ML19322C771
| ML19322C771 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 01/15/1980 |
| From: | Metropolitan Edison Co |
| To: | |
| Shared Package | |
| ML19322C770 | List: |
| References | |
| NUDOCS 8001230154 | |
| Download: ML19322C771 (70) | |
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i FIFTH INTERIM REPORT ON THE i
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THREE MILE ISLAND NUCLEAR STATION UNIT II (TMI-2) ACCIDENT
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JANUARY 15, 1980
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CONTENTS j
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SEQUENCE OF EVENTS l
II.
RECOVERY ORGANIZATION III.
PLANT MODIFICATIONS IV.
DECONTAMINATION OF AUXILIARY AND FUEL HANDLING 3UILDING r
V.
RADIOLOGICAL MONITORING
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RECOVERY PLANNING j
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SPECIAL PROJECTS VIII.
REPORTABLE OCCURRENCES i
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SEQUENCE OF EVENTS i
- r Since this section is presently undergoing updating it is not available at_this time. An updated copy will~be forwarded in a future submittal.
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O II.
RECOVERY ORGANIZATION Included in this section are organization charts representing the TMI Unit II Recovery Organization for the period of October 1 through December 31, 1979.
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-Rev, 4
Sr. V.P. MET-ED V.P. CPUSC F
V.P. !!et-Ed/ l Director Ha8" Direct r Director Manager TMI-II Director Director THI-Ii Env/ifealth/
5 e Vii'es fes'h. Functions Reliability Eng.
Rad. Control TMI-II ll Safety Special Manager Projects Training Director Manager Site Operations THI-I Manager Manager Project Ops.
Rad. Control Manager Manager Admin & Serv.
Admin. & Serv.
'lifI GENERATION GROUP Manager Manager Recovery Eng.
Plant Eng'pr'ng Manager Manager QA/QC Task Mgmt.
III. PLANT MODIFICATIONS Included in this section are updated and anended subsections from previous Interim Reports.
Changes from the previous reports are denoted by change bars in the right hand margin and Rev. 4 on the bottom right hand corner of the page. Subsections from the previous Interim Reports which have not had any changes are not included in this report.
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Rev. 4 L
A.
Hydrogen Recombiners 1.0 System Function and Design Objectives i
In anticipation of having to process substantial amounts of hydrogen to prevent a hydrogen explosion in the reactor building, and because of the uncertainty of the quantity of hydrogen being generated, the available hydrogen recombiner capacity was increased.
Operations shall not permit an uncontrolled release of reactor building atmosphere to the environment.
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2.0 System Description
A thermal-type hydrogen recombiner is installed in the fuel r
handling building at the. spent fuel pool operating floor and is connected to the reactor building ventilation and purge system as originally intended (see FSAR Figure 6.2-30).
In addition, a skid-mounted, thermal-type hydrogen recombiner has been installed next to, and has been connected in paralled with, the i
first hydrogen recombiner. The integrity of the system is as originally installed and will ensure that there is no uncontrolled radioactive release to the environment.
3.0 System Operation The hydrogen recombiners are only operated as required to control reactor building hydrogen concentrations. Periodic samples are taken from the containment atmosphere to monitor hydrogen concentration levels.
Recombiner operation is monitored and controlled manually from
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a local panel. A recombiner " trouble" alarm is annunciated in the control room.
4.0 Status The Hydrogen Recombiners are installed and functional. a progran is in process to decontaminate one of the units and remove for eventual installation in Unit I.
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Rev. 4
B.
Auxiliary and Fuel Handling Building Supplementary Air Filtration Systems 1.0 System Function and Design Objectives Radioactive iodine, released from the Reactor Coolant System during the TMI Unit 2 accident, was transferred into the Unit 2 Auxiliary and Fuel Handling Buildings.
I==ediate change out of the Auxiliary and Fuel Handling Building charcoal filter trains was not feasible because of the high radiation and contamination levels in the filter areas. As a consequence of the I-131 release rate, it was decided to construct a supplementary air filtration system to reduce off-site releases. The function of the system is to remove radioactive particulates and radioiodine in the exhaust air from the Auxiliary and Fuel Handling Buildings.
2.0 System Description
The system interfaces with the Auxiliary, Fuel Handling and Service Building HVAC Systems, and is installed on the roof of the Auxiliary Building.
The system consists of four parallel filter trains, each comprised of an exhaust fan and a filter unit.
3.0 System Operation Air is drawn from the duct connecting the penthouse plenium to the station vent through a co= mon inlet duct which connects the four parallel filter trains.
The inlet duct is provided with an in-line radiation monitor.
Air then flows through the operating filter trains, through a radiation
=onitor in each outlet duct and is exhausted to the atmosphere through the fans.
4.0 System Status Eng?neering is Complete.
Constrw-tion is Complete.
System description, flow diagrams, and operating procedures, are complete.
An operating and failure modes analysis has been prepared.
All four (4) trains are operable. The station vent is capped. Present operation is with three (3) trains.
Rev. 4
D.
Fuel Pool Waste Storage System 1.0 System Function and Design Objectives This Fuel Pool Waste Storage System is used for temporary storage of liquid waste. These tanks add approximately 110,000 gallons to the present storage capacity of the plant, and are located within the "A" spent fuel pool.
These tanks can be filled with liquid waste from the Reactor Building Sump and the Miscellaneous Waste Hold-Up Tank.
This system enhances the capability of the plant to move and process radioactive waste.
2.0 System Description
The system consists basically of upper (4 at 15,000 gallons each) and lower (2 at 25,000 gallons each) tanks, forming two separate storage areas.
Either storage area is capable of being filled from either the Reactor Building Su=p or the Miscellaneous Waste Hold-Up Tank, and each has level indication.
The tanks are protected from over-filling by automatically closing the feed valve when the storage area is nearly full. Provisions have been made to both flush the piping system after completion of the pumping operation, and to drain the piping system as required.
The vents from the tanks and the scand pipes are directed through a dryer and a charcoal filter to remove moisture and iodine before proceeding to the fuel pool ventilation system. The tanks and vent system is protected by a relief valve which vents through a parallel set of dryers and charcoal filters.
The tanks will be emptied as necessary by steam eductors. Two eductors are permanently installed in each stand pipe.
3.0 System Operation Water is transferred from the Reactor Building Sump or the Miscellaneous Waste Storage Tank to the tank f arm..
After either the lower set of tanks or upper set of tanks is full the level controllers automatically close the air ;trated inlet valves.
Air forced from the tanks during the filling process is vented to a charcoal filter & dryer to remove noisture and iodine.
This air is then piped to the Fuel Fool Ventilation System.
The steam eductors give the capability to transfer waste water from the tank farm to the Miscellaneous Waste Storage Tank or Epicor II Rad Waste System, from the upper tanks to the lower tanks in the tank farm (or vice versa) or to recirculate the water in the tanks.
Rev. 4 I
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Fuel Pool Waste Storage System (continued)
A high temperature alarm and temperature switch to close the steam control valve, is installed in the tank vent line to prevent damage to the filter / dryer skids during use of the eductors.
4.0 System Status The system is being used to store 93,000 gallons from the Unit 2 Miscellaneous Waste System.
The steam eductors have not been used since no water has been pumped out of the tanks to date.
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Upgraded Decay Heat Removal System 1.0 System Function and Design Objectives Future operation of the existing decay heat removal (DBR) system may result in radiation levels possibly ranging up to 500 Rads per hour in the vicinity of the system fluid components. This condition would severely limit personnel access for routine surveillance, operation, and maintenance. The upgraded DHR system consists of a program intended to identify, evaluate, and implement modifications necessary to ensure the integrity and reliability of the system in a radiation environment, substantially exceeding the original design basis, for up to one year of operation.
2.0 System Description
Proposed DHR system modifications include additional decay heat vault shielding, a remote TV conitoring system, modified DHR pump and motor bearing oilers, a vibration monitoring system, and associated operating and testing procedures.
Vault shielding will be provided by lead bricks assembled in a steel support frame.
This will reduce the ambient personnel radiation exposure levels to "as low as reasonably achievable" (ALARA) in the accessible area above the vault. Radiation surveys will be made during initial DHR system operation aad periodically thereafter to determine shield effectiveness.
The TV monitoring system will provide remote surveillance capability for DHR system operation and maintenance.
Two independent systems are provided, one for each vault. Each system includes a radiation-tolerant, closed-circuit television with remote controls.
Specific operations to be monitored include pump and motor bearing oil level, pump packing leak-off, remote oil fill, and pump venting.
DHR pump and =otor bearing oiler modifications will provide for increased oil storage capacity, a means for re=otely reading oil levels, and to permit feeding of oil to the bearings.
Provision for remote venting of the pumps is also provided.
Provisions will be made for monitoring pump vibration and loose parts in the system.
This is intended to provide early indication of pump and =otor degradation, loose parts in the system (particularly at the heat exchanger tube inlet), and changes in flow patterns due to partial line blockages.
Monitoring and control for these modifications will be provided i
from the fan room at elevation 322 in the service building.
3.0 System Operation These modifications to the DHR will not appreciably alter system operation.
Rev. 4 f
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1 4.0 Status The TV monitoring system,.the bearing oil tanks and piping, I
and pump venting arrangement are installed and operational.
t The installation of vault hatch shielding plugging is in I
process in support' of the Mini Decay Heat System.
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Steam Generator "B" Closed Loop Cooling System 1.0 System Function and Design Objectives In order to provide a high pressure, closed cooling loop for water-solid steam generator "B", a system utilizing new equip-ment must be installed.
The closed loop must remove the decay heat.from the core plus the added heat load from one reactor coolant pump. To minimize the possibility for contamination of the closed loop, the system r.ust be operated at a higher pressure than the reactor coolant system.
The heat transferred to the closed loop will ultimately be rejected to the river.
The system is intended to provide backup decay heat removal capability should the present steaming from steam generator "A" be discontinued.
2.0 System Description
The system consists of a new heat exchanger, pump, surge tank, piping and valves. The hot water leaving the steam generator will pass through the tube side of the new heat exchanger and return to the steam generator via the new pump. A pressurizer surge tank will maintain the steam generator secondary side pressure above the primary coolant system pressure.
The shell side of the heat exchanger is supplied with cooling water from the secondary services closed cooling water system which, in turn, will be cooled by water from the nuclear services river water pumps piped to the turbine building via the secondary services river water piping.
The new pump discharge piping is connected to the existing feedwater piping downstream of :he main feedwater pumps, and the heat exchanger inlet piping is connected to the drain pot on the main stems line between the main steam isolation valve and main turbine stop valves, h
3.0 System Status System is operational and in standby.
Rev. 4 vy
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Steam Generator "A" Closed Loop Cooling System 1.0 System Function and Design Objectives In order to provide a high pressure, closed cooling loop for water-solid steam generator "A", a cooling system utilizing new equipment has been proposed. The closed loop would remove the decay heat from the core plus the added heat load from one reactor coolant pump. To minimize the possibility for contamination of the closed loop, the system would be operated at a higher pressure than the reactor coolant system. The heat transferred to the closed loop would be rejected to the river. The system would be intended to provide primary decay heat removal capability redundant to the steam generator "B" closed loop cooling system.
