ML23213A234

Submittal of Defueled Safety Analysis Report and Decommissioning Quality Assurance Plan

ML23213A234
Person / Time
Site:	Crane
Issue date:	08/01/2023
From:	Devik T TMI-2 Energy Solutions
To:	Office of Nuclear Material Safety and Safeguards, Document Control Desk
References
TMI2-RA-COR-2023-0019
Download: ML23213A234 (169)

Text

August 1, 2023 TMI2-RA-COR-2023-0019 10 CFR 50.71(e) 10 CFR 50.54(a)(3) 10 CFR 71.106(b)

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Three Mile Island Nuclear Station, Unit 2 (TMI-2)

NRC Possession Only License No. DPR 73 NRC Docket No. 50-320

Subject:

Submittal of Defueled Safety Analysis Report and Decommissioning Quality Assurance Plan

Reference:

Letter from Gerard P. Van Noordennen, EnergySolutions, to Document Control Desk, USNRC, Three Mile Island Nuclear Station, Unit 2, Update 14 of Post-Defueling Monitored Storage Safety Analysis Report, dated August 10, 2021 (ML21236A288)

In accordance with 10 CFR 50.71, Maintenance of records, making of reports, and 10 CFR 50.54, Conditions of licenses, TMI-2 Solutions, LLC (TMI-2S) is providing the updated Defueled Safety Analysis Report (DSAR), Revision 0, and Decommissioning Quality Assurance Plan (DQAP), Revision 20.

The last revision of the Post-Defueling Monitored Storage (PDMS) Safety Analysis Report (SAR) was submitted to the Nuclear Regulatory Commission (NRC) on August 10, 2021. The current revision supports the transition from PDMS to active decommissioning. This update reflects the current plant configuration and administrative processes. This revision replaces the PDMS SAR in its entirety.

The last revision of the PDMS Quality Assurance Plan (QAP) was submitted on August 10, 2021. The current revision supports the transition from PDMS to active decommissioning. The changes from the PDMS QAP to the DQAP were evaluated in accordance with 10 CFR 50.54(a)(3) and 10 CFR 71.106(b). These evaluations determined that there was no reduction in the TMI-2S QA Program commitments requiring prior NRC approval. This revision to the PDMS QAP includes the requirements that were relocated to the DQAP from the PDMS Technical Specifications. Changes within this revision are marked by vertical lines in the right-hand margin of the document.

This letter contains no new regulatory commitments.

TMI2-RA-COR-2023-0019 Page 2 In accordance with 10 CFR 50.91(b)(1), a copy of this submittal has been sent to the Commonwealth of Pennsylvania.

As required by 10 CFR 50.71(e)(2)(i), I certify that to the best of my knowledge, the information contained in the Enclosures to this letter accurately reflect information and analyses submitted to the NRC, or prepared pursuant to the NRC requirements as described above.

If you have any questions with respect to the contents of this letter, please contact me at (603) 384-0239 or trdevik@energysolutions.com.

Respectfully, Timothy R. Devik Licensing Manager TMI2Solutions cc:

NRC Regional Administrator - Region I NRC Lead Inspector - Three Mile Island Nuclear Station - Unit 2 NRC Project Manager - Three Mile Island Nuclear Station - Unit 2 Director, Bureau of Radiation Protection - PA Dept of Environmental Protection Chief, Division of Nuclear Safety, Bureau of Radiation Protection - PA Dept of Environmental Protection Chairman, Board of County Commissioners - Dauphin County Manager - Londonderry Township

Enclosures:

1. Defueled Safety Analysis Report, Revision 0

2. Decommissioning Quality Assurance Plan, Revision 20 Digitally signed by Timothy Devik DN: C=US, OU=Licensing, O=TMI-2 Solutions, CN=Timothy Devik, E=trdevik@energysolutions.com Reason: Approved Location: your signing location here Date: 2023-08-01 12:58:36 Foxit PhantomPDF Version: 9.7.5 Timothy Devik

118 Pages Follow Defueled Safety Analysis Report, Revision 0

TMI-2 DEFUELED SAFETY ANALYSIS REPORT TMI2-RA-LBD-2023-0003

TMI-2 Defueled Safety Analysis Report Revision 0 ii TABLE OF CONTENTS INTRODUCTION AND GENERAL DESCRIPTION OF PLANT............................................................................................................. 1-1

1.1 INTRODUCTION

.................................................................................................... 1-2 1.1.1 Post-Defueling Monitored Storage............................................................. 1-2 1.1.2 Decommissioning.......................................................................................... 1-2 1.1.3 Safety-Related Structures, Systems, and Components............................. 1-3 1.1.4 Important to Safety.................................................................................... 1-4 1.1.5 Terms Used to Describe the Damaged Special Nuclear Material at TMI-2

........................................................................................................................ 1-5 1.1.6 Relation of the TMI-2 DSAR to the Existing TMI-1 DSAR and TMI-2 FSAR............................................................................................................. 1-5 1.2 GENERAL PLANT DESCRIPTION.................................................................... 1-5 1.2.1 Site Characteristics...................................................................................... 1-6 1.2.2 Reactor Building.......................................................................................... 1-6 1.2.3 Fire Protection, Service, and Suppression................................................. 1-6 1.2.4 Radioactive Waste Management................................................................ 1-6 1.2.5 Radiation Monitoring.................................................................................. 1-7 1.2.6 Electrical Systems........................................................................................ 1-7 1.2.7 Cork Seam.................................................................................................... 1-7 1.2.8 Decommissioning Support Systems............................................................ 1-7 1.3 MATERIAL REFERENCED................................................................................. 1-8 1.4 PRESENTATION.................................................................................................. 1-11 Table 1.4-1, DSAR Figures.................................................................................... 1-11 Table 1.4-2, Abbreviations and Acronyms.......................................................... 1-12 Figure 1.2-1, General Arrangement of TMI-2 Cork Seam Area................................... 1-13 SITE CHARACTERISTICS...................................................... 2-1 2.1 GEOGRAPHY AND DEMOGRAPHY................................................................. 2-2 2.1.1 Site Location................................................................................................. 2-2 2.1.2 Site Description............................................................................................ 2-2 2.1.3 Population and Population Distribution.................................................... 2-2 2.2 NEARBY INDUSTRIAL, TRANSPORTATION, AND MILITARY FACILITIES............................................................................................................ 2-2

TMI-2 Defueled Safety Analysis Report Revision 0 iii 2.3 METEOROLOGY................................................................................................... 2-2 2.4 HYDROLOGIC ENGINEERING......................................................................... 2-3 2.4.1 Hydrologic Description................................................................................ 2-3 2.4.2 Floods............................................................................................................ 2-3 2.4.3 Flooding Protection Requirements............................................................. 2-3 2.4.4 Environmental Acceptance of Effluents.................................................... 2-4 2.4.5 Groundwater................................................................................................ 2-5 2.4.6 Flood Protection........................................................................................... 2-5 2.5 GEOLOGY AND SEISMOLOGY......................................................................... 2-5 2.6 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM........... 2-6 Figure 2.1-1, General Area Map......................................................................................... 2-7 Figure 2.1-2, Site Topographical Map, 5-Mile Radius..................................................... 2-8 Figure 2.1-3, Extended Plot Plan........................................................................................ 2-9 Figure 2.4-1, Flood Water Surface Profiles..................................................................... 2-10 Figure 2.4-2, Details of Outfall 001 Discharge System................................................... 2-11 DESIGN CRITERIA - STRUCTURES, SYSTEMS, AND COMPONENTS.............................................................................................. 3-1 3.1 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA................... 3-2 3.1.1 Criterion 1 - Quality Standards and Records........................................... 3-2 3.1.2 Criterion 2 - Design Bases for Protection Against Natural Phenomena 3-3 3.1.3 Criterion 3 - Fire Protection....................................................................... 3-3 3.1.4 Criterion 60 - Control of Releases of Radioactive Materials to the Environment................................................................................................. 3-4 3.1.5 Criterion 64 - Monitoring Radioactivity Releases.................................... 3-4 3.2 CLASSIFICATION OF STRUCTURES, SYSTEMS, AND COMPONENTS.. 3-4 3.2.1 Seismic Classification................................................................................... 3-4 Table 3.2-1, Category II Systems............................................................................ 3-7 3.3 MISSILE PROTECTION....................................................................................... 3-8 3.3.1 Missiles Generated by Natural Phenomena.............................................. 3-8 Table 3.3-1, Tornado-Generated Missiles.............................................................. 3-8 Table 3.3-2, Summary of Results.......................................................................... 3-10 3.3.2 Criteria and Design.................................................................................... 3-10 3.4 WATER LEVEL (FLOOD) DESIGN.................................................................. 3-11

TMI-2 Defueled Safety Analysis Report Revision 0 iv 3.4.1 Flood Elevation........................................................................................... 3-11 3.4.2 Phenomena Considered in Design Load Calculations............................ 3-11 3.4.3 Flood Force Application............................................................................ 3-11 3.4.4 Flood Protection......................................................................................... 3-11 3.5 HEAVY LOADS PROGRAM.............................................................................. 3-12 FUEL............................................................................................ 4-1

4.1 INTRODUCTION

.................................................................................................... 4-2

4.2 BACKGROUND

INFORMATION....................................................................... 4-2 4.2.1 Description of the March 1979 Accident................................................... 4-2 4.2.2 Post-Accident Conditions............................................................................ 4-3 Table 4.2-1, Core Material Inventory.................................................................... 4-5 4.3 SPECIAL NUCLEAR MATERIAL ACCOUNTABILITY AND CRITICALITY SAFETY ANALYSIS............................................................................................... 4-6 4.3.1 Introduction.................................................................................................. 4-6 4.3.2 Background.................................................................................................. 4-6 4.3.3 Historical SNM Accountability Process..................................................... 4-7 4.3.4 Post-Defueling SNM Accountability........................................................ 4-10 4.3.5 Criticality Analysis..................................................................................... 4-11 4.3.6 Control of SNM During PDMS................................................................ 4-11 4.3.7 Control of SNM During Decommissioning.............................................. 4-11 Table 4.3-1, Final SNM Inventory by Location - Reactor Building................. 4-13 Table 4.3-2, Final SNM Inventory by Location - Auxiliary and Fuel Handling Buildings..................................................................................................... 4-14 Table 4.3-3, Auxiliary and Fuel Handling Building Cubicle Designation........ 4-15 Figure 4.3-1, SNM Accountability Locations, Reactor Building, 282-6 Elevation... 4-19 Figure 4.3-2, SNM Accountability Locations, Reactor Building, 305-0 Elevation... 4-20 Figure 4.3-3, SNM Accountability Locations, Reactor Building, 347-6 Elevation... 4-21 Figure 4.3-4, Fuel Handling and Auxiliary Building, 280-6 Elevation...................... 4-22 Figure 4.3-5, Fuel Handling and Auxiliary Building, 305-0 Elevation...................... 4-23 Figure 4.3-6, Fuel Handling and Auxiliary Building, 328-0 Elevation...................... 4-24 Figure 4.3-7, Fuel Handling and Auxiliary Building, 347-6 Elevation...................... 4-25 RADIOLOGICAL CONDITIONS........................................... 5-1

5.1 INTRODUCTION

.................................................................................................... 5-2

TMI-2 Defueled Safety Analysis Report Revision 0 v

5.2 ENSURING OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLY ACHIEVABLE....................................................................... 5-2 5.2.1 ALARA Program......................................................................................... 5-2 5.3 PDMS RADIOLOGICAL SURVEY...................................................................... 5-3 5.3.1 Radiological Assessment.............................................................................. 5-3 5.3.2 Radiological Conditions at Beginning of PDMS....................................... 5-4 Table 5.3-1, PDMS Radiological Conditions - Reactor Building........................ 5-5 Table 5.3-2, PDMS Radiological Conditions - Auxiliary and Fuel Handling Buildings....................................................................................................... 5-6 Table 5.3-3, PDMS Radiological Conditions - Other Buildings........................ 5-12 Table 5.3-4, Surface Contamination - Reactor Building................................... 5-13 Table 5.3-5, Surface Contamination - Auxiliary and Fuel Handling Buildings.. 5-14 Table 5.3-6, Surface Contamination - Other Buildings..................................... 5-18 5.3.3 Reactor Building Drone Survey................................................................ 5-18 5.4 RADIATION PROTECTION PROGRAM........................................................ 5-19 5.4.1 Equipment, Instrumentation, and Facilities............................................ 5-19 5.4.2 Procedures.................................................................................................. 5-19 5.5 RADIOACTIVE WASTE MANAGEMENT...................................................... 5-20 5.5.1 Radioactive Waste - Miscellaneous Liquids (WDL) System................. 5-20 5.5.2 Sump Pump Discharge and Miscellaneous Sumps System.................... 5-21 Table 5.5-1, Operational Sump Systems.............................................................. 5-21 5.6 RADIATION MONITORING.............................................................................. 5-21 5.6.1 Function...................................................................................................... 5-21 5.6.2 Radiological Surveys.................................................................................. 5-22 5.6.3 Effluent Monitoring................................................................................... 5-22 5.6.4 General Radiological Monitoring............................................................. 5-22 5.7 SEALED SOURCES.............................................................................................. 5-22 Appendix 5A, Results of TMI-2 Reactor Building Drone Surveys............................... 5-23 Appendix 5B, Sealed Source Requirements.................................................................... 5-26 DEACTIVATED SYSTEMS AND FACILITIES................... 6-2 OPERATIONAL SYSTEMS AND FACILITIES................... 7-1

7.1 INTRODUCTION

.................................................................................................... 7-2

TMI-2 Defueled Safety Analysis Report Revision 0 vi 7.2 OPERATIONAL FACILITIES.............................................................................. 7-2 7.2.1 Reactor Building.......................................................................................... 7-2 7.2.2 Auxiliary Building........................................................................................ 7-2 7.2.3 Fuel Handling Building............................................................................... 7-3 7.2.4 Flood Protection........................................................................................... 7-3 7.2.5 Control and Service Buildings.................................................................... 7-3 7.2.6 Turbine Building.......................................................................................... 7-3 7.2.7 Containment Air Control Envelope........................................................... 7-3 7.3 OPERATIONAL SYSTEMS.................................................................................. 7-4 7.3.1 Reactor Building Systems............................................................................ 7-4 7.3.2 Fire Protection, Service, and Suppression................................................. 7-4 7.3.3 Electrical Systems........................................................................................ 7-5 7.3.4 Support Systems........................................................................................... 7-5 Table 7.3-1, Operational Systems........................................................................... 7-7 ROUTINE AND UNANTICIPATED RELEASES................. 8-1 8.1 GENERAL................................................................................................................ 8-2 8.1.1 Routine Releases........................................................................................... 8-2 8.1.2 Source Terms................................................................................................ 8-3 8.1.3 Unanticipated Releases................................................................................ 8-4 Table 8.1-1, Ci Fractions in Residual Fuel............................................................ 8-4 8.2 UNANTICIPATED EVENTS ANALYSIS........................................................... 8-4 8.2.1 Introduction.................................................................................................. 8-4 8.2.2 Dose Rate Limits for Fires and HEPA Filter Failure............................... 8-5 Table 8.2-1, Container Dose Rates for Each Release Scenario and Various Distances....................................................................................................... 8-6 Table 8.2-2, Building HEPA Filter Dose Rate Limits........................................... 8-6 8.2.3 Waste Handling Events............................................................................... 8-7 Table 8.2-3, 1 EC at Nearest Drinking Water PWST Activity Limits................ 8-8 8.2.4 Oxyacetylene Explosion............................................................................... 8-8 8.2.5 Other Events................................................................................................. 8-8 TECHNICAL SPECIFICATIONS........................................... 9-2 ADMINISTRATIVE FUNCTIONS...................................... 10-1

TMI-2 Defueled Safety Analysis Report Revision 0 vii

10.1 INTRODUCTION

.................................................................................................. 10-2 10.2 QUALITY ASSURANCE PLAN.......................................................................... 10-2 10.3 SECURITY PLAN................................................................................................. 10-3 10.4 EMERGENCY PLAN........................................................................................... 10-3 10.5 RADIATION PROTECTION PROGRAM........................................................ 10-4 10.6 ORGANIZATION................................................................................................. 10-4

TMI-2 Defueled Safety Analysis Report Chapter 1:

Revision 0 Introduction and General Description of Plant INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

TMI-2 Defueled Safety Analysis Report Chapter 1:

Revision 0 Introduction and General Description of Plant 1-2

1.1 INTRODUCTION

The Three Mile Island Unit 2 (TMI-2) operating license was issued on February 8, 1978, and commercial operation was declared on December 30, 1978. On March 28, 1979, the unit experienced an accident that resulted in severe damage to the reactor core. TMI-2 has been in a non-operating status since that time. GPU Nuclear (GPUN) conducted a substantial program to defuel the Reactor Vessel (RV) and decontaminate the facility. As a result, TMI-2 was defueled and decontaminated to the extent that the plant was placed in a safe, inherently stable condition suitable for long-term management, and any threat to public health and safety had been minimized. This long-term management condition, termed Post-Defueling Monitored Storage (PDMS), was entered in December 1993.

In December 2020, the Nuclear Regulatory Commission (NRC) approved the transfer of GPUN Possession Only License No. DPR-73 for Three Mile Island Nuclear Station (TMINS) Unit 2 to TMI-2 Solutions. In February 2021, TMI-2 Solutions submitted a License Amendment Request to the NRC to modify the TMI-2 Technical Specifications to permit the completion of the decommissioning of TMI-2.

1.1.1 POST-DEFUELING MONITORED STORAGE PDMS was established on December 28, 1993, based on three principal considerations:

A. The RV and the Reactor Coolant System (RCS) have been defueled and the core material has been shipped offsite.

B. Decontamination has been completed to the extent that further major decontamination programs are not justified on the basis of worker dose.

C. A condition of stability and safety has been established such that there is no risk to public health and safety.

1.1.2 DECOMMISSIONING Decommissioning efforts are largely divided into two phases. Phase 1 has been broken into two subphases. Phase 1A represents a continuation of PDMS, during which preparations for decommissioning will occur. Phase 1B focuses on remediation of the areas subject to core damage from the 1979 accident, with the overall goal to reduce the radiological source term at TMI-2 and the TMI-2 site to levels that are generally consistent with a nuclear plant toward the end of its operational life that has not experienced a core-damage accident.

A. Physical dismantlement and decontamination activities will begin in 2023.

Specific Phase 1B decommissioning objectives include but are not limited to:

1. Reduce the reactor building source term by removing the remaining Fuel Bearing Material (FBM) and remediating the most highly contaminated areas.

2. Package, transport, and store FBM at the ISFSI1.

1 The dry cask vendor for TMI-2 is pursuing a licensing designation for the proposed dry storage canisters which is not Spent Nuclear Fuel or Greater Than Class C (GTCC), but instead uses the term Fuel Bearing Material.

TMI-2 Defueled Safety Analysis Report Chapter 1:

Revision 0 Introduction and General Description of Plant 1-3

3. Reduce the source term of large components.

4. Remove, package, and dispose of Class A, Class B, and Class C radioactive waste.

B. The overall goal of Phase 2 is decommissioning of the TMI-2 site to a level that permits the release of the site for unrestricted use, except for an area to be set aside for long-term storage of Fuel Bearing Material (FBM). Specific Phase 2 decommissioning objectives include but are not limited to:

1. Remove, package, and dispose of all remaining systems and equipment in preparation for structural demolition.

2. Demolish and dispose of all plant structures.

3. Prepare and execute the License Termination Plan.

4. Backfill the site as needed.

1.1.3 SAFETY-RELATED STRUCTURES, SYSTEMS, AND COMPONENTS Due to the non-operating and defueled status of TMI-2 during decommissioning, there are no Structures, Systems, or Components (SSCs) classified as Safety-related at TMI-2. Safety-related SSCs are those which are necessary to ensure:

A. The integrity of the reactor coolant pressure boundary, B. The capability to shut down the reactor and to maintain it in a safe shutdown condition, or C. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guidelines of 10 CFR 100, Reactor Site Criteria (Ref. 1.3.9).

Criterion A requires maintenance of the reactor coolant pressure boundary. Due to the defueled condition of TMI-2, there is no reactor coolant or reactor coolant pressure boundary required.

Criterion B requires a capability to shut down the reactor and maintains it in a safe shutdown condition. In its current defueled state, there are no SSCs required to maintain a safe shutdown condition.

Criterion C requires a capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guidelines of 10 CFR 100, Reactor Site Criteria (Ref. 1.3.9). The analysis detailed in Chapter 8 demonstrates that there are no postulated events which could result in exposures comparable to the guidelines of 10 CFR 100, Reactor Site Criteria (Ref. 1.3.9).

TMI-2 Defueled Safety Analysis Report Chapter 1:

Revision 0 Introduction and General Description of Plant 1-4 1.1.4 IMPORTANT TO SAFETY Another commonly used term is Important to Safety. The most complete definition of what constitutes an Important to Safety SSC as defined in 10 CFR 50, Domestic Licensing of Production and Utilization Facilities (Ref. 1.3.3), is found in 10 CFR 50.49 (Ref. 1.3.3), which can be extended to all SSCs. Based on the discussion provided below, there are no Important to Safety SSCs at TMI-2. From this, Important to Safety can be defined as the following, taken from 10 CFR 50.49 (Ref. 1.3.3):

(1) Safety-related electric equipment (i) This equipment is that relied upon to remain functional during and following design basis events to ensure (A) The integrity of the reactor coolant pressure boundary; (B) The capability to shut down the reactor and maintain it in a safe shutdown condition; or (C) The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guidelines in § 50.34(a)(1), § 50.67(b)(2), or § 100.11 of this chapter, as applicable.

(ii) Design basis events are defined as conditions of normal operation, including anticipated operational occurrences, design basis accidents, external events, and natural phenomena for which the plant must be designed to ensure functions (b)(1)(i) (A) through (C) of this section.

(2) Nonsafety-related electric equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions specified in subparagraphs (b)(1)(i) (A) through (C) of paragraph (b)(1) of this section by the safety-related equipment.

(3) Certain post-accident monitoring equipment.

With respect to paragraph (1)(i) and as described in Section 1.1.4 of TMI-2 Post-Defueling Monitored Storage Safety Analysis Report (PDMS SAR) (Ref. 1.3.52),

there is no safety-related SSCs at TMI-2.

With respect to paragraph (1)(ii), as there is no Safety-related equipment at TMI-2, there are no applicable events for which TMI-2 needs to be designed to ensure the function of safety-related equipment.

With respect to paragraph (2), as Paragraph (1)(i) does not apply, there is no need for safety-related electrical equipment.

With respect to paragraph (3), it refers to specific guidance concerning the types of variables to be monitored is provided in Revision 2 of NRC Regulatory Guide 1.97, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident (Ref. 1.3.12). In review of this regulatory guide, the instrumentation referenced is that need to monitor design basis events, or in other words, events that could challenge items (1)(i)(A) through (1)(i)(C) above. There is no requirement for post-accident monitoring equipment.

TMI-2 Defueled Safety Analysis Report Chapter 1:

Revision 0 Introduction and General Description of Plant 1-5 Another measure of whether an SSC is Important to Safety is whether it is required by Technical Specifications. With NRC approval of the PDMS to DECON License Amendment, there will be no systems required by Technical Specifications; thus, this measure of Important to Safety is not satisfied.

As described in TMI2-QA-PG-001, Three Mile Island Unit 2 Decommissioning Quality Assurance Plan (DQAP) (Ref. 1.3.40), the concept of what is Important to Safety in 10 CFR 71, Packaging and Transportation of Radioactive Material (Ref. 1.3.5) differs from the concept in 10 CFR 50, Domestic Licensing of Production and Utilization Facilities, Appendix B (Ref. 1.3.3), described above.

When applying the criteria of 10 CFR 50, Appendix B (Ref. 1.3.3), at TMI-2, the unique shutdown and defueled status of the facility is considered.

1.1.5 TERMS USED TO DESCRIBE THE DAMAGED SPECIAL NUCLEAR MATERIAL AT TMI-2 Several terms are used to describe the damaged Special Nuclear Material (SNM) at TMI-2. Core Debris is used to describe approximately 99% of the damaged SNM removed during the cleanup program, which ended when TMI-2 was placed in PDMS. Fuel Bearing Material (FBM) is used to describe the remaining 1% of SNM and any accompanying material that will be placed in canisters and stored at the TMI-2 ISFSI. Residual Fuel is used to describe the UO2 portion of the FBM at TMI-

2.

1.1.6 RELATION OF THE TMI-2 DSAR TO THE EXISTING TMI-1 DSAR AND TMI-2 FSAR This TMI-2 DSAR makes reference to relevant portions of the TMI-1 Defueled Safety Analysis Report (TMI-1 DSAR) (Ref. 1.3.49) or the TMI-2 Final Safety Analysis Report (TMI-2 FSAR) (Ref. 1.3.51). The TMI-1 DSAR will continue to be updated as required, and the updated document will be applicable for those changing site-related conditions that have a bearing on TMI-2. The TMI-2 FSAR will not be updated but will continue to be applied as appropriate to TMI-2.

1.2 GENERAL PLANT DESCRIPTION On March 28, 1979, TMI-2 experienced an accident which severely damaged the reactor core. From late 1985 to early 1990, the reactor core was removed and shipped to the Idaho National Engineering Laboratory for analysis and long-term storage; the last shipment took place in April 1990. In addition, the facility was substantially decontaminated and was placed in a stable and benign condition suitable for long-term management.

TMI-2 was originally designed to comply with the seventy General Design Criteria of 10 CFR 50, Domestic Licensing of Production and Utilization Facilities, Appendix A, dated July 11, 1967 (Ref. 1.3.3), and addressed plant design with respect to the Revised General Design Criteria dated July 15, 1971, and similarly detailed in 10 CFR 50, Domestic Licensing of Production and Utilization Facilities, Appendix A (Ref. 1.3.3). Due to the present defueled status of TMI-2, many of these criteria no longer apply to the facility. A description of the applicable General Design Criteria, revised as of January 1, 1987, and listed in 10 CFR 50, Domestic Licensing of Production and Utilization Facilities, Appendix A (Ref. 1.3.3), is included in Section 3.1.

TMI-2 Defueled Safety Analysis Report Chapter 1:

Revision 0 Introduction and General Description of Plant 1-6 1.2.1 SITE CHARACTERISTICS As described in Chapter 2, the site is located on the Susquehanna River about ten miles southeast of Harrisburg, Pennsylvania. It is characterized by a 2,000-foot minimum exclusion distance; a two-mile radius low-population zone; sound bedrock as a structural foundation; an ample supply of emergency offsite power and favorable conditions of hydrology, geology, seismology, and meteorology. The land within a ten-mile radius of the site is used primarily for farming.

There are two airports within ten miles of the site. Harrisburg International Airport (formerly Olmsted State Airport) is located approximately two and one-half miles northwest of the site, and the Capitol City Airport is located approximately eight miles west-northwest of the site.

1.2.2 REACTOR BUILDING The Reactor Building is a reinforced concrete structure composed of cylindrical walls with a flat foundation mat and a dome roof lined with carbon steel. The structure provides biological shielding for normal and unanticipated conditions.

The steel liner encloses the equipment and systems which remain inside the Reactor Building.

1.2.3 FIRE PROTECTION, SERVICE, AND SUPPRESSION TMI-2 has been permanently defueled. The Fire Protection requirements of 10 CFR 50.48(f) (Ref. 1.3.3) apply; these requirements are described more thoroughly in NRC Regulatory Guide 1.191, Fire Protection Program for Nuclear Power Plants During Decommissioning and Permanent Shutdown (Ref. 1.3.11).

The TMI-2 Fire Protection Program meets the requirements of NRC Regulatory Guide 1.191, Fire Protection Program for Nuclear Power Plants During Decommissioning and Permanent Shutdown (Ref. 1.3.11). TMI2-FP-EVA-0001, TMI-2 Fire Protection Plan Evaluation (FPPE) (Ref. 1.3.37), describes how the program satisfies these requirements.

1.2.4 RADIOACTIVE WASTE MANAGEMENT A major component of the decommissioning work scope for TMI-2 is the packaging, transportation, and disposal of contaminated/activated equipment, piping, concrete, and soil. A waste management plan that is consistent with regulatory requirements and disposal/processing options for each waste type at the time of the decommissioning activities will be used.

Low-Level Radioactive Waste (LLRW) will be disposed at a site that is Licensed to receive it for disposal. LLRW from TMI-2 will be packaged to meet Department of Transportation (DOT) criteria for shipment and transported by licensed transporters. The waste management plan is based on the evaluation of available methods and strategies for processing, packaging, and transporting radioactive waste in conjunction with the available disposal facility options and associated Waste Acceptance Criteria (WAC).

Waste is planned to be transported largely by railroad in standard and specialty bulk packages, such as intermodal containers, and gondola-type rail cars.

TMI-2 Defueled Safety Analysis Report Chapter 1:

Revision 0 Introduction and General Description of Plant 1-7 1.2.5 RADIATION MONITORING Radiation monitors will be maintained operational to provide for evaluation of airborne radiological conditions. This requires monitoring the Reactor Building exhaust ventilation and the station vent during periods when a ventilation system is operating. The monitors will provide the necessary information to evaluate environmental releases and air quality conditions in the plant.

Monitoring and survey data will provide a basis for a trend analysis to ensure that radioactive contamination is controlled in the plant and enables timely corrective actions, if necessary.

1.2.6 ELECTRICAL SYSTEMS Portions of the TMI-2 AC and DC electrical systems will be maintained operational to provide reliable power to decommissioning support systems, controls, and instrumentation.

1.2.7 CORK SEAM The TMI-2 Cork Seam is a construction joint located between the various major structures at TMI-2 (see Figure 1.2-1, General Arrangement of TMI-2 Cork Seam Area). During the TMI-2 accident, the cork seam located in the Auxiliary Building Seal Injection Valve Room (SIVR) was contaminated with radioactive water. Since the accident, radioactive material has spread along the joint in one direction into the Annulus and in the other direction into the Auxiliary Building, Service Building, and Control Building. The radioactive contamination is prevented from entering the groundwater table by a Poly-Vinyl Chloride (PVC) water stop and thus minimizes the threat to the health and safety of the public.

A program is in place for continued monitoring of the water level in the cork seam by the plant staff. If the water levels begin to increase, the water will be pumped out for processing as needed.

1.2.8 DECOMMISSIONING SUPPORT SYSTEMS Other systems necessary to support decommissioning activities have been provided.

The ventilation systems for the Auxiliary Building, Fuel Handling Building, Control Building, and Service Building will remain operational to the extent needed to maintain an acceptable level of habitability in the areas and to mitigate the effects of temperature and humidity extremes. The systems may be replaced with suitable temporary local ventilation equipment if age and deterioration lead to failure of the original installed plant equipment. Building in-leakage will be collected in the various sumps and handled through the appropriate liquid waste disposal process.

TMI-2 Defueled Safety Analysis Report Chapter 1:

Revision 0 Introduction and General Description of Plant 1-8 1.3 MATERIAL REFERENCED The following documents are referenced as part of this application.

1.3.1 10 CFR 20, Standards for Protection Against Radiation.

1.3.2 10 CFR 37, Physical Protection of Category 1 and Category 2 Quantities of Radioactive Material.

1.3.3 10 CFR 50, Domestic Licensing of Production and Utilization Facilities.

1.3.4 10 CFR 70, Domestic Licensing of Special Nuclear Material.

1.3.5 10 CFR 71, Packaging and Transportation of Radioactive Material.

1.3.6 10 CFR 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste.

1.3.7 10 CFR 73.67, Licensee fixed site and in-transit requirements for the physical protection of special nuclear material of moderate and low strategic significance.

1.3.8 10 CFR 74, Material Control and Accounting of Special Nuclear Material.

1.3.9 10 CFR 100, Reactor Site Criteria.

1.3.10 NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors, Revision 6, dated November 2012.

1.3.11 NRC Regulatory Guide 1.191, Fire Protection Program for Nuclear Power Plants During Decommissioning and Permanent Shutdown.

1.3.12 NRC Regulatory Guide 1.97, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident, Revision 2.

1.3.13 NRC Regulatory Guide 7.10, Establishing Quality Assurance Programs for Packaging Used in Transport of Radioactive Material.

1.3.14 NRC Regulatory Guide 8.8, Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable.

1.3.15 NSAC 80-1, Analysis of Three Mile Island Unit 2 Accident, Electrical Power Research Institute, March 1980.

1.3.16 NUREG-0683, Final Programmatic Environmental Impact Statement Related to Decontamination and Disposal of Radioactive Wastes Resulting from March 28, 1979, Accident, Three Mile Island Nuclear Station, Unit 2, March 1981.

1.3.17 NUREG-0683, Supplement 3, Programmatic Environmental Impact Statement Related to Decontamination and Disposal of Radioactive Wastes Resulting from March 28, 1979 Accident, Three Mile Island Nuclear Station, Unit 2, August 1989.

1.3.18 NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition.

TMI-2 Defueled Safety Analysis Report Chapter 1:

Revision 0 Introduction and General Description of Plant 1-9 1.3.19 NUREG/CR-6407, Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety.

1.3.20 Rogovin, M., et al., Three Mile Island: A Report to the Commissioners and the Public. U.S. Nuclear Regulatory Commission, January 1980.

1.3.21 U.S. Department of Transportation, Bureau of Transportation Statistics, U.S.

General Aviation Safety Data and U.S. Air Carrier Safety Data; Website:

https://www.bts.gov/topics/national-transportation-statistics. Accessed 1/17/2023.

1.3.22 U.S. Department of Transportation, Federal Aviation Administration (FAA), The Operations Network (OPSNET). Website: https://aspm.faa.gov/opsnet/sys/main.asp.

Accessed 1/17/2023.

1.3.23 U.S. Environmental Protection Agency, EPA-400/R-17/001, PAG Manual:

Protective Action Guides and Planning Guidance for Radiological Incidents.

1.3.24 U.S. Environmental Protection Agency, National Pollution Discharge Elimination System (NPDES) Permit No. PA0009920, issued December 30, 1974.

1.3.25 U.S. Nuclear Regulatory Commission, Humboldt Bay Independent Spent Fuel Storage Installation Safety Evaluation Report, Docket No. 72-27, November 2005 (NRC Document ML053140041).

1.3.26 164090-EN-CALC-003, TMI-2 Source Term Limitations and Administrative Controls to Prevent Exceeding the Emergency Action Levels (EALs) for Zeolite Liner Drop and Processed Water Storage Tank (PWST) Rupture.

1.3.27 164090-EN-CALC-004, Source Term Limitations and Administrative Controls for the TMI-2 Decommissioning Emergency Plan Action Levels, Revision 0.

1.3.28 Calculation 6612-93-021, TMI-2 Waste Stream Update, Revision 0, dated June 25, 1993.

1.3.29 CY-AA-170-100, Radiological Environmental Monitoring Program (REMP).

1.3.30 CY-TM-170-300, Offsite Dose Calculation Manual (ODCM).

1.3.31 EP-TM-1002, Three Mile Island Independent Spent Fuel Storage Installation (ISFSI) Only Emergency Plan (IOEP).

1.3.32 GPU Nuclear 4000-ADM-4420.03, Review and Qualification of Selected Preliminary Calculations and Characterization Measurements for SNM Documentation.

1.3.33 GPU Nuclear 4000-PLN-4420.02, SNM Accountability Plan.

1.3.34 TMI2-EN-CALC-S-00-0011, TMI-2 Aircraft Impact of Reactor Building, Revision 0.

1.3.35 TMI2-EN-PN-102, TMI-2 Flood Protection Plan 1.3.36 TMI2-EN-RPT-001, Determination of the Safe Fuel Mass Limit for Decommissioning TMI-2, Revision 1.

1.3.37 TMI2-FP-EVA-0001, TMI-2 Fire Protection Plan Evaluation (FPPE).

1.3.38 TMI2-FP-PN-001, TMI-2 Fire Protection Plan.

TMI-2 Defueled Safety Analysis Report Chapter 1:

Revision 0 Introduction and General Description of Plant 1-10 1.3.39 TMI2-PM-PN-001, TMI-2 Project Management Plan.

1.3.40 TMI2-QA-PG-001, Three Mile Island Unit 2 Decommissioning Quality Assurance Plan (DQAP).

1.3.41 TMI2-QA-PG-001, Three Mile Island Unit 2 Post Defueling Monitored Storage (PDMS) Quality Assurance Plan (QAP).

1.3.42 TMI2-QA-PN-001, TMI-2 QA Program Implementation.

1.3.43 TMI2-RA-COR-2022-0008, Supplemental Information for License Amendment Request - Three Mile Island Unit 2 Decommissioning Technical Specifications, dated April 7, 2022.

1.3.44 TMI2-RA-COR-2022-0019, from Lackey, M. L., License Amendment Request -

Three Mile Island Unit 2 Decommissioning Technical Specifications, Response to Request for Additional Information, dated September 29, 2022.

1.3.45 TMI2-RA-COR-2023-0002, from Hazelhoff, A. C., Supplement to License Amendment Request - Proposed Changes to TMI-2 Possession Only License and Technical Specifications Attachment 1, dated January 27, 2023.

1.3.46 TMI2-RP1-PG-001, Radiological Protection Program.

1.3.47 TMI2-SE-PN-001, TMI-2 Materials Security Plan (SUNSI) 1.3.48 C-1101-122-E410-003, RO River Stage Discharge and Discharge Frequency Analysis, April 2012.

1.3.49 TMI-1 Defueled Safety Analysis Report (TMI-1 DSAR).

1.3.50 TMI-2 Defueling Completion Report.

1.3.51 TMI-2 Final Safety Analysis Report (TMI-2 FSAR).

1.3.52 TMI-2 Post-Defueling Monitored Storage Safety Analysis Report (PDMS SAR).

1.3.53 TMI-2 Technical Planning Bulletin 85-03, Reactor Building Basement Measurements, Revision 0, February 11, 1985.

1.3.54 TMI-2 Technical Specifications.

1.3.55 DOE Letter WWB-100-85, Bixby, W. W. (DOE) to Burton, H. M. (EG&G),

Accountability for the TMI-2 Core, dated October 8, 1985.

1.3.56 GPU Nuclear Letter C312-91-2045, SNM Accountability, transmitting the Auxiliary and Fuel Handling Buildings PDSR, dated June 7, 1991.

1.3.57 GPU Nuclear Letter C312-91-2052, SNM Accountability, transmitting the Reactor Building Miscellaneous Components PDSR, dated June 18, 1991.

1.3.58 GPU Nuclear Letter C312-91-2055, SNM Accountability, transmitting the Reactor Coolant System PDSR, dated July 3, 1991.

1.3.59 GPU Nuclear Letter C312-91-2064, SNM Accountability, transmitting the 'A' and 'B' Once-Through Steam Generators PDSR, Revision 1, dated July 3, 1991.

TMI-2 Defueled Safety Analysis Report Chapter 1:

Revision 0 Introduction and General Description of Plant 1-11 1.3.60 GPU Nuclear Letter C312-93-2004, SNM Accountability, transmitting the Reactor Vessel PDSR, dated February 1, 1993.

1.3.61 Letter, Snyder, B. J. (NRC) to Standerfer, F. R. (GPUNC), Approval of Exemption from 10 CFR 30.51,40.61,70.51(d) and 70.53, dated October 17, 1985.

1.3.62 NRC Letter, Request for Additional Information for Requested Licensing Action Regarding Decommissioning Technical Specifications, dated July 29, 2022.

1.3.63 TPO/TMI-051, Location and Characterization of Fuel Debris in TMI-2, Revision 0, April 1984.

1.3.64 TPO/TMI-124, Ex-Vessel Fuel Characterization, Revision 0, July 1984.

1.3.65 TPO/TMI-187, Instrument Selection for Residual Fuel Measurements, Revision 0, January 1987.

1.3.66 NRC Form 741, Nuclear Materials Transaction Report.

1.3.67 NRC Form 742, Materials Balance Report.

1.4 PRESENTATION DSAR figures used in this document are listed in Table 1.4-1, below.

TABLE 1.4-1, DSAR FIGURES Figure No.

Title 1.2-1 General Arrangement of TMI-2 Cork Seam Area 2.1-1 General Area Map 2.1-2 Site Topographical Map, 5-Mile Radius 2.1-3 Extended Plot Plan 2.4-1 Flood Water Surface Profiles 2.4-2 Details of Outfall 001 Discharge System 4.3-1 SNM Accountability Locations, Reactor Building 282-6 Elevation 4.3-2 SNM Accountability Locations, Reactor Building 305-0 Elevation 4.3-3 SNM Accountability Locations, Reactor Building 347-6 Elevation 4.3-4 Auxiliary Building, 280-6 Elevation 4.3-5 Auxiliary Building, 305-0 Elevation 4.3-6 Auxiliary Building, 328-0 Elevation 4.3-7 Auxiliary Building, 347 6 Elevation

TMI-2 Defueled Safety Analysis Report Chapter 1:

Revision 0 Introduction and General Description of Plant 1-12 Abbreviations and acronyms used in this document are listed in Table 1.4-2.

TABLE 1.4-2, ABBREVIATIONS AND ACRONYMS ABST Auxiliary Building Sump Tank MU Makeup and Purification System ALARA As Low As Is Reasonably Achievable MWHT Miscellaneous Waste Holdup Tank AX Auxiliary Building NPDES National Pollutant Discharge Elimination System BWST Borated Water Storage Tank NRC Nuclear Regulatory Commission CACE Containment Air Control Envelope NSAC Nuclear Safety Analysis Center CFR Code of Federal Regulations ODCM Offsite Dose Calculation Manual cfs/cfm Cubic Feet per Second/Cubic Feet per Minute OTSG Once Through Steam Generator Ci Curie PDMS Post-Defueling Monitored Storage CSA Core Support Assembly PDSR Post-Defueling Survey Report D&D Decontamination and Dismantlement PEIS Programmatic Environmental Impact Statement DCR Defueling Completion Report PMF Probable Maximum Flood DHR Decay Heat Removal PORV Pilot Operated Relief Valve DOE Department of Energy psi Pounds per Square Inch DQAP Decommissioning Quality Assurance Program PVC Poly-Vinyl Chloride DSAR Defueled Safety Analysis Report PWST Processed Water Storage Tank DWCS Defueling Water Cleanup System QA Quality Assurance FBM Fuel Bearing Material RB Reactor Building FHB Fuel Handling Building RC Reactor Coolant FPPE Fire Protection Program Evaluation RCS Reactor Coolant System FSAR Final Safety Analysis Report REMP Radiological Environmental Monitoring Program GPU General Public Utilities RV Reactor Vessel GPUN GPU Nuclear SAR Safety Analysis Report GPUNC General Public Utilities Nuclear Corporation SD Sump Discharge GTCC Greater Than Class C SDS Submerged Demineralizer System HEPA High Efficiency Particulate Air SFML Safe Fuel Mass Limit INEL Idaho National Engineering Laboratory SIVR Seal Injection Valve Room ISFSI Independent Spent Fuel Storage Installation SNM Special Nuclear Material IWTS Industrial Waste Treatment System SPC Standby Pressure Control LLEA Local Law Enforcement Agency SSC Structure, System, or Component LLRW Low Level Radioactive Waste TMINS Three Mile Island Nuclear Station LOCA Loss of Coolant Accident TRVFS Temporary Reactor Vessel Filtration System MCC Motor Control Center UTM Universal Transmeridian MDL Minimum Detectable level WAC Waste Acceptance Criteria mph Miles per Hour WDL Waste Disposal Liquid

TMI-2 Defueled Safety Analysis Report Chapter 1:

Revision 0 Introduction and General Description of Plant 1-13 FIGURE 1.2-1, GENERAL ARRANGEMENT OF TMI-2 CORK SEAM AREA

TMI-2 Defueled Safety Analysis Report Revision 0 SITE CHARACTERISTICS

TMI-2 Defueled Safety Analysis Report Chapter 2:

Revision 0 Site Characteristics 2-2 2.1 GEOGRAPHY AND DEMOGRAPHY 2.1.1 SITE LOCATION Three Mile Island Nuclear Station (TMINS) is located approximately 2.5 miles south of Middletown, Pennsylvania, at latitude 40°915 N and at longitude 76°4330 W. TMINS is located in Londonderry Township of Dauphin County, Pennsylvania, about 2.5 miles north of the southern tip of Dauphin County, where Dauphin is conterminal with York and Lancaster counties. Its location with respect to regional topographic and cultural features is shown on Figure 2.1-1, General Area Map, and with respect to local features on Figure 2.1-2, Site Topographical Map, 5 Mile Radius. The station is located on Three Mile Island (TMI), situated in the Susquehanna River upstream from York Haven Dam.

