ML14080A192

Final Status Survey Report for the Leslie C. Wilbur Nuclear Reactor Facility at the Worcester Polytechnic Institute

ML14080A192
Person / Time
Site:	05000134
Issue date:	03/11/2014
From:	James Adler TLG Services
To:	NRC/FSME, Worcester Polytechnic Institute
References
W19-1579-005, Rev. 0
Download: ML14080A192 (238)

Text

Document W19-1579-005, Rev. 0 FINAL STATUS SURVEY REPORT for the LESLIE C. WILBUR NUCLEAR REACTOR FACILITY at the WORCESTER POLYTECHNIC INSTITUTE Operating License No. R-61 Docket No. 50-134 Prepared for:

WORCESTER POLYTECHNIC INSTITUTE Prepared by:

TLG Services, Inc. Bridgewater, Connecticut March 2014

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Page iii of viii TLG Services, Inc. TABLE OF CONTENTS SECTION-PAGE EXECUTIVE

SUMMARY

.............................................................................

vi i

1.0 INTRODUCTION

........................................................................................... 1-1 2.0 FACILITY DESCRIPTION RELEVANT TO FSS .................................... 2-1 3.0 REMEDIATION WORK AND RADIOLOGICAL CONDITIONS ENCOUNTERED ........................................................................................... 3-1 4.0 RADIOLOGICAL CONTAMINANTS AND CRITERIA

........................... 4-1 4.1 Radioactive Contaminants Identified During Decommissioning

........ 4-1 4.2 Characteristics of the Radioactive Contaminants that Potentially Could be Present During FSS ...............................................................

4-4 4.3 FSS Criteria ........................................................................................... 4-6

5.0 FINAL

STATUS SURVEY PROCESS ........................................................ 5-1 5.1 FSS Plan and Changes to the Plan ...................................................... 5-1 5.2 FSS Process ........................................................................................... 5-2 5.2.1 Identification of Survey Units

.................................................... 5-2 5.2.2 Classification by Contamination Potential

................................ 5-5 5.2.3 Survey Reference Systems ......................................................... 5-6 5.2.4 Placement of Survey Locations

.................................................. 5-9 5.2.5 Survey Instrumentation ........................................................... 5-10 5.2.6 Survey Techniques ................................................................... 5-12 5.3 Survey Design Packages ..................................................................... 5-21 5.4 Background and Reference Area Measurements

............................... 5-22 5.5 Quality Control .................................................................................... 5-26 5.6 Data Testing to Demonstrate Compliance with License Termination Cr iteria ............................................................. 5-28 6.0 FSS RESULTS ............................................................................................... 6-1 6.1 Overview of FSS Data ........................................................................... 6-1 6.2 Discussion Regarding Survey Units with Detectable Radioactivity ..................................................................... 6-23

7.0 CONCLUSION

S ............................................................................................. 7-1

8.0 BIBLIOGRAPHY

........................................................................................... 8-1 Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Page iv of viii TLG Services, Inc. TABLE OF CONTENTS (Continued)

SECTION-PAGE TABLES 3.1 Radiological Conditions Encountered During Dismantling

........................... 3-5 4.1 Summary of Waste Stream and Pre-FSS Sample Characterization Results ................................................................................. 4-2 4.2 Summary of Analytical Results for Pre-FSS Samples from the Biological Shield ................................................................................ 4-2 4.3 Summary o f the Radionuclides Mixtures Representative of Residual Radioactivity .................................................................................

4-5 4.4 License Termination Screening Values for Building Surface Contamination ............................................................... 4-7 4.5 License Termination Screening Values for Surface Soil

................................ 4-8 5.1 FSS Survey Units and MARSSIM Classifications ......................................... 5-3 5.2 Instrumentation for WPI Final Status Survey

............................................. 5-11 5.3 Direct Beta Reference Area Background Data Summary

............................ 5-23 5.4 Embedded Pipe Detector Background Data

................................................. 5-25 6.1 Total Beta Surface Contamination Direct Measurement Results Summary ............................................................................................. 6-2 6.2 Beta Surface Contamination Scan Results Summary .................................... 6-6 6.3 Removable Beta Surface Contamination Smear Sample Results Summary ........................................................................................... 6-10 6.4 Volumetric Radioactivity Sample Results Summary ................................... 6-14 6.5A Embedded Pipe / Conduit Internal Contamination Assessment Summary: Direct Beta Contamination Scan at Line Ends .......................................................................................... 6-15 6.5B Embedded Pipe / Conduit Internal Contamination Assessment Summary: Removable Beta Contamination at Line Ends - Smear Samples ...................................................................... 6-17 6.5C Embedded Pipe / Conduit Internal Contamination Assessment Summary: Removable Beta Contamination Interior Line Surfaces - Swab Samples ......................................................... 6-19 6.5D Embedded Pipe / Conduit Internal Contamination Assessment Summary: Surface Contamination Interior Line Surfaces - NaI Gamma Logging

............................................. 6-21 6.6 Survey Units with Detectable Radioactivity

................................................ 6-25

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Page v of viii TLG Services, Inc. TABLE OF CONTENTS (Continued)

SECTION-PAGE FIGURES 2.1 Exterior View of Washburn Shops a nd Stoddard Laboratory Building ......................................................................... 2-2 2.2 Ground Floor WPI Reactor Facility Layout Major Features Relevant to FSS ..................................................................... 2-3 2.3 First Floor WPI Reactor Facility Layout Major Features Relevant to FSS ..................................................................... 2-4 2.4 Artist Rendering of Biological Shield / Reactor Pool Structure (Pre-D&D) ................................................................................

2-5 3.1 Locations of Removed Radioactive Items ........................................................

3-3 3.2 Post Removal View of Former Reactor Pool Water Treatment System Area ...................................................................... 3-4 3.3 Mid-Remediation View of Reactor Pool Floor (After Liner Removal Before Concrete Removal) ........................................ 3-6 3.4 Post Remediation View of Remaining Portion of Thermal Column Liner

... . 3-7 3.5 Post Remediation View o f Beam Port Tube / Shutter Housing Area

.............

3-8 3.6 Post Remediation View of Reactor Pool Floor Area ........................................ 3-9 3.7 Post Remediation View of Remaining Beam Port Vent Stub, Water Treatment Return Line and Scupper Drain Outlet Pipe

.................. 3-10 3.8 Post Remediation View of Remaining Thermal Column Vent and Drain Stubs ............................................................................................. 3-11 3.9a Post Remediation View of Remaining Ground Floor Exhaust Duct and Plenum............................................................................. 3-12 3.9b View of First Floor Exhaust Duct Plenum .................................................... 3-13 3.10 Post Remediation Configuration of Biological Shield ................................... 3-14 4.1 Location of Waste Stream Characterization and Pre-FSS Samples

.............. 4-3 5.1 Photographic Example Reference Grid System (Interior Reactor Pool: Survey Units 1.1, 1.3, 1.4 and 1.6)

............................ 5-7 5.2 Example Grid Location Map (Interior Reactor Pool: Survey Units 1.1, 1.3, 1.4 and 1.6) ............................ 5-8 5.3 Photographic Example of Triangular Grid Placement o f Measurement Locations ................................................................................. 5-10 5.4 Photographic View of Surface Beta Contamination Scanning Process ............................................................................................ 5-13 5.5 Photographic View o f the Direct Beta Contamination Measurement Process .................................................................................... 5-14 Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Page vi of viii TLG Services, Inc. TABLE OF CONTENTS (Continued)

SECTION-PAGE FIGURES (Continued)

5.6 Sampling

Locations for SU 1.6: Soil under Former Reactor Core Area .......................................................................................... 5-16 5.7 Photographic View of Concrete Core Sampling (SU 1.4: Reactor Pool Interior Concrete Walls)

............................................ 5-17 5.8 Photographic Views of Aluminum Liner Sampling (SU 1.2: Thermal Column Liner) ................................................................... 5-18 5.9 Photographic View o f an Embedded Pipe Swab Sampling (SU 1.8.4: Drain Line from Beam Port Shutter Housing)

............................ 5-20 5.10 Photographic View of an Embedded Pipe Being Gamma Logged

................ 5-21 5.11 View of Embedded Pipe Reference Background Area .................................. 5-26 APPENDICES A GEL Laboratory Analysis Report for Waste Stream and Pre-FSS Samples (Extracted from the FSS Plan, embedded file)

................................. A-1 B Gross Beta DCGL for Radionuclide Mixture at WPI (Extracted from the FSS Plan) ....................................................................... B-1 C Waste Shipment Manifest May 2013 ......................................................... C-1 D Waste Shipment Manifest December 2013

................................................ D-1 E Decommissioning Work Procedure (DWP)-10, Performance of Final Status Survey (embedded file) ......................................................... E-1 F FSS S F-1 G GEL Laboratory Analysis Reports for FSS Samples (embedded files)

......... G-1 Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Page vii of viii TLG Services, Inc. EXECUTIVE

SUMMARY

The physical dismantling work for the Leslie C. Wilbur Nuclear Reactor Facility was completed in late 2012, after which a Final Status Survey (FSS) Plan was submitted to the USNRC for approval, early in 2013. That plan was approved in June 2013: implementation occurred during the summer and fall of 2013. All resulting radioactive wastes have been removed from the facility.

The results of the FSS indicate that the facility meets the criteria for termination of the R-61 license, as establ2 Decommissioning Plan for the Leslie C. Wilbur Nuclear Reactor Facility , Rev. 2 (DP). All individual radiological measurement determinations made throughout the facility for surface contamination (both total and removable) were found to be less than the criteria established in the DP. Likewise, sample results for concrete, metallic liners, soil and sediment were found to be less than the volumetric radionuclide concentration criteria established in the DP. As such, the WPI reactor facility (License R-61, Docket 50-134) meets the criteria established in its DP for termination of the NRC license.

The report provided herein documents and presents the resulting data, as well as the process and protocols that were used to conduct the FSS. This report was prepared to su The WPI facility that was the subject of the FSS is a two-floor classroom area, that

has 1, 237 m 2 of structural surfaces (walls, floors, ceilings , and the interior and exterior surfaces of the reactor pool and biological shield), a concrete biological shield / pool structure having a mass on the order of 500 tons, plus associated fixtures and equipment such as lights, heating and air conditioning units, an exhaust ventilation system, and associated service piping and electrical conduits. The FSS was conducted using the Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) process, with the facility divided into 43 individual survey units (SU). The overall scope of the FSS included the following determinations: 791 direct beta contamination measurements were made on structural surfaces and remaining equipment. 808 smear samples for removable beta contamination were taken on structures and remaining equipment. 1,017 m 2 of structural surfaces were scanned for evidence of elevated beta surface contamination.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Page viii of viii TLG Services, Inc. The interior surfaces of three heating and air conditioning units, and reactor rooms exhaust ventilation system (duct work, filter plenums, blower and roof surfaces at the discharge point) were subjected to direct and smear sample beta contamination measurements, as well as scans for evidence of elevated beta surface contamination. Eight samples were collected, for laboratory analysis, from the reactor pool and thermal column liners, where those materials were suspected of having residual radioactivity due to neutron activation. Six concrete cores were collected, for laboratory analysis, from locations within the reactor pool/biological shield suspected of potentially being neutron activated. Four soil samples were collected, for laboratory analysis, from beneath the reactor core area. One sediment sample was collected, for laboratory analysis, from an embedded pipe sewer line trap. 210 gamma logging measurements were made within 13 pipes/conduits (11 of which were embedded in massive concrete structures) to provide evidence of the absence of internal contamination and/or to provide a semi-quantitative estimate of interior surface contamination levels. The 13 pipes/conduits were also swab sampled to provide additional evidence of the absence of internal contamination. Twenty-one pipe/conduit ends were checked for evidence of contamination (by beta scanning and smear sampling for removable beta contamination). Remaining supplies, furniture , fixtures and other miscellaneous items within the facility were checked for radioactive contamination.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 1, Page 1 of 1 TLG Services, Inc.

1.0 INTRODUCTION

This report presents the results of the Final Status Survey (FSS) that was conducted for the Leslie C. Wilbur Nuclear Reactor FacilWorcester Polytechnic Institute (WPI), from July 2013 through November 20

13. This FSS was conducted in accordance with a Final Status Survey Plan for the Leslie C. Wilbur Nuclear Reactor Facility at the Worcester Polytechnic Institute, Rev. 0 (FSS Plan) that was submitted to the NRC in January 2013, and approved in June 2013. That FSS Plan was prepared in accordance with the guidelines and recommendation presented in NUREG-1757, Consolidated NMSS Decommissioning Guidance, and NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM).

WPI discontinued routine operation of the Reactor on June 2007, and submitted an application to the NRC for a Possession-only license, which was granted on August 26, 2008. A Decommissioning Plan (DP) was prepared and submitted to the NRC for approval: approval to decommission was granted in December 2010. Fuel was removed from the facility in July 2011, which allowed initiation of decommissioning activities. During the fall of 2011, the reactor pool was drained, the reactor systems

de-energized and the facility cleared of furniture and classroom equipment. Preparations were then made for decommissioning the reactor facility

dismantling commenced in July of 2012. Dismantling work was completed in October of 2012.

O ff-site shipment of the resulting decommissioning wastes occurred in May of 2013. An additional small amount of radioactive waste was shipped off-site in December of 2013. final DP commitment; to perform a FSS to demonstrate that the license termination criteria specified in the DP have been satisfied. This report presents the results of that FSS and supports WPI request for termination of USNRC License R-

61.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 2, Page 1 of 5 TLG Services, Inc. 2.0 FACILITY DESCRIPTION RELEVANT TO FSS The Reactor is located within a portion of the Washburn Shops and Stoddard Laboratories Building, located on the WPI campus, between West and Boynton Streets. Figure 2.1 provides an exterior photograph of the Washburn Shops and Stoddard Laboratories Building. It is a four-story brick building, constructed in the

. The reactor room is located within a portion of the lower two levels of the building. The r walls are generally constructed of concrete block, with interior partition walls constructed of wood framing and sheet rock. The footprint of the reactor room is approximately 90 feet long and 21 feet wide; the ground floor has a ceiling height of approximately 10 feet, and the first floo r (located above the ground floor) has a ceiling height of approximately 12 feet). The first floor of the facility includes a 9 foot by 21 foot office room, and a small (12 foot by 8 foot) reactor tool closet. The remainder of the room is an open area, housing class instruction areas and the upper portion of the reactor biological shield (situated in the center of the room). The ground floor facility includes two partitioned areas a 6 foot by 8 foot dark room and a 6 foot by 16 foot radioactive material storage area and the lower portion of the reactor biological shield. The ground level floor is constructed of poured concrete covered with vinyl tile, and the first floor level is supported by large timbers, with wooden planks covered with vinyl tile (except the reactor office, which is carpeted). Figures 2.2 and 2.3 provide the layouts of the ground and first floors, respectively, indicating major features relevant to the FSS.

The reactor biological shield is an 18 foot by 16 foot by 14 foot high reinforced concrete structure, which contained the reactor pool and experimental facilities (a beam port and a thermal column). The reactor core was located in an 8 foot by 8 foot by 15 foot deep open topped, water-filled pool within the biological shield, lined with 1/4 inch aluminum. The beam port consisted of an 8 inch diameter steel and aluminum tube embedded in the east wall of the biological shield. The Thermal column consisted of a 40 inch square cross-section cavity th at penetrated the West wall of the biological shield, which was lined with 3/4 inch thick aluminum. The Thermal column was filed with graphite blocks prior to the conclusion of remedial activities. The biological shield also contains a number of embedded, small diameter

(1 to 2 inch ID) aluminum pipes that were used for beam port and thermal column ventilation, supplying and returning pool water for filtration and demineralization, and draining pool scuppers. cut-away rendering of the biological shield / pool structure is provided in Figure 2.4. Reactor room air is exhausted via a blower intake is via a duct from the roof blower, with two intakes located on the North wall on each of the reactor room floors. Two drains to the sanitary sewer system are located on the ground floor.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 2, Page 2 of 5 TLG Services, Inc. FIGURE 2.1 EXTERIOR VIEW OF WASHBURN SHOPS AND STODDARD LABORATORY BUILDING

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 2, Page 3 of 5 TLG Services, Inc. FIGURE 2.2 GROUND FLOOR WPI REACTOR FACILITY LAYOUT MAJOR FEATURES RELEVANT TO FSS

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 2, Page 4 of 5 TLG Services, Inc. FIGURE 2.3 FIRST FLOOR WPI REACTOR FACILITY LAYOUT MAJOR FEATURES RELEVANT TO FSS

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 2, Page 5 of 5 TLG Services, Inc. FIGURE 2.4 ARTIST RENDERING OF BIOLOGICAL SHIELD / REACTOR POOL STRUCTURE (PRE-D&D)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 3, Page 1 of 14 TLG Services, Inc.

3.0 REMEDIATION

WORK AND RADIOLOGICAL CONDITIONS ENCOUNTERED The remediation work described in Section 2.3.1.2 of the DP has either been completed or was not required. Based upon in-process radiological surveys conducted during and following the remedial work activities, it is believed that radioactivity has been sufficiently removed from the facility to meet license termination criteria.

Approximately 300 ft 3 of equipment and materials was removed during the remediation process. The resulting wastes were removed from the facility in May of 2013, with the wastes being shipped to the Toxco facility in Oak Ridge, TN. A copy of the receipt manifest for that shipment is provided in Appendix C. An additional 0.66 ft 3 of waste was removed from the facility in December 2013 after conclusion of the site FSS work. This additional waste was generated du ring performance of the FSS, when a small amount radioactive material was discovered while surveying laboratory supplies and equipment. A copy of the receipt manifest for that shipment is provided in Appendix D.

During the decommissioning process, radioactive surface contamination was not found on any structural surface at the reactor facility. Surface contamination within equipment and systems, which had a theoretical potential for being in contact with reactor-produced radioactivity, was found to be extremely limited, with radioactive filter media. The majority of the radioactivity that was encountered during the decommissioning process was found in the various materials that were in close proximity to the reactor core (i.e., those materials subjected to neutron irradiation during operations).

Those materials were located in the bottom of the reactor pool (i.e., the reactor core box and internalsor liner directly under the reactor core box and the underlying concrete bio-shield floor) and the adjacent experimental features (i.e., the beam port shutter housing, thermal column graphite and aluminum thermal column liner). Those items essentially comprised all of the radioactive materials that were found during conduct of the dismantling work. Figure 3.1 shows the locations of the items that were found to contain a majority of the detectable radioactivity. Table 3.1 summarizes the radiological conditions that were encountered during the removal of the radioactive equipment and structural materials.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 3, Page 2 of 14 TLG Services, Inc. The following are the areas within the facility that were dismantled or demolished and removed as waste, where there was a potential for encountering residual radioactivity during the FSS:

1.Reactor pool water treatment system area (see Figure 3.2 for a photographic view of the area from which this system was removed) 2.Portions of the Reactor pool and thermal column aluminum liners (that were neutron activated) a.An approximate 6 foot by 5 foot section of the pool floor liner under the Reactor core area (see Figure 3.3 for a photographic view of this area after the liner was removed, but before removal of the underlying activated concrete).

b.Approximately 3 feet of the 5 foot long thermal column liner (closest to the reactor) (see Figure 3.4 for a post-remediation photographic view of this area).

c.An approximate 2 foot by 3 foot section of the pool wall liner surrounding the beam port tube and shutter housing (see Figure 3.5 for a post-remediation photographic view of this area).

3.Portions of the concrete biological shield (that were neutron activated) a.An approximate 5 foot by 4 foot by 1 foot deep section of the pool floor concrete under the reactor core area , which exposed the underlying soil (see Figure 3.6 for a photographic view of this area).

b.An approximate 1.5 foot by 2 foot by 0.75 foot deep section of the pool wall concrete surrounding the beam port shutter housing area (see Figure 3.5 for a photographic view of this area).

4.Exhaust vent a.Beam port vent pipes, booster fans and duct work (see Figure 3

.7 for a photographic view of the remaining vent stubs embedded in the biological shield). b.Thermal column vent pipes, booster fans and duct work (see Figure 3.8 for a photographic view of the remaining vent stubs embedded in the biological shield). c.Main exhaust duct header (see Figures 3.9a and 3.9b for photographic views of the remaining portions of this duct header on the ground and first floors, respectively).

Figure 3.10, a cross-sectional view of the biological shield, shows the post-remediation configuration of the reactor / biological shield, as it was encountered at the beginning of the FSS. This figure specifically indicates the location of remaining materials and where activated aluminum liners and concrete have been removed.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 3, Page 3 of 14 TLG Services, Inc. FIGURE 3.1 LOCATIONS OF REMOVED RADIOACTIVE ITEMS

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 3, Page 4 of 14 TLG Services, Inc. FIGURE 3.2 POST REMOVAL VIEW OF FORMER REACTOR POOL WATER TREATMENT SYSTEM AREA

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 3, Page 5 of 14 TLG Services, Inc.

TABLE 3.1 RADIOLOGICAL CONDITIONS ENCOUNTERED DURING DISMANTLING AREA RADIOLOGICAL CONDITION Reactor Pool Aluminum Liner (floor and walls)

Activation on the pool floor liner was encountered directly under the reactor core box area. At that location the maximum contact exposure rate was approximately 0.015 mr/hr, with detectable gamma activity observed over approximate ly a 1 m 2 area. Activation on the pool wall liner was encountered within a limited area around the beam port, with a maximum contact exposure rate of approximately 0.018 mr/hr being found inside the aluminum beam port shutter housing.

No other indications of activation or surface contamination were observed on the pool liner. Thermal Column Aluminum Liner The maximum activation was encountered on the liner closest to the reactor core, with a contact exposure rate of approximately 1.5 mr/hr. Indications of activation on the thermal column liner decreased to non

- detectable at about 2.5 feet out from the edge of the reactor core.

Biological Shield Concrete Floor After the activated portion of the aluminum floor liner was removed, activated concrete was encountered below the liner. At that location the maximum contact exposure rate was approximately 0.035 mr/hr, with detectable gamma activity observed over an approxim ate ly 1 m 2 area. Biological Shield Concrete Walls Concrete on the east wall surrounding the embedded beam port shutter housing was found to exhibit maximum contact exposure rates of 0.018 mr/hr (after removal of the beam port shutter housing). There were no other indications of activation or surface contamination elsewhere on the portions of the interior biological shield walls (where the aluminum liner had been removed).

Beam Port Tube Shutter Housing Activation of the beam port structure was encountered within the aluminum beam port shutter housing to which the beam port tube was connected; the shutter housing had a maximum contact exposure rate of approximately 0.

015 mr/hr. Reactor Water Treatment System A maximum contact exposure rate of 0.030 mr/hr was observed on the demineralizer while filled with resin that had been used to purify reactor pool water over the operating life of the reactor. With the exception of used filter cartridges, all internal surfaces of the system exposed by the dismantling process were found to be free of detectable surface contamination.

Ground Floor, East of Biological Shield Centerline Two small radioactive items were (possibly pieces of an irradiated flux wire) were found on the floor near the stairs area. The pool water treatment system was also located in this area, which contained reactor pool water and known radioactive resin. No other radioactive contamination or unexplained elevated radioactivity was encountered in this area.

Radioactive Material Storage Room, Ground Floor Bags of contaminated operational era trash had been stored in this room. Historically radioactive sealed sources were also stored in this area.

First Floor Reactor Room Historically a classroom area. Sealed sample vials containing irradiated materials were handled on a laboratory bench in this area. No other radioactive contamination or unexplained elevated radioactivity was encountered in this area.

Reactor Tool Closet , First Floor Reactor Room This room was used to store tools and supplies for reactor maintenance, and contained a radioactive material disposal sink. However, no radioactive contamination or unexplained elevated radioactivity was encountered in this area.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 3, Page 6 of 14 TLG Services, Inc. FIGURE 3.3 MID-REMEDIATION VIEW OF REACTOR POOL FLOOR (AFTER LINER REMOVAL BEFORE CONCRETE REMOVAL)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 3, Page 7 of 14 TLG Services, Inc. FIGURE 3.4 POST REMEDIATION VIEW OF REMAINING PORTION OF THERMAL COLUMN LINER

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 3, Page 8 of 14 TLG Services, Inc. FIGURE 3.5 POST REMEDIATION VIEW OF BEAM PORT TUBE / SHUTTER HOUSING AREA

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 3, Page 9 of 14 TLG Services, Inc. FIGURE 3.6 POST REMEDIATION VIEW OF REACTOR POOL FLOOR AREA

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 3, Page 10 of 14 TLG Services, Inc. FIGURE 3.7 POST REMEDIATION VIEW OF REMAINING BEAM PORT VENT STUB, WATER TREATMENT RETURN LINE AND SCUPPER DRAIN OUTLET PIPE

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 3, Page 11 of 14 TLG Services, Inc. FIGURE 3.8 POST REMEDIATION VIEW OF REMAINING THERMAL COLUMN VENT AND DRAIN STUBS

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 3, Page 12 of 14 TLG Services, Inc. FIGURE 3.9a POST REMEDIATION VIEW OF REMAINING GROUND FLOOR EXHAUST DUCT AND PLENUM

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 3, Page 13 of 14 TLG Services, Inc. FIGURE 3.9b VIEW OF FIRST FLOOR EXHAUST DUCT PLENUM

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 3, Page 14 of 14 TLG Services, Inc. FIGURE 3.10 POST REMEDIATION CONFIGURATION OF BIOLOGICAL SHIELD

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 4, Page 1 of 9 TLG Services, Inc. 4.0 RADIOLOGICAL CONTAMINANTS AND CRITERIA

4.1 RADIOACTIVE

CONTAMINANTS IDENTIFIED DURING DECOMMISSIONING Samples of waste materials were obtained during the remediation process for 10CFR61 characterization purposes. As a minimum, each unique waste stream was sampled that had the potential for containing different mixtures of radionuclides. This data is useful for identifying the potential radiological contaminants that might remain in various structural materials that could be present at the time of the FSS.

Five samples were analyzed by an outside laboratory: the samples consisted of the following materials: Stainless Steel Regulating Blade: Neutron Activated Aluminum from the Thermal Column liner: Neutron Activated Concrete from the Biological Shield: Neutron Activated Graphite from the Thermal Column: Neutron Activated Resin from the Reactor Pool Water Treatment System Table 4.1 summarizes the analytical results for the radionuclides in these waste stream samples that were positively identified and attributable to reactor operation.

In addition, two pre-FSS concrete samples were taken from the remaining biological shield, at locations likely to represent peak radionuclide concentrations present anywhere in the biological shield concrete. These samples were obtained at the end of remediation work for the purpose of providing a preliminary indication that sufficient materials had been removed. The samples were obtained from the remaining edge of the poowall adjacent to the beam port tube area. Table 4.2 provides a summary of the analytical results for the radionuclides that were positively identified in these samples and which are attributable to reactor operation.

A complete presentation of the laboratory results is presented in Appendix A. The locations from which the seven samples (five waste stream and two pre-FSS) were obtained are shown in Figure

4.1. Worcester

Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 4, Page 2 of 9 TLG Services, Inc. TABLE 4.1

SUMMARY

OF WASTE STREAM AND P RE-FSS SAMPLE CHARACTERIZATION RESULTS

• Sample ID No.

WPI #1 WPI #2 WPI #3 WPI # 4 WPI #5 WPI #8 WPI #9 Material Stainless Steel (neutron activated)

Aluminum (neutron activated)

Concrete (neutron activated)

Graphite (neutron activated)

Ion-exchang e Resin Concrete Concrete Item Sampled Regulating control blade Thermal column liner Biological shield -floor and wall Contents of thermal column Contents of reactor pool water treatment system Remaining biological shield wall

- near beam port Remaining biological shield floor

- near reactor core area Purpose of Sample Waste stream char. Waste stream char. Waste stream char. Waste stream char. Waste stream char. Pre-FSS Pre-FSS Radionuclide

µCi/g µCi/g µCi/g µCi/g µCi/g µCi/g µCi/g Fe-55 4.38E-01 8.69E - - - - Ni-63 3.47E - - - - - Co - - 2.27E - - Co-60 2.24E-01 6.70E-04 3.39E-06 3.38E-06 1.55E-07 7.79E-07 5.85E-07 Zn 3.36E - - - - Cs-134 - - 3.92E - 1.85E Cs-137 - - - - 3.01E - Eu-152 - - 1.52E-05 4.0 2E 2.65E-06 2.26E-06 Eu-154 - - 8.06E-07 2.57E - -

• Includes detected radionuclides only and excludes naturally occurring radionuclides - see Appendix A for complete reporting of analytical results TABLE 4.2

SUMMARY

OF ANALYTICAL RESULTS FOR PRE-FSS SAMPLES FROM THE BIOLOGICAL SHIELD Sample Location Sample ID (Location shown on Figure 4.1) Co-6 0 (pCi/g) Cs-134 (pCi/g) Eu-152 (pCi/g) Interior East Bio

-Shield Wall, Adjacent to Beam Port WPI # 8 0.78 0.19 2.65 Interior Bio

-Shield Floor, Edge of Demolition Area WPI # 9 0.59 no t detecta ble 2.26 Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 4, Page 3 of 9 TLG Services, Inc. FIGURE 4.1 LOCATION OF WASTE STREAM CHARACTERIZATION AND PRE-FSS SAMPLES

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 4, Page 4 of 9 TLG Services, Inc. 4.2 CHARACTERISTICS OF THE RADIOACTIVE CONTAMINANTS THAT POTENTIALLY COULD BE PRESENT DURING FSS The radionuclide mixtures represented by four of the five waste streams and the two pre-FSS samples are likely to be representative of any residual radioactivity that might have be en present at the time the FSS was performed. All potentially activated stainless steel was dismantled intact (i.e., without segmentation), and removed from the facility. Accordingly, no residual radioactivity due to this material was expected to be present. The activated aluminum radionuclide mixture had the potential for being found in the remaining portions of floor and lower walls of reactor pool liner and thermal column liner that were in close proximity to the reactor core. Likewise, the activated concrete radionuclide mixture (three samples) had the potential for being found in the remaining portions of the biological shield walls and floor that were in close proximity to the reactor core. All irradiated graphite was removed from the reactor facility, and was not present during the FSS. The contaminated resin was removed along with the reactor pool water likely be representative of any contamination found on surfaces that had been wetted by pool water. It was also slightly possible that if surface contamination was found on structural surfaces that it could be due to dust or debris created during the process of removing the activated graphite, activated aluminum liners or activated bio-shield concrete. With the exception of the Fe-55 contained in activated aluminum, all sources of radioactivity at WPI (i.e., activated concrete and surfaces wetted with pool water) were comprised entirely of one or more gamma emitters (i.e., Co-60, Cs-134, Cs-137, Eu-152 and Eu-154) that were readily identified using portable detection equipment and the mixture readily quantified by gamma spectroscopy methods. The one exception (i.e., the radioactivity that may be found in aluminum liners) does contain a hard-to-detect radionuclide (Fe-55); however, Co-60 which represents 7.13% of the total radioactivity in activated aluminum, can be readily detected and was used to scale the hard-to-detect radionuclide Fe-55.

