ML20197A992
| ML20197A992 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 12/15/1997 |
| From: | DETROIT EDISON CO. |
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| References | |
| NUDOCS 9712230268 | |
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Text
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l U.S. Nuclear Regulatory Commission Docket No. 50-16 NRC License No. DPR-9 i
t Enrico Fermi Atomic Power Plant, Unit 1 Fermi 1 Safety Analysis Report O
Detroit Edison 9712230268 971215 PDR ADOCK 05000016 P
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Fermi 1 Safety Analysis Report Table of Contents PAGE LIST OF TABLES iit LIST OF FIGURES iv 1.0
SUMMARY
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2.0 INTRODUCTION
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3.0 DESCRIPTION
OF FERMI l' 31 3.1 Description of Plant 31 3.2 Decommissioning Activities 3-6 3.3 Current Plant Condition 3 12 3.4 Access Control 3 14 4.0 RADIOLOGICAL CONDITIONS 4-1 4.1 Total Nuclide Inventory 4-1 4.2 Radiation and Surface Contamination Levels 41 5.0 SURVEILLANCE 51 5.1 Surveys 51 5.2 Inspections and Testing 5-1 6.0 ADMINISTRATIVE CONTROLS 6-1 6.1 Organization and Responsibilities 61 6.2 Quality Assurance Program 6-2 6.3 Review Committee 6-5 7.0 FIRE PROTECTION PROGRAM 71 O
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l 8.0 ENVIRONMENTAL ASSESSMENT 81 8.1 Distribution of Estimated Population 81 8.2 Non RadiologicalImpacts 81 8.3 RadiologicalImpacts 83 8.4 Postulated R9diological Accidents 84 8.5 Alternatives Considered 8 10 t
8.6 Conclusions 8 11 9.0 References 91 O
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LIST OF TAHLES Inbh Il11s Pett 4.1 Calculated Activity in the Reactor Vessel,its 45 Intemals, the Primary Shield Tank, and the Secondary Shield, July 1986 4.2 Summary of Radiological Survey Data 1997 4-6 4.3 Radiation and Suiface Contamination Levels 47 Associated with Decay Pool and Cutup Pool 1986 8.1 1980 Distribution of Estimated Population 8-13 8.2 1980 Distribution of Projected Population in Year 2000 8 14 4
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LIST OF FIGURES ram na em 3.1 Aerial View of Enrico Fenni Units 1 and 2 3-15 3.2 Site Location 3-16 3.3 Cutaway View of Reactor Plant Layout 3-17 3.4 Fermi 1 Buildit.gs 3-18 3.5 Perspective View of Reactor 3-19 3.6 Reactor Vessel (Elevation) 3 20 3.7 Primary Shield System 3-21 3.8 Plan View of Reactor Plant Below Operating Floor 3 22 3.9 Fuel and Repair Building (Plan) 3-23 3.10 Sodium Service, Inert Gas, and Waste Gas Buildings 3-24 4.1 Fermi 1 Area Buildings 43 4.2 Fermi 1 FARB Basement 44 O
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/l SECTION 1: SUhth1ARY V
The Enrico Fenni Atomic Power Plant, Unit 1 (Fermi 1) was a fast breeder reactor power plant cooled by sodium and operated at essentially atmospheric pressure. The reactor pit.nt was designed for a maximum capability of 430 hiw1; however, the maximum reactvr power with the 6rst core loadmg (Core A) was 200 htwt. The primary system was 611td with sodium in December of 1960 and enticahty was achieved in August 1963.
The reactor was tested at low power in its first couple years of operation. Power ascension testing above I htwt commenced in December 1965, immediately followi0g r:ceipt of the high power operatmg license. In October 1966, dunng a power ascension, zirconium plates at the bottom of the reactor vessel became loose and blocked sodium coolant Gow to some fuel subassemblies. Two subassemblies started to melt. Radiation monitors alarmed and the operators manually shut down the reactor. No abnormal releases to the environment occurred. Three years and nine months later, the cause had been determined, cleanup completed, fuel replaced, and Fermi 1 was restarted.
In 1972, the core was approaching the bumup limit.
In November,1972, the Power Reactor Development Company (PRDC) made the decision to decommission Fermi 1. The fuel and blanket subassemblies were shipped offsite in 1973. The non radioactive secondary sodium system was drained and the sodium sent to Fike Chemical Company. The radioactive primary sodium was stored in storage tanks and in 55 gallon drums until the sodium was shipped offsite in 1984. I)ecommissioning of the Fermi i plant was originally completcd in December of 1975. Ilased on cunent regulatory requirements, Fermi 1 is identined as being in a SAFSTOR status, awaiting Anal decommissioning. The SAFSTOR O
license for Fermi 1 expires m 2025.
V The environmental assessment of potential radiological releases dunng SAFSTOR was performed for three different postulated accidents: (1) the liquid waste tanks rupture causing an airbor.ie release (2) a liquid radwaste release into Lake Enc, and (3) a residual sodium airborne release to the environment.
The results of the postulated radiological accidents were well below the hfPC values in 10 CFR 20, Appendix 11, Table ;I. Ilased on the low potential radiological exposures, there is no need to take any further actions to protect the health and safety of the pubbe dur'nr the SAFSTOR of Fenni 1.
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SECTION 2: INTRODUCTION G
This Fermi i Safety Analysis Report (FISAR) is the licensing basis document for Fermi l in the permanently shutdown condition. Revision 0 was prepared in 1997 based on information in References 1 through 17 (see Section 9) and the results of an evaluation performed in 1997 on the status of Fermi 1.
The references include information submitted A the NRC to support the issuance of the SAFSTOR license in 1989 that extended the Fenni i license to 2025. The accident analyses performed for the environmental assessment to support the SAFSTOR license were transferred to this document without update.
in the mterim period, additional radioactive decay has occurred, so the analyses are conservative.
1his Fermi ! Safety Analysis Report meets the requirement of 10 CFR 50.71 (e)(4) as modified by the 1996 Decommissioning Rule. Included in this document are the Fermi 1 Quality Assurance Program and Fire Protection Program, which have been prepared commensurate with the remaining radiological risk pot ~1 by Fermi 1.
1his document describes the cunent status of the facility, as well as providing some decornmissioning history. It does not include the design basis of the systems when the reactor was operating. Some sketches and general building layout drawings are included. Detailed system drawings are not included because drawings were not mamtained up-to-date durmg or following plant operation. The emphasis of the Fermi 1 Safety Analysis Repcrt is on matters impo:1 ant to the SAFSTOR status of the facility and future decommissioning efforts.
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SECTION 3: DESCRIPTION OF FERMI 1 V
3.1 DESCRIPTION
OF Tile PLANT 3.1.1 Plant 1,ocation Fermi i is located on the same site as Fermi 2, within the same owner controlled ana and outside the Fer ni 2 protected area. The site is on the western shore of Lake Erie at Lag ona lleach, Frenchtown Township, Monroe County, hiichigan (Figure 3.1 and Figure 3.2). The plant is approximately 6 miles cast-northeast of Monroe, Michigan; 30 miles southwest of downtown Detroit, Michigan; and 25 miles northeast of downtown Toledo, Ohio.
3.1.2 General Description of Plant Fermi 1 was a 200 Mw(t), sodium. cooled, fast breeder reactor that operated at essentially atmosphe-ic pressure. The general layout of the plant at tho time of operation is shown in Figure 3.3, Fermi 1 *, s last operated at low power in September 1972. Power Reactor Development Company (PRDC) decided to decommission the facility in November 1972. The effort was completed in December 1975 with the dismantling and shipping offsite of the radioactive fuel, mechanical components, and blanket subassemblies; also, contaminated areas were sealed, a Protected Area (testricted area) boundal, (Figure 3.4) was established, and surveillance began in accordance with Technical Specifications. In 1989, an amendment to the Fermi i license reclassified th<: plant as being in SAFSTOR status based on current regulations, it remains in this condition at the present time.
The following is a general desenption of some major components and systems still of interest.
3.1.2.1 l(cactor Yessel and Associaird Structures The stainless steel rtactor vessel, shown in Figures 3.5 and 3,6, is composed of the following parts: the upper and lower vessel, the transfer rotor container, und the rotating shield plug container. The cylindrical lower reactor vessel, which contained the core and blanket, is 114 inches in diameter and has a dished elliptical bottom head. The transfer rotor cor.tainer, used for fuel storage and transfer, is attached to the lower reactor vessel. The upper reactor vessel, which is eccentric with the lower vessel, is also cylindrical and is 174 inches in diameter.1he upper portion of the vessel is scaled at the top by the rotating shield plug, which supports the control mechanism, the fuel holddown mechanism (llDM),
and the offset handimg mechanism (OllM). The plug comainer is an extension of the upper reactor s essel and is stepped to mamtain the biological shielding effectiveness of the rotating plug.
The vessel as a whole contamed the reactor and the primary sodium coolant which flowed upward through the core and blanket. Sodium coolant for the core and inner radial blanket ertered the lower vessel through three equispaced nonles 14 mehes m diameter; sodium coolant for the outer blanket entered the lower reactor vessel through three equispaced nozzles 6 inches in diaineter Sodium from the core and blanket discharged into a common pool and left the upper reactor vessel through three equispaced nonles 30 inches m diameter.
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3.1.2.2 l'rlmary Shield Tank V
'lhe reactor vessel was surrounded by a graphite neutron shield located in a nitrogen atmosphere inside the primary shield tank. The graphite consisted of an inner 6 inch layer of 5% borated graphite next to the reactor vessel wall, a layer of heat insulatior, a region of unborated graphite, and a 6 inch layer of borated graphite which lined the inside of the primary shield tank (Figure 3.7).
1he upper portion of the primary shie'id tank served as a biological shield and was integral in purpose and shieldmg effectiveness with the totatmg shield plug and the biological shield function of the operating floor.
3.1.2.3 Core and Illanket Components The rer.ctor was an assembly of 870 removabk subassembly units located on a square lattice spacmg.
The core whassemblies contair ed the upper and lower axial blankets and occupied a roughly cyhndrical region in the center of the lattice. The active fuel region of each core subassembly was made up of 140 rirconium clad uramum molybdenum alloy pins ennched to 25.6 w/o (weight percent) in uranium 235.
The entire core region was about 31 inches in diameter and 31 inches high, while the axial blanket regions were each 17 inches high.
The subassembly lattice positions imrnediately surrounding the core region comprised the inner radial blanket (IRil) region; the subassembly lattice positions surrounding the IRB comprised the outer radial blanket region (ORB).
Surroundmg the ORB region were 198 lattice positions used for steel O
subassemblies that provided thennal and irradiation shielding for the reactor vessel. Together the core U
und blanket regions approximated a cylinder 80 inches in diameter and 70 inches high.
3.1.2.4 Ileat Removal Systems The heat removal systems consisted of e.cee primary and three secondary coolant loops. The sodium pumps, one per loop, were all single stage centrifugal mechanical pumps. Ileat was removed from the reactor core and blanket by the pnmary sodmm coolant, transfened to the secondary sodium coolant by three parallel imemiediate heat exchangers located m the Reactor Building, and finally transferred to water and steam in three once through steam generators located in the Steam Generator Building. Figure 3.8 shows the layout of the heat removal systems in the Reactor Building during the period of plant operation.
Ihe pnmary sodium Cowed through the three 30 mch pipes from the upper plenum of the reactor vessel to the shell side of the mtermediate heat exchangers and then to the primary pumps. Sodium was then pumped into a 16 inch header and through a tee mto the 6 mch and 14 inch lines. The 14 inch lines dehrered approumately 87% of the total pnmary now to the plenum serving the reactor core, and the 6 meh hnes dehvered the remamder to the radial blanket The pnmary system Dows w'ere measured by electromagnetic Cowmeters mstalled on each of the 14 meh core and 6 mch radial blanket lines.
3.1.2.5 1.Iquid Waste Disposal Sptem The hquid waste disposal system is located m the Fuel and Repair Building (FARB). Its major
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components were three liquid waste surge tanks. a hquid waste test tank, a liquid waste dump tank, ion
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eschange umts, a hquid waste metenng pump, and the associated piping and valving.
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ne liquid waste disposal system provided for substantial holdup capacity, permitting discontinuous
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discharge if necessary. The major sources of liquid waste were effluents from subassembly cleaning, miscellaneous decontamination operations, and laboratory sinks. Afler process and decay when i
necessary, wastes were diluted and discharged to the lagoon from which they reached Lake Eric via Swan Creek.
i All drains inside the building are routed to the hot sump. This system is currently used as a liquid waste holdup system, not a discharge system. Discharging radioactive waste will require installing a monitoring system and written procedures.
