ML12145A490

Updated Final Safety Analysis Report

ML12145A490
Person / Time
Site:	Peach Bottom
Issue date:	04/30/2012
From:	Exelon Nuclear, Exelon Generation Co
To:	NRC/FSME/DWMEP
References
U000DOCKET
Download: ML12145A490 (73)

Text

U000DOCKET 50-171 UPDATED FINAL SAFETY ANALYSIS REPORT PEACH BOTTOM ATOMIC POWER STATION UNIT 1 Revision 7, April 2012

NOTE: Chapters and sections marked as Historical are to be used as information only. Other chapters and sections not marked as Historical comprise the safety analysis and have been updated to include all changes made in the facility and to incorporate the conclusions of all safety evaluations since July, 1974. The Updated Final Safety Analysis Report (UFSAR) will be revised on a replacement-page basis to include the effects of all changes made in the facility or procedures as described in the UFSAR; all safety evaluations performed in support of requested license amendments or in support of conclusions that changes did not require prior NRC approval pursuant to 10 CFR 50.59; and all analyses of new safety issues performed at NRC request.

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TABLE OF CONTENTS Page No.

1. Introduction ...................................................... 1

2. Decommissioning Summary ......................................... 2-5

3. Decommissioned Plant Description .................................. 6 3.1 Disposition of Special Nuclear Material ...................... 6 3.2 Disposition of By-Product Material ......................... 6-9 3.3 Exclusion Area .............................................. 10 3.4 Containment Vessel ....................................... 10-14 3.5 Fuel Pool Building ....................................... 14-15 3.6 Liquid Radwaste Area ........................................ 15 3.7 Administration Building ..................................... 16 3.8 Remaining Portions of the Main Building Complex .......... 16-17 3.9 Incidental Tanks and Buildings .............................. 17 3.10 Decommissioned Status of Systems and Components .......... 18-27 3.11 Inventory of Radioactive Materials Left on Site .......... 27-30

4. Decommissioned Plant Safety Analysis ............................. 31 4.1 Thermal Analysis of Heat Generated by Activation Product Decay ............................................ 31-37 4.2 Site Flooding ............................................ 38-44 4.3 Containment Pressure Transient ........................... 45-47 4.4 Safeguards and Radiological Safety ....................... 48-57 Revision 7, April 2012

TABLE OF CONTENTS (Con't)

Page No.

5. Radiological Safety During Decommissioning ....................... 58 5.1 Responsibilities ............................................ 58 5.2 Personnel ................................................... 58 5.3 Regulations ................................................. 59 5.4 Procedures ............................................... 59-61 5.5 Radiological Surveys ........................................ 62 5.6 Emergency Plans .......................................... 62-63 5.7 Conclusions ................................................. 63

6. Administration ................................................... 64 6.1 Licensing ................................................ 64-65 6.2 Manning ..................................................... 65 6.3 Records ..................................................... 66 6.4 Inspection ............................................... 67-68 6.5 Reports ..................................................... 69 Appendix A - Technical Specifications Revision 7, April 2012

1. INTRODUCTION This document describes the plan for decommissioning the Peach Bottom- 1 Atomic Power Station, and presented a safety analysis which demonstrated that the facility was placed in a status which is not hazardous to the health and safety of the public. Fuel was removed from the reactor and shipped to an offsite storage facility (Idaho National Labs). The decommissioning was completed, and a Part 50 Operating License exists and includes periodic inspections of the facilities within the Exclusion Area*

to assure that the decommissioned facility will not be hazardous to the health and safety of the public.

• Definitions:

Exclusion Area - The area of the facility to which access will be restricted by a locked enclosure.

Controlled Area - The area within the Exclusion Area in which access is restricted by appropriate locked barriers and in which radiation levels exceed 1 mr/hr or contamination levels exceed those acceptable (as defined in Figure 3.2-1) for release for unrestricted use.

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NOTE Chapter 2 is considered Historical information and involves planning for placing Peach Bottom Atomic Power Station, Unit 1 in its present SAFSTOR status.

2. DECOMMISSIONING

SUMMARY

The Peach Bottom - 1 Atomic Power Station will he decommissioned by the Philadelphia Electric Company. Residual activity Contained within the Exclusion Area (which will be created as part of the decommissioning plan) will be licensed under a Part 50 Possession Only license. Philadelphia Electric Company will be solely and completely responsible for the residual activity and for the provisions of the Part 50 Possession Only License, The Peach Bottom - I decommissioning and the final decommissioned plant configuration will resemble that of the Carolinas Virginia Tube Reactor, the Pathfinder Reactor and the Saxton Reactor. All residual activity will be contained within the Containment and Spent Fuel Pool Buildings. Within the Containment Building, more than 99 percent of the estimated 3 megacuries of activity will be contained inside of the- reactor vessel in the form of induced activity in the vessel walls, reactor internals and control rod couplings. The reactor vessel will be contained inside the reactor vessel cavity and is accessible only by removing the concrete missile shields, the refueling port flanges and the refueling port shield plugs. The missile shields can only be removed with the building crane which will be electrically deactivated once decommissioning is completed.

The delay beds, various filters and traps and any contaminated equipment or piping removed from the facility will be transferred to Units 2-3 or will be shipped off-site to either a licensed facility for post-radiation experiments or a licensed burial facility. All such shipments will be made in accordance with applicable AEC/DOT regulations. No contaminated equipment removed from areas outside the Exclusion Area will be stored within the Exclusion Area.

The decommissioning will be accomplished using personnel from Philadelphia Electric and, where appropriate, Personnel from a Private contractor under direction of Philadelphia Electric. All decommissioning activities will be carried out under the existing Part 50 license, DPR-12, or a Part 50 Possession Only license. Where appropriate, written procedures, approved by Revision 7, April 2012

Philadelphia Electric Company will be used for any decommissioning work which could affect the nuclear safety of the plant, could result in release of activity or could result in significant radiological hazards to personnel. Philadelphia Electric Company will be responsible for all decommissioning activities including those performed by contractors.

The general decommissioning plan will involve the unloading, canning and transfer of fuel to the Spent Fuel Pool. Defueling could start approximately eight weeks after reactor shutdown and is expected to take from 20 to 40 weeks depending on whether the operation is performed on a one or two shift per day basis. During the defueling period, preliminary work on decommissioning systems and components which do not affect nuclear safety will be started. (Typical work will involve layup of the turbine-generator, decontamination of the new fuel storage area, etc.) Once the reactor is defueled and during the fuel shipping period, additional systems including reactor related systems will be decommissioned and the Proposed Post Operational Sampling Program, if to be done, will be completed. The fission product trapping system delay beds will be degassed and the gases released in accordance with the facility Technical Specifications. (Until all the fuel has been accepted at the fuel reprocessing facility, the containment isolation systems will be kept operable so that fuel can be brought back into the containment for recanning if required.)

Once the fuel has been accepted at the reprocessing facility, the remaining plant systems will be decommissioned (see Figure 5.4-1 for the expected schedule for decommissioning activities). As part of this work, the Spent Fuel Pool will be drained, decontaminated and prepared for inclusion in the Exclusion Area.

The accessible areas of the Exclusion Area will be decontaminated. All contaminated equipment and piping outside the Exclusion Area Will either be decontaminated or removed for disposal at an Off-site licensed burial facility. All contaminated areas outside of the Exclusion Area will be decontaminated.

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Near the end of the decommissioning operation, the radwaste facility will be drained, partially dismantled and decontaminated. All solid waste from the decommissioning operation will be shipped off-site for disposal at a licensed burial facility.

An exclusion fence will be installed around the Containment and Spent Fuel Pool Buildings as shown in Figure 4.4-2 to establish the Exclusion Area.

The existing perimeter fence around the entire Peach Bottom facility (Units 1, 2 and 3) will be maintained so that access to the site will be controlled by Philadelphia Electric Company.

Entry to the Exclusion Area will require approved entry through the perimeter fence, and unlocking a posted gate in the Exclusion Area Fence (either on Elevation 116'-0" or Elevation 176'-6")*. Entry to the Containment Building or the Spent Fuel Pool Building will require the use of an additional key to unlock the doors to those buildings. Entry to the Controlled Areas, where radiation levels may be greater than 1 mR/hr will require an additional key to open the grating over the southwest stairwell in the Containment. Entry to the High Radiation Area around the reactor vessel will require a physical restoration of the electrical supply to the containment crane to move the three-feet thick concrete missile beams.

The Technical Specifications of the plant will be reduced in steps which are consistent with the continuing health and safety of the public and the plant staff, but which are practical in the sense that they are applicable to the changing plant conditions. A revised set of Technical Specifications are included in this report (Appendix A) which are proposed for use throughout the decommissioning. These Technical Specifications are similar to those used by the Carolinas Virginia Nuclear Power Associates and those used by the Saxton Nuclear Experimental Corporation during the decommissioning of the CVTR and the Saxton reactors, in that they phase out certain requirements of License DPR-12 at key points in the decommissioning.

• Elevations refer to Conowingo Datum (C.D.)

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Philadelphia Electric requests that a decommissioning authorization and an amendment to License DPR-12 be issued simultaneously which will place the facility under the revised Technical Specifications (Appendix A) when the reactor is shut down* for decommissioning and will place the facility in Part 50 Possession Only status when all the fuel is removed from the reactor and it is not to be refueled. Following this authorization, the decommissioning will proceed as described in this plan.