Table 4.3 provides a summary of the radionuclides mixtures that would be representative of residual radioactivity that could have been encountered at the WPI facility during the FSS.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 4, Page 5 of 9 TLG Services, Inc. TABLE 4.3

SUMMARY

OF THE RADIONUCLIDES MIXTURES REPRESENTATIVE OF RESIDUAL RADIOACTIVITY Original Contaminant Source Potential Locations at WPI Radionuclide Constituents (fractional abundance)

Aluminum (neutron activated)

Aluminum liners within the reactor pool in close proximity to the reactor core, thermal column liner and beam port tube.

Note: a remote possibility exist ed for loose surface contamination being present at locations where (or near where) liners where segmented (such as the pool floor and underlying soil surface beneath the reactor core area that became exposed when the liner and concrete above it was removed it was also remotely possible that this soil could have become directly neutron activated)

. Fe-55 (0.9251), Co-60 (0.0713), Zn-65 (0.0036)

Concrete (neutron activated) Concrete Biological Shield, in close proximity to the reactor core and beam port shutter and tube.

Note: a remote possibility existed for loose surface contamination being present at locations where (or near where) the activated concrete that was demolished and removed (such as the pool floor and underlying soil surface beneath the reactor core area that became exposed when the concrete above it was removed it was also remotely possible that this soil could have become directly neutron activated)

. Co-60 (0.1717), Cs-134 (0.0202), Eu-152 (0.7677), Eu-154 (0.0404)

Graphite (neutron activated) None expected all graphite within the thermal column was removed during decommissioning

. Note: a remote possibility exist ed for loose surface contamination being present at locations where (or near where) the activated graphite was handled and removed, such as the remaining thermal column liner and adjacent floor surfaces

. Co-60 (0.0078), Eu-152 (0.9325), Eu-154 (0.0596)

Radioactivity in Reactor Pool Water Surface s that were wetted by reactor pool water (such as interior pool liner surfaces, drain and water treatment system piping embedded within the biological shield, scupper drains and operating floor surfaces , and floor wall surfaces adjacent to the removed pool water treatment system).

Co-60 (0.0005), Cs-137 (0.9995)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 4, Page 6 of 9 TLG Services, Inc. 4.3 FSS CRITERIA Future uses of the former reactor facility at WPI have not yet been completely defined. However, it is likely that the facility will continue being used in some capacity for academic programs

as such, license termination will be based on the un-restricted release scenario. In its DP, WPI committed to the use of NRC screening values in lieu of development of site-specific Derived Concentration Guideline Levels (DCGLs). These screening values were presented Tables 2.2 and 2.3 of the DP for building surface and surface soil, respectively, and are presented herein as Tables 4.4 and 4.5. (Note: soil screening values for Nb-94 and Eu-152 were revised for DP Table 2.3 on September 25, 2012 (DP revision 2) to meet the radionuclide concentration limits in the MOU between NRC and EPA regarding consultation levels).

In order to satisfy the surface contamination criteria, measurement of gross beta surface activity is based upon application of the Sum-of-Fractions rule for the radionuclides potentially present. The radionuclide mixture found in the reactor e representative of any radioactivity that might have become dispersed and deposited on the reactor facility structural or equipment surfaces. As such, that radionuclide mixture is the basis for application of a gross beta DCGL equivalent to the DP Table 2.2 surface contamination criteria.

Based upon that mixture (99.95% Cs-137 and 0.05% Co-60), the gross beta DCGL is 23,700 DPM/100 cm 2, based upon the S um-of-Fractions (SOF) rule. The derivation of this value is provided in Appendix B. This value was applied to all the total surface contamination determinations that were made. Additionally, one-tenth of this value (i.e., 2,370 DPM/100 cm 2) was the criteria for the removable fraction of any surface contamination.

Satisfying the soil contamination criteria (either for soil or activated materials) will be demonstrated by the SOF approach. The SOF of contaminant concentrations divided by their respective Default Screening Values must therefore be <Unity (i.e., <1.0).

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 4, Page 7 of 9 TLG Services, Inc. TABLE 4.4 LICENSE TERMINATION SCREENING VALUES FOR BUILDING SURFACE CONTAMINATION Radionuclide Symbol Acceptable Screening Levels

• f or Unrestricted Release (DPM/100 cm 2)** Hydrogen-3 (Tritium) 3 H 1.2 E+08 Carbon-14 14 C 3.7E+0 6 Sodium-22 22 Na 9.5E+03 Sulfur-35 35 S 1.3E+07 Chlorine-36 36 Cl 5.03 E+05 Manganese-54 54 Mn 3.2E+04 Iron-55 55 Fe 4.5E+06 Cobalt-60 60 Co 7.1E+03 Nickel-63 63 Ni 1.8E+06 Strontium-90 90 Sr 8.7E+03 Technetium

-99 99 Tc 1.3E+06 Iodine-129 129 I 3.5E+04 Cesium-137 137 Cs 2.8E+04 Iridium-192 192 Ir 7.4E+04

• Screening levels are based on the assumption that the fraction of removable surface contamination is equal to 0.1. For cases when the fraction of removable contamination is undetermined or higher than 0.1, users may assume, for screening purposes, that 100 percent of surface contamination is removable, and therefore the screening levels should be decreased by a factor of 10. Alternatively, users having site-specific data on the fraction of removable contamination, based on site-specific resuspension factors (e.g., within 10-to-100 percent range), may calculate site-specific screening levels using RESRAD-BUILD Version 3.0. ** Units are disintegrations per minute (DPM) per 100 square centimeters (DPM/100 cm 2). One DPM is equivalent to 0.0167 Becquerel (Bq). Therefore, to convert to units of Bq/m 2 multiply each value by 1.67. The screening values represent surface concentrations of individual radionuclides that would be deemed in compliance with the 0.25 mSv/yr (25 mrem / year) unrestricted release dose limit in 10 CFR 20.1402. For radionuclides in a Sum-of-F.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 4, Page 8 of 9 TLG Services, Inc. TABLE 4.5 LICENSE TERMINATION SCREENING VALUES FOR SURFACE SOIL Radionuclide

• Symbol Surface Soil Screening Values** for Unrestricted Release (pCi/g)*** Hydrogen-3 (Tritium) 3 H 1.1E+02 Carbon-14 14 C 1.2E+0 1 Sodium-22 22 Na 4.3E+00 Sulfur-35 35 S 2.7E+02 Chlorine-36 36 Cl 3.6E-01 Calcium-45 45 Ca 5.7E+01 Scandium-46 46 Sc 1.5E+01 Manganese-54 54 Mn 1.5E+01 Iron-55 55 Fe 1.0E+0 4 Cobalt-57 57 Co 1.5E+02 Cobalt-60 60 Co 3.8E+00 Nickel-59 59 Ni 5.5E+03 Nickel-63 63 Ni 2.1E+03 Strontium-90 90 Sr 1.7E+00 Niobium-94 94 Nb 3.0E+00 **** Technetium

-99 99 Tc 1.9E+01 Iodine-129 129 I 5.0E-01 Cesium-134 134 Cs 5.7E+00 Cesium-137 137 Cs 1.1E+01 Europium-152 152 E u 7.0E+00 ****

Europium-154 154 Eu 8.0E+00 Iridium-192 192 Ir 4.1E+01 Lead-210 210 Pb 9.0E-01 Radium-226 226 Ra 7.0E-01 Radium-226+C 226Ra+C 6.0E-01 Actinium-227 227 Ac 5.0E-01 Actinium-227+C 227Ac+C 5.0E-01 Thorium-228 228 Th 4.7E+00 Thorium-228+C 228 T h+C 4.7E+00 Thorium-230 230 Th 1.8E+00 Thorium-230+C 230Th+C 6.0E-01 (Note: Table 4.5 is continued on next page)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 4, Page 9 of 9 TLG Services, Inc. TABLE 4.5 (continued) LICENSE TERMINATION SCREENING VALUES FOR SURFACE SOIL Radionuclide*

Symbol Surface Soil Screening Values** for Unrestricted Release (pCi/g)***

Thorium-232 232 TH 1.1E+00 Thorium-232+C 232TH+C 1.1E+00 Protactinium

-231 231 Pa 3.0E-01 Protactinium

-231+C 231Pa+C 3.0E-01 Uranium-234 234 U 1.3E+01 Uranium-235 235 U 8.0E+00 Uranium-235+C 235 U+C 2.9E-01 Uranium-238 238 U 1.4E+01 Uranium-238+C 238 U+C 5.0E-01 Plutonium-238 238 Pu 2.5E+00 Plutonium-239 239 Pu 2.3E+00 Plutonium-241 241 Pu 7.2E+01 Americium 241 Am 2.1E+00 Curium-242 241 Cm 1.6E+02 Curium-243 243 Cm 3.2E+00

• Plus Chain (+C) indicates a value for a radionuclide with its decay progeny present in equilibrium. The values care concentrations of the parent radionuclide, but account for contributions from the complete chain of progeny in equilibrium with the parent radionuclide (NUREG/CR-5512 Volumes 1, 2, and 3).

• • These values represent surface soil concentrations of individual radionuclides that would be deemed in compliance with the 25 mrem/year (0.25 mSv/year) unrestricted release dose limit in 10 CFR 20.1402. For radionuclides in a mixture, m-of-Fraction *** Screening values are in units of (pCi/g) equivalent to 25 mrem/year (0.25 mSv/year). To convert from pCi/g to units of Becquerel per kilogram (Bq/kg) divide each value by 0.027. These values were derived based on selection of the 90 th percentile of the output dose distribution for each specific radionuclide (or radionuclide with the specific decay chain). Behavioral parameters were set at the mean of the distribution of the assumed critical group. The metabolic parameters **** From: MOU Table 1: Consultation Triggers for Residential and Commercial / Industrial Soil Contamination, between the Environmental Protection Agency and the Nuclear R Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 1 of 28 TLG Services, Inc.

5.0 FINAL

STATUS SURVEY PROCESS 5.1 FSS PLAN AND CHANGES TO THE PLAN The objective of the FSS was to demonstrate that remedial actions have been effective in removal/reduction of radiological materials and contamination, and that the post-remediation radiological conditions satisfy the NRC-approved criteria for termination of the WPI Reactor License and enable future use of the WPI facility without radiological restrictions. The FSS w as performed using the guidelines and recommendations presented in NUREG-1757 and MARSSIM. The scope and methodology by which the planned FSS work was performed is documented in the FSS Plan which was approved by NRC.

With very few exceptions, the FSS work was performed as described in the FSS Plan. The following are changes that were implemented in an effort to enhance the FSS process:

1.Four of the survey units specified in the FSS plan (SU 1.7 Top of Biological Shield, SU 2.3 Ground Floor - South Wall and Floor East of Biological Shield and SU 2.1/3.5 Ground/First Floors - Lower Walls and Floors) were subdivided into multiple smaller SUs in order to facilitate measurement location placement and/or allow segregation structural material make-up to enhance ability to make better comparison of the SU direct surface contamination measurement data with that of background reference area data. These subdivisions were as follows:

a.SU 1.7 Top of Biological Shield: This SU was subdivided to facilitate placement of random-systematic survey measurement locations. This subdivision resulted in two SUs; 1.7V for the vertical surface of the reactor pool stub walls and 1.7H for the horizontal surfaces of the reactor pool bridge, operating area floor, lower half of the south wall adjacent to the reactor pool and the scupper drain area surrounding the reactor pool. b.SU 2.3 Ground Floor Reactor Facility, South Wall and Floor, East of Reactor Centerline: This survey unit originally included a dark room, built against the south wall, constructed of fiber board and plywood over wood framing. The north, east and west walls, and ceiling / roof of the dark room were assigned to a new SU 2.9 Dark Room. This subdivision was primarily made to facilitate location mapping and segregate differing construction materials that exhibited differing background results (wood / Fiber board vs. Concrete block).

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 2 of 28 TLG Services, Inc.

c.SU 2.1 and 3.5 Ground Floor / First Floor Reactor Facility, Lower Walls and Floor, East of Reactor Centerline: These steel framed steel grate stair structure. This structure was removed from SU 2.1 and 3.5 and assigned a unique SU (SU 2.10) to facilitate location mapping and segregate differing construction materials that exhibited differing background results (i.e., steel vs. concrete slab and block).

2.Two additional pipe stubs were discovered that were added to the embedded pipe survey work scope; they are in SUs 3.5 and 3.7.

3.Provisions were made for surveying incidental miscellaneous materials that were present within the facility that were not directly included with the structures, systems and component FSS Plan work scope. Examples of such miscellaneous materials include: contents within desk drawers and file cabinets, small tools , laboratory equipment and supplies (such as glassware), and classroom furniture. The process for performing these surveys was added by revision to the decommissioning work procedure (DWP-10; copy provided in Appendix E) that specified performance of the FSS work activities.

4.Due to the small amount of MARSSIM Class 3 areas, for purposes of simplicity, they were generally surveyed on the same basis as MARSSIM Class 2 areas.

The following sub-sections describe the specific details of the FSS work that was performed. Data results are presented in Section 6.0.

5.2 FSS PROCESS The following subsections provide a description of the methodology that was used to perform the FSS. These subsections follow the general topics provided in the FSS Plan, but are enhanced with increased specificity and finality with regard to information made available at the conclusion of the FSS site work.

5.2.1 Identification

of Survey Units The WPI reactor facility was divided into manageable areas called survey units.

A survey unit is defined as a contiguous areas or portion of a facility with common radiological and physical characteristics. The WPI reactor facility was divided into 43 survey units for the FSS, representing approximate ly 1,237 m 2 of structural area (floors, walls and ceilings). This division was based on historical assessments made during preparation of the DP, and information gained from radiological monitoring that was conducted during remedial D&D activities. A listing of the survey units for the WPI reactor facility areas that were used for conducting the FSS is presented in Table 5.1.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 3 of 28 TLG Services, Inc. TABLE 5.1 FSS SURVEY UNITS AND MARSSIM CLASSIFICATIONS Survey Unit No. SU Description Approx. Surface Area (m 2) MARSSIM Classification

1.0 BIOLOGICAL

SHIELD / REACTOR POOL

1.1 Reactor

Pool Aluminum Liner 48 1 1.2 Thermal Column Aluminum Liner 2 1 1.3 Interior Reactor Pool Concrete Floor na 1 1.4 Interior Reactor Pool Concrete Walls na 1 1.5 Beam Port Tube 1 1 1.6 Soil Under Removed Concrete Floor Area 1.3 1 1.7 Top of Bio

-Shield 1.7V Top of Bio

-Shield, Vertical Surfaces 9 1 1.7H Top of Bio

-Shield, Horizontal Surfaces 29 1 1.8 Embedded Piping (In Biological Shield)

1.8.1 Drain

Lines from Reactor Pool Scupper Drains na 1 1.8.2 Return Line from Pool Water Treatment System na 1 1.8.3 Vent line above Beam Port na 1 1.8.4 Drain Line from Beam Port Shutter Housing na 1 1.8.5 Drain from Reactor Pool (Feed to Rx Water Treatment System) na 1 1.8.6 Vent Line from Thermal Column (Upper

-Left) na 1 1.8.7 Vent Line from Thermal Column (Lower

-Left) na 1 1.8.8 Vent Line from Thermal Column (Upper

-Right) na 1 1.8.9 Vent Line from Thermal Column (Lower

-Right) na 1 1.8.10 Vent Line from Thermal Column (Top

- Center) na 1 1.9 Exterior Walls of Biological Shield

1.9.1 North

Exterior Bio

-shield Wall 21 2 1.9.2 West Exterior Bio

-Shield Wall 18 1 1.9.3 Thermal Column Shield Door 5 1 Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 4 of 28 TLG Services, Inc. TABLE 5.1 FSS SURVEY UNITS AND MARSSIM CLASSIFICATIONS Survey Unit No. SU Description Approx. Surface Area (m 2) MARSSIM Classification 1.9.4 East Exterior Bio

-Shield Wall 18 1 2.0 GROUND FLOOR REACTOR FACILITY

2.1 Ground

Floor of Reactor Room; Floor and South Wall (West of Reactor Centerline) 73 1 2.2 Ground Floor of Reactor Room; Lower North & West Walls (West of Reactor Centerline) 35 2 2.3 Ground Floor of Reactor Room; Floor and South Walls (East of Reactor Centerline) 92 1 2.4 Ground Floor Reactor Room; Lower North and East Walls (East of Reactor Centerline) 29 2 2.5 Ground Floor Reactor Room; Upper Walls, Ceiling and Exterior of Suspended Equipment (West of Reactor Centerline) 84 2 2.6 Groun d Floor Reactor Room; Upper Walls, Ceiling, and Exterior of Suspended Equipment (East of Reactor Centerline) 93 2 2.7 Ground Floor Reactor Room; RAM Storage Room (Lower Walls & Floor) 28 1 2.8 Ground Floor Reactor Room; RAM Storage Room (Upper Walls, Ceiling, and exterior of suspended equipment) 28 1 2.9 Groun d Floor Reactor Room; Interior and Exterior Walls, Ceiling and Roof of Dark Room 41 1 2.10 Reactor Room Stairs 4.3 1 3.0 FIRST FLOOR REACTOR FACILITY

3.1 Reactor

Office Lower Walls and Floor 54 3 3.2 Reactor Office Upper Walls and Ceiling 54 3 3.3 First Floor Reactor Room; Lower Walls and Floor (West of Reactor Centerline) 80 2 Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 5 of 28 TLG Services, Inc. TABLE 5.1 FSS SURVEY UNITS AND MARSSIM CLASSIFICATIONS Survey Unit No. SU Description Approx. Surface Area (m 2) MARSSIM Classification

3.4 First

Floor Reactor Room; Upper Walls and Ceiling (West of Reactor Centerline) 99 2 3.5 First Floor Reactor Room; Lower Walls and Floor (East of Reactor Centerline) 91 1 3.6 First Floor Reactor Room; Upper Walls and Ceiling (East of Reactor Centerline) 115 2 3.7 First Floor Reactor Room; Tool Closet

-Lower Walls and Floor 42 1 3.8 First Floor Reactor Room; Tool Closet (Upper Walls and Ceiling) 42 1 4.0 OTHER AREAS / ITEMS

4.1 Exhaust

Ventilation System (Interior Duct and Equipment Surfaces) na 1 4.2 Reactor Water Treatment System Drain Piping to Sanitary Sewer System na 1 4.3 HVAC Units (3 Free

-Standing Units) na 1 5.2.2 Classification by Contamination Potential The survey units designated for the WPI reactor f acility were classified as one of the three MARSSIM classification areas, according to contamination potential, as follows: Class 1 Areas that have, or had prior to remediation, a potential for radioactive contamination (based on site operating history) or known contamination (based on previous radiation surveys) above the DCGL. Examples include site areas previously subjected to remedial actions (e.g., locations where leaks or spills are known to have occurred) and former waste storage areas.

Class 2 Areas that have, or had prior to remediation, a potential for radioactive contamination or known contamination, but are not expected to exceed the DCGL. Examples include locations where radioactive materials were present in an unsealed form , potentially contaminated transport routes, areas handling low Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 6 of 28 TLG Services, Inc. concentrations of radioactive materials, and areas on the perimeter of former contamination control areas.

Class 3 Any impacted areas that were not expected to contain any residual radioactivity, or were expected to contain levels of residual radioactivity at a small fraction of the DCGL, based on site operating history and previous radiation surveys. Examples include buffer zones around Class 1 and Class 2 areas, and areas with a very low potential for residual contamination, but having insufficient information to justify a non-impacted classification.

The bases for classification was the facility history (including the initial Historical Site Assessment and radiological monitoring conducted during characterization) and remedial activities. The classifications delineated in the FSS Plan were used and not changed during performance of the FSS

however, the estimated surface areas of each were updated using new information obtained during installation the reference grids (see Section 5.2.3) to provide better accuracy. Table 5.1 indicates the designated MARSSIM classifications for the various survey unit areas within the reactor facility.

5.2.3 Survey

Reference Systems A one meter by one meter grid system was established on structural surfaces to provide a means for referencing measurement and sampling locations. Grids were assigned Cartesian coordinate indicators to enable survey location identification.

Structure grids were setup so that they could be referenced to permanent building features. Figures 5.1 and 5.2 provide photographic view of the grid system used for SUs 1.1, 1.3, 1.4 and 1.6 and a grid map corresponding to that area. Systems and irregular surfaces of less than 20 m 2 were not gridded, but survey locations were referenced to prominent facility features on survey maps or photographs. Survey maps indicating the coordinate system for the applicable survey units are provided in Section 6.0, along with presentation of the FSS data.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 7 of 28 TLG Services, Inc. FIGURE 5.1 PHOTOGRAPHIC EXAMPLE REFERENCE GRID SYSTEM (INTERIOR REACTOR POOL: SURVEY UNITS 1.1, 1.3, 1.4 and 1.6)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 8 of 28 TLG Services, Inc. FIGURE 5.2 EXAMPLE GRID LOCATION MAP (INTERIOR REACTOR POOL: SURVEY UNITS 1.1, 1.3, 1.4 and 1.6)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 9 of 28 TLG Services, Inc. 5.2.4 Placement of Survey Locations A random-start, systematic placement pattern was used for placement of the radiological measurement / sampling locations, where open and regular surfaces existed (e.g., floors, walls, ceilings and interior / exterior reactor pool surfaces). To accomplish this, a triangular grid pattern was used for most survey units, except where dimensions and/or other factors related to a specific survey unit required use of an alternate pattern. The spacing (L) between data points on the triangular grid pattern was estimated by:

L = [(Survey Unit Area) / (0.866 x number of data points)]

1/2 The minimum number of datum to be obtained from each survey unit was 17. This number was based upon using a relative shift of 1.67 and Type I and Type II decision errors of 0.05, to estimate the minimum number of datum to perform a Wilcoxon Rank Sum (WRS) test, as obtained from Table 5.5 of NUREG-1575 (MARRSIM). The relative shift component was based upon MARSSIM guidance for situations where final sample data were not yet available.

To simpl ify the placement of the triangular grid pattern, while assuring a sufficient number of data points were obtained for statistical purposes, the value of L was generally rounded down to the nearest whole meter. However, if the systematic pattern did not provide sufficient data points (or, conversely, provided for too many data points), the dimensions of the triangular grid were adjusted accordingly.

The designated random-systematic survey locations were marked onto each of the

SUs surfaces with numbered brightly colored adhesive tags (to assure proper measurement locating by the FSS personnel and to provide reproducible re-locating of measurement points). Figure 5.3 provides a photographic example of triangular grid placement of measurement locations.

However, survey units with areas < 10 m2 and/or with known asymmetrical radioactive material distribution patterns (such as SU 1.5 - Beam Port Tube, and SU 1.3 - Interior Concrete Floor - Reactor Pool), were not evaluated on a statistical basis as previously described. Instead, a minimum of four measurements (or samples) were obtained from such areas at locations having the highest probability of exhibiting residual contamination, based on judgment using prior knowledge of contaminant distribution, with a bias towards finding peak concentrations. The resulting da ta were then compared individually with the applicable radionuclide concentration limits on a S um-of-Fractions basis.

For survey units with irregular geometries, such as the SU 4.1 - Exhaust Ventilation Ducts and the SU 4.3 - HVAC Units , placement of the measurement or sampling points w as judgmental, biased to locations considered to have the highest Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 10 of 28 TLG Services, Inc. probability of residual contamination, with additional locations chosen to provide well-distributed survey unit coverage.

FIGURE 5.3 PHOTOGRAPHIC EXAMPLE OF TRIANGULAR GRID PLACEMENT O F MEASUREMENT LOCATIONS (Note: Orange tags indicate triangular grid location placements, while pink tags indicate completion of surface beta contamination scans within a rectangular grid box

.) 5.2.5 Survey Instrumentation Table 5.2 provides a listing of radiological survey instrumentation that were used to implement the WPI Reactor FSS. This table specifies the application of each instrument, along with the actual range of detection sensitivities determined on an area-by-area basis, and the corresponding percentages when compared to the applicable license termination criteria. Detection sensitivities specific to each Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 11 of 28 TLG Services, Inc. survey unit are provided in Tables

6.1 through

6.5A-D, which provides a summary of the overall results for the FSS. The instruments were maintained and calibrated in accordance with WPI procedure RPP-13, Calibration and Quality Control of Portable Radiological Survey Instruments. TABLE 5.2 INSTRUMENTATION FOR WPI FINAL STATUS SURVEY Detector

• Meter

• Make: Model Application Sensitivity Range , DPM/100 cm 2 (% of LT criteria)

Type Make: Model Scanning Static Count (1 minute)

Beta Scintillation Bicron: DP

-6A Bicron: Labtech Surface contamination scan and measurement 1 , 730-2 , 400 (7 to 10%) 320-430 (1.4 to 1.8%) Geiger-Mueller Ludlum: 44-40 Ludlum: 2200 Smear and Pipe Swab sample counting na 100-120 (4.2 to 5.1%) Ludlum: 44-10 Ludlum: 12 Area gamma scanning na na NaI Ludlum: 44-62 Ludlum: 12 Gamma scanning- small diameter pipe na 7 , 770-15 ,540 (33 to 66%) Alpha Spectra: 818/2 Ludlum 732

-1 PC-Based Gamma Ray Spectroscopy System In-situ and sample screening radionuclide screening na na For simplicity in application to the FSS, instrument response (efficiency) w as based on NIST-traceable sources, with Cl-36 (beta Eave. = 252 keV and Emax = 714 keV) being the source of choice to best represent the various radionuclide mixtures. The beta energies of Cl-36 ar e representative of the dominant potential contaminants at WPI (i.e., Cs-137 @ beta Eave. = 195 keV and Emax. = 1167 keV and Eu-152 @ beta Eave. = 288 keV and Emax. = 1840 keV) and thus provided conservative overestimates Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 12 of 28 TLG Services, Inc. of the contaminant mixture gross beta activities. The 4 pi efficiency for the 100 cm 2 DPS detectors used for direct beta surface contamination measurements was 26.3

% and ~20.5% for the 15.5 cm 2 GM pancake detectors that were used count smear and pipe swab samples. Detection sensitivities were estimated using the guidance in MARSSIM and NUREG-1507, Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Fields Conditions (NRC, 1997b). Measuring instruments were calibrated on an annual basis and whenever the accuracy of the equipment might have been suspect. Calibration was performed by a commercial vendor (RSA Laboratories, of Hebron, CT) using standards traceable to NIST or an equivalent standard organization. Instruments were marked with calibration tags to indicate calibration status.

Check source and background count operational checks were performed in accordance with procedure RPP-13 at the beginning and end of each day of FSS activity and whenever there might have been a reason to question instrument performance.

5.2.6 Survey

Techniques Data collected for FSS of structural surfaces consisted of scans to identify locations of elevated residual contamination, direct measurements of beta surface activity, and measurements of removable beta surface activity. FSS of survey units with potential for volumetric radioactivity (i.e., neutron activated materials areas, soil and sediment) were scanned for both gamma and direct surface beta radiation to identify locations of highest potential residual contamination, followed by obtaining samples of the structural or residual materials, and then analyzed for potential contaminant concentrations of concern by gamma spectroscopy, with scaling-in of hard-to-detect radionuclides based upon previously determined scaling factors. Small diameter embedded pipes and conduits were evaluated for residual internal contamination by checking accessible ends for fixed and removable contamination, and by gamma logging accessible interior surfaces with a small diameter (1.3 cm)

NaI gamma scintillation detector. Survey techniques are described in more detail in the following sub-sections. Detection sensitivities for the various contamination quantifying techniques were presented in Table 5.2, which indicated an ability to detect contamination levels at very small fractions of the gross beta surface contamination criterion for all techniques, except interior pipe/conduit contamination determinations.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 13 of 28 TLG Services, Inc. Beta Surface Scans Beta scanning of structure surfaces w as performed to identify locations of residual surface activity. Hand-held, 100 cm2 thin plastic beta scintillation detectors were used for most surfaces. Scanning w as performed with the detector within 0.5 cm of the surface with a scanning speed no greater than 1 detector width per second. Audible signals were monitored and locations with elevated levels (if identified) subjected to further investigation. Detection sensitivity for scanning for ranged from 1,730 to 2,400 DPM/100 cm

2. Minimum scan coverage was 100 percent for Class 1 surfaces, and 25 percent for Class 2 and 3 surfaces. Scan Coverage for Class 2 and Class 3 surfaces were biased towards areas considered by professional judgment to have highest potential for contamination. Figure 5.

4 provides a photographic view of this scanning process being performed.

FIGURE 5.4 PHOTOGRAPHIC VIEW OF SURFACE BETA CONTAMINATION SCANNING PROCESS

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 14 of 28 TLG Services, Inc. Gamma Surface Scans The primary purpose for gamma scanning of areas and surfaces w as to identify locations of highest potential residual surface activity due to neutron activation. NaI gamma scintillation detectors (2 inch x 2 inch) were used for these scans. These scans were performed inside the reactor pool biological shield structure (including the small exposed soil surfaces under the reactor core box area). Scanning w as performed by moving the detector in a serpentine pattern over the structural surfaces in question, while advancing at a rate of approximately 0.5 m per second. The distance between the detector and the surface were maintained within 5 cm of the surface. Audible signals and digital readouts were monitored for indications of elevated radioactivity.

General area gamma scanning w as performed throughout the remainder of the facility to provide additional assurance that unforeseen sources of radioactivity were not present.

Direct Surface Beta Contamination Measurements Direct measurement of beta surface activity w as performed at designated locations using a 100 cm2 thin plastic scintillation detector. Measurements were conducted by integrating the count over a timed one-minute counting period. Figure 5.5 provides a photographic view of the direct beta contamination measurement process. Detection sensitivities for these removable contaminations determinations ranged from 320 to 430 DPM/100 cm

2. FIGURE 5.5 PHOTOGRAPHIC VIEW OF THE DIRECT BETA CONTAMINATION MEASUREMENT PROCESS

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 15 of 28 TLG Services, Inc. Removable Surface Beta Contamination Measurements Smear sampling for removable radioactivity w as performed at each direct surface measurement location. A 100 cm2 surface area w as wiped with a ~ 2 inch diameter paper filter or cloth, using moderate pressure. Smear samples were analyzed onsite for gross beta activity using a shielded 15.5 cm 2 GM pancake detector, sample holder and scaler. Detection sensitivities for these removable contaminations determinations ranged from 100-to-1 20 DPM/100 cm

2.