3.1.3 General Description of Hulldings in the Protected Area Refer to Figure 3.4 for relative locations of buildings.
3.1.3.1 Reactor Building The Reactor fluilding is a cylindrical vertical steel vessel 72 feet in diameter and 120 feet high with the lower 51 feet below finished grade elevation. The inside of the Reactor Building is divided into two regions by a 5 foot thick steel and concrete operating floor, he above floor region is normally accessible to personnel and houses the machinery dome, containment crane, and other miscellaneous nonradioactive equipment. De below floor region houses the reactor vessel and intemals, the primary shield tank, the secondary shield, the intermediate heat exchangers, primary sodium pumps, the decay tanks, the primary sodium overflow tank and associated equipment and piping for the primary and secondar'; sodium coolant systems. Figure 3.8 shows the layout of the b - u floor region during plant operation. The Reactor Huilding is surrounded by an approximately 3 foot wide annulus that is located belo v floor level to a depth of about 3 feet below the concrete pedestal on which the steel Reactor
-lluildi.'g stands. De annulus contams an access hole to the northwest sodium tunnel, and four floor drains tot drain into a collectmn tank and sump pump system in the basement of the Steam Generator lluilding.
3.1.3.2 Fuel and Repair ilullding (FARil)
De FARH located approximately-100 feet north of the Reactor Buildmg, is connected to the Reactor lluitdmg by a covered transport car track (trestle). The substructure of the FARB consists of heavy remforced concrete walls and rests on bedrock. The superstructure consists of two different types of
- construction. The walls awe the operatmg Door m the new fuel receiving and storage area and the irradiated fuel decay and cu;.up pool areas are remforced concrete. All other superstructure walls consist
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of structural steel with corrugated asbestos sidmg.
The FAR!l contained process cells, water 611ed decay and cut up pools, a new fuel handling and storage area, a central control room for fuel handlmg and waste system operations, a 75 ton crane, and a transport car access area for the performance of fuel handhng functions. Figure 3.9 shows the building layout during plant operation. Space was prouded for a repair and cleaning facility for maintenance of contaminated equ pment. De fuel transport machme, or cask car, unloaded irradiated fuel from the reactor via the transfer rotor, transported the irradiated fuel in finned pots from the Reactor Building to the FARil via the trestle and unloaded the pots mio the transfer tank rotor. The pot was transferred to a v
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(V) position under a steam cleaning machin: that removed the fuel from the pot and positioned the fuel so that the sodium was cleaned from the subassembly by steaming, followed by a water rinse using an automatically programmed cycle. The subassembly was then placed in a container in the cut up pool, tested for fission product leakage, and transfened to the decay pool for a decay period of approximately 180 days per design before further processing.
Le FARll contained a cold trap system (punfication system) for the transfer tank sodium in a separate substructure room diagonally adjacent to the fuel transfer tank room. He sodium lines and equipment were shrouded in a welded carbon steel secondary stmeture which was inert with nitrogen. The piping outside the walled areas was in the repair pit area and was contained in a concrete vault with a removable cover slab.
3.1.3.3 Ilealth Physics Hullding The llealth Physics Building was dismantled and removed in 1980. The radioactive drain line to the FARll was previously plugged. Only the concrete basement and the plugged, buried drain system remain.
3.1.3.4 Sodium Hullding The Sodium iluilding and the Reactor Building are connected by an underground concrete tunnel. The Sodium Huilding housed the equipment used for storing and purifying the primary sodium. The Sodium Duildmg, Waste Gas Building, and Ine:1 Gas Building formed one structural complex. Figure 3.10
(~T shows the layout of the complex during operation. The building is divided into four sections: the storage V
tank room contains the three 15,000 gallon primary sodium storage tanks; the cold trap room contained a cold trap cell and the equipment required to determine and maintain the purity of the primary sodium; the sodium.potassiam (NaK) room contamed the ventilation equipment and the air to NaK heat exchanger equipment for the cold trap; and the valve control room contained the sodium service hand wheels and motors for the valves, electne panels for the induction heating for the piping, and the control panel. The NaK was used to cool the cold trap. The storage tank room has 30-inch thick cast concrete walls and a 30 inch thick combinatmn of pre-cast and poured concrete roof to provide shielding.
3.1.3.5 Waste Gas ilulldirg The Waste Gas lluilding housed the waste gas disposal system that removed waste gases from the plant by u process which included storage until the gases decayed to a suitable level, dilution below the maximum permissible concentration m nit, and dispersmn mto the atmosphere through a stack. Piping, valves, and mechanical equipnient are housed in chambers below grade; the holdup tanks are housed above grade in shielded cells of the buildmg. Piping transported the waste gas to the FARD where it exited to the atmosphere via a waste pas stack. The holdup tank chambers are inside the Protected Area, while the below grade chamber and the grade level valve operating room are outside the Protected Area boundary. A person cannot enter the portion of the buildmg inside the Protected Area from the portion outside the Protected Area.
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3.1.3.6 Inert Gas Building 4
De Inert Gas Huilding housed the compressors, vt.por trap, hold up and vacuum tanks, valves, piping and other associated equipment for the purification and distribution of the argon cover gas system to the pnmary, secondary, and liARB cover gas systems. Entry into the inert Gas Building chambers is only through hatches in the roof and in the Sodium Building floor.
3.1.3.7 Sodium Tunnel A steel lined tunnel runs from the northwest comer of the Reactor Building to the cold trap room of the Sodium iluilding. It contained primary sodium service system piping. The piping in this tunnel has been 6
drained and capped at the Reactor fluilding and either capped or isolated with shut disconnected valves in the cold trap room. Access to this tunnel is via a manhole between the cold trap room and trestleway.
3.1.3.8 Ventilation Building The Ventilation Building housed equipment for the Reactor Huilding ventilation system. This building has been emptied of almost all equipment and the fence has been modified to be continuous past the east doors of the building.
3.l.3.9 Fission Product Detector (FPD) Hullding This is a small building partly below ground level, to the east of the Reactor Building. The building contained the gaseous fission product detector and piping. The gaseous FPD monitored the fission product concentration to detect the failure of a fuel element. The FPD received primary argon cover gas samples from the reactor vessel and the No.1 and No. 3 primary pump tanks. Access is through a manhole in the roof of the buildmg.
3.1.3.10 East and West Sodium Gallery The cast and west sodium gallenes consist of chambers which hold the secondary sodium lines. %e east gallery supplied the No, I and 2 steam generators and the west gallery supplied the No. 3 steam generator. The sodium lines have been capped where they exited and entered the Reactor Huilding.
Access to the three cast sodium gallery chambers and to the south compartment of the west sodium gallery chamber are via honiontal steel doors just above ground level. Access to the north compartment is via a tunnel from the Reactor Huilding annulus or a sealed manhole.
3.l A General Description of flulldings Outside of Protected Area The buildings outside of the l'rotected Area include the Steam Generator Building, the Turbine Building, the Control Building and the Office lluildmg. The Steam Generator Building houses'two of the three onginal steam generators, and the three 12,000 gallon secondary sodium storage tanks. The three secondary sodium system pumps and moton, and No. I steam ger.erator were removed as part of the original decommissioning effort. The No. I steam generator was shipped to Japan for destructive testing.
The secondary sodium tanks, cold traps. plugging indicator, a collection tank and sump pump system are all contained in the basement area. Stean' produced within the three steam generators was passed from the Steam Generator fluildmg to the adjacent lurbme Building and was used to operate the turbine.
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1he turbine is a 150 Mw(e) tandem-compound, single flow machine in which four stages of feedwater heating were used. The main condenser is a single Dow, divided water box, welded steel unit. Four feedwater heaters, a drain cooler, and a reheated drain cooler were included in the cycle. Three feedwater pumps were provided, two of which were capable of pumping the flow required for the ultimate 430 Mw(t) conditions.
The Of0cc Building is located at the front of the plant and directly connected to the Control Buildmg, which connects to the Steam Generator Building and the Turbine Building. Other buildings associated with Fenni 1 are: the Water Tower, the Potable Water Building, the General Service Water Building, and the Boile house. The Boilerhouse contains a retired oil fired boiler which produced steam to feed the same turbine generator. ~1his building is currently awaiting remont in the near future. AS mentioned in Section 3.1.3,5, sections of the Waste Gas Building are also outi.c the Protected Area.
3,2 DECOMMISSIONING ACTIVITIES 3.2.1 Disposal of Fuel and Blanket Subassemblies Fuel for the Fermi reactor was defined as the 25.6% enriched uraniunVmolybdenum alloy pins contained in the mid-portion of the fuel subass,mblies. This material contained approximately 4J00 kilograms of uranium, which was assigned to the project under an Atomic Energy Commission (AEC) lease agreement. Under terms of the agreement, the uranium was to be retumed to the AEC in the form of UF meeting diffusion plant purity specifications, if a private commercial company in the U.S. could do 6
the work; if no company were available, the AEC would accept the irradiated materials under the spent (O
fuels chemical processing and conversion provisions of the Atomic Energy Act of 1954, using specified 7
hypot!.'ical plant costs. At the time of decommissioning, no private company was available, so the AEC aveed to accept the matenal at its Savannah River Project (SRP).
Two shipping casks were dengned, fabricated and beensed for Fermi fuel. Subassembly cutting and cask loading were accomplished on a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> day,6-day week basis using the FARB cut up pool. The casks were loaded manually through the top with the cask '.pright on the floor of the pool. The cask lid was set m place pnor to lifting the cask out of the pool. After the cask was loaded, it was set in a tray adjacent to the pool assembly, leak checked, and decontammated. Dunng these operations, the spread of contamination was readily controlled and limited to the cut up pool area, where laboratory coats ano shoc covers were sequired. The activity of the pool was typically 5 x 10 5pCikm3; this was principally 137 s (cesium) from the surface of the subassemblies. The first shipment of fuel 60 o (cobalt) and C
C from the site was made on i ebruary 6.1973. On hiay 15, 1973, the last shipment of fuel, which contamed fuel segments located at the hot cell facihty at Battelle hiemorial Institute in Columbus, Ohio at the time of pemianent shutdown, arnved at the Si P.
Disposal of all blanket subassembhes was accomphshed by shipment to the Idaho Chemical Processing Plant (ICPP). 1his involved 962 subassembhes or segments which included a total of 6524 grams of 239Pu. Of the 962 subassembhes or segments 318 were uncut outer radial blankets,168 were cut outer radial blankets,73 were cut inner radial blankets,202 were upper axial blankets,132 were cut lower aua' blankets, and 69 were uncut lower axial blankets. The 'erm " cut" indicates that the nozzle was removed from the subassembly.
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All the blanket shipments were made using two Philadelphia Electric Company Model PII l shipping casks, one of which was owned by General Atomic Company. To utilize these casks to handle Fermi l blanket subassemblies, the Columbus Ohio 1.aboratories of flattelle Memorial Institute designed 14 baskets and basket cor"ainers, which were then fabncated by the Central Ohio Welding Company under 11attelle's supervision. 'the subassembly baskets were constructed of carbon steel gridwork, sized to fit into a cylinder 25 inches O.D. x 158 inches long, containing 37 individual storage holes. Toe basket containers were fabricated under rigid specifications and quality assurance provisions and were made of Type 316 stainless steel,1/4 inch thick in the walls and with a 3/4 inch thick top and bottom.
Five dry mns were made to develop handhng procedures and to establish time requirements before the casks for shipment were loaded. The casks had spacers due to unloading procedure at ICPP which required no more than 1/2 mch axial subassembly movement. A 2. inch diameter, carbon steel mill tube of appropriate length 5 to 62 inches, weighing approximately 0.735 lb/fl was placed over each subassembly as a spacer. Once the subassembly and spacer were in place, the basket was raised from the pool and nilowed to dry ovemight. Radiation levels up to 50 R/hr at the basket surface necessitated isolation of the building and red health physics controls of personnel while the basket was unshielded.
A helium mass spectrometer Ic A test of the container weld followed, prior to closing the cask for shipment.