2.0 Description The system will consist of a new heat exchanger, pump, surge tank, and piping and valves.
The hot water leaving the steam generator would be cooled in the shell side of the heat exchanger and returned to the steam generator by a new pump.
A pressurized surge tank would maintain the steam generator secondary side at a minimum pressure greater than the primary coolant system pressure.
The tube side of the heat exchanger would be supplied with cooling water from the nuclear services river water pumps piped to the turbine building via installed secondary services river water piping.
The new pu=p discharge piping would be connected to the existing feedwater piping downstream of the main feedwater pumps. The heat exchanger inlet process piping would be connected to the main steam turbine bypass line between the isolation valve and the control valve at the condenser.
3.0 System Status This system will not be installed.
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Alternate Decay Heat Removal System 1.0 System Function and Design Objectives The proposed Alternate Decay Heat Removal (ADHR) system augments the two existing DHR systems and the proposed water solid secondary / natural circulation system as backup to steam generator "A" steaming.
An integral Decay Heat Closed Cooling Water (DHCCW) system is included to transport heat from the ADHR cooler and the ADER pump seal coolers to the nuclear services river water system. Connection points are also provided outside the fuel handling building to connect other dedicated liquid waste processing systems.
The specific function of the ADER system is to remove decay heat such that the reactor coolant system can be brought to and maintained at a cold shutdown condition. With the exception of gross core flow restrictions, this system is intended to provide sufficient core flow to maintain reactor coolant subcooled.
2.0 System Description
s The two ADHR pumps and a new heat exchanger will be mounted on a skid located outside the west wall of the fuel handling building and penetrate the fuel handling building west wall of a valve vault. The pipe runs will terminate in the valve vault by capping each line. Hook-up to the ADHR skid will be made later if needed. In addition, three capped taps will be provided on the ADHR piping installed outside the fuel handling building. These taps may be used later to connect other dedicated liquid waste processing syste=s.
Motor control centers and I&C panels for operation of all ADER system pumps and motor operated valves will be =ounted in a control trailer located near the ADER skid.
The DHCCW system provides cooling water to the ADER system heat exchanger and pump seal coolers.
It utilizes a closed loop system to provide a double barrier between the ADER system and the river water to prevent the direct release of radioactivity to the environ-ment. A radiation detector is provided to monitor the level of radioactivity in the DHCCW system at the outlet of the DHR cooler.
A radiation level indicator with high radiation level alarm is located in the ADHR system remote control room.
If radioactivity is detected, operation of the decay heat re= oval loop and its associated DHCCW loop can be halted and the affected decay heat removal cooler isolated. The DHCCW system is mounted on a second skid and consists of the DHCCW pump, heat exchanger, and surge tank.
Both skids will be located outdoors at grade level near the west vall of the fuel handling building and adjacent to each other.
Rev.
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3.0 System Operation A detailed description of this system is in the Westinghouse turnover document.
1 4.0 System Status The piping for the ADHR system has been designed, fabricated, and received on site. The skid for the ADER system with its components, two pumps, heat exchanger, valves and piping is' completed.
Motor control centers are on site.
The valve vault excavation is completed and piping installt. tion up to the second isolation valves is completed. The electrical trailer is completed.
EAectrical power and service water connections would not be made unless the system were put into service. Tie-in of the ADHR system to the existing plant DHR system has been completed. Disposition of the valve pit is under evaluation.
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Standby Reactor Coolant Pressure Control System 1.0 Systems Tunction and Design Objectives High radiation levels and flooding in the reactor building have or could potentially render much of the reactor coolant (RC) system electrical equipment and instrumentation inoperable. With much of the instrumentation inoperable, the RCS should be main-tained water " solid".
An alternate system of pressure control is required to ensure safe and reliable cooling of the reactor core, should control of the existing system become unmanageable. The standby reactor coolant pressure control (SRCPC) system will ensure reliable core cooling by performing the following function:
Maintain the RC system in a water-solid condition for a.
natural circulation core cooling.
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Maintain sufficient available N?SH should RC pump operation be required.
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Control the quality of the =akeup fluid.
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Maintain pressure within control limits while accomodating ther=al and volumetric contractions in RC system inventory, i
2.0 System Description
The SRCPC system ties into the existing High Pressure Injection lines (see FSAR Figure 9.3-6). RC system pressure is maintained by three surge tanks arranged in series with a pressurized nitrogen blanket over the last tank. A fluid inventory of approximately two thirds of the total tank capacity is sufficient to maintain RC system pressure during sudden RC system inventory reduction transients. A level control valve at the tanks' discharge will prevent nitrogen from entering the RC system.
Long term makeup will be provided by the charging pump taking suction from an atmospheric storage tank. Makeup fluid conditions are adjusted by chemical addition and heating to meet RC system water quality requirements.
The RC system pressure will normally be maintained between 50 and 600 psig during the intended cooldown process.
The SRCPC makeup system will be operated manually from a local panel during initial operation and from the control room after system automation is complete. Makeup is provided in response to decreasing pressure in the RC system. An alarm will annunciate at the control station when the pressure differential between the RC and SRCPC makeup system reaches or exceeds 50 psi.
Rev. 4 l
The SRCPC makeup system will prevent gross depressurization on the RC system when operating in a water-solid mode. Over-press'irization protection can be provided by increased letdown resulting directly from RC system pressure increase, letdown with concurrent termination of RC pump seal injection or makeup, opening the pressurizer vent valve, opening the pressurizer electromatic safety relief block valves, or lifting the pressurizer safety relief valves (the latter two methods are undesirable and will only be considered as a last resort).
3.0 Status Phase II is operational, i
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Liquid Radioactive Waste Processing System Title "EPICOR II" 1.0 System Function and Design Criteria The system is designed to cleanup radioactive liquids so as to produce water capable of being released from Three Mile Island. Cleanup includes removal of radioisotopes and chemical constitutents to comply with Plant Technical Specifications for Water Relesses to the Susquehanna River. The design is optimized with respect to ALARA considerations.
Instrumentation and controls are provided for monitoring of system performance. Water flows are monitored where the values l
are critical to the process and or system safety.
Inline monitoring and a comprehensive sampling system are provided for thorough analyses of system water cleanup performance.
Radiation and airborne monitoring equipment is provided for analysis of activity levels.
Shielding is being provided to minimize expost related to the operation of this systa=.
An EVAC subsystem is utilized to cleanup and monitor any gases that might be released from the liquid processing system.
It is the goal to minimize gas releases from the system, however, should they occur, they will be cleaned to reduce any releases to the environment. Monitoring of the air exhaust will continue to detect any potential radioactive gas. A slight negative pressure is maintained to ensure building inleakage is maintained.
The system is being optimized with respect to ALARA considerations.
2.0 System Descriotion Liquid Processing The TMI Station Chemical Cleaning Building is used to house the l
system along with the existing tankage and sump existing in that building. Piping and pumps are provided for water movement through cleanup vessels.
The system is composed of a pre-filter, two demineralizers and an after filter.
The pre-filter and demineralizers are designed for ease of hookup and disconnect to allow for quick l
installation and remote, reliable removal.
Gas Processing The primary components are a fan, an air cleanup filter train, and necessary ducting.
The =ain EVAC components located ex-ternal to the Station Chemical Cleaning Building, but are enclosed in their own shelter.
Rev. 4 L
3.0 System Operation The Auxiliary Building Emergency Liquid Cleanup System consists of a vendor supplied liquid radwaste process system which is located in the Chemical Cleaning 3uilding. The system is designed to decontaminate by filtration and ion exchange approximately 400,000 gallons of radio-active waste water contained in the Auxiliary Building of TMI Unit 2.
Contaminated water is being pumped from a connection located on the l
Miscellaneous Waste Holdup Tank (WDL-T-2) by a pump located in the Chemical Cleaning Building through the yard and into the process system. Yard piping is enclosed within a guard pipe, the-end of which j
terminates inside the Chemical Cleaning Building.
Decontaminated water is delivered to the Clean Water Receiving Tank (CC-T-2) for sampling and analysis and pumped to the Liquid Waste Disposal System of TMI Unit II for storage if within specs, or l
transferred to the Off Spec Water Receiving Batch Tank (CC-T-1) for recycling through the process system. Capability also exists to discharge to a tank truck.
CC-T-1 may also be used for storage.
l The Chemical Cleaning Suilding (CCB) has been made into a low leakage confinement building and provided with an exhaust ventilation system to maintain the building at a negative pressure.
HEPA and charcoal filtering is provided on the ventilation system which discharges to a local stack at the roof line of the CC3 where all effluent air is monitored for radioactivity.
4 Normal operation of the processing Lystem is by remote means except for infrequent operations, such as sampling and chemical addition.
All re=ote system operations are controlled from the TV Monitor Control Building located outside the northwest corner of the Chemical Cleaning Suilding.
Re=ote handling of spent resin containers from their position inside the Chemical Cleaning Building to the transport cask and truck are provided.
The system interface with the TMI Unit 2 Radwaste Disposal Miscellaneous Liquids System, the IMI Unit 1 Liquid Waste Disposal System, De-mineralized Water System and the Service Air System.
4.0 -Status The system is operating successfully, and has processed approximately 90,000 gallons. A total of sixteen (16) spent' resin liners have been used and are in the vaste staging area.
All processed water is stored in EPICOR II storage tanks.
Rev. 4
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Trash Compactor i
1.0 System Function and Design Criteria Additional compaction facilities were required due to the amount of compactible waste being generated.
Therefore, a system was needed to compact low level solid waste into 55 gallon drums for storage or shipment.
The system shall compact waste into 55 gallon Department of Transportation (DOT) drums meeting requirements for shipping LSA material.
2.0 System Description
A Stock Equipment Company Model 2407 compactor was installed in the Unit 2 Model Room. A drawing of the unit is l
attached.
The unit includes roughing and REPA filters. The discharge of the unit is vented to the Aux. Building ventilation system which contains charcoal filters.
3.0 System Operation Trash is compacted into 55 gallon DOT approved drums.
Standard plant operating procedures have been revised to include the use of the new compactor. Use is limited to compactible dry waste only. No wood, metal or liquids are permitted.
All bags of trash are surveyed prior to compaccion; bags in excess of 500 mr.
are not compacted.
4.0 Status Compaction of trash for LSA shipments continues.
Drums, that have been compacted, are stored in the compacted waste staging facility until ready for shipment to a radwaste burial ground in Hanford, Washington.
307 drums were shipped from Oct. through December 1979 with 107 drums in storage at the end of this period.
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Staging Facilities for Dewatered Resins and Evaporator Bottoms A.
WG 21 - Interim Solid Waste Staging Facility 1.0 System Function and Design Criteria Facilit.ies are needed to stage dewatered radioactive resin and filters generated by EPICOR I and EPICOR II until they can be shipped to a burial site.
WG-21 provides space for this staging.
2.0 System Description
The facility consists of 16-54" diameter cells and 12-84" diameter cells to receive 4' x 6' and 6' x 6' resin liners.
