ML20202J426
| ML20202J426 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 02/16/1998 |
| From: | DAIRYLAND POWER COOPERATIVE |
| To: | NRC |
| References | |
| NUDOCS 9802230149 | |
| Download: ML20202J426 (2) | |
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0 YLAND a . NRC Docket No.30-409 COOPERATIVE LA enoSSE DoluNG WATER nEAcToR (LACDWR)
- route 1. DoX Ms GENOA, WISCONSIN 64832 9738 * (008) 609 2331 TO: A)R C U l i CONTROLLED DISTRIBUTION NO,33 FROM: LACBWR Plant Manager February 16,1998
SUBJECT:
Changes to LACBWR Controlling DocumtDb
- 1. The following documents have been revised or issued new.
DECOMMISSIONING PLAN, revised January 1998 lastructions:
Remove Title Page Replace with new Title Page Remove page 3 2 Replace with new page 3 2 Remove page 5 21 Replace with new page 5-21 Remove page 5-31 Replace with new page 5 31 Remove page 5 38 Replace with new page 5-38 Remove page 5-45 Replace with new page 5-45 Remove Figures 5.1 & 5.2 (previously deleted)
Remove page 6-1 Replace with new page 61 Remove page 6-4/ 6-5 Replace with new pages 6-4 & 6 5 Remove pages 6-14 & 615 Replace with new pages 6-14 & 615 Remove Figure 6.2 Replace with new Figure 6.2 Remove page 71/ 7 2 Replace with new pages 71 & 7 2 Remove page 7-3 Replace with new page 7-3 Remove page 7-5 Replace w;th new page 7 5 Remove page 8-8 Replace with page 8 8 N_ git Pages that were previously printed on both sides are being reissued separately.
INITIAL SITE CIIARACTERIZATION SURVEY FOR SAFSTOR, LAC-TR-138, revised January 1998 Remove Title Page Replace with new Title Page Remove page 2 Replace with new page 2 Remove pages 7 - 11 Replace with new pages 7 11 f Remove page 15 Remove pages 24 - 29 Replace with new page 15 Replace with new pages 24 - 29 hy D/i p i 9802230149 900216 hDR ADOCK 0500 9 ll l l lllllll ll
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{ NRC Docket No.30-409
$ The material listed above is transmitted herewith. Please verify rece pt of all listed material, i destroy superseded material, and sign below to acknowledge receipt, O The material listed above has been placed in your binder.
O Please review listed material, notify your personnel of changes, and sign below to acknowledge your review and notification of personnel. [To be checked for sunervisors for department specific procedures and LACBWR Technical Specifications.)
i O The material listed above has been changed. [To be checked for supervisors when materials -
i applicable to other departments are issued to them.)
/S/ DATE l
Please return this notification to the LACBWR Secretary within ten (10) working days.
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I 1 LA CROSSE IlOILING WATER REACTOR (LACIlWR)
DECOMMISSIONING PLAN I
Revised January 1998 DAIRYLAND POWER COOPERATIVE LA CROSSE BOILING WATER REACTOR (LACDWR) 4601 State Road 35 Genoa, WI 54632-8846 l
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, 3. FACILITY SITE CilARACTERISTICS -(cont'd)
St. Louis, Missouri, is on the opposite side of the Mississippi River from the plant and was abandoned from 1980 to 1981. The line has since been restored to service but is not frequently used. State Trunk liighway 56 originates in the village of Genoa and runs East towards Viroqua, the county seat. The origin poir t for Ilighway $6 is approximately 1 1/2 miles north of the -
reactor plant.
On the Iowa and Minnesota side of the river, State Trunk liighway 26 runs within 4 miles of the original exclusion area. All the mentioned highway facilities are two-lane paved roadways with
- unlimited access.
The car count on the road (liighway 35) passing through the nuclear F "lity original exclusion area is 2,950 cars per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, as determined by the Vernon County Wisconsin Ilighway Department in 1984.
There does exist north of the plant, approximately .9 mile, a U.S. Army Corps of Enginects Lock and Dam on the Mississippi River. This lock is not classified as an industrial facility, although it employs approximately 11 individuals.
3.3 b1ETEOROLOGY 3.3.1 Meteorolog'tcal MeasutcInent Progam
/~N V The LACBWR meteorological measurernent program consists of onsite equipment located within the Mississippi River valley. Meteorological parameters monitored are wind speed, wind direction, and temperature. Data is also available from the National Weather Service (NWS), -
approximately 35 k.m (21.7 mi.) north of LACBWR.- l 3.3.2 Gentuil Climatology '
- The plant site area exhibits a typical continental type of climate. Temperature extremes in the La Crosse /LACBWR region are more marked because of the river-valley location. Average temperatures vary from -7.l'C (19.2T) in the three months of winter to 21.9'C (71.4T) in the
. summer months.' A maximum temperature of 42.2*C (108.0") was recorded in July 1936, with a minimum low of-41.7'C (-43.07) recorded in January 1873, both in La Crosse. Monthly precipitation in the area averages between 5.1 cm (2.0 in.) and 10.7 cm (4.2 in.) from March through October and 2.5 cm'(1 in.) and 5.1 cm (2 in.) for the rest of the year. Average annual precipitation is 79.2 cm (31.2 in.). Monthly snow and sleet averages between 12.7 cm (5 in.) and
- 35.6 cm (14 in.) from November through March, the largest amount normally occurring during March. The normal annual amount of snow and sleet is 110.5 cm (43.5 in.).
O D PLAN 32 January 1998
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, 5. PLANT STATUS -(cont'd)
(j 5.2.19 Slallon and Control Air System There are two station air compressors, a single-stage compressor and a 2 stage compressor. The l single-stage compressor is a positive displacement lubricated type compressor which includes encapsulated compressor, fluid management system, motor section, :md compressor cooling system. The 2 stage compressor consists of essentially three parts: the low-pressure unit, the l high pressure unit and the motor. Air from this compressor passes through an afler-cooler and an oil separator to cool the air and to remove moisture and oil from the air before permitting it to enter the air receiver. One compressor is normally running, and the other compressor can be started when necessary. The air receivers act as a volume storage unit for the station.
The air ieceiver outlet linesjoin to form a header for supply to the station and the control air systems. Station air is provided to the Cribhouse, where it is piped to near the suction of the Low Pressure Service Water pumps; to the High Pressure Service Water tank to charge the tank; and to the generator and reactor plants at all floor levels, for station usage as needed.
Control air is supplied from the receiver discharge header through a control air prefilter, air filter, and Sullair air dryer, or through Deltech filters and Trinity air dryers to various instruments and valves in the reactor and generator plants.
Alarms are provided in the Control Room to warn orlow control air header pressure, compressor n breaker trip or low oil level in IB Air Compressor.
U System Status This system is maintained and in continuous operation.