No prAlems were encountered other t aan when the bottom supports of one basket failed, resulting in km;e pms and basket damage the,; were no radiation exposure incidents, and contamination levels were low. Several pins which wer6 not located by the time the last shipment was made to Idaho were shipped to Nuclear Engineering company's Sheffield, Ilknois, burial site on September 23,1975.
f 3.2.2 Sodium and NaK Removal 3.2.2.1 Nonradioactive Sodlum At the beginning of the retirement program, about 34.600 gallons of nonradioactive sodium was stored in three systems: the secondary system, service system, and storage tanks. The secondary sodium system had been dramed for steam generator mamtenance in 1972, before the decision was made to retire the plant. The major volume of secondary sodium was drained into the three 12,000 gallon storage tanks via the service system by normal operatmg procedures. A complete drain, except about 3600 gallons in the tilX tube bundles, was accomphshed by gravity drain supplemented by evacuating the storage tanks and pressutirmg the mirogen cover pas system. In November 1973, the sodium from the lilX tube bundles u as dramed into the pnmary sodium system.
The secondary sodium service system was dramed directly to a barrel fill station.
This was accomplished by removmg the old secondary sodium system fill line and pressurizing the system with miropen cover gas. A sodium barrel fill station was constructed in the NaK room located in the Sodium limidmg. adjacent to the fonner pnmary and secondary sodium outdoor tank car unloading station. A barrel storage area was built m another adjacent area between the Sodium and Waste Gas lluildmgs. The sodium banels used were USA Standard Department of Transportation No.17 E. Type 1, 55-gallon drums, All drums were visually mspected and leak tested before use.
The transfer of secondary sodium to drums began on May 31,1973, and was completed July 24,1973.
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A total of 569 drums (about 30.400 gallons) were filled All filled drums were sold and shipped to Fike U
Chemical Company of Nitro. West Virgima, to be processed into sodium methylate.
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liased on the above, there is an assumed 600 gallons of non radioactive sodium remaining in the secondary sodium system, the secondary sodium service system and the secondary sodium storage tanks.
1he sodium is a solid at room temperature, but the term gallons is used for historical consistency and consistency with the size of tanks and drums for comparative purposes. Of this 600 gallons, approximately 200 gallons is in the storage tanks as verified by visual inspection (Spring 1997). The semaining 400 gallons is considered a high estimate based on a walkdown of the piping low points which may have prevented drainage. The volume is assumed to be closer to 50 gallons.
3.2.2.2 Radioactive Sodium At the begmning of the retirement program, about 77,000 gallons of radioactive sodium was stored essentially in five major systems: the primary system, service system, storage tanks, transfer tank and FARil service system. Prior to draining the pnmary sodium, samples were taken from the primary semce system No. I sodium storage tank and the FARil transfer tank for analysis. The level of 22Na,137 s, and 90 r.
C S
radioactivity in this sodium was 0.022 pCilg. and was primarily of All nondrainable components had to be siphoned by opening the components and installing special siphon pipes connected separately to the service system. It was necessary to drain the primary system in steps, beginning May 8,1973. First,3800 cubic feet of sodium from the primary system and the lilX and pump tanks was transferred to the storage tanks via the overflow pumps. All remaining sodium required removal via special siphon pipes in individual components.
About 15,000 gallons of sodium was drained and siphoned from the reactor vessel to the storage tanks, O
except for the plenum, in three individual efforts. The plenum was drained separately via the service V
system recirculating line. About 400 gallons of sodium from the overflow tank was removed by drilling mto the bottom of the tank and siphoning. 'lhe FARil transfer tank contained about 8000 gallons of sodium with another 280 m the adjacent ove 110w tank. Iloth tanks were siphon drained with individual siphon pipes. The two storage tanks m the FARil cold trap room contained about 775 gallons of sodium which was transferred to the pnmary storage tanks with the rest of the primary sodium.
There was more primary sodium than could be stored m the three primary storage tanks. The remainder was placed m drums. Drummmg primary sodium was accomplished using the equipment for barreling secondary sodium, with some mmor modifications. A total of 630 steel drums were filled with about 32.000 gallons of primaiy sodium m November 1973. The filled drums were coded and moved to temporary storage in the fuel transport trestleway and Reactor lluilding. Each drum measured about 5 mRhr at a distance of two feet. 'lhe drums were stored four to a pallet, two and tree tiers high withm controlled areas with restncted access for penodic mspection and smear checks.
The remainmg pnmary sodium. about 45.000 gallons, was temporarily stored frozen in the three storage tanks m the Sodium ilutidmg. When the sodium was ready to be shipped offsite, the sodium from the three storage tanks was transferred to steel drumt Dunng this drumming operatioh, samples were collected from representative drums. The maumum levels of activity were: 1.07E 3 pCilgm of 22Na, 137 s and 1,15E 3 pri gm gross F A tatal of 1347 drums, including the drums in h83E 4 pCi'gm of C
temporary storage, were shipped offsite m 1984. The sodium was sent to Idaho to the Department of
- Energy,
,V
.L 8 Rev 0,12/97
_ _ _ _ _ _ _ ~ _ _ _ _ _ _ -
l l
i 9
i 3.2.2.3 NaK At the btcinning of the retirement program. about 940 pounds of NaK was stored essentially in four major kications: the primary cold trap cooling loop, the recirculating cover gas vapor trap, the clean gas purincation unit, and the fission products detector vapor trap. About 2350 pounds of the NaK was t'
nonradioactive and was given to Mine Safety Appliance Company who provided the shipping containers and transportation costs. All radioactive NaK, $90 pounds. containing trace arr.ounts of 137 s was j
C shipped for burial offsite.
Prior to draining the NaK systems, samples were taken and analyzed by PRDC for radioactive constituents.1he results showed that the two contaminated NaK systems contained 137 s activity. NaK C
is aliquid at room temperature.
3.2.2.3.1 Primary Cold Trap NaK loop Drain About 550 pounds of clean NaK.78 (7F% potassium) from the primary cold trap cooling loop was drained to the NaK sump tank prior to '.emoval of the cold trap tank in 1973. The remaining 250 pounds of NaK in the cold trap jacket was drained directly into special Mine Safety Appliance 200 pound capacity NaK drums. The $50 pounds of NaK m the sump tank was transferred to three MSA NaK drums. The residual NaK lefl in the system was passivated with carbon dioxide.
3.2.2.3.* Tlear Argon Puriflestion Unit Drain About 1000 pounds of elem NaK.78 was drained f om the clean argon purification unit into an MSA drum. A siphon now was estabbshed and the NnK was shipped to Mine Safety Appliance Company.
inspection of the argon entramment trap showed that no NaK had carried over from the purification unit.
as the trap was empty. 'the residual NaK m the system was passivated with carbon dioxide.
3.2.2.3.3 Recirculating Cmcr Gas Sodium Yapor Trap System About 540 pounds of contammated NaK 78 was dramed from the sodium vapor trap in the primary
[
recuculating cover gas system. The residual system NaK was passivated with CO. No details were 2
given in the reviewed documents as to how the trop was dramed, and no drain line for the trap could be located on the trap drawings which would allow complete drainage of the NaK.
3.2.2.3.4 l'ission Products Cmcr Gas Sampling Vapor Trap Drain About 50 pounds of contammated NaK.78 was dramed from the sodium vapor trap.for the fission products cover gas sampimg system. Iloth contammated vapor trap tanks were purged with CO to 2
passivate the NaK residuals.
3.2.3 -
Decontamination and Scaling of Contaminated Areas
- Generally, each item slated for dnposal was rmsed to remove loose surfaca contamination and s. < eyed to detennine contamination lesels. In general radioactive or contaminated items were sent for offsite disposal to Nuclear !!ngmeermg Company's Morehead, Kentucky and Beatty, Nevada sites. Nuclear lingmeeting Company's Sheftield. Ilknon site w as also used 39 Rev 0,12/97 s
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3.2.3.1 FARH A significant effort was exerted in decontaminating the FAlm decay and cut up pools. After all equipment was removed from the decay pool and cutup pool, the pool walls were sembbed with a detergent and soapy water. The nnal wash consisted of a 10% nitric acid solution and a demineralized water rinse.
Subsequent to the Gnal cleaning and drying, a 20. mil thick layer of Cooks Spray Booth Shield White strippable paint was applied in several coats to all surfaces of both pool walls, Door, and tunnel.
Approximately 50 gallons of paint were applied using hand rollers with personnel working from a hangmg cage.
The FAltll cold trap piping and equipment was removed and sent offsite for disposal.
The liquid waste and sump pump system has been deactivated, but Ien intact so any potential groundwater leakage can be pumped from the. sump to the FARB liquid waste storage tanks for later disposition. He waste gas stack and its associated equipment were dismantled and removed from the FAltli.
3.2.3.2 Heaclor ilullding ADer shipping offsite all the radioactive material that r<
won w -d..
recoverable, and decontaminating to the extent practicable, the only remainmg '..ajc*
L.......... ming task was sealing (d
the reactor buildmg, the primary sodium system, and the secondary sodium system. It was decided to 7
seal the reactor vessel within the pnmary shield tank and to seal the outlying components directly, using the Reactor iluildmg as an isolation structure agamst personnel eccess to the primary systern.
De iteactor fluilding superstn.eture is of a design conducive to long life with little maintenance. It was decided that since the primary system would be completely sealed and there was no other source of radiatmn or contamination in the buildmg. closute would constitute only the prevention of personnel access to radiation areas. The buildmg itself would be permitted to breathe. No sealing of building penetrations were made, except as related to closure of the radioactive pnmary sodium system, service system, and cover gas system. These systems were then scaled and maintained at slightly positive inert pas pressure to prevent the entrance of water or moisture and to minimize dispersal of any remaining radioactive matenal.
Alter removal of the sodmm. the pnmary system was cooled. A gastight system was then established consisting of the pnmary sodmm system plus the primary shield tank, the machinery dome, and portions of the pnmary and secondary service systems extendmg out to welded pipe caps. This system was nlled and passivated with carbon dioude (CO ) which reacts with residual sodium to fami an inert sohd 2
compound sodmm carbonate (Na2CO ). A CO2 gas atmosphere is maintained under positive pressure 3
withm the sealed system by use of a CO2 bottle gas system. In addition to intemal components such as the OllM, llDM, etc., the retired reactor vessel contams seven safety rods, one control rod,10 lower guide tubes, one antimony.be yhum neutron source, four dumrny sebassemblies, and 198 stainless-steel thermal shield bars, fuel support plates, but no fuel or blanket subassemblies.
7 i
)
v 3 10 Rev 0,12'97
b o
The 3 inch machinery dome exhaust outlet pipe to the waste gas stack was cut nearly flush with the f
shield tank, and the opentag was seal welded. A l/2 inch stainless steel plate was welded inside the exit pipe md a 1/2 inch hat type cover welded over the exit port. All of the argon gas lines were cut and seal-weld capped or partially removed, with the exception of the former 10 pound argon supply line for emergency exit port cooling. This argon gas line was rerouted with 1/4 inch stainless tubing to permanently supply caibon dioxide to ti,e reactor vessel via the exit pipe.1he modified supply Ime was installed and seal welded through the side of the hat type cover, lhe bodium service system piping was cut in the Sodium 11uilding cold trap room :md the Reactor fluilding and caps were welded onto the pipes where they connect to the primary system. Sodium service piping between the Reactor fluilding and the cold trap roorn was closed by valves at one end and by welded caps at the other end, wherever they were cut, Radiation levels in the tunnel have not been measured, but contamination levels in December 1973 were less than 100 dpm/100 cm2, lhe auxiliary fuel storage facility was sealed aner CO2 was added to passivate any residual sodium that may have dripped fmrn the fuel storage pots.
All six neutron detector racks containing four Assion counters, three absolute proportional chambers, four gamma compensated ion chambers, and seven uncotopensated ion chambers were removed from the counter tubes. The neutron counter tube to-operating floor seal sleeves beneath the counter tube boxes were removed and a plate was seal welded over the top of the lower tube beneath the operating floor, 4
1he three primary pump shafts were sealed by removmg the pump drive motors and motor barrels, and O
seal welded stainless steel Deld fabncated covers over the shaft stubs to the top of the shield plugs. The inert gas supply and vent liner to the pump seals and all electrical lines were cut and removed to permit the dismantling and sealing desenbed.1he Taylor pressure transducers and conduits were removed, and the NaK Diled Isinch sensor imes were pinched off, scaled, and stuffed into the shield plig. All three primary sodium pump main dove motors motor banels, shaft seal housings, and motor speed control liquid theostats were sold, The 6ssion pmducts detector 1/2 inch sample imes to the No. I and No 3 pump and the reactor exit port were cut and capped with socket weld 6ttmps m the east sodium pipe gallery. The cover gas equalizing Imes are interconnected, and the siphon breaker lines have no openings, and they remained unmodined.
