ML20197C107

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Decomm Plan Will Progress in 4 Phases.Phase I Incl Loading Unloading & Transfering Fuel to Spent Fuel Pool,Phase II Incl Shipping Fuel to Reprocessing Plant in Id, & Phase III Initiated After Fuel Is Accepted
ML20197C107
Person / Time
Site: Peach Bottom Constellation icon.png
Issue date: 07/31/1978
From:
PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML20197C104 List:
References
NUDOCS 7811150119
Download: ML20197C107 (118)
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DECOMMISSIONING PEACH BOTTON UNIT 1 FINAL REPORT Prepared by Catalytic, Inc. for Philadelphia Electric Company Catalytic Contract July 1978 No. 35930

CONTENTS 1. INTRODUCTION. 1 2. BACKGROUND. 2 8 3. SUMM ARY. 4. DECOMMISSIONING PROGRAM DEVELOPMENT.. .. 12 4.1. Decommissioning Plan and Safety Analysis Report. .12 4.2. Specifications .13 4.3. Control Work Packages. 15 5. PROGRAM ADMINISTRATION. .17 5.1. Project Approach. .17 5.2. Contracte. Organization. .18 5.3. Schedule. . 18 5.4. Manpower. .22 j 5.5. Health Physics. .23 l 6. IMPLEMENTATION OF DECOMMISSIONING ACTIVITY.. . 24 l 6.1. Fuel Disposal. .24 6.1.1. Reactor Defueling. . 24 ] 6.1.2. Spent Fuel Shipping. 25 j 6.2. Primary System Lay Up. . 25 6.2.1. Post Defueling Temperature Monitoring. 25 6.2.2. Fission Product Trapping System Degassing. .25 6.3. End of Life Sampling Program.. .26 6.4. Disposition of By-Product Material. .. 27 6.5. Contalnment Vessel .28 6.5.1. Refueling Floor Area . 28 6.5.2. Reactor Vessel Access .29 6.5.3. Refueling Equipment. . 29 6.5.4. Containment Sump Area. ... 36 6.5.5. Reactor Vessel Internals.. .36 6.5.6. Main Coolant System. . 36 6.5.7, External Fission Product Trapping System. .38 6.5.8. Purified Helium Handling System .42 6.5.9. Nonpurified Helium Handling System. .42 6.5.10. Helium Chemical Cleanup Systerr.. .44 6.5.11. Helium' System Integrity .44 6.5.12. Containment Penetrations. .45 6.5.13. Containment Vessel Access Doors. .45 6.5.13. Flammable Materials in Containment .49 6.5.15. Containment Exhaust System. . 49 1 li

~ - 6.6. Fuel Pool Building.. .49 6.6.1. Fuel Pool. . 49 6.6.2. Spent Fuel Chute and Elevator. .54 6.6.3. Traveling Bridge Hoist. . 54 6.6.4. Fuel Pool Building Sump. .56 6.6.5. Fuel Pool Piping. .56 6.6.6. Ventilation Systems. . 56 6.6.7. Traveling Holsts ; . 57 6.6.8. Access Control. . 57 6.6.9. Decontamination . 57 6.6.10. Fuel Pool Cooling System.. ... 57 6.7. Liquid Naste Area. .60 6.7.1. Component Remova!. .60 6.7.2. Embedded Liquid Waste Piping. .61 6.7.3. Decontamination of ' iquid Waste Area. . 62 6.7.4. Access Control. .62 6.8. Exclusion Area. .62 6.9, Administration Building . 65 6.10. Remaining Portions of the Main Building Complex . 71 6.10.1. Turbine and Auxiliary Buildings. .71

6. P 1 New Fuel Storage and Vault..

. 72 6 W 2. Source Storage Vault. . 72 MD 4 Shield Cooling Water System. .72 .72 6 30.5. Spent Fuel Cask Travel Area. e 10 S. Ventilation System and Stack. .73 0 10.7. Radiation Monitoring System. .73 6.11. 4r.cidental Tanks and Buildings. . 73 6.12. Waste Disposal and Transport. .73 .73 6.12.1. Liquid Waste. 6.12.2. Uncontaminated Waste. .74 0.12.3. Solid Radioactive Waste. .76 6.12.4. Transport anL Burial. .77 7. HEALTH PHYSICS AND SAFETY, .78 7.1. Radiation Protection Training . 78 . 78 7.2. Radiation Expot.ure Control. 7.2.1. Expc3'4re Limits 79 7.2.2. Expos ure Control Planning. . 79 7.2.3. Expos ure to Personnel.. .80 7.2.4. Perso 1nel Monitoring. . 80 7.2.5. Internel Exposure. .80 7.2.6. Radiological Work Permits .80 7.3. Contamination Control. . 81 7.3.1. Control Areas,.. .81 7.3.2. Radioactive Material Control, .82 7.3.3. Decontamination.. .83 t lii ~ -

7.4. Surveys for t adioactive Materials.. .- 85 . 85 7.4.1. Fudiation Surveys. . 86 7.4.2. Contamination Surveys. 7.4.3. Airborne Radioactivity. .. 86 7.4.4. Embedded Pipe Surveys 87 .. 88 7.4.5. Shipment and Transport Surveys. 7.5. Facility Final Survey 88 . 88 7.5.1. Radiation and Contamination Levels. 7.5.2. Survey Techniques. .. 88 . 90 7.5.3. Final Survey of Exclusion Areas. ... 91 7.5.4. Final Survey of Unrestricted Areas. 7.5.5. Final Survey Summary. 92 7.6. Instrumentation. 92 .. 92 7.6.1. Dose Rate Measuring. 7.6.2. Contamination Monitoring .. 93 . 93 7,6.3. Air Sampiers. 7.6.4. Calibration of instruments. . 93 .. 94 7.7. Safety 8. STATUS OF RETIRED PLANT. . 95 8.1. Contaminated Systems. . 95 .. 95 8.1.1. Primary Helium System.. ..... 95 8.1.2. Hydraulic Control Rod Drive System. .. 95 8.1.3. Shield Cooling System. 8.1.4. Fission Products Trapping System 96 . 96 8.1.5. Fuel Handling Purge System 8.1.6. Chemical Cleanup System 96 . 96 8.1.7. Fuel Pool Cooling System. . 96 8.1.8. Radiation and Process Monitors. 8.1.9. Decontamination System 96 . 97 8.1.10. Liquid Waste Disposal System. . 97 8.1.11. Ventilation System. . 97 8.1.12. Nonpurified Helium Handling System. . 97 8.1.13. Purified Helium System. .. 97 8.1.14 Containment Equipment Cooling Water System .. 98 8.2. Noncontaminated Systems . 98 8.2.1. Feedwater System . 98 8.2.2. Inert Gas Generator. . 98 8.2.3. Circulating Water System. . 98 8.2.4. Turbine Generator and Auxiliaries. 8.2.5. Emergency Cooling Water System. 98 .. 98 8.2.6. Chilled Water System. 8.2.7. Refrigeration and Brine Systems. 98 8.2.8. Nitrogen Recondensers. 99 .... 99 8.2.9. Containment Hot Water Heating System ... 99 8.2.10. Electrical System. ... 99 8.2.11. Containrnent Cathodic Protection System. ... 99 8.2.12. Support Systerns iv ..m.

8.3. Safeguards and Radiological Safety. . 100 8.3.1. High Radiation Areas. .100 8.3.2. Controlled Areas... .100 8.3.3. Accessible inspection Areas of Exclusion Area. .100 8.3.4. Llquid Waste Area 130 8.3.5. Remaining Portions of Main Building Complex. .101 8.3.6. Mi cellaneous Radiological Safeguards. .101 9. COST OF DECOMMISSIONING . 102

10. INSPECTION AND REPORTS.

.103 10.1. Inspection. .103 10.2. Reportc. .103 APPENDIX A. Historical Photographs. .A APPENDIX B. Control Work Packages .8 APPENDIX C. Liquid Waste Hazards Summary. .C FIGURES Plant Perspective. . Frontispiece 21 Isometric View of Reactor as installed in Reactor Cavity. 3 22 Main Coolant System - Simplified Process Flow Diagram 5 23 Isometric of Primary Coolant System. 6 5-1 Field Organization. .19 52 Phase 11 Schedule .20 53 Phase ill Schedule. .21 6-1 Activity Rem' ning in Primary System by Principal Decay Mode. .30 62 Gate and Mar a y to Refueling Floor. 31 6-3 Refueling Floor - Controlled A.ea Barricades. . 31 64 Refueling Floor . 33 6-5 Fuel Charging Machine. .34 66 Storage Port with Viewing Device. .35 67 Containment Sump with Grating. . 37 6-8 Monorsil and Electric Hoists. .39 6-9 Purge Condensibles Vessel Shipment. . 41 6 10 Steel Plate over LTDB Cavity. 43 6 11 Typical Capped Penetration to Containment. .46 6-12 Filter on Containment Equipment Door. .47 6 13 Containment Pressure Equalization .. 48 6-14 Ground Floor Exclusion Fence and Controlled Area Barricades.. 50 6 15 Upper Basement - Controlled Area Barricades. . 51 6 16 Lower Basement - Controlled Area Barricades. .52 6 17 Spent Fuel Elevator Arrangement. .55 6 18 Spent Fuel Pool Cover .58 6 19 Spent Fuel Pool Grating and Grapple. .59 6 20 Embedded Pipes .63 6 21 Liquid Waste Area Sump Grating 64 y . -.. - _ _. - -.., ~ _...

.66 6-22 Exclusion Fence at Personnel Lock 6 23 Exclusion Fence Around Containment Vessel. .67 6 24 Exclusion Fence - West Elevation .68 6 25 Plant Layout - Plan of Ground Floor. .69 6-26 Plant Layout - Plan of Plant Operating Floor.. .70 6-27 Solidifying Tritiated Water. . 75 TABLES 61 Radionuclides in Byproduct Material Discharged from Peach Bottom Unit 1. .27 71 Acceptable Surface Contamination Levels .83 9-1 Project Cost. . 102 vi

1. INTRODUCTION This report documents the activi;ies of decommissioning Peach Bottom 1, the first high-temperature gas cooled reactor (HTGR) to operate in the United States. The report also presents the status of systems in the reactor facility;it will be a resource for the facility until such time as dismantlement and disposal of the remaining systems is carried out.

The followhg sections of the report show compliance with the decommissioning work for regulatory purposes. For future inspectors of the facility the report details the status of systems and describes the condition of the plant when mothballed. This material should be of use to those involved in the ultimate dismantling of the plant. Information is provided for those respon-sible for decommissioning other power reactor f acilities, particularly HTGRs. A portion of this report covers the radiological conditions encountered and their impact on work progress. However, one of the principal observations during the progress of the decommission-ing was that the plant was relatively uncontaminated compared with a pressurized water reactor or a boiling water reactor. In addition, the piping and systems were relatively easy to remove and dispose of. The guide for decommissioning activities is contained in Decommission /ng Plan and Sa/ety Analysis Report. Peach Bottom Atomic Power Station Unit 1. This report was prepared by Suntac Nuclear Corporation for Philadelphia Electric Company. It was submitted to the AEC in August, 1974 and was revised in its final form in May 1975. The plan provided for changing technical specifications as the decommissioning proceeded. Based upon the plan, an amendment to the operating license was obtained. This amendment implemented the revised schedule of technical specifications as the decommissioning progressed and culminated in a possession-only license for the facility. The decommissioning plan called for work to progress in four phases. Phase i of the plan i included unloading, canning and transfer of the fuel to the spent fuel pool, and degassing of the helium purification' system delay beds. During this period, preliminary work that did not affect nuclear safety was undertaken. Phase 11 of the plan included shipping fuel to the reprocessing facilities in Idaho. Substantial mechanical decommissioning work was also performed during this period. However, all fuel handling and necessary nuclear safety systems were maintained intact. One.e all fuel was accepted at the reprocessing facility, Phase lit was initiated.-These tasks included removal and decommissioning of all fuel handling systems and the radwaste facilities, as well as final decommissioning of all systems. Phase IV concerns the mothballed status hnd involves continued surveillance of the facility on a periodic basis. 1

2. BACKGROUND Peach Bottom Unit 1 was the first prototype high temperature gas cooled reactor built in the United States, the result of an AEC invitation to industry for development of a prototype HTGR.

In August 1959, the Atomic Energy Commission signed contracts with Philadelphia Electric Company and General Atomic Company for the development, construction, and operation of the Peach Bottom Atomic Power Station. Construction began in 1962; initial criticality took place on March 3,1966, and commercial operation started in June 1967. The plant is owned by Philadel-phia Electric Company and operated successfully 'or over seven years until October 31,1974, at which time the plant was shut down for decommissioning. Peach Bottom Unit 1 is located about 80 miles southwest of Philadelphia on the west shore of the Susquehanna River. The purpose of this prototype HTGR was to demonstrate the feasibility of a high performance helium cooled nuclear power plant and to develop the necessary technical data for application to larger HTGR plante. This 40-MW(e) plant generated 6 total of 1,385,919 gross electrical megawatt hours with a gross plant capacity factor of 74% for a lifetime of 1349 equivalent full power days. Peach Bottom 1 produced superheated steam at 10000F and 1450 psig; overall steam supply system availability was C6% and gross thermal efficiency was 39%. The heart of the nuclear steam supply systt a was a helium-cooled, graphite-moderated 115 MWt reactor core operating at high temperature on a thorium uranium fuel cycle. The reactor core contained 804 fuel elements, each containing 30 graphite fuel compacts. The fuel cycle concept was based on continuous operation of the reactor until core depletion, at which time the spent fuel would be replaced v.ith a new core. ' 1 isu 7etric view of the reactor vessel is shown in Figure 21. Two cores were utilized in Peach Bottom Unit 1. Fuel particles in Core 1 were coated with a single layer of pyrolytic graphite. Fast neutron-induced dimensional changes due to fission product recoils resulted in cracking and distorting of the coatings on the fuel particles. The broken coatings, in turn, caused the compacts to distort and swell. In addition, the radial expan-sion resulting from the swelling cracked the graphite sleeve containing the compacts. A total of 90 fuel elements in Core 1 developed cracked sleeves. The existence of cracked fuel elements was detocted by an increase in total primary system activity. However, plant operation was not impaired; reactor operation "/as continued, and pri-mary loop activity reached 270 curies, well below the design activity level of 4225 curies. Core 1 accumulated 452 equivalent full power days before it was prematurely replaced with ' Core 2. This core contained improved Buffer Isotropic (BISO) coatings on the fuel particles and operated successfully for its entire design lifetime. However, in June 1974, it was discovered 2

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that cesium 134 was being released into the primary system as the core reached its full lifetime. This condition did not occur in Core 1 because of its early replacement. The 134Cs activity increased the primary system activity by a factor of about 4; however, it did not have any significant impact on the decommissionlag work since most of the primary helium system was untouched. The increased activity did, however, increase the radiological controls and radiation exposure during the General Atomic End of Life sampling program. The Peach Bottom 1 steam generation system consisted of two helium coolant loops that transterred heat from the reactor core to two shell and tube steam generators. Superheated steam (10000F and 1450 psig) frcm the two steam generators was combined at the throttle of the turbine generator to produce 45 MW of gross electrical power. Gross thermal efficiency at design operating conditions was approximately 39E Figure 2 2 shows the main coolant system process flow; Figure 2 3 is an isometric view of the primary coolant system. The primary helium system contained a purge system for the fuel elements th'at trapped fission products before they were released into the primary system. Componentsincluded a trap in each fuel element. a condensibles trap, water cooled delay beds, dust removal filter, low temperature delay beds, and a second dust removal filter. Purified helium system components, included oil removal filters, liquid nitrogen traps for 85 r removal, tanks, valves and manifolds. The compo-K nents of the purge system are unique to Peach Bottom 1 and are not common in later HTGR designs. During initial operation, high steam genern shell tamperatures were encountered; these resulted in a slight reduction of operating steam conditiMs. Tne steam generator shell is cooled by a cold t.elium return flow system. The insufficient ficv/ in the bottom head area caused shell temperatures to reach maximum at about 60% rated power. To correct the problem, additional baffles were installed in the steam generator interiors to direct coolant flow to the bottom head. An external bottom head cooling unit was also added to each steam generator. Reactor power was increaseo to 100%, but operating steam temperature and pressure were reduced slightly as a result of the limiting steam generator shell temperatures. No steam tube leaks developed during the operating lifetime of the plant. Dif ficulties encountered in operation involved moisture cell f ailures and carryover of purified helium compressor oil. Moisture monitor cell failures resulted from exposure to dry helium and hydrocarbon impurity levels in the main loops. This proWm was allevia:ed by adding standby moisture monitors to provide at power testing capability and permit replacement of cells before f ailure. It was determined that the purified helium compressor oil carryover contributed to the hydrocarbon impurity levels in the main loops. Replacement of the internal components of the oil separation section of the compressor, the charcoal filters and the oil demister minimized oil leakage into the inain loons. A research and development program was conducted at Peach Bottom 1 throughout the reactor lifetime. This program, conducted by General Atomic and Oak Ridge National Laboratories, included core component performance, fission product release and plat 60ut, circulating activity, coolant chemistry, and other important features of reactor operation. Post irradiation examina-tion of test fuel elements was also performed as part of the program. In addition, a post-4 w ., - -..~--m4.-w -.e-,-. -, = w,,- -+r -=cmm,-- r ,,e,-- --nw -~ y

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shutdown program was conducted by General Atomic to validate generic HTGR design codes and predictions by comparison of actual to predicted physics, thermal, fission product, and materials behavior. This sampling program, as it af fects the decommissioned status of the plant, is discLssed in Section 6.3. Throug: out Peach Bottom 1 operation, excellent agreement was found between predicted and actual core physics characteristics, thus verifying the methods used and providing a reference data base for application to larger HTGR plants. The reactor control system functioned exceptionally well and the plant was operated in a load-following manner during the majority of its li'etime, demonstrating the ability of an HTGR to run in this manner. The decision to decommission the Peach Bottom HTGR was based on a study of the benefits to be derived from further operation beyond depletion of Core 2, relative to the investment necessary to satisfy the AEC's requirements for a full term license. The study indicated that any gains from continued operation of this relatively small plant were not sufficient to justify the large expenses that would be incurred in satisfying changes needed for a permanent license. With results of the study in hand, contracts were let by Philadephia Electric for preliminary studies to determine the best method of decommissioning. The method chosen, mothballing, is the subject of this report. 1 I l l 1 l 7 ~_ _ _,-..~__._, i

3.