Soil Sampling Soil sampling w as limited to the floor area of the biological shield where the concrete floor was removed, which exposed the underlying soil surface (SU 1.6). The exposed soil surface was scanned for both gamma and direct beta radioactivity, with no apparent elevated results observed. As such, four sampling locations were evenly distributed over the soil surface. Samples of surface soil (from approximately the upper 15 cm) were obtained from selected locations using a hand trowel. Approximately 500-to-1,000 g of soil was collected at each sampling location. Soil samples were analyzed by gamma spectroscopy (screened on-site with a shielded NaI d etector and then sent to GEL Laboratories for analysis. As discussed in Section 4, hard-to-detect radionuclide concentrations were scaled-in, as applicable, using ratios obtained from waste characterization data. It was originally planned that if reactor-originated radioactivity was detected in the first sample layer , additional samples at a greater depth (e.g., 15-to-30 cm) would be obtained to verify that radionuclide concentrations were not increasing / or were limited to the top surface layer. However, bedrock was encountered at a depth of 14-to-15 cm below the surface; therefore, no additional deeper soil samples could be obtained. Figure

5.6 provides

a photographic view of the SU 1.6 soil sampling locations (note: the fifth location that can be observed at the center of the soil area is the location from which a pre-FSS characterization sample was obtained).

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 16 of 28 TLG Services, Inc. FIGURE 5.6 SAMPLING LOCATIONS FOR SU 1.6: SOIL UNDER FORMER REACTOR CORE AREA Structural Media Sampling Sampling of structural materials w as performed in and around the reactor pool / biological shield to determine radionuclide concentrations in media potentially made radioactive by neutron irradiation. This included the aluminum floor / wall liner within the reactor pool, the aluminum liner within the thermal column and beam port tube, and underlying concrete of the biological shield, thermal column and beam port areas. Sampling locations were to be selected based on gamma scan results (to find the location of highest potential activity) or at the theoretical points of maximum activation based on media with the closest proximity to the reactor core if no apparent elevated gamma scan activity is found. No apparent elevated gamma scanning results were observed anywhere within the reactor pool / Thermal column structure. However, a slight elevation in direct beta levels (approximately Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 17 of 28 TLG Services, Inc. twice background) was observed at the theoretical point of maximum neutron activation within the thermal column (i.e., at the leading edge of the remaining thermal column liner closest to the former reactor core area).

Samples were obtained from the selected locations using core drills. Approximately 3-inch diameter cores were collected at each sampling location (complete thickness in the case of aluminum liner material, and ~15 cm long increments for concrete. In the case of the concrete biological shield, additional samples at a greater depth (e.g.,

15-to-30 cm or beyond) were obtained to verify that radionuclide concentrations were not increasing / or are limited to the top surface layer if reactor-originated radioactivity were detected in the first core sample increments. Samples were analyzed by gamma spectroscopy (screened on-site with a shielded NaI detector and then sent to GEL Laboratories for analysis. As applicable, hard-to-detect radionuclide concentrations were scaled-in using the ratios presented in Section 4.

Figure 5.7 provides a photographic view of the sampling process for concrete and Figure 5.8 provides a photographic view of the sampling process for aluminum liner materials.

FIGURE 5.7 PHOTOGRAPHIC VIEW OF CONCRETE CORE SAMPLING (SU 1.4: REACTOR POOL INTERIOR CONCRETE WALLS)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 18 of 28 TLG Services, Inc. FIGURE 5.8 PHOTOGRAPHIC VIEWS OF ALUMINUM LINER SAMPLING (SU 1.2: THERMAL COLUMN LINER)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 19 of 28 TLG Services, Inc. Embedded Pipes Embedded pipes (and conduits, vent lines and similar items) within and around the biological shield were evaluated on a semi-quantitative base to rule out the presence of residual radioactivity deposited on their internal surfaces. Based upon knowledge obtained while removing the non-embedded portions of these pipes and similar items during the remediation work, these items were believed to not likely to contain any detectable radioactivity. During the remediation work, detectable radioactivity was not found to be present in any of the items wetted by pool water (except where it became concentrated on filters and the de-mineralizer resin). As previously discussed, any potential radioactivity within these pipes would likely result from being wetted with reactor pool water, having the readily detectable 99.95% Cs-137 and 0.05% Co-60 radionuclide mixture. Similarly, removed non-embedded portions of thermal column and beam port drains, vents, booster fans and duct work were all found to be free of detectable radioactivity. Also, as previously discussed, any potential radioactivity within the thermal column related items would likely result from the presence of activated graphite dust, having the readily detectable 0.78% Co-60, 93.25% Eu-152 and 5.96% Eu-154 radionuclide mixture.

Multiple techniques were used to confirm that residual radioactivity within these pipes was not present, checking the open ends of the pipes for evidence of contamination, swabbing the interior of the pipes to detect removable radioactive residues, and direct gamma measurements within the pipes.

The ends of each pipe were directly scanned and smear sampled for beta radioactivity, and then swabs were pushed / pulled through the embedded pipes to collect any residue on their internal surfaces. Swabs were then checked for the presence of radioactivity by direct beta counting. Figure 5.9 provides a photographic view of an embedded pipe being swab sampled.

T he interior pipe surfaces were then checked for the presence of radioactive contamination by inserting a small diameter NaI gamma detector (0.5 inch diameter-by- one inch long crystal) into the pipe and making a series of one-minute timed counts along the accessible length of the pipes. This detector was determined to be capable of detecting approximately 15,600 DPM/100 cm 2 (66% of the gross surface contamination criterion), for a one-inch diameter pipe, with 1 minute count times and counts being made at 10 cm intervals within each pipe. It was assumed that this MDL would be similar for slightly larger diameter pipe, where increased source term (due to increased surface area per unit length), would be offset by decreased detection efficiency from partial increased source distance from increased pipe diameter. Overall, accessible for gamma logging. Figure 5.

10 provides a photographic view of the gamma logging process.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 20 of 28 TLG Services, Inc. An intermediate conversion factor of 25,760 net CPM per Micro Curie (Cs-137) was derived for the NaI detector, used to convert net gross gamma count rates at a point of interest within an embedded pipe, to estimate of the amount of residual radioactive contamination within a defined interval within the pipe. This was done by making a series of counts using the NaI detector and a Cs-137 point source at varying distances (contact out to 50 cm) along an imaginary line to simulate radioactive contamination within a pipe, to determine detection efficiencies along that line. A final conversion factor was then derived to estimate surface contamination levels (in units of DPM/100 cm

2) using the assumption that all net counts at a point of interest within an embedded were due to residual radioactivity in front of, and 6.27 cm in back of the center of the detector.

FIGURE 5.9 PHOTOGRAPHIC VIEW OF AN EMBEDDED PIPE SWAB SAMPLING (SU 1.8.4: DRAIN LINE FROM BEAM PORT SHUTTER HOUSING)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 21 of 28 TLG Services, Inc. FIGURE 5.10 PHOTOGRAPHIC VIEW OF AN EMBEDDED PIPE BEING GAMMA LOGGED 5.3 SURVEY DESIGN PACKAGES The FSS survey activities for each of the survey unit designs were prepared and the implemented in accordance with a written and approved FSS procedure (provided in Appendix E). This procedure included specifications for the following activities: Preparation of the Facility for FSS and Checks for Extraneous Radioactive Materials Performance of the FSS Field Work Performance of Data Reduction and Evaluation Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 22 of 28 TLG Services, Inc. 5.4 BACKGROUND AND REFERENCE AREA MEASUREMENTS Background measurements were made to quantify levels of natural radioactivity emanating from construction materials. This included determination of beta surface radioactivity emanating from structural surfaces and gross gamma radioactivity levels within piping / conduit embedded within concrete structures. Sampling of structural materials for determination of the concentration of naturally occurring radionuclides was deemed not necessary, as all contamination of interest to the FSS (activation and fission product radionuclides) generally would not be observed naturally occurring in construction materials. Direct Beta Surface Radioactivity Eight sets of background measurements for direct beta surface radioactivity were made to represent each of the major types of material from which the reactor facility was constructed. This included the following types of materials: Sheet Rock / Wall Board Cement Block Linoleum Floor Tile (placed over wood, metal decking and concrete slab) Poured / Cast Concrete Structural Wood (planking and beam s) Structural Steel Aluminum Plate (placed over poured / cast concrete)

Seventeen timed direct beta surface contamination measurements were made for each type of material. Materials for these background measurements were found in and around the Washburn Shops building, sufficiently remote from the reactor facility to be unaffected by its operation, but close enough to be representative of the A single survey meter was used to make these background measurements to avoid introduction of any discrepancies from mixing of data derived from instruments with differing responses. Data resulting from these measurements is presented in Table 5.3, with the count data statistics, DPM/100 cm 2 values , detection sensitivities and detector characteristics provided for each background reference area.

The mean levels were subtracted from each of the individual gross direct beta surface contamination measurements within a particular survey unit of appropriate material type, to negate the beta contribution from naturally occurring radioactivity, as well as any inherent instrument background. Some survey units were comprised of mixed materials; in those cases, data from multiple material Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 23 of 28 TLG Services, Inc. background sets were combined to derive mean count rates to represent those mixed material survey units.

These direct beta reference background data sets were also established for use as reference areas, for use in performing WRS event that one or more individual measurement results exceeded the gross beta criterion for surface contamination, with the survey unit mean result being less than or equal to the 23,700 DPM/100 cm 2 gross beta criterion. However, upon evaluation of the data, it was determined that all individual surface contamination measurement data, for all survey units, did not exceed 5% of the total surface or removable contamination criterion, thus making the WRS test unnecessary for all survey units.

Mean direct beta levels for the various materials were observed to vary by as much as a factor of two, with minimum to maximum individual measurement data varying by as much as a factor of four. As would be expected, data sets comprised of natural earthen materials (i.e., concrete block, poured / cast concrete and ceramic tile) exhibited higher results. It should be noted that the magnitude of the background reference area results were small when compared to the 23,700 DPM

/100 cm 2 gross beta criteria for total surface contamination. The count rates for the 136 individual reference area direct beta measurements have an equivalent range of only 3.4 to 13.2% of the direct beta surface contamination criterion, and the range of the mean count rates comprising the eight material data sets have an equivalent range of approximately 5 to 10% of the direct beta surface contamination criterion. As will be observed from the FSS survey unit data presented in Section 6.0, subtraction of the contributions from natural radioactivity was not decisive in determining that all the FSS survey units met the gross beta criterion for total surface contamination.

TABLE 5.3 DIRECT BETA REFERENCE AREA BACKGROUND DATA

SUMMARY

REFERENCE AREA COUNT DATA Mean (CPM) 334 540 468 614 309 435 482 642 Median (CPM) 331 538 465 605 297 427 447 639 Range (CPM) 154 64 168 343 219 388 230 157 Min. (CPM) 248 506 380 481 218 298 402 576 Max. (CPM) 402 570 548 824 437 686 632 733 Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 24 of 28 TLG Services, Inc. TABLE 5.3 (continued)

DIRECT BETA REFERENCE AREA BACKGROUND DATA

SUMMARY

Statistic Sheet Rock Cement Block Linoleum Tile Poured Concrete Structural Wood Structural Steel Aluminum Plate (Over Concrete) Ceramic Tile Standard Deviation (CPM) 41 20 54 93 53 90 68 48 M (BKG sample population) 17 17 17 17 17 17 17 17 REFERENCE AREA DPM/100 cm 2 EQUIVALENT Mean (DPM/100 cm 2) 1 , 271 2 , 053 1 , 779 2 , 334 1 , 173 1 , 654 1 , 833 2 , 441 Maximum (DPM/100 cm 2) 1 , 529 2 , 167 2 , 084 3 , 133 1 , 662 2 , 608 2 , 403 2 , 787 Median (DPM/100 cm 2) 1 , 259 2 , 046 1 , 768 2 , 300 1 , 129 1 , 624 1 , 700 2 , 430 RESULTING DETECTION SENSITIVITIES FOR SURVEY UNITS WITH SIMILAR MATERIALS MDL, 1 min. direct (DPM/100 cm 2) using mean BKG 326 414 386 441 314 372 391 451 MDL, 1 min. direct (DPM/100 cm 2) using max. BKG 358 425 417 511 373 466 448 482 MDL, scanning (DPM/100 cm 2) using mean BKG 1 , 766 2 , 245 2 , 090 2 , 394 1 , 697 2 , 015 2 , 121 2 , 447 DETECTOR CHARACTERISTICS Beta detection efficiency (fractional

) 0.263 0.263 0.263 0.263 0.263 0.263 0.263 0.263 Detector Area (cm 2) 100 100 100 100 100 100 100 100 Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 25 of 28 TLG Services, Inc. Embedded Pipe Internal Radioactivity Gross gamma background levels were determined for the NaI detector that was used to gamma log piping / conduit embedded within concrete structures (e.g. the reactor biological shield). A background reference area was created by core drilling a 12 inch deep by two inch diameter hole into the top northwest corner of the biological shield, and inserting a 6 inch long section of unused, one-inch inside diameter aluminum pipe. This area was chosen as the background reference area because of its extreme distance from the reactor core and the known absence of any surface contamination in the area. Figure 5.11 provides a photographic image of this core hole with the Ludlum 44-62 0.5 inch diameter by 1 inch long NaI detector inserted into it. A series of one minute counts were made at various positions within the core hole (i.e., inserted at 3, 10 and 30 cm). Table 5.4 presents the results of those measurements, along with corresponding detection levels (as counts per minute above background). As can be observed, the count rate increased as the detector became more surrounded by the concrete material. The 1,453 count per minute mean value for the 30 cm in position was subtracted from all of the gross measurements made on pipe embedded in concrete. However, two survey units (3.5 and 3.7) had pipe that was not embedded in concrete; in that situation, the lower 961 count per minute mean value for the 3 cm position was used for background subtraction and MDL determination.

TABLE 5.4 EMBEDDED PIPE DETECTOR BACKGROUND DATA Location Count Distance From Start (cm) BKG (CPM) Group Means (CPM) Min. Detection Level (NCPM) 1 3 1 ,004 961 147 2 3 942 3 3 936 1 10 1 ,239 1 ,225 166 2 10 1 ,241 3 10 1 ,194 1 30 1 ,467 1 ,453 180 2 30 1 ,497 3 30 1 ,394 Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 26 of 28 TLG Services, Inc. FIGURE 5.11 VIEW OF EMBEDDED PIPE REFERENCE BACKGROUND AREA

5.5 QUALITY

CONTROL The FSS survey process was controlled by written procedure to assure that proper measurement techniques and protocols were used, such that the data generated by the FSS process remained valid. The following highlights the requirements that were imposed on the FSS measurement process to control the quality of the data.

Instrument Calibration Survey Instruments were calibrated by a third-party NRC / Agreement State-licensed vendor per requirements of WPI RPP-13, Calibration and Quality Control of Portable Radiological Survey Instruments

.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 27 of 28 TLG Services, Inc. Instrument Operability Checks Survey Instruments were function checked per RPP-13, Calibration and Quality Control of Portable Radiological Survey Instruments

. QC Limits were established for check source and background to establish an acceptable instrument response range to permit verification that the portable survey instrument is functioning appropriately during daily use. Radiological function checks were performed before and after use each shift. Any instrument(s) failing initial function tests were not used to generate FSS data until repaired and/or recalibrated and proper function verified. Data generated by any instrument failing post-use function checks was not be considered valid, back to the last passing function check.

Data and Calculation Checking All calculation and data entry operations were checked for errors by an individual that did not perform the original work. Work sheets were signed and dated by the individual performing the independent checks.

Replicate Measurements Five percent of the FSS surface contamination measurements (direct beta contamination, removable contamination smears and embedded pipe / conduit gamma logging) were repeated by an individual that did not perform the original measurements. These replicate measurements were distributed across the FSS work scope, with measurement locations selected on a random basis using a random number generator. Original and replicate results were compared. Results were expected to match within expected statistical variations. All results were determined to match; had they not, the discrepancy would have been investigated, and if necessary the survey results deemed invalid and new measurements made.

Control of Samples Collected FSS samples were logged onto sample data sheets, and the samples labeled with collection date, sample location and depth, survey unit and name of sample collector. Samples were kept in locked cabinets when not personally maintained by FSS personnel.

Chain of custody forms were used when samples were transferred to a third party for analysis. Receipt chain of custody forms were checked to assure that all samples were accounted for at the outside laboratory, had proper laboratory identifications assigned, and had properly assigned analysis parameters.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 5, Page 28 of 28 TLG Services, Inc.

5.6 DATA TESTING TO DEMONSTRATE COMPLIANCE WITH LICENSE TERMINATION CR ITERIA The WPI FSS plan, as guided by MARSSIM, provided for a graded approach in evaluating and testing the FSS data to determine if the license termination criteria had been met. This approach was two-tiered, which is described below:

Tier one: If all the individual datum, for all the applicable types of radiological determinations (e.g., total surface contamination, removable surface contamination, or volumetric concentration ) within a survey unit were found to be less than the applicable surface contamination or volumetric concentration limit, the survey unit would be deemed as having met the license termination criterion; and, if all the survey units were found to have met the applicable license termination criterion, the overall reactor facility would be deemed to meet the requirements for termination of the NRC license.

Tier two: If a survey unit was found to have one or more of its individual radiological determinations exceeding an applicable surface contamination or volumetric concentration population was less than or equal to the applicable surface contamination or volumetric concentration limit, the survey unit could be deemed as having met the applicable license termination criterion on a probabilistic basis. That is, a Null Hypothesis is statistical ly tested and rejected, which would demonstrate compliance with project criteria. The Null Hypothesis being that the residual radiological contamination levels exceed project criteria. Further, if this type of statistical testing were successful in demonstrating rejection of the Null Hypothesis, a population contained a sufficient amount of data to make the statistical testing valid. However, as demonstrated by the FSS data presented in Section 6.0, all individual radiological determinations, within all survey units, were less than the applicable surface contamination or volumetric concentration limits. As such, statistical testing of the data was not necessary, and no further explanation of the planned Null Hypothesis testing was needed or warranted.

Because there were multiple potential contaminants in the volumetric materials that were tested, compliance with concentration criteria for soil and activated materials (concrete and aluminum) were evaluated using the Sum-of-Fractions approach; where the resulting fractions of each of the radionuclides concentrations divided by its radionuclide specific criterion, were summed. The acceptance criterion was a sum-of-fractions less than or equal to one.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 1 of 26 TLG Services, Inc. 6.0 FSS RESULTS

6.1 OVERVIEW

OF FSS DATA Tables 6.1 through 6.5A-D present a summary of the results of the various types of measurements and sample analyses that were performed, with results summarized on a survey unit basis. As a minimum, these summary tables identify the maximum results found in each survey unit (in appropriate units and as a percent of the applicable criteria), detection sensitivity associated with the maximum results (in appropriate units and as a percent of the applicable criteria), and pertinent SU parameters, such as MARSSIM contamination potential class, surface area, and number of measurements / samples. Appendix F provides the complete data for each of the survey units, including one or more of the following to provide additional understanding and measurement point locations for the survey units: Grid maps, measurement point plots and photographs. As can be observed from these summary tables, the results of all individual measurements / sample analyses were less than the applicable license termination criteria previously presented in Tables 4.4 and 4.5, using the evaluation criteria discussed in Section 5.9. Of the 43 survey units included in this FSS, 35 had no detectable radioactivity. The other eight survey units had at least one individual result with detectable radioactivity, albeit with all individual results less than the license termination criteria. Detectable radioactivity being defined as having a result greater than the highest MDL or MDC for each of the various FSS measurement attributes. These maximum MDLs are indicated below, except for the volumetric concentration MDC determinations which are provided in the GEL report (presented in Appendix G).

Measurement Attribute Maximum MDL (percent of the 23,700 DPM/100 cm 2 surface contamination criterion) Total Beta Surface Contamination (direct timed measurements

) 2% Total Beta Surface Contamination (scanning) 10% Removable Beta Surface Contamination (smear / swab sample) 5% (of 10% of the surface contamination criterion)

Internal Pipe / Conduit Surface Contamination (NaI Gamma Logging) 66%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 2 of 26 TLG Services, Inc. Of those eight survey units with detectable radioactivity, seven (SUs 1.1, 1.2, 1.3, 1.4, 1.5, 1.6 and 1.8.4) were located in very close proximity to the former reactor core region within the biological shield; and the other remaining survey unit (SU 4.2) being the reactor water drain system. The results of these survey units are discussed in more detail in Section 6.2.

TABLE 6.1 TOTAL BETA SURFACE CONTAMINATION DIRECT MEASUREMENT RESULTS

SUMMARY

SU N o. SU Description Structural Surface Area (m 2) MARSSIM Classification Number of Measurements M ax. Result (DPM/100 cm 2) M ax. Result (% of Criteria) MDL (DPM/100 cm 2) MDL (% of Criteria) 1.0 BIOLOGICAL SHIELD / REACTOR POOL

1.1 Reactor

Pool Aluminum Liner 48 1 54 99 0.4% 391 1.7% 1.2 Thermal Column Aluminum Liner 2 1 21 1,779 7.5% 391 1.7% 1.3 Interior Reactor Pool Concrete Floor <1 1 na na na na na 1.4 Interior Reactor Pool Concrete Walls <1 1 na na na na na 1.5 Beam Port Tube

<1 1 12 0.3% 391 1.7% 1.6 Soil Under Removed Concrete Floor Area 1.3 1 na na na na na 1.7V Top of Bio

-Shield, Vertical Surfaces 9 1 15 0.1% 441 1.9% 1.7H Top of Bio

-Shield, Horizontal Surfaces 29 1 39 481 2.0% 423 1.8% 1.8 Embedded Piping / Conduits (in Biological Shield)

1.8.1 Drain

Lines from Reactor Pool Scupper Drains na 1 See embedded pipe measurements summary 1.8.2 Return Line from Pool Water Treatment System na 1 See embedded pipe measurements summary 1.8.3 Vent line above Beam Port na 1 See embedded pipe measurements summary 1.8.4 Drain Line from Beam Port Shutter Housing na 1 See embedded pipe measurements summary Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 3 of 26 TLG Services, Inc. TABLE 6.1 TOTAL BETA SURFACE CONTAMINATION DIRECT MEASUREMENT RESULTS

SUMMARY

SU N o. SU Description Structural Surface Area (m 2) MARSSIM Classification Number of Measurements M ax. Result (DPM/100 cm 2) M ax. Result (% of Criteria) MDL (DPM/100 cm 2) MDL (% of Criteria) 1.8.5 Drain from Reactor Pool (Feed to Rx Water Treatment System) na 1 See embedded pipe measurements summary 1.8.6 Vent Line from Thermal Column (Upper-Left) na 1 See embedded pipe measurements summary 1.8.7 Vent Line from Thermal Column (Lower-Left) na 1 See embedded pipe measurements summary 1.8.8 Vent Line from Thermal Column (Upper-Right) na 1 See embedded pipe measurements summary 1.8.9 Vent Line from Thermal Column (Lower-Right) na 1 See embedded pipe measurements summary 1.8.10 Vent Line from Thermal Column (Top- Center) na 1 See embedded pipe measurements summary 1.9 Exterior Walls of Biological Shield

1.9.1 North

Exterior Bio

-Shield Wall 21 2 22 -327 -1.4% 441 1.9% 1.9.2 West Exterior Bio

-Shield Wall 18 1 17 -220 -0.9% 441 1.9% 1.9.3 Thermal Column Shield Door 5 1 24 0.4% 372 1.6% 1.9.4 East Exterior Bio

-Shield Wall 18 1 22 -171 -0.7% 441 1.9% 2.0 GROUND FLOOR REACTOR FACILITY

2.1 Ground

Floor of Reactor Room; Floor and South Wall (West of Reactor Centerline) 73 1 26 247 1.0% 400 1.7% 2.2 Ground Floor of Reactor Room; Lower North & West Walls (West of Reactor Centerline) 35 2 45 410 1.7% 373 1.6% 2.3 Ground Floor of Reactor Room; Floor and South Walls (East of Reactor Centerline) 92 1 31 209 0.9% 400 1.7%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 4 of 26 TLG Services, Inc. TABLE 6.1 TOTAL BETA SURFACE CONTAMINATION DIRECT MEASUREMENT RESULTS

SUMMARY

SU N o. SU Description Structural Surface Area (m 2) MARSSIM Classification Number of Measurements M ax. Result (DPM/100 cm 2) M ax. Result (% of Criteria) MDL (DPM/100 cm 2) MDL (% of Criteria) 2.4 Ground Floor Reactor Room; Lower North and East Walls (East of Reactor Centerline) 29 2 37 376 1.6% 373 1.6% 2.5 Ground Floor Reactor Room; Upper Walls, Ceiling and Exterior of Suspended Equipment (West of Reactor Centerline) 84 2 28 657 2.8% 354 1.5% 2.6 Ground Floor Reactor Room; Upper Walls, Ceiling, and Exterior of Suspended Equipment (East of Reactor Centerline) 93 2 25 474 2.0% 35 4 1.5% 2.7 Ground Floor Reactor Room; RAM Storage Room (Lower Walls & Floor) 28 1 29 557 2.4% 377 1.6% 2.8 Ground Floor Reactor Room; RAM Storage Room (Upper Walls, Ceiling, and exterior of suspended equipment) 28 1 31 919 3.9% 354 1.5% 2.9 Ground Floor Reactor Room; Interior and Exterior Walls, Ceiling and Roof of Dark Room 41 1 19 626 2.6% 320 1.4% 2.10 Reactor Room Stairs 4.3 1 30 -11 0.0% 372 1.6% 3.0 FIRST FLOOR REACTOR FACILITY

3.1 Reactor

Office Lower Walls and Floor 54 3 18 695 2.9% 326 1.4% 3.2 Reactor Office Upper Walls and Ceiling 54 3 17 934 3.9% 326 1.4% 3.3 First Floor Reactor Room; Lower Walls and Floor (West of Reactor Centerline) 80 2 21 216 0.9% 357 1.5% 3.4 First Floor Reactor Room; Upper Walls and Ceiling (West of Reactor Centerline) 99 2 28 303 1.3% 377 1.6%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 5 of 26 TLG Services, Inc. TABLE 6.1 TOTAL BETA SURFACE CONTAMINATION DIRECT MEASUREMENT RESULTS

SUMMARY

SU N o. SU Description Structural Surface Area (m 2) MARSSIM Classification Number of Measurements M ax. Result (DPM/100 cm 2) M ax. Result (% of Criteria) MDL (DPM/100 cm 2) MDL (% of Criteria) 3.5 First Floor Reactor Room; Lower Walls and Floor (East of Reactor Centerline) 91 1 27 308 1.3% 357 1.5% 3.6 First Floor Reactor Room; Upper Walls and Ceiling (East of Reactor Centerline) 115 2 32 276 1.2% 377 1.6% 3.7 First Floor Reactor Room; Tool Closet-Lower Walls and Floor 42 1 22 64 0.3% 357 1.5% 3.8 First Floor Reactor Room; Tool Closet (Upper Walls and Ceiling) 42 1 23 0.3% 350 1.5% 4.0 OTHER AREAS / ITEMS

4.1 Exhaust

Ventilation System (Interior Duct and Equipment Surfaces) na 1 37 760 3.2% 372 1.6% 4.2 Reactor Water Treatment System Drain Piping to Sanitary Sewer System na 1 See embedded pipe measurements summary 4.3 HVAC Units (3 Free

-Standing Units) na 1 39 255 1.1% 372 1.6%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 6 of 26 TLG Services, Inc.