When the 6.meh overSow Ime between the reactor vessel and the primary overflow tank was cut and capped, the overSow tank was separated from the pnmary system Subsequently, the overSow tank was purged separately to accomplish a CO passivation of the residual sodium deposits. When passivation of 2
the sodium in the overSow tank was completed the 4 mch gas supply and pressure equalizing line in the top of the tank was cut to encourage contmued oudatmn of the sodium deposits and to prevent any unexpected buildup of hydrogen as a result of moisture leaking into a sealed tank. The overSow tank was left open indefinitely to the air atmosphere of the lower Reactor iluilding througli the 4 inch line.
The breech in this 4 inch hne is the only remammg opemng in what was formerly the primary sodium system.
G ).
All Rev 0,12M
~..
O 3.2.3.3 Sodium llullding t/
1he primary sodium cold trap and most of the associated equipment were removed during the initial retirement.
'the primary sodium service system was secured by cutting the service piping at the tunnel ends. The lines to the overnow tank and the intermediate heat exchangers (IllX) drain line were cut. The nor.zles b
of the tank were capped. The residual sodium on the walls of the piping was convened to Na2CO3 y purging the piping with CO. T.ie remainder of the primary sodium service system including the hot 2
trap economizer, heat exchanger and pump, and all piping except the storage tank hne have been removed from the Sodium lluildmg.
The nitrogen-to water cooling system for the sodium storage tank room atmosphere was dismantled, and the blower and heat exchanger equipment scrapped. The nitrogen atmosphere in the sod!um storage tank room was replaced with air. The sodium tanks were passivated with CO2, welded shut, and a N2 blanket provided.
3.2.4 Protected Area lloundary 1hc Fermi 1 Protected Area lloundary was revised to exclude many nonradioactive areas such as the Steam Generetor iluilding, the Control and Office lluddings, and Turbine 11uilding. The Protected Area lloundary as shown in Figure 3.4 is marked by a seven foot high chain link fence and building walls that enclose the FARil, the Reactor lluilding, the Sodium 11uilding, and the Waste Gas and Inert Gas n
fluildings. The llealth Physics lluildmg has been dismantled.
L) 3.3 CUltRI:NT PI, ANT CONI)lTIONS 3.1.1 Primary System Retnement of the primary system meluded drammg all the primary sodium and renlling with an inert gas (mtrogen) to which CO was added to reduce the residual sodium deposits to inactive solids. The system 2
was then sealed and maintamed at shghtly positive inert gas pressure to prevent the entrance of water and'or moisture and to minimize dispersal of any remammg radmactive material. A gastight system was estabhshed consisting of the pnmary sodium system plus the primary shield tank, the machinery dome, and portions of the primary sodium service and secondary sodium systems extending out to welded pipe capt 'this system was filled with carbon dioside, which reacts with the residual selium to form an inert CO ). The pnmary system is presently being maintained under a positive pressure sohd compound (Na2 3
by use of bottled CO.
2 The machmery dome has been seal welded m place. The machinery dome glass observation ports.,
gaskets and gasket compressors were replaced with carbon steel plates which weis seal: welded in place.
Its access door was also seal welded m place.
3.3.2 -
s'rlmary Sodium Storage Tanks Nitrogen cover gas is mamtamed on the pnmary sodium storage tanks. When N pressure decreases.
2
(]
additional N2 is manually added The door to the sodium storage tank room remains locked except C/
dunny entnes.
.L12 Rev 0,12/97
i 3.3.3 L6 quid Waste Disposal System All potentially contaminated drains and sumps collect in the hot sump in the Fuel and Repair fluilding.
[
The radioactive liquid waste (LRW) collected in the hot sump can be transferred to the liquid waste tanks,(MK 7,8,9,15). Liquid quantities in tnese tanks are monitored and recorded. Currently, there are no operational capabilities for discharge. Discharges can be performed if monitored in accordance with the Technical Specifications. The capability to discharge LRW would have to be established and procedure pit, sted if discharges were to be required in the future.
3.3A Electrical Supply Power is supplied to Fermi 1 through the 120KV switchyard located south of Fermi 1, Two offsite power lines supply the switchyard. The switchyard provides power through the switchgear room in the i
Fermi i Turbine lluilding to the motor control center supplying, in part, loads in the Protected Area. The motor control center is located on the first floor of the Fermi 1 Control Building. A control battery h>cated on the second floor of the Control Building provides DC power to Fermi 1.
i 3.3.5 Sodium Memnining in Plant Systems l
As part of the initial decommissioning of Fenni 1 following shutdown, the sodium (Na) in the primary l
system, secondary system and supporting systems was drained. Due to the configuration of some of the i
equipment and systems, a small fraction (less than 1%) of the total Na volume could not be removed.
This residual Na plated out on the walls of the equipment and piping, or pooled in the low points of the
~O equipment and systems forming heels. ~the majonty of the equipment and piping which was abandoned in place was blanketed with CO gas to try to react any residual Na (passivate) to fen non reactive 2
CO ). lixperimentation at the time showed that due to the permeability of CO2 sodium carbonate (Na2 3
CO crust to form less than 1 " thick, beyond which CO, the potential exists for a Na2 3
/4 gas through Na2 3
no further reaction between the CO and the Na may occur.
2 The sodium is a solid at room temperature, so the remaining sodium is in a solid form. The unit
" gallons" is used to describe the amount of remaining sodium for histotical consistency and consistency with the size of tanks and drums for comparative purpc'ses, llased on information documented in References 3 -and 4, periodic technical reports covering the original decommissioning period, and observations of some tanks in 1997, approximately 700 gallors of residual sodium is estimated to be contamed in Fermi i systems, of uhich approsimately 435 gallons is slightly radioactive. Some of this sodmm has been converted to sodium carb(mate or bicarbonate, but some is still sodium metal.
The residual primary sodium is expe ted to be located ir the following buildings:
L lleactor fluilding and sodium tunnt, approsimately 374 gallons e-
- FARil. approximi,tely i1 gallons.
Sodium iluilding approximately 50 gallons t
Approximately 225 2fs5 gallons of secondary sodium residuals are estimated to be in the Steam
.(" j-9r iluilding, though one report.would lead to an estirnate of 725 gallons of secondary sodium.
Gem
- ach tank.
3 13 Rev 0,12/97
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Od 3.3.6 Taylar Instrume itation Taylor Pressure Instrumentation wa: used throughout the plant and contained NaK as a transmitting medium w'tich is contained within armor covered capillary lines. These capillary lines are expected tc contain unreacted NaK since they have been sealed. A total of 33 of these instrument lines were identified. He 15 NaK fuled instrument lines that were instalkd in the 5 team Generator Building were removed or disconnected in 1997.
3.3.7 Monitoring and Alarms 3.3.7.1 Water Intrusion System Monitoring detectors for water intrusion are located in two e.reas: the Fuel and Repair Building basement hot sump and the Reactor Building biological shield annulus. Accumulation of water in these areas activates an alarm in the control room at Fenni 2. The water intrusion alarm in the Reactor Building biological shicit annulus will sound if the water level exceeds an accumulation of approximately six inches. The water intrusion alarm will sound for the FARB waste water sump if the water level is greater than the lower grating level over the sump pit.
3.3.7.2 Cover Gas Monitoring The primary system cover gas pressure is monitored with high and low pressure alarms. The monitors O
and the alarm circuitry are periodically checked and calibrated in accordance with the Technical U
Specifications and written procedures.
3.4 ACCESS CONTitOI.
The area encompassed by physical bamers and to which access is controlled is the Protected Area. The Protected Area is enclosed by either a chain hnk fence or building walls which piovide equivalent degree of resistance to penetration. The fence is topped by three or more strands of barbed wire or brackets angled outward with an overall height of no less than seven feet. Normal entry to the Protected Area is through a normally locked gate m the fence adjacent to the Sodium Building. Other doors that are a part of extemal walls, which act as a part of the Protected Area boundary, are locked or permanently sealed.
Access to the Protected Area is controlled, hmited, and recorded. The access key is at the Fermi 2 Itadiation Protecuon Control Pomt. A second key is held in sale keeping by the Custod.an for use only m extenuating circumstances.
Wntten procedures delineate the requirements associated with entry into the Protected Area and specific areas withm the Protected Area to prevent unauthonzed entnes and to protect the safe'ty and health of authonzed personnel.
G l
3-14 Rev 0,12/97
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I O
SECTION 4: RADIOLOGICAL CONDITIONS
/
4.1 Total Nuclide Inventory 4.1.1 Radionuellde Inventory Activation analyses were performed for the reactor vessel, the primary shield tank, and the concrete, to determine the 1986 radionuclide inventory and project the inventory at the end of the 40 year S.
1 TOR period. Reference Numbers 11 and 13 provide the details for Table 4.1, a summary of the resuits of analyzing the reactor vessel, primary shield tank, secondary shield wall (concrete), reactor vessel internals, sodium residuals, and liquid waste samples.
The 40 year SAFSTOR will result in the following estimated reductions between 1986 and 2025:
85% reJuction in total activity 90% reduction in liquid waste activity 89% to 90% reduction in exposure ne inventorv of radionuclides in the waste water was based on measurements of samples from the waste water sump, tank dose rates and activity measurements for the final total discharge after the 1973 to 1975 decommissioning work. The 1986 total activity for 60 o and 37 s was estimated at 6.0E+3 pCi cach.
C 1 C This activity will decrease to 3.1E+1pCi for 60 o and 2.4E+3 Ci for 137 s in forty years.
C C
]v Dunng the 1983 sodium drummmg operations, primary sodium samples were taken and analyzed for isotopic concentrations and activities. These results were used to perform total activity calculations for the remaining residual sodium in July 1986. The residual sodium isotopic activity was found to be 9.8E+02 pCi for 22 a and 4.8E+3 pCi for the 137 s rspectively. After 40 years, the activity would N
C reduce to 2.5E-2 pCi and 2.0E+3 pCi respectively. This amounts to a 65% reduction in sodium activity.
4.2 Radiation and Surface Contamination Levels 4.2,1 Dose Rates As indicated on the building maps, Figures 4.1 and 4.2, most of the general areas in the Reactor Building, the Fuel and Repair fluildmg (FARil). and the Sodium Building were surveyed as part of the scoping effort perfonned in 1997. The observed dose rates were typically < 0.1 mrem /hr with dose rates in many areas on the pR/hr order of magmtude. Table 4.2 summanzo the results of the 1997 scoping surveys.
The highest dose rate observed m the Reactor Buildmg basement was located on the No. I lilX (3 mrenVhr on contact and 1.5 mremihr at 30 cm). The general area dose rates in the remainder of the basement were en the order of 0.1 mrent'hr. Other areas where general area dose rates exceeded 0.1 mrem /hr include the primary sodmm storage tank room (0.4 to 1.2 mrem /hr), the liquid waste tank room (0.1 to 5 mrem'hr), and the steam cleaning chamber (0.4 to 2 mrem /hr).
A portable gamma spectroscopy umt was used to identify the isotopic mix of the crud causing the hotsr9 on the lilX. The only isotope detected was 137Cs (30.07 year half-life).
!v 41 Rev 0,12/97
[]
Dose rate surveys in the Turbine Building, Office Building and portions of the Steam Generator Building were performed using a Ludlum 12S Micro-R instrument to verify that these areas were unaffected by the operation of Fermi 1. No dose rates in excess of normal background levels were obsen ed.
4.2.2 Contaminated Areas ne scoping surveys performed in 1997 focused primarily on areas assumed to be radiologienlly uncontaminated. No contamination was detected in these areas. Significant contamination was found in the steam cleaning chamber (up to 200 mrad /h smearable). This area was expected to be contaminated based on survey information contained 4. ihe 1972 Fermi 1 Retirement Report (140,000 dpm/100 cm2).
The highest contamination levels v.ere on the fuel transfer ports to the cut-up pool and to the transfer
- tank. Uamma spectroscopy analysis of an air sample from the chamber identified 137 s as the only C
radionuclide present.
A room ;ocated in the basement of the FARB was used as a temporary storage area for depleted uranium.
He urar.ium was removed before or during thc retirement of the plant. Low levels of beta gamma and alpha contamination were detected during the 1997 scoping survey of this area (106 dpm/100 cm2 and 34 dpm/100 cm2, respectively).
The other known contaminated areas include the FARB hot sump, the cut-up pool, and the decay pool.