The cells are installed in the Unit 2 cooling tower desilting basin, backfilled for shielding and capped with 3' thick concrete plugs.
3.0 System Operation Eight (8) EPICOR I Resin 1.iners, one (1) EPICOR I Pref 11ter, and one (1) smaller resin liner (used to remove trace activity and fluorescein dye) are staged in the facility.
Sixteen (16)
EPICOR II resin liners and one (1) Unit i used precoat liner are also staged in the facility.
i 4.0 System Status Construction of the interim solid waste staging facility has been completed and is operational. Additional shielding (lead bricks) were installed along the interface between the cell cover and facility top to provide shielding due to streaming on some of the cells that are loaded.
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are below the 5 mr/hr design criteria.
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WG-22 Solid Waste Staging Facility 1.0 System Function and Design Criteria Facilities are required to stage the following radioactive wastes until they can be shipped to a burial site:
1.1 Dewatered radioactive resins from EPICOR I.
1.2 Dewatered radioactive resins from EPICOR II.
1.3 'Dewatered radioactive resins or solidified evaporator bottoms from systems used to process water more radio-active than that processed by EPICOR I or EPICOR II.
The sump meets the seismic requirements of Reg. Guide 1.143.
Contact readings on the sides of the facility will be less than 0.5 mr/hr and less than 2.5 mr/hr on the top.
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2.0 System Description
The facility is designed as a modular one.
Each module consists of 60" - 84" diameter cells imbedded in concrete and capped with 3' thick concrete plugs.
Each cell has a drain line to a sump which will serve three modules.
The sump is designed to collect any leakage from liners installed in the cells and meets the seismic requirements of Reg. Guide 1.143.
3.0 System Operation Module A is near completion and is expected to be operational in early January 1980.
4.0 System Status Work is continuing on the facility sump.
Painting on some of the cells is complete.
Some backfilling is complete and will be cocpleted in early 1980.
i The mudmat has been poured for Module 3 and construction schedules are being prepared for Modules 3 & C.
Rev, 4
S.
Nuclear Sampling System 1.0 System Function and Design Objectives This nuclear sampling system is to be used as a temporary liquid waste sampling facility to allow IMI Unit 2 recovery operations to continue without interfering in the normal operations of Unit I when that unit is returned to service.
It will provide a single contralled station whereby fluid samples may be taken from tanks other viae innaccessible for local sampling and/or from tanks that require frequent sampling for analyses of chemical and radiochemical content.
Included in tha sampling scope will be capability for. representative samples of Unit 2 Reactor Coolant from the pressurizer steam or water space or upstream of letdown coolers, and from the Mini-Decay Heat System, samples from the three Unit 2 Reactor Coolant Bleed Tanks, Uait 2 Miscellaneous Waste Hold-up Tank and the new Fuel Pool Waste Storage System containing liquid waste from both the Unit 2 Reactor Building Sump and Miscellaneous Waste Hold-up Tank.
Provisions have also been provided in the system for monitoring of boron concentration in the reactor coolant.
2.0 Syste= Description Unit 2 Sample Lines which presently run into Unit I sampling area shall be rerouted to a new sample sink to be located in the Fuel Handling Building 305' elevation of Unit 2.
In an adjacent room, the so-called "=odel room" a boronometer shall be installed.
The system shall provide for adequate recycle, purge and return of waste liquids. Purging of radioactive piping shall be performed prior to installation of new sa=ple lines.
Draicage from the sa=ple sink will be routed to the Fuel Pool Waste Storage System. A shielded bottle to collect drainage will also be provided.
i All piping, valves and components of the sa=pling system will meet the design conditions of the system with which they are associated or will =eet 150 psig and 200 F.
Primary coolant sampling points will have the design condition of 2500 psig and 670 F up to valve SNS-V-70.
Air exhausted from the sample hood will be filtered through charcoal and HEPA filters and discharged to the Auxiliary Building ventilation system exhaust ductwork.
3.0 System Operation i
A detailed description of the systems operation is not yet available as design changes are still being =ade.
This description shall be incorporated in a subsequent report.
I 4.0 System Status l
The system design is essentially complete.
Construction is in progress' and will be completed in early 1980.
Rev. 4
T.
Mini-Decay Heat Removal System 1.0 System Function and Design Objectives The specific function of the MDHR system is to remove decay heat such that the reactor coolant system can be brought to and-maintained at a cold shutdown condition. The system is intended to provide sufficient core flow to maintain reactor coolant subcooled.
2.0 System Description
The two MDHR pumps and two heat exchangers will be mounted at the south end of the 280'-6" elevation in the fuel handling building. New pipe runs will be installed from the existing DHR system piping to the new equipment. Cooling water to the heat exchangers is provided by the existing Nuclear Services Closed Cooling System by means of new piping.
One pump and one heat exchanger can acco=modate the current decay heat load from the core.
The system will be capable of being monitored and controlled from a new control panel in the control room or a local control panel.
The system piping and components are small to minimize the volume of reactor coolant outside of the reactor building.
3.0 System Operation A detail system description (Rev. 0) was issued November 16, 1979. The operating procedure has been written and. is being reviewed.
It will be available by January 31, 1980.
4.0 System Status The engineering is approximately 95% complete.
Piping fabrication and installation is about 95% complete. The electrical and HVAC installation is approximately 80% complete.
The current schedule for completion is February 8, 1980 for the piping system and February 28, 1980 for the HVAC system.
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Alternate Condensate Pumps Subsystems 1.0 System Function and Design Objectives The alternate condensate pumps are intended to provide backup to the existing condensate pumps to supply feedwater to the steam generators for decay heat removal and/or provide feed-water to the temporary auxiliary boiler (see separate section for temporary auxiliary boiler).
2.0 Description The two 50 gpm alternate condensate pumps are piped to take suction from the condenser hot well and discharge to the stern generator through either of two new condensate demineralizers.
3.0 Status The system is installed and is being used to provide feedwater to the Temporary Auxiliary Boiler.
Rev. 4
p V.
Temporary Auxiliary Boiler System 1.0 System Function and Design Objectives The temporary auxiliary boiler systen is intended to furnish steam to the Unit II turbine gland seals so that the existing auxiliary boilers (Unit I) can be shutdown and serviced.
i 4
2.0 Description The temporary (skid mounted) auxiliary boiler is designed to receive feedwater from the alternate condensate pu=ps and deliver 185 psig steam to the Unit II auxiliary steam header.
l 3.0 Status I
The boiler skid is in place and operational.
i i
r I
i t
t l
Rev.:<4 9
W.
TMI-II Low Level Waste Processing System (EPICOR I Relocation) 1.0 System Function and Design Obj ectives The low level liquid waste processing system for 4.1 pc/ml wastes shall utilize the discharge from the Contaminated Drain Pumps (WDL-P15A and B, Crane Company -Deming Division-Single Stage Centrifugal Pumps, provided as serial numbers DC - 551976 and DC - 551977 under Deming Division Customer Order No. C-0066) as a feed source. Cartridge filtration equipment (WDL - F7A and B) between the pumps and the processing system is available for use in series with the processing system prefilter.
Effluents from the process system will be collected in a WASTE PROCESS MONITOR TANK for sampling and discharge to the plant Evaporator Condensate Test Tanks (WDL - T-9A and/or B) or recycle cleanup, as indicated by the sample results.
2.0 System Status Relocation of EPICOR I.
t Rev. 4 l
4 e
X.
EPICOR II Solidification System 1.0 System Function and Design Objectives Pursuant to USNRC order, EPICOR II system solid wastes (i.e.-resins) =ust be contained in a solid, free standing monolith, for shipment and burial. The EPICOR II solidification system is being provided to accomplish solidification of these wastes.
2.0 System Description
The system remains in the conceptual design and development phase.
3.0 System Operation Later.
r 4.0 System Status Conceptual design phase.
4 3
.h j
Rev. 4
Y.
Containment Service Building 1.0 System Function and Design Objet*ives
.a.
Provides contamination and airbocne particulate control envelope at the containment equipmet.* hatch.
b.
Provides for efficient personnel access to containment.
c.
Allows passage of large pieces of equipment and bulk i
radwaste.
d.
Provides a waste staging and temporary storage area.
e.
Provides a decontamination area for equipment removed from containment.
f.
Provides space to handle containment se-vice systems.
g.
Allows for maintaining a hot tool crib in v.
- of containment.
I a
2.0 System Description
2 Building area of approximately 20,000 FT height approximately 35 ft located adjacent to the equipment hatch.
3.0 System Status Design criteria develop =ent phase.
i f
4 I
i l
l Y
?
i i
Rev. 4 l
a Z.
Evaporator / Solidification Facility 1.0 System Function and Design Objectives The Evaporator / Solidification Facility shall provide for the collection, treatment, storage and disposal of liquid radio-active wastes generated during decontamination of TMI Unit II and for the collection, treatment, storage, solidification and disposal of spent resins generated within the facility.
2.0 System Description
The Evaporator / Solidification facility occupies an area of 112' x 62' and is 67' high. It is located next to the west wall of the diesel generator building. '4ater from the various tanks in Unit II and from the containment service building is piped to the Facility, where it is chemically treated prict to entering the HPD Evaporator.
The distillate from the Evaporator is passed through a demineralizer and stored in RC Evaporator Condensate test tanks prior to discharge to the river or recycling.
Concentrates from the Evaporator are stored in a concentrated waste storage tank from where they can be slurried to the solidification subsystem, as can be the Unit II spent resins.
3.0 System Status Engineering of the facility is approximately 60% complete.
Preliminary flew diagrams and equipment specifications have been reviewed. Final documents to be issued in February.
No fabricatian or purchase of equipment other than the Evaporator has been started. The Evaporator should be on site July 1980.
Rev. 4
IV.
DECONTAMINATION OF AUXILIARY AND FUEL HANDLING BUILDING. THREE MILE ISLAND, UNIT II.
A.
Function and Objectives 1.
The decontamination effort in the Auxiliary and Fuel Handling Building is to decontaminate all areas of these buildings to less than 1,000 DPM, end to reduce radiation levels to design levels.
B.
Decontamination Activities 1.
Decontamination of open areas (corridors, stairwells, etc.)
is 89% complete. Cubicle decontamination is 73 complete.
General radiation levels in both buildings has been reduced to less than one (1) MR/hr. except for isolated tank cubicles and valve alleys. No tank, sump or internal piping decon-tamination has been accomplished to date. This effort is scheduled and will commence when the details for the flushing water to be used have been finalized.
2.
Decontamination efforts are restricted by the use of water in the cleanup effort. The only liquids being used for decontamination are radiac wash (decon solution) which must be contained in drums and solidified. 'Jhen an acceptable source of watar has been made available, the use of a hydro-laser for decontamination will be put into effect. This will aid in the decon effort.
3.
Solidification of twenty-two hundred (2200) gallons of radiac wash, the only liquids generated to date during the decon effort, has been accomplished.
This solidification generated seventy-three (73) drums of solidified waste.
All solidified drums were less than 100 MR/hr. on contact.
C.
Decontamination Status 1.
General Radiation levels have been reduced from 1 R/hr.
except for isolated cubicles and valve alleys.