• The reactor will be considered to be shutdown when at least 52 of the 55 control rods are fully inserted in the core, the reactor is kept in the Low Pressure Mode, the outlet temperature is less than 500ºF and the decay heat is less than 1 Mw. (The Low Pressure mode of operation will insert a 0.01% scram from the intermediate range channels of the reactor protection system.)

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NOTE Chapter 3 (except for Figure 3.2-1) is considered Historical information and involves planning for placing Peach Bottom Atomic Power Station, Unit 1 in its present SAFSTOR status.

3. DECOMMISSIONED PIANT DESCRIPTION 3.1 Disposition of Special Nuclear Material All fuel will be removed from the reactor and canned; and then shipped to a fuel reprocessing facility off-site after the cooling time required for shipment.

3.2 Disposition of By-Product Material Essentially all by-product material with the exception of that contained within the primary system, within the annulus of the fuel pool and low level residual surface contamination in Controlled Areas will be removed from the Peach Bottom Unit 1 facility.

3.2.1 Certain components of the primary system (such as certain primary coolant isolation valves, samples of the primary piping, parts of one of the primary coolant compressors and steam generator internals) may be removed as part of a proposed Post Operative Sampling Program, packaged in appropriate shipping containers, and shipped to licensed laboratories for analysis.

3.2.2 All drainable liquids (water, oils and refrigerants) will be drained from the plant systems and disposed of.

All liquids which are contaminated, including all liquid waste generated during the defueling and decommissioning operations, will be processed in the radwaste facility.

The resulting liquids, will be discharged (in the case of low activity water) in accordance with plant Technical Specifications, or will be converted into solid waste for off-site licensed burial. Contaminated liquid wastes generated after the Unit 1 radwaste system is decommissioned will be converted into solid waste or will be transported to Unit 2 for processing. There will be no liquid radioactive wastes (other than Revision 7, April 2012

residual liquids in drained systems) stored in the decommissioned plant.

3.2.3 All radioactive solid waste such as ion exchange resins, filter socks, solidified liquid waste, contaminated equipment and trash resulting from the defueling and decommissioning work will be packaged in appropriate radioactive waste containers and will be shipped off-site to a licensed burial facility for disposal.

3.2.4 All gaseous activity, including noble gas activity desorbed from the delay beds, will be collected in a holdup tank prior to release. The gas in the holdup tank will be sampled and the gas will then be released under controlled conditions and in accordance with Technical Specifications through the plant stack. There will be no radioactive gaseous waste stored in the decommissioned plant.

3.2.5 Components and piping in systems which are located outside the Exclusion Area (containment and fuel pool buildings) and which have become contaminated will be decontaminated to less than the levels given in Figure 3.2-1 or will be dismantled, removed, packaged for shipment and shipped to an off-site licensed burial facility.

3.2.6 The primary system will be sealed to prevent escape of the radioactivity contained within the system. All openings which may be made as part of the post operative sampling program will be seal welded closed.

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3.2.7 Approximately 3 mega curies of activity will remain within the Peach Bottom 1 facility when the decommissioning is completed. More than 94 percent of this activity exists in the form of induced radioactivity in the irradiated reactor vessel and metal vessel internals. (See Figure 3.11-1) The remaining major fraction of activity is accounted for by 0.16 mega curies in the Stellite springs on the control rod couplings (which will also be contained within the reactor vessel) and by 30 Ci of surface contamination within the primary system. The activity remaining in the annulus of the fuel pool is estimated to be 0.1 mCi.

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Figure 3.2-1 ACCEPTABLE SURFACE CONTAMINATION LEVELSf a b,c b,d b,e NUCLIDE AVERAGE MAXIMUM REMOVABLE U-nat, U-235, U-238, and 5,000 dpma/100 cm2 15,000 dpma/100 cm2 1,000 dpma/100 cm2 associated decay products Transuranics, Ra-226, Ra-228, Th-230, Th-228, Pa- 100 dpm/100 cm2 100 dpm/100 cm2 20 dpm/100 cm2 231, Ac-227, I-125, I-129 Th-nat, Th-232, Sr-90, Ra-223, Ra-224, U-232, I-126, 1,000 dpm/100 cm2 3,000 dpm/100 cm2 200 dpm/100 cm2 I-131, I-133 Beta-gamma emitters (nuclides with decay modes other than alpha emission 5,000 dpm -/100 cm2 15,000 dpm -/100 cm2 1,000 dpm -/100 cm2 or spontaneous fission) except Sr-90 and others noted above.

a Where surface contamination by both alpha- and beta-gamma-emitting nuclides exists the limits established for alpha- and beta-gamma-emitting nuclides should apply independently.

b As used in this table, dpm (disintegrations per minute) means the rate of emission by radioactive material as determined by correcting the counts per minute observed by an appropriate detector for background, efficiency, and geometric factors associated with the instrumentation c

Measurements of average contaminant should not be averaged over more than 1 square meter. For objects of less surface area, the average should be derived for each such object.

d The maximum contamination level applies to an area of not more than 100 cm squared e

The amount of removable radioactive material of surface area should be determined by wiping that area with dry filter or soft 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 proportionally and the entire surface should be wiped.

f Figure 3.2-1 taken from Regulatory Guide 1.86 Termination of Operating Licenses for Nuclear Reactors, June 1974 Revision 7, April 2012

3.3 Exclusion Area The area which includes the Containment and Spent Fuel Pool Buildings will constitute the decommissioned exclusion area.

This area will be enclosed by an eight foot high security fence with locked gate as described in Section 4.4.

3.4 Containment Vessel 3.4.1 Elevation 176'-6" (Refueling Floor) of the Containment will be made accessible for periodic inspections through the access lock on that elevation. Accessible areas on or above 176'-6" will be decontaminated to levels less than the release limits given in Figure 3.2-1.

Equipment with internal contamination levels higher than the release levels will be sealed with blank flanges to enclose the activity. Radiation levels within the accessible area will be less than 1 mR/hr. Access to areas below 176'-6" will be controlled as described in Section 4.4.

3.4.2 Elevation 90' in the vicinity of the Containment Sump will be made accessible for periodic inspections through the personnel lock on elevation 116'-0". Access from the personnel lock to the containment sump is controlled as described in Section 4.4. All areas within the access path will be decontaminated to levels less than those given in Figure 3.2-1. The containment sump itself will be decontaminated to as low a level as practicable with the intent of achieving levels within the Figure 3.2-1 release limits. Radiation levels within the access path will be less than 1 mR/hr.

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3.4.3 The control rods, permanent reflector blocks, hexagonal reflector elements and dummy spacer assemblies will be left in the reactor vessel. (Several control rods may be removed as part of the post operative inspection.

Any rods removed may not be replaced in the reactor vessel. If not replaced, they will be shipped offsite to a licensed laboratory for analysis or to a licensed burial facility for disposal).

The control rod drive systems (control and Emergency Shutdown) will be left intact in the fully inserted position. Oil will be drained and accumulator gas vented from the hydraulically operated rods. Power will be disconnected from the electrically driven rods. The drive exteriors and the sub-pile room will be cleaned if necessary to eliminate any fire hazards from residual oil.

After all fuel is removed from the reactor vessel, the helium cooling system will be shut down and in-vessel temperatures will be monitored by using existing thermocouples to measure the actual equilibrium temperatures produced by activation product decay. Once acceptable, equilibrium temperatures are established, helium will be displaced with nitrogen as discussed in Section 3.10.1.

3.4.4 The shield plugs and blind flanges will be installed on the refueling Ports. Caps will be welded onto the discharge lines from the vessel safety valves and the relief valves on the primary coolant piping loops. The three foot thick concrete missile shields will be installed over the reactor vessel. The Reactor jib Crane and Reactor Service Crane will be positioned so that they cannot be used to move any of the shield plugs. Power to these cranes will be disconnected.

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3.4.5 The refueling equipment will be decommissioned by:

a. Canning Machine The isolation valve at the top of the machine will be closed and secured. Penetrations to the machine cavity will be sealed.

b. Charge Machine The valve at the bottom of the machine will be closed and secured. The machine will be positioned over the canning machine and the gas lock sleeve will be lowered.

c. Transfer Machine A blank flange will be installed on the bottom of the machine and the machine stored in the refueling area.

All extra grapples and equipment for the Transfer Machine will be decontaminated, shipped offsite to a licensed burial facility or placed in a sealed container.

d. Transfer Cask The valve at the bottom of the cask will be closed and secured. A blank flange will be installed on the bottom of the cask and the cask will be stored in the refueling area.

e. Viewing Device The viewing device will be secured to a Fuel Handling Equipment Storage Port.

f. Isolation Valves Blank flanges will be installed on the topside of the Isolation Valves and the valves will be secured to Fuel Handling Equipment Storage Ports.

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3.4.6 The noble gas activity will be purged from the delay beds and released under controlled conditions (after sampling) through the stack. The purge condenser, water-cooled delay beds, and low temperature delay beds will be removed and shipped off-site to a licensed burial facility. All filter cartridges and granulated charcoal beds will be removed as cartridges or as an entire vessel and will be shipped off-site for disposal.

(All granulated charcoal, and filter cartridges will be removed to minimize potential fire hazards.) Because the fission product inventory experienced during actual operating conditions is low and because of the resulting low decay heat generation rate, neither purge flow or cooling water is required to prevent overheating of the delay beds. Therefore, the emergency cooling system for the water cooled delay beds will be removed from service soon after reactor shutdown to permit decommissioning work on this system early in the schedule.