D PLAN 5-21 January 1998
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i
. 5. PLANT STATUS -(cont'd)
(g) 5.2.29 llcaling,XcatilatioA.and Air-CODslitioning Sysicm3 The Reactor Building ventilation system utilizes two 30 ton,12,000-cfm air conditioning units for drawing fresh air into the building and for circulating the air throughout the building. The air enters the Reactor Building through two 20 inch isolation dampers in series, and is exhausted
- from the building by a centrifugal exhaust fan which has a capacity of 6000 cfm at 4 inches of water static pressure. The exhaust fan discharges through two 20 inch isolation dampers in series to the tunnel.
A 20-inch damper is also provided for reciret.lation of the exhaust fan discharge a'.r. The exhaust system is provided with conventional and high-efliciency fdters and with a gaseous and particulate radiation monitor system.
The Waste Treatment Building ventilation is provided by a 2000-cfm exhaust fan that draws air from the shielded vault areas of the building and exhausts the air through a duct out the floor of the building to the waste gas storage vault. The stack blowers then exhaust the air from the waste gas storage vault through the connecting tunnel and discharge the air up the stack.
The exhaust air from the Reactor Building and from the Waste Treatment Building are discharged into the tunnel connecting the Waste Treatment Building, the Reactor Building, and the Turbine Building to a plenum at the base of the stack. The stack is 350 feet high and is of structural
^ concrete with an aluminum nozzle at the top. The nozzle tapers to 4 feet 6 inches at the discharge, providing a stack exit velocity of approximately 70 fps with the two 35,000 cfm stack blowers in operation.
The Turbine Building heating system provides heat to the turbine and machine shop areas through unit heaters and through automatic steam heating units.
The Control Room Ileating and Air Conditioning unit serves the Control Room, Electrical Equipment Room, Shill Supervisor's area, and adjacent oflice. l The office area and laboratory are provided with a separate multi zone heating and air-conditioning unit.
The heating boiler is a Cleaver ilrooks, Type 100 Model CD-189,150-hp unit. At 150 psig, the boiler will deliver 6,275,000 Btu /hr. The boiler fuel is No. 2 fuel oil. The oil is supplied by and atomized in a Type CB-1 burner which will deliver 45 gph.
Two 14.7 kW resistance heaters with power supplied from the essential busses are available to heat the Containment Buildig in the event normal heating is lost.
Syltcm.Sinhts O These systems are maintained cperational and used as conditions require.
V D-PLAN 5-31 January 1998 t
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, 5. PLANT STATUS -(cont'd)
(q g 5.2.34.2 StackDas PASS Syskem Descriplion The Stack Gas Post Accident Sampling System makes use of the same equipment that provides the normal stack gas sample flow. The vacuum pump for stack gas sampling draws the extra flow, above what the stack monitors draw, to make the total flow isokinetic to the stack discharge. This flow can be diverted through the post accident sample canister by opening manual isolation valves. The sample canister is connected to the system by two quick disconnects, and, therefore, can be easily removed from the system and taken to the laboratory for analysis. The sample canister diversion valve is controlled from the local control panel in the No. 3 Feedwater Heater area.
5.2.34.3 Reaclor Coolant PASS System Descripjlon The Reactor Coolant Post-Accident Sampling System takes primary coolant from an incore flux monitofing flushing connection, through 2 solenoid operated isolation valves with a heat exchanger between them, to a motor-operated pressure reducing valve. Downstream of the pressure reducing valve, the coolant sample can be diluted with demineralized water which then flows through the sample cylinder or its bypass valve, through another solenoid isolation valve, and back to the Containment 11uilding basement or to the waste water tanks.
System 1tattu (O) The Stack Gas PASS System is maintained in continuous operation. The Reactor Coolant PASS System is no longer needed. The Containment Atmosphere Pl.SS System is retained in place.
5.2.35 Containtuent integrityEystenu With the plant in the sal:STOR condition, there is no longer a postulated accident that would result in containment pressurization or that takes credit for Containment integrity.
SystentStattu Containment integrity systema are not required to be operable. l
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D PLAN 5-38 January 1998
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, 5. PLANT STATUS -(cont'd) t 5.7.1.2 CantaintncnLJluilding Air Exhau1LGasconumd Partic_ulate Monitet. A monitor is located on the Containment Building mezzanine level. This monitor has a fixed filter particulate detector and a gaseous detector. It takes its suction from the outlet of the C.D. ventilation filters.
5.7.1.3 Sinck_Monitat. A monitor is installed to sample the stack emissions. This monitor draws air from the stack through an isokinetic nozzle. This morator detects particulate and gaseous activity released to the stack. This monitor alenns locally and in the control room.
5.7.1.4 Eixedlacation Moniton. Area radiation monitors are used to detect and measure gamma radiation fields at various remote locations. There are filleen iemote units located throughout the plant, The measured dose rate is displayed on meters located in the Control Room.
5.7.2 EQI1able Moniton Portable instruments are located throughout the plant. Instruments are available to detect various levels of 0 eta, gamma, and alpha radiation.
5.7.3 Lahoratorv-Type Monitors p Laboratory instmments are available to detennine contamination levels and radtoisotope d concentrations. These instruments consist ofinternal proportional counters, gamma analyzers, and liquid scintillation counters.
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- (v D-PLAN 5-45 January 1998
( ) 6. DECOMMISSIONING PROGRAM 6.1 QDUICIlyES The primary objective of the Decommissioning Program at LACHWR wih be to safely monitor the facility and prevent any unplanned release of radioactivity to the environment. Some of the goals during the SAFSTOR period are n follows:
+ To sarely store activated fuel entilit can be removed from the site. l l
- To establish a monitoring and surveillance program for comparison to baseline conditions. l
+ To maintain systems required dunng the SAFSTOR period.
+ To lay up non-operating systems.
- To salvage equipment that is no longer being used.
+ To handle radioactive waste generated during the SAFSTOR period in accordance with plant procedures and applicable requirements.
+ To reduce general area radiation levels in the vicinity of equipment operated or maintained during the S AFSTOR period to limit personnel dose to as low as reasonably achievable.
b + To start decontaminating and dismantling unused systems while minimizing the generation of radioactive waste and personnel dose from this activity.
+ Maintain qualified and trained staff to fulfill these goals.
6.2 ORGANIZATION AND RESPONSIBILIIIES The organization of the SAFSTOR staff at LACBWR is as indicated in Figure 6-1. The staff may change as activities bebg performed vary and stafling needs change. The organization is directed by a Plant Manager, who reports directly to the Dairyland Power Cooperative Assistant General Manager for Generation. The individuals who report directly to the Plant Manager each have distinct functions in insuring the safety of the facility during the SAFSTOR mode.
The Plant Manager is responsible for the safety of the facility, its daily operation and surveillance, long range planning, licensing and any other responsibilities which may come to light in long term SAFSTOR operation. Quality assurance activities and security control and support are provided by a Cooperative-wide quality assurance and security program. The Plant Manager is responsible for opeation of any onsite security required as well as insuring compliance with the quality
, assurance program.