The TMI-2 RV coordinates are N300, 324.40, E2, 286, 366.04, based on the Pennsylvania State coordinate system (UTM coordinates are Zone 18, 4, 446, 020 meters north, 353, 070 meters east). It is one of the largest of a group of several islands in the Susquehanna River and is situated about 900 feet from the east bank.

It is elongated parallel to the flow of the river, with its 11,000-foot length and 1700-foot width. TMI-2 is located adjacent to TMI-1 in the northern one-third of the island.

The southeasterly-flowing Susquehanna River makes a sharp change in direction to nearly due south, in the vicinity of Middletown. After this directional change just north of TMINS, the channel widens to approximately 1.5 miles.

2.1.2 SITE DESCRIPTION Figure 2.1-3, Extended Plot Plan, shows the site marked to indicate the site and the minimum exclusion distance. For accident evaluations, the distance to the site boundary in each direction is used. Those distances may be derived from Figure 2.1-3.

2.1.3 POPULATION AND POPULATION DISTRIBUTION The population and population demographics are given in Section 2.2 of the TMI-1 DSAR (Ref. 1.3.49). This information is updated as appropriate with TMI-1 DSAR updates.

2.2 NEARBY INDUSTRIAL, TRANSPORTATION, AND MILITARY FACILITIES The nearby industrial, transportation, and military facilities are described in Chapter 2 of the TMI-1 DSAR (Ref. 1.3.49).

2.3 METEOROLOGY The meteorology for the TMINS site is given in Section 2.5 of the TMI-1 DSAR (Ref. 1.3.49). Since TMI-1 and TMI-2 are on the same site, the meteorological information for the two units is the same. The TMI-1 DSAR is updated periodically in accordance with NRC regulations and contains the currently applicable meteorological information for the TMINS site.

TMI-2 Defueled Safety Analysis Report Chapter 2:

Revision 0 Site Characteristics 2-3 2.4 HYDROLOGIC ENGINEERING 2.4.1 HYDROLOGIC DESCRIPTION A. Site and Facilities TMI-2 is located on Three Mile Island in the Susquehanna River, about ten miles southeast of Harrisburg. Existing grade elevation of TMI-2 is approximately 304 feet, based upon U.S.G.S. datum.

All necessary features of TMI-2 are protected from floods and wave action associated with these floods, up to and including the Probable Maximum Flood (1,625,000 cfs), by the installation of barriers as described in Section 2.4.3.

2.4.2 FLOODS The description of flooding conditions which were considered in the design of the site and facilities on the site is given in Section 2.4 of the TMI-2 FSAR (Ref. 1.3.51). These flooding conditions include the Probable Maximum Flood with coincident wind wave activity, the Probable Maximum Precipitation, potential dam failures, ice flooding, and channel diversions.

2.4.3 FLOODING PROTECTION REQUIREMENTS TMINS is situated on a portion of the island that is, under natural conditions, above the level of a Hurricane Agnes-magnitude flood. Natural topography in the main station area is above an elevation of 300.5 feet, the crest of the Agnes flood at approximately one million cubic feet per second.

The original design flood for the site was established at 1,100,000 cfs, based upon the provisional Probable Maximum Flood established by the Army Corps of Engineers prior to 1969.

In June of 1969, the Corps of Engineers issued a revised and provisional value of the Probable Maximum Flood for the Susquehanna River at Harrisburg, which was established as 1,600,000 cfs, as the result of upstream reservoir regulation. This design flood at TMINS was established at 1,625,000 cfs, and the water surface profiles on Figure 2.4-1, Flood Water Surface Profiles, were extended by computation to cover this flood magnitude. Unit computations established a PMF elevation of 313.3 feet; flood protection is designed to protect against this water level, per C-1101-122-E410-003, RO River Stage Discharge and Discharge Frequency Analysis, April 2012 (Ref. 1.3.48).

Structures are provided with complete protection at the entrance faces, rather than attempting to protect individual equipment or systems. Ground-level doors and entrances to the Concrete Power Block Buildings are either watertight or are provided with flood panels. All openings that are potential leaks (e.g., ducts, pipes, conduits, cable trays) are configured to minimize water intrusion. The water stops between adjacent building walls and mats were designed to be capable of withstanding a maximum water head of 45 feet, which is in excess of the maximum head associated with the flood level.

TMI-2 Defueled Safety Analysis Report Chapter 2:

Revision 0 Site Characteristics 2-4 Unit flood protection will be achieved by instituting operational procedures and actions predicated upon monitoring of upstream river stages and precipitation reports through the National Weather Service Advance Hydrologic Prediction service. Operational procedures to establish flood protection for TMI-2 are outlined in Section 2.4.6.

2.4.4 ENVIRONMENTAL ACCEPTANCE OF EFFLUENTS TMI-2 will continue discharging stormwater collected from non-radiological areas of TMINS through the NPDES Discharge Outfalls (e.g., Outfall 005, SWRO-1, SWRO-2, and SWRO-3). Also, TMI-2 will collect stormwater/groundwater from various (non-radiological) sumps and store it in a holding tank(s), where it will be sampled and analyzed to confirm the collected water meets the NPDES discharge criteria. Once confirmed acceptable, the collected water will be discharged through Outfall 001, which enters the middle channel of the Susquehanna River through the station discharge pipe located approximately 640 feet downstream from the previous intake structures, as shown in Figure 2.4-2, Details of Outfall 001 Discharge System.

Processed liquid waste will be discharged on a batch basis. Prior to release, each batch will be sampled and analyzed to determine its radioactivity content. Based upon the activity analysis, the wastes will either be released under controlled conditions or recycled for further processing. The flowrate of waste discharge will be a function of NPDES Permit No. PA0009920 (Ref. 1.3.24).

A section of the York Haven Dam blocks the east channel of the Susquehanna River at Three Mile Island, approximately one mile downstream from TMI-2. The York Haven Dam forms a pool extending approximately 3.5 miles upstream, containing a volume of about 10,000 acre-feet. As long as the river flow is 20,000 cfs or less, all flow discharges through the York Haven Hydro Plant tailrace into the lower section of Conewago Falls. When the river flow is above 20,000 cfs, the excess flow spills over the portion of the main dam upstream of the headrace wall and flows down through Conewago Falls, joining the flow from the tailrace at the foot of the dam; the full river flow then continues through the lower section of the falls.

The exact extent of the mixing of TMINS effluents with the river depends on such factors as station discharge flowrate and the river flowrate. In 1980, an experiment was conducted which tracked the dispersion and dilution of a dye from the TMINS discharge. This study showed that the plant discharge water and Susquehanna River water are typically greater than or equal to () 99% mixed before intake by downstream users.

All users of water downstream of TMINS are also downstream of York Haven Dam. Therefore, it is assured that mixing of station effluent and river water flow will occur prior to use.

TMI-2 Defueled Safety Analysis Report Chapter 2:

Revision 0 Site Characteristics 2-5 2.4.5 GROUNDWATER TMINS is located in the Triassic lowland of Pennsylvania, a region often referred to as the Gettysburg Basin. The island was formed as a result of fluvial deposition by the Susquehanna River. It is composed of subrounded to rounded sand and gravel, containing varying amounts of silt and clay. Soil depths vary from approximately 6 feet at the south of the island to about 30 feet at the center of island.

The site is underlain by Gettysburg shale, at an elevation of approximately 277 feet.

There are two different water-bearing zones at TMINS. One is comprised of the Gettysburg shale (i.e., bedrock), and the other is comprised of the unconsolidated material overlying the bedrock. Permeabilities in the geological materials on TMINS vary; however, groundwater discharges into the Susquehanna River and does not communicate with offsite groundwater supplies.

Hydrostatic pressure of the water table on the east and west shores of the river should prevent the island groundwater and the station discharge from communications with onshore groundwater. Therefore, groundwater effluents from TMINS cannot impact the quality of groundwater offsite. Additionally, the tritium concentrations in TMINS groundwater are well below the regulatory limits of 10 CFR 20, Standards for Protection Against Radiation (Ref. 1.3.1), and will not adversely affect the Susquehanna River.

A more thorough description of the groundwater characteristics and related features are given in Section 2.4.13 of the TMI-2 FSAR (Ref. 1.3.51).

2.4.6 FLOOD PROTECTION The emergency flood protection procedure, TMI2-EN-PN-102, TMI-2 Flood Protection Plan (Ref. 1.3.35), is entered when Susquehanna River flow at Harrisburg exceeds 200,000 cfs or the National Weather Service (NWS) Forecast Center forecasts a Susquehanna River flow greater than 350,000 cfs within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The Susquehanna River forecast for flow and river level at Harrisburg are obtained from the National Weather Service Advance Hydrologic Prediction Service. The forecast should always provide a minimum of a 48-hour forecast and is updated as required every six (6) hours during a potential flood condition. For this purpose, the flow at TMINS is assumed to be equal to the flow at Harrisburg. The river flow and level forecast are based upon both measured rainfall and rainfall forecasts across the Susquehanna watershed.

When a 36-hour forecast of 900,000 cfs or greater is received, flood panels will be moved into place.

2.5 GEOLOGY AND SEISMOLOGY The geology and seismology for the TMINS site has been reviewed and accepted by the NRC, based on the information presented in Section 2.5 of the TMI-2 FSAR (Ref. 1.3.51).

TMI-2 Defueled Safety Analysis Report Chapter 2:

Revision 0 Site Characteristics 2-6 2.6 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM The Radiological Environmental Monitoring Program (REMP) is being conducted in the vicinity of TMINS. The REMP provides representative measurements of radiation and radioactive materials in exposure pathways and for those radionuclides that lead to the highest potential radiation exposures of members of the public, resulting from facility activities. This monitoring program complies with 10 CFR 50, Domestic Licensing of Production and Utilization Facilities, Appendix I (Ref. 1.3.3), and thereby supplements the radiological effluent monitoring program by verifying measurable concentrations of radioactive materials and levels of radiation are not higher than expected based on effluent measurements and models of the environmental exposure pathways.

TMI-2 Defueled Safety Analysis Report Chapter 2:

Revision 0 Site Characteristics 2-7 FIGURE 2.1-1, GENERAL AREA MAP (BEST COPY AVAILABLE)

TMI-2 Defueled Safety Analysis Report Chapter 2:

Revision 0 Site Characteristics 2-8 FIGURE 2.1-2, SITE TOPOGRAPHICAL MAP, 5-MILE RADIUS (BEST COPY AVAILABLE)

TMI-2 Defueled Safety Analysis Report Chapter 2:

Revision 0 Site Characteristics 2-9 FIGURE 2.1-3, EXTENDED PLOT PLAN (BEST COPY AVAILABLE)

TMI-2 Defueled Safety Analysis Report Chapter 2:

Revision 0 Site Characteristics 2-10 FIGURE 2.4-1, FLOOD WATER SURFACE PROFILES (BEST COPY AVAILABLE)

TMI-2 Defueled Safety Analysis Report Chapter 2:

Revision 0 Site Characteristics 2-11 FIGURE 2.4-2, DETAILS OF OUTFALL 001 DISCHARGE SYSTEM (BEST COPY AVAILABLE)

TMI-2 Defueled Safety Analysis Report Revision 0 DESIGN CRITERIA -

STRUCTURES, SYSTEMS, AND COMPONENTS

TMI-2 Defueled Safety Analysis Report Chapter 3: Design Criteria -

Revision 0 Structures, Systems, and Components 3-2 3.1 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA During the Decontamination and Decommissioning (DECON) period, fulfillment of many of the general design criteria in 10 CFR 50, Domestic Licensing of Production and Utilization Facilities, Appendix A (Ref. 1.3.3), is not necessary or appropriate for TMI-2, given that the criteria that addresses such requirements as quality standards, radiological monitoring, and natural phenomena only apply to a limited degree. Since the plant was originally designed and constructed in accordance with these criteria, and since neither the accident nor activities during the recovery period significantly degraded the plant with respect to the capabilities required, the facility, as it currently exists, is designed and constructed to standards which far exceed the requirements for decommissioning. These capabilities will be removed over time as source term is removed in accordance with 10 CFR 50.59 (Ref. 1.3.3) and proper procedural controls.

Compliance with TMI-2 Solutions understanding of the general design criteria in 10 CFR 50, Domestic Licensing of Production and Utilization Facilities, Appendix A, as revised on January 1, 1987 (Ref. 1.3.3), that remain applicable in the defueled condition are summarized in the following sections.

3.1.1 CRITERION 1 - QUALITY STANDARDS AND RECORDS Structures, Systems, and Components (SSCs) important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety function to be performed. Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keeping with the required safety function. A Quality Assurance (QA) program shall be established and implemented in order to provide adequate assurance that these SSCs will satisfactorily perform their safety functions. Appropriate records of the design, fabrication, erection, and testing of SSCs important to safety shall be maintained by or under control of the nuclear power unit licensee throughout the life of the unit.

Discussion Due to the unique condition of TMI-2, the specific requirements of Criterion 1 are not applicable; however, the intent of Criterion 1 has been addressed, recognizing that the degree of quality assurance necessary to assure that the required capabilities are maintained is far less extensive than that which was originally required for TMI-2. A QA program has been established and will be maintained commensurate with the functional requirements of decommissioning. The QA plan for decommissioning is described in Section 10.1 and detailed in the DQAP (Ref. 1.3.40).

TMI-2 Defueled Safety Analysis Report Chapter 3: Design Criteria -

Revision 0 Structures, Systems, and Components 3-3 3.1.2 CRITERION 2 - DESIGN BASES FOR PROTECTION AGAINST NATURAL PHENOMENA SSCs important to safety shall be designed to withstand the effect of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunamis, and seiches without loss of capability to perform their safety functions. The design bases for these SSCs shall reflect:

A. Appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data has been accumulated, B. Appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena, and C. The importance of the safety functions to be performed.

Discussion Due to the unique condition of TMI-2, the specific requirements of Criterion 2 are not applicable; however, the intent of Criterion 2 has been addressed by recognizing the level of protection from natural phenomena required during decommissioning.

There are no active functions required to be performed by any system to provide protection from natural phenomena. Those SSCs necessary for the level of protection required were originally designed and constructed to criteria which exceed the current requirements. This level of protection is more than adequate to meet the functional requirements for protection from natural phenomena.

3.1.3 CRITERION 3 - FIRE PROTECTION SSCs important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions.

Noncombustible and heat-resistant materials shall be used whenever practical throughout the unit, particularly in locations such as the containment and control room. Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fire on SSCs important to safety. Fire-fighting systems shall be designed to assure that their rupture or inadvertent operation does not significantly impair the safety capability of these SSCs.

Discussion In the permanently defueled condition of TMI-2, the specific requirements of Criterion 3 are not applicable. TMI-2 has been permanently defueled. The Fire Protection requirements of 10 CFR 50.48(f) (Ref. 1.3.3) apply. The requirements of 10 CFR 50.48(f) are described more thoroughly in NRC Regulatory Guide 1.191, Fire Protection Program for Nuclear Power Plants During Decommissioning and Permanent Shutdown (Ref. 1.3.11). TMI2-FP-PN-001, TMI-2 Fire Protection Plan (Ref. 1.3.38), meets the requirements of NRC Regulatory Guide 1.191, Fire Protection Program for Nuclear Power Plants During Decommissioning and Permanent Shutdown (Ref. 1.3.11). The FPPE (Ref. 1.3.37), describes how the program satisfies these requirements.

TMI-2 Defueled Safety Analysis Report Chapter 3: Design Criteria -

Revision 0 Structures, Systems, and Components 3-4 3.1.4 CRITERION 60 - CONTROL OF RELEASES OF RADIOACTIVE MATERIALS TO THE ENVIRONMENT The nuclear power unit design shall include means to suitably control the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced during normal reactor operation, including anticipated operational occurrences. Sufficient holdup capacity shall be provided for retention of gaseous and liquid effluents containing radioactive materials, particularly where unfavorable site environmental conditions can be expected to impose unusual operational limitations upon the release of such effluents to the environment.

Discussion Provision is included in the design and processing of radioactive waste and the release of such waste under controls adequate to prevent exceeding the limits of 10 CFR 20, Standards for Protection Against Radiation (Ref. 1.3.1). TMI-2 also includes programmatic controls to prevent radioactivity releases during accidents from exceeding the limits of the Emergency Plan. A description of the function of the Radioactive Waste Disposal System is included in Chapter 5. The effects of potential accidents are analyzed in Chapter 8.

3.1.5 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 operation (including anticipated operational occurrences) and from postulated accidents.

Discussion The facility contains means for monitoring the Reactor Building atmosphere, effluent discharge paths, and the facility environs for radioactivity which could be released under any conditions. The details of the effluent discharge monitoring are contained in Chapter 5, while the REMP (Ref. 1.3.29) is discussed in Chapter 2.

3.2 CLASSIFICATION OF STRUCTURES, SYSTEMS, AND COMPONENTS 3.2.1 SEISMIC CLASSIFICATION Unit SSCs have been classified according to their function and the degree of integrity required to protect the public. SSCs are classified for seismic design purposes as either Category I or Category II. The original seismic design criteria for SSCs are given in Section 3.2 of the TMI-2 FSAR (Ref. 1.3.51).

TMI-2 Defueled Safety Analysis Report Chapter 3: Design Criteria -

Revision 0 Structures, Systems, and Components 3-5 A. Seismic Category I Seismic Category I SSCs, including instruments and controls, are those which are necessary to ensure:

1. The integrity of the reactor coolant pressure boundary,

2. The capability to shut down the reactor and to maintain it in a safe shutdown condition, or

3. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline exposures of 10 CFR 100, Reactor Site Criteria (Ref. 1.3.9).

The first criterion requires the reactor coolant pressure boundary to be ensured.

Due to the non-operating and defueled condition of TMI-2, there is no reactor coolant pressure boundary.

The second criterion requires the capability to shut down the reactor and maintain it in a safe shutdown condition. Due to the non-operating and defueled condition of TMI-2, there is no reactor to shut down and maintain in a safe shutdown condition. In addition, the criticality analysis discussed in Chapter 4 demonstrates that criticality has been precluded at TMI-2.

The third criterion requires the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guidelines of 10 CFR 100, Reactor Site Criteria (Ref. 1.3.9).

The analysis detailed in Chapter 8 demonstrates that there are no postulated events which could result in exposures comparable to the guidelines of 10 CFR 100, Reactor Site Criteria (Ref. 1.3.9).

Due to the non-operating and defueled status of TMI-2, there are no SSCs which are required to meet the criteria of Seismic Category I. Therefore, there are no SSCs classified as Seismic Category I at TMI-2. Those SSCs which were originally designed as Seismic Category I and the applicable design criteria are described in Section 3.2.1 of the TMI-2 FSAR (Ref. 1.3.51).

B. Category II Those SSCs which are relied upon for the isolation of residual contamination from the environment and for the prevention of an uncontrolled release of radioactivity during a seismic event have been designated as Category II.

1. Design Basis All SSCs which were designed to Seismic Category I or Category II standards are in conformance with the seismic loading requirements of the Uniform Building Code for Zone 1. Those SSCs that were originally designed and fabricated to Seismic Category I or Seismic Category II requirements currently have not had their original structural design capability significantly modified nor have they significantly degraded and will meet the seismic loading requirements of the Uniform Building Code for Zone 1. In the future however, changes will be made to structures as decommissioning progresses and engineering modifications will be prepared.

TMI-2 Defueled Safety Analysis Report Chapter 3: Design Criteria -

Revision 0 Structures, Systems, and Components 3-6

2. Category II Structures Structures designated as Category II are the Reactor Building, the Control and Service Buildings, the Auxiliary Building, and the Fuel Handling Building. These five structures were originally designed in accordance with Seismic Category I requirements. During structure removal, control measures will be put in place in accordance with TMI2-RP1-PG-001, Radiological Protection Program (Ref. 1.3.46), to control the spread of contamination.

3. Category II Systems As a result of the accident on March 28, 1979, various plant systems were contaminated with radioactive materials from the reactor core. These systems have been decontaminated to the extent practical. However, a number of these systems still contain some degree of residual contamination and provide an initial barrier for control of the contamination which remains within the respective system. The Reactor Building provides a second barrier for those systems or partial systems located within that structure.

The structural capabilities of the systems as originally designed in accordance with Seismic Category I and Seismic Category II criteria have not been degraded significantly, and these capabilities provide the necessary degree of contamination control until they are removed during decommissioning. During system removal, control measures will be put in place in accordance with the Radiation Protection Program to control the spread of contamination. These Category II systems and their original seismic design criteria are listed in Table 3.2-1, Category II Systems.

Other piping systems designated as Category II which were not designed to Seismic Category I or Seismic Category II criteria also provide an initial barrier for control of internal contamination. Since these systems were designed to function under equal or higher pressure and temperature conditions, they will maintain structural integrity under the conditions which exist. In addition, deactivated systems have been drained to the extent practical, thus minimizing contaminated leakage from these systems until they are removed during decommissioning. These Category II systems are listed in Table 3.2-1 and are identified as not being designed to seismic criteria (i.e., Non-Seismic [N/S]). During system removal. Control measures will be put in place in accordance with the Radiation Protection Program to control the spread of contamination.

TMI-2 Defueled Safety Analysis Report Chapter 3: Design Criteria -

Revision 0 Structures, Systems, and Components 3-7 TABLE 3.2-1, CATEGORY II SYSTEMS System FSAR Seismic Class In Reactor Building Operational/

Deactivated Canister Loading Decon N/S No Deactivated Core Flood I (1)

Partially (2)

Deactivated Decant Filter Skid N/S No Deactivated Decay Heat Removal I

Partially Deactivated Decon Process Water N/S Partially Deactivated Defueling (3)

N/S Yes Deactivated Defueling Water Cleanup N/S Partially Deactivated Dewatering Station N/S No Deactivated Fuel Handling I

Partially Deactivated Fuel Transfer Canal Fill and Drain N/S Partially Deactivated Main Steam I (4)

Partially Deactivated Makeup and Purification I

Partially Deactivated Nuclear Plant Nitrogen II Partially Deactivated Nuclear Sampling I (5)

Partially Deactivated Processed Water Storage and Distribution N/S No Deactivated Reactor Building Purge Exhaust Upstream of HEPA Filter II No Operational Reactor Building Spray I

Partially Deactivated Reactor Building Sump (Level Measurement)

N/S Partially Operational Reactor Coolant I

Yes Deactivated Sludge Transfer N/S Partially Deactivated Spent Fuel I

Partially Deactivated Steam Generator Recirculation N/S No Deactivated Steam Generator Secondary Vent and Drain inside Reactor Building I/II Yes Deactivated Submerged Demineralizer N/S No Deactivated Sump Water Sucker N/S Partially Deactivated Temporary Nuclear Sampling N/S (6)

No Deactivated Waste Disposal Gas I (7)

Partially Deactivated Waste Disposal Liquid I/II Partially Operational Waste Disposal Solid I

Partially Deactivated Notes:

(1)

Except for N2 supply lines from NM-U-26/NM-U-27 to CF-V-114/CF-V-115.

(2)

All of system is in the Reactor Building except for 1-inch fill and drain lines.

(3)

System consists of the defueling equipment that remains in the RV.

(4)

Classified as Seismic Category I up to MSWs and EF-P-2. All else is classified as Seismic Category

2. Portions of Seismic Category II piping are contaminated.

(5)

Majority of system is Seismic Category I. Seismic Category II portion is very small and contains a small pump and associated 1.5-inch tubing.

(6)

System consists of mainly 1/2-inch and 3/8-inch diameters.

(7)

System was designed and constructed as Seismic Category I, except for two flexible connections to the secondary vent and drain system. A few N/S modifications were made during the course of the cleanup effort; however, the majority of the system remains seismically qualified.

TMI-2 Defueled Safety Analysis Report Chapter 3: Design Criteria -

Revision 0 Structures, Systems, and Components 3-8 3.3 MISSILE PROTECTION 3.3.1 MISSILES GENERATED BY NATURAL PHENOMENA A. Applicable Design Parameters The forces due to tornado loading have been assumed as the forces associated with a wind having a velocity of 360 mph. The 360-mph velocity is considered as a resultant of a 300-mph tangential and 60-mph translational velocity of the storm. All other structures designed for tornado loadings are provided either with adequate areas of openings to relieve the differential pressure of 3 psi in three (3) seconds or are designed to withstand an external pressure of 3 psi.

B. Tornado Missiles The missiles assumed to have been generated by the tornado event are listed in Table 3.3-1, Tornado-Generated Missiles. They include items such as siding panel, pipe, and steel plate, which could be detached from structures under tornado-associated loadings. All these missiles have been investigated for their ability to penetrate exterior concrete walls and slabs. The maximum penetration of a concrete barrier was found to be 30 inches. The minimum thickness of structures designed to withstand tornado-generated missiles is 36 inches. Their effects on the buildings designed for tornado loadings were considered together with the effects of loadings described in Section 3.3.1(A). Thus, with the limited exceptions described in Section 3.3.2, the capability of buildings designed for tornado loadings will not be jeopardized as a result of flying objects from other structures under a tornado event.

TABLE 3.3-1, TORNADO-GENERATED MISSILES Missile Type Weight (lb.)

Impact Area (sq. ft.)

Impact Elevation Above Grade (ft)

Impact Velocity (mph)

Utility Pole 1200 1.0 25 200 Passenger Auto 2000 25 25 100 Passenger Auto 4000 30 3

100 Concrete Fragment (10x 3 x 4) 4500 30 5

60 Steel Plate 1000 1.0 10 200 Crated Motor 1000 15 10 200 Pipe (4 x 12) 250 0.15 25 200 Wood Plank (4 x 12 x 12) 110 0.33 Any Height 360 Street Light Fixtures 25 0.5 Any Height 360 Crushed Rock (1-1/2) 0.25 0.01 Any Height 360 Siding Panels (1-6 x 50) 400 0.04 Any Height 360 Tools 125 0.1 Any Height 360 Ductwork 150 0.3 Any Height 360 8 Handrail Section 50 0.05 Any Height 360

TMI-2 Defueled Safety Analysis Report Chapter 3: Design Criteria -

Revision 0 Structures, Systems, and Components 3-9 C. Aircraft Impact The TMI-2 FSAR (Ref. 1.3.51) determined the probability of an aircraft impact on the TMI-2 reactor site considering three aircraft categories: a large aircraft, a very large aircraft, and a small aircraft. This calculation was based on estimates made in the 1968 Evaluation, which examined historical flight data collected over a ten-year period from 1956 through 1965. This data was surveyed periodically to determine if there is any need to revise the original estimates, but this evaluation has not been updated recently.

TMI2-EN-CALC-S-00-0011, TMI-2 Aircraft Impact of Reactor Building (Ref. 1.3.34), reproduced those calculations using the same methodology as the 1968 Evaluation but replaced historical flight data using information from recent operational reports (see U.S. Department of Transportation, Federal Aviation Administration [FAA], The Operations Network [OPSNET]

[Ref. 1.3.22]) and safety statistics from the U.S. Department of Transportation, Bureau of Transportation Statistics, U.S. General Aviation Safety Data and U.S.

Air Carrier Safety Data (Ref. 1.3.21).

An overview of the original analysis of record is contained in the TMI-2 FSAR (Ref. 1.3.51); however, key portions of that analysis were reproduced within TMI2-EN-CALC-S-00-0011, TMI-2 Aircraft Impact of Reactor Building (Ref. 1.3.34), as applicable, and for completeness and clarity. Flight information used in this calculation, contain flight data within the last ten years to determine an accurate representation of operational and accident data from all U.S. safety data and local airport operations.

The probability of an aircraft impact determined in this evaluation is then compared to the acceptance criteria for aircraft hazards as presented within NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (Ref. 1.3.18), and the U.S. NRC Humboldt Bay Independent Spent Fuel Storage Installation Safety Evaluation Report, Docket No. 72-27 (NRC Document ML053140041) (Ref. 1.3.25).

The Humboldt Bay Safety Evaluation Report (SER) (Ref. 1.3.25) concluded that, when compared to a nuclear reactor facility, an ISFSI is a relatively passive system that does not have complex control requirements and has contents with relatively low thermal energy. Consequently, potential fuel damage and the associated radioactive source terms from a potential accident at an ISFSI are significantly less than those expected from a potential accident at an operating nuclear reactor facility, and as a result, the estimated consequences are less severe. The staff, therefore, concludes that a probability of 1 x 10-6 crashes per year is an acceptable threshold probability criterion for evaluating aircraft crash hazards at the ISFSI.

TMI-2 Defueled Safety Analysis Report Chapter 3: Design Criteria -

Revision 0 Structures, Systems, and Components 3-10 Similar to the Humboldt Bay ISFSI, the TMI-2 Reactor Building is considered a passive system, as it is no longer in operation and its primary function is to prevent any radiological release excess of the dose guidelines found in 10 CFR 100, Reactor Site Criteria (Ref. 1.3.9). Therefore, an acceptance criterion of 1 x 10-6 crashes per year is used as the acceptance criteria when evaluating the cumulative probability of an aircraft impact on the TMI-2 Reactor Building.

Estimates for the probability of various types of aircraft accident scenarios and the approximate recurrence interval are summarized below in Table 3.3-2, Summary of Results.

TABLE 3.3-2,

SUMMARY

OF RESULTS Accident Scenario Calculation Section Approximate Impact Probability/year (Unit 2 Reactor Building)

Approximate Recurrence Interval/year (Unit 2 Reactor Building)

Probability of Impact by Large Aircraft 6.1 5.1 x 10-8 19.6 million years Probability of Impact by a Very Large Aircraft 6.2 5.0 x 10-9 200 million years Probability of Impact by Small Aircraft 6.3 9.091 x 10-7 1.1 million years Cumulative Probability of Aircraft Impact 6.4 9.651 x 10-7 1.036 million years For each of the three accident scenarios shown above, and the cumulative probability of an aircraft impact, the probability of occurrence was less than the acceptance criteria of 1.0 x 10-6 and is therefore a non-credible hazard.

3.3.2 CRITERIA AND DESIGN The missile protection criterion is based on precluding damage to SSCs important to safety. During decommissioning, there is no equipment that is important to safety; therefore, the consequences of a missile impact are limited to physical damage. With limited exceptions as described below, all contamination isolation structures are protected from a loss of function due to damage from potential external missiles. Due to the non-operating and defueled status of TMI-2, there are no postulated internally generated missiles which would have any effect on the capability of structures to provide their required contamination isolation function.

The Auxiliary Building is rectangular in plan and has three main floors. At the east exterior wall, a large door opening is located at the grade level. This door opening is not protected from external missiles. Similarly, the Reactor Building Equipment Hatch will be replaced with a large door opening at grade level that is not protected from external missiles.

TMI-2 Defueled Safety Analysis Report Chapter 3: Design Criteria -

Revision 0 Structures, Systems, and Components 3-11 In addition, concrete drilling operations in a small exterior wall section of the Auxiliary Building damaged six (6) reinforcing steel bars. Nonetheless, the wall presently meets normal non-operating load requirements. These two areas comprise the limited exceptions to wind and tornado loadings and missile protection.

3.4 WATER LEVEL (FLOOD) DESIGN Flood history, flood design consideration, design of hydraulic facilities, and emergency operation requirements for the unit facilities are further discussed in Section 2.4.

3.4.1 FLOOD ELEVATION The foundation mats, exposed walls, and flood panels of those structures designed to withstand floods are designed to withstand the hydrostatic pressures associated with a water level of 312 ft. on the east side and 312.5 ft. on the west side.

3.4.2 PHENOMENA CONSIDERED IN DESIGN LOAD CALCULATIONS Due to wave action, the maximum water level considered is 4 ft. higher than the PMF elevation expected around the station facilities. No other phenomena have been considered in the design load calculations. For more information, refer to Section 2.4.3 of the TMI-2 FSAR (Ref. 1.3.51) 3.4.3 FLOOD FORCE APPLICATION Static forces on the flood panels and vertical walls are calculated considering hydrostatic pressures. Dynamic forces on the flood panels and vertical walls are calculated considering wave action. With the exception of the Decommissioning Support Building (DSB), the buoyancy forces on the mats are calculated considering the uplift due to the height of water above the bottom of the mats.

3.4.4 FLOOD PROTECTION Unit design is based on a water elevation of 308.5 ft. on the west side and 308 ft.

on the east side under flood conditions. Structures which originally contained Engineering Safety Feature equipment are sealed against entry of flood water to an elevation of 312.5 ft. on the west side and 312 ft. on the east side. Complete protection has been provided at the exterior faces of these structures. The water stops between adjacent building walls and mats are capable of withstanding a maximum water head of 45 feet, which is in excess of the maximum associated head for the flood level. The exterior sliding doors and flood panels are provided with watertight seals.

TMI-2 Defueled Safety Analysis Report Chapter 3: Design Criteria -

Revision 0 Structures, Systems, and Components 3-12 3.5 HEAVY LOADS PROGRAM To provide a high degree of assurance that a load drop into the RV will not occur, TMI-2 Solutions has developed a hoisting and rigging program that addresses movement of loads at TMI-2. The purpose of the hoisting and rigging program is to define the minimum requirements for the safe operations of cranes and hoists. The hoisting and rigging program will provide, as applicable, detailed requirements for training and qualification of personnel, inspection and maintenance of cranes or hoists, the safe use of rigging equipment, as well as direction for Non-Standard Lifts in order to ensure that lifting operations are performed in a safe manner. A Non-Standard Lift has characteristics that require additional planning and performance efforts to ensure that the lift is performed in a safe manner. A lift plan will be developed for all lifts as directed by the hoisting and rigging program, where a load drop or load impingement could contribute to release or dispersal of radioactive material to the environment, which could exceed the threshold for an unusual event.

Implementation of the hoisting and rigging program provides a defensive, in-depth approach to preventing a load drop from occurring. Crane design features (e.g., load cells, travel stops) will be employed as required in order to ensure safe travel paths. Barriers will be provided as required by the lift plan to preclude the effects of a load drop.

TMI-2 Defueled Safety Analysis Report Revision 0 FUEL

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-2

4.1 INTRODUCTION

This chapter summarizes the conditions and activities associated with the TMI-2 core subsequent to the accident on March 29, 1979.

Section 4.2 provides a brief description of the accident and the post-accident core conditions as they became known through the defueling process. Section 4.3 provides summary descriptions of the Special Nuclear Material Accountability Programs and Criticality Analyses performed for TMI-2.

This discussion is relevant because the Fuel Bearing Material (FBM) that remains is directly dependent on the core conditions after the accident and the level of success achieved by the activities undertaken to remove core debris during the cleanup program.

The residual fuel conditions are provided in the Post-Defueling Survey Reports (PDSRs);

this information is summarized on Table 4.3-1, Final SNM Inventory by Location -

Reactor Building, and Table 4.3-2, Final SNM Inventory by Location - Auxiliary and Fuel Handling Buildings. The analysis demonstrating assured subcriticality is summarized in Section 4.3. With the exception of Sections 4.3.5 and 4.3.7, this information is historical and will not be updated during decommissioning.

4.2 BACKGROUND

INFORMATION 4.

2.1 DESCRIPTION

OF THE MARCH 1979 ACCIDENT The March 1979 accident was initiated by cessation of secondary feedwater flow.

The steam generator boiled dry, and the resultant reduction of primary-to-secondary heat transfer caused the primary coolant to heat up and increase the primary system pressure. The Pilot Operated Relief Valve (PORV) opened to relieve pressure. Primary system pressure continued to rise, causing a reactor trip, and the PORV failed to close when the pressure decreased. The first 100 minutes of the accident can be characterized as a small break loss-of-coolant accident with resultant loss of primary coolant and decreasing pressure. It differed from the scenario expected during a Loss of Coolant Accident (LOCA) in that the pressurizer liquid level indication remained high. This was interpreted by the reactor operation as indicating that the RCS was full of water when, in fact, the RCS was continually voiding. Up to about 100 minutes into the accident, the core was still covered with sufficient water to be cooled.

The Reactor Coolant Pumps were turned off at 100 minutes and core heat-up began as the water level decreased to elevations below the top of the core. By 150 minutes, a Zircaloy-steam exothermic reaction was occurring, which increased the core heat-up rate. Consequently, Zircaloy melting temperatures were exceeded, resulting in relocation of the molten Zircaloy and some liquefied fuel to the lower core regions, solidifying near the coolant interface. This condition prevailed until 174 minutes, at which time a large region of consolidated, degraded core material existed in the lower, central regions of the core. Coolant flow through this consolidated material was probably negligible. The intact fuel rod stubs in the lower core region indicate that only the lower portion of the core remained cool.

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-3 A Reactor Coolant Pump was turned on briefly at 174 minutes, and coolant was pumped into the RV. The resultant thermal-mechanical forces generated from the rapid steam formation are believed to have shattered the oxidized fuel rod remnants in the upper regions of the core, forming a rubble bed on top of the consolidated core materials. The consolidated core materials continued to heat up during the next 50 minutes (174-224 minutes into the accident), even though coolant delivery to the RV from the pump transient and emergency core cooling injection is estimated to have covered the core by approximately 210 minutes. By 224 minutes, much of the non-cooled consolidated region had reached temperatures sufficient to melt the U-Zr-O ternary mixture.

Online TMI-2 data recorded during the accident suggests that the crust surrounding the consolidated core failed between 224 and 226 minutes into the accident. Based on the end-state core and core support assembly configuration and supporting analysis of the degraded core heat-up, it is believed that as the crust failed, molten core material flowed around the bypass region and migrated down into the lower internals and lower head region. Limited damage to the Core Support Assembly (CSA) occurred as the core material flowed to the lower plenum. It is estimated that about 17-20 tons of material relocated to the lower internals and lower head region.

Several in-core instrument guide tubes were melted, but overall, vessel integrity was maintained throughout the accident.

4.2.2 POST-ACCIDENT CONDITIONS The post-accident conditions inside the TMI-2 RV are summarized in the following sections.

A. Core Inventory The original core inventory included approximately 207,300 lb. of fuel (i.e.,

UO2) and 75,400 lb. of structural and absorber material for a total of 282,700 lb. Including the material added as a result of damage to the RV internal components due to the dynamics of the accident and the material generated by defueling activities, the total post-accident core debris material was estimated to be 296,100 lb. This estimate does not include any portions of the CSA and grid structures beneath the core that may have partially melted and mixed with the core material (see Table 4.2-1, Core Material Inventory).

Due to the accident progression, some of this core material was relocated within and outside of the RV. Each of the relevant areas is discussed separately. During the defueling program, approximately 99% of this material was removed.

B. Upper Core Region This region covers approximately the top 8 ft. of the original core height (13.8 ft.). The bottom boundary of this region was the layer of resolidified material identified during debris bed probing and observed during the core bore activities.

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-4 The material in this region consisted of partial, standing peripheral assemblies, partial assemblies hanging from the underside of the plenum, loose debris, fuel assembly, and control element structural components such as end fittings and spiders, fuel rods of various sizes, and loose debris consisting of resolidified materials and fractured fuel rods and pellets.

C. Mid-Core Region This region included the layer of resolidified material referenced in Section 4.2.2(B). This region was surrounded by standing peripheral assemblies and supported by partial fuel assemblies. This region consisted of a large, resolidified mass that was relatively thin at the periphery and very thick near the center. The thickness varied from about 13 in. at the periphery to about 60 in. at the center. Composition of this damaged material included a monolithic porous resolidified mass of rubble that was fused together by once-molten material, buried end fittings, and other structural material.

D. Lower Core Region The geometrically intact partial assemblies located at the bottom of the core varied in length from 9 in. near the center to full length at the periphery outboard of the resolidified material. Only two assemblies were full length and with better than 90% of the fuel rod cross-section. Most of the lower portions of these fuel assemblies were ductile, while upper portions were brittle, indicating a higher degree of oxidation. On the eastern side of the core, near the major relocation path, a number of the partial assemblies were brittle near the bottom, with ductile fuel rods on the upper portions of the stub assemblies. Several assemblies on the east side were resolidified masses with no identifiable fuel rods.

E. Lower CSA Region This region consisted of the area between the bottom of the lower end fittings and the 2-inch-thick flow distributor. The majority of the region just beneath the core region contained only fine, loose debris with no structural damage. The region outboard of the fuel region under the flow bypass region contained a large number of resolidified columns which were created by the downward flow of melted core material from the bypass region. The eastern part of the lower internals contained a large amount of resolidified material under several fuel assemblies and under the bypass region.

This area was the major relocation path for the melted core material which flowed to the lower head region. There was some structural damage in this area with some melting of in-core guide tubes and support posts.

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-5 F. Lower Head Region The lower head region encompassed the space between the flow distributor and the spherical lower head. A large quantity of material was relocated to this region during the accident. The material appeared to be distributed non-uniformly. Particle sizes in the loose portion ranged from fine dust to nodules approximately eight inches in diameter. There was much less loose material in the north quadrant of this region than in the other areas.

G. Core Former Region This region includes the area between the baffle plates and the thermal shield in the upper CSA. Some damage (melting) to the baffle plates occurred on the eastern side. Inspections of this region revealed a large mass of material between the baffle plates and the core barrel. Material was observed on the various core former plates around the core circumference with most of the material accumulated in the east and the north. No physical damage of the thermal shield and core barrel was observed. Resolidified material was seen at a number of locations just below and penetrating through flow holes below the core former plates.

H. Ex-Vessel This region consisted of any area outside the boundaries of the RV where core material had been transported. This included the RCS and associated components, the Reactor Building sump, and Auxiliary and Fuel Handling Buildings sump. During the accident, small quantities of core debris were relocated throughout the RCS and support systems.