TABLE 6.2 BETA SURFACE CONTAMINATION SCAN RESULTS

SUMMARY

SU N o. SU Description Structural Surface Area (m 2) MARSSIM Classification Scan Coverage

(%) Scanning MDL (DPM/100 cm 2) Areas with Elevated Radioactivity M ax. Result (DPM/100 cm 2) 1.0 BIOLOGICAL SHIELD / REACTOR POOL 1.1 Reactor Pool Aluminum Liner 48 1 100% 2 ,121 None na 1.2 Thermal Column Aluminum Liner 2 1 100% 2 ,121 Yes, see note 1 below. 1 ,600 1.3 Interior Reactor Pool Concrete Floor

<1 1 100% 2,394 None na 1.4 Interior Reactor Pool Concrete Walls

<1 1 100% of exposed concrete 2 ,394 None na 1.5 Beam Port Tube

<1 1 100% 2 ,015 None na 1.6 Soil Under Removed Concrete Floor Area 1.3 1 100% of exposed soil 2 ,394 None na 1.7V Top of Bio

-Shield, Vertical Surfaces 9 1 100% 2 ,293 None na 1.7H Top of Bio

-Shield, Horizontal Surfaces 29 1 100% 2 ,293 None na 1.8 Embedded Piping / Conduit (in Biological Shield)

1.8.1 Drain

Lines from Reactor Pool Scupper Drains na 1 See embedded pipe measurements summary 1.8.2 Return Line from Pool Water Treatment System na 1 See embedded pipe measurements summary 1.8.3 Vent line above Beam Port na 1 See embedded pipe measurements summary 1.8.4 Drain Line from Beam Port Shutter Housing na 1 See embedded pipe measurements summary Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 7 of 26 TLG Services, Inc. TABLE 6.2 BETA SURFACE CONTAMINATION SCAN RESULTS

SUMMARY

SU N o. SU Description Structural Surface Area (m 2) MARSSIM Classification Scan Coverage

(%) Scanning MDL (DPM/100 cm 2) Areas with Elevated Radioactivity M ax. Result (DPM/100 cm 2) 1.8.5 Drain from Reactor Pool (Feed to Rx Water Treatment System) na 1 See embedded pipe measurements summary 1.8.6 Vent Line from Thermal Column (Upper

-Left) na 1 See embedded pipe measurements summary 1.8.7 Vent Line from Thermal Column (Lower

-Left) na 1 See embedded pipe measurements summary 1.8.8 Vent Line from Thermal Column (Upper

-Right) na 1 See embedded pipe measurements summary 1.8.9 Vent Line from Thermal Column (Lower

-Right) na 1 See embedded pipe measurements summary 1.8.10 Vent Line from Thermal Column (Top

- Center) na 1 See embedded pipe measurements summary 1.9 Exterior Walls of Biological Shield

1.9.1 North

Exterior Bio

-Shield Wall 21 2 100% 2 ,394 None na 1.9.2 West Exterior Bio

-Shield Wall 18 1 100% 2 ,394 None na 1.9.3 Thermal Column Shield Door 5 1 100% 2 ,015 None na 1.9.4 East Exterior Bio

-Shield Wall 18 1 100% 2 ,394 None na 2.0 GROUND FLOOR REACTOR FACILITY

2.1 Ground

Floor of Reactor Room; Floor and South Wall (West of Reactor Centerline) 73 1 100% 2 ,169 None na 2.2 Ground Floor of Reactor Room; Lower North & West Walls (West of Reactor Centerline) 35 2 100% 2 ,020 None na Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 8 of 26 TLG Services, Inc. TABLE 6.2 BETA SURFACE CONTAMINATION SCAN RESULTS

SUMMARY

SU N o. SU Description Structural Surface Area (m 2) MARSSIM Classification Scan Coverage

(%) Scanning MDL (DPM/100 cm 2) Areas with Elevated Radioactivity M ax. Result (DPM/100 cm 2) 2.3 Ground Floor of Reactor Room; Floor and South Walls (East of Reactor Centerline) 92 1 100% 2 ,169 None na 2.4 Ground Floor Reactor Room; Lower North and East Walls (East of Reactor Centerline) 29 2 100% 2 ,020 None na 2.5 Ground Floor Reactor Room; Upper Walls, Ceiling and Exterior of Suspended Equipment (West of Reactor Centerline) 84 2 100% 1 ,918 None na 2.6 Ground Floor Reactor Room; Upper Walls, Ceiling, and Exterior of Suspended Equipment (East of Reactor Centerline) 93 2 100% 1 ,918 None na 2.7 Ground Floor Reactor Room; RAM Storage Room (Lower Walls & Floor) 28 1 100% 2 ,015 None na 2.8 Ground Floor Reactor Room; RAM Storage Room (Upper Walls, Ceiling, and exterior of suspended equipment) 28 1 100% 1 ,918 None na 2.9 Ground Floor Reactor Room; Interior and Exterior Walls, Ceiling and Roof of Dark Room 41 1 100% 1 ,732 None na 2.10 Reactor Room Stairs 4.3 1 100% 2 ,015 None na 3.0 FIRST FLOOR REACTOR FACILITY

3.1 Reactor

Office Lower Walls and Floor 54 3 100% 1 ,766 None na Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 9 of 26 TLG Services, Inc. TABLE 6.2 BETA SURFACE CONTAMINATION SCAN RESULTS

SUMMARY

SU N o. SU Description Structural Surface Area (m 2) MARSSIM Classification Scan Coverage

(%) Scanning MDL (DPM/100 cm 2) Areas with Elevated Radioactivity M ax. Result (DPM/100 cm 2) 3.2 Reactor Office Upper Walls and Ceiling 54 3 100% 1 ,766 None na 3.3 First Floor Reactor Room; Lower Walls and Floor (West of Reactor Centerline) 80 2 25% 1 ,935 None na 3.4 First Floor Reactor Room; Upper Walls and Ceiling (West of Reactor Centerline) 99 2 25% 2 ,043 None na 3.5 First Floor Reactor Room; Lower Walls and Floor (East of Reactor Centerline) 91 1 100% 1,935 None na 3.6 First Floor Reactor Room; Upper Walls and Ceiling (East of Reactor Centerline) 115 2 25% 2 ,043 None na 3.7 First Floor Reactor Room; Tool Closet

-Lower Walls and Floor 42 1 100% 1 ,935 None na 3.8 First Floor Reactor Room; Tool Closet (Upper Walls and Ceiling) 42 1 100% 1 ,895 None na 4.0 OTHER AREAS / ITEMS

4.1 Exhaust

Ventilation System (Interior Duct and Equipment Surfaces) na 1 100% of available access locations 2 ,015 None na 4.2 Reactor Water Treatment System Drain Piping to Sanitary Sewer System na 1 See embedded pipe measurements summary 4.3 HVAC Units (3 Free

-Standing Units) na 1 100% 2 ,015 None na Note 1: The thermal column liner was observed to have detectable surface beta radioactivity within a 30 cm wide section of the remaining liner, closest to the former reactor core area. Results were similar for the liner floor, ceiling and two side walls. The maximum scanning results in these areas was approximately twice that of the background results observed in the reference measurement areas.

Four volumetric samples of the liner material were collected from the areas where the maximum scan results were observed, with one sample collected from each of the surfaces.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 10 of 26 TLG Services, Inc. TABLE 6.3 REMOVABLE BETA SURFACE CONTAMINATION SMEAR SAMPLE RESULTS

SUMMARY

SU N o. SU Description Structural Surface Area (m 2) MARSSIM Classification number of Samples M ax. Result (DPM/100 cm 2) M ax. Result (% of Criteria) MDL (DPM/100 cm 2) MDL (% of Criteria) 1.0 BIOLOGICAL SHIELD / REACTOR POOL

1.1 Reactor

Pool Aluminum Liner 48 1 54 54 2.3% 112 4.7% 1.2 Thermal Column Aluminum Liner 2 1 20 73 3.1% 133 5.6% 1.3 Interior Reactor Pool Concrete Floor

<1 1 na na na na na 1.4 Interior Reactor Pool Concrete Walls

<1 1 na na na na na 1.5 Beam Port Tube

<1 1 12 49 2.1% 109 4.6% 1.6 Soil Under Removed Concrete Floor Area 1.3 1 na na na na na 1.7V Top of Bio

-Shield, Vertic al Surfaces 9 1 15 10 0.4% 112 4.7% 1.7H Top of Bio

-Shield, Horizontal Surfaces 29 1 39 78 3.3% 104 4.4% 1.8 Embedded Piping /Conduit (in Biological Shield)

1.8.1 Drain

Lines from Reactor Pool Scupper Drains na 1 See embedded pipe measurements summary 1.8.2 Return Line from Pool Water Treatment System na 1 See embedded pipe measurements summary 1.8.3 Vent line above Beam Port na 1 See embedded pipe measurements summary 1.8.4 Drain Line from Beam Port Shutter Housing na 1 See embedded pipe measurement s summary 1.8.5 Drain from Reactor Pool (Feed to Rx Water Treatment System) na 1 See embedded pipe measurements summary Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 11 of 26 TLG Services, Inc. TABLE 6.3 REMOVABLE BETA SURFACE CONTAMINATION SMEAR SAMPLE RESULTS

SUMMARY

SU N o. SU Description Structural Surface Area (m 2) MARSSIM Classification number of Samples M ax. Result (DPM/100 cm 2) M ax. Result (% of Criteria) MDL (DPM/100 cm 2) MDL (% of Criteria) 1.8.6 Vent Line from Thermal Column (Upper

-Left) na 1 See embedded pipe measurements summary 1.8.7 Vent Line from Thermal Column (L ower-Left) na 1 See embedded pipe measurements summary 1.8.8 Vent Line from Thermal Column (Upper

-Right) na 1 See embedded pipe measurements summary 1.8.9 Vent Line from Thermal Column (Lower

-Right) na 1 See embedded pipe measurements summary 1.8.1 0 Vent Line from Thermal Column (Top

- Center) na 1 See embedded pipe measurements summary 1.9 Exterior Walls of Biological Shield

1.9.1 North

Exterior Bio

-Shield Wall 21 2 22 54 2.3% 104 4.4% 1.9.2 West Exterior Bio

-Shield Wall 18 1 17 34 1.4% 114 4.8% 1.9.3 Thermal Column Shield Door 5 1 24 15 0.6% 109 4.6% 1.9.4 East Exterior Bio

-Shield Wall 18 1 22 34 1.4% 112 4.7% 2.0 GROUND FLOOR REACTOR FACILITY

2.1 Ground

Floor of Reactor Room; Floor and South Wall (West of Reactor Centerline) 73 1 26 44 1.9% 1 16 4.9% 2.2 Ground Floor of Reactor Room; Lower North & West Walls (West of Reactor Centerline) 35 2 45 44 1.9% 109 4.6% 2.3 Ground Floor of Reactor Room; Floor and South Walls (East of Reactor Centerline) 92 1 31 58 2.5% 107 4.5%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 12 of 26 TLG Services, Inc. TABLE 6.3 REMOVABLE BETA SURFACE CONTAMINATION SMEAR SAMPLE RESULTS

SUMMARY

SU N o. SU Description Structural Surface Area (m 2) MARSSIM Classification number of Samples M ax. Result (DPM/100 cm 2) M ax. Result (% of Criteria) MDL (DPM/100 cm 2) MDL (% of Criteria) 2.4 Ground Floor Reactor Room; Lower North and East Walls (East of Reactor Centerline) 29 2 37 63 2.7% 102 4.3% 2.5 Ground Floor Reactor Room; Upper Walls, Ceiling and Exterior of Suspended Equipment (West of Reactor Centerline) 84 2 28 39 1.6% 107 4.5% 2.6 Ground Floor Reactor Room; Upper Walls, Ceiling, and Exterior of Suspended Equipment (East of Reactor Centerline) 93 2 34 34 1.4% 114 4.8% 2.7 Ground Floor Reactor Room; RAM Storage Room (Lower Walls & Floor) 28 1 29 39 1.6% 112 4.7% 2.8 Ground Floor Reactor Room; RAM Storage Room (Upper Walls, Ceiling, and exterior of suspended equipment) 28 1 31 49 2.1% 114 4.8% 2.9 Ground Floor Reactor Room; Interior and Exterior Walls, Ceiling and Roof of Dark Room 41 1 19 24 1.0% 114 4.8% 2.10 Reactor Room Stairs 4.3 1 30 73 3.1% 1 04 4.4% 3.0 FIRST FLOOR REACTOR FACILITY

3.1 Reactor

Office Lower Walls and Floor 54 3 18 44 1.9% 114 4.8% 3.2 Reactor Office Upper Walls and Ceiling 54 3 17 58 2.5% 107 4.5%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 13 of 26 TLG Services, Inc. TABLE 6.3 REMOVABLE BETA SURFACE CONTAMINATION SMEAR SAMPLE RESULTS

SUMMARY

SU N o. SU Description Structural Surface Area (m 2) MARSSIM Classification number of Samples M ax. Result (DPM/100 cm 2) M ax. Result (% of Criteria) MDL (DPM/100 cm 2) MDL (% of Criteria) 3.3 First Floor Reactor Room; Lower Walls and Floor (West of Reactor Centerline) 80 2 21 78 3.3% 112 4.7% 3.4 First Floor Reactor Room; Upper Walls and Ceiling (West of Reactor Centerline) 99 2 33 63 2.7% 112 4.7% 3.5 First Floor Reactor Room; Lower Walls and Floor (East of Reactor Centerline) 91 1 27 63 2.7% 112 4.7% 3.6 First Floor Reactor Room; Upper Walls and Ceiling (East of Reactor Centerline) 115 2 35 54 2.3% 109 4.6% 3.7 First Floor Reactor Room; Tool Closet

-Lower Walls and Floor 42 1 22 39 1.6% 107 4.5% 3.8 First Floor Reactor Room, Tool Closet (Upper Walls and Ceiling

) 42 1 23 34 1.4% 119 5.0% 4.0 OTHER AREAS / ITEMS

4.1 Exhaust

Ventilation System (Interior Duct and Equipment Surfaces) na 1 37 78 3.3% 107 4.5% 4.2 Reactor Water Treatment System Drain Piping to Sanitary Sewer System na 1 See embedded pipe measurements summary 4.3 HVAC Units (3 Free

-Standing Units) na 1 39 58 2.5% 104 4.4%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 14 of 26 TLG Services, Inc. TABLE 6.4 VOLUMETRIC RADIOACTIVITY SAMPLE RESULTS

SUMMARY

(a) SU N o. SU Description Sample Locations Sample Media SU Mean SOF (Top Layer) M ax Sample Result, SOF Top layer)

MAX Sample Result, SOF (Underlying Layer)

Radionuclide Makeup 1.1 Reactor Pool Aluminum Liner 4 Aluminum 0.03 0.11 na Fe-55: 93%, Co

-60: 7% 1.2 Thermal Column Aluminum Liner 4 Aluminum 0.24 0.27 na Fe-55: 92%, Co

-60: 7%, Eu

-152: 1% 1.3 Interior Reactor Pool Concrete Floor 2 Concrete 0.18 0.18 0.00 Co-60: 12%, Cs

-134: 2%, Eu

-152: 82%, Eu-154: 4% 1.4 Interior Reactor Pool Concrete Walls 4 Concrete 0.06 0.22 0.08 Co-60: 27%, Cs

-134: 2%, Eu

-152: 68%, Eu-154: 3% 1.5 Beam Port Tube Sampled and included as a part of SU 1.1, as sample #3 (SOF= 0.11) (b) 1.6 Soil Under Removed Concrete Floor Area 4 Soil 0.16 0.19 na (c) Co-60: 16%, Cs

-134: 2%, Eu

-152: 78%, Eu-154: 4% 4.2 Reactor Water Treatment System Drain Piping to Sanitary Sewer System 1 Sediment (d) 0.40 0.4 0 na Co-60: 3%, Cs

-137: 97% (a)Table only shows survey units that were subject to volumetric sampling. (b)This sample was obtained at the approximate location where the beam port tube penetrated the east wall of the aluminum pool liner. Surrounding pool liner had been previously removed as a part of the decommissioning work scope. (c)Soil was not available; bedrock was encountered at an approximate15 cm depth. (d) treatment system valve sump; sediment was not observed up- or down-stream of this location. Sampling removed all of the sediment that was present; therefore, this sample is no longer representative of this remaining piping system.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 15 of 26 TLG Services, Inc.

TABLE 6.5A EMBEDDED PIPE / CONDUIT INTERNAL CONTAMINATION ASSESSMENT

SUMMARY

DIRECT BETA CONTAMINATION SCAN AT LINE ENDS (a) SU N o. SU Description Approximate Line Length, c m Line ID, cm Maximum (DPM/100 cm 2) Maximum Result (% of Criteria) Scanning MDL (DPM/100 cm 2) MDL (% of C riteria) 1.8.1 Drain Lines from Reactor Pool Scupper Drains 850 2.54 None Detected None Detected 1 ,932 8% 1.8.2 Return Line from Pool Water Treatment System 370 2.54 None Detected None Detected 1 ,932 8% 1.8.3 Vent line above Beam Port 180 2.54 None Detected None Detect ed 1 ,932 8% 1.8.4 Drain Line from Beam Port Shutter Housing 150 2.54 None Detected None Detected 1 ,932 8% 1.8.5 Drain from Reactor Pool (Feed to Rx Water Treatment System) 326 5.08 None Detected None Detected 2 ,160 9% 1.8.6 Vent Line from Thermal Column (Upper-Left) 99 2.54 None Detected None Detected 1 ,932 8% 1.8.7 Vent Line from Thermal Column (Lower

-Left) 99 2.54 None Detected None Detected 1 ,932 8% 1.8.8 Vent Line from Thermal Column (Upper

-Right) 99 2.54 None Detected None Detected 1 ,932 8% 1.8.9 Vent Line from Thermal Column (Lower

-Right) 99 2.54 None Detected None Detected 1 ,932 8% 1.8.10 Vent Line from Thermal Column (Top

- Center) 140 5.08 None Detected None Detected 1 ,932 8% 3.5 First Floor Reactor Room; Lower Walls and Floor, East of Reactor Centerline (Includes pipe stub in floor and line running to former hold

-up tank) 500 2.54 and 3.2 None Detected None Detected 2 ,116 9%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 16 of 26 TLG Services, Inc.

TABLE 6.5A EMBEDDED PIPE / CONDUIT INTERNAL CONTAMINATION ASSESSMENT

SUMMARY

DIRECT BETA CONTAMINATION SCAN AT LINE ENDS (a) SU N o. SU Description Approximate Line Length, c m Line ID, cm Maximum (DPM/100 cm 2) Maximum Result (% of Criteria) Scanning MDL (DPM/100 cm 2) MDL (% of C riteria) 3.7 First Floor Reactor Room; Tool Closet

-Lower Walls and Floor (Includes remaining safety shower floor drain and pipe stub) 63 5.08 None Detected None Detected 1 ,807 8% 4.2 Reactor Water Treatment System Drain Piping to Sanitary Sewer System (includes 3 connecting pipe segments) 126 3.8 and 10.2 None Detected None Detected 2 ,327 10% (a)Table only shows survey units that contained embedded piping / conduits or other piping remaining in place that required assessment.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 17 of 26 TLG Services, Inc.

TABLE 6.5B EMBEDDED PIPE / CONDUIT INTERNAL CONTAMINATION ASSESSMENT

SUMMARY

REMOVABLE BETA CONTAMINATION AT LINE ENDS - SMEAR SAMPLES (a) SU N o. S U Description Maximum (DPM/100 cm 2) Maximum Result ,  % of Criteria (Removable) MDL, (DPM/100 cm 2) MDL % of Criteria (Removable) 1.8.1 Drain Lines from Reactor Pool Scupper Drains 29 1.2% 104 4.4% 1.8.2 Return Line from Pool Water Treatment System 15 0.6% 104 4.4% 1.8.3 Vent line above Beam Port 34 1.4% 104 4.4% 1.8.4 Drain Line from Beam Port Shutter Housing 15 0.6% 104 4.4% 1.8.5 Drain from Reactor Pool (Feed to Rx Water Treatment System) 1.2% 104 4.4% 1.8.6 Vent Line from Thermal Column (Upper-Left) 15 0.6% 104 4.4% 1.8.7 Vent Line from Thermal Column (Lower-Left) 0.6% 104 4.4% 1.8.8 Vent Line from Thermal Column (Upper-Right) 44 1.9% 104 4.4% 1.8.9 Vent Line from Thermal Column (Lower-Right) 1.0% 104 4.4% 1.8.10 Vent Line from Thermal Column (Top- Center) 5 0.2% 104 4.4% 3.5 First Floor Reactor Room; Lower Walls and Floor, East of Reactor Centerline (Includes pipe stub in floor and line running to former hold-up tank) 0.2% 91 3.8%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 18 of 26 TLG Services, Inc. TABLE 6.5B EMBEDDED PIPE / CONDUIT INTERNAL CONTAMINATION ASSESSMENT

SUMMARY

REMOVABLE BETA CONTAMINATION AT LINE ENDS - SMEAR SAMPLES (a) SU N o. S U Description Maximum (DPM/100 cm 2) Maximum Result ,  % of Criteria (Removable) MDL, (DPM/100 cm 2) MDL % of Criteria (Removable) 3.7 First Floor Reactor Room; Tool Closet-Lower Walls and Floor (Includes remaining safety shower floor drain and pipe stub) 19 0.8% 96 4.1% 4.2 Reactor Water Treatment System Drain Piping to Sanitary Sewer System (includes 3 connecting pipe segments) 0.2% 127 5.4% (a)Table only shows survey units that contained embedded piping / conduits or other piping remaining in place that required assessment.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 19 of 26 TLG Services, Inc.

TABLE 6.5C EMBEDDED PIPE / CONDUIT INTERNAL CONTAMINATION ASSESSMENT

SUMMARY

REMOVABLE BETA CONTAMINATION INTERIOR LINE SURFAC ES - SWAB SAMPLES (a) SU N o. SU Description Maximum (DPM/S wab) Maximum Result (% of Criteria (b)) MDL, (DPM/S wab) MDL (% of Criteria (b)) 1.8.1 Drain Lines from Reactor Pool Scupper Drains 44 1.9% 114 4.8% 1.8.2 Return Line from Pool Water Treatment System 15 0.6% 121 5.1% 1.8.3 Vent line above Beam Port 0.8% 121 5.1% 1.8.4 Drain Line from Beam Port Shutter Housing 19 0.8% 102 4.3% 1.8.5 Drain from Reactor Pool (Feed to Rx Water Treatment System) 93 3.9% 135 5.7% 1.8.6 Vent Line from Thermal Column (Upper-Left) 1.2% 125 5.3% 1.8.7 Vent Line from Thermal Column (Lower-Left) 58 2.5% 109 4.6% 1.8.8 Vent Line from Thermal Column (Upper-Right) 10 0.4% 109 4.6% 1.8.9 Vent Line from Thermal Column (Lower-Right) 49 2.1% 135 5.7% 1.8.10 Vent Line from Thermal Column (Top- Center) 73 3.1% 135 5.7%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 20 of 26 TLG Services, Inc. TABLE 6.5C EMBEDDED PIPE / CONDUIT INTERNAL CONTAMINATION ASSESSMENT

SUMMARY

REMOVABLE BETA CONTAMINATION INTERIOR LINE SURFAC ES - SWAB SAMPLES (a) SU N o. SU Description Maximum (DPM/S wab) Maximum Result (% of Criteria (b)) MDL, (DPM/S wab) MDL (% of Criteria (b)) 3.5 First Floor Reactor Room; Lower Walls and Floor, East of Reactor Centerline (Includes pipe stub in floor and line running to former hold-up tank) 0.2% 148 6.2% 3.7 First Floor Reactor Room; Tool Closet-Lower Walls and Floor (Includes remaining safety shower floor drain and pipe stub) 0 0.0% 148 6.2% 4.2 Reactor Water Treatment System Drain Piping to Sanitary Sewer System (includes 3 connecting pipe segments) na na na na (a) Table only shows survey units that contained embedded piping / conduits or other piping remaining in place that required assessment. (b) Conservatively assumes result is compared to the removable criterion of 10% of the 23,700 DPM/100 cm 2, whereas any detected removable contamination would have been collected from a much larger surface area.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 21 of 26 TLG Services, Inc. TABLE 6.5D EMBEDDED PIPE / CONDUIT INTERNAL CONTAMINATION ASSESSMENT

SUMMARY

SURFACE CONTAMINATION INTERIOR LINE SURFACES

- N aI GAMMA LOGGING (a) SU N o. SU Description L ine Mean (DPM/100 cm 2) Maximum (DPM/100 cm 2) Maximum Result (% of Criteria) MDL (DPM/100 cm 2) MDL (% of Criteria) % of Line Accessible for Logging 1.8.1 Drain Lines from Reactor Pool Scupper Drains

-4 , 397 3 ,303 13.9% 15 ,540 65.6% 55% 1.8.2 Return Line from Pool Water Treatment System

-7 , 861 5 ,199 21.9% 15 ,540 65.6% 42% 1.8.3 Vent line above Beam Port

-18 ,814 -1 , 867 -7.9% 15 ,540 65.6% 12% 1.8.4 Drain Line from Beam Port Shutter Housing

-4 , 428 20,451 (b) 86.3% 15 ,540 65.6% 34% 1.8.5 Drain from Reactor Poo l (Feed to Rx Water Treatment System) 1 ,730 10 ,183 43.0% 15 ,540 65.6% 100% 1.8.6 Vent Line from Thermal Column (Upper

-Left) -1 , 068 15 ,626 (c) 65.9% 15 ,540 65.6% 90% 1.8.7 Vent Line from Thermal Column (Lower

-Left) -1 , 795 14 ,678 (c) 61.9% 15 ,540 65.6% 92% 1.8.8 Vent Line from Thermal Column (Upper

-Right) -2 , 677 6,319 (d) 26.7% 15 ,540 65.6% 86% 1.8.9 Vent Line from Thermal Column (Lower

-Right) -3 , 091 9,938 (d) 41.9% 15 ,540 65.6% 85% 1.8.10 Vent Line from Thermal Column (Top

- Center) 1 ,497 9 ,407 (e) 39.7% 15 ,540 65.6% 100% 3.5 First Floor Reactor Room; Lower Walls and Floor, East of Reactor Centerline (Includes pipe stub in floor and line running to former hold-up tank) -17 ,840 (f) -10 ,829 (f) -45.7% 15 ,540 65.6% 18% 3.7 First Floor Reactor Room; Tool Clos et-Lower Walls and Floor (Includes remaining safety shower floor drain and pipe stub)

-19 ,811 (f) -13 ,213 (f) -55.8% 15 ,540 65.6% 100%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 22 of 26 TLG Services, Inc. TABLE 6.5D EMBEDDED PIPE / CONDUIT INTERNAL CONTAMINATION ASSESSMENT

SUMMARY

SURFACE CONTAMINATION INTERIOR LINE SURFACES

- N aI GAMMA LOGGING (a) SU N o. SU Description L ine Mean (DPM/100 cm 2) Maximum (DPM/100 cm 2) Maximum Result (% of Criteria) MDL (DPM/100 cm 2) MDL (% of Criteria) % of Line Accessible for Logging 4.2 Reactor Water Treatment System Drain Piping to Sanitary Sewer System (includes 3 connecting pipe segments)

-8 , 166 13 , 5 86 57.3% 15 ,540 65.6% 100% (a)Table only shows survey units that contained embedded piping / conduits or other piping remaining in-place that required assessment. (b)This section of the pipe is located in close proximity to an area the experienced a slight amount of neutron activation. Volumetric Aluminum sample SU 1.1 #3 was collected from the remaining end of the beam port tube, which is in close proximity to the measurement location for the maximum result. Results of that aluminum sample had a SOF result of 0.11. This end of the drain line is also embedded in concrete that experienced neutron activation, and is in close proximity to where volumetric concrete sample SU 1.4 #4 was obtained, which had a SOF result of 0.202. (c)These maximum results occurred where the vent lines join ed an area that was determined to be slightly neutron activated, and the maximum result locations are in close proximity to where volumetric sample SU 1.2 #2 was obtained. This sample had a SOF result of 0

.273. (d)These maximum results occurred where the vent lines joined an area that was determined to be slightly neutron activated, and the maximum result locations are in close proximity to where volumetric sample SU 1.2 #3 was obtained. This sample had a SOF result of 0.23. This maximum result occurred where the vent line join ed the thermal column ceiling liner, an area that was determined to be neutron activated, and is in close proximity to the aluminum samples that were obtained for SU 1.2. (e)This maximum result occurred where the vent line joined the thermal column ceiling liner, an area that was determined to be neutron activated; it was in close proximity to the location where volumetric sample SU 1.2 no. 4 was obtained. This sample had a SOF result of 0.2.2. (f)These pipes are not embedded in, nor located near, large amounts of concrete, and/or consist of heavy-walled cast iron which is likely to have reduced background radiation levels; hence the extreme negative values.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 23 of 26 TLG Services, Inc.

6.2 DISCUSSION

REGARDING SURVEY UNITS WITH DETECTABLE RADIOACTIVITY Eight of the 43 survey units were found to have detectable radioactivity; they are listed in Table 6.6. As can be seen from this table, none of these survey units had any individual results that exceeded the license termination criteria; thus these survey units met the acceptance criteria and d id not require further remediation to meet the criteria for license termination.

Survey units 1.1, 1.2, 1.3, 1.4, 1.5, 1.6 and 1.8.4 are located within a small, isolated area within the WPI facility; all are located within the reactor pool and represent the various materials and structures that were in close proximity to the former reactor core, and thus exposed to the resulting neutron flux. Direct beta scan results throughout this area of the reactor pool interior were negative, except in the region where the four aluminum samples were taken from the thermal column liner (SU 1.4), which had a mean sample SOF of 0.24 and a maximum sample SOF of 0.27. In that region, direct beta scanning appeared to reveal detectable radioactivity at a level equivalent to 1,600 DPM/100 cm 2, even though this level is less than the calculated MDL of 2,400 DPM/100 cm2 for surface radioactivity.

This indicates that direct beta scanning should be capable of detecting volumetric radioactivity at a level equivalent to a SOF of approximately 0.36 (i.e. 2,400/1,600 x 0.24 = 0.36). This should be valid for all of the materials (aluminum, concrete and soil), as they all have somewhat similar radionuclide makeups (see Table 6.4). As the entire interior of the reactor pool liner and the exposed underlying concrete / soil surfaces , the thermal column liner, and the interior surfaces of the beam port tube, the volumetric sample results represent the worst case neutron activation within the biological shield.

Additionally, the concrete cores taken from the floor and walls of the reactor pool were subdivided into upper and lower increments (i.e., ~0-15 cm and

~15-30 cm increments), with all upper increments being subjected to laboratory analysis. Three lower core increments that corresponded to the three upper core increments having the highest observed SOF results were then selected for laboratory analysis.

Two were from the reactor pool floor and the other east wall, in close proximity to the embedded beam port tube. When matched against the upper core results, a decreasing trend with depth is observed. This indicates that the neutron induced concentration of radioactivity decreases with depth, and that the volumetric sample results represent the worst case neutron activation within the biological shield.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 24 of 26 TLG Services, Inc. The matched SOF results for the three concrete cores are as follows:

Concrete Core ID/Location SOF Result Top Increment (0-15 cm) Lower Increment (15-30 cm) SU 1.3 no. 1 (X= 6.11 m, Y= -0.34 m) / Reactor Pool Floor, Between Rx Core and Thermal Column Areas 0.1 73 0 SU 1.3 no. 3 (X= 5.60 m, Y= -1.70 m) / SE of Reactor Core Area 0.180 0 SU 1.4 n o. 3 (X= 1.37 m, Y= 0.67 m) / East Wall, Next to South Side of Beam Port Tube 0.221 0.081 Survey unit 4.2 consists of remnants of various segments of plastic or cast iron pipe (1.25-to-4 inch diameter) embedded in concrete, which drained water from the reactor pool / reactor water treatment system, to the sanitary sewer system. The only indication of detectable radioactivity found within these remnants was from sediment found in a drain trap located in the bottom of a sump pit. This drain trap was where the outlet from reactor pool / reactor water treatment system tied into the sewer system. This trap was found to be plugged with sediment, and was completely removed during the sampling process. That sample was found to have a SOF result of 0.4, with a majority of the radioactivity comprised of Cs-137. After sampling, the trap was subjected to NaI gamma logging and direct beta scans, with no indications of there being any residual radioactivity. The sewer line was checked approximately 4 meters downstream at a cleanout, by NaI Gamma logging and visual observation, with there being no visible indications of sediment and no detectable results from the gamma logging.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 25 of 26 TLG Services, Inc. TABLE 6.6 SURVEY UNITS WITH DETECTABLE RADIOACTIVITY SU ID No. SU Description Detectable FSS Attribute Datum with Detectable Results Number of Datum Fraction of LT Criteria (Mean) Fraction of LT Criteria (M ax. Result) 1.1 Reactor Pool Aluminum Liner Volumetric Sample-Aluminum 4 4 0.03 (SOF) 0.11 (SOF)

1.2 Thermal

Column Aluminum Liner Volumetric Sample-Aluminum 4 4 0.24 (SOF) 0.27 (SOF) 1.3 Interior Reactor Pool Concrete Floor (top layer 0-15 cm depth)

Volumetric Sample-Concrete 2 2 0.18 (SOF) 0.18 (SOF) 1.4 Interior Reactor Pool Concrete Walls Volumetric Sample-Concrete 3 4 0.06 (SOF) 0.22 (SOF) 1.5 Beam Port Tube Volumetric Sample-Aluminum (end of BP tube protruding into reactor pool) 1 (Included as one of the four samples in SU 1.4) 1 0.11 (SOF) 0.11 (SOF) 1.6 Soil Under Removed Reactor Pool Floor Area Volumetric Sample- Soil 4 4 0.16 (SOF) 0.19 (SOF)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 6, Page 26 of 26 TLG Services, Inc. TABLE 6.6 SURVEY UNITS WITH DETECTABLE RADIOACTIVITY SU ID No. SU Description Detectable FSS Attribute Datum with Detectable Results Number of Datum Fraction of LT Criteria (Mean) Fraction of LT Criteria (M ax. Result) 1.8.4 Drain Line from Beam Port Shutter Housing Interior Pipe Gamma Logging Result (Note: this location is in close proximity to the volumetric sample- aluminum shown in SU 1.5 and a concrete volumetric sample in SU 1.4 that had a SOF of 0.22) 1 7 na 86.3 (%) 4.2 Reactor Water Treatment System Drain Piping to Sanitary Sewer System Volumetric Sample-Sediment (Note: sampling process removed all observed sediment. Gamma logging results were all less than 66% of surface contamination criteria) 1 1 0.4 (SOF) 0.4 (SOF)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 7, Page 1 of 2 TLG Services, Inc.