SUMMARY

siter operating for seven years and with two cores, the Peach Bottom 1 HTGR reactor was shut down. This report discusses the decommissioning of the facility to mothball status. A study was performed to determine the best method for decommissioning. Based on technical and eco-nomic evaluation of several options, mothballing of the facility under a Part 50 Possession Only license was selected. Based on the mothballing concept, a plan for decommissioning was prepared that included a safety analysis of the facility during and after the decommissioning. The plan was submitted to the AEC and us accepted with one amendment. Included were revised facility technical specifi-cations that were applicable at various milestones during particular phases of decommissioning. As an initial step, the plan called for removal of all fuel and degassing of the primary system, followed by removal of flammable and radioactive materials from accessible areas. The decom-missioning plan also required establishing exclusion areas around the containment and the fuel pool building, with provision for periodic inspection during the duration of the mothball status. Engineering of the decommissioning plan was performed by Catalytic,Inc. under subcontract to SUNTAC Nuclear Corporation, a joint venture company between Catalytic and NUS Corporation of Rockville, Md. (The joint venture ended before decommissioning work began.) Management, engineering, and procurement were also the responsibility of Catalytic. All work that was safety related, or that could result in release of significant radioactivity from the plant, was performed using written procedures and specifications. A total of 16 Specifications and 71 step by step Controlled Work Packages (CWP) were developed for specific work to be done during decommissioning. The CWPs contained quality and radiological sign-offs to provide control of all tasks. The work of decommissioning was performed during three phases. Phase i included all defuel-ing operations and degasification of the primary and purification systems. Phase il included - work on systems not affecting nuclear safety while fuel was on site during the fuel shipping phase of the work. Phase lli involved decommissioning of the fuel pool, radwaste system and final clean up and surveys. Work was performed by a combination of utility and Catalytic per-sonnei. Construction management services and radiation protection services were contracted to Catalytic, Inc. Catalytic also provided necessary laborers and carpenters for the job, in addition to subcontractor services. Philadelphia Electric provided the construction forces and foremen. The original schedule for the work allowed 97 weeks for decommissioning. Actual elapsed time was 170 weeks. Delays were caused by inability to receive fuel on schedule at the site in Idaho; a labor strike, again at the fuel receiving site in Idaho; and delays caused by outage work at Peach Bottom Units 2 and 3. 8 u. eea y w s a a-- ~ < .-e -\\w.., - - s am

Defueling began immediately af ter shutsown. Following defueling, the General Atomic Company End of Life sampling program was conducted to obtain samples of pnmary loop and steam gen-erator components for metallurgical analysis. Actual work on decommissioning began on Janu-ary 20,1976; final site werk was completed February 16,1978. At peak,39 men were utilized for the project. This work force included staff and four health physics personnel. Approximately 28,600 man hours, including 5200 manhours of health physics, were expended on the actual work of decommissioning. The total cost for decommissioning Peach Bottom 1 was $3,524,000. Radiation protection was provided by Catalytic for all contracted work, with Peach Bottom facil-ity H.P. personnel responsible for work involving defueling and gaseous waste. Control was maintained by the use of radiological work permits for controlled areas. Routine radiation pro-tection training was conducted by the Unit 2-3 staff and special training for specific work was performed by the Catalytic H.P. staff. Some meck ups were used and all supervisors vsere trained in management of work involving radiation. The reactor vessel, primary system piping and steam generators remain in place. Except for elec-trical insulation and graphite components within the reactor vessel, all flammable materials were removed from the systems. These included all charcoal traps fiom the helium purification system, as well as all oils and other flammable liquids and solids. All potentially radioactive liquids were removed from the system. According to the decommissioning plan, refrigerants were drained, cooling water was drained, the fuel pool was drained, and the liquid waste system was decommissioned. Gaseous radioactive wastes were monitored and released from the helium purification system. All spent fuel was removed from the reactor and shipped to a facility in Idaho; all 804 fuel ele-ments were accounted for. The reactor vessel and primary system remain intact. Dummy fuel elements were placed inside the reactor vessel to provide support for fuel elements during defueling. Openings in the primary system made during the End-of Life sampling program were welded closed. The missile beams were replaced over the reactor vessel access ports and the crane was deactivated. Fuel handl-ing equipment was deactivated and secured in place en the refueling floor. Access to high radia-tion areas was closed with multiple bolted or welded barricades. Access to clean inspection areas was provided through locked gates or doors as required by the decommissioning plan. Provision was made for visual inspection of the accessible areas, including the subpile room and the containment sump. All penetrations to the containment were cut and capped. A ventila-tion filter was installed in the equipment hatch for atmospheric pressure equalization of the containment vessel nd as a check on the presence of airborne radioactive materials. The fission product delay beds and their associated f;lters were removed. The most radioactive of these vessels, the purge condensibles trap, read 30 R/hr on contact; the part was removed following a planned course of action to minimize radiation exposure. No other component had any appre-ciable contact radiation. The spent fuel pool was drained and cecontaminated. The fuel racks were cut and disposed of as radioactive waste. During this work, some low levels of alpha airborne radioactivity were 9

238 u for P found. The source of this radioactivity was not identified, but was considered to be MPC calculations. Other fuel handling equipment was deactiviated, and the fuel transfer chute was removed and sealed with a blank flange. It was found that the spool piece removed from the transfer tube was more contaminated than other components. The fuel pool cooling system was removed and a locked exclusion area was established for the fuel pool building. The liquid waste area was decommissioned by removing all equipment except some sections of embedded piping. Low levels of contamination exist within these pipes and, as a result, this area was also established as a locked exclusion area. Other areas of the plant were cleaned and contaminated equipment was removed. Floor drains were either cleaned or removed. Some ducting was removed, as well all filters. The stack was found to be not contaminated. The machine shop area was in active use by Philadelphis Electric Company supporting work crews at Units 2 and 3 and was not surveyed, in the turbine and auxili-ary buildings, services necessary to support use of the shop and administration areas were lef t intact. The turbine-generator and the diesel-generator were retained for future use. Radioactive liquids were either monitored and released, or solidified and shipped as solid waste. A total of 235 gallons of oil were disposed of in this manner, as were 72 gallons of tritiated water. Tritiated water solidification was performed by workers protected by bubble suits. Surveys were performed on potentially contaminated material and only contaminated waste was disposed of by licensed burial. Much of the material removed was not radioactive; clean materials were salvaged as scrap or disposed of as trash. Except for sealed sources, all waste was shipped as LSA (Iow specific activity) material. Drums, wooden boxes and tanks removed during decommissioning were used as shipping containers. A special box containing lead and concrete shielding was used for shipping the purge condensi-bles trap. The spent fuel shipping cask was used to ship graphite reflector elements and a con-trol rod guide tube to the burial ground. A total of 490 containers containing 14,000 ft3 and 380 curies was shipped for burial as radioactive waste. Exposure control planning was not necessary for most work. However, the work involved in removing the purge condensibles trap was carefully planned and actual removal of this vessel was accomplished without incident. Total exposure for the decommissioning work was 8.95 man-rem, with the maximum does to an individual 740 mrem. Dosimetry was by commercial TLD service supplemented by in-house TLD badges and self. reading dosimeters. No internal exposures occurred on this work. Whole body counts were performed before an indi-vidual started work and on completion of work. In addition, before and after urine specimens were collected from workers handling tritium. Spread of contamination was minimized by establishing controlled access zones around specific work sites. Where necessary, tents were used to prevent the spread of chips or splatter from cutting or chipping operations. Major controlled areas were established in the fuel pool building, in the liquid waste area, and on the refueling floor. Fuel pool and refueling floor con-trolled areas were maintained clean by papering. The liquid v.aste basement was designated a controlled area until final decontamination was accomplished. 10

7 dpm/100 cm2 were found on primary system Interior surface contamination levels up to 10 6 2 on fuel pool components. These components were cut components and up to 10 dpm/100 cm by reciprocating saws or portable band saws. Flame cutting was used on vessels whose surface 4 dpm/100 cm2. Internally contaminated components were bag-contamination was less than 10 ged or taped closed as they were cut out. Decontamination was required for the spent fuel pool walls and various local areas as well as for i some embedded piping. Commercial wall cleaner was found to be very effective for cleaning the fuel pool walls. Foaming bathroom cleaner, solvents and rags, and an absolute filter vacuum cleaner were also used where applicable. High pressure water was used to clean embedded ~ piping. Surveys for radiation and radioactive contamination were routinely performed during all decom-missioning tasks. Surveys were made before starting work in any area. Airborne radioactivity was monitored while operating in highly contaminated areas or cutting into systems that were con-sidered to be potentially contaminated. In addition, special surveys were taken on embedded pipes for direct and removable contamination. When decommissioning activities were com-pleted, a final radiation and contamination survey was performed to provide baseline data for the decommissioned facility and to show conformance with the rquirements of the decom-missioning plan. lonization chamber instruments were used for dose rate measurements. A teletector with a collapsing probe was also used for dose rate measurements, especially in higher radiation areas or for remote areas. In addition, alpha scintillation detectors were used during the final survey and for surveying the new fuel storage area. All radiation monitoring and counting instruments were calibrated and traceable to the National Bureau of Standards. By the end of the decommissioning work, Peach Bottom 1 had been retired or decommissioned according to the plan. A few systems outside the exclusion areas were kept in service to provide utilities at Unit 1 All radioactivity and radiation are within exclusion areas which are locked with controlled keys. High radiation areas exist wl thin the loop cavities and within the reactor vessel cavity; however, access to these areas is blocked. All accessible spaces within the exclusion areas had, at completion of the project, radiation levels less than 1.0 mR/hr. In all other areas of the plant, the radiation levels were less than 0.01 mR!hr and no contamination existed. 1 l 11 ..-... ~. ~ _., _.. ~ _. _ - -.. _.,, _..,, _. -. _.. _ -. i

4, DECOMMISSIONING PROGRAM DEVELOPMENT 4.1. DECOMMISSION PLAN AND SAFETY ANALYSIS REPORT The basis for all decommissioning work on Peach Bottom Unit 1 was the " Decommissioning Plan and Safety Analysis Report." This report was prepared for the Philadelphia Electric Com-pany by SUNTAC Nuclear Corporation. Philadelphia Ele tric Company (PECO) submitted this report to the AEC in August 1974. During the review by the AEC, one revision was made in May 1975, SUNTAC was directed by Philadelphia Electric to investigate the decommissioning alternatives; i.e., dismantlement, entombment, or mothballing. Each alternative was evaluated for public safety and economic feasibility. The results of this evaluation were submitted to PECO which made the final choice of mothballing. Once Philadelphia Electric elected to pursue this ap-proach, SUNTAC was again contracted to prepare the plan for decommissioning that was ultimately submitted to the AEC The decommisslor'ing plan contained a detailed overview of all of the radiological and mechani-cal procedures required to convert Peach Bottom Unit 1 to a mothballed status. The document consisted of a number of sections, the first of which, the Introduction and the Decommissioning Summary, briefly described the purpose of the report, defined the overall plan and summarized the proposed decommissioned plant status. Other sections discussed the interrelationship between spent fuel shipping and all other decommissioning activities and established the cri-teria for the exclusion areas. The plan specifica ly defined either the work to be done, or the status of a system or area upon completion of decommissioning. As part of the plan, a safety analysis was performed to demonstrate that the facility, during and af ter completion of decommissioning, will continue to remain in a status which is not hazardous to the health and safety of the public. Also, thermal analysis was performed to evaluate the ef-fects of heat generation from the decay of activiation products on ignition of the graphite reflector blocks in the reactor vessel and on the integrity of the concrete vessel enclosure. The analysis concluded that the reactor vessel containing the graphite reflector blocks may be safely placed in lay-up under an air environment and that no degradation of the concrete will occur. The potential for site flooding and the radiological safety hazards that would result were also investigated. the analysis concluded that flooding at the site to the containment vessel's grade elevation is very unkkely. Furthermore, even l' the site were flooded, the effects would be less for the decommissioned plant than for the operational reactor because:

  • All radioactive material remaining in the plant will be contained in the containment ves-sel and the fuel pool building.

12

  • The containment vessel will not be made significantly more buoyant under flooding con-ditions in the decommissioned status than it was in the operational status.

The plan defined and evaluated the decommissioning design basis accident. An analysis was performed to evaluate the maximum containment pressure rise resulting from the maximum accident condition that could conceivably occur during the decommissioning activities. With the reactor in the low pressure shutdown mode, th9 maximum pressure rise possible from the release of 1800 lb of helium coolant is 4.0 psig. This rise would not cause a safety hazarc be-cause it is less than the design basis accident postulated pressure of 8.0 psig for the operating plant. Requirements for site security, exclusion area boundaries and access points, periodic Inspec-tion, and key control for access into the exclusion area were also covered in the plan. The administration of radiological safety during the decommissioning was defined. The plan specifically referred to Phnadelphia Electric Company's responsbility concerning personnel, generation of written proceo';res, performance of radiological surveys, and establishment of emergency procedures. This section concluded that because the radiation safety aspects of the decommissioning and the performance of these activities will be monitored by experienced health physics personnel, neither the decommissioning work nor the final decommissio ied facility will represent a risk to the healt'i and safety of the public. Requirements for licensing, mannin0 records, inspection, and reports both during and after the decommissioning were also included in the plan. 4.2. SPECIFICATIONS The decommissioning plan required use of written procedures or specifications approved by PECO for any decommissioning work that could affect the nuclear safety of the plant, result in release of activity, or result in significant radiological hazard to personnel. To meet this require-ment, Catalytic generated 16 engineering specifications to control the physical removal of equip-ment and piping during decommissioning activities. A brief description of the purpose and sccpe for each of these procedures follows:

  • Nondestructive Examination of Seal Welds Using the Liquid Penetrant Method This procedure established the techniques to be used to perform liquid penetrant examination of welds. The method was utilized to examine pipe cap and vessel noz-zie welds. The acceptance criteria was that of ANSI 31.1.

= Visual Examination of Welds This specification established acceptance standards used during visual examina-tion of welds. The acceptance criteria was based on ANSI B31.1. This procedure was used to examine all seal welds made during the decommissioning before perform-ing liquid penetrant examination. 13 i

  • Handling Contaminated Liquid Waste This procedure described the required precautions to solidify contaminated liquid waste for shipment to a licensed burial site. The procedures were used for solidify.

ing contaminated oil and tritiated water.

  • Erection of Protective Tents and Ventilation This specification provided various methods for building tents and for containing contamination when cutting into or disassembling contaminated systems.
  • Field Welding This specification covered all welding performed during decommissioning. These.

procedures had been qualified in accordance with ASME Section IX.

  • Removal of Insulation Containing Asbestos Fibers This specification described the method used to prevent personnel exposure to asbestos dusts produced by the removal of insulation materials to meet OSHA requirements.
  • Decontamination of Radioactive Contaminated Surfaces This specification described methods proved successful in the decontamination of radioactive contaminated surfaces. Toe specifications gave guidelines based on both the material to be decontaminated and the levels of allowable contamination.
  • Packaging of Solid Radioactive Contaminated Waste This procedure described methods for packaging contaminated radioactive solid waste. It provided guidelines for the use of corrugated cardboard cartons, wooden boxes, and steel drums.
  • Chemical Cleaning Using Hydro Blitz Hi Pressure Washer This specification established the method and techniques to be used for cleaning surfaces using chemicals and a Hydro Blitz Hi Pressure Washer.
  • Health Physics Manual This document was the manual of radiation protection procedures used by Catalytic personnel during the decommissioning of Peach Bottom Unit 1. The proce-dure gave specific guidance on all aspects of the Health Physics control program. It was the primary document used by the Health Physics group in controlling work practices and procedures to ensure compliance with 10 CFR 20, Standards for Fro.

tection Against Radiation.

  • Verification of Helium System Integrity During Pressurization This procedure was written for performing a pneumatic test to demonstrate that contaminated material would not migrate from the helium system into general 14

.~

containment as a result of helium system modifications made during decommis-sioning and the End-of Life sampling program.

  • Operation of Minuteman X 100 Wet Dry Vacuum Cleaner This procedure describes the operation of the vacuum cleaner used for radioactive materials.
  • Wipe Surveys of Embedded Pipes and Drains This procedure describes performance of wipe surveys on embedded pipes and drains. Its main applications were the decommissioning of the containment sumps and drains and the radwaste facility.
  • Direct Surveys of Embedded Pipes and Drains This procedure describes the in pipe direct surveys performed on embedded pipes and drains.
  • Cleaning and Decontamination of Embedded Pipes and Drains This procedure described methods for decontamination of embedded pipes and drains. Guidelines based on the levels of contamination were included.
  • Wipe Testing and Disposal of Sealed Sources in Radiation Monitors This procedure was written to ensure accountability and to ensure that radiation safety practices were followed in the removal, leak test and dispo. sal of sealed sources.

4.3. CONTROL WORK PACKAGES To propeity implement the decommissioning according to the plan and the approved engi-neering specifications, 71 Control Work Packages (CWPs) were prepared. A CWP is a step by-step procedure that contains the specific work steps to complete a particular task or series of tasks. A typical CWP will contain a listing of required materials and special tools required for the work. It contains references to prints and specifications, technical data, sketches, and data sheets as needed. It lists prerequisites that must be completed before beginning the particular work steps and it contains specific work instructions and sign offs for work completion. Special instructions such as cleanliness, quality control, or radiological safety are also part of the CWP. For work in radiation areas or for work requiring contamination control, the health physics con-trols and hold points are inserted as work steps in the CWP with appropriate sign-offs. Where work requires inspection, the quality control inspector requirements are listed and inspector signatures are required to verify work acceptance. The CWP is a per nanent record of all work accomplished when it is complete. The CWPs prepared for the various phases provided specific work steps for decommissioning activities, including removal of radioactive components and the sealing of specific openings, pipes and penetrations that required inspection. CWPs were developed for special cleaning and decontamination work including the stack, new fuel storage area, spent fuel pool and sump, 15 - =

cleaning the control rod drives, and cleaning drains and embedded pipes where specific health physics coverage was required. CWPs were also used for installing the access controls and - barricades required by the plan, for surveying and decontaminating the inspection access areas, and for specific waste handling and disposal such as solidification of tritiated water. A complete listing of the CWPs prepared for the decommissioning is contained in Appnndix B. Y 4 W i 16 ,. _., _.. _.,, l

5. PROGRAM ADMINISTRATION 5.1. PROJECT APPROACH Philadelphia Electric Company was responsible for all aspects of the decommissioning. Engi-neering and planning prior to implementation of decommissioning had been contracted by Phila-delphia Electric to SUNTAC. For implementation of decommissioning, Catalytic, as contractor, provided engineering, planning, procurement, health physics and construction management services.

The three phases of the decommissioning activities were divided as follows:

  • Phase I - performed primarily by Philadelphia Electric n
  • Phases 11 and Ill - Implemented jointly by Catalytic and Philadelphia Electric Company.

l Phase I consisted of reactor defueling, degasification and system draining. Phase il consisted of work that could be performed during the fuel shipping period that did not affect the spent fuel building equipment, the fuel handling equipment, and the general containment integrity. Phase Ill could not be started until all fuel was shipped off site and received at the reprocessing facility. Phase til involved removing the nuclear safety-related systems outside containment end then sealing the containment vessel. Catalytic's responsibilities were to provide planning, engi-neering, construction management, and health physics services to Philadelphia Electric Com-pany. The Catalytic management function involved developing detailed plans, updating estim-ates, projecting actual costs, revising the project schedule and issuing daily schedules. Catalytic's engineering assistance consisted of preparing all Control Work Packages (CWPs), resolving field problems, revising the CWPs accordingly, and performing inspection. A part time home office construction manager and an on site project superintendent were assigned by Catalytic to the project. It was the responsibility of the project superintendent to coordinate the activities of his staf f with the Philadelphia Electric construction forces, in addi-tion, he reviewed job progress, coordinated the work efforts of the subcontractors, arid reported iob status to both PECO and Catalytic management. The construction manager was responsible for the overall accomplishment of the Catalytic ef fort. Catalytic assisted PECO by purchasing wme of the required material, supplies, equipment and tools. Catalytic also issued subcontracts for the transportation and burial of radioactive waste, erection of the perimeter security fence, high pressure water decontamic,ation, and weld inspection. 17

in adddion, Catalytic provided a health physics supervisor and technicians, as well as furnishing radiation monitoring instrumentation. The health physics supervisor was responsible for imple-menting the health physics program on a day to-da" basis in accordance with manuals devel-oped for this project. PECO was ultimately responsible for radiological safety and performed periodic audits of the Catalytic program. Construction forces and supervision provided by Philadelphia Electric Company performed the work of Phases ll and lit. Catalytic provided laborers and carpenters as support personnel for the PECO construction forces. 5.2. CONTRACTOR ORGANIZ.ATION The Catalytic field organization was set up to provide all required functions wlth minimal tech-nical assistance from the home office. The key personnel in the field were the project superin-tendent, project engineers and health physics supervisor. This staff was supported by field ad-ministration support and services shared with the Catalytic Contract Maintenance organization, also present on the site. Home of fice functions were accounting, estimating, cost analysis and some purchasing. An organization chart, Figure 51, shows the interrelationship of the Catalytic field staff with Philadelphia Electric Company personnel and with home office support. 5.3. SCHEDULE The original decommissioning scheduled allowed 97 weeks from the reactor shutdown date to completion of decommissioning activities. The actual deccmmissioning duration was approxi-mately 170 weeks elapsed time. Three major reasons were responsible for the expanded timefrarne. The fuel processing plant in Idaho was not prepered to receive fuel shipments when the reactor was ready to be defueled; this set back the decommissioning approximately 24 weeks. During fuel shipping operations, a labor strike occurred at the Idaho processing facility which caused an eight week delay. The third major delay was caused by the unavailability of PECO manpower to start Phase til af ter the fuel shipment was completed. This situation arose because Philadelphia Electric's construction forces were being utilized on Peach Bottom Units 2 and 3 outages. This delay added an additionat 28 weeks. The reactor was shut down on October 31,1974. The first fuel shipment lef t for idaho on June 26,1975, and the last fuel shipment left on February 17,1977. The End-of Life sampling program was conducted by Catalytic for General Atomic Company between October 1975 and February 1976. Phas'e 11 work activities began on January 20,1976, and ended on July 27,1976. Work was stopped for a three month period during Phase il for an outage at Peach Bottom Unit 2. Phase lli activities started on July 11,1977, and were completed by February 16,1978. Figures 5 2 and 5 3 show the actual schedule of Phase il and Phase til mechanical and radiological work. 18 w.