O D PLAN 6-1 January 1998
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. 6. DECOMMISSIONING PROGRAM -(cont'd)
( responsible for coordinating the development in-house of the procedures necessary to totally dismantle the facility once the fuel is shipped from site.
The Radiation Protection Engineer will be responsible for radiation protection, projections and trending. This engineer will be responsible for working with the Health and Safety Supervisor in preparing long term prognosis for exposures and procedures necessary for decon, waste management, chemical control and fuel shipment. The Radiation Protection Engineer will assist in ensuring that an aggressive ALARA program is canied out and that contamination and background radiation exposure is reduced as low as reasonably achievable during the SAFSTOR period.
The Reactor Engineer will be responsible for all activities involving the stored fuel and will assist with plans for eventual decommissioning of the facility. This engineer will be responsible for any required reports to be generated on the stored special nuclear material.
The Safety Review Committee will remain the Offsite Review Group responsible for overtight of facility activities. It will have a quorum of 4 persons including the chairman. No more than a minority of the quomm shall have line responsibility for opere' ion of the facility. The SRC shall meet at least onw per year.
The Operations Review Committee (the Onsite Review Committee) will remain responsible for p the review of day-to-day operations. It will consist of a quorum of at least 4 individuals drawn d from the management stafrat the site. It is chaired by the Plant Manager. The Safety Review Committee and the Operations Review Committee will review all material as required by Technical Specifications including, but not limited to, facility changes, license amendinents, and plan changes in Emergency Plan and Security Plan. The committees will also review any special tests.
6.3 CONTRACIQR.UJE The use of contractors at LACBWR will continue as required throughout the SAFSTOR and DECON periods.
The use of contractors will be minimir.ed and generally limited to areas of specialty which cannot be accomplished by Dairyland stafTpersonnel. The use of contractors will be complementary in nature. It will highlight areas where DPC expertise or stafTmg is inadequate to perfonn specific tasks without outside help.
Contractor employment for specific tasks, possibly including monitoring or evaluating the facility during the S AFSTOR or aiding in dismantlement or cleanup during the DECON, will continue to be governed by the requirements of the LACBWR Quality Assurance Program. l
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\.l.J D-PLAN 6-4 January 1998
. 6. DECOMMISSIONING PROGRAM - (cont'd)
Contractors will be selected in each case on a basis of ability, price, past w perforinance and regulatory requirements.
The licensee. Dairyland Power Cooperative, will retain full responsibility for the performance of contractor tasks and will provido the supervision necessary to ensure that the tasks performed by contractors are in full compliance with the Quality Assurance Program, the purchase agreement and other appropriate regulations.
The use of contractors has the potential of aiding the IACBWR Decommissioning Project over the next 20+ years in certain select areas of unique expertise.
The ability to maximize the benefit from contractors will be closely tied to adherence to the principles stated in the Quality Assurance Program and other DPC purchasing policies and procedures.
6.4 TRAINING PROGRAM 6.4.1 Trainine Procram Descrintion 6,4.1.1 1ACBWR has established General Employee Training (CET) requirements for all personnel who may be assigned to perform work at 1ACBWR.
6.4.1.2 In addition to CET, programs have been designed to initially qualify personnel, and maintain their proficiency, in the following areas:
- a) llealth Physics Technician (llPT)
(
b) Operator c) Certified Fuel llandler (CFil) 6.4.1.3 Special infrequently performed evolutions relattag to decommission-ing activities may be included for training as they approach. These evolutions may typically be:
a) Cask llandling b) Systems Internals and Equipment Decontamination and Dismantling c) Special Tests d) Any other evolution deterrnined by plant management to require special training.
6.4,2 General Emplovec Trainine (CET1 6.4.2.1 All personnel either assigned to IACBWR, or who may be assigned duties at IACBWR, will receive GET commensurate with their assignment. This training will include, as appropriate:
a) Emergency Plan Training b) Security Plan Training c) Radiation Protection Training d) Quality Assurance Training e) Respiratory Protection Training f) Industrial Safety, First Aid, and Fire Protection O
V D PLAN 65 May 1991
. 6. DECOMMISSIONING PROGRAM -(cont'd) q b The LACllWR Spent Fuel (333 assemblics)is stored under water in the high density spent fuel storage racks in the LACllWR Fuel Storage Well which is located adjacent to the reactor in the LACBWR containment building.
Additional small quantitles of SNM are contained in neutron and calibration sources and in fission detectors which are appi >priately stored at various locations in the LACl3WR plant.
All fuel kndling and all shipment and receipt of SNM is accomplished according to approved written procedures. Appropriate accounting records will be maintained c.nd appropriate inventories, reports and documentation will be accomplished by or under the direction of the LACilWR Accoumability Representative in accordance with the requirements set forth in 10 CFR 70,10 CFR 73 and 10 CFR 74.
6.9 SAFSTOR FIRE PROTECTION PROGRAM 6.9.1 hegtam_6dministration 6.9.1.1 The LACllWR Plant Manager is responsible for the fire protection program. A member of the Dairyland technical staffis responsible for annual evaluation of couipment provided for fire fighting, training, and maintaining a current and effective fire protection program.
6.9.1.2 The training program for the Fire Response Team will be maintained under the direction of a designated staff member and will meet or execcci the requirements of Section 27 of the NFPA Code 1976.
6.9.l.3 The Fire Response Team will consist of a minimum of two (2) members. These l individuals will be available to respond in the event of a fire emerger.cy at the LACBWR Unit.
The Fire Response Team Leader will be a member of the Operations Department. The Fire Response Team will not include any personnel requimd for other essential functions during a fire emergency.
6.9.1.4 Implementing procedures for surveillance testing and inspection, to assure that necessary equipment is in place and operable, have been established. Four fire drills, conducted under the direction of the Fire Protection Supervisor, will be held each quarter, with the intent of maximizing the number of fire brigade members to be drilled.
6.9.l.5 Self-contained breathing apparatus will be supplied for each member of the Fire l Response Team and for any control room personnel. One hour of breathing air spare bottles for each of the above required masks will be available within the confines of the unit with cascade recharging facilities located on the Genoa site.
I 6.9.1.6 A section of the Fire Protection Plan delineates inspection and surveillance test p frequency, reports necessary, and statements of actions.
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l l D PLAN 6 14 January 1998
- 6. DECOhihilSSIONING PROGRAh! -(cont'd) 6.9.2 SAFSTOR Analysis LACBWR can safely maintain and control the FESW in the case of the worst postulated fire in each fire area of the plant.