TABLE 4.2-1, CORE MATERIAL INVENTORY Original Loading Weight (lb.)

Total UO2 (177 assemblies) 207,300 Total Zircaloy (4) 51,100 Absorber and other structural material 24,300 Total Original Core Material 282,700 Post-Accident Weight (lb.)

Original Mass 282,700 Material melted from plenum upper core tie plate and grid pads 500 Material melted from baffle plates, core former, and in-core assemblies 600 O2 due to Zr oxidation 7,700 Defueling-generated material 4,600 Total Estimated Post-Accident Core Debris Material Inventory 296,100

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-6 4.3 SPECIAL NUCLEAR MATERIAL ACCOUNTABILITY AND CRITICALITY SAFETY ANALYSIS 4.

3.1 INTRODUCTION

The purpose of this section is to describe the current TMI-2 Special Nuclear Material (SNM) Accountability Program and summarize the criticality safety analyses. This section identifies the methods and sequence of events for residual SNM accountability; the QA program that was applied during the cleanup program to the SNM measurements; the areas, systems, and components that were assessed for residual quantities of SNM; and the areas, systems, and components that did not require SNM assessment.

The quantity of fuel (i.e., UO2) remaining at TMI-2 is a small fraction of the initial fuel load. As a result of TMI-2 defueling and decontamination activities, approximately 99% of the fuel was removed and transferred to DOE and/or licensed burial facilities.

The final results of the SNM Accountability Program are based on a comprehensive post-defueling survey of the TMI-2 facility. The post-defueling survey consisted of a review of the TMI-2 plant to identify areas that could contain SNM and areas unlikely to contain SNM. The quantity of SNM was determined in each area that was identified to have SNM present. This section describes the process by which the post-defueling survey was conducted and summarizes the results of the survey.

Finally, this section presents a summary of the criticality safety analyses presented in TMI2-RA-COR-2022-0008, Supplemental Information to License Amendment Request - Three Mile Island Unit 2 Decommissioning Technical Specifications, dated April 7, 2022 (Ref. 1.3.43), which demonstrate that a criticality event could not occur at TMI-2. Further details of SNM Control during PDMS are provided in TMI2-RA-COR-2022-0019, License Amendment Request - Three Mile Island Unit 2 Decommissioning Technical Specifications, Response to Request for Additional Information, dated September 29, 2022 (Ref. 1.3.44).

4.

3.2 BACKGROUND

The March 1979 accident resulted in significant damage to the reactor core with a subsequent release of fuel and fission products into the RCS and other connected systems. The core was reduced to fractured fuel pellets, resolidified fuel masses, structural metal components, loose rubble, and partial fuel assemblies. The generic term used to refer to the post-accident core material is core debris. The core debris removed from the TMI-2 facility was loaded into special canisters for shipment to the DOE INEL facility in Idaho. Each shipment was accompanied by a completed copy of NRC Form 741, Nuclear Materials Transaction Report (Ref. 1.3.66), which recorded the net weight of the contents of each canister.

Fuel accountability by the normal method (i.e., accounting for individual fuel assemblies) was not possible. Since the canisters were filled with a mixture of SNM, other materials, and water, there was no practical or feasible method to determine the exact SNM content in each canister. A statement to that effect was included on each copy of NRC Form 741, Nuclear Materials Transaction Report (Ref. 1.3.66).

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-7 In October 1985, GPU Nuclear (GPUN), the U.S. Department of Energy (DOE),

and the U.S. Nuclear Regulatory Commission (NRC) entered into an agreement that final SNM accountability for TMI-2 would be performed after defueling was completed. This agreement is detailed in DOE Letter WWB-100-85, Bixby, W. W.

(DOE) to Burton, H. M. (EG&G), Accountability for the TMI-2 Core (Ref. 1.3.55), and Letter, Snyder, B. J. (NRC) to Standerfer, F. R. (GPUNC),

Approval of Exemption from 10 CFR 30.51, 40.61, 70.51(d), and 70.53 (Ref. 1.3.61). The accountability would be based upon a thorough post-defueling survey of TMI-2, which would quantify the amount of residual SNM in plant systems and components. Implied in this agreement was an understanding that the post-defueling survey will involve all areas and SSCs where SNM could reasonably be suspected to have been deposited as a result of the 1979 accident and subsequent cleanup activities.

4.3.3 HISTORICAL SNM ACCOUNTABILITY PROCESS A. Classification of Plant Areas The entire TMI-2 Plant was reviewed to determine where SNM could have been deposited as a result of the 1979 accident and subsequent cleanup activities.

Each area was classified into one of three categories:

1. Category 1: Locations where SNM was highly probable.

2. Category 2: Locations where it was possible that SNM could be deposited.

3. Category 3: Locations where it was unlikely that SNM was deposited.

Category 1 locations required that measurements or, in selected cases, analysis, be performed for SNM. Category 2 areas were considered to have a lower probability of fuel deposits but were assessed in the same manner as the Category 1 areas. Category 3 areas were determined not to require SNM assessment based on analyses of the TMI-2 accident and review of cleanup activities, found in NSAC 80-1, Analysis of Three Mile Island Unit 2 Accident, Electrical Power Research Institute (Ref. 1.3.15), and Rogovin, M., et al., Three Mile Island: A Report to the Commissioners and the Public (Ref. 1.3.20).

B. SNM Accountability Methods SNM accountability for TMI-2 was completed in accordance with GPU Nuclear 4000-PLN-4420.02, SNM Accountability Plan (Ref. 1.3.33). Several plant areas and components were characterized for SNM deposits prior to initiation of the formal SNM Accountability Program. In some cases, ALARA considerations, the quality of the previous measurements, and lack of actions potentially affecting SNM deposits warranted their use. These measurements were independently reviewed in accordance with GPU Nuclear 4000-ADM-4420.03, Review and Qualification of Selected Preliminary Calculations and Characterization Measurements for SNM Documentation (Ref. 1.3.32), to ensure sufficient data existed to meet SNM accountability QA standards. In all cases, the quantity of residual SNM was determined through measurements, sampling, inspection, or engineering analysis.

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-8

1. Measurements In most cases, measurements were performed in individual locations after planned cleanup activities were completed within the area. In some areas, as stated above, it was determined that the cleanup activities did not materially affect the original SNM measurements which were then used for SNM accountability. The post-defueling survey required the application of several measurement techniques. Technique selection for an individual measurement depended upon the geometry of the component/system or area to be assayed, physical access limitations, radiological conditions, personnel exposure considerations, and the probable quantity of SNM in the area. Where required or desirable, the measurements also involved the use of more than one measurement technique. Since the final SNM accountability activities were classified as Important to Safety, measurements conducted for SNM accountability were performed using QA-approved procedures.

Gamma scintillation spectrometry using sodium iodide detectors accounted for the majority of the early work. Later measurements involved the use of high-purity germanium detectors, which allowed greater resolution for the tracer isotopes of interest. Other measurement techniques included alpha scans using proportional detectors and gross gamma measurement techniques using collimated Geiger-Mueller detectors. The end fitting and dry-vessel measurements were completed using neutron interrogation techniques. Detailed descriptions of the measurement techniques and selection criteria can be found in TPO/TMI-051, Location and Characterization of Fuel Debris in TMI-2 (Ref. 1.3.63); TPO/TMI-124, Ex-Vessel Fuel Characterization (Ref. 1.3.64); and TPO/TMI-187, Instrument Selection for Residual Fuel Measurements (Ref. 1.3.65).

2. Sampling To obtain additional isotopic and volumetric information for use with the other analysis techniques, sampling of suspected fuel locations was performed. Solid and liquid samples were obtained from various areas and components to obtain isotopic, composition, and density data for use with measurements and visual inspections. Scrape samples were taken of metal surfaces (i.e., manways, piping, filter housings) to determine film depositions. These samples were analyzed using either onsite or offsite facilities, applying QA-approved procedures.

3. Visual Inspection In areas where measurement was not practical, video camera probes were used to estimate the volume of material remaining in the subject area. Using the volumetric data generated through sampling, a fuel quantity was assigned.

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-9

4. Engineering Analysis In the latter part of the project, several areas that had not been measured were estimated using a flow-path analysis. The flow-path analysis was performed by examination of possible SNM introduction pathways into an area through plant systems during the accident or subsequent cleanup activities.

C. Documentation The quantity of residual SNM in each location was documented in a GPUN engineering calculation. The overall results are provided in the following documents:

• GPU Nuclear Letter C312-91-2045, SNM Accountability, transmitting the Auxiliary and Fuel Handling Buildings PDSR (Ref. 1.3.56).

• GPU Nuclear Letter C312-91-2052, SNM Accountability, transmitting the Reactor Building Miscellaneous Components PDSR (Ref. 1.3.57).

• GPU Nuclear Letter C312-91-2055, SNM Accountability, transmitting the Reactor Coolant System PDSR (Ref. 1.3.58).

• GPU Nuclear Letter C312-91-2064, SNM Accountability, transmitting the A and B Once-Through Steam Generators PDSR (Ref. 1.3.59).

• GPU Nuclear Letter C312-93-2004, SNM Accountability, transmitting the Reactor Vessel PDSR (Ref. 1.3.60).

These results are summarized in Table 4.3-1, Final SNM Inventory by Location - Reactor Building, and Table 4.3-2, Final SNM Inventory by Location - Auxiliary and Fuel Handling Buildings, and detailed in Table 4.3-3, Auxiliary and Fuel Handling Building Cubicle Designation.

The locations of residual SNM in the Reactor, Auxiliary, and Fuel Handling Buildings are provided in the following figures:

• Figure 4.3-1, SNM Accountability Locations, Reactor Building, 282-6 Elevation.

• Figure 4.3-2, SNM Accountability Locations, Reactor Building, 305-0 Elevation.

• Figure 4.3-3, SNM Accountability Locations, Reactor Building, 347-6 Elevation.

• Figure 4.3-4, Auxiliary Building, 280-6 Elevation.

• Figure 4.3-5, Auxiliary Building, 305-0 Elevation.

The engineering calculations were based on geometric configuration, analysis of the measurement data, instrument calibrations, capabilities, and performance. Also included in the calculations were any specific assumptions made based on review of earlier measurements and the relevant history of that location during the accident and cleanup. All SNM engineering calculations were produced and approved in accordance with approved procedures.

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-10 The engineering calculations, in turn, provide the quantity of SNM for a specific area, system, or component that is outlined in the PDSR. Each PDSR contains:

• A detailed description of the area, system, or component.

• Its role in the accident and/or cleanup activities.

• The rationale supporting a conclusion as to whether contained residual SNM exists and, if so, a summary of the appropriate SNM engineering calculations.

• Applicable photographs and/or drawings of the area.

• An assessment of residual fuel.

The PDSRs were forwarded to the NRC, as evidenced in the following documents:

• GPU Nuclear Letter C312-91-2045, SNM Accountability, transmitting the Auxiliary and Fuel Handling Buildings PDSR (Ref. 1.3.56).

• GPU Nuclear Letter C312-91-2052, SNM Accountability, transmitting the Reactor Building Miscellaneous Components PDSR (Ref. 1.3.57).

• GPU Nuclear Letter C312-91-2055, SNM Accountability, transmitting the Reactor Coolant System PDSR (Ref. 1.3.58).

• GPU Nuclear Letter C312-91-2064, SNM Accountability, transmitting the A and B Once-Through Steam Generators PDSR (Ref. 1.3.59).

• GPU Nuclear Letter C312-93-2004, SNM Accountability, transmitting the Reactor Vessel PDSR (Ref. 1.3.60).

The completed PDSRs formed the basis for the final TMI-2 SNM inventory detailed on Table 4.3-1, Final SNM Inventory by Location - Reactor Building, and Table 4.3-2, Final SNM Inventory by Location - Auxiliary and Fuel Handling Buildings.

4.3.4 POST-DEFUELING SNM ACCOUNTABILITY Accountability following the previous defueling program was performed by summing the residual fuel quantities identified in the PDSRs and reporting the results as the remaining plant inventory of SNM. The amount of fuel shipped to the DOE INEL was determined by subtracting the sum of the final plant inventory and the amount of SNM shipped as radioactive waste from the pre-accident plant inventory of SNM, as corrected for decay in the most recent SNM Material Balance Report.

Pre-Accident Reported Inventory

- Final In-Plant Inventory

- SNM Shipped as Samples/Radwaste

= SNM Shipped to INEL (in canisters)

The resulting SNM inventory was reported on NRC Form 742, Materials Balance Report (Ref. 1.3.67).

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-11 4.3.5 CRITICALITY ANALYSIS A Safe Fuel Mass Limit (SFML) for the planned TMI-2 decommissioning was developed by applying realistic but bounding assumptions to the material remaining with the reactor vessel and outside the reactor vessel, per TMI2-EN-RPT-001, Determination of the Safe Fuel Mass Limit for Decommissioning TMI-2 (Ref. 1.3.36). The single 1200 kg uranium (or 1361 kg UO2) SFML can be utilized for all work in TMI-2 decommissioning and bound all credible operational upsets.

Through examination and conservative application of historic material sampling results, several key factors are credited:

A. 10 wt.% of the impurities identified in Mix 3 from the TMI-2 Defueling Completion Report (DCR) (Ref. 1.3.50), which is considered representative of the debris that remains in the RV.

B. Boron and iron are further reduced by 50% and 10% respectively, in the 10 wt.% impurity percentage identified in Mix 3.

C. Actual exposure histories applied to the average enrichment for all three batches of original TMI-2 fuel (burn-up credit).

The derived SFML bounds the entire expected fissile mass inventory throughout all physically separated areas within the reactor building. The bounding fissile mass used to produce the SFML is modeled as assembled in idealized conditions that cannot credibly exist during decommissioning operations.

4.3.6 CONTROL OF SNM DURING PDMS Control of SNM at TMI-2 during PDMS relied upon isolation boundaries and control of access to components which contain SNM. Isolation boundaries were maintained, as necessary, to prevent relocation of significant SNM quantities. The RCS, which contains the largest quantity of SNM, was drained to the extent practical and isolated within the Reactor Building. There was no physical inventory of SNM quantities at TMI-2 during PDMS because the remaining materials are of low enrichment, highly radioactive, and relatively inaccessible. The NRC granted TMI-2 an exception from the physical inventory requirements of 10 CFR 70.51(d) (Ref. 1.3.4). However, all shipments of accountable quantities of SNM from TMI-2 during PDMS were reported as required on NRC Form 741, Nuclear Materials Transaction Report (Ref. 1.3.66).

10 CFR numbering was revised. The exemption applies to current regulation 10 CFR 74.19(c) (Ref. 1.3.8).

4.3.7 CONTROL OF SNM DURING DECOMMISSIONING As discussed above, specific quantities of SNM exist in multiple plant components and cubicles. Decommissioning planning activities will specifically evaluate each discrete quantity of SNM to determine the method of retrieval and waste disposition path of all portions of the discrete component or cubicle under evaluation. Most SNM as a constituent of Fuel Bearing Material (i.e., a term used to describe the fuel and fission product contaminated plant components in the TMI-2-specific 10 CFR 72 license amendment to the NAC International MAGNASTOR dry cask storage system) will be loaded into a dry cask storage system designed to meet applicable 10 CFR 72 requirements (Ref. 1.3.6).

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-12 The dry cask storage system consists of multiple components:

A. The Transportation Storage Container (TSC) is a right circular cylinder which acts as the containment boundary for Fuel Bearing Material. Following loading, the TSC is seal-welded and backfilled with an inert gas prior to closure.

B. The Waste Basket Liner (WBL) is an open-topped right circular cylinder into which larger segments of Fuel Bearing Material are loaded and then placed into a TSC.

C. Internal support components, including the Debris Material Container and Segmented Tube Assembly, in which smaller segments and fines of Fuel Bearing Material are placed and then loaded into a WBL. Loaded and closed TSCs are stored within a concrete overpack and placed on an ISFSI for long-term dry storage.

Large quantities of SNM as a constituent of Fuel Bearing Material which are not readily removable from plant components will be segmented into smaller pieces and placed within WBLs. Segmentation and loading of WBLs is a wet process which will occur in the fuel transfer canal. Small fines and granular Fuel Bearing Material will be collected and placed into WBLs in specifically designed Debris Material Containers.

Small quantities of SNM are expected to be present in low-level radioactive waste shipments and sample materials. Masses greater than 1 gram will be reported via NRC Form 741, Nuclear Materials Transaction Report (Ref. 1.3.66), accompanying the waste shipment and be validated against waste acceptance criteria of the recipient disposal facility.

Based on existing exemptions (see Letter, Snyder, B. J. [NRC] to Standerfer, F. R.

[GPUNC], Approval of Exemption from 10 CFR 30.51,40.61,70.51[d] and 70.53

[Ref. 1.3.61]), to physical inventory requirements and annual reporting, SNM accountability and control will only consider transfer of SNM (as reported on NRC Form 742, Materials Balance Report [Ref. 1.3.67]) via characterized Low-Level Radioactive Waste (LLRW) packages and sample materials. SNM remaining in dry cask storage will be determined by subtracting the SNM quantities present from the quantities shipped as LLRW and/or samples. Final material balance reports will be provided when all SNM has been confirmed by Final Status Survey to have been removed.

Final In-Plant Inventory as determined by PDSRs

- SNM Shipped as Samples/Radwaste

= SNM in Dry Cask Storage (in canisters)

Further details of SNM Control during PDMS are provided in TMI2-RA-COR-2022-0019, from Lackey, M.L., License Amendment Request -

Three Mile Island Unit 2 Decommissioning Technical Specifications, Response to Request for Additional Information, dated September 29, 2022 (Ref. 1.3.44).

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-13 TABLE 4.3-1, FINAL SNM INVENTORY BY LOCATION - REACTOR BUILDING Location Description Fuel (kg)

RB Reactor Vessel 925 RB Reactor Head 1.3 RB Plenum 2.1 RB Pressurizer (including surge line) 0.5 RB OTSG A Tube Sheet 1.4 RB OTSG A Tube Bundle 1.7 RB OTSG A Lower Head and J-Legs 4.0 RB A Hot Legs 0.9 RB A Cold Legs 7.2 RB A Core Flood Line 0.6 RB OSTG B Tube Sheet 36.0 RB OSTG B Tube Bundle 9.1 RB OSTG B Lower Head and J-Legs 10.1 RB B Hot Legs 1.8 RB B Cold Legs 2.4 RB B Core Flood Line 0.4 RB Reactor Coolant Pumps 6.2 RB Decay Heat Drop Line 1.5 RB Basement Letdown Coolers 3.7*

RB Basement Reactor Coolant Drain Tank 0.1 RB Basement RB Basement and Sump 1.3 RB Fuel Transfer Canal 18.9 RB Core Flood System 4.9 RB In-Core Instrument Guide Tubes in the A D-ring 21.0 RB Upper End Fitting Storage Area 5.9 RB Tool Decontamination Facility 0.1 RB DWCS 3.7 RB Defueling Tools 0.6 RB TRVFS 4.4 RB RB Drains 4.4 RB RCS Surface Films 4.6 Total SNM Inventory

< 1086

• Minimum Detectable Limit (MDL)

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-14 TABLE 4.3-2, FINAL SNM INVENTORY BY LOCATION - AUXILIARY AND FUEL HANDLING BUILDINGS Location Description Fuel (kg)

AX004 Seal Injection Valve Room 0.03 AX006 Make Up Pump Room - 1B 0.07*

AX007 Make Up Pump Room - 1A 0.23*

AX012 Auxiliary Building Sump Tank Room 0.10 AX015a/b Cleanup Filters 0.10*

AX019 Waste Disposal Liquid Valve Room 0.01*

AX020 Reactor Coolant Bleed Tank Room - 1B & 1C 3.50 AX021 Reactor Coolant Bleed Tank Room - 1A 0.31 AX024 Auxiliary Building Sump Filters 0.02 AX112 Seal Return Coolers and Filter Room 0.30*

AX114 MU & P Demineralizer Room - 1A 1.06 AX115 MU & P Demineralizer Room - 1B 0.13 AX116 Make Up Tank Room 0.31 AX117 MU & P Filter Room 0.06 AX128 Instrument and Valve Room 0.01 AX102 RB Sump Pump Filter Room 0.10 AX131 Miscellaneous Waste Tank Room AX134 Miscellaneous Waste Tank Pump Room AX218 Concentrated Waste Storage Tank Room 0.01 AX501 RB Spray Pump - 1A 0.01 AX502 RB Spray Pump - 1B 0.01 AX503 DHR Cooler & Pump - 1A 0.01 AX504 DHR Cooler & Pump - 1B 0.01 FH001 MU Suction Valve Room 0.46 FH003a Make Up Discharge Valve Room 0.01 FH003b Make Up Discharge Valve Room 0.10 FH002 Access Corridor 0.16 FH004 Westinghouse Valve Room FH014 Annulus FH101 MU & P Valve Room 0.32 FH109 Spent Fuel Pool A 3.80 FH112 Annulus 0.01 Multiple Embedded Valves and Piping (MU System) 0.17 Multiple Embedded Valves and Piping (WDL System) 0.04 Total SNM Inventory 11.46

• Minimum Detectable Limit (MDL)

Note: All other locations contain less than 0.005 kg UO2 per area.

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-15 TABLE 4.3-3, AUXILIARY AND FUEL HANDLING BUILDING CUBICLE DESIGNATION Cubicle No.

Level Area Designation FH001 280-6 Make-Up Suction Valve Room FH002 280-6 Access Corridor FH003a/b 280-6 Make-Up Discharge Valve Rooms FH004 280-6 Westinghouse Valve Room FH005 280-6 Mini Decay Heat Vault FH006 280-6 Decay Heat Service Coolers FH007 280-6 Neutralizer & Reclaimed Boric Acid Access Area FH008 280-6 Neutralizer Tank Pumps Room FH009 280-6 Neutralizer Tank Room FH010 280-6 Reclaimed Boric Acid Tank Room FH011 280-6 Reclaimed Boric Acid Pump Room FH012 280-6 Neutralizer Tank Filter Room FH013 280-6 Oil Drum Storage Room FH014 280-6 Annulus from 280-6 to 305 FH101 305 Make-Up & Purification Valve Room FH102 305 East Corridor FH103 305 Sample Room FH104 305 West Corridor FH105 305 Model Room FH106 305 Monitor Tanks & Sample Sink Area FH107 305 Trash Compactor Area FH108 305 Truck Bay FH109 305 Spent Fuel Pool A (Under Cover)

FH110 305 SDS Spent Fuel Pool B (Under Cover)

FH111 305 Fuel Cask Storage (Under Cover)

FH112 305 Annulus from 305 to 328 FH201 328 East Corridor FH202 328 West Corridor FH203 328 Surge Tank FH204 328 New Fuel Storage - SPC Pit FH205 328 Annulus from 328 to 347-6 FH301 347-6 A Spent Fuel Pool (Above Cover)

FH302 347-6 B Spent Fuel Pool (Above Cover)

FH303 347-6 Upper New Fuel Storage Area FH304 347-6 Annulus Above 347-6 FH305 347-6 Spent Fuel Pool Access Area AX001 280-6 Reactor Building Emergency Cooling Booster Pumps Area AX002 250-6 Access Corridor

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-16 TABLE 4.3-3, AUXILIARY AND FUEL HANDLING BUILDING CUBICLE DESIGNATION (cont.)

Cubicle No.

Level Area Designation AX002a 280-6 Nitrogen Piping AX003 280-6 Access Area AX004 280-6 Seal Injection Valve Room AX005 280-6 Make-Up & Purification Pump 1C Room AX006 280-6 Make-Up & Purification Pump 1B Room AX007 280-6 Make-Up & Purification Pump 1A Room AX008 280-6 Spent Resin Storage Tank 1B Room & Access Area AX009 280-6 Spent Resin Storage Tank 1A Room AX010 280-6 Spent Resin Transfer Pump Room AX011 280-6 Auxiliary Building Sump Tank Pump & Valve Room AX012 280-6 Auxiliary Building Sump, Pump, & Tank Room AX013 280-6 Evaporation Tanks, Pumps, and Demineralization Area AX014 280-6 Reactor Coolant Evaporator Room AX015a/b 280-6 Clean Up Filters Area AX016 280-6 Clean Up Demineralizer 2A Room AX017 280-6 Clean Up Demineralizer 2B Room AX018 280-6 Waste Transfer Pumps Room AX019 280-6 Waste Disposal Liquid Valve Room AX020 280-6 Reactor Coolant Bleed Hold Up Tank 1B & 1C Room AX021 280-6 Reactor Coolant Bleed Hold Up Tank 1A Room AX022 280-6 North Stairwell Between 280-6 & 309 AX023 280-6 Elevator Shaft from Sump to 305 AX024 280-6 Auxiliary Building Sump Filters Room AX025 280-6 Area Between Service, Control, and Reactor Buildings AX026 280-6 Seal Injection Filters Area AX027 280-6 South Stairwell AX101 305 Radwaste Disposal CNTL PNL Access Area AX102 305 Reactor Building Sump Pump Filters Area AX103 305 MCC 2-11EB Room AX104 305 MCC 2-21EB Room AX105 305 Substation 2-11E Room AX106 305 Substation 2-21E Room AX107 305 MCC 2-11EA Room AX108 305 MCC 2-21EA Room AX109 305 Nuclear Service Coolers & Pumps Area AX110 305 Intermediate Coolers Area AX111 305 Intermediate Cooling Pumps & Filters AX112 305 Seal Return Coolers and Filter Area

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-17 TABLE 4.3-3, AUXILIARY AND FUEL HANDLING BUILDING CUBICLE DESIGNATION (cont.)

Cubicle No.

Level Area Designation AX113 305 Waste Gas Analyzer Room AX114 305 Make Up & Purification Demineralizer 1A Room AX115 305 Make Up & Purification Demineralizer 1B Room AX116 305 Make Up Tank Room AX117 305 Make Up & Purification Filter Room AX118 305 Spent Fuel Coolers Area AX119 305 Spent Fuel Demineralizer Room AX120 305 Spent Fuel Filters Area AX121 305 Elevator Shaft from 305 to 328 AX122 305 North Stairwell from 305 to 328 AX123 305 Access Area AX124 305 Concentrated Liquid Waste Pump Room AX125 305 Waste Gas Decay Tank 1B Room AX126 305 Waste Gas Filter Room AX127 305 Waste Gas Decay Tank 1A Room & Access Corridor AX128 305 Vent Header, Valve, and Instrument Room AX129 305 Deborating Demineralizer 1B Room AX130 305 Deborating Demineralizer 1A Room AX131 305 Miscellaneous Waste Hold Up Tank Room AX132 305 Corridor Between Unit 1 and Unit 2 AX133 305 South Stairwell Between 305 and 328 AX134 305 MWHT Pump Room AX135 305 Radwaste Disposal Control Panels AX136 305 Hot Tool Room (Part AX-123)

AX201 328 North Stairwell from 328 to 347 AX202 328 Elevator Shaft from 328 to 347 AX203 328 4160V Switchgear 2-1E Room AX204 328 4160V Switchgear 2-2E Room AX205 328 Reactor Building Purge and Breather Area AX206 328 Reactor Building Purge Air Exhaust Unit 1B AX207 328 Reactor Building Purge Air Exhaust Unit 1A AX208 328 Auxiliary Building Exhaust Unit B AX209 328 Auxiliary Building Exhaust Unit A AX210 328 Fuel Handling Building Exhaust Unit B AX211 328 Fuel Handling Building Exhaust Unit B AX212 328 Decay Heat Surge Tank & Unit Substation Area AX213 328 Unit Substations & Access Area AX214 328 Electro-Con Facility

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-18 TABLE 4.3-3, AUXILIARY AND FUEL HANDLING BUILDING CUBICLE DESIGNATION (cont.)

Cubicle No.

Level Area Designation AX215 328 Fuel Handling Building Supply Unit AX216 328 Auxiliary Building Supply Unit AX217 328 Access Area AX218 328 Concentrated Waste Tank AX219 328 Instrument Racks & Atmospheric Monitor Cabinet Area AX220 328 Chemical Addition Area AX221 328 Chemical Addition Area AX222 328 South Stairwell Above 379 AX223 328 Air Handling Units General Area AX301 363-9 Elevator Shaft Above 347-6 & Machinery Room 363-9 AX302 328 North Stairwell Above 347-6 AX303 347-6 Elevator and Stairwell Access AX401 353-6 Roof AX402 356-6 Cooling Water Surge Tank Room AX403 353-6 Damper Room (Penthouse)

AX501 258-6 Reactor Building Spray Pump 1A Room AX502 258-6 Reactor Building Spray Pump 1B Room AX503 258-6 Decay Heat Removal Cooler & Pump 1A Room AX504 258-6 Decay Heat Removal Cooler & Pump 1B Room

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-19 FIGURE 4.3-1, SNM ACCOUNTABILITY LOCATIONS, REACTOR BUILDING, 282-6 ELEVATION (BEST AVAILABLE COPY)

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-20 FIGURE 4.3-2, SNM ACCOUNTABILITY LOCATIONS, REACTOR BUILDING, 305-0 ELEVATION (BEST AVAILABLE COPY)

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-21 FIGURE 4.3-3, SNM ACCOUNTABILITY LOCATIONS, REACTOR BUILDING, 347-6 ELEVATION (BEST AVAILABLE COPY)

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-22 FIGURE 4.3-4, FUEL HANDLING AND AUXILIARY BUILDING, 280-6 ELEVATION (BEST AVAILABLE COPY)

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-23 FIGURE 4.3-5, FUEL HANDLING AND AUXILIARY BUILDING, 305-0 ELEVATION (BEST AVAILABLE COPY)

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-24 FIGURE 4.3-6, FUEL HANDLING AND AUXILIARY BUILDING, 328-0 ELEVATION (BEST AVAILABLE COPY)

TMI-2 Defueled Safety Analysis Report Chapter 4:

Revision 0 Fuel 4-25 FIGURE 4.3-7, FUEL HANDLING AND AUXILIARY BUILDING, 347-6 ELEVATION (BEST AVAILABLE COPY)

TMI-2 Defueled Safety Analysis Report Revision 0 RADIOLOGICAL CONDITIONS

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-2

5.1 INTRODUCTION

One consequence of the March 1979 accident was widespread radioactive contamination of the Reactor, Fuel Handling, and Auxiliary Buildings. RCS water was released to the Reactor Building and overflowed to the Auxiliary and Fuel Handling Buildings. These areas required extraordinary decontamination efforts to achieve the cleanup program objectives. Contamination of areas outside the Reactor Building and Auxiliary and Fuel Handling Buildings were minor and limited. These areas outside the Reactor Building and Auxiliary and Fuel Handling Buildings were decontaminated and either released for unrestricted use or configured such that the contamination is suitably contained.

The objectives of the decontamination program were to remove and/or stabilize the contamination to reduce occupational exposure to workers and to prevent release of contamination to the environment during recovery and cleanup activities. In addition, a final decontamination objective was to ensure that any remaining contamination was stable and suitably isolated.

Section 5.3 provides a description of the contamination levels remaining in the various areas of the plant and systems at the start of PDMS for use in planning decontamination and decommissioning activities and to describe ongoing radiological controls. This information is provided for information only and will not be updated.

Sections 5.2 and 5.4 describe aspects of the Radiation Protection Program at TMI-2 that will control activities during decommissioning. Sections 5.5 and 5.6 describe radioactive waste management and monitoring systems. Section 5.7 describes the Sealed Source Program as transferred from the TMI-2 Technical Specifications to this DSAR.

5.2 ENSURING OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLY ACHIEVABLE 5.2.1 ALARA PROGRAM Consistent with station modification, maintenance, operational requirements, and economic and social considerations, the policy of TMI2S is to:

A. Maintain the occupational dose equivalent to the individual As Low As Is Reasonably Achievable (ALARA);

B. Maintain the sum of occupational dose equivalents received by all exposed workers ALARA; and C. Limit the number of workers authorized to receive exposure to radiation.

NRC Regulatory Guide 8.8, Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable, Revision 3, Sections C.1, C.3, and C.4 (Ref. 1.3.14), is used as a basis for developing the ALARA and Radiation Protection (RP) programs.

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-3 Station managements commitment to this policy is reflected in RP procedures and programs. RP staff provides the radiological conditions and protection requirements necessary to complete work safely. The responsibility of each individual to adhere to these requirements and the procedures governing their work is key to the success of the program.

5.3 PDMS RADIOLOGICAL SURVEY This section describes the radiological conditions at the TMI-2 facility upon entering PDMS. These conditions are expressed in terms of general area dose rate, loose surface contamination, and general isotopic distribution. These conditions will guide initial work planning for decommissioning until surveys based on current conditions are completed.

5.3.1 RADIOLOGICAL ASSESSMENT Upon completion of cleanup activities (including decontamination) in a given area or cubicle, the area was isolated to prohibit uncontrolled access. Deactivated systems traversing the area of cubicle were drained, vented, and isolated. The subject area was, at that point, configured for long-term monitored storage and available for a final PDMS radiological assessment. This assessment was performed utilizing radiological surveys (in this case, radiation, contamination, and air activity surveys performed by Radiation Protection Technicians [RPTs]) as a basis for determining whether the established decontamination program endpoints were achieved as well as to document the radiological conditions that existed upon entering PDMS.

A. Pre-PDMS Radiological Survey Methodology Radiation contamination and air activity surveys were routinely performed during the course of the cleanup program in support of work activities. These surveys were performed in accordance with regulatory and industry standards and practices to verify and document radiation and contamination levels for use in controlling personnel exposure. These surveys were then evaluated as to whether or not they supported the conclusion that decontamination endpoints had been achieved. In those instances where existing surveys were judged unsuitable for substantiating decontamination endpoints, additional surveys were conducted.

B. PDMS Period Radiological Survey Methodology During the PDMS period, radiological conditions within the facility were monitored through sampling and periodic surveillance. Surveillance activities for the Reactor Building consisted of radiation surveys in conjunction with planned Reactor Building inspections. The purpose for conducting these surveys was to provide assurance that conditions were either stable or to provide early indication of any changing conditions that may require corrective action.

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-4 5.3.2 RADIOLOGICAL CONDITIONS AT BEGINNING OF PDMS Table 5.3-1, PDMS Radiological Conditions - Reactor Building, lists the specific radiological goals for the TMI-2 Reactor Building and the corresponding radiological conditions as of the radiological surveys conducted in September 1992.

The Reactor Building radiological conditions listed in Table 5.3-1 reflect rounded-off, average PDMS survey data for the entire cubicle/area in question. This data was compiled in the manner described in Section 5.3.1(A).

Table 5.3-2, PDMS Radiological Conditions - Auxiliary and Fuel Handling Buildings, lists the equivalent information for the Auxiliary and Fuel Handling Buildings as of November 1993.

Table 5.3-3, PDMS Radiological Conditions - Other Buildings, provides a summary of the radiological conditions for the balance-of-plant areas not covered by Table 5.3-1 and Table 5.3-2.

A. Surface Contamination at Beginning of PDMS To establish a baseline at the beginning of PDMS, the radioactivity present as surface contamination in various areas of the facility has been evaluated. This information serves as an initial reference for the evaluation of future activities in the respective areas.

In order to appraise the radioactivity present as loose surface contamination upon entry into PDMS, an analytical model was constructed utilizing available loose surface contamination data, generalized waste stream isotopic distributions, and estimates of surface area. This information is formatted in a manner similar to the general area dose rate and loose surface contamination data presented in Table 5.3-1, Table 5.3-2, and Table 5.3-3. Only surface contamination was considered; fixed contamination or contamination internal to piping systems or equipment was omitted. The generalized waste streams or distribution of principal isotopes are referenced on each of the tables.

Table 5.3-4, Surface Contamination - Reactor Building, lists the data obtained from the analytical model described above for the TMI-2 Reactor Building. Table 5.3-5, Surface Contamination - Auxiliary and Fuel Handling Buildings, lists the equivalent data for the Auxiliary and Fuel Handling Buildings, and Table 5.3-6, Surface Contamination - Other Buildings, provides a similar summary for the balance-of-plant areas. All calculations of the quantities of curies listed are based on the specific decontamination goals given on Table 5.3-2 and Table 5.3-3.

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-5 TABLE 5.3-1, PDMS RADIOLOGICAL CONDITIONS - REACTOR BUILDING Area Description Specific Decontamination Goals PDMS Radiological Conditions*

General Area Dose Rate (mR/hr.)

Surface Contamination (dpm/100 cm2)

General Area Dose Rate (mR/hr.)

Surface Contamination (dpm/100 cm2)

Elevation 305 to 347

<100

<50,000 150 2,000,000 Elevation 347 and above

<30

<50,000 50 710,000 Refueling Canal

<100

<50,000 120 670,000 D-Ring Interior, Elevation 349 and above As Is (A D-Ring)

As Is (B D-Ring)

As Is (A D-Ring)

As Is (B D-Ring) 300 (A D-Ring) 200 (B D-Ring) 280,000 (A D-Ring) 220,000 (B D-Ring)

Top of D-Rings

<100

<50,000 40 (A D-Ring) 50 (B D-Ring) 270,000 190,000 Basement, Elevation 282 As Is As Is 56,000**

See note below.

• The radiological conditions in this table reflect rounded-off, average PDMS survey data.

• • This is the decay-corrected dose rate taken by the ROVER Robot, per TMI-2 Technical Planning Bulletin 85-03, Reactor Building Basement Measurements, Revision 0, February 11, 1985 (Ref. 1.3.53).

Note: This area is inaccessible; no meaningful data exists.

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-6 TABLE 5.3-2, PDMS RADIOLOGICAL CONDITIONS - AUXILIARY AND FUEL HANDLING BUILDINGS Cubicle No.

Area Description Specific Decontamination Goals PDMS Radiological Conditions*

General Area Dose Rate (mR/hr.)

Surface Contamination

<7/Overheads (dpm/100 cm2)

General Area Dose Rate (mR/hr.)

Surface Contamination

<7/Overheads (dpm/100 cm2)

AX001 Reactor Building Emergency Cooling Booster Pumps Area (WGP-1 Shielded Enclosure)

<2.5

<2.5

<1,000/<10,000

<1,000/10,000 1.5 2.4 590/2,500 81,000/6,500 AX002 Access Corridor

<2.5

<1,000/<10,000 1.5 110/1,300 AX002a N2 Piping System

<2.5

<1,000/N/A 1.8 110/N/A AX003 Access Area

<2.5

<1,000/<10,000 0.8 2,900/2,300 AX004 Seal Injection Valve Room

<1000

<50,000/<50,000 120 68,000/750,000 AX005 MUP Pump 1C Room

<500

<50,000/<50,000 8

40,000/30,000 AX006 MUP Pump 1B Room

<500

<50,000/<50,000 60 88,000/29,000 AX007 MUP Pump 1A Room

<500

<50,000/<50,000 40 9,200/21,000 AX008 Spent Resin Storage Tank 1B Room As Is As Is 710 960,000/3,400,000 AX009 Spent Resin Storage Tank 1A Room As Is As Is 1,700 3,000,000/6,000,000 AX010 Spent Resin Transfer Pump Room As Is As Is 180 1,400,000/5,100,000 AX011 Auxiliary Building Sump Tank Pump/Valve Room

<50

<5,000/<50,000 8

3,200/4,900 AX012 Auxiliary Building Sump & Tank Room

<50

<5,000/<50,000 560 390,000/68,000 AX013 Evaporator Condensate Tanks, Pumps Demineralizer Room

<500

<1,000/<10,000 5

130/140 AX014 RC Evaporator Room

<500

<50,000/<50,000 18 20,000/8,100 AX015a Cleanup Filters Room

<500

<50,000/<50,000 110 10,000/8,700 AX015b Cleanup After Filters Room

<500

<50,000/<50,000 38 21,000/13,000 AX016 Cleanup Demineralizer 2A Room

<500

<50,000/<50,000 38 21,000/13,000 AX017 Cleanup Demineralizer 2B Room

<500

<50,000/<50,000 110 10,000/8,700 AX018 Waste Transfer Pumps Room

<500

<50,000/<50,000 10 17,000/14,000 AX019 Waste Disposal Liquid Valve Room

<500

<50,000/<50,000 19 7,800/5,700 AX020 RC Bleed Holdup Tanks 1B & 1C Room

<500

<50,000/<50,000 160 520,000/200,000 AX021 RC Bleed Holdup Tank 1A Room

<500

<50,000/<50,000 18 1,800/12,000 AX022 North Stairwell

<2.5

<1,000/N/A 1.3 440/N/A AX023 Elevator Pit

<10

<50,000/N/A 14 25,000/N/A

• The radiological conditions in this table reflect rounded-off, average PDMS survey data.

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-7 TABLE 5.3-2, PDMS RADIOLOGICAL CONDITIONS - AUXILIARY AND FUEL HANDLINGBUILDINGS (cont.)

Cubicle No.

Area Description Specific Decontamination Goals PDMS Radiological Conditions General Area Dose Rate (mR/hr.)

Surface Contamination

<7/Overheads (dpm/100 cm2)

General Area Dose Rate (mR/hr.)

Surface Contamination

<7/Overheads (dpm/100 cm2)

AX024 Auxiliary Building Sump Filters Room

<500

<50,000/<50,000 15 156,400/11,000 AX025 Area between Service, Control, and Reactor Buildings

<500

<1,000/<10,000 3.5 1,000/310 AX026 Seal Injection Filters Room

<500

<50,000/<50,000 12 9,000/1,100 AX027 South Stairwell

<2.5

<1,000/N/A 0.2 480/N/A AX101 Radwaste Disposal Control Panel Area

<2.5

<1,000/<10,000 0.2 90/110 AX102 Reactor Building Sump Pumps Filters Room

<1000

<50,000/<50,000 47 9,300/4,200 AX103 MCC 2-11EB Room

<2.5

<1,000/<10,000 0.2 470/480 AX104 MCC 2-21EB Room

<2.5

<1,000/<10,000 0.2 470/470 AX105 Substation 2-11E Room

<2.5

<1,000/<10,000 0.2 480/530 AX106 Substation 2-21E Room

<2.5

<1,000/<10,000 0.2 480/530 AX107 MCC 2-11EA Room

<2.5

<1,000/<10,000 0.2 460/530 AX108 MCC 2-21EA Room

<2.5

<1,000/<10,000 0.2 480/500 AX109 Nuclear Services Coolers and Pumps Area

<2.5

<1,000/<10,000 0.2 100/130 AX110 Intermediate Coolers Area

<2.5

<1,000/<10,000 0.2 100/130 AX111 Intermediate Cooling Pumps & Filters Room

<50

<1,000/<10,000 0.7 440/410 AX112 Seal Return Coolers & Filter Room

<1000

<50,000/<50,000 99 350,000/38,000 AX113 Waste as Analyzer Room

<50

<50,000/<50,000 19 22,000/6,400 AX114 MUP Demineralizer 1A Room As Is As Is 73,000 9,800/34,000 AX115 MUP Demineralizer 1B Room As Is As Is 68,000 31,000/280,000 AX116 Makeup Tank Room

<500

<50,000/<50,000 60 310,000/23,000 AX117 MUP Filters Room

<1000 As Is 940 330,000,000/2,400 AX118 Spent Fuel Coolers and Pumps Area

<2.5

<1,000/<10,000 1.1 1,000/3,000 AX119 Spent Fuel Demineralizer Room

<2.5

<1,000/<10,000 0.4 480/330 AX120 Spent Fuel Filter Room

<2.5

<1,000/<10,000 0.6 360/1,000 AX121 Inside Elevator Cab

<2.5

<1,000/N/A 0.3 250/N/A AX122 North Stairwell

<2.5

<1,000/N/A 0.2 470/N/A

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-8 TABLE 5.3-2, PDMS RADIOLOGICAL CONDITIONS - AUXILIARY AND FUEL HANDLING BUILDINGS (cont.)