3.4.7 All piping connections which penetrate the containment will be severed capped and welded outside the containment, except for the containment vacuum breaker assembly and associated penetration. Electrical penetrations will be left as is. The ventilation supply and exhaust ducts will be severed and welded shut outside the containment. A six inch pressure equalization line equipped with a replaceable absolute filter will be installed on the containment to prevent a pressure differential from developing between the inside and outside of the containment. In addition, the containment vacuum breaker assembly was left as-is per NCR 95-00043.

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3.4.8 The outer door of the emergency air lock and the equipment door on Elevation 116' will be welded closed.

The inner doors of the personnel air lock on Elevation 116' and the access lock on Elevation 176'-6" will be welded in the open position. The interlock mechanisms in these two latter locks will be disabled and the outer door on each lock will be secured with a heavy duty padlock.

3.4.9 Flammable materials other than electrical cables and solid graphite left in the sealed reactor vessel will be removed from the containment.

3.5 Fuel Pool Building 3.5.1 The fuel pool will be drained to radwaste and the fuel pool walls and floor decontaminated. The fuel grid will be decontaminated and released, or will be removed and shipped offsite for disposal at a licensed burial facility. The Fuel Pit Sump Pump will be removed and the sump decontaminated. The spool piece in the Spent Fuel Pit Tube will be removed, the Spent Fuel Elevator will be removed and the blank flange will be welded onto the containment end of the Spent Fuel Pit Tube. The Spent Fuel Pit Tube outside of containment will be decontaminated or removed. (All equipment and tube sections removed will be shipped off-site to a licensed burial facility for disposal).

3.5.2 The fuel pool will be covered with steel mesh similar to fence wire. This wire will be fastened or welded to prevent easy entry to the pool, but will permit visual inspection of the pool during periodic inspections. The doors which permit the cask monorail to move the cask from the pool to the cask storage area will be welded -

closed. The personnel door on Elevation 116'-O" will be closed and locked. This personnel door will be used for access for periodic inspections. The area from this Revision 7, April 2012

door up to and around the fuel pool will be decontaminated to levels within the Figure 3.2-1 release limits.

3.5.3 The Spent Fuel Grapple Crane and the Spent Fuel Cask Traveling Hoist will be decontaminated and retired in place. Power for the crane and hoist will be disconnected, 3.5.4 Fuel Grapples and tools Used in the fuel pool will be decontaminated Or will be packaged for shipment off-site to a licensed burial facility for disposal.

3.6 Liquid Radwaste Area After most of the facility decommissioning is completed, the Liquid Radwaste Area will be decommissioned.

3.6.1 Filter socks, filter media, spent resins and all contaminated radwaste equipment will be removed, packaged for shipment and shipped off-site to a licensed burial facility for disposal. Piping (including waste discharge piping and waste piping between the containment and the radwaste facility) will either be decontaminated to levels less than the Figure 3.2-1 release levels or will be removed, packaged and shipped off-site to a licensed burial facility.

3.6.2 The radwaste facility will be decontaminated to levels less than the Figure 3.2-1 release levels, and will be released for unrestricted use.

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3.7 Administration Building 3.7.1 The drains and exhaust systems in the laboratory and laundry will be decontaminated to levels less than the Figure 3 .2-1 release limits or will be removed and shipped off-site for disposal at a licensed burial facility. All radioactive material (such as check sources) and all instrumentation and supplies will be removed from the health physics and instrument repair facilities, the laboratory, the counting room, and the storage vault; and taken to Unit 2.

3.7.2 Any contaminated portions of the building will be decontaminated to levels less than the Figure 3.2-1 release limits. Surveys will be made to assure that whole body radiation levels in the building are less than 0.04 mrem/hr. The building will be released for unrestricted use. The electrical system for this building will be left operational or will be left in a condition where it can easily be reactivated so that the building can be used in the future for storage or as office space. (All utility services required for this building will be arranged so that they are independent of the Exclusion Area portion of the facility).

3.8 Remaining Portions of the Main Building Complex 3.8.1 All systems in the remaining portions of the Turbine and Auxiliary Buildings will be drained. All oil or other flammable material will be removed.

3.8.2 The new fuel storage area will be decontaminated if required. The exhaust ventilation duct, if contaminated, will be decontaminated or removed for shipment off-site to a licensed burial facility.

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3.8.3 The ventilation stack will be Surveyed and if contaminated will be decontaminated or will be dismantled and shipped off-site for disposal at a licensed burial facility.

3.8.4 The Shield Cooling System will be decontaminated or dismantled and shipped off-site for disposal.

3.8.5 All sources stored in the Source Storage Vault will be transferred to other licensed facilities.

3.8.6 Surveys will be made to assure that contamination levels on equipment and/or building surfaces are less than the Figure 3.2-1 release limits and that whole body radiation levels are less than 0.08 mrem/hr. The building areas and equipment will be released from future controls.

3.8.7 Equipment in these areas may be dismantled for salvage or scrap or may be retired in place. Certain key items such as the turbine-generator and the diesel-generator will be removed to other facilities or will be laid up so they can be salvaged at a later date.

3.9 Incidental Tanks and Buildings The incidental tanks and buildings will be surveyed to assure they meet the Figure 3.2-1 limits for release and that whole body radiation levels are less than 0.08 mrem/hr. These tanks and buildings will be razed to the degree which is practicable to reduce on-going maintenance for cosmetic purposes.

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3.10 Decommissioned Status of Systems and Components At the completion of decommissioning, the status of the various systems will be:

3.10.1 Primary Helium System

a. Proposed Post Operational Sampling Program Helium in the primary system will be displaced by two system volumes of nitrogen to carry the helium through the delay beds and into the storage tanks.

The primary helium system will be isolated from the purification system with caps welded over the severed intertie lines. This isolation will be done to permit the sampling program, if it is to be done, to begin on the main coolant loop while the purification system is still being used to degas the delay beds).

In the number 1 loop, the hot valve and portions of the concentric pipe and helium return line will be removed. All resulting open pipe ends will be covered with welded closures. The number 1 steam generator tube bundle or portions of it will be welded closed. The number 1 helium circulator impeller and turning vanes will be removed. All openings made in the circulator case, including the shaft opening, will be welded closed. Approximately 80 trepanned samples will be taken from the primary piping in loops 1 and 2. All holes produced will be covered with welded plates.

b. Additional Decommissioning The shaft penetration on the number 2 helium circulator will be seal welded. All discharge pipes from the reactor vessel safety valves and the coolant piping relief valves will be closed with welded caps.

Thus, the primary system will be sealed off so that all residual activity is inside it.

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The seal oil and lube oil will be drained from the helium circulator oil systems and the filter cartridges will be removed from the filters in these systems to reduce fire hazard potential. Cleaning will be performed in the oil system areas of the containment to remove residual oil from the outer surfaces of the equipment and from the floor, in order to eliminate a potential fire hazard.

3.10.2 Hydraulic Control Rod Drive System As discussed in Section 3.4.3, the control rods will be left in the reactor vessel, and all drives will be deactivated in the fully inserted position. When the hydraulic fluid is drained from the drives, the fluid will also be drained from the reservoir and Pumps. The outside of the reservoir and Pumps and the floor area around them will be cleaned (along with the drives and the subpile room) to remove any residual oil to eliminate a potential fire hazard.

3.10.3 Shield Cooling System The inhibitor solution will be drained to radwaste and processed. This entire system (including the Shield Cooling Water Return Tank, Heat Exchanger, pumps and out of containment piping will either be decontaminated to levels less than those given in Figure 3.2-1 or the equipment will be removed, packaged and shipped off-site for disposal.

3.10.4 Feedwater System The Feedwater System will be drained. The system may be disassembled for salvage or scrap, or may be retired in-place.

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3.10.5 Inert Gas Generator The propane line from the propane storage tank to the generator has been purged with nitrogen and the storage tank has been purged with nitrogen. The system may be dismantled for salvage or scrap or may be retired in-place.

3.10.6 Circulating Water System The circulating water system will be drained. The equipment may be disassembled for salvage or scrap or may be retired in-place.

3.10.7 Turbine Generator and Auxiliaries The entire Turbine Generator System, including auxiliaries, may be removed and moved to a fossil-fired station to replace older equipment.

If the turbine generator and auxiliaries are not moved early after shutdown, they may be layed up to protect them for possible future use. This layup will be done in accordance with Philadelphia Electric procedures.

3.10.8 Emergency Cooling Water Systems The systems will be vented (under normal conditions, there is not any water in these systems). The Emergency Cooling Water Piping will be cut and capped at the containment.

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3.10.9 Gas Delay System The beds will be degassed by shutting down the cooling systems to the delay beds and then passing the helium Stored in the storage tanks through the purification system to heat it up and then through the beds to the liquid nitrogen traps. The liquid nitrogen traps will be regenerated periodically and the activity from regeneration transferred to the holdup tank. The gas in the holdup tank will be sampled and the activity released via the stack in accordance with the plant Technical Specifications.

The Purge Condensibles Traps, degassed delay beds, first and second dust collectors and the pre and after filters on the liquid nitrogen traps will be removed from the facility, packaged and shipped to an off-site licensed burial facility for disposal. All piping which is cut for equipment removal will be capped and welded. The shield plugs or decking plates will be reinstalled over the equipment cubicles.

The dust will be emptied out of the Bypass Filter dust collector and packaged as solid waste for off-site disposal at a licensed burial facility.