These are highlighted in Figure 4.2. All other contamination is located inside system components.
Asbestos samples taken from the Reactor Building and NaK room were not contaminated.
(3 Surveys of the decay pool and cutup pool were last performed in 1986. Radiation and surface
\\s' contamination levels measured are shown in Table 4.3.
/
)
4-2 Rev 0,12/97
-.~
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B = contaminated areas
~
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= surveyed areas
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Pit Building 8 10 (Dismantled) Slab (FARB) uR/hr
.i Sodium Ilullding A
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Key:
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O Access Manhole Readings in areas outside the Protected Area are in the background range.
No contamination was detected outside the Piotected Area.
O..
Figure 4.1 l'ermi 1 Area fluildings Rev 0, I?/97 4-3
L Liquid Waste Test Tank i
<0.1 mR/hr 0.2 mR/hr Uranium _
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- 4
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= surveyed areas
(
Figure 4.2 Fermi 1 FARB Basement Rev 0.12/97 44
E') '
'TAHLE-4.1 CALCULATED ACTIVITY IN THE REACTOR VESSEL, ITS INTERNALS, THE PRIMARY SHIELD TANK, AND.THE SECONDARY SHIELD
- JULY 1986
-
- REALTOR VESSEL TOTAL ACTIVITY. CURIES ISOTOPE JULY 1986 JULY 2026 JULY 2086
" Nb 1.87 E 06 1.87 E-06 1.87 E-06
" Co 1.29 E-00 6.73 E-03 2.53 E-06
" Ni 2.12 E-03 2.12 E-03 2,12 E-03 63 Ni 2.50 E-01 1.87 E-01 1.21 E-01
C 9.50 E-10 9.50 E 10 9.50 E-10 ss Fe 5.09 E-01 1.77 E-05 3.61 -12E TOTAL 20.5 E-08 1.96 E-08 1.23 E-08 e REACTOR VESSEL INTERNALS TOTAL ACTIVITY CURIES
" Nb 4.06 E 02 4.06 E-02 4.06 E-02 O.
" Co 274.
1.41 5.2 E-04 sv Ni 3.92 3.92 3.92
Ni 87.
64.
41.
" Fe 110.
2.5 E-03 2.92 E 10
'IOTAL 415 20.
45, ERIMARY SIIIELD TANK TOTAL ACTIVITY. CURIES
C 3.92 E-10 3.92 E-10 3.92 E-10
" Fe 8.13 E-02 2.82 E-06 5.77 E-13 TOTAL 8.13 E 02 2.82 E-06 3.93 E-10
_ CONCRETE SECONDARY SillELD TOTAL ACTIVITY. CURIES e
'" Eu 2J5 E 03 3 44 E-04 1.52 E-05 Reference 11 4-5 Rev 0,12/97
J.;\\ -
' TABLE 4.2 j
SUMMARY
OF RADIOLOGICAL SURVEY DATA -- 1997!
=1 CONTAMINATION -
GENERAL AREA -
CONTACT.
LEVEIRe DOSE RATES
- DOSE RATES dpm'100 cm*
p.etor BuilAino
- Operating Floor.
3 5 pR/hr
<1(C / <20 e
e-Basement Elevations --
0.1 mR/hr 0.5 3.0 mR/hr
<!CO / <20 -
- - - Atmulus Ares 10-30 pR/hr
<100 / <20 6
- Fuel and Renair Bui1Aing Steam Cleaning Chamber 0.4 2.0 mR/hr 6 mR/hr 200 r.iradh" '
e
- Fuel Transfer Tank Room -
. 0.2 mR/hr 0.2 mrem
<10) / <20 Uranium Storage Room 14 pR/hr
' 14 pR/hr 106/34 e
- 15 Liquid waste tank rm.
0.1 2.0 mR/hr 0.3-4.0 mR/hr
<100 / <20 e
. #7,#8,#9 liquid waste tank rm.
0.15 mR/hr 0.416 mR/hr
<100 / <20 Liquid waste pump room 8-40 pR/hr 50-80 pR/hr
<100 ' <20
<100 i <20 e-Mezzanine area 610 pR/hr Decay pool room.
1115 pR/hr
<100 / <20 e
- . Cut up pool room 10-12 pR/hr
<100 / :20
<100 / <20
- - New Fuel transfer chute
- 20 pR/ht_
<100 / < 20
- Transfer tank cold trap rm..
<0.1 mR/hr
<100 / < 20 Repair Pit 810 pR/ht-e-
<100 / <.10 590' genersi walkways -
6-10 pR/hr
+
<100 / <20 576* general walkways 510 pR/hr o
l
[~
Sodium Buildine i
- Na storage tank room
<0.2 1.2 mR/hr.
<100 / <20 t
e Cold trap room (0.1 mR/hr
<0.1 mR/hr
<100 / <20 e
Waste Gas Building e Valve room 10-14 pR/hr
<100 / <20 Inert Gas Building
- Tank room i5-60 pR/hr 120 R/hr
<100 / <20 e - Inert Gas Tunnel 1416 pR/hr
<100 / <20 Steam Generator Building
+ -Annulus sump room :
7 8 pR/hr 7-8 p R/hr
<100 / <20 Office rurbine Buildines r
-* All elevatio~ s 412 plUhr
<100 / <20 -
n Mieellaneous Northeast Na gallery -
F 12 plUhr 8-12 pR/hr
<100 / <20
- <100 / <20 means <100 dpm /100 cmi beta gamma and <20 dpm /100 cm alpha contamination 2
Due to the high beta gamma contanunation levels, the smears were scanned with a PAC-4G alpha -
meter in lieu of a Irboratory analysis of the smears. No alpha activity was detected.
Note: Eberime RO-2f Eberkne E 520. and Ludlum 12S survey meters were used for dore rate measurements. Ludium 177, Tennelec LD5100, and Eberline PAC-4G survey meters were used for
.(g contamination measurements.
4-6 Rev 0,12/97
- , c L
Table 4.3
- Radiation 'and Surface Contamination Levels Associated with Decay Pool and Cutup Pool-1986 Contamination Levels,
- Radiation I.evels dpm/100 cm' L( cation mR/hr.
Beta-Gamma Alpha
~i I'
' Decay pool and cutup area -
1 E-2 to 2 E-2 Floor dra'ns. pool area 1.5 E-2 to 2.5 E-1 Decay pool,inside 1 E-2 to 8 E-2 Decay pool, tunnel
'l E-1 to 4 E-0 1500 to 3500
<20 l
l
. Cutup pool, inside 8 E-2 to 2 E-1 1500 to 4000
<20.
Cutup pool, tunnel 1E1 f
2 1
8;.
! O.
l
[.
l A
4 i
O 47
- Rev 0,12/97
SECTION S: SURVEILIANCE 5.1 Surveys Two types of surveys are identified in the Fermi 1 Technical Specifications, environmental and radiological, in addition to periodic facility inspections and instrumentation testing.
For the environmental surveys, stations hwe been established where it is estimated that maximum concentrations of radioactive material discharged from the facility may occur.
Environmental surveillance is not required until discharge of radiological liquid effluents has commenced; this is in compliance with Amendment No. I1 of the Possession-Only License No. DPR-9. The water and sediment sample points and frequency are contahed in the Fermi 1 Technical Specifications.
Periodic radiation surveys are performed to check for the presence of gamma radiation and transferable contamination at the frequency specified in the Technical Speciiications.
Gamma radiation measurements using portable survey instruments and contamination checks using smears are made in the following areas:
Reactor Building - Operating floor, doors and seats around machinery dome, breather pipe, sump pump serving Reactor Building annulus.
Fuel and Repair Building - Pool area, operating floor access peints to contamination areas, steam cleaning room access plug.
(v) 5.2 Inspections and Tests A weekly general walk-through and inspection of the Protected Area is conducted. A mc,nthly visual mspection of the Protected Area is also perfonned. The monthly inspection consists of visual inspection of the fence, gates and exterior doors of the Protected Area, a check and recording of the level ofliquid in the liquid waste tanks, and a visual water level check from the top access of all active sumps which servt the Protected Area. All abnormal conditions observed are recorded and reported to the Custodian.
Weekiy, a continuity test of the FARB waste water sump and Reactor Building biological shield annulus water ietrusion alarm circuits is performed. Testing of the water intrusion monitors, and the primary cover gas pressure alarms is performed every six months. Testing of the carbon dioxide pressure relief valve is performed annually. All testmg is performed in accordance with wTitten and approved procedures.
Techmcal Specifications requirements also have been estabbshed for weekly surveillance of the priniary system cover gas pressure and ti.e mtrogen cover gas pressure for the primary sodium tanks. This surveillance frequency is adequate because a tempoiary loss of pressure of the cover gas will not cause significant water reacticns with the residual sodium since it is expected that the exposed sodium has been passivated at this time and water is not likely to enter the systems because of cover gas pressure loss.
The primary system is located w: thin the Reactor Building.nd the primary sodium tanks are located in the Sodium Building complex. All components containing significant residual primary sodium are therefore inside buildings and protected from the elements.
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!!f discharges.were to resume, during periods when radioactive effluents'are being discharged, the
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', : discharge radiation monitor is required to be checked with a source once a week and calibrated at least once every six inonths or before each discharge batch..L The radiation monitor surveillances would be
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SECTION 6:
ADMINISTRATIVE CONTROLS The Detroit Edison Company (Edison), as the licensee for Fermi 1, has the responsibility for trair.'.aining the License and Technical Specifications. Administraive controls have been established to ensure that management and administration of Fermi i is perfonced in a con;istent manner that complies with regulatory requirements.
Responsibility for Fermi 1 is delegated through the line organization of the Senior Vice President,-
Nuclear Generation. A Review Committee functions to advise the Custodian on all matters relating to nuclear safety and to review and approve procedures, design changes, Licensee Event Reports, and other activities.
6.1 Organization and Responsibilities 6.1.1 Senior Vice President, Nuclear Generation The Senior Vice President,- Nuclear Generation is responsible for overall plant safety of Edison nuclear
- power plants, ine'~iing Fermi 1.
6.1.2 Nuclear Assessment Manager The Nuclear Assessment Manager reports to the Senior Vice President, Nuclear Generation, and is responsible for the Fermi i facility. Reporting to the Nuclear Assessment Manag:1 is the Fermi 1 Custodian.
6.1.3 -
Other Support Fermi 2 organizations provide suppet for the Fermi i facility as needed.
6.1.4 Custodian The Custodian or Cuaodial Delegates shall be responsible for direEig all Fermi 1 activities, seeing that the activities are done in a safe manner and in compliance with the Technical Specifications, and reporting these activities to the NRC. Key responsibilities include:
Coordinate, approve, and assign work done at the facility.
e
. Maintain the physical facility as defined by Techmcal Specification B.1, Figure B-1.
Comply.vith the Technical Specifications; admmistrative controls; and local, state, and Federal Regulations.
- Plan, control, an,i monitor decommissionmg activities.
Assign duties to the Custodial Delegates and Cu.,todial Agents as required.
e Control access to the Protected Area via a reserve access key that shall be used in extenuating circumstances only.
Approve temporary changes to the Fermi 1 Manual.
Review Fermi i b lanual and design changes.
Maintain records in accordance with the type and retention period stated in the Technical A
Specifications, Section 1.9.
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The Custodian shall be appointed in wTiting by the Nuclear Assessment Manager, and approved by the Senior' Vice President, Nuclear Generation. The Custodian shall, as a minimum, have a basic
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understanding of surveillance and llealth Physics procedures and the Fermi 1 Manual.
6.1.4.1 Custodial Delegates Custodial Delegates shall act in the absence or on behalf of the Custodian and shall assume such duties and responsibilities listed in Section 6.1.4 as required or assigned by the Custodian, Custodial Delegates shall be appointed in wTiting.
6.1.4.2 IIcalth Physicist The llealth Physicist shall review all procedures and limits involving the handling of radioactive materials. This individual shall be responsible for ensuring that all plant discharges and shipments are within the limitations set forth in the Code of Federal Regulations.
The llealth Physicist shall be appointed in writing by the Custodian and shall have two years of specialized training in health physics or equivalent and three years work experience related to radiological health and safety.
6.1.4.3 Custodlal Agents Custodial Agents shall have unescorted access to the Protected Area and may perform activities such as O
surveillances, health physics activities maintenance, escort duties or other activities. Custodial Agents shall be authorized in writing by the Custodian or Custodial Delegates.
6.1.4.4 IIcalth Physics Technician A person who has received training in health physics techniques and procedures shall be on site and may direct health physics activities whenever radioactive materials are being moved.
6.2 Quality Assurance Program 6'*
Introduction The purpose of the Fermi I Quahty Assurance Program is to provide assurance that work is performed at the Fermi l nealcar facility m a quality manner. The program meets the requirement of 10 CFR 50.54a.
It has been est.lished for the S AFSTOR condition, based on the condition, status, and history of the facihty. This program was initially developed in 1997. Prior to the program's inception, there were admmistrative contiols and review requirements established, but not a Quality Assurance Program. The Qup'ity Assurance Program applies to work in the nuclear portion of the Fermi i facihty, which is the pon m it. side the Protected Area. The area conesponds to the description of the facility in Fermi 1 Technical Specifications. Note that this Quahty Assurance Program does not implement Appendix 13 of 10 CFR 50. There are no systems or components at Fermi I that are used to prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public.
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6.2.2 Organization t
The Senior Vice President, Nuclear Generation is responsible for Detroit Edison's nuclear power facilities. The Fermi 1 Custodian is responsible for the Fermi i facility. The Custodian's responsibilities are discussed in Section 6.1.
Appointed Custodial Delegates can fulGil the responsibilities of the Custodian. Other Fermi 1 workers are responsible for the quality of their work. Fermi 1 workers may have concurrent responsibilities at Fermi 2 or elsewhere in the Detroit Edison Company.
6.2.3 Procedures Procedures ensure that the requirements of the Technical Specifications are carried out in a proper and timely manner. They also serve as training and reference units for future Custodians, Custodial Delegates and Custodial Agents. Administrative proadures include Custodial responsibilities and authority, Procedure Manual control, Custodial Delegate and Custodial Agent selection and function.
reporting procedures and Review Committee functions. In addition, there are appropriate technical procedures for details ofinspections, surveillances and operation.
Procedures are reviewed and approved by the Review Committee. The Custodian may temporarily change a procedure by written document following a determination that the change does not constitute a signi6 cant increase in hazards associated with the facility.
6.2.4 Design Control Design change documents are used to modify installed Fermi i systems. Modincations to systems or components previously disconnected from the Fermi i current systems and/or abandoned may be performed with a design change or other work control document. Design change documents are maintained as records to preserve infomaation on the con 0guration of the plant. Additionally, the Fermi i Safety Analysis Report will be updated at least once per 24 months to provide a current mtegrated description of the facility.
Design change documents are reviewed by the Custodian, Technical Engineer, and Heaith Physicist, or their delegates, and approved by the Fermi 1 Review Committee.
6.2.5 Work Control Repair, mamtenance and modi 6 cation of mstalled systems, structures and components shall be performed using a work control document. Approval to start work shall be noted on the document by the Custodian or delegate. Any post maintenance testmg. inspection and Technical Specincation out-of-service requirements apphcable to the work activity shat! be speci0ed on the work control document.
6.2.6 Document Control The Fermi i Manual has been issued as a controlled document to a speci6ed users list. Procedures for use are obtained by copying from one of the controlled manuals.
Copies of records are obtamed from the site records management organization, nN 6-3 Rev C 12/97
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'A cha'nge cont ol process has been established which requires specific reviews and approvals of change
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documents, such as' procedure changes, design changes, and license changes.. Revisions to approved change documents require the same review and approval process, with the ~ exception of temporary changes to procedures, as addressed in Section 6.2.3.
Typically, the Fermi 1 Review Committee -
approves the change documents.
- In the future, as design changes are made, an appropriaie drawing (s), as applicable and available, will be marked up and attached to the design change. Sketches may be used as an attemate.
6.2,7 _ ' Special Processes
-Welding, heat treating, and non destructive testing ofinservic( systems will be controlled and performed by qualified workers. Non. destructive testing performed fot information only purposes does not fall within the controls of this Quality Assurance Program, nor does cutting using welding equipment.
-6.2.8 Measuring and Test Eq4ipment
- Measuring and tert equipment (M & TE) used to perform Technical Specification surveillances shall be calibrated periodically.
6.2.9 Records Records required by Fermi i Technica: Specifications are maintained as Quality Assurance recoids for at least five years, or duration o the license, as specified in the Technical Specifications.- _These records r
include. records required by 10 CFR 50.75(g). As. built drawings for Fermi I did not exist at the time 10 CFR 50.75(g) was issued, so as allowed by the regulation, other available information conceming areas and locations in which radioactive materials v.>ere stored and used were substituted. This attemate information currently includes a special list of where radioactive material was used, some dr, wings, and current design changes after issuance.
The quality assurance records are maintained in Fermi 2 facilities for records.
E 6.2.10 Corrective Action Signi0 cant conditions adverse to quahty shall be documented, a cause analysis perfotmed. and corrective action to prevent recarrence implemented. Conditions meeting the Technical Specification reportability
- erjteria shall be considered significant conditions adverse to quality.
-6.2.11 Review The Fermi 1-. Review Committee reviews performance at the Fermi i facility. Their* reviews melude
- procedures, design changes, and.10 CFR 50.59 safety evaluations. See Section 6.3 for further discussion ion the Fermi l' Review Committee.
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Audits of facility activities and Quality Assurance Program adherence will be perfermed at least two times per year. If performed by an individual, tre individual will be independent of line responsibility for Fermi 1. If performed by a team, a minority of the team may be responsible for line activities at Fermi i to provide Fermi 1 expertise to the team. In this case, the audit team leader must be independent of the line responsibility for Fermi 1. Review responsibilities (e.g., a Review Committee member) are not considered line responsibilities for determination ofindependence.
The audit results shall be documented and presented to the Fenni 1 Review Committee and Custodian.
6,3 Review Committee The Review Committee shall function to advise the Custodian on all matters relating to nuclear safety.
The Review Committee shall be responsible for review and approval of the following:
Procedures for activities at the facility.
Annual report to the NRC Licensee Event Reports e
Design changes e
10 CFR 50.59 Safety Evaluations License Amendments, including Technical Specification changes Facility monitoring results e
Audit Subcommittee Reports e
The Review Committee meets at intervals not exceeding 13 months and prepares and distributes formal mmutes of its meetings. Special meetmgs may be called by the Chairman, Custodian, or one of the Custodial Delegates, as required.
The Review Committee is composed of five or more personnel from within the Detroit Edison organization or consultants, at least three of whom have had two years or more of experience in a
responsible position at an operatmg nuclear power facility and have had basic health 1 ysics training, 6-5 Rev 0,12'97
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7.0L F1RE PROTECTION PROGRAM -
.The Fermi 1 Fire Protection Program consists of Se following provisions:
i le' Fire extinguishers are-located at selected areas within the facility.
Most are dry chemical
- extinguishers, Class D extinguishers, which are used for metal fires, are also provided in some areas, since water, CO and dry chemical extinguishers cannot be used for sodium or NaK fires. Fire 2
extinguishers are checked periodically.
e = In case of a fire at Fermi 1 the Fermi 2 fire brigade responds. Frenchtown Fire Department responds to fires requiring offsite assistance.
. ; Welding and other hot work is controlled by permit. Fire watches are appointed during hot work.
f The Fermi 1 General Service Water System provides water to fire hydrants outside the Fermi i Protected Area. Two electric pumps and a diesel driven fire pump are installed in the system. When
. the system is out of service, reliance is on the fire extinguishers and on the Frenchtown Fire
- Department's ability to pump from Lake Erie.
The fire hazards at Fermi 1 include:
. i Office and record storage areas Training areas r
-e Residual sodium and NaK e
Residual oils located in plant equipment e.
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= SECTION 8: - ENVIRONMENTAL ASSESSMENT c
ne SAFSTOR status of Fermi 1 will not cause a significant environmental impact. The Fertni 1 original
- decommissioning effort was completed in 1975. Since this involved dismantling and shipping the radioactive fuel, blanket assemblies, and mechanical components offsite, the highest potential for impact -
due to decommissioning has already occurred.'
On April 29,1969, Detroit Edison filed an application with the AEC for a permit to construct Fermi 2:
. Construction Permit No. CPR 87 was issued on September 26,1972 following reviews by the AEC staff, _
Advisory Committee on Reactor Safeguards, and public hearings dealing with environmental matters before an Atomic Safety and Lic-nsing Board. The staff's conclusions were issued as a Final Environmental Statement (CP FES) in July 1972. On April 4,1975, Detroit Edison docketed the Environmental Report (ER OL) in support of the application for an operating license. In August 1981, tl.e NRC issued the " Fir.,1 Environmental Statement Related to the Openttion of Enrico Fermi Atomic Power Plant, Unit No. 2," NUREG 0769. The OL FES presents assessments that supplement those described in the CP FES. The report is written in accordance with 10 CFR 51 which implements the
. requirements of the National Environmental Policy Act of 1%9 (NEPA).
As discussed in Section 3.1.1 Fermi i shares the ll20 acre site with Fermi 2. The environmental information presented in this section is mostly based on relevant information and studies presented in the -
Fenni 2 ER OL, Updated Final Safety Analysis Report (UFSAR), and the NRC's OL-FES. The environmental information is applicable since much of the data was collected at the time of
-decommissioning or after Fermi 1 was decommissioned. Information on personnel working at Fermi i has been updated.
l-8,1 Distribution of Estimated Population Tables 8.1 and 8.2 represen' the population data around the site during the year 1987 and the predicted
_ population in the year 2000, respectively.
8.2 Non RadiologicalImpacts The regional demography and land use, site water use, site ecology, geology and meteorology hase not j;
changed significantly since described in Section 2 of the Fermi 2 ER OL and UFSAR.
8.2.1 Socioeconomic and Cultural Resources Few personnel are required for mamtenance of the facility in SAFSTOR to perform surveillances, rounds, maintenance, audits, and other activities. Some of the individuals performing work at Fermi 1 are employed by. Edison in other capacities. Specific activitics require additional resources, but the numbers are small compared to the total number of site workers. Because of the relativqly small number -
of people involved, there is virtually no impact on the community and traffic patterns.
8.2.2 Land Use The SAFSTOR of thd facility will not affect land use onsite or offsite.
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8.2.3 -
' Hy'drology V
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LThe hydrology of the site and its envitons has not changed significantly since the Fenni 2 ER-OL and UFSAR. It is not anticipated there will be any significant changes over the '. AFSTOR period.
8.21 Water Use.
. While in SAFSTOR status, Fumi 1 only uses water systems, such as service water and potable water.
- that also directly support Fermi 2.-- Liquids collected in the FARB sump system are mostly due tc the intrusion of underground or rain water. The FARB waste water sump can be transferred to the liquid waste tanks. Currently, there is no active State of Michigan NPDES Permit covering the discharge of waste waters from the Fermi l facility. Stonn and sanitary sewers associated with the Fermi i facility are connected to the Fermi 2 site storm anc' 3anitary sewer systems.
8.2.5 Aquatic and Terrestrial Resources The aquatic and terrestrial ecology of the site and its environs is presented in the Fermi 2 ER OL and further discusseo in the NRC OL-FES. No activities are currently anticipated that will impact the aquatic and terrestrial resources.
8.2.6 Unavoidable impacts Fermi l occupies a small restricted area of the Fermi site over the SAFSTOR period.
Some of the unrestricted portions of the Fermi i facility are effectively used by Fermi 2 such as:
. - The Office and Turbine Buildings house training and storage facilities.
The potable water system supplies the potable water for the Fermi 2.
The general senice water intake structure er Fermi 2 is located on the Fermi 1 intake canal.
The Fenni I switchyard provides one division of offsite power for Fermi 2.
8.2.7 Local Short-Term Uses Versus Long-Term Productivity The site is presently being used for production of electricity by Fermi 2 and there are no plans for the site other than electrical power generation.
!8.2.8 Commitment of Resources The 40 years of SAFSTOR will not involve commitment of a significant account of resources above what would be required for dismantling. It can be reasonably assumed that there will be less volume of radioact.ve waste to dispose of at the end of SAFSTOR than in 1985 due to the additional period of radioactive decay. Immediate dismantling m 1985 would have required offsite shipment of radioactive
. material and a larger burial area at a waste disposal site.
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8.3 ItadiologicalImpacts
{)T Section 4 covexs radiological conditions at Fermi 1. On July 23,1986 Edison submitted supplemental information in response to NRC questions. Relevant information has been included in this section; for the detailed data see Reference 11.