Cubicle No.

Area Description Specific Decontamination Goals PDMS Radiological Conditions General Area Dose Rate (mR/hr.)

Surface Contamination

<7/Overheads (dpm/100 cm2)

General Area Dose Rate (mR/hr.)

Surface Contamination

<7/Overheads (dpm/100 cm2)

AX123 Access Area (includes AX136)

<2.5

<1,000/<10,000 0.2 160/140 AX124 Concentrated Liquid Waste Pump Room

<500

<50,000/<50,000 3.3 3,300/1,800 AX125 Waste Gas Decay Tank 1B Room

<500

<50,000/<50,000 0.2 1,000/1,000 AX126 Waste Gas Filter Room

<500

<50,000/<50,000 0.2 100/160 AX127 Waste Gas Decay Tank 1A Room

<500

<50,000/<50,000 0.6 6,400/690 AX128 Valve & Instrument Room

<500

<50,000/<50,000 2.7 1,000/1,300 AX129 Deborating Demineralizer 1B Room

<500

<50,000/<50,000 0.3 1,000/600 AX130 Deborating Demineralizer 1A Room

<500

<50,000/<50,000 0.5 540/520 AX131 Miscellaneous Waste Holdup Tank Room

<50

<50,000/<50,000 120 1,900/6,500 AX132 Unit 1 and Unit 2 Corridor

<2.5

<1,000/<10,000 0.2 100/100 AX133 South Stairwell

<2.5

<1,000/N/A 0.2 500/N/A AX134 Miscellaneous Waste Tank Pumps Room

<50

<50,000/<50,000 13 13,000/45,000 AX135 Radwaste Disposal Control Panels

<2.5

<1,000/<10,000 0.2 130/120 AX201 North Stairwell

<2.5

<1,000/N/A 0.2 450/N/A AX202 Elevator Shaft

<2.5

<1,000/<10,000 0.2 480/480 AX203 4160V Switchgear 2-1E Room

<2.5

<1,000/<10,000 0.2 480/460 AX204 4160V Switchgear 2-2E Room

<2.5

<1,000/<10,000 0.2 480/480 AX205 Reactor Building Purge and H2 Control Area

<2.5

<1,000/<10,000 0.7 100/130 AX206 Reactor Building Purge Air Exhaust Unit B

<50 N/A 10 200,000/N/A AX207 Reactor Building Purge Air Exhaust Unit A

<50 N/A 13 200,000/N/A AX208 Auxiliary Building Exhaust Unit B

<50 N/A 0.4 3,900/N/A AX209 Auxiliary Building Exhaust Unit A

<50 N/A 0.7 10,000/N/A AX210 Fuel Handling Building Exhaust Unit B

<50 N/A 0.9 12,000/N/A AX211 Fuel Handling Building Exhaust Unit A

<50 N/A 0.3 7,200/N/A AX212 Decay Heat Surge Tank and Substation Area

<2.5

<1,000/<10,000 0.2 100/90 AX213 Unit Substations and Access Area

<2.5

<1,000/<10,000 0.2 130/120

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-9 TABLE 5.3-2, PDMS RADIOLOGICAL CONDITIONS - AUXILIARY AND FUEL HANDLING BUILDINGS (cont.)

Cubicle No.

Area Description Specific Decontamination Goals PDMS Radiological Conditions General Area Dose Rate (mR/hr.)

Surface Contamination

<7/Overheads (dpm/100 cm2)

General Area Dose Rate (mR/hr.)

Surface Contamination

<7/Overheads (dpm/100 cm2)

AX214 Decontamination Facility (Internal Area of Decontamination Facility Tanks)

<2.5

<2.5

<1,000/<10,000

<1,000/N/A 0.3 0.4 150/100 4,400/N/A AX215 Fuel Handling Building Supply

<2.5 N/A 0.2 450/N/A AX216 Auxiliary Building Supply Unit

<2.5 N/A 0.2 450/N/A AX217 Access Area

<2.5

<1,000/<10,000 0.2 120/370 AX218 Concentrated Waste Storage Tank Room

<500

<50,000/<50,000 15 1,900/1,000 AX219 Instrument Racks and Atmospheric Monitor Area

<2.5

<1,000/<10,000 0.3 390/5,900 AX220 Caustic Liquids Mixing Area

<500

<1,000/<10,000 1.4 440/360 AX221 Caustic Liquids Mixing Area Corridor

<500

<1,000/<10,000 0.8 450/880 AX222 South Stairwell

<2.5

<1,000/N/A 0.2 100/N/A AX223 Air Handling Units General Area

<2.5

<5,000/<10,000 0.8 490/450 AX301 Elevator Machine Room

<2.5

<1,000/<10,000 0.2 100/170 AX302 North Stairwell

<2.5

<1,000/<10,000 0.2 480/480 AX303 Elevator and Stairwell Access

<2.5

<1,000/<10,000 0.2 510/510 AX304 Auxiliary Building Exhaust Fans

<2.5

<1,000/<10,000 0.6 750/510 AX305 Fuel Handling Building Exhaust Fans

<2.5

<1,000/<10,000 0.2 650/390 AX401 Roof

<2.5

<1,000/N/A 0.2 90/N/A AX402 Cooling Water Surge Tanks Room

<500

<50,000/<50,000 0.2 110/230 AX403 Damper Room

<500

<50,000/<50,000 0.2 120/130 AX501 Reactor Building Spray Pump 1A Room

<25

<5,000/<50,000 17 370,000/2,100,000 AX502 Reactor Building Spray Pump 1B Room

<25 As Is 31 110,000/540,000 AX503 Decay Heat Removal Cooler & Pump 1A Room

<25

<50,000/<50,000 11 43,000/280,000 AX504 Decay Heat Removal Cooler & 1B Room

<2.5

<50,000/<50,000 6.1 15,000/89,000 FH001 Makeup Suction Valve Room

<500 As Is 19 70,000/89,000 FH002 Access Corridor

<2.5

<1,000/<10,000 1.5 1,000/1,500 FH003a Makeup Discharge Valve Room

<1000

<50,000/<100,000 69 140,000/77,000

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-10 TABLE 5.3-2, PDMS RADIOLOGICAL CONDITIONS - AUXILIARY AND FUEL HANDLING BUILDINGS (cont.)

Cubicle No.

Area Description Specific Decontamination Goals PDMS Radiological Conditions General Area Dose Rate (mR/hr.)

Surface Contamination

<7/Overheads (dpm/100 cm2)

General Area Dose Rate (mR/hr.)

Surface Contamination

<7/Overheads (dpm/100 cm2)

FH003b Makeup Discharge Valve Room

<1000

<50,000/<100,000 220 510,000/180,000 FH004 Westinghouse Valve Room

<500

<50,000/<50,000 59 38,000/1,100,000 FH005 Mini Decay Heat Vault

<500

<50,000/<50,000 2.4 2,700/1,500 FH006 Decay Heat Service Coolers Area

<500

<1,000/<10,000 6.4 1,100/1,400 FH007 Neutralizer and Reclaimed Boric Acid Access Area

<500

<1,000/<10,000 0.8 100/460 FH008 Neutralizer Tanks Pumps Room

<500

<50,000/<50,000 180 22,000/4,700 FH009 Neutralizer Tanks Room

<500

<50,000/<50,000 150 21,000/8,300 FH010 Reclaimed Boric Acid Tank Room

<500

<50,000/<50,000 4.3 2,400/6,300 FH011 Reclaimed Boric Acid Pump Room

<500

<50,000/<50,000 9.1 14,000/20,000 FH012 Neutralizer Tanks Filters Room

<500

<50,000/<50,000 31 2,800/1,200 FH013 Oil Drum Storage Area

<500

<1,000/<10,000 0.2 100/100 FH014 Annulus

<500

<50,000/<50,000 110 35,000/7,500 FH101 MUP Valve Room

<500

<50,000/<50,000 200 850,000/14,000,000 FH102 East Corridor (Chemistry Sample Shielded Storage Cage)

<2.5

<2.5

<1,000/<10,000

<1,000/N/A 1.1 4

200/31,000 380,000/N/A FH103 Sample Room

<50

<50,000/<50,000 1.2 4,000/1,600 FH104 West Corridor

<2.5

<1,000/<10,000 0.2 120/100 FH105 Model Room (Boronometer Shielded Enclosure)

<2.5

<2.5

<1,000/<10,000

<1,000/N/A 0.2 21 100/100 22,000/N/A FH106 Monitor Tanks and Sample Sink Area

<2.5

<1,000/<10,000 0.7 330/110 FH107 Trash Compactor Area

<2.5

<1,000/<10,000 0.2 100/130 FH108 Truck Bay

<2.5

<1,000/<10,000 0.2 70/70 FH109 Spent Fuel Pool A (Under Fuel Pool Cover)

As Is As Is/N/A 230 55,000,000/N/A FH110 SDS Spent Fuel Pool B (Under Fuel Pool Cover)

<2.5

<1,000/N/A 6.9 180,000/N/A FH111 Fuel Cask Storage (Under Fuel Pool Cover)

<1000

<1,000/N/A 0.3 150,000/N/A FH112 Annulus

<100

<50,000/<50,000 19 3,500/840

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-11 TABLE 5.3-2, PDMS RADIOLOGICAL CONDITIONS - AUXILIARY AND FUEL HANDLING BUILDINGS (cont.)

Cubicle No.

Area Description Specific Decontamination Goals PDMS Radiological Conditions General Area Dose Rate (mR/hr.)

Surface Contamination

<7/Overheads (dpm/100 cm2)

General Area Dose Rate (mR/hr.)

Surface Contamination

<7/Overheads (dpm/100 cm2)

FH201 East Corridor

<2.5

<1,000/<10,000 1.0 420/9,400 FH202 West Corridor

<2.5

<1,000/<10,000 0.2 480/490 FH203 Surge Tank Area

<500

<50,000/<50,000 28 1,000/Inaccessible FH204 Standby Pressure Control Area

<500

<1,000/<10,000 0.2 1,000/1,000 FH205 Annulus

<100

<50,000/<50,000 8.7 700/8,500 FH301 Upper Spent Fuel Pool A Area (Above Fuel Pool Cover)

<2.5

<1,000/<10,000 3.9 240/300 FH302 SDS Operating Area

<2.5

<1,000/<10,000 1.2 480/400 FH303 Upper Standby Pressure Control Area

<2.5

<1,000/<10,000 0.2 300/160 FH304 Annulus

<500

<50,000/<50,000 0.6 2,200/2,200 FH305 Spent Fuel Pool Access Area

<2.5

<1,000/<10,000 1.3 390/900

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-12 TABLE 5.3-3, PDMS RADIOLOGICAL CONDITIONS - OTHER BUILDINGS Cubicle No.

Area Description Specific Decontamination Goals PDMS Radiological Conditions*

General Area Dose Rate (mR/hr.)

Surface Contamination

<7/Overheads (dpm/100 cm2)

General Area Dose Rate (mR/hr.)

Surface Contamination

<7/Overheads (dpm/100 cm2)

SB000 Service Building, Elevation 281

<2.5

<1,000/<10,000 0.4 100**/330 SB002 M-20 Area

<2.5

<1,000/<10,000 0.3 110/110 SB002 M-20 Area Sump

<2.5

<1,000/N/A 0.4 1,000/N/A SB100 Service Building Elevation 305 (Reactor Building Containment Control)

(Cubicle Secondary Chemistry Lab)

<2.5

<2.5

<2.5

<1,000/<10,000

<1,000/<10,000

<1,000/<10,000 0.2 0.2 0.2 110/380 1,300/4,000 1,700/140 SB500 Tendon Access Gallery

<2.5

<1,000/<10,000 0.4 110/110 TB000 Turbine Building, Elevation 281

<2.5

<1,000/<10,000 0.2 100/100 PA108 CACE Building

<2.5

<1,000/<10,000 0.4 100/100 RA101 PWST House (PWST Sump)

<2.5

<2.5

<1,000/<10,000

<1,000/N/A 0.2 0.2 120/110 120/N/A RA104 BWST Area

<2.5

<1,000/N/A 0.3 90/N/A

• The radiological conditions in this table reflect rounded-off, average PDMS survey data.

• • These values do not include surface contamination on the cork seam.

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-13 TABLE 5.3-4, SURFACE CONTAMINATION - REACTOR BUILDING Area Description Principal Isotopes Curies Elevations 305 to 347 Cs-137 Sr-90 9.7 E-1 1.9 E-1 Elevations 347 and above Cs-137 Sr-90 6.7 E-1 3.3 E-1 Refueling Canal Cs-137 Sr-90 2.8 E-2 2.2 E-2 D-Ring Interior Elevation 349 and above Cs-137 Sr-90 3.2 E-2 2.0 E-3 Basement, Elevation 282 Cs-137 Sr-90 6.5 E+2 5.9 E+2

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-14 TABLE 5.3-5, SURFACE CONTAMINATION - AUXILIARY AND FUEL HANDLING BUILDINGS Cubicle No.

Area Description Principal Isotopes1 Curies2 AX001 Reactor Building Emergency Cooling Booster Pumps Area C

1.82E-3 AX002 Access Corridor B

5.45E-5 AX002a N2 Piping System C

6.67E-6 AX003 Access Area C

9.35E-4 AX004 Seal Injection Valve Room B

9.33E-3 AX005 MUP Pump 1C Room B

4.36E-3 AX006 MUP Pump 1B Room A

9.54E-3 AX007 MUP Pump 1A Room B

1.00E-3 AX008 Spent Resin Storage Tank 1B Room B

1.30E-1 AX009 Spent Resin Storage Tank 1A Room B

2.54E-1 AX010 Spent Resin Transfer Pump Room B

6.74E-2 AX011 Auxiliary Building Sump Tank Pumps and Valve Room B

1.18E-4 AX012 Auxiliary Building Sump and Tank Room B

3.97E-2 AX013 Evaporator Condensate Tanks, Pumps, and Demineralizer Room B

2.31E-5 AX014 RC Evaporator Room A

2.27E-3 AX015a Cleanup Filters Room A

3.78E-4 AX015b Cleanup After Filters Room A

7.71E-4 AX016 Cleanup Demineralizer 2A Room A

1.23E-3 AX017 Cleanup Demineralizer 2B Room A

6.03E-4 AX018 Waste Transfer Pumps Room B

1.06E-3 AX019 Waste Disposal Liquid Valve Room A

7.94E-4 AX020 RC Bleed Holdup Tanks 1B and 1C Room A

3.05E-1 AX021 RC Bleed Holdup Tank 1A Room B

7.97E-4 AX022 North Stairwell B

2.80E-5 AX023 Elevator Pit and Associated Equipment B

1.01E-3 AX024 Auxiliary Building Sump Filters Room B

1.78E-4 AX025 Area Between Service, Control, and Reactor Buildings B

1.73E-4 AX026 Seal Injection Filters Room C

1.90E-4 AX027 South Stairwell B

1.71E-5 AX101 Radwaste Disposal Control Panel Area B

1.96E-5 AX102 Reactor Building Sump Pumps Filters Room B

2.61E-4 AX103 MCC 2-11EB Room C

2.88E-5 AX104 MCC 2-21EB Room B

3.03E-5 AX105 Substation 2-11E Room B

7.10E-5 AX106 Substation 2-21E Room B

8.04E-5 AX107 MCC 2-11EA Room B

8.94E-5 AX108 MCC 2-21EA Room A

6.46E-5 AX109 Nuclear Services Coolers and Pumps Area B

3.42E-5 AX110 Intermediate Coolers Area C

3.54E-5

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-15 TABLE 5.3-5, SURFACE CONTAMINATION - AUXILIARY AND FUEL HANDLING BUILDINGS (cont.)

Cubicle No.

Area Description Principal Isotopes1 Curies2 AX111 Intermediate Cooling Pumps and Filters Room B

7.16E-5 AX112 Seal Return Coolers and Filter Room B

2.43E-2 AX113 Waste Gas Analyzer Room B

2.97E-3 AX114 MUP Demineralizer 1A Room B

5.99E-4 AX115 MUP Demineralizer 1B Room B

2.78E-3 AX116 Makeup Tank Room A

2.15E-2 AX117 MUP Filters Room C

2.58E+1 AX118 Spent Fuel Coolers and Pumps Area C

2.89E-4 AX119 Spent Fuel Demineralizer Room B

2.81E-5 AX120 Spent Fuel Filters Room A

9.33E-6 AX121 Inside Elevator Cab B

1.14E-5 AX122 North Stairwell B

2.95E-5 AX123 Access Area (includes AX-136, Hot Tool Room)

B 9.45E-5 AX124 Concentrated Liquid Waste Pump Room B

1.09E-4 AX125 Waste Gas Decay Tank 1B Room B

1.92E-4 AX126 Waste Gas Filter Room B

4.75E-6 AX127 Waste Gas Decay Tank 1A Room B

1.17E-3 AX128 Valve and Instrument Room B

8.02E-5 AX129 Deborating Demineralizer 1B Room B

5.80E-5 AX130 Deborating Demineralizer 1A Room B

3.21E-5 AX131 Miscellaneous Waste Holdup Tank Room B

2.69E-4 AX132 Corridor Between Unit 1 and Unit 2 B

1.47E-4 AX133 South Stairwell B

2.41E-5 AX134 Miscellaneous Waste Tank Pumps Room A

1.38E-3 AX135 Radwaste Disposal Control Panels B

4.61E-6 AX201 North Stairwell B

2.54E-5 AX202 Elevator Shaft B

2.16E-5 AX203 4160V Switchgear 2-1E Room B

1.05E-4 AX204 4160V Switchgear 2-2E Room B

1.09E-4 AX205 Reactor Building Purge and H2 Control Area A

3.95E-5 AX206 Reactor Building Purge Air Exhaust Unit B B

1.41E-2 AX207 Reactor Building Purge Air Exhaust Unit A B

1.56E-2 AX208 Auxiliary Building Exhaust Unit B B

2.42E-4 AX209 Auxiliary Building Exhaust Unit A B

6.24E-4 AX210 Fuel Handling Building Exhaust Unit B B

6.06E-4 AX211 Fuel Handling Building Exhaust Unit A B

3.56E-4 AX212 Decay Heat Surge Tank and Substation Area B

6.33E-5 AX213 Unit Substations and Access Area C

8.11E-5 AX214 Decontamination Facility C

1.68E-4

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-16 TABLE 5.3-5, SURFACE CONTAMINATION - AUXILIARY AND FUEL HANDLING BUILDINGS (cont.)

Cubicle No.

Area Description Principal Isotopes1 Curies2 AX215 Fuel Handling Building Supply Unit C

2.80E-5 AX216 Auxiliary Building Supply Unit B

3.36E-5 AX217 Access Area B

6.94E-5 AX218 Concentrated Waste Storage Tank Room B

1.44E-4 AX219 Instrument Racks and Atmospheric Monitor Area B

6.39E-5 AX220 Caustic Liquids Mixing Area B

4.28E-5 AX221 Caustic Liquids Mixing Area Corridor B

1.12E-4 AX222 South Stairwell B

4.89E-6 AX223 Air Handling Units General Area C

3.79E-4 AX301 Elevator Machine Room C

5.67E-6 AX302 North Stairwell B

2.40E-5 AX303 Elevator and Stairwell Access C

4.83E-5 AX304 Auxiliary Building Exhaust Fans C

2.22E-5 AX305 Fuel Handling Building Exhaust Fans A

1.94E-5 AX401 Roof A

1.54E-4 AX402 Cooling Water Surge Tanks Room C

1.55E-5 AX403 Damper Room B

1.26E-5 AX501 Reactor Building Spray Pump 1A Room A

2.98E-2 AX502 Reactor Building Spray Pump 1B Room B

8.56E-3 AX503 Decay Heat Removal Cooler and Pump 1A Room A

8.08E-3 AX504 Decay Heat Removal Cooler and Pump 1B Room A

2.67E-3 FH001 Makeup Suction Valve Room C

1.57E-2 FH002 Access Corridor C

2.17E-4 FH003a Makeup Discharge Valve Room B

3.45E-3 FH003b Makeup Discharge Valve Room B

1.75E-2 FH004 Westinghouse Valve Room C

6.66E-3 FH005 Mini Decay Heat Vault B

8.61E-5 FH006 Decay Heat Service Coolers Area B

4.48E-4 FH007 Neutralizer and Reclaimed Boric Acid Access Area B

4.50E-5 FH008 Neutralizer Tanks Pumps Room B

2.17E-3 FH009 Neutralizer Tanks Room B

2.94E-3 FH010 Reclaimed Boric Acid Tank Room A

1.97E-4 FH011 Reclaimed Boric Acid Pump Room A

9.16E-4 FH012 Neutralizer Tanks Filters Room B

4.29E-5 FH013 Oil Drum Storage Area B

4.49E-6 FH014 Annulus A

5.22E-3 FH101 MUP Valve Room B

1.14E-1 FH102 East Corridor B

5.67E-4 FH103 Sample Room B

2.40E-4

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-17 TABLE 5.3-5, SURFACE CONTAMINATION - AUXILIARY AND FUEL HANDLING BUILDINGS (cont.)

Cubicle No.

Area Description Principal Isotopes1 Curies2 FH104 West Corridor C

3.20E-5 FH105 Model Room B

4.38E-4 FH106 Monitor Tanks and Sample Sink Area C

7.08E-5 FH107 Trash Compactor Area B

5.62E-6 FH108 Truck Bay A

5.46E-5 FH109 Spent Fuel Pool A3 C

1.35E+2 FH110 SDS Spent Fuel Pool3 C

4.62E-2 FH111 Fuel Cask Storage3 C

1.14E-2 FH112 Annulus B

4.39E-4 FH201 East Corridor B

1.51E-4 FH202 West Corridor B

1.07E-4 FH203 Surge Tank Area B

4.35E-5 FH204 Standby Pressure Control Area C

2.36E-4 FH205 Annulus B

7.24E-5 FH301 Upper Spent Fuel Pool A Area C

5.30E-5 FH302 SDS Operating Area B

1.53E-4 FH303 Upper Standby Pressure Control Area C

1.92E-4 FH304 Annulus B

6.05E-4 FH305 Spent Fuel Pool Access Area C

2.65E-4 1The principal isotopes and their relative distribution are defined below. The Sr-90 value represents the sum of the Sr-90 and Y-90 isotopes which are in equilibrium; the Cs-137 value represents the sum of the Cs-137 and Ba-137m isotopes which are in equilibrium. The A, B, and C categories relate to normal, makeup, and defueling waste streams, respectively. Only those isotopes important from an offsite dose perspective are included.

A B

C Sr-90 0.08 Sr-90 0.29 Sr-90 0.63 Cs-137 0.92 Cs-137 0.71 Cs-137 0.28 Pu-238 4.43E-6 Pu-238 1.67E-5 Pu-238 4.25E-4 Pu-239 5.39E-5 Pu-239 2.04E-4 Pu-239 5.18E-3 Pu-240 1.43E-5 Pu-240 5.41E-5 Pu-240 1.37E-3 Pu-241 4.86E-4 Pu-241 1.84E-3 Pu-241 0.04 Am-241 1.56E-5 Am-141 5.92E-5 Am-141 1.50E-3 Pm-147 0.04 2These are calculated values based on the specific decontamination values given in Table 5.3-2, PDMS Radiological Conditions - Auxiliary and Fuel Handling Buildings.

3A metal cover with an access door was placed over these areas to prevent spread of contamination.

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-18 TABLE 5.3-6, SURFACE CONTAMINATION - OTHER BUILDINGS Cubicle No.

Area Description Principal Isotopes1 Curies2 SB000 Service Building Elevation 281 A

1.31E-4 SB002 M-20 Area A

1.16E-4 SB002 M-20 Area Sump A

1.04E-5 SB100 Service Building Elevation 305 (Reactor Building Containment Control Cubicle)

(Secondary Chemistry Lab)

A B

B 7.23E-5 5.72E-5 3.81E-4 SB500 Tendon Access Gallery A

1.22E-4 TB000 Turbine Building Elevation 281' A

5.90E-4 PA108 CACE Building B

2.69E-5 RA101 PWST Pump House (PWST Sump)

A A

2.72E-5 2.53E-6 RA104 BWST Area A

2.92E-5 1The principal isotopes and their relative distribution are defined below. The Sr-90 value represents the sum of the Sr-90 and Y-90 isotopes which are in equilibrium; the Cs-137 value represents the sum of the Cs-137 and Ba-137m isotopes which are in equilibrium.

The A and B categories relate to normal and makeup waste streams, respectively. Only those isotopes important from an offsite dose perspective are included.

A B

Sr-90 0.08 Sr-90 0.29 Cs-137 0.92 Cs-137 0.71 Pu-238 4.43E-6 Pu-238 1.67E-5 Pu-239 5.39E-5 Pu-239 2.04E-4 Pu-240 1.43E-5 Pu-240 5.41E-5 Pu-241 4.86E-4 Pu-241 1.84E-3 Am-241 1.56E-5 Am-241 5.92E-5 2These are calculated values based on the specific decontamination values presented in Table 5.3-3, PDMS Radiological Conditions - Other Buildings.

5.3.3 REACTOR BUILDING DRONE SURVEY On June 23, 2021, an aerial drone was used to assess conditions in the TMI-2 Reactor Building. As part of that survey, radiological conditions were measured.

The results of the survey are provided in Appendix 5A, Results of TMI-2 Reactor Building Drone Surveys.

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-19 5.4 RADIATION PROTECTION PROGRAM 5.4.1 EQUIPMENT, INSTRUMENTATION, AND FACILITIES A. Personnel Monitoring Administrative controls provide the requirements for personnel entering radiologically posted areas onsite and the need to wear personnel monitoring devices based on radiological hazards. These administrative controls also provide for the use of portal personnel radiation monitors.

A portal personnel radiation monitor is provided at the plant exit for monitoring of surface and internal activity of people leaving the plant. The portal monitor provides for complete head-to-foot coverage. The portal console monitor located on the portal frame includes status lights, as well as a contamination alarm. The contamination signal from the console alerts personnel to the contamination condition so that the proper action can be taken if necessary.

B. Protective Clothing The nature of the decommissioning work and the radiological conditions are the governing factors in the selection of protective clothing. Example of protection apparel available are shoe covers, rubber overshoes, head covers, beanies, gloves (cotton liners and rubber gloves), and coveralls or lab coats. Additional items of specialized apparel such as plastic or rubber suits, face shields, and respirators are available for activities involving high-level contamination and airborne radioactivity areas. In all cases, administrative controls provide the requirements for RP to evaluate the radiological conditions and specify the required items of protective clothing.

C. Physical Barriers for Access to High Radiation Areas Controlled areas are posted as radiation areas, high radiation areas, radioactive material areas, airborne radioactivity areas, or combinations thereof. Access to controlled areas for all work is authorized in accordance with RP procedures.

High radiation areas are controlled as required by TMI-2 Technical Specifications, Section 6.11 (Ref. 1.3.54).

D. Records RP-related records are maintained in accordance with administrative procedures and a retention and retrieval system.

5.4.2 PROCEDURES TMI-2 maintains a Radiation Protection Program (RPP) (see TMI2-RP1-PG-001, Radiological Protection Program [Ref. 1.3.46]) that meets or exceeds standards for protection against exposures to radiation and radioactivity. The implementation of the RPP ensures that the facility will be managed and maintained during decommissioning in a manner that minimizes risk to employees, contractors, visitors, and the public of exposure to radiation and radioactivity at the facility. The implementation of the RPP also ensures a radiologically safe working environment for employees and visitors at TMI-2.

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-20 The RPP and related procedures are designed to provide protection of personnel against exposure to radiation and RAM in a manner consistent with applicable regulations. The policy of TMI-2 Solutions is to maintain personnel radiation exposure As Low As Is Reasonably Achievable (ALARA). Therefore, each individual is trained to minimize his or her exposure consistent with discharging his or her duties. Each individual is responsible for observing rules adopted for his or her safety and that of others. RP procedures are in place which factor in ALARA controls.

RP personnel evaluate radiological conditions during decommissioning in accordance with established and required RP procedures. RP personnel will ensure compliance with all applicable regulations and will ensure that the required records are adequately maintained.

Training of site personnel in RP principles and procedures is given at the beginning of their work assignments and periodically retraining thereafter as defined in RP procedures.

Procedures are in place which require performance of ALARA reviews as necessary as decommissioning progresses.

5.5 RADIOACTIVE WASTE MANAGEMENT Liquid radwaste management systems that are operational during DECON are the Radioactive Waste - Miscellaneous Liquids (WDL) System and the Sump Pump Discharge and Miscellaneous Sumps System. Major portions of these two systems are operational to prevent localized flooding and to provide proper disposal of effluents.

5.5.1 RADIOACTIVE WASTE - MISCELLANEOUS LIQUIDS (WDL) SYSTEM A. Function Portions of the WDL System remain operational. This status provides assurance that significant quantities of liquid wastes will not accumulate in an uncontrolled manner in the Auxiliary and Reactor Buildings. Liquid radwaste in these buildings may result from either rainwater in-leakage or decommissioning activities. The WDL System currently achieves its objective by meeting the following criteria:

1. Existing sumps in the Auxiliary and Reactor Buildings will be monitored and pumped, as required.

2. Liquid storage capabilities have been provided for accumulation until sufficient quantities are available for batch processing, as necessary.

3. A Receiving Tank (T-3) is in place in the Kelly Enclosure (Auxiliary Building, 281-ft. Elevation) to accept water from the Miscellaneous Waste Holdup Tank (MWHT) to recirculate, sample, and transfer liquid radioactive waste to a vendor truck for offsite processing/disposal or released in accordance with 10 CFR 20, Standards for Protection Against Radiation (Ref. 1.3.1); 10 CFR 50, Domestic Licensing of Production and Utilization Facilities (Ref. 1.3.3); and applicable NPDES regulations.

Different provisions for handling liquid wastes will be employed as progress is made in decommissioning.

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-21 5.5.2 SUMP PUMP DISCHARGE AND MISCELLANEOUS SUMPS SYSTEM A. Function There are a number of sumps in TMI-2 that will be maintained in an operational condition. The various sumps and their locations are listed in Table 5.5-1, Operational Sump Systems.

Maintaining the various building sumps operational assures that water buildup does not cause adverse localized flooding. These sumps will contain water that is either clean or slightly radioactive. Clean water is presently routed to the Industrial Waste Treatment System (IWTS). Radioactive water will be processed and discharged via approved pathways; slightly radioactive water will be pumped to the IWTS and released in accordance with 10 CFR 20, Standards for Protection Against Radiation (Ref. 1.3.1); 10 CFR 50, Domestic Licensing of Production and Utilization Facilities (Ref. 1.3.3); and applicable NPDES regulations. The discharge from the IWTS is monitored for radiation in accordance with the ODCM (Ref. 1.3.30).

TABLE 5.5-1, OPERATIONAL SUMP SYSTEMS Sumps Associated with SD System Sump Location Turbine Building Sump Turbine Building Control Building Area Sump M-20 Area West Control and Service Buildings Sump Service Building Tendon Access Gallery Sump Tendon Access Gallery Air Intake Tunnel Normal Sump Air Intake Tunnel Sumps Associated with WDL System Sump Location Reactor Building Basement Sump Reactor Building Auxiliary Building Sump Auxiliary Building Decay Heat Removal Pump Room Sumps (2)

Auxiliary Building Reactor Building Spray Pump Room Sumps (2)

Auxiliary Building Contaminated Drain Tank Room Sump Service Building 5.6 RADIATION MONITORING 5.6.1 FUNCTION The radiation monitoring requirements for the facility are primarily those associated with the radiological conditions in the facility and effluent monitoring. The offsite dose calculations for normal time periods and unanticipated events (refer to Chapter 8, Routine and Unanticipated Releases) are based on assumed and measured radiological conditions associated with the various areas of the facility. In order to assure that the offsite dose calculations for the various events remain bounding, the radiological conditions are periodically monitored to assure they remain within acceptable bounds. In addition, all effluents are monitored for effluent reporting.

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-22 Broader radiological conditions monitoring will be conducted throughout the facility to assure compliance with good radiological condition practices and 10 CFR 20, Standards for Protection Against Radiation (Ref. 1.3.1). These radiological monitoring activities are required to support activities such as decontamination, dismantlement, visual inspections, preventive maintenance, or other routine tasks.

5.6.2 RADIOLOGICAL SURVEYS A. Auxiliary and Fuel Handling Buildings Radiological Surveys Radiological surveys will be conducted on a periodic basis to monitor radiological conditions in the Auxiliary and Fuel Handling Buildings. These radiological surveys will consist of air sampling and loose surface contamination and radiation dose rate surveys.

B. Reactor Building Radiological Surveys Radiological surveys will be conducted on a periodic basis to monitor radiological conditions in the Reactor Building. These radiological surveys will consist of air sampling and loose surface contamination and radiation dose rate surveys.

5.6.3 EFFLUENT MONITORING Airborne effluents will be monitored during ventilation of the Reactor Building.

During Reactor Building Ventilation System operation, the station ventilation stack monitor (HP-R-219/HP-R-219A) will provide real-time monitoring of releases.

In the Auxiliary and Fuel Handling Building, the ventilation system exhaust pathway is also monitored by the station ventilation stack monitor (HP-R-219/

HP-R-219A), thus assuring a monitored effluent release.

5.6.4 GENERAL RADIOLOGICAL MONITORING Throughout decommissioning, an ongoing radiological surveillance program will monitor radiological conditions within the Reactor, Auxiliary and Fuel Handling Buildings. By means of various surveys, potential degradation of radiological conditions will be identified in order for appropriate remedial actions to be taken.

Radiological surveys will be performed in areas requiring access for decontamination, decommissioning, visual inspection, preventive maintenance, or other routine tasks. Radiological support of work will be conducted in accordance with RP procedures and good radiological work practices.

5.7 SEALED SOURCES Sealed Source requirements from TMI-2 Technical Specifications (Ref. 1.3.54) have been transferred to the DSAR as provided in Appendix 5B, Sealed Source Requirements.

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-23 APPENDIX 5A, RESULTS OF TMI-2 REACTOR BUILDING DRONE SURVEYS (PAGE 1 OF 3)

On June 23, 2021, an aerial drone was used to access conditions inside the TMI-2 Reactor Building. As part of this assessment, a radiation survey was performed. The results of the radiation survey are provided below for each Reactor Building Elevation. Legends for each area are modified to give the best visual representation of the dose rates in the area. Dose surveys are shown with bands of consistent dose rates seen across the areas. A table was constructed with these ranges of dose rates and each area was given a roman numeral identifier. The table shows the corresponding measured dose rate, as seen on the dosimeter, and a corrected dose rate, that was adjusted based on Exelon Powerlabs testing.

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-24 APPENDIX 5A RESULTS OF TMI-2 REACTOR BUILDING DRONE SURVEYS (PAGE 2 OF 3)

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-25 APPENDIX 5A RESULTS OF TMI-2 REACTOR BUILDING DRONE SURVEYS (PAGE 3 OF 3)

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-26 APPENDIX 5B, SEALED SOURCE REQUIREMENTS (PAGE 1 OF 2) 3/4.4 SEALED SOURCES 3/4.4.1 SEALED SOURCE INTEGRITY LIMITING CONDITIONS 3.4.1 Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material (except as noted in 4.4.1.2) shall be free of 0.005 microcuries of removable contamination.

APPLICABLE: DURING DECON ACTION:

a.

Each sealed source with removable contamination in excess of the above limit shall be immediately withdrawn from use and:

1.

Either decontaminate and repair, or

2.

Dispose in accordance with Commission Regulations.

SURVEILLANCE REQUIREMENTS TEST REQUIREMENTS 4.4.1.1 Each sealed source shall be tested for leakage and/or contamination by:

a.

The licensee, or

b.

Other persons specifically authorized by the Commission or an Agreement State.

The test method shall have a detection sensitivity of at least 0.005 microcuries per test sample.

TEST FREQUENCIES 4.4.1.2 Each category of sealed source shall be tested at the frequency described below.

a.

Source in use (excluding fission detectors previously subjected to core flux) - At least once per six months for all sealed sources containing radioactive material:

1.

With half-life greater than 30 days (excluding Hydrogen 3) and

2.

In any form other than gas.

b.

Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous six months. Sealed sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to being placed into use.

c.

Fission detectors - Each sealed fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair or maintenance to the source.

TMI-2 Defueled Safety Analysis Report Chapter 5:

Revision 0 Radiological Conditions 5-27 APPENDIX 5B SEALED SOURCE REQUIREMENTS (PAGE 2 OF 2)

REPORTS 4.4.1.3 A report shall be prepared and submitted to the Commission on an annual basis if sealed source or fission detector leakage tests reveal the presence of 0.005 microcuries of removable contamination.

3/4.4 SEALED SOURCES BASES 3/4.4.1 SEALED SOURCE INTEGRITY The limitation on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. This limitation will ensure that leakage from byproduct, source, and Special Nuclear Material sources will not exceed allowable intake values.

TMI-2 Defueled Safety Analysis Report Revision 0 DEACTIVATED SYSTEMS AND FACILITIES DELETED

TMI-2 Defueled Safety Analysis Report Revision 0 OPERATIONAL SYSTEMS AND FACILITIES

TMI-2 Defueled Safety Analysis Report Chapter 7:

Revision 0 Operational Systems and Facilities 7-2

7.1 INTRODUCTION

This chapter describes those systems and facilities which will be maintained in an operational condition at the start of decommissioning. Generally, those facilities which are maintained operational are those buildings or areas that contain operational systems or partially operational systems.

Operational facilities and systems serve several functions within the scope of decommissioning activities, including support of site operations, maintenance activities, and surveillance activities. Table 7.2-1, Operational Facilities, and Table 7.3-1, Operational Systems, provide a listing of those facilities and systems which will be maintained in an operational condition at the start of DECON. These tables also provide other relevant information concerning the status of the listed facilities and systems.

The following systems and facilities discussed in Chapter 7 provide reasonable assurance that TMI-2 can be maintained during DECON with minimal risk to the health and safety of the public: 1) Reactor Building structure; 2) Reactor Building ventilation; 3) fire protection; 4) Auxiliary and Fuel Handling Buildings ventilation systems; 5) associated support and monitoring systems; and 6) Unit 2s flood protection capabilities.

7.2 OPERATIONAL FACILITIES TMI-2 facilities required to be operational are described in this section. Facilities are required to be operational to support operational systems within those facilities and/or to isolate internal contamination from the environment.

7.2.1 REACTOR BUILDING The Reactor Building provides shielding of the environment from the contained radiation. It also provides the means to assure that any effluents from the Reactor Building will be controlled, filtered, and monitored. The Reactor Building was designed to withstand airplane crashes, a safe-shutdown earthquake, tornados, floods, and other natural phenomena. The Reactor Building was also designed with the capability to isolate any radioactive materials produced as a result of accidents or other unplanned events.

Although modifications were made to several of the Reactor Building penetrations, the structural capabilities of the Reactor Building were not significantly diminished by the accident or any of the cleanup activities. The Containment Air Control Envelope (CACE) will provide a barrier to the environment for the probable natural phenomena as required by the local building codes.

7.2.2 AUXILIARY BUILDING The Auxiliary Building will serve primarily to support operation of the liquid radwaste, Auxiliary Building sump, HEPA-filtered ventilation, and effluent monitoring systems required for decommissioning activities.

TMI-2 Defueled Safety Analysis Report Chapter 7:

Revision 0 Operational Systems and Facilities 7-3 7.2.3 FUEL HANDLING BUILDING During decommissioning, the Fuel Handling Building is not required for storage of new or spent fuel. However, it may be utilized for the temporary staging of site-generated radwaste or other appropriate uses. During decommissioning, the capability to ventilate the Fuel Handling Building with a HEPA-filtered pathway to the environment will be maintained.

7.2.4 FLOOD PROTECTION Flood Protection for TMI-2 is described in Section 2.4.

7.2.5 CONTROL AND SERVICE BUILDINGS The Control Building houses the TMI-2 Control Room, a relay room, two deactivated inverter battery rooms, a cable spreading room, and a mechanical equipment room. Although the Alarm Monitoring System directs instrument alarm outputs to the TMI-1 Control Room, the TMI-2 Control Room annunciators/panel indications will be relied upon to provide specific information should the need arise.

The Service Building houses the operational Compressed Air System compressors, receiver tanks, and associated piping. In addition, the Service Building provides access to the Reactor Building, the Auxiliary Building, the Control Building, and the Air Intake Tunnel. The Control and Service Buildings sumps will maintain functionality as needed for collection of water intrusion, and means will be provided to pump contaminated water out to a storage or processing facility.

7.2.6 TURBINE BUILDING The Turbine Building houses operational support systems. Operational systems located in the Turbine Building include:

A. The Turbine Building 13.2KV power feed load disconnects, which feeds a portion of the 480/120VAC distribution system and the 125VDC distribution system.

B. Sump pump and Sump Discharge (SD) lines from TMI-2.

C. Turbine Building crane.

7.2.7 CONTAINMENT AIR CONTROL ENVELOPE The CACE is currently used as a staging/packing area for materials and equipment requiring transfer into or out of the reactor building, while helping to control airborne releases from the reactor building. The CACE will also permit the removal of the equipment hatch without exposing the Reactor Building contents directly to the atmosphere.

The CACE is designed for the probable natural phenomena as required by the local building codes. It does not have as part of its design basis the severe natural phenomena used for permanent nuclear power plant structures. These severe natural phenomena (e.g., tornadoes, Safe Shutdown Earthquakes [SSE], and probable maximum floods) are not postulated to occur during the short-term design life of the CACE.

TMI-2 Defueled Safety Analysis Report Chapter 7:

Revision 0 Operational Systems and Facilities 7-4 7.3 OPERATIONAL SYSTEMS The following sections describe the systems that will maintain functionality as needed. Table 7.3-1, Operational Systems, provides a listing of those systems. Also listed are the system codes, internal contamination, and any relevant remarks regarding operation.

Each of the following sections addresses the function of the specific system. A list of references providing more detailed information is included in Section 7.1. As decommissioning progresses, these systems will be modified, replaced by temporary systems, and eventually entirely removed when they are no longer required.

As required by the DQAP (Ref. 1.3.40), procedures are in place to control the operations, surveillance, testing, modifications, and maintenance of the operational systems.