2.
Air borne activity has been reduced from 6 x 10-7 pc/cc to 3 x10-ll pc/ce.
6 3.
Surface contamination has been reduced from 15 x 10 DPM to less than 1,000 DPM in all accessible areas.
Rev. 4
V.
RADIOLOGICAL MONITORING This section contains an Executive Summary of TMI Lnits I & II Liquid and Gaseous Releases as a Result of the Incident of March 28, 1979, and continuing through October, 1979. Tables 1-9 l
provide additional data for liquid discharges to the Susquehanna River.
Also attached are updated running tables that contain the results of analyses performed on water samples taken in the vicinity of the Three Mile Island Nuclear Station.
9 F
V 1
h Rev. 4
Page 1 of 3
~
EXECUTIVE SUHFIARY TilREE HILE ISLAND UNITS I and II LIQUID and CASEOUS RELEASES 1st incident 2nd Quarter Period Quarter 1/1/79 to 3/28/79 to 4/1/79 to 5/1/79 to 6/1/79 to 4/1/79 to l
DIScilARGE PATilWAYS 3/31/79 3/31/79 4/30/79 5/31/79 6/30/79 6/30/79 I.
Liquid Released:
I a) Discharge less Tritium:
- 1) Concentration (PC1/cc) 1.29E-8 (a) 7.44E-8 (a) 1.75E-7 (a) 2.89E-8 (a) 2.84E-8 (a) 8.63E-8 (a)
- 2) Total Activity (C1) 0.277E0 (b) 1.00E-1 (b) 1.62E0 (b) 2.21E-1 (b) 1.88E-1 (b) 2.03E0 (b) b)
Iodine-131 Released:
- 1) Concentration (UCi/cc) 4.97E-9 (a) 7.16E-8 (a) 1.70E-8 (a) 2.25E-9 (n) 5.608-10 (d) 7.57E-9 (a)
- 2) Total Activity (C1) 0.107E0 9.62E-2 1.57E-1 1.72E-2 3.70E-3 1.78E-1 c) Tritium Released:
- 1) Concentration (pci/cc) 4,83r.6 (a) 5.13E-7 (a) 8.45E-7 (a) 7.05E-7 (a) 4.60E-7 (a) 6.77E-7 (a)
- 2) Total Activity (C1) 104.lE0 0.69E0 7.80E0 5.38E0 3.04E0 1.59El II.
Airborne Iodine Released:
a) Quarterly Release Rate (pC1/sec) 5.8E-1 5.8E-1 1.20E0 9.89E-3 2.12E-5 1.22E0 b) Total Iodine-131 Released (C1) 4.57E0 4.57E0 9.48E0 7.8E-2 1.67E-4 9.6E0 III.
Noble Cases Released:
.[.
a) Quarterly Release Rate (Ci/sec) 1.12E0 1.12E0 1.41E-1 1.81E-4 9.5E-5 1.41E-1 k) b) Total Noble Cases Released (C1) 8.83E6 8.'83E6 1.llE6 1.43E3 7.50E2 1.llE6 FOOTNOTES:
' c) Concentrations are based upon actual HDCT flows. These are concentrations in the effluent averaged over the period.
' b) This data includes Todine-131 released to the Susquehanna River as a regult of the TM1 Unit II accident on Ebrch 28, 1979.
,. l
. --- J
l Page 2 of 3 t
0 EXECUTIVE SUIDIARY TilREE IIILE ISLAND UllITS I and II I.lQUID and CASEOUS RELEASES 3rd Quarter 7/1/79 to 8/1/79 to 9/1/79 to 7/1/79 to DISCIIARGE PATilWAYS 7/31/79 8/31/79 9/30/79 9/30/79 3
i I.
I.iquid Released:
a) Discharge less Tritium:
- 1) Concentration (pC1/cc) 1.12E-8 (a) 2.66E-9 (a) 2.33E-9 (a) 5.18E-9 (a)
- 2) Total Activity (C1) 7.85E-2 (b) 1.89E-2 1.78E-2 1.13E-1 b)
Iodine-131 Released:
- 1) Concentration (pC1/cc) 4.57E-10 (a) 9.10E-II (a) 8.63E-Il (a) 2.07E-10 (a)
- 2) Total Activity (C1) 3.20E-3 6.46E-4 6.59E-4 4.51E-3 c) Tritium Released :
- 1) Concentration (pcl/cc) 7.20E-7 (a) 3.20E-7 (a) 3.34E-7 (a) 4.53E-7 (a)
- 2) Total Activity (C1) 5.04E0 2.2'E0 2.55E0 9.86E0 II.
Airborne Iodine Released:
a) Quarterly Release Rate (pCi/sec) 1.58E-6
<MDA
<MDA 1.58E-6 b) Total Iodine-131 Realease (C1) 1.24E-5
<MDA
<MDA 1.24E-5
- III, Noble Cases Released:
a) Quarterly Release Rate (C1/sec) 1.27E-5 1.146-5 8.88E-6 3.30E-5 o
i b) Total Noble' Gases Released (C1) 100 90 70 260 FOOTNOTES:
a) Concentrations are based upon actual HDCT flows. These are concentrations in the effluent averaged over the period.
h b) This data includes Indine-131 released to the Susquehanna River as a result of the TM1 Unit II accident on March 28, 1979.
I 3:
L
Page 3 c2 3
?
.b EXECllTIVE
SUMMARY
l TilREE Mll ISI.AND UNITS I and II 1.lQUID and GASEOUS REl. EASES 4th Quarter j
f 10/1/79 to 11/1/79 to 12/1/79 to 10/1/79 to
- i DISCllARCE PATilWAYS 10/31/79 11/30/79 12/31/79 12/31/79 I.
I.iquid Released:
J a) Discharge less Tritium
- 1) Concentration ( C1/cc) 1.25E-9
- 2) Total Activity (C1) 9.29E-3 t ~
b)
Iodine-131 Released-
- 1) Concentration ( C1/cc) 4.89E-Il
')
- 2) Total Activity (C1) 3.62E-4 c) Tritium Released:
l
't
- 1) Concentration ( C1/cc) 7.61E-7 l
- 2) Total Activity (C1) 5.64E0 d) MDCT Flow For Month (cc) 7.41E-12 i
II.
Airborne Iodine Released:
a) Quarterly Release Rate ( C1/sec)
<MDA b) Total lodine-131 released (C1)
<MDA III. Noble Cases Released:
I a) Quarterly Release Rate (Ci/sec) 9.51E-6 b) Total Noble Cases released (C1) 75 FOOTNOTES :
a) Concentrations are baseil upon actual MDCT flows. These are concentrations in the effluent averaged over the period.
d.f.
0
,5 k
.z
~t
- s..
TABLE 1 LIOUID RADIONUCLIDE DISCHARGES FROM UNIT 1 BY ISOTOPE 1/1/79 - 3/27/79
')
Activity Radionuclide (C1) 3H 2.54E+1 51Cr 1.65E-3 54Mn 3.36E-4 58Co 2.13E-2 59Fe 1.33E-4 60Co 1.19E-3 65zn 3.94E-5 95Nb 1.43E-3 95z 7.71E-5 97Zr 8.88E-5 99Mo 8.56E-6 103Ru 7.37E-5 110Ag 8.32E,-4 122Sb 5.78E-5 124Sb 3.77E-5 131I i
2.54E-4 131he 2.60E-5 1321 133 7 133Se 2.60E-5' 133Xe 9.95E-3 134Cs 3.21E-3 136Cs 1.22E-5 137Cs 4.55E-3 140Ba 2.88E-5 140La 3.94E-4 e
s
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.. u.,s.
eyyo.... _..t....e m;a e p - r.i._..m o
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s.ssw s.-
w..
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- .. ?4f<t, J.z-
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..... r ;.-.. s.
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'k's u -:?.h? "i& :.n. ~~ditu:s-a,i::: :
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ix. s,.
- a
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TABLE 2 LIQUID RADIONUCLIDE DISCH.UtGES TROM UNIT 2'BY ISOTOPE 1/1/79 - 3/27/79 Activity Radionuclide
' (C1) - ---
3E 7.81E+1 24Na 1.82E-2 41Ar 1.19E-5 51 'r 2.10E-3 C
54Ma 1.13E-2 58Co
.2.11E-1 59Co 2.29E-4 i
59Fe 1.39E-3 60Co
-3. 88E-3 95Nb 4.2E-4 95Zr 1.59E-4 99Mo 3.85E-5 103Ru 2.10E-4 110.Ag
-1.07E-3, 113m Ag 1.983-4 12:sb 1.012-4 1:4 Sb 9.26E-5 131'r 8.82E-4 133I ^
6.92E-5 133Xe 3.13E-2 133 1.34E-4 e
134Cs 1.94E-3 135,
3.89g-4 3
137Cs 2.18E-3 140La 6.98E-4 187g 3.43E-4 s
l c
s l
~
j l
.. ~..
.. j.a. - ;.. -
j '.- -
- ,.. : - "-* 2 : '
' TABLE 3_
LIOUID RADIONUCLIDE DISCHARGES
'FROM UNITS 1 AND 2 BY ISOTCPE 1/1/79 - 3/27/79 Accivity Radionuclide (C1)'
311 1.04E+2 24Na
-1.82E-2 41 'r
- 1.19E-5 A
51Cr 3.75E-3 54th 1.16E-2 58Co 2.32E-1 59Co
-2.29E-4 59Fe 1.52E-3 60Co 5.07E-3 65Zs 3.94E-5 95Nb 1.85E-3 95Zn 2.36E-4 l
l 97Zr
.8.88E-5,
99Mo 4.71E-5' 103Ru
.2.84E-4 110.Ag 1.9E-3 1103Ag 1.98E-4 122Sb 1.59E-4
~
124Sb 1.3E-4 131I 3.47E-4 1313Xe 2.60E-5 133I 6.92E-5 133Xe 4.13E-2 133 Xe 1.6E-4 1
134Cs 5.15E-3 l
13sXe 3.89E-4 l
136Cs 1.22E-5 '
137Cs 6.73E-3 140Ba 2.88E-5 140La 1.09E-3 187W 3.43E-4
-M.
g
.1-
' s s ' - [*
...-4
.J
t Paga 1 of 4 TABLE 4 LIQUID RADIONUCLIDE DISCHARGES By ISOTOPE 3/28/79 - 4/30/79 5/1/79 - 5/31/79 Activity Activity RADIONUCLIDE (C1)
(C1)
H 8.49E0 5.38E O 3
P 1.1E-3 4.1E-3 32 cr 3.56E 4 9'.43E 4 Sl 8"Mn 3.75E-4 1.21E-4 6.29E-3 88 Co 2.15E-2
" Co 4.60E-3
- 1. 23E-3 "Sr 1.38E 0 1.53E-1 i
" Sr 3.32E-2 9.16E-3
" Sr 4.51E.4
'Nb 1.79E-4 4.49E 4 "Zr 4.92E.5 5.58E-5 1 ""Ag 1.14E-3 7.63E 4 122 Sb 4.03E-5 181 I*
2.53E-1 1.72E-2 182 I 2.98E-3 183 I
- 1. 23E-4 I'l"Xe 6.11E-4 l
188 Xe 1.12E-2 6.27E-5 13'Cs 1.28E-3 2.19E-3 i
1"Cs 1.43E-3 1 "Cs 5.39E-3
- 4. 00E-3 1 " Ba 4.23E 4 3.95E 3 1 " La 1.09E-3 4.24E-3 "Zn 1.05E-2 Total less H 1.72E0 2.21E-1 3
a I is the only radionuclide of significance released to the river from 1s1 Unit II accident of 3/28/79. Other radionuclides came primarily from Unit I.