7.0CONCLUSION

S The results of the FSS indicate that the WPI reactor facility meets the criteria set forth in the decommissioning plan for termination of its NRC license. All individual determinations for radioactive surface contamination and volumetric radionuclide concentrations are below the criteria set forth in the decommissioning plan, and provided in this report in Tables 4.3 and 4.4 respectively.

The majority of the facility was found to be free of any detectable licensed radioactive material. Radioactive surface contamination (fixed and removable) was not found anywhere within the facility. The licensed radioactivity that was found was volumetric in nature, almost entirely due to neutron activation of aluminum and concrete construction materials. This neutron activation was limited to a small region within the interior of the reactor pool / biological shield / thermal column (and a small area of underlying soil), all in very close proximity to the former reactor core. In all cases, the radionuclide concentrations in all samples were less than the concentration limits specified in the decommissioning plan (see table 4.4), with the S um-of-Fractions for all the detected radionuclides being less than one. Depth profiling of the more activated concrete areas indicated that the degree of activation decreased with depth; indicating that the residual neutron induced radioactivity is limited to the near surface layers and limited in overall volume.

The only sample result with detectable radioactivity outside of the interior of the reactor pool / biological shield was a single sediment sample that was collected from a sewer line drain trap that serviced the former reactor water treatment system. In that case, the resulting SOF for the sample was 0.4. However, the sampling process removed all of the sediment material that was present.

The following lists the small amount of residual radioactivity remaining within the facility, which in all cases is well below the criteria for license termination specified in the approved DP: The concrete samples from the area of maximum neutron activation, within the first 15 cm of the reactor pool0.18 SOF of the allowable radionuclide concentration limits. No detectable radionuclides of interest were found in the samples from the underlying 15 to 30 cm layers. Aluminum liner samples from the area of maximum neutron activation within the remaining thermal column liner had a maximum result of 0.27 SOF of the allowable concentration limits. The mean of all the samples was 0.24 SOF of the allowable concentration limits.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Section 7, Page 2 of 2 TLG Services, Inc. Concrete samples from the areas of maximum neutron activation within the biological shield walls, were determined to have a maximum SOF result of 0.22 for all radionuclides (concrete immediately surrounding the beam port tube), and a mean SOF result of 0.06 for all the samples representing the four walls of the biological shield. A deeper sample, behind that of the sample found with the SOF of 0.22, was found to have significantly lower radionuclide concentration, with a SOF of only 0.08. Aluminum liner samples, from the areas of maximum neutron activation, had a mean SOF of 0.03, with a maximum sample result coming from the beam port tube end, with a SOF result of 0.11. Soil under the former reactor core had a mean SOF of 0.16, with the maximum individual sample having a SOF of 0.19. The internal pipe gamma logging measurements only revealed one spot with detectable radioactivity, and a few borderline results, within a limited number of the embedded pipes / conduits. The one detectable result was found to be at 86%

of the surface contamination criterion , limited to a few centimeters in length at the end of the beam port drain line. Th is result is unlikely to be due to surface contamination within the lines. The detectable and borderline results were found where vent / drain lines either joined the removed beam port shutter housing, or the remaining aluminum thermal column liner. Those locations are known to have been influenced by neutron activation. These locations are in very close proximity to the neutron activation samples discussed above, that had SOF results well below one.

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8.0 BIBLIOGRAPHY

1.Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), NUREG-1575 (Rev. 1), U.S. Nuclear Regulatory Commission, 2000.

2.Decommissioning Plan for the Leslie C. Wilbur Nuclear Reactor Facility (Rev. 2),

Worcester Polytechnic Institute , September 25, 2012.

3.Consolidated NMSS Decommissioning Guidance. NUREG-1757, U.S. Nuclear Regulatory Commission, 2000.

4.Manual for Conducting Radiological Surveys in Support of License Termination, NUREG/CR-5849 (draft), U.S. Nuclear Regulatory Commission, 1992.

5.Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions, NUREG/CR-1507, U.S. Nuclear Regulatory Commission, 1997.

6.Final Status Survey Plan for the Leslie C. Wilbur Nuclear Reactor Facility at the Worcester Polytechnic Institute Rev. 0, January 2013

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix A, Page 1 of 1 TLG Services, Inc. APPENDIX A GEL Laboratory Analysis Report For Waste Stream and Pre-FSS Samples (Extracted from the FSS Plan)

(NOTE: Click on the Adobe link above for a full copy of the GEL Laboratory Analysis Report)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix B, Page 1 of 1 TLG Services, Inc. APPENDIX B GROSS BETA DCGL FOR RADIONUCLIDE MIXTURE AT WPI (Extracted from the FSS Plan)

Section 4.0 indicated that the reactor pool water radionuclide mixture that would be representative of surface contamination at the WPI facility. The fractional contributions of radionuclides in that mixture are, 0.9995 for Cs-137 and 0.0005 for Co-60. Each of the radionuclides in this mixture decays to some extent by emitting beta particles. The abundance (A) of beta emissions per decay is 1.0 for Co-60 and 0.85 for Cs-137.

To develop a gross-beta DCGL for the structural surfaces, the fractional contribution (f) of each of the radionuclide contaminants to the total mix is divided by the Default Screening DCGL for that radionuclide, 7,100 DPM/100 cm 2 for Co-60 and 28,000 DPM/100 cm 2 for Cs-137. The gross DCGL was then calculated by:

Gross Beta DCGL = fraction of beta emitters (i.e., f x A) (f/DCGL) Gross Beta DCGL = __________0.850075___________ (0.0005/7,100) + (0.9995/28,000

) The resulting gross beta DCGL value is 23,765 DPM/100 cm2 (23,700 rounded).

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissionin g Final Status Survey Report Appendix C, Page 1 of 4 TLG Services, Inc. APPENDIX C WASTE SHIPMENT MANIFEST MAY 2013

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Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix D, Page 1 of 3 TLG Services, Inc.

APPENDIX D WASTE SHIPMENT MANIFEST DECEMBER 2013

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix D, Page 2 of 3 TLG Services, Inc.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix D, Page 3 of 3 TLG Services, Inc.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix E, Page 1 of 1 TLG Services, Inc. APPENDIX E DECOMMISSIONING WORK PROCEDURE (DWP) - 10, PERFORMANCE OF FINAL STATUS SURVEY

(NOTE: Click on the Adobe link above for a full copy of DWP-10)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F, Page 1 of 2 TLG Services, Inc. APPENDIX F FSS SURVEY UNIT DATA TABLE OF CONTENTS Number of Pages F-1.1 Reactor Pool Aluminum Liner ......................................................................... 6 F-1.2 Thermal Column Aluminum Liner ................................................................. 4 F-1.3 Interior Reactor Pool Concrete Floor............................................................... 2 F-1.4 Interior Reactor Pool Concrete Walls .............................................................. 3 F-1.5 Beam Port Tube ............................................................................................... 2 F-1.6 Soil under Removed Concrete Floor Area ....................................................... 2 F-1.7H Top of Biological Shield - Horizontal Surfaces

................................................ 4 F-1.7V Top of Biological Shield - Vertical Surfaces

.................................................... 2 F-1.8.1 Drain Lines from Reactor Pool Scupper Drains ............................................. 3 F-1.8.2 Return Line from Pool Water Treatment System .......................................... 3 F-1.8.3 Vent Line above Beam Port ............................................................................. 2 F-1.8.4 Drain Line from Beam Port Shutter Housing

................................................ 2 F-1.8.5 Feed Line from Pool to Pool Water Treatment System .................................. 3 F-1.8.6 Vent Line from Thermal Column (Upper Left)

............................................... 2 F-1.8.7 Vent Line from Thermal Column (Lower Left)

............................................... 1 F-1.8.8 Vent Line from Thermal Column (Upper Right) ............................................ 1 F-1.8.9 Vent Line from Thermal Column Lower Right) ............................................. 1 F-1.8.10 Vent Line from Thermal Column (Center Top)

.............................................. 1 F-l.9.1 North Exterior Bio logical Shield Wall ............................................................ 2 F-1.9.2 West Exterior Biological Shield Wall .............................................................. 2 F-1.9.3 Thermal Column Shield Door .......................................................................... 3 F-1.9.4 East Exterior Biological Shield Wall ............................................................... 3 F-2.1 Ground Floor Reactor Room - Floor and South Wall - West of Reactor Centerline ..............................................................................

3 F-2.2 Ground Floor Reactor Room - Lower North and West Walls - West of Reactor Centerline ..............................................................................

4 F-2.3 Ground Floor Reactor Room - Floor and South Wall - East of Reactor Centerline ............................................................................... 3 F-2.4 Ground Floor Reactor Room - Lower North and East Walls -

East of Reactor Centerline ...............................................................................

3 F-2.5 Ground Floor Reactor Room - Upper Walls, Ceiling and Exterior of Suspended Equipment - West of Reactor Centerline

................... 3 Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F, Page 2 of 2 TLG Services, Inc. TABLE OF CONTENTS (Continued)

Number of Pages F-2.6 Ground Floor Reactor Room - Upper Walls, Ceiling and Exterior of Suspended Equipment - East of Reactor Centerline .................................. 4 F-2.7 Ground Floor Reactor Room - Radioactive Material Storage Room -

Lower Walls and Floor ..................................................................................... 2 F-2.8 Ground Floor Reactor Room - Radioactive Material Storage Room - Upper Walls, Ceiling and Exterior of Suspended Equipment

........................ 2 F-2.9 Ground Floor Reactor Room - Interior and Exterior Walls, Ceiling and Roof of Dark Room ....................................................................... 4 F-2.10 Ground Floor Reactor Room - Reactor Room Stairs ....................................... 3 F-3.1 Reactor Office - Lower Walls and Floor .......................................................... 3 F-3.2 Reactor Office - Upper Walls and Ceiling ....................................................... 3 F-3.3 First Floor Reactor Room - Lower Walls and Floor -

West of Reactor Centerline ..............................................................................

3 F-3.4 First Floor Reactor Room - Upper Walls and Ceiling - West of Reactor Centerline .............................................................................. 3 F-3.5 First Floor Reactor Room - Lower Walls and Floor - East of Reactor Centerline ............................................................................... 5 F-3.6 First Floor Reactor Room - Upper Walls and Ceiling - East of Reactor Centerline ............................................................................... 3 F-3.7 First Floor Reactor Room - Tool Closet - Lower Walls and Floor

.................. 5 F-3.8 First Floor Reactor Room - Tool Closet - Upper Walls and Ceiling

............... 3 F-4.1 Exhaust Ventilation System - Interior Duct and Equipment Surfaces (and Discharge Point Surrounding Roof Surfaces) .......................................

11 F-4.2 Reactor Water Treatment System Drain Piping (Leading to Sanitary Sewer System) ............................................................... 4 F-4.3 Heating, Ventilation and Air Conditioning Units (Three Units)

................... 3

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.1, Pag e 1 of 6 TLG Services, Inc. APPENDIX F-1.1 SURVEY UNIT DATA REACTOR POOL ALUMINUM LINER STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 0.07 0.61 -186 -0.8% 0.8% 2 2.07 0.61 -103 -0.4% 29 1.2% 3 3.07 0.61 -179 -0.8% 2.3% 4 4.07 0.61 0.4% 24 1.0% 5 5.07 0.61 0.2% 54 2.3% 6 7.07 0.61 -209 -0.9% 15 0.6% 7 8.07 0.61 91 0.4% 0.2% 8 9.07 0.61 15 0.1% 1.0% 9 0.57 1.51 -167 -0.7% 0.8% 10 1.57 1.51 -122 -0.5% 49 2.1% 11 2.57 1.51 -213 -0.9% 0.4% 12 3.57 1.51 -133 -0.6% 2.3% 13 4.57 1.51 -186 -0.8% 0.4% 14 5.57 1.51 -300 -1.3% 0.6% 15 6.57 1.51 -110 -0.5% 19 0.8% 16 7.57 1.51 -205 -0.9% 29 1.2% 17 8.57 1.51 -122 -0.5% 44 1.9% 18 9.57 1.51 -289 -1.2% 44 1.9% 19 0.07 2.41 -342 -1.4% 10 0.4% 20 1.07 2.41 -338 -1.4% 0.6% 21 2.07 2.41 -205 -0.9% 1.4% 22 3.07 2.41 -300 -1.3% 2.1% 23 4.07 2.41 -380 -1.6% 5 0.2% 24 5.07 2.41 -357 -1.5% 1.6% 25 6.07 2.41 -213 -0.9% 19 0.8% 26 7.07 2.41 -179 -0.8% 0 0.0% 27 8.07 2.41 -270 -1.1% 34 1.4% 28 9.07 2.41 0.4% 44 1.9% 29 0.57 3.31 -133 -0.6% 0.2% 30 1.57 3.31 -110 -0.5% 2.3% 31 2.57 3.31 -464 -2.0% 1.2% 32 3.57 3.31 -221 -0.9% 0.6% 33 4.57 3.31 -209 -0.9% 15 0.6% 34 5.57 3.31 -205 -0.9% 29 1.2% 35 6.57 3.31 -236 -1.0% 5 0.2% 36 7.57 3.31 -354 -1.5% 29 1.2% 37 8.57 3.31 -144 -0.6% 0.8% 38 9.57 3.31 -357 -1.5% 29 1.2% 39 0.07 4.21 0.4% 39 1.6% 40 1.07 4.21 -350 -1.5% 0.2%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.1, Pag e 2 of 6 TLG Services, Inc. STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 41 2.07 4.21 -236 -1.0% 0.4% 42 3.07 4.21 -327 -1.4% 1.0% 43 4.07 4.21 -293 -1.2% 1.2% 44 5.07 4.21 -346 -1.5% 0.2% 45 6.07 4.21 -342 -1.4% 0 0.0% 46 7.07 4.21 -1 67 -0.7% 1.0% 47 8.07 4.21 -167 -0.7% 5 0.2% 48 9.07 4.21 -221 -0.9% 1.2% 49 5.57 -0.29 11 0.0% 10 0.4% 50 6.57 -0.29 99 0.4% 29 1.2% 51 5.07 -1.19 -144 -0.6% 0.8% 52 7.07 -1.19 -137 -0.6% 19 0.8% 53 5.57 -2.09 0.3% 10 0.4% 54 6.57 -2.09 -167 -0.7% 10 0.4% SU Mean: -197 0% 1 0.04% SU Max.: 99 0.4% 54 2.3% SU MDL: 391 1.7% 112 4.7% Criteria: 23,700 - 2,370 - SU 1.1 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.1, Pag e 3 of 6 TLG Services, Inc. View of SU 1.1 Aluminum pool liner

VOLUMETRIC CONTAMINATION DATA Sample No.

Grid Coord., m Sample Media / Descript. Sample Data X Y Radionuclides:

Fe-55 Co-60 Ni-63 Cs-134 Cs-137 Eu-152 Eu-154 SOF Criteria, pCi/g:

1E4 3.8 2.1E4 5.7 11 7 8 -------- 1 8.37 0.67 Al liner-North Wall Sample, pCi/g 0.318 0.025 0.000 0.000 0.000 0.000 0.000 0.006 Fraction of Criteria 0.000 0.006 0.000 0.000 0.000 0.000 0.000 2 3.56 0.67 Al liner-South Wall Sample, pCi/g 0.447 0.035 0.000 0.000 0.000 0.000 0.000 0.009 Fraction of Criteria 0.0 00 0.009 0.000 0.000 0.000 0.000 0.000 3 1.2 0.77 Al-End of Beam Port Tube Sample, pCi/g 5.201 0.401 0.000 0.000 0.000 0.000 0.000 0.106 Fraction of Criteria 0.001 0.106 0.000 0.000 0.000 0.000 0.000 4 6.1 1.57 Al liner-West Wall Sample, pCi/g 0.6 50 0.050 0.000 0.000 0.000 0.000 0.000 0.013 Fraction of Criteria 0.000 0.013 0.000 0.000 0.000 0.000 0.000 SU Mean SOF:

0.034 Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.1, Pag e 4 of 6 TLG Services, Inc. INTERIOR REACTOR POOL GRID MAP View of Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.1, Pag e 5 of 6 TLG Services, Inc. Aluminum liner sample no. 1 location, on the north wall of the reactor pool View of aluminum liner sample no. 2 location, on the south wall of the reactor pool

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.1, Pag e 6 of 6 TLG Services, Inc. View of the cut-notch at the top of the tube is the location of the aluminum sample no. 3 View of aluminum liner sample no. 4 location, on the west wall of the reactor pool, above the thermal column Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.2, Pag e 1 of 4 TLG Services, Inc. APPENDIX F-1.2 SURVEY UNIT DATA THERMAL COLUMN ALUMINUM LINER STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 See Location Map See Location Map 1,624 6.9% 19 0.8% 2 992 4.2% 10 0.4% 3 947 4.0% 0.6% 4 392 1.7% 73 3.1% 5 259 1.1% 1.9% 6 1,422 6.0% 34 1.4% 7 1,076 4.5% 29 1.2% 8 411 1.7% 0.6% 9 0.2% 15 0.6% 10 202 0.9% 49 2.1% 11 1,779 7.5% 49 2.1% 12 125 0.5% 63 2.7% 13 449 1.9% 63 2.7% 14 0.3% 19 0.8% 15 -217 -0.9% 19 0.8% 16 1,323 5.6% 5 0.2% 17 1,247 5.3% 0.4% 18 601 2.5% 49 2.1% 19 -521 -2.2% 54 2.3% 20 -255 -1.1% 29 1.2% 21 734 3.1% - - SU Mean: 594 2.5% 25 1.1% SU Max.: 1,779 7.5% 73 3.1% SU MDL: 391 1.7% 133 5.6% Criteria: 23,700 - 2,370 -

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.2, Pag e 2 of 4 TLG Services, Inc. SU 1.2 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.2, Pag e 3 of 4 TLG Services, Inc. VOLUMETRIC CONTAMINATION DATA Sample No.

Grid Coord., m Sample Media / Descript. Sample Data X Y Radionuclides:

Fe-55 Co-60 Ni-63 Cs-134 Cs-137 Eu-152 Eu-154 SOF Criteria, pCi/g:

1E4 3.8 2.1E4 5.7 11 7 8 -------- 1 (see map) Floor Sample, pCi/g 12.34 0.951 0.000 0.000 0.000 0.000 0.000 0.251 Fraction of Criteria 0.001 0.250 0.000 0.000 0.000 0.000 0.000 2 (see map) North Wall Sample, pCi/g 10.01 0.772 0.000 0.00 0 0.000 0.481 0.000 0.273 Fraction of Criteria 0.001 0.203 0.000 0.000 0.000 0.069 0.000 3 (see map) South Wall Sample, pCi/g 11.26 0.868 0.000 0.000 0.000 0.000 0.000 0.230 Fraction of Criteria 0.001 0.228 0.000 0.000 0.000 0.000 0.000 4 (see map) Ceiling Sample, pCi/g 9.909 0.764 0.000 0.000 0.000 0.000 0.000 0.202 Fraction of Criteria 0.001 0.201 0.000 0.000 0.000 0.000 0.000 SU Mean SOF:

0.239 View of thermal column aluminum liner floor (north is left side of photo)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.2, Pag e 4 of 4 TLG Services, Inc. View of sample no. 1 location

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.3, Pag e 1 of 2 TLG Services, Inc. APPENDIX F-1.3 SURVEY UNIT DATA INTERIOR REACTOR POOL CONCRETE FLOOR VOLUMETRIC CONTAMINATION DATA Sample No. Grid Coord., m Sample Media / Descript. Sample Data X Y Radionuclides:

Fe-55 Co-60 Ni-63 Cs-134 Cs-137 Eu-152 Eu-154 SOF Criteria, pCi/g:

1.0E4 3.8 2.1E4 5.7 11 7 8 -------- 1 6.1 -0.34 Concrete core, floor between Rx core and thermal column areas, 0-15 cm depth Sample, pCi/g 0.000 0.121 0.000 0.024 0.000 0.914 0.048 0.173 Fraction of Criteria 0.000 0.032 0.000 0.0 04 0.000 0.131 0.006 1 6.11 -0.34 Concrete core, floor between Rx core and thermal column areas, 15-30 cm depth Sample, pCi/g 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 Fraction of Criteria 0.000 0.000 0.000 0.000 0.000 0.000 0.000 3 5.6 1.7 Concrete core, floor SE of Rx core area, 0-15 cm depth Sample, pCi/g 0.000 0.169 0.000 0.023 0.000 0.880 0.046 0.180 Fraction of Criteria 0.000 0.044 0.000 0.004 0.000 0.126 0.006 3 5.6 1.7 Concrete core, floor SE of Rx core area, 15-30 cm depth Sample, pCi/g 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 Fraction of Criteria 0.000 0.000 0.000 0.000 0.000 0.000 0.000 SU Mean SOF:

0.176 (0-15 cm) 0.000 (15-30 cm)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.3, Pag e 2 of 2 TLG Services, Inc. View of exposed concrete pool floor prior to sampling View of reactor pool floor concrete core sample locations; sample no. 1 is the upper right core hole and sample no. 3 is the lower right core hole. Samples nos. 2 and 4 were provided to the NRC, and are located-upper left and lower-left respectively

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.4, Pag e 1 of 3 TLG Services, Inc. APPENDIX F-1.4 SURVEY UNIT DATA INTERIOR REACTOR POOL CONCRETE WALLS VOLUMETRIC CONTAMINATION DATA Sample No. Grid Coord., m Sample Media / Descript. Sample Data X Y Radionuclides:

Fe-55 Co-60 Ni-63 Cs-134 Cs-137 Eu-152 Eu-154 SOF Criteria, pCi/g:

1E4 3.8 2.1E4 5.7 11 7 8 -------- 1 8.37 0.67 North wall, 0-15 cm Sample, pCi/g 0.000 0.002 0.000 0.000 0.000 0.007 0.000 0.001 Fraction of Criteria 0.000 0.000 0.000 0.000 0.000 0.001 0.000 2 3.56 0.67 South wall, 0-15 cm Sample, pCi/g 0.000 0.005 0.000 0.001 0.000 0.02 2 0.001 0.005 Fraction of Criteria 0.000 0.001 0.000 0.000 0.000 0.003 0.000 3 1.37 0.67 East wall, next to south side of beam port tube, 0-15 cm Sample, pCi/g 0.000 0.344 0.000 0.022 0.000 0.850 0.045 0.221 Fraction of Criteria 0.000 0.091 0.0 00 0.004 0.000 0.121 0.006 3 1.37 0.67 East wall, next to south side of beam port tube, 15-30 cm Sample, pCi/g 0.000 0.152 0.000 0.007 0.000 0.266 0.014 0.081 Fraction of Criteria 0.000 0.040 0.000 0.001 0.000 0.038 0.002 4 8.37 1.57 West wall, ab ove center of thermal column, 0-15 cm Sample, pCi/g 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 Fraction of Criteria 0.000 0.000 0.000 0.000 0.000 0.000 0.000 SU Mean SOF:

0.057 (0-15 cm)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.4, Pag e 2 of 3 TLG Services, Inc. View of concrete sample no. 1 location, prior to coring View of cut-end of beam port tube inside reactor pool, the hole to the right of the beam port tube is the location of concrete core sample no. 3

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.4, Pag e 3 of 3 TLG Services, Inc. View of concrete sample no. 4 location, above the thermal column, prior to sample extraction

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.5, Pag e 1 of 2 TLG Services, Inc.

APPENDIX F-1.5 SURVEY UNIT DATA BEAM PORT TUBE STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 See diagram below

-859 -3.6% 24 1.0% 2 -791 -3.3% 15 0.6% 3 -844 -3.6% 10 0.4% 4 -810 -3.4% 15 0.6% 5 -726 -3.1% 15 0.6% 6 -635 -2.7% 1.2% 7 -639 -2.7% 10 0.4% 8 -719 -3.0% 1.9% 9 0.3% 0.6% 10 -209 -0.9% 49 2.1% 11 -240 -1.0% 1.6% 12 -228 -1.0% 0.2% SU Mean: -564 -2.4% 0 0.0% SU Max.: 0.3% 49 2.1% SU MDL: 391 1.7% 109 4.6% Criteria: 23,700 - 2,370 - Longitudinal and cross- sectional diagram of the beam port tube, indicating the location of the surface contamination measurements

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.5, Pag e 2 of 2 TLG Services, Inc. VOLUMETRIC CONTAMINATION DATA See SU 1.1, Sample no. 3

- End of BP Tube (SOF= 0.11)

View of cut end of beam port tube inside reactor pool; cut notch at the top of the tube is the location of the aluminum sample no. 3 from SU 1.1 and the hole to the right of the beam port tube is the location of concrete core sample no. 3 from SU 1.3 View of beam port flange-end outside of biological shield (photo is pre-D&D)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.6, Pag e 1 of 2 TLG Services, Inc. APPENDIX F-1.6 SURVEY UNIT DATA SOIL UNDER REMOVED CONCRETE FLOOR AREA VOLUMETRIC CONTAMINATION DATA Sample No. Grid Coord., m Sample Media / Descrip. Sample Data X Y Radionuclides:

Fe-55 Co-60 Ni-63 Cs-134 Cs-137 Eu-152 Eu-154 SOF Criteria, pCi/g:

1E4 3.8 2.1E4 5.7 11 7 8 -------- 1 5.7 -0.9 Soil, SW Quadr ant, 0 to 14 cm Sample, pCi/g 0.000 0.171 0.000 0.015 0.000 0.553 0.029 0.130 Fraction of Criteria 0.000 0.045 0.000 0.003 0.000 0.079 0.004 2 6.27 -0.9 Soil, NW Quadrant, 0-15 cm Sample, pCi/g 0.000 0.107 0.000 0.016 0.000 0.615 0.032 0.123 Fraction of Criteria 0.000 0.028 0.000 0.003 0.000 0.088 0.004 3 5.7 -1.28 Soil, SE Quadrant, 0-11 cm Sample, pCi/g 0.000 0.204 0.000 0.024 0.000 0.911 0.048 0.194 Fraction of Criteria 0.000 0.054 0.000 0.004 0.000 0.130 0.006 4 6.27 -1.28 Soil, NE Quadrant, 0-14 cm Sample, pCi/g 0.000 0.143 0.000 0.025 0.000 0.951 0.050 0.184 Fraction of Criteria 0.000 0.038 0.000 0.004 0.000 0.136 0.006 SU Mean SOF:

0.158 View of soil sample locations (right side of photo is north)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.6, Pag e 2 of 2 TLG Services, Inc. INTERIOR REACTOR POOL GRID MAP

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.7H, Pag e 1 of 4 TLG Services, Inc. APPENDIX F-1.7H SURVEY UNIT DATA (a) TOP OF BIOLOGICAL SHIELD - HORIZONTAL SURFACES STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 -1.34 0.28 -1,052 -4.4% 0.4% 2 -0.34 0.28 -915 -3.9% 5 0.2% 3 0.66 0.28 -10 0.0% 24 1.0% 4 1.66 0.28 -6 0.0% 5 0.2% 5 2.66 0.28 -2 0.0% 24 1.0% 6 3.66 0.28 13 0.1% 44 1.9% 7 4.66 0.28 0.4% 0.8% 8 5.66 0.28 -884 -3.7% 24 1.0% 9 6.66 0.28 -881 -3.7% 49 2.1% 10 0.16 1.18 0.2% 0.4% 11 1.16 1.18 -303 -1.3% 0.2% 12 2.16 1.18 -242 -1.0% 0 0.0% 13 3.16 1.18 -367 -1.5% 39 1.6% 14 4.16 1.18 -287 -1.2% 29 1.2% 15 5.16 1.18 24 0.1% 0 0.0% 16 0.66 2.08 62 0.3% 78 3.3% 17 4.66 2.08 116 0.5% 1.0% 18 0.16 2.98 2 0.0% 15 0.6% 19 1.16 2.98 -637 -2.7% 39 1.6% 20 2.16 2.98 -854 -3.6% 24 1.0% 21 3.16 2.98 -1,014 -4.3% 0.2% 22 4.16 2.98 -1,036 -4.4% 1.2% 23 5.16 2.98 188 0.8% 0 0.0% 24 0.66 3.88 100 0.4% 0.2% 25 4.66 3.88 427 1.8% 1.2% 26 0.16 4.78 264 1.1% 1.0% 27 1.16 4.78 226 1.0% 0.6% 28 2.16 4.78 481 2.0% 5 0.2% 29 3.16 4.78 321 1.4% 0 0.0% 30 4.16 4.78 393 1.7% 24 1.0% 31 5.16 4.78 359 1.5% 15 0.6% 32 0.66 5.68 -466 -2.0% 5 0.2% 33 1.66 5.68 -253 -1.1% 44 1.9% 34 2.66 5.68 -280 -1.2% 0.2% 35 3.66 5.68 -204 -0.9% 1.9% 36 4.66 5.68 -523 -2.2% 15 0.6% 37 2.16 2.98 -846 -3.6% 49 2.1% 38 3.16 2.98 -926 -3.9% 0 0.0% 39 4.16 2.98 -888 -3.7% 0.2%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.7H, Pag e 2 of 4 TLG Services, Inc. STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria SU Mean: -257 -1.1% 8 0.34% SU Max.: 481 2.0% 78 3.3% SU MDL: 423 1.8% 104 4.4% Criteria: 23,700 - 2,370 - (a) Survey units 1.7V and 1.7H are sub-sets of a single survey unit SU 1.7H STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.7H, Pag e 3 of 4 TLG Services, Inc. TOP OF BIOLOGICAL SHIELD HORIZONTAL SURFACE GRID MAP

