safety evaluation appropriate for each specific facility. 4.3.2. Each stored waste container shall be clearly identified.
4.4. VERIFY that remote monitoring systems, if applicable, are operable. 4.5. When waste containers are stored in a Storage Facility, periodic (normally quarterly) inspections must be performed per the following: 
: 1. DOCUMENT all observations on Attachment 1, Radwaste Storage Facility/DAW Waste Container Integrity Inspection. 
: 2. INSPECT accessible stored waste containers for the following: A. Free standing water on surfaces. B. Swollen, expanded, or deformed waste containers. C. Cracks in waste container. D. Corrosion which could affect waste container integrity. 
: 3. If applicable, CHECK for water in Storage Facility sump. A. If water is found in the sump, have chemistry sample for activity prior to determining disposal options. 
: 4. Visibly INSPECT interior and exterior of the building, if applicable, for holes, cracks, or other physical damage. 
: 5. If any abnormal observations are noted, then PERFORM the following:
A. RECORD building location for abnormal observation on Attachment 1.
B. RECORD specific abnormality on Attachment 1.
C. NOTIFY Radwaste Specialist. 
: 5. DOCUMENTATION 5.1. Attachment 1, Radwaste Storage Facility/DAW Waste Container Integrity Inspection RW-AA-102 Revision 3 Page 3 of 4 
: 6. REFERENCES 6.1. Commitments - None  6.2. Writers References 6.2.1. NRC Reg. SECY- 94-198, Review of Existin g Guidance Concerning the Extended Storage of Low-Level Radioactive Waste  6.2.2. 10 CFR 50.59 Final Safety Evaluation, applicable to the specific Storage Facility 6.2.3. EPRI Storage Guideline 
: 7. ATTACHMENTS 7.1. Attachment 1, Radwaste Storage Facility/DAW Waste Container Integrity Inspection RW-AA-102 Revision 3 Page 4 of 4 ATTACHMENT 1 Radwaste Storage Facility/DAW Waste Container Integrity Inspection Page 1 of 1 Inspection Date: ______________________ Inspection Performed By: __________________________________ 

Number of Containers in Storage ______,  Minimum Containers to inspect (>
10%) _______ Inspection Techniques (i.e. visual via camera or local visual inspection) _________________________________
Inspection Description Identify Container ID that is leaking Location 1 Free standing water on the top of any waste container or on the storage area floor 2 Evidence of deterioration (e.g., corrosion, bulging of container, leaks, cracks surface contamination) 3 Water in the Storage Facility sump (if applicable) 4 Physical damage to the building, if applicable Yes/No If yes, then describe details in the Additional Comments or Descriptions Section below.
5 Evidence of unstable package stacking (e.g., tilting, crushing of lower containers) 6 Details of any handling damage or other defects (e.g., dents, scratches, crushing) If yes, then describe details in the Additional Comments or Descriptions Section below. Additional Comments or Descriptions: __________________________________________________________________________ ___________________________________________________________________________
___________________________________________________________________________ Inspection Reviewed by: ___________________________________/___________________                          Radwaste Specialist or Designee           /           Date 

RW-AA-104Revision 2 Page 1 of 5 Level 2 - Reference Use RADWASTE STORAGE FACILITY/WASTE CONTAINER INSPECTIONS
: 1. PURPOSE 1.1. Provide guidance on periodic (normally quarterly) inspections of waste containers  (i.e. Liners and HICs that contain water purification media, sludges, filters, DAW) 

that are placed into long-term storage. 1.2. Periodic (normally quarterly) inspections of DAW waste containers other than liners and HICs that are placed into long-term storage are covered per RW-AA-102. 
: 2. TERMS AND DEFINITIONS 2.1. Storage Facility:  A building, facility, or location on-site in which waste containers are placed into long-term storage. A Storage Facility does not include outside areas 

used for the temporary storage (staging) of waste containers.
These include:
Interim Radwaste Storage Facility (IRSF) - Radwaste storage building at Dresden, LaSalle, Quad Cities, and Zion.
Interim Solid Waste Staging Facility (ISWSF) - Three Mile Island.
Low-Level Waste Storage Facility (LLRWSF) - building outside of the protected area boundary at Peach Bottom and at Oyster Creek designed for the interim storage of waste containers.
In-plant Shielded Storage Area at Clinton, Byron and Braidwood.
Shielded Storage Area at Dresden, LaSalle, Oyster Creek, Quad Cities and Three Mile Island.
Limerick High Level Storage Area (HLSA) for storage of waste containers.
2.2. Waste Container
:  High Integrity Container (HIC), Steel Liner, Liner. 2.2.1. To minimize the possibility of a chemical reaction that would lead to gas generation, the chemical control program, which prevents mixing of the waste with highly oxidative or other undesirable chemicals should be in place. 2.2.2. All containers stored for a period in excess of 1 year shall have a passive vent installed. 2.2.3. Containers should generally comply with the criteria of 10 CFR Part 71, 49 CFR Part 170 to minimize the need for repackaging for shipment, and 40 CFR 173.410. 2.2.4. Radioactive wastes and materials should be repackaged when containers are degraded or leaking.
RW-AA-104Revision 2 Page 2 of 5 2.2.5. All containers should be selected and stored to prevent container degradation due to corrosives, environment, and physical/mechanical stresses. 
: 3. RESPONSIBILITIES 3.1. Radwaste Specialists are responsible for the storage operation of site Storage Facilities. 3.2. Radwaste Specialists are responsible for maintaining specific site procedures for placement and tracking of waste containers in Storage Facilities. 3.3. The Radiation Protection Department is responsible for performing the surveys and postings of the Storage Facility as required. 3.4. The storage facility will be maintained in accordance with each sites specific procedures. 
: 4. MAIN BODY 4.1. All waste containers transferred between buildings are regulated by the site Radiation Protection procedures. 4.2. Storage Facilities should be operated in accordance with the 50.59 or UFSAR. 4.3. Data Storage and Retrieval 4.3.1. Storage/shipping database information and supporting data shall be reasonably retrievable. 4.3.2. The Radiation Protection Department should maintain the shipping classification/quantification paperwork for each container placed into storage. A hard copy of the records should be archived after two years. 4.3.3. It is recommended that storage records be maintained in electronic format and/or software. 1. Where possible, the database should be located on a network level computer, which offers greater protection from loss and greater security from theft. 
: 2. RETAIN a backup copy of all software applications used to create and manage the database.  (Data could be stored longer than such applications 

are available on the market.) 
: 3. MAINTAIN the data hardware and storage media in a reasonably current technology. This may require periodic migration of data to new hardware. 
: 4. MAINTAIN current data by container or waste package number as if the container is ready for shipment to a processor or burial site.
RW-AA-104Revision 2 Page 3 of 5 4.4. ALARA/Radiological Guidance 4.4.1. Waste should not be stacked in such a way that it will increase the hazard of damaging the container and spilling the contents. 4.4.2. Increased container handling and personnel exposure can be anticipated during storage. Consequently, the methodology for maintaining exposures as low as reasonably achievable (ALARA) should be consistent with USNRC Regulatory 

Guides 8.8 and 8.10. 

===4.5. Inspections===
4.5.1. Inspections can be accomplished by use of television monitors; by walk-throughs, if storage facility layout, shielding, and the container storage array permits. 4.6. Containers and Contents 4.6.1. Waste containers stored in the Storage Facility will be stored in accordance with site procedures. Site storage procedures are written in accordance with the individual safety evaluation appropriate for each specific facility. 4.6.2. Each stored waste container shall be clearly identified. 

4.6.3. VERIFY that remote monitoring systems, if applicable, are operable. 4.6.4. When waste containers are stored in a Storage Facility, periodic (normally quarterly) inspections of accessible surfaces must be performed per the following:
: 1. DOCUMENT all observations on Attachment 1, Radwaste Storage Facility/Waste Liner Container Integrity Inspection. 2. Utilizing remote monitoring systems where possible, MONITOR all stored waste containers and the storage area floor for the following: A. Free-standing water on surfaces. B. Swollen, expanded, or deformed waste containers.
C. Cracks or holes in any waste container. D. Corrosion that could affect waste container integrity. 
: 3. If applicable, CHECK for water in the Storage Facility sump.
A. If water is found in the sump, HAVE Chemistry sample for activity prior to determining where to dispose. 
: 4. VISIBLY INSPECT the interior and exterior of the building, if applicable, for holes, cracks, or other physical damage.
RW-AA-104Revision 2 Page 4 of 5
: 5. If any abnormal observations are noted, then PERFORM the following:
A. RECORD the building location for the abnormal observation on  .
B. RECORD the specific abnormality on Attachment 1.
C. NOTIFY the Radwaste Specialist. 
: 5. DOCUMENTATION 5.1. Attachment 1, Radwaste Storage Facility/Waste Liner Container Integrity Inspection 
: 6. REFERENCES 6.1. Commitments - None 6.2. Writers References 6.2.1. NRC Reg SECY- 94-198, Review of Ex isting Guidance Concerning the Extended Storage of Low-Level Radioactive Waste. 6.2.2. 10 CFR 50.59 Final Safety Evaluation, applicable to the specific Storage Facility.
6.2.3. EPRI Guidelines for Operating an Interim On-site Low-Level Radioactive Waste Storage Facility 1002764. 6.2.4. USNRC Regulatory Guides 8.8 and 8.10. 
: 7. ATTACHMENTS 7.1. Attachment 1, Radwaste Storage Facility/Waste Liner Container Integrity Inspection RW-AA-104Revision 2 Page 5 of 5 ATTACHMENT 1 Radwaste Storage Facility/Waste Liner Container Integrity Inspection Page 1 of 1 Inspection Date:                                              

Inspection Performed By:                                                                

Number of Containers in Storage             ,  Minimum Containers to inspect (> 10%)                 

Inspection Techniques (i.e. visual via crane camera)
Inspection Description Identify Container ID that is leaking Location 1 Free standing water on the top of any waste container or on the storage area floor 2 Evidence of deterioration (e.g., corrosion, bulging of container, leaks, cracks surface contamination) 3 Water in the Storage Facility sump (if applicable) 4 Physical damage to the building, if applicable Yes/No If yes, then describe details in the Additional Comments or Descriptions Section below.
5 Evidence of unstable package stacking (e.g., tilting, crushing of lower containers) 6 Details of any handling damage or other defects (e.g., dents, scratches, crushing) If yes, then describe details in the Additional Comments or Descriptions Section below.
Additional Comments or Descriptions: 

Inspection Reviewed by:                                                                       
/
Radwaste Specialist or Designee           /           Date

RW-AA-105Revision 2 Page 1 of 8 Level 2 - Reference Use GUIDELINES FOR OPERATING AN INTERIM ON SITE LOW LEVEL RADIOACTIVE WASTE STORAGE FACILITY
: 1. PURPOSE 1.1. This procedure provides general guidelines for the safe, efficient, and routine operation of an on-site Low Level Waste (LLW) Storage Facility. 
: 2. TERMS AND DEFINITIONS 2.1. Storage Facility:  A building, facility, or location on-site in which waste containers are placed into long-term storage. A Storage Facility does not include outside areas 

used for the temporary storage (staging) of waste containers.
These include:
Interim Radwaste Storage Facility (IRSF) - Radwaste storage building at Dresden, LaSalle, Quad Cities, and Zion.
Solid Waste Staging Facility (SWSF) - Three Mile Island.
DAW Facility - Radwaste storage facility at Braidwood, Byron, Dresden, LaSalle, and Quad Cities.
Interim Solid Waste Staging Facility (ISWSF) - Three Mile Island.
Low-Level Waste Storage Facility (LLRWSF) - Building outside of the protected area boundary at Peach Bottom and at Oyster Creek designed for 

the interim storage of waste containers.
Radwaste HLSA - For the interim storage of waste containers at Limerick.
In-plant Shielded Storage Area at Byron, Braidwood and Clinton.
Shielded Storage Area at Dresden, LaSalle, Oyster Creek, Quad Cities and Three Mile Island.
2.2. Low level radioactive waste (LLW)
:  is a general term for a wide variety of radioactively contaminated materials that are deemed to be waste.
2.3. Contaminated Wastes
:  these wastes include protective clothing, machinery and related components, processed solids, and other substances that have been contaminated with varying levels of radioactivity.
2.4. Dry Solid LLW
:  is solid radioactive waste which was not generated as a result of liquid treatment process.
2.5. Treatment Processes:  this includes combustible solids, compactable solids, metal, plastics, concrete, and similar dry wastes.
RW-AA-105Revision 2 Page 2 of 8 2.6. Wet Solid LLW:  is any radioactive waste arising from liquid treatment processes (e.g., spent ion exchange resin, spent cartridge filters, evaporator concentrates, sludge). 2.7. Liquid LLW
:  is defined as low-level radioactive liquid (e.g., oil, decontamination solutions, aqueous liquids). For interim storage considerations, liquid waste is further defined as any waste that contains free liquid in amounts which exceed the 

requirements for disposal as established by the disposal facility licensing authority.
2.8. Solidified LLW
:  for storage purposes is liquid waste or wet solid waste that has been converted into a solid waste form to meet the waste acceptance criteria for disposal.
2.9. Waste Container:  A box, drum, liner, High Integrity Container (HIC), or other package holding radioactive waste or mixed waste.
2.10. Mixed Waste:  A hazardous waste (as defined by 40 CFR 261.3) which is also radioactively contaminated. 
: 3. RESPONSIBILITIES 3.1. Radwaste Specialists/Radiation Protection Department are responsible for the storage operation of site Storage Facilities. 3.2. Radwaste Specialists are responsible for maintaining specific site procedures for placement and tracking of waste containers in Storage Facilities. 3.3. The Radiation Protection Department is responsible for surveys and postings of the Storage Facility. 
: 4. MAIN BODY 4.1. General Guidance on Waste Forms 4.1.1. Container integrity should be protected against corrosion from the external environment weather protection should be included where necessary and practical. 4.1.2. All containers should be selected and stored to prevent container degradation due to corrosives, environment, and physical/mechanical stresses.
4.1.3. If liquids exist that are corrosive, then proven provisions should be made to protect the container (i.e., special liners or coatings) and/or to neutralize the excess liquids. 4.1.4. Storage containers should be raised off storage pads, where water accumulation can be expected to cause external corrosion and possible degradation of container 

integrity. 4.1.5. The container should be resistant to degradation caused by radiation effects.
RW-AA-105Revision 2 Page 3 of 8 4.1.6. The container should be resistant to biodegradation. 4.1.7. The container should remain stable under the compressive loads inherent in the disposal environment. 4.1.8. The container should remain stable if exposed to moisture or water after disposal.
4.1.9. The as-generated waste should be compatible with the container.
4.1.10. The waste container should be designed to ensure radioactive material containment during normal and abnormal occurrences. 4.1.11. All containers should be selected and stored to prevent container degradation due to corrosives, environment, and physical/mechanical stresses. 4.1.12. The following items should be included in inspection documentation:
REFER to Attachment 1 of RW-AA-102 or RW-AA-104, as appropriate. 4.1.13. Storage of hazardous waste, as specified under the Resource Conservation and Recovery Act (RCRA), is not addressed in this document. 4.1.14. Where possible, waste should be processed before storage, packaged in a form ready for transport and disposal at the end of the storage period in accordance with the requirements in 49 CFR Parts 170-189 and 10 CFR Part 61 respectively. 4.1.15. Some waste forms (i.e., liquids) are not appropriate for long-term storage. 4.1.16. Industrial waste forms (e.g., corrosives, hazardous materials, flammables, etc.)
should not be stored with radioactive wastes/materials. 4.1.17. Raw (untreated, unprocessed) radioactive waste or unpackaged radioactive materials should not be placed in storage. 4.1.18. Unless storage containers are equipped with special vent designs that allow depressurization and do not permit the migration of radioactive materials, resins highly loaded with radioactive material, such as boiling water reactor water cleanup system resins, should not be stored for a period in excess of approximately 1 year. 4.1.19. Biocides may be applied to control biological growths in certain instances, but only as a last resort and when there is a known, well understood problem. 4.1.20. Biocides are relatively short term inhibitors of biological growths and need to be reapplied periodically. 4.1.21. Some biocides should be avoided since they may contain components qualified as hazardous by the US EPA (i.e., result in a mixed waste). 4.1.22. Oxidizers (i.e. chlorine, chlorites, peroxides) must not be used without full consideration given to potential reactions with the waste forms, containers, seals 

and gaskets.
RW-AA-105Revision 2 Page 4 of 8 4.1.23. Solid waste containing liquid shall contain as little free-standing and non-corrosive liquid as is reasonably achievable. Applicable to resins, evaporator bottoms, sludges, and filters. 4.2. General Guidance for Adding Storage Capacity 4.2.1. To increase storage capacities authorized in the SAR, or to construct new storage facilities, perform an evaluation of the safety of LLW storage. 4.2.2. Document that evaluation, and make it available for USNRC staff inspections. Then either: 1. Amend your licenses where necessary to allow storage of LLW; 
: 2. Perform a 50.59 evaluation, document the evaluation, and report it to the Commission annually; or 3. Conduct an evaluation under 20.1501 and maintain a record of the results in accordance with 20.2103(a). 4.2.3. It also is possible to store waste from one nuclear plant at another nuclear plant, if formal NRC approval has been received. 