P 4
~
7
~ ~.
...,i.,;..- s e.r.....;.,r.,-
z
... A - --: -u-. ;'
PKg2 2 of 4 TABLE 4 LIQUID RADIONUCLIDE DISCHARGES By ISOTOPE 6/1/79 - 6/30/79 7/1/79 - 7/31/79 Activity Activity RADIONUCLIDE (C1)
(C1) s 3.04E0 5.04E0 u
32 6.5E 4 P
3'Mn 7.69E-3 2,37E 4 ssCo 1.57E-3 1.87E-3 8'Co 6.81E.4 1.'30E-3 asSr 1.45E-1 1.04E-2
Sr 9. 25 E-3 5.66E.4
Nb 3.83E 4 2.46E 4 3.05E 5 38 Zr Zr 1.21E-5
'7 Ru 4.41E 4 1.63E 4 183 II'Ag 7.79E-5 118"Ag 1.22E-4 6.34E.4 12: Sb t2'Sb 5.37E.6 12s b S
II8 Sb 2.95E-5 131 I*
- 3. 70E-3 3.2E-3 131"Xe 9.60E 4 18'Cs 1.70E-3
- 5. 30E-3 5.16E-2 137C3 t,41 -2 iwoBa 2.74E-3 8.08E 4 lLa 6.79E-3 1.'93E 3 Ce 2.76E-5 1.63E-5 t61 1"'Ce 1.62E 4 Total less H 1.'88E-1
- 7. 85 E-2 3
- 131 I is the only radionuclide of significance released to the river from
~
Unit II accident of 3/28/79 Other radionuclides came primarily from Unit I.
- .i+
...a.
0 '
~
.C....
- -, w. - -
Pegs 3 of 4
~~
' TABLE 4 LIQUID RADIONUCLIDE DISCHARGE By ISOTOPE 8/1/79 8/31/79 9/1/79 - 9/30/79 t
Activity Activity RADIONUCLIDE (C1)
(C1)
H 2.27E0 2.55E0 3
P 2.96E 4 3.'52E 4 82 5'Mn 6;40E-5 2.14E 5 8'Co 1.24E-3 1,'81E.4
Co 5.88E 4 2.16E.4 s'Sr 1.70E-3 1.17E-3
Sr 1'27E 4 1 '19E.4 I
Nb 5.87E.4
- 1. tAE-5
'8
Zr
'87 2r Ru 6.28E-5 183
~II'Ag 1l'"Ag 4.95E 4 1.00E 4 Sb 1.84E.4 2.02E 4 122 12*Sb
- 7. 05E-5 Sb 5.30E 4 2.10E-3 128 1.02E-5 128 Sb I
6.46E-4 6.59E 4 13I 131 Xe 13'Cs 2.46E-3 2.13E-3 iss Cs 1.93E-5 Cs 9.06E-3 1.03E-2 187 IBa 2.53E.4 i
lLa 5.36E 4 1.23E 4 l'I Ce 3.55E 6 I
lCe
- 8. 71E-5 l
l Total less H 1.89E 1.78E-2 8
l
..a. '
.. n n s. -
-. - +.hi
~C G
... [ ;...
Pego 4 of 4 TABLE (4)
LIQUID RADIONUCLIDE DISCHARGE By ISOTOPE 10/1/79-10/31/79 11/1/79-11/30/79 Activity Activity RADIONUCLIDE (Ci)
(C1) 3 H 5.64E0 32 2.77E-4 p
s'Mn 6.96E-5 s7 Co sa Co 1.04E-3
" Co 8.37E-4 ss 4.56E-4 Sr
Sr 8.21E-5 ss Nb 5.57E-5 1I # g 2.34E-4 A
122 Sb 6.85E-5 12s sb 3.15E-5 181 I 3.62E-4 is'Cs 2.01E-3 lCs 3.77E-3 8
Total less 8 9.29E-3 O
m 4
,m e@e 48-
- S g
-e 4
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~
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~ ~^*'-
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~
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- =_g; -
s.
~
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- 4'**
'J,. =..',".rh
- ...w w '.
.'.[w*..
.. A **
1.m.-
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6**
- -====~*****- '
t N 'h
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s
[
l' age 1 of 2
~
t I
f,;;8 TABI.E 5 s
[:
VOI.UME OF LIQUID WASTE DIScilARGE 1/1/79 to 3/27/79
~
i';
UNIT I - 293,262 gallons UNIT II 238,308 gallons g,
(.O k.
TABI.E 6
SUMMARY
OF LIQUID Vol.UME DISCl!ARCES (CA1.1.ONS) 3/28/79-4/30/79 5/1/79-5/30/79 6/1/79-6/30/79 7/1/79-7/31/79 8/1/79-8/31/79 9/1/79-9/30/79 IWTS 2,776,600 2,348,910 1,776,070 1,821,030 1,801,030 2,429,620 IWFS 616,110 505,820 682,320 733,150 625,140 497,740 WECST (A&B) 93,903 112,229 41,888 125,827 56,800 58,048
' UNIT I See, Neut, 860,037 904,694 802,475 881,262 829,303 730,819 MDCT = TOTA 1. - (IWrS + IWFS + WECST (A&B) + UNIT I Sec, Neut,)
i 2,793,000,000 2,017,600,000 1,745,100,000 1,848~800,000 1,875,600,000 2,019,900,000 To m y
TOTAI.S FOR ACCIDEKr TO 9/30/79 n "
- t' '
IWrS = 12,953,260 gallons WECST = 488,695 gallons 1WFS = 3,660,280 gallons See, Neut = 5,008,590 gallons g.;1j Total Effluent = 1,23E 10 v.n k.
1c 1 l'
Page 2 of 2
- tl
',,. vlt -
1-x s..-
]i :ij
,,.)
TABI.E 6
,.f ]
[.
SUt91ARY OF LIQUID Vol.UME DISCllARGES
" " i t, l'.'d (CALI.ONS) t-t gj
,.::3 10/1/79-10/31/79 11/1/79-11/30/79 12/1/79-12/31/79
<a
+
'1 h.
IWTS 2,304,070 i}
IWFS 558,320 i
WECST (A&B) 80,189
- )
Unit I Sec. Neut.
752,893
.i g
4
.}
HDCT = TOTAL - (IWTS + IWFS + WECST (A&B) + UNIT I Sec. Neut.)
.i.
TOTALS 1,957,600,000 f
TOTAI.S FOR ACCIDENT TO 10/30/79 IWTS = 1,526E 7 gallons WECST = 5.689E 5 gallons 5.761E 6 gallons
'j IllFS = 4.219E 6 gallons Sec. Neut.
=
Total Effluent = 1.426E 10 II. ',
.s e
..i l ?.
Il i
lI
.4 I
I
..,)
TABLE 7 SUSOUEHANNA RIVER FLOW RATES i
1st Ouarter January 8.9 E+4 cfs or 5.34 E+6 cfm February 3.43 E+4 cfs or 2.'06 E+6 cfm 4
March 1,20 E+5 cfs or 7.2 E+6 cfm Average 8.11 E+4 cfs 4.87 E+6 cfm 2nd Quarter April 5.7 E+4 efs or 3.42 E+6 cfm May 3.86 E+4 cfs or 2.32 E+6 cfm June 2.78 EM cfs or 1.67 E+6 cfm Average 4.1 E+4 cfs 2.47 E+6 cfm i
i.
3rd Ouarter July 1.05 E+4 cfs or 6.3 E+5 cfm August
- 2.0 E+4 cfs or 1.0 E+5 cfm September
- 2.34 E+4 cfs or 1.41 E+6 cfm 4th Quarter October
- 3.90 E+4 cfs or 2.34 E+6 cfm i
November
- 4.38 E+4 cfs or 2.63 E+6 cfm L
i
- Estimate by U. S. Geological Survey l
., :.:. ~
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t,,.
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. z.
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d
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TABLE 8 N#'J e
!,"I
')
'5 Tnt LIQUID 't! DISC 1tARGE FOR 1979
[
i :
.l 1
UNIT 11 UNIT I & 11 UNIT I WECST - TANK llA 6 IRB IWFS WETT-TANK 9A&98 10S1 COMI'OSITE.,
TOTAL.
?, !.
DIScilARCED j
IWTS &
S P*
TO river
? ",1?. :
l Sua of Releases (each SEC NEtrT. TEST TANK V01.UNE
)ates
FOR HONTl!
- !3 I *',
VOLUME DIScitARCED Composite Sample Release SampleJ)
NEUT.
8A&88 DISC 11 ARCED 12 pct /cc fCl I
Cl s
pct /cc Cl Cl C1 C1 cc x 10 cc a lo
. TAN.
3.2 2.6tE-2 8.35 8.52 T lic s,e 1.36El 6.97 3.lE-7 2.16 t/3-l/31 22.8 FEE.
3.73 1.96E-2 7.11 7.85 2.87El 6.25 1.54E-6 9.63 2/7-2/28 16.6 8
3.49El 7.23 1.93E-5
'140 3/7-3/21 140
,.j HAR.
4.18 10.7 analgzed for !!
l.09 5.0E-7 0.55 3/29 onl) 0.55 i
APR.
3.55 1.89E-2 6.75 5.47 No !!qu!J release 9.23 8 lE-7 7.48 4/1-4/30 7.60 II ;
7.64 6.2E-7 4.74
- /1-6/1 5.38 e
om[ silt HAY 4.25 7.02E-3 2.98 5.38 1979 **
JUNE I.59 4.23E-3 0.67 0.69 acc!Jent on 6.61 4.6E-7 3.04 6/l-6/30 3.04 3/28/79 7.00 7.2E-7 5.04 6/30-8/l 5,o4 4.72 7.72E-3 3.64 4.23 f
JULY 3.2E-7 2.27 8/1-8/31 2.27 AUG.
2.33 6.15E-3
- 1. 30 1.52
<MnA 7.10 I
SEPT.
2.20 1.16E-2 2.55 1.75
< HO.;
7.65 1.5E-7
.15 8/31-9/30 2.55 I
6 OCT.
3.04 1.71E-2 5.26 5.64
- HinA 7.41 6.3E-7 4.67 9/30 5 t,4 NOV.
e e
DLC.
l g
s
.it
.f:,
e
.S.
g t.
28 e
I e Sample 16st Jus to {cclJent onilarch 28, 1979 s ~ t.'
l
- These rule.nes were meanttur=J via the '.10SI composite.'