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.7H, Pag e 4 of 4 TLG Services, Inc. View of top of biological shield horizontal survey locations; reactor bridge View of top of biological shield horizontal survey locations; operating area floor

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.7V, Pag e 1 of 2 TLG Services, Inc. APPENDIX F-1.7V SURVEY UNIT DATA (a) TOP OF BIOLOGICAL SHIELD - VERTICAL SURFACES STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 0.41 0.29 -570 -2.4% 0.4% 2 1.3 1 0.29 -570 -2.4% 1.2% 3 2.21 0.29 -589 -2.5% 1.4% 4 3.11 0.29 -669 -2.8% 0.4% 5 4.01 0.29 -577 -2.4% 0.4% 6 4.91 0.29 -471 -2.0% 1.0% 7 5.81 0.29 -361 -1.5% 1.0% 8 6.71 0.29 -182 -0.8% 0.2% 9 7.61 0.29 0.3% -2 4 -1.0% 10 8.51 0.29 -148 -0.6% 0.4% 11 9.41 0.29 0.1% 1.0% 12 10.31 0.29 -239 -1.0% 1.9% 13 11.21 0.29 -749 -3.2% 5 0.2% 14 12.11 0.29 -1,034 -4.4% 1.0% 15 13.01 0.29 -376 -1.6% 10 0.4% SU Mean: -442 -1.9% 0.7% SU Max.: 0.1% 10 0.4% SU MDL: 441 1.9% 112 4.7% Criteria: 23,700 - 2,370 - (a)Survey units 1.7V and 1.7H are sub-sets of a single survey unit SU 1.7V STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.7V, Pag e 2 of 2 TLG Services, Inc. TOP OF BIOLOGICAL SHIELD VERTICAL SURFACE GRID MAP View of top of biological shield vertical surface scanning survey, near location no. 15 Worcester Polytechnic Institute Document W19-1 579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.8.1, Pag e 1 of 3 TLG Services, Inc. APPENDIX F-1.8.1 SURVEY UNIT DATA DRAIN LINES FROM REACTOR POOL SCUPPER DRAINS EMBEDDED PIPE / CONDUIT DATA Beta Surface Contamination Checks at Line Ends Location Beta Scan Results, DPM/100 cm 2 Scan MDL, DPM/100 cm 2 Scan MDL, % Criteria Removable, DPM/100 cm 2 % of Criteria MDL, DPM/100 cm 2 MDL, % of Criteria North drain bowl 0 1,673 7.1% 0.4% 104 4.4% South drain bowl 0 1,673 7.1% 29 1.2% 10 4 4.4% Cut end at East Biological shield wall 0 1,932 8.2% 0.8% 104 4.4% Beta Surface Contamination: Internal Line Swabs Line Segment Length Swabbed, cm DPM/Swab MDL, DPM/Swab From North drain bowl to first elbow 170 -19 114 From South drain bowl to first elbow 170 10 114 From cut end at East biological shield wall to elbow 140 44 114 Internal Surface Contamination: NaI Gamma Log Results Line Segment 1: From north scupper drain basin to refusal at elbow transition from vertical to horizontal pipe run Dist. from start, cm DPM/100 cm 2 % of Criteria 3 -30,562 -129% 13 -7,727 -33% 23 804 3.4% 33 -3,763 -15.9% 43 -6,520 -27.5% 53 -7,813 -33% 63 -8,071 -34.1% 73 -9,450 -39.87% 83 -3,073 -13% 93 -2,470 -10.4% 103 -833 -3.5% 113 -4,883 -20.6% 123 2,011 8.5% 133 -4,625 -19.5% 143 0.2% 153 -3,763 -15.9% 158 2,528 10.67%

Worcester Polytechnic Institute Document W19-1 579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.8.1, Pag e 2 of 3 TLG Services, Inc. Line Segment 2:

From south scupper drain basin to refusal at elbow transition from vertical to horizontal pipe run. Dist. from start, cm DPM/100 cm 2 % of Criter ia 3 -39,783 -167.9% 13 -5,055 -21.3% 23 -2,384 -10.1% 33 -2,039 -8.6% 43 -2,212 -9.3% 53 -6,348 -26.8% 63 -661 -2.8% 73 3,045 12.9% 83 -3,590 -15.2% 93 -3,073 -13% 103 718 3% 113 -1,867 -7.9% 123 -5,572 -23.5% 133 1,063 4.5% 143 -2,212 -9.3% 153 -3,763 -15.9% 163 -2,901 -12.2% 165 -4,366 -18.4% Line Segment 3:

From cut pipe end at east biological shield wall to refusal at elbow at transition from horizontal to vertical pipe run

. Dist. from start, cm DPM/100 cm 2 % of Criteria 3 -19,274 -81.3% 13 -12,380 -52.2% 23 -10,398 -43.9% 33 -9,709 -40.9% 43 -1,867 -7.9% 53 -316 -1.3% 63 -4,538 -19.2% 73 -833 -3.5% 83 373 1.6% 93 -2,212 -9.3% 103 2,097 8.9% 113 3,303 13.9% 123 804 3.4% 133 1,580 6.7% 141 2,786 11.8% Internal Surface Contamination: Summary SU Mean Result, DPM/100 cm 2: -4,397 SU Maximum Result, DPM/100 cm 2: 3,303 SU Maximum Result, % of Criteria:

14% Criteria, DPM/100 cm 2: 23,700 MDL, DPM/100 cm 2: 15,540 MDL, % of Criteria:

66%

Worcester Polytechnic Institute Document W19-1 579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.8.1, Pag e 3 of 3 TLG Services, Inc. View of SU 1.8.1 scupper drain line swab sampling View of SU 1.8.1 scupper drain line gamma logging Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.8.2, Pag e 1 of 3 TLG Services, Inc. APPENDIX F-1.8.2 SURVEY UNIT DATA RETURN LINE FROM POOL WATER TREATMENT SYSTEM EMBEDDED PIPE / CONDUIT DATA Beta Surface Contamination Checks at Line Ends Location Beta Scan Results, DPM/100 cm 2 Scan MDL, DPM/100 cm 2 Scan MDL, % Crite ria Removable, DPM/100 cm 2 % of Criteria MDL, DPM/100 cm 2 MDL, % of Criteria Cut end at East Biological shield wall 0 1,932 8.2% 15 0.6% 104 4.4% Discharge Nozzle at South Reactor Pool 0 1,673 7.1% 0.6% 104 4.4% Beta Surface Contamination: Internal Line Swabs Line Segment Length Swabbed, cm DPM/Swab MDL, DPM/Swab From cut end at East biological shield wall to elbow 142 -5 121 From discharge nozzle at South Reactor Pool wall to elbow 18 15 121 Internal Surface Contamination: NaI Gamma Log Results Line Segment 1: From cut end at east biological shield wall to refusal at elbow transition from horizontal to vertical pipe run

. Dist. from start, cm DPM/100 cm 2 % of Criteria 3 -28,063 -118.4% 1 3 -11,260 -47.5% 23 -4,797 -20.2% 33 -8,330 -35.2% 43 -919 -3.9% 53 -3,246 -13.7% 63 977 4.1% 73 287 1.2% 83 -2,470 -10.4% 93 1,494 6.3% 103 2,355 9.9% 113 5,199 21.9% 123 -5,572 -23.5% 133 977 4.1% 138 2,442 10.3%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.8.2, Pag e 2 of 3 TLG Services, Inc. Line Segment 2:

From discharge nozzle at reactor pool wall to refusal at first elbow transition from horizontal to vertical pipe run

. (Connects to opposite end vertical run at end of segment 1).

Dist. from start, cm DPM/100 cm 2 % of Criteria 3 -39,524 -166.8% 13 -26,512 -111.9% 16 -24,530 -103.5% Internal Surface Contamination Summary SU Mean Result, DPM/100 cm 2: -7,861 SU Maximum Result, DPM/100 cm 2: 5,199 SU Maximum Result, % of Criteria:

22% Criteria, DPM/100 cm 2: 23,700 MDL, DPM/100 cm 2: 15,540 MDL, % of Criteria:

66% View of the cut end side at the east biological shield wall (segment no. one)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.8.2, Pag e 3 of 3 TLG Services, Inc. View of discharge nozzle at reactor pool wall (segment no. two side)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.8.3, Pag e 1 of 2 TLG Services, Inc. APPENDIX F-1.8.3 SURVEY UNIT DATA VENT LINE ABOVE BEAM PORT EMBEDDED PIPE / CONDUIT DATA Beta Surface Contamination Checks at Line Ends Location Beta Scan Results, DPM/100 cm 2 Scan MDL, DPM/100 cm 2 Scan MDL, % Criteria Removable, DPM/100 cm 2 % of Criteria MDL, DPM/100 cm 2 MDL, % of Criteria Cut end, at east biological shield wall 0 1,932 8.2% 34 1.4% 104 4.4% Beta Surface Contamination: Internal Line Swabs Line Segment Length Swabbed, cm DPM/Swab MDL, DPM/Swab From cut end at east biological shield wall 41 -19 121 Internal Surface Contamination: NaI Gamma Log Results Line Segment 1:

Start Location: From cut end at east biological shield wall to refusal at pipe curve

. Dist. from start, cm DPM/100 cm 2 % of Criteria 3 -36,767 -155.1% 13 -17,809 -75.1% 21 -1,867 -7.9% Internal Surface Contamination: Summary SU Mean Result, DPM/100 cm 2: -18,814 SU Maximum Result, DPM/100 cm 2: -1,867 SU Maximum Result, % of Criteria:

-8% Criteria, DPM/100 cm 2: 23,700 MDL, DPM/100 cm 2: 15,540 MDL, % of Criteria:

66%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.8.3, Pag e 2 of 2 TLG Services, Inc. View of beam port vent line being swab sampled

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.8.4, Pag e 1 of 2 TLG Services, Inc. APPENDIX F-1.8.4 SURVEY UNIT DATA DRAIN LINE FROM BEAM PORT SHUTTER HOUSING EMBEDDED PIPE / CONDUIT DATA Beta Surface Contamination Checks at Line Ends Location Beta Scan Results, DPM/100 cm 2 Scan MDL, DPM/100 cm 2 Scan MDL, % Criteria Removable, DPM/100 cm 2 % of Criteria MDL, DPM/100 c m 2 MDL, % of Criteria Cut end at east biological shield wall 0 1,932 8.2% 15 6.3% 104 4.4 Beta Surface Contamination: Internal Line Swabs Line Segment Length Swabbed, cm DPM/Swab MDL, DPM/Swab From cut end at east biological shield wall to elbow below shutter housing 144 19 102 Internal Surface Contamination: NaI Gamma Log Results Line Segment 1: Start Location: From cut end at east biological shield wall to refusal at elbow transition from horizontal to vertical.

Dist. from start, cm DPM/100 cm 2 % of Criteria 3 -35,388 -149.3% 13 -12,208 -51.5% 23 -11,604 -49% 33 -833 -3.5% 43 6,664 28.1% 48 1,925 8.1% Line Segment 2:

At cut end where concrete was removed around Beam Port inside reactor pool, inside elbow embedded in concrete below former shutter housing. (see photograph below))

Dist. from start, cm DPM/100 cm 2 % of Criteria 3 20,451 86.3% Internal Surface Contamination Summary SU Mean Result, DPM/100 cm 2: -4,428 SU Maximum Result, DPM/100 cm 2: 20,451* SU Maximum Result, % of Criteria

86.3% Criteria, DPM/100 cm 2: 23,700 MDL, DPM/100 cm 2: 15,540 MDL, % of Criteria:

66% *This section of pipe having the maximum result is located in close proximity to the beam port, an area with a slight amount of known neutron activation. Volumetric aluminum sample SU 1.1 no. 1 was collected from the remaining end of the Beam Port tube in this area, with a SOF result of 0.11 (see view of sample location below- cut out from top of tube). This section of the pipe is also embedded in concrete in close proximity to where volumetric concrete sample SU 1.4 #4 was obtained, which had a SOF result of 0.202 (see view of sample location below-core hole to right of beam port tube).

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.8.4, Pag e 2 of 2 TLG Services, Inc. View of gamma logging detector at 3 cm location within segment two; note close proximity to location of concrete sample SU 1.4 no. 3 and SU1.1 no.1

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.8.5, Pag e 1 of 3 TLG Services, Inc. APPENDIX F-1.8.5 SURVEY UNIT DATA FEED LINE FROM POOL TO POOL WATER TREATMENT SYSTEM EMBEDDED PIPE / CONDUIT DATA Beta Surface Contamination Checks at Line Ends Location Beta Scan Results, DPM/100 cm 2 Scan MDL, DPM/100 cm 2 Scan MDL, % Criteria Removable, DPM/100 cm 2 % of Criteria MDL, DPM/100 cm 2 MDL, % of Criteria Cut end outside biological shield in excavated pit next to water treatment system 0 2,160 9.1% 1.2% 104 4.4% Cut e nd inside reactor pool 0 1,932 8.2% -44 1.9% 104 4.4% Beta Surface Contamination: Internal Line Swabs Line Segment Length Swabbed, cm DPM/Swab MDL, DPM/Swab From cut end inside reactor pool to cut end outside biological shield 330 93 135 Internal Surface Contamination: NaI Gamma Log Results Line Segment 1: From cut end outside biological shield (in excavated pit next to water treatment system) to obstruction (approximately even with the exterior east wall of biological shield)

Dist. from start, cm DPM/100 cm 2 % of Criteria 3 402 1.7% 13 7,511 31.7% 23 10,183 43% 33 9,019 38.1% 43 4,021 17% 53 3,892 16.4% 63 9,062 38.2% 73 5,314 22.4% 83 4,668 19.7% 93 4,840 20.4% 103 4,323 18.2% 113 6,391 27% 123 4,840 20.4%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.8.5, Pag e 2 of 3 TLG Services, Inc. Line Segment 2:

From inside reactor pool at the location of removed floor drain elbow, to the west side of the obstruction noted for at segment one.

Dist. from start, cm DPM/100 cm 2 % of Criteria 3 -7,095 -29.9% 13 -6,319 -26.7% 23 -8,258 -34.8% 33 -4,467 -18.8% 43 -3,906 -1 6.5% 53 -1,795 -7.6% 63 -2,312 -9.8% 73 -804 -3.4% 83 2,169 9.2% 93 -675 -2.8% 103 1,479 6.2% 113 3,978 16.8% 123 4,021 17.0% 133 -718 -3.0% 143 1,565 6.6% 153 2,341 9.9% 163 4,064 17.1% 173 747 3.2% 183 -3,346 -14.1% 193 704 3.0% 203 2,987 12.6% Internal Surface Contamination Summary SU Mean Result, DPM/100 cm 2: 1,730 SU Maximum Result, DPM/100 cm 2: 10,183 SU Maximum Result, % of Criteria:

43.0% Criteria, DPM/100 cm 2: 23,700 MDL, DPM/100 cm 2: 15,540 MDL, % of Criteria:

66%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.8.5, Pag e 3 of 3 TLG Services, Inc. View of swab sample being retrieved (at the segment 1 side of the line)

View of gamma logging detector in line segment two, at the 3 cm position

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.8.6, Pag e 1 of 2 TLG Services, Inc. APPENDIX F-1.8.6 SURVEY UNIT DATA VENT LINE FROM THERMAL COLUMN (UPPER LEFT)

EMBEDDED PIPE / CONDUIT DATA Beta Surface Contamination Checks at Line Ends Location Beta Scan Results, DPM/100 cm 2 Scan MDL, DPM/100 cm 2 Scan MDL, % Criteria Removable, DPM/100 cm 2 % of Criteria MDL, DPM/100 cm 2 MDL, % of Criteria Cut end at West Biological shield wall 0 1,932 8.2% 15 6.3% 104 4.4 Beta Surface Contamination: Internal Line Swabs Line Segment Length Swabbed, cm DPM/Swab MDL, DPM/Swab From cut end at west Biological shield wall to elbow, and from thermal column liner to elbow 99 -29 125 Internal Surface Contamination: NaI Gamma Log Results Line Segment 1:

Start Location: From cut end at west biological shield wall to elbow Dist. from start, cm DPM/100 cm 2 % of Criteria 3 -22,807 -96.2% 13 -12,552 -53.0% 23 -1,436 -6.1% 33 -5,917 -25.0% 43 -2,126 -9.0% 53 -1,092 -4.6% 63 1,235 5.2% 66 -2,729 -11.5% Line Segment 2: From thermal column liner to opposite side of elbow ( the segment one stopping point)

Dist. from start, cm DPM/100 cm 2 % of Criteria 3 8,215 34.7% 13 15,626 65.9% 23 11,834 49.9% Internal Surface Contamination: Summary SU MEAN RESULT, DPM/100 cm 2: -1,068 SU MAXIMUM RESULT, DPM/100 cm 2: 15,626 SU MAXIMUM RESULT, % of CRITERIA:

66% CRITERIA, DPM/100 cm 2: 23,700 MDL, DPM/100 cm 2: 15,540 MDL, % of CRITERIA:

66%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.8.6, Pag e 2 of 2 TLG Services, Inc. View of swab sampling from the segment two side of the line View of gamma logging detector inserted into the segment one side of the line Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.8.7, Pag e 1 of 1 TLG Services, Inc. APPENDIX F-1.8.7 SURVEY UNIT DATA VENT LINE FROM THERMAL COLUMN (LOWER LEFT)

EMBEDDED PIPE / CONDUIT DATA Beta Surface Contamination Checks at Line Ends Location Beta Scan Results, DPM/100 cm 2 Scan MDL, DPM/100 cm 2 Scan MDL, % Criteria Removable, DPM/100 cm 2 % of Criteria MDL, DPM/100 cm 2 MDL, % of Criteria Cut end at West Biological shield wall 0 1,932 8.2% 0.6% 104 4.4% Beta Surface Contamination: Internal Line Swabs Line Segment Length Swabbed, cm DPM/Swab MDL, DPM/Swab From cut end at west biological shield wall to elbow, and from thermal column liner to elbow 99 58 109 Internal Surface Contamination: NaI Gamma Log Resul ts Line Segment 1:

Start Location: From cut end at west biological shield wall to elbow Dist. from start, cm DPM/100 cm 2 % of Criteria 3 -18,843 -79.5% 13 -15,913 -67.1% 23 -144 -0.6% 33 -10,398 -43.9% 43 -3,677 -15.5% 53 -6,951 -29.3% 63 2,011 8.5% 66 7,095 29.9% Line Segment 2:

From north wall thermal column liner, to elbow (segment one stopping point)

Dist. from start, cm DPM/100 cm 2 % of Criteria 3 13,385 56.5% 13 14,678 61.9% 23 -1,350 -5.7% 25 -1,436 -6.1% Internal Surface Contamination Summary SU Mean Result, DPM/100 cm 2: -1,795 SU Maximum Result, DPM/100 cm 2: 14,678 SU Maximum Result, % of Criteria:

62% Criteria, DPM/100 cm 2: 23,700 MDL, DPM/100 cm 2: 15,540 MDL, % of Criteria:

66%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.8.8, Pag e 1 of 1 TLG Services, Inc.

APPENDIX F-1.8.8 SURVEY UNIT DATA VENT LINE FROM THERMAL COLUMN (UPPER RIGHT) EMBEDDED PIPE / CONDUIT DATA Beta Surface Contamination Checks at Line Ends Location Beta Scan Results, DPM/100 cm 2 Scan MDL, DPM/100 cm 2 Scan MDL, % Criteria Removable, DPM/100 cm 2 % of Criteria MDL, DPM/100 cm 2 MDL, % of Criteria Cut end at West Biological shield wall 0 1932 8.2% 44 1.9% 104 4.4% Beta Surface Contamination: Internal Line Swabs Line Segment Length Swabbed, cm DPM/Swab MDL, DPM/Swab From cut end at west biological shield wall to elbow and from thermal column liner to elbow 99 10 109 Internal Surface Contamination: NaI Gamma Log Results Line Segment 1: From cut end at west biological shield wall to elbow Dist. from start, cm DPM/100 cm 2 % of Criteria 3 -13,156 -55.5% 13 -8,588 -36.2% 23 -6,434 -27.1% 33 -3,763 -15.9% 43 -9,881 -41.7% 53 546 2.3% 63 1,063 4.5% Line Segment 2: From south wall thermal column liner, to elbow (segment one stopping point)

Dist. from start, cm DPM/100 cm 2 % of Criteria 3 2,442 10.3% 13 6,319 26.7% 22 4,682 19.8% Internal Surface Contamination Summary SU MEAN RESULT, DPM/100 cm 2: -2,677 SU MAXIMUM RESULT, DPM/100 cm 2: 6,319 SU MAXIMUM RESULT, % of CRITERIA:

27% CRITERIA, DPM/100 cm 2: 23,700 MDL, DPM/100 cm 2: 15,540 MDL, % of CRITERIA:

66%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.8.9, Pag e 1 of 1 TLG Services, Inc. APPENDIX F-1.8.9 SURVEY UNIT DATA VENT LINE FROM THERMAL COLUMN (LOWER RIGHT) EMBEDDED PIPE / CONDUIT DATA Beta Surface Contamination Checks at Line Ends Location Beta Scan Results, DPM/100 cm 2 Scan MDL, DPM/100 cm 2 Scan MDL, % Criteria Removable, DPM/100 cm 2 % of Criteria MDL, DPM/100 cm 2 MDL, % of Criteria Cut end at West Biological shield wall 0 1932 8.2% 1.0% 104 4.4% Beta Surface Contamination: Internal Line Swabs Line Segment Length Swabbed, cm DPM/Swab MDL, DPM/Swab From cut end at west biological shield wall to elbow, and from thermal column liner to elbow 99 49 135 Internal Surface Contamination: NaI Gamma Log Results Line Segment 1:

From cut end at west biological shield wall to elbow Dist. from start, cm DPM/100cm 2 % of Criteria 3 -19,188 -81.0% 13 -3,160 -13.3% 23 -5,142 -21.7% 33 -7,899 -33.3% 43 -6,434 -27.1% 53 -2,298 -9.7% 63 -2,126 -9.0% Line Segment 2: From south wall thermal column liner, to elbow (segment one stopping point)

Dist. from start, cm DPM/100 cm 2 % of Criteria 3 7,009 29.6% 13 9,938 41.9% 21 -1,609 -6.8% Internal Surface Contamination Summary SU Mean Result, DPM/100 cm 2: -3,091 SU Maximum Result, DPM/100 cm 2: 9,938 SU Maximum Result, % of Criteria:

42% Criteria, DPM/100 cm 2: 23,700 MDL, DPM/100 cm 2: 15,540 MDL, % of Criteria:

66%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.8.10, Pag e 1 of 1 TLG Services, Inc. APPENDIX F-1.8.10 SURVEY UNIT DATA VENT LINE FROM THERMAL COLUMN (CENTER TOP)

EMBEDDED PIPE / CONDUIT DATA Beta Surface Contamination Checks at Line Ends Location Beta Scan Results, DPM/100 cm 2 Scan MDL, DPM/100 cm 2 Scan MDL, % Criteria Removable, DPM/100 cm 2 % of Criteria MDL, DPM/100 cm 2 MDL, % of Criteria Cut end at west biological shield wall 0 1,932 8.2% 5 0.21% 104 4.4% Beta Surface Contamination: Internal Line Swabs Line Segment Length Swabbed, cm DPM/Swab MDL, DPM/Swab From cut end at west biological shield wall to elbow, and from thermal column liner to elbow 140 73 135 Internal Surface Contamination: NaI Gamma Log Results Line Segment 1: Entire line Dist. from start, cm DPM/100 cm 2 % of Criteria 3 -10,11 1 -42.7% 13 -7,138 -30.1% 23 -2,398 -10.1% 33 2,643 11.2% 43 2,212 9.3% 53 747 3.2% 63 -2,183 -9.2% 73 3,461 14.6% 83 2,169 9.2% 93 4,280 18.1% 103 2,427 10.2% 113 2,212 9.3% 123 6,176 26.1% 133 9,407 39.7% 140 8,545 36.1% Internal Surface Contamination Summary SU Mean Result, DPM/100 cm 2: 1,497 SU Maximum Result, DPM/100 cm 2: 9,407 SU Maximum Result, % of Criteria:

40% Criteria, / DPM/100 cm 2: 23,700 MDL, DPM/100 cm 2: 15,540 MDL, % of Criteria:

66%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.9.1, Pag e 1 of 2 TLG Services, Inc. APPENDIX F-1.9.1 SURVEY UNIT DATA NORTH EXTERIOR BIOLOGICAL SHIELD WALL STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 5.27 0.75 -528 -2.2% 5 0.2% 2 6.27 0.75 -353 -1.5% 5 0.2% 3 7.27 0.75 -444 -1.9% 0.8% 4 8.27 0.75 -327 -1.4% 0.6% 5 9.27 0.75 -410 -1.7% 0.8% 6 4.77 1.65 -490 -2.1% 10 0.4% 7 5.77 1.65 -661 -2.8% 1.2% 8 6.77 1.65 -585 -2.5% 29 1.2% 9 7.77 1.65 -406 -1.7% 19 0.8% 10 8.77 1.65 -463 -2.0% 0 0.0% 11 9.77 1.65 -452 -1.9% 0.4% 12 5.27 2.55 -726 -3.1% 15 0.6% 13 6.27 2.55 -566 -2.4% 0.4% 14 7.27 2.55 -494 -2.1% 0 0.0% 15 8.27 2.55 -589 -2.5% 19 0.8% 16 9.27 2.55 -612 -2.6% 0.8% 17 4.77 3.45 -844 -3.6% 19 0.8% 18 5.77 3.45 -745 -3.1% 0 0.0% 19 6.77 3.45 -711 -3.0% 10 0.4% 20 7.77 3.45 -806 -3.4% 49 2.1% 21 8.77 3.45 -627 -2.6% 54 2.3% 22 9.77 3.45 -657 -2.8% 1.0% SU Mean: -568 -2.4% 4 0.2% SU Max.: -327 -1.4% 54 2.3% SU MDL: 441 1.9% 104 4.4% Criteria: 23,700 - 2,370 -

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.9.1, Pag e 2 of 2 TLG Services, Inc. SU 1.9.1 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS EXTERIOR OF BIOLOGICAL SHIELD WALL GRID MAP Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.9.2, Pag e 1 of 2 TLG Services, Inc. APPENDIX F-1.9.2 SURVEY UNIT DATA WEST EXTERIOR BIOLOGICAL SHIELD WALL STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 10.56 0.51 -539 -2.3% 0.4% 2 11.56 0.51 -433 -1.8% 0.6% 3 13.56 0.51 -410 -1.7% 1.0% 4 14.56 0.51 -330 -1.4% 34 1.4% 5 11.06 1.41 -304 -1.3% 2.1% 6 12.06 1.41 -494 -2.1% 5 0.2% 7 13.06 1.41 -414 -1.7% 15 0.6% 8 14.06 1.41 -422 -1.8% 0.4% 9 10.56 2.31 -498 -2.1% 1.4% 10 11.56 2.31 -270 -1.1% 0 0.0% 11 12.56 2.31 -403 -1.7% 1.4% 12 13.56 2.31 -296 -1.2% 19 0.8% 13 14.56 2.31 -342 -1.4% 10 0.4% 14 11.06 3.21 -505 -2.1% 0.6% 15 12.06 3.21 -220 -0.9% 1.0% 16 13.06 3.21 -380 -1.6% 0.4% 17 14.06 3.21 -395 -1.7% 10 0.4% SU Mean: -391 -1.7% 0.3% SU Max.: -220 0.9% 34 1.4% SU MDL: 441 1.9% 114 4.8% Criteria: 23,700 - 2,370 - Partial view of west exterior wall of biological shield

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.9.2, Pag e 2 of 2 TLG Services, Inc. SU 1.9.2 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS EXTERIOR OF BIOLOGICAL SHIELD WALL GRID MAP

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.9.3, Pag e 1 of 3 TLG Services, Inc. APPENDIX F-1.9.3 SURVEY UNIT DATA THERMAL COLUMN SHIELD DOOR STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 0.18 -0.5 3 -627 -2.6% 0 0.0% 2 0.78 -0.53 -433 -1.8% 0 0.0% 3 1.38 -0.53 -650 -2.7% 15 0.6% 4 0.48 -0.03 -521 -2.2% 15 0.6% 5 1.08 -0.03 -517 -2.2% 5 0.2% 6 0.18 0.47 -601 -2.5% 0 0.0% 7 0.78 0.47 -707 -3.0% 5 0.2% 8 1.38 0.47 -608 -2.6% 0.6% 9 0.48 0.97 -707 -3.0% 10 0.4% 10 1.08 0.97 -601 -2.5% 0.8% 11 -1.02 1.47 -722 -3.0% 0.2% 12 -0.42 1.47 -730 -3.1% 10 0.4% 13 0.18 1.47 -232 -1.0% 0.6% 14 0.78 1.47 -262 -1.1% 1.9% 15 1.38 1.47 0.4% 1.2% 16 1.98 1.47 -544 -2.3% 0 0.0% 17 2.58 1.47 -479 -2.0% 0 0.0% 18 0.48 1.97 -586 -2.5% 0.8% 19 1.08 1.97 -441 -1.9% 1.4% 20 0.18 2.47 -540 -2.3% 10 0.4% 21 0.78 2.47 -555 -2.3% 0.6% 22 1.38 2.47 -464 -2.0% 0.4% 23 0.48 2.97 -673 -2.8% 0.6% 24 1.08 2.97 -662 -2.8% 1.4% SU Mean: -539 -2.3% -8 0.3% SU Max.: -84 0.4% 15 0.6% SU MDL: 372 1.6% 109 4.6% Criteria: 23,700 - 2,370 -

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.9.3, Pag e 2 of 3 TLG Services, Inc. SU 1.9.3 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS THERMAL COLUMN SHIELD DOOR GRID MAP

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.9.3, Pag e 3 of 3 TLG Services, Inc.