7.3.1 REACTOR BUILDING SYSTEMS The following systems or portions of systems maintain functionality as needed to support decommissioning.

A. Reactor Building Ventilation The Reactor Building Ventilation System ensures that uncontrolled atmospheric migration of radioactive contamination will not create a hazard to either the public or site personnel. The Reactor Building Ventilation System consists of supply and exhaust units and associated ductwork, dampers, and filters.

B. Reactor Building Airlock The personnel airlock doors will be used for Reactor Building ingress and egress. The airlock is designed as a double-door system with one of the doors always closed during routine entry into the Reactor Building. Under these conditions, the Reactor Building remains isolated (an enclosed volume) and the amount of Reactor Building air which could be released to or from the Reactor Building is limited to the volume inside of the airlock assembly.

There are situations when it is necessary to open both doors of the airlock assembly simultaneously, as is the case of movement of a long piece of equipment into or out of the Reactor Building. These entries will be controlled in accordance with TMI2-RP1-PG-001, Radiological Protection Program (Ref. 1.3.46).

7.3.2 FIRE PROTECTION, SERVICE, AND SUPPRESSION TMI-2 has been permanently defueled. The Fire Protection requirements of 10 CFR 50.48(f) (Ref. 1.3.3) apply. The requirements of 10 CFR 50.48(f) are described more thoroughly in NRC Regulatory Guide 1.191, Fire Protection Program for Nuclear Power Plants During Decommissioning and Permanent Shutdown (Ref. 1.3.11). The TMI-2 Fire Protection Program meets the requirements of NRC Regulatory Guide 1.191, Fire Protection Program for Nuclear Power Plants During Decommissioning and Permanent Shutdown (Ref. 1.3.11).

The FPPE (Ref. 1.3.37) describes how the program satisfies these requirements.

TMI-2 Defueled Safety Analysis Report Chapter 7:

Revision 0 Operational Systems and Facilities 7-5 7.3.3 ELECTRICAL SYSTEMS Various plant systems are required to maintain functionality as needed to support the dismantlement, monitoring, protection, and surveillance activities associated with decommissioning. Some systems require continuous operation, while others require only intermittent operation. Due to the need for electrical power support for these activities, the Electrical Distribution System will maintain functionality as needed and remain energized.

7.3.4 SUPPORT SYSTEMS The operational systems discussed in this section provide the necessary measures to support activities at TMI-2. Although they do not directly ensure protective functions, their operation is necessary to carry out anticipated operation, inspection, surveillance, maintenance, and decommissioning activities.

A. Auxiliary Building Ventilation System The ability to ventilate the Auxiliary Building will be maintained to support decommissioning activities. When in operation, this system performs the following functions:

• Provides fresh air in a sufficient quantity to maintain room temperatures compatible for personnel and equipment.

• Minimizes the spread of contamination by providing air flow from clean areas to potentially contaminated areas and to the exhaust.

• Filters exhaust air.

B. Fuel Handling Building Ventilation System The ability to ventilate the Fuel Handling Building will be maintained to support decommissioning activities. When in operation, this system performs the following functions:

• Provides fresh air in a sufficient quantity to maintain room temperatures suitable for personnel and equipment.

• Minimizes the spread of contamination by providing air flow from clean areas to potentially contaminated areas and then to the exhaust.

• Filters exhaust air.

C. Compressed Air Supply System Portions of the original plant Instrument and Service Air Systems are to be utilized to provide compressed air to operational pneumatic devices.

D. Control Room Ventilation System The Control Room Ventilation System will be maintained in an operational condition to support decommissioning activities. This system provides fresh, heated or cooled air in a sufficient quantity to support personnel occupancy and equipment protection.

TMI-2 Defueled Safety Analysis Report Chapter 7:

Revision 0 Operational Systems and Facilities 7-6 E. Service Building Ventilation System The Service Building Ventilation System will be maintained in an operational condition to support decommissioning activities. When in operation, this system performs the following functions:

• Provides fresh, heated air in a sufficient quantity to maintain room temperatures suitable for personnel and equipment.

• Minimizes the spread of contamination by providing air flow from clean areas to potentially contaminated areas and then to the exhaust.

• Filters exhaust air.

F. Alarm Monitoring System The function of the alarm monitoring system is to notify plant operations personnel of an abnormal plant condition which requires operator action to correct or which represents a threat to plant, personnel, or equipment safety.

The Alarm Monitoring System provides the means to remotely monitor select TMI-2 alarms and TMI-2 station vent monitor signals. As required by EP-TM-1002, Three Mile Island Independent Spent Fuel Storage Installation (ISFSI) Only Emergency Plan (IOEP) (Ref. 1.3.31), the Alarm Monitoring System is designed such that if the remote monitoring of the alarms becomes inoperable, the TMI-2 Control Room alarms and station vent monitor signals can be monitored from the annunciators and other recorders/equipment in the TMI-2 Control Room.

TMI-2 Defueled Safety Analysis Report Chapter 7:

Revision 0 Operational Systems and Facilities 7-7 TABLE 7.3-1, OPERATIONAL SYSTEMS

System Description

System Code Remarks Reactor Building Ventilation AH Ventilation will be operational as determined by the RP program.

Reactor Building Airlock and Equipment Hatch RBA Airlocks for personnel/equipment access.

Fire Protection FP See TMI2-FP-EVA-0001, TMI-2 Fire Protection Program Evaluation (FPPE) (Ref. 1.3.37).

Waste Disposal - Liquid (Miscellaneous)

WDL Necessary equipment/tanks to process water will be maintained operational. The building sump pumps, the miscellaneous waste holdup tank (WDL-T-2), the ABST (WDL-T-5), and interconnecting pipe shall maintain functionality as needed for water removal functions.

Sump Pump Discharge and Miscellaneous SD Facilities are sealed to limit exterior water ingress. Periodic sump pump operations will prevent sump accumulation of drainage and inadvertent in-leakage.

Radiation Monitoring HP Radiation monitors and alarms maintain functionality as needed -selected HP monitoring, and survey programs are also continued.

Electric Distribution EE Electrical equipment that supports operable systems and facilities shall maintain functionality as needed.

Auxiliary Building Ventilation AH Ventilation will be operational to the extent necessary Fuel Handling Building Ventilation AH Ventilation will be operational to the extent necessary.

Compressed Air Supply IA/SA Air-cooled air compressors use portions of Instrument and Service Air Systems.

Control Building Ventilation AH Ventilation will be operational to the extent necessary.

Service Building Ventilation AH Ventilation will be operational to the extent necessary. Includes Control Building Area Ventilation.

Alarm Monitoring System

TMI-2 Defueled Safety Analysis Report Revision 0 ROUTINE AND UNANTICIPATED RELEASES

TMI-2 Defueled Safety Analysis Report Chapter 8:

Revision 0 Routine and Unanticipated Releases 8-2 8.1 GENERAL The primary objective of the TMI-2 Cleanup Program was the elimination of radiological hazards to the public resulting from the accident on March 28, 1979, and minimization of onsite worker exposure. The program progressed from the initial efforts to stabilize the plant conditions through the final major cleanup efforts, including the removal and shipment of fuel and decontamination of major portions of the Auxiliary, Fuel Handling, and Reactor Buildings.

The potential for release of significant quantities of radionuclides during decommissioning is substantially reduced from the potential for release during normal power plant operation.

This decrease results from the reduced radionuclide inventory (shown in Table 5.3-4, Surface Contamination - Reactor Building, and Table 5.3-5, Surface Contamination -

Auxiliary and Fuel Handling Buildings). Therefore, the assessment of any radionuclide release during DECON hinges on the identification of processes or events that could either alter the potential for transport of the remaining radionuclide inventory or provide unanticipated transport mechanisms to the environment. A range of potential unanticipated events has been postulated to establish the bounding conditions of potential offsite releases.

In addition to routine releases, the environmental effects for unanticipated events that result in offsite radiation exposures in excess of those which result from routine releases are bounded by the ODCM (Ref. 1.3.30) and the TMI IOEP (Ref. 1.3.31).

8.1.1 ROUTINE RELEASES Atmospheric releases to the environment during routine decommissioning operations will be limited to airborne contamination released as a result of operating the building ventilation systems. Ventilation discharges will be through controlled, filtered, and monitored paths. As described in the ODCM (Ref. 1.3.30), this control requires that the dose to offsite personnel be limited to the design objectives of 10 CFR 50, Domestic Licensing of Production and Utilization Facilities, Appendix I (Ref. 1.3.3). This will assure the dose received by the public during DECON is equivalent to or less than that from a normal operating reactor. The limits also assure that the environmental impacts are consistent with those assessed in NUREG-0683, Final Programmatic Environmental Impact Statement Related to Decontamination and Disposal of Radioactive Wastes Resulting from March 28, 1979 Accident, Three Mile Island Nuclear Station, Unit 2 (PEIS) (Ref. 1.3.16).

Liquid systems, except for systems needed to occasionally process batches of contaminated liquids, were drained to the extent practical and deactivated for decommissioning. The major sources of contaminated liquids requiring processing during decommissioning are expected to be water used to provide shielding for core debris material removal and any necessary component sectioning, groundwater in-leakage, collected precipitation, and occasional small quantities of fluids used for local decontamination. Rainwater and groundwater in-leakage will be collected and analyzed for any contamination. The capability of processing this liquid will be available to ensure that discharges are well within regulatory requirements.

TMI-2 Defueled Safety Analysis Report Chapter 8:

Revision 0 Routine and Unanticipated Releases 8-3 8.1.2 SOURCE TERMS The inventory of radionuclides remaining onsite is greatly reduced from that existing prior to the accident or during any of the phases of the recovery operations.

This results primarily from the following:

A. Removal of the fuel, which represents the largest concentration of radionuclides, B. Processing and shipping radioactive waste, and C. Natural decay.

The remaining radioactivity can be characterized as residual contamination located primarily in either closed piping systems that were drained but not aggressively decontaminated or surface films that are tightly adherent to equipment or structural surfaces. An exception is the Reactor Building basement (282-ft. elevation).

The largest source of radioactivity in the Reactor Building basement is the block wall enclosing the stairwell and elevator. Radionuclides (primarily cesium and strontium) have been absorbed into the concrete structure of the blocks during the period when the wall was partially submerged in highly contaminated water that collected in the Reactor Building basement during and following the accident.

Since the Radioactive Material (RAM) is embedded in the concrete, it is not readily available as a source for airborne release in the near term. However, over longer periods of time, mechanisms related to diffusion and leaching by cyclic changes in moisture content may transport a fraction of the radionuclides in the block wall to the surface where it can become available for suspension. Even though this fraction is expected to be small, the large inventory of the block wall (i.e., an estimated 19,000 Ci of Cs and 750 Ci of Sr prior to decay during PDMS) could make any suspension of radionuclides reaching the surface a significant airborne source term.

Other major sources of radioactivity in the Reactor Building that could make a significant contribution to the airborne source term include the remaining wall and floor areas that were submerged in the highly contaminated water located in the Reactor Building basement following the accident, the interior of the D-rings, and the sediment remaining in the Reactor Building basement after the completion of sediment removal activities.

Another important factor in the consideration of residual contamination is the transuranic content. Although the quantity of fuel remaining after completion of defueling is insufficient to be of concern with respect to criticality, it is necessary to examine the potential contribution it could make to radiological source terms.

TMI-2 Defueled Safety Analysis Report Chapter 8:

Revision 0 Routine and Unanticipated Releases 8-4 The relative fractions of the significant transuranic elements remaining in the residual fuel are given in Table 8.1-1, Ci Fractions in Residual Fuel. The Ci fractions for the residual fuel were calculated based on the original core inventory corrected for 8 years decay. On the basis of the samples analyzed to date, as well as the analyses of the course of the accident, the transuranic elements can be assumed to be associated with residual fuel. Most of the residual fuel remaining during decommissioning will be fixed in the form of fine and granular debris that was inaccessible to defueling during earlier clean-up operations, tightly adherent surface deposits not readily removable by available dynamic defueling techniques, and resolidified material that is tightly adherent to the RV components.

8.1.3 UNANTICIPATED RELEASES The offsite doses from unanticipated events are bounded by the ODCM (Ref. 1.3.30) and the TMI IOEP (Ref. 1.3.31), as discussed in Section 8.2.

TABLE 8.1-1, CI FRACTIONS IN RESIDUAL FUEL Isotope Ci Fraction Pu-238 1.12 E-2 Pu-239 6.81 E-2 Pu-240 3.32 E-2 Pu-241 7.73 E-1 Am-241 1.15 E-1 Total 1.00 8.2 UNANTICIPATED EVENTS ANALYSIS 8.

2.1 INTRODUCTION

Upon entry into DECON, major decommissioning activities will begin with Phase 1B, which entails activities necessary to complete the cleanup from the accident that occurred on March 28, 1979 (i.e., source term reduction and debris material removal). The objective of Phase 1B is to achieve building and equipment decontamination to the point where general area dose rates approximate those of an undamaged reactor nearing the end of its operating life. At the completion of Phase 1B, TMI-2 will prepare for Phase 2 decommissioning, which entails typical D&D activities.

As described in TMI2-RA-COR-2023-0002, Supplement to License Amendment Request - Proposed Changes to TMI-2 Possession Only License and Technical Specifications (Ref. 1.3.45), a decision was made to use administrative controls to limit any accidental offsite release to less than the criteria for initiating an Unusual Event, as defined by NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors (Ref. 1.3.10), for permanently defueled reactors.

TMI-2 Defueled Safety Analysis Report Chapter 8:

Revision 0 Routine and Unanticipated Releases 8-5 As part of evaluating potential offsite releases, an airborne source term based on known loose surface contamination isotopic breakdown was derived. This derivation used decay-corrected 1993 10 CFR Part 61 waste stream data for Dry Active Waste (DAW) found in Calculation 6612-93-021, TMI-2 Waste Stream Update (Ref. 1.3.28), and removable contamination smears on a drone used to survey the Reactor Building in June of 2021, as well as recent smears associated with the Reactor Building elevator and the polar crane.

In addition, events that could exceed the threshold of a Notification of an Unusual Event (NOUE) were selected for review based on TMI2-RA-COR-2023-0002, from Hazelhoff, A. C., Supplement to License Amendment Request - Proposed Changes to TMI-2 Possession Only License and Technical Specifications, dated January 27, 2023 (Ref. 1.3.45). As described, the events that could exceed the threshold of an NOUE were determined to be:

• A fire involving DAW,

• A failure of a HEPA Filter,

• A drop of a spent zeolite liner,

• A rupture of a Processed Water Storage Tank (PWST), and

• An oxyacetylene explosion.

8.2.2 DOSE RATE LIMITS FOR FIRES AND HEPA FILTER FAILURE A fire involving DAW and a failure of a HEPA Filter was addressed in 164090-EN-CALC-004, Source Term Limitations and Administrative Controls for the TMI-2 Decommissioning Emergency Plan Action Levels (Ref. 1.3.27).

Based on this calculation, administrative limits will be placed on the activity content of combustibles available for a fire to limit the fire severity and to prevent having a fire that will cause a release corresponding to the dose associated with two times the ODCM limit at a location corresponding to the highest atmospheric dispersion factor listed in the ODCM or above. This will ensure that the release levels will remain below the NOUE limit for the site. Procedures will be developed in accordance with TMI2-RP1-PG-001, Radiological Protection Program (Ref. 1.3.46), to perform monitoring and to implement the source term administrative limits to ensure potential events do not exceed site boundary dose limits.

TMI-2 Defueled Safety Analysis Report Chapter 8:

Revision 0 Routine and Unanticipated Releases 8-6 TABLE 8.2-1, CONTAINER DOSE RATES FOR EACH RELEASE SCENARIO AND VARIOUS DISTANCES Container Release Height HEPA Cs-137 Activity Limit (Ci)

Dose Rate, mR hr-1 1.27 cm 30 cm 100 cm B-25 Elevated N

1.04E+00 5.32E+02 2.91E+02 1.00E+02 20 yd roll-off Elevated N

1.04E+00 1.03E+02 6.67E+01 3.34E+01 30 yd roll-off Elevated N

1.04E+00 8.23E+01 5.96E+01 3.21E+01 40 yd roll-off Elevated N

1.04E+00 6.73E+01 5.27E+01 3.05E+01 B-25 Elevated Y

1.04E+02 5.32E+04 2.91E+04 1.00E+04 20 yd roll-off Elevated Y

1.04E+02 1.03E+04 6.67E+03 3.34E+03 30 yd roll-off Elevated Y

1.04E+02 8.23E+03 5.96E+03 3.21E+03 40 yd roll-off Elevated Y

1.04E+02 6.73E+03 5.27E+03 3.05E+03 B-25 Ground N

5.26E-01 2.68E+02 1.47E+02 5.06E+01 20 yd roll-off Ground N

5.26E-01 5.19E+01 3.37E+01 1.68E+01 30 yd roll-off Ground N

5.26E-01 4.15E+01 3.01E+01 1.62E+01 40 yd roll-off Ground N

5.26E-01 3.39E+01 2.66E+01 1.54E+01 B-25 Ground Y

5.26E+01 2.68E+04 1.47E+04 5.06E+03 20 yd roll-off Ground Y

5.26E+01 5.19E+03 3.37E+03 1.68E+03 30 yd roll-off Ground Y

5.26E+01 4.15E+03 3.01E+03 1.62E+03 40 yd roll-off Ground Y

5.26E+01 3.39E+03 2.66E+03 1.54E+03 In addition, 164090-EN-CALC-004, Source Term Limitations and Administrative Controls for the TMI-2 Decommissioning Emergency Plan Action Levels (Ref. 1.3.27), established cut-off dose rate limits for building installed HEPA Filters such that in the event of a 100% HEPA Filter failure and release of entrained contamination due to a fire, the criteria for an NOUE would not be exceeded.

These limits are:

TABLE 8.2-2, BUILDING HEPA FILTER DOSE RATE LIMITS Distance from External Surface of HEPA Filter Array (cm)

Installed HEPA Filters Dose Rate Limit (mR hr-1) 1.27 104 30 42 100 14 Procedures prepared in accordance with TMI2-RP1-PG-001, Radiological Protection Program (Ref. 1.3.46), will be developed to perform this monitoring and to implement the administrative limits described above.

TMI-2 Defueled Safety Analysis Report Chapter 8:

Revision 0 Routine and Unanticipated Releases 8-7 8.2.3 WASTE HANDLING EVENTS A zeolite liner drop and a Processed Water Storage Tank (PWST) Rupture were addressed in 164090-EN-CALC-003, TMI-2 Source Term Limitations and Administrative Controls to Prevent Exceeding the Emergency Action Levels (EALs) for Zeolite Liner Drop and Processed Water Storage Tank (PWST) Rupture (Ref. 1.3.26). This calculation evaluated potential liquid storage and processing accidents during active decommissioning.

Due to the high concentration of radionuclides on an expended zeolite liner, an accident involving spent zeolite demineralizer media was determined to have the highest potential for airborne dose consequences off-site. To calculate potential airborne off-site dose, the average activity on January 1, 2021, was divided by the zeolite mass to calculate the concentrations of activity on the zeolite. Based on this calculation, the activities required to exceed 1 mrem TEDE or the NOUE levels at the EAB are greater than the dispersed Cesium-137 and Strontium-90 activities remaining onsite as decayed from Table 2.4 in NUREG-0683, Supplement 3, Programmatic Environmental Impact Statement Related to Decontamination and Disposal of Radioactive Wastes Resulting from March 28, 1979 Accident, Three Mile Island Nuclear Station, Unit 2 (Ref. 1.3.17).

A PWST rupture accident was previously analyzed in NUREG-0683, Final Programmatic Environmental Impact Statement Related to Decontamination and Disposal of Radioactive Wastes Resulting from March 28, 1979, Accident, Three Mile Island Nuclear Station, Unit 2, Vol. 1, Section 10.4.2.1, Failure of Processed Water Storage Tank (Ref. 1.3.16). NUREG-0683 states that during water processing operations, processed water will be temporarily stored in two holding tanks located outdoors, each of 500,000-gallon capacity. If one of these tanks ruptured and 50% of its entire contents were released, storm drains would transport the water to the east channel of the river. A criterion for storage of processed accident water in these tanks is that the content of radioactivity stored in each outside processed water storage tank should be limited such that a tank failure would not result in greater than 10 CFR 20 (Table II, Col. 2) concentrations (Ref. 1.3.1) at the nearest drinking water intake for combined radionuclides as a function of actual tank volume. The effluent concentration limit per nuclide is listed below. The concentration limit for the PWST is based on a sum of the fractions for each radionuclide.

TMI-2 Defueled Safety Analysis Report Chapter 8:

Revision 0 Routine and Unanticipated Releases 8-8 TABLE 8.2-3, 1 EC AT NEAREST DRINKING WATER PWST ACTIVITY LIMITS Nuclide Ci EC (Ci/ml)

Tank Limit (Ci)

H-3 1.00E-03 6.52E+09 C-14 3.00E-05 1.96E+08 Mn-54 3.00E-05 1.96E+08 Fe-55 1.00E-04 6.52E+08 Ni-63 1.00E-04 6.52E+08 Co-60 3.00E-06 1.96E+07 Sr-90 5.00E-07 3.26E+06 Tc-99 1.00E-05 6.52E+07 Ru-106 3.00E-06 1.96E+07 Sb-125 3.00E-05 1.96E+08 I-129 2.00E-07 6.89E-10 Cs-134 9.00E-06 5.87E+07 Cs-137 1.00E-06 6.52E+06 Ce-144 3.00E-06 1.96E+07 Pu-238 2.00E-08 1.30E+05 Pu-241 1.00E-06 6.52E+06 Am-241 2.00E-08 1.30E+05 8.2.4 OXYACETYLENE EXPLOSION With respect to an oxyacetylene explosion, there are no current plans to use oxyacetylene in the vicinity of high source term components in the Reactor Building (i.e., the Reactor Vessel Internals and the Reactor Coolant System and associated components). For these high source term components, TMI-2 Solutions plans to use mechanical means for size reduction. Additionally, any use of oxyacetylene in the vicinity of high source term components in the Reactor Building would be evaluated and controlled by the TMI-2 Engineering Program and be subject to a review in accordance with 10 CFR 50.59 (Ref. 1.3.3) to confirm prior NRC approval is not required.

8.2.5 OTHER EVENTS In addition to the above, NRC Letter, Request for Additional Information for Requested Licensing Action Regarding Decommissioning Technical Specifications (Ref. 1.3.62), requested information on three other potential events: Buildup of Radiolytic Gases, Dust Explosion and Exothermic Reaction Hazard, and a potential fire in the cork seam. These events were addressed in TMI2-RA-COR-2022-0019, License Amendment Request - Three Mile Island Unit 2 Decommissioning Technical Specifications, Response to Request for Additional Information (Ref. 1.3.44).

TMI-2 Defueled Safety Analysis Report Chapter 8:

Revision 0 Routine and Unanticipated Releases 8-9 A. Buildup of Radiolytic Gases With respect to buildup of radiolytic gases, TMI2-RA-COR-2022-0019, License Amendment Request - Three Mile Island Unit 2 Decommissioning Technical Specifications, Response to Request for Additional Information (Ref. 1.3.44), states:

In preparation for entry into PDMS, the plant systems were vented, drained, and the remaining water volumes were processed for disposal. As a result, there are no significant water volumes remaining in TMI-2.

However, TMI-2 Solutions recognizes there may be small, localized hydrogen gas pockets remaining within the highly contaminated portions of plant systems and components that could lead to hydrogen production.

TMI-2 Solutions will establish a work planning instruction which will evaluate specific hydrogen concerns relevant to a given scope of work and include hydrogen mitigation measures appropriate for that work.

B. Dust Explosion and Exothermic Reaction Hazard With respect to dust explosion and exothermic reaction hazard, TMI2-RA-COR-2022-0019, License Amendment Request - Three Mile Island Unit 2 Decommissioning Technical Specifications, Response to Request for Additional Information (Ref. 1.3.44), concluded this was not a hazard for decommissioning based on the following:

1. If fines were released from the reactor vessel to the Reactor Building Basement via pressurizer relief valves, they were exposed to oxygen in the water for several years after the accident and then to the atmosphere after water was removed and are thoroughly oxidized.

2. If any unoxidized fines exist, they would be mixed with river water sediment, concrete dust, and dirt which would act as a diluent and would minimize any potential for ignition and propagation.

3. Pyrophoricity of TMI-2 sediment was not a safety concern during cleanup operations and a further 30 years of oxidation has occurred.

C. Fire in the Cork Seam With respect to a potential fire in the cork seam, TMI2-RA-COR-2022-0019, License Amendment Request - Three Mile Island Unit 2 Decommissioning Technical Specifications, Response to Request for Additional Information (Ref. 1.3.44), concluded that the cork seam construction joint consists of combustible materials (polysulfide sealant, polyurethane foam, and the cork itself), but the following measures minimize the possibility of an adverse fire involving the cork seam:

1. The work process controls are in place in accordance with NRC Regulatory Guide 1.191, Fire Protection Program for Nuclear Power Plants During Decommissioning and Permanent Shutdown (Ref. 1.3.11),

2. The cork seam itself is partially saturated with water,

TMI-2 Defueled Safety Analysis Report Chapter 8:

Revision 0 Routine and Unanticipated Releases 8-10

3. The cork seam is only present in the basement elevation, which has concrete walls and ceilings as barriers,

4. The majority of the cork seam location is isolated from the major decommissioning impacted areas (e.g., inside Locked High Radiation Area cubicles in the Auxiliary Building, etc.),

5. A portion of the cork seam is in an area with ventilation that exhausts through HEPA filtration,

6. Of the total volume of cork seam, only a narrow ~1 inch width is exposed and available to a potential fire, and

7. There are presently no major combustibles within the areas containing the cork seam.

TMI-2 Defueled Safety Analysis Report Revision 0 TECHNICAL SPECIFICATIONS DELETED

TMI-2 Defueled Safety Analysis Report Revision 0 ADMINISTRATIVE FUNCTIONS

TMI-2 Defueled Safety Analysis Report Chapter 10:

Revision 0 Administrative Functions 10-2

10.1 INTRODUCTION

The primary administrative functions for the management of TMI-2 during Decommissioning are referenced in this chapter.

10.2 QUALITY ASSURANCE PLAN TMI2-QA-PG-001, Three Mile Island Unit 2 Decommissioning Quality Assurance Plan (DQAP) (Ref. 1.3.40), has been developed to provide TMI-2 with a limited scope QA Program that is structured to correspond to the 18-criteria format of 10 CFR 50, Domestic Licensing of Production and Utilization Facilities, Appendix B (Ref. 1.3.3). It is applied as necessary to assure decommissioning activities subject to the plan are in compliance with applicable regulatory requirements, assure compliance with the conditions of the NRC License, and assure conformance to the descriptions in this DSAR.

The DQAP (Ref. 1.3.40) is implemented in a graded and customized approach which is based on the Importance to Safety and safety significance of Systems, Structures and Components (SSCs) and activities, as well as on an evaluation of regulations, risks, complexity, and history of previous implementation. The application of a graded approach only allows grading of rigor in implementing these requirements and does not relieve TMI-2 Solutions of its responsibility to maintain compliance with associated regulatory codes and standards.

The concept of what is Important to Safety in 10 CFR 71, Packaging and Transportation of Radioactive Material (Ref. 1.3.5), differs from the concept in 10 CFR 50, Domestic Licensing of Production and Utilization Facilities, Appendix B (Ref. 1.3.3). When applying the criteria of 10 CFR 50, Domestic Licensing of Production and Utilization Facilities, Appendix B (Ref. 1.3.3), at TMI-2, the unique shutdown and defueled status of the facility is considered.

These considerations have little applicability to the activities governed by the guidelines of 10 CFR 71, Packaging and Transportation of Radioactive Material (Ref. 1.3.5). The criteria, as specified in Appendix A of the DQAP (Ref. 1.3.40), considers the guidance in NUREG/CR-6407, Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety (Ref. 1.3.19), and NRC Regulatory Guide 7.10, Establishing Quality Assurance Programs for Packaging Used in Transport of Radioactive Material (Ref. 1.3.13), for applying a graded approach based on the SSCs Importance to Safety.

TMI-2 Defueled Safety Analysis Report Chapter 10:

Revision 0 Administrative Functions 10-3 10.3 SECURITY PLAN The TMI-2 specific security measures are detailed in TMI2-SE-PN-001, TMI-2 Materials Security Plan (SUNSI) (Ref. 1.3.47), and include provisions for 10 CFR 37, Physical Protection of Category 1 and Category 2 Quantities of Radioactive Material (Ref. 1.3.2);

10 CFR 73.67, Licensee fixed site and in-transit requirements for the physical protection of special nuclear material of moderate and low strategic significance (Ref. 1.3.7); and considerations for 10 CFR 20, Standards for Protection Against Radiation (Ref. 1.3.1), and rely on some controls from the TMI-1 NRC Approved ISFSI Physical Security Plan (PSP).

In accordance with these plans, security zones contain the radiological materials of concern. These security zones can only be accessed through controlled access points. The Restricted Area is also controlled. The plan relies on the detection of Sabotage, Diversion, and Theft of the radiological materials of concern and a Local Law Enforcement Agency (LLEA) response if Sabotage, Diversion, or Theft is detected. Once TMI-2 begins to retrieve, aggregate, package, and load storage canisters for placement at the ISFSI, The loaded canisters will also be subject to the requirements of 10 CFR 37, Physical Protection of Category 1 and Category 2 Quantities of Radioactive Material (Ref. 1.3.2) and 10 CFR 73.67, Licensee fixed site and in-transit requirements for the physical protection of special nuclear material of moderate and low strategic significance (Ref. 1.3.7) as they are transported on site to the TMI-1 ISFSI. Current planning is to pour a new pad inside the TMI-1 ISFSI fence to store the canisters.

10.4 EMERGENCY PLAN A radiological emergency response plan (found in EP-TM-1002, Three Mile Island Independent Spent Fuel Storage Installation [ISFSI] Only Emergency Plan [IOEP]

[Ref. 1.3.31]) is maintained for TMI that assures the safety of the public and site personnel from radiological incidents that may occur during TMI-2 Phase 1 activities. The Three Mile Island Nuclear Station (TMINS) IOEP encompasses incidents at TMI-1 and TMI-2, ensuring a coordinated response to any incident onsite. The IOEP implements the planning standards contained in 10 CFR 50.47(b) (Ref. 1.3.3), as amended by the NRC and includes the plan contents required by 10 CFR 50, Domestic Licensing of Production and Utilization Facilities, Appendix E (Ref. 1.3.3).

The IOEP is based upon accident analyses that conclude that the consequences of any radiological incidents that may occur at the station are limited to significantly less than the guidelines found in EPA-400/R-17/001, PAG Manual: Protective Action Guides and Planning Guidance for Radiological Incidents (Ref. 1.3.23). The IOEP describes all commitments and resources needed to assess, classify, and respond to any incident at TMI, including those required for notification of local and state governments, mobilization of the site emergency response organization and corporate resources, and assessment and analysis of radiological conditions and mitigation of any adverse conditions. The IOEP also contains description of the administrative programs in place to maintain readiness to implement the plan, including the identification and training of required resources and the maintenance of the plans, procedures, facilities, equipment, and supplies needed for response.

TMI-2 Defueled Safety Analysis Report Chapter 10:

Revision 0 Administrative Functions 10-4 10.5 RADIATION PROTECTION PROGRAM TMI-2 maintains a Radiation Protection Program (RPP) (see TMI2-RP1-PG-001, Radiological Protection Program [Ref. 1.3.46]) as described in Section 5.4.2 of the DSAR.

10.6 ORGANIZATION As specified in TMI2-QA-PN-001, TMI-2 QA Program Implementation, Section 4.1.2 (Ref. 1.3.42):

An official project Organization Chart signed by the Project Director is maintained to define lines of authority for the highest management levels as described in the PDMS QAP (Ref. 1.3.41) through intermediate levels to and including all organizational positions responsible for decommissioning activities.

As a minimum, this Organization Chart shall be updated at times when there are changes made that affect positions responsible for implementation of the PDMS QAP (Ref. 1.3.41).

The latest approved revision to the Organization Chart shall be accessible to the project organization.

The official project Organization Chart is contained in TMI2-PM-PN-001, TMI-2 Project Management Plan (Ref. 1.3.39), and describes the basic functions to be performed by the organization. It is the responsibility of the Project Director to ensure that functions described in this procedure are assigned to the appropriate organizational element of the Organization Chart.

47 Pages Follow Decommissioning Quality Assurance Plan, Revision 20

Title

Signature
and
Date

Authored
By:

TMI-2
Solutions
Quality
Assurance

Manager

Reviewed
By:

TMI-2
Solutions
Licensing
Manager

Approved
By:

TMI-2
Solutions
Project
Director

Approved
By:

EnergySolutions
Group
Director
QA,

D&D

Approved
By:

TMI-2
Solutions
Cognizant
Officer

Approved
By:

Technical
Reviewer/PORC

Effective
Date:
At
start
of
DECON

Three
Mile
Island
Unit
2

Decommissioning

Quality
Assurance
Plan
(DQAP)

TMI2-QA-PG-001,
Revision
20

DigitallysignedbyMichaelJanus DN:OU=QualityAssurance,O=QualityAssurance,

CN=MichaelJanus,E=mtjanus@energysolutions.com Reason:Iamtheauthorofthisdocument Location:yoursigninglocationhere Date:2023-04-1815:39:19 FoxitPhantomPDFVersion:9.7.0 MichaelJanus DigitallysignedbyTimothyDevik DN:C=US,OU=LicensingManager,O=TMI-2Solutions,

CN=TimothyDevik,E=trdevik@energysolutions.com Reason:Reviewed Location:

Date:2023-04-1815:55:30 FoxitPhantomPDFVersion:9.4.1 TimothyDevik David Del Vecchio Digitally signed by David Del Vecchio Date: 2023.04.18 16:32:49 -04'00' David Del Vecchio Digitally signed by David Del Vecchio Date: 2023.04.18 16:33:19 -04'00' Anthony R. Bejma Digitally signed by Anthony R.

Bejma Date: 2023.04.18 16:58:54 -04'00' Frank Eppler Digitally signed by Frank Eppler Date: 2023.04.18 17:02:48 -04'00' 5/2/2023

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
2
of
47

Record
of
Changes

Revision
19:
Initial
revision
of
PDMS
QAP
for
TMI-2
Solutions
following
License
Transfer
from
the

FirstEnergy
Companies
(GPU
Nuclear,
Inc.,
Metropolitan
Edison
Company,
Jersey
Central
Power
&
Light

Company,
and
Pennsylvania
Electric
Company).
Changes
were
generally
limited
to
those
necessary
to
depict

TMI-2
Solutions
as
the
licensee
and
the
associated
organizational
titles.
Section
2.0,
Quality
Policy,
was

revised
to
include
the
approach
to
revising
this
plan
as
stated
in
the
License
Transfer
Application.
Section
3.0,

Scope,
was
revised
to
recognize
the
continuity
with
Exelon
by
assuming
the
previous
services
agreement.

Section
5.1
was
revised
to
delete
reference
to
the
PDMS
SAR
for
a
description
of
the
organization
to
align
with

a
parallel
change
to
the
PDMS
SAR.
Section
5.7,
Control
of
Purchased
Material,
Equipment,
and
Services,
was

revised
to
add
EnergySolutions
as
an
addition
to
other
clients.
Sections
6.0
and
7.0
were
revised
to
recognize

that
TMI-2
Solutions
may
prepare,
review,
and
approve
documents
required
by
the
activities
described
in

Section
6.0
and
perform
Independent
Safety
Reviews
required
in
Section
7.0
in
addition
to
the
TMI-1
License

holder.
All
other
changes
were
administrative
in
nature
and
had
no
impact
on
technical
content
(e.g.,
deletion

of
historical
Record
of
Changes
and
List
of
Effective
Pages,
spelling,
punctuation,
format,
and
consistency
in

how
this
plan
is
referenced).

Revision
20:
Revision
prepared
primarily
to
coincide
with
the
transition
from
PDMS
to
DECON;
incorporate

requirements
previously
contained
in
the
Technical
Specifications
as
specified
in
License
Amendment
Request

-
Three
Mile
Island,
Unit
2,
Technical
Specifications
(License
Amendment
67);
remove
details
relative
to

Exelons
contract
scope
to
maintain
Unit
2
in
PDMS/SAFSTOR;
and
incorporate
additional
requirements
in

Appendix
A
relative
to
implementation
of
10
CFR
71,
Subpart
H.
Details
of
the
revision
include:
Table
of

Contents
-
Reconciled
to
align
with
content
(Sectional/Appendix
titles
and
page
references);
General
-

Replaced
PDMS/SAFSTOR
and
PDMS
with
decommissioning
as
applicable,
minor
non-intent
administrative

enhancements;
Section
1.0
-
Deleted
bases
for
there
not
being
any
SSCs
that
perform
a
safety
related
function

to
eliminate
duplication
addressed
in
other
license
bases
documents
and
corresponding
changes,
and
revised

applicability
of
the
plan
from
as
necessary
to
assure
the
continued
safe

and
stable
PDMS/SAFSTOR

conditions,
assure
compliance
with
applicable
regulatory
requirements
to
as
necessary
to
assure

decommissioning
activities
subject
to
this
plan
are
in
compliance
with
applicable
regulatory
requirements;

Section
2.0
-
Replaced
organizational
responsibility
for
issuance
and
approval
of
this
plan
to
align
with
current

the
EnergySolutions
organization
and
responsibilities
of
the
Cognizant
Officer
as
described
within
this
plan;

Section
3.0
-
Replaced
historical
information
relative
to
Exelons
contract
scope
to
maintain
Unit
2
in

PDMS/SAFSTOR
with
recognition
that
TMI-2
Solutions
may
elect
to
subcontract
decommissioning
activities

subject
to
this
plan,
added
summary
of
requirements
relocated
from
the
Technical
Specifications,
added

definitions
for

SUBSTANTIVE
CHANGES,
SITE
BOUNDARY
and
NPDES
PERMIT
that
were
relocated

from
the
Technical
Specifications,
replaced
Technical
Specification
references
with
the
DSAR
in
association

with
the
SSCs
that
are
Within
QA
Plan
Scope,
revised
the
description
for
Lifting
and
Handling
activities
to

align
with
License
Amendment,
enhanced
description
of
10
CFR
71
activities
Within
QA
Plan
Scope,

deleted
historical
rationale
for
separating
the
10
CFR
71
requirements
into
Appendix
A,
added
NUREG/CR-6407
to
the
dialogue,
revised
the
revision
level
of
RG
7.10
that
coincides
with
the
rule
change,
and
deleted
the

quote
from
RG
7.10,
and
added
reference
to
Section
5.16
for
controls
of
Conditions
Adverse
to
Quality;

Section
4.0
-
Added
facility
staff
qualification
requirements
that
were
relocated
from
the
Technical

Specifications;
Section
5.0:
Subsection
5.1
-
Deleted
reference
to
the
requirements
specified
in
the

administrative
section
of
the
Technical
Specifications,
expanded
to
include
purpose
of
providing
authority
and

organizational
freedom
for
the
QA
function,
replaced
historical
information
relative
to
Exelons
contract
scope

to
maintain
Unit
2
in
PDMS/SAFSTOR
in
5.1.2
with
recognition
that
TMI-2
may
elect
to
subcontract

decommissioning
services,
changed
the
QAM
description
and
reporting
relationship
to
the
Cognizant
Officer

and
described
the
direct
reporting
of
an
onsite
project
QA
organization
responsible
for
the
day
to
day
oversight

of
plan
implementation
in
5.1.3,
edited
the
EngerySolutions,
Regulatory
Affairs
title
in
5.1.4
deleted
in
the

future
and
added
former
Technical
Specification
reference
in
5.1.5,
established
reporting
relationship
to
the

Cognizant
Officer
and
modified
responsibilities
to
incorporate
requirements
relocated
from
the
Technical

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
3
of
47

Record
of
Changes

Specifications
in
5.1.6,
deleted
PDMS
Manager
position
and
responsibilities
in
5.1.7,
and
added
requirement

for
Radioactive
Waste
Management
activities
that
was
relocated
from
the
Technical
Specifications;
Subsection

5.2
-
Revised
of
critical
activities
to
be
more
relevant
to
decommissioning
activities;
Subsection
5.3
-
Replaced

project
personnel
with
on-site
workers
for
consistency,
deleted
information
relative
to
Exelons
contract
scope

and
processes,
and
deleted
reference
to
the
Technical
Specifications
in
the
description
of
SSCs
required
to

maintain
the
plant
in
a
safe
and
stable
condition;
Subsection
5.4
-
Deleted
reference
to
the
current
PDMS

Services
Contractor,
and
replaced
release
with
execution
in
reference
to
the
procurement
process;
Subsection

5.5
-
Separated
examples
of
activities
performed
during
decommissioning
to
assure
conformance
to
the
DSAR

that
were
specified
in
the
Technical
Specifications
as
5.5.1/5.5.1.a
from
the
examples
of
additional
procedures

required
for
decommissioning,
and
inserted
requirements
for
the
ODCM,
Radioactive
Effluent
Controls

Program,
Radioactive
Monitoring
Program
and
Radiation
Protection
Program
that
were
relocated
from
the

Technical
Specifications
as
5.5.2.a
through
d;
Subsection
5.6
-
Reformatted
existing
content,
and
added

requirements
related
to
procedure
reviews,
substantive
changes
to
the
ODCM
and
temporary
changes
to

procedures
that
were
relocated
from
the
Technical
Specifications
as
5.6.2.a,
b
and
5.6.3.a
through
c;
Subsection

5.10
-
Enhanced
sentence
regarding
inspection
hold
points
by
inserting
are
to
be
established;
Subsection
5.11
-

Deleted
reference
to
Technical
Specifications
in
reference
to
performance
criteria;
Subsection
5.17
-

Reformatted
existing
content,
and
added
requirements
related
to
specific
records
retention
periods
that
were

relocated
from
the
Technical
Specifications
as
5.17.5.a
and

b;
Subsection
5.18
-
Replaced

audit
requirement

for
performance
of
QA
Assessment
activities
provided
by
the
Decommissioning
Services

contractor,
and

replaced
PDMS
Manager
with
TMI-2
Solutions
Project
Director;
Section
6.0
-
Deleted
reference
to
the
TMI
1

License
holder
and
TMI
Review
and
Approval
Matrix,
replaced
references
to
Section
6.7
of
the
Technical

Specifications
with
DQAP
references
in
6.1
based
on
the
relocation
of
the
requirements
from
the
Technical

Specifications,
and
replaced
requirement
to
maintain
the
PDMS
condition
in
the
PDMS
SAR
in
6.4
with

requirement
to
assure
conformance
to
the
descriptions
in
the
Decommissioning
SAR,;
Section
7.0
-
Deleted

reference
to
the
TMI
1
License
holder
and
TMI
Review
and
Approval
Matrix,
and
deleted
7.6
related
to

independent
reviews
performed
for
Exelon
generated
documents;
Appendix
A:
Retitled
Section
A1.0
to

General
to
distinguish
content
from
Section
A4.0;
refined
and
added
specific
requirements
throughout

Appendix
in
response
to
10
CFR
71,
Subpart
H
Quality
Assurance
requirements;
distinguished
applicability
of

the
requirements
for
each
of

the
Sections
as
they
relate
to
Important
to
Safety
activities
performed
by
TMI-2

Solutions,
its
contractors
and
subcontractors,
or
the
certificate
holders,
with
conditional
requirements
for
use,
as

applicable,
throughout
Appendix;
Section
A1.0
-
Deleted
reference
to
transport
performed
by
Exelon
under
the

Services
Agreement
and
historical
information
regarding
reactor
fuel;
Section
A3.0
-
Deleted
training
and

indoctrination
requirement
and
relocated
it
to
Section
A4.3;
Section
A4.0
-
Deleted
reference
to
Section
1.0

statement
in
A4.1
specific
to
there
being
no
SSCs
performing
a
safety
related
function,
and
revised
the
revision

level
of
RG
7.10
that
coincides
with
the
rule
change;
Sections
A18.0
and
A19.0
-
Added
evaluation
and

reporting
requirements
in
reference
to
10
CFR
21
and
10
CFR
71.95;
Appendix
B
-
Established
appendix
to

incorporate
reporting
requirements
relocated
from
the
Technical
Specifications

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
4
of
47

TABLE
OF
CONTENTS

Section

Title

Page

1.0

Introduction
and
Background

6

2.0

Quality
Policy

6

3.0

Scope

7

4.0

Training

10

5.0

Quality
Assurance
Criteria

11

5.1

Criterion
I,
Organization

11

5.2

Criterion
II,
Quality
Assurance
Program
Description

13

5.3

Criterion
III,
Design
Control

14

5.4

Criterion
IV,
Procurement
Document
Control

15

5.5

Criterion
V,
Instructions,
Procedures,
Drawings

16

5.6

Criterion
VI,
Document
Control

19

5.7

Criterion
VII,
Control
of
Purchased
Material,
Equipment,
and
Services

21

5.8

Criterion
VIII,
Identification
and
Control
of
Materials,
Parts,
and
Components

21

5.9

Criterion
IX,
Control
of
Special
Processes

21

5.10

Criterion
X,
Inspection

22

5.11

Criterion
XI,
Test
Control

22

5.12

Criterion
XII,
Control
of
Measuring
and
Test
Equipment

22

5.13

Criterion
XIII,
Handling,
Storage,
and
Shipping

23

5.14

Criterion
XIV,
Inspection
Test,
and
Operating
Status

23

5.15

Criterion
XV,
Nonconforming
Materials,
Parts,
or
Components

23

5.16

Criterion
XVI,
Corrective
Action

24

5.17

Criterion
XVII,
Quality
Assurance
Records

24

5.18

Criterion
XVIII,
Audits

26

6.0

Independent
Technical
Review

27

7.0

Independent
Safety
Review

28

Appendix
A

10
CFR
71
Subpart
H
Quality
Assurance
Plan

30

A1.0

General

30

A2.0

Responsibilities

30

A3.0

Organization
(10
CFR
71.103)

30

A4.0

Quality
Assurance
Program
(10
CFR
71.105)

31

A5.0

Changes
to
Quality
Assurance
Program
(10
CFR
71.106)

31

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
5
of
47

TABLE
OF
CONTENTS

A6.0

Design
Control
(10
CFR
71.107)

32

A7.0

Procurement
Document
Control
(10
CFR
71.109)

32

A8.0

Instructions,
Procedures,
and
Drawings
(10
CFR
71.111)

33

A9.0

Document
Control
(10
CFR
71.113)

34

A10.0

Control
Of
Purchased
Material,
Equipment,
and
Services
(10
CFR
71.115)

34

A11.0

Identification
and
Control
of
Materials,
Parts,
and
Components
(10
CFR

71.117)

35

A12.0

Control
of
Special
Processes
(10
CFR
71.119)

36

A13.0

Inspection
(10
CFR
71.121)

36

A14.0

Test
Control
(10
CFR
71.123)

38

A15.0

Control
of
Measuring
and
Test
Equipment
(10
CFR
71.125)

39

A16.0

Handling,
Storage,
and
Shipping
Control
(10
CFR
71.127)

40

A17.0

Inspection,
Test
and
Operating
Status
(10
CFR
71.129)

41

A18.0

Control
of
Nonconforming
Materials,
Parts
or
Components
(10
CFR
71.131)

41

A19.0

Corrective
Action
(10
CFR
71.133)

42

A20.0

Quality
Assurance
Records
(10
CFR
71.135)

42

A21.0

Audits
(10
CFR
71.137)

44

Appendix
B

Reporting
Requirements

45

B.1

Routine
Reports

45

B.2

Special
Reports

45

B.3

Nonroutine
Reports

46

B.4

Exceptional
Occurrences

47

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
6
of
47

1.0

INTRODUCTION
AND
BACKGROUND

Upon
completion
of
the
Three
Mile
Island
Unit
2
Cleanup
Program,
the
facility
was
placed
in
a

safe
stable
condition
posing
no
significant
risk
to
the
health
and
safety
of
the
public.
This

condition
was
called
Post
Defueling
Monitored
Storage
or
PDMS.
It
is
comparable
to
the

condition
described
by
the
Nuclear
Regulatory
Commission
as
"SAFSTOR".
All
reactor
and

power
generating
systems
were
deactivated
and
the
reactor
vessel
was
defueled
to
the
extent

practicable.
There
is
no
fuel
in
the
spent
fuel
pools,
the
pools
are
drained,
there
is
no
fuel
stored

on
site
in
dry
storage,
and
there
is
no
need
or
requirement
for
NRC
Licensed
Operators
or
NRC

Licensed
Fuel
Handlers
at
Three
Mile
Island
Unit
2.
There
are
no
SSCs
that
perform
a
safety

related
function,
as
defined
in
10
CFR
50.2,
where
the
SSC
must
perform
an
active
function
to

prevent
or
mitigate
the
consequences
of
postulated
accidents
that
could
cause
undue
risk
to
the

health
and
safety
of
the
public.