,.e
- ; j ',
Calculated uslug Ja!!y concentrations obtained f rom
';}/.
mas i.'MP) 10S1 (Radiological Environmental MonttorLig Program -
8.
',s.
. i t
data.
\\
p s
e y y u[. ghi A fi
[.f.9,.
L n s ;h.(()(.kd,.'yp ayh(
a w,kd'[ ' k ',
),/.f.Y n p;.y,h[,.!.k.
b; [hj !
3.f.I.7
(
f y:m.3.,p.v,'.vayd y
- g Q g.mf ;.S ghe.
.m Q~ M. ; :
- q # },$,.p p. j g Pp),.y(,t.gN6.W.
1 M.o. J E W; i, '. i '.(a v :. ;; gi,1. yff t.l B.
?,tyf g
,y e
gs.; c '.
- r
.;r;.
t.,
lji M
,.g.
- g..;..
i y,
. -.s.
,a m
.. m.x,7 eg a
s
.s
.2
+q m asspgegggggg gt('h'.ggg h.
j j
v.
,a
... r.
8 8
Fase lof 2 e,
TABLE 9
- s a
IMI LIQUID RAD 10STRONT112t DISCllAkCE FOR 1979 f
UNIT II UNIT I t
N W WK 8A & 85
.-l' M - Cumposite WECST - Composite Tank 11A & 118 a
Tank 9A & 98 N
TANE TANK TANK SOSr
'I-88Sr YO N I'S r 90 VOI13HE 8'$r 90$r
'V01.UME i
Sr I
DISC 11 ARCED DISC 11AkCED DISCHARCED cc x 100 pC1/cc C1 pC1/cc C1 cc x los pC1/cc C1 pC1/cc C1
'cc,gos pC1/cc C1 pC1/cc cI j,,,
~. f JAN.
3.2 8.3E-8 2.66E-3 7.8E-9 2.50E-6 0.401
<HDA
<MDA
<MDA (HDA 2.31' 3.5E-7 8.09E-5 8.4E-8 1.94E-5
'.0 FES.
3.73 5.1g-7 1.90g-4 4.8E-8 1.79E-5 3.49 4.3E-7 1.50E-4 6.3E-8 2.20E-5 1.68 5.6E-8 9.41E-6
<MDA l
1 HAR.
4.18 e
a 3.89 8
2.07
~
APR.
3.55 3.8E-7 1.35E-4 2.2E-8 7.80E-6 Nona None HAY 4.25 9.1E-6 3.87E-3 2.2E-7 9.35E-5 None None
~
JUNE 1.59 4.8E-5 7.61E-3 1.6E-6 2.54E-4 None Nune N#"*
~
~
~
~
'None
. JUI.Y 4.72 2.2E-5 1.04E-2 1.2E-6
- 5. 6tE-4 None yo,,
AUG.
2.11 8.0E-6 1.7 0E-3 6.0E-7 1.27E-4 i
i SEPT
- 2.20 5.3E-6 1.17E-3 6.3E-7 1.39Ee4 None None 8
- OCT, 3.04 1.5E-6 4.56E-4 2.7E-7 5.2tE-5 None None
-s NOV.
4 I
0*Sr = 2E-8 pct /cc EEC.
Naimus HDA for 1979
.* S.mple lost due to accident on Nrch 28, 1979.
S0hr = SE-9 pel/cc ie:i Special sample taken for perioJ of 3/28/79 - 3/31/79 i
s 8
(
4w, sw
.1 e:m h w, e*s'.
p.4 m m. w a
- e r g o g %a
', h*, e' a w' ?.
4r o s' g 6 +h h. a@ h )
5 s![' 7..
g e Q
- w
~y
<e e gg.gw
.g g p p maa e n w e,s.ga m e p m[% u@h, hh.
fij
{.ha.k..}
r o.
y9g. h..,y fph h 6
n
- i. s.P Q44%
((li$ g ;.p,,gg@ @gi,$Qj;,
g p ggjg vg1 i$a ig.
r
s' )
si t
- i. ' )
a*-
,,, i, '
TABLE 9 (Continued) rygs 2.of 2 THI LlyillD RADIOSTRot4TIDH DISCHARGE FOR 1979 UNIT I & II
- I k.!.
-t 10-S RML Composite Total Ci for the Hunth t 3
.y j !
volume 90 89Sr + 90Sr l
l 8SSr Sr 90Sr 88Sr DISCIIARCED 12 pC1/cc C1 pC1/cc C1 C t.
C1 C1
- g cc x 10 JAN.
6.97
~
Analysis for Sr v.sa started.
1.08E-4 2.19E-5 1.29E-4 on N rch 2*,
1979
.,k
, FEO.
6.25 3.49E-4 3.99E-5 3.89E-4 HAR.
8.32
<pDA
<HDA
< b)Aa 6
<h)An a
<b)A8 4
'f, s '
l APR.
9.23 1.5E-7 1.38E O 3.6E-9 3.32E-2 1.38E0 3.32E-2 1.41EO y
.l HAY 7.64 2.0E-8 1.53E-1
- 1. 2E-9 9.16E-3 1.53E-1 9.16E-3 1.62E-1 i, O,
~~
,s.
to
-[
~
{
6.61 2.2E-8 1.45E-1 1.4E-9 9.25E-3 1.45E-1 9.25E-3 1.54E-1
. ' JUNE
~
4
..]
t JULY 7.00
<HDA
<HDA 1.04E-2 5.66E-4 1.10E-2
,1
,g AUG.
7.10
<HDA
<!D)A I
- IO E~
I"
~4
~
't "l
i
.'+<
SEPT.
7.65
<HDA
<HDA l.17E-3 1.39E-4 1.31E-3 t
..I l OCT.
7.41
<3E-9
<6E-10 4.St E-4 8,21E,5 5,38E-4
' *i.
- t. w NOV.
i
' r,' 'f,
,[ DEC.
Sample lost due to ' accident oni H.srcle 28, 1979.
,,Sr a 2E-8 )cl/cc
't
' -' "('
Special sample taken for period of 3/28/79 - 3/31/79.
m agouis HDA for 1979
'6' t
90Sr " SE-9 pct /cc l-li
- i.,,
. l..'.f.
i j
e L
f
. }E
(
8- '*}'.
t' e,
),
4 l
)
j
.,r
- q.
M; i,
l 8 womw
. gen t O dhsER @q p. a m n e. p qh_ egg sMMsgg.gg
g-----.-_-.-._===--...-.=_._
M I
(
)
)
~
o
, 8 tiote: Composite comples are taken at. 8E1, 802, 7G1, 7G2, Mater lj
- Indicates Finiched (Treated) Wat.er
'IG3, &l5F1.
141/t TRITIUli o
d 1
__1 I
i i
'6/8-e
'6/[1.
'6/15
" 6/22-
._ /25___ _ _5 / 2 6 5
5
. _ /27, 5/28 5/29 5/30 5/31 6/1 6/ _ 6/14 6L21_,_6L28
_q 27 i
Suatara creek 103 270 160 210 130 190
<120
<130 150 160 180 160 3
t j 27 y swatara creek 1C34
<183
<197
<197
< 266 196
<179
<268
<260
<187
<259
<255 30 Brunner' Island 8E1 130 170 200 160 '
_220
<100 100 130 140__
160__
110 30 f Brunner Island llE1*
100
_200 200
__L10
_230
_ 120 _ 1100 1pn 150 130
<100 i
s' Discharne Pit 10S1 1510 950 240 550 1530 540 1550 Discharae pit 10S10
_1100 381
<182 472 925 142 1040 i
j 38 __colu nbia Uuter Plant 7G1 190 170
_160 210 210
.5130 100 160 160 160 150 120.
- 38 'l coltrabia Water Plant Tula 180 160 130 i 38
. __ ColumbinJ3ater Plant 7cio
<267
<197
<280
<266
<180
<179
<268
<260
<187
<259
<255
-190 120 150
<130
<130 160
_170 160
- 140 *
's Steelton Water Vorks ISFl*
100 170 35 33 steelton Wa'ter Works 15F1Q"
<267
<197
<280
<266
<180
<179
<268
<260
< 187__
259*
<249*
2
._ygos 8c2 190 150 180 120 150 110
<120 130 110 J30 180 100 14 8 "
i.
t 1,8 g yiics 8c20
<267
<197
<280
<266
<180
<179
<268
<260
<187
<259
<255 f
u
~ n Lancaster 703"
~
160 120 n
l 32 York 902*
210 140 160 120 190 290
<100 130 110 180 180 u
York 902Q"
<267
<197
<280
<266
<180 192
<268
<260
<276
<259
<255
?
,i.
IIr19htsv111e 7G2 1.40-160 160
_160
_2 10
_ 100 _ _ 170
_lan.__
100 14 0 150 l
Uriglitsville 7G2*
<100 150 150 160 100
<120
<100 140 130 100 160
- - - -. -. - - _ _ _ = _ _... _. _. _.. _ _ _ _. _ - _.. _.. _.
S I
43 g
g
!!ote: Coupocite samples are taken at 8El, 8C2, 7G1. 7G2, Unter j
" Indicates Finished (Treated) Dater 703, &l')F1.
IC i / t.
TR$IIUM
[
. E
' 6/29 I-O 7/6 l.- --.I*~772TI 77D-IT/3-I 8/10 1 8/16
' 8/23 ' 8/30 57H-MIIJ-
~771F Pace 3--- or sm ple. 7/5.,_7/12
,]LIL.,Jg(. 0/2_
8/1__, 3/]6_._ _B/2L _0/30_ _9/6 _. 9/13 9/20 27 Suutara crech ic.s
<120 160 200 290 230 180 -
250 200 230 300 410 170 3
27 y
Swatara creek 1C3Q
<195 147 143 160
<135 157
<177 _
<283
<170
<273
<136
<304 a
30 Brunner Island 8el
<130 180
_ 24 0 180 240 180 160 210 180 190 130 180
- 30 f Brunner Island BEL" 120 200_.
210-'
210_
230 120 300 220 250 320 110 160 j
Is
'8 '<
columbia Water Plant, 701 110 210-~~
150 220 200 240 230 140 160
~
3 J70 250 240 38 colmnbi.et_Uater Plant 7Gl*
<l50 270 220 190 160 210 250 210
_ 220 190 160 F9T)
- 38
-__Calumbin_.hter Plant
_]G14
<l84 177
<177 129
<185
<l71
<177
<283
<170
<278
<128
<304 l
.35 o
Steelton Unter Vorks ISF1" 100*
280
_2_10 250 210 180 300 210 260 160 130 170.
i T.
, 35 'f Steelton Wa'ter Works ISF1Q*
<184*
<173
<177
<177
<135
<171
<177
<233
<170
<278
<128
<209
._ ylics 8c2 120 250 220 24 0 290 19 F 200
- f1Tf-7 10 - - 200 -
15Fj 7 30
- 10 i
i ylics 8caQ 122 155
<177
<1Z7
<185 31Zl_
14 6
<283
<170
<278 18
_. 128 93.8 4
i n
Lancaster 7G3" 230 220 230 230 240 170 140
~~1 tid-170 170 170 160 l
I n 3]
York 9G25
<l50
~E20 220 170 220 270 230 180 7 70 18F 170 220 i
York 902Q*
148
<l78
.117
<177
<185
_283 5177
< 281
<170
<278
<128
<125 1.