View of SU 1.9.3, thermal column shield door, with west side face (outside of biological shield) to the left and south side face to the right View of underside of thermal column shield door being surveyed Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.9.4, Pag e 1 of 3 TLG Services, Inc. APPENDIX F-1.9.4 SURVEY UNIT DATA EAST EXTERIOR BIOLOGICAL SHIELD WALL STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 0.74 0.31 -395 -1.7% 34 1.4% 2 1.74 0.31 -399 -1.7% 29 1.2% 3 2.74 0.31 -342 -1.4% 1.6% 4 3.74 0.31 -171 -0.7% 15 0.6% 5 0.24 1.21 -193 -0.8% 10 0.4% 6 1.24 1.21 -463 -2.0% 2.5% 7 2.24 1.21 -376 -1.6% 15 0.6% 8 3.24 1.21 -509 -2.1% 10 0.4% 9 4.24 1.21 -467 -2.0% 2.1% 10 0.74 2.11 -353 -1.5% 5 0.2% 11 1.74 2.11 -418 -1.8% 1.0% 12 2.74 2.11 -733 -3.1% 0 0.0% 13 3.74 2.11 -597 -2.5% 1.2% 14 0.24 3.01 -281 -1.2% 5 0.2% 15 1.24 3.01 -357 -1.5% 0.6% 16 2.24 3.01 -612 -2.6% 0.2% 17 3.24 3.01 -532 -2.2% 2.5% 18 4.24 3.01 -490 -2.1% 15 0.6% 19 0.74 3.91 -866 -3.7% 2.7% 20 1.74 3.91 -1,038 -4.4% 0 0.0% 21 2.74 3.91 -996 -4.2% 0 0.0% 22 3.74 3.91 -1,121 -4.7% 10 0.4% SU Mean: -532 -2.2% 0.4% SU Max.: -171 -0.7% 34 1.4% SU MDL: 441 1.9% 112 4.7% Criteria: 23,700 - 2,370 -

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.9.4, Pag e 2 of 3 TLG Services, Inc. SU 1.9.4 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS EXTERIOR OF BIOLOGICAL SHIELD WALL GRID MAP

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-1.9.4, Pag e 3 of 3 TLG Services, Inc. Partial view of SU 1.9.4, east exterior wall of biological shield (wall to the right of center)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.1, Pag e 1 of 3 TLG Services, Inc. APPENDIX F-2.1 SURVEY UNIT DATA GROUND FLOOR REACTOR ROOM - FLOOR AND SOUTH WALL - WEST OF REACTOR CENTERLINE STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 c m 2 % of Criteria 1 -1.95 -2.14 -160 -0.7% 2.3% 2 -3.95 -2.14 0.2% 0.6% 3 -5.95 -2.14 0.4% 2.1% 4 -7.95 -2.14 0.2% 0.6% 5 -9.95 -2.14 72 0.3% 5 0.2% 6 -11.95 -2.14 99 0.4% 0.2% 7 -2.95 -3.84 46 0.2% 1.6% 8 -4.95 -3.84 -217 -0.9% 1.6% 9 -6.95 -3.84 -179 -0.8% 0.2% 10 -8.95 -3.84 95 0.4% 44 1.9% 11 -3.95 -5.54 -152 -0.6% 19 0.8% 12 -5.95 -5.54 145 0.6% 0.8% 13 -7.95 -5.54 -7 0.0% 2.1% 14 -9.95 -5.54 88 0.4% 0.8% 15 -2.95 -7.24 -106 -0.4% 29 1.2% 16 -4.95 -7.24 0.1% 1.2% 17 -6.95 -7.24 0 0.0% 0.2% 18 -8.95 -7.24 -4 0.0% 1.2% 19 -3.95 -8.94 133 0.6% 0.2% 20 -5.95 -8.94 247 1.0% 0.4% 21 -7.95 -8.94 164 0.7% 5 0.2% 22 -9.95 -8.94 126 0.5% 2.5% 23 -2.95 -10.64 -7 0.0% 2.1% 24 -4.95 -10.62 0.3% 0.4% 25 -6.95 -10.62 -125 -0.5% 1.2% 26 -8.95 -10.62 0.2% 1.0% SU Mean: -4 0.0% 0.7% SU Max.: 247 1.0% 44 1.9% SU MDL: 400 1.7% 116 4.9% Criteria: 23,700 - 2,370 -

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.1, Pag e 2 of 3 TLG Services, Inc. SU 2.1 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS Partial view of SU 2.1, floor and south wall (SU 1.9.3, thermal column shield door is shown in the center of the photo)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.1, Pag e 3 of 3 TLG Services, Inc. Alternate view of SU 2.1 floor and south wall (to the right of center), with ceiling tiles removed to gain access to upper portions of the south wall; the biological shield area is shown in the left upper two thirds of the photograph

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.2, Pag e 1 of 4 TLG Services, Inc. APPENDIX F-2.2 SURVEY UNIT DATA GROUND FLOOR REACTOR ROOM - LOWER NORTH AND WEST WALLS - WEST OF REACTOR CENTERLINE STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 -0.09 -1.54 368 1.6% 1.4% 2 -0.59 -0.64 410 1.7% 5 0.2% 3 -1.09 -1.54 308 1.3% 19 0.8% 4 -1.59 -0.64 376 1.6% 0.4% 5 -2.09 -1.54 243 1.0% 0 0.0% 6 -2.59 -0.64 247 1.0% 1.0% 7 -3.09 -1.54 91 0.4% 10 0.4% 8 -3.59 -0.64 262 1.1% 0.8% 9 -4.09 -1.54 384 1.6% 0.8% 10 -4.59 -0.64 2 85 1.2% 0 0.0% 11 -5.09 -1.54 118 0.5% 19 0.8% 12 -5.59 -0.64 175 0.7% 19 0.8% 13 -6.09 -1.54 289 1.2% 15 0.6% 14 -6.59 -0.64 209 0.9% 1.0% 15 -7.09 -1.54 319 1.3% 15 0.6% 16 -7.59 -0.64 53 0.2% 10 0.4% 17 -8.09 -1.54 49 0.2% 0.8% 18 -8.5 9 -0.64 194 0.8% 0.2% 19 -9.09 -1.54 -358 -1.5% 44 1.9% 20 -9.59 -0.64 -156 -0.7% 1.0% 21 -10.09 -1.54 232 1.0% 0.8% 22 -10.59 -0.64 372 1.6% 19 0.8% 23 -11.09 -1.54 376 1.6% 1.4% 24 -11.59 -0.64 380 1.6% 1.2% 25 -12.09 -1.54 0.3% 1.4% 26 -12.59 -0.64 0.1% 0.2% 27 -13.09 -1.54 -103 -0.4% 10 0.4% 28 -13.59 -0.64 407 1.7% 1.0% 29 -14.09 -1.54 338 1.4% 15 0.6% 30 -14.59 -0.64 281 1.2% 0.6% 31 -15.09 -1.54 175 0.7% 0.8% 32 -15.59 -0.64 41 0.2% 29 1.2% 33 -16.09 -1.54 0.1% 24 1.0% 34 -16.59 -0.64 -225 -0.9% 1.6% 35 -17.09 -1.54 -324 -1.4% 0.2% 36 -17.59 -0.64 -171 -0.7% 1.4% 37 -18.09 -1.54 0.3% 1.0% 38 -18.59 -0.64 -122 -0.5% 0.6% 39 -19.09 -1.54 -145 -0.6% 0.8%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.2, Pag e 2 of 4 TLG Services, Inc. STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 40 -19.59 -0.64 0.3% 0.4% 41 -20.09 -1.54 11 0.0% 34 1.4% 42 -20.59 -0.64 0.3% 1.6% 43 -21.09 -1.54 114 0.5% 10 0.4% 44 -21.59 -0.64 0.3% 1.4% 45 -22.09 -1.54 53 0.2% 0.4% SU Mean: 115 0.5% 0.3% SU Max.: 410 1.7% 44 -1.9% SU MDL: 373 1.6% 109 4.6% Criteria: 23,700 - 2,370 - PARTIAL SECTION OF SU 2.2 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS (WESTERN SECTION)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.2, Pag e 3 of 4 TLG Services, Inc. PARTIAL SECTION OF SU 2.2 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS (MIDDLE SECTION)

PARTIAL SECTION OF SU 2.2 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS (EASTERN SECTION)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.2, Pag e 4 of 4 TLG Services, Inc. Partial view of SU 2.2, western and middle sections shown (lower west and north walls are at the center and right side of the photograph - south wall is shown on the left side of the photograph)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.3, Pag e 1 of 3 TLG Services, Inc. APPENDIX F-2.3 SURVEY UNIT DATA GROUND FLOOR REACTOR ROOM - FLOOR AND SOUTH WALL - EAST OF REACTOR CENTERLI NE STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 0.81 -2.1 0.1% 0.2% 2 2.81 -2.1 -141 -0.6% 15 0.6% 3 4.81 -2.1 19 0.1% 0 0.0% 4 6.81 -2.1 65 0.3% 0.2% 5 8.81 -2.1 -103 -0.4% 15 0.6% 6 10.81 -2.1 -171 -0.7% 19 0.8% 7 3.81 -3.8 -190 -0.8% 0.6% 8 5.81 -3.8 31 0.1% 0.4% 9 7.81 -3.8 0.1% 58 2.5% 10 9.81 -3.8 -137 -0.6% 0.4% 11 11.81 -3.8 -114 -0.5% 19 0.8% 12 2.81 -5.5 209 0.9% 10 0.4% 13 4.81 -5.5 0.2% 10 0.4% 14 6.81 -5.5 19 0.1% 0.6% 15 8.81 -5.5 72 0.3% 0.6% 16 10.81 -5.5 0.2% 0.4% 17 12.81 -5.5 103 0.4% 10 0.4% 18 3.81 -7.2 103 0.4% 0.8% 19 5.81 -7.2 0.1% 0.2% 20 7.81 7.2 50 0.2% 2.1% 21 9.81 -7.2 80 0.3% 24 1.0% 22 11.81 -7.2 0.1% 1.9% 23 2.81 -8.9 107 0.4% 0.2% 24 4.81 -8.9 0.2% 10 0.4% 25 6.81 -8.9 175 0.7% 1.6% 26 8.81 -8.9 194 0.8% 0.8% 27 10.81 -8.9 0.1% 1.0% 28 3.81 -10.6 0.4% 34 1.4% 29 5.81 -10.6 -152 -0.6% 0.8% 30 7.81 -10.6 -175 -0.7% 1.4% 31 9.81 -10.6 -190 -0.8% 0.4% SU Mean: 0.1% 0.2% SU Max.: 209 0.9% 58 2.5% SU MDL: 400 1.7% 107 4.5% Criteria: 23,700 - 2,370 -

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.3, Pag e 2 of 3 TLG Services, Inc. SU 2.3 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.3, Pag e 3 of 3 TLG Services, Inc. o the Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.4, Pag e 1 of 3 TLG Services, Inc. APPENDIX F-2.4 SURVEY UNIT DATA GROUND FLOOR REACTOR ROOM - LOWER NORTH AND EAST WALLS - EAST OF REACTOR CENTERLINE STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 0.14 -0.67 376 1.6% 15 0.6% 2 0.64 -1.57 349 1.5% 5 0.2% 3 1.14 -0.67 315 1.3% 44 1.9% 4 1.64 -1.57 304 1.3% 34 1.4% 5 2.14 -0.67 311 1.3% 0.8% 6 2.64 -1.57 323 1.4% 10 0.4% 7 3.14 -0.67 -445 -1.9% 15 0.6% 8 3.64 -1.57 216 0.9% 34 1.4% 9 4.14 -0.67 300 1.3% 39 1.6% 10 4.64 -1.57 125 0.5% 0.4% 11 5.14 -0.67 144 0.6% 0.2% 12 5.64 -1.57 311 1.3% 0.8% 13 6.14 -0.67 319 1.3% 34 1.4% 14 6.64 -1.57 243 1.0% 44 1.9% 15 7.14 -0.67 190 0.8% 10 0.4% 16 7.64 -1.57 346 1.5% 19 0.8% 17 8.14 -0.67 220 0.9% 3.5% 18 8.64 -1.57 209 0.9% 0.6% 19 9.14 -0.67 -206 -0.9% 29 1.2% 20 9.64 -1.57 34 6 1.5% 0.4% 21 10.14 -0.67 319 1.3% 0.6% 22 10.64 -1.57 26 0.1% 0.4% 23 11.14 -0.67 61 0.3% 0 0.0% 24 11.64 -1.57 0.3% 0.2% 25 12.14 -0.67 -240 -1.0% 5 0.2% 26 12.64 -1.57 -282 -1.2% 0 0.0% 27 13.14 -0.67 -540 -2.3% 44 1.9% 28 13.64 -1.57 -171 -0.7% 63 2.7% 29 14.14 -0.67 -392 -1.7% 1.0% 30 14.64 -1.57 -187 -0.8% 19 0.8% 31 15.14 -0.67 -552 -2.3% 0 0.0% 32 15.64 -1.57 -445 -1.9% 0.6% 33 16.14 -0.67 -548 -2.3% 44 1.9% 34 16.64 -1.57 -415 -1.8% 0.2% 35 17.1 4 -0.67 -430 -1.8% 5 0.2% 36 17.64 -1.57 -464 -2.0% 1.2% 37 18.14 -0.67 -126 -0.5% 0 0.0%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.4, Pag e 2 of 3 TLG Services, Inc. GROUND FLOOR REACTOR ROOM

- LOWER NORTH AND EAST WALLS

- EAST OF REACTOR CENTERLINE (Continued)

STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria SU Mean: 0.02% 7 0.3% SU Max.: 376 1.6% 63 2.7% SU MDL: 373 1.6% 102 4.3% Criteria: 23,700 - 2,370 - SU 2.4 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS (WESTERN PORTION)

SU 2.4 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS (EASTERN PORTION)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.4, Pag e 3 of 3 TLG Services, Inc. Partial view of SU 2.4 (eastern portion) showing lower north and east wall (below red horizontal dividing line); partially shown in the foreground is the west exterior wall of SU 2.9, dark room

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.5, Pag e 1 of 3 TLG Services, Inc. APPENDIX F-2.5 SURVEY UNIT DATA GROUND FLOOR REACTOR ROOM - UPPER WALLS, CEILING AND EXTERIOR OF SUSPENDED EQUIPMENT - WEST OF REACTOR CENTERLINE STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 -0.3 2.17 -126 -0.5% 0.2% 2 -3.3 2.17 -275 -1.2% 0.4% 3 -5.3 2.17 -377 -1.6% 1.6% 4 -7.3 2.17 -278 -1.2% 1 9 0.8% 5 -9.3 2.17 0.2% 5 0.2% 6 -11.3 2.17 0.1% 15 0.6% 7 -4.3 3.87 0.3% 0.4% 8 -6.3 3.87 -202 -0.9% 0.2% 9 -8.3 3.87 -305 -1.3% 5 0.2% 10 -10.3 3.87 0.3% 0 0.0% 11 -5.3 5.57 -199 -0.8% 29 1.2% 12 -7.3 5.57 -8 0.0% 0.2% 13 -9.3 5.57 37 0.2% 1.4% 14 -4.3 7.27 52 0.2% 5 0.2% 15 -6.3 7.27 41 0.2% 0 0.0% 16 -8.3 7.27 0.2% 34 1.4% 17 -10.3 7.27 -221 -0.9% 19 0.8% 18 -1.69 1.39 265 1.1% 34 1.4% 19 -3.69 1.39 273 1.2% 29 1.2% 20 -5.69 1.39 292 1.2% 19 0.8% 21 -7.69 1.39 227 1.0% 0.8% 22 -9.69 1.39 258 1.1% 15 0.6% 23 -11.69 1.39 657 2.8% 39 1.6% 24 -13.69 1.39 535 2.3% 0.6% 25 -15.69 1.39 345 1.5% 0 0.0% 26 -17.69 1.39 -301 -1.3% 0.4% 27 -19.69 1.39 -370 -1.6% 15 0.6% 28 -21.69 1.39 -107 -0.5% 10 0.4% SU Mean: 0.0% 5 0.2% SU Max.: 657 2.8% 39 1.6% SU MDL: 354 1.5% 107 4.5% Criteria: 23,700 - 2,370 -

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.5, Pag e 2 of 3 TLG Services, Inc. SU 2.5 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS (CEILING AND SUSPENDED EQUIPMENT SURFACES)

SU 2.5 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS (UPPER WALL WESTERN SECTION)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.5, Pag e 3 of 3 TLG Services, Inc. SU 2.5 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS (UPPER WALL EASTERN SECTION)

Partial view of SU 2.5, ceiling, west upper wall and western sections of the north upper wall are shown; upper walls are above the red horizontal dividing line (right side of photo begins at X = 4.5 m and faces west)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.6, Pag e 1 of 4 TLG Services, Inc. APPENDIX F-2.6 SURVEY UNIT DATA GROUND FLOOR REACTOR ROOM - UPPER WALLS, CEILING AND EXTERIOR OF SUSPENDED EQUIPMENT - EAST OF REACTOR CENTERLINE STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 1.91 1.86 -271 -1.1% 5 0.2% 2 3.91 1.86 -446 -1.9% 0.8% 3 5.91 1.86 -149 -0.6% 1.4% 4 7.91 1.86 -218 -0.9% 0.4% 5 9.91 1.86 -324 -1.4% 1.0% 6 11.91 1.86 0.1% 0.8% 7 4.91 3.56 0.3% 0.8% 8 6.91 3.56 -210 -0.9% 0.4% 9 8.91 3.56 -282 -1.2% 0.8% 1 0 10.91 3.56 0.3% 1.9% 11 5.91 5.26 -149 -0.6% 0.8% 12 7.91 5.26 0.2% 19 0.8% 13 9.91 5.26 -252 -1.1% 34 1.4% 14 11.91 5.26 0.2% 0.6% 15 4.91 6.96 94 0.4% 1.9% 16 6.91 6.96 -145 -0.6% 2.3% 17 8.91 6.96 -214 -0.9% 15 0.6% 18 10.91 6.96 364 1.5% 5 0.2% 19 1.05 0.88 474 2.0% 0 0.0% 20 3.05 0.88 223 0.9% 0 0.0% 21 5.05 0.88 326 1.4% 1.4% 22 7.05 0.88 360 1.5% 0.4% 23 13.05 0.88 0.4% 0.8% 24 15.05 0.88 -180 -0.8% 1.0% 25 17.05 0.88 0.2% 15 0.6% 26 Biased Location: Ext. of Suspended Equip.: Top of conduit, near #21 0.2% 27 Biased Location: Ext. of Suspended Equip.: Top of conduit near #2 15 0.6% 28 Biased Location: Ext. of Suspended Equip.: Top of light fixture, near #15

-1 0 -0.4% 29 Biased Location: Ext. of Suspended Equip.: Top of conduit, next to #15 1.2% 30 Biased Location: Ext. of Suspended Equip.: Top of wall ledge, next to #18 1.4% 31 Biased Location: Ext. of Suspended Equip.: Top of ventilation duct, below #5 10 0.4% 32 Biased Location: Ext. of Suspended Equip.: Top of light fixture, near #3 0.8%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.6, Pag e 2 of 4 TLG Services, Inc. STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 33 Biased Location: Ext. of Suspended Equip.: Top of circuit breaker cabinet, below #20 1.0% 34 Biased Location: Ext. of Suspended Equip.: Top of pipe, below #19 1.0% SU Mean: 0.2% 0.6% SU Max.: 474 2.0% 34 1.4% SU MDL: 354 1.5% 114 4.8% Criteria: 23,700 - 2,370 - SU 2.6 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS (CEILING AND SUSPENDED EQUIPMENT SURFACES)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.6, Pag e 3 of 4 TLG Services, Inc. SU 2.6 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS (UPPER WALL)

Partial view of SU 2.6, facing southeast, showing ceiling and suspended equipment surfaces above, and a partial view of the east wall (left of center)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.6, Pag e 4 of 4 TLG Services, Inc. Partial view of SU 2.6, facing northeast, showing ceiling and suspended equipment surfaces above, a partial view of the north wall (left of center), and the east wall right of center (upper walls are above the horizontal red-line)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.7, Pag e 1 of 2 TLG Services, Inc. APPENDIX F-2.7 SURVEY UNIT DATA GROUND FLOOR REACTOR ROOM - RADIOACTIVE MATERIAL STORAGE ROOM - LOWER WALLS AND FLOOR STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 0.76 -0.72 0.2% 19 0.8% 2 1.76 -0.72 -199 -0.8% 15 0.6% 3 2.76 -0.72 -283 -1.2% 0.2% 4 3.76 -0.72 -146 -0.6% 10 0.4% 5 -0.74 -1.62 -325 -1.4% 24 1.0% 6 0.26 -1.62 78 0.3% 0.4% 7 1.26 -1.62 154 0.7% 10 0.4% 8 2.26 -1.62 170 0.7% 19 0.8% 9 3.26 -1.62 162 0.7% 0 0.0% 10 4.26 -1.62 261 1.1% 1.0% 11 5.26 -1.62 322 1.4% 1.4% 12 -1.24 -2.52 -222 -0.9% 2.3% 13 -0.24 -2.52 -142 -0.6% 15 0.6% 14 0.76 -2.52 215 0.9% 0 0.0% 15 1.76 -2.52 284 1.2% 0.4% 16 2.76 -2.52 348 1.5% 29 1.2% 17 3.76 -2.52 295 1.2% 2.1% 18 4.76 -2.52 557 2.4% 10 0.4% 19 5.76 -2.52 379 1.6% 0.6% 20 -0.74 -3.42 0.3% 5 0.2% 21 0.26 -3.42 272 1.1% 5 0.2% 22 1.26 -3.42 386 1.6% 1.9% 23 2.26 -3.42 253 1.1% 0 0.0% 24 3.26 -3.42 386 1.6% 3 9 1.6% 25 5.26 -3.42 497 2.1% 5 0.2% 26 0.76 -4.32 52 0.2% 0.4% 27 1.76 -4.32 78 0.3% 0.2% 28 2.76 -4.32 493 2.1% 1.2% 29 3.76 -4.32 436 1.8% 0.2% SU Mean: 160 0.7% 0.1% SU Max.: 557 2.4% 39 1.6% SU MDL: 377 1.6% 112 4.7% Criteria: 23,700 - 2,370 -

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.7, Pag e 2 of 2 TLG Services, Inc. SU 2.7 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.8, Pag e 1 of 2 TLG Services, Inc. APPENDIX F-2.8 SURVEY UNIT DATA GROUND FLOOR REACTOR ROOM - RADIOACTIVE MATERIAL STORAGE ROOM - UPPER WALLS, CEILING AND EXTERIOR OF SUSPENDED EQUIPMENT STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 0.78 0.31 -149 -0.6% 5 0.2% 2 1.78 0.31 75 0.3% 0 0.0% 3 2.78 0.31 60 0.3% 1.2% 4 3.78 0.31 125 0.5% 15 0.6% 5 0.28 1.21 -339 -1.4% 0.4% 6 1.28 1.21 -271 -1.1% 1.6% 7 2.28 1.21 -221 -0.9% 19 0.8% 8 3.28 1.21 -259 -1.1% 49 2.1% 9 4.28 1.21 0.1% 0.2% 10 -1.22 2.11 121 0.5% 0 0.0% 11 0.78 2.11 0.2% 0 0.0% 12 1.78 2.11 -161 -0.7% 0.4% 13 2.78 2.11 -183 -0.8% 0.6% 14 3.78 2.11 197 0.8% 0.8% 15 4.78 2.11 421 1.8% 0.2% 16 5.78 2.11 661 2.8% 24 1.0% 17 0.28 3.01 -214 -0.9% 1.0% 18 1.28 3.01 -320 -1.4% 0.6% 19 2.28 3.01 -263 -1.1% 0.8% 20 3.28 3.01 0.3% 1.4% 21 4.28 3.01 68 0.3% 0.2% 22 5.28 3.01 661 2.8% 5 0.2% 23 0.78 3.91 592 2.5% 1.4% 24 1.78 3.91 391 1.6% 10 0.4% 25 2.78 3.91 528 2.2% 1.6% 26 3.78 3.91 645 2.7% 1.0% 27 0.28 4.81 288 1.2% 2.1% 28 1.28 4.81 14 0.1% 1.2% 29 2.28 4.81 547 2.3% 0.2% 30 3.28 4.81 695 2.9% 0.2% 31 4.28 4.81 919 3.9% 0.4% SU Mean: 145 0.6% 0.4% SU Max.: 919 3.9% 49 2.1% SU MDL: 354 1.5% 114 4.8% Criteria: 23,700 - 2,370 -

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.8, Pag e 2 of 2 TLG Services, Inc. SU 2.8 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.9, Pag e 1 of 4 TLG Services, Inc. APPENDIX F-2.9 SURVEY UNIT DATA GROUND FLOOR REACTOR ROOM - INTERIOR AND EXTERIOR WALLS, CEILING AND ROOF OF DARK ROOM STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 -5.19 0.11 459 1.9% 0.2% 2 -7.19 0.11 424 1.8% 0.4% 3 -9.19 0.11 67 0.3% 5 0.2% 4 -6.19 1.81 135 0.6% 24 1.0% 5 -8.19 1.81 154 0.7% 0.4% 6 -10.19 1.81 352 1.5% 1.6% 7 -7.19 3.51 542 2.3% 0.2% 8 -9.19 3.51 386 1.6% 0.4% 9 6.11 0.22 413 1.7% 1.2% 10 7.61 0.22 402 1.7% 0 0.0% 11 9.11 0.22 269 1.1% 0.4% 12 10.61 0.22 288 1.2% 0.2% 13 5.36 1.52 299 1.3% 15 0.6% 14 6.86 1.52 413 1.7% 0.4% 15 8.36 1.52 113 0.5% 0.6% 16 9.86 1.52 128 0.5% 10 0.4% 17 11.36 1.52 626 2.6% 0 0.0% 18 7.61 2.82 307 1.3% 1.0% 19 9.11 2.82 269 1.1% 1.6% SU Mean: 318 1.3% 0.3% SU Max.: 626 2.6% 24 1.0% SU MD L: 320 1.4% 114 4.8% Criteria: 23,700 - 2,370 -

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.9, Pag e 2 of 4 TLG Services, Inc. SU 2.9 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS (EXTERIOR WALLS AND ROOF)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.9, Pag e 3 of 4 TLG Services, Inc. SU 2.9 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS (INTERIOR WALLS AND CEILING)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.9, Pag e 4 of 4 TLG Services, Inc. View Interior view of dark room

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.10, Pag e 1 of 3 TLG Services, Inc. APPENDIX F-2.10 SURVEY UNIT DATA GROUND FLOOR REACTOR ROOM - REACTOR ROOM STAIRS STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 See location on map

-430 -1.8% 0.8% 2 -460 -1.9% 19 0.8% 3 -338 -1.4% 44 1.9% 4 -449 -1.9% 15 0.6% 5 -418 -1.8% 44 1.9% 6 -357 -1.5% 1.0% 7 -289 -1.2% 0.4% 8 -422 -1.8% 24 1.0% 9 -346 -1.5% 0 0.0% 10 -616 -2.6% 0.2% 11 -373 -1.6% 19 0.8% 12 -479 -2.0% 10 0.4% 13 -338 -1.4% 15 0.6% 14 -475 -2.0% 15 0.6% 15 -308 -1.3% 29 1.2% 16 0.2% 29 1.2% 17 -240 -1.0% 0 0.0% 18 -243 -1.0% 15 0.6% 19 -11 0.0% 5 0.2% 20 -190 -0.8% 5 0.2% 21 0.3% 49 2.1% 22 -106 -0.4% 19 0.8% 23 -361 -1.5% 73 3.1% 24 -384 -1.6% 10 0.4% 25 -133 -0.6% 0.4% 26 -384 -1.6% 39 1.6% 27 -251 -1.1% 10 0.4% 28 -426 -1.8% 0.2% 29 -449 -1.9% 44 1.9% 30 0.2% 15 0.6% SU Mean: -315 -1.3% 16 0.7% SU Max.: -11 0.0% 73 3.1% SU MDL: 372 1.6% 104 4.4% Criteria: 23,700 - 2,370 -

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.10, Pag e 2 of 3 TLG Services, Inc. SU 2.10 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-2.10, Pag e 3 of 3 TLG Services, Inc. Side view of reactor room stairs Downward view of reactor room stairs

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.1, Pag e 1 of 3 TLG Services, Inc. APPENDIX F-3.1 SURVEY UNIT DATA REACTOR OFFICE - LOWER WALLS AND FLOOR STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 -1.15 -0.34 452 1.9% 0.8% 2 -3.15 -0.34 585 2.5% 1.2% 3 -5.15 -0.34 695 2.9% 0.6% 4 -0.15 -1.84 436 1.8% 5 0.2% 5 -2.15 -1.84 322 1.4% 15 0.6% 6 -4.15 -1.84 128 0.5% 0.6% 7 -6.15 -1.84 436 1.8% 0 0.0% 8 0.85 -3.34 0.1% 0.4% 9 -1.15 -3.34 22 0.1% 0.6% 10 -3.15 -3.34 235 1.0% 19 0.8% 11 -5.15 -3.34 444 1.9% 44 1.9% 12 -7.15 -3.34 -142 -0.6% 15 0.6% 13 -0.15 -4.84 inaccessible inaccessible inaccessible inaccessible 14 -2.15 -4.84 178 0.8% 10 0.4% 15 -4.15 -4.84 14 0.1% 0.2% 16 -6.15 -4.84 322 1.4% 29 1.2% 17 -1.15 -6.34 0.4% 0.4% 18 -3.15 -6.34 41 0.2% 1.9% 19 -5.15 -6.34 235 1.0% 1.0% SU Mean: 239 1.0% 0.1% SU Max.: 695 2.9% 44 1.9% SU MDL: 326 1.4% 114 4.8% Criteria: 23,700 - 2,370 -