3.10.10 Chilled System The water and glycol will be drained from this system.

3.10.11 Refrigeration and Brine Systems The refrigerant will be drained or vented (one refrigerant is a liquid at room temperatures, the other is a gas). The lubricating oil system will be drained and the oil heaters will be disconnected.

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3.10.12 Nitrogen Recondensers The seal and lubricating oil will be drained and the helium (non radioactive) will be vented.

3.10.13 Fuel Handling Purge System The system will be vented. The oil filter cartridge and exhaust filter cartridges will be removed. The oil will be drained from the vacuum pumps. The filter cartridges will be processed as solid radioactive waste.

3.10.14 Non Purified Helium Handling System Once the delay beds have been degassed, the filter cartridge will be removed from the oil adsorber and processed as solid radioactive waste.

The lubricating oil, primary oil and oil injection Systems on the transfer compressors will be drained.

The line from the helium makeup bottles will be cut and capped at the containment.

3.10.15 Purified Helium System The oil and cooling water will be drained from the Purified Helium Compressors. The filter cartridges from the oil filters will be removed and processed as solid radioactive waste. The Pump Down Plate Out Adsorber will be removed from the system and shipped off-site for licensed disposal. The piping cut for its removal will be capped and welded.

3.10.16 Chemical Cleanup System The Steam Generator Purge Plate Out Trap will be removed and packaged for off-site licensed disposal. All piping cut for the removal of this equipment will be capped and welded. The copper in the Oxidizer will be regenerated to the copper oxide form and left in the Oxidizer Revision 7, April 2012

vessel. The Purge Water Condenser, Water Separator Caustic Scrubber and Water Scrubber will be drained.

3.10.17 Fuel Pool Cooling System The fuel pool filter socks will be removed and shipped off-site as radioactive solid waste. The fuel pool filters, heat exchangers, pumps and booster pump will be removed, packaged and shipped off-site for licensed disposal. The Fuel Pool Cooling System piping which is outside the Fuel Pool Building will be decontaminated to levels less than those in Figure 3.2-1 or the piping will be removed.

3.10.18 Radiation and Process Monitors Check sources will be removed from all monitors outside the Containment and Fuel Pool Buildings. The circuits for these monitors will be de-energized as permitted by the Technical Specifications. Some of these monitors may be moved to Units 2-3 for use in those facilities.

3.10.19 Containment Equipment Cooling Water This system will be drained. Out of containment equipment will be surveyed. If contaminated, it will be decontaminated or removed, packaged and shipped off-site for licensed disposal. If not contaminated, the out of containment equipment may be disassembled for salvage or scrap.

3.10.20 Containment Hot Water Heating System This system will be drained. Out of containment equipment may be disassembled for salvage or scrap.

3.10.21 Decontamination System The system will be drained and a blank flange installed over the top of the Decontamination Tank.

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3.10.22 Liquid Waste Disposal System The oil in the Contaminated Oil Storage Tank will be drained, mixed with an adsorbent, or solidified, packaged and shipped off-site for licensed disposal.

The waste resins will be sluiced to the drumming area so they can be drummed and shipped off-site for licensed disposal. The liquid waste system will be drained and the water either discharged in accordance with the Technical Specifications or trucked to Units 2-3 for processing. The filter socks will be removed and shipped off-site for licensed disposal. The waste system piping (including the discharge piping) will be decontaminated to levels less than those in Figure 3.2-1 or will be removed and shipped off-site for licensed disposal. The Waste Building, including the Waste Building Sump, will be decontaminated to levels less than those given in Figure 3.2-1.

3.10.23 Ventilation System All exhaust filters from all plant systems feeding the

1. 1 plenum and those in the exhaust of the #3 plenum will be removed, packaged and shipped off-site for licensed disposal. All contaminated ducts outside the Containment or Fuel Pool Buildings will be decontaminated to levels less than those in Figure 3.2-1 or will be removed, packaged and shipped off-site for licensed disposal. Covers will be welded over the supply and exhaust duct penetrations of the containment.

The stack will be surveyed and if contaminated, will be decontaminated or will be disassembled, packaged and shipped off-site for licensed disposal.

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3.10.24 Turbine Building Cooling Water System This system will be drained and may be disassembled for salvage or scrap 3.10.25 Critical Service Water System This system will be drained and may be disassembled for salvage or scrap.

3.10.26 Electrical System The electrical system will be modified so that all electrical service is disconnected except that needed to:

a. Supply lighting in the containment and Fuel Pool Building for periodic inspections.

b. Supply power to the Containment Cathodic Protection System rectifier units.

c. Supply 110/220 service for the Administration Building.

d. Supply power as needed for layup of balance of plant equipment.

This modification will be done in such a way that the power supplied to areas outside the exclusion area is independent from that supplied to areas within the exclusion area so that circuits within the exclusion area cannot accidentally be energized.

Most of the electrical system will be retired in place. Wiring will be cut as required to remove other equipment.

Some key pieces of equipment such as main transformers may be Salvaged for Use at other facilities.

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3.10.27 Containment Cathodic Protection This system will be kept Operational to help protect the containment vessel from Corrosion. The rectifier units are located inside the Exclusion Area so they can be checked during the periodic inspections.

3.10.28 Diesel Generator Fuel oil for the diesel generator will be drained. The diesel generator may be moved to another facility or may be layed up on site so that it can be salvaged at a later date.

3.10.29 Firefighting and Alarm Systems The carbon dioxide will be removed from the Cardox System. The alarm systems will be deactivated.

3.10.30 Makeup Water System This system will be drained. The system may be disassembled for salvage or scrap.

3.10.31 Service Water System This system will be drained. It may be disassembled for salvage or scrap.

3.10.32 Fire Water System This system may be retained or may be disassembled for salvage or scrap.

3.10.33 Domestic Water System The Domestic Water System will be drained. If the Administration Building is to be used, domestic water will be piped to the Administration Building from either the Information Center Well or the Unit 2-3 Domestic Water System.

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3.10.34 Service and Instrument Air Systems These systems will be vented and drained. The systems may be disassembled for salvage or scrap.

3.10.35 Instrument Systems All instrument systems will be deactivated. Those systems which contain fluids will be drained and/ or vented. The instrument systems will generally be retired in place. Cables will be cut and removed as is required to remove other equipment. Certain components of the instrument systems may be removed for use at other facilities.

3.11 Inventory Of Radioactive Materials Left On-Site The only significant Source of radioactive materials left in the Unit-1 facility will be the neutron activation products contained in the reactor vessel. Some fission products left on the internal surfaces of the primary coolant system and some fission product activity in the annulus of the Spent Fuel Pool.

The results of an activation analysis study on the reactor vessel are given in Figure 3.11-1.

The activity remaining on the Surface of the primary coolant system is based on analyses performed by Oak Ridge National Laboratories. The activity is expected to be about 1.25 Ci/cm2 at the outlet from the steam generator. Although the studies made show the activity to be progressively less than this further downstream, the assumption that the 1.25 Ci/cm2 activity covers the entire 2.4 x 107 cm2 area of the primary system results in a total residual activity in the primary coolant system of 30 curies. A breakdown of this activity is given in Figure 3.11-2.

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The activity remaining in the annulus of the fuel pool is a result of a leak which occurred in the pool in 9/71. This crack was repaired in 9/73 and the fact that no water has collected in the Fuel Pool Sump since then shows that the leak was properly repaired. The residual activity in the Spent Fuel Pool annulus is estimated to be 0.1 mCi which is primarily made up of Cs-137.

Although there is residual activity in systems other than the primary coolant system in the containment, the total activity in such systems (once the traps and delay beds are removed) is expected to be less than ten percent of the activity in the primary coolant system or 3 curies.

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Figure 3.11-1 Activation Source Strengths Six Months After Shutdown Activity in Curies Region Mn-54 Fe-55 Fe-59 Co-58 Co-60 Ni-59 Ni-63 Total Lower Reactor Vessel 2.22 (3)* 1.76 (5) 2.12 (-7) 3.89 (1) 8.91 (3) 2.75 (0) 4.18 (2) 1.88 (5)

Upper Reactor Vessel 2.12 (3) 2.08 (5) 2.52 (-7) 3.73 (1) 1.05 (4) 3.25 (0) 4.95 (2) 2.21 (5)

Upper Reactor Head 1.87 (2) 1.31 (4) 1.58 (-8) 3.29 (0) 6.64 (2) 2.04 (-1) 3.12 (1) 1.40 (4)

Lower Reactor Head 6.62 (2) 5.24 (0) 6.34 (-12) 1.16 (1) 2.66(-1) 8.2 (-5) 1.25 (-2) 6.79 (2)

Metallic Insulation 1.09 (2) 1.30 (4) 1.57 (-8) 5.37 (1) 9.15 (3) 5.67 (0) 8.61 (2) 2.32 (4)

Metallic Insulation Lining 1.39 (1) 1.66 (3) 2.08 (-9) 3.32 (2) 1.58 (4) 3.51 (1) 5.33 (3) 2.32 (4)

Plenum Shroud 1.17 (0) 7.91 (4) 9.58 (-9) 2.0 (-2) 3.90 (3) 1.20 (0) 1.84 (2) 8.32 (4)

Upper Thermal Shield 5.44 (2) 5.99 (4) 7.22 (-8) 9.32 (0) 2.94 (3) 9.10 (-1) 1.38 (2) 6.35 (4)