Project Managar (Home Office) e ~~ 7 PECO l Home Office l Field PECO Field // Construction t Project Superintendent ngi e E ngineer s 8 _J Health Physics Field Field Quality Control PECO Supervisor Project Enciaeer Office Manager Engineer Sut> Foreman 2 i f Health Physics Clerical Carpenters and PECO Technicians Personnel Laborers Craf t Personnet i /hClient interface Part Time 1 FIGURE 5-1. Field Organization

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j 03 / 7 3 2 / 7 6 1 / 7 7 7 7/ 2 eN$ w$,8 O[s@ 2Wr<E,** gC" 0wOZwh2$C' c 62 / 3 9 1 !3 l I 2 1 i 3 1 le 5 3 u I de 7 h 2 / c 2 .!f S 1 0 1 2 e U sa h 3 E I P 1 L 2 U D 2 -5 E 6 H / E 2 C R S U 0 G 1 3 I 1 / E F 1 SA 3 H 2 / P t me ts s y rs m S o e 2 s m s ts 0 e v 4 u r b p S l A e i H m O re o b d C e i. e n i p e i m. S rs f 5 a r t 0 r t b u n d A p T fe ru e a t e 3 n e m e O H t e r F u d n t n b d t t tr R a e a e s r n e u h e fi w a l a S E G a p t t n t d t u t 0 to n v s r G N O 4 P r I m m n e O e P o e a r o e 't s 3 g u tr A D 't o r d m s r i O t; o A u p b r m b F C o r K N h t ts o n d C' s p. g p fr s P v C E S s R i e a d p n A K ro d d. !o ta B 'e ra ra u e t r 3 r a u n v D W e o i 3 m io. S e c t. e k m e g te tw 0 to u s s n r o a P E s t n r C te ru s a ta ta M i Pla F ra ri ra k F re C n d d s W o. r a o D e m a u m e P a E r C C C ta t n T i n a n r a r K e d n s t e a o t r d r r k t e m C w F d G n e t, m R n r t S e e tc s an a P a m i F s r n o e re u p ia p T r T C u r D !a t a te o n a a O C t d g W io a r o m y u o o m lo e O f A t p s. D S S a f S W m G F m C C e a e s l d a r T C Tr n' o fo s o H u e fa n W e H i t u i s E h o. i v p L t O O te s n n f m r. fo te !e. e o e w u y o p i e p ta a u h H h f O 6 H t i C a a u N p t C o P C a o C W l D L ia r a ra la e e la ta v e d h tc s R 7 la la la n c s c t i d d e e e e O t f. n o o o o o V e c n v v v v v h T v v v v i e a i a v v u n o o o o m t d d t e r e e e e a r C 9 o o m a m m m m m la N la p e t m m m m a e h m m + tu e e e e h e e t h u 1 o c u O S S C R R R R h S C E e C S R R C C R R R R C oo f 4 ? a li il

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l PHASE lit SCHEDU i CONTROL WORK PACKAGE 4 19j Week Endmq 7/22 7/29 86 8/12 8!19 8/26 9/2 9/9 9/16 9/23 9/30 10/7 10/1d Shield Cochng System - Cut and Hernove Pipe Shield Coolmq Systeg Remove Tank T 56 Exhaust Pienum Oil hitei Removal I _ Exhaust FHters Remova! Decon. System: Install Blank Flange + Liquid Waste Disposal System Remove Eq.and Tanks I' Liquid Waste Disposa; System Remove Sys. Pipmg Decon. bquid Waste Bldg and Sump Vent. System - Removal of Enhaust Fdters Vent. System - Gurtmg Removal Vent. Svstem - Weld Covers to Ducts Vent:1ation Stack Survey and Decon. Decont. of New Fuel Storage Area Remove Spent Fuel Pit F dters Remove Scent Fuel Pit Heat Exchanger Remove Spent Fael Pit Coolmg Water Pumps Remave Spent Fuel Pit Externai Piping ~~I Remove Spent Fuel Pit Coolino System and Cap Pipmg Cont. E quip. Coohng Water System Removal I Adm Bldg lab. Laundry Drams and Exh. I I I I I Cecca Waus and floor of Spent f uel P.t Remove Spent Fuel Pit Sump Pump Decan Soent Fuel Pit Sumo Pump Remove S.F.P Tube Spool Pwce Install Cap on S F P. Tutu Remove S F P. Tube and S F. Elev Install S F. Pit. Gratmg Weid Cask Manorad Door Closed Decon. Spent Fuel Pit Access Area Decon and Deact. Cask Travehnq Hoists Deact & Dycon S F. Grappte Crane Remove $ F. Pit Tools Survey & Decon. Inso Access Arcas l Decom. Containment Samp Decom Canmng & Chargmq Machs Decom Transfer Machine l Decom. Fransf er Machme Decom. Isolation Vanes install AbsoNte Fdter and Wetd Doois l l 1 Derom Personnel Air Lock l a-i Emtmn ut Gates and Barricades j Cut and Cop Misc Penetrations Remove Rad Mannor Chech Sources j j i l Post incident filter and Vent. Purge Filter j Cut and Cap Steam Gen. Pipe ,1 4 I i i

k LE l l 197g '7 10/21 10/28 11/4 11/11'11/18 11/25! 12/2 12/9 12/16 12/23 12'30 1/6 1/13 1/20 1/27 2/3 2/10 2!17 2/24 4 I i t t i i ~~t t-t- i jt 1-- j j '1 l 1--t i i i t t j i- ~I i i 7 i i i i [ i I_~T_ { i~ -a F l I i r-~7 T~ i i i t-i i ^i 1 [ 4 l 1 zI I A -a 1 I + i .L.___ } J_1 1 (q.- -- y -.__y..-- 7-j__. _. n--4. 7-_- L_- _ _4 _..____.,___.____4 i __t_._,_._.._) r -+ l t, _ - l i i.r i i i i .._..._ __7_. _.. _ _ t _.. q _ __ i _ _. -.._ _ _ t 4 l 7 i 1 i i i l + FIGURE 5 3. Phase ill Schedule { l 21 3 mmmmmes

? Decommissioning activity in net weeks of mechanical work totaled 65 as shown: Weeks 18 End of Life Sampling 15 Phase 11 32 Phase lli 5.4. MANPOWER The workforce peaked at 39 men during Phase 11 operations. This force was composed mainly of Philadelphia Electric construction mechanics and union carpenters and laborers. A total of 14,420 manual manhours were expended, including health physics technicians and clerical personnel. The Phase til workforce peaked at 33 people for a total of 14,140 manual manhours. Again, the major component was Philadelphia Electric construction mechanics and union carpenters and laborers. Clerical personnel were on a part time basis, and labor hours for health physics tech-nicians were accrued as part manual and part non-manual. A distribution of maximum and average workforce personnel for Phases !! and lllis shown in the following tables: PH ASE 11 PERSONNEL MAXIMUM AVERAGE Foremen 2 2 Construction 20 13 8 5 Carpenter Labor Field Staf f 6 4 3 3 Health Physics Total 39 27 PH ASE Ill PERSONN EL MAXIMUM AVERAGE Forernen 2 1 Construction 18 9 6 4 Carpenter Labor Field Staff 4 3 Health Physics 3 2 Total 33 19 During both Phases 11 and Ill, Catalytic maintained a etaff of one project superintendent, two engineers, one health physics supervisor and a maximum of three health physics technicians. 22

5.5. HEALTH PHYSICS The Health Physics effort was coordinated between Philadelphia Electric Company and Cataly-tic, Inc. As licensee, PECO was responsible for compliance with state and federal regulations and the requirements of the Unit 1 Technical Specifications. Catalytic provided the Health Physics Supervisor, technicians and instrumentation during Pnases ll and 111 of decom-missioning. Philadelphia Electric Company performed all health physics activities during defueling and fuel shipping as well as those periods of Phase ll during which decommissioning activities were stopped. The Health Physics Supervisor was responsible to the Project Superintendent for integrating I health physics planning, controls, and operations with the decommissioning activities. He was also responsible to the Health Physics and Chemistry Engineer of Philadelphia Electric Com-pany for ensuring compliance with state and federal regulations, PECO health physics proce-dures, and Unit 1 Technical Specifications. During Phase 11 and Phase Ill, Catalytic health phvsics personnel performed the necessary radia-tion and contamination surveys, provided radiation work permits, maintained radiological con-trolled areas, and supervised the final decontamination of the facility. The Philadelphia Electric Company health physics staff was responsible for personnel monitoring, bio-assay services, and monitoring of all ef fluent discharges. l 23

l

6. IMPLEMENTATION OF DECOMMISSIONING ACTIVITY Four general categories of material were to be removed from the facility under the decommis-sioning plan: certain components of the fission product trapping system and helium purification j

system; all drainable liquids (water, olls, and refrigerants); all flammable materials except elec-trical cables and solid graphite in the reactor vessel; and other items outside the primary system that were contaminated with radioactive material in excess of the amounts specified in NRC Regulatory Guide 1.86. In some situations, other components, piping, and material were removed to obtain access and working clearances around equipment. Additional steps taken included sealing certain systems to prevent unrestricted release of radioactivity, erecting barriers to restrict access to radiation areas, and decontamination and cleanup of all accessible areas within the Unit 1 facility. l 6.1. FUEL DISPOSAL 6.1.1. Reactor Defueling Preparations for defueling the reactor commenced immediately following final shutdown on October 31,1974. Removal of Core 2 from the Peach Bottom Unit No.1 reactor was completed l on June 11,1975. All 804 fuel elements were canned, leak tested, and stored in the spent fuel pool. Fuel inventory was maintained by the use of individual record cards for each fuel element. Several logs and core maps were kept to enhance fuel inventory contrcls. A total of 513 dummy elements were inserted into the core to maintain lateral support of the core during defueling. Dummy insertion control sheets were utilized to document the loading of dummy elements into the core. With the exception of one control rod absorber and three hex reflector elements (GAC surveillance program), no other components were removed from the reactor. During defueling, the primary coolant system, purification system, helium transfer systems, closed coolant sys-t9ms, and emergency power systems remained in service to provide core cooling and control of impurity levels. Philadelphia electric records, as well as an inventory count, accounted for all 804 elements. In addition, an alternate method was used to verify the defuelin0 status of the reactor. The fact that all spent fuel canisters are sufficiently welghted to compensate for buoyancy eliminates the possibility that a canister was stored empty. One postulated occurrence was a mix up between a fuel element and a dummy, resulting in the canning of the dummy while leaving a fuel element still in the core. While this would require four errors in the fuel Inventory controls, this possibil-ity had to be considered. To eliminate any such error, radiation readings were taken on 640 of the 804 elements at the canning machine, verifying the presence of fuel. Because these 24

measurements were not taken on all the fuel elements, another verification method was de-veloped as follows:

  • The core was mapped using the viewing device to Jdentify the location of the dummy elements. All dummies were painted white and were therefore easily identi.

fiable. Two independent surveys using different personnel were conducted to en-sure accuracy.

  • At those core locations not containing a dummy, the absence of the fuel element was verified by partially lowering a dummy into the void. An individual control sheet was used for each void to document this operation.

I i l 6.1.2. Spent Fuel Shipping Starting on June 24,1975, spent fuel was shipped to Aerojet Nuclear Company in Idaho (now EG&G), utilizing two shipping casks. A total of 44 fuel shipments were made by truck in the j 18 element cask. These shipments were overweight because of the heavy fuel shipping cask. l The necessity of obtaining overweight permits caused considerable delay in shipping all the fuel from the site. The last fuel shipment was made on February 17,1977 and was received in Idaho on February 26,1977. One nonfuel shipment was made in a fuel shipping cask to dispose of a control rod guide tube and reflectors removed from the reactor vessel. In addition to the normal fuel shipments,27 fuel shipments were made in the single element Hallam fuel shipping cask; 25 contained fuel elements, one a control rod and one a core reflector. These fuel ship-ments were made in support of the Peach Bottom Post Irradiation Examination Program con-ducted by General Atomic Company. Following shipment of all fuel, the spent fuel pool was drained and the water processed through the liquid waste system prior to release. 6.2. PRIMARY SYSTEM LAY UP 6.2.1. Post.Defueling Temperature Monitoring Following defueling of the reactor, a temperature monitoring test program was conducted to ensure that heat generation within the reactor vessel would not be excessive. The test was conducted in accordance with the procedure presented in Appendix B of the decommissioning plan. All forced and convection cooling was terminated and the reactor vessel was allowed to heat up from activation product decay. The test revealed negligible decay heat levels within the vessel, resulting in no significant rise above ambient temperatures, It was, therefore, concluded i that there is sufficient dissipation of the activation product decay heat to allow safe lay-up of the reactor vessel under an atmospheric environment. 6.2.2. Fission Product Trapping System Degassing Subsequent to defueling of the reactor, the helium purification system delay beds were degas-sed to desorb all gaseous activity. The helium purification system, or external fission product 25

trapping system, consisted of a series of water and brine-cooled charcoal traps. The purpose of degassing the delay beds was to establish a controlled release of all removable gaseous activity from the site. The Charcoal was heated to an average of 1100F, well above normal operating temperatures. The delay bed effluent was collected in a holdup tank, sampled, and then released. At the conclusion of the heating and purging operations, all helium was discharged. These releases were made under controlled conditions and in accordance with the Technical Specifications. At the conclusion of the heating and purging operation, all helium systems were purged with nitrogen. The degassing and purging of the helium systems were completed on July 24,1975. The removable radioactive gases released from the purification system contained 3.5 curies of 85 r and 0.25 curies of tritium. K Before the removal of the fission product trapping system, senior radiochemists from General Atomic reviewed the results of the degassing operations. They concluded that virtually all re-leasable gaseous activity had been effectively purged from the beds and no additional release would be expected to occur during physical removal of the beds. The low radiation and airborne activity levels experienced during the actual removal of the beds substantiated their conclu-sions. 6.3. END OF LIFE SAMPLING PROGRAM j The End of Life sampling program was performed by Catalytic for General Atomic Company following final shutdown of the facility and before decommissioning work began. The objective of the sampling program was to obtain specimens of various components of the high tempera-ture helium circulation system for subsequent laboratory examination and HTGR design methods verification. The planning and performance of this work was performed by Catalytic and NUS Corporation through the SUNTAC Nuclear Corporation joint venture. The steam generator tubing samples were obtained by cutting windows through the side of the No.1 steam generator, then removing the tubing with grinders. When all samples were removed, the steam generator wall section was replaced and welded into place. Samples from the primary system piping were obtained by trepanning. The technique employed a hole saw to cut sections from the piping. The saw blades, together with the trepan sample they contained, were packaged for shipment to General Atomic. The trepanning holes were covered with welded plates. Piping on the inlet to the steam generator was opened to obtain trepan specimens from the concentric piping; these large openings were also covered with welded plates. A total of 148 samples of piping, steam generator tubing and internals were obtained, packaged, and shipped to General Atomic Company for subsequent analysis. 26 l l l l

Upon completion of the End-of Life sampling program, the Loop 1 steam generator cavity was cleaned of debris and decontaminated. Removable sunace contamination levels were reduced to less than 1000 dpm/100 cm2, Phase lli of decommissioning had started and openings to the helium system had ben made before the End of Life sampling program was completed. Because of this overlap, the helium system integrity was not tested until all cuts into the helium system had been made and re-sealed. The helium system integrity test is described in Section 6.5.11. 6.4. DISPOSITION OF BY. PRODUCT MATERIAL Most byproduct material outside of the primary system was removed from the Peach Bottom Unit 1 facility. The exceptions are byproduct material in some primary system sample piping, material in the annulus of the spent fuel pool, and material in embedded pipes within the exclusion area and liquid radwaste area. Contaminated water, including liquid waste generated during the defueling and decommission-ing operations, was processed in the radwaste facility. Contaminated oil and water from the tri-tium holding tank were handled separately. Oil from contaminated systems was drained to the contaminated oil tank, analyzed for radioactivity, and then solidified and shipped to a licensed burial ground. This oil contained 420 pCi of tritium and less than 0.1 Ci of byproduct material. The water in the tritium hold up tank contained 260 pCilmi. This is a relatively high concentra-tion of tritium and the water was solidified using special handling and solidifiction procedures. The total quantity of tritium solidified was approximately 71 curies. During the defueling and decommissioning, 195,000 gallons of water were processed through the liquid radioactive waste system and discharged in accordance with the plant Technical Specifications. The water contained 0.014 curies of byproduct material and 0.30 curies of tritium. Percentages of itotopes of byproduct material discharged are shown in Table 6-1. TABLE 61 RADIONUCLtDES IN BYPRODUCT MATERIAL DISCHARGED FROM PEACH BOTTOM UNIT 1 Radionuclide Percent 134Cs 22 137Cs 54 90 r 3 S 59pe 1 60 o 4 C 65 n 8 Z 14C 8 27

Approximately 70 gallons of contaminated water and oil were transported to Unit 2 for process-log af ter the radwaste system was decommissioned. Radioactive ralid wastes were packaged in appropriate radioactive waste containers and ship-ped to a licensed burial facility for ultimate disposal. Thirteen shipments were made which had a total volume of 14.000 f t3 and contained 380 curies of radioactive material. Of the 380 curies shipped,300 curies were from six graphite reflector elements and a control rod guide tube, 71 curies were solidified tritiated water, and the remaining 9 curies were from all of the other material. General Atomic Company conducted an activation analysis study based on the volume weighted thermal neutron flux. The study indicated that approximately 1.4 megacuries of activity re-mained within the Peach Bottom Unit 1 facility af ter decommissioning was completed. Most of this activity is from activated components in the reactor vessel internals. The principle nuclide is 55 e in the steel pressure vessel. The remaining major sources of activity are the stellite l F 1 60 o) and surface contamination C springs on the control rod drive ccoplings (0.1 megacuries of 137 s and 134Cs). This activity will be reduced sub-C within the primtry system (1.05 curies of stantially by radioactive decay while the plant is in mothball status A composite decay curve for the activated components and fission product surface contamination was prepared to indi-cate the amount of activity (and principal decay mode) which will be presented through the year 2075. This composite decay curve is shown in Figure 61, The activity remaining in the annulus of the fuel pool is estimated to be 100 microcuries (prin-137 s is contained in embedded pipes within the cipally 137Cs). Another 15 microcuries of C radwaste basement and exclusion area. 6.5. CONTAINMENT VESSEL 6.5.1 Refueling Floor Area The decommissioning plan required that the refueling floor would be accessible for periodic inspections. Under the plan, the refueling floor and elevations above were required to meet acceptable surface contamination levels for decommissioned facilities. Another requirement was that radiation leveis in the accessible area be less than 1 mRlhr, whole body. I The refueling floor contained a large quantity of equipment and material that had been used for refueling operations and the General Atomic End-of Life sampling program. Except for the re-j fueling equipment, this material was removed during decommissioning. Removable equipment and material were monitored for radioactive contamination and segregated for disposal as contaminated or noncontaminated waste. Af ter waste removal and completion of other decommissioning activities in the area, the refuel-ing floor was surveyed for fixed and removable radioactive contamination. Many small spots of fixed contamination were detected on the refueling floor as well as on equipment and floors on 5 elevations above the refueling floor. Some of the contamination levels were as high as 10 dpm/ 100 cm. All accessible spots were successfully decontaminated or removed. 2 28 . ~

The refueling floor is now accessible only through the locked gate and manway (Figure 6 2) on the 176 ft elevation. All stairwells leading down from the refueling floor were barricaded to pre-vent entry to the controlled areas of the exclusion area. The southwest stairwell has'a locked barricade to allow access, when necessary, to the controlled areas of the containment building. Barricades for the refueling floor are shown in Figure 6 3. Radiation levels on the refueling floor at the end of decommis ning were between 0.01 mR/hr to 0.9 mR/hr.The highest radiation level is above the isolation valves in the vicinity of Survey Point Number 22. 6.5.2. Reactor Vessel Access Access to the reactor vessel was prevented by installing blind flanges on the refueling ports. The 3 f t thick concrete missile beams, shield plugs and shield blocks were replaced over the reactor vessel for permanent shielding and access control. (See Figure 6 4.) Af ter the missile beams were replaced, the electrical control wire to the reactor service crane motor was cut and the lif ting cables for the shields were disposed of. 6.5.3. Refueling Equipment At the completion of decommissioning, most of the refueling equipment was internally con. 4 and 106 dpm/100 taminated with radioactive material. Contamination levels were between 10 cm ; external radiation levels on these items were all less than 1 mRlbr. This equipment was 2 decommissioned as described in the following paragraphs. 6.5.3.1. Canning and Charging Machines. The isolation valve at the top of the canning machine was closed and all penetrations to the machine cavity were sealed. As shown in Figure 6-5, the charging machine was positioned over the canning machine and the valve at the bottom of the charging machine was closed. The gas lock sleeve was lowered and the isolation valve handle was removed. 6.5.3.2. Transfer Machine. The transfer machine was placed in its normal storage location adja-cent to the northwest stairwell. A steel plate was welded to the bottom flange of this machine, and the vacuum line and purge pipe connections were all sealed. 6.5.3.3. Transfer Cask. The valve at the bottom of the cask shown in Figure 6-4 was closed and a steel plate was welded to the bottom flange. All vacuum, air, and helium line connections were also sealed. The machine was placed in its normal storage position on the west side of the re-actor head on the refuel floor and secured by welding the latch and welding a support to a build-ing column. 6.5.3.4. Vlowing Device. The viewing device was secured to the fuel handling equipment storage port shown in Figure 6-6. 29

ACTIVITY REMAINING'IN PRIMARY SYSTEM BY PRINCIPAL DECAY' MODE i l 106 \\ \\ \\, 55pe, 63Ni (Low Energy Beta and Electron Capture Nuclides) 105_ \\ \\. 104 - Ge 103 - cc j g 60Co (Principal Gamma Emitter) N E 102 h ~~~

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G.S.3.5. Isolation Valves. The three reactor isolation valves ware secured over the fuel handling equipment storage ports. As shown in Figure 6 6, steel plates were welded on top of each valve. All purge connections were sealed with blind flanges. G.5.3.6. Fuel Handling Purge System. The exhaust filters and oil filter were removed from the fuel handling purge system. 6.5.3.7. Jin Crane Electrical service to the Jib crane was cut and the crane was retired in place. 6.5.4. Containment Sump Area The containment sump area (elevation 90 f t) is accessible for periodic inspections. Liquids in the contaminated oil tank, tritium holding tank, and the sump were removed and disposed of as described in Section 6.12 The tritium holding tank and associated piping were removed and disposed of as radioactive waste. Before work on the sump began, a radiation survey of the sump area indicated less than 0.5 mRlhr. The contamination survey of the sump detected fixed contamination levels of 40,000 dpm/100 cm2 and removable contamination of 3000 dpm/100 cm2, The pipes and drains connected to the sump were cut and sealed. The sump pumps and associa-ted piping were removed as radioactive waste. The sump pit was then decontaminated, after which fixed contamination was found to be less than 5000 dpm/100 cm2 and removable contam-ination less i. n 500 dpm/100 cm2, The sump pit was then covered with metal grating, as shown in Figure 6-7, to allow visual inspec. tion of the pit, Any ground water seepage or other water laakage into the containment vessel will eventually collect in the sump pit and be detected during the regular inspections. Access areas in the sump pit room and the access path to the room were decontaminated to the accep-table surface contamination levels for decommissioned facilities. At completion of work, radia-tion levels within the access path were less than 0.01 mR/hr. 6.5.5. Reactor Vesselinternals Both the control and emergency shutdown systems were left intact and locked in the fully inserted position. The oil was drained from the hydraulically operated rods and the gas was vented from the accumulator. The electrical power was disconnected from the electrically driven rods and the batteries were removed. Residual oil was removed from the contro :M drive exterior surfaces and the sub-pile room surfaces. 4 6.5.6. Main Coolant System Decommissioning activities involving the main coolant system can best be described by con-sidering the helium system as five subsystems: the main coolant system, fission product delay 36 -. ~.