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.1, Pag e 2 of 3 TLG Services, Inc. SU 3.1 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.1, Pag e 3 of 3 TLG Services, Inc. View of reactor office facing south; SU 3.1 is below the mid-height, red horizontal dividing line Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.2, Pag e 1 of 3 TLG Services, Inc. APPENDIX F-3.2 SURVEY UNIT DATA REACTOR OFFICE - UPPER WALLS AND CEILING STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 -0.92 1.15 710 3.0% 0.4% 2 -2.92 1.15 535 2.3% 5 0.2% 3 -4.92 1.15 588 2.5% 0.8% 4 -6.92 1.15 665 2.8% 0.4% 5 0.08 2.65 326 1.4% 0 0.0% 6 -1.92 2.65 934 3.9% 0.6% 7 -3.92 2.65 779 3.3% 24 1.0% 8 -5.92 2.65 855 3.6% 44 1.9% 9 -7.92 2.65 566 2.4% 10 0.4% 10 1.08 4.15 125 0.5% 44 1.9% 11 -0.92 4.15 543 2.3% 1.2% 12 -2.92 4.15 98 0.4% 58 2.5% 13 -4.92 4.15 197 0.8% 10 0.4% 14 -6.92 4.15 919 3.9% 1.9% 15 -1.92 5.65 -153 -0.6% 0 0.0% 16 -3.92 5.65 0.1% 24 1.0% 17 -5.92 5.65 212 0.9% 0.6% SU Mean: 463 2.0% 5 0.2% SU Max.: 934 3.9% 58 2.5% SU MDL: 326 1.4% 107 4.5% Criteria: 23,700 - 2,370 -

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.2, Pag e 2 of 3 TLG Services, Inc. SU 3.2 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.2, Pag e 3 of 3 TLG Services, Inc. View of reactor office facing south; SU 3.2 is above the mid-height, red horizontal dividing line Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.3, Pag e 1 of 3 TLG Services, Inc. APPENDIX F-3.3 SURVEY UNIT DATA FIRST FLOOR REACTOR ROOM - LOWER WALLS AND FLOOR - WEST OF REACTOR CENTERLINE STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 -0.89 -0.39 -502 -2.1% 15 0.6% 2 -2.89 -0.39 -263 -1.1% 19 0.8% 3 -4.89 -0.39 -289 -1.2% 1.9% 4 -6.89 -0.39 -563 -2.4% 24 1.0% 5 -1.89 -2.09 -263 -1.1% 15 0.6% 6 -3.89 -2.09 -168 -0.7% 0 0.0% 7 -5.89 -2.09 -229 -1.0% 10 0.4% 8 -7.89 -2.09 -122 -0.5% 39 1.6% 9 -9.89 -2.09 -335 -1.4% 78 3.3% 10 -4.89 -3.79 0.1% 0.2% 11 -6.89 -3.79 -179 -0.8% 1.0% 12 -8.89 -3.79 -350 -1.5% 1.2% 13 -5.89 -5.49 0.4% 0.8% 14 -7.89 -5.49 -145 -0.6% 0.8% 15 -9.89 -5.49 -263 -1.1% 0.8% 16 -4.89 -7.19 26 0.1% 44 1.9% 17 -6.89 -7.19 87 0.4% 34 1.4% 18 -8.89 -7.19 -335 -1.4% 0.4% 19 -3.89 -8.89 197 0.8% 0 0.0% 20 -5.89 -8.89 216 0.9% 24 1.0% 21 -7.89 -8.89 163 0.7% 0.4% SU Mean: -163 -0.7% 6 0.3% SU Max.: 216 0.9% 78 3.3% SU MDL: 357 1.5% 112 4.7% Criteria: 23,700 - 2,370 -

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.3, Pag e 2 of 3 TLG Services, Inc. SU 3.3 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.3, Pag e 3 of 3 TLG Services, Inc. West facing view of floor and west and north lower walls (below the mid-height, red horizontal dividing line)

Southeast facing view of lower south wall (below the mid-height, red horizontal dividing line)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.4, Pag e 1 of 3 TLG Services, Inc. APPENDIX F-3.4 SURVEY UNIT DATA FIRST FLOOR REACTOR ROOM - UPPER WALLS AND CEILING - WEST OF REACTOR CENTERLINE STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 -0.67 0.59 -636 -2.7% 0.4% 2 -2.67 0.59 -446 -1.9% 5 0.2% 3 -4.67 0.59 -351 -1.5% 1.4% 4 -6.67 0.59 -678 -2.9% 10 0.4% 5 -1.67 2.29 -268 -1.1% 1.2% 6 -3.67 2.29 -397 -1.7% 0.2% 7 -5.67 2.29 52 0.2% 0.6% 8 -7.67 2.29 0.4% 2.3% 9 -9.67 2.29 -515 -2.2% 1.2% 10 -0.67 3.99 -9 0.0% 1.4% 11 -2.67 3.99 40 0.2% 0.2% 12 -4.67 3.99 -142 -0.6% 0.2% 13 -6.67 3.99 -317 -1.3% 1.0% 14 -8.67 3.99 -405 -1.7% 1.2% 15 -1.67 5.69 -587 -2.5% 1.0% 16 -3.67 5.69 -294 -1.2% 10 0.4% 17 -5.67 5.69 0.3% 24 1.0% 18 -7.67 5.6 9 -108 -0.5% 2.5% 19 -9.67 5.69 -237 -1.0% 0.4% 20 -0.67 7.29 -165 -0.7% 0.8% 21 -2.67 7.29 265 1.1% 1.0% 22 -4.67 7.29 0.1% 0.8% 23 -6.67 7.29 -192 -0.8% 29 1.2% 24 -8.67 7.29 -328 -1.4% 1.2% 25 -1.67 9.09 257 1.1% 24 1.0% 26 -3.67 9.09 227 1.0% 1.9% 27 -5.67 9.09 303 1.3% 10 0.4% 28 -7.67 9.09 -439 -1.9% 0.2% 29 Biased Location: Ext. of Suspended Equip.: Top of pipe, below no. 16 0.2% 30 Biased Location: Ext. of Suspended Equip.: Side of light fixture, next to no. 22 15 0.6% 31 Biased Location: Ext. of Suspended Equip.: Ventilation grate, ~1m above no. 3 63 2.7% 32 Biased Location: Ext. of Suspended Equip.: Top of AC duct, ~1m above #9 5 0.2% 33 Biased Location: Ext. of Suspended Equip.: Top of fire sprinkler main, near no. 27 24 1.0%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.4, Pag e 2 of 3 TLG Services, Inc. STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria SU Mean: -199 -0.8% 0.4% SU Max.: 303 1.3% 63 2.7% SU MDL: 377 1.6% 112 4.7% Criteria: 23,700 - 2,370 - SU 3.4 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.4, Pag e 3 of 3 TLG Services, Inc. View facing northeast of upper north wall (above mid-height, red horizontal dividing line) and ceiling View facing south east, of ceiling and upper south and west walls

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.5, Pag e 1 of 5 TLG Services, Inc. APPENDIX F-3.5 SURVEY UNIT DATA FIRST FLOOR REACTOR ROOM - LOWER WALLS AND FLOOR - EAST OF REACTOR CENTERLINE STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 0.32 -0.77 -556 -2.3% 1.9% 2 2.32 -0.77 -521 -2.2% 19 0.8% 3 4.32 -0.77 -578 -2.4% 0.8% 4 6.32 -0.77 -578 -2.4% 1.2% 5 8.32 -0.77 -442 -1.9% 1.4% 6 1.32 -2.47 -179 -0.8% 1.0% 7 3.32 -2.47 -251 -1.1% 0.2% 8 5.32 -2.47 -324 -1.4% 0.6% 9 7.32 -2.47 -168 -0.7% 0 0.0% 10 9.32 -2.47 -229 -1.0% 44 1.9% 11 11.32 -2.47 -430 -1.8% 24 1.0% 12 4.32 -4.17 -274 -1.2% 0.2% 13 6.32 -4.17 -289 -1.2% 1.6% 14 8.32 -4.17 -187 -0.8% 10 0.4% 15 10.32 -4.17 -472 -2.0% 0.4% 16 5.32 -5.87 0.3% 15 0.6% 17 7.32 -5.87 -8 0.0% 0.8% 18 9.32 -5.87 -134 -0.6% 24 1.0% 19 11.32 -5.87 -415 -1.8% 19 0.8% 20 4.32 -7.57 76 0.3% 63 2.7% 21 6.32 -7.57 -4 0.0% 0.4% 22 8.32 -7.57 0.1% 0.2% 23 10.32 -7.57 -251 -1.1% 0.8% 24 3.32 -9.27 308 1.3% 0.2% 25 5.32 -9.27 95 0.4% 0.6% 26 7.32 -9.27 98 0.4% 19 0.8% 27 9.32 -9.27 114 0.5% 0.8% SU Mean: -211 -0.9% 0.1% SU Max.: 308 1.3% 63 2.7% SU MDL: 357 1.5% 112 4.7% Criteria: 23,700 - 2,370 -

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.5, Pag e 2 of 5 TLG Services, Inc. SU 3.5 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.5, Pag e 3 of 5 TLG Services, Inc.

EMBEDDED PIPE / CONDUIT DATA Beta Surface Contamination Checks at Line Ends Location Beta Scan Results, DPM/100 cm 2 Scan MDL, DPM/100cm 2 Scan MDL, % Criteria Removable, DPM/100 cm 2 % of Criteria MDL, DPM/100cm 2 MDL, % of Criteria Pipe Stub in Floor at SE Corner of Room 0 1,932 8.2% 0.6% 116 4.9% Cut end of Pipe, above former hold-up tank (opposite end of pipe run) 0 2,116 8.9% 0.2% 91 3.8% Beta Surface Contamination: Internal Line Swabs Line Segment Length Swabbed, cm DPM/Swab MDL, DPM/Swab 52 -5 148 From cut end (above former hold up tank), to blockage in pipe 40 -24 148 Internal Surface Contamination: NaI Gamma Log Results Line Segment 1: Dist. from start, cm DPM/100 cm 2 % of Criteria 3 -14,620 -61.69% 13 -19,101 -80.60% 23 -19,791 -83.51% 33 -21,687 -91.50% 43 -23,324 -98.41% 52 -27,891 -117.68% Line Segment 2: From cut end (above former hold up tank), to blockage in pipe (this segment is the opposite end if the segment 1 line)

Dist. from start, cm DPM/100 cm 2 % of Criteria 3 -13,414 -56.60% 13 -12,983 -54.78% 23 -10,829 -45.69% 33 -18,067 -76.23% 40 -14,534 -61.33%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.5, Pag e 4 of 5 TLG Services, Inc.

Internal Surface Contamination: Summary SU Mean Result, DPM/100 cm 2: -17,840 SU Maximum Result, DPM/100 cm 2: -10,829 SU Maximum Result, % of Criteria:

-45.7% Criteria, DPM/100 cm 2: 23,700 MDL, DPM/100 cm 2: 15,540 MDL, % of Criteria:

66% West facing view of SU 3.5 (the lower walls are below the mid-height, red horizontal dividing line)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.5, Pag e 5 of 5 TLG Services, Inc. View of pipe sub end at floor level; starting point of segment no. 1 View of segment no. 2; starting point is the bottom cut end

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.6, Pag e 1 of 3 TLG Services, Inc. APPENDIX F-3.6 SURVEY UNIT DATA FIRST FLOOR REACTOR ROOM - UPPER WALLS AND CEILING - EAST OF REACTOR CENTERLINE STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 1.89 1.03 -694 -2.9% 0.2% 2 3.89 1.03 -192 -0.8% 19 0.8% 3 5.89 1.03 -648 -2.7% 1.2% 4 7.89 1.03 -625 -2.6% 39 1.6% 5 9.89 1.03 -469 -2.0% 34 1.4% 6 0.89 2.73 -446 -1.9% 1.2% 7 2.89 2.73 -503 -2.1% 0.8% 8 4.89 2.73 -515 -2.2% 24 1.0% 9 6.89 2.73 -416 -1.8% -3 9 -1.6% 10 8.89 2.73 -408 -1.7% 0.6% 11 10.89 2.73 -713 -3.0% 0.8% 12 1.89 4.43 -169 -0.7% 19 0.8% 13 3.89 4.43 0.2% 0.4% 14 5.89 4.43 0.1% 19 0.8% 15 7.89 4.43 48 0.2% 24 1.0% 16 9.89 4.43 -230 -1.0% 54 2.3% 17 0.89 6.13 -648 -2.7% 0.8% 18 2.89 6.13 204 0.9% 19 0.8% 19 4.89 6.13 0.1% 0.2% 20 6.89 6.13 162 0.7% 5 0.2% 21 8.89 6.13 120 0.5% 0.2% 22 10.89 6.13 -610 -2.6% 10 0.4% 23 1.89 7.83 0.3% 0.6% 24 3.89 7.83 56 0.2% 1.2% 25 5.89 7.83 242 1.0% 1.0% 26 7.89 7.83 21 0.1% 0.2% 27 8.89 7.83 -154 -0.6% 1.0% 28 0.89 9.553 94 0.4% 0.2% 29 2.89 9.53 0.3% 1.9% 30 4.89 9.53 154 0.7% 10 0.4% 31 6.89 9.53 276 1.2% 0.2% 32 8.89 9.53 -199 -0.8% 1.0% 33 Biased Location: Ext. of Suspended Equip.: Top of water pipe, near no. 18 0 0.0% 34 Biased Location: Ext. of Suspended Equip.: Top of clock, near no. 22 24 1.0% 35 Biased Location: Ext. of Suspended Equip.: Top of electric box, between nos. 30 and 31

-1 9 -0.8%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.6, Pag e 2 of 3 TLG Services, Inc.

STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria SU Mean: -203 -0.9% 0.1% SU Max.: 276 1.2% 54 2.3% SU MDL: 377 1.6% 109 4.6% Criteria: 23,700 - 2,370 - SU 3.6 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.6, Pag e 3 of 3 TLG Services, Inc. East facing view of SU 3.6 (upper walls start above the mid-height, red horizontal dividing line)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.7, Pag e 1 of 5 TLG Services, Inc. APPENDIX F-3.7 SURVEY UNIT DATA FIRST FLOOR REACTOR ROOM - TOOL CLOSET - LOWER WALLS AND FLOOR STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 1.42 0 -499 -2.1% 0.2% 2 0.67 -1.3 -335 -1.4% 5 0.2% 3 2.17 -1.3 -365 -1.5% 34 1.4% 4 -1.58 -2.6 -510 -2.2% 1.4% 5 -0.08 -2.6 -251 -1.1% 0.4% 6 1.42 -2.6 19 0.1% 0.2% 7 2.92 -2.6 64 0.3% 0.8% 8 -0.83 -3.9 -434 -1.8% 5 0.2% 9 0.67 -3.9 -464 -2.0% 29 1.2% 10 2.17 -3.9 -324 -1.4% 1.6% 11 3.67 -3.9 -400 -1.7% 10 0.4% 12 5.17 -3.9 -384 -1.6% 39 1.6% 13 1.42 -5.2 -423 -1.8% 1.6% 14 0.67 -6.5 -426 -1.8% 5 0.2% 15 2.17 -6.5 -343 -1.4% 0.4% 16 3.67 -6.5 -365 -1.5% 0.4% 17 -1.24 0.05 -175 -0.7% 15 0.6% 18 -2.74 0.05 -137 -0.6% 15 0.6% 19 -4.24 0.05 -392 -1.7% 19 0.8% 20 -0.49 1.35 -354 -1.5% 1.0% 21 -1.99 1.35 -590 -2.5% 15 0.6% 22 -3.49 1.35 -514 -2.2% na na SU Mean: -346 -1.5% 0 0.0% SU Max.: 64 0.3% 39 1.6% SU MDL: 357 1.5% 107 4.5% Criteria: 23,700 - 2,370 -

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.7, Pag e 2 of 5 TLG Services, Inc. SU 3.7 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS (WITHOUT CENTER WALL POINTS)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.7, Pag e 3 of 5 TLG Services, Inc. SU 3.7 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS (CENTER WALL POINTS)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.7, Pag e 4 of 5 TLG Services, Inc.

EMBEDDED PIPE / CONDUIT DATA Beta Surface Contamination Checks at Line Ends Location Beta Scan Results, DPM/100 cm 2 Scan MDL, DPM/100 cm 2 Scan MDL, % Criteria Removable, DPM/100 cm 2 % of Criteria MDL, DPM/100 cm 2 MDL, % of Criteria Drain Bowl and ~2 ft. long pipe stub 0 1,807 7.6% 19 0.8% 96 4.1% Beta Surface Contamination: Internal Line Swabs Line Segment Length Swabbed, cm DPM/Swab MDL, DPM/Swab Drain Bowl and ~2 ft. long pipe stub 63 0 1 48 Internal Surface Contamination: NaI Gamma Log Results Line Segment 1: From floor drain bowl at floor level, to cut end of pipe (in room below)

Dist. from start, cm DPM/100 cm 2 % of Criteria 3 -23,166 -97.75% 13 -24,458 -103.20% 23 -21,744 -91.75% 33 -18,211 -76.84% 43 -18,986 -80.11% 53 -18,900 -79.75% 63 -13,213 -55.75% Internal Surface Contamination: Summary SU Mean Result, DPM/100 cm 2: -19,811 SU Maximum Result, DPM/100 cm 2: -13,213 SU Maximum Result, % of Criteria:

-55.8% Criteria, DPM/100 cm 2: 23,700 MDL, DPM/100 cm 2: 15,540 MDL, % of Criteria:

66% NOTE: Pipe is thick cast iron, which is likely to be shielding the detector from external sources of ambient background; hence the large negative results.

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.7, Pag e 5 of 5 TLG Services, Inc. Partial view of lower and upper sections of the center wall and south wall (at right side of photo); SU 3.7 data points are located below the horizontal red line View of line segment no. 1 drain bowl, and surrounding floor

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.8, Pag e 1 of 3 TLG Services, Inc. APPENDIX F-3.8 SURVEY UNIT DATA FIRST FLOOR REACTOR ROOM - TOOL CLOSET - UPPER WALLS AND CEILING STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 0.91 0.09 -139 -0.6% 1.2% 2 1.91 0.09 -276 -1.2% 2.5% 3 2.91 0.09 -394 -1.7% 0.8% 4 0.41 0.99 0.3% 1.0% 5 1.41 0.99 -261 -1.1% 0.8% 6 2.41 0.99 -360 -1.5% 2.1% 7 3.41 0.99 -299 -1.3% 24 1.0% 8 0.91 1.89 -265 -1.1% 1.0% 9 1.91 1.89 -348 -1.5% 0.4% 10 2.91 1.89 -417 -1.8% 0.8% 11 3.91 1.89 -288 -1.2% 1.9% 12 4.91 1.89 -276 -1.2% 1.0% 13 5.91 1.89 -177 -0.7% 1.6% 14 0.41 2.79 -375 -1.6% 1.0% 15 1.41 2.79 -398 -1.7% 19 0.8% 16 2.41 2.79 -333 -1.4% 0 0.0% 17 3.41 2.79 -440 -1.9% 2.1% 18 4.41 2.79 -375 -1.6% 1.4% 19 5.41 2.79 -326 -1.4% 0 0.0% 20 3.91 3.69 -322 -1.4% 5 0.2% 21 2.41 4.59 -364 -1.5% 1.0% 22 -1.24 2.05 -371 -1.6% 1.6% 23 -0.49 3.35 -333 -1.4% 34 1.4% SU Mean: -313 -1.3% 0.8% SU Max.: 0.3% 34 1.4% SU MDL: 350 1.5% 119 5.0% Criteria: 23,700 - 2,370 -

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.8, Pag e 2 of 3 TLG Services, Inc. SU 3.8 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS (WITHOUT CENTER WALL POINTS)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-3.8, Pag e 3 of 3 TLG Services, Inc. SU 3.8 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS (CENTER WALL POINTS)

Partial view of upper sections of the center wall; SU 3.8 data points 22 and 23 are located above the horizontal red line

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-4.1, Pag e 1 of 11 TLG Services, Inc. APPENDIX F-4.1 SURVEY UNIT DATA EXHAUST VENTILATION SYSTEM - INTERIOR DUCT AND EQUIPMENT SURFACES (AND DISCHARGE POINT SURROUNDING ROOF SURFACES)

STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 See location on maps -715 -3.0% 15 0.6% 2 -426 -1.8% 73 3.1% 3 -551 -2.3% 0.4% 4 -525 -2.2% 24 1.0% 5 -422 -1.8% 1 0 0.4% 6 -373 -1.6% 19 0.8% 7 -167 -0.7% 0.2% 8 38 0.2% 5 0.2% 9 84 0.4% 0.8% 10 262 1.1% 0.4% 11 0.2% 15 0.6% 12 0.1% 0.4% 13 0.1% 34 1.4% 14 760 3.2% 24 1.0% 15 -449 -1.9% 0.2% 16 -34 2 -1.4% 0.2% 17 0.3% 0.4% 18 0.3% 5 0.2% 19 243 1.0% 44 1.9% 20 -217 -0.9% 0.4% 21 -114 -0.5% 10 0.4% 22 0.3% 0 0.0% 23 -175 -0.7% 29 1.2% 24 217 0.9% 19 0.8% 25 -190 -0.8% 78 3.3% 26 -103 -0.4% 29 1.2% 27 -475 -2.0% 24 1.0% 28 -551 -2.3% 5 0.2% 29 -285 -1.2% 1.0% 30 529 2.2% 10 0.4% 31 -236 -1.0% 78 3.3% 32 -498 -2.1% 24 1.0% 33 144 0.6% 0.4% 34 213 0.9% 44 1.9% 35 -129 -0.5% 1.2% 36 612 2.6% 0.2% 3 7 -582 -2.5% 0.4%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-4.1, Pag e 2 of 11 TLG Services, Inc. STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria SU Mean: -127 -0.5% 12 0.5% SU Max.: 760 3.2% 78 3.3% SU MDL: 372 1.6% 107 4.5% Criteria: 23,700 - 2,370 - SU 1.2 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS (Drawings not to scale see photographs for additional detail)

EXHAUST VENTILATION SYSTEM: INTAKE PLENUM

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-4.1, Pag e 3 of 11 TLG Services, Inc. EXHAUST VENTILATION SYSTEM: PLENUM EXHUAST DUCT TO VERTICAL TRANSITION

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-4.1, Pag e 4 of 11 TLG Services, Inc. EXHAUST VENTILATION SYSTEM: PLAN VIEW OF AIR DISCHARGE POINT ON ROOF

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-4.1, Pag e 5 of 11 TLG Services, Inc. EXHAUST VENTILATION SYSTEM: HEATING RECIRCULATION AIR INTAKE PLENUM

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-4.1, Pag e 6 of 11 TLG Services, Inc. View of first floor exhaust ventilation system intake plenum

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-4.1, Pag e 7 of 11 TLG Services, Inc. View of ground floor exhaust ventilation system intake plenum (at floor level) and thermal column / beam port intake duct (round - at ceiling level)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-4.1, Pag e 8 of 11 TLG Services, Inc. View of roof blower and discharge point

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-4.1, Pag e 9 of 11 TLG Services, Inc. View of interior surface of exhaust duct through access panel; duct is located above drop ceiling of first floor reactor room

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-4.1, Pag e 10 of 11 TLG Services, Inc. View of roof drain survey location (no. 37)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-4.1, Pag e 11 of 11 TLG Services, Inc. View of heating recirculation air intake plenum

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-4.2, Pag e 1 of 4 TLG Services, Inc. APPENDIX F-4.2 SURVEY UNIT DATA REACTOR WATER TREATMENT SYSTEM DRAIN PIPING (LEADING TO SANITARY SEWER SYSTEM) VOLUMETRIC CONTAMINATION DATA Sample No.

Grid Coord., m Sample Media / Descript. Sample Data X Y Radionuclid es Fe-55 Co-60 Ni-63 Cs-134 Cs-137 Eu-152 Eu-154 SOF Criteria, pCi/g: 1E4 3.8 2.1E4 5.7 11 7 8 -------- 1 See attached photos below Sediment sample SU 4.2 #1 / collected from inside drain trap, at bottom of valve sump Sample, pCi/g 0.000 0.109 0.00 0 4.070 0.000 0.000 0.000 0.4 Fraction of Criteria 0.000 0.029 0.000 0.000 0.370 0.000 0.000 SU Mean SOF:

0.4 (a) View of inside of drain trap at bottom of sump, before collection of sediment sample

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-4.2, Pag e 2 of 4 TLG Services, Inc. EMBEDDED PIPE / CONDUIT DATA Beta Surface Contamination Checks at Line Ends Location Beta Scan Results, DPM/100 cm 2 Scan MDL, DPM/100 cm 2 Scan MDL, % Criteria Removable, DPM/100 cm 2 % of Criteria MDL, DPM/100 cm 2 MDL, % of Criteria 4" dia. trap to sewer line embedded in bottom of sump 0 2,327 9.8% 0.6% 127 5.4% 1.5 " dia. line from water treatment system to sump, embedded inside wall of sump 0 2,327 9.8% 2.5% 127 5.4% Sewer line clean out, 4.3 meters north of sump 0 2,327 9.8% 3.3% 127 5.4% 1.5 " dia. line from water treatment system to sump, embedded in floor near former water treatment system 0 2,327 9.8% 0.2% 127 5.4%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-4.2, Pag e 3 of 4 TLG Services, Inc.

Internal Surface Contamination: NaI Gamma Log Results Line Segment 1:

Horizontal section of 1.5" dia. plastic pipe (Drain from reactor water treatment system to sewer trap)

From cut end in sump pit, to elbow at transition to vertical section of pipe.

Dist. from start, cm DPM/100 cm 2 % of Criteria 3 2,513 10.6% 13 3,030 12.8% 23 4,797 20.2% 33 11,992 50.6% 43 13,586 57.3% 53 12,595 53.2% Line Segment 2: Vertical section of 1.5" dia. plastic pipe (Drain from reactor water treatment system to sewer trap): From cut end at floor, near former reactor water treatment system

, to elbow at transition to horizontal section (ending location of segment 1)

. Dist. from start, cm DPM/100 cm 2 % of Criteria 3 -3,346 -14.12% 13 -6,233 -26.30% 23 1,781 7.51% 33 8,718 36.78% 42 9,666 40.78% Line Segment 3: 4" dia. Sewer line (at bottom of sump): From opening at bottom of sump, to back side of pipe trap.(b) Dist. from start, cm DPM/100 cm 2 % of Criteria 0 -15,755 -66.48% 3 -20,710 -87.38% 13 -24,760 -104.47% 23 -16,488 -69.57% 31 -18,986 -80.11% Line Segment 4:

4" dia. sewer line cleanout riser (~4.3 meters down stream of line segment 3 trap): From cleanout cap at floor level, to bottom of C/O riser at transition to horizontal section of sewer line Start Location: Cleanout Cap at Floor Level Ending Location: Bottom of C/O Riser at Transition to Horizontal Section of Sewer Line Dist. from start, cm DPM/100 cm2 % of Criteria 3 -30,016 -126.65% 13 -28,207 -119.02% 23 -19,805 -83.57% 33 -27,173 -114.65% 43 -19,891 -83.93% 54 -16,961 -71.57%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-4.2, Pag e 4 of 4 TLG Services, Inc.

Internal Surface Contamination Summary SU Mean Result, DPM/100 cm 2: -8,166 SU Maximum Result, DPM/100 cm 2: 13,5 86 SU Maximum Result, % of Criteria:

57.3% Criteria, DPM/100 cm 2: 23,700 MDL, DPM/100 cm 2: 15,540 MDL, % of Criteria:

66% (a) Collection of this sample removed all visible sediment from the trap location. (b) Segment 3 location 13 to 23 cm is the approximate section of the trap from which sediment sample 4.2 no. 1 was collected (see volumetric sample results above).

View of line segment no. 1 with NaI detector inserted at cut end in sump; segment no. 2 (vertical section) cut end is shown at top center of photograph

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-4.3, Pag e 1 of 3 TLG Services, Inc. APPENDIX F-4.3 SURVEY UNIT DATA HEATING, VENTILATION AND AIR CONDITIONING UNITS (THREE UNITS)

STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 1 Unit 1- See map for location -331 -1.4% 44 1.9% 2 -593 -2.5% 10 0.4% 3 -627 -2.6% 58 2.5% 4 -779 -3.3% 10 0.4% 5 -779 -3.3% 29 1.2% 6 -631 -2.7% 0.6% 7 -738 -3.1% 29 1.2% 8 -757 -3.2% 5 0.2% 9 -529 -2.2% 15 0.6% 10 -593 -2.5% 0 0.0% 11 -555 -2.3% 0 0.0% 12 -468 -2.0% 34 1.4% 13 -365 -1.5% 10 0.4% 14 Unit 2- See map for location -502 -2.1% 0.4% 15 -487 -2.1% 19 0.8% 16 -684 -2.9% 39 1.6% 17 -548 -2.3% 58 2.5% 18 -559 -2.4% 0.6% 19 -639 -2.7% 39 1.6% 20 -772 -3.3% 0.2% 21 -764 -3.2% 34 1.4% 22 -388 -1.6% 1.2% 23 -703 -3.0% 15 0.6% 24 -430 -1.8% 24 1.0% 25 -365 -1.5% 24 1.0% 26 -589 -2.5% 44 1.9% 27 Unit 3- See map for location -186 -0.8% 15 0.6% 28 0.4% 5 0.2% 29 0.2% 10 0.4% 30 -110 -0.5% 10 0.4% 31 -106 -0.4% 5 0.2% 32 -179 -0.8% 0 0.0% 33 -411 -1.7% 5 0.2% 34 -338 -1.4% 34 1.4% 35 49 0.2% 0.8%

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-4.3, Pag e 2 of 3 TLG Services, Inc. STRUCTURAL SURFACE BETA CONTAMINATION DATA Item No. Grid Coordinates Total Removable X meters Y meters DPM/100 cm 2 % of Criteria DPM/100 cm 2 % of Criteria 36 0.3% 0.6% 37 255 1.1% 5 0.2% 38 Unit 3- See map for location 129 0.5% 10 0.4% 39 -179 -0.8% 39 1.6% SU Mean: -422 -1.8% 15 0.6% SU Max.: 255 1.1% 58 2.5% SU MDL: 372 1.6% 104 4.4% Criteria: 23,700 - 2,370 - Interior view of HVAC unit No. 2, with front panel removed (units 1 and 3 are identical to unit 2)

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix F-4.3, Pag e 3 of 3 TLG Services, Inc. SU 4.3 STRUCTURAL SURFACE BETA CONTAMINATION DATA POINT LOCATIONS

Worcester Polytechnic Institute Document W19-1579-005, Rev. 0 Reactor Decommissioning Final Status Survey Report Appendix G, Page 1 of 1 TLG Services, Inc. APPENDIX G GEL LABORATORY ANALYSIS REPORTS FOR FSS SAMPLES

(NOTE: Click on the Adobe links above for a full copy of the GEL Laboratory Analysis Reports)