1he occupational radiation exposure of workers involved in maintenance and surveillance.
lhe environmental impacts of teleases ofliquid and gaseous effluents.
S e
The :rnpact of postulated accidents.
8.3.1 Personnri Dose illstory 1he table below represents personnel exposure history from 1973 through 1985.
Termi i Personnel Dose Illstory. Personrem YEAR DOSli YEAR DOSE 1973 5.79 1980 0
1974 5.05 1981 0.011 1975 9.70 1982 0.166 1976 0.025 1983 0.285 1977 0.041 1954 0.020 1978 0.043 1985 0.110 m
1979 0.081 (G
1he total cumulr.tive dose received at I:ermi i for the period 1973 through 1985 was 21.3 personrem.
Of the total cumulative dose,20.5 personrem was recorded during the three year period (19731975) in which fuel, blanket equipment, cold trap components, and equipment contaminated in the cut up, shipment, and clean up operations wcre remm ed from the site.
Approximately 0.2 persontem was ruorded durmg the intenm six year period of surve;llance and maintenance activities. (1976-1981)
Approumately 0.6 persontem was recorded during the last four year period (19821985) which includes the sodm-drummmg and shipment operations. Much of the dose received in 1985 (0.1 persontem) was frc" averatmn of a 'l1.D cahbratioi. facihty located :n the Fuel and Repair Building (FARB), a Fermi 2 activity.
The total for all personriel esposures attnbuted to Fermi 1 recorded from 1986 through 1996 were lower than 10 mrem per year.
8.3.2 OccupationalItadiation Esposure Espected Mairdenance, repair, and sun eillance operations over the period 1985 2025 is expected to average about 0 035 persontem per year, rr U
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Personnel dose resu ang from decontammation and decommissioning activities will depend on future V
factors not prer,ently available; i.e., availability oflow level burial sites, adoption of de minimus levels, development of new dismantlement and decontammation techniques and acceptance in burial sites of material that has been contaminated with sodium. All of these factors, if realized, along with normal decay wdl provide for lower personnel wposure in 2025 as opposed to immediate dismantlement in 1985. When decommissioning activities are performed during the SAFSTOR period, the exposure will be experienced earlier, however, the cumulative exposure will be less than if immediate dismantlement was performed.
8.4 Postulated Radiologleal Accidents There are three postulated radiological accidents that could occur dunng SAFSTOR. These are desenbed in the following sections. Additionally, a hypothetical secondary sodium scenario is addressed which potentially could lead to the liquid waste or sodium releases, though it is highly unlikely.
8.4.1 1.lquid Releases it is assumed that two liquid waste tanks in the Fuel and Repair Iluilding rupture. For the analysis, it is assumed that the tanks contain a total of 7550 gallons of radioactive radwaste containing 6 mC' of 60co and 6 mci of 137 s. This assumption is based on 1986 activity.
C Sffngio A Airborne Releng Assumptions:
(q Tanks rupture / malfunction and radioactive inventory is spilled on Door,
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25% of the inventory is assumed to be released through a vent to the environment.
e Release occurs over a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> penod and an individual is exposed for the entire time at the exclusion area boundary (EAll).
Dose factors from Regulatory Guide 1.109 and ICRP Publication 30 are used.
X/Q = 155 E 5 sec/m3 (Fermi 2 UFSAR. Chapter 15, Table 15A 2) 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 50th percentile value at Fermi 2 tixelusion Area lloundary (EAll) of 915 meters NW. This distance is conservative smee the Fermi 1 EAll is approximately 1211 meter NW.
Maximum Permissible Concentratmn (MPC) from 10 CFR 20, Appendix II, Table 11.
(Values used are based on reguiations in existence in 1986)
Results:
1 mmd Water Tank Source Airborne Release Concentratic.1 pCi ml
/
Bishds in Tank At fab MPC (air)
C/MPC*
00Co 2.10 E 4 3.23 E 12 1E8 3.23 E 4 137 s 2.10 E 4 3.23 E-12 2E9 1.62 E 3 C
'C/MPC - ratio of EAll concentration to MPC 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Dose at F All. mrem Adp11 Child Whole llody 3.27 E-4 1.11E-4 1.ung*
t 46 E 3 5.29 E 3
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'I.ung is most entical organ.
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Ss.otn.rio 11 I inuid Release to 1.ake Frie o
Assumptions:
Since the hquid radwaste tanks are located in the basement of the FARD, mir.or. racking of the structure could occur in the event of an earthquake. The tanks could under;o stress cracking and leaking to allow fluid flow be.wcen the interior of the structure and the surrounding carth, initially, liquid would be retained within the structure and diluted by inflowing ground water from the dolomite aquifer. There would also be a slow inflow of grour.d water and the water level inside the structure would rise until it reached the elevation of the piemmetric level of the ground water. At that time the radioactive liquid may be diluted by as much os 10:1; however, no credit is taken for dilution via the influx of water.
Tanks are approximately 450 fl. from the Lake Eric shoreline.
Flow rate within the aquifer is 0.24 ft day.
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Delay time in traveling from the tank to Lake Eric is 1875 days plus 40 days to move upward through till and !ake bottom sediments, (Fermi 1 UFSAR, Section 15.7.3.2).
Dilution factor of 77 at Monroe City Water intake 3200 meters south of Fermi 2 (Fermi 2 UFSAR, Appendix ll A).
Radioactivity decay with delay time is assumed.
Individual consumed water, fish, and invertebrates for 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.
e Dose factors from Regulatory Guide 1,109, MPC from 10 CFR 20, Appendix 11, Table 11.
e Results:
pd 1.iouid Waste Tank Not.rce Monroe city Water intake Concentration, pCi/ml Nuclide lolank 1 nterine 1 ake At intake MPC (water)
C/M PC*
boCo 2.10 E 4 1.04 E 4 1.35 E 6 5.00 E 5 0.03 137Cs 2.10 E 4 1.86 E-4 2.41 E-6 2.00 ti 5 0.12
- C/MPC n ratio of concentration at intake to MPC.
Incestion Dose. mrem Water
[nh invertebrate Total Adult Whole Body 0%
29.90 2.38 32.64 Adult, Liver 0.53 30.30 3.62 34.45 Child, Whole llody 0.19 4.24 0.!4 4.97 Child Done 1.10 29.90 3.68 34.68
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8.4.2 Airborne Releases from Sodium xd It is assumed that a fire or other catastrophic event results in the release to the environment of the 22 a and N
residual sodium including the entire radionuclide inventory which contains a total of 0.98 mci 137 s (per Section 4.1).
4.84 mci C
Assumptions:
100% ofinventary becomes airbome, e
Release occurs over a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period and individual is exposed for the entire time at the EAB.
e Dose factors from Regulatory Guide 1.109 and ICRP Publication 30.
X/Q = 1.55 E 5 sec/m3 htC from 10 CFR 20. Appendix B, Table 11.
Results:
Airbome Release Concentration pCi/ml
'pshds At EAll MPC(air)
C/hiPC' 2a 2.11 E 12 6E9 3.52 E 4 N
137 s 1.04 E 11 2E9 5.21 E 3 C
- C/hiPC a ratio of EAD concentration to h1PC
& Nit Dost mrem
,q Whole llody 1.31 E 3 LJ Liver
- 1.48 E 3
- Liver is most entical organ.
H4.3 Discuulon lioth Scenario A and B and the releases from the residual sodium result in concentration levels that are well below the h1PC values in 10 CFR 2' Appendis ll 'lable 11 for releases to unrestricted areas.
The doses associated with Scenario B are below the hmits at which precautionary measures would be taken for on accident type release. The doses associated with the fish and invertebrates are the result of the concentration faciors and the models in Iteputatory Guide 1.109. In this Scenano, the radioactive hquid is relmed to the aquifer and groundwater. The results are extremely conservative since no credit was assumed for:
- Dilution from the mittal in0us of water into the FARil.
Removal of suspended particulates by tillering action of the soil and clay, Removal of some fonus through adsorption by the soil and clay.
e The 1875 day travel time to the shorehne of 1 ake 1.ne provides ample time to sink wells, follow the progress to the lake, and take remedial action should it become necessary.
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8.4.4 Ilypothetici secondary Sodium Scenario
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There are three 12,000 gallon secondary sodium storage tanks located in three separate rooms in the Tenni i Steam Generator fluildii g llasement. The Steam Generator lluilding is located outside of the Fermi I protected Area. Due te the siphoning method used for draining the tanks, there is a maximum estimated 210 gallons of solid sos....m remaining in the three tanks (approximately 70 gallons per tank based on visual inspection). In order to prevent any hazard from fire or hydrogen formation, the three storage tanks were purged with 7 8 psig CO on January 9, l'774 in an attempt to passivate the residual 2
sodium deposits by con <crting the sodium to sodium carbona c. Ilased on visual inspection of the tanks dunng 1997, the residual sodium appears to be unreacted.
H.4,4.1 Sodium Water lleaction / II,sdrogen lleaction When sodium (Na) reacts with water (11 0). the exothermic process produces caustic sodium hydroxide 2
(NaOll), hydrogen (11 ), and heat according to the following reaction:
2 2Na + 2110 6 2NaOli + 112 + heat 2
1he presence of water given a tank breach is plausible since the water table elevation at Fermi is approximately $75 feet, and the top of the secondary sodium storage tanks is at approximately $70 feet.
There has been intermittent inleakage of rain water into the storage tank bays over the past 25 years, however, recent drain tile installation has prevented further substantial inleakage, 1his postulated scenario is highly unlikely due to the structural integrity of the tanks, layout of the tank p/
bays and equipment, and the mechamsm of the reaction. Sodium is a solid at room temperature and t
would therefore not drip out of a hole or crack in a tank. A more realistic scenario would involve water being miroduced into the tank through a postulated crack, with the resulting reaction generating hydrogen as a byproduct. The amount of hydrogen generated and contained in the tank would be limited by the hydrogen pressure buildup. The hydrogen which escapes the tank would dissipate in the larger areas of the buildmg. The worst case hypothetical scenario is evaluated here involving instantaneous reaction of all the sadium m a secondary sodium storage tank.
Assumptions:
Sufficient wat(r is present to completely react the 70 gallons of Na estimated to be in the tank.
I e
Total and complete Na Il O reaction O.e. maximum amount of hydrogen and caustic 2
generated).
The reaction O.e., hydrogen and causoe formation) occurs instantaneously, lhe gererated hydrogen concentrates u ithm the specific tank bay, e
The air to hydrogen ratio is conducive for complete combustion.
e The center of combustion is at the center of the respective Na storage tank.
All of the hydrogen combusts mstantaneously 0 c.. no partial bum during hydrogen gene:ation).
llydrogen combustion efficiency oOO"o e
i l
OL) l S7 Rev 0,12/97
h 8.4.4.2 Structuralllazard V
ne postulated hydrogen explosion results in the formation of a detonation wave which impacts the respective tank bay structure and equipment, along with creating a ground shock wave. The effect of the detonation and reflected waves will result in failure of some Steam Generator fluildmg structural members. Ilowever, since the building is outside of the Protected Area and does not contain radioactive sodium, buildmg failure itself does not result in radiological release. He Steam Generator fluilding rests on bedrock, with the outer walls made of reinforced coneretc. The north wall is 4' thick at the bottom and 3'2" thick at the Protected Area grade (590'). The east and west walls are 2'9" at the bottom and l*4" at the Steam Generator fluilding grade (582'). A radiological release might potentially occur if the ground shock were of sufficient amplitude and frequency to caase failure of equipment and ctructures outside of the Steam Generator fluilding which contain radioactive contam;aation, such as the primary systeni in the Reactor lluilding or primary sodium storage tanks, but this is highly unlikely.
Assumptions:
Detonation Wavn Any leakage pressure from the tarik bay *into the upper levels of the Steam Generator 11uilding e
will result in blowout of the transite siding prior to failure of the concrete walls.
Primary and secondary fragments will remain within the building, or if they penetrate the building structure will have lost most of their kinetic energy (i.e., will not penetrate outlying structures) 8.4.4.3 Caustic itelease llazard Ob The generation of hydiogen and Naoll, with the potential subsequent explosion of the hydrogen, could provide a means for sendmg the NaOli airborne. With the postulated Steam Generator lluilding structural failure or transite siding blowout. a vent path of the coustic NaOli from the Steam Generator iluildmg to the atmosphere is available. The caustic plume was evaluated for the onsite (i.e., Rrmi 2 operations) and offsite impact.
Assumptions:
The NaOli is not diluted by any water m the room.