Lower Thermal Shield 7.39 (3) 2.01 (6) 2.46 (-6) 1.26 (2) 9.78 (4) 3.06 (1) 4.64 (3) 2.12 (6)

Core plate Thermal Shield 6.03 (2) 1.08 (4) 1.30 (-8) 1.03 (1) 5.32 (2) 1.64 (-1) 2.50 (1) 1.20 (4)

Core Support Plate 1.96 (3) 5.25 (3) 6.34 (-9) 3.36 (1) 2.59 (2) 8.0 (-2) 1.22 (1) 7.51 (3)

Standoff Pins 1.86 (2) 2.0 (4) 2.43 (-8) 9.16 (1) 1.4 (4) 8.7 (0) 1.32 (3) 3.56 (4)

Control Rod Couplings 1.68 (-2) 3.97 (1) 5.02 (-11) 5.58 (-2) 1.64 (5) 1.18 (-1) 1.76 (1) 1.64 (50 TOTAL 1.60 (4) 2.60 (4) 3.17 (-6) 7.48 (2) 3.28 (5) 8.87 (1) 1.35 (4) 2.96 (6)

• 2.22 (3) = 2.22 x 103 Revision 7, April 2012

Figure 3.11-2 Deposited Activity on Interior Surfaces of the Reactor Coolant System Activity Radionuclide (curies)

Cs-137 20.2 Cs-134 5.8 Ce-144 2.0 Sr-90 2.0 30.0 Revision 7, April 2012

NOTE Chapter 4, Sections 4.1 and 4.3 are considered Historical information and involve planning for placing Peach Bottom Atomic Power Station, Unit 1 in its present SAFSTOR status.

4. DECOMMISSIONED PLANT SAFETY ANALYSIS This chapter provides the analyses performed for placing Peach Bottom Unit 1 in a SAFSTOR status with all spent fuel removed from the site, spent fuel pool drained and decontaminated, all radioactive liquids removed, and accessible areas of the facility decontaminated. These analyses continue to bound the possible events and their consequences for Peach Bottom Unit 1.

4.1 Thermal Analysis of Heat Generated by Activation Product Decay 4.1.1 Summary The Peach Bottom Unit 1 Facility is in a SAFSTOR status with the reactor vessel and its internals, not including fuel, remaining on-site in the containment. The effect of the heat generation due to the decay of activation products in the reactor vessel on the susceptibility to ignition of the graphite reflector blocks in the reactor vessel and on the integrity of the concrete vessel enclosure was evaluated. The heat generation rate due to activation product decay, six months after shutdown, was calculated to be 29,100 BTU/hr resulting in a maximum graphite temperature of 286 F. At this temperature, a self sustaining release of stored energy in the graphite irradiated at 550F could not occur.

Since the graphite ignition temperature is approximately 1200F, it was concluded that the reactor vessel containing the graphite reflector blocks may be safely laid up under an atmospheric environment.

The thermal analysis also indicated that the maximum concrete temperatures will be less than 60 above ambient and therefore no degradation of the concrete physical properties will occur.

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4.1.2 Safety Consideration The reactor vessel was defueled and dummy elements were placed in the core cavity. (This was necessary to prevent the core from collapsing during the defueling operation since the fuel elements were not self supporting). The graphite reflector blocks which were left in place are combustible but have an ignition temperature of 1200F. Even with this high ignition temperature, a high degree of confidence is required to assure that they will not ignite during long term storage. It is desirable to leave the reactor vessel under an atmospheric environment, relying on passive mechanisms to dissipate the activation product decay heat, so that the peak temperature in the graphite is below that necessary for a self sustaining release of stored energy.

The second area of concern is the effect of elevated temperatures on the physical properties of concrete, especially the tensile and compressive strength.

After defueling, the emergency vessel cooling system as well as the biological shield cooling system were deactivated. These systems were provided to cool the reactor vessel, the vessel cavity and the biological shielding during normal reactor operation and during emergency conditions.

4.1.3 Reactor Vessel/Reactor Vessel Cavity Heat Transfer The system temperatures were conservatively calculated using a one-dimensional model of an isothermal heat source, the reactor vessel, located within a grey body enclosure. After the reactor vessel temperature was determined, analyses were performed to calculate the temperature of the graphite within the vessel, considering the thermal resistances of the air spaces Revision 7, April 2012

between the reactor vessel, the thermal shield, and the reflector blocks. The thermal model accounts for conduction heat transfer across the concrete enclosure, the cavity internal insulation and the reactor vessel internal structures, and considers combined radiation and convection heat transfer across all air spaces within the vessel and the reactor cavity.

The uninsulated reactor vessel is surrounded by a 3/8 inch thick carbon steel shroud which supports the reactor vessel emergency cooling coils. The shroud forms an eight inch annular air space around the reactor vessel. The shroud and reactor vessel are located in a 20 foot diameter concrete cavity lined with four inches of insulation. The insulation combined with cooling coils embedded in the biological shielding, protected the concrete from excessive temperatures during reactor operation. Heat transfer is therefore retarded due to a combination of thick insulated concrete walls and air spaces. Annular air spaces within the complex range from 2 to 11 inches and as such transfer heat by combined radiation and natural convection. A major conservatism in the analysis is the assumption that the air in the reactor cavity heats up but does not transfer heat by mass transport to the containment atmosphere.

In reality convective mass transport will occur since the reactor cavity is not air tight.

An activation analysis using the ORNL developed code ORIGEN,

• was performed to determine the neutron activation product inventory and the magnitude of the heat generation from the decay of these activation products in the reactor vessel complex.

• ORIGEN - The ORNL Isotope Generation and Depletion Code, M. J. Bell ORNL - 4628, May, 1973 Revision 7, April 2012

The decay heat energies of the important nuclides formed by fast and thermal neutron activations 6 months after shutdown were calculated considering gamma and beta decay energy. Six months elapsed from reactor shutdown until the reactor was defueled and forced convection cooling was terminated.

The maximum calculated system temperatures were as follows:

Peak temperature in graphite 286.2F Thermal shield 278.4F Reactor vessel 269.2F Emergency cooling shroud 262.5F Inner surface of insulation 256.4F Inner surface of concrete 145.1F The heat sink for the activation product decay heat was assumed to be at a temperature of 90F. It consists of the surrounding concrete and metal structures which interact with the reactor cavity exterior walls through radiant energy exchange. Although the containment ambient air may reach 100-110F on certain summer days, a 90F sink temperature was considered to be appropriate because of the ambient temperature's transient nature (diurnal) and the large thermal inertia of the concrete structures.

4.1.4 Stored Energy When graphite is irradiated, an increase in internal energy takes place due to atomic dislocations within the carbon lattice. The net rate of stored energy accumulation is determined by the difference between energy accumulation due to irradiation and that of the Revision 7, April 2012

material annealing at the irradiation temperature. The consequence of this stored energy is the potential for its rapid release during a subsequent heating resulting in potentially high temperature excursions. Under extremely high stored energy conditions, the release of all of this energy could cause the graphite temperatures to reach the graphite ignition temperature, and in the presence of air result in a self sustaining combustion reaction.

The total stored energy in the reflector blocks in the Peach Bottom 1 reactor is small due to the fact that the lowest operating temperature in the reflection region was >550F, and the rate of dislocation annealing at this temperature is quite large. Some of the residual stored energy within the graphite would be released as heat if the cold shutdown temperature was raised above the irradiation temperature of 550F. A conservative thermal analysis of the hottest point in the reflector elements indicated that the temperature will not exceed 287F. Therefore, there is a substantial margin for possible stored energy release. Even if this could occur, the stored energy release would not be self sustaining since the rate of energy release with increasing temperature is well below the specific heat of unirradiated graphite(1). For a self sustaining stored energy release the graphite would have to be heated above approximately 900F.

(1)

R. E. Nightingale, Nuclear Graphite, Academic Press 1962, Pages 325, 341 Revision 7, April 2012

4.1.5 Post Defueling Temperature Monitoring To insure that heat generation within the reactor vessel was not excessive, a temperature monitoring test program was conducted before inerting of the reactor vessel was discontinued. After the reactor was defueled, forced convection cooling was terminated. The reactor vessel was allowed to heat up from activation product decay.

The existing core and vessel thermocouples were monitored to determine the magnitude of the layup temperatures and to verify the results of the thermal analysis. As expected, the predicted temperatures are not approached.

4.1.6 Concrete Structural Integrity The significant properties of concern with respect to the integrity of concrete are the concrete tensile and compressive strengths and thermal conductivity, as it affects its heat dissipation ability. It is reported by ORNL(2) that the tensile strength of concrete is affected only about 5% by temperatures up to 482F.

Exposure to 572F will produce a 14% to 16% reduction.

The effect of temperature on the compressive strength is not detrimental; in fact, a slight increase in strength is observed at 392F. Both properties were tested at temperatures significantly higher than expected in the Peach Bottom Unit 1 structures where the peak temperature is below 150F. It is further reported in ORNL-4227 that the thermal conductivity of concrete decreases by only approximately 5 percent with a temperature increase from 50F to 150F.

(2)

ORNL-4227 - "Prestressed Concrete in Nuclear Pressure Vessels, A Critical Review of Current Literature", Chen Pang Tan, May 1968. Pages 249, 253-256 Revision 7, April 2012

Therefore ability of the concrete to dissipate heat will not be significantly affected by the temperature expected following placement of Peach Bottom Unit 1 into SAFSTOR status.