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system, purified helium handling system, nonpurified helium handling systems, and the chemi-cLI cleanup system. The main coolant system was opened in Primary Loop No.1 during the End of Life sampling program. During this program,148 samples of piping and steam generator tubing were removed for metallurgical and fission product plateout analysis. Upon completion of the sampling pro-gram, the openings in the main coolant system were covered and seal welded. The main helium compressor shaf ts and penetrations were disconnected and sealed during de-commissioning to maintain system integrity. At that time, radiation surveys at the shaft open-ings indicated radiation levels of less than 5 mR/hr. The seal oil and lobe oil were drained from the helium compressor oil systems and the associated olf filter cartridges were removed. The main coolant bypass filters and their associated dust collectors were emoved entirely to reduce the inventory of byproduct material left on site. The maximum radiation level on these filters was a contact reading of 400 mR/hr, showing a hot spot on the Loop No. 2 dust collector. The general radiation level on the bypass filters and dust collector was approximately 45 mR/hr on contact. The dust collection pipe vent filter was also removed. All of these filters were packaged and shipped to the burial facility as radioactive waste. The piping, which was cut to remove the filters, was seal welded closed af ter the work was completed. 6.5.7. External Fission Product Trapping System Fission products were removed from the primary coolant helium by an external charcoaltrapping and delay system. A purge stream of helium trom the reactor was passed through the delay sys-tem to trap halogen and noble gas fission products. the decommissioning f..an required the removal of the water cooled delay beds, low temperature delay beds, and dust collectors from the system. The system components were located in shielded cavities and were inaccessible for radiation surveys. The components were assumed, however, to contain a large inventory of fission pro-ducts. Radiation exposure to personnel was, therefore, a major concern in planning the removal of these components. Since the vessels were too heavy to be removed manually (between 1% and 7 tons) and radiation exposure had to be minimized, a monorail was erected above both sys-tems. The monorail and electric holsts with long control cables provided a good method for rigging out the components while minimizing personnel exposure (See Figure 6 8). The plan for disposal of the large components of the system was to use each vessel as its own shipping container. All penetrations would be seal-welded closed and inspected. Small compo-nents and vessels with high radiation levels would require separate shipping containers. 6.5.7.1. Water Cooled Delay Beds. The work steps required to remove a water cooled delay bed were:

  • Lif t the access plates from over the vessel cavities.
  • Transfer the magnetite out of the vessel cavity.

38

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  • Cut the inlet and outlet pipes and cap them on each side of the cut.
  • Lif t the vessel (using the monorail) out of its cavity.
  • Transfer the vessel to the equipment door for shipment preparation.

Two methods were considered for the removal of the magnetite shielding above the vessels. The first method was to shovel it into containers that were raised and lowered by the electric hoist on the monorail. This r.iethod could be used when there was adequate physical space for a man to work in the cavity and when radiation levels were minimal. The second methods used an electromagnet attached to an electric holst. This procedure eliminated the problem of physical confinement and radiation exposure to the workers. Using the electromagnet, however, was somewhat slower than digging ble magnetite with shovels. The first component to be removed was the water separator. After this vessel was removed, radiation surveys could be made of the remaining vessels from the bottom of the water cooled delay bed cavity. The radiation surveys determined that of all the components to be removed from the water-cooled delay bed system, only the purge condensibles trap (A 301) presented major radiological hazards. The dose rate on contact with A 301 was 30 R/hr. All other com-ponents had contact dose rates of 150 mR/hr or less. After lifting the access plate above the purge condensibles trap, the upper layers of magnetite were removed by shoveling. This was continued until the radiation level in the vicinity of the worker's feet reached approximately 100 mR/hr. Only a few inches of magnetite made the dif-ference between less than 5 and 100 mRlhr. The remainder of the magnetite was removed using the electromagnet as described above. Af ter the magnetite was removed and a detailed radiation survey conducted, the disassembly operation was planned on a step by step basis. The top of the vessel was covered with lead sheets to shield personnel during cutting of the inlet and outlet piping. Pipe cuts were made with reciprocating saws to minimize airborne radioactivity and the saw was secured to the pipe to permit the operator to stand back while cutting was in progress. A special shipping box was designed for shipping the vessel to the burial site. The box was large enough to allow one foot of concrete to surround the vessel for shielding during shipment. Lead sheeting was used as a liner for the box to minimize exposure to personnel during trans-port and while filling the box with concrete. The vessel was removed, shielded and then transported to the equipment door using the mono-rail. The vessel was then set on a dolly and pulled through the equipment door where it was lifted with the fuel cask hoist and placed in itu shipping box. (See Figure 6 9.) The loaded box was then moved with a fork lift to the truck access way; here, a ready mix concrete truck was waiting to fill the box with concrete. Af ter filling, the highest contact reading on the exterior of the box was 125 mR/hr. Shipment was subsequently made in accordance with the DOT regula-tions. The removal of the remaining components in the water cooled delay bed system (first water-cooled delay bed, second water-cooled delay bed, and aerosol filter) did not present any special 40 1 l

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radiological hazards. The plan called for continuous monitoring during removal of the magnetite until each vessel was uncovered. Detailed radiation surveys of the vessels indicated that the ~ maximum dose rate was 150 mR/hr on contact. Each of these vessels was sealed and used as its own shipping container. 6.5.7.2. Low Temperature Delay Beds. A radiation and contaminat!on survey of the low-ternperature delay bed (LTDB) system indicated no radiation levels above 2 mRlhr and no detec-table surface contamination insulation was removed from the components of this system, sur. veyed for contamination, and disposed cf as nonradioactive waste. Pad eyes were installed into the ceiling of the LTDB cavity for rigging to allow the vessels to be lowered and laid on their sides prior to raising them with the monorail. The aftercoolers were removed first because they interfered with the removal of the low temperature delay beds. The five beds and aftercoolers were removed and moved to the equipment door on the monorail. The openings were seal welded closed and the vessels were shipped to the burial site as radioactive waste. A steel plate was installed over the LTOB access opening at the 116 ft. elevation to close the cavity. (See Figure 610.) The dust removal filters following the water cooled delay beds (WCDBs) and LTOBs in the fis-sion product system were removed and placed in shipping containers for disposal as radioactive waste. Maximum radiation levels on these filters were 10 mRlhr. 6.5.8. Purified Helium Handling System Under the decommissicning plan, the components of the purified helium handling system that required removal were the charcoal canisters from the oil removal filter. the liquid nitrogen traps, and the filter cartridges from the purified helium compressors. Radiation and contamination surveys of these components indicated less than 1 mR/hr and no external surface contamination. The liquid nitrogen traps were removed, seal welded, and ship-ped as their own containers. The remaining components were removed and placed in radwaste shipping containers. All openings were sealed after removal of the components. 6.5.9. Nonpurified Helium Handling System The components of the nonpurified helium handling system which required removal under the decommissioning plan were the steam generator plateout adsorber and the oil adsorber. The plateout adsorber was located behind a solid block wall in the Loop 1 cavity. No radiological sur-vey could be performed until the wall was opened. A small hole was punched through the wall to permit a long survey rMter probe to be inserted into the cavity for a radiation survey. The sur. vey indicated dose rates of less than 5 mR/hr on the surface of the vessel. An opening was made in the wall and the vessel was cut out and placed in a shipping container. l 42

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The oil adsorber had been used to remove oil from the moisture delcctor vacuum pump and fuel handling purge vacuum pump exhausts. The filter cartridge was removed and packaged as radio-active waste. Radiation levels were less than 1 mR/hr on contact. All openings to this part of the helium system were sealed af ter the components were removed. l 6.5.10. Helium Chemical Cleanup System Helium purges from the baffle area tielow the tube sheet of the steam generators passed through the chemical cleanup system to remove moisture and chemical impurities. The first in-line piece of equipment, the steam generator purge plateout trap, was isolated behind a solid concrete block wall. No radiological data were available and it was believed that this equipment could conceivably have radiation levels as high as 5 Rlhr. As a precaution, a small hole was punched through the block wall to obtain radiation data. The radiation survey indicated a dose rate of 15 mRlhr at 1 ft. An opening was then jackhammered in the wall large enough to permit removal of the vessel, and a detailed radiation survey was made that showed a 150 mRlbr con-tact does rate near the top of the vessel. This reading indicated that the connecting piping would l [ be internally contaminated. The inlet and outlet pipes were cut with reciprocating saws to mini-l mize airborne radioactivity. Af ter the vessel was removed, the inlet and outlet pipes were l capped and seal welded. The vessel was capped, welded closed and shipped to the licensed burial facility as its own container. 6.5.11. Helium System Integrity A preliminary pressure test was performed on the helium system af ter all components were removed and the piping was welded closed. During the preliminary test, a number of leaks were l identified and welded. Other leaks remained, however, and at least part of the leakage was attri-buted to cut tubes that had not been plugged in the No.1 steam generator. To limit this leakage, all connections to the head of the steam generator were cut and cappei This procedure com-pletely isolated No.1 steam generator from the secondary system. Af ter these connections were sealed, a pressure test of the primary system was conducted to determine whether fission products would be released into containment by atmospheric pres-sure changes on the system. To perform this test, air samples and surface contamination sam-pies were collected in representative areas including the loop cavity for No.1 steam generator. 2 n all areas and did.not change during i Contamination levels were all less than 100 dpm/100 cm or af ter the testing. Air samples in all areas both before and af ter the test showed readings of less than 1012 pCilmi. However, during the testing, the airborne radioactivity in the steam gen-erator loop cavity increased to 3 x 1012 Ci/ml. This concentration is less than one percent of the applicable permiasible concentration for effluents to uncontrolled areas for 37 s, the prin-1 C cipal radionuclide in the system. Testing duration was 48 hours and the sample duration was about 24 hours during the second half of the test. Primary system test pressure was approxi. mately 10 psi. 44

i 6.5.12. Containment Penetrations All piping connections which penttrated the containment were cut and capped ( ee Figure 611) on the outside of the' containment with one exception; the emergency cooling water lines were cut and capped inside containment. A portion of these lines had to be cut and removed inside containment to permit removal of the liquid nitrogen traps. Rather than cut the lines again out-side, they were capped inside. Each penetration was surveyed for surface contamination.The only penetrations with removable contamination greater than 100 dpm/100 cm were the shield cooling water, the cooling water to 2 the fuel canning machine, and the radioactive waste line from the containment sump pumps. The external surfaces of these pipes were decontaminated before capping and welding. A six inch pressure equalization line, equipped with a replaceable absolute filter, was installed on the containment equipment door to permit equalization of the pressure differential between the inside and outside of the containment. The filter is shown in Figure 612; installation in Figure 6-13. 6.5.13. Containment Vessel Access Doors Although the radiation levels in the containment vessel were very low, access control measures were installed in various locations to prevent inadvertent entry to spacr;s beyond the specified inspection areas. The outer door of the emergency air lock and the equipment door on the groun'd floor were welded closed. The inner doors of the personnel air lock on the ground floor and the access lock on the refueling floor were welded in the open position. The interlock mechanisms in these two air locks were disabled, and the outer door on each air lock was secured with a padlock. All stairwells and ladders leading down from refueling floor were barricaded as shown in Fig. ure 6 3. The southwest stairwell has a locked barricade to allow access, when necessary, to the controlled areas of the containment building between the refueling floor and the ground floor. I Three barricades were installed on the ground floor to prevent accidental entry to controlled areas as shown in Figure 614. An expanded metal wall was placed in the vicinity of the refrig-erant-brine exchanger to prevent access to the north. A gate of expanded metal was bolted over the entrance to the control rod hydraulic erpipment area. The lhlrd barricade is an expanded metal wall, 8 f t high, between the containment wall and the concrete wall near the south stanway. Barricades were installed in the containment upper and lower basements to restrict access to only the path leading to the inspection area at the sump. These barricades are indicated in Figures 6-15 and 616. The doors leading to valve alley, the chemical clean-up system dehydrator-oxidizer area, the subpile room, and the contaminated oil tank room were welded closed. An <nspection viewing port was cut into each of these doors ano covered with expanded metal. An expanded metal gate was installed over the openin;j in the wall where the steam generator purge plateout trap was removed. 45 l t

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CONTAINMENT PRESSURE EOUAll2ATION REPLACEABLE ABSOLUTE FILTER CAMBRIDGE MODEL 10E 1 6" C.S. PIPE l F ~ ,l n 2z I/4" SEAL WELD a > US 2 1 TACK WELD HEAD O z = 0F BOLTS .W / J l l.., I r-k a = r...L........ '_ [ M:M i L EQUIPMENT DOOR / / I/16" NEOPRENE ' / \\ GASKET / \\ 4 - 1/2" 4 x I" LONG 1/4" C.S. FL ANGE ' \\ N C. BOLT S EQUAL LY S PACED ON 9 t/2"DIA. BOLT CIRCLE r N a w e a 8 8 FIGURE 613. Containment Pressure Equalization 48

The 3 to containment and each barricade are posted with a sign which bears the stan-dard a symbol and the words " Caution Radioactive Material Beyond This Point" 6.5,14. Flammable Materials in Containment All flammable materials other than electrical cable and the solid graphite components within the reactor vessel were removed from the containment vessel 6.5.15. Containment Exhaust System The filters from the containment exhaust plenum and post incident recirculation system were removed and disposed of as radioactive waste. The radiation levels of these filters were less than 0.5 mR/hr. 6,6. FUEL POOL BUILDING The decommissioning plan required specific tasks to be performed in the fuel pool building as follows:

  • Remove allliquids from the pool.
  • Remove or decontaminate all tools and equipment from the pool and sump.
  • Remove spent fuel cooling system piping.
  • Remove the spent fuel pit elevator and elevator tube spool piece.
  • Decontaminate the fuel pool walls and floor.
  • Decontaminate or remove the crane, grapple and hoist.

Provide access control measures. 6.6.1. Fuel Pool Af ter all the spent fuel was received at the Idaho facility, the pool was drained and the water was processed through the liquid waste system. Approximately three inches of water was left on the bottom of the pool to decrease airborne radioactivity during the work to follow, Initial radia-tion and contamination surveys in the pool indicated radiation levels of less than 1.0 mR/hr, 2 5 dpm/100 cm, and a removable contamination maximum fixed contamination levels of 4 x 10 level of 1 x 10 dpm/100 cm. These contamination levels were found mostly on the spent fuel 6 2 rack. Air samples taken during work procedures showed readings of 5 x 10-10 pCl/ml and indi-cated the presence of long lived icotopes. Although these air concentrations did not require l 49 ~.

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respiratory protection equipment. the workers wore full-f ace respirators due to the potential airborne contamination from the high removable contamination levels. Initial access to the pool was made using the 15 ton cask hoist and a lift basket. A 19-ft high scaffold was set up as a work platform for the crane at the elevation of the top of the spent fuel rack. A 40-f t wood extension ladder was also placed into the pool to enable workers to climb in and out. To prevent the spread of contamination, workers wore double sets of boots and gloves and removed the outer set prior to exiting from the pool onto the walkway. 6.6.1.1. Spent Fuel Tools. Wipe surveys of spent fuel tools and handling devices indicated re-dpm/100 cm. The contamination was in the form of loose 2 5 movable contamination levels of 10 dust and dirt and was hosed off with water before removing the items from the pool. Wipe sur. 2 residual con-veys following this decontamination typically indicated about 7500 dpm/100 cm tamination. These tools were then transferred by the 40-ton traveling hoist to a laydown area next to the cask wash area. A temporary containment was constructed of poly sheeting to pre-vent the spread of contamination during cutting of the tools. The spent fuel basket support and the cask support were cut with saws into sections small enough to fit into the radwaste boxes. However, the base plate for the cask support was 3 in. thick stainless steel. which made saw cutting infeasible. A special shipping container was therefore built for this plate for shipment to the burial site. the fuel cell basket was sealed up and shipped as an integral unit. 6.6.1.2. Spent Fuel Rack. The spent fuel rack was constructed of upper and lower horizontal grids with supporting columns between the grids. Each square in the grid had % in, square rod guide tubes between the grids, which were f astened with machine screws and acorn nuts. The guide tubes were removed first to create work space to cut the upper horizontal grid. The top suppod nut was chiseled off to permit lif ting tubes from a bottom support notch. The upper grid and supporting columns were cut with reciprocating saws to minimize airborne radioactivi-ty. All material removed from the fuel rack was cut to packaging dimensions and placed directly into radioactive waste containers. The lower grid was lef t in place and used as a work platform to remove the necessary sections of the fuel chute and spent fuel elevator. After these components were removed, the lower grid was cut and removed. Full-face respirators and protective clothing were used by all workers while cutting the fuel rack, spent fuel chute, and spent fuel elevator. Continuous air sampling was conducted during this work. All air samples collected indicated less than maximum permissible concentration of airborne radioactivity. 6.6.1.3. Decontamination of Fuel Pool. Air samples taken af ter the remaining trash, metal scrap and tools were removed from the bottom of the pool were less than MPC. These readings indicated that the major souices of contamination had been removed, eliminating the need for respirator use during work in the pool. 53 .. ~. - -

The spent fuel pool wa!I decontamination effort used a mobile scaffold platform which could be lowered by removing 5-f t. sections. This enabled the decontamination crew to reach all areas of the spent f uel pool walls. The use of a commercial liquid cleaner and rags provided the best decontamination and ens-n etic cleaning. A single application of cleaner usually reduced the activity on the pool walls from 30,000 to less than 300 dpm/100 cm. Some of the pool surfaces exhibited fissures in thh 2 i paint which required more than one attempt at decontamination. At completion of work, the 2 highest removable contamination levels remaining in these areas was 1300 dpm/100 cm. l The pool was cleaned of the residual water by vacuuming the water into 55 gallon drums. The j water was pumped from this drum to a drum at the top of the pool. The water was then trans-l I ferred to the liquid waste system and the pool floor decontaminated. At this time, the remaining removable contamination level was 1000 dpm/100 cm2, i G.ti.2. Spent Fuel Chute and Elevator After the spent fuel chute elevator cables were disconnected, the elevator and cables were lowered to the floor of the pool and transferred to a radwaste box. The spent fuel elevator is shown in Figure 617. f The spent fuel chute was constructed with a removable spool piece flanged at both ends. This section was located inside the spent fuel chute enclosure with access from the ground floor near the equipment hatch. After the lower flange was unbolted, a contamination survey was per-2 were detected, although general radiation 8 dpml100 cm formed. Contamination levels of 10 levels in the eclosure were only 1 mR!hr. The ends of the chute and spool piece were double l poly bagged to maintain contamination control. 1 When the spool piece had been disconnected, the bottom section of the chute was lowered into the spent fuel pool through the pool wall penetration. The chute was cut into small sections and placed in a radwaste box. The upper section of the shute was also cut and removed to radioac-tive waste. The flanges at the penetrations to the pool and containment wall were decontami-nated and a 3/8 in. thick blanking flange was welded to each chute opening. The chute enclosure was cleaned and decontaminated to acceptable surface contamination levels. Af ter the final radiation survey was completed, an expanded metal barricade was bceted over the access door to the enclosure. 6.6.3. Traveling Bridge Holst. The traveling bridge hoist (grapple crane) was moved to the south end of the pool. The 8 in. guide pipe,4 in. hoist pipe, and drum cable were cut off under the base of the crane. Water and pressure from the accumulator on the hydraulic indexing system were removed. Because the 2 removable contamination level on the tube was 25,000 dpm/100 cm, the pipes and cable were disposed of as radioactive waste. The hoist was then moved to the north end of the pool, de-54

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activated and as final step, decontaminated to acceptable surface contamination levels for decommissioned racilities. 6.6.4. Fuel Pool Building Sump The head tank and support steel were removed from above the sump pit and the sump pump ar<d piping were removed from the pit. A direct reading of the pump sludge was 40,000 dpm. A bosun's chair was rigged over the sump pit to lower an individual into the pit to determine the water level, assist in pumping the remaining water from the pit, and perform wipe surveys of the pit walls. Before entering the pit, a 200 mR self reading dosimeter was lowered into the area; after one hour the dosimeter read zero mR. The water was found to be 19 inches deep by means of a dip stick. A worker dressed in full protective clothing and an air line respirator was lowered into the sump pit in the bosun's chair to guide the pump suction nozzle to pump as much water as possible from the pit. While in the pit, he also measured the residual water depth and took wipes of the pit walls; residual water depth in the pit was found to be 8 inches, and the maxi-mum removable contamination on the walls was 500 dpm/100 cm2. After the area around the top of the sump pit was decontaminated to less than 250 dpm/100 cm2 of removable contamination, grating was welded over the top of the sump. 6.6.5. Fuel Pool Piping The spent fuel cooling water pipes and the slulce pipe were removed, cut to packaged dimen-sions and disposed of as radioactive waste. Contamination levels on this piping were 60,000 dpm/100 cm. Embedded portions of the supply and return lines were cut and capped at the 2 exterior wall of the building. 6.6.6. Ventilation Systems The prefilter and absolute filter were removed from the cleanup ventilation system and disposed of as radioactive waste..A contamination survey of the pipe at the exhaust of the filter indicated less than 250 dpm/100 cm2 removable contamination and less than 5000 dpm/100 cm2 fixed contamination. The supply and exhaust ducts of the main ventilation system to the fuel building were cut and capped at the exterior of the building. A contamination survey of these pipes indicated less than 100 dpm/100 cm2 removable and less than 5000 dpm/100 cm2 fixed contamination. When the isolation valve was removed from the cleanup line to cap the pipe, a contamination survey indicated 40,000 dpm/100 cm2 removable contamination on the exhaust side of the valve 2 on the inlet side of the valve. An examination of this system revealed and 1000 dpm/100 cm that a purge line for the spent fuel cask had been installed downstream of the isolation valve. The purge line <as cut and removed as radioactive waste. The cleanup line duct was then cut of f in 6 f t sections and contamination surveys were performed. Approximately 30 f t. of duct was 56

removed before contamination was found to meet acceptable surface contamination levels for decommissioned facilities. 6.6.7. Traveling Holsts Contamination surveys of the 15-ton cask service hoist and the 40 ton hoist indicated con. 2 on the hook and block assemblies of both cranes. The tamination levels of 105 dpm/100 cm blocks and cables were lowered directly into radioactive waste containers for disposal. The drums, drum housings, and associated equipment for the hoists were decontaminated to accep. table surface contamination levels. the electrical power to each hoist was disconnected and the hoists were placed in lay-up in the spent f uel building. 6.6.8. Access Control The top of the fuel pool wcs covered with welded rectangular steel grating to permit visual in. spections of the pool. Support beams were welded to the steel li surrounding the pool and the grating was welded to the support beam. (See Figures 6-18 and 619). The access door from the new fuel storage area was welded closed. The cask monorail doors were sealed by welding a steel bar to the inside of the doors. Access to the pool area for in-spections is available through the locked door on the ground floor (elevation 116 ft). 6.6.9. Decontamination All accessible areas for inspection were decontaminated. Fixed contamination levels were de-contaminated to less than 5000 dpm/100 cm2. Several areas of the concrete walkway required removal of the top surface of the concrete to remove stubborn spots of fixed contamination. Removable contamination levels were reduced to less than 1000 dpm/100 cm2. At conclusion of this work, all radiation levels within the building were less than 0.2 mR/hr. 6.6.10. Fuel Pool Cooling System The fuel poo! cooling system components were located in the liquid waste basement. The de-commissioning plan required all components and piping of this system outside of the exclusion area to be removed. 2 Contamination surveys of the heat exchanger indicated the presence of 50,000 dpm/100 cm i fixed contamination and 2000 dpm/100 cm2 removable contamination. The contamination levels 6 2 fixed, and 10 dpm/100 cm2 removable. Whole body ) 4 of the fuel pit filters were 10 dpm/100 cm radiation from these components was less than 1 mRlhr. l 1 57

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The system piping was cut with saws to reduce airborne radioactivity. In areas where saw cut-ting was impractical, the pipes were cut using an oxyacetylene torch. Respiratory protective equipment was used by workers while torch cutting the contaminated pipes. The spent fuel heat exchanger was drained, disconnected from the piping and resealed. The exterior surfaces were decontaminated and the exchanger was removed.through the access hatch;it was used as its own shipping container for transport to the burial facility. The filters from the filter tanks were removed and disposed of as radioactive waste. The filter tanks were then disconnected from the system piping and removed to radioactive v.aste containers. The two spent fuel cooling water pumps were disconnected and removed to Ursit 2-3 for salvage as contaminated pumps. The spent fuel cooling water booster pump was removed and packaged as radioactive waste. Piping removed from the system was cut to packaging dimensions and disposed of as radioactive waste. Decontamination of the area was postponed until the liquid waste system was removed. The area around the system was tnen decontaminated along with the liquid waste basement. 6.7. LIQUID WASTE AREA The liquid radwaste area and liquid waste processing systems were located in the basement of the auxiliary building. The decommissioning plan specified that all radwaste system equipment, _ such as filters, demineralizer resins, pumps, etc., would be removed and shipped to a licensed burial facility. Piping in the radwaste system would either be decontaminated or removed as radioactive waste. Preliminary contamination surveys indicated that the contamination levels within the system were from 10 to 10 dpm/100 cm. This level of contamination was considered to be impractical 4 6 2 to decontaminate; therefore, the components and associated piping were removed. The entire radwaste basement was maintained as a contamination controlled area until final decontamina-tion of the facility. When decommissioning activities on the liquid radwaste area began, radia-tion levels within the basement were less than 2 mRlhr. 6.7.1. Component Removal Accessible piping in the radwaste system was cut, using saws when practical, to reduce air-borne radioactivity. As the piping was removed,it was cut to packaging dimensions for disposal. The waste receiver tanks and monitoring tanks were 4 5 f t.1.D. x 9 f t. long. Removal of the tanks as whole units would have required extensive demolition of some walls and ceiling of the waste facility, These tanks and the laundry tank contained low level radioactive sludge on their bot-toms. The method chosen for disposal was to cut the tanks into sections small enough to fit into the normal radioactive waste shipping boxes. The tops of the tanks were removed first. 60

Workers then removed the sludge from the bottom of the tanks. Af ter sludge removal, segments of the tanks were cut and removed. One waste monitoring tank was retained for use until the completion of work that generated liquid radioactive waste. These steel tanks were cut by oxyacetylene torches. Workers involved in this operation wore full face respirators with supplied breathing air. Continuous air samples were taken in the area during cutting operations. All samples indicated that the airborne radioactivity was below maxi-mum permissible concentrations. The spent resins from the domineralizers were removed, drummed, dewatered and shipped to the burial facility. The demineralizer shells were disconnected from the piping and removed to shipping containers. To enable removal of the shells, the demineralizer room wall portal was enlarged. Other contaminated components removed were the liquid waste filters and tanks, the waste re-ceiver tank pump, the sump pumps, and the liquid waste controlled discharge pump. In each case,,the component was disconnected and removed to the radwaste shipping container. The general method for removal of this equipment and piping was to cut the material to packaging dimensions, move it to the equipment hatch area, and lif t the material to the radwaste container on the ground floor above. The 40 ton traveling hoist and smaller electric hoists were used for this rigging. 6.7.2. Embedded Liquid Waste Piping The liquid waste system received effluents from more than twenty sources of liquid radioactive l waste. Most of the sources were outside the liquid radwaste area and the connecting piping was j embedded in the concrete floors and walls of the auxiliary building and liquid radwaste area. In addition, the floors of the liquid radwaste area contained a network of equipment and floor drains which drained to the sump. The labyrinth of shield walls within the area also contained many embedded pipes which connected various components of the liquid waste system. More than 100 embedded pipes required decor tamination or removal from the liquid waste system. Other items removed included embedded pipes in grouted concrete block walls and drains in the auxiliary building ground floor. The embedded pipes in poured concrete walls and long runs of piping in poured concrete floors were decontaminated. Each pipe was identified and monitored for fixed contamination before decontamination began. The floor drain traps were removed by Jackhammering away the floor area around the drain and cutting the trap from the adjoining pipe. Various methods were used to decontaminate these pipes and drains. Short straight pipes were decontaminated with flue brushes. Organic and acid soluble coatings were removed or loosened by acid leaching and flushing, using sulfuric and nitric acids. Cleaning with a 6500 psi water nozzle was the last method used for pipes still contaminated af ter other treatment. This high pressure water cleaning was a very effective technique for decontamination of the pipes. The high pressure nozzle is self propelled and will pass through any pipe over 1%" diameter without short 90 degree turns. The pipes were monitored for fixed and removable contamination af ter the final decontamination effort. 61 ~ .~

Of the 82 pipes and rains which are embedded in the walls and floors of the liquid radwaste basement,11 remain contaminated to levels exceeding those acceptable for release for unre. stricted use. These 11 contaminated embedded pipes are shown in Figure 6 20. A hazards evaluation was prepared to determine the radiological status and to recommend alternatives to removal of these pipes. The hazards evaluation showed that radiological hazards would be mini. mal even under conditions of flooding, fire or earthquake and that the quantitative amounts of radioactive material in the pipes are exempt from licensing. The hazards evaluation is contained in Appendix C. All embedded pipes above floor level or in the sump were sealed by capping the welding. Embedded pipes in the floo. were plugged and covered with concrete to floor level. Each con-taminated pipe has been identified with a sign reading. " Buried Radioactive Pipe" 6.7.3. Decontamination of Liquid Waste Area The radwaste f acility had large areas of fixed contamination with levels up to 10 dpm/100 cm2, 6 The walls aho had many spots of fixed contamination. Each square meter of the facility walls and floors was surveyed for fixed and removable contamination. Of the 711 square meters sur-2 veyed,122 had fixed contamination greater than 5000 dpm/100 cm. All of these areas were de-contaminated using solvents and rags or by removing the concrete with Jackhammers. After decontamination, all areas were monitored for fixed and removable radioactivity. Fixed con-tamination was less than 5000 dpm/100 cm, and removable contamination was less than 500 2 dpm/100 cm2 Although the liquid radwaste area sump was decontaminated with high pressure water, a few stubborn spots of fixed contamination remained. These spots were removed with sandpaper to acceptable surf ace contamination levels for decommissioned facilities. 6.7.4. Access Control The original plug for the equipment access opening to the radwaste basement was disposed of as radioactive waste. A 4" steel plate was welded over the equipment access opening. Rec-tangular metal grating was welded over the sump, as shown in Figure 6 21, to permit periodic visual inspections. An inspection port was cut in the access door and a hasp and padlock were l installed on the door. 6.8. EXCLUSION AREA A chain link fence was installed around the containment and spent fuel pool building, as shown in Figure 614, to establish the exclusion area specified in the decommissioning plan. Access to e the exclusion area requires authorized entry through the perimeter security fence and unlocking a posted gate in the exclusion area fence. Entry to the containment structure, the spent fuel pool building or the hquid radwaste area requires the use of an additional key to the doors of l these buildings. The controlled areas, where radiation levels are greater than 1 mR!hr, can be I 62 i

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entered only by using an additional key to open the grating over the southwest stairwell in the containment. Access to the reactor vessel requires physical restoration of the electrical supply to the containment crane to move the 3 f t thick concrete missile beams; special tooling is also needed to remove the reactor nozzle shield plugs. Different keys are provided for opening the locked gates and doors of the decommissioned fac-ility. All keys are under the control of the General Superintendent, Generation Division, of the Philadelphia Electric Company. Periodic inspection parties will be given the keys for the exclu-sion area fence, the containment, the liquid radwaste area and the spent fuel pool personnel door. These parties will not. however, be given the key to the grating over the southwest stair-well. Any entry to the controlled areas in the containment will be made with the approval of a supervisory representative of Philadelphia Electric Company. Several views of the fence are shown in Figures 6 22,6 23 and 6 24. The exclusion fence was installed by a local subcontractor. The material used was aluminum fencing,7 ft high, with 3-strand barbed wire,450 for ene foot above the fence fabric. For practical layout, some minor adjustments were made to the fence location shown in tho decommissioning plan; however, there was no departure from security or enclosure intent. Two ground floor areas were covered on top with fencing: outside the personnel door to the spent fuel pool, and over the projection I of the personnellock through containment shown earlier in Figure 6-14. l An expanded metal wall with a locked gate was placed outside the access lock on the refueling floor to keep the containment inaccessible without authorization at this point as shown earlier in figure 6-2. The liquid waste system basement entrance was secured. This was based on a recommendation in the hazards evaluation for radioactivity in embedded pipes that were left in place. This closure was accomplished by welding a % in. carbon steel plate over the basement access hatch and converting the basement door into a locked inspection door. All of the exclusion areas were posted with requisite no trespassing and radioactive material signs. In addition to installing physical locked barriers to prevent entry, the radiological safeguards include the stipulation that the area within the exclusion area fence will not be used for any pur-pose. It would be occupied, under controlled conditions, only when a periodic inspection is made or during maintenance of the decommissioned facility. 6.9. ADMINISTRATION BUILDING The administration building contained the offices, control room, laboratories, machine shop, laundry, and personnel decontamination area as shown in the plant layouts of Figures 6 25 and 6 26. All areas were required to meet acceptable surface contamination levels for decommis-sioned facilities. According to the decommissioning plan, radiation levels were required to be l less than 0.04 mR/hr. I 65

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All radioactive materials were removed from the radiochemistry laboratory, counting room, personnel decontamination area and laundry. Obsolete lab equipment and materials were dis-posed of as radioactive waste. Reusable equipment was removed to Unit 2 3 for reuse. The drains from the sinks and floors of the laobratory, decontamination area and laundry were successfully decontaminated with acid or high pressure water. Af ter decontamination, the floor, sink and decontamination shower drains were connected to the clean drain system and the laundry drain was capped. 4 dpm/100 cm2. This duct The ventilation duct from the laundry area was contaminated to 10 was removed and disposed of as radioactive waste. The ventilation system from the radiochem-istry laboratory was surveyed for fixed and removable contamination and determined to be within acceptable surface contamination levels for decommissioned facilities. After the pre. filters were removed, the system was retired in place. 1 All laboratory fixtures, sinks, and floors were decontaminated to acceptable levels. Several l spots of fixed contamination on the floor required removal of the top layer of concrete for suc-cessf ul decontamination. The machine shop, machine shop office, tool room and store room are being used as electrical maintenance facilities for Unit 2 3. Contaminated electrical motors and pumps in this area are controlled by the operating licensa of Unit 2 3 and receive radiological surveillance. These areas were not subject to radiological surveys under decommissioning. Upon completion of decommissioning, areas of the administration building were surveyed for fixed and removable contamination and whole body radiation dose rates. Contamination levels 2 fixed and less than 100 dpm/100 cm2 removable; whole body were less than 5000 dpm/100 cm radiation levels were less than 0.01 mP/hr above background. 6.10. REMAINING PORTIONS OF THE MAIN BUILDING COMPLEX 6.10.1. Turbine and Auxiliary Buildings f The oil was drained from equipment that was retired in place. Equipment in the turbine and auxillary buildings necessary to support the of fice and shop f acilities remain in service. The turbine and generator are maintained for potential use or sale elsewhere. All flammable and obsolete equipment was removed. After decommissioning work was complete, a final radiation and contamination survey was conducted in these areas. Fixed and removable contamination levels are within the acceptable i surface contamination levels for decommissioned facilities. Radiation levels are less than 0.0t mRlhr above background. 71

6.10.2. New Fuel Storage Vault The lower level of the new fuel storage area was cleaned and decontaminated. The upror level was used as a contractor weld shop during Phase lli of decommissioning. Both levels of the vault were surveyed for fixed and removable contamination from beta gamma emitters and from uranium. The surveys indicated contamination levels were below the levels allowed for decommissioned facilities. i 6.10.3. Source Storage VM All radioactive 3ources were removed from the source storage vault. Obsolete sources were wipe tested, inventoried and packaged for shipment to the licensed burial facility. Usable sources were transferred to a storage vault at Unit 2 3. A radiation and contamination survey of the vault indicated no fixed or removable contamination above acceptable surf ace contamination levels; radiation levels were less than 0.01 mR/hr. 6.10.4. Shield Cooling Water System Contamination surveys of the shielu cooling water system indicated fixed contamination levels of 104 dpm/100 cm2 on the interior of the system. This contamination level required that all components of the system exterior to containment be removed. A contamination control area was established around equipment to be removed. Piping in the system was cut with saws to reduce the spread of contamination. The heat exchangers and shield cooling tank openings were seal welded and the vessels were used as their own ship-ping containers. All other piping and equipment removed from the shield cooling water system were packaged as radioactive waste. The penetrations at containment were capped and seal-welded. System interconnection to the containment equipment cooling water system were surveyed for contamination. The containment equipment cooling water system was found to be within acceptable surf ace contamination levels. Af ter components and piping from the system were removed, the area was decontarr.inated to acceptable surf ace contamination levels. 6.10.5. Spent Fuel Cask Travel Area The entire cask travel area was decontaminated. Most of the cask wash area, including the curbs, required removal of concrete before acceptable contamination levels were observed. The drain and drain pipe were also removed. Af ter deconMmination, the resulting holes were patched with new concrete. 72

6.10.6. Ventilation System and Stack Contamination surveys were performed on each branch of the ventilation system at the intakes and at the containment penetrations. All ventilation ducts had acceptable surface contamina-tion levels except the spent fuel pit cleanup system. (See Section 6.6.6.) The prefilters and absolute filters from the ventilation system plenums were removed and dis-posed of as radioactive waste. The plenums and base of the stack were cleaned of debris after filter removal. Final contamination survey of these areas indicated no contamination above acceptable surf ace contamination levels for decommissioned f acilities. 6.10.7. Radiation Monitoring System The radiation monitors were removed from service at the end of Phase Ill of decommissioning. The only monitor remaining in service is in the lobby of the administration building. The radioactive check sources in the monitors were removed, inventoried, wipe tested, and packaged for shipment to the licensed burial facility. No leaking sources were detected during wipe testing. 6.11. INCIDENTAL TANKS AND BUILDINGS The steam generator blowdowr tank, condensate storage tanks, neutralizer tank, and settling basin were surveyed for contamination and radiation. All fixed and removable contamination levels were within accer. table surface contamination levels for decommissioned facilities. Whole body radiation levels were less than 0.01 mRlhr above background. The liquid nnrogen supply facility was removed by the vendor who owned this equipment. No other outside tanks or buildings were removed during decommissioning. 6.12. WASTE DISPOSAL AND TRANSPORT 6.12.1. Liquid Waste Liquids such as water, oils, and refrigerants were drained from plant systems and disposed of according to their contamination levels. Refrigerants and uncontaminated oils were discarded j as clean waste. Most contaminated liquids, including liquid waste generated during the defuel-ing and decommissioning operations, were processed in the radwaste facility. The exceptions, contaminated oil, water from the tritium holding tank and waste generated after decom-missioning the radwaste f acility, were handled differently. Oils from contaminated systems were drained to the contaminated oil tank, analyzed for radio-activity, and then solidified and shipped to a licensed burial ground. the activity of the oil from pCilmi for 137 s and 4 x 10 3 all of the potentially contaminated systems was less than 9 x 10-8 C 73 i l l l

Cilml for tritium. These low levels of activity did not require special handling during solidifica-tion. The oil was mixed with an absorbent in a concrete mixer until the absorbent was dry. After adding cement and water, the mixture was poured into a 55-gallon drum to harden. When solidi-fled, this substance resembled a hard mortar. A total of 235 gallons of oil containing 3800 micro-curies of tritium and less than 0.008 microcuries of byproduct material were solidified in this manner. The water in the tritium holding tank contained 260 pCi/ml, a relatively high concentration of tritium, and required special solidification procedures. To prevent creation of significant air-borne radioactivity, all liquids were transferred in a closed system. A special drum adapter was fabricated to allow regulation of flow during filling, and a vent duct was connected from the drum adapter to the plant ventilation system. A fine mesh wire tube was placed in the center of each drum and a cement absorbent mixture was added. The water was pumped from the tritium holding tank to the drum adapter and was allowed to diffuse into the mixture through the center tube. This procedure was tested using un-contaminated water after which the drum was cut away from the solidified mass; solidification was positive and no liquid reached the outer periphery of the mixture, All personnel involved in handling the trituim wore positive pressure plastic suits (bubble suits) during liquid transfer and capping of the drums, as shown in Figure 6 27. A practice training was held during testing with clean water, prior to actual solidification. Bio assay samples indicated no uptake of tritium by the personnel involved the total quantity of tritium processed was approximately 71 curies. Af ter the Unit 1 radwaste facility was decommissioned, approximately 70 gallons of contamina-ted oil and water were transferred to Peach Bottom Unit 2 for processing. 6.12.2. Uncontaminated Waste A large quantity of material that was disassembled during decommissioning was not con-taminated with radioactivity, including insulation from the exterior of piping and vessels, struc-tural steel and other substances removed to gain access around components, piping from un-contaminated systems, and other parts taken from uncontaminated areas. These materials were surveyed for fixed and removable contamination. Only those items which 2 2 fixed and less than 100 dpm/100 cm had no detectable activity (less than 500 dpm/100 cm removable) were released to scrap salvage or clean waste. The general method used for removal of clean waste was:

  • The area where work was to be performed was tested for removable contamination and decontaminated if necessary.
  • External surfaces of the items to be disassembled were surveyed for radioactive contamination.

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  • The item was removed from the system and tested for internal contamination.

The exceptions to this method were removal of insulation and small pipe (less than 1 in. diam-eter). Insulation was first surveyed for external contamination, then removed, after which the pipe (or vessel) was surveyed for contamination. If both surfaces were uncontaminated, the in-sulation was discarded as clean waste. All piping smaller than one inch in diameter was treated as contaminated waste because interior surfaces could not be surveyed for contamination. Of the approximately 20,000 ft3 of clean 3 of waste removed during decommissioning,14,000 f t3 of metal scrap was salvaged and 6,000 ft debris was disposed of as trash, 6.12.3. Solid Radioactive Waste Special handling, packaging, and transportation to a Ilcensed burial ground were required for radioactive solid waste such as ion exchange resins, filter socks, solidified liquid waste, con-taminated equipment, and contaminated trash resulting from the defueling and decommission. ing operations. Most solid radioactive waste met the criteria of 49 CFR 173 for " low specific activity" (LSA) material, The only exceptions were reflector elements, a control rod guide tube, and radioactive sources from instrumentation, lon exchange resins were dewatered and packaged in 55-gallon drums (Specification DOT 17H). Low radiation exposure rates (less than 5 mR/hr) from most of the solid radioactive waste sim-plified the handling and packaging problems. As a usual procedure, a shipping container (box or drum) was placed in the vicinity of the working area. As solid waste was removed, it was cut to fit packaging dimensions and placed directly into the container. In areas where containers could not be placed and moved out conveniently, the material was sealed or bagged in plastic (to prevent release of contamination) and taken to the nearest container. The shipping containers used were DOT 17H drums and wooden boxes built to DOT Specification 19A. The usual box sixe was 7 f t x 4 f t x 3.5 f t. Other packaging methods were used for large components. Vessels too large for the usual boxes were seal we!%d and shipped as their own containers. Contaminated components, which were too large for usual boxes and too difficult to cut, were placed in specially fabricated boxes. Examples of material handled in this matter are the spent fuel cask support plate and fuel ele-ment grid plate. I Graphite reflector elements and a control rod guide tube were packaged in the spent fuel shipping cask for transport to the burial ground. ] l A total of 490 containers of solid radioactive waste with a volt,me of 14,000 ft3 were packaged and shipped during decommissioning. The total radioactivity of the so!id waste was 380 curies. 76

6.12.4. Transport and Burial Solid radioactive waste was transported to licensed burial grounds at either Morehead, Ky. or Barnwell, S.C. Shipments were made in compliance with 49 CFR 173 in accordance with regula-tions governing solo use vehicles. Subcontractors for transportation and burial were Nuclear Engineering, Chem Nuclear Company, Home Transport Company and Tri-State Transportation Company. No abnormal occurrences resulted during or af ter these shipments. = I l i l 77 -..... _...... ~ _, _ -. _, _...... _. -_

7. HEALT.H PHYSICS AND SAFETY The provision of health physics and safety services was a coordinated effort between Philadel-phia Electric Company and Catalytic, Inc. Philadephia Electric Company provided health physics services during defueling, fuel shipment, and periods during which decommissioning activities were stopped on Phase 11. Catalytic, Inc. provided the health physics services while Phase 11 and Phase lit were in progress.

7.1. RADIATION PROTECTION TRAINING Personnel who worked in radiologically controlled areas received approximately six hours of training in radiological safety. Training included such subjects as methods for minimizing radia-tion exposure, exposure limits, biological effects of radiation, use of protective clothing, facility security training and contamination control. A quiz was given to each trainee to ascertain that the important points of radiation safety had been understood. These personnel also received respiratory protection training. This course included the proper use of full face respirators and full face pressure demand masks. The training program included the advantages and disadvantages of each type of mask, inspection, protection factors and prop-er use. Each person was required to demonstrate putting on the mask and checking the seal for each type of mask to be used. Catalytic supervisors and engineers were given additional training in the management of work involving radiation exposure and contamination control procedures. This course was a multi-media (video tape, workbook, practice and lecture) series, and stressed proper radiation protec-tion orocedures. Special personnel training was conducted for the tritium solidification process. Subjects cover-ed use of positive pressure suits, biological hazards, and use of the special drum adapter. A mock up operation was conducted using full protective equipment prior to the tritium solidifi-cation operation. 7.2. RADIATION EXPOSURE CONTROL Radiation exposure control was accomplished by setting exposure limits for workers, planning the tasks to minimize radiation exposure, and instituting personnel monitoring methods capable of observing actual exposure on a daily basis. 78

7.2.1. Exposure Limits Administrative exposure limits were established before the start of decommissioning activities. Permissible exposures are out!!' ed below: WORK DUTY WORK DUTY Nornul decommission activities Decommission activities conducted in the steam generator cavities and the water cooled delay bed cavities. Limits (Whole Body) Limits (Whole Body) 100 mremiday 500 mremiday 200 mrem / week 2400 mrem / week 400 mrem / month 2400 mrem / quarter 1000 mrem / quarter 3000 mremlyear Extended exposure limits allowing exposures greater than these amounts required approval by l health physics personnel and the station Plant Superintendent. However, these extended expo-sure limits were not required. the maximum exposure to any indi idual did not exceed 50 mreml v day for normal decommissioning activities or 200 mremiday for work in the steam generator cavities or water cooled delay beds. 7.2.2. Exposure Control Planning Radiation exposure rates in Peach Bottom Unit 1 were very low. The only radiation areas greater than 4 mR/hr were in the vicinity of the primary system or fuel purge system. Component re-movai in these areas was relatively simple and no extensive exposure control planning was necessary. The only work performed during decommissioning which required extensive ex-posure control planning was the removal of the ndensibles trap vessel A 301. The radiation reading from this vessel was 15 Rlhr at one foot from,qe surface. Before the condensibles trap was removed, a planning Tieeting was held to develop exact pro-cedures for rigging out and moving the vessel to the shipping container Subjects discussed included: assignment of specific personnel to specifi; tasks; temporary shielding of the vesset during transport; rigging and transoort methods to t+, used; timing; staging areas (Iow dose rate) for workers while waiting to perform tasks; and health physics control. Removal was accompl(shed according to the procedures and without deviation. The vessel was lif ted with electric hoists from the overhead monorail, moved to a transport dolly and shielded, moved out of containment to the auxiliary building crane access area and lif ted into the ship-ping container. Then the container was moved to a waiting concrete truck at *' roll up door and filled with concrete. I 79

The total procedure was accomplished in less than three hours, with 16 workers involved in the operation. The maximum radiation exposure received by any of the workers was 180 mrem and the total exposure to all workers was 1040 mrem. 7.2.3. Exposure to Personnel The total exposure to all personnel during the decommissioning operation was 8.95 man rem. Of this total, approximately 2.0 man-rem was received by personnel during removal of the con-densibles trap. The maximum exposure received by any individual during decommissioning was 750 mrem. All personnel exposures were less than 200 mrem / week during all phases of decommissioning. 7.2.4. Personnel Monitoring Thermoluminescent dosimeters (TLDs) were used as the primary method for determining per-sonnel radiation exposures. These dosimeters were changed monthly to enable them to be processed and their exposures to be recorded. The Harshaw TLD system was used as supple-mentary dosimetry on all decommissioning activities after June 1976. These dosimeters pro-vided daily exposure readout on all personnel. Direct reading dosimeters were also used by all personnel. During Phase 11, when work was being performed in steam generator cavities and water cooled delay beds, these dosimeters were recorded on a daily basis. During Phase lit, dosimeter readings were recorded only on Radiological Work Permits (RWPs) as exposures did not exceed 100 mrem / week and the Har-shaw TLDs were read on a daily basis. 7.2.5. Internal Exposure Decommissioning personnel were monitored for internal uptake of radioactive material. Whole body counts were performed on workers before starting and upon completion of work. No uptake was detected in any of the workers. Urine samples were obtained from workers involved in the solidification of tritiated water and removal of the tritium holding tank. A base line sample was obtained before work started and a sample was obtained 24 hours after the work was completed. No detectable uptake was observ-ed in any of these workers. 7.2.6. Radiological Work Permits Radiological work permits (RWPs) were used to control the working conditions in rad!ological controlled areas. The RWP contains information on radiation exposure rates, airborne racioac. 80

tivity and contamination levels, as well as required protective clothing and equipment and re-quired monitoring. It also documerits health physics and supervisory approval to perform the work under the conditions specified. RWPs were issued daily for each controlled area where work was to be performed. During the component removal phases, a total of 280 permits were issued. 7.3. CONTAMINATION CONTROL Although radiation exposure rates and contamination levels were relatively low for most areas of Peach Bottom Unit 1, contamination potentials were a major concern. Contamination within 7 dpm/100 cm2. Other systems contained signifi-the primary system was estimrited to be 10 cantly lower levels of contamination but stili sufficiently high to contaminate plant areas in excess of the allowed limits for decommissioning. Each individual job was evaluated for potential and actual contamination. Contamination control areas, protective clothing requirements and material control procedures were specified prior to the start of work on each job. These controls significantly reduced final decontamination activities as very few areas were con-taminated by the decommissioning ef fort. 7.3.1. Control Areas Three major contamination control areas were established for decommissioning work: the spent fuel pit building, the radwaste f acility, and the refueling ficor. Other minor control areas were established for local work but were removed upon completion of the individual job. The refueling floor contamination control area was established before the start of decommis-sioning. This section contained the contaminated equipment associated with fuel transfer and 5 dpm fuel canning operations; contamination levels of some of this equipment were 10 /100 cm2. The control area was used while cutting and packaging most of the contaminated equipment on the refueling floor, and included a roped off section, paper covering on the floor, and a step of f pad. Protective clothing was worn in the area and removed before leaving. The spent fuel building was a major contamination control area. The floor around the top of the pool was covered with paper, and step-off pads were placed at the entrance to the pool and at the exit from the building. Protective clothing was worn in the area and removed before leaving. Additional shoe covers and gloves were required for entrance to the bottom of the pool; these were removed before returning to the upper pool area. The radwaste basement contamination control area required a dif ferent approach. The basement contained five areas where equipment was removed. Because the crane access hatch to the basement was at one end of the facility, large quantities of material were moved across the 81

rooms and lif ted through the hatch to the radwaste contaners. Most of this material was con-2 taminated with low level radioactivity (less than 10,000 dpm/100 cm ). Since bagging or wrapping this material would have been costly and time consuming, it was decided to maintain the whole basement area as a contamination control area. This arrangement allowed unrestricted move-ment of all material within the basement to the radwaste containers located above the equip-ment hatch. The potential for wet contamination was high, so the floor was kept uncovered and decontaminated as necessary. Protective clothing was worn in the area and removed before exiting at the step-of f pad. A contamination control area was also set up in the vicinity of the cask wash area (ground floor) for cutting and packaging of f arge spent fuel handling equipment. The area was enclosed on the bottom and all sides with plastic sheet f astened to an outside frame of scaffold supports. Pro-tective clothing was worn in the area and removed before leaving. Other contamination control areas were set up for individual tasks. As a usual procedure, paper was used to cover the floor; other components of the control area were a roped-off section and a step off pad. 7.3.2. Radioactive Material Control The decommissioning plan specified removal of the majority of contaminated systems except for the primary system and piping within the exclusion area. Usually these systems were uncon-7 taminated on the exterior but were highly contaminated on the interior surf aces (as high as 10 dpm/100 cm2). The spent fuel pool equipment was also contaminated on the exterior surfaces dpm/100 cm ). Contamination control during removal of these systems and equipment 2 0 (to 10 required observance of good material control procedures. Most cutting on contaminated equipment was done with either reciprocating saws or band saws. This method of cutting produces relatively large chips of material which fall in the vicinity of the cut and are too large to become airborne. Flame cutting and arc-gouging were allowed only 2 when contamination levels were less than 10,000 dpm/100 cm, When contaminated material was removed from a system or location, the material was either bagged or taped closed and moved to a radwaste container or a laydown area. In many situa. tions, the radwaste containers were placed adjacent to the contamination control area to permit material to be placed directly into the container without requiring wrapping or bagging. Contaminated tools and equipment were double bagged before leaving a ccntamination control area for transfer either t: contaminated tool storage or to another controlled area. Upon com-pletion of decommissioning, all useful contaminated tools and equipment were removed to the l contaminated tool storage area in Ur.it 2. 2 All tools and equipment which had no detectable radioactivity (less than 500 dpm/100 cm fixed and less than 100 dpm/100 cm removable) were released from further control. 2 i 82

7.3.3. Decontamination A major effort of decommissioning was the decontamination of the entire facility. The spent fuel pool walls (600 f t ) were contaminated, as were more than 80 embedded pipes in the rad-2 waste system. Also, more than 10,000 ft2 of the 75,000 ft2 of floor space within the facility were known or suspected to be contaminated. All areas outside the specified exclusion area were required to meet the acceptable surface contamination levels shown in Table 71. The decommissioning plan also specified that all inspection access areas within the exclusion area v.are to be decontaminated to these levels. In addition, contamination levels within the exclusion area were to be reduced to the lowest practicable levels. TABLE 71 ACCEPTABLE SURFACE CONTAMINATION LEVELS b NUCLIDEa AVERAGE c MAXIMUMbd REMOVABLE c b U nat, U 235, U 238, and 2 2 1,000 dpm a/100 cm 2 15,000 dpm a 100 cm associated decay products 5,000 dpm a/100 cm Transuranics, Ra-226, Ra 228, Th 230, Th 228, Pa-231, Ac-227, 2 2 2 300 dpm/100 cm 20 dpm/100 cm 1-125,1129 100 dpm/100 cm Th nat, Th 232, Sr 90, Ra 223, 2 2 200 dpm/10n 2 3000 dpm!100 cm Ra-224, U-232, l 126,1131,1 133 1000 dpm/100 cm Beta-gamma emitters (nuclides with decay modes other than alpha emission or spontaneous fission) except Sr 90 and 2 1000 dpm 09/100 cm2 2 15,000 dpm #9/100 cm of hers noted above 5000 dpm #1/100 cm aWhere surface contamination by both alpha and beta gamma-emitting nuclides exists, the limits estab-l lished for alpha and beta gamma-emitting nuclides should apply independently. As used in this table, dpm (disintegrations per minute) means the rate of emission by radioactive material i b as determined by correcting the counts per minute observed by an appropriate detector for background, j ef ficiency, and geometric f actors associated with the instrumentation. cMeasurements of average contaminant should not be averaged over more than 1 square meter. For objects of less surf ace area, the average should be derived for each such object. 2 dThe maximum contamination level applies to an area of not more than 100 cm - 2 of surface area should be determined by eThe amount of removable radioactive material per 100 cm wiping that area with dry filter or sof t absorbent paper, applying moderate pressure, and assessing the amount of radioactive material on the wipe with an appropriate instrument of known efficiency. When removable contamination on objects of less surface area is determined, the pertinent levels should be reduced proportionaffy and the entire surface should be wiped. l l t 83

Various methods of decontamination were used, depending on surface porosity, type of con-tamination (fixed or removable), level of contamination, and accessibility. Removable con-tamination was disposed of with an absolute filtered vacuum cleaner or by washing with sol-vents and rags. Fixed contamination on metal surfaces was usually cleaned with foaming bath-room cleaner and rags. Painted walls in the spent fuel pool responded best to the use of a com-mercial liquid cleaner. Some stubborn fixed contamination on concrete required abrasives (sandpaper and wire brushes), and when this method was ineffective, a jackhammer was used. The refueling floor, refueling equipmcqt and areas above the refueling floor contained many 6 2 small spots of fixed contaminatinr. :.ith levels up to 10 dpm/100 cm. Eacn of these spots was 2 decontaminated to less tt an 50L0 dpm/100 cm (average) or was physically removed. A mobile scaffold platform, which could be lowered by removing 5 ft sections, enabled the de-contamination crew to mach all areas of the spent fuel pool walls. Three methods of cleaning the walls were tested; results showed that a commercial liquid cleaner and rags produced the best decontamination and cosmetic cleaning. One application of cleaner to the pool walls usually reduced the original removable contamination of 30,000 dpm/100 cm2 to less than 300 dpm/100 cm. Some of the pool surfaces had fissures in the paint surface which required 2 more than one attempt at decontamination. After cleaning, the highest removable contamination areas was 1300 dpm/100 cm. The floors and walls around the top of the pool and the fuel 2 handling crane (grapple crane) were decontaminated to less than 500 dpm/100 cm2 (fixed) and 2 less than 1000 dpm/100 cm (removable). All areas in the containment vessel, including the controlled areas such as loop cavities, low-temperature de!ay bed cavity, sub pile room, helium dump tank. area, and other areas behind barricades, were decontaminated as necessary to remove loose contamination; removable con-2 tamination was reduced to less than 1000 dpm/100 cm, All inspection access areas within the exclusion areas were decontaminated to the acceptatAe surf ace cont?mination levels. The fuel cask wash and loading area contained many local spots of fixed contamination with levels up to 105 dpm/100 cm2 After other methods proved unsuccessful, these contaminated spots and areas were decontaminated by removing the first inch of concrete. The radwaste system and drains leading to the radwaste systems included 82 pipes which were embedded within the floors and walls of the facility. Most of these pipes were contaminated iri excess of the allowable surface contamination levels. Various methods were used to decontarm nate these pipes and drains: pipes which were readily removable were taken out and disposed of; short straight pipes were decontaminated with flue brushes; acid leaching and flushing using sulfuric acid removed or loosened organic and acid soluble coatings. The last method used for pipes still contaminated was cleaning with high pressure water err-ploying a 6500 psi water nozzle; this proved a very effective decontamination technique for the pipes. The high pressure nozzle is self-propelled and will pass through any pipe over 1% in, diameter without any short 90 degree turns. Only eleven pipes had any residual low level radio-activity af ter decontamination by this method. 84 g e v. -.-w

6 dpm/ The liquid waste basement had large areas of fixed contamination with levels up to 10 100 cm ; the walls also had many spots of fixed contamination. Each square meter of the facility 2 I was surveyed for fixed and removable contamination. Of the 711 square meters surveyed,122 2 had fixed contamination greater than 5000 dpm/100 cm. All of these areas were decontamina-ted, using solvents and rags or by removing the concrete with jackhammers. After decontamina-2 tion, monitoring of all areas showed total fixed contamination to be less than 5000 dpm/100 cm, and total removable contamination less than 500 dom /100 cm2 Local spots of fixed contamination requiring decontamination were also located in the laundry area, laboratory, small tool decontamination room, auxiliary stairs, and decontamiration sink. All of these areas were successfully decontaminated. 7.4. SURVEYS FOR RADIOACTIVE MATERIAL Various types of surveys for radioactive material and radiation were routinely performed throughout the decommissioning of Peach Bottom Unit 1. Radiation and contamination surveys were conducted before starting work in any area. Airborno radioactivity levels were determined l during any task which involved work in highly contaminated areas or cutting into potentially con- ) taminated systems. Special surveys were taken for embedded pipes for direct and removable contamination. Finally, when decommissioning activities were completed in an area, a radiation and contamination survey was performed to provide baseline. data for the decommissioned facility. 7.4.1. Radiation Surveys Radiation dose rate surveys (whole body) were made of all accessible areas of the facility before the beginning of dacommissioning activities. In addition, each Control Work Package (CWP) contained prerequisites for survey requirements and review by the Health Pnysics Supervisor before work activities could begin. Areas in which personnel could receive 5 mRlhr (radiation areas) were monitored intermittently during work activities to detect any changes of radiation levels. Areas where whole tody dose rates were greater than 100 mR'hr (high sadiation areas) were monitored continuously during shielding changes such as removing magnetite f rom the water cooled delay beds. The only accessible areas in the plant with radibtion levels over 100 mRlbr (whole body) were in the loop cavities and in the water cooled delay beds. Most areas withir' the plant were below 5 mR/hr. Special radiation surveys were gsrformed in areas where components were located in com-pletely closed shielded enclosures. A hole was cut in the shield wall large enough to insert a Teletector probe for surveys before an entrance was opened. t Radiation surveys for beta radiation were performed when highly contaminated systems were opened.The beta dose rates were highest on the spent f uel chute (1400 mrad!hr at 4 cm) and the primary systi:m piping (300 mrad /hr at 4 cm), 85

Whole body radiation dose rate surveys were also performed as part of the facility final survey. This final survey is detailed in Section 7.5 of this report. 7.4.2. Contamination Surveys Contamination surveys comprised the bulk of the work for the health physics staff during de-commissioning. In addition to routine wipe surveys to detect removable contamination, contami-nation surveys were required on each piece of material or equipment leaving the facility. During decommissioning, more than 20,000 ft3 of material were released from the site as uncontami-nated waste and 14,000 f t3 of material were shipped to a licensed burial ground as radioactive waste. All of this material was monitored for fixed and removable contamination. Thin window (1.8 mg/cm ) pancake G M detectors were used for monitoring all thed and re-2 movable beta gamma radioactivity. For alpha radioactivity, ZnS(Ag) scintillation detectors were employed. Removable beta gamma contamination was determined by two techniques. Disc smears were taken over a 100 cm2 area and analyzed with a shielded G M detector and laboratory scaler. Paper towel wipes were taken over a 1 ft2 area and analyzed with a portable G M instrument. The relationship of 100 dpm/100 cm2 approximately equals 100 cpm /ft2 was used for towel i wipes. Removable contamination surveys were made throughout the facility while decommissioning activities were in progress. Areas outside contamination control zones were limited to 100 dpm/ 100 cm. Areas within contamination control zones were monitored for removable contamination 2 and decontaminated as necessary. Before their release from the facility, equipment and material were required to be surveyed for fixed and removable contamination. Only those items which had no detectable activity (less 2 or fixed and less than 100 dpm/100 cm2 or removable contamination) f f than 500 dpm/100 cm were allowed to be released for unrestricted use. Af ter decommissioning activities were complete. the entire facility was tested for contamination as a part of the facility final survey. ) 7.4.3. Airborno Radioactivity Airborne radioactivity was monitored during work procedures in areas of high potential or actual surf ace contamination. Air samples were also taken in isolation or enclosed areas before entry. With the exception of cutting out the dust collector in the loop cavity and for some work in the spent fuel pool, all airborne activity was less than 1 x 10-9 pCilml. Airborne activity during cutting of the dust collector was 2 x 10-8 gCilml. Isotopic analysis of this sample and a compos-ite MPC analysis indicated this sample to be 51% of maximum permissible concentration. 86 1

An air sample taken in the breathing zone of workers sawing spent fuel rack guide rods in the spent fuel pit room detected airborne radioactivity containing alpha emitters. The sample indicated 2 x 1011 pCilmi of long lived alpha radioactivity in addition to the beta radioactivity normally found. Based on an analysis of potential alpha emitters in this reactor, the MPC for 238 u was used for all work in this area. Air sampling for alpha radioactivity was con. P insoluble tinued for the duration of work in the area. Air sampies with sufficient volume to provide a minimum detectable activity of 7 x 10-13 pCi/mi t 100% were collected and analyzed while work continued in the fuel pool. Although samples 238 u, respiratory 137 s and P C indicated levels less than maximum permissible concentrations of protection was used and the area posted as required by 10 CFR Part 20.203 as an " Airborne Radioactivity Area" 7.4.4. Embedded Pipe Surveys Peach Bottom Unit 1 contained contaminated piping which was embedded in concrete floors and walls outside of the specified exclusion areas. Acccrding to the plan, these pipes were required to meet accep'able surface contamination levels for unrestricted release; therefore, special survey techniques were applied to determine fixed and removable contamination levels. A sample of the drain pipe residue was obtained and analyzec k>r isotopic content by gamma 137 s C spectroscopy. The results of this analysis indicated that 98% of the radioactivity was from 60Co and other isotopes. and 134Cs, and remaining 2% was from All embedded pipes were surveyed internally for fixed contamination using a small end window 2 2 137Cs in a cylindrical (2.2 mg/cm ) G M tube.This tube was calibrated for 100 cm geometry with configuration of the pipe sizes to be surveyed. Readings were obtained at six-inch intervals inside of the pipes at each opening for 60 inches or until further movement of the probe was obstructed. The readings were then analyzed to determine average and high contamination for each pipe monitored. Wlpe surveys were performed on the pipes when final decontamination was complete. The wipe was a single pass over the internal circumferences of the end of the pipe. Of the 82 pipes and drains which were embedded in the walls and floors of the radwaste base-ment,11 remained contaminated to levels exceeding those acceptable for release for unrestrict-ed use even af ter extensive decontamination efforts including mechanical cleaning, acid leach-ing, and Hydrolaser cleaning.- A hazards evaluation was prepared to determine the radiological status and to recommend al-ternatives to removal of these pipes. The hazards evaluation shows that rat vlogical hazards would be minimal even under conditions of flooding, firo or earthquake and tb t the quantitative amounts of radioactive material in the pipes are exempt from licensing. The liquio waste hazards summary is contained in Appendix C. 87 ~ -~

7.4.5. Shipment and Transport Surveys A Shipment inventory Data Sheet and Radioactive Materials Shipment Record was completed for each shipment. All drums, boxes or other radwaste containers were surveyed and labeled before loading. A radiation and removable contamination survey was performed on the truck before it was loaded. Af ter loading, the truck was again surveyed for radiation dose rates and the results of both surveys logged on a special shipment survey sheet. All surveys showed that shipments met the radiation levels specified in 49 CFR. 7.5. FACILITY FINAL SURVEY The facility final radiation and contamination survey served two major functions: it ensured that the various areas of the plant conformed to the criteria of the decommissioning plan; and it identified any radiological hazards. Final surveys in an area were conducted af ter decommission-ing activities were completed and before the area was barricaded. After all decommissioning work was completed, a final radiation and contamination survey was conducted in the unrestrict-ed areas. 7.5.1. Radiation and Contamination Levels The decommissioning plan criteria for radiation levelt are:

  • Less than 0.04 mrem /hr in all areas of unrestricted use in the administration building
  • Less than 0.08 mremlhr in all areas of unrestricted use in the turbine and auxiliary buildings
  • Less than 1.0 mR/hr in the accessible inspection areas of the exclusion areas.