The
facility
remains
licensed
under
10
CFR
50,
however,
the
license
is
"to
possess
but
not
to

operate"
the
facility.
Therefore,
the
regulations
and
quality
criteria
that
are
designed
to
protect

the
public
against
reactor
operation
related
design
basis
events
are
no
longer
applicable
at
this

permanently
shut
down
and
defueled
facility.
The
licensing
and
regulatory
basis
of
the
Three

Mile
Island
Unit
2
facility
at
this
time
is
more
representative
of
a
materials
licensee
storing
and

processing
radioactive
waste
than
a
commercial
nuclear
power
reactor
authorized
to
operate.

This
plan
has
been
developed
to
provide
Three
Mile
Island
Unit
2
with
a
limited
scope
Quality

Assurance
(QA)
Program
using
the
criteria
of
10
CFR
50
Appendix
B
as
guidelines
and
is

structured
to
correspond
to
the
18
criteria
format
of
10
CFR
50,
Appendix
B.
It
is
applied
as

necessary
to
assure
decommissioning
activities
subject
to
this
plan
are
in
compliance
with

applicable
regulatory
requirements;
assure
compliance
with
the
conditions
of
the
NRC
License;

and
assure
conformance
to
the
descriptions
in
the
Defueled
Safety
Analysis
Report.

2.0

QUALITY
POLICY

TMI-2
Solutions
policy
is
to
conduct
decommissioning
activities
in
a
manner
that
ensures

protection
of
the
health
and
safety
of
the
public
and
the
personnel
on
site.
TMI-2
Solutions
will

adhere
to
the
applicable
Quality
Assurance
requirements
of
10
CFR
50
Appendix
B,
to

appropriate
Regulatory
Guides,
and
to
appropriate
industrial
codes
and
standards,
that
are

applicable
to
the
shutdown,
defueled,
and
depressurized
status
of
the
plant
systems
and

components.

This
plan
has
been
developed
to
provide
direction
on
those
aspects
of
10
CFR
50
Appendix
B

that
will
be
applied
to
decommissioning
activities
and
is
issued
under
the
authority
of
the

EnergySolutions
Chief
Nuclear
Officer
and
Group
Director
QA,
D&D.
Revisions
to
this
plan

will
be
implemented
in
accordance
with
10
CFR
50.54
(a)
as
stated
in
the
License
Transfer

Application
in
reference
to
previous
commitments
established
by
GPU
Nuclear
and
accepted
by

the
NRC.
Revisions
will
be
written
as
needed,
subjected
to
independent
Technical
Review,
and

shall
be
approved
by
the
EnergySolutions
Group
Director
QA,
D&D,
and
the
TMI-2
Solutions

Cognizant
Officer.
Revisions
to
this
plan
will
be
submitted
to
the
NRC
for
approval
in

accordance
with
10
CFR
50.54
(a)(4)
prior
to
the
implementation
of
the
changes
if
they
involve
a

reduction
in
commitments
previously
accepted
by
the
NRC.
Changes
will
be
submitted
to
the

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
7
of
47

NRC
in
accordance
with
10
CFR
50.54
(a)(3)
after
implementation
if
the
changes
do
not
involve

a
reduction
in
commitment.
Copies
will
be
sent
to
the
NRC
Region
1
TMI-2
Inspector
and
the

TMI-2
Project
Manager
at
NRC
Headquarters.
Revisions
requiring
approval
shall
be
regarded
as

approved
by
the
NRC
upon
receipt
of
an
NRC
correspondence
to
this
effect
from
the
appropriate

reviewing
office,
or
60
days
after
submittal
to
the
NRC,
whichever
comes
first.

Implementation
of
this
plan
involves
the
participation
of
all
personnel
performing

decommissioning
activities
and
is
implemented
primarily
through
a
policy
of
adherence
to

written
instructions,
procedures,
and
drawings.
Involvement
by
all
personnel
in
a
program
of

reporting
quality
issues
and
safety
concerns
to
the
supervisory
and
management
staff
assures

continuous
review
and
assessment
of
activities,
the
encouragement
of
a
questioning
attitude,
and

the
fostering
of
a
Safety
Conscience
Work
Environment.
These
aspects
will
provide
assurance

that
quality
deficiencies
are
identified
and
corrected.

This
plan
is
implemented
in
a
graded
and
customized
approach
which
is
based
on
the
importance

to
safety
and
safety
significance
of
SSCs
and
activities,
and
on
an
evaluation
of
regulations,

risks,
complexity,
and
history
of
previous
implementation.
The
application
of
a
graded
approach

only
allows
grading
of
rigor
in
implementing
these
requirements
and
does
not
relieve
TMI-2

Solutions
of
its
responsibility
to
maintain
compliance
with
associated
regulatory
codes
and

standards.

3.0

SCOPE

TMI-2
Solutions
may
elect
to
subcontract
decommissioning
activities
subject
to
this
plan.
Even

though
activities
may
be
performed
by
others
under
contractual
agreements,
the
final
decision

making
authority
regarding
direction
and
control
of
TMI-2
decommissioning
activities
and

responsibility
for
the
effective
implementation
of
the
Three
Mile
Island
Unit
2
DQAP
remains

with
the
TMI-2
Solutions
Cognizant
Officer.

Quality
Assurance
requirements
associated
with
review
and
audit
functions,
the
qualification
of

staff
performing
those
functions,
and
the
records
associated
with
those
functions,
originally

contained
in
Section
6.5
of
the
Technical
Specifications
of
the
NRC
License
Number
DPR-73,

were
transferred
verbatim
to
Revision
12
of
the
PDMS
Quality
Assurance
Plan,
as
stated
in

License
Amendment
63.
Additional
requirements,
previously
contained
in
Section
1.0,

Definitions,
and
Section
6.0,
Administrative
Controls
of
the
Technical
Specifications
of
the
NRC

License
Number
DPR-73
were
transferred
to
Revision
20
of
this
plan
in
accordance
with
License

Amendment
67.
These
requirements
were
transferred
to
an
appropriate
section
of
the
main
body

and
Appendix
B
of
this
plan,
with
citations
added
to
refer
to
the
original
section
of
the
Technical

Specifications.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
8
of
47

The
following
standalone
definitions
were
incorporated
here
as
part
of
the
transfer
of
Section

1.0,
Definitions
from
the
Technical
Specifications.
An
additional
definition
for
the
Off-Site
Dose

Calculation
Manual
(ODCM)
was
incorporated
into
section
5.5.2
as
appropriate.:

SUBSTANTIVE
CHANGES
(FormerlyTechnicalSpecification1.15)
-
Substantive
Changes
are

those
which
affect
the
activities
associated
with
a
document
or
the
documents
meaning
or
intent.

Examples
of
non-substantive
changes
are:
(1)
correcting
spelling;
(2)
adding
(but
not
deleting)

sign-off
spaces;
(3)
blocking
in
notes,
cautions,
etc.;
(4)
changes
in
corporate
and
personnel
titles

which
do
not
reassign
responsibilities
and
which
are
not
referenced
in
the
Technical

Specifications;
and
(5)
changes
in
nomenclature
or
editorial
changes
which
clearly
do
not
change

function,
meaning
or
intent.

SITE
BOUNDARY
(FormerlyTechnicalSpecification1.18)
-
The
SITE
BOUNDARY
shall
be

that
line
beyond
which
the
land
is
neither
owned,
nor
leased,
nor
otherwise
controlled
by
TMI-2

Solutions,
LLC.
The
SITE
BOUNDARY
for
gaseous
and
liquid
effluents
shall
be
as
shown
in

the
Offsite
Dose
Calculation
Manual
(ODCM).

NPDES
PERMIT
(FormerlyTechnicalSpecification1.19)
-
The
NPDES
PERMIT
is
the

National
Pollutant
Discharge
Elimination
System
(NPDES)
Permit
No.
PA0009920,
effective

June
1,
2010,
issued
by
the
Pennsylvania
Department
of
Environmental
Protection
and
assigned

to
Constellation
Energy
Generation,

LLC.
This
permit
authorizes
discharge
to
the
Susquehanna

River
with
effluent
limitations,
monitoring
requirements
and
other
conditions
within
the
permit.

Appendix
B
to
10
CFR
50
states
in
part,
"Nuclearpowerplants
includestructures,systems,

andcomponentsthatpreventormitigatetheconsequencesofpostulatedaccidentsthatcould

causeunduerisktothehealthandsafetyofthepublic.Thisappendixestablishesquality

assurancerequirementsforthedesign,manufacture,construction,andoperationofthose

structures,systems,andcomponents.Thepertinentrequirementsofthisappendixapplytoall

activitiesaffectingthesafety-related
functions(emphasisadded)ofthosestructures,systems,

andcomponents."

As
previously
stated,
there
are
no
SSCs
that
perform
"safety-related"
functions.
Therefore,
this

plan
is
of
limited
scope
and
the
criteria
discussed
below
will
be
applied
to
items
and
activities

determined
to
be
"Within
QA
Plan
Scope"
considering
an
activity's
or
item's
potential
impact
on

safety.

Impact
on
safety
(or
"Importance
to
Safety")
in
this
context
is
a
consideration
of
the
potential
for

radiation
exposure
to
or
intake
of
radioactive
material
by
a
member
of
the
off-site
public
or
on-site
worker,
or
a
release
of
radioactive
material
to
the
environment
in
an
unplanned
or

unmonitored
manner.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
9
of
47

Activities
and
items
that
are
designated
"Within
QA
Plan
Scope",
or
"in-scope",
include
the

following:

• 
  SSCs
  required
  for
  the
  control
  or
  monitoring
  of
  radioactive
  releases
  to
  the
  environment,
  as
  

described
in
the
Defueled
Safety
Analysis
Report,
and
activities
and
equipment
that
involve

their
operation,
testing,
maintenance,
surveillance,
modification,
and
analysis
are
Within
QA

Plan
Scope.
The
degree
to
which
the
Quality
criteria
are
applied
is
based
on
the
potential
for

failures
or
deficiencies
involving
such
activities,
systems
or
components
to
contribute
to
a

violation
of
Technical
Specifications,
or
an
unplanned
or
unmonitored
release
of
radioactive

material
to
the
environment
and
unplanned
radiation
exposure
to
a
worker
or
a
member
of

the
public.

• 
  Radiological
  protection
  and
  radiological
  monitoring
  activities
  including
  activities
  related
  to
  

occupational
dose
control;
offsite
exposure
control,
analysis,
and
evaluation;
environmental

radiological
measurement
and
analysis;
control
of
instrumentation
used
in
these
activities;

and
documentation
and
reporting
of
required
radiological
and
environmental
data
are
Within

QA
Plan
Scope.
The
degree
to
which
the
criteria
are
applied
to
equipment
used
in
these

activities
is
based
on
a
graded
approach
considering
the
uniqueness
and
complexity
of
the

equipment,
the
impact
on
safety
of
failures,
and
the
importance
of
the
end
use
of
the
data

collected.

• 
  Chemistry
  and
  radiochemistry
  analyses
  needed
  for
  control
  of
  radioactive
  effluents
  to
  the
  

environment,
evaluation
of
waste
forms,
and
control
of
personnel
exposure
to
potential

radiological
hazards
is
Within
QA
Plan
Scope.
Equipment
used
in
these
analyses
that
could

affect
the
outcome
of
analyses
are
Within
QA
Plan
Scope
where
testing
or
calibration
of
the

equipment
is
required
or
where
the
equipment
needs
certification
of
accuracy
beyond
that

specified
in
the
manufacturers
catalogues
or
standard
literature.

• 
  Management
  of
  radioactive
  waste
  is
  Within
  QA
  Plan
  Scope
  where
  such
  management
  

activities
involve
determination
of
characteristics
and
classification
of
waste
for
transport
or

disposal,
controls
to
prevent
unplanned
or
unauthorized
dispersal,
or
controls
intended
to

assure
proper
packaging
for
transport
or
disposal.

• 
  Lifting
  and
  Handling
  activities
  for
  all
  lifts
  where
  a
  load
  drop
  or
  load
  impingement
  could
  

contribute
to
release
or
dispersal
of
radioactive
material
to
the
environment
which
could

exceed
the
threshold
for
an
Unusual
Event
are
Within
QA
Plan
Scope..

• 
  Assurance
  activities
  including
  audit,
  inspection,
  non-destructive
  examination,
  monitoring,
  

and
surveillance
of
the
areas
listed
above,
are
Within
the
QA
Plan
Scope.

• 
  Preparation,
  independent
  review,
  approval,
  and
  distribution
  of
  instructions,
  procedures,
  and
  

drawings
related
to
"in-scope"
activities
are
activities
that
are
Within
QA
Plan
Scope.

• 
  Training,
  Emergency
  Planning,
  Security,
  and
  Fire
  Protection
  activities
  as
  they
  relate
  to
  

preventing
or
minimizing
uncontrolled
release
of
radioactive
material
to
the
environment
or

unplanned
radiation
exposure
to
workers
or
members
of
the
public
are
activities
Within
QA

Plan
Scope.

• 
  Packaging,
  preparation
  for
  shipment
  and
  transportation
  of
  licensed
  material
  in
  excess
  of
  a
  

Type
A
quantity
are
activities
Within
QA
Plan
Scope.
However,
the
specific
quality

assurance
criteria
of
10
CFR
71,
Subpart
H
that
are
applied
to
these
activities
are
specified
in

Appendix
A
of
this
plan.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
10
of
47

The
concept
of
what
is
"Important
to
Safety"
in
10
CFR
71
differs
from
the
concept
in
10
CFR

50
Appendix
B.
When
applying
the
criteria
of
10
CFR
50
Appendix
B
at
Three
Mile
Island
Unit

2,
the
unique
shutdown
and
defueled
status
of
the
facility
is
considered.
These
considerations

have
little
applicability
to
the
activities
governed
by
10
CFR
71.
The
criteria,
as
specified
in

Appendix
A
of
this
plan,
considers
the
guidance
in
NUREG/CR-6407,
Classification
of

Transportation
Packaging
and
Dry
Spent
Fuel
Storage
System
Components
According
to

Importance
to
Safety
and
NRC
Regulatory
Guide
7.10,
"Establishing
Quality
Assurance

Programs
for
Packaging
used
in
Transport
of
Radioactive
Material",
Revision
3
for
applying
a

graded
approach
based
on
the
SSCs
importance
to
safety.

The
items
discussed
above
are
considered
"Activities
Affecting
Quality"
in
the
context
of
this

plan.
Errors,
omissions,
deviations,
and
non-conformances
of
materials
and
documents

associated
with
these
activities
that
could
have
an
impact
on
radiological
exposures
or
releases

are
considered
"Conditions
Adverse
to
Quality"
in
the
context
of
this
plan
and
shall
be
controlled

in
accordance
with
Section
5.16
of
this
plan.

Achievement
of
quality
is
the
responsibility
of
everyone
involved
in
the
performance
of,

direction
of,
supervision
of,
and
assessment
and
oversight
of
decommissioning
activities.
This

involves
a
multi-level
approach
that
includes
independent
review
at
the
document
preparation

stage,
self-checking
and
independent
verification
at
the
performance
stage,
and
review
and
audit

of
in-scope
activities.

4.0

TRAINING

Personnel
performing
or
verifying
activities
affecting
quality
shall
receive
indoctrination
and

training
as
necessary
to
assure
suitable
proficiency.
Training
programs
shall
be
established
for

those
personnel
such
that
they
are
knowledgeable
in
the
quality
assurance
program
and
proficient

in
implementing
these
requirements.
Training
may
be
conducted
by
on-site
staff
or
third-party

organizations
as
appropriate.
These
training
programs
shall
assure
the
following:

• 
  Personnel
  responsible
  for
  performing
  activities
  affecting
  quality
  understand
  the
  purpose,
  

scope,
and
implementation
of
applicable
procedures,

• 
  Personnel
  performing
  activities
  affecting
  quality
  are
  trained
  and
  qualified,
  as
  appropriate,
  in
  

the
principles
and
techniques
of
the
activity
being
conducted,
and,

• 
  The
  training
  is
  documented
  and
  the
  documentation
  describes
  the
  content
  and
  method
  of
  the
  

training,
attendance,
date
of
attendance,
and
results
of
the
training
session.

The
program
takes
into
account
the
need
for
special
controls,
processes,
test
equipment,
tools,

and
skills
to
attain
the
required
quality,
and
the
need
for
verification
of
satisfactory

implementation.

Facility
Staff
Qualifications

Each
member
of
the
facility
staff
shall
meet
or
exceed
the
minimum
qualifications
of
ANSI

N18.1-1971
for
comparable
positions
unless
otherwise
noted.
The
requirements
of
ANSI
N18.1-1971
that
pertain
to
operator
license
qualifications
for
facility
staff
shall
not
apply.
(Formerly

TechnicalSpecification6.3and6.3.1)

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
11
of
47

The
management
position
responsible
for
radiological
control
or
his
deputy
shall
meet
or
exceed

the
qualifications
of
Regulatory
Guide
1.8
of
1977.
Each
Radiological
Controls
Technician
in
a

responsible
position
shall
meet
or
exceed
the
qualifications
of
ANSI
N18.1-1971,
paragraph

4.3.2
or
4.5.2,
or
be
formally
qualified
through
an
NRC-approved
TMI
Radiation
Controls

training
program.
All
Radiological
Controls
Technicians
will
be
qualified
through
training
and

examination
in
each
area
or
specific
task
related
to
their
radiological
control
functions
prior
to

their
performance
of
those
tasks.
(FormerlyTechnicalSpecification6.3.2)

5.0

QUALITY
ASSURANCE
CRITERIA

5.1

Criterion
I,
Organization

TMI-2
Solutions
shall
establish
an
organization
for
the
management,
conduct,
and
oversight
of

decommissioning
activities
within
the
scope
of
this
plan.
The
organization
shall
conform
to
the

descriptions
provided
below.

TMI-2
Solutions
shall
exercise
overall
responsibility
for
activities
affecting
quality
but
may

delegate
to
others,
such
as
third-party
contractors,
agents,
or
consultants,
the
work
of
establishing

and
executing
the
quality
assurance
program,
or
any
part
of
it,
provided
TMI-2
Solutions
retains

responsibility
for
all
such
activities.
The
persons
and
organizations
performing
QA
functions

shall
report
to
a
management
level
that
assures
that
the
required
authority
and
organizational

freedom
are
provided
to
identify
quality
problems;
to
initiate,
recommend,
or
provide
solutions;

and
to
verify
implementation
of
solutions;
and
shall
include
sufficient
independence
from
cost

and
schedule
when
opposed
to
safety.
The
individuals
assigned
the
responsibility
for
assuring

effective
execution
of
any
portion
of
this
plan
shall
have
direct
access
to
the
levels
of

management
necessary
to
perform
QA
functions.

The
principle
organizational
positions
are
as
follows:

5.1.1
TMI-2
Solutions
President/TMI-2
Solutions
Chief
Nuclear
Officer:
The
TMI-2
Solutions

Chief
Nuclear
Officer
is
located
off-site
and
is
responsible
to
provide
corporate
direction

on
activities
associated
with
the
safe
and
efficient
management
and
oversight
of
Three

Mile
Island
Unit
2.
The
individual
performing
this
function
fulfills
the
combined
roles
of

TMI-2
Solutions
Chief
Nuclear
Officer
and
TMI-2
Solutions
President,
and
is
responsible

to
the
EnergySolutions,
LLCPresident
and
Chief
Executive
Officer.

5.1.2
TMI-2
Solutions
Cognizant
Officer:
TMI-2
Solutions
Cognizant
Officer
is
located
off-site
and
is
responsible
for
the
overall
implementation,
and
effectiveness
of
the
TMI-2

Decommissioning
QA
Plan.
The
TMI-2
Solutions
Cognizant
Officer
is
responsible
to
the

TMI-2
SolutionsPresident/TMI-2
Solutions
Chief
Nuclear
Officer.
This
individual
is

responsible
for
the
overall
direction
of
decommissioning
activities
including
Operations

and
Maintenance,
Engineering
and
Engineering
Support,
Radiological
and
Environmental

Controls,
Waste
Management
and
regulatory
compliance
functions
including

implementation
of
this
plan.
The
TMI-2
Solutions
Cognizant
Officer
shall
have
a

minimum
of
a
baccalaureate
degree
in
a
technical
discipline
and
five
years
of
experience

in
a
managerial
position.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
12
of
47

TMI-2
Solutions
may
elect
to
subcontract
decommissioning
services.
These
services
are

required
to
meet
the
requirements
of
the
Defueled
Safety
Analysis
Report,
Technical

Specifications,
and
this
plan,
as
applicable.
Even
though
the
services
are
provided
by

TMI-2
Solutions
subcontractors,
the
final
decision-making
authority
regarding
direction

and
control
of
TMI-2
decommissioning
activities
remains
with
the
TMI-2
Solutions

Cognizant
Officer.
Also,
responsibility
for
the
implementation,
and
effectiveness
of
the

TMI-2
Decommissioning
QA
Plan
remains
with
the
TMI-2
Solutions
Cognizant
Officer.

5.1.3
TMI2
Solutions
Quality
Assurance
(QA)
Manager:

The
TMI-2
Solutions
QA
Manager

reports
directly
to
the
TMI-2
Solutions
Cognizant
Officer
and
is
subject
to
oversight
by

the
corporate
EnergySolutions
Group
Director
QA,
D&D.
The
TMI-2
Solutions
QA

Manager
is
responsible
for
the
independent
oversight
of
TMI-2
Decommissioning

Quality
Assurance
Plan
implementation
by
the
project
organization
reporting
to
the
TMI-2
Solutions
Cognizant
Officer.
This
responsibility
includes
the
performance
of
reviews

and
audits
to
verify
that
this
plan,
applicable
policies
and
procedures,
applicable
laws
and

regulations,
and
applicable
codes
and
standards
are
effectively
implemented
for
activities

within
the
scope
of
this
plan.
The
TMI-2
Solutions
QA
Manager
is
supported
by
a

permanent
onsite
QA
organization
for
decommission
activities
who
are
responsible
for

conducting
the
day-to-day
independent
oversight
of
the
implementation
of
the
TMI-2

Solutions
Decommissioning
QA
Plan,
however
the
TMI-2
Solutions
QA
Manager
retains

overall
responsibility
for
the
independent
oversight
of
this
plans
implementation.
The

TMI-2
Solutions
QA
Manager
shall
be
the
ultimate
authority
for
resolution
of
questions

regarding
the
applicability
and
scope
of
this
plan.
The
TMI-2
Solutions
QA
Manager
is

located
offsite
but
supports
onsite
activities
in
accordance
with
project
needs.

5.1.4
TMI-2
Solutions
Employee
Concerns
Program
(ECP)
Manager:
The
TMI-2
Solutions

ECP
Manager
is
located
on-site
for
dismantlement
and
reports
to
the
TMI-2
Solutions

Project
Director
and
is
subject
to
oversight
by
the
EnergySolutions
VP
Regulatory

Affairs.
The
TMI-2
Solutions
ECP
Manager
shall
assure
the
implementation
of
an

Employee
Concerns
Program.
The
TMI-2
Solutions
ECP
Manager
is
accessible
on
a

confidential
basis
and
has
access
to
the
highest
level
of
corporate
management
as
needed,

who
may
be
contacted
by
any
company
or
contracted
employee
who
has
a
safety
concern

that
he/she
considers
to
be
not
adequately
addressed.
The
TMI-2
Solutions
ECP
Manager

is
empowered
to
investigate
such
matters
and
identify
needed
corrective
actions.

5.1.5
Decommissioning
Nuclear
Safety
Review
Board
(DNSRB):
This
board
is
an
independent

panel
located
off-site
that
serves
to
assure
that
the
Three
Mile
Island
Unit
2
SSCs
are

maintained
so
as
to
protect
the
health
and
safety
of
the
workers,
the
public
and
the

environment,
and
to
enable
effective
and
efficient
dismantlement
and
decommissioning.

This
board
assists
and
reports
to
the
TMI-2
Solutions
Cognizant
Officer
in
carrying
out

the
oversight
functions
of
that
office
and
serves
an
advisory
role
to
the
TMI-2
Solutions

Cognizant
Officer.
The
board
reviews
overall
operating
trends
at
Three
Mile
Island
Unit

2,
monitors
general
plant
conditions,
and
performs
periodic
tours
of
the
facility
and
meets

with
site
staff
to
assess
the
overall
safety
condition
and
performance
of
the
facility.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
13
of
47

The
Decommissioning
Nuclear
Safety
Review
Board
shall
collectively
have
experience

in
the
areas
of
Quality
Assurance,
Radiological
Safety,
Environmental
Controls,

Decommissioning,
Civil
and
Structural
Engineering,
Industrial
Safety,
and
Facility

Management.(FormerlyTechnicalSpecification6.2.2)

5.1.6
TMI-2
Solutions
Project
Director:
The
TMI-2
Solutions
Project
Director
reports
to
the

Cognizant
Officer
and
is
located
on-site
for
dismantlement.
The
TMI-2
Solutions
Project

Director
has
responsibility
for
the
management
of
overall
facility
operations
at
Three

Mile
Island
Unit
2.
and
shall
delegate
in
writing
the
succession
to
this
responsibility

during
absence.
(FormerlyTechnicalSpecification6.1.1)
This
overall
responsibility
may

be
shared
by
more
than
one
individual.

The
TMI-2
Solutions
facility
organization
shall
ensure
an
individual
qualified
in
radiation

protection
procedures
is
on
site
whenever
Radioactive
Waste
Management
activities
are
in

progress.
(FormerlyTechnicalSpecification6.2.2)

5.2

Criterion
II,
Quality
Assurance
Program

This
plan
provides
the
base
document
for
the
implementation
of
a
quality
program
that
complies

with
the
applicable
criteria
of
10
CFR
50
Appendix
B.
This
plan
shall
be
implemented
through

written
policies
and
procedures.
Appendix
A
of
this
plan
provides
specific
Quality
Assurance

Criteria
to
be
applied
to
NRC
Licensed
transportation
SSCs
under
the
auspices
of
10
CFR
71,

and
specifies
how
the
criteria
of
10
CFR
71,
Subpart
H
will
be
applied
at
Three
Mile
Island
Unit

2
during
decommissioning.

Quality
Assurance
is
the
process
of
performing
planned
and
systematic
actions
that
will
provide

a
high
level
of
confidence
that
an
SSC,
or
program
will
perform
its
intended
actions
or
achieve

its
stated
goals
in
a
manner
consistent
with
the
intended
or
design
safety
function.
Quality

Assurance
involves
individual
efforts,
teamwork,
interfaces,
processes,
and
programs
that
are

carried
out
in
accordance
with
the
10
CFR
50,
Appendix
B
principles.
Staff
and
contractors

performing
Activities
Affecting
Quality
are
required
to
adhere
to
this
plan
and
are
required
to

report
any
deficiencies
or
concerns
regarding
potential
safety
issues
through
their
supervisory

and
management
chains.

Quality
Assurance
includes
Quality
Control,
which
encompasses
the

measurable
processes
by
which
programs,
materials,
and
SSCs
are
verified
to
be
in
conformance

to
the
predetermined
requirements
and
specifications.
In
addition,
management
review
and

oversight
of
activities
is
a
major
aspect
of
implementing
this
plan.

Assuring
quality
begins
with
clearly
describing
the
activity
affecting
quality
in
appropriate
work

control
documents
containing
understandable
and
concise
implementing
directions,
clearly

defined
performance
criteria
and
acceptance
criteria
where
applicable,
and
performing
the

described
activity
in
accordance
with
the
work
control
documents.
Where
appropriate,

completion
of
the
work
is
documented
through
such
means
as
recorded
data
and
documentary

evidence,
check
off
lists,
and
signed
off
procedural
steps,
followed
by
review
and
verification
as

needed.
Records
of
performance
of
completion
are
reviewed
by
appropriate
levels
of

management
and
maintained
in
a
retrievable
and
auditable
manner.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
14
of
47

Personnel
at
all
levels
of
the
organization
perform
quality
functions
through
adherence
to

instructions,
procedures,
and
drawings,
performance
of
self-checking,
looking
for
deficiencies

and
errors,
and
identification
and
reporting
of
errors,
omissions,
and
deficiencies.
The

management
and
supervisory
staff
encourage
and
support
workers
below
them
to
foster
a

questioning
attitude.

Quality
verification
includes
self-checking
while
preparing
written
documents
and
during

performance
of
prescribed
activities,
peer
review,
and
independent
verification
of
performance

of
critical
activities.
Another
level
of
quality
verification
includes
observation
of
the

performance
of
activities
by
independent
monitors
who
may
periodically
assess
performance
and

independently
verify
data
collected,
calibration
of
instruments,
and
performance
of
prescribed

actions
when
deemed
necessary
to
assure
safety.
The
highest
tier
of
verification
is
the
formal

audit
of
programmatic
areas.
Audits
are
formal
reviews
performed
in
accordance
with
written

procedures
or
checklists,
are
documented,
and
are
reported
to
senior
management.
Audits
are

generally
performed
by
staff
not
directly
responsible
for
the
area
being
audited
and
are
therefore

independent
reviews
of
the
implementation
of
programs.

Written
reports
of
audits,
and
reviews
by
independent
monitors
are
prepared
and
submitted
to

appropriate
levels
of
management
for
review.

This
approach
to
quality
verification
assures
that
deficiencies
are
identified
and
reported
as
soon

as
possible
by
the
staff
performing
the
activities
so
that
a
sense
of
personal
responsibility
and

ownership
of
the
tasks
being
performed
and
the
errors
that
will
occur
is
established.
This

contributes
to
the
effective
implementation
of
the
defense
in
depth
concept
where
multiple

barriers
exist
to
prevent
a
single
error
or
deficiency
resulting
in
a
significant
quality
defect.

5.3

Criterion
III,
Design
Control

When
decommissioning
activities
require
design
or
modification
of
SSCs
within
the
scope
of

this
plan,
controls
shall
be
applied
commensurate
with
the
potential
impact
on
safety
of
on-site

workers
and
the
general
public.

A
design
control
process
shall
be
implemented
to
assure
that
new
installations
and
modifications

to
existing
installations
within
the
scope
of
this
plan
conform
to
applicable
regulatory

requirements,
appropriate
building
codes,
and
the
design
bases
described
in
the
Defueled
Safety

Analysis
Report.
Design
activities
may
be
performed
by
an
outside
contractor
using
a
TMI-2

Solutions
reviewed
Vendor
QA
Plan.
Regardless
of
the
program
used,
TMI-2
Solutions
retains

final
approval
authority
for
design
activities
and
maintains
responsibility
for
the
adequacy
of

those
design
activities.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
15
of
47

Design
control
measures
shall
be
implemented
for
SSCs
within
the
scope
of
this
plan
by

controlled
written
procedures
commensurate
with
the
potential
impact
on
safety
addressing
the

following
considerations:

• 
  Describes
  how
  design
  inputs
  are
  specified
  and
  translated
  into
  design
  documents,
  the
  

verification
of
design
by
persons
other
than
the
originator,
and
the
assurance
that
changes
to

the
design
are
properly
reviewed
and
controlled;

• 
  Prescribes
  provisions
  for
  independent
  design
  verification
  using
  methods
  such
  as
  design
  

review
or
alternate
calculations.
Verification
and
independent
reviews,
including
records
of

these
activities
and
qualifications
of
reviewer,
shall
be
performed
in
a
manner
that
meets
the

requirements
Section
6.4
of
this
plan;

• 
  Prescribes
  an
  approval
  process
  for
  designs
  and
  design
  changes
  and
  requires
  approval
  of
  field
  

changes
by
the
same
level
of
staff
approving
the
original
design;

• 
  Prescribe
  provisions
  for
  design
  and
  specification
  changes,
  including
  field
  changes,
  and
  

assurance
that
independent
design
verification
methodology
is
applied
to
changes,
and
that

changes
to
designs
or
specifications
receive
at
a
minimum
the
same
level
of
review
as
the

original
design;

• 
  Provides
  a
  mechanism
  for
  maintaining
  records
  of
  design
  activities
  with
  sufficient
  detail
  to
  

permit
assessment
and
auditing
as
required
by
this
plan;
and

• 
  Provides
  methods
  to
  ensure
  that
  adequate
  precautions
  or
  evaluations
  are
  in
  place
  to
  preclude
  

decommissioning
activities,
including
installation,
modification,
or
removal
of
SSC's,
from

damaging
or
adversely
impacting
the
ability
of
items
required
by
the
Technical

Specifications.
to
maintain
the
plant
in
a
safe
and
stable
condition,
or
from
creating

conditions
that
could
preclude
future
unconditional
release
of
the
facility
from
license

control.

In
addition,
any
design
control
process
implemented
during
decommissioning
shall
contain

provisions
to
assure
that
the
reviews
and
evaluations
required
by
the
regulations
in
10
CFR
50.59

are
performed
by
qualified
individuals
and
documented
in
an
auditable
and
retrievable
manner.

There
shall
be
provisions
to
assure
that
responsible
plant
personnel
are
made
aware
of
any
design

changes
or
modifications
that
may
affect
the
performance
of
their
duties.

5.4

CriterionIV,ProcurementDocumentControl

Procurement
documents
for
materials
and
services
that
are
within
the
scope
of
this
plan
shall

contain
sufficient
information
to
assure
that
quality
requirements
are
clearly
communicated
to

vendors
and
that
sufficient
verification
of
quality
requirements
is
provided
by
the
vendors.
This

requirement
is
applicable
to
procurement
activities
performed
either
by
TMI-2
Solutions
or
any

contractor/sub-contractor
performing
decommissioning
activities.

Procurement
documents
for
materials
and
services
that
have
a
potential
impact
on
safety
shall

contain
specific
technical
and
quality
requirements.
Reference
to
manufacturer's
catalogue

specifications
is
adequate
where
such
specifications
adequately
address
the
critical
attributes

necessary
to
satisfy
the
intended
safety
function
of
the
procured
item
or
service.
Critical

attributes
are
those
important
design,
material,
and
performance
characteristics
of
an
item
that

provide
reasonable
assurance
that
it
will
perform
its
intended
function.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
16
of
47

Procurement
documents
for
materials
and
services
that
have
a
potential
impact
on
safety
shall

identify
those
records
which
suppliers
shall
retain,
maintain,
and
control
and
those
which
shall
be

delivered
prior
to
use
or
delivery
of
an
item
or
service.

If
considered
necessary
by
TMI-2
Solutions
Management
due
to
the
unique
characteristics
or

safety
significance
of
an
SSC
or
service,
procurement
documents
shall
specify
that
TMI-2

Solutions
or
a
designee
has
the
right
to
access
the
supplier
and
lower-tier
supplier
facilities
for

inspection,
audit,
or
review
of
records.

Procurement
processes
shall
be
specified
by
written
procedures
which
provide
for
review
by

appropriate
management
personnel
and
require
that
procurement
documents
for
materials
and

services
within
the
scope
of
this
plan
be
subject
to
a
quality
review
to
ensure
applicable

regulatory
requirements,
design
bases,
quality
assurance,
and
other
requirements
are
adequately

satisfied
prior
to
execution.

5.5

CriterionV,Instructions,Procedures,andDrawings



Activities
within
the
scope
of
this
plan
shall
be
conducted
in
accordance
with
approved
written

instruction,
procedures,
and/or
drawings
appropriate
to
the
potential
impact
on
safety
of
the

activity.
Administrative
procedures
and
instructions
shall
be
implemented
that
describe
the

minimum
requirements
for
the
preparation,
review,
approval,
and
control
of
instructions,

procedures,
and
drawings.

Where
applicable,
instructions,
procedures
and
drawings
shall
contain
appropriate
qualitative
and

quantitative
acceptance
criteria
to
ensure
activities
affecting
quality
are
satisfactorily
completed

and
shall
assure
that
appropriate
records
are
maintained
to
document
that
important
activities
are

satisfactorily
completed.

Instructions,
procedures,
and
drawings
shall
be
sufficiently
detailed
for
a
qualified
individual
to

perform
the
required
function
without
direct
supervision.
The
level
of
detail
in
the
instructions,

procedures,
and
drawings
should
be
commensurate
with
the
complexity
of
the
task
and
the
level

of
qualifications
and
training
of
the
individuals
normally
expected
to
perform
the
function.