.-:n j
34
\\
85 Wrighnv_ille 7G2
<110 wrtahtcville 7G2*
160 150 160 170 200 210 240 210 170 200 170j 210 t
j 1
l5 O
S i
l~
l ll I
veR 5
29 22 1
8 8
8 8
//
6 4
4 4
4 11 2
2 2
2 2
1 1_
i
?
2 2
2 2
52 12
?
2 2
2 2
//
?
2 2
2 2
11 11 i
- 5 91 4
1 1
1 1
/L 9
9 9
9 9
11 4
2 2
2 2
M 1 1_
U i
I T
-28 0
9 0
0 0
0 9
0 9
0 9
0 0
9 0
0 I
z.
R
//
9 2
}5 2
3 2
0 2
1 2
7 7
2 7
3 1
1 2
1 2
1 2
1 2
2 1
3_1 1_
T 11 x
1_
2 1 1 n.
iI-u
__2 5
B D
g o
o O
n m
9
)
?
a y>/
)
5 5
m j/
5
@/
5 21 9
g 9
4 o
9 o
h 9
//
2 J
2 2
2 j
J J J_ 2 u
01 l
t 1i 7,
(
f r1 i
e/
0 0
0 0
0 0
5 0
0 0
5 5
5 8
1 7
t1 95
_5 0
_0 9
9 3
3 7
0 0
0 0
0 a C.
T2 1
2 1
1 1
2 2
1 2
2 2
0
._ 2 2
2 7/
i 2
\\
1 T1 0p I
1;10 0
6 6
6 6
6 0
0 0
0 0
0 0
0 8
11 9
9 9
9 9
8 7
6 8
8 7
1 4
1
/
2 2
2 2
2 W0 1
1 1
1 1
1 2
1 2
01 7F Q1 5
, l 6
- 1 0
0 0
0 6
9 1&
0 0
0 0
0 6
41 6
6 9
8 8
G 0
6 6
5 5
1 6
9 2
//
9 9
9 7
1 1
1 1
2 1
1 1
]
T 3
00 2
2 2
2
, G 1_
1 1
27 I
C 0
0 0
0 0
0 8
4 5
0 0
0 B
8 8
8 4
2 3
6 1
7 8
4 0
0 8
8 8
8 2
84
,r 2
1 1
1 1
1 1
1 1
2/
2 2
2 2
1 l e 70 1
Et 91 8 a 1
. W 5
4 7
5 2
5 5
ta)
}7 2
2 9
2 1
d 72 1
1 7
1 ne 7/
et:
99 k
t a e i
t r "l
1 l
Q Q
2 "2.
0 3
E E
G G
l 1
l c
0 "3
"2 Q
2 T
3 4
1 D'
1 C
B 8
7 7
G F
F 8
2 G
G 2
S 0
e(
5 5
C 7
9 0
3 7
r J_
1 8
9 1
ad 1
1 l
e sh i i l
f en l i pn t
t s
s t
n n
n k
k mi aF a
a a
r r
l l
l o
o s
s P.
P P
W W
ee t t d
d r
r r
r r
i a k
k nn e
e g
e e
e sc e
e aa t
t t
t t
l oi e
e l l a
a q
a
'a l
pd r
r ss W W U
W W
i r
oI a
n a n
n e
v.
nn f
C C
I I
r a
a r r i
i i o
o t
e C"
~
r r
e e b
b.
b t
t s
k h
t A
a a
nn n
r l
l a
a c
t t
n n u
u u
e e
g s
c k
k t
a a
uu l
l l
e e
g m
n r
r i
~
n r
e a
o o
6" j
w w
rr o
o o
t t
i I
i Y
Y t
t S
s BB C
C c
S S
P" J
i a
~
g" l
? ,.. [2-. [] '1'l . gd _*l
?.. 82L
' fy?>
l i
S a
8 i
o a
s
- j t
gi it N:k l:i
,4 t,l.Il4$
i
- [-
a.-u.......
1
$j' llote: Composite samplen are taken at, 8E1, BC2, 701. 7G2, Water j
- Indicates Fininhed (Treated) Unter
"(G3, &l5Pl.
pC1/1 IODINE - 131 E
=
I l~ ~
l l
i l
i i
i i
8 C
Page _20 or Sample 11/9' 11/10 11/11 11/12 11/13 11/14 11/15 11/16 11/17 11/18
_-_...t
-- 11/19,- 11/20-- -
7 i
Swatara creek 10.1_ _<0.3
<0.3
<0.3
<0.3
<0.3
<0.4
<0.4 J.3
<0.3
<0.4
<n t
.co.4 7
,a Swat. ara Creek IC3q k0.5
<0.3
<0.5
<0.4
<0.3
<0.4
<0.5
<0.4
<0.4
<0.4
<0.3_
<0.2 0
Brunner Island 8E1
<0.1
<0.3
<0.4
<0.3
<0.3
<0.3
<0.3
<0.5
< 0. 4 _
<0.5
<0.4
<0.4 Brunner Island 8Ela
<0.2
<0.3
<0.2
<0.3
<0.3
<0.3
<0.2
<0.3
<0.4
<0.5
<0.3
< 0.4
~0 3
i, Colunbla Water Plant 701 70 7 7 073' 70 2- - ~ - 7DT
<0.3 0.5 0.9 g3_
<g g
<0.3
<0.5
< 0.4 lS
~
- 3 i
colunbia Water Plant.
7G15
<0.3
<0.3
<0.5
<0.2
<0.4
<0.4
<0.3
<0.4
<0.3
< 0.4
< 0. 5
< 0. 5 15 18 colmabin_lMnt_l'.lnnt 7Glo
<0.4
<0.4
<0.4
<0.4
<0.5 0.8 1.2
< 0 5_
<0.4_
<0.4
<0.4
<0.4 a
Steelton Unter Marks 15F1
<0.4
<0.5
<0.3
<0.3
<0.2 1.0
<0.2
<0.i_
<0.4
< 0. 5
< 0.4
< 0.4 15 '"
15 steelton unter Works ISy1q
<0.4
<0.3 70.4 70.4
<0.4
~70.4 7.3
< 0.4
<0 1
<0.4
<0.1__
<0.4 0
i
~
._IiEl Oca
<0.4
<0.3
<0.3
<0.3
<0.5
<0.3
<0.3
<0.5
<0.5
< 0. 6
< 0. 3 1 < 0.4 28 : "
u I
iS ? _ YiiGG 8C2q
<0.3
<0.3
<0.4
<0.4
<0.5--
<0.3
<0.3
_ 0.3
<0.5
<0.3
_ 0.4
<0.4 f
I,a nc as t.e r 703"
<0.4
<0.5
<0.2
<0.4
<0.3
_ 0.4 1.1
<0.5
<0.3
< 0. 5
< 0. 5
< 0.4 u
l 1t York 902"
<0.5
<0.3
<0.3
<0.3
<0.2
<0.4
<0.4
.50 3
<0.4
< 0. 5 s_0 3_._
< 0. 5 York 9c2q.
7074 - 70'3-70 ~4- ~iO~J-70 3-70~4-~ 70 T-
. 0.3
<0.4
_ 0.5
< 0.4
<0.3 i
i li Intake 1352 70.3
<0.4
<0.3
< 0. T 1.6 3T2- <0.3
<0.3
<0.4
<0.4
<0.6
<0.3 i?
.l Wrinhtsv111e 7028
<0.4
<n 4 0.5
<0.3
<0.3
<0.4
<0.5
<0,5
<0.4
<0.5
<0.5
<0.5
>f t
3
./
9 4
3 2
t.
l g
3 (o
(o n n o n
o y,
o o
' c y
r e
e
(
e c
g u i li I
l
/
I.
g 3
a1 3
3 3 5 3
t 2
- f. 4
/
g o
o 0
o<.
o 0
o i
o o
t 1 7
o r
3 u
i y
M 0
3 i
t.
/. 6 3
5 2,
h l
[3 0
o o
o (o
o u
o o
0, oA t
(
(
(
c t
i 1
9 3
3 4
4 3
4 3
3 3
3 3
2 3
1
/
0
~
0 0
0 0
0 0
0 0
o 0
H
~
E I
l l
4 4
3 3
4 3
4 4
8 4
4 2
i D
2 0
/
0 0
0 0
0 0
0 0
0 0
0 1
1 1
1
- 3. _
4 3
5 3
4 4
3 3
4 5
7 2
/
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VI.
RECOVERY PLANNING During the past quarter the_" Summary Technical Plan for TMI-II Decontamination and Defueling" was issued. More detailed and comprehensive plans are under preparation covering TMI-II decon-tamination and defueli'ag. The technical plans are being developed from the many and varfous topical studies underway for recovery.
The technical plans v.11 contain a summary of the sequence of events considered nee sssary to accomplish specific recovery objectives, including the options and alternatives being evaluated.
Y c
f 1
1 Rev. 4
VII.
SPECIAL PROJECTS Special Projects programs consist of three areas:
1.
Containment surveillance experimental program 2.
Reactor Building purge program 3.
Reactor Building reentry program Containment Building Surveillance Experimental Program: The major experi=ent conducted during the fourth quarter of 1979 was the cutting of the hole through penetration R-626.
This hole was used to insert a camera and radiation monitors and to take air samples and swipe surveys. Additionally, a TLD probe was inserted through the 1. ole and radiation readings on TLD's, file badges and dosimeters ware obtained.
The major results from the experiments through R-626 are as follows.
Visual inspection of the containment building above the operating floor through R-626 showed no significant damage to the containment building.
Condensation 'sas noted in the building and dust or dirt was noted on the floor.
No other significant findings were obtained using the camera.
Radiation readings inside pen 0tration R-626 have varied due to problems using standard radiation monitoring i
equipment in the Krypton-85 atmoaphere.
The initial readings have been determined to be in error (high) due to interference in the instruments by the Krypton-85 cloud.
t These initial readings showed constant beta readings of 390 rad /hr regardless of position inside the building and i
garna readings of 2.7 to 4.6 R/hr.
Subsequent readings taken inside ecne**=edan R-626 have invalidated the initial gamma readings and Metropolitan Edison now considers the proper reading to b% 150 to 400 mr/hr general area gamma radiation readings inside reactor building.
Although beta radiation data is still inconclusive, the best estimate as determined through penetration R-626 is 300 to 400 rad /hr general area beta radiation levels.
i Air sample data taken through penetration R-626 1s roughly the same as sample data taken through the normal sample path, i.e. through EPR-227.
4 Particulate levels taken through penetration R-526 are in tha range of 10-9 to 10-10 microcuries per millileter.
The major particulate isotope determined was Cesium 134 and Cesium 137.
There was no detectable iodine in the sa=ples.
Rev. 4
Kr -
2-85 concentrations in air samples from penetration R-L s were lower than those taken from other sources.
Readings of Krypton-85 ranged from.4 to.7 microcuries per millileter. These readings are lower than the current values being obtained through the normal sample path of
.8 to 1 microcurie per millileter of Krypton-85.