No specifications for radiation levels within the locked or barricaded portions of the exclusion areas were set by the plan, but reasonable efforts were made to reduce radiation to the lowest possible levels. The acceptable surface contamination levels specified in NRC Regulatory Guide 1.86 are given in Table 71. The decommissioning plan also required that the accessible inspection areas of the exclusion area meet these levels. In addition, a reasonable effort was made to reduce all contamination within the plant to the lowest prscticable levels. 7.5.2. Survey Techniques Several different techniques were used to perform the various required surveys for gamma radia-tion, fixed beta gamma contamination, removable beta gamma contamination, fixed uranium contamination, fixed unidentified alpha contamination, and removable alpha contamination. 88

Beta-gamma radiation surveys in areas above 0.2 mR/hr were performed with calibrated ion chamber instruments. Gamma dose rates were read directly with this Instrument. Gamma radiation surveys in areas less than 0.2 mRlbr were performed with a portable gamma scintillation detector and analyzer. This instrument was calibrated to measure gamma radiation above an integal bias of 100 kev. This technique was used to differentiate between radiation from byproduct material and natural background radiation. Fixed beta-gamma contamination was surveyed with calibrated pancake G M detectors and 2 portable count rate meters. To convert counts per minute to dpm/100 cm, detectors were tested 137 s source. All detectors tested 25% or higher efficiencies C for efficiency to an NBS traceable at 1 cm from the source. The detector background was determined to be 30 cpm. With a known detector area of 15.5 cm2 and a minimum detector efficiency of 25%, the average fixed con-tamination of 5000 dpm/100 cm2 was converted to maximum permissible cpm by the following calculation: 5000 dpm 1 cpm 15.5 cm2 + 30 cpm background = 224 cpm. 100 cm2 4 dpm The surveys were performed by placing the probe approximately 1 cm from the surface to be monitored. Areas with rendings above 200 cpm were designated for decontamination. Removable beta gamma contamination was surveyed by wiping an area of 100 cm2 with a disc wipe. The wipes were analyzed in a calibrated laboratory counting system. Removable beta gamma surveys were also performed on large sections of tne facility. Most of the floor surface of the area of concern was wiped with a Masslinn broom. The surface of the broom wipe was then monitored using a pancake G M detector. These surveys were performed on an area to evaluate possible gross contamination before the final wipe survey with disc wipes. Fixed alpha radioactivity from uranium contamination was surveyed with a ZnS(Ag) scintillation 2 detector and a portable count rate meter. To convert counts per minute to dpm/100 cm, the de-230Th source; the value was determined to be 31%. With a tector was tested for efficiency to a known detector area of 59 cm, the average fixed contamination of 5000 dpm/100 cm2 was con-2 verted to maximum permissible cpm by the following calculation: l 1 cpm 500 dpm x 59 cm2x = 922 cpm. 100 cm2 3.2 dpm Surveys for uranium contamination in the fuel vault area were made by placing the probe face in contact with the surface to be monitored. Maximum allowabin reading was 500 cpm on the in-strument, or less than 3000 dpm/100 cm2, Fixed alpha radioactivity from unidentified alpha contamination was monitored with the alpha scintillation detector as described above, but a portable scaler was used for counting. The back-89

i ground of the detector was tested and determined to be 6 cpm. With a known efficiency of 31% and a detector area of 59 cm, the contamination level of 100 dpm/100 cm2 was converted to 2 maximum permissible cpm by the following calculation: 1 cpm 100 dpm x 59 cm2 x + 6 cpm instrument background = 24 cpm. 100 cni2 3.2 dpm This low count rate introduces statistical error. To avoid statistical evaluation of each reading, all monitored points were limited to 10 counts in one minute. The surveys were performed by placing the probe in contact with the surface to be monitored and making a one minute count with the scaler. Removable alpha radioactivity for uranium and unidentified alpha contamina:lon was surveyed by wiping an ar 'a of 100 cm2 with a disc wipe. The wipes were than analyzed in a calibrated laboratory counting system. 2 or removable uranium contamination did not introduce statis-f The limit of 1000 dpm/100 cm 2 or re-f tical error problems for short counting times. However, the limit of 20 dpm/100 cm movable unidentified alpha contamination requires long counting times and testing the count rate for significance. In addition, the population of this Icw count rate is distributed as the Pois-son distribution. A statistical evrluation was developed for use with this low count rate. Another survey technique was used for monitoring fixtd contamination ' embedded pipes. This technique is described in Section 7.4.4. of this report. All sanitary system drains and traps in the auxl'iary builcing and 1dministo Pon building were surveyed with the gamma scintillation probe.The probe was placo :n contav. with the external surface of the pipe or drain and a reading obtained. 7.5.3. Final Survey of Exclusion Areas At the time of final survey of the exclusion areas, high radiatier. areas existed in the reactor pres-sure vessel cavity and primary loop cavities. The maximum reading in the loop cavities was 200 mR/hr at the bottom of steam generator No.1, and general radiation levels in the loop cavi-ties ranged from 5 mRlhr to 80 mR/hr. A radiation area is locateo on elevation 111 f t in the sub-pele ioom. This radiation derives from material in an access port near the control rod drives. At the,me of survey, the hot spot on the access port read 1 Rlhr at contact. Radiation levels at the rail around the walkway read between 4 and 20 mR/hr. { Radiation dose rates up to 60 mR/hr on contact were detected on piping in the fission products trapping system. This piping is located in the pipeway and in the wster-cooled delay bed area. The only other area within the exclusion area with radiation levels above 1 mR/hr is located on the southwest stairwell of containment at elevation 141 f t, at the door to loop cavity No.1; the radiation level in this area read 2 mRlbr. 90

The final survey showed that, with the exception of some fissures in the paint of the spent fuel pit walls, all surface areas within the exclusion area had removable beta. gamma contamination levels of less than 1000 dpm/100 cm2; the fissures had removable contamination levels up to 2 1500 dpm/100 cm, At the end of decommissioning work, radiation dose rates within the inspection areas of the containment vessel, spent fuel pool building, and liquid waste facility were less than 1 mRlhr: 2 fixed beta gamma contamination levels were less than 5000 dpm/100 cm ; and removable contamination levels were less than 1000 dpm/100 cm2 Embedded pipes which were potentially contaminated were monitored for fixed and removable contamination. Piping in the administration and auxiliary buildings outside of the liquid waste 2 2 area shcwed readings of less than 5000 dpm/cm (fixed) and less than 1000 dpm/100 cm (remov-able). The liquid waste area contains eleven pipes which are above the acceptable contamination levels. A hazards summary (Appendix C) has been prepared for these pipes. 7.5.4. Final Survey of Unrestricted Areas At final survey, radiation dose rates within the auxiliary building and administration building were less than 0.01 mR!hr. A radiation level of 0.03 mR/hr was detected in the record room on the second floor of the administration building. During the final survey, an investigation determined 226 a source in a storage container on the floor above the R that the radiation was from a 2 mC record room. After additional shielding was placed under the source container, the radiation level was reduced to less than 0.01 mR/hr. Fixed contamination surveys for beta gamma radiation indicated levels of less than 5000 dpm/ 100 cm for the auxiliary and administration buildings; removable beta gamma in these buildings 2 2 was less than 1000 dpm/100 cm, Although the fuel used in Peach Bottom 1 contained thorium as a fertile material. the activity-mass re!ationship of the uranium thorium mixture shows that uranium provides the predominant alpha activity. Both levels of the vault were monitored for removable and fixed alpha contami-2 nation, and showed fixed contamination levels of less than 5000 dpm/100 cm, with removable 2 contamination levels less than 100 dpm/100 cm, An 80 point survey was performed to determine alpha radioactivity contamination from uniden-2 tified alpha emitters. All 80 points indicated less than 100 dpml100 cm fixed contamination and 2 less than 20 dpm/100 cm removable contamination. Sanitary sewer traps and floor drains within the auxiliary and administration buildings were sur-veyed for contained radioactive material. All readings were less than 0.01 mR/hr, indicating a content of less than 0.1 pClof137 s. C Twenty five survey locations have been identified for future radiation and contamination sur-veys. These locations are shown in Figures 6 3,614,6-15 and 616. Contamination and radiation 91

readings were taken on the surface of the 5 !c.. x 7 in. survey location signs. Survey locations 1 through 19 were less than 0.01 mRlhr gamma. At the time of survey, the following locations had radiation dose rates 0.01 mRlbr or greater: Survey Point mR/hr Survey Point m R/5r 20 0.01 23 0.04 21 0.01 24 0.06 22 0.07 25 0.01 2 fixed contamination (beta gamma) and less than All locations had less than 5000 dpm/100 cm 1000 dpm/100 cm2 removable contamination (beta gamma). Seventy five of the final survey areas were tested for removable contamination by gross area wipes. This method consisted of wiping most of the floor surface with a Masslinn broom and testing the wipe for radioactivity. 7.5.5. Final Survey Summary The final facility survey was performed in more than 80 areas of the facility. Direct surveys for 2 of building surf aces. The removable con-contamination were performed on over 40,000 f t tamination surveys involved the collecting and analysis of more than 5000 wipe samples. Dose rate measurements were performed in more than 500 locations. I All of these surveys indicate a very low level of residual radiation and contamination within the facility. Radiation levels above 1 mRlhr and contamination levels above the acceptable surface f contamination levels exist only beyond the barricaded locations of the exclusion area. The ( facility in its present status presents no radiological hazard to workers outside the exclusion area or in the unrestructed areas of the facility. 7.6. INSTRUMENTATION 7.6.1. Dose Rate Measuring Beta and gamma dose rate measurements for personnel protection were taken primarily with ionization type instruments. The maximum dose rate capability of the ion chamber instruments was 500 R/hr. In areas of restricted access or high dose rates, a G-M type Teletector was used. This instrument has a telescoping probe and a maximum range of 1000 R/hr. Final survey measurements outside exclusion areas were taken with a gamma scintillation detector with a portable gamma analyzer; calibrated range was between 0.01 mRlhr and 0.2 mR/hr. 92

7.6.2. Contamination Monitoring Fixed beta-gamma contamination was monitored using portable count rate instruments with pancake G-M probes. Fixed alpha contamination was monitored with both portable count rate instruments and a semi portable scaler with an alpha scintillation probe. Removable beta and gamma contamination was monitored with disc or towel wipes. Towel wipes were counted with a pancake G.M probe and a portable count rate instrument. Disc wipes were read with a shielded pancake G M probe and scaler or a laboratory beta counter. Removable alpha contaminaticn was monitored with disc wipes and read with a laboratory alpha counter or alpha scintillation probe and scaler. Additional contamination surveys and isotopic analyses were performed on water, air, sludge and wipe samples with a gamma spectroscopy system. Tritiated water samples were measured with a liquid scintillation counter. Contarninated oil was analyzed for tritium and isotopic con-tent by Radiation Management Corp. of P.hiladelphia. A special survey technique was used to determine in pipe contamination and the level of resid-ual activity in embedded pipes. A one-inch end window G M probe with laboratory scaler was used for this purpose. 7.6.3. Air Samplers Air sampling was performed with either high volume or low volume samplers. A 47 mm glass fiber filter was used in either unit and read with a shielded pancake type G M probe and labora-tory scaler or laboratory beta counter. Alpha determinations were made with an alpha scintilla-tion detector and laboratory scaler. 7.6.4. Calibration of instruments All portable radiation monitoring instruments were calibrated initially and at three-month inter. vals thereafter by an outside vendor. The calibrations are traceable to the National Bureau of Standards. 20 %. Dose rate measuring instruments were calibrated to two points on each range within No correction factors for dose rates were required for these instruments. Count rate meters with pancake G M detectors were calibrated to correct count rate by the ven-dor. Detector efficiencies were tested with a source traceable to the National Bureau of S t 137 s source; C dards. For final facility surveys, the detectors were tested to an NBS traceable all detectorr 'ested to 25% or higher efficiency. Laboratory counting equipment was calibrated daily during decommissioning operations by test-ing for efficiency to an NBS traceable source. Air samplers were tested for air volume by mea-suring dif ferential pressure across a calibrated orifice. 93

A special calibration was performed on the portable gamma analyzer system to determine low dose rates for final surveys. The high voltage and threshold set points were adjusted to cut out 137Cs traceable to NBS gamma radiation below 100 kev. The counter was then calibrated with within +20% 0%. This method was used to dillerentiate between radiation from byproduct material and natural background material, and to reduce the low energy contribution to the spec-trum typical of small Nal(TI) detectors. The detection system for embedded pipe surveys (G M detector and scaler) was calibrated for various pipe size geometries. The procedure for the calibration and survey technique is given in Section 7.4.4. Alpha radiation detection instruments were calibrated to NBS traceable 230Th and 239 u P sources. Portable survey instruments were calibrated at three month intervals by the outside vendor. Laboratory counting systems were calibrated daily during the alpha monitoring activities. i 7.7. SAFETY Overall safety responsibility was derived from the Philadelphia Electric General Superintendent and administered by the Catalytic Field Superintendent and Philadelphla Electric Construction Management. Catalytic complied with its standard Corporate Safety Program for field work. This program is in compliance with all OSH A requirements. Safety considerations related to a specific task were discussed and controlled by Control Work Package instructions. Philadelphia Electric safety rules and practices were in force for both Phase 11 and Phase ill work. Weekly on-the job safety meetings were held with craft labor. Unit 2 3 safety personnel made monthly safety presentations. Six minor injuries occurred during the combined 47-week work periods which took place in two calendar years. First aid was administered to all injured personnel. There were no,ost time injuries. A further discussion of radiological safety is presented in Sections 7.2.2. throt gh 7.2.5. 94

8. STATUS OF RETIRED PLANT 8.1. CONTAMINATED SYSTEMS 8.1.1. Primary Helium System The decommissioning plan specified that the fission product trapping system be isolated from th_ main coolant loop to permit degassing of the purification system during the End of Life sampling program. However, isolation was found to be not necessary, as degassing of the purifi-cation system was completed before the sampling program began.

During the End-of. Life sampling program,148 samples of steam generator tubing, primary piping, and steam generator internals were removed f rom Primary Loop No.1. Openings in the primary helium system, which were made during the decommissioning and End of. Life sampling program, were welded closed. Discharge pipes from the reactor safety valves and the coolant piping relief valves were welded closed. Loop 1 and 2 main compressor drive shaf ts were disconnected and the shaf t openings were sealed. The seal oil and tube oil were drained and the filters were removed from the helium circulator oil systems. Oil residues were cleaned up to eliminate fire hazards. An integrity test was performed on the primary system; this is discussed in Section 6.5.11. The primary helium system is internally contaminated from fission product material. Removable 6 2 contamination was determined to be 10 dpm/100 cm in January 1976. 8.1.2. Hydraulle Control Rod Drive System The control rod drives were deactivated fully inserted. Hydraulic fluid from the drives, pumps and reservoir were drained. Oil residues from system components were removed to climinate fire hazards. Surface contamination levels on the exterior control rod drive system are within acceptable sur-f ace contamination levels for decommissioned f acilities. 8.1.3. Shield Cooling System The inhibitor solution was drained from the shield cooling water system and processed tnrough the liquid waste system. All compcnents of the shield cooling system outside containment were 95

removed. Supply and return lines were cut and capped at containment. Contamination in this 4 2 system was about 10 dpm/100 cm at completion of work. 8.1.4. Fission Products Trapping System Af ter degassing, components of the fission product trapping and delay system were removed as discussed in Section 6.5.7. Piping cut for equipment removal was seal welded closed. Shield plugs and decking plates were reinstalled over equipment cubicles. External contamination of than 10 dpm/100 cm, and internal contamination was approximately 3 2 this area was lest 6 2 10 dpm/100 cm at time of testing. 8.1.5. Fuel Handling Purge System The fuel handling purge system was vented and oil was drained from the vacuum pumps. The oil filter and exhaust filter cartridges were removed as radioactive waste. 8.1.6. Chemk al Cleanup System The purge water condenser, water separator caustic scrubber and water scrubber in the chemical cleanup system were drained. The copper in the oxidizer was regenerated to copper oxide and lef t in the vessel. The steam generator purge plateout trap was removed to con-taminated waste. Cuts made in the pipe were welded closed. No other pipe or components were removed. 8.1.7. Fuel Pool Cooling System This entire fuel pool cooling system 'was removed, except for the inlet and outlet pool pipe penetration stubs which were seal welded closed to make them inaccessible. 8.1.8. Radiation and Process Monitors Except for one, all radiation and process monitors were deactivated in place and all radiation test sources were removed. The active radiation monitor is in the Unit 1 administration building lobby on the ground floor. 8.1.9. Decontamination System Liquid from the decontamination system was drained and processed as contaminated liquid waste. A blank flange was welded on the top of the decontamination tank, and the system was retired in piace. The discharge line was capped off at containment. This system was not contaminated. 96

8.1.10. Liquid Waste Disposal System Oil in the contaminated oil storage tank was drained, solidified, and disposed of as radioactive waste. lon exchange resins froni the liquid waste system were drummed, dewatered and shipped off site for licensed disposal. Liquid waste was discharged in accordance with the technical specifications. This entire system was removed as discussed in Section 6.7. Eleven embedded pipe segments remain as shown in Figure 617. They are identified with " Buried Radioactive Pipe" signs. At completion of work, these pipes contained about 8 #Ci of radioactive material, 2 and contamination levels up to 25,000 dpm/100 crn exists in the sealed piping. 8.1.11. Ventilation System Filters from the ver.tilation system Exhaust Plenums 1 and 2 were removed, as were the labora-tory vent filters. The dryer duct and laundry hood ducts were also removed, as was a contami-nated portion of the fuel pool cleanup line. The stack is not contaminated and is retired in place. Supply and exhaust ducts were seal welded at the containment and the fuel pool. The liquid waste and laboratory vent systems aro within contamination limits for unrestricted use, and remain in place. 8.1.12. Nonpurified Hellur,) Handling System The plateout absorber was removed as radioactive waste and the charcoal cartridge was re-moved from the oil adsorber. Oil was drained from the lubricating priming, injection, and pulsing oil systems of the helium transfer compressors. The helium supply line is capped at contain-ment and all openings made to the system were welded closed. There was no further work per-formed on this system. 8.1.13. Purified Helium Handling System The charcoal cartridges f rom the oil removal filters were removed and the vessels were resealed. The liquid nitrogen traps were removed and the associated pipes were sealed. The helium supply was cut and capped at the containment penetration. The cartridges were removed from the oil filters of the purified helium compressors. Oil and cooling water from the purified helium com-pressors were drained. 8.1.14. Containment Equipment Cooling Water System The containment equipment cooling water system was processed through the liquid waste sys-tem. Contamination surveys of this system indicated that contamination was within acceptable surf ace contamination levels; the system was therefore retired in place. 97

8.2 NONCONTAMINATED SYSTEMS Piping and ventilation penetrations to the conta:nment vessel were cut and sealed during decommissioning. When cutting and sealing a penetration, it.was required that a fixed and removable contamination survey be made of the interior of the piping. The following systems were determined to be noncontaminated based upon these surveys. 8.2.1, Feedwater System The feedwater system was drained and lef t in place without further decommissioning. The feed, water supply lines were seal welded at containment. 8.2.2. Inert Gas Generator The inert gas system was retired in place for potential salvage. The propane tank and supply line had earlier been purged with nitrogen. Supply and return lines were seal welded at con-tainment. 8.2.3. Circulating Water System Part of the circulating water system is self draining. The system was left in standby condition without additional work. 8.2.4. Turbine Generator and Auxillaries The entire system is in a lay up condition for potential sale or use elsewhere. 8.2.5. Emergency Cooling Water System The emergency cooling water system was tented, capped inside containment, and retired in place. 8.2.6. Chilled Water System Water and glycol were drained from the chilled water system, and the system was retired in place. Supply and return lines were cut and capped at containment. 8.2.7. Refrigeration and Brine Systems Refrigerants were removed from these systems, the lubricating oil system was drained and the oil heater breakers opened. Supply and return lines were cut and capped at containment. The refrigeration and brine systems were retired in place. 98