The SAFSTOR fire protection progism and systems were reviewed using the criteria and
[ guidelines of Branch Technical Position 9.5-1 for general guidance. It was concluded that the LACBWR Fire Protection Program and detection and extinguishing systems are adequate, considering the reduced risk due to plant being in the SAFSTOR mode. The installed fire -
I protection equipment being maintained during th) SAFSTOR period is the same as that used during plant operation.
e Fire protection practices include isolation of fire areas via sealed penetratior.s; detection of potential fires and location identification for the plant operators, coverage by automatic extinguishing systems in the plant, protecting cables with fire re,i.; tant coverings, and installed emergency lighting systems.
}, 6.9.3 Plant Fire Layout The LACBWR plant is divided into fire areas. These areas are separated from each other by one or more of the following:
- 1) 3-hour, or better, fire walls.
- 2) Walls and ceilings with ratings well in excess of the combustibles involved.
6.9.4 Eire Protection Systems
-t The LACBWR Fire Protection System and equipment provide the means to quickly combat all types of fires that might occur at the plant and to maintain the plant in a safe condition. The Fire Protection System consists of a CO2 flooding syttem for the IB Emergency Diesel Generator (EDG), Halon dooding system, portable extinguishers, sprinkler systems, hose stations and fire hydrants, transfarmer deluge systems, portable smoke ejectors, and a fire and smoke detection l system. .
6.9.4.1 Fire Suppression Water. The fire suppression water system is a combined usage water system and is called the High Pressure Service Water System (HPSW). Water is supplied from the hiississippi River which is the west boundary of the plant. A reinforced concrete flumejuts out from the cribhouse to channel water to the pumps.
Two 125 psi net head, vertical turbine, diesel fire pumps are connected in parallel and take suction from the well supplied by the flume.
O
- D-PLAN 6-15 January 1998
U,, a -
Year 1987 1988 1989 1990 - 2039 2nd 3rd 4th Activities Qtr Qtr Qtr
- +
Reactor Shutdewn x File for Possession-Only License x Reactor Defueling x Receive Possession-Only License x File Technical Specifications for Interim Period Submit Decommissioning Plan x Submit SAFSTOR Technical Specifications x Perform Baseline Radiation Survey Perform System Modifications Decommissioning Plan Approval x SAFSTOR Period
- Limited Dismantlement Shipment of Fuel Offsite ** x Modification to Decommissioning Plan for SAFSTOR ** x R Update DECON Plan
- iCommence DECON *
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- SAFSTOR period expected to last 30-50 years. A detailed DECON Plan will be submitted prior to end of that period.
- Dependent on schedule of federal repository.
Tentative Schedule for LACBWR Decommissioning D-Plan FIGURE 6.2 January 1998
h 7. DECOMMISSIONING ACTIVITIES 7.1 }$E.PARATiON FOR SAFSTOR T1.e plant was shut down on April 30,1987. Reactor defueling was completed June 11,1987.
Since the plant shut down, some systems have been secured. Additional systems will be shut
.down following determination oflayup methodology. Others are awaiting changes to plant Technical Specifications. Section 5.2 discussed the plant systems and their status.
In addition to preparation of this Decommissioning Plan, proposed revisions to Technical Specifications, the Security Plan, the Emergency Plan, and the Quality Assurance Program Description have been completed. An addendum to the Environmental Report and a preliminar; DECON plan have also been submitted.
7.2 SAFSTOR MODIELCATIONS The LACBWR staff reviewed the facility to determine if any modifications shoulc be implemented to enhance safety or improve monitoring during the SAFSTOR period while fuel is stored onsite.
Some modifications were evaluated as being beneficial and therefore have been performed.
The majority involve the Fuel Element Storage Well System (FESW). A redundant FESW level e indicator has been added. A second remote manually- operated FESW makeup line has been installed, which supplies water from the Overhead Storage Tank. Also, a local direct means of measuring FESW water level has been installed.
Even though credit is not taken in the safety analyses (Section 9) for containment integrity, the automatic closure signals for containment isolation valves which will still be used have been modified. The valves close on either a high Containment Building activity signal or a low FESW level signal, which has been set below the normal water level range. An FESW level indicator is used to gencate the low FESW level signal. A new Containment Building activity monitor has been installed, which will generate the high activity signal.
The gas activity monitors have been recalibrated to a Kr-85 equivalent Kr-85 will be the predominant gaseous isotope during the S AFSTOR period.
O D-PLAN 74 January 1998
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-- 7.. DECOMMISSIONING ACTIVITIES - (cont'd)
() ~7.3 _ ACTIVITIES DURING SAFSTOR PERIOD _
7.3.1- Flushine Systems and Decontamination Durine SAFSTOR During the SAFSTOR period, selected systems and components, especially those
.in accessible areas, will be flushed or decontaminated.- Surface areas in-accecsible areas will continue to be decontaminated. The principal reasons for a selected-flushing and decontamination program are:
1)- To reduce the-contamination levels and radiation dose rates in areas that will be accessible for periodic maintenance and surveillance activities during the SAFSTOR period.
- 2) To reduce radiological surveillance requirements.
- 3) To reduce ~the need for. protective equipment _for personnel conducting maintenance and surveillance.
- 4) To reduce the inventory of radioactive material and the potential for the transfer of radioactive material to non controlled areas.
7.3.1.1 Internal-System Flushine. -Various closed plant systems, which-contain water, will be flushed by recirculating water through the system's piping,_ vessels,= tanks and other components, with subsequent removal of suspended solids and radioactive ions by' filtration and/or domineralization.
Some of these systems or components will be drained after the flushing-O- operations. indicate that further reductions in radioactivity by this method are impractical. These systems may be maintained in a' dry layup condition.
Other systems will not be drained and may be maintained in a wet layup condition to reduce radiation dose rates in accessible locations. In some cases, the installation of additional shielding to reduce radiution dose-rates. near the accessible areas of previously flushed systems or components may be necessary for ALARA. ;
7.3.1.2 Area and System Decontamination. The decontamination program during the SAFSTOR period will be a continuation of routine decontamination werk performed at LACBWR. Plant areas and component outer surfaces will be decon-taminated 1 o reduce the requirements - for protective equipment use and to reduce _ the potential for the translocation of radioactive material. Decon-tamination methods that are used are dependent upon a number of variables, O
D-PLAN 7-2 March 1992
, 7. DECOMMISSIONING ACTIVITIES -(cont'd) n such as surface texture, material type, contamination levels, and the tenacity with which the
( -) radioactive material clings to the contaminated surfaces.
Surface areas are primarily decontaminated using hand wiping, wet mopping, and wet vacuuming techniques. Detergents and other mild chemicals may be used with any of these techniques. The residual water cleaning solutions are collected by floor drains and processed through the liquid waste system. Most areas are routinely decentaminated to levels below 2000 dpm/R2 (about 500 dpm/100 cm'). Many areas are maintained below the Lower Limit of Detection (LLD).
Efforts will be made to maintain all accessible areas in the plant as free of surface contunination as is reasonably achievable.