Metropolitan Edison is continuing its weekly sampling program through panel HPR-227 and intends to use HPR-227 data to control the reactor building purge program.
Swipe samples taken from the wall and penetration flange of penetration R-626 showed mainly Cesium 134 and Cesium 137 as the plateout sources. Cesium 134 was found in the range of 4 to 7 x 10-2 microcuries per swipe. Cesium 137 was found in the range of 2 to 4 x 10-1 microcuries per swipe.
Cobalt 58, Cobalt 60 and Niocium 95 were detected in the range of 10-4 to 10-5 microcuries per swipe.
Data from the TLD probe is still being evaluated and is therefore not being included in this report.
Temperature and humidity inside the reactor ouilding were 840F and 100%, respectively.
Metropolitan Edison has sent the flange cutout from penetration R-626 and the inlet hydrogen recombiner spoolpiece to Oak Rid;;
for analysis. Analysis results have not been received.
As part of its initial entry program, Metropolitan Edison expects to conduct an entry into #2 personnel airlock in January. This entry will provide better radiation information on the 305' elevation of containment.
This entry will require opening the outer airlock door only.
Reactor Building Purge: Metropolitan Edison submitted its request to purge the reactor building to the Nuclear Regulatory Com=ission on November 13th. A reactor building purge program safety analysis and environmental report accompanied this purge request.
The Nuclear Regulatory Commission has asked 33 questions concerning the initial submittal and Metropolitan Edison has answered those questions in writing to the Nuclear Regulatory Commission. Procedure preparation and engineering change modifications required to support the reactor building purge are proceeding and preparations to support the reactor building purge should be completed in February.
Containment Entrv: Metropolitan Edison is proceeding with plans to conduct the initial entry into the reactor containment building. Entry team training, entry procedure preparation and engineering change modi-fications required to support the initial entry are proceeding. Metro-politan Edison is reviewing experimental data to determine if reactor building entry prior to building purge is feasible. Metropolitan Edison expects preparations to conduct the initial rea: tor building entry to be completed by the end of February, 1980.
Rev. 4
VIII.
REPORTABLE OCCURRENCES This section addresses all LER's whose period of reportability f allsin the time interval subsequent to the March 28, 1979 incident. The occurrences that fall in this category have been assigned LER numbers 79-12 through 79-22.
These LER's have been reported to NRC in varying degrees ranging from LER completed to not reported. All LER's that have not been previously submitted are attached to this report. Table VIII.1 is a status summary of all LER's that fall in this time interval.
h l
I l
t Rev. 4
.=
TABLE VIII. 1 REPORTABLE OCCURRENCE
SUMMARY
DATE OF
- 1) ATE LER NO.
DESCRIPTION OCCURRENCE REPORTEI)
COMMFNTS 79-12 Failure of Fire Barrler Seals 3/5/79 9/11/79 LER Transmitted via CQL #1163 79-13 Station RW discharge a T exceeded 3/21/79 4/6/79 Instrumentation was faulty.
FTS limit A T remained within Ifmits.
Non reportability transmitted via CQL #0474.
79-14 THI-II Incident 3/28/79 Will be submitted in the final report.
79-15 Personnel Overexposure 3/28/79 1/15/80 Previously reported via CQL
- 0620. LER attached.
79-16 Overheating of fire pumps FS-P1 4/28/79 1/15/80 LER attached.
79-17 Personnel Overexposure 3/29/79 1/15/80 Previously reported via CQL
- 1094. LER attached.
79-18 Personnel overexposure 3/28/79 1/15/80 Previously reported via CQL
- 1188 and CQL #1499. LER attached.
79-19 Overheating of fire pump FS-P1 Non reportability was determined by Plant Operations Review Committee (PORC).
79-20 Fish monitoring studies terminated 11/1/79 12/21/79 LER transmitted via CQL 1561.
79-21 Emergency diesel failed to start 12/5/79 1/15/80 LER attached.
79-22 NR pump failure 12/19/79 1/15/80 LER attached.
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8 60 61 COCKET NUMSER 68 EVENT QATE 74 75 REPORT DATE 40 EVENT DESCRIPTION AND PROBA8LE CONSEQUENCES h gl Fire pump 75-P2 had heen restored to servica Eut successful surveillance testing i
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t i l o ( l This event occurred as the result of system leakaze in excess of the 4ceka r oc c's 1 I$ i i l I capacity which resulted in the diesel driven fi-e oms runnine continuous 1v.
I
- A cooling syste= reservoir was added to 75-P3 hut determined not to Se needed on I
,,,7j i,,3, l FS-P1 (Unit 2).
OP 2104-o.1 (Unit 21 and 0.P.1104-45 (17 nit 11 have 5een revised I
,,,,i l to require operator attendance during diesel runs and to outline neviv CContinuing'3 l 30 7
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onnor NAME OF PREPARER PHON E:
January 15', 1980 LER 79-016/03L-0 Event Description and Probably Consequences:
(10) with untested capacity. This is considered reportable under T.S. 6.9.1, equipment failure as described in T.S. action statement 3.7.10.
Cause Description and Corrective Actions:
(27) developed cooldown procedures at the termination of each run.
The Unit II diesel driven pump (FS-P1) overheating was the result of several intermittant runs to raise system pressure with no cooldown performed between runs. Full operational status was restored on May 5,1979 for all four fire pumps which 4
is within the action statement requirements of T.S. 3.7.10.
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in AE7 Cat : ATE as EVENT DEScalPTION ANO PACSASLE CCNSECUENCES O'o l o 121 i While manipulating valves controllinz coolant flow and handline laEcratore 1
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bottles and flasks cdntaining coolant samples station personnel received l
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extremity exposures in excess of allowable limits. This is reports.la under I
i o i,i t 10CFR 20.405b. For further information, see Met Ed letter to V. Stello 1
l
!o !s i t dated August 21. 1971 (c0L 1094).
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For subsequent samples of reactor coolant, detailed coolant sampling orocedures l l
!i;i, i were developed and intensive training was given to station sateline eersonnel.
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shielding was properly positioned and better dosimetry coverage was provided, I
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C. F. McPhatter (717) 948-8552 s
NAME CF PREPARER PwCN E:
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NRC PCIM 366 U. S. NUCLEAR REtULATO2Y COMMIS$11N 87 77)
LICENSEE EVENT REPORT CONTROL BLCCK: l l
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8 60 68 CCCKET NUV8ER 68 69 EVENT DATE 74 75 REPORT QATE 80 EVENT DESCRIPTION AND PROBABLE CONSEQUENCES h Io l2l l While attempting to isolate leakage from the makeup system in the fuel handline I
l o I 31 1 Suilding,atation personnel received whole body and extre=ity exeosures in excess l
I o I.s l l of allowable limits. This is reportable under 10CFR 20.405b.
For further I
Io Isl l infor=ation see Met Ed letters dated September 28.1979 (COL 11881 and I
io is i l December 5, 1979 (GQL 1499).
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23 34 35 36 31 40 41 42 4J 44 47 CAUSE CESCRIPTION AND CORRECTIVE ACTIONS li lo l i Detailed reviews by radiological engineering personnel and radiological control I
g ; supervision are conducted for any job involvine access to hi2hiv contene.4 I
,,,,,iareas.
These administrative controls vill remain in effect until further l
l investigation indicates the proper precautions to Be taken to minimize further l
,,,,i; exposures.
I= proved beta detecting equipment will be utilized.
(Con tinuind i
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i C. E. MC? hatter (717) 948-8552 2,
NAVE OF PREPARER PHONE:
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January /$, 1980 LER 79-018/12L-0 Cause Description and Corrective Actions: (27)
Extra protective clothing and dosimetry will be used to properly protect and monitor workers beyond normal requirement.
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Discoverv
Description:
(32) 82, 3SR whole body doses 166, 161, 29, 13, 26, 40R.
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8 60 61 QQCK ET NUMSER 68 69 EV ENT DATE 74 7S REPCRT OATE 80 EVENT DESCRIPTION AND PROB ABLE CCNSEQUENCES hio I o i 2 I I on 12/5/77 Diesel Generator CTF-2,131 tripped 20. seconds af ter a :aanual start I
i o i 3, q attenpt due to lov lu8e oil pressure. The failure posed no threat to continued i
lo141 I core cooling; since the redundant diesel generator was verified ooera51e". offsite I l o i s i I power was available, the BOP' diesel generators were operahle, and the 13,2K!T l
t o is t I alternate feed was availahle. This event is-recortahle under T.T.
6.9.1.9.e i
1 l
- o;7; ; as descri5ed in T.S. 3.8.1.1.
I l o is i i 80 t
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JJ 34 35 36 37 40 el 42 43 44 47 CAUSE DESORIPTICN AND CORRECTIVE ACTICNS g i i o, l The failure appearra to be due to lov lube oil level.
The lube oil level was I
ii i,i l restored to the full level and the diesel was satisfactorily tested.
Acerooriate I
,,,7, g procedures are being revised to insure adecuate lube oil level ortar to startine I
i, ;3l l the diesel generators.
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7 8 9 80 sA
%#CwtR OTkER iTATUS is RY Ol5COVERY DESCRiPTfCN NA l
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NAME OF PREPARER PHONE:
C'RC PORM 384 U. S. NUCLE A?. KECUL ATO RY COMMIS$10N if WI LICENSEE EVENT FiEPORT CONTROL BLOCK: l l
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4 60 El COCK ET NUMSER 68 69 EVENT QATg 74 75 REPCRT DATE 40 I
EVENT DESCRIPTION AND PRC8ABLE CONSEQUENCES h io i2l l During disassembly of NR-P-1.3 as a result of high punp vibration, it was l
t discovered that a to'p lug on a split ring coupling which joins two sections l
,o,3, l
olAi l of the pump shaft had sheared. Therefore, although the shaft was still intact, l
lo i3l l the pump's ability to pump water over an extended period of time is unlikely.
l lo is i 1 This event is considered reportable under T.S. 6.9.1.9.5 ecuipment failure.
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33 34 33 36 31 do 41 42 43 44 47 CAUSE DESCRIPTION AND CORRECTIVE ACTICNS h iiiol l The cause for the coupling failure is most likely the result of high vibration l
li ii i i induced by badiv worn bearings. The defective comeonent and the worn bearines I
g,i,il are being replaced.
The cause for the bearinz wear is under investi2stion.
1 l i l 31 l I
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7 s 9 80 STA
% POWER OTHER STATUS Das Q RV OtSCOVERY CESCRIPTION lt I 5 I (fG,,,,j@ l 01 0 lo l@l NA I
I A l@l High cume vibration observed I
Cc'OTENT AMOUNT CF ACTIVITY @ l LOCATION CF RELEASE @
ACTIVITY RELEASED OF RELEASE NA l
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4 9 10 tt 44 45 80 PERSONNEL EXPCSVRES NbveER TYPE DESCRIPTION FTTTl 101010 l@l z l@l NA i
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7 d 3 to es 69 ao. ;;
Ron Warren 017) 943~$l38 NAME CF PREPARER PHCNE: