ML092650291

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Post De-Fueling Monitored Storage Safety Analysis Report, Update 8, August 2009
ML092650291
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 08/31/2009
From:
GPU Nuclear
To:
Office of Nuclear Reactor Regulation
References
Download: ML092650291 (40)
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TMI-2 POST DE-FUELING MONITORED STORAGE SAFETY ANALYSIS REPORT UPDATE 8 August 2009

TMI-2 POST-DEFUELING MONITORED STORAGE SAFETY ANALYSIS REPORT UPDATE 8 AUGUST 2009

TMI-2 POST DE-FUELING MONITORED STORAGE SAFETY ANALYSIS REPORT UPDATE 8 August 2009

TMI-2 POST-DEFI JETT.ITN MONITORED STOR AGE SAFETY ANAT YSTS REPORT List of Effective Pages Page # Ilpdatey# Date TC-1 4 8/01 1-i 4 8/01 1.1-1 2 8/97 1.1-2 2 8/97 1.1-3 2 8/97 1.1-4 2 8/97 1.1-5 2 8/97 1.1-6 4 8/01 1.2-1 4 8/01 1.2-2 2 8/97 1.2-3 4 8/01 1.3-1 2 8/97 1.3-2 2 8/97 1.3-3 2 8/97 1.3-4 2 8/97 1.3-5 2 8/97 1.4-1 4 8/01 1.4-2 4 8/01 1.4-3 4 8/01 L.4-4 5 8/03 1.4-5 2 8/97 1.4-6 2 8/97 1.4-7 2 8/97 1.4-8 2 8/97 1.4-9 2 8/97 1.5-1 2 8/97

- i- Update-8 8/09

TMI-2 POST-DEFI.FI NG MONITORFD STORAGE SA-FTY ANAl .YS*S RPORT List of Effective Pages Pag# Ulndatef Dale 2-i 2 8/97 2-ii 2 8/97 2.1-1 4 8/01 2.1-2 2 8/97 2.1-3 2 8/97 2.1-4 4 8/01 2.2-1 2 8/97 2.3-1 2 8/97 2.4-1 2 8/97 2.4-2 3 8/99 2.4-3 2 8/97 2.4-4 8 8/09 2.4-5 2 8/97 2.4-6 2 8/97 2.5-1 2 8/97 3-i 2 8/97 3-ii 2 8/97 3-iii 6 8/05 3-iv 6 8/05 3-v 6 8/05 3-vi 2 8/97 3-vii 2 8/97 3-viii 2 8/97 3-ix 2 8/97 3-x 2 8/97 3-xi 2 8/97 Update-8 8/09

TMI-2 POST-DERFTL TNG MO* TORFD STOR AGE sAF-Ty ANATYSIS REPORT

-List of Effective Pages Page # U pdale # Dale 3-xii 2 8/97 3-xiii 2 8/97 3.1-1 2 8/97 3.1-2 3 8/99 3.1-3 2 8/97 3.1-4 2 8/97 3.1-5 2 8/97 3.1-6 2 8/97 3.1-7 4 8/01 3.1-8 2 8/97 3.1-9 6 8/05 3.1-10 3 8/99 3.1-11 2 8/97 3.1-12 6 8/05 3.1-13 3 8/99 3.1-14 2 8/97 3.1-15 4 8/01 3.1-16 4 8/01 3.1-17 2 8/97 3.1-18 2 8/97 3.1-19 2 8/97 3.1-20 4 8/01 3.1-20a 6 8/05 3.1-21 4 8/01 3.1-22 6 8/05 3.1-22a 3 8/99

,3- Update-8 8/09

TM1-2 PO-ST-DER IFUTNG MONiTORED STOR AGE SAFFTY ANAl .Y,ql.q RFPOIRT List of Effective Pages Page # I lpdate# Date 3.1-23 5 8/03 3.1-24 2 8/97 3.1-25 2 8/97 3.1-26 2 8/97 3.1-27 2 8/97 3.1-28 2 8/97 3.1-29 2 8/97 3.1-30 2 8/97 3.1-31 2 8/97 3.1-32 2 8/97 3.1-33 2 8/97 3.1-34 2 8/97 3.1-35 2 8/97 3.1-36 2 8/97 3.1-37 2 8/97 3.1-38 2 8/97 3.1-39 2 8/97 3.1-40 2 8/97 3.1-41 2 8/97 3.1-42 2 8/97 3.1-43 4 8/01 3.2-1 2 8/97 3.2-2 2 8/97 3.2-3 2 8/97 3.2-4 2 8/97 3.3-1 2 8/97 Update-8 8/09

TMI-2 POST-DFFT.TI INO MONITRF.I) STORAGE, SAFETY ANAlY-SIS REPORT List of Effective Pages Page # U1pdatp..# Dale 3.3-2 2 8/97 3.4-1 2 8/97 3.5-1 2 8/97 3.5-2 2 8/97 3.5-3 2 8/97 3.6-1 2 8/97 3.7-1 2 8/97 3.7-2 2 8/97 3.7-3 2 8/97 3.7-4 2 8/97 3.7-5 2 8/97 3.7-6 2 8/97 3.7-7 2 8/97 3.7-8 2 8/97 3.7-9 2 8/97 3.7-10 2 8/97 3.7-11 2 8/97 3.7-12 2 8/97 3.7-13 2 8/97 4-i 2 8/97 4-ii 2 8/97 4-iii 2 8/97 4-iv 2 8/97 4.0-1 2 8/97 4.1-1 2 8/97 4.1-2 2 8/97 Update-8 8/09

,TMI-2 POST-DFFIT ETI.NG MONITORED STORAGE SAFETY ANATLYSIS REPORT List of Effective Pages Page # Dute 81pd9te 4.1-3 2 8/97 4.1-4 2 8/97 4.1-5 2 8/97 4.2-1 2 8/97 4.3-1 2 8/97 4.3-2 2 8/97 4.3-3 2 8/97 4.3-4 2 8/97 4.3-5 2 8/97 4.3-5a 2 8/97 4.3-5b 2 8/97 4.3-6 2 8/97 4.3-7 2 8/97 4.3-8 1 6/95 4.3-9 1 6/95 4.3-10 1 6/95 4.3-11 1 6/95 4.3-12 1 6/95 4.3-13 1 6/95 4.3-14 1 6/95 4.3-15 1 6/95 4.3-16 1 6/95 4.3-17 2 6/95 4A-1 2 8/97 4A-2 2 8/97 4A-3 2 8/97 Update-8 8/09

TMI-2 POST-DEFRITE INCNMONITORED STORAGE SAFETY ANAlYSES RF.PORT List of Effective Pages Pape # Updale. Date 4A-4 2 8/97 4A-5 8 8/09 4A-6 2 8/97 4A-7 2 8/97 4A-8 2 8/97 5-i 2 8/97 5-ii 2 8/97 5-iii 2 8/97 5-iv 2 8/97 5-v 2 8/97 5-vi 2 8/97 5-vii 2 8/97 5.0-1 2 8/97 5.1-1 2 8/97 5.1-2 2 8/97 5.1-3 2 / 8/97 5.1-4 2 8/97 5.1-5 2 8/97 5.2-1 2 8/97 5.2-2 2 8/97 5.3-1 2 8/97 5.3-2 2 8/97 5.3-3 2 8/97 5.3-4 2 8/97 5.3-5 2 8/97 5.3-6 2 8/97 Update-8 8/09

TMI-2 POST-DEFI lE ING MONITORED STOCR AE.SAFETY ANAL YSIS REPORT List of Effective Pages Pa # IpdateA Date 5.3-7 2 8/97 5.3-8 2 8/97 5.3-9 2 8/97 5.3-10 2 8/97 5.3-11 2 8/97 5.3-12 2 8/97 5.3-13 2 8/97 5.3-14 2 8/97 5.3-15 2 8/97 5.3-16 2 8/97 5.3-17 2 8/97 5.3-18 2 8/97 5.3-19 2 8/97 5.3-20 2 8/97 5.3-21 2 8/97 5.3-22 .2 8/97 5.3-23 2 8/97 5.3-24 2 8/97 5.A-1 2 8/97 5.A-2 2 8/97 5.A-3 2 8/97 5.A-4 2 8/97 5.A-5 2 8/97 5.A-6 2 8/97 5.A-7 2 8/97 5.A-8 2 8/97 Update-8 8/09

TMI-2 POST-DFPFIJFI,1NG MONITORID',STORACGP SA-ETY ANALYSIS RIPORT List of Effective Pages Page # Date 5.A-9 2 8/97 5,A- 10 2 8/97 5.A4 1 2 8/97 5.A-12 2 8/97 5.A-13 2 8/97 5.A-14 2 8/97 5.A-15 2 8/97 5.A-16 2 8/97 5.A-17 2 8/97 5.A-18 2 8/97 5.A-19 2 8/97 5.A-20 2 8/97 5.A-21 2 8/97 5.A-22 2 8/97 5.A-23 2 8/97 5.A-24 2 8/97 5.A-25 2 8/97 5.A-26 2 8/97 5.A-27 2 8/97 5.A-28 2 8/97 5.A-29 2 8/97 5.B-1 2 8/97 5.B-2 2 8/97 5.B-3 2 8/97 5.B-4 2 8/97 6-i 2 8/97 Update-8 8/09

TMI-2 POST-DF.FT I ING MONITORED STOR AGE S AFFTY ANAL YSES REPORT List of Effective Pages Pag # Uipdate.# Date 6-ii 2 8/97 6-iii 2 8/97 6-iv 2 8/97 6-v 2 8/97 6-vi 2 8/97 6-vii 2 8/97 6-viii 2 8/97 6-ix 2 8/97 6-x 2 8/97 6-xi 2 8/97 6-xii 2 8/97 6.0-1 5 8/03 6.1-1 3 8/99 6.1-2 2 8/97 6.1-3 3 8/99 6.1-4 3 8/99 6.1-5 2 8/97 6.2-1 3 8/99 6.2-2 2 8/97 6.2-3 2 8/97 6.2-4 2 8/97 6.2-5 2 8/97 6.2-6 2 8/97 6.2-7 3 8/99 6.2-8 2 8/97 6.2-9 2 8/97 Update-8 8/09

TMI-2 POST--DF-IF, ING MONITORED STORAGESAFETY ANAL.YSIS REPORT List of Effective Pages Pagp # Tjpdaite #Dak 6.2-10 2 8/97 6.2-11 2 8/97 6.2-12 2 8/97 6.2-13 2 8/97 6.2-14 2 8/97 6.2-15 2 8/97 6.2-16 2 8/97 6.2-17 2 8/97 6.2-18 2 8/97 6.2-19 3 8/99 6.2-20 2 8/97 6.2-21 3 8/99 6.2-22 -3 8/99 6.2-23 2 8/97 6.3-1 2 8/97 6.3-2 2 8/97 6.3-3 2 8/97 6.3-4 2 8/97 6.3-5 2 8/97 6.3-6 2 8/97 6.3-7 2 8/97 6.4-1 2 8/97 6.5-1 2 8/97 6.5-2 2 8/97 6.5-3 3 8/99 6.5-4 3 8/99 Update-8 8/09

TMI-2 POST-DEFI 1ELING MONITORED -STORAGE SAFETY ANAl ,YSRS REPORT List of Effective Pages Page # p /97 6.5-5 2 8/97 6.5-6 2 8/97 6.5-7 2 8/97 6.5-8 3 8/99 6.5-9 3 8/99 6.5-10 3 8/99 7-i 2 8/97 7-iii 2 8/97 7-iv 2 8/97 7-jv 2 8/97 7-v 2 8/97 7-vi 2 8/97 7-vii 2 8/97 7-viii 2 8/97 7.0-1 2 8/97 7.1-1 8 8/09 7.1-2 2 8/97 7.1-3 2 8/97 7.1-4 4 8/01 7.1-5 4 8/01 7.1-6 6 8/05 7.1-7 2 8/97 7.1-8 2 8/97 7.1-9 2 8/97 7.2-1 2 8/97 7.2-2 2 8/97 Update-8 8/09

TMI-2 POST-DEFT TEO ING MONITORED STOR AGE SAFETY ANAl YSIS REPORT List of Effective Pages Page # Updxate.. Date 7.2-3 6 8/05 7.2-4 8 S8/09 7.2-5 8 8/09 7.2-6 6 8/05 7.2-7 2 8/97 7.2-8 4 8/01 7.2-9 8 8/09 7.2-10 2 8/97 7.2-11 2 8/97 7.2-12 2 8/97 7.2-13 6 8/05 7.2-14 4 8/01 7.2-15 6 8/05 7.2-16 2 8/97 7.2-17 2 8/97 7.2-18 4 8/01 7.2-19 6 8/05 7.2-20 4 8/01 7.2-21 2 8/97 7.2-22 5 8/03 7.2-23 5 8/03 7.2-24 6 8/05 7.2-25 8 8/09 7.2-26 7 8/07 7.2-27 2 8/97 7.2-28 6 8/05 Update-8 8/09