===4.3. Facility===
4.3.1. On-Site Storage Facilities should be located inside a fenced security area or a locked building. 4.3.2. On-Site Storage Facilities should not be located close to the site boundaries (fence line exposure issues, potential offsite releases) or in areas that are susceptible to flooding. 4.3.3. On-Site Storage Facilities (buildings) should be provided with fire/smoke detectors. 4.3.4. On-Site Storage Facilities should have provisions for collecting liquid drainage, including provisions for sampling all collected liquids. 4.3.5. 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. 4.3.6. The collection system should be sized such that no leakage can escape the facility.
4.3.7. The collection system should contain leak detection capabilities (i.e., sump high level alarms). 4.3.8. The sump should be inspected on a routine maintenance and surveillance schedule and kept dry.
RW-AA-105Revision 2 Page 5 of 8 4.3.9. If outdoor storage is necessary, the pad should be adequately bermed to allow for the collection of rainwater and/or leakage from the stored containers. 4.3.10. Facility lighting is adequate for the intended purpose. Over time, as the number of stored waste containers increases. The available light to some storage areas and 

containers will decrease. 4.3.11. Radiation surveys of individual packages and the storage area are required. ENSURE effluent sampling is being performed; and security inspections are being performed.
4.3.12. MAINTAIN records of tests and inspections of installed fire protection systems.
4.3.13. ENSURE testing and inspections of installed ventilation systems, radiation alarms, and continuous air monitors are being performed.
4.3.14. MAINTAIN records of the results of all surveys, thermoluminescent dosimeter (TLD) readings, and other methods of radiological monitoring. 4.4. Data Storage and Retrieval 4.4.1. Database information and supporting data shall be reasonably retrievable.
4.4.2. Database information elect/hard copied information should be maintained by the Radiation Protection Department. 4.4.3. It is recommended that storage records be maintained in electronic format. 1. Where possible, the database should be located on a network level computer, which offers greater protection from loss and greater security from theft.  
(Such data can be used by terrorist organizations.) 
: 2. RETAIN a backup copy of all software applications used to create and manage the database.  (Data could be stored longer than such applications are available on the market.) 
: 3. MAINTAIN the data hardware and storage media in a reasonably current technology. This may require periodic migration of data to new hardware. 
: 4. MAINTAIN current data by container or waste package number as if the container is ready for shipment to a processor or burial site. 4.5. ALARA/Radiological Guidance 4.5.1. Waste should not be stacked in such a way that it will increase the hazard of damaging the container and spilling the contents.
RW-AA-105Revision 2 Page 6 of 8 4.5.2. Increased container handling and personnel exposure can be anticipated during storage. Consequently, the methodology for maintaining exposures as low as reasonably achievable (ALARA) should be consistent with USNRC Regulatory 

Guides 8.8 and 8.10. 

===4.6. Inspections===
4.6.1. All inspection procedures developed should minimize occupational exposure.
4.6.2. A program of at least periodic (quarterly) visual inspection of container integrity should be performed for a representative number of packages. 4.6.3. Visual inspections should include an evaluation of container integrity/breach, damage, swelling, corrosion products, seals, latches, retaining clips, markings/labels. 4.6.4. Inspections can be accomplished by use of television monitors; by walk-throughs if storage facility layout, shielding, and the container storage array permit; or by 

selecting waste containers that are representative of the types of waste and containers stored in the facility and placing them in a location specifically designed 

for inspection purposes. 4.6.5. Extending inspection frequencies should be done only when supported by a review of historical trending data based on careful inspections which are well documented. 4.6.6. All waste containers transferred between buildings are regulated by the site radiation protection procedures. 4.6.7. Storage Facilities will not be used to process, package or decontaminate radioactive material unless such activity is allowed by the Storage Facility safety evaluation. 4.7. Containers and Contents 4.7.1. Waste containers stored in the Storage Facility will be stored in accordance with site procedures. Site storage procedures are written in accordance with the individual safety evaluation appropriate for each specific facility.
4.7.2. VERIFY that remote monitoring systems, if applicable, are operable. 4.7.3. When waste containers are stored in a Storage Facility, periodic (normally quarterly) inspections of accessible surfaces must be performed per the following: 
: 1. DOCUMENT all observations on Attachment 1 of RW-AA-102, Radwaste Storage Facility/DAW Waste Container Integrity Inspection or RW-AA-104, Radwaste Storage Facility/Waste Liner Container Integrity Inspection as appropriate.
RW-AA-105Revision 2 Page 7 of 8
: 2. Utilizing remote monitoring systems where possible, MONITOR all stored waste containers and the storage area floor for the following: A. Free-standing water on surfaces.
B. Swollen, expanded, or deformed waste containers.
C. Cracks in any waste container.
D. Corrosion which could affect waste container integrity. 
: 3. If applicable, CHECK for water in the Storage Facility sump.
A. If water is found in the sump, HAVE Chemistry sample for activity prior to determining where to dispose. 
: 4. VISIBLY INSPECT the interior and exterior of the building, if applicable, for holes, cracks, or other physical damage. 
: 5. If any abnormal observations are noted, then PERFORM the following using   of RW-AA-102 or RW-AA-104, as appropriate:
A. RECORD the building location for the abnormal observation.
B. RECORD the specific abnormality.
C. NOTIFY the Radwaste Specialist. 
: 5. DOCUMENTATION - None 6. REFERENCES 6.1. Commitments - None 6.2. Writers References 6.2.1. NRC Generic Letter 81-38, Storage of Low Level Radioactive Wastes at Power Reactor Sites. 6.2.2. 10 CFR 50.59 Final Safety Evaluation, applicable to the specific Storage Facility. 6.2.3. USNRC Regulatory Guides 8.8 and 8.10. 6.2.4. EPRI Guidelines for Operating on Interim On-Site Low-Level Radioactive Waste Storage Facility. 6.2.5. USNRC SECY-94-198, Review of Existing Guidance Concerning the Extended Storage of Low-Level Radioactive Waste, August 1, 1994.
RW-AA-105Revision 2 Page 8 of 8 6.2.6. USNRC Inspection Manual, Inspection Procedure 84900, Low-Level Radioactive Waste Storage, October 20, 2000. 6.2.7. USNRC Inspection Manual, Inspection Procedure 84850, Radioactive Waste Management - Inspection of Waste Generator Requirements of 10 CFR Par 20 and 

10 CFR Part 61, June 6, 2002. 
: 7. ATTACHMENTS - None TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 20 of 67 5.0 RADIOACTIVE WASTE CHARACTERISTICS (RECEIVING STATION AND FORWARDING STATION)
Radwaste in storage at the LaSalle County Station (Receiving Station) and stations projected to ship waste for long-term storage is provided below. 

Table 5.1 

Station HIC Size cu. Ft. Contact Dose R/hr.
Contents Status Braidwood No Class B/C waste currently in storage Byron  No Class B/C waste currently in storage Clinton No Class B/C waste currently in storage LaSalle PL14-170 FEDX 170.8 25 ALPS Resin 
& Charcoal On-site Storage LaSalle PL14-170 FEDX 170.8 50 ALPS Resin 
& Charcoal On-site Storage Waste projected to be stored in the LaSalle IRSF for long-term storage will be similar to the types of material currently in storage. Container contact dose rates may be in the range of 25-150 R/Hr (with a few containers potentially higher).
Upper layer stored containers are expected to average about 50 R/hr in a filled IRSF. To the greatest extent practicable, future containers shipped to the LaSalle IRSF will be Energy Solutions 8-120 HPDE HICs or equivalent. All containers received from off-site will be in compliance with 10 and 49 CFR transportation requirements, as applicable, and the stations WAC (see Section 7.3). 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 21 of 67 6.0 INITIAL DURATION OF STORAGE The expected initial duration of storage of Radwaste in the existing LaSalle IRSF is established as 80 years. This is anticipated to be the maximum storage duration required to accommodate life-of-plant plus 40 years of projected radioactive waste of Class B/C type. At this point in time, a permanent disposal facility in the United States is not available for out-of-compact Class B/C radioactive waste. Never-the-less, Exelon Nuclear will endeavor to identify a suitable permanent disposal facility in the near future and will work to establish a contract with a new permanent disposal facility for accepting shipment of stored Class B/C Radioactive waste from the LaSalle IRSF as soon as 

feasible.
TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 22 of 67 7.0 COMMON TECHNICAL ASSESSMENTS 7.1 Dewatered Resin Flammable Gas Generation Assessment The Dewatered Resin Flammable Gas Generation Assessment evaluates the hydrogen gas generated in bead resin Radwaste based on dose rates due to the concern of hydrogen gas Lower Flammability Limit (LFL) of 4.0% in the IRSF storage area. An Administrative Limit of 2.0 % of hydrogen gas accumulated is stipulated per IRSF storage bay volume for analysis purposes. The LaSalle storage bay Administrative Limit is 109,730 L, and the combustible gas generated per 270 HICs is 21,287 L/yr is conservatively based on 500 R/hr hot containers. It is not anticipated that the administrative limit would be reached due to ambient air exchanges even without the forced air ventilation system in 

operation.
It is reasonable to conclude that flammable gas generation in the LaSalle IRSF storage bay for the proposed waste forms, even at maximum theoretical container quantities and dose rates, is not of concern as long as the containers (and shield bells, if used) are adequately vented. This conclusion holds even without storage bay forced ventilation in 

operation for extended periods. Appendix B provides the detailed evaluation of flammable gas generation impacts on the IRSF storage system.  

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 23 of 67 7.2 HDPE (Poly) High Integrity Containers (HIC) Container Integrity Assessment HDPE 8-120B containers for the storage of Class B/C Radwaste for extended periods will be exposed to air, gamma radiation, and stacking stresses under storage conditions in the LaSalle IRSF. These containers were reviewed to ensure the containers will maintain their integrity for the duration of the extended storage period (Exelon has established the extended storage period to be 80 years) and will not rupture when subjected to handling 

for transportation to a future disposal site. 

Evaluation indicates that creep is not of concern for HDPE containers storage in air for the 80 year established storage duration. The design of the HDPE container assembly isolates the HDPE container from service loads during stacking and lifting.  

Oxidative Induction Time (OIT) testing of actual samples from the ES 8-120B HDPE container will be used to conservatively predict overall service life of this container for storage in air. Test results indicate that the estimated service life for HDPE containers in air for extended storage inside an ambient temperature storage bay is at least 100 years. 

Appendix C provides the container integrity assessment for the 8-120 poly HICs expected 

to be stored in the LaSalle IRSF for a service life up to 100 years. 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 24 of 67 7.3 Waste Acceptance Criteria (WAC)
Proposed Waste Acceptance Criteria based on the Barnwell Waste Management Facility (Barnwell, South Carolina) License Amendment No. 49, have been developed. 

Appendix D provides the proposed Waste Acceptance Criteria (WAC) for LaSalle IRSF. 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 25 of 67 7.4 Effluent Release Monitoring Assessment 7.4.1 Purpose The purpose of this activity is to review monitoring needs and capabilities for potential releases (effluents) specifically at the LaSalle County IRSF against regulatory requirements/guidance and industry standards. The conclusions of the assessment, however, have been performed on a generic basis, and are applicable to all Exelon Interim Radwaste Storage Facilities. This monitoring may or may not be continuous. Additionally, the purpose of this activity is to provide resolution options to identified issues. 

These facilities were designed for storage of radioactive waste that has been packaged for shipment and disposal at licensed disposal sites. Such waste could be staged in preparation for shipment, held for decay prior to shipment if necessary, or stored (temporarily, but for long-term periods, in excess of 5 years) in the event that disposal site access is unavailable. These buildings have no waste processing or packaging provisions now or planned. Its staging and storage-only missions will continue to be the design basis, but with the assumption that waste may be received from not only the generating station but from other designated and NRC approved Exelon stations.
This review indicates that the LaSalle County Station IRSF fully complies with all identified monitoring requirements/guidance. With regards to continuous air effluent release monitoring, the facilities Continuous Air Monitors (CAMs) are not required.
7.4.2 Current Facility Configurations The LaSalle County IRSF HVAC system exhausts through louvers in the truck bay. No effluent radioactivity monitors are provided, but the truck bay contains two Continuous Air Monitors (CAMs) that provide a local alarm function in the unexpected event of detectable airborne radioactivity. One of these CAMs is in the vicinity of the exhaust louver. For liquids, separate sump systems (hot and cold) are provided and no liquid effluents are expected. 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 26 of 67 7.4.3 Waste Characteristics and Airborne Activity Monitoring Practicalities The principal waste form anticipated to be stored in the LSCS IRSF is projected to be dewatered resins packaged in a manner suitable for shipment to a disposal facility. The packaged radioactivity will be at or near ambient temperature and mostly non-volatile particulate and/or adhered to the resin. The free volume in 

the container will be in a quiescent condition that would yield little airborne activity. 

Passive vents with HEPA filtration are provided to prevent container pressurization from any gases formed radiolytically; these gases are not radioactive. Flow through these vents would be expected to minimally impact air flow conditions in the facility. Container handling might increase flows as a result of flexure during container set down. Handling velocities are very slow and minimal contained radioactivity re-suspension is expected to result. Container vent filters would contain even this small radioactive release. 

Containers are selected for the possibility of extended storage such that integrity is not a significant concern. However, minor failures such as container body leaks or defective vent filters have little potential for generating detectable airborne activity. Only accident conditions such as a postulated container drop could result in some initial airborne particulate if a container is breached.
Container handling is a manned operation, so such an accident would be immediately detected and airborne activity monitoring is not relied upon to initiate any emergency actions. 

7.4.4 NRC Regulatory Positions on Release (Effluent) Monitoring NRCs regulatory documents that have been reviewed with regards to their positions on monitoring of release (effluents) are presented along with 

observation discussions in this section, with relevant Electric Power Research Institute (EPRI) and International Atomic Energy Agency (IAEA) documents 

evaluated in the next sections.   

The NRC regulatory documents are listed with the latest one (Regulatory Issue Summary RIS, 2008-32) first, as this RIS addresses the current Exelon IRSF situation of needing to store Class B and C radwaste for an extended periods of time, and consolidates and clarifies past NRC positions related to interim radwaste storage. Therefore, the relevant positions in this current RIS are given the most weight for the recommendations contained herein. 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 27 of 67 7.4.4.1 Regulatory Issue Summary 2008-32 / 10 CFR 50 Appendix A The RIS Interim Low-Level Radioactive Waste Storage at Reactor Sites, December 2008 (Reference 3) refers directly to the over-riding regulations (as opposed to guidance) of Appendix A to 10 CFR Part 50, as follows: 

When evaluating interim long-term on-site LLRW storage, Part 50 licensees must consider the applicability of the general design criteria 

listed in Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50, specifically Criteria 61, 63 and 64. Criterion 61, Fuel Storage and Handling and Radioactivity Control, specifies that fuel storage and handling, radioactive waste and other systems that may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions. Criterion 63, Monitoring Fuel and Waste Storage, states that appropriate systems shall be provided in fuel storage, radioactive waste systems, and associated handling areas to (1) detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions. Criterion 64, Monitoring Radioactivity Releases, specifies that there must be a method for monitoring the level of 

radioactivity in effluent release pathways and to the plant environs. 

In short, per this RIS, the NRC staff position regarding effluent monitoring for extended storage is that licensees must consider the applicability of the general design criteria listed in Appendix A, General Design Criteria for Nuclear Power Plants, to 10CFR Part 50, specifically Criteria 61, 63, and 64 (emphasis added -
consider is interpreted as not a requirement, and applicability of credible effluent release pathways is considered below). Criteria 63 and 64 are the pertinent criteria for consideration, including related to effluent monitoring. 

URS - Washington Division has considered the applicability of the specified general design criteria under Part 50 Appendix A, and our assessment concludes that for long term IRSF storage of solidified or dewatered resin LLRW [with no processing or repackaging in the IRSF], no credible potential release pathways exist for liquid, gaseous, or solid effluent release, and therefore, monitoring is not required. This is because the packages are closed except for the filtered passive vent incorporated into each container for gas release purposes. No volatile radioactive species are available for release from this vent or from the building. This would also be the case for a postulated container integrity failure under normal storage conditions. Surveillance programs should include adequate methods for detecting failure of container integrity. 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 28 of 67 Observation discussions of these 10 CFR 50 Appendix A criteria is provided following the full text of the criteria below.
Criterion 61--Fuel storage and handling and radioactivity control. The fuel storage and handling, radioactive waste, and other systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions. These systems shall be designed (1) with a capability to permit appropriate periodic inspection and testing of components important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, confinement, and filtering systems, (4) with a residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat and other residual heat removal, and (5) to prevent significant reduction in fuel storage coolant inventory under accident conditions.
Observation Discussion - With regards to the LaSalle County IRSF, Criterion 61 does not address effluent monitoring, and the issues mentioned or implied such as inspection provisions, appropriate shielding, containment, confinement and filtration are not factors in determining the need for such monitoring  Therefore, there are no issues with regards to this criterion requiring resolution.
Criterion 63--Monitoring fuel and waste storage. Appropriate systems shall be provided in fuel storage and radioactive waste systems and associated handling areas (1) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions.  
: 1. Criterion 63 - LaSalle County IRSF Observation Discussion Continuous air (airborne) monitors are provided in the 

following two locations: IRSF Truck Bay and the Ventilation System - on the mezzanine level near the exhaust louver (Reference 12). Except for active movement of radwaste containers, these CAMs are always 

in operation. As noted in Section 7.4.3, only accident conditions such as a postulated container drop could result in some initial airborne activity if a container was breached. Container handling is a manned operation, so such an accident would be immediately detected and 

airborne monitoring is not relied upon to initiate emergency actions. Additionally, as stated in earlier in this subsection, the LaSalle CAMs are considered a 

plus, but not required.

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 29 of 67 Criterion 64--Monitoring radioactivity releases. Means shall be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents.  
: 1. Criterion 64 - LaSalle County IRSF Observations The LaSalle County Station has systems and procedures in place that actively and continuously monitor the Stations 

environs for radioactivity.
No radioactive effluent discharges are expected from the IRSF during normal operation or any plant or IRSF anticipated operational occurrences. Nevertheless, CAMS are used to monitor the airborne activity environment in the IRSF for indication of unexpected releases from waste 

packages.   

Design basis accidents and external events have been reviewed for their release potential and detection means.   

Postulated design basis accidents associated with personnel operations in the IRSF are: (1) container handling accidents, bounded by a container drop from the expected lift height for transfer between the truckbay and storage 

bay; and (2) a postulated fire caused by bringing a postulated ignition source (crane electrical equipment) into the storage bay during waste handling. Container drop would be detected by the operator since the operation is monitored by camera. A fire would be detected either by the operator or by the provided smoke detectors as shown 

below, Figure 7.4.4.1. 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 30 of 67 Figure 7.4.4.1 Design basis events not involving personnel occupancy have minimal potential for releases from the passive closed packages. No ignition sources would be in the storage area, which is unlighted with the crane and wiring stowed in the truck bay. Nevertheless, smoke detectors are positioned between the truck and storage bays as shown below and would be alarmed in the control room. Seismic overturn of a waste package has not been ruled out analytically. The 

CAMs in the truckbay would collect indications of any associated airborne activity, and are in the vicinity of the IRSF exhaust vents. The seismic event would be detected by plant seismic monitors  

Releases from even these postulated accidents and events are expected to be small, and can be estimated based on post-event surveys as necessary. 

7.4.5 NUREG-0800 - 11.4 (SRP Appendix 11.4-A) Design Guidance for Temporary Storage of Low-Level Radioactive Waste, Reference 4.
7.4.5.1 Item 6A, page 11.4 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."  Smoke DetectorStorage/Truck Bay Interface Wall TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 31 of 67  Observation Discussions
: 1. Appendix A to 10 CFR Part 50 items were discussed above; additional discussion here is unnecessary. 
: 2. Failure of Container Integrity Liquid Monitoring - LaSalle County Station Free liquids storage is not allowed in the LaSalle County 

IRSF therefore failure of container integrity would produce only small amounts of liquids, mainly from dewatered resin filled HICs. 

Two separate sump systems (hot and cold) are provided in the LaSalle County IRSF. The hot sump (located in the truck bay) provides gravity drainage from the storage bay and decon area. Its high-level alarm provides annunciation 

in the IRSF. 
: 3. Failure of Container Integrity Radioactive Particulate Monitoring - LaSalle County Station All containers destined for LaSalle County IRSF storage will be fitted with HEPA filter type gaskets that allow internal generated gases to escape without loss of any radioactive contents. Additionally, these containers are not expected to contain radioactive gases. The air space within these containers is also expected to be extremely quiescent with gases leaving the containers only due to thermal breathing due to temperature changes, and slow radiolytic gas generation. Under these conditions, airborne particulate in the container will be minimal and even in a vent HEPA failure it is considered unlikely that detectable airborne radioactivity in IRSF air space could be expected.
Container handling could result in a momentary more rapid 

airflow through the vent due to container flexure, e.g. when it is set down. This is also anticipated to result in a minimal radioactivity release potential, even with a vent HEPA failure. Postulate accident conditions could result in a transient airborne activity event. Those associated with attended operations such as container handling accidents 

would be continuously monitored and therefore, would not use CAMs as the trouble indicator. Unattended operation events could include container failure due to degradation.
This is unlikely to be detected by CAM monitoring of TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 32 of 67  airborne radioactivity. Periodic container inspection is the method to be used. 

The only other event that could be postulated under unattended conditions would be events such as seismic events would be detected by means other than CAM related alarms. Even though event initiation monitoring would be provided by CAMs (which is a 

plus), failure of container integrity release of radioactive material from stored HICs that would go unnoticed is considered incredible.
7.4.5.2 Item 6E, page 11.4 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).
Observation Discussion - Station off-site doses are monitored in accordance with the Off-Site Dose Calculation (ODCM) requirements; other discussion on this issue is not necessary. Curie limits are not necessary on a technical basis, however, controls to assure that dose limits are being met WILL be based on container dose rates rather than inventory limits. In any case, this continuous monitoring requirement is met by the LaSalle County IRSF.
7.4.5.3 Item 6G, page 11.4 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.
Observation Discussions
: 1. Ventilation Exhaust and Airborne Monitoring Systems - LaSalle County Station 

As stated in Section 2.2.1, continuous airborne monitors are provided in the IRSF Truck Bay and on the mezzanine 

level near the exhaust louver; they are always in operation, except during loading and unloading operations (note, that the alarm function may need to be overridden when spurious alarms could result from background radiation from suspended waste packages). Additionally, to limit the TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 33 of 67 possibility of radioactivity release, the truck bay diesel exhaust fan is left off at all times (Reference 12).
This regulatory requirement is considered to be met.
7.4.6 NRC Inspection Manual, Inspection Procedure 65051 (Reference 7), Subsection 03.01, paragraphs g and h, indicate the following:  
: g. Inspection Requirements 02.07, Startup and Operating Procedures. Item 3 states regarding routine radiological survey and monitoring requirements, states No installed air monitoring is required for dry-waste storage-only facilities. Occasional air monitoring, with portable air samplers, may be required for compliance with 10 CFR 

20.103 and 20.201 when there is a potential for airborne 

radioactivity. 

Occasional is not defined in the Inspection Procedure; for the purpose of this technical report the term means quarterly as defined 

in Section 6.2 of Reference 10. 
: h. Inspection Requirement 02.08, Effluent Monitoring Changes. States: A low-level radioactive waste storage facility may have been originally designed as a storage-only facility that would involve no effluent monitoring requirements. However, the licensee may decide at a later date to conduct such activities as dry-active waste segregation, solid-waste compaction, repackaging, or waste processing that were not considered in the original evaluation. These changes could result in a new effluent release pathway that would necessitate monitoring requirements being added to the technical specifications in order to comply with 10 CFR 50, Appendix A, GDC 63 and 64. Specific questions as to whether effluent monitoring is 

required for a particular operation should be referred to NRR. 

Observation Discussion - The LaSalle Radwaste Storage Facilities will continue to be utilized for dry-waste storage-only, in this case, dewatered resin. The only operational changes with respect to the original design relate to: (a) potential duration of storage; (b) the predominant waste streams; (c) source strength for shielding purposes; and (d) waste receipt from stations other than LaSalle. No dry-waste segregation, compaction, processing or repackaging will be performed in these facilities. 

It should be noted that the cited 10CFR20 sections are no longer codified.
Similar language and requirements are found in 20.1703 (used here replacing 20.103) and 20.1501 (used here replacing 20.201). For the 

purposes of this activity, occasional is defined as intended for supplementary use when needed. It should be noted that this might be TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 34 of 67  considered the most appropriate system for an IRSF. Currently, LaSalle has two CAMs which have a local alarm function. A CAM system that simply collects airborne activity over an extended period of time, which can then be read in a laboratory would provide the highest sensitivity to the expected extremely small airborne activity condition. Such a cartridge system could also be used for post event diagnosis of releases.   

7.4.6.1 10 CFR 20.1703 Use of Individual Respiratory Protection Equipment - the action portion of this regulatory requirement for the IRSFs is Subsection c (1), which says in part, The licensee shall implement and maintain a respiratory protection program that includes: (1) Air sampling sufficient to identify the potential hazard, permit proper equipment selection, and (2) estimate doses. 
: 1. Ventilation Exhaust and Airborne Monitoring Systems - LaSalle County Station The LaSalle County Station has Respiratory Protection Programs that includes: (1) air sampling sufficient to identify the potential hazard, permit proper equipment selection, and (2) estimate IRSF doses. This is not a continuous monitoring requirement.
7.4.6.2 10 CFR 20.1501 General - the action portion of this regulatory requirement for the IRSFs is Subsection (a),
which states, Each licensee shall make or cause to be made, surveys that - (1) May be necessary for the licensee to comply with the regulations in this part; and (2) Are reasonable under the circumstances to evaluate - (i) The magnitude and extent of radiation levels; and (ii) concentrations or quantities of radioactive material; and (iii) The potential radiological hazards. 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 35 of 67  Observation Discussion - LAS Station have Respiratory Protection Programs that adheres to this subsection requirement.
7.4.6.3 Information Notice No. 90-09 Extended Interim Storage for Low-Level Radioactive Waste by Fuel Cycle and Material Licensees (Reference 9) - Relevant data was not found in this document. 7.4.6.4 Generic Letter 81-38 Storage of Low Level Radioactive Wastes at Power Reactor Sites (Reference 1) says, in part, Potential release pathways of all radionuclides present in the solidified waste form shall be monitored as per 10 CFR 

50, Appendix A (Reference 2).
The referenced release pathway monitoring requirements of the General Design Criteria were addressed above in Section 7.4.4.1.
7.4.7 EPRI Guidelines for Operating an Interim Onsite Low Level Radioactive Waste Storage Facility - Revision 1, Reference 10,  This latest EPRI Final Report has several pertinent guidelines with regards to this issue. These are quoted and discussed below.
7.4.7.1 Surveys should include general area radiation and contamination surveys as well as the monitoring of the radioactive waste or radioactive material containers for surface contamination.
Observation Discussion - Procedures, such as RW-AA-105 Guidelines for Operating an Interim On Site Low Level Radioactive Waste Facility, and protocols are in place to make sure general area radiation and contamination surveys as well as the monitoring of the radioactive waste or radioactive material containers for surface contamination are performed at the LaSalle 

Station IRSF.

7.4.7.2 Storage facilities (buildings) should be monitored by Continuous Air Monitors.  
 - Ensure that any required monitoring is being performed. - Verify that the monitoring equipment is operational.
TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 36 of 67  - Verify that the results of monitoring are being saved and evaluated, and that any such evaluations are being documented.
Observation Discussion  Since the IRSF only has storage functions, CAMs are currently available but their usage is determined by Station Radiation 

Protection.
7.4.7.3 Verify that monitoring is being performed at the site boundary for the storage facility and that records are being maintained for all site boundary dose measurements (e.g.,
thermoluminescent dosimeter (TLD) readings).
Observation Discussion LaSalle County Station site boundary monitoring is performed according to the ODCM. 

7.4.7.4 Alarm systems should be monitored in a constantly manned location such as the control room or guard station.
Observation Discussions Only local alarms are provided. This is considered sufficient given the limited potential for airborne activity during unmanned IRSF 

operations.
TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 37 of 67 7.4.8 IAEA  Only one IAEA guideline was identified with regards to this issue. Its identity and 

related discussion is provided below:
7.4.8.1 IAEA Technical Reports Series No. 390 (Reference 11), Interim Storage of Radioactive Waste Packages Section 

5.1 (Operational Control of Storage Conditions) - says, in part, To verify the predicted performance of a storage facility, on the basis of the safety assessment and design criteria, a number of operational control measures must be applied. They may be applied systematically or randomly, or they may involve continuous monitoring. Control measures may be accomplished locally for contact handled waste, or may be remote (e.g. cameras) in areas where personnel are not permitted routine access. - Relevant data was not found in this document.
7.4.9 Conclusions The purpose of this activity was to review release (effluent) monitoring capability at the LaSalle County Interim Radwaste Storage Facility against regulatory requirements/guidance and industry standards. A review of such monitoring capability at each follows: 

7.4.9.1 URS - Washington Division has considered the applicability of the specified general design criteria under Part 50 Appendix A, and our assessment concludes that for long term IRSF storage of solidified or dewatered resin LLRW [with no processing or repackaging in the IRSF], no credible potential release pathways exist for liquid, gaseous, or solid effluent release, and therefore, continuous effluent monitoring is not required.
7.4.9.2 Applicability to Exelon Fleet Interim Radwaste Storage Facilities The conclusions derived for the LaSalle IRSF regarding continuous effluent monitoring are also applicable to other Exelon 

IRSFs Fleet.

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 38 of 67 7.4.10 References 
 [1] Generic Letter 81-38 Storage of Low Level Radioactive Wastes at Power Reactor Sites, November 10, 1981 

[2] 10 CFR 50, Appendix A General Provisions 

[3] NRC Regulatory Issue Summary 2008-32 Interim Low-Level Radioactive Waste Storage at Reactor Sites, December 2008 
 [4] NUREG-0800 Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, March 2007 

[5] Intentionally Left Blank 

[6] 50.59 Safety Evaluation (Final Report) Low Level Waste Storage, LaSalle Interim Radwaste Storage Facility, July 1992 

[7] NRC Inspection Manual, Inspection Procedure 65051 Low-Level Radioactive Waste Storage Facilities, June 13, 1986 

[8] NRC Regulatory Issue Summary -2008-12, Considerations for Extended Interim Storage of Low-Level Radioactive Waste By Fuel Cycle and Materials Licensees May 

9, 2008  
[9] Information Notice No. 90-09 Extended Interim Storage for Low-Level Radioactive Waste by Fuel Cycle and Material Licensees  

[10] EPRI Guidelines of Operating an Interim On Site Low Level Radioactive Waste Storage Facility, June 6, 2008 

[11] IAEA Technical Reports Series No. 390 Interim Storage of Radioactive Waste Packages, 1998 

[12] LaSalle County Station Procedure # LOP-WX-32, Revision 10 IRSF General Use Procedure, March 18, 2008 TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 39 of 67 8.0 LASALLE IRSF TECHNICAL ASSESSMENT 8.1 IRSF Facility Description The LaSalle Interim Radwaste Storage Facility (IRSF) is located within the protected area (PA) of LaSalle County Nuclear Station, in the agricultural area of Brookfield Township, LaSalle County, Illinois. The LaSalle IRSF was designed to provide temporary storage for LLRW generated by Unit 1 and Unit 2 if disposal capability became unavailable. 

The IRSF provides a LLRW storage system including radiation shielding, remote loading and unloading of containers, HVAC systems, and containment of potential spills, as 

further described below: 

8.1.1 Physical Description A separate physical area provides contingency for storing Radwaste at the site location.
The building is approximately 65 feet wide by 134.5 feet long and 49 feet high. See Attachment A - IRSF LaSalle County Site and Attachment B - Google Satellite IRSF Location. The IRSF is composed of four (4) areas: truck bay, control room, mechanical/electric equipment room, and storage area.
8.1.1.1 Truck Bay The truck bay area is utilized for receiving Radwaste and is equipped with an overhead crane. The truck bay area includes a container monitoring system and swipe area, and sumps for the truck and storage bays. All loading and unloading of containers is performed after the truck cab is removed from the truck bay. Personnel are restricted from the truck bay during radiologically hot container movement to ensure ALARA.
Subsequently, the container is remotely removed from the truck bay area to the storage area through a 7 by 7 notch opening. Table 8.1 provides the physical dimension list of 

the truck bay area.
TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 40 of 67  Table 8.1 - IRSF Truck Bay Physical Dimensions Boundaries Direction Mat Wide Long Height Thick Access To Area   32 60 49 Wall Thickness (concrete)         North Conc. 2 6   East Conc. 3 2   West Conc. 3 2   South Conc. 2 6 Notch Opening for HIC South   7 7  Storage area Upper Opening  South   60 15 from ceiling  Storage area Floor  Conc. 30  Roof  Conc. 1 3  Doors West   14 17  Exterior  West (emergency exit)   3 7  Exterior  North   8 10  Exterior  Southeast   3 6 8  Control Room Mezzanine Southeast   30 11 6   Ladder South   2 6 10   Louvers Southeast   52 68  Exterior 8.1.1.2 Control Room The control room provides remote monitoring/supervision on the IRSF overhead crane operations via closed circuit television (CCTV) monitors. A record board mounted on the control room wall provides a location record of all containers by placing a circular disc with container identification number, weight, date it was stored, and the radiation level on the date stored. Table 8.2 provides the physical dimension list of the control room area.
TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 41 of 67 Table 8.2 - IRSF Control Room Physical Dimensions Boundaries Direction Mat Wide Long Height Thick Access To Area   14 19  16 4 Wall Thickness (concrete)         North CMU    12   East CMU    6   West CMU    30   South CMU    6  Floor   Conc. 6   Roof  Conc. 12  Doors West    3 6 8  Truck Bay  North    3 7 2  Exterior   South    3 7 2  Mech/Elect Room  South   3 7 2  Mech/Elect Room 8.1.1.3 Equipment Room The equipment room contains the facilitys utilities and HVAC system. Table 8.3 provides the physical dimension of the components in the equipment room. 

8.3 Equipment Room Physical Dimensions Boundaries Direction Mat Wide Long Height Thick Access To Area   19 43 6  16 4   Wall Thickness (concrete)         North CMU    6   East CMU    12   West Conc. 30   South CMU    12  Floor   Conc. 6   Roof  Conc. 12  Doors North    3 7 2  Control Room  North   3 7 2  Control Room  South    3 7 2  Exterior  South   3 7 2  Exterior TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 42 of 67 8.1.1.4 Storage Bay The storage bay area is strictly used to store Class B/C Radwaste, usually as dewatered resins. There is no routine accessibility in this area, with storage movement provided by a remotely operated crane to load and unload the waste containers through the wall notch.
The storage area has a triangular pitch arrangement, which accommodates 135 containers per layer (270 containers double stacked). Table 8.4 provides the physical dimensions of the storage bay room.
Table 8.4 Storage Bay Physical Dimensions Boundaries Direction Mat Wide Long HeightThick Access ToArea   60 95  49 Wall Thickness (concrete)         North Conc. 30 (lower section) and 15 (upper section)   East Conc. 30 (lower section) and 15 (upper section)   West Conc. 30 (lower section) and 15 (upper section)  South Conc. 30 (lower section) and 15 (upper section)  Floor   Conc. 30   Roof  Conc.
15  Doors None 8.1.1.5 Fire Detection System The IRSF fire alarm system is designed as Class A in accordance with NFPA 72, 1990 and NFPA 72E, 1990 and UBC, 1990. The IRSF fire detection system design incorporates good engineering features and capabilities for early detection, prevention and mitigation of fires. The IRSF local fire alarm panel is connected to annunciators in the IRSF control room and also in the Stations Main Control Room. Three IRSF areas 

receive continuous fire detection monitoring, 1) the truck bay/storage bay (see Section 7.4.4), 2) the control room, and 3) the equipment room. Within these areas, there are 18 TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 43 of 67  smoke detectors spread throughout the duct and ceiling. See Interim Radwaste Storage Facility (IRSF) Fire Protection Functional Test, LES-FP-23, Rev 8. The three IRSF fire 

detection areas are discussed below: 

A. Truck Bay 

The truck bay area contains two (9) pyrotronics thermal fire detectors mounted on the ceiling, two (2) on the adjacent corners of the room manual pull stations - 4.5 feet from floor with horn-strobe located above at 10 feet high, and one  See Attachment E -LaSalle IRSF Fire Detection System Equipment Location.
The IRSF fire detection system is an alarm system only. An annual fire protection functional test is performed on the IRSF verifying that the smoke detectors, duct smoke detectors, manual-electrical actuation station, fire alarm system, and supervisory panels are working properly. Fire hydrants are provided for IRSF fire fighting; with spacing provided as specified by NFPA 10 allowable distance criteria. 

B. Control Room 

The control room area contains one (1) ionization smoke detector and one (1) manual pull 

station.
C. Equipment Room 

In the equipment room, there are five (5) duct smoke detectors, three (3) ionization smoke detectors, and one manual pull station. Installed within the HVAC ductwork are five dampers that are actuated by fusible links. The fire dampers are not operated by the fire alarm system. The fire dampers operate only when the fusible link is subjected to sufficient heat to melt the fusible links. The automatic dampers installed in the HVAC 

provide no fire protection features. 

D. Storage Bay 

There is one local smoke detector located directly over the storage bay/truck bay interface wall, see figure 7.4.4.1. This detector provides some level of detection for fires in that area. Additionally, there are duct fire detectors and ionization detectors located in the truck bay and equipment room recirculation air ducts that are capable of detecting smoke in the storage bay area during all operational fan modes. 

Fire detectors in the storage area are extremely difficult to maintain since there is no routine human accessibility in that area.
TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 44 of 67 8.1.1.6 Heating and Ventilation System The IRSFs contains two patterns of ventilation and heating system for cooling and heating. The HVAC provides a 19,500 cfm fan capacity with an electric heating coil. 

There are four (4) exhaust ducts of 1,662 cfm covering the storage area and one (1) exhaust duct of 1,662 cfm covering the truck bay area maintaining a temperature range of 

50 o F to 120 o F. There is one air intake in the storage bay area of 16,050 cfm. 

During the cooling season, the system operates on 100% external air. And during the heating season, the outside air dampers move to the minimum position, where only 3,450 cfm of external air is allowed into the building, and the remaining 16,050 cfm of ventilation air gets recirculated. In the truck bay area, a positive pressure duct-mounted electrical coil provides an additional heating supply located at the mezzanine level. 

In the mechanical/electrical equipment room, a 4,000 cfm rooftop exhaust fan and motor operated lovers maintain the room at a temperature range of 70 o F to 120 o F. During heating season, general area unit heaters provide sufficient heat in the equipment room. 

In the control room, a ventilation, air conditioning, and heating system are provided to maintain a temperature range of 70 oF to 78 oF. The air conditioning system provides 1,600 cfm of controlled air temperature controlled by a louver with motor operated damper connected to the return air duct in the mechanical equipment room, which is then exhausted through the rooftop exhaust fan. The heating system is provided by duct-

mounted electric coils.    

There are fire dampers in all the duct penetrations providing fire barriers in case of a fire 

event.
8.1.1.7 Overhead Bridge Crane The truck and storage bay is covered by a 20-ton trolley-mounted overhead crane. Five (5) closed circuit television (CCTV) cameras fixed to the crane in several orientations provide real time monitoring of the container loading/unloading processes from the IRSF control room under ALARA conditions. Two (2) cameras are fixed to observe grid system coordinates, and the other three (3) can be remotely moved to several orientations facilitating container placement and surveillance routines. A grid indexing system is used to maximize the use of the storage area and to meticulously control and confirm 

container location. 

In addition, lighting located at the top of the crane hook provides visibility in the storage bay area, since there is no fixed lighting in this area.
TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 45 of 67 8.1.1.8 Surveillance System IRSF operations personnel perform a periodic visual inspection and surveillance of the storage bay with CCTV monitors. Inspections performed include surveillance for container swelling, corrosion products and signs of breaching. Inspections and surveillances are performed in accordance with written procedures. 

In addition, several routine maintenance activities are performed for the IRSF equipment such as crane surveillance testing, rad monitor surveillance testing, sump level 

surveillance testing, and HVAC testing.
Dry runs of the IRSF crane and grid indexing system (CCTV) are performed 

periodically to ensure properly operation. 

Periodic radiation monitoring of IRSF external areas are performed to ensure continued compliance with 10 CFR 20 and 40 CFR 190 requirements. 

Procedures and equipment necessary for sampling and emptying truck bay and storage bay sumps are in place. 

8.1.1.9 

==References:==
: 1. LaSalle County Nuclear Station Interim Radwaste Storage Facility 10 CFR 50.59 Safety Analysis, Commonwealth Edison Company, 1992. 
: 2. Google Earth. LaSalle County Nuclear Power Station, 05/12/09. 
: 3. Interim Radwaste Storage Facility (IRSF) Fire Protection Functional Test, QLES-FP-23, Rev 8. 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 46 of 67   .A - IRSF LaSalle County Site  

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 47 of 67   .B - Google Satellite IRSF Location

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 48 of 67  Attachment 8.C -IRSF Architecture Drawing

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 49 of 67  Attachment 8.D -LaSalle IRSF Container Arrangement TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 50 of 67  Attachment 8.E -LaSalle IRSF Fire Detection System Equipment Location (Reference 3)

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 51 of 67 8.2 Site Dose Criteria Values, Location, and Bases 8.2.1 Onsite Radiation Protection Considerations 

Radiation protection design for the IRSF meets 10CFR20 requirements including maintenance of personnel doses ALARA. The IRSF is located wholly inside of the site 

boundary, owner controlled area, and the protected area of the LSCS. Locations inside and outside of the IRSF may be restricted areas with access controlled for the purpose of protecting individuals from exposure to radiation. These are summarized as follows:
When in use, the storage bay will be at least a high radiation area (10CFR20.1601 and possibly a very high radiation area. Waste is placed in and removed from the storage bay using a crane controlled from a non-radiation area. Access to the 

storage bay is not required. During waste package handling, the truck bay will be a restricted, potentially high radiation area, particularly when a waste container is outside of the transport cask. The truck bay may also become a radiation area if significant quantities of radioactive waste are in storage, due to roof scatter. The IRSF Control Room and Mechanical Equipment Room are designed to be non-radiation areas during waste storage and handling, with dose rates limited to less than 1 mR/hr as an ALARA practice for operating personnel. Areas outside of the truck bay doors may become temporary radiation areas when containers are being handled. Occupancy is controlled by HP personnel coverage and station PA announcements when waste is being moved. During storage only conditions, locations outside of the IRSF are non-radiation areas. While this would require dose rates to be maintained at less than 5 mrem/hr, a tighter criterion of ~ 2 mrem/hr is used herein. This provides added assurance that dose rates at non-owner controlled unrestricted area locations will be within 10CFR20.1301 (a) dose limits, as well as 40CFR190 limits.
8.2.2 Offsite Radiation Protection Considerations Offsite dose rates are controlled in conformance with 10CFR20.1301 (a), where dose rates to a member of the public in an unrestricted are is limited to 2 mrem/hr or 100 mrem/year. The restricted area boundary nearest to the IRSF (i.e., to the west of the IRSF) is at 395 meters from the center of IRSF. As shown in L-003388, Rev.1, doses will be well within these limits. 

Annual doses to offsite public dose receptors are also required to be within 25 mrem/year for all nuclear fuel cycle activities. The unrestricted area boundary dose limits selected TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/21/09 Page 52 of 67  for the analysis in Reference 60 of the ODCM is 10% of this value or 2.5 mrem/year. Additionally, a limit of 1 mrem/year is applied, in L-003388, Rev.1, for at the nearest resident, with 24/7 occupancy assumed. 

8.3 Physical Security Program Assessment The IRSF for LaSalle is located inside the plant protected area. Access to the IRSF facility is controlled in accordance with the plants physical security program. Material accountability and control is maintained. The impact of the proposed change to allow LSCS to receive LLRW generated at Byron, Braidwood, and Clinton on the physical security program will be completed in accordance with current plant procedures, in accordance with the requirements of 10CFR50.54(p).
8.4 Design Basis Event Assessments 8.4.1 Fire Assessment No additional active fire detection or suppression is needed for the LaSalle IRSF if Design Basis Fire Hazard Analysis Storage Area Case 2 (Appendix E, Section 6.4.2 Case 2- 6 HDPE HICS Arrayed with Adequate Separation Distance to Other HICs to Eliminate Fire Spread) is adhered to. In this case, 100 double stacked poly HICs are interspersed according to the LaSalle existing crane index system given in Figure 8.4.1.1 below and Appendix E, Section 6.4-B, Case 2 HDPE HICS Arrayed with Adequate Separation Distance to Other HICs to Eliminate Fire Spread.  

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/21/09 Page 53 of 67 Figure 8.4.1.1 This should give approximately 13 years of capacity based on an average container 

storage rate of 8 HICs per year.
After this period, if necessary, steel shell poly HICs can be placed in the open residual spaces, and stacked two high if needed. Using this proposed approach the full capacity of 270 HIC's is attainable, and, additional fire protection systems in the 

IRSF is not considered necessary.
Current fire protection systems provided in the Truck Bay, Mechanical Equipment Room, and IRSF Control Room are acceptable.
The design basis fire protection and life safety factors analysis is provided in Appendix E. 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 54 of 67 8.4.2 Tornado Assessment The LaSalle County Station IRSF was designed to Uniform Building Code requirements, with wind design requirements for standard wind loads at a 50 year recurrence interval. 

As per the 10 CFR 50.59 Safety Evaluation Final Report of July 1992, although the facility is not designed to (then) tornado loads and pressures, its inherent stability provides protection, with the mass of radiological shielding built into the structures walls and roof resulting in wind and missile protection above the code requirements. This 1992 

evaluation indicated no realistic accident scenarios regarding tornado impacts can be postulated regarding structural failure leading to radiological releases greater than 10% of 10CFR100 limits, i.e., 2.5 Rem. In March 2007, NRC issued a Revision I to its Regulatory Guide 1.76 on Design-Basis Tornado and Tornado Missiles for Nuclear 

Power Plants. This Regulatory Guide Revision lowered the design-basis tornado characteristics, on a tornado intensity region basis. LaSalle is in tornado intensity Region I, with a tornado pressure drop of 1.2 psi, maximum wind speed of 230 miles per hour (mph), translational speed of 46 mph, and maximum rotational speed of 184 mph. This compares to a pressure drop of 3.0 psi, 360 mph maximum wind speed, 70 mph maximum translational speed, and 290 mph rotational speed in the previous Revision 0 of the Regulatory Guide, dated April 1974. Thus, this significant lowering of applicable tornado parameters for structural design implies that the original IRSF design basis and the 1992 50.59 statements concerning tornado adequacy of the LaSalle IRSF can be 

considered conservative today.
With regard to tornado missiles (which were not addressed in the original version of the Regulatory Guide), it should be noted that the largest missile to be assumed, an automobile missile, is considered to impact no higher than 30 feet above grade level, which is below the crane rail of the IRSF and a region where the structural capacity of the 30 thick concrete walls would be expected to be sufficient for protection. Other tornado missiles as defined in Revision 1 of Regulatory Guide 1.76 are a 6.625 inch diameter by 15 foot long Schedule 40 pipe of mass 130 kilograms (287 pounds) and a solid steel sphere of one inch diameter and mass 0.0669 kilograms (0.147 pounds). The Region I maximum horizontal velocities of these missiles are 135 feet per second for the pipe and 26 feet per second for the sphere. Testing by Sandia Laboratories and reported in the 

April 1976 EPRI Report NP-148 on Full-Scale Tornado-Missile Impact Tests at velocities higher than these (corresponding to those consistent with Revision 0 of the Regulatory Guide) and including Schedule 40 pipes of 3 inch diameter weighing 78 pounds and 12 inch diameter weighing 743 pounds impacting into reinforced concrete panels between 12 and 24 inches in thickness showed limited frontface penetration depths (less than 40% of the panel thickness, in all cases), and backface damage limited to slight cracks or at most limited scabbing; testing of 1 inch rebar weighing 8 pounds (bounding the solid steel sphere case) impacting an 18 inch thick panel at 50% higher velocities showed 20% frontface penetration depth, and no backface damage. The report concluded that a thickness of 24 inches is sufficient to prevent backface scabbing. Accordingly, it can be concluded that none of the postulated missiles would cause any significant effects to the 30 inch thick concrete walls up to 30 feet above grade level, and only limited damage in terms of potential spalling for the 15 inch thick walls above the crane rail (a TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 55 of 67  postulated impact on the roof of 12 to 15 inches thickness would be prevented from significant scabbing or penetration damage by the metal roof deck). Missile penetration would not occur based on these tests, and the concrete fragments resulting from any scabbing would not be expected to produce any significant damage to HICs stored below.
Any damage that would occur is considered bounded by the conservatively modeled 

container drop. 

Similarly, a postulated HIC drop onto the storage bay floor would not produce significant floor damage and no through-floor cracking for leakage, as the energy of such a drop is comparable to the worst energy from the missile impact tests, and the floor concrete thickness is greater than the damage depths of any of the tests. 

In the event that a tornado could tear open a portion of the roof or thinner walls above the crane rail, tornado suctioning of any resin from a postulated damaged HIC is also considered incredible, based on the General Electric Company report APED-5696 of November 1968, Tornado Protection for the Spent Fuel Storage Pool. This report established that the vertical walls of the spent fuel storage pool would shield the water at the pool surface and result in no significant water removal. For resins (with density roughly equivalent to water) located at elevations below the bottom half of a building with 34 feet tall vertical walls remaining below the crane rail, similar shielding of the resins from the winds would apply rendering removal of any significant amount of resins by the tornado an incredible event.
8.4.3 Flood Assessment As per the 1992 50.59, flooding of the IRSF site is not a consideration, as the Probable Maximum Flood elevation is 497 feet, a full 212 feet below the floor elevation of the LaSalle IRSF. Also, LSCS UFSAR Section 2.4.3 indicates the station site is considered flood-proof with regard to the Probable Maximum Flood (to elevation 522.5 feet 

including wave runup, more than 180 feet below the floor elevation of the IRSF.
8.4.4 Seismic Assessment As per the 1992 50.59, the LaSalle IRSF is designed to Uniform Building Code requirements for seismic design meeting Zone I. Per the 50.59, no realistic seismic accident scenarios can be postulated regarding structural failure leading to radiological releases greater than 10% of 10CFR100 limits, i.e., 2.5 Rem. If an earthquake is assumed to result in an IRSF structural failure that leads to loss of the HICs containment capability and subsequent release of their dispersible radioactive contents, there is no apparent mechanism for any subsequent release of these contents to the environment or 

the public, so post-event clean-up activities should assure essentially no radiological release consequences. In any case, the 1992 50.59 consequences of not greater than 10%
of 10CFR100 limits, i.e., 2.5 Rem, would still be expected to apply. 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 56 of 67 8.5 Shielding Assessment The design function of the IRSF is to store low level radioactive waste, either for staging in preparation for shipping to a disposal site, or for storage on a more extended basis if a disposal site is not available. The IRSF could also be used on a storage for decay basis for some high activity waste. For shielding purposes, these functions must be met with shielding and waste handling processes sufficient to meet regulated exposure limits.  

8.5.1 Facility and Shielding Description The LSCS IRSF consists primarily of: 
: 1. A storage bay area that is physically 60 wide by 95 long. Waste could be stored up to a height of about 27 high. The wall between the storage bay and the truckbay is 34 high with a 7 by 7 notch through which waste containers can be 

conveyed.    
: 2. A truckbay adjacent to the storage bay is for receiving or shipping waste using truck mounted shipping casks. 
: 3. A control room and HVAC room are located outside of the storage bay. Waste movement uses a bridge crane that is remotely controlled from the shielded 

control room. 

Wall shielding below the crane rail 30 thick concrete. Wall shielding above the crane rail is 15 concrete. The roof shielding is 12 to 15 on a metal deck supported by I-beams. 

This arrangement provides flexibility in the types and arrangements for container storage and a number of configurations have been evaluated over the life of the IRSF.
8.5.2 Regulatory Requirement for Which Shielding is Necessary Shielding assessments must demonstrate the following: 
: 1. That the annual dose to the nearest real offsite receptor, due to radiation from waste stored in the IRSF, is controlled to less than 1 mrem/year. This receptor is assumed to be at the nearest residence location and is based on a worst case 24/7 occupancy. This criterion is consistent with guidance in Standard Review Plan, Appendix 11.4-A such that it is not likely to cause the 40 CFR 190 limits to be 

exceeded. 
: 2. 10 CFR 20.1301 limits exposure rates in unrestricted areas to less than 100 mrem/year and 2 mrem/hr. To meet these criteria doses rates are evaluated at the nearest unrestricted area boundary with an occupancy by a member of the public assumed at 20 hours per year (From Reference 60 to the ODCM).    
: 3. Occupational exposure in operation of the IRSF and its general vicinity will be maintained ALARA, and with 10CFR20.1201 limits. Areas outside IRSF shielded structure, including the IRSF control room and mechanical equipment TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 57 of 67  room, will be non-radiation areas, with a design goal that maximum dose rates will be less than 2 mrem/hr. The storage bay will likely be a high radiation area when waste is present, with no personnel access required. The truck bay may become a radiation area as waste accumulates in the storage bay, due to roof 

scatter.        

Items 1 and 2 above are addressed as part of the ODCM related to Radiological Effluent 

Controls, and the related accounting for direct radiation from contained radiation sources for purposes of 40 CFR 190 and 10 CFR 20.1301 compliance.     

8.5.3 Shielding Analysis Methodology Analyses of direct radiation transport from the IRSF are performed using the point-kernel methods in MicroShield to a limited extent, for example for shielding effectiveness for a single container. Monte Carlo methods in MCNP are used for the complex geometries involved in assessment of skyshine, roof scatter, etc.   

8.5.4 IRSF Loading Strategies Current loading plans are based on the use of Energy Solutions 8-120 HICs, stacked up to two high. The currently indexed layout uses a triangular pitch providing 135 floor storage locations, or a total storage of 270 containers. The present plans reflect the current unavailability of disposal sites for Class B and C waste. These wastes are principally from PWR CVCS or BWR RWCU systems, and are expected to have average radioactivity levels higher than originally assumed in IRSF design, though high activity wastes such as RWCU resin were always expected to be included. Therefore, the shielding design has been reanalyzed to credit available design margins and to establish new operational limitations. Loading and placement restrictions are applied to assure that the radiation dose requirements are met. These are summarized below: 
: 1. Handling is controlled based on container contact dose rates.   
: 2. Shielding analyses are done conservatively, assuming an all Co-60 isotopic mix. 3. Containers on the periphery of the IRSF are limited to a 25 R/hr contact. 
: 4. The average contact dose rate limit for upwardly exposed containers (i.e., the upper layer of containers), is 50 R/hr. 
: 5. Exposure from higher activity containers can be minimized by having intervening containers nearer to outside walls, and by stacking lower activity containers over them.   
: 6. It is recognized that dose rates will build up slowly, particularly for skyshine and scatter into the truckbay due to the cumulative effect of many containers being required. Nonetheless, radiation surveys should be used to confirm that dose limits are being met in the near vicinity of the IRSF as waste packages 

are introduced.
TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 58 of 67  
: 7. Analyses, see calculation L-003388 LaSalle and Quad Cities IRSF Shielding Evaluation for Class B and C Waste, Rev. 1, indicate that if the dose rates in the immediate vicinity of the IRSF remain below 2 mrem/hr, then the limits to 40 CFR 190 and 10 CFR 20.1301 receptors will be met. Design Analysis L-003388, Pages 16 and 17 summarize the demonstration of compliance with theses requirements.
: 8. 9. Operational credit for decay in storage can be used, provided it is controlled by procedure.
This loading strategy is not the only approach that can be used to meet the required radiation exposure control. Additionally, as noted above, storage of Class B and C waste 

is not the sole waste storage function of the IRSF.      

8.5.5 Impacts of Accepting Clinton, Byron and Braidwood Waste at the LaSalle IRSF for shielding purposes As analyzed for shielding purposes, the source of radioactive waste is not a significant issue, for the following reasons: 
: 1. The assumption of an all Co-60 source isotopic mix bounds all of the waste from these facilities. 
: 2. Control of waste handling based on container contact dose rates is a site independent approach. 
: 3. Waste from LaSalle must be prepared suitable for transportation and disposal before placement in the IRSF, with the possible exception of a hold for decay mission. Waste from Clinton, Byron and Braidwood would be prepared to the same standard, except that a higher activity requiring hold for decay would not be allowed. Therefore, LaSalle waste bounds waste from the other stations.   
: 4. The capacity of the LaSalle IRSF is ample for an extended period of storage of waste from these 4 stations and the added waste will, therefore, not constrain 

LaSalle operation.   
: 5. From an IRSF operations perspective, the waste handling would be identical for all stations. Only frequency would be impacted. Even this frequency can be considered less than originally assumed for the IRSF when it was anticipated that 

Class A, B, and C waste would be stored. 
: 6. The LaSalle ODCM process for handling 40 CFR 190 and 10 CFR 20.1301 compliance with respect to the IRSF will continue to be applicable. 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 59 of 67 8.6 IRSF Decontamination Capability GL 81-38 requires provisions be made for contamination isolation and decontamination capabilities when significant handling and personnel exposure can be anticipated.  

Containers stored at the LaSalle County IRSF identified as requiring isolation and decontamination will be isolated according to Station protocols and forwarded to the Radwaste Building for decontamination or transferred offsite for that purpose. Radwaste Building facilities are designed to ensure that the ALARA methodology is incorporated 

as per Regulatory Guides 8.8 and 8.10. 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 60 of 67 8.7 Container Drop Assessment A container drop is considered a very unlikely event given crane features, operator training, and procedural controls. However, this has historically been treated as a bounding accident event with a dose limit criterion of 10% of 10CFR100 limits. LSCS practices of closing the IRSF openings and stopping ventilation when handling HICs make it unlikely that significant fractions of any potentially airborne activity will escape before settling. This provides some possibility of a qualitative dismissal of this event.
However, a new container drop assessment for the IRSF was completed in Calculation L-003430, Revision 0. The approach utilizes NUREG-1320 (Reference 8.7-1) formulations 

as endorsed in Reference 8.7-2, conservatively treating the resins that could potentially be released from the drop as a dry powder with much smaller than expected particle sizes. The calculation shows the regulatory limit is met. 

Ref.:
8.7-1    NUREG-1320, Nuclear Fuel Cycle Facility Accident Analysis 

Handbook, May 1988 

8.7-2 NRC Draft Document for Trial Use, Risk-Informed Decision-Making for Nuclear Material and Waste Applications, May 2005 8.8 IRSF Container Repackaging Capacity GL 81-38 and SRP 11.4A requires provisions be made for additional reprocessing or repackaging due to container failure and/or as required for final transport and burial per DOT and disposal site criteria. Containers stored at the LaSalle County IRSF identified as requiring reprocessing or repackaging to meet transportation and disposal requirements will be isolated according to Station protocols and forwarded to the Radwaste Building for the necessary treatment, or transferred offsite for that purpose. The LaSalle County 

IRSF does not feature reprocessing or repackaging facilities. 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 61 of 67 8.9 IRSF Liquid Identification Capability Liquid storage is not allowed in the LaSalle County IRSF; therefore, failure of container integrity would produce only small amounts of liquid, mainly from dewatered resin filled 

HICs. 

Two separate sump systems (hot and cold) are provided in the LaSalle County IRSF:  1)
The hot sump (located in the truck bay) provides gravity drainage from the storage bay and 2) the decon area (which will not be utilized per Exelon procedural controls). Sump high-level alarms provide annunciation in the IRSF and Radwaste Control Rooms.
8.10 IRSF Annunciation to Continuously Monitored Area Fire detector alarms and sump high-level alarms are annunciated in the IRSF control room. CAMs are alarmed locally only. Further discussion of the LaSalle County IRSF annunciation to continuously monitored areas was provided in Section 7.4 

8.11 UFSAR and Technical Specifications Review 8.11.1 Technical Specifications and Facility Operating Licenses The Unit 1 and 2 LSCS Technical Specifications have been reviewed and no changes are required. The engineering change (EC) will require a License Amendment to the LSCS Unit 1 and 2 Facility Operating Licenses (FOLs-NPF-11 and NPF- 18) to allow receipt of LLRW shipped from Byron, Braidwood and Clinton. Facility Operating Licenses NPF-11 and NPF-18, Section 2.B(5) indicates that  pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but separate, such byproduct and special nuclear materials as may be produced by the operation of LaSalle County Station, Units 1 and 2. The FOLs will require revision to allow receipt of LLRW generated at Byron, Braidwood and Clinton. 

Exelon will be responsible for the preparation of the License Amendment Request (LAR). 8.11.2 UFSAR Revisions UFSAR Section 11.4.1.1 may require revision to address statement that the IRSF provides storage space for approximately 2 years of abnormal output of processed waste. 

UFSAR Table 11.4-1 may require revision to include increased average contact dose at the surface of the HICs and the type and design capacity of the storage facility. Drums will no longer be stored in the IRSF.  

UFSAR Section 11.4.1.4 and Table 11.4-2 may require revision to clarify expected annual volumes of waste, or a new table added to address expected volumes of waste to 

generated or shipped from the other plants.
TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 62 of 67 Section 11.4.2.7 may require revision to allow for the storage of processed, solidified waste and dewatered resins. 

UFSAR Markups will be provided after LAR approach is defined and implemented by 

Exelon. 8.11.3 Other Licensing Basis Documents The following programs and manuals may need to be evaluated to determine if revision and update is required:  (Refer to Attached 50.59 Applicability Determination Form) 

 - Offsite Dose Calculation Manual - Radiation Protection Program - Fire Protection Program - Environmental Protection Plan - Emergency Plan - Process Control Program - Security Program TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE  Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 63 of 67  

==9.0 CONCLUSION==
S This Technical Report was developed to evaluate the existing IRSF at LaSalle Station for operating in an extended storage (i.e., beyond 5 years) mode and receiving Class B/C LLRW for storage from the Byron, Braidwood, and Clinton Stations. The report also supports a potential License Amendment Request (LAR) for LaSalle since the LaSalle Docket License will need revision to accept LLRW from the Illinois sister stations.
Current NRC regulatory guidance was reviewed to ensure that all regulatory criteria are met.
This Technical Report concludes, after assessing both normal design requirements and potential design basis events, that the existing LaSalle Station IRSF facility 

meets current NRC regulatory requirements and performs acceptably in an 

extended storage operational mode (up to 80 years in duration) for LaSalle generated Class B/C LLRW in HDPE (high-density polyethylene, or simply "poly") 
"High Integrity Containers" (HICs). This conclusion is equally valid for extended storage of Class B/C LLRW from Byron, Braidwood, and/or Clinton Station at the 

LaSalle IRSF.
In coming to this determination, the following key operational aspects were considered: 

Characterization of probable Class B/C LLRW to be received and stored. Assessment of potential container flammable gas generation. Assessment of poly HIC integrity for extended storage. Development requirements of applicable Waste Acceptance Criteria (WAC). Evaluation of the need for continuous effluent monitoring. Evaluation of design basis events, including: fire, seismic, flood, tornado, and dropped container. Updated shielding and dose acceptability assessments.
Some of the key findings of this Technical Report are: 
: 1. With proper container venting, flammable gas generation for Class B/C dewatered resin waste is not of concern, even in the event the storage bay ventilation system 

is not operating for extended periods. 
: 2. Poly HICs (8-120 units) are expected to retain the physical properties associated with their integrity for extended storage for 100 years or more. Dewatering may be required prior to shipping but would be accomplished in a plant facility away from the IRSF. With low HIC stresses expected because of the two high stacking structural systems provided with the poly HIC, container embrittlement even with high radiation doses is not expected to be an issue. The addition of anti oxidant compounds (AO) also substantially reduces poly HIC physical property degradation impacts associated with extended storage and radioactivity. Poly Page 64 of 67 HIC container storage inside the storage bay in the as-designed controlled temperature environment, with negligible UV exposure, enhances longevity of the 

poly HICs. 
: 3. Continuous effluent monitoring is not required based on technical assessment. No credible release mechanism for volatile or liquid container contents has been found that would require continuous monitoring. 
: 4. Seismic, flood, and tornado design basis events are found to be bounded by the previous 50.59 analyses. Dropped container assessment conservatively shows that less than 10% of 10CRF100 dose limits would result. Fire hazard assessments show that no credible fire initiation scenario exists, but a non-mechanistic assumed fire evaluation shows that when containers are spaced according to the specified arrangement, a fire in a poly HIC or group of 6 poly HICs would not 

spread to other non-steel shelled poly HICs, and the potential dose consequences of radiological releases from this event would be less than 10% of 10CRF100 dose limits. Up to 122 non-steel shelled poly HICs can be stored without requiring 

active fire suppression in the Storage Bay. When steel shelled poly HICs are 

interspersed as specified in this report, full storage capacity (270 HICs in a 

double-stacked configuration) would result. Therefore no active fire detection or suppression is required in the storage bay under these conditions. 
: 5. Shielding evaluations using state-of-the-art Monte Carlo analytical tools indicate that: (a) containers on the periphery of the IRSF are limited to 25 R/hr contact, and (b) the average contact dose rate limit for upwardly exposed containers (i.e., the upper layer of containers), is 50 R/hr. These operational limits assure that 10CFR20 ALARA and 40CRF190 dose requirements would be met. Containers with substantially higher contact dose rates (limited to 380 R/hr per the ODCM) can be accommodated in the storage bay as long as they are emplaced such that they are located in the interior of the array such that they have no line of site to an outside wall. Furthermore, if necessary to meet (b) above, such a container can be placed on the bottom layer with a lower dose (e.g., 50 R/hr) container on the top 

layer. 6. A 50.59 Applicability Review Form was completed per LS-AA-104-1002 and determined that a 50.59 screening is not required. A License Amendment Request (LAR) is needed, since the current LaSalle Operating License does not allow receipt of LLRW from other stations. 

This Technical Report is appended to an EC DCR (No. 375636) and establishes a solid technical foundation for supporting the proposed LAR submittal to be prepared by 

Exelon Regulatory Assurance. 

Page 65 of 67 10.0     10 CFR 50.59 REVIEW-LASALLE 

See draft 50.59 Applicability Review form below:
50.59 APPLICABILITY REVIEW FORM     LS-AA-104-1002 Revision 2    Page 66 of 67   Page 66 of 1 Activity/Document Number
:   EC No. - 375636, Technical Report Supporting                                   Revision Number:______
0______            LaSalle Interim Radwaste Storage Facility (IRSF)
Address the questions below for all aspects of the Activity. If the answer is yes for any portion of the Activity, apply the identified process(es) to that portion of the Activity. Note that it is not unusual to have more than one process apply to a given Activi ty. See Section 4 of the Resource Manual (RM) for additional guidance.
I. Does the proposed Activity involve a change:  
: 1. Technical Specifications or Operating License (10CFR50.90)?
NO    YES See Section 4.2.1.1 of the RM 
: 2. Conditions of License Quality Assurance program (10CFR50.54(a))?
Security Plan (10CFR50.54(p))?
Emergency Plan (10CFR50.54(q))?
NO    YES  NO    YES  NO    YES  See Section 4.2.1.2 of the RM 
: 3. Codes and Standards IST Program Plan (10CFR50.55a(f))?
ISI Program Plan (10CFR50.55a(g))? 

NO    YES  NO    YES  See Section 4.2.1.3 of the RM 
: 4. ECCS Acceptance Criteria (10CFR50.46)?
NO    YES See Section 4.2.1.4 of the RM 
: 5. Specific Exemptions (10CFR50.12)?
NO    YES See Section 4.2.1.5 of the RM 
: 6. Radiation Protection Program (10CFR20)?
NO    YES See Section 4.2.1.6 of the RM 
: 7. Fire Protection Program (applicable UFSAR or operating license condition)?
NO    YES See Section 4.2.1.7 of the RM 
: 8. Programs controlled by the Operating License or the Technical Specifications (such as the ODCM).
NO     YESSee Section 4.2.1.7 of the RM 
: 9. Environmental Protection Program NO    YES See Section 4.2.1.7 of the RM 
: 10. Other programs controlled by other regulations.
NO    YES See Section 4.2.1 of the RM II. Does the proposed Activity involve maintenance which restores SSCs to their original condition or involve a temporary alteration supporting maintenance that will be in effect during at-power operations for 90 days or less?   NO    YES See Section 4.2.2 of the RM  III. Does the proposed Activity involve a change to the:  
: 1. UFSAR (including documents incorporated by reference) that is excluded from the requirement to perform a 50.59 Review by NEI96-07 or NEI 98-03?
NO    YES See Section 4.2.3 of the RM 
: 2. Managerial or administrative procedures governing the conduct of facility operations (subject to the control of 10CFR50, Appendix B)
NO   YES See Section 4.2.4 of the RM  3. Procedures for performing maintenance activities (subject to 10CFR50, Appendix B)?
NO    YES See Section 4.2.4 of the RM 
: 4. Regulatory commitment not covered by another regulation based change process (see NEI 99-04)?
NO    YES See Section 4.2.3/4.2.4 of the RMIV. Does the proposed Activity involve a change to the Independent Spent Fuel Storage Installation (ISFSI) (subject to control by 10 CFR 72.48)
NO    YES See Section 4.2.6 of the RM 

Check one of the following:
If all aspects of the Activity are controlled by one or more of the above processes, thena 50.59 Screening is not required and the Activity may be implemented in accordance with its governing procedure.
If any portion of the Activity is not controlled by one or more of the above processes, then process a 50.59 Screening for the portion not covered by any of the above processes. The remaining portion of the activity should be implemented in accordance with its governing procedure.
Signoff:   50.59 Screener/50.59 Evaluator:   Sign:    Date: ___/___/___  (Circle One) (Print name) (Signature)
TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Page 67 of 67  APPENDICES Appendix A.
Discussion Of Prior IRSF 50.59 Safety Assessments Appendix B.
Dewatered Resin Flammable Gas Generation Assessment Appendix C.
HDPE (Poly) High Integrity Containers (HIC) Container Integrity Assessment Appendix D.
Waste Acceptance Criteria (WAC)
Appendix E.
Fire Hazard Analysis Report for Interim Radwaste Storage Facility at the LaSalle County Nuclear Station 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Appendix A - Page 1 of 12  APPENDIX A DISCUSSION OF PRIOR IRSF 50.59 SAFETY ASSESSMENTS 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Appendix A - Page 2 of 12 INTRODUCTION:
In support of Exelons 50.59 LAR Support task to perform technical assessment of transshipment of radioactive waste to LaSalle County Station from Byron, Braidwood, and Clinton Stations, the two previously developed 1992 and 1994 Interim Radwaste Storage Facility (IRSF) Safety Evaluations (50.59) were reviewed, observations made, and recommendations developed. Detailed overviews of these activities are provided in the following sections. 
: 1. LASALLE IRSF 50.59 1992 The 1992 prepared 50.59 focused on mostly Class A radioactive waste, with some Class B/C containers stored in the IRSF. Additionally, this 1992 document focused exclusively on waste solidified in concrete, bitumen, and/or DOW polymer binders. An overview of this documents observations and conclusions follow: 

1.1. Reference - Letter, Carlos Diaz (Fire Protection Engineer) to B. Rybak dated January 23, 1990; subject: Waste Forms Permissible for 

Storage in the IRSFs" Observations - Two of the three items discussed in the letter addresses high density polyethylene (HDPE) high integrity containers (HICs) or its contents, as follows: 

Item 1:  High density polyethylene, although not easily ignited, presents a severe fire hazard, exceeded only by polystyrene and ABS thermoplastics....In the future, should the stations determined that the amount of polyethylene HICs stored in the IRSFs will increase, consideration should be given to the installation of suppression systems. 

Item 2:  Resin, sometimes composed of polystyrene (a thermoplastic) presents the most severe fire hazard in this category (see above).....Therefore, resins should be stored in non-combustible (metal) containers and kept to a minimum. 

Discussion:  This 18 year old letter validates current thinking that 1) HDPE HICs and 2) dewatered resins are fire hazards and additional consideration should be given to 

equipping IRSFs with a fire suppression system. 

1.2. Refer to QE-06.1 Checklist, item F4 and F5.
Observations
:  F4 Relevant Issue Response - Combustible wastes are stored in metal containers. Poly HICs contain only non-combustible waste. IRSF materials of construction minimize heat sources. Refer to Attachment B.
TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Appendix A - Page 3 of 12 F3 Relevant Issue Response - The potential exists for propagation and crossing of fire zone boundaries within the IRSF. Refer to Attachment B. 

Discussion:  Additional concerns with regards to combustible material in IRSF storage are presented here in F3 and F4. Attachment B of the 1992 report provides a layout of the IRSF noting its fire zones. Attachment A of the 1992 report, however, provides the fire assessment. Since the current mission for the LaSalle County Stations IRSF 

includes storing of dewatered resins, a new fire assessment will be needed. 1.3. Refer to the UE&C 50.59 Section 1.0 (Introduction), last paragraph.
Observation - .....The scope of this safety evaluation includes issues related to storage of DAW, in addition to processed waste solidified in cement, bitumen, and/or Dow binder (vinyl ester styrene). Additionally, it provides for the use of High Integrity Containers (HICs) as storage containers for waste solidified in cement. 

Discussion:  The title 50.59 did not provide for IRSF storage of dewatered resins in HDPE HICs. The new assessment will cover its storage. 

1.4. Refer to Sections 4.5.2 (Gas Generation Rates from Biodegradation) and 4.5.3 (Gas Generation Rates from Chemical Reaction)
Observations
:  Section 4.5.2 - Gas generation from biodegradation is not expected to occur based on experience with processed waste solidification technology and implementation of the Process Control Program.......In addition, per CECo administration requirements, passive vents will be featured on all waste containers 

in the IRSF. 

Section 4.5.3 - Radwaste containing reactive components will not be authorized for IRSF storage unless their reactive component is removed....Therefore, gas generation as a result of chemical reaction is not expected to occur. 

Discussions
:  Section 4.5.2 - A review of current Exelon utilized HICs drawing (based on, proprietary drawing -Duratek drawing C-038-166003-001, Rev. 1, sheet 2 of 3) verifies that gas vent holes  are feature. Discussions with Exelon and Energy Solutions (the current HIC vendor) ensure that all HICs will continue to feature 

HEPA filtered vents. 

Section 4.5.3 - A reactive materials prohibition will need to be placed in the IRSF waste acceptance criteria (WAC).

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Appendix A - Page 4 of 12 1.5. Refer to Section 5.6 (Consideration of 300 Year Design Life)
Observation - This section states, The HICs intended for use in the IRSF will comply with the NRC Technical Position of Waste Form Revision 1, dated January 1, 1991 regarding the 300 year design life consideration. 