8.2.8. Nitrogen Recondensers The seal and lubricating oil were drained from the nitrogen recondensers and the helium was vented. The system was retired in place. 8.2.9. Containment Hot Water Heating System The containment hot water system was drained, then cut and capped at containment along with all other containment penetrations. 8.2.10. Electrical System All 480 volt breakers to equipment in the exclusion area were locked open. Electric service is available for lighting in containment, fuel pool buildings, and liquid waste basement. Power is available for the containment cathodic protection system and 110/220 service is available for the administration building and other equipment in lay.up status. The original design of the electri-cal system provides that power supplied outside the exclusion area is independent from that supplied inside the exclusion area. 8.2.11. Containment Cathodic Protection System This system is in operation to reduce corrosion of the containment vessel. The power supply panel is located inside the exclusion area and will be inspected periodically. 8.2.12. Support Systems The following systems are in standby or in service condition:

  • Turbine Building Cooling Water System
  • Critical Service Water System

. Diesel Generator j

  • Firefighting and Alarm Systems
  • Makeup Water System j

I

  • Service Water System
  • Fire Water System
  • Domestic Water System
  • Service and instrument Air Systems
  • Instrument Systems (Outside Containment)

These noncontaminated systems are outside the exclusion area and will be operated to provide office and shop facilities in the former Unit 1 administration building. i 99

8.3. SAFEGUARDS AND RADIOLOGICAL SAFETY Radiological hazards and radioactive material are present only in the controlled areas of the exclusion areas. These areas are locked and barricaded; key control of the locked areas is by the General Superintendent, Generation Division of Philadelphia Electric Company. Dif ferent keys to the various locked barriers are provided to further control inadvertent entry or entry for vandalism. 8.3.1. High Radiation Areas High radiation areas are located within the loop cavities and reactor vessel cavity. The maximum radiation level in the loop cavities is 200 mR/hr. Removable contamination levels in the loop 2 cavities are less than 100 dpm/100 cm. Radiation and contamination levels in the reactor vessel cavity ware not determined. Access to the loop cavities is from the ground floor, and requires passing through two locked barriers and removing two barricades. Gaining entry to the loop cavi ties from the refueling floor involves passing through two locked barriers, a locked stair hatch and removing a barricade. Access to the reactor vessel cavity first requires passage through two locked barriers, each keyed differently; any further entry involves activating and mobilizing the 40-ton crane, removing the missile beams, and providing special tooling to remove the nozzle shield plugs. The cables needed to remove the missile beam were disposed of during decommissioning. 8.3.2. Controlled Areas The maximum radiation level in any controlled area is 20 mR/hr, and is present only in the upper sub-pile room (elevation 111 ft). This area is accessible only by passing through two locked bar-riers, each keyed differently, cind opening a metal door which has been welded closed. Remov-able contamination levels in any controlled area are less than 100 dpm/100 cm2, 8.3.3. Accessible Inspection Areas of the Exclusion Area All radiation levels in any accessible inspection areas are less than 1.0 mR/hr. All removable cointamination levels are less than 100 dpm/100 cm2 and all fixed contamination levels are less 2 than 500 dpm/100 cm. Access to these areas requires passing through at least two locked barriers. 8.3.4. Lic. aid Waste Area All radiation levels within the liquid radwaste area are less than 1.0 mR/hr. All removable con-2 tamination levels are less than 100 dpm/100 cm. Access to this area is through a locked door. 100

8.3.5. Remainir.g Portions of Main,uilt %x in areas of unrestricted access, all radiat el 'e less than 0.01 mRlhr above background. 2 1100 cm, fixed contamination levels are Removable contamination levels are less .c 0t less than 5000 dpm/100 cm. Areas outside ti exclusion areas wil! provide of fice and shop 2 f acil:Ges associated with Unit 2 and 3 activities. Control of these areas will be in accordance wtih the Peach Bottom Unit 2 and 3 operatmg license. 8.3.6. Miscellaneous Radiological Safeguards Entry into the controlled areas in the containment will be made in the presence of at least one supervisory representative of Philadelphia Electric Company. A policy has been established that the area within the exclusion area fences will not be used for any purpose. A perimeter security fence with guarded accesses surrounds the total area of Units 1,2 and 3. Units 2 and 3 are opera-tional and access t;/ the general public remains controlled. Since the decommissioned clant is within the Unit 2 and 3 complex, any radiological problem caused by fire, flood or storm damage will be acted upon immediately. 101

9. COST OF DECOMMISSIONING and Phase Ill. Circum-The main decommissiong work periods were referred to as Phase 1f stances prevented a continuous work flow during these phases, in addition, each phase was interrupted, Phase 11 more so than Phase 111. These incidents of stopping and starting extended the overall duration of work, which added to increased cost and a greater time span of cost inflation.

The total cost for Phase ll was $1,586,000 and for Phase lli $1,938,000, resulting in a grand The total cost for physical removal of components from the facility was total of $3,524,000. Fuel removal costs were accrued against the fuel owner and are not included in $1,517,000. decommissioning costs. Functional cost distribution is shown in Table 9-1. TABLE 91 PROJECT COST FUNCTIONAL BREAKDOWN COST $ Subcontracts 53,000 Engineering Management 547,000 j 1 PECO Supervision 70,000 PECO Craf t 900,000 Heating Oil 79,000 129,000 Insurance Auxiliary Power 254,000 Store Supplies 77,000 Purchase Order Supplies 296,000 Security and Operations Labor 985,000 PECO Travel and Per Diem 14,000 PECO Engineering 90,000 Undistributed Overhead 30,000 Total $3,524,000 102 f l

10. INSPECTION AND REPORTS 10.1. INSPECTION The physical barriers to unauthorized entrance into the facility, e.g., fences, buildings, welded doors, and access openings, will be inspected semiannually to ensure that these barriers have not deteriorated and that locks and locking apparatus are intact.

A facliity radiation survey will also be performed semiannually to verify that radioactive material is not escaping or migrating through the containment barriers in the facility. This inspection includes a survey of radiation levels and surface contamination, as well as air particulate activity in accessible areas of the containment vessel. The filter on the containment pressure equaliza-tion tube will also be changed semiannually and its radioactivity will be measured. Should work be required in the contro!!ed area of the containment, radiation and airborne radio-activity surveys will be performed before beginning, and all work "/ill be performed under appro. priate control to minimize radiation exposure of personnel and to prevent the release of radioactivity to the environment. All radiation surveys, tests, counting work and radiation exposure control measures will be done l according to written instructions and procedures that meet the requirements of the Peach Bottom Atomic Power Station Units 2 and 3 radiation control procedures. An environmental radiation survey will be performed as part of the Peach Bottom site environ-mental program as described in the Units 2 and 3 Technical Specifications. Security of the exclu-sion area will be under control of the shif t supervisor and site security personnel. The NRC will be notified of any abnormal occurrences such as (1) the entrance of an unauthori. zed person or persons into the f acility, or (2) a significant chan0e in the radiation or contamina-tion levelt in the facility or the of f site enviror ment. 10.2. RFPORTS An annual report will be made to the Director of Licensing, Washington, D.C., describing the re-sults of facility radiation surveys. Status of the facility and an evaluation of the performance of security and surveillance measures will also be covered. 1 i An abnormal occurrence report will be submitted to the Regulatory Operations Regional Office by telephone within 24 hours of discovery of an abnormal occurrence,The abnormal occurrence will also be reported in the annual report described just above. j i 103 -~

Records or logs relative to the following items will be kept and retained:

  • Environmental surveys
  • Facility radiation surveys
  • Inspections of the physical barriers
  • Abnormal occurrences.

I J I 104

APPENDIX A HISTORICAL PHOTOGRAPHS DECOMMISSIONING OF PEACH BOTTOM 1 l l l l Contract 35930 Philadelphia Electric Company

APPENDIX A CONTENTS Figure A 1 Construction Progress,1962 Figure A 2 Reactor Installation (3/26/64) Figure A 3 Refueling Floor During Operations Figure A 4 Clean Transfer Cask Adapter Plate Figure A 5 Early "O" Ring Placement l

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APPENDIX B CONTROL WORK PACKAGES DECOMMISSIONING OF PEACH BOTTOM 1 Contract 35930 Philadelphia Electric Company f } 1

PHASE 11 CONTROL WORK PACKAGES CWP Number Description 35930 001-002 Seal Main Helium Compressor Shaft Openings 35930-001 004 Removal of Oil Filter Cartridges from Lube and Seal Oil Systems 35930-001 005 Removal of Oil Filter Cartridges from the Oil Absorber A 402 35930-002-001 Chemical Cleaning of Control Rod Drives l 35930 006-001 Cut and Cap E.C.W.S. at Containment 35930-007 001 Remove Water Cooled Delay Beds 35930 007-002 Remc/o Low Temperature Delay Beds 35930-007-003 Remove Dust Collectors 35930-007 005 Remove LN2 Traps 35930-009-002 Chemical Mixing Tanks 35930-010 004 Waste Disposal Drain Tank 35930 010 005 Solidification of Tritiated Water 35930-010 006 Solidification of Contaminated Oil l 35930 017-001 Cut and Cap Helium Make up Bottles to Nonpurified Helium System 35930-018-001 Removal of Filter Cartridges from Purified Helium Compressors 35930 018 002 Remove Pump Down Plateout Absorber 35930 018-003 Remove Charcoal f rom Filters F 304 and F 305 1 35930-018-004 Removal of Dust Filter F 303 35930-019-001 Removal of Steam Generator Purge Plateout Trap 35930-025 003 Seal Vessel Safety and Primary Coolant Relief Lines 35930-025 012 Cap Miscellaneous Penetrations 35930-026-002 Erection of Monorall 35930-027-001 Cut and Ship Hallam Cask Support Structure 35930 027 002 Ship Radioactive Waste 35930 024-010 Decontaminate Spent Fuel Pit Access Area 35930-024-011 Decontaminate and Deactivate Cask Traveling Hoists 35930 024-012 Deactivate Spent Fuel Grapple Crane and Decontaminate the Spent Fuel Grapple Crane 35930-024 013 Remove Spent Fuel Pit Tools 35930-025 001 Survey and Decontaminate Inspection Access Areas B1

PHASE Ill CONTROL WORK PACKAGES CWP Number Description 35930-003 001 Shield Cooling System - Cut and Remove Pipe 35930 003 002 Remove Shield Cooling System Tank T-58 35930-008-001 Exhaust Plenum Oil Filter Removal 35930-008 002 Exhaust Filters Removal 35930 009 001 Decontaminate System - Install Blank Flange 35930 010-001 Removing the Contaminated Tanks and Equipment of the Liquid Waste System 35930-010 002 Liquid Waste Disposal System - Remove System Pipir.g 35930 010-003 Decontaminate Liquid Waste Building and Liquid Waste Sump 35930 011 001 Ventilation System - Removal of Exhaust Filters 35930 011 002 Ventilation System Ducting Removal 35930 011-003 Ventilation System - Weld Covers to Ducts 35930-011 004 Ventilation Stack - Survey and Decontaminate 35930-016 001 Decontamination of New Fuel Storage Area 35930 020-001 Remove Spent Fuel Pit Filters 35930-020 002 Remove Spent Fuel Pit Heat Exchanger 35930-020 003 Remove Spent Fuel Pit Cooling Water Pumps 35930-020 004 Remove Spent Fuel Pit Cooling System External Piping 35930-020 005 Cap and Weld Piping on Spent Fuel Pit Cooling System 35930 021 002 Containment Equipment Cooling Water System Removal 35930 022 001 Cutting and Capping of Containment Hot Water Heating System Piping 35930-023 001 Administration Building Laboratory and Laundry Drains and Exhausts 35930-024 001 Decontaminate the Walls and Floor of the Spent Fuel Pit 35930-024-002 Remove Spent Fuel Pit Sump Pump 35930-024 003 Decontaminate Spent Fuel Pit Sump 35930-024 004 Remove Spent Fuel Pit Tube Spool Piece 35930-024 006 Install Cap on Spent Fuel Pit Tube 35930-024 007 Remove Spent Fuel Pit Tube and Spent Fuel Elevator 35930-024 008 Installation of Steel Grating over Spent Fuel Pit 35930 024 009 Weld Cask Monorail Door Closed B2

PH ASE lli CONTROL WORK PACKAGES (Continued) CWP Number Description 35930 025-002 Decontaminate the Containment Sump 35930-025 004 Decommission Canning and Charging Machines 35930 025-006 Decommission Transfer Machine 35930-025-007 Decommission Transf er Cask 35930-025-009 Decommissioning of isolation Valves Installation of the Absolute Filter and Welding the Equipment Door and 35930-025 010 Personnel Escape Lock Outer Door Closed 35930-025-011 Decommissioning the Personnel Air Lock and Access Lock 35930-026-001 Erection of Gates and Barricades 35930-028-001 Cut and Cap Miscellaneous Penetrations i 35930-028-002 Remove Check Sources From Radiation Monitors 35930 028 003 Remove Post incident Filters and Vent Purge Plenum Recirculation Filter 35930 028-004 Cut and Cap Lines at the Head of Steam Generator No.1 35930-028 005 Liquid Waste Disposal System - Decommission System Drains and Embedded Pipes B3

APPENDIX C HAZARDS EVALUATION FOR EMBEDDED PIPES AT PEACH BOTTOM UNIT 1 l l Contract 35930 Philadelphia Electric Company Work Performed by: Prepared by: J.P. Andrews Catalytic, Inc. E. E. Clements 1500 Market Street J. Schmidt Philadelphia, PA 19102

1.0 INTRODUCTION

This document discusses the potential hazards associated with leaving contaminated embedded piping in place in the radwaste basement of Peach Bottom Unit 1. The Decommissioning Plan and Safety Analysis Report for Peach Bottom Atom 1 Power Station Unit 1 originally specified that all contaminated material and piping in the Liquid Radwaste Area would be either decontaminated or removed and shipped of f site to a licensed burial facility. I A rumber of embedded pipes could not be successfully decontaminated and, since the levels of contamination are only slightly above the permissible levels, a decision was made, based on this analysis, to leave the pipes in place. Of the eighty two pipes and drains which are embedded in the walls and floors of the radwaste I basement, eleven remain contaminated to levels exceeding that acceptable for release for un-restricted use following extensive decontamination efforts including mechanical cleaning, acid leaching and hydrolaser (6,500 psi) cleaning. A summary of the radiological status of the remain-ing pipes is shown in Table 1. 2.0 REVIEW OF HAZARDS 2.1 Total Quantity of Material The quantity of material is estimated based on the measured contamination levels and on the surface area of the pipes. The estimated quantity is about 8 microcuries, separated into two sources of 5 and 3 microcuries each. These quantities are listed by nuclides in Table 11 and are exempt from licensing requirements of Part 34 and from the labeling requirements of Part 20.304. This is based on the fractional quantity of each nuclide present in the mixture of nuclides ex-pected to be present. In all cases, the total f ractional quantity of each nuclide is less than 1. 2.2. Type of Material Remaining The remaining contamination in the pipes is in the form of an adherent film that was not re-moved by the decontamination efforts used. These efforts included the following in order; (a) Acid leaching and flushing using sulfuric acid to remove or loosen organic and acid soluble coatings. (b) Hydrolaser cleaning using 6,500 psi water nozzle to clean adherent scale and crud accumu-lations in the piping. The results of these cleaning activities, as shown in Table I,is that the removable contamination is very low and the remaining contamination, although above the release limits, is relatively low. All of the pipes in the floor and equipment drains are Type 304 Stainless Steel. Pipe I 33 is Car-bon Steel. C-1

2.3 Degree of Continement The contaminated piping is embedded in the radwaste floor and in the structural concrete of the building. (The accessible sections of the adjacent pipe runs were removed).The drains and traps of the embedded floor drains were removed to eliminate the accumulated activity at the traps, to permit contamination surveys of representative sections of the imbedded pipe, and to permit decontamination. Following final surveys, the contamination was evaluated and the pipes either capped, or plugged and covered with concrete to bring the surface back to the floor elevation. The pipes are. therefore, either capped with welded caps or sealed by concrete. Access to the piping may only be obtained by mechanical removal of the concrete or removal l of the welded caps. 2.4 Physical Security The pipes embedded in the basement floor do not require physical security measures beyond that provided by the welded caps and concrete covering. However, the current plan is to pro. vide locked access to this area similar to that provided to the accessible areas within the con-tainment and in the fuel pool area. 2.5 Susceptibility to Release of Radiation f 2.5.1 Flooding l The worst condition for release of the radioactive material into water would be by simultaneous release of all contamination contained in the embedded pipes into a small volume of water. If all of the material contained in the embedded pipes was released simultaneously into the Rad-waste Basement Sump, which has a capacity of 342 gallons, the resulting concentration would be 6.3 x 10-6 pCi/ml. For cesium-137, this concentration is less than the allowable concentration for unrestricted release. 2.5.2 Airborne Radioactivity The worst condition for release of the radioactive material into the air would be by simultaneous relcase of all contamination contained in the embedded pipes into the volume of air in the Liquid Radwaste Area (28,000 f t ). The resulting concentration would be 1.0 x 10 8 pCilml. For cesium-3 137, this concentration does not exceed the allowable concentration permitted for occupational airborne radioactivity exposure for a forty hour week. 2.5.3 Whole Body Radiation The total amount of the radioactive material in the embedded pipes in the Liquid Radwaste Area is 8.14 Ci, primarily cesium 137. If all of this radioactive material was concentrated into a point source, the whole body radiation exposure rate in air at 1 foot would be 0.03 mR/hr. For any other source geometry, the exposure rate will be less than 0.03 mR/hr. C-2

2.6 Required Surveillance Because the amounts of the materials in the contamination are exempt from licensing require-ments of 10 CFR Part 34 and from the labeling requirements of 10 CFR Part 20.304, no surveil-lance is required. 3.0 ALTERNATIVES CONSIDERED 3.1 Removal of Embedded Pipes 3.1.1 Radwaste Basement Floor There are approximately 105 feet of embedded piping under the radwaste basement floor. This floor is reinforced concrete and the pipes are embedded to a depth of about 1 foot. Experience was gained in the removal of the piping from this matrix when the traps were removed. The con-crete was very difficult to remove. An average rate of 2 to 3 feet of pipe could be removed per j manday. Based on this experience, the complete removal of these pipes was considered to be uneconomical. 3.1.2 Pipe 133 Pipe 133 is the resin sluice pipe from the fuel pool area to the radwaste basement. It is em-bedded in a horizontal structural concrete beam for shielding purposes. It is cor,sidered highly undesirable to remove this pipe at this time with the consequent structural damag' to the build-ing. It was, therefore, considered uneconomic to remove this pipe. 3.2 Additional Decontamination The alternative of recalling the hydrolaser contractor was considered. A significant scheduling delay would result if this course were taken with no significant guarantee that fixed contamina-tion levels would be reduced further. Acid leaching was also considered. Handling acids is hazardous and additional cleaning seemed unwarranted based on the risk to the workers and the potential to reduce contamination further. 4.0 RECOMMENDED SURVEY POINTS A survey point should be established at the low point for all pipes and should be checked for contamination as a regular part of the six month regultr inspection of the decommissioned plant. This survey pcint should be in the radwaste sump area. C-3

TABLE C 1 CONTAMINATED PIPES IN LIQUID RADWASTE AREA Contamination Total

Level, Radioactive 2

Material Pipe I.D. Pipe Length 04 dpm/100 cm Pipe No. (Inches) (Feet) Average Maximum Removable (Microcuiles) l l33 1.5 80 24,000 24,000 100 3.18 D 11 2 12 37,000 60,000 300 .96 D-12 2 6 9,400 12,000 2,700 .13 D-24 2 4 8,700 17,000 860 .08 D 27 2 12 24,000 45,000 230 .64 D 29 2 9 22,000 23,000 100 .43 D 32 2 4 5,500 8,200 110 .05 D 33 2 4 9,000 12,000 200 .08 l D 37 3 12 11,000 17,000 580 .43 i l RW2 3 30 19,000 46,000 680 1.88 RW4 4 12 5,400 7,100 160 .28 TOTAL 8.14 Acceptable surf ace contamination levels are: axirku 000 0-7 dprn/100 cm2, Removable 1,000 dpm/100 cm2 C4

l TABLE C 2 ISOTOPIC CONTENT OF CONTAMINATED PIPES IN LIQUID RADWASTE AREA Exemnt Quantity (pCI) l Isotope % of Total p Ci of isotope 10 CFR % of Exempt Quantity l PIPE l 33 134Cs 5.2 0.16 1 16 137 s 92.5 2.95 10 29.5 C 95 r 0.6 0.02 10 0.2 Z 60 o 1.6 0.05 1 5 C 90 r 0.2 0.006 0.1 6 S Total %of Exempt Quantity 56.7 i PIPES CONNECTED TO RADWASTE SUMP 134Cs 5.2 0.26 1 26 137Cs 92.5 4.59 10 45.9 95 r 0.6 0.03 10 0.3 Z 60 o 1.6 0.08 1 8 C 90 r 0.2 0.009 0.1 9 S Total %of Exempt Quantity 89.2 C5

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