Small tools and components will be periodically decontaminated by wiping with cleaning agents, steam cleaning, abrasive blasting, dishwasher, ultrasonic cleaning, electropolishing or other methods. Some unused equipment may be decon'.aminated as a prior step to removal for disposal as commercial or radioactive solid waste. Some unused equipment may be decontaminated prior to continued use in unrestricted areas.
Larger systems and components in accessible areas may be decontaminated using hydrolazers, abrasives, chemicals or other methods, aRer appropriate ALARA and economic evaluations are conducted.
n 7.3.2 Removal of Unused Equipment During S AFSTOR During the SAFSTOR period, some equipment and plant components will no longer be considered useful or necessary to maintain the p.at in the SAFSTOR condition. Some equipment located in unrestricted areas may be transferred directly for use at another location or disposed of as commercial solid waste.
Some unused equipment or components located within restricted areas, which have not previously been used for applications involving radioactive materials will be thoroughly surveyed and documented as having no detectable radioactive material (less than LLD) prior to transfer to another user or disposal as commercial solid waste.
Other unused equipment or plant system components which have previously been used for applications involving radioacti ; mterials may be removed, thoroughly surveyed and transferred to another licensed user, or disposed of as low level solid radioactive waste material. Some equipment may be decontaminated and will be surveyed to verify that it contains no detectable radioactive material (less than LLD), prior to transfer to an unlicensed user, or for disposal as commercial solid waste.
Removal of plant equipment will be performed only aner review. A 10 CFR 50.59 safety analysis l will be conducted prior to dismantling any system.
O V
D-PLAN 7-3 January 1998
.. ' 7; DECOMMISSIONING ACTIVITIES -(cont'd);
- 7,4.2 - _ In-Plant Monitoring -
b Routine radiation dose rate and contamination surveys will be taken of plant areas along with
- more specific surveys needed to support maintenance at the site. A pre-established location contact dose rate survey will be routinely performed to assist in plant radionuclide trending.
- These points are located throughout the plant on systems that contained radioactive liquid / gases during plant operation.
7.4.3 Release Point /Emuent Monitoring During the SAFSTOR period, effluent release points for radionuclides will be monitored during all periods of potential discharge, as in the past; The two potential discharge points are the stack ?
and the liquid waste line.
a) stack - the emuents of the stack will be continuously monitored for particulate and gaseous' activity. The r.a.ic gas detector (s) have been recalibrated to an equivalent --
- Kr-85 energy. The stack monitor will be capable of detecting the maximum Kr.85 concentration postulated from any accident during the SAFSTOR period, Filters for this monitor will be changed and analyzed for radionuclides on a routine basis established in the ODCM. l b) - Liquid dischargo - the liquid emuents will be' monitored during the time of release.
i Each batch release will be gamma analyzed before discharge to ensure ODCM requirements will not be exceeded.
All data collected concerning emuent releases will be maintained and will be included in the .
annual emuent report. l 7.4.4 -- Environmental Monitoring Offsite area dose rates as well as fish, air, liquid, and earth samples will continue to be taken and analyzed to ensure the plant is not adversely affecting the surrounding environment during SAFSTOR. The necessary samples and sample frequencies will be specified in the ODCM.
All data collected will be submitted in the annual environmental report.
O D-PLAN 7-5 January 1998
. 8. HEALTIl PIIYSICS -(cont'd)
A dry filter paper or cloth disc will be wiped over approximately one square foot (12"x12" square or 12'-long S-shaped) of the surface being monitored. Swipes will be counted for beta-gamma activity in a gas-flow proportional detector or with a 2 x GM probe or equivalent in fixed geometry sample holder as necessary. Alpha activity of a swipe will be determined by means of a windowless gas-flow proportional detector and a scaler or equivalent, when alpha radioectivity is suspected of being present.
8.4.4 Liquid Activity Surveys Samples of water containing radioactivity are collect ~l and analyzed on a routine basis. Spent fuel pool water is analyzed to detect indications of ,.adation of the fuel stored in the pool.
Samples ofliquid radioactive wastes and processed wastes are analyzed to ensure :evels of radioactivity are below the levels permitted for release. Samples are analyzed by Health and Safety Department personnel in accordance with established procedures.
8.4.5 Envirormental Surveys Environmental samples will be taken within the surrounding areas of the plant. These samples will be analyzed to determine any efTects plant efiluent releases may have on the environment.
This program will be conducted as per the ODCM. l 8.5 RADI ATION PROTECTION EOUIPMENT AND INSTRUMENTATION A variety of equipment and instruments are used as part of the radiation protection program.
Equipment and instrumentation are selected to perform a particular function. Sensitivity, ease of operation and maintenance, and reliability are factcrs that are considered in the selection of a particular instrument. As the technology of radiation detection instrumentation improves, new instruments are obtained to more accurately measure radioactivity and ensure an effective radiation protection program.
This equipment can be broken down into several specific groups each with its own dedicated functions. These groups are:
a) Portable Instruments b) Installed Instruments c) Personnel Monitoring Instruments d) Counting Room Instruments This equipment will be used, checked and calibrated by trained personnel according to in-plant procedures.
O v D-PLAN 8-8 January 1998
t LAC-TR-138
- LACBWR INITIAL SITE CHARACTERIZATION' SURVEY FOR SAFSTOR 4
O By:
Larry Nelson Health and Safety Supervisor October 1995 Revised: January 1998 Dairyland Power Cooperative 3200 East Avenue South La Crosse, WI 54601 o
+, o s, 14 , g .
LAC-TR-138 PAGE2
,- 2.0 OPERATING EVENTS WHICH COULD AFFECT PLANT CLEANUP
\
(S) (1) Egiled Fuel l
l l Dwing refueling operations following the first few fuel cycles, several fuel elements were
[ observed to have failed fuel rods. These fuel failures were severe enough to have allowed l fission products to escape into the Fuel Element Storage Well and reactor coolant. These fission product particles then entered, or had the potential to enter and lodge in or plate out in, the following systems:
- 1) Forced Circulation
- 2) Purification
- 3) Decay Heat
- 5) Fuel Element Storage Well
- 6) Overhead Storage Tank
- 7) Emergency Core Spray
- 8) Condensate system between main condenser and condensate demineralizer resin beds
- 9) Reactor Vessel
- 10) SealInjection
- 11) Waste Water g 12) Reactor Coolan: Post-Accident Sampling System Q 13) Control Rod Drive System (2) Egel Element Storage Well Leakage The stainless steel liner in the Fuel Element Storage Well (FESW) has had a history of leakage. From the date ofinitial service until 1980, the leakage increased from
- approximately 2 gallons per hour (gph) to just over 14 gph. In 1980, epoxy was injected behind the liner and leakage was reduced to approximately 2 gph. Since then, the leakage rate has slightly increased.
This leakage traveled through flaws in the structural concrete of the containment building and through the biological shield. This leakage may require the r emoval of additional concrete from LACBWR to accomplish the final site release.
(3) Release of Contaminated Water To The Controlled Area. July 2.1982 at 0630 The failure to close the resin inlet valve in the resin addition line to the number one full flow demineralizer following the addition of resins subsequently caused the release of water to the Turbine Building Floor through the gasket on a bulls-eye sight glass in that line. Approximately 75 gallons of contaminated water could not be accounted for in the waste water storage tanks. Approximately 20 gallons of this water is estimated to have p entered the ground in the radiological controlled area outside the west turbine hall door V and the turbine hall truck bay door. Contaminated ground was removed over a 3 sq. f1 area by the west turbine hall door and 2 sq. yard area by the truck bay door. It was placed in waste storage barrels and later sent to burial at Barnwell, S.C. l
LAC-TR-138 PAGE7
_ Spent Fuel Radioactivity Inventorv
~
l January 1988 l l Halflife Activity - Half Life Activity I
Radionuclide (Years) (Curies) Radionuclide (Years) (Curies)
Ce-144 .. 7.801 E-1 2.636 E+6 Sr-90 2.770 E+1 1.147 E+6 l Cs-137 3.014 E+1 1.666 E+6 Pu-241 1.440 E+1 1.138 E+6 L Ru-106 1.008 E40 1.524 E+6 Fe-55 2.700 E+0 5.254 E+5 Zr(Nb)-95 1.754E-1(9.58E-2) 3.555 E+5 Zr-95 1.750 E-1 3.52 E+2 Cs-134 2.070 E+0 3.291 E+5 Ni-59 8.000 E+4 2.P7 E+2 '
Kr-85 1.072 E+1 1.160 E+5 Tc-99 2.120 E+5 2.76 E+2 Ag-110m 6.990 E-1 1.018 E+5 Sb-125 - 2,760 E+0 2.73 E+2 Sr-89 1.385 E-1 1.009 E+5 Eu-155 4.960 E+0 1.68 E+2 Te-127m 2.990 E-1 8.238 E+4 U-234 2.440 E+5 6.37 E+1 Co-60 5.270 E+0 6.395 E+4 Am-243 7.380 E+3 6.31 E+1 :
Ru-103 1.075 E-1 6.334 E+4 Cd-l13m 1.359 E+1 1.78 E+1 Pm-147 2.67 to 4 129 E+4 Nb-94 2.000 E+4 1.59 E+1 Ni 1.0w E+2 3.540 E+4 Cs-135 3.000 E+6 1.40 E+1 Cc-141 ~ 8.890 E-2 2.638 E+4 U-238 4.470 E+9 1.22 E+1 Cm-242 4.459 E-1 1.858 E+4 Eu-156 4.160 E-2 8.63 Am-241 4.329 E+2 1.474 E+4 Pu-242 3.760 E+5 8.58 Pu-238 8.774 E+1 1.262 E+4 U-236 2.340 E+7 6.32-Pu-239 2.410 E+4 - 8.837 E+3 Sn-121m 7.600 E+1 4.44 Pu-240 6.550 E+3 7.165 E+3 Np-237 2.140 E+6 2.19 Eu-154 8.750 E+0 4.020 E+3 U-235 7.040 E+8 1.89 Cm-244 1.812 E+1 3.603 E+3 Sm-151 9.316 E+1 1.51 Cr-51 '7.590 E-2 3.002 E+3 Sn-126 1.000 E+5 7.01 E-1 Te-129m - 9.340 E-2 1.170 E+3 Se-79 = 6.500 E+4 5.52 E-1
- s. H-3 1.226 E+1 5.510 E+2 I-129 1.570 E+7 3.90 E-1 Fe-59 ~ 1.220 E-1 5.120 E+2 Zr-93 1.500 E+6 1.I1 E-1 Eu-l52 1.360 E+1 5.110 E+2 I-l 31 2.200 E-2 2.00 E-3 Am-242m 1.505 E+2 4.900 E+2 Total Activity = 1.00 E7 curies NOTP Attachment 1 is an inventory of the Spent Fuel Radioactivity decay corrected. l
f 9
LA C-TR-138 PAGE8 rm 4.2 Core Intemals/RX Comp 2nsnidadionuclide Inventory
(
Reactor components in and near the reactor core during power operation become radioactive due to nuclear interaction with the large neutron flux present in this region. Most of the residual radioactivity is produced by n,y react:ons with the atomic nuclei of the target material although n,P reactions, for example the l production of C-14 from N-14, are also significant.
The residual radioactivity in the materials of the various LACBWR reactor components has been estimated using activation analysis theory and, where available, actual data from laboratory analyses ofirradiated metal samples. Best estimates of neutron fluxes in and irradiation histories of specific components were used in these calculations. Original material chemical compositions were obtained from actual material certification records when readily available but standard compositions for specified materials were used in some cases, iladioactive decay of the activation product nuclides has been taken into account to obtain the best estimate values for the residual radioactivity as of January 1,1988.
As of January 1,1988, the largest contribution (210,000 Ci) to the dioactivity inventory in LACBWR activated metal components is the isotope Fe-55 (HL = 2.7y). The next largest contributor (101,000 Ci) is Co-60 (HL = 5.27y).
The activity of these two relatively short-lived isotopes will decrease very Q
V significantly during the proposed SAFSTOR period. The major long-livu contributor to the radioactivity inventory (10,700 Ci) is Ni-63 (HL = 00y). The activity of other activation product nuclides have been lumped together in two categories, those with halflives less than 5 years and those with halflives greater ,
than 5 ycars. The group with HL <5y consists mostly of Zr-95 (64d)in Zircaloy components and Cr-51 (27.7d) and Fe-59 (44.6d) in stainless steel components along with smail quantities of En-113 (115d), Sn-119 (293d), Sn-123 (129d),
Hf-175 (70d), Hf-181 (42.4d), W-181 (121d) and W-185 (75.1d). The group with d
HL > Sy consists mostly of Ni-59 (8x10 y) with small quantities of C-14 (5730y),
6 4 Zr-93 (1.5x10 y), Sn-121 (50y), Cd-l 13 (14.6y), Nb-94 (2x10 y) and Tc-99 (2.13x10'y).
r
\
c O LAC-TR-138 O~ -
PAGE 9 Core.lnternal/RX Comoonent Radionuclide Inventory - January 1.1988 Estimated Curie Content Other Nuclides Components Co-60 Fe-55 Ni-63 Total T,y < Sy T,2 > Sy In Reactor 22,109 63,22! 1,352 2,810 15 89,507 Fuel Shrouds (72 Zr,8 SS) .
4,886 J,526 817 24 15 10,568 Control Rods (29) 1,270 594 63 2,396 4 4,327 Core Vertical Posts (52)
Core Lateral Support 21,477 770 105 8 31,468 Structure 9,108 78,851 2,826 386 30 115,532 Steam Separators (16) 33,439 1,443 3,402 123 17 1 4,968 Thermal Shield 1,029 10 2 ~0 1,388 Pmsure Vessel 347 6,458 15,230 546 75 6 22,315 Core Support Stmeture 408 15 2 ~0 598 Horizontal Grid Bars (7) 173 Incore Monitor Guide Tubes 307 188 611 7 J 1.118 189,226 7,133 5,824 84 281,307 Total 79,540 In FESW 14,988 2,384 ~0 26 31,065 Fuel Shrouds (24 SS) 13,667 1,007 95 27 3 2,050 Fuel Shrouas (73 Zr) 918 3,456 2,386 910 ~0 17 6,769 Control Rods (10)
Start-up Sources (2) 3.177 2.285 156 j ._3 5.626 20,666 3,545 32 49 45,510 Total 21,218 NOTE: Attachment 2 is an iaventory of the' Core Internals /RX Component Radioactivity decay corrected. l L_-- m
- LAC-TR-138 PAGE 10 l 4.3 ' Plant Loose Surface Radionuclide Inventorv A plant smear survey was performed of all accessible interior building surfaces in an attempt to determine the amount ofloose surface contamination in the plant.
The specific isotopic identification of the contamination was also determined.'
Each smear was gamma scanned to determine not only a correlation factor in Ci per DPM/100 cm 2but also the percentile of each radioisotope present in the mixture. From previous, part 61, analysis of plant smears, it has been detennined 4
that Fe-55 is the major beta emitter in the plant and is in approximately the same 4
percentage as Co-60, Fe-55 will be the only beta emitting isotope listed as a contaminant. Alpha activity on the surfaces has been checked by the use of an Internal Proportional Counter and has been' found to be negligible and so will not be considered. The survey did indicate that the major isotopes present in the plant's loose surface contamination are the following isotopes:
t l hoto_ps ' 1/2 Life-
- Co - 5,27 years -
- Cs-137 30.1 years Mn-54 312.2 days Fe-55 2,7 years It must be realized when reviewing the r sults of this survey that a 100% smearing of plant surfaces was not performed and therefore the following data is subject to significant error, The majority of the loose smface contamination throughout the plant is found in
- . the plant contaminated areas. The plant areas that were classified as contaminated areas at plant shutdowa in 1987 are listed below, a) Waste Treatment Building
- (1) decon area .
(2) basement (3) resin liner room
- (4) high level storage pit-b) Turbine Building (1) stop valve area
. (2) full flow room -
(3)- condensate bay (4) area under main condenser 4 (5) feed pump bed plates (6) tunnel-includes 4500 WT cubicle and crawlway to stack I
c) Containment Building (1) 701 south by FESW (2) mezz around core spray pumps (3) west NI platform l y. (4) purification platform
! (5) FESW IX cubicle (6) purification pump area (7) basement -includes UCRD platform, purification IX cubicle, retention tank cubicle, subbasement, and FCP cubicles. l
LAC-TR-138 -
PAGEI1.
PLANT LOOSE SURFACE CONTAMINATION - JANUARY 1988 Isotopes Present, in pCi - Total Area pCi Co-60) . Cs-137, J Mn-54 ' LCe-144 l Co-57; 1Cs-134 Fe Content Location i
Turbine Building (TB) 0.83 ' O.07 - - - - 0.83 ' I.73 -
a) Main Floor 0.49 0.14 0.04 - -
0.49 1.16 b) Mezzanine -including stop valve area O!42 0.06 0.02 0.42 0.92 c) Grade Floor -includes feedwater heater area 0.81 0.18 0.06 - - - 0.81 1.86 d) Tunnel Containment Building (CB) 3.16. 0.2 0.39 - - -
3.16 6.91 a) Above grade 31.44 7.40 2.36 0.04 0.04 0.08 31.44- 72.80 b) Below grade 7.57 0.48 0.66 - .-- -- - 7.57 16.28 Waste Treatment Building
~ 44.72' 8.53 3.53 0.04 0.04 0.08 44.72 101.66 Totals NOTE: Attachment 3 is an inventory of LACBWR's Loose Surface Contamination Inventory decay corree'.ed. l u.
O o Q.j
/~
%/
LAC-TR-138 PAGE 15 PLANT SYSTEMS INTERNAL RADIONUCLIDE INVENTORY - JANUARY 1988 - (cont'd)
Nuclide Activity,in pCi System Total pCi Content Plant System Fe-55 Alpha Co-60 Mn-54 Cs-137 Ce-144 Zn-65 Other Fuel Element Storage 1.4 E4 1.0 E4 1.7 E6 Well System 8.5 ES 3.9 E2 8.5 E5 - -
Fuel Element Storags 6.0 E2 4.6 E3 4.5 E2 8.3 E?
Well - all but floor 1.3 E3 4.9 1.3 E3 -
Fuel Element Storage 4.1 E4 1.1 ES ' Cs-134 = 1.3 E2 5.3 E7 Well floor 2.6 E7 7.6 E3 2.6 E7 5.0 ES -
Co-58 = 1.3 E2 4.2 E4 4.0 E2 2.2 E3 Fe-5) = 1.7 E3 3.1 E5 Resin Lines 1.3 E5 1.0 E2 1.3 E5 -
Co-57 = 4.8 El Co-58 = 2.1 E3 Nb-95 = 3.5 E2 Ru 103 = 1.6 E2 3.6 E6 3.4 E4 1.9 E5 Fe-59 = 1.4 ES 2.6 E7 Main Condenser 1.1 E7 8.5 E3 1.1 E7 -
Co-57 = 4.1 E3 Co-58 = 1.7 E5 Nb-95 = 3.0 E4 Ru-103 = 1.4 E4 l
NOTE: See attachment 4 is an inventory of the plant system - Internal Radionuclide Inventory decay corrected.
l
> LAC-TR-138 PAGE 24 ATTACHMENT 1-n V SPENT FUEL RADIOACTIVITY INVIiNTORY Decay Corrected to Januar,,1998 Halflife Activity Halflife Radionuclide (Years) (Curies) Radionuclide (Years) (Curies)
Ce-144 7.801 E-1 365 Sr 90 2.770 E + 1 8.93 E5 Cs 137 3.014 E+1 1,32 E6 Pu-241 1.440 E+1 7.03 E5 Ru-106 1.008 E+0 1.58 E3 Fe-55 2.700 E+0 4.03 E4 Cs-134 2.070 E+0 1.16 E4 Ni-59 8.000 E+4 287 Kr-85 1.072 E+1 6,08 E4 Tc-99 2.120 E+5 276 Ag-110m - 6.990 E-1 5 Sb-125 2.760 E+0 22 Co-60 5.270 E+0 1,72 E4 Eu-155 4.960 E+0 42 Pm 147 2.620 E+0 2.93 E3 U-234 2.440 E+5 64 l Ni 1.000 E+2 3.30 E4 Am-243 7 380 E+3 63 Am-241 4.329 E+2 1.45 E4 Cd-l13m 1.359 E+1 11 Pu-238 8.774 E+1 1.17 E4 Nb-94 2.000 E+4 16 Pu-239 2.410 E+4 8.84 E3 Cs-135 3.000 E+6 14 Pu-240 6.550 E+3 7.16 E3 U-238 4.470 E+9 12 Eu-154 8.750 E+0 1.82 E3 Pu-242 3,760 E+5 8.58 Cm 244 1.812 E+1 2.46 E3 U-236 2.340 E+7 6.32 H-3 1.226 E+1 313 Sn-121m 7.600 E+1 4-Eu-152 1.360 E+1 307 Np-237 2.140 E+6 2,19 Am-742m 1.503 E+2 468 U-235 7.040 E+8 1.89 Sm-151 9.316 E+1 1.4 Sn-126 1.000 E*5 7.01 E - 1 Se-79 6.500 E+4 5.52 E - 1 1-129 1.570 E+7 3.90 E - 1 Zr-93 1.500 E+6 1.11 E - 1 Total Activity = 3.13 E6 Curies O
LAC-TR-138 PAGE 25 ATTACHMENT 2 CORE INTERNAllRX COMPONENT RADIONUCLIDE INVENTORY - JANUARY 1998 Estimated Curie Content Other Nuclides b Components Co-60 Fe-55 Ni-63 Total Tn > Sy In Reactor 4,855 1,261 8 12,060 Fuel Shrouds (72 Zr,8 SS) 5,936 371 762 8 2,453 Control Rods (29) 1,312 46 59 2 443 Core Vertical Posts (52) 341 2,445 1,649 718 4 4,816 Core Lateral Support Structure 6,055 2,637 15 17,685 Steam Separators (16) 8,978 261 115 0.5 764 Thnaal Shield 387 79 9 - 181 Pressure Vessel 93 1,170 509 3 3,416 Core Support Structure 1,734 31 14 - 91 llorizontal Grid Bars (7) 46 82 14 570 __3_ 669 Incore Monitor Guide Tubes 21,355 14,531 6,655 43.5 42,585 Total l
In FESW 7 1,151 2,224 13 7,057 Fuel Shrouds (24 SS) 3,669 246 77 89 2 514 Fuel Shrouds (73 Zr) 183 849 9 1,969 Control Rods (10) 928 146 _2 1,176 Start-up Sources (2) 853 175 1,587 3,308 26 10,618 Total 5,697 f
t'% n .
~
LM LAC-TR-138 .
PAGE 26:
ATTACHMENT 3 PLANT LOOSE SURFACE CONTAMINATION - JANUARY 1998 Total Area Isotopes Present,in Ci pCi Content Location Co-60 Cs-137 Fe-55 Turbine Buildine (TB)
U.22 0.06 0.06 0.34 a) Main Floor b) Mezzanine-including 0.13 0.11 0.04 0.28 stop valve area c) Grade Floor-includes 0.19 0.11 0.05 0.03 feedwater heater area 0.22 0. I4 0.06 0.42 d) Tunnel Co_ntainment Buildine (CB) 0.85 0.16 0.24 1.25 a) Abovegrade S.44 -5.88 2.41 16.73 b) Belowgrade 2.03 0.38 0.58 2.99 Waste Treatment Building 12.01 6.78 3.43 22.22 Totals E . . ._ . . . . . . . _ _ _ _ . _ _ . . . _ . _ _ _ . . . . . . . . . . . . . . . _ _ . . _ _ _
LAC-TR-138 . -
PAGE 27 ATTACHNENT4 PLANT SYSTEMS INTERNAL RADIONUCLIDE INVENTORY - JAI 4UARY 125L8 -
-Nuclide Activity,in pCi- System Total Plant System. Others pCi Content Mn-54 Fe-55 Alpha Co-60 Cs-137 430 135 688 CB Ventilation 123 --
OfTgas -
upstream of filters MMI REA/OJE OfTgas -
MEI REA/OJE downstream of filters 1,305 4.0 El 4,564 3,973 0.24 9,882 1 TB drains 2,918 3.2 10,202 1,907 0.79 15,031 CB drains 6.8 967 95 - 1,345 TB Waste Water 276 16,126 7.9 EI ..56,380 1,828 5.15 74,418 CB War Water 19,966 2.9 E2 69,804 - 6.06 90,066 Main St. .
71 1.8 ~250 159 -- .482 Turbine 1.2 El ;23,894 2.'67 30,743 Primary Purification -6,834 -
Emergency Core Spray - SYSTEAI REAf0VED - ,
3.4 El 3,490 :620 0.29 5,142 Overhead Storage Tank 998 44 60I, Seal Inject 123 3.8 430 --
l,
. LAC-TR-138 I-PAGE 28 ATTACIBfENT4 PLANT SYSTEMS INTERNAL RADIONUCLIDE INVENTORY - JANUARY 1998 Nuclide Activity,in pCi System Total pCi Content Plant System Fe-55 Alpha Co-60 Mn-54 Other 7,679 4.9 E2 26,848 9.39 - Zn 65 = 0.11 . 35027 Decay Heat Baron Inject SYSTEM RE\f0VED 760 4.6 El 2,658 0.88 '3,465 Reactor Coolant PASS 1,536 9.4 El 5,370 1.79 7,002 .
Alternate Core Spray 17,662 ' l.1 E3 61,750 20.90 Zn 65 = 0.24 80,533 Shutdown Condenser 40,272 13.93 . Zn 65 = 0.16 52,525 -
Control Rod Drive Efiluent i1,519 7.2 E2 115,187 7.0 E3 402,715 133 Co 57 = 0.20 525,037 Forced Circulation Zn 65 = 1.56 191,978 1.2 E4 671,192 230 Co 57 = 0.33 875,403 Reactor Vessel and Internals .
Zn 65 = 2.68 16,126 2.8 E2 .56,380 9.69 Zn 65 = 0.10 72,796 Condensate aller beds & Feedwater 3.1El 10,471 3.94 13,501 Condensate to beds 2,995
LAC-TR-138 PAGE 29 -
' ATTACHMENT 4 PLANT SYSTEMS INTERNAL RADIONUCLIDE INVENTORY - JANUARY 1998 '
Nuclide Activity,in Ci SystemTotal '
Plant System Other - pCi Content :
Fe-55 Alpha Co-60 Mn-54 Cs-137 -
Fuel Element Storage Well System 65,272 3.9 E2 228,205 4.24 -- 293,871 Fuel Element Storage Well
- all but floor 100 4.9 349 0.18 3,655 4,109 Fuel Element Storage Well . Cs-134 = 4.57 floor 1,996,568 7.6 E3 6,980,399 151 32,578 Zn-65 = 3.43 - 9,017,304' Resin lines 9,983 1.0 E2 34,902 12.72 - 44,998- -
Zn-65 = 5.93 Main Condenser 844,702 8.5 E3 2,953,246 1,091 -
Co-57 =' 0.35 3,807,545-l
'l i
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