Discussion - HICs projected to receive long-term IRSF storage do not conform to the identified NRC position. An assessment of this containers design life is needed for extended storage purposes.
1.6. Refer to Section 6.2.7 (Ventilation System)
Observation - This section states that the storage bay HVAC is heated positive pressure and the Truck Bay HVAC is heated negative pressure. 

Discussion - Neither HEPA nor charcoal filtration is mentioned. An assessment will be necessary to determine whether pressure direction and lack of HEPA or charcoal filtration is adequate for projected increased radiation dose and curie container 

loadings. 1.7. Refer to Section 6.2.8 (Container Placement Grid System)
Observation - The first paragraph of this subsection states in the IRSF is limited to accepting containers no larger than a standard size 195 cu. ft. liner. This is due to the 

seven by seven foot notch opening to the storage bay. 

Discussion - Each Station that will utilize the LaSalle IRSF will need to limit their container size to no larger than a standard size 195 cu. ft. liner. Specific dimensions will need to be established in the WAC.
1.8. Refer to Section 7.0 (IRSF Shielding and ALARA Considerations)
Observation - This section states in part, Based on filling the IRSF with 416 15 R/hr 195 cu. Ft. containers with and average isotopic activity of 0.467 curies/cubic foot, approximately 105 curies of radioactive waste material will be stored in the LaSalle 

County Station IRSF for five years. 

Discussion - Increase radiation dose and container average curies is expected and therefore, updated 10 CFR 20 and 40 CFR 190 assessments are needed. 

1.9. Refer to Section 7.5 (Conclusion of Design as ALARA)
Observation - The expected dose rate contribution from stored waste in the truck bay area is on the order of one-two mR/hr......These values are comparable to does rates TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Appendix A - Page 5 of 12  which might be expected from the exterior of full shipping containers and is considered ALARA. 

Discussion- Due to highly elevated container dose rates expected for storage, it is unclear whether wastes currently destined for IRSF storage will maintain truck bay radiation dose as ALARA without engineering evaluations and fixes. An updated assessment will be performed to make sure that truck bay ALARA is maintained. 

1.10. Refer to Section 9.2 (Impact of Dropped Container in the IRSF and Container Hitting an IRSF Wall)
Discussion - With the higher projected curie loadings, an updated assessment is necessary to determine whether this design basis accident produces results different than 

that identified 1992.  

1.11. LaSalle Station UFSAR, Refer to Section 11.4 (Solid Waste Management System)
Note:  A very limited review of the Station's UFSAR was performed for this activity; a detailed review is required.
Discussions and Observations:
Section 11.4.1.1 has time limitations on the IRSF, which will need to be removed. 

Section 11.4.2.7 says that the IRSF provides storage for processed, solidified waste; dewatered resins are not mentioned. This will require correction. 
: 2. LASALLE IRSF 50.59 1994 A review of the 1994 LaSalle County Station IRSF 50.59 was performed. The following observations were made.  

2.1. Reference QE-06.1 (6) checklist - page 6 of 21 has the following entries. Similar observations are in this "F4" section.
Checklist Statement - In the Fire Protection Assessment Section, 9.1, it is stated that Interim storage of dewatered wastes in the IRSF does not change that assessment, referring to the 50.59 performed in 1992. 

Discussion - A new combustion source (potentially large quantities of high-density polyethylene HICs loaded with high contact dose rate dewatered bead resins) was introduced. The previous 1992 50.59 had a smaller quantity of HICs  and resin was solidified in cement, bitumen, or Dow binder.   

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Appendix A - Page 6 of 12  In a storage bay conflagration event, dewatered resins can become dry and ignite. This position was confirmed by CECO Fire Protection assessment in a memo included in the 1992 50.59 report (see Section 3.1.1, above). This memo recommended that HDPE HICs and especially with dewatered resins be stored in limited quantities and in steel liners.
Checklist Statement - Processed, dewatered waste is stored in high integrity, vented, corrosion resistant containers. Spontaneous combustion of this material is extremely unlikely. IRSF design and materials of construction minimizes sources of energy.
Storage of dewatered waste in the IRSF does not cause the capacity of the fire zone to be exceeded. Although high-density polyethylene and ion exchange resins can be incinerated, they are not combustible materials that would cause the capacity of the fire zone to be exceeded. See Attachment B for further discussion of this issue.
Discussion
: 1. The fire zone limit was not provided. 
: 2. According to Attachment 1, resins utilized at Exelon Stations have an auto ignition temperature of 230 &#xba;C. 
: 3. HDPE is definitely combustible, according to Reference 2 (Attachment 2), containing 18,690 btu/lb. 
: 4. The Attachment B of the 1994 document discussion was limited 
: 5. The new fire assessment, identified in Subsections 3.1.2 and 3.1.3, above will need to resolve these issues.
2.2. Reference QE-06.1 (6) checklist - page 19 of 21 has the following entry. Checklist Statement - The storage of dewatered waste in the IRSF does not affect the site radioactive material inventory.
Discussion - The receiving of off-site waste will affect the Stations radioactive material inventory.
Changes in the UFSAR, at a minimum may be necessary.
2.3. Reference Attachment B of the 1994 document, page 5 and 6 Observation - Section 5.1 (General Description of Container) states that drop tests at 16,000 lbs were performed on HICs and that they passed.
TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Appendix A - Page 7 of 12 2.4. Reference Attachment B of the 1994 document, page 11 Observation - Section 6.3 (Ventilation) says in part, This ventilation system is adequate to prevent the buildup of potential combustible gases being vented from containers excessive temperature extremes. Storage bay ventilation adequacy with long-term storage requirements needs to be verified. 2.5. Reference Attachment B of the 1994 document, page 12 Observation - Section 6.5.2 (Airborne):  A discussion of storage and truck bay monitoring capability is addressed. 

Discussion - The storage bay does not currently employ radiation monitors. The need for this feature requires assessment. 

2.6. Attachment B of the 1994 document, page 18 - Section 9.3 (Ventilation System Failure)
Observation - This subsection says in part, Although there may be somewhat more gas generation with the interim storage of dewatered waste, it is still well below the flammability limits for several months and consequently, this event is considered highly 

unlikely.  

Discussion - Desk top testing or analysis of a full IRSF with containers gassing needs to be performed. 

2.7. Attachment B of the 1994 document, page 19 (Crane Failure to Loss of Off-Site Power)
Observation - The LaSalle IRSF is fitted with a ha nd crank that can be utilized in the event of power loss. 

Discussion - Utilizing crane failure truck bay hand crank may be non-ALARA. An assessment will be required. 

2.8. Letter, Steven M. Abrams (M/S Group), dated March 9, 1994; subject:  Addressing Stacking in the Interim Radwaste Storage Facility (IRSF)
Observation - The letter states that automatic fire suppression system should be installed in the Storage Bay and Nuclear Mutual Limited (NML) told CECo (now Exelon) that they would be penalized unless one is installed. Exelon chose, however, to accept the penalty, because it was less costly than the fire suppression installation cost. 

Discussion - With very high container rad levels expected, this Exelon decision will need to be revisited. Unless HICs are stored with a steel shell, active fire suppression may be TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Appendix A - Page 8 of 12  needed, at least in the truck bay. Storage bay active fire suppression should also be considered.  

2.9. Letter, Chron # 209876, dated June 15, 1994 Observation - The letter says in part, The Station can allow minimal personnel inside the truck bay during container moves to aide in the move process. With expected high container radiation levels, this philosophy may not be ALARA and therefore additional evaluation is necessary.
TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Appendix A - Page 9 of 12  
: 3. CONCLUSIONS Reviews of the 1992 and 1994 LaSalle County IRSF Safety Evaluations identified the following assessment requirements.
Subsection No.Required Assessment 1.2 New fire assessment is required. 1.3 New fire assessment will cover dewatered resins in HDPE HICs.
1.4 A reactive materials prohibition will need to be placed in the IRSF waste acceptance criteria (WAC). 1.5 A HIC design life assessment for extended storage is required. 1.6 Additional assessments are not required per this subsection discussion.
1.7 Ventilation pressure direction and lack of HEPA storage bay impact assessment is required. 1.8 Updated 10 CFR 20 and 40 CFR 190 assessments are required.
1.9 New radiation assessment must include truck bay ALARA assessment. 1.10 New design basis assessment (container hitting IRSF wall) is required.
1.11 Detailed UFSAR review is needed to evaluate removal/additions of text regarding radwaste storage. 2.1 Additional assessments are not required per this subsection discussion.
2.2 Additional UFSAR reviews are not required per this subsection discussion.
2.3 New drop container evaluation will be performed to assess radiological consequences.
2.4 Storage bay ventilation adequacy with long-term storage requirements needs to be verified. This will be provided in the Subsection 3.1.6 assessment.
2.5 Radiation monitor assessments will be performed in conjunction with the Subsection 3.1.7 assessment. 2.6 Container flammable gas assessment is required.
2.7 Use of truck bay hand crank ALARA assessment will be performed in conjunction with the Subsection 3.17 assessment 2.8 Storage bay fire suppression needs assessment will be performed in conjunction with the Subsection 3.1.2 fire assessment.
2.9 Personnel in truck bay during container movements ALARA assessment will be performed in conjunction with the Subsection 3.1.7 assessment. 

TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Appendix A - Page 10 of 12  
: 4. REFERENCES 
 [1] Final 10 CFR 50.59 Q6 Checklist Plus Safety Evaluation for the LaSalle IRSF 

[2] Chaz Miller, Profiles in Garbage: High-Density Polyethylene, Waste Age July 2001 TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Appendix A - Page 11 of 12  ATTACHMENT 1  Siemens Material Safety Data Sheets for Product Names USF NR-39LC Ion Exchange Resin  USF C-471C H Ion Exchange Resin  USF C-550C H Ion Exchange Resin SIEMENS Material Safety Data Sheet Water Technologies SECTION1-CHEMICAL PRODUCT AND COMPANY INFORMATION Product Name: USF NR-39LC Ion Exchange Resin Part Number: multiple Chemical Family: cation and anion exchange polymers Manufacturer's Name: Siemens Water Technologies Corp.Address: 181 Thorn Hill Road, Warrendale, PA 15086 ProductfTechnlcallnformation Phone Number: (815)877-3041 Medical/Handling Emergency Phone Number: Call CHEMTREC at 800/424-9300 24 hours a day Transportation Emergency Phone Number: Call CHEMTREC at 800/424-9300 24 hours a day Issue Date: June 23, 2009 Revision Date/Revision Number: June 23, 2009/Rev 1 SECTION2-COMPOSITION INFORMATION Chemical Name%By Weight CAS#Sulfonated copolymer of styrene and divinylbenzene in lithium?20-30 068584-06-5 form Trimethylamine functionalized, chloromethylated copolymer of 15-30 069011-18-3 styrene and divinylbenzene in hydroxide form Water 30-60 007732-18-5 SECTION3-HAZARDS IDENTIFICATION Appearance
&Odor: Spherical beads/Odorless to slight amine odor Emergency Overview:*May cause eye and skin irritation.
*May cause toxic fumes/vapors if burned.*May react violently when exposed to oxidizing agents such as Nitric Acid (HN0 3).Fire&Explosion Hazards: This material will not burn until moisture is removed, then resin starts to burn in flame at 230 0 C.Under fire conditions some components of this product may decompose.
The smoke may contain unidentified toxic and/or irritating compounds.
Nitric acid and other strong oxidizing agents can cause explosive-type reactions when mixed with ion exchange resins.Proper design of equipment to prevent build up of pressure is necessary if use of an oxidizing agent such as nitric acid is contemplated.
Primary Route(s)of Exposure: skin and eye contact Inhalation
-Acute Effects: Vapors are unlikely due to physical properties.
Skin Contact-Acute Effects: Skin contact may cause mild irritation and redness.Eye Contact-Acute Effects: May cause severe eye irritation and redness.May cause moderate corneal injury.Effects are likely to heal.
SIEMENS Material Safety Data Sheet Water Technologies Ingestion-Acute Effects: Single dose oral toxicity is considered to be low.No hazards anticipated from swallowing small amounts incidental to normal handling operation.
Swallowing large amounts may cause irritation to the gastrointestinal tract SECTION4-FIRST AID MEASURES Inhalation First Aid: Remove affected person from area to fresh air and provide oxygen if breathing is difficult.
Give artificial respiration ONLY if breathing has stopped and give CPR ONLY if there is no breathing and no pulse.Obtain medical attention.
No adverse effects anticipated by this route of exposure.Skin Contact First Aid: Immediately remove clothing from affected area and wash skin vigorously with flowing water.Clothing should be washed before reuse.Seek medical attention if irritation occurs.DO NOT instruct person to neutralize affected skin area.Eye Contact First Aid: Immediately irrigate eyes with flowing water continuously for 15 minutes while holding eyes open.Contacts should be removed before or dUring flushing.Obtain medical attention.
DO NOT instruct person to neutralize.
Ingestion First Aid: No adverse effects anticipated by this route of exposure incidental to proper industrial handling.If ingestion does occur, if victim is alert and not convulsing rinse mouth with water and give plenty of water to drink.If spontaneous vomiting occurs, have affected person lean forward with head down to avoid breathing in of vomitus.Rinse mouth again and give more water to drink.Obtain medical attention.
Medical Conditions Aggravated:
There are no known conditions aggravated by exposure.Note to Physician:
No specific antidote.Supportive care.Treatment based on judgment of the physician in response to reactions of the patient.SECTION5-FIRE FIGHTING MEASURES Flash Point/Method:
N/A Auto Ignition Temperature:
Above 500 0 C (900&deg;F)Upper/Lower Explosion Limits: N/A Extinguishing Media: Water, carbon dioxide, dry chemical Appendix A Attachment I PageofProject No.29487-NCS0097.N.LAS EC DCR#375636 Fire Fighting Procedures:
Keep people away.Isolate fire area and deny unnecessary entry.Cool surrounding area with water to localize fire zone.Soak thoroughly with water to cool and prevent reignition.
Fire-Fighting Equipment:
NIOSH approved positive-pressure self-contained breathing apparatus (SCBA)and protective fire fighting clothing (includes fire fighting helmet, coat, pants, boots and gloves).If protective equipment is not available or not used, fight fire from a protected location or a safe distance.USF NR-39LC Ion Exchange Resin Page 2 of6 SIEMENS Material Safety Data Sheet Water Technologies Fire&Explosion Hazards: This material will not burn until moisture is removed, then resin starts to burn in flame at 230&deg;C.Under fire conditions some components of this product may decompose.
The smoke may contain unidentified toxic and/or irritating compounds.
Nitric acid and other strong oxidizing agents can cause explosive-type reactions when mixed with ion exchange resins.Proper design of equipment to prevent build up of pressure is necessary if use of an oxidizing agent such as nitric acid is contemplated.
Hazardous Products of Decomposition and/or Combustion:
May include but not limited to hydrocarbons, sulfur oxides, organic sulfonates, carbon monoxide, carbon dioxide and benzene compounds.
NFPA Ratings: HEAL TH-1 FLAMMABILlTY-1 REACTIVITY-1 OTHER-none SECTION6-ACCIDENTAL RELEASE MEASURES Spill/Leak Procedures:
Isolate spill area to prevent fa.lls as material can be a slipping hazard.Avoid contact with eyes and skin.Material is heavier than water and has limited water solubility.
It will collect on the lowest surface.Cleanup: Clean up floor area.Sweep up.Avoid generation of dust.Regulatory Requirements:
Follow all applicable Federal, State, Local, or Provincial regulations.
Disposal: DO NOT DUMP INTO ANY SEWERS, ON THE GROUND, OR INTO ANY BODY OF WATER.All disposal methods must be in compliance with all Federal, State, Local and Provincial laws and regulations.
Regulations may vary in different locations.
Waste characterizations and compliance with applicable laws are the responsibility solely of the waste generator.
SECTION7-HANDLING AND STORAGE Handling: Practice reasonable care and caution.Metal equipment should be compatible with feed, regenerant, resin form and effluent of that process.Storage: Keep containers tightly closed when not in use.Store between 2&deg;-38&deg;C100&deg;F).General Comments: Containers of this material may be hazardous when empty since they retain product residues (dust, solids);observe all warnings and precautions listed for the product.SECTION 8-PERSONAL PROTECTION!
EXPOSURE CONTROL Respiratory Protection:
No respiratory protection should be needed.Skin Protection:
Wear gloves impervious to this material to prevent skin contact.USF NR-39LC Ion Exchange Resin Page 3 of6 SIEMENS Material Safety Data Sheet Water Technologies Eye Protection:
Wear protective eyeglasses or chemical safety goggles.Contact lenses are not eye protective devices.Appropriate eye protecting must be worn instead of, or in conjunction with contact lenses.Ventilation Protection:
Good general ventilation should be sufficient.
Other Protection:
Never eat, drink, or smoke in work areas.Practice good personal hygiene after using this material, especially before eating, drinking, smoking, using the toilet, or applying cosmetics.
Safety showers, with quick opening valves which stay open, and eye wash fountains, or other means of washing the eyes with a gently flow of cool to tepid tap water, should be readily available in all areas where this material is handled or stored.Water should be supplied through insulated and heat-traced lines to prevent freeze-ups in cold weather.Exposure Limits: Exposure limits have not been developed.
SECTION9-PHYSICAL AND CHEMICAL PROPERTIES Appearance
&Odor: Spherical beads/Odorless to slight amine odor Vapor Pressure: N/A*Boiling Point: N/A Specific Gravity: N/D**Volatile Percentage:
N/A Flash Point/method:
N/A Vapor Density (Air=1): N/A Melting Point: N/A Solubility in Water: Insoluble pH: N/A Auto Ignition Temperature:
Above 500 0 C (900 0 F)Upper/Lower Explosion Limits: N/A Other: N/D*N/A=Not applicable
**N/D=Not determined SECTION 10-STABILITY AND REACTIVITY Stability:
Stable under normal handling and storage conditions.
Appendix A Attachment L Page:i of J!{Project No.29487-NCS0097.N.LAS EC DCR#375636 Incompatibilities:
Oxidizing agents such as nitric acid attack organic ion exchange under certain conditions and could result is slightly degraded resin up to an explosive reaction.Before using strong oXidizing agents, consult sources knowledgeable in handling such materials.
Polymerization:
Hazardous polymerization cannot occur.Decomposition:
Hazardous decomposition products depend upon temperature, air supply, and the presence of other materials.
Hazardous decomposition products may include and are not limited to: aromatic compounds, hydrocarbons, organic sultonates, sultur oxides.USF NR-39LC Ion Exchange Resin Page 4 of 6 SIEMENS Material Safety Data Sheet Water Technologies Conditions to Avoid: Resin can decompose at temperatures greater than 90 0 C (194 0 F).Do not pack column with dry ion exchange resins.Dry beads expand when wet.This expansion can cause a glass column to shatter.SECTION 11-TOXICOLOGICAL INFORMATION Inhalation
-Acute: Vapors are unlikely due to physical properties.
Inhalation
-Chronic: There are no known chronic inhalation effects.Skin Contact-Acute: Skin contact may cause mild irritation and redness.Skin Contact-Chronic: There are no known chronic dermal effects.Eye Contact-Acute: May cause severe eye irritation and redness.May cause moderate corneal injury.Effects are likely to heal.Ingestion-Acute: Single dose oral toxicity is considered to be low.No hazards anticipated from swallowing small amounts incidental to normal handling operation.
Swallowing large amounts may cause irritation to the gastrointestinal tract.Swallowing extremely large amounts may produce gastrointestinal disturbances.
Ingestion-Chronic: There are no known chronic ingestion effects.Carcinogenicity/Mutagenicity:
There are no known carcinogenic/mutagenic effects.Reproductive Effects: There are no known reproductive effects.Neurotoxicity:
There are no known neurotoxic effects.Other Effects: There are no other known toxic effects.Target Organs: Target organs include the eyes and skin.SECTION 12-ECOLOGICAL INFORMATION The environmental fate and ecological toxicity are not known.Appendix A Attachment L Page r-ofProject No.29487-NCS0097.N.LAS EC DCR#375636 SECTION 13-DISPOSAL CONSIDERATIONS Spill/Leak Procedures:
Isolate spill area to prevent falls as material can be a slipping hazard.Avoid contact with eyes and skin.Material is heavier than water and has limited water solubility.
It will collect on the lowest surface.Cleanup: Clean up floor area.Sweep up.Avoid generation of dust.Regulatory Requirements:
Follow all applicable Federal, State, Local, or Provincial regulations.
USF NR-39LC Ion Exchange Resin Page 50f6 SIEMENS Material Safety Data Sheet Water Technologies Disposal: DO NOT DUMP INTO ANY SEWERS, ON THE GROUND, OR INTO ANY BODY OF WATER.All disposal methods must be in compliance with all Federal, State Local and Provincial laws and regulations.
Regulations may vary in different locations.
Waste characterizations and compliance with applicable laws are the responsibility solely of the waste generator.
SECTION 14-TRANSPORTATION INFORMATION DOT Shipping