TMI-2 POST-DEFR IET .NC MONITORERD STORAGE SAFETY ANAl YSIIS REPORT List of Effective Pages Pag # Ilpdate..# Date 7.2-29 2 8/97 7.2-30 2 8197 7.2-31 2 8/97 7.2-32 6 8/05 7.2-33 2 8/97 7.2-34 2 8/97 7.2-35 2 8/97 7.2-36 2 8/97 7.2-37 3 8/99 7.2-38 2 8/97 7.2-39 2 8/97 7.2-40 2 8/97 7.2-41 2 8/97 7.2-42 2 8/97 7.2-43 2 8/97 7.2-44 2 8/97 7.2-45 2 8/97 7.2-46 2 8/97 7.2-47 2 8/97 7.2-48 2 8/97 7.2-49 2 8/97 7.2-50 2 8/97 7.2-51 2 8/97 7.2-52 2 8/97 8-i 1 6/95 8-ii 1 6/95 Update-8 8/09

TMI-2 POST-DFFIJFLING, MONETORED STOR AGE SAFETY ANAIlYSIS REPORT List of Effective Pages Pagp # I Ipdatef# Date 8.1-1 2 8/97 8.1-2 2 8/97 8.1-3 2 8/97 8.1-4 2 8/97 8.1-5 2 8/97 8.1-6 2 8/97 8.1-7 2 8/97 8.1-8 2 8/97 8.1-9 2 8/97 8.1-10 2 8/97 8.1-11 2 8/97 8.1-12 2 8/97 8.1-13 2 8/97 8.2-1 2 8/97 8.2-2 2 8/97 8.2-3 2 8/97 8.2-4 2 8/97 8.2-5 2 8/97 8.2-6 2 8/97 8.2-7 2 8/97 8.2-8 2 8/97 8.2-9 2 8/97 8.2-10 2 8/97 8.2-11 2 8/97 8.2-12 2 8/97 Update-8 8/09

TMI-2 POST-DEFI JET ING MONLTORED STORAGE SAFETY ANAlYSIS REPORT List of Effective Pages Page # I Indate~ Date ChapterL9 Deleted 10-i 7 8/07 10-ii (Deleted) 4 8/01 10.0-1 8 8/09 10.1-1 2 8/97 10.2-1 8 8/09 10.3-1 8 8/09 10.4-1 8 8/09 10.5-1 8 8/09 10.5-2 7 8/07 10.5-3 8 8/09 10.5-4 8 8/09 Update-8 8/09

A section of the York Haven Dam blocks the east channel, of the Susquehaina River at Three Mile Island approximatelyone mile downstream from TMI-2,. The York Haven Dam forms ýa pool, extending approximately.341/2 miles upstream containing a volume of about 10,000 acre-feet. As long as the river flow is 20,000 cfs or less, all the flow discharges through the York Havein Hydro Plaiit tailrace into the lower section of Conewago Falls. When the river flow is above 20,000 cfs, the excess flow spills over the portion of the main dam upstream of the headrace wall and flows down through the ConeWago Falls joining the flow from the tailrace at the foot of the dam; the full river flow then continues through the lower section of the falls. The exact extent of the mixing of TMINS effluents with the river depends on such factors as station discharge flowrate and the river flowrate. In 1980, an experiment was conducted which tracked the dispersion and dilution of a dye from the TMINS discharge. This study showed that plant discharge water and Susquehanna River water are typically 99% or greater mixed before intake by downstream users.

TMINS effluent releases are diluted by the effluent of the mechanical draft cooling towers by factors typically ranging from 167 times to 7600 times, depending upon'the wateruintake from the river. The diluted radiological waste is further diluted by the flowof the Susquehanna River. At low flow (1700 cfs), the river would dilute the effluent by~a factorof 20. However, at normal river flow (34,000 cfs), the dilution from the river would be 403 times. For effluent releases a minimum flow rate of 5000 gpm is maintained from the Mechanical Draft Cooling Towers, however, 8,000 gpm MDCT flow is always maintained during IWTS and/or IWFS releases (continuous -releases)., Dilution-credit up to 38,000 gpm is taken for Mechanical Draft Cooling Tower effluent ratesin FSAR calculations.

All users of surface water downstream of Three Mile Is.land are"also downstream of York Haven Dam.

Therefore, it is assured that mixing of station effluent and river water flow will occur prior to use.

2.4.5 GROUNDWATER

Three Mile Island Nuclear Station is located in the Triassic lowland of Pennsylvania, a region often referred to as the Gettysburg Basin. The island was formed as a result of fluvial deposition by the Susquehanna River.

It is composed of sub-rounded to rounded sand and gravel, containing varying amounts of silt and clay. Soil depths vary from approximately 6 feet at the south of the island to about 30 feet at the center of the island.

The site is underlain by Gettysburg shale which is at approximately 277 feet elevation.

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

2.4-3 UPDATE 2 - AUGUST 1997

I Hydrostatic pressure of the water table on the east and west shores of the river should prevent the island groundwater and the station discharge from communications with onshore groundwater. Therefore, groundwater effluents from TMINS cannot impact the quality of groundwater off-site. Additionally, the tritium concentrations in the TMINS groundwater are well below 10 CFR 20 regulatory limits and will not adversely affect the Susquehanna River.

A more thorough description of the groundwater characteristics and related features are given in Section 2.4.13 of the Unit 2 FSAR.

2.4.6 EMERGENCY OPERATIONS REQUIREMENTS 2.4.6.1 Flood Protection Although the flood protection design features of the station are based'on the PMF, the emergency operational procedures are based on forecasts received from the Federal-State River Forecast Center, National Weather Service, State College, Pennsylvania. Communications are normally by phone-or through civil defense radio as a backup.

The emergency procedure for the station will go into effect.when the'Federal-State River Forecast Center forecasts in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, a river flow of 350,000 cfs. A flood ALERT will be initiated when ad.36hour forecast of 640,000 cfs is received and a EMERGENCY CLOSURE when a 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> forecast of 900,000 cfs or greater is received. During the EMERGENCY CLOSURE, flood panels will be moved into place' 2.4-4 UPDATE 8 - AUGUST 2009

Shielded Work Platform Assembly UPDATE 8- AUGUST 2009 SHIELDED WORK PLATFORMASSEMBLY NOTE: For POMS the handrage have bean removed and a niolslwelcontenlnaalon barrer consisting of a lead brick FIGURE 4A-1 eftmney.wee intalled on top of the platfarm. Thi ahhnney also provides accese for a W to 200' F RTI) PAGE 4A-5 ouspended In the flReotor Vessel. The RTD hseea remote digllaireadout in 1heunit 2 Control Room, I

7.1 OPERATIONAL FACILITIES TMI-2 facilities required to be operational during PDMS are describedin this section. Facilities are required to be operational to support operational systems within those facilities and/or to isolate internal contamination from the environment. Table 7.1-1 provides a listing of operational facilities for TMIl-2 during PDMS. Each internally contaminated facility is identified-along with any relevant remarks regarding the final layup of the facility.

Each of the following sections addresses the PDMS function of the facility, the facility description, and applicable evaluations. Additional reference information is listed in Section 7.3.

7.1.1 CONTAINMENTY(REACTOR-BUILDING) 7.1.1.1 PDMS Function The primary function of the Containment during PDMS is as a contamination barrier. The Containment provides shielding of the environment from the contained radiation. It also-provides the means to assure that any effluents from the Containment will be controlled,- filtered, and monitored.

The Containment was designed to withstand approximately 60 psi of internal pressure, airplane crashes, a safe-shutdown-earthquake, tornado's, floods and other natural phenomena;' The Containment Was also designed with the capability to isolate any radioactive materials produced as a result of accidents or other unplanned events...

Although modifications were made to several of the Containment penetrations, (See Table 3.7-1) the structural capabilities of the Containment were not significantly diminished by the accident or any of the cleanup activities and are expected to be retained through the PDMS period. The Containment design and isolation capabilities relied upon during PDMS are described in the following sections..

7.1.1.2 Containment Structure The Containment is a reinforced concrete structure composed, of cylindrical walls with a: flat foundation mat and a dome roof and lined with a carbon steel liner. The structure provides biological shielding for normal and unanticipated conditions. The steel liner encloses the equipment and systems which remain inside the Containment and ensures that an acceptable upper limit of leakage of radioactive material will not be exceeded under the worst unanticipated event. - - .

The foundation slab is reinforced with conventional, carbonrsteel reinforcing. The cylindrical walls are prestressed!with a grouted tendon post-tensioning system in-the vertical and horizontal directions. The dome roof is prestressed utilizinga three-way grouted, tendon post-tensioning system. The-insidesurface of the Reactor Building is lined with carbon~steel to ensure a high degree of leak tightness. The thickness of the liner plate is 3/8 in. for the cylinder, 1/2 in. for the dome and 114 in.-for the base. .

7.1-1 UPDATE 8 - AUGUST 2009

The foundation mat bears on the bedrock and is 11 feet 6 inches thick with an additional 2 foot thick concrete slab above the liner plate. The cylindrical portion has an inside diameter of 130 feet, wall thickness of 4 feet, and a height of 157 feet from the topof the foundation slab to the spring line. The roof is a shallow dome which has a large radius of 110 feet and a transition radius of 20 feet 6 inches.;

7.1.1.3 Containment Functional Design During PDMS, the Containment serves primarily as a contamination barrier and provides shielding from the radiation due to contained contamination., All effluents will be controlled, filtered, and monitored; The functional requirements for the Reactor, Building during-PDMS are listedbelow:,

a. The Containment pressure will be maintained at equilibrium withiatmospheric pressure by utilizing a passive ventilation system (see Containment Atmospheric Breather, Section 7.2.1.2) via the Auxiliary Building.
b. Containment-isolation willbe maintained -by.a&single passive barrier either inside or outside of Containment on each Containment penetration. Active isolation capability is not required for PDMS except forthe Containment Atmospheric&Breather and the RB Purge Containment valves. Various passive means, or their equivalent, are acceptable for piping systems and include locked closed valves, closed and deactivated remote manual valves, closed and deactivated automatic valves, and blind flanges.
c. Monitoring of effluent releases will be provided by existing and/or additional monitoring equipment as designated in Section 7.2.4.

7.1.1.4 Facility Description Systems within the Containment-not required to be maintained in an operational condition during PDMS have been deactivated. The electric power circuits in the Containment have been deenergized except for those necessary for PDMS monitoring, inspection, and surveillance equipment, anid other PDMS support requirements. Prior to each inspection inside the Containment, circuits will be energized to provide lighting and power for required equipment.

7.1.1.5 Evaluation:

The Containment was originally designed to withstand airplane crashes,-seismic even!ts, tornadoes, floods, and other natural phenomena. Although there were modifications made during the cleanup period to several of the piping penetrationsi,these modifications were"performed'so that the striutural integrity of the Containment has been maintained. Neither the accident nor any activity during the cleanup period has, significantly degraded any of the structural capabilities of theContainment. Therefore, the Containment is structurally capable of,withstanding the original design basis events (except internalpressurization)'during the PDMS period without further analysis.

The internal pressure of the Containment during PDMS is controlled by a passive breather system (see Section 7.2.1.2). This system will maintain the Reactor Building in equilibrium with atmospheric pressure (via the Auxiliary Building) at all times during its use. In addition, a range of postulated events has been investigated (see Chapter 8) and none of these events could result in any significant pressurization of the Containment. Therefore, even with the reduced pressure capabilities due to the 7.1-2 UPDATE 2 - AUGUST 1997

Containment and the Breather HEPA filter that will automatically closeupon receipt of a Containment pressure increase of 1/4/psi. The purpose of this isolation is to protect the Breather HEPA filter in the event of a significant fire in the Reactor Building.1 The Breather is operated in the following modes:

  • PassiveBreathing - AH-V-3A, AH-V-52, AH-V-153, AH-V-154 and AH-V-25 are open and AH-V-4A and AH-V-120A are closed. A filter housing door downstream of AH-F-33 is opened. In this configuration, the Reactor Building is allowed to naturally aspirate via a HEPA-filtered pathway to the Auxiliary Building which, in,turn, either naturally aspirates or is ventilated to the environment through yet another set of HEPA filters. In the event of loss of air or loss of power, AH-V52 will fail closed.
  • DOP Testing - DOP testing of the HEPA filter is performed without the sample filter paper frames in place. ANSI N510-1980, Testing of Nuclear Air-CleaningSystems will provide guidance in the performance of DOP testing of the HEPA filter.

Prior to operation of the RB Purge System, the RB Breather will be isolated. In this configuration, valve AH-V-120A, AH-V153, AH-V154, AH-V-25 and AH-V-52 will be closed and valves AH-V-3A and AH-V-4A will be open.

Provisions have been made to allow annual sample'filter paper removal and assay and reinstallation or replacement of the HEPA filter.

7.2.1':2.3 Evaluation.,

This section demonstrates that the Containment Atmospheric Breather is the,"most probable pathway" by which the Containment can discharge air to (or intake air from) the environment. It is presumed that the Containment Atmospheric Breather can be deemed, the "most probable pathway". if the mass flowrate through the breather system in response to an atmospheric pressure change is orders of magnitude larger than the mass flowrates through all other pathways in respohse to the same pressure changes.

The mass flowrate through the Breather in response .to a pressure.differential can be calculated from its flow resistance. Similarly, the mass flowrate throutigh "all oth'r paths" can be calculated from the flow resistance for "all other paths." The flow resistance for "all-other paths" can be calculated from the rate it which the pressluire i the Containmuent attempted to achieve atmospheric equilibrium when the Containment was sealed and pressurized.

For the purpose of calculating the flowrates, theContainment is visualized as shown.in Figure 7.2-1.

The variou's known or potential leaks have been lumped together as an "equivalent" leak. The Containment Atmospheric Breather has been modeled as a 30 ft. straight length of 6-inch diameter pipe with 6ne HEPA.filter,.

t1he maximum overpressure in Containment from the postulated worst case fire is estimated to be approximately 5 psig.

7.2-3 UPDATE 6 - AUGUST 2005

The flow through-the Breather vent and the "equivalent" leak will be calculated usingthe extended Bernoulli equation. For the Breather:-

(P./p)+ (V, /2g1 )= (P./P)+ [f (ld)] (Vv/2g1 +

)+(DPp) I where:

P= the pressure inthe Containment p density of Containment air the (negligible) velocity in the Containment g, gravitational constant = 32.17 Lbm-ftLbf-sec.

Pa = the ambient pressure ,

I = pipe length d = pipe diameter f = friction factor for flow through the 30 ft. pipe.

V, = velocity in the Breather vent pipe DPf pressure drop across the HEPA filter Since the breather system is designed to allow the Containment to respond to small changesin, atmospheric pressure, the pressure differences will be small and the flow in the 6 inch diameter pipe will be assumed to be laminar. In that case, f = 641R. 64/(d Vyp/1P) , where p is the absolute Viscosity of air.

The velocity in the pipe caw be written interms of the masis flowrate in the pipe:

(2))

VvI=` (Oal" ,-" (2) "** >"

where:

=the mass flow rate in the Breather vent pipe d=' the diameter of the Breather vent pipe" The pressure drop across the HEPA filter can 'also written in terms of the mass flowrate in the pipe:

(DPf 1p)=(Khv/1p2) (3) where:

K is the rated pressure drop across the'filter of i-inch of water af 1i0006CFM which equals 0.312 Lbf-sec/ft. (Note that this assumes the HEPA filter pressure drop is linear with flow. This is conservative since the actual pressure drop will be less at the expected lower than 1000 CFM flow rate which in turn would allow more flow through the Breather.)

7.2-4 UPDATE 8 - AUGUST 2009

Substituting into equation (1):

my =(Pý - Pa)! [(128p/(wpgý)) 11d")+ (KIP)] (4)

For the "equivalent" leak, the result is the same except that the term for the HEPA filter is absent.

7, =(P - P./ [(128p.(yrpgc ))( l/d4)] (5)

In this case, the quantity (MVd') must be determined. To find an equivalent valtie of (lid') for the leaks, data from the leak test of the Containment were used. In the test, the proportional leak rate was calculated as 0.0852% per day when the Containment was' held at 70.6 psia. Since the pressure in the Containment is proportional to the air mass in the Containment, the proportional leak rate is the leak path mass flowrate, ml, divided by the air mass in the Containment at the time of the measurement.

0.0852%/day = M,/M= I ,M 1/p V)

'where:

M = the Containment air mass V = the Containment free volume = 2.1E6ft3 Converting the leak rate from % per day to inverse seconds, and combining with equation (5):

(0.000852/(24X 3600))= (P. - P. )/ [(pvX128M/(prr g. ))(I/0d )4 (6)

As a result:

(I/do )Ieak = - )(/[.(128Ur/rTg 0 )(o.000852/(24 x 3600))]

Which, in turn, leads to the determination that the equivalent value of (lid') for the leak paths is:

3 (Idlid)Mk =2.61E10ft-

[If the length of the leak path is on the order of the Containment wall thickness (i.e., 4 ft), the total leak diameter would be 0.042 inches.]

The ratio of leak flow to Breather vent flow can then be written as:

mi/m,,t. = [(128./7rg.)(/1d4).t.+ K] [(121,p/rrg.X/d4),,,] (7) with:

(I/d 4).n = 480ft-3 K=0.312 Ib- sec/ft5 This gives the ratio of mass flow rates as:

(m1/mh )= (0.00717 + 0.312)/ 3.89E5 =0.000001 Therefore, the Containment Atmospheric Breather clearly is "the most probable pathway".

7.2-5 UPDATE 8 - AUGUST 2009

As stated in Section 7.2.1.2.2, there is a welded plate installed downstream of the HEPA filter that holds four sample filter paper frames; each frame holds a set of two filter papers. The air flow into and out of the containment via the Breather also passes through each set of two sample filters. (For the purposes of this discussion, the sample filter papers closest to the Breather HEPA will be referred to as the No. 1 Filters and the sample filter paperstfarthest from the Breather -EPA andclosest to tlhe AuxiliaW Building atmosphere as the No. 2 Filters.) The Breather HEPA filters the air leaving the Containment intothe Auxiliary Building. Filter No. 1 collects the material that may pass through the HEPA filter. Filter No. 2 filters and samples the air coming back into the Containment from-the AuxiliaryBuilding.

All four of the No. I Filters are removed annually and one is assayed for radioactivity. Ifany activity is t found on the filter, it wl be assumed that, for the sampling time period, a like amount ofactivity was released from the Containment into the Auxiliary Building (ize, an assumed efficiency of 50%). This is a very conservative approach since the sample filter papers used have a collection efficiency of greater than 90%. Using this methodology, any activity assumed to be released is captured onthe filter and will be assumed to have been released over the six month time pe*rid. Since the filterdeJposition is cumulative, this method provides determinative (but not real time) monitoring to verify that effluents through the Breather are within the calculated values in Chapter 8. Due to the extremely low releases calculated for PDMS, the sample filter paper is deemed adequate for determining the releases anticipated during PDMS.

7.2.1.3 Containment Ventilation and Purge 7.2.1.3.lPDMS Function, During-PDMS, the Containment Ventilation and Purge System ensures that uncontrolled atmospheric migration of radioactive contamination will not create a hazard to either the public or site personnel.

7.2.1.3.2System*Description,  :; *  :

The Containment Ventilation andtiPurge System will be maintained in an operational condition to support activities in the Containment (e.g., surveillance entries, maintenance) during PDMS.: Testing to ensure operability of the Containment Ventilation and Purge includes HEPA filter pressure drop, exhaust flow rate, DOP testing (guidance provided by ANSI N5 10-1980), and visual inspection ofthe filter train. The Containment Ventilation and Purge System consists of a single operational Containmentpurge exhaust unit, make-up air supply associated ductwork, dampers, and filters. The purge exhaust unit (maximum flow 25,000 cfm) draws air from the D-rings through HEPA filters, and discharges to the station vent. The PDMS configuration is shown on GPUN Drawing 302-2041'.

7.2.1.3.3Evaluition Operation of the Containment Vent and Purge System provides fresh air to the Containment while providing a filtered, monitored exhaust path. Atmospheric radiation monitoring, as described in Section 7.2.4, provides for monitoring of airborne releases from the system by using monitorsloc'at6d in the exhaust duct and in the station vent. This ensures that releases from the Containment to the environment are minimal.

7.2.1.4 Containment Airlocks and Equipment'Hatch 7.2.1.4.1.1 PDMS Function 7.2-6 UPDATE 6 - AUGUST 2005

d. The Halon systems protecting the Air Intake Tunnel have been deactivated by removing the Halon cylinders and deenergizing the ultraviolet and pressure detectors.

.The Halon system protecting the relay roomhas been deactivated by removing the cylinders. The heat sensitive wire fire detectionsystem will remain operational to monitor these areas.

e. Portable fire extinguishers and self-contained breathing apparatus are staged with emergency response crew equipment.: Additional. portable fire extinguishers are located throughout the plant as needed to support work activities.
f. Transient combustibles inside the Containment and the AFHB have been removed to the maximum extent practical.
g. Deleted. " .
h. The charcoal filters havebeen removed from all HVAC systems in Unit 2.
i. The 12 in. fire service loop, which runs through the Diesel Generator Building, AFHB, Control Building area and Turbine Building (east and west), has been cut and capped off. The Diesel Generator Building has been turned over to TMI-1. Fire Service Water System standpipes have been configured to the East and West side of the Turbine Building which permits connection of the local fire hydrants to the 331' elevation of the Turbine Deck by way of staged fire hoses. This will allow responsive action by the Station Fire Brigade and or local Fire Departments.
j. ' DELETED
k. The station fire brigade is fully trained to assure that the personnel are familiar with system configurations, plant layout, and the procedures in Unit 2.
1. The Fire Protection Program and housekeeping inspections and their frequency are addressed in plant procedures.

2

'The air intake tunnel halon system was removed because the probability of an airplane crash in the vicinity of the air intake tunnel was estimated to be less than 2E-7/year and because of the presence of heat activated detectors.

7.2-9 UPDATE 8 - AUGUST 2009

7.2.2.3 Evaluation The scope of fireprotection has been reduced for areas in which systems have been deactivated and combustibles have been significantly reduced, so that the corresponding fire hazards have been minimized.

Deluge systems in the air/intake tunnel, the Auxiliary, the Turbine, and-Control Buildings have been deactivated for PDMS. There are no deluge systems in the Containment.

Detection devices provide contacts for supervisory indication that each device is operational and, in the event of detector actuation, indicates,the.location wherethe..wiredetector actuated the alarm.

The station fire brigade is under the supervisory control of Unit 1. Upon detection of a fire in Unit 2, the station fire brigade will respond to the specific location in Unit.2:. This response in accordance with ongoing station fire brigade training and procedures will ensure mitigation of a fire in Unit 2 during PDMS. The fire protection,and, suppression systems areconfigured to provide adequate capability to extinguish any potential fire during PDMS.

7.2-10 UPDATE 2 - AUGUST 1997

7.2.6.5.3 Evaluation In preparation forPDMS, various building seams, link seals, and major cracks have been repaired to the extent practical to minimize expected inleakage from storms and high groundwater levels. The inleakage rates and flowpaths experienced to date do not affect plant equipment required for PDMS.

Additionally, the Sump Pump Discharge and WDL system are operational to transfer accumulated water to minimize potential spread of contamination due to localized flooding.

7.2.6.6 Sewers 7.2.6.6.1 PDMS Function The basic function of the sewage collection system is to transport sewage from TMIvl-2 structures to the Sewage Treatment Plant. The PDMS configuration is shown on GPUN Drawing 302-151.

7.2.6.6.2 System Description Sewage from the personnel access facility (PAF) in the Turbine Building is routed to'the Sewage Treatment Plant (STP) which serves both TMI-land TMI-2. The major operational portion of the Sewer System is underground gravity flow piping that provides for the-transport of sewage from the Unit 2 support facilities to the STP.

7.2.6.6.3 Evaluation The Sewage Treatment Plant will process sewage from the PAP. The majority of TMI-2 sewage piping is underground below the frost line. The original plant sanitary waste/sewage system is deactivated.

7.2.6.7 Domestic Water System 7.2.6.7.1 PDMS Function During PDMS, portions of the existing domestic water system will remain operational to provide domestic water services required during PDMS.

7.2.6.7.2 System Description The domestic water system is maintained as a modified operational system., Unit 2 is supplied with domestic water from Unit I which is then distributed to Unit 2 support facilities. Domestic water is provided to the radwaste seal water unit in the Auxiliary Building, to the PAF in the Turbine Building, and to several outbuildings. The PDMS configuration is shown on GPUN Drawings302-158 Sht. 4.

7.2-25 UPDATE 8 - AUGUST 2009

7.2.6.7.3 Evaluation Since personnel access into the plant will be infrequent, only one source of domestic water is required in the Turbine Building. The Auxiliary;Building header supplies domestic waterto the seal water unit. Unit 1 and Unit 2 support facilities will remain operational; therefore, domestic water will continue to be supplied.

7.2.6.8 Control Room Ventilation System 7.2.6.8.1 PDMS Function The Control Room Ventilation System will be maintained in an operational condition to support PDMS activities. This system provides fresh, filtered; heated or cooled air in~sufficient quantity to support personnel-occupancy and equipment protection.

7.2.6.8.2 System Description The Control Room Ventilation System consists of one supply fan (AH-C-16B) running in a forced ventilation mode during normal year round conditions. The supply fan willprimarily recirculate the control room air as it is heated/cooled. A small amount of fresh, air (outside air),willbe&force supplied by bypass booster fan (AH-C-16X). Exhaust fan (AH-E-35) will return control'room airto the suction of supply fan (AH-C-16B). A small amount of the control room air will be "exhausted" out of this recirc mode, primarily by exfiltration dampers in the control room and via the kitchen and toilet fans.

This provides for a small amount of air change per day.

Control Room air temperature is monitored by a temperature element locatedin theControl.Room return air duct. The temperature element provides Signals to a programmable thermostat, which controls heating or cooling as conditions dictate. Two steps of heating are available for freeze protection and cooling is available from the 10 ton air conditioner Which also reduces Control Room humidity.

Neither cooling or heating functions will operate unless supply fan (AH-C-16B) is running and satisfying a flow switch in the supply air duct.

Additional outside air can be provided by performing special operations if the chiller malfunctions and/or additional cool outside air is desired.

7.2.6;8.3 , Evaluation.

During PDMS,, Control Room, ventilation and air handling equipment provides a filtered pathway for active operationto meet industrial and radiological requirements. The Control Room Ventilation System is maintained operational for the maintenance and surveillance entries into the TMI-2 Control Room and in response to off-normal conditions.

7.2-26 UPDATE 7 - AUGUST 2007

CHAPTER 10 ADMINISTRATIVE FUNCTIONS

10.0 INTRODUCTION

The primary administrative functions necessary for the management of TMI-2 during PDMS are referenced in this chapter.

On December 23, 2008, the NRC issued an Order approving the transfer of the operating license for Three Mile Nuclear Station, Unit 1, held by AmerGen Energy Company, LLC, to Exelon Generation Company, LLC (Exelon Generation Company). Thus, any reference in this Section to Exelon Generation Company is made based on this Order.

10.0-1 UPDATE 8 - AUGUST 2009

10.2 SECURITY PLAN The Code of Federal Regulations 10 CFR 50 and 10 CFR 73 define the security requirements for nuclear power plants. Due to the defueled and non-operating condition of TMI-2 during PDMS, the security requirements applicable to the facility are less than those that are applicable to an operating nuclear power plant. TMJ-2 complies with all applicable security requirements. TMI-2 utilizes site physical security, guard training and qualification, and safeguards contingency plans maintained by TMI-1. These plans are administered and are under the authority of Exelon Generation Company, the TMI-I License holder. The specific security provisions for TMI-2 are in the "TMI Modified Amended Physical Security Plan."

10.2-1 UPDATE 8 - AUGUST 2009

" I 10.3 EMERGENCY PLAN 10 CFR 50.47 establishes requirements for the content and criteria for acceptance of emergency plans. Emergency planning requirements are based on the assumption of the potential necessity to notify the public of the existence of, or potential for significant off-site releases. 10 CFR 50 Appendix E recognizes that emergency planning needs are different for facilities that present less risk to the public. Due to the non-operating and defueled status of TMI-2 during PDMS, there is no potential for any significant off-site radioactive releases and, due to the existence of TMiI-1 on the same site, emergency planning requirements for the site are dominated by TMI-1. Therefore, the limited emergency planning necessary to accommodate the existence of TMI-2 on the same site as TMI-1 has been incorporated into one integrated emergency plan. There exists only one Emergency Preparedness Plan for the TMI station. The Plan encompasses both TMI-1 and TMI-2 and is under the authority of Exelon Generation Company, the TMI-1 License holder.

The emergency plan for the TMI site incorporates all of the essential emergency planning requirements established by 10 CFR 50 Appendix E and other regulatory guidance. Since there are no events associated with TMI-2 which could result in a release approaching the levels established in the Protection Action Guide, the site emergency action levels are based on potential events which could occur at TMI- 1. The site emergency facilities, such as the Emergency Control Center, the Technical Support Center, and the Operations Support Center are located in or in convenient proximity to TMI-1. All site personnel are trained and drilled to respond to any declared site emergency event 10.3-1 UPDATE 8- AUGUST 2009

,, 1 4 10.4 RADIATION PROTECTION PROGRAM TMI-1 maintains a Radiation Protection Program which meets or exceeds standards for protection against exposures to radiation and radioactivity at the TMI site. There exists only one Radiation Protection Program for the TMI station. The Program encompasses both TMI-1 and TMI-2 and is under the authority of Exelon Generation Company, the TMI-1 License holder. The implementation of the Radiation Protection Program ensures that the facility will be managed and maintained during PDMS in a manner which minimizes risks to employees, contractors, visitors, and the public from exposure to radiation and radioactivity at the facility. The implementation of the program also ensures a radiologically safe working environment for employees and visitors at TMI-2.

10.4-1 UPDATE 8- AUGUST 2009

10.5 ORGANIZATION The organizational elements responsibleifor the PDMS phaseý of TMI-2 are shown on Figure 10.5-1. The specific:responsibilities are discussed below. Additionally, the PDMS Technicall Specifications prescribe specific requirementsfor staff qualifications, training, and the review and audit of TMI-2 activities.

As part of the sale of TMI-1, GPU!Nuclear entered into an 'agreementwith AmerrGen, now Exelon Generation Company, for TMI-2 services. Under this agreement and as a contractorsubjectto GPU Nuclear's ultimate direction-and control, Exelon Generation Company will provide albservices;, materials and equipment required~to maintain TMI-2 in Post-Defueling Monitored -Storage{PDMS).. Services provided by Exelon Generation Company. will meet ,all the requirements of the Safety Aialysis Report, Technical Specifications and Quality Assurance Program: " Servi'cessin*Clude:

  • Management services;
  • Operations, maintenance and testing;
  • Radwaste operations;' .
  • Quality Assurance;
  • Radiation controls 'and health phyisics;';
  • Environmental 'controls;
  • Security;
  • Safety;
  • Administrative services, including logistical support, information technology support and procurement services;
  • Engineering and Licensing; SWarehousing and.housekeeping;
  • Support services 'required in coxinection with the performance.of routine corrective and. preventative miaimtenaun"e; 0 Interface with the NRC as necessary in connection with inspections, audits, site visits or meetings;
  • Maintain required NRC licensing documents for TMI-2; and prepare regulatory correspondence for GPU Nuclear signature or file on behalf of P

GPU Nucleai, to the extent permitted under applicable NRC regulations, all documents required in connection with PDMS of TMI-2.

On March 7, 2001 the NRC issued an Order approving the application regarding the proposed merger of GPU, Inc.'and FirstEnergy Corp. Aspart of this order it was recognized that the holders of the TMI-2 license, GPU Nuclear, Metropolitan Edison Company, Jersey Central Power & Light Company and Pennsylvania Electric Company would become subsidiaries of FirstEnergy Corp.

Thus any reference to FirstEnergy is made based on this relationship.

Figure 10.5-1 also shows the Exelon Generation Company organization which will provide the above services.

10.5-1 UPDATE 8 - AUGUST 2009

10.5.1. President and Chief Nuclear Officer The President and Chief Nuclear Officer is responsible'tfothe FirstEiiergy Nuclear Committee of the Board to provide top level directioni on-all activities associated with the safe and efficient management-and oversight of all TMI-2 activities. This position also serves as the GPU NuclearCognizant Officer.

10.5.2 Vice President GPU Nuclear Oversight The Vice President, GPU Nuclear Oversight is responsible to ensure the TMI-2 PDMS Quality,Assurance Program is maintained andimplemented in: accordance with the PDMS Quality Assurance Plan, and, applicable policies, and procedures, applicable laws, regulations, licenses and technical requirements. Additionally, the Vice President,

,GPUNuclearOversight :is responsible to manage, direct andprovide support to the GPU Nuclear Employee Concerns Program and is the sponsor of the TMI-2 Company Nuclear Review Board (CNRB).

10.5.3 GPU Nuclear Responsible Engineer Three Mile Island Unit 2 (TMI-2)

The GPU Nuclear Responsible Engineer, Three Mile Island Unit 2 (TMI-2) has the overall responsibility for the management of TMI-2 during PDMS. This overall responsibility may be shared by more than one individual.

10.5.4 Employee Concerns Program An Employee Concerns Program is provided for GPU Nuclear.. The Vice President, GPU Nuclear Oversight, is responsiblefor the program and will appoint an individual to administer the progiam. If necessary tis individual will have access to the Chief Nuclear Officer and FirstEnergy Nuclear Committeeof the Board.

This individual is accessibleon a confidential basis, if desired, to anyone in the company or its contracted employees having a nuclear orradiation safety concern he or she considers is not being adequately'addressýle; This individual is empowered to investigate such matters, identify any needed actibns and seek'its resolution. This indiidualI willtreply'o the Iperson who rised the concern.

10.5.5 TMI-2 Company Nuclear Review Board (CNRB)

Independent oversight is provided by the TMI-2 CNRB. The CNRB serves to independently assure that the TMI-2 structures, systems anid components are maintained so as to protect the health and safety of the workers, the publii and the environment and to enable effective and efficient dismantlement and decommissioning in the future. The CNRB is sponsored by the Vice Presidefit' GPU Nuclear Oversight.

10.5-2 UPDATE 7 - AUGUST 2007

A, 10.5.6 Manager, PDMS The Manager, PDMS has the first-level management responsibility for maintaining the TMI-2 PDMS condition. The Manager, PDMS is directly responsible for the operations and maintenance activities associated with the TMI-2 PDMS.

10.5.7 Organizational Commitments TMI-2 License Amendment and Technical Specification Change Request No. 78, submitted to the NRC on April 6, 2000, requested organizational and administrative changes that will exist following the sale of the Oyster Creek Nuclear Generating Station to AmerGen. Attachment 3 to that submittal listed a number of commitments for TMI-2 and a general commitment to list the commitments in the PDMS SAR. The listing, issued in TMI-2 Technical Specification Amendment No. 54 as modified to reflect current conditions, is as follows:

1. The GPU Nuclear Cognizant Officer will have overall responsibility for TMI-2. A description of responsibilities and qualifications for this position is addressed in the PDMS Quality Assurance (QA) Plan.
2. A FirstEnergy employee or third party contractor will be assigned to the TMI site.
3. The individual responsible for the Employee Concerns Program will have access, if necessary, to the FirstEnergy Nuclear Committee of the Board. This function is described in the PDMS Quality Assurance Plan.
4. GPU Nuclear will periodically assess Exelon Generation Company performance with support from other GPU (owners group) organizations as needed (e.g. GPU Internal Audits, Contracts, Legal, etc.).
5. GPU Nuclear will establish a TMI-2 Company Nuclear Review Board that will advise the GPU Nuclear Cognizant Officer. A description of responsibilities and qualifications is addressed in the PDMS Quality Assurance Plan.
6. AlI Quality Assurance audit reports prepared by Exelon Generation Company for TMI-2 will be provided to the GPU Nuclear Cognizant Officer.
7. GPU Nuclear will conduct a periodic QA Plan audit of Exelon Generation Company. The audit and frequency is specified in the GPU Nuclear PDMS Quality Assurance Plan.
8. A GPU Nuclear employee or third party contractor (ultimately responsible to GPU Nuclear) will review and approve all 10 CFR 50.59 evaluations unique to TMI-2 and all evaluations involving a TMI-2 facility change.

This is incorporated in the TMI Review and Approval Matrix.

9. A GPU Nuclear employee or third party contractor (ultimately responsible to GPU Nuclear) will review and approve proposed changes to the emergency preparedness program that are unique to TMI-2.

10.5-3 UPDATE 8 - AUGUST 2009

GPU NUCLEAR CORP.

ORGANIZATIONAL CHART i -* Employee Concerns Program Manager I

rf-GPU Nuclear Organization Exelon Generation Company TMI Nuclear Oversight, Manager (TMI-2 matters)

NUCLEAR TMI-2 GPU NUCLEAR CORP.

ORGANIZATIONAL CHART CAD FILE: SIA.SKOO0,.1000,001-,0203 PDMS SAR UPDATE 8 - Aug. 2009 THIS IS A COMPUTER GEN*IEAtD DRAVANG, DO NOT REVISE MANUALLY FIGURE: 10.5-1 PAGE 10.5-4

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