5.5.1
Written
procedures
shall
be
established,
implemented,
and
maintained
for
the
activities
to

be
performed
during
decommissioning
(Phase
1b
and
Phase
2).

a.
Examples
of
activities
to
assure
conformance
to
the
descriptions
in
the
Defueled

Safety
Analysis
Report
(FormerlyTechnicalSpecification6.7.1)
are:

• 
  Procedures
  for
  implementation
  of
  the
  Technical
  Specifications
  (Formerly

TechnicalSpecification6.7.1.a),

• 
  Procedures
  for
  Radioactive
  Waste
  management
  and
  shipment
  (Formerly

TechnicalSpecification6.7.1.b),

• 
  Procedures
  for
  implementation
  of
  the
  Radiation
  Protection
  Plan
  (Formerly

TechnicalSpecification6.7.1.c),

• 
  Procedures
  for
  the
  implementation
  of
  the
  Fire
  Protection
  Program
  (Formerly

TechnicalSpecification6.7.1.d),

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
17
of
47

• 
  Procedures
  for
  the
  implementation
  of
  the
  Flood
  Protection
  Program
  (Formerly

TechnicalSpecification6.7.1.e)

b.
Examples
of
additional
procedures
required
for
decommissioning
are:

• 
  Administrative
  Procedures
  prescribing
  organizational
  responsibilities
  and
  

prescribing
the
processes
and
responsibilities
for
preparation,
review,
and

approval
of
instructions,
procedures,
and
drawings
within
the
scope
of
this
plan,

• 
  Operating
  Procedures
  for
  SSCs
  within
  the
  scope
  of
  this
  plan
  that
  provide
  

sufficient
detail
to
assure
safe
operation
in
compliance
with
License
Conditions,

• 
  Surveillance,
  Test,
  and
  Preventive
  Maintenance
  Procedures
  that
  provide
  sufficient
  

detail
for
a
qualified
technician
to
perform
those
surveillances,
inspections,
tests,

and
calibrations
to
assure
implementation
of
the
requirements
of
the
License,

• 
  Radiation
  Protection
  and
  Chemistry
  Procedures
  that
  assure
  compliance
  with
  

applicable
radiological
safety,
sampling,
and
analysis
requirements
of
the
License

and
of
the
environmental
monitoring
programs
and
the
Off-Site
Dose
Calculation

Manual,

• 
  Procedures
  for
  engineering
  designs
  and
  calculations,
  procurement
  of
  materials
  

and
services,
and
independent
technical
and
safety
reviews
of
activities
within
the

scope
of
this
plan,

• 
  Procedures
  for
  the
  implementation
  of
  Quality
  Assurance
  monitoring,
  audit,
  

receipt
inspection
and,
and
corrective
action
processes.

5.5.2
The
following
manual
and
programs
shall
be
established,
implemented,
and
maintained:

a.
Off-Site
Dose
Calculation
Manual
(ODCM)
(Ref.TechnicalSpecification1.12)

The
Off-Site
Dose
Calculation
Manual
(ODCM)
shall
contain
the
methodology
and

parameters
used
in
the
calculation
of
off-site
doses
resulting
from
radioactive
gaseous

and
liquid
effluents,
in
the
calculation
of
gaseous
and
liquid
effluent
monitoring

alarm/trip
setpoints,
and
in
the
conduct
of
the
Radiological
Environmental

Monitoring
Program.
The
ODCM
shall
also
contain:

• 
  The
  programs
  required
  by
  Section
  5.5.2.b
  and
  5.5.2.c
  below,
  and
  
  

• 
  Descriptions
  of
  the
  information
  that
  should
  be
  included
  in
  the
  Annual
  

Radiological
Environmental
Operating
and
Annual
Radioactive
Effluent
Release

Reports
required
by
Appendix
B,
B.1.1
of
this
plan
and
TechnicalSpecification

6.8.1.2,
respectively.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
18
of
47

b.
Radioactive
Effluent
Controls
Program
(FormerlyTechnicalSpecification6.7.4.a)

A
program
shall
be
provided
conforming
with
10
CFR
50.36a
for
the
control
of

radioactive
effluents
and
for
maintaining
the
doses
to
members
of
the
public
from

radioactive
effluents
as
low
as
reasonably
achievable.
The
program
shall
be
contained

in
the
ODCM,
(2)
shall
be
implemented
by
operating
procedures,
and
(3)
shall

include
remedial
actions
to
be
taken
whenever
the
program
limits
are
exceeded.
The

program
shall
include
the
following
elements:

• 
  Limitations
  on
  the
  operability
  of
  radioactive
  liquid
  and
  gaseous
  monitoring
  

instrumentation
including
surveillance
tests
and
setpoint
determination
in

accordance
with
the
methodology
in
the
ODCM,

• 
  Limitations
  on
  the
  concentrations
  of
  radioactive
  material
  released
  in
  liquid
  

effluents
to
unrestricted
area
conforming
to
10
times
the
concentrations
specified

in
10
CFR
Part
20.1001
-
20.2402,
Appendix
B,
Table
2,
Column
2,

• 
  Monitoring,
  sampling,
  and
  analysis
  of
  radioactive
  liquid
  and
  gaseous
  effluents
  in
  

accordance
with
10
CFR
20.1302
and
with
the
methodology
and
parameters
in
the

ODCM,

• 
  Limitations
  on
  the
  annual
  and
  quarterly
  doses
  or
  dose
  commitment
  to
  a
  member
  

of
the
public
from
radioactive
materials
in
liquid
effluents
released
from
the

facility
to
the
SITE
BOUNDARY
conforming
to
Appendix
I
to
10
CFR
Part
50,

• 
  Determination
  of
  cumulative
  and
  projected
  dose
  contributions
  from
  radioactive
  

effluents
for
the
current
calendar
quarter
and
current
calendar
year
in
accordance

with
the
methodology
and
parameters
in
the
ODCM
at
least
every
31
days,

• 
  Limitations
  on
  the
  operability
  and
  use
  of
  the
  liquid
  and
  gaseous
  effluent
  treatment
  

systems
to
ensure
that
the
appropriate
systems
are
used
to
reduce
releases
of

radioactivity
when
the
projected
doses
in
a
31-day
period
would
exceed
2
percent

of
the
guidelines
for
the
annual
dose
or
dose
commitment
conforming
to

Appendix
I
to
10
CFR
Part
50,

• 
  Limitations
  on
  the
  dose
  rate
  resulting
  from
  radioactive
  material
  released
  in
  

gaseous
effluents
to
areas
at
or
beyond
the
SITE
BOUNDARY.
The
limits
are
as

follows:

-

For
noble
gases:
less
than
or
equal
to
500
mrem/yr
to
the
total
body
and
less

than
or
equal
to
3000
mrem/yr
to
the
skin,
and

-

For
tritium
and
all
radionuclides
in
particulate
form
with
half-lives
greater

than
8
days:
less
than
or
equal
to
1500
mrem/yr
to
any
organ

• 
  Limitations
  on
  the
  annual
  and
  quarterly
  air
  doses
  resulting
  from
  noble
  gases
  

released
in
gaseous
effluents
from
the
facility
to
areas
beyond
the
SITE

BOUNDARY
conforming
to
Appendix
I
to
10
CFR
Part
50,

• 
  Limitations
  on
  the
  annual
  and
  quarterly
  doses
  to
  a
  member
  of
  the
  public
  from
  

tritium
and
all
radionuclides
in
particulate
form
with
half-lives
greater
than
8
days

in
gaseous
effluents
released
from
each
facility
to
areas
beyond
the
SITE

BOUNDARY
conforming
to
Appendix
I
to
10
CFR
Part
50,

• 
  Limitations
  on
  the
  annual
  dose
  or
  dose
  commitment
  to
  any
  member
  of
  the
  public
  

due
to
releases
of
radioactivity
and
to
radiation
from
uranium
fuel
cycle
sources

conforming
to
40
CFR
Part
190.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
19
of
47

c.
Radiological
Environmental
Monitoring
Program
(FormerlyTechnicalSpecification

6.7.4.b)

A
program
will
be
provided
to
monitor
the
radiation
and
radionuclides
in
the
environs

of
the
plant.
The
program
shall
provide
(1)
representative
measurements
of

radioactivity
in
the
highest
potential
exposure
pathways,
and
(2)
verification
of
the

accuracy
of
the
effluent
monitoring
program
and
modeling
of
environmental
exposure

pathways.
The
program
shall
(1)
be
contained
in
the
ODCM,
(2)
conform
to
the

guidance
of
Appendix
I
to
10
CFR
Part
50,
and
(3)
Include
the
following:

• 
  Monitoring,
  sampling,
  analysis,
  and
  reporting
  of
  radiation
  and
  radionuclides
  in
  

the
environment
in
accordance
with
the
methodology
and
parameters
in
the

ODCM.

d.
Radiation
Protection
Program
(FormerlyTechnicalSpecification6.10)

Procedures
for
personnel
radiation
protection
shall
be
prepared
consistent
with
the

requirements
of
10
CFR
Part
20
and
shall
be
approved,
maintained,
and
adhered
to

for
all
operations
involving
personnel
radiation
exposure.

5.6

Criterion
VI,
Document
Control

5.6.1
Measures
shall
be
established
to
control
the
issuance
of
documents,
such
as
instructions,

procedures,
drawings,
and
other
documents,
including
changes
thereto,
which
prescribe

activities
within
the
scope
of
this
plan.
These
measures
shall
assure
that
applicable

documents
are
available
for
use
by
the
individuals
performing
the
activities
and
at
the

locations
where
the
activities
are
being
performed.

a.
The
process
for
controlling
documents
shall
specify
the
organizational
elements

responsible
for
approval
and
issuance
of
documents.

b.
Documents
and
changes
to
documents
within
the
scope
of
this
plan
shall
be

controlled
in
a
manner
that
precludes
the
use
of
inappropriate
or
outdated
documents

and
there
shall
be
provisions
to
assure
that
documents
in
use
are
the
currently

approved
revisions.

c.
Changes
to
documents
shall
be
reviewed
and
approved
by
the
same
organizations
that

performed
the
original
review
unless
another
responsible
organization
is
designated

by
management.

5.6.2
Documents
and
substantive
changes
thereto
controlling
activities
within
the
scope
of
this

plan
shall
be
prepared
by
a
staff
or
contractors
knowledgeable
in
the
area
affected
by
the

procedure
and
shall
be
reviewed
for
technical
adequacy
by
an
independent
individual(s)

or
group
other
than
the
preparer,
but
who
may
be
from
the
same
organization
as
the

individual
who
prepared
the
procedure
or
change.
(Refer
to
Sections
6.0
and
7.0
of
this

plan
for
requirements
related
to
independent
Technical
and
Independent
Safety
Review).

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
20
of
47

a.
Each
of
the
procedures
required
for
decommissioning
(Ref.
5.5.1
of
this
plan),
and

SUBSTANTIVE
CHANGES
thereto,
shall
be
reviewed
and
approved
prior
to

implementation
and
shall
be
reviewed
periodically
as
follows
(FormerlyTechnical

Specification6.7.2):

• 
  At
  least
  every
  two
  years,
  the
  group
  responsible
  for
  Quality
  Assurance
  will
  assess
  

a
representative
sample
of
plant
procedures
that
are
used
more
frequently
than

every
two
years


(FormerlyTechnicalSpecification6.7.2.a).

• 
  Plant
  procedures
  that
  have
  been
  used
  at
  least
  biennially
  receive
  scrutiny
  by
  

individuals
knowledgeable
in
procedures
and
are
updated
as
necessary
to
ensure

adequacy
during
suitable
controlled
activities
(FormerlyTechnicalSpecification

6.7.2.b).

• 
  Plant
  procedures
  that
  have
  not
  been
  used
  for
  two
  years
  will
  be
  reviewed
  before
  

use
or
biennially
to
determine
if
changes
are
necessary
or
desirable
(Formerly

TechnicalSpecification6.7.2.c).

b.
SUBSTANTIVE
CHANGES
to
the
ODCM
(FormerlyTechnicalSpecification6.12):

• 
  Shall
  be
  documented
  and
  records
  of
  reviews
  performed
  shall
  be
  realized
  as
  

required
by
5.17.5.b
of
this
plan.
This
documentation
shall
contain:

-

Sufficient
information
to
support
the
change
together
with
the
appropriate

analyses
or
evaluations
justifying
the
change(s),
and

-

A
determination
that
the
change
will
maintain
the
level
of
radioactive
effluent

control
required
by
10
CFR
20.1301,
40
CFR
Part
190,
10
CFR
50.36a
and

Appendix
I
to
10
CFR
Part
50
and
not
adversely
impact
the
accuracy
or

reliability
of
effluent,
dose,
or
setpoint
calculations
(FormerlyTechnical

Specification6.12.a).

• 
  Shall
  become
  effective
  after
  review
  and
  acceptance
  by
  the
  TMI-2
  Solutions
  

Project
Director
(FormerlyTechnicalSpecification6.12.b).

• 
  Shall
  be
  submitted
  to
  the
  Commission
  in
  the
  form
  of
  a
  complete,
  legible
  copy
  of
  

the
entire
ODCM
as
part
of
or
concurrent
with
the
Annual
Radioactive
Effluent

Release
Report
for
the
period
of
the
report
in
which
any
change
to
the
ODCM

was
made.
Each
change
shall
be
identified
by
markings
in
the
margin
of
the

affected
pages,
clearly
indicating
the
area
of
the
page
that
was
changed,
and
shall

indicate
the
date
(e.g.,
month/year)
the
change
was
implemented
(Formerly

TechnicalSpecification6.12.c).

5.6.3
Temporary
changes
to
procedures
required
for
the
activities
to
be
performed
during

decommissioning
(Phase
1b
and
Phase
2)
may
be
made
provided
(FormerlyTechnical

Specification6.7.3):

a.
The
intent
of
the
original
procedures
is
not
altered
(FormerlyTechnicalSpecification

6.7.3.a)

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
21
of
47

b.
The
change
is
approved
by
two
members
of
the
responsible
organization

knowledgeable
in
the
area
affected
by
the
procedure.
For
changes
which
may
affect

the
operational
status
of
facility
systems
or
equipment,
at
least
one
of
these

individuals
shall
be
a
member
of
facility
management
or
supervision
(Formerly

TechnicalSpecification6.7.3.b);
and

c.
The
change
is
documented,
reviewed,
and
approved
within
14
days
of

implementation
(FormerlyTechnicalSpecification6.7.3.c).

5.7

Criterion
VII,
ControlofPurchasedMaterial,Equipment,&Services

A
documented
process
will
be
implemented
to
verify
that
purchased
material
within
the
scope
of

this
plan
conforms
to
the
requirements
specified
in
the
procurement
documents.
The
process
will

include
measures
such
as
physical
inspection
of
materials
upon
receipt,
review
of
documentary

evidence
provided
by
the
vendor,
or
testing
and
analysis
upon
receipt,
as
appropriate,
to

determine
that
the
material
meets
the
specifications
in
the
procurement
documents.
The

processes
shall
use
a
graded
approach
applying
considerations
of
a
material's
or
component's

importance
to
safety,
complexity,
history
of
previous
use,
and
the
ability
to
be
readily
assessed

by
review
of
vendor
published
commercial
documents
such
as
catalogues
and
published

specification
sheets.
Applying
the
graded
approach,
the
procurement
documents
may
also
require

TMI-2
Solutions
review
and
approval
of
the
vendor's
drawings,
procedures,
manufacturing

plans,
and
Quality
Assurance
Plans
as
needed.

When
procurement
documents
require
implementation
of
a
Quality
Assurance
Plan
by
the

vendor,
the
adequacy
of
vendor's
quality
assurance
program
shall
be
verified
and,
if
appropriate,

the
vendors'
adherence
to
their
quality
assurance
program
shall
be
verified.
This
may
be

accomplished
by
direct
audit
of
the
vendor,
review
of
the
vendor's
documented
program,
or
by

review
of
audit
history
performed
by
EnergySolutions
or
other
clients.

Measures
shall
be
established
to
identify
nonconforming
items
to
assure
that
such
items
do
not

get
placed
in
service
until
the
nonconforming
issues
are
corrected.

5.8

Criterion
VIII,
IdentificationandControlofMaterials,Parts,andComponents

For
materials
within
the
scope
of
this
plan,
measures
shall
be
established
to
assure
that
only

correct
and
accepted
items
are
used
or
installed.
Identification
of
items
shall
be
maintained
on

the
items
or
in
documents
traceable
to
the
items,
or
in
a
manner
that
assures
that
identification
is

established
and
maintained.

5.9

Criterion
IX,
Control
of
Special
Processes

Measures
shall
be
established
to
assure
that
special
processes,
including
welding,
heat
treating,

chemical
cleaning,
and
nondestructive
examination
are
controlled
and
accomplished
by
qualified

personnel
using
qualified
procedures
in
accordance
with
applicable
codes,
standards,

specifications,
criteria,
and
other
special
requirements.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
22
of
47

Procedures
for
the
performance
of
special
processes
shall
have
provisions
for
the
recording
of

the
identity
and
qualification
of
personnel
performing
special
processes.

5.10

Criterion
X,
Inspection

Appropriate
inspections
shall
be
performed
to
verify
that
activities
within
the
scope
of
this
plan

have
been
conducted
in
conformance
with
the
applicable
instructions,
procedures,
and
drawings

for
accomplishing
the
activity.
The
frequency
and
degree
of
inspection
shall
be
determined

through
application
of
the
graded
approach
considering
the
safety
significance
and
complexity
of

the
SSC
or
activity.

Such
inspections
shall
be
performed
by
individuals
other
than
those
who
performed
the
activity

being
inspected
and
shall
be
performed
in
accordance
with
appropriate
procedures
containing
a

description
of
objectives,
acceptance
criteria
and
prerequisites
for
performing
the
inspections.

These
procedures
shall
also
specify
any
special
equipment
or
calibrations
required
to
conduct
the

inspection.

Personnel
performing
required
inspections
shall
be
qualified
based
upon
experience
and
training

in
inspection
methods.

If
mandatory
inspection
hold
points
are
to
be
established,
which
require
witnessing
or
inspection

and
beyond
which
work
shall
not
proceed
without
prior
consent
are
required,
the
specific
hold

points
shall
be
indicated
in
the
work
control
documents
and
the
work
control
documents
shall
be

reviewed
prior
to
implementation
by
the
organization
required
to
perform
the
inspection.

5.11

Criterion
XI,
Test
Control

Appropriate
testing
shall
be
performed
to
demonstrate
that
the
SSCs
within
the
scope
of
this
plan

will
perform
satisfactorily
in
service,
and
that
performance
criteria
specified
in
License
Basis

Documents
are
met.

Tests
within
the
scope
of
this
plan
shall
be
prescribed
by
written
procedures
that
contain
detailed

performance
steps
and
checklists
for
conducting
the
test.
Procedures
shall
include
appropriate

inspection
hold
points
and
provide
documentation
of
performance,
results,
and
satisfactory

completion
of
the
prescribed
tests.
Such
documentation
shall
include
an
evaluation
to
verify
that

test
requirements
have
been
met.

Where
necessary,
procedures
shall
specify
test
prerequisites
such
as
calibrated
instrumentation,

required
test
equipment,
controlled
and
suitable
environmental
conditions,
and
provisions
for

data
collection
and
storage,
and
shall
contain
appropriate
quantitative
and
qualitative
acceptance

criteria
that
provide
evidence
that
applicable
design
specifications
and
procurement

specifications
have
been
met.

5.12

Criterion
XII,
Control
of
Measuring
and
Test
Equipment

Procedures
shall
be
implemented
that
establish
measures
to
assure
that
tools,
gauges,

instruments,
and
other
measuring
and
testing
devices
used
in
activities
within
the
scope
of
this

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
23
of
47

plan
are
properly
controlled,
calibrated,
and
adjusted
at
specified
periods
to
maintain
accuracy

within
necessary
limits.
The
procedures
shall
assure
that
appropriate
records
are
maintained
to

provide
objective
evidence
that
the
control
program
is
implemented.

When
Measuring
and
Test
Equipment
is
found
to
be
out
of
calibration,
the
validity
of
previous

tests
and
calibrations
where
the
item
had
been
used
since
its
last
calibration
shall
be
evaluated.

Such
evaluations
shall
be
documented
and
corrective
actions
shall
be
taken,
as
necessary.

Rulers,
tape
measures,
levels,
and
other
such
devices
need
not
be
included
in
the
Measuring
and

Test
Equipment
control
program
if
normal
commercial
practices
provide
for
adequate
accuracy.

5.13

Criterion
XIII,
Handling,
Storage
and
Shipping

Procedures
will
be
developed
to
properly
control
the
handling,
storage,
and
shipping
of
items

within
the
scope
of
this
plan
considering
the
safety
significance,
complexity,
and
physical

characteristics
of
the
items
and
material.
The
procedures
will
include
requirements
as
needed
to

specify
cleanliness,
packaging,
and
preservation
methods
to
prevent
damage,
deterioration,
or

loss.

Cleaning
and
preservation
requirements
shall
be
included
in
design
and
procurement
documents,

specifications,
and
procedures,
if
appropriate.

Controls
shall
be
established
to
ensure
that
items
whose
shelf
life
has
expired
will
not
be
used.

5.14

Criterion
XIV,
Inspection,
Test,
and
Operating
Status

Measures
shall
be
established
to
indicate
the
status
of
inspections
and
tests
performed
on

individual
items
within
the
scope
of
this
plan
by
markings
such
as
stamps,
tags,
labels,
routing

cards,
inspection
records,
or
other
suitable
means.
The
status
of
inspection
and
test
activities

shall
be
identified
either
on
the
items
or
in
documents
traceable
to
the
items.
These
measures
will

provide
for
the
identification
and
status
of
items
where
it
is
necessary
to
ensure
that
required

inspections
and
tests
are
performed
and
to
ensure
that
items
that
have
not
passed
the
required

inspections
and
tests
are
not
inadvertently
installed,
used,
or
operated.

Measures
will
also
be
established
for
indicating
the
operating
status
of
SSCs
and
to
prevent

inadvertent
operation
by
such
methods
as
tagging
or
locking
valves
and
switches.Such
measures

will
include
provisions
for
permission
to
release
equipment
or
systems
for
maintenance
or

modification,
documentation
of
the
mechanisms
employed
to
isolate
such
systems
and

components
for
protection
of
equipment
and
personnel,
and
release
of
the
controls
for
return
to

service
when
necessary.
Control
measures
shall
include
such
items
as
switching
and
tagging,

control
of
lifted
leads
and
jumpers,
and
temporary
bypasses.

5.15

Criterion
XV,
Nonconforming
Materials,
Parts,
or
Components

For
items
within
the
scope
of
this
plan,
measures
shall
be
established
to
control
materials,
parts,

or
components
which
do
not
conform
to
requirements
in
order
to
prevent
their
inadvertent
use.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
24
of
47

These
measures
shall
include,
as
appropriate,
procedures
for
identification,
documentation,

segregation,
disposition,
and
notification
to
affected
organizations.

Measures
shall
be
established
to
control
further
processing,
delivery
or
installation
of
a

nonconforming
item
or
continuation
of
a
nonconforming
service
or
activity,
pending
a
decision

on
its
disposition.

Nonconforming
items
shall
be
reviewed
and
accepted,
rejected,
repaired,
or
reworked
in

accordance
with
documented
procedures.
Procedures
will
provide
for
documentation
of
the

review
and
disposition
and
will
require
documented
technical
justification
for
the
action
taken.

5.16

Criterion
XVI,
Corrective
Action

A
process
shall
be
implemented
to
assure
that
conditions
adverse
to
quality,
such
as
failures,

malfunctions,
discrepancies,
deviations
from
specifications,
defective
material,
and
equipment,

and
nonconformances
are
promptly
identified
and
corrected.

For
significant
conditions
adverse
to
quality,
such
as
Technical
Specification
or
License

Condition
violations
or
events
reportable
to
the
Nuclear
Regulatory
Commission,
the
process

will
include
provisions
to
assure
that
the
cause
of
the
condition
is
determined
and
that
corrective

actions
are
implemented
to
correct
the
condition
and
to
prevent
recurrence
of
the
condition.

The
identification
of
the
condition
adverse
to
quality,
the
causes
of
the
condition,
and
the

corrective
actions
shall
be
documented
and
reported
to
the
appropriate
management
personnel

for
review.

5.17

Criterion
XVII,
Quality
Assurance
Records

Sufficient
records
shall
be
maintained
to
provide
documentary
evidence
that
activities
affecting

quality
have
been
adequately
performed
in
compliance
with
this
plan.
These
records
shall

include
all
those
required
by
5.17.5
of
this
plan,
Technical
Specifications
and
any
Environmental

Permits
and
programs,
records
of
required
independent
Technical
and
Independent
Safety

Reviews,
and
records
of
procedure
and
plan
revisions
that
are
within
the
scope
of
this
plan.

5.17.1
These
records
shall
be
identifiable
and
retrievable
and
shall
document,
when
applicable,

the
identification
of
the
individual
generating
the
document
or
the
data
recorder,
the
type

of
observation,
the
results,
the
acceptability,
and
the
action
taken
in
connection
with
any

deficiencies
noted.

a.

Records
of
calibration
of
measuring
and
test
equipment
and
instrumentation
within

the
scope
of
this
plan
shall
contain
documentary
evidence
of
traceability
to
national

standards.

5.17.2
Records
stored
for
retention
shall
be
protected
from
damage
or
deterioration
from

environmental
and
catastrophic
events
through
either
protected
storage
areas
or
devices

or
separate
duplicate
file
storage
methods.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
25
of
47

5.17.3
Written
records
(FormerlyTechnicalSpecification6.5.1.8)
of
activities
performed
in

accordance
with
Sections
6.1
through
6.7
of
this
plan
(FormerlyTechnicalSpecification

6.5.1.1through6.5.1.7)
shall
be
maintained
in
accordance
with
5.17.5
of
this
plan.

5.17.4
Written
records
of
reports
of
reviews
encompassed
in
Section
7.4
of
this
plan
(Formerly

TechnicalSpecification6.5.2.5)
shall
be
maintained
in
accordance
with
5.17.5
of
this

plan.

5.17.5
Records
Retention

Procedures
will
be
implemented
to
specify
responsibility,
duration,
and
location
of

records
retention.

a.
The
following
records
shall
be
retained
for
at
least
five
years
(FormerlyTechnical

Specification6.9.1):

• 
  Records
  of
  sealed
  source
  and
  fission
  detection
  leak
  tests
  and
  results.
  

• 
  Records
  of
  annual
  physical
  inventory
  of
  all
  sealed
  source
  material
  of
  record.
  

b.
The
following
records
shall
be
retained
as
long
as
the
Licensee
has
an
NRC
license
to

possess
at
the
Three
Mile
Island
facility
(FormerlyTechnicalSpecification6.9.2).

• 
  Records
  and
  logs
  of
  unit
  operation
  covering
  time
  interval
  at
  each
  power
  level.
  

• 
  Records
  and
  logs
  of
  principal
  maintenance
  activities,
  inspections,
  repair,
  and
  

replacement
of
principal
items
of
equipment
related
to
nuclear
safety
and

radioactive
waste
systems.

• 
  All
  reportable
  events
  submitted
  to
  the
  Commission.
  

• 
  Records
  of
  surveillance
  activities,
  inspections
  and
  calibrations
  required
  by
  

Technical
Specifications.

• 
  Records
  of
  changes
  made
  to
  the
  procedures
  required
  by
  Recovery
  Technical
  

Specification
6.8.1,
records
of
changes
to
programs
and
procedures
required
by

PDMS
Technical
Specification
6.7.1,
and
records
required
by
5.5.1
of
this
plan.

• 
  Radiation
  Safety
  Program
  Reports
  and
  Quarterly
  Recovery
  Progress
  Reports
  on
  

the
March
28,
1979
incident.

• 
  Records
  of
  radioactive
  shipments.
  

• 
  Records
  and
  logs
  of
  radioactive
  waste
  systems
  operations.
  

• 
  Records
  and
  drawing
  changes
  reflecting
  facility
  design
  modifications
  made
  to
  

systems
and
equipment
described
in
the
Safety
Analysis
Report,
Technical

Evaluation
Report
(TER),
System
Descriptions
(SD),
or
Safety
Evaluation

previously
submitted
to
NRC.

• 
  Records
  of
  new
  and
  irradiated
  fuel
  inventory,
  fuel
  transfers
  and
  assembly
  burnup
  

histories.

• 
  Records
  of
  transient
  or
  operational
  cycles
  for
  those
  unit
  components
  designed
  for
  

a
limited
number
of
transients
or
cycles.

• 
  Records
  of
  reactor
  tests
  and
  experiments.
  

• 
  Records
  of
  training
  and
  qualification
  for
  current
  members
  of
  the
  facility
  staff.
  

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
26
of
47

• 
  Records
  of
  in-service
  inspections
  previously
  required
  by
  the
  Technical
  

Specifications.

• 
  Records
  of
  Quality
  Assurance
  activities
  required
  by
  the
  Operating,
  Recovery,
  

PDMS
or
Decommissioning
Quality
Assurance
Plans.

• 
  Records
  of
  reviews
  performed
  for
  changes
  made
  to
  procedures
  or
  equipment
  or
  

reviews
of
tests
and
experiments
pursuant
to
10
CFR
50.59.

• 
  Records
  of
  meetings
  of
  the
  Plant
  Operation
  Review
  Committee
  (PORC)
  and
  the
  

Generation
Review
Committee
(GRC),
and
reports
of
evaluations
prepared
by
the

Independent
Onsite
Safety
Review
Group
(IOSRG),
if
applicable
to
Three
Mile

Island
Unit
2.

• 
  Records
  of
  the
  incident
  which
  occurred
  on
  March
  28,
  1979.
  

• 
  Records
  of
  facility
  radiation
  and
  contamination
  surveys.
  

• 
  Records
  of
  radiation
  exposure
  received
  by
  all
  individuals
  for
  whom
  monitoring
  

was
required.

• 
  Records
  of
  gaseous
  and
  liquid
  radioactive
  material
  released
  to
  the
  environs.
  

• 
  Records
  of
  reviews
  performed
  for
  changes
  made
  to
  the
  Offsite
  Dose
  Calculation
  

Manual
(ODCM).

5.18

Criterion
XVIII,
Audits

Planned
audits
shall
be
conducted
to
verify
compliance
with
appropriate
requirements
of
this

plan
and
to
determine
its
effectiveness.
The
audits
shall
be
performed
in
accordance
with
written

procedures
or
checklists
by
appropriately
qualified
personnel
not
having
direct
responsibility
in

the
areas
being
audited.
Audit
results
shall
be
documented
and
reviewed
by
management
having

responsibility
in
the
area
audited.
Follow-up
action,
including
re-audit
of
discrepant
areas,
shall

be
taken
where
indicated.

Audits
of
facility
activities
(FormerlyTechnicalSpecification6.5.3.1)
shall
be
performed
in

accordance
with
this
plan
and
shall
encompass
the
following
at
the
specified
interval:

a.
The
conformance
of
facility
operations
to
provisions
contained
within
the
Technical

Specifications
and
applicable
license
conditions.
(Biennial)

b.
The
performance
of
activities
required
by
the
Decommissioning
QA
Plan.
(Biennial)

c.
The
Radiation
Protection
Plan
and
applicable
implementing
procedures.
(Biennial)

d.
The
Fire
Protection
Program
and
implementing
procedures.
(Biennial)

e.
An
independent
fire
protection
and
loss
prevention
program
inspection
and
technical
audit

shall
be
performed
utilizing
either
qualified
licensee
personnel
or
an
outside
fire
protection

firm.
(Biennial)

f.
An
inspection
and
audit
of
the
fire
protection
and
loss
prevention
program
by
an
outside

qualified
fire
consultant.
(Triennial)

g.
The
Off-site
Dose
Calculation
Manual
and
implementing
procedures.
(Biennial)

h.
The
performance
of
the
QA
Assessment
activities
provided
by
the
Decommissioning

Services
Agreement
contractor
(Biennial).
Any
other
area
of
facility
operation
considered

appropriate
by
the
TMI-2
Solutions
Project
Director
or
the
TMI-2
Solutions
Cognizant

Officer.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
27
of
47

Audit
reports
(FormerlyTechnicalSpecification6.5.3.2)
encompassed
by
this
section
shall
be

forwarded
for
action
to
the
management
positions
responsible
for
the
areas
audited
and
the
TMI-2
Solutions
Cognizant
Officer
within
60
days
after
completion
of
the
audit.
Upper
management

shall
be
informed
in
accordance
with
this
plan.

Deficiencies
in
program
implementation
shall
be
regarded
as
Conditions
Adverse
to
Quality
and

shall
be
subject
to
the
appropriate
Corrective
Actions
in
conformance
with
Section
5.16
of
this

plan.

Single
audits
may
encompass
multiple
specified
audit
areas.
Audit
interval
schedules
are
based

on
the
month
in
which
the
audit
begins
and
the
frequency
of
the
audits
as
specified
may
be

increased
by
no
more
than
twenty
five
percent.
If
intervals
are
extended,
the
succeeding
audit
is

scheduled
based
on
the
original
anniversary
month
rather
than
the
extended
start
date,
and

combined
interval
for
any
three
consecutive
audits
shall
not
exceed
3.25
times
the
specified
audit

interval.



6.0

TECHNICALREVIEWANDCONTROL



The
TMI-2
Solutions
Cognizant
Officer
shall
be
responsible
for
ensuring
the
preparation,
review,

and
approval
of
documents
required
by
the
activities
described
in
Sections
6.1
through
6.7
of
this

plan
(FormerlyTechnicalSpecificationsSections6.5.1.1through6.5.1.7).
Preparation,
review

and
approval
of
these
documents
may
be
performed
by
TMI-2
Solutions
or
by
TMI-2
Solutions

contractor.
Implementing
approvals
shall
be
performed
at
the
cognizant
manager
level
or
above.



6.1

(FormerlyTechnicalSpecification6.5.1.1)
Each
procedure
required
by
5.5.1,
as
described

5.5.1.a,
5.5.2.b
and
5.5.2.c
of
this
plan,
and
other
procedures
including
those
for
tests
and

experiments
and
SUBSTANTIVE
CHANGES
thereto
shall
be
prepared
by
a
designated

individual(s)
or
group
knowledgeable
in
the
area
affected
by
the
procedure.

Each
such

procedure,
and
SUBSTANTIVE
CHANGES
thereto,
shall
be
given
a
technical
review
by

an
individual(s)
or
group
other
than
the
preparer,
but
who
may
be
from
the
same

organization
as
the
individual
who
prepared
the
procedure
or
change.

6.2

(FormerlyTechnicalSpecification6.5.1.2)
Proposed
changes
to
the
Technical

Specifications
shall
be
reviewed
by
a
knowledgeable
individual(s)
or
group
other
than
the

individual(s)
or
group
who
prepared
the
change.

6.3

(FormerlyTechnicalSpecification6.5.1.3)
Proposed
tests
and
experiments
shall
be

reviewed
by
a
knowledgeable
individual(s)
or
group
other
than
the
preparer
but
who
may

be
from
the
same
division
as
the
individual
who
prepared
the
tests
and
experiments.

6.4

(FormerlyTechnicalSpecification6.5.1.4)
Proposed
modifications
to
facility
structures,

systems,
and
components
necessary
to
assure
conformance
to
the
descriptions
in
the

Defueled
SAR
shall
be
designed
by
an
individual/organization
knowledgeable
in
the

areas
affected
by
the
proposed
modification.

Each
such
modification
shall
be
technically

reviewed
by
an
individual(s)/group
other
than
the
individual/group
which
designed
the

modification
but
may
be
from
the
same
group
as
the
individual
who
designed
the

modification.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
28
of
47

6.5

(FormerlyTechnicalSpecification6.5.1.5)
Investigation
of
all
violations
of
the
Technical

Specifications
including
the
preparation
and
forwarding
of
reports
covering
evaluation

and
recommendations
to
prevent
recurrence,
shall
be
reviewed
by
a
knowledgeable

individual(s)/group
other
than
the
individual(s)/group
which
performed
the
investigation.

6.6

(FormerlyTechnicalSpecification6.5.1.6)
All
reportable
events
shall
be
reviewed
by
an

individual/group
other
than
the
individual/group
which
prepared
the
report.

6.7

(FormerlyTechnicalSpecification6.5.1.7)
Individuals
responsible
for
reviews
performed

in
accordance
with
Section
6.1
through
6.6
(FormerlyTechnicalSpecification6.5.1.1

through6.5.1.6)
shall
include
a
determination
of
whether
or
not
additional
cross

disciplinary
review
is
necessary.

If
deemed
necessary,
such
review
shall
be
performed
by

the
appropriate
personnel.
Individuals
responsible
for
reviews
considered
under
Sections

6.1,
6.3,
and
6.4
(FormerlyTechnicalSpecifications6.5.1.1,6.5.1.3,and6.5.1.4)
shall

render
determinations
in
writing
with
regard
to
whether
or
not
NRC
approval
is
required

pursuant
to
10
CFR
50.59.



6.8

Responsible
Technical
Reviewers
(FormerlyTechnicalSpecification6.5.1.9)
shall
meet

or
exceed
the
qualifications
of
ANSI/ANS
3.1
of
1978
Section
4.6
or
4.4
for
applicable

disciplines
or
have
7
years
of
appropriate
experience
in
the
field
of
his
or
her
specialty.

Credit
toward
experience
will
be
given
for
advanced
degrees
on
a
one
to
one
basis
up
to
a

maximum
of
two
years.

Responsible
Technical
Reviewers
shall
be
designated
in
writing.

7.0

INDEPENDENT
SAFETY
REVIEW

The
TMI-2
Solutions
Cognizant
Officer
(FormerlyTechnicalSpecification6.5.2.1)
shall
be

responsible
for
ensuring
the
Independent
Safety
Review
of
the
subjects
described
in
Section
7.4

of
this
plan
(FormerlyTechnicalSpecificationSection6.5.2.5).
Independent
Safety
Reviews
may

be
performed
by
TMI-2
Solutions
or
by
TMI-2
Solutions
contractor.

7.1

(FormerlyTechnicalSpecification6.5.2.2)
Independent
safety
review
shall
be
completed

by
an
individual
or
group
not
having
direct
responsibility
for
the
performance
of
the

activities
under
review,
but
who
may
be
from
the
same
functionally
cognizant

organization
as
the
individual
or
group
performing
the
original
work.

7.2

(FormerlyTechnicalSpecification6.5.2.3)
TMI-2
Solutions
shall
collectively
have

access
to
the
experience
and
competence
required
to
independently
review
subjects
in
the

following
areas:

a.

Nuclear
facility
operations

b.

Nuclear
engineering

c.

Chemistry
and
radiochemistry

d.

Metallurgy

e.

Instrumentation
and
control

f.

Radiological
safety

g.

Mechanical
engineering

h.

Electrical
engineering

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
29
of
47

i.

Administrative
controls
and
quality
assurance
practices

j.

Other
appropriate
fields
such
as
radioactive
waste
management
operations

associated
with
the
unique
characteristics
of
Three
Mile
Island
Unit
2.

7.3

(FormerlyTechnicalSpecification6.5.2.4)Consultants
may
be
utilized
as
determined
by

the
TMI-2
Solutions
Cognizant
Officer
to
provide
expert
advice.

7.4

(FormerlyTechnicalSpecification6.5.2.5)
The
following
subjects
shall
be
independently

reviewed
by
Independent
Safety
Reviewers
(ISRs):

a.

Written
evaluations
of
changes
in
the
facility
as
described
in
the
Defueled
Safety

Analysis
Report,
of
changes
in
procedures
as
described
in
the
Defueled
Safety

Analysis
Report,
and
of
tests
or
experiments
not
described
in
the
Defueled
Safety

Analysis
Report,
which
are
completed
without
prior
NRC
approval
under
the

provisions
of
10
CFR
50.59(c)(1).

This
review
is
to
verify
that
such
changes,

tests,
or
experiments
did
not
involve
a
change
in
the
Technical
Specifications
or

require
NRC
approval
pursuant
to
10
CFR
50.59.

Such
reviews
need
not
be

performed
prior
to
implementation.

b.

Proposed
changes
in
procedures,
proposed
changes
in
the
facility,
or
proposed

tests
or
experiments,
any
of
which
involves
a
change
in
the
Technical

Specifications
or
requires
NRC
approval
pursuant
to
10
CFR
50.59.

Matters
of

this
kind
shall
be
reviewed
prior
to
submittal
to
the
NRC.

c.

Proposed
changes
to
Technical
Specifications
or
license
amendments
shall
be

reviewed
prior
to
submittal
to
the
NRC
for
approval.

d.

Violations,
deviations,
and
reportable
events
which
require
reporting
to
the
NRC

in
writing.
Such
reviews
are
performed
after
the
fact.
Review
of
events
covered

under
this
subsection
shall
include
results
of
any
investigations
made
and
the

recommendations
resulting
from
such
investigations
to
prevent
or
reduce
the

probability
of
recurrence
of
the
event.

e.

Written
summaries
of
audit
reports
in
the
areas
specified
in
Section
5.18
of
this

plan
(FormerlyTechnicalSpecification6.5.3).

f.

Any
other
matters
involving
the
plant
which
a
reviewer
deems
appropriate
for

consideration
or
which
is
referred
to
the
independent
reviewers.

7.5

Independent
Safety
Reviewers
or
ISR's
(FormerlyTechnicalSpecification6.5.2.6)
shall

either
have
a
Bachelors
Degree
in
Engineering
or
the
Physical
Sciences
and
five
years
of

professional
level
experience
in
the
area
being
reviewed
or
have
nine
years
of
appropriate

experience
in
the
field
of
his
or
her
specialty.

An
individual
performing
reviews
may

possess
competence
in
more
than
one
specialty
area.
Credit
toward
experience
will
be

given
for
advanced
degrees
on
a
one
for
one
basis
up
to
a
maximum
of
two
years.



Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
30
of
47

APPENDIX
A

10
CFR
71
SUBPART
H
QUALITY
ASSURANCE
PLAN

A1.0
GENERAL

A1.1
This
Appendix
to
the
Decommissioning
Quality
Assurance
Plan
addresses
the
Quality

Assurance
(QA)
requirements
applicable
to
packaging,
preparation
for
shipment,

transportation
of
licensed
material
in
excess
of
a
Type
A
quantity.
It
is
designed
to
assure

compliance
with
10
CFR
71,
Subpart
H,
and
the
applicable
transportation
regulations

contained
in
49
CFR.

This
Appendix
discusses
the
application
of
the
Quality
Assurance
Criteria
of
10
CFR,

Part
71,
Subpart
H,
as
they
relate
to
Important
to
Safety
activities
performed
by
TMI-2

Solutions
or
its
contractors
and
subcontractors.
TMI-2
Solutions,
as
the
Licensee
for

Three
Mile
Island
Unit
2,
retains
the
responsibility
for
the
overall
effectiveness
of
this

plan.
This
plan
will
be
implemented
through
the
use
of
controlled
and
approved

procedures,
and
will
be
applied
in
a
graded
approach
to
an
extent
that
is
commensurate

with
the
quality
assurance
requirements
importance
to
safety.

A1.2
Each
certificate
holder
shall
be
responsible,
as
specified
in
procurement
documents,
for

satisfying
the
quality
assurance
requirements
that
apply
to
components
of
packaging

subject
to
10
CFR,
Part
71,
Subpart
H.
TMI-2
Solutions
is
simultaneously
responsible
for

these
quality
assurance
requirements
through
the
oversight
of
contractors
and

subcontractors.
TMI-2
Solutions
is
also
responsible
for
satisfying
the
quality
assurance

requirements
that
apply
to
its
use
of
packaging
for
the
shipment
of
licensed
material

subject
to
10
CFR,
Part
71,
Subpart
H.