== Description:==

This product is not regulated by DOT when shipped domestically by land.Canadian TOG Information:
For TDG regulatory information, if required, consult transportation regulations, or product shipping.SECTION 15-REGULATORY INFORMATION US Regulations:
SARA HAZARD CATEGORY: This product has been reviewed according to the EPA"Hazard Categories" promUlgated under Sections 311 and 312 of the Superfund Amendment and Reauthorization Act of 1986 (SRA Title III)and is considered, under applicable definitions, to meet the following categories:
An immediate health hazard TSCA Considerations:
Every different salt or ionic form of an ion exchange resin is a separate chemical.If you use an ion exchange resin for ion exchange purposes and then remove theproduct resin from its vessel or container prior to recovery of the original or another form of the resin or of another chemical, the by-product resin must be listed on the TSCA Inventory (Unless an exemption is applicable).
It is the responsibility of the customer to ensure that such isolated, recycled by-product resins are in compliance with TSCA.Failure to comply could result in substantial civil or criminal penalties being assessed by the EPA.State RegUlations:
Consult individual state agency for further information.
Canadian RegUlations:
WHMIS INFORMATION:
The Canadian Workplace Hazardous Materials Information System (WHMIS)Classification for this product is: D2B-eye or skin irritant.Refer elsewhere in the MSDS for specific warnings and safe handling infonnation.
CPR Statement:
This product has been classified in accordance with the hazard criteria of the Canadian Controlled Products Regulations (CPR)and the MSDS contains all the information required by the CPR.SECTION 16-OTHER INFORMATION Disclaimer:
The information contained herein is based on data considered accurate.However, no warranty is expressed or implied regarding the accuracy of these data or the results to be obtained from the user thereof.It is the buyer's responsibility to ensure that its activities comply with federal, state, provincial and local laws.USF NR-39LC Ion Exchange Resin Appendix A Attachment L Page fL-of a Project No.29487-NCS0097.N.LAS EC DCR#375636 Page 6 of6 SIEMENS Material Safety Data Sheet Water Technologies SECTION1-CHEMICAL PRODUCT AND COMPANY INFORMATION Product Name: USF C-471C H Ion Exchange Resin Part Number: multiple Chemical Family: cation exchange polymer Manufacturer's Name: Siemens Water Technologies Corp.Address: 181 Thorn Hill Road, Warrendale, PA 15086 Productrrechnicallnformation Phone Number: (815)877-3041 Medical/Handling Emergency Phone Number: Call CHEMTREC at 800/424-9300 24 hours a day Transportation Emergency Phone Number: Call CHEMTREC at 800/424-9300 24 hours a day Issue Date: March 27, 2009 Revision Date/Revision Number: March 27, 2009/Rev 0 SECTION2-COMPOSITION INFORMATION Chemical Name Sulfonated copolymer of styrene and divinylbenzene in hydrogen form Water%By Weight CAS#40-70 069011-20-7 30-60 007732-18-5 SECTION3-HAZARDS IDENTIFICATION Appearance
&Odor: Spherical beads/Odorless to slight amine odor Emergency Overview:*May cause eye and skin irritation.
*May cause toxic fumeslvapors if burned.*May react violently when exposed to oxidizing agents such as Nitric Acid (HN0 3).Fire&Explosion Hazards: This material will not burn until moisture is removed, then resin starts to bum in flame at 230 0 C.Under fire conditions some components of this product may decompose.
The smoke may contain unidentified toxic and/or irritating compounds.
Nitric acid and other strong oxidizing agents can cause explosive-type reactions when mixed with ion exchange resins.Proper design of equipment to prevent build up of pressure is necessary if use of an oxidizing agent such as nitric acid is contemplated.
Primary Route(s)of Exposure: skin and eye contact Inhalation
-Acute Effects: Vapors are unlikely due to physical properties.
Skin Contact-Acute Effects: Skin contact may cause mild irritation and redness.Eye Contact-Acute Effects: May cause severe eye irritation and redness.May cause moderate corneal injury.Effects are likely to heal.Appendix A Attaclunent
_I Page'7 ofProject No.29487-NCS0097.N.LAS EC DCR#375636 SIEMENS Material Safety Data Sheet Water Technologies Ingestion-Acute Effects: Single dose oral toxicity is considered to be low.No hazards anticipated from swallowing small amounts incidental to normal handling operation.
Swallowing large amounts may cause irritation to the gastrointestinal tract.SECTION4-FIRST AID MEASURES Inhalation First Aid: Remove affected person from area to fresh air and provide oxygen if breathing is difficult.
Give artificial respiration ONLY if breathing has stopped and give CPR ONLY if there is no breathing and no pulse.Obtain medical attention.
No adverse effects anticipated by this route of exposure.Skin Contact First Aid: Immediately remove clothing from affected area and wash skin vigorously with flowing water.Clothing should be washed before reuse.Seek medical attention if irritation occurs.DO NOT instruct person to neutralize affected skin area.Eye Contact First Aid: Immediately irrigate eyes with flowing water continuously for 15 minutes while holding eyes open.Contacts should be removed before or during flushing.Obtain medical attention.
DO NOT instruct person to neutralize.
Ingestion First Aid: No adverse effects anticipated by this route of exposure incidental to proper industrial handling.If ingestion does occur, if victim is alert and not convulsing rinse mouth with water and give plenty of water to drink.If spontaneous vomiting occurs, have affected person lean forward with head down to avoid breathing in of vomitus.Rinse mouth again and give more water to drink.Obtain medical attention.
Medical Conditions Aggravated:
There are no known conditions aggravated by exposure.Note to Physician:
No specific antidote.Supportive care.Treatment based on judgment of the physician in response to reactions of the patient.SECTION5-FIRE FIGHTING MEASURES Flash Point/Method:
N/A Auto Ignition Temperature:
Above 500 0 C (900 0 F)Upper/Lower Explosion Limits: N/A Extinguishing Media: Water, carbon dioxide, dry chemical Appendix A Attachment
_l_Page L of--.!..X Project No.29487-NCS0097.N.LAS EC DCR#375636 Fire Fighting Procedures:
Keep people away.Isolate fire area and deny unnecessary entry.Cool surrounding area with water to localize fire zone.Soak thoroughly with water to cool and prevent reignition.
Fire-Fighting Equipment:
NIOSH approved positive-pressure self-contained breathing apparatus (SCBA)and protective fire fighting clothing (includes fire fighting helmet, coat, pants, boots and gloves).If protective equipment is not available or not used, fight fire from a protected location or a safe distance.USF C-471C H Ion Exchange Resin Page 2 of6 SIEMENS Material Safety Data Sheet Water Technologies Fire&Explosion Hazards: This material will not burn until moisture is removed, then resin starts to bum in flame at 230&deg;C.Under fire conditions some components of this product may decompose.
The smoke may contain unidentified toxic and/or irritating compounds.
Nitric acid and other strong oxidizing agents can cause explosive-type reactions when mixed with ion exchange resins.Proper design of equipment to prevent build up of pressure is necessary if use of an oxidizing agent such as nitric acid is contemplated.
Hazardous Products of Decomposition and/or Combustion:
May include but not limited to hydrocarbons, sulfur oxides, organic sulfonates, carbon monoxide, carbon dioxide and benzene compounds.
NFPA Ratings: HEALTH-1 FLAMMABILlTY-1 REACTIVITY-1 OTHER-none SECTION6-ACCIDENTAL RELEASE MEASURES Spill/Leak Procedures:
Isolate spill area to prevent falls as material can be a slipping hazard.Avoid contact with eyes and skin.Material is heavier than water and has limited water solubility.
It will collect on the lowest surface.Cleanup: Clean up floor area.Sweep up.Avoid generation of dust.Regulatory Requirements:
Follow all applicable Federal, State, Local, or Provincial regulations.
Disposal: DO NOT DUMP INTO ANY SEWERS, ON THE GROUND, OR INTO ANY BODY OF WATER.All disposal methods must be in compliance with all Federal, State, Local and Provincial laws and regulations.
Regulations may vary in different locations.
Waste characterizations and compliance with applicable laws are the responsibility solely of the waste generator.
SECTION7-HANDLING AND STORAGE Handling: Practice reasonable care and caution.Metal equipment should be compatible with feed, regenerant, resin form and effluent of that process.Storage: Keep containers tightly closed when not in use.Store between 2&deg;-38&deg;C (35&deg;100&deg;F).General Comments: Containers of this material may be hazardous when empty since they retain product residues (dust, solids);observe all warnings and precautions listed for the prodUCt.SECTION 8-PERSONAL PROTECTIONI EXPOSURE CONTROL Respiratory Protection:
No respiratory protection should be needed.Skin Protection:
Wear gloves impervious to this material to prevent skin contact.USF C-471C H Ion Exchange Resin Page 3 of6 SIEMENS Material Safety Data Sheet Water Technologies Eye Protection:
Wear protective eyeglasses or chemical safety goggles.Contact lenses are not eye protective devices.Appropriate eye protecting must be worn instead of, or in conjunction with contact lenses.Ventilation Protection:
Good general ventilation should be sufficient.
Other Protection:
Never eat, drink, or smoke in work areas.Practice good personal hygiene after using this material, especially before eating, drinking, smoking, using the toilet, or applying cosmetics.
Safety showers, with quick opening valves which stay open, and eye wash fountains, or other means of washing the eyes with a gently flow of cool to tepid tap water, should be readily available in all areas where this material is handled or stored.Water should be supplied through insulated and heat-traced lines topreventfreeze-ups in cold weather.Exposure Limits: Exposure limits have not been developed.
SECTION9-PHYSICAL AND CHEMICAL PROPERTIES Appearance
&Odor: Spherical beads/Odorless to slight amine odor Vapor Pressure: N/A*Boiling Point: N1A Specific Gravity: N/D**Volatile Percentage:
N/A Flash Point/method:
N/A Vapor Density (Air=1): N/A Melting Point: N/A Solubility in Water: Insoluble pH: N/A Auto Ignition Temperature:
Above 500 0 C (900 0 F)Upper/Lower Explosion Limits: N/A*N/A=Not applicable
**N/D=Not determined Other: N/D Appendix A Attachment L Page.LS:!of 1k.Project No.29487-NCS0097.N.LAS EC DCR#375636.SECTION 10-STABILITY AND REACTIVITY Stability:
Stable under normal handling and storage conditions.
Incompatibilities:
OxidiZing agents such as nitric acid attack organic ion exchange under certain conditions and could result is slightly degraded resin up to an explosive reaction.Before using strong oxidizing agents, consult sources knowledgeable in handling such materials.
Polymerization:
Hazardous polymerization cannot occur.Decomposition:
Hazardous decomposition products depend upon temperature, air supply, and the presence of other materials.
Hazardous decomposition products may include and are not limited to:aromaticcompounds, hydrocarbons, organic sulfonates, sulfur oxides.USF C-471C H Ion Exchange Resin Page 4 of6 SIEMENS Material Safety Data Sheet Water Technologies Conditions to Avoid: Resin can decompose at temperatures greater than 90&deg;C (194&deg;F).Do not pack column with dry ion exchange resins.Dry beads expand when wet.This expansion can cause a glass column to shatter.SECTION 11-TOXICOLOGICAL INFORMATION Inhalation
-Acute: Vapors are unlikely due to physical properties.
Inhalation
-Chronic: There are no known chronic inhalation effects.Skin Contact-Acute: Skin contact may cause mild irritation and redness.Skin Contact-Chronic: There are no known chronic dermal effects.Eye Contact-Acute: May cause severe eye irritation and redness.May cause moderate corneal injury.Effects are likely to heal.Ingestion-Acute: Single dose oral toxicity is considered to be low.No hazards anticipated from swallowing small amounts incidental to normalhandlingoperation.
Swallowing large amounts may cause irritation to the gastrointestinal tract.Swallowing extremely large amounts may produce gastrointestinal disturbances.
Ingestion-Chronic: There are no known chronic ingestion effects.Carcinogenicity/Mutagenicity:
There are no known carcinogenic/mutagenic effects.Reproductive Effects: There are no known reproductive effects.Neurotoxicity:
There are no known neurotoxic effects.Other Effects: There are no other known toxic effects.Target Organs: Target organs include the eyes and skin.SECTION 12-ECOLOGICAL INFORMATION The environmental fate and ecological toxicity are not known.Appendix A Attachment
-.L..-Page!..L of I f(Project No.29487-NCS0097.N.LAS EC DCR#375636 SECTION 13-DISPOSAL CONSIDERATIONS Spill/Leak Procedures:
Isolate spill area to prevent falls as material can be a slipping hazard.Avoid contact with eyes and skin.Material is heavier than water and has limited water solubility.
It will collect on the lowest surface.Cleanup: Clean up floor area.Sweep up.Avoid generation of dust.Regulatory Requirements:
Follow all applicable Federal, State, Local, or Provincial regulations.
USFC-471CH Ion Exchange Resin Page 50f6 SIEMENS Material Safety Data Sheet Water Technologies Disposal: DO NOT DUMP INTO ANY SEWERS, ON THE GROUND, OR INTO ANY BODY OF WATER.All disposal methods must be in compliance with all Federal, State Local and Provincial laws and regUlations.
RegUlations may vary in different locations.
Waste characterizations and compliance with applicable laws are the responsibility solely of the waste generator.
SECTION 14-TRANSPORTATION INFORMATION DOT Shipping

== Description:==

This product is not regulated by DOT when shipped domestically by land.Canadian TOG Information:
For TOG regulatory information, if required, consult transportation regulations, or product shipping.SECTION 15-REGULATORY INFORMATION US RegUlations:
SARA HAZARD CATEGORY: This product has beenreviewedaccording to the EPA"Hazard Categories" promUlgated under Sections 311 and 312 of the Superfund Amendment and Reauthorization Act of 1986 (SRA Title III)and is considered, under applicable definitions, to meet the following categories:
An immediate health hazard TSCA Considerations:
Every different salt or ionic form of an ion exchange resin is a separate chemical.If you use an ion exchange resin for ion exchange purposes and then remove theproduct resin from its vessel or container prior to recovery of the original or another form of the resin or of another chemical, the by-product resin must be listed on the TSCA Inventory (Unless an exemption is applicable).
It is the responsibility of the customer to ensure that such isolated, recycled by-product resins are in compliance with TSCA.Failure to comply could result in substantial civil or criminal penalties being assessed by the EPA.State RegUlations:
Consult individual state agency for further information.
Canadian RegUlations:
WHMIS INFORMATION:
The Canadian Workplace Hazardous Materials Information System (WHMIS)Classification for this product is: D2B-eye or skin irritant.Refer elsewhere in the MSDS for specific warnings and safe handling information.
CPR Statement:
This product has been classified in accordance with the hazard criteria of the Canadian Controlled Products RegUlations (CPR)and the MSDS contains all the information required by the CPR.SECTION 16-OTHER INFORMATION Disclaimer:
The information contained herein is based on data considered accurate.However, no warranty is expressed or implied regarding the accuracy of these data or the results to be obtained from the user thereof.It is the buyer's responsibility to ensure that its activities comply with federal, state, provincial and local laws.USF C-471C H Ion Exchange Resin Appendix A Attachment
----L Page I;;lof II Project No.29487-NCS0097.N.LAS EC DCR#375636 Page 6 of6 SIEMENS Material Safety Data Sheet Water Technologies SECTION1-CHEMICAL PRODUCT AND COMPANY INFORMATION Product Name: USF C-550C H Ion Exchange Resin Part Number: multiple Chemical Family: cation exchange polymer Manufacturer's Name: Siemens Water Technologies Corp.Address: 181 Thorn Hill Road, Warrendale, PA 15086 ProductlTechnicallnformation Phone Number: (815)877-3041 Medical/Handling Emergency Phone Number: Call CHEMTREC at 800/424-9300 24 hours a day Transportation Emergency Phone Number: Call CHEMTREC at 800/424-9300 24 hours a day Issue Date: June 23, 2009 RevisionDate/RevisionNumber:
June 23, 2009/Rev 1 SECTION2-COMPOSITION INFORMATION Chemical Name Sulfonated copolymer of styrene and divinylbenzene in hydrogen form Water%By Weight CAS#40-70 069011-20-7 30-60 007732-18-5 SECTION3-HAZARDS IDENTIFICATION Appearance
&Odor: Spherical beads/Odorless to slight amine odor Emergency Overview:*May cause eye and skin irritation.
*May cause toxic fumes/vapors if burned.*May react violently when exposed to oxidizing agents such as Nitric Acid (HN0 3).Fire&Explosion Hazards: This material will not burn until moisture is removed, then resin starts to burn in flame at 230 0 C.Under fire conditions some components of this product may decompose.Thesmoke may contain unidentified toxic and/or irritating compounds.
Nitric acid and other strong oXidizing agents can cause explosive-type reactions when mixed with ion exchange resins.Proper design of equipment to prevent build up of pressure is necessary if use of an oxidizing agent such as nitric acid is contemplated.
Primary Route(s)of Exposure: skin and eye contact Inhalation
-Acute Effects: Vapors are unlikely due to physical properties.
Skin Contact-Acute Effects: Skin contact may cause mild irritation and redness.Eye Contact-Acute Effects: May cause severe eye irritation and redness.May cause moderate corneal injury.Effects are likely to heal.Appendix A Attachment J Page 2.5o fProject No.29487-NCS0097.N.LAS EC DCR#375636 SIEMENS Material Safety Data Sheet Water Technologies Ingestion-Acute Effects: Single dose oral toxicity is considered to be low.No hazards anticipated from swallowing small amounts incidental to normal handling operation.
Swallowing large amounts may cause irritation to the gastrointestinal tract.SECTION4-FIRST AID MEASURES Inhalation First Aid: Remove affected person from area to fresh air and provide oxygen if breathing is difficult.
Give artificial respiration ONLY if breathing has stopped and give CPR ONLY if there is no breathing and no pulse.Obtain medical attention.
No adverse effects anticipated by this route of exposure.Skin Contact First Aid: Immediately remove clothing from affected area and wash skin vigorously with flowing water.Clothing should be washed before reuse.Seek medical attention if irritation occurs.DO NOT instruct person to neutralize affected skin area.Eye Contact First Aid: Immediately irrigate eyes with flowing water continuously for 15 minutes while holding eyes open.Contacts should be removed before or during flushing.Obtain medical attention.
DO NOT instruct person to neutralize.
Ingestion First Aid: No adverse effects anticipated by this route of exposure incidental to proper industrial handling.If ingestion does occur, if victim is alert and not convulsing rinse mouth with water and give plenty of water to drink.If spontaneous vomiting occurs, have affected person lean forward with head down to avoid breathing in of vomitus.Rinse mouth again and give more water to drink.Obtain medical attention.
Medical Conditions Aggravated:
There are no known conditions aggravated by exposure.Note to Physician:
No specific antidote.Supportive care.Treatment based on judgment of the physician in response to reactions of the patient.SECTION5-FIRE FIGHTING MEASURES Flash Point/Method:
N/A Auto Ignition Temperature:
Above 500 0 C (900 0 F)Upper/Lower Explosion Limits: N/A Extinguishing Media: Water, carbon dioxide, dry chemical Appendix A Attachment
_,_Page I Yof.1.Y Project No.29487-NCS0097.N.LAS EC DCR#375636 Fire Fighting Procedures:
Keep people away.Isolate fire area and deny unnecessary entry.Cool surrounding area with water to localize fire zone.Soak thoroughly with water to cool and prevent reignition.
Fire-Fighting Equipment:
NIOSH approved positive-pressure self-contained breathing apparatus (SCBA)and protective fire fighting clothing (includes fire fighting helmet, coat, pants, boots and gloves).If protective equipment is not available or not used, fight fire from a protected location or a safe distance.USF C-550C H Ion Exchange Resin Page 2 of6 SIEMENS Material Safety Data Sheet Water Technologies Fire&Explosion Hazards: This material will not burn until moisture is removed, then resin starts to burn in flame at 230&deg;C.Under fire conditions some components of this product may decompose.
The smoke may contain unidentified toxic and/or irritating compounds.
Nitric acid and other strong oxidizing agents can cause explosive-type reactions when mixed with ion exchange resins.Proper design of equipment to prevent build up of pressure is necessary if use of an oxidizing agent such as nitric acid is contemplated.
Hazardous Products of Decomposition and/or Combustion:
May include but not limited to hydrocarbons, sulfur oxides, organic sulfonates, carbon monoxide, carbon dioxide and benzene compounds.
NFPA Ratings: HEAL TH-1 FLAMMABILITY-1 REACTIVITY-1 OTHER-none SECTION6-ACCIDENTAL RELEASE MEASURES Spill/Leak Procedures:
Isolate spill area to prevent falls as material can be a slipping hazard.Avoid contact with eyes and skin.Material is heavier than water and has limited water solubility.
It will collect on the lowest surface.Cleanup: Clean up floor area.Sweep up.Avoid generation of dust.Regulatory Requirements:
Follow all applicable Federal, State, Local, or Provincial regulations.
Disposal: DO NOT DUMP INTO ANY SEWERS, ON THE GROUND, OR INTO ANY BODY OF WATER.All disposal methods must be in compliance with all Federal, State, Local and Provincial laws and regulations.
Regulations may vary in different locations.
Waste characterizations and compliance with applicable laws are the responsibility solely of the waste generator.
SECTION7-HANDLING AND STORAGE Handling: Practice reasonable care and caution.Metal equipment should be compatible with feed, regenerant, resin form and effluent of that process.Storage: Keep containers tightly closed when not in use.Store between 2&deg;-38&deg;C (35&deg;100&deg;F).General Comments: Containers of this material may be hazardous when empty since they retain product residues (dust, solids);observe all warnings and precautions listed for the product.SECTION B-PERSONAL PROTECTIONI EXPOSURE CONTROL Respiratory Protection:
No respiratory protection should be needed.Appendix A Attachment L Page/-.C ofProject No.29487-NCS0097.N.LAS EC DCR#375636 Skin Protection:
Wear gloves impervious to this material to prevent skin contact.USF C-550C H Ion Exchange Resin Page 3 of6 SIEMENS Material Safety Data Sheet Water Technologies Eye Protection:
Wear protective eyeglasses or chemical safety goggles.Contact lenses are not eye protective devices.Appropriate eye protecting must be worn instead of, or in conjunction with contact lenses.Ventilation Protection:
Good general ventilation should be sufficient.
Other Protection:
Never eat, drink, or smoke in work areas.Practice good personal hygiene after using this material, especially before eating, drinking, smoking, using the toilet, or applying cosmetics.
Safety showers, with quick opening valves which stay open, and eye wash fountains, or other means of washing the eyes with a gently flow of cool to tepid tap water, should be readily available in all areas where this material is handled or stored.Water should be supplied through insulated and heat-traced lines to prevent freeze-ups in cold weather.Exposure Limits: Exposure limits have not been developed.
SECTION9-PHYSICAL AND CHEMICAL PROPERTIES Appearance
&Odor: Spherical beads/Odorless to slight amine odor Vapor Pressure: N/A*Boiling Point: N/A Specific Gravity: N/D**Volatile Percentage:
N/A Flash Point/method:
N/A Vapor Density (Air=1): N/A Melting Point: N/A Solubility in Water: Insoluble pH: N/A Auto Ignition Temperature:
Above 500 0 C (900 0 F)Upper/Lower Explosion Limits: N/A Other: N/D*N/A=Not applicable
**N/D=Not determined SECTION 10-STABILITY AND REACTIVITY Stability:
Stable under normal handling and storage conditions.
Appendix A Attachment L Page IProject No.29487-NCS0097.N.LAS EC DCR#375636 Incompatibilities:
OXidiZing agents such as nitric acid attack organic ion exchange under certain conditions and could result is slightly degraded resin up to an explosive reaction.Before using strong oxidizing agents, consult sources knowledgeable in handling such materials.
Polymerization:
Hazardous polymerization cannot occur.Decomposition:
Hazardous decomposition products depend upon temperature, air supply, and the presence of other materials.
Hazardous decomposition products may include and are not limited to: aromatic compounds, hydrocarbons, organic sulfonates, sulfur oxides.USF C-550C H Ion Exchange Resin Page 4 of6 SIEMENS Material Safety Data Sheet Water Technologies Conditions to Avoid: Resin can decompose at temperatures greater than 90&deg;C (194&deg;F).Do not pack column with dry ion exchange resins.Dry beads expand when wet.This expansion can cause a glass column to shatter.SECTION 11-TOXICOLOGICAL INFORMATION Inhalation
-Acute: Vapors are unlikely due to physical properties.
Inhalation
-Chronic: There are no known chronic inhalation effects.Skin Contact-Acute: Skin contact may cause mild irritation and redness.Skin Contact-Chronic: There are no known chronic dermal effects.Eye Contact-Acute: May cause severe eye irritation and redness.May cause moderate corneal injury.Effects are likely to heal.Ingestion-Acute: Single dose oral toxicity is considered to be low.No hazards anticipated from swallowing small amounts incidental to normal handling operation.
Swallowing large amounts may cause irritation to the gastrointestinal tract.Swallowing extremely large amounts may produce gastrointestinal disturbances.
Ingestion-Chronic: There are no known chronic ingestion effects.Carcinogenicity/Mutagenicity:
There are no known carcinogenic/mutagenic effects.Reproductive Effects: There are no known reproductive effects.Neurotoxicity:
There are no known neurotoxic effects.Other Effects: There are no other known toxic effects.Target Organs: Target organs include the eyes and skin.SECTION 12-ECOLOGICAL INFORMATION The environmental fate and ecological toxicity are not known.Appendix A Attachment-I-Page Lr Project No.29487-NCS0097.N.LAS EC nCR#375636 SECTION 13-DISPOSAL CONSIDERATIONS Spill/Leak Procedures:
Isolate spill area to prevent falls as material can be a slipping hazard.Avoid contact with eyes and skin.Material is heavier than water and has limited water solUbility.
It will collect on the lowest surface.Cleanup: Clean up floor area.Sweep up.Avoid generation of dust.Regulatory Requirements:
Follow all applicable Federal, State, Local, or Provincial regulations.
USF C-550C H Ion Exchange Resin.Page 50f6 SIEMENS Material Safety Data Sheet Water Technologies Disposal: DO NOT DUMP INTO ANY SEWERS, ON THE GROUND, OR INTO ANY BODY OF WATER.All disposal methods must be in compliance with all Federal, State Local and Provincial laws and regulations.
Regulations may vary in different locations.
Waste characterizations and compliance with applicable laws are the responsibility solely of the waste generator.
SECTION 14-TRANSPORTATION INFORMATION DOT Shipping

== Description:==

This product is not regulated by DOT when shipped domestically by land.Canadian TOG Information:
For TOG regulatory information, if required, consult transportation regulations, or product shipping.SECTION 15-REGULATORY INFORMATION US Regulations:
SARA HAZARD CATEGORY: This product has been reviewed according to the EPA"Hazard Categories" promulgated under Sections 311 and 312 of the Superfund Amendment and Reauthorization Act of 1986 (SRA Title III)and is considered, under applicable definitions, to meet the following categories:
An immediate health hazard TSCA Considerations:
Every different salt or ionic form of an ion exchange resin is a separate chemical.If you use an ion exchange resin for ion exchange purposes and then remove theproduct resin from its vessel or container prior to recovery of the original or another form of the resin or of another chemical, the by-product resin must be listed on the TSCA Inventory (Unless an exemption is applicable).
It is the responsibility of the customer to ensure that such isolated, recycled by-product resins are in compliance with TSCA.Failure to comply could result in substantial civil or criminal penalties being assessed by the EPA.State Regulations:
Consult individual state agency for further information.
Canadian Regulations:
WHMIS INFORMATION:
The Canadian Workplace Hazardous Materials Information System (WHMIS)Classification for this product is: D2B-eye or skin irritant.Refer elsewhere in the MSDS for specific warnings and safe handling information.
CPR Statement:
This product has been classified in accordance with the hazard criteria of the Canadian Controlled Products RegUlations (CPR)and the MSDS contains all the information required by the CPR.SECTION 16-OTHER INFORMATION Disclaimer:
The information contained herein is based on data considered accurate.However, no warranty is expressed or implied regarding the accuracy of these data or the results to be obtained from the user thereof.It is the buyer's responsibility to ensure that its activities comply with federal, state, provincial and local laws.USF C-550C H Ion Exchange Resin Appendix A Attachment
----'-Page.L.X'Of..Lr Project No.29487-NCS0097.N.LAS EC DCR#375636 Page 6 of 6 TECHINICAL REPORT SUPPORTING ENGINEERING CHANGE Project No. 29487-NCS0097 EC No. 375636 Rev. 0 Date: 9/28/09 Appendix A - Page 12 of 12  ATTACHMENT 2 Miller, Chaz, "Profiles in Garbage:  High-Density Polyethylene, Waste Age", July 1, 2001 Profiles in Garbage: High-Density Polyethylene IOick to Print I http://www.printthis.clickability.com/pt/cpt?action=cpt&title=Profiles.PRINTTHIS Powered by t'i Oldcabillty SAVE lHlS I EMAIL lHlS I Close Profiles in Garbage: High-Density Polyethylene Jul1,2001 12:00 PM, By Chaz Miller Ilml&1 10 MY"YANoOl.I 10 u new59tltor I High-density polyethylene (HDPE)resin is produced from the chemical compound ethylene.Bottles are blow-molded while containers are injection-molded.
Milk bottles are the most common HDPE package.Most milk and water bottles use a natural-colored HDPE resin, while detergents, shampoos, motor oils and other products usually have added colorants.
Injection-molded HPDE containers also are used for products such as margarine and yogurt.HDPE bottles and containers use the No.2 plastic resin code.Additionally, HDPE resin can be used to make packaging products such as bottle caps, sacks and trash bags.HDPE bottles and containers began displacing heavier metal, glass and paper packages in the 1970s.Although the amount of HDPE used in packages has almost tripled since 1980, HDPE's solid waste market-share is less than 1 percent.HDPE bottles and containers are almost half of all HDPE packaging products.This profile only covers HDPE bottles and containers.
HDPE Municipal Solid Waste (MSW)Facts: Generated:
*1.92 million tons or 0.87%of total MSW by weight.**14.2 pounds per person in 1998.*85%are generated at home, 15%at businesses.
lof4 Recycled:*370,000 tons for a 19.27%recycling rate.**220,000 tons of liquid bottles for a 31.4%recycling rate.*Appendix A AttachmentPage-..l..of LI Project No.29487-NCS0097.N.LAS BC DCR#375636 8/3/2009 12:51 Prv Profiles in Garbage: High-Density Polyethylene http://www.printthis.clickability.com/pt/cpt?action=cpt&title=Profiles.
*150,000 tons of other HDPE bottles and containers for a 12.3%recycling rate.*Recycled Content:*Some nonfood HDPE containers include limited amounts of recycled HDPE.Composted:
*HDPE does not compost.Incinerated or Landf"illed:
*1.55 million tons or1%of discarded MSW by weight.**HDPE is highly combustible with 18,690 Btus in a pound of HDPE compared to 4,500 Btus to 5,000 Btus for a pound of MSW.This high Btu rate causes problems for boilers with low per-pound Btu ratings.*HDPE is not biodegradable in landfills.
Landf"ill Volume:*6.3 million cubic yards or 1.5%oflandfilled MSW.Density: Appendix A Attachment
:;J Pageof f-/Project No.29487-NCS0097.N.LAS EC DCR#375636 20f4*Landfilled milk jugs have a density of355 pounds per cubic yard.**Loose milk jugs have a density of 24 pounds per cubic yard.*Flattened milk jugs have a density of 65 pounds per cubic yard.*Loose,colored HDPE bottles have a density of 45 pounds per cubic yard.*Bales ofHDPE generally weigh 500 pounds to 800 pounds.Source Reduction:
*An empty 1 gallon milk jug decreased in weight from 95 grams in the early'70s to less than 60 grams today.*1,000 gallons of fruit juice can be packaged in 213 pounds ofHDPE containers.
This is less weight than that of competing package types, including other plastics.Recycling Markets: The packaging industry, which uses post-consumer recycled HDPE for bottles, is the largest HDPE recycling market.Drainage pipe, film, pallets and plastic lumber are other uses.HDPE also is exported, usually in bales with other plastics, to Pacific Rim processors.
Markets for clear (translucent)
HDPE milk bottles are the highest priced markets for HDPE bottles.8/3/2009 12:51 PlY Profiles in Garbage: High-Density Polyethylene End-Market Specifications:
http://www.printthis.clickability.com/pt/cpt?action=cpt&title=Profiles.
HDPE bottles fall under the Institute of Scrap Recycling Industries (I SRI), Washington, D.C., Baled Recycled Plastic Standard P-200 (HDPE Mixed), P-201 (HDPE Natural)or P-202 (HDPE Pigmented).
These specifications allow for 2%total contamination, no free flowing liquid and less than one month of outdoor storage unless the HDPE is covered with ultraviolet protective materials.
HDPE containers are not covered by any ISRI specifications.
These injection-molded containers can be incompatible with blow-molded bottles in reprocessing operations because the two types of packages have different melt flow indexes.Plastic processors take baled HDPE and separate the bottle components, i.e caps, labels and their adhesives.
They then are washed, dried and ground into HDPE flakes.Some processors produce pellets from the flakes.Recycling Cost and Value:*Collection costs range from$987 per ton to$1,401 per ton.*Processing costs range from$121.58 per ton to$256.15 per ton.Chaz Miller is director of state programs for the Environmental Industry Associations, Washington, D.C.E-mail the author at:cmiller@envasns.org.
To view additional Profiles in Garbage, visitwww.wasteage.com Sources: American Plastics Council, Washington, D.C.Website: www.ameriplas.org or www.plasticsresource.org"Design for Recycling, A Plastic Bottle Recyclers Perspective," Society of Plastics Industries, February 1992"Measurement Standards and Reporting Guidelines," National Recycling Coalition, Alexandria, Va.Website: www.nrc-recycle.org Modern Plastics, New York, February 2001.Website: www.modplas.com."Municipal Solid Waste Generation, Recycling and Disposal in the United States: Facts and Figures for 1998," U.S.Environmental Protection Agency (EPA), Washington, D.C.Website: www.epa.gov"Scrap Specifications Circular 1998," Institute of Scrap Recycling Industries, Washington, D.C.Website: www.lsn.org Waste Recyclers Council of the National Solid Wastes Management Association, Washington, D.C.Website: www.envasns.org/nswma 30f4*1998 U.S.EPA estimates.
Appendix A Attachment';;'
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40f4 Find this article at: http://WNN.wasteage.comlmag/waste....Profi les..Jjarbage
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/index html IOick to Print I o Check the box to include the list of links referenced in the article.to 2008 Penton Media, Inc.AU rights reserved.SAVE THIS I EMAIL THIS I Close Appendix A Attachment
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