Substantially similar conclusion of temperature effects on concrete strength properties was presented by Bertero and Polivka(3) where exposure of concrete to a sustained temperature of 300F produced no significant changes in mechanical characteristics. Weigler and Fisher(4) examined the effects of temperature on conductivity in the range up to 140F and concluded that "for practical considerations the influence of temperature on the thermal conductivity can be ignored.

(3)

Influence of Thermal Exposure on Mechanical Characteristics of Concrete" V. V.

Bertero and M. Polivka. Concrete for Nuclear Reactors, Special Publication SP-34, American Concrete Institution, 1970 (4)

"Influence of High Temperatures on Strength and Deformations of Concrete." H.

Weigler and R. Fischer, Concrete for Nuclear Reactors, Special Publication SP-34, American Concrete Institute, 1970 Revision 7, April 2012

4.2 Site Flooding 4.2.1 Summary The potential for site flooding was analyzed and the consequences of such flooding were evaluated with respect to any radiological safety hazards. Based on the historical hydrological data for the Peach Bottom site and the most recent flood levels experienced in 1972 as a result of tropical storm Agnes, flooding at the site to the containment vessel's grade elevation of 116'-0" (C.D.)(1) is judged to be remote. Furthermore, even if the site were inundated, the effects would be less deleterious for Unit 1 in SAFSTOR status than for the operational site. This is due to the following facts:

1. All radioactive material remaining on site will be contained in the containment vessel and the Fuel Pool Building.

2. The containment vessel will not be made more buoyant under flooding conditions in the SAFSTOR status than it was in the operational status as the only weight which will be removed from it will be the fuel, charcoal delay beds, miscellaneous filters, Krypton traps, water and other plant liquids which constitute less than 1% of the gross weight.

Depending on the components or sections of components to be removed for metallurgical examination under the Post Operational Sampling Program, the removed weight will still be less than 2% of the gross weight. For this worst case, the neutral buoyancy is at an elevation of 162 ft. (C.D.) which is 48 ft. above the highest historically observed flood level at the Peach Bottom site.

(1)

Conowingo Datum (C. D.) is MSL - 0.7 ft.

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4.2.2 Historical Flood Data The Peach Bottom site is located on the west bank of the Conowingo Pond which is impounded by Conowingo Dam located nine miles downstream. Holtwood Dam located about six miles upstream from the Peach Bottom site, forms the upper limit of Conowingo Pond.

Data on historic floods of the Susquehanna River at Harrisburg, Pa. have been compiled in several reports by the Commonwealth of Pennsylvania and the United States Geological Survey. Continuous gage-height records are available since 1874. The previous maximum flood of record occurred on March 19, 1936 when the peak flow of 740,000 cubic feet per second was recorded at Harrisburg, Pa.

The peak flow at Conowingo was estimated at 839,000 cfs by Exelon Generation Company, LLC (formerly PECO Energy Company formerly Philadelphia Electric Company) by utilizing recently determined Conowingo spillway discharge coefficients as obtained from a model study by the Alden Research Laboratory of Worcester Polytechnic Institute. The water elevation attained in the vicinity of the Peach Bottom site during the 1936 flood is estimated to be Elevation 113.0 ft. (C.D.). The peak flow occurring during the 1936 flood is the basis for the present Standard Project Flood (SPF)(2) as defined by the U. S. Army Corp. of Engineers, Baltimore District.

(2)

The Standard Project Flood is defined by the Corp. of Engineers as a synthetic flood that represents the critical concentrations of runoff from the most severe combination of precipitation and snow melt that is considered "reasonably characteristic" of the drainage basin involved and as such is the maximum level of protection usually considered practical for local flood protection facilities. The Standard Project Flood is used as a measure of a reservoir's adequacy to control large floods.

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The maximum flood in the Susquehanna River occurred on June 24, 1972 as a result of tropical storm Agnes. The U. S. Geological Survey measured the peak flow of 1.13 x 106 cfs downstream of Conowingo Dam where Interstate 95 crosses the Susquehanna River; 1.08 x 106 cfs at the tailwater of Holtwood Dam; and 1.02 x 106 cfs at 6

Harrisburg. A peak flow of 0.972 x 10 cfs was calculated at the Conowingo Dam by Exelon Generation Company, LLC Exelon Generation Company, LLC personnel based on actual elevation measurements and the spillway discharge coefficients obtained from the model study of Alden. Figure 4.2-1 lists the flood levels measured by the Exelon Generation Company, LLC near and about the Peach Bottom Site. This data compares well with data published by the U. S. Geological Survey in Harrisburg and is slightly conservative (i.e., flood elevations shown are higher).

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FIGURE 4.2-1 Exelon Generation Company, LLC Flood Heights June 22-24, 1972 Susquehanna River Basin Location on Susquehanna River Elevation (feet C.D.)

Conowingo Dam 111.4 Peach Bottom (screen structure) 114.2 Peach Bottom with wind generated waves(4) 116.0 Muddy Run Pumped Storage(5) 128.5 (4)

The significant wave height defined as the average height of the highest one-third of all waves generated is estimated at 2.7 ft. tip to trough.

For height above still water, 2/3 of the value is used.

(5)

The Muddy Run Pumped Storage Generating Plant is located on the east side of the Reservoir, about four miles upstream from Peach Bottom.

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4.2.3 Site Flooding Effects The general grade elevation of the Peach Bottom 1 facility is 116.0 (C.D.). Based on historical flood data, inundation and safety margins associated with the decommissioned status of the facility have been examined using Agnes flood elevations and superimposed wind generated waves.(3)

In the SAFSTOR status the only radio-activity remaining on site will be within the containment and the Fuel Pool Building. The containment vessel is a 100 foot diameter thin shell steel cylindrical structure. The shell near grade elevation is 15/32 inch thick and is reinforced by stiffening rings at elevations 115'-6", 130'-6" and 143'-9". A stress analysis was made of the ability of the containment vessel above grade level to withstand an extended flood loading without buckling and it was concluded that the vessel could satisfactorily withstand flood elevations to 122.8 ft. (C.D.)

This is 6.8 feet above the levels associated with tropical storm Agnes and assumed wind generated waves.

The bottom of the equipment access door, the personnel access door and the personnel emergency air lock are at elevation 116 ft. (C.D.). The equipment door and emergency air lock is welded shut so that inleakage of water during a flood is considered highly improbable.

Potential leakage at the normal personnel access door, which is secured, will be limited to seal leakage.

(3)

UFSAR Peach Bottom Atomic Power Station Units 2 and 3, Section 2.4 Revision 7, April 2012

An analysis of the containment vessel also indicated that the neutral buoyancy point, at which the containment would undergo upheaval with no artificial downward restraining forces such as adhesion of the lower areas to the surrounding soil, is 47 feet above the highest recorded flood level. No data is available to substantiate the effects of soil adhesion in preventing upheaval, but the force required to continue driving "frozen" or "set" pilings is usually many times that required for "free" or "running" pilings due to soil adhesion. Such high buoyancy margin is due to the fact that the reactor vessel and most of the concrete structure inside the containment are located above grade level.

Should the containment vessel be breached by flood waters, the quantity of radionuclides which could be released would be limited to the small amount of contamination that is not sealed within the Primary Reactor Coolant boundary. Mixing and release would be further inhibited by the internal compartmentation of the plant systems. More than 99 percent of the residual activity estimated to have decayed to less than 0.2 megacuries is contained in the reactor vessel and an additional 20 curies is estimated to be contained in the sealed primary coolant system and will not be dispersible. It is conservatively estimated that less than 3 curies of activity is retained in the remainder of the facility. Since most of this activity will be retained in sealed auxiliary systems, only a small fraction of it would be dispersible by any flood waters which could be postulated to penetrate the Containment Building.

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The spent fuel pool is located outside the containment, housed in a separate concrete building. The building will not be more buoyant under flooding conditions than the containment vessel and the five foot thick walls will withstand hydrostatic forces in excess of those that the containment vessel can tolerate. The residual activity in the Spent Fuel Pool Building is estimated to be less than 0.1 millicurie and would not present any appreciable activity if dispersed in flood waters.

Additionally, for flood waters to enter the spent fuel pool, the water elevation would have to rise in excess of 6 feet above grade as this is the lowest access opening to the fuel pool.

4.2.4 Conclusions Based on these extensive measurements made by the Exelon Generation Company, LLC and the U. S. Geological Survey, it is concluded that the decommissioned Peach Bottom 1 containment vessel will not be subjected to hydrostatic forces which could compromise the vessel integrity nor will the vessel be subjected to upheaval due to buoyant forces resulting from abnormal water levels. Dispersal to the environment of the small amounts of radioactivity remaining in the containment is considered to be highly unlikely considering the history of the region's flood levels. In view of the inherent safety of the site and the attendant safety margins associated with the containment structure, the potential for accidental release of by-product material as a consequence of storm induced floods in the Susquehanna River is not considered inimical to the health and safety of the public.

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4.3 Containment Pressure Transient 4.3.1 Summary An analysis was performed to evaluate the maximum containment pressure rise resulting from an accident condition that may conceivably occur during the decommissioning activities. With the reactor in the Low Pressure shutdown mode(1), the maximum pressure rise possible due to uncontrolled release of 1800 lbs. of helium coolant is 4.0 psig.