A2.0
RESPONSIBILITIES

The
TMI-2
Solutions
Cognizant
Officer
is
responsible
for
establishing
the
facility's
quality

assurance
policies,
goals,
and
objectives,
and
retains
overall
responsibility
and
the
authority
for

implementing
those
policies.

A3.0
ORGANIZATION
(10
CFR
71.103)

The
TMI-2
Solutions
organization
responsible
for
Three
Mile
Island
Unit
2
is
described
in

Section
5.1
of
the
base
Decommissioning
Quality
Assurance
Plan.
Any
member
of
the
staff

performing
activities
that
are
under
the
purview
of
this
plan
has
the
responsibility
and
authority

to
stop
work
and
the
delivery
or
installation
of
nonconforming
materials
related
to
this
Appendix

when
quality
or
safety
concerns
are
identified.
These
individuals
will
have
direct
access
to
a

level
of
TMI-2
Solutions
Management
that
can
ensure
proper
evaluation
and
correction
of

conditions
adverse
to
quality.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
31
of
47

APPENDIX
A

10
CFR
71
SUBPART
H
QUALITY
ASSURANCE
PLAN

A4.0
QUALITY
ASSURANCE
PROGRAM
(10
CFR
71.105)

A4.1
This
Appendix
constitutes
a
program
to
supplement
the
base
Decommissioning
Quality

Assurance
Plan
for
performance
of
activities
associated
with
the
packaging
and

transportation
of
radioactive
material
under
the
provisions
of
10
CFR
71.

A4.2
This
Appendix
recognizes
the
description
of
"important
to
safety"
as
used
in
NRC

Regulatory
Guide
7.10,
"Establishing
Quality
Assurance
Programs
for
Packaging
used
in

Transport
of
Radioactive
Material",
Revision
3.
The
Regulatory
Guide
states,
"Forthe

purposesofthisRegulatoryGuide,structures,systems,andcomponentsimportantto

safetymeansthefeaturesofaTypeBorfissilematerialpackagethatareintendedto(1)

maintaintheconditionsrequiredtosafelytransportthepackagecontents;(2)prevent

damagetothepackageduringtransport;or(3)providereasonableassurancethatthe

radioactivecontentscanbereceived,handled,transported,andretrievedwithoutundue

risktothehealthandsafetyofthepublicortheenvironment".

a.
This
Appendix
will
be
applied
to
activities
related
to
the
criteria
of
10
CFR
71

involving
SSCs
meeting
this
definition
of
Important
to
Safety.
Important
to
Safety
is

further
categorized
into
one
of
three
quality
assurance
classification
categories
(ITS-A,
ITS-B,
or
ITS
C)
or
Not
Important
to
Safety
(NITS)
consistent
with
the
NRC

guidance
given
in
NUREG/CR-6407,
Classification
of
Transportation
Packaging

and
Dry
Spent
Fuel
Storage
System
Components
According
to
Importance
to
Safety

and
Regulatory
Guide
7.10.
These
classifications
will
provide
the
bases
for
applying

the
criteria
to
activities
conducted
by
TMI-2
Solutions,
including
the
oversight
of

contractors
and
subcontractors,
in
a
graded
approach
to
an
extent
commensurate
with

the
SSCs
importance
to
safety,
as
necessary,
to
ensure
conformance
with
the

approved
design.

A4.3
Personnel
performing
activities
affecting
quality
will
receive
training
and
indoctrination

on
this
plan
and
associated
procedures
as
necessary
to
ensure
that
suitable
proficiency
is

achieved
and
maintained.

A4.4
Management
of
organizations
responsible
for
implementation
of
the
criteria
in
this

Appendix
shall
regularly
review
the
status
and
adequacy
of
that
part
which
they
are

executing.

A5.0
CHANGES
TO
QUALITY
ASSURANCE
PROGRAM
(10
CFR
71.106)

Any
changes
to
this
Appendix
will
be
considered
a
change
to
the
Decommissioning
Quality

Assurance
Plan
and
will
be
reviewed
and
approved
in
the
same
manner
as
the
base
plan
and
will

be
submitted
to
the
Nuclear
Regulatory
Commission
in
compliance
with
the
requirements
of
10

CFR
71.106.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
32
of
47

APPENDIX
A

10
CFR
71
SUBPART
H
QUALITY
ASSURANCE
PLAN

A6.0
DESIGN
CONTROL
(10
CFR
71.107)

A6.1
Design
controls
for
transportation
packaging
will
be
subcontracted
to
the
respective

certificate
holders.
Consequently,
the
criteria
of
10
CFR
Part
71.107
are
not
applicable.

Assurance
that
the
design
control
activities
were
accomplished
under
control
of
an
NRC

approved
QA
Program
is
required
and
will
be
accomplished
by
requiring
the
certificate

holders
to
submit
documented
proof
of
package
design
under
an
NRC
approved
QA

Program.
Documented
proof
will
include
NRC
Certificates
of
Compliance,
controlled

distribution
copies
of
the
vendor's
transportation
cask
manuals.
Documents
will
be
kept

on
file
at
the
Three
Mile
Island
facility.

A7.0
PROCUREMENT
DOCUMENT
CONTROL
(10
CFR
71.109)

A7.1
TMI-2
Solutions
will
procure
transportation
packaging
from
contractors
and

subcontractors
with
QA
Programs
that
comply
with
10
CFR
71.
Certificate
holders
will

be
required
to
provide
evidence
of
NRC
approval
of
their
respective
QA
Programs.

A7.2
Procurement
processes
shall
be
specified
by
written
procedures
which
provide
for
review

by
appropriate
management
personnel
and
require
that
procurement
documents
be
subject

to
a
quality
review
to
assure
applicable
regulatory
requirements,
design
bases,
quality

assurance,
and
other
requirements
which
are
necessary
to
assure
adequate
quality
are

included
or
referenced
in
the
documents
prior
to
execution.
Provisions
for
assuring

sufficient
verification
of
quality
requirements
is
provided
by
the
contractors
and

subcontractors
shall
also
be
included.

a.
If
considered
necessary
by
TMI-2
Solutions
Management
due
to
the
unique

characteristics
or
importance
to
safety
of
an
SSC
or
service,
procurement
documents

shall
specify
that
TMI-2
Solutions
or
a
designee
has
the
right
to
access
the
supplier

and
lower-tier
supplier
facilities
for
inspection,
audit,
or
review
of
records.

A7.3
The
certificate
holders
will
be
required
to
supply
appropriate
certifications
verifying
that

the
items
and
services
being
supplied
were
controlled
under
an
approved
NRC
QA

Program.
All
pertinent
documentation
requirements
shall
be
determined
and
furnished
by

the
contractors
and
subcontractors,
including
available
documentation
that
provides

objective
evidence
of
the
quality
of
the
items
and
services
and
identification
of
the
type

of
verification
activities
required
during
use
and
maintenance.
Records
which
contractors

and
subcontractors
shall
retain,
maintain,
and
control
and
those
which
shall
be
delivered

prior
to
use
or
delivery
of
an
item
or
service
will
be
identified.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
33
of
47

APPENDIX
A

10
CFR
71
SUBPART
H
QUALITY
ASSURANCE
PLAN

A7.4
Procurement
of
licensed
package
services
will
also
entail
procurement
of
any
required

Important
to
Safety
replacement
parts
directly
from
the
package
certificate
holder,
as

applicable.
The
certificate
holder
will
be
contractually
required
to
provide
any
necessary

certification
documents
for
purchased
material
and
to
provide
the
services
of
qualified

staff
to
either
perform
or
oversee
the
installation
of
any
replacement
parts.
The
certificate

holder
will
perform
such
activities
under
the
control
of
certified
procedures
and
in

accordance
with
the
certificate
holders
NRC
approved
QA
Program.
Documentation
of

the
performance
of
these
activities
will
be
provided
to
TMI-2
Solutions
for
retention
in

project
files.

A8.0
INSTRUCTIONS,
PROCEDURES,
AND
DRAWINGS
(10
CFR
71.111)

A8.1
Activities
that
are
Important
to
Safety
will
be
prescribed
by
documented
instructions,

procedure,
or
drawings
of
a
type
appropriate
to
the
circumstances.
These
instructions,

procedures,
or
drawings
shall
be
followed
and
complied
with
verbatim
for
all
activities

performed
at
Three
Mile
Island
Unit
2.
Procedures
prepared
by
TMI-2
Solutions
shall
be

written,
reviewed,
and
approved
in
accordance
with
administrative
procedures
in
effect

for
Three
Mile
Island
Unit
2.
Administrative
program
procedures
shall
identify
the

approving
organization
for
instructions
and
procedures
and
shall
require
that
changes
to

approved
procedures
be
reviewed
and
approved
by
the
same
organizations
that
approved

the
original
document.
When
appropriate,
TMI-2
Solutions
will
invoke
the
use
of

certificate
holder
procedures
that
have
been
prepared,
reviewed,
and
approved
in

accordance
with
the
certificate
holders
NRC
approved
QA
Program.
Instructions,

procedures,
and
drawings
shall
contain
sufficient
quantitative
and
qualitative
acceptance

criteria
to
allow
determination
that
activities
prescribed
have
been
satisfactorily

accomplished.
The
requirements
for
preparation,
review,
independent
review,
and

approval
of
instructions,
procedures,
and
drawings
specified
in
the
base

Decommissioning
Quality
Assurance
Plan
shall
also
apply
to
10
CFR
71
related

activities.

A8.2
If
repair,
rework,
or
maintenance
is
required
to
be
performed
on
an
NRC
Licensed

transportation
package
at
the
Three
Mile
Island
facility,
the
work
will
be
performed
either

by
or
under
the
direct
supervision
of
qualified
certificate
holders
staff.
Such
work
will
be

performed
in
accordance
with
written
procedures
that
have
been
prepared,
reviewed,
and

approved
in
accordance
with
the
requirements
of
the
certificate
holders
NRC
approved

QA
Program.
The
work
will
be
coordinated
with
certificate
holder's
quality
assurance

personnel
to
ensure
that
appropriate
inspection
and
test
points
are
incorporated
in
the

procedure
and
that
effective
repairs
or
rework
have
been
satisfactorily
performed.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
34
of
47

APPENDIX
A

10
CFR
71
SUBPART
H
QUALITY
ASSURANCE
PLAN

A9.0
DOCUMENT
CONTROL
(10
CFR
71.113)

A9.1
Procedures
shall
be
in
effect
to
control
the
issuance
of
documents
such
as
instructions,

procedures,
and
drawings,
and
changes
thereto.
The
requirements
for
review,
approval,

release,
distribution,
and
use
specified
in
the
base
Decommissioning
Quality
Assurance

Plan
shall
also
apply
to
10
CFR
71
related
activities.

A9.2
Documents
prepared
by
certificate
holders
shall
be
prepared
and
issued
in
accordance

with
the
certificate
holder's
NRC
approved
Quality
Assurance
Program.
Certificate

holder
documents
used
at
the
Three
Mile
Island
facility
shall
be
processed
in
such
a

manner
that
TMI-2
Solutions
staff
and
their
contractor
staff
are
on
controlled
distribution

for
such
documents,
and
there
is
an
established
method
for
site
staff
to
verify
that
the

documents
in
use
are
current
and
properly
approved.

A10.0
CONTROL
OF
PURCHASED
MATERIAL,
EQUIPMENT,
AND
SERVICES

(10
CFR
71.115)

Procedures
will
be
in
effect
to
verify
that
purchased
Important
to
Safety
material,
equipment,
and

services,
whether
purchased
directly
or
through
contractors
and
subcontractors,
conforms
to
the

requirements
specified
in
the
procurement
documents.
The
procedures
will
include
provisions,
as

appropriate,
for
supplier
evaluation
and
selection,
review,
and
acceptance
of
objective
evidence

of
quality
furnished
by
the
supplier,
source
surveillance
and
inspection,
and
receipt
inspection

upon
delivery,
as
appropriate.
The
processes
shall
use
a
graded
approach
applying
considerations

of
a
material's,
component's
or
services
importance
to
safety,
complexity,
and
history
of

previous
use.
Applying
the
graded
approach,
the
procurement
documents
may
also
require
TMI-2
Solutions
review
and
approval
of
contractor
and
subcontractor
drawings,
procedures,

manufacturing
plans,
and
Quality
Assurance
Plans
as
needed.

A10.1
When
procurement
documents
require
implementation
of
a
QA
program
compliant
with

10
CFR
71,
Subpart
H,
the
adequacy
of
supplier's
QA
program
shall
be
verified
and,
if

appropriate,
the
suppliers
adherence
to
their
approved
QA
program
shall
be
verified.

This
may
be
accomplished
by
direct
audit
of
the
supplier,
in-process
surveillance
of
the

supplier's
documented
program,
or
by
review
of
audit
history
performed
by

EnergySolutions
or
an
approved
third-party
source.
Effectiveness
of
the
control
of
quality

by
contractors
and
subcontractors
will
be
assessed
at
intervals
consistent
with
the

importance,
complexity,
quantity
of
the
product
or
services,
and
performance
history.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
35
of
47

APPENDIX
A

10
CFR
71
SUBPART
H
QUALITY
ASSURANCE
PLAN

A10.2
Documentary
evidence
that
the
material,
equipment
and
services
conform
to
the

procurement
documents,
including
identification
of
the
specific
requirements
met
by
the

purchased
material,
equipment
and
services,
shall
be
verified
and
available
prior
to

installation
or
use
of
the
item
or
service.
These
records
shall
be
retained
for
the
life
of
the

transportation
package
to
which
it
applies
in
accordance
with
Section
A20.0
of
this

Appendix.

A10.3
For
transportation
packages,
such
documentation
should
be
referenced
in
the
certificate

of
compliances,
relate
to
the
use
and
maintenance
of
the
package,
and
identify
the

necessary
actions
to
be
taken
prior
to
delivery
of
the
package
to
a
carrier
for

transportation.

A11.0
IDENTIFICATION
AND
CONTROL
OF
MATERIALS,
PARTS,
AND

COMPONENTS
(10
CFR
71.117)

A11.1
Identification
and
control
of
Important
to
Safety
materials,
parts,
and
components
during

fabrication
of
transportation
components
will
be
performed
by
the
contractors
and

subcontractors
under
the
controls
of
the
certificate
holders
NRC
approved
QA
programs.

These
same
controls
may
be
extended
to
installation
and
use,
as
applicable,
when

included
in
the
contractors
or
subcontractors
scope
of
services.

A11.2
TMI-2
Solutions
use
of
certificate
holder
supplied
components
will
be
controlled
to

ensure
that
the
components
unique
identification
(serial
number
or
other
appropriate

identifier)
is
maintained
on
the
item
or
on
records
traceable
to
the
item
throughout

installation
and
use.
The
unique
identification
shall
be
verified
and
traceable
to
the

documentary
evidence
of
conformance
addressed
in
Section
A10.0
of
this
Appendix.

Verification
shall
also
include
conformance
of
the
components
unique
identification
to

design
documents
and/or
Certificate
of
Conformance,
as
applicable,
to
prevent
the
use
of

incorrect
components.

A11.3
Replacement
parts
for
components
will
controlled
in
the
same
manner
as
described
above

for
components,
and
will
include
measures
to
preclude
use
of
items
whose
shelf
life
or

operation
times
have
expired.

A11.4
Nonconforming
material,
parts,
and
components
will
be
identified
and
segregated
to

prevent
the
use
of
incorrect
or
defective
items
in
accordance
with
Section
A18.0
of
this

Appendix.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
36
of
47

APPENDIX
A

10
CFR
71
SUBPART
H
QUALITY
ASSURANCE
PLAN

A12.0
CONTROL
OF
SPECIAL
PROCESSES
(10
CFR
71.119)

Special
process
controls
for
transportation
packaging
will
be
subcontracted
to
the
respective

certificate
holders.
Those
processes
will
be
controlled
by
the
certificate
holder
and
will
be

performed
in
accordance
with
the
requirements
of
the
certificate
holders
NRC
approved
QA

Program
and
not
performed
by
TMI-2
Solutions.

A13.0
INSPECTION
(10
CFR
71.121)

A13.1
Inspection
of
Important
to
Safety
transportation
materials,
parts,
components,
and

systems
during
fabrication
will
be
performed
by
the
contractors
and
subcontractors
under

the
controls
of
the
certificate
holders
NRC
approved
QA
programs.
These
same
controls

may
be
extended
to
installation
and
use,
as
applicable,
when
included
in
the
contractors

or
subcontractors
scope
of
services.

A13.2
Appropriate
inspections
for
fabrication,
installation,
or
use
of
Important
to
Safety
SSC

performed
by
TMI-2
Solutions,
its
contractors,
or
subcontractors
in
accordance
with
this

plan,
will
be
performed
to
verify
conformance
with
the
documented
instructions,

procedures,
and
drawings
for
accomplishing
the
activity.
The
frequency
and
degree
of

inspection
shall
be
determined
through
application
of
the
graded
approach
considering

the
safety
significance
and
complexity
of
the
SSC
or
activity.

a.
Inspection
requirements
will
be
prescribed
in
procedures
or
work
control
documents

controlling
the
activity.
Inspection
planning
will
be
performed
to
identify:

• 
  Applicable
  instructions,
  procedures
  and
  drawings
  required
  for
  the
  scope
  of
  the
  

inspection;

• 
  Characteristics
  and
  activities
  to
  be
  inspected;
  
  

• 
  Appropriate
  work
  operations
  requiring
  examinations,
  measurements,
  or
  tests
  

where
necessary
to
assure
quality,
including
required
hold
points;

• 
  Prerequisites
  for
  performing
  the
  inspections,
  including
  any
  special
  equipment
  

required
to
perform
the
inspection;

• 
  Inspection
  technique/method;
  and
  
  

• 
  Acceptance
  criteria
  and
  tolerances
  specified
  in
  applicable
  design
  documents,
  

procedures,
or
approved
vendor
documents.

b.
When
direct
inspection
is
not
possible
or
practical,
alternate
methods
shall
be

implemented.
These
may
include
review
of
processes,
equipment,
and
personnel

qualification,
and
monitoring
of
the
process
parameters
or
activity.
A
combination
of

direct
and
indirect
methods
shall
be
used
when
one
method
alone
is
not
adequate
or

sufficient
to
determine
compliance
with
applicable
acceptance
criteria.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
37
of
47

APPENDIX
A

10
CFR
71
SUBPART
H
QUALITY
ASSURANCE
PLAN

A13.3
If
mandatory
inspection
hold
points
are
to
be
established,
which
require
witnessing
or

inspection
and
beyond
which
work
shall
not
proceed
without
prior
consent
are
required,

the
specific
hold
points
shall
be
indicated
in
the
work
control
documents
and
the
work

control
documents
shall
be
reviewed
prior
to
implementation
by
the
organization
required

to
perform
the
inspection.

A13.4
Inspections
shall
be
performed
by
individuals
other
than
those
who
performed
the
activity

being
inspected.
Procedures
will
be
in
effect
to
ensure
that
inspectors
are
qualified
in

accordance
with
applicable
codes,
standards,
and
TMI-2
Solutions
training
programs,

and
that
qualifications
and
certifications
of
inspection
personnel
are
kept
current.

A13.5
Visual
inspections
by
designated
personnel
will
be
performed
upon
receipt
of
Important

to
Safety
transportation
packaging
to
ensure
compliance
with
procurement

documentation.
The
criteria
for
acceptance
of
each
of
these
inspections
and
action
to
be

taken
if
noncompliance
is
encountered
will
be
specified
in
written
documents
conforming

to
the
Vendor's
Cask
Manual.

a.
These
visual
inspections
should
include
an
inspection
of
the
following:

• 
  Surface
  conditions
  

• 
  Weld
  and
  structural
  integrity
  

• 
  Condition
  of
  flange
  or
  sealing
  faces
  

• 
  Gaskets
  and
  seals
  

• 
  Gauges,
  rupture
  disks,
  valves,
  pressure
  relief
  devices
  

• 
  Condition
  of
  tie-down
  members
  

• 
  Labeling
  and
  marking
  

• 
  Leak-tightness
  of
  the
  packaging
  

b.
Final
inspections
will
be
performed
with
a
checklist
to
verify
as
a
minimum
that
the

following
items
are
complied
with:

• 
  Transportation
  packages
  are
  properly
  assembled
  

• 
  All
  shipping
  papers
  are
  properly
  completed
  

• 
  Transportation
  packages
  are
  conspicuously
  and
  durably
  marked
  as
  required
  by
  

DOT
regulations.

• 
  Individuals
  designated
  by
  TMI-2
  Solutions
  Management
  have
  given
  authorization
  

for
shipment
of
the
transportation
package.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
38
of
47

APPENDIX
A

10
CFR
71
SUBPART
H
QUALITY
ASSURANCE
PLAN

c.
The
inspection
program
shall
ensure
adequate
maintenance
of
packages.
The

certificate
holder
of
the
transportation
package
shall
identify
all
items
to
be

maintained,
criteria
for
acceptability
or
replacement,
and
the
frequencies
of
inspection

assigned
to
each
item
during
use
of
the
transportation
package.

A14.0
TEST
CONTROL
(10
CFR
71.123)

A14.1
Testing
of
Important
to
Safety
transportation
materials,
parts,
components,
and
systems

during
fabrication
will
be
performed
under
the
controls
of
the
certificate
holders
NRC

approved
QA
programs.
These
same
controls
may
be
extended
to
installation
and
use,
as

applicable,
when
included
in
the
contractors
or
subcontractors
scope
of
services.

A14.2
Testing
of
Important
to
Safety
structures,
systems
and
components
performed
by
TMI-2

Solutions,
its
contractors,
or
subcontractors
in
accordance
with
this
plan,
will
be

performed
under
a
test
program
that
will
demonstrate
that
the
items
will
perform

satisfactorily
in
service.
The
testing
will
be
identified
and
performed
in
accordance
with

written
test
plans
that
incorporate
requirements
and
acceptance
limits
in
the
Certificate
of

Compliance.

a.
Test
requirements
will
be
prescribed
in
test
plans
that
are
incorporated
into

procedures
or
work
control
documents
controlling
the
activity.
Test
plans
will
include

provisions
to
ensure:

• 
  All
  prerequisites
  for
  the
  given
  test
  are
  specified
  (suitable
  environmental
  testing
  

conditions,
test
instrumentation
is
available
and
used,
personnel
qualifications);

• 
  Sufficient
  instruction
  for
  the
  performance
  of
  the
  test,
  acceptance/rejection
  criteria
  

and
limits;
and

• 
  Test
  results
  are
  documented
  and
  evaluated
  to
  ensure
  that
  the
  acceptance
  criteria
  

has
been
satisfied

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
39
of
47

APPENDIX
A

10
CFR
71
SUBPART
H
QUALITY
ASSURANCE
PLAN

A14.3
The
transportation
package
certificate
holder
shall
provide
for
a
test
program
to

demonstrate
that
packaging
components
will
perform
satisfactorily
in
service.
The
testing

will
be
identified
and
performed
in
accordance
with
written
test
procedures
that
are

prepared,
reviewed,
and
approved
in
accordance
with
the
certificate
holders
NRC

approved
QA
Program.

a.
These
written
procedures
will
incorporate
the
acceptance
limits
contained
in
the

package
approval,
provisions
for
assuring
that
all
prerequisites
for
a
given
test
have

been
met,
that
adequate
test
instrumentation
is
available
and
used
and
that
the
test
is

performed
under
suitable
environmental
conditions.
Test
results
shall
be
documented

and
evaluated
to
assure
that
test
requirements
have
been
satisfied
prior
to
delivering

packages
for
transport
to
a
carrier.
The
following
items
shall
be
included
in
typical

tests:

• 
  Structural
  integrity
  

• 
  Leak
  tightness
  of
  the
  package,
  gaskets,
  and
  seals
  

• 
  Shielding
  integrity
  

b.
Records
of
tests
shall
be
retained
in
the
Project
files.

A15.0
CONTROL
OF
MEASURING
AND
TEST
EQUIPMENT
(10
CFR
71.125)

A15.1
Control
of
measuring
and
test
equipment
(M&TE)
used
in
activities
affecting
Important

to
Safety
transportation
materials,
parts,
components,
and
systems
during
fabrication
will

be
performed
by
the
contractors
and
subcontractors
under
the
controls
of
the
certificate

holders
NRC
approved
QA
programs.
These
same
controls
may
be
extended
to

installation
and
use,
as
applicable,
when
included
in
the
contractors
or
subcontractors

scope
of
services.

A15.2
M&TE
used
in
activities
affecting
Important
to
Safety
structures,
systems,
and

components
by
TMI-2
Solutions,
its
contractors,
or
subcontractors
in
accordance
with

this
plan,
will
be
controlled
under
an
M&TE
program.
Written
procedures
will
be

established
requiring
that
all
instruments,
gauges,
and
other
measuring
and
testing

devices
are
properly
controlled,
calibrated
(if
necessary),
and
adjusted
at
specific
times
to

maintain
accuracy
within
necessary
limits.
This
includes
M&TE
used
for
maintenance
of

Important
to
Safety
items.
M&TE
will
be
tagged
or
labeled
to
indicate
the
last
calibration

date
and
the
calibration
due
date.
All
calibration
test
data
shall
be
maintained
with
project

records
or
will
be
readily
traceable
to
vendors
records.

When
M&TE
is
found
to
be
out
of
calibration,
the
validity
of
previous
inspections,
tests,

and
calibrations
where
the
item
had
been
used
since
its
last
calibration
shall
be
evaluated.

Such
evaluations
shall
be
documented
and
corrective
actions
shall
be
taken,
as
necessary.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
40
of
47

APPENDIX
A

10
CFR
71
SUBPART
H
QUALITY
ASSURANCE
PLAN

A16.0
HANDLING,
STORAGE,
AND
SHIPPING
CONTROL
(10
CFR
71.127)

A16.1
Handling,
storage,
and
shipping
controls
for
Important
to
Safety
transportation
materials,

parts,
components,
and
systems
during
fabrication
will
be
performed
by
the
contractors

and
subcontractors
under
the
controls
of
the
certificate
holders
NRC
approved
QA

programs.
These
same
controls
may
be
extended
to
installation
and
use,
as
applicable,

when
included
in
the
contractors
or
subcontractors
scope
of
services.

A16.2
Handling,
storage,
and
shipping
controls
for
Important
to
Safety
transportation
materials,

parts,
components,
systems,
and
structures
by
TMI-2
Solutions,
its
contractors,
or

subcontractors
in
accordance
with
this
plan,
will
be
controlled
by
written
procedures.

These
procedures
will
establish
measures
that
will
control
the
handling,
storage,
shipping,

cleaning,
and
preservation
of
materials
and
equipment
to
prevent
damage
or
deterioration

and
to
ensure
adherence
to
any
specific
conditions
identified
in
the
certificate
holders

certificate
of
compliance.
When
necessary
for
particular
items,
special
protective

environments,
such
as
inert
gas
atmosphere,
and
specific
humidity
and
temperature
limits

must
be
specified
and
provided.

a.
The
following
additional
measures
will
be
taken
when
handling
and
storing

transportation
packages:

• 
  If
  a
  package
  requires
  special
  handling
  and
  lifting
  equipment
  such
  equipment
  will
  

be
provided
by
or
specified
by
the
certificate
holder
and
will
be
documented
in

the
certificate
holders
handling
procedures
supplied
with
the
package.
Site

personnel
will
review
and
verify
that
appropriate
certification
documents
are

provided
prior
to
the
use
of
such
special
lifting
equipment.

• 
  Special
  handling
  or
  storage
  provisions
  for
  packages
  will
  be
  as
  specified
  in
  the
  

certificate
holders
NRC
approved
QA
Program
and
controlled
cask
manuals.

Controlled
distribution
copies
of
these
documents
will
be
provided
to
the
site,
will

be
verified
current,
and
will
be
used
by
site
personnel.

• 
  Special
  protective
  environments
  will
  be
  verified
  and
  monitored
  at
  the
  site
  where
  

required.

• 
  All
  conditions
  identified
  in
  a
  certificate
  of
  compliance
  when
  unloading
  packaging
  

will
be
adhered
to.
This
will
be
controlled
by
written
instructions
and
procedures

that
will
implement
these
requirements
as
specified
in
the
controlled
distribution

copies
of
the
certificates
of
compliance
and
cask
manuals
provided
by
the

certificate
holder.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
41
of
47

APPENDIX
A

10
CFR
71
SUBPART
H
QUALITY
ASSURANCE
PLAN

A16.3
When
preparing
a
transportation
package
for
shipment
the
following
applicable
measures

will
be
taken:

a.
Specified
operations,
inspections,
and
tests
will
be
completed
on
site
in
accordance

with
certificate
holders
documents
and
documentary
evidence
of
their
completion

retained
in
project
files.

b.
NRC
and
DOT
requirements
will
be
verified
to
have
been
satisfied
prior
to
delivery

to
a
carrier.

c.
Necessary
shipping
papers
will
be
prepared
and
approved
prior
to
delivery
of
package

to
a
carrier.

A17.0
INSPECTION,
TEST
AND
OPERATING
STATUS
(10
CFR
71.129)

A17.1
Inspection,
test,
and
operating
status
for
Important
to
Safety
transportation
materials,

parts,
components,
and
systems
during
fabrication
will
be
controlled
by
the
contractors

and
subcontractors
under
the
controls
of
the
certificate
holders
NRC
approved
QA

programs.
These
same
controls
may
be
extended
to
installation
and
use,
as
applicable,

when
included
in
the
contractors
or
subcontractors
scope
of
services.

A17.2
Inspection,
test,
and
operating
status
for
Important
to
Safety
items
of
packaging
by
TMI-2
Solutions,
its
contractors,
or
subcontractors
in
accordance
with
this
plan,
will
be

controlled
by
written
procedures.

a.
The
procedures
will
require
the
status
of
inspections
and
tests
to
be
indicated
by

the
use
of
markings
such
as
stamps,
tags,
labels,
checklists,
written
traveler

documents,
or
other
suitable
means
that
provide
for
the
identification
of
items
that

have
satisfactorily
passed
required
inspections
and
tests.

b.
Additionally,
the
procedures
will
require
the
identification
of
the
operating
status

of
components
of
the
packaging
such
as
tagging
valves.

c.
Application
and
removal
of
such
status
indicators
shall
be
specified
and

controlled
where
necessary
to
preclude
inadvertent
by-passing
of
such
inspections

and
tests,
and
to
prevent
inadvertent
operation.

A18.0
CONTROL
OF
NONCONFORMING
MATERIALS,
PARTS,
OR
COMPONENTS

(10
CFR
71.131)

Procedures
shall
be
in
effect
to
control
nonconforming
Important
to
Safety
materials,
parts,
and

components
in
order
to
prevent
inadvertent
use
or
installation.
The
requirements
specified
for
the

control
of
nonconforming
materials,
parts
or
components
in
the
base
Decommissioning
Quality

Assurance
Plan
shall
also
apply
to
10
CFR
71
related
activities.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
42
of
47

APPENDIX
A

10
CFR
71
SUBPART
H
QUALITY
ASSURANCE
PLAN

Nonconforming
items
will
be
segregated
and
labeled
in
such
a
manner
to
prevent
inadvertent
use

until
proper
disposition
is
completed.
Nonconforming
items
shall
be
reviewed
and
accepted,

rejected,
repaired,
or
reworked
in
accordance
with
documented
procedures.
The
acceptability
of

nonconforming
items
after
designated
repair
or
rework
will
be
verified
by
designated
personnel

by
re-inspecting
or
retesting
the
item
against
the
original
requirements.
Affected
organizations

will
be
notified
of
the
status
of
nonconforming
items.
Records
of
such
inspections,
evaluations,

and
re-inspections
and
disposition
will
be
retained
in
project
files.

Nonconforming
conditions
will
be
evaluated
and
reported
in
accordance
with
10
CFR
21
and

10
CFR
71.95.

A19.0
CORRECTIVE
ACTION
(10
CFR
71.133)

Procedures
will
be
in
effect
and
implemented
to
ensure
that
conditions
adverse
to
quality,
such
as

failures,
malfunctions,
deficiencies,
deviations,
defects
in
material
and
equipment,
and

nonconformances
are
promptly
identified
and
corrected.
In
the
case
of
a
significant
condition

adverse
to
quality,
the
procedures
shall
ensure
that
the
cause
of
the
condition
is
determined
and

corrective
action
taken
to
preclude
repetition.
The
identification
of
the
significant
condition

adverse
to
quality,
the
cause
of
the
condition,
and
the
corrective
action
taken
shall
be

documented
and
reported
to
appropriate
levels
of
management
for
review.

Conditions
adverse
to
quality
will
be
evaluated
and
reported
in
accordance
with
10
CFR
21
and

10
CFR
71.95.

A20.0
QUALITY
ASSURANCE
RECORDS
(10
CFR
71.135)

A20.1
Sufficient
written
records
shall
be
maintained
to
furnish
evidence
that
Important
to
Safety

activities
have
been
performed
in
accordance
with
the
written
instructions,
procedures,

and
drawings.

a.
The
records
shall
include
the
following:

• 
  

Changes
to
the
quality
assurance
program
as
required
by
Section
A5.0
of
this

Appendix

• 
  

Instructions,
procedures,
and
drawings
required
by
Section
A8.0
of
this

Appendix
to
prescribe
quality
assurance
activities,
and
closely
related
data
such

as
required
qualifications
of
personnel,
procedures,
and
equipment

• 
  

Instructions
or
procedures
that
establish
a
records
retention
program
that
is

consistent
with
applicable
regulations
and
designates
factors
such
as
duration,

location,
and
assigned
responsibility

• 
  

Design,
fabrication,
erection,
testing,
and
maintenance
records

• 
  

Records
of
use

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
43
of
47

APPENDIX
A

10
CFR
71
SUBPART
H
QUALITY
ASSURANCE
PLAN

• 
  

Results
of
reviews,
inspections,
tests,
audits,
monitoring
of
work
performance,

and
material
analysis

• 
  

Maintenance
records

• 
  

Delivery
of
package
to
a
carrier
(including
proof
that
applicable
NRC
and
DOT

requirements
have
been
satisfied).

A20.2
These
records
shall
be
identifiable
and
retrievable
and
shall
document,
when
applicable,

the
identification
of
the
individual
generating
the
document
or
the
data
recorder,
the
type

of
observation,
the
results,
the
acceptability,
and
the
action
taken
in
connection
with
any

deficiencies
noted.

a.
Records
of
calibration
of
measuring
and
test
equipment
and
instrumentation
within

the
scope
of
this
Appendix
shall
contain,
or
be
traceable
to,
documentary
evidence
of

traceability
to
national
standards.

A20.3
Records
stored
for
retention
shall
be
protected
from
damage
or
deterioration
from

environmental
and
catastrophic
events
through
either
protected
storage
areas
or
devices

or
separate
duplicate
file
storage
methods.

A20.4
Shipping
records
required
by
10
CFR
71
and
49
CFR
shall
be
maintained
by
the

shipper/licensee
for
radioactive
material
and
copies
will
be
maintained
by
the
project
in

an
auditable
and
retrievable
manner
in
the
files.

A20.5
Records
Retention

a.
Records
identified
in
Section
A20.1
of
this
Appendix
related
to
Important
to
Safety

transportation
packages
shall
be
retained
for
a
minimum
of
3
years
beyond
the
date

when
the
licensee,
certificate
holder
last
engage
in
the
activity
for
which
the
quality

assurance
program
was
developed.
Any
portions
of
the
quality
assurance
program,

written
procedures
or
instructions
that
are
superseded
shall
be
retained
for
a
minimum

of
3
years
after
it
is
superseded.

• 
  No
  repairs
  or
  modifications
  to
  licensed
  transportation
  packages
  will
  be
  performed
  

by
TMI-2
Solutions.
These
activities
will
be
performed
only
under
the
direct

control
of
the
certificate
holder
and
any
required
records
of
these
activities
will
be

maintained
by
the
certificate
holder
in
accordance
with
the
certificate
holders

NRC
approved
QA
Program.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
44
of
47

APPENDIX
A

10
CFR
71
SUBPART
H
QUALITY
ASSURANCE
PLAN

A21.0
AUDITS
(10
CFR
71.137)

Procedures
shall
be
in
effect
to
ensure
that
a
comprehensive
system
for
planned
and
periodic

audits
to
verify
compliance
with
all
aspects
of
this
Appendix
and
to
determine
the
effectiveness

of
the
program
are
carried
out.
The
requirements
for
planning,
conduct,
review,
and
follow-up
of

audits
specified
in
the
base
Decommissioning
Quality
Assurance
Plan
shall
also
apply
to
10
CFR

71
related
activities.

A21.1
These
audits
shall
be
performed
in
accordance
with
this
Appendix
and
shall
be
performed

at
the
following
frequencies:

a.
For
activities
subject
to
the
requirements
of
10
CFR
71,
Subpart
H,
the
frequency
of

audits
should
be
based
on
each
activitys
importance
to
safety;
however,
each
quality

criterion
should
be
audited
at
least
once
each
year.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
45
of
47

APPENDIX
B

REPORTING
REQUIREMENTS

This
Appendix
to
the
Decommissioning
Quality
Assurance
Plan
addresses
the
requirements

previously
contained
in
Section
6.8
of
the
Technical
Specifications
of
the
NRC
License
Number

DPR-73
that
were
transferred
to
Revision
20
of
this
plan
in
accordance
with
License
Amendment

67.
This
Appendix
applies
to
TMI-2
Solutions
and
TMI-2
Solutions
contractors
for
reporting

requirements.

B.1

Routine
Reports
(FormerlyTechnicalSpecification6.8.1)

In
addition
to
the
applicable
reporting
requirements
of
Title
10,
Code
of
Federal

Regulations,
the
following
reports
shall
be
in
accordance
with
10
CFR
50.4
unless

otherwise
noted.
Some
of
the
reporting
requirements
of
Title
10,
Code
of
Federal

Regulations
are
repeated
below.

B.1.1
Annual
Radiological
Environmental
Operating
Report
(FormerlyTechnical

Specification6.8.1.1)

The
Annual
Radiological
Environmental
Operating
Report
covering
the
operation

of
the
facility
during
the
previous
calendar
year
shall
be
submitted
before
May
1

of
each
year.
The
report
shall
include
summaries,
interpretations,
and
analysis
of

trends
of
the
results
of
the
Radiological
Environmental
Monitoring
Program
for

the
reporting
period.
The
material
provided
shall
be
consistent
with
the
objectives

outlined
in
(1)
the
ODCM
and
(2)
Sections
IV.B.2,
IV.B.3,
and
IV.C
of
Appendix

I
to
10
CFR
Part
50.

B.1.2
Biennial
Reports
(FormerlyTechnicalSpecification6.8.1.4)

A.
Reports
required
on
a
biennial
basis
shall
be
submitted
on
a
frequency
not
to

exceed
once
every
two
years
(24
months).
The
reports
shall
cover
the

activities
of
the
facility
as
described
below
up
to
a
minimum
of
6
months

prior
to
the
date
of
the
filing.

B.
Reports
required
on
a
biennial
basis
shall
include:

1.
All
changes
to
the
PDMS
SAR
during
the
previous
update
and
all
changes

to
the
Decommissioning
SAR
thereafter.

2.
All
changes,
tests,
or
experiments
meeting
the
requirements
of
10
CFR

50.59.

B.2

Special
Reports
(FormerlyTechnicalSpecification6.8.2)

Special
reports
shall
be
submitted
in
accordance
with
10
CFR
50.4
within
the
time
period

specified
for
each
report.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
46
of
47

APPENDIX
B

REPORTING
REQUIREMENTS

B.3

Nonroutine
Reports
(FormerlyTechnicalSpecification6.8.3)

A
report
shall
be
submitted
in
the
event
that
an
Exceptional
Occurrence
as
specified
in

B.4
of
this
Appendix
occurs.
The
report
shall
be
submitted
under
one
of
the
report

schedules
described
below.

B.3.1
Prompt
Reports
(FormerlyTechnicalSpecification6.8.3.1)

Those
events
specified
as
prompt
report
occurrences
shall
be
reported
within
24

hours
by
telephone,
telegraph,
or
facsimile
transmission
to
the
NRC
followed
by
a

written
report
to
the
NRC
within
30
days.

B.3.2
Thirty
Day
Event
Reports
(FormerlyTechnicalSpecification6.8.3.2)

Nonroutine
events
not
requiring
a
prompt
report,
as
described
in
B.3.1
of
this

Appendix,
shall
be
reported
to
the
NRC
either
within
30
days
of
their
occurrence

or
within
the
time
limit
specified
by
the
reporting
requirement
of
the

corresponding
certification
or
permit
issued
pursuant
to
Sections
401
or
402
of
PL

92-500,
the
Federal
Water
Pollution
Control
Act
(FWPCA)
Amendment
of
1972,

whichever
time
duration
following
the
nonroutine
event
shall
result
in
the
earlier

submittal.

B.3.3
Content
of
Nonroutine
Reports
(FormerlyTechnicalSpecification6.8.3.3)

Written
30-day
reports
and,
to
the
extent
possible,
the
preliminary
telephone,

telegraph,
or
facsimile
reports
shall
(a)
describe,
analyze,
and
evaluate
the

occurrence,
including
extent
and
magnitude
of
the
impact,
(b)
describe
the
cause

of
the
occurrence,
and
(c)
indicate
the
corrective
action
(including
any
significant

changes
made
in
procedures)
taken
to
preclude
repetition
of
the
occurrence
and
to

prevent
similar
occurrences
involving
similar
components
or
systems.

Three
Mile
Island
Unit
2
DQAP

Number:
TMI2-QA-PG-001

Revision:
20

Page:
47
of
47

APPENDIX
B

REPORTING
REQUIREMENTS

B.4

Exceptional
Occurrences
(FormerlyTechnicalSpecification6.13)

B.4.1
Unusual
or
Important
Environmental
Events
(FormerlyTechnicalSpecification

6.13.1)

Any
occurrence
of
an
unusual
or
important
event
that
causes
or
could
potentially

cause
significant
environmental
impact
causally
related
with
station
operation

shall
be
recorded
and
reported
to
the
NRC
per
B.3.1
of
this
Appendix.
The

following
are
examples
of
such
events:
excessive
bird
impaction
events
on

cooling
tower
structures
or
meteorological
towers
(i.e.,
more
than
100
in
any
one

day);
onsite
plant
or
animal
disease
outbreaks;
unusual
mortality
of
any
species

protected
by
the
Endangered
Species
Act
of
1973;
fish
kills
near
or
downstream

of
the
site.

B.4.2
Exceeding
Limits
of
Relevant
Permits
(FormerlyTechnicalSpecification6.13.2)

Any
occurrence
of
exceeding
the
limits
specified
in
relevant
permits
and

certificates
issued
by
other
Federal
and
State
agencies
which
are
reportable
to
the

agency
which
issued
the
permit
shall
be
reported
to
the
NRC
in
accordance
with

the
provisions
of
B.3.2
of
this
Appendix.
This
requirement
shall
apply
only
to

topics
of
National
Environmental
Policy
Act
(NEPA)
concern
within
the

requirements
of
the
Station
NPDES
permit
as
related
to
Three
Mile
Island
Unit
2

discharges.