4.3.2 Decommissioning Design Basis Accident For the operating reactor the design basis accident for which the containment is designed is based on the following series of postulated failures:

1. a primary system rupture

2. rupture of one steam generator tube

3. failure of the helium loop valves to isolate the ruptured steam generator from the reactor

4. loss of all forced circulation cooling of the core The highest peak containment pressure resulting from this multiple accident is 8.0 psig. The analysis assumes that:

1) 926 lbs. of primary coolant loop helium at 790F is released to containment,

2) 2,200 lbs. of water and steam from the faulty steam generator system enters the primary loop, (1)

The reactor will be considered to be shutdown when at least 52 of the 55 control rods are fully inserted in the core, the reactor is kept in the Low Pressure Mode, the outlet temperature is less than 500F and the decay heat is less than 1 Mw. (The Low Pressure mode of operation will insert a 0.01% scram from the intermediate range channels of the reactor protection system.)

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3) total immediate chemical reaction of 1100 lbs. of the released water and steam with the reactor core graphite assuming its complete conversion to carbon monoxide and hydrogen which are then released at 660F to the containment,

4) subsequent chemical reaction of the remaining water, steam, oxygen, and carbon dioxide in the containment atmosphere with the core graphite at a rate dictated by natural convection of the containment gas mixture through the reactor vessel.

During the defueling phase, this design basis accident is not possible. The reactor will be shutdown and in the Low Pressure operating mode. The energy of helium will be reduced due to lower operating temperatures and the heat source required to produce the graphite-water reaction which contributes to the containment pressure rise is no longer available.

Calculations show that the peak pressure transient possible is 4.0 psig. This is based on the following assumptions:

1) the plants helium coolant inventory is instantaneously released into the containment.

2) maximum inventory of helium in the primary loop and in coolant holdup tanks is 1800 lbs. at 200F average temperature.

3) no heat transfer to passive heat sinks.

4) free containment volume of 720,000 ft3.

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Immediately after shutdown, the plant helium inventory was 1900 pounds. Due to normal system leakage and/or controlled release, at the end of Phase I (fuel removed from the reactor and delay beds degassed), there was no more than 1800 lbs. of helium within the plant.

Therefore, a containment pressure of 4 psig cannot be exceeded under any postulated accident conditions.

The plant helium inventory was eliminated to place Unit 1 in SAFSTOR status.

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4.4 Safeguards and Radiological Safety Although the radiation levels in Peach Bottom Unit 1 in SAFSTOR status are very low, precautions are taken to assure that people are not exposed to the radiation or radioactivity.

4.4.1 Site Security of the facility shall be included as part of the Peach Bottom Atomic Power Station Security Plan.

4.4.2 Exclusion Area Access to the Exclusion Area is made through a locked gate. See Figures 4.4-2 and 4.4-3 for containment entry barriers.

The area outside the Containment, Radwaste area and Fuel Pool Buildings is decontaminated to less than the release levels given in Figure 3.2-1. The radiation levels at the Exclusion Area boundary are less than 0.08 mrem/hr.

4.4.3 Containment Various areas in the containment are accessible for periodic inspections. The refueling floor and the areas above it are available for inspection so that radiological surveys can be made to evaluate any migration of activity within the SAFSTOR facility. The Containment below grade areas are available for inspection so that radiological surveys can be made on a lower elevation in the Containment and so that any water leakage can be detected visually. Appropriate barricades are installed in the Containment to prevent accidental access to controlled areas.

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Elevation 176'-6" (Refueling Floor) is accessible for periodic inspections through the access lock on that elevation. The access lock outer door is padlocked in the closed position. This lock requires a key different from the one for the gate in the expanded metal wall to make entry to the Containment more difficult.

Accessible areas on or above Elevation 176'-6" are decontaminated to levels less than the release levels given in Figure 3.2-1. Equipment with internal contamination levels higher than the release levels are sealed to enclose the activity. Radiation levels within the accessible area are less than 1 mrem/hr. All four stairways leading down from the 176'-6" elevation are blocked with expanded metal grating to limit access during the periodic inspections (see Figure 4.4-3). The grating on the southwest stairway is hinged and locked in place. The key for this lock is different than those for access to the Containment to prevent unintentional access to the controlled portions of the Containment.

The southwest stairway is used for access to lower levels of the Containment for special inspection if such inspection is indicated by the periodic inspection on elevation 176'-6" and in the vicinity of the containment sump. The grating on the other three stairways is welded or bolted in place. Expanded metal grating is also placed over the ladder access opening to prevent access to the chilled water head tank platform.

Exposure to radiation levels from the reactor would require that the missile beams be removed from above the reactor vessel. This will require the replacement of power supply breakers to activate the building crane.

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Elevation 90' in the vicinity of the Containment Sump is made accessible for periodic inspections through the personnel lock on Elevation 116'-0". The outer door of the personnel lock is locked closed. This lock requires a key different from the one for the gate in the Exclusion Area fence. Entry into the controlled areas is controlled by the following barricades: (see Figures 4.4-2, 4.4-4 and 4.4-5) a) An eight foot expanded metal wall in the vicinity of the Freon-brine exchanger on Elevation 116'-0" to prevent access to the North.

b) A metal gate at the entrance to the control rod drive hydraulic equipment area on Elevation 116'-0".

c) An eight foot expanded metal wall from a point near the entry to the south stairway on Elevation 116-0" to the Containment wall to prevent access to the western portion of the Containment.

D) An expanded metal gate over the opening made in the wall to remove the Steam Generator Purge Plate-out Trap.

The installation of these barriers permit access from the personnel lock down the south stairs to Elevation 104'-0" and then down the southeast stairs to Elevation 93'-0" and the containment sump. All areas within this access path are decontaminated to levels less than those given in Figure 3.2-1. The Containment sump itself is decontaminated to as low a level as is practical with the intent of achieving levels within the Figure 3.2-1 release limits. Radiation levels within the access path are less than 1 mrem/hr.

Radiological postings are in place as required by radiological control procedures.

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NOTE FIGURE 4.4-1 has been deleted Revision 7, April 2012

FIGURE 4.4-2 CONTAINMENT GROUND FLOOR PLAN EL. 116'-0" B-# = Barricade Revision 7, April 2012

FIGURE 4.4-3 REACTOR REFUELING FLOOR PLAN EL. 176'-6" B-# = Barricade Revision 7, April 2012

FIGURE 4.4-4 CONTAINMENT UPPER BASEMENT PLAN Revision 7, April 2012

FIGURE 4.4-5 CONTAINMENT LOWER BASEMENT PLAN B-# = Barricade Revision 7, April 2012

4.4.4 Spent Fuel Pool Building The Spent Fuel Pool operating level is made accessible for periodic inspections of the pool and sump. Access to the Fuel Pool area is controlled by:

a) The doors on Elevation 122'-0" which permit the cask monorail to move the cask from the pool to the cask storage area is welded closed.

b) The hatch leading from the New Fuel Vault into the Fuel Pit Sump area on Elevation 116'-0" is welded closed.

c) An expanded metal gate is installed over the entrance on Elevation 116'-0" to the Fuel Pit Tube Area.

The personnel door on Elevation 116'-0" which leads into the Spent Fuel Pool Area is closed and locked. The key for this door is different from the one required to open the gate in the Exclusion Area fence. This personnel door is used for access to the Fuel Pool area during periodic inspections. The area from this door up to and around the pool is decontaminated to levels within the Figure 3.2-1 release levels. The Spent Fuel Pool is covered with steel mesh similar to fence wire. This wire is fastened or welded to prevent easy entry to the pool, but permits visual inspection of the pool during periodic inspections.

4.4.5 Key Control Since inadvertent entry (or entry for vandalism) to these decommissioned areas is being prevented by locked barriers, key control is important.

As noted earlier, the keys for opening each of these barriers is different. All keys are under the control of the Exelon Generation Company, LLC. Periodic inspection parties are given the keys for the locked barriers, but are not given the key to the grating over Revision 7, April 2012

the southwest stairwell. Any entry into a controlled area in Containment is made with at least one supervisory representative.

4.4.6 Miscellaneous Radiological Safeguards In addition to the physical locked barriers to prevent entry, the radiological safeguards include the policy that the area within the Exclusion Area fence will not be used for any purpose. It will be occupied (under controlled conditions) only when a periodic inspection is made or when maintenance is being performed to maintain the condition of the SAFSTOR facility.

Other possible radiological hazards could occur in the event of a fire, flooding, or severe local storm damage.

Since all readily flammable material has been removed as appropriate from the SAFSTOR facility, the chance of fire has been reduced considerably.

Flooding has been discussed in Section 4.2. Since this SAFSTOR facility is in the immediate vicinity of Units 2-3, which are operating, any local condition which may require emergency action will be apparent to the Unit 2-3 operating staff so that they can take appropriate, timely action.

4.4.7 Processing of Unit 1 Liquid Radwaste Any liquid waste created or found at Unit 1 will be designated as Unit 1 waste and appropriately reported to the NRC in accordance with Technical Specifications.

This waste, which may be temporarily staged in the Exclusion Area, will be transferred to the PBAPS radwaste facility located between Units 2 and 3 for processing. This waste will be clearly designated as belonging to the Unit 1 facility. This process will be controlled by written procedures.

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NOTE Chapter 5 is considered Historical information and involves planning for placing the Peach Bottom Atomic Power Station, Unit 1 in its present SAFSTOR status.

5. RADIOLOGICAL SAFETY DURING DECOMMISSIONING 5.1 Responsibilities Exelon Generation Company, LLC has full responsibility as the licensee for any obligations to the Atomic Energy Commission, the Department of Health of the Commonwealth of Pennsylvania and any other regulatory agencies. As the licensee, Exelon Generation Company, LLC retains the right and responsibility for final approval of all work done in the decommissioning activities and to assure that all such work is done in compliance with all license requirements.

The Peach Bottom operating staff will participate directly in the decommissioning operation. The Peach Bottom 1 POR (Plant Operations Review) Committee and the OSR (Operations and Safety Review) Committee will review any safety questions that may arise in the course of decommissioning.

5.2 Personnel The Peach Bottom Superintendent shall have the responsibility for the administration of all functions of the Peach Bottom 1 facility. He has the responsibility for safely maintaining the reactor facility and for safely conducting those activities necessary to carry out the decommissioning program. As the Superintendent of Units 2 and 3 as well as Unit 1, the Peach Bottom Superintendent has a technical operating staff to call on for assistance in the Unit 1 decommissioning as he deems necessary.

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5.3 Regulations All work associated with decommissioning of the Unit 1 facility will be performed in accordance with the requirements of 10CFR20, the Technical Specifications and the conditions of the decommissioning authorization.

5.4 Procedures The decommissioning work is expected to follow a schedule similar to the one presented in Figure 5.4-1. A similar schedule will be used to help determine manpower requirements during the decommissioning operation, to assure a logical sequence of work and to assure that written procedures are available for certain portions of the work. General or specific procedures shall be written as appropriate for any work to be performed which could affect the nuclear safety of the plant, could result in release of activity or could result in significant radiological hazards to personnel. The need for a procedure and the type of procedure (general or specific) shall be determined by the POR Committee. All procedures shall be reviewed by the POR Committee and approved by the Peach Bottom superintendent.

All decommissioning work will be done in accordance with the procedures set forth in the Peach Bottom Health Physics Manual and the Peach Bottom 1 Technical Specifications (Appendix A).

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FIGURE 5.4-1 PEACH BOTTOM 1 DECOMMISSIONING SCHEDULE Page 1 of 2 Revision 7, April 2012

FIGURE 5.4-1 PEACH BOTTOM 1 DECOMMISSIONING SCHEDULE Page 2 of 2 Revision 7, April 2012

5.5 Radiological Surveys All radiological (radiation, contamination and airborne activity) surveys will be performed under the cognizance of the Exelon Generation Company, LLC.

In addition to surveys performed during the decommissioning to evaluate radiological conditions encountered, to monitor equipment for shipment to a licensed burial facility off-site or to release equipment for uncontrolled use, a through survey of the facility will be made after decommissioning is complete to document the radiation and contamination levels in the decommissioned plant. This final survey documentation will be available as a baseline survey to help evaluate survey results obtained in future facility inspections.

5.6 Emergency Plans The emergency plans and procedures prepared for use during operation of Peach Bottom 1, or revisions thereto, will be continued in effect during the decommissioning. These procedures include:

a) Procedure Following Personal Injury or Contamination b) Injuries in Suspected or Known Radiation Areas c) Medical Services and Hospitalization Procedure d) Health Physics Emergency Procedures - Evacuation e) Health Physics Emergency Procedures - In-Plant Radioactivity and Contamination Control f) Health Physics Emergency Procedures - Leaking Spent Fuel Can in the Fuel Pool g) Health Physics Emergency Procedures - Search and Rescue h) Health Physics Emergency Procedures - Fire in a Radiation Area Revision 7, April 2012

i) Protection Plan Procedure - Intrusion j) Protection Plan Procedure - Bomb Threat k) Protection Plan Procedure - Civil Disturbance 5.7 Conclusions The Peach Bottom 1 facility is not now a risk to the health and safety of the public. The proposed decommissioning of the reactor facility will be performed in accordance with this decommissioning plan and will be administered by supervisory personnel who are experienced with nuclear plant work. The radiation safety aspects of the decommissioning activities will be evaluated by, and the performance of these activities will be monitored by experienced health physics personnel.

It is therefore concluded that neither the decommissioning process nor the final decommissioned facility will represent a risk to the health and safety of the public.

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NOTE Chapter 6 (except for Section 6.4) is considered Historical. Historical sections involved planning for placing Peach Bottom Atomic Power Station, Unit 1 in its present SAFSTOR status.

6. ADMINISTRATION The Exelon Generation Company, LLC will be responsible for the administration of the Unit 1 facility including periodic inspection.

6.1 Licensing This amendment to license DPR-12 requests permission to place the facility under revised technical specifications (Attachment A) for the duration of the decommissioning once the authorization to decommission is issued by DOL. Once all fuel is removed from the reactor, this same amendment requests permission to reduce the license status from Part 50-Utilization to Part 50-Possession only.

The Technical Specifications (Appendix A) will administratively assure the safety of the plant following defueling of the reactor by assuring that no fuel is replaced in the reactor and that certain minimum manning and equipment levels will be maintained throughout the decommissioning. These Technical Specifications are written so that as the decommissioning progresses, some of the Technical Specification requirements are dropped as key milestones in the decommissioning plan are reached.

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After the decommissioning of Peach Bottom 1 is completed, a Decommissioned Facility Report will be prepared and submitted to the AEC. The AEC will be asked to perform a post-decommissioning inspection of the facility to assure that its final condition is in compliance with the authorized decommissioning plan. The Technical Specifications (Appendix A) include a section of specifications to remain in force once the decommissioned facility has been inspected and approved by the AEC.

Changes in the license, if required, will be formally requested through the AEC.

6.2 Manning During Phase I of the decommissioning, the minimum shift crew shall be three persons, including one senior licensed operator, and one licensed operator. The senior licensed operator may be shared with the Boiling Water Reactor Facility when fuel handling operations are not being conducted.

Once all fuel is removed from the reactor and the delay beds are degassed, the minimum shift crew shall be one person.

One licensed senior reactor operator or formerly licensed senior operator (HTGR) shall be present at the Peach Bottom site anytime decommissioning activities or fuel handling operations are in progress. Once the decommissioning work is completed, responsibility for the Peach Bottom Unit 1 facility will be transferred to Unit 2 & 3 personnel.

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6.3 Records Normal plant operating records which are related to safety, personnel exposure, radioactive material discharge or shipment of radioactive material will be maintained throughout the facility decommissioning. In addition, plant records, health physics records and log books related to the decommissioning will be maintained.

The decommissioned facility status will be documented in a Decommissioned Facility Report. At the completion of the decommissioning, all appropriate operating and decommissioning records will be retained by Philadelphia Electric Company.

After decommissioning is completed, records and logs will be maintained relative to the following items:

(1) Facility radiation surveys (2) Inspections of the physical barriers (3) Abnormal occurrences Revision 7, April 2012

6.4 Inspection Exelon Generation Company, LLC will perform a routine inspection of the decommissioned facility in accordance with Technical Specifications. In addition, any time local conditions (fire, flood, etc.) indicate a possible change in facility condition, a non- routine inspection will be performed.

Inspections will normally be performed by Exelon Generation Company personnel, but could be performed by personnel from qualified firms who have contracts with the licensee. All inspections will be carried out by not less than two people, at least one of whom shall be experienced in health physics operations, surveys, field work and monitoring. The inspections will be carried out in accordance with written procedures which will cover the entry, inspection, inspection equipment removal, and reporting of the results.

Environmental Surveys will be performed as part of the Peach Bottom site environmental program and will not be reported separately for Unit 1.

The routine inspection will include the following:

a) At least 20 radiation and/or contamination survey points will be established on various elevations in the Containment and in the Spent Fuel Pool Building. These survey points will be selected to represent the areas within the accessible areas of Exclusion Area which have the highest potential to become contaminated if any activity is transported from the controlled areas of the Exclusion Area. These points will be surveyed during routine inspections.

b) Additional smears will be taken in random locations in accessible areas of the Exclusion Area.

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c) A general radiation survey will be performed in the Exclusion Area.

d) At least two air samples in the Containment and at least one in the Spent Fuel Pool Building will be taken during each inspection.

e) The high efficiency particulate filter on the Contain ment breather line will be replaced and surveyed in accordance with Technical Specifications.

f) Deleted g) Visual inspections will be made which will include:

1) The inside and outside of the Containment Vessel.

2) The Containment Sump to assure that ground water has not seeped into the Containment.

3) The barricades and locks within the Containment to assure that the integrity of the barrier to entry to controlled areas is intact.

4) (This step is "HISTORICAL".) The inside and outside of the Spent Fuel Pool Building.

5) The Spent Fuel Pit Sump to assure that groundwater has not seeped into the building.

6) (This step is "HISTORICAL".) The barricades within the Spent Fuel Pool Building to assure that the integrity of the barrier to entry to controlled areas is intact.

7) The doors and locks at the two entry points to the Containment and the one entry point to the Spent Fuel Pool Building.

8) Deleted Revision 7, April 2012

6.5 Reports a) An annual report will be submitted to the Director of Licensing, U. S. Atomic Energy Commission, Washington, D. C. 20545, describing the results of facility radiation surveys, the status of the facility, and an evaluation of the performance of security and surveillance measures. This report will be included as a section of the annual report submitted for Units 2 and 3.

b) An abnormal occurrence report will be submitted to the Regulatory Operations Regional Office by telephone within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of discovery of an abnormal occurrence.

The abnormal occurrence will also be reported in the annual report described in Section 6.5 (a).

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