The release is at ground level.
1.osses of Naoll from the plume due to settimg and reactions are neglected.
The puff travels directly towards the Fermi 2 south entrol room air intake, though this is not the prevathng wind direction.
The puff does diffuse.
The NaOli chmbs durmg dispersion to reach the contral room intake, even though NaOli is heavier than air.
The highest concentration at the control room mtake is when the cloud has trave'ed the distance to the mtake.
Smce the release is mstantanco is, the puiT is only bne0y at the intake.
e it is assumed that the control roon, normal sentilation remains in service, and that the control toom is not manua'ly isolated based on an explosion at Fermi 1.
NaOli is deposited on buildmgs m the pathw ay between Fermi 1 and the Fermi 2 control room, decreasmg the amount of Naoll m the plume.
l S-S Rev 0.12/9 /
4 b
8.4.4.4 Results t
i v
Fadiolocical liarard Conseouences
'lhere is no radioactive sodium associated with the secondary sodium storage tank residuals; therefore, there will be no radiclogical releases resulting from the initiating explosion. Any subsequent failure of the Steam Generator 13uildmg or the production of fragments due to the explosion is not expected to impact systems and'or equipment which may contain radioactive contamination. If they do, the effects would be bounded by the following analysis.
The effect of the detonatic,n wave induced ground motion was not quantitatively evaluated due to the expected high values for the response spectra and due to the lack of available information on buildmg and system response. 'Ihe assumption was made, though considered unlikely, that thc grot.nd shock could lead to damage of the primary system, primary sodium storage tanks, other primary sodim containing tanks or piping, and/or ligeid waste tanks. The possible consequences were evaluated qualitatively and were considered to be bounded by the following existing analyses for Fermi 1 and 2.
The concem at Fermi 1 would be with the priniary system or primary Na containing tanks or piping or liquid waste tanks being ruptured by the blast or ground shock, leading to the release of radioactively contaminated Na anNor waste water. Section b.4 already addresses the radiological effects of a total release of the primary Na inventory and ofliquid waste tankt rupture.
The concern at Fermi 2 is with the safety integrity of the plant being jeopardized by the blast or ground shock. A previous analysis summarized in the Fermi 2 Updated Final Safety Analysis Report which p
evaluated 'he results of an uplosion of the 20,00J gallon liquid hydrogen tanks located near the natural
(
draft coolmg towers is considered to bound the effects due to the smaller hydrogen explosion postulated in the Fermi 1 Steam Generator fluilding and similar distance to safety related equipment. The same unctusion applies if all thra secondary sodium storage tanks are affected by the event.
Chemical lla7ard Conscouences A previously prepared Umversity of hiichigan study (iteference 17) concluded that if 307 gallons of sodium were totally combusted and dispersed, the concentration at the site boundary (1000 meters was used for this study) for the few second: that the vapor cloud passed would just be at the respiratory protection limit of five times the Threshold Limit Value (0.002 g/m3), or at 0.01 g/m3. This amount bounds the 70 gallons which is postulated to react m this event. The conclusion reached from an evaluation of the consequences of a reaction of 70 gallons of sodium contained in a secondary sodium storage tank is that the maumum NaOli concentration m the Fermi 2 control room due to the reaction of 70 gallons of Na and the subsequent release of the p,enerated NaOli is less than 0.002 g/m3 If the total sodium inventory of all three secondary sodium storage tanks at Fermi I (~ 210 gallons) were to react and form airbome NaOll, the concentration m the Fermi 2 control room would increase by at most a factor of three compared to the postulated one tank event. During this highly unlikely scenario, the NaOli concentratmn could be above the NaOli limit of 0.002 g/m3, but less than the 0.010 g/m3 that the University of Michigan study desenbed as the respiratory protection needed concentration, if credit were taken for NaOli reactions with the air and settling, the control room concentration of NaOli followmg even a three tank scenano w ould hkely be within the Threshold 1.i nit Value.
V 1he NaOli release could affect people at i ermi i or elsewhere onsite, but this is an industrial safety concem vs. a nuclear salety impact.
09 Rev 0,12 97
I 4
h a
J Conclusion l
De worst case radiological consequences of the postulated hypothetical secondary sodium scenario are bounded by the liquid waste tank release and sodium release events analyzed in Section F.4.1 S.4.3.:
The postulated secondary sodium scenario does not casse damage at Fermi 2 that could lead to greater
}
radiological consequences.
- 8.3 :
Alters
.s Conaldered Once a nuclear nacility has reached the end ofi's useful life, it must be placed in a condition such that there is no unreasonable risk form the decommissioned facility to' the health and safety of the public.
Several attematives are available: DECON, ENTOM9, and SAFSTOR. De no action attemative is not viable for Fe mi-1 since it is already in a decommissioned state The three alternatives are discussed
- below, f
8.5.1 DECON i
DECON is defined as immediiitely removing all radioactive materials to levels which are considered acceptable to permit the property to be released for unrestricted use. DECON is the only one of the decomm'issioning attematives which leads to termination of the facility license and release of the facility and site for unrestricted use shortly aller cessation of facility operations. DECON would involve the temoval'or decontamination of all equipment, structures, and those portions of the facility containing radioactivity. Although the fuel hrs be removed from the Fermi I site, the reactor vessel, its internals t
n and most of the sodmm pipmg remam.
V.
Clearing the Fenni i site for unrestricted ute is of!ittle environmental val'te since Fermi i lies within the site boundary of Fenni 2 and could not be used for other purposes. A major effort would be involved in the complete tcmoval of the reactor vessel and its internals and the sodium piping. Because of size and induced radioactivity, this would have required the removal and cutup into sections of some of the various piping and equipment und shipment in commercially available licensed shipping casks to an offsite licensed burial site. This was considered undesirable due to personnel exposure to additional radioactivity.
This ahemative is not considered preferable since little or no improvement would be realized in personnel exposure, land use, aesthetics. or value.
i8.5.2.
ENTOMH ENTOMil means to encase and maintain property in a strong and structurally long lived material (e.g.,
concrete) to assure retention until radioactivity decays to a level acceptable for releasing the facility for unrestricted use. ENTOMils is mtended for use where the residual radioactivity will decay to levels
- permitting unrestricted rdcase of the facihty within a reasonable time period of continued structural-integri'y of the entombing structure
- approximately 100_ years is considered to be consistent with recommended epa policy on institutional centrol reliance for radioactivity containment.
i
, Primary considerations for retiring the~ reactor and primary system were: (1) removal of all core and-
~
, W
- blanket thel, (2) removal of sodium?(3) pastight seal of the primary system. and (4) passivation and
' i j
. maintenance of the entire primary sodmm system with carbon dioxide! The reactor vessel was sealed s
~ ~'
Lwithin the primary shield ta..k and the outlying components wcre sealed directly,.using the Reactor =
Building as an isolation structiire against personnel access to the primary system.
?
S 10?
' Rev 0,12/97 s
+ _,
e u
+w.
r E,,,
m, _.
__._iym-m.a.,....-_...h._
.__~,_4e m.m m _,. _ _, - -
m-.,_
?
l AU The primary system was Elled with nitrogen to which CO2 was added to reduce the residual sodium deposits to inactive solids. The sys'em was then sealed and maintained et slightly positive inert gas pressure to prevent the entrance of water or moisture and to minimize dispersal of any remaining tadioactive material.
To ENTOMB the Fermi i facihty at the present time would not result in any enhancements over the present decommissioned status.
There would be increased personnel exposure due to removal of radioactive equipment to accomplish the task.
The nickel-63 and niobium 94 in the reactor vessel would not decay to levels permitting the release of the facilities for unrestricted use within the guidelines of 100 years.
Limited surveillance activities would have to be maintained.
8.5.3 SAFSTOlt SAFSTOR is dermed as those activities required to place and maintain a radioactive facility in such condition that the risk to safety is within acceptable bounds and that the facility can be safely stored and sub>equently decontammated to levels which permit release of the facility for unrestricted use.
SAFSTOR consists of a short penod of preparation for safe storage, a variable safe storage period of contmuing care consisting of security, surveillance, and maintenance and a short period of final decontamination (DECON).
Fenni 1 was decommissioned accordmg the NRC (AEC) rules and directives in effect at the time and was considered as bemg left in a decommissioned state. In accordance with present dennitiors,it is in a SAFSTOR condition and a Prehmmary Fmal Decommissioning Plan 6r Fermi 1 was prepared and submitted to the NRC in Reference 16.
The puipose of the Preliminary Fmal Decommissionmg Plan for Fermi 1 was to state Detroit Edison's mtentions regarding the ultimate decommissioning cf Fermi 1.
Detroit Edison is planning on a SAFSIOR period to be followed by a Onal decommissionmg to achieve a radiologically releasable site and tennination of the Nuclear Regulatory Commission (NRC) license. It is expected that Onal decommissioning for the Fenm
- plant will be perfonned concurrently with that for Fenni 2.
The Prehmmary Fmal Decommissmmne Plan is stnctly an overuew of the fmal decommissioning effort.
Pnor to commencement of the Gnal decommissmnmg effort, an updated Post Shutdown Decommissioning Actinties Report (or equivalent) will be prepared and submitted to the NRC in accordance with 10 CFR 50 82.
H.6 Conclusions SAI STOR was determined to be the most nable decommissioning alternative for Fermi 1. DECON would result in litJe improvement over 5AFS' LOR, and ENTOMB is not a viable choice because of the presence oflong lived radioisotopes.
O i
h 11 Rev 0,12/97
l Retaining Fermi 1 in a SAFSTOR status for a 40 year period (1985 2025) will result in the following:
1(eduction in dose rate of more than 90%.
Reduction in personnel exposure at the time of final action.
Reduction in volume of radioactive wastes at time of fmal action, potential increased availability of repository sites for radioactive materials.
e Continued compatibility with the long term use of the Fermi 2 site since Fermi i buildings are being user' for Fermi 2 activities.
Nominal expense and impact on the community because use of Fermi 2 personnel for Fenni i surveillance activities provides readily available manpower resources.
Integration of Fermi 1 into the Fermi 2 decommissiomng progiam.
Continued minimization of the risk to the health and safety of the public.
O O
b 12 Rev 0,12/97
2
~
g g
o-TABl.E 8.1 ' 1990 DISTRIBUTION OF ESTIMATED POPULATION Distance from Site (miles)*
4 Ihrection m
- from ste 01-I'-2
' 2-3 '
34 4-5 5-10 10-20 20-30 30-40
' 40-50 0-50 '
N 29
-263 177 79 197 14.912 105.823' 530,643 686.014 312,537 1,650,674-NNii
.O I02 12 90 81 7.618 100,866 615,106
'l.026.606 363.659 2.113.866
. NIi 0
259 131.
12 0
0
~
11.180 60,183 18.877 L549 41,191.
1:Nii
'o-u.
0 0
0 0
6.960 16.547 14.243
-16.899
-54,690 i-Is 0
0 0
0 0
0 610 7.056
'17,294 3.207 29,067
[liSI.
- D:
0 0
0, 0
0 0
0-2,349 0
2.849 Sli 0-o 0
0 0
0 0
401.
6.713 47.673 54,787
$I SSli
.0' o
.0-0 0
0 0
1,052 16.653
~ 21.920 39,625 S
-41
'576 51 0
0 0
0 6.568 15.655 35,130 58.021 l:
SSW G
710 21 0
0 0
3.004 107,943 22,580 38,523 172.781 SW 0
208 9
0 117 936 11,008 319,037 73,578
' 23,552.
433.445 WSW
'O-24 846 2,236 1,779 34,474 6,715 9,531 10.0 M 9.468' 160,923 W
0 58 29 165 600 4,49I 5,640 I1.222
.27,702 29,887 79,794
[-
WNW 0
18 -
31
,52 109 3,8 %
6.195 17,271
'11,078 12.4% -
51,056-l; NW 3
76 353 639 313 4.942 7,398 98,185 116,185 37,802 265.901 I
NNW O'
140 243 64 77 2.621 19.545-120.357 77.607 69.070 289.724
- . :c.
-Q-TOTAL 73 -
2,434 1.903 3,337 3,278 73.800 284,944 1,957,514 2.138,739 1.022,372 5,498,394 i; P i
Source-ER-OL te l8
- To convert miles to lilometers, multiply by 1.6093.
g i
~
__OL O
O~ ' '
TABLEE2 1980 DISTRIBUTION OF PROJECTED POPULATION IN YEAR 2009 1
Distance from Site (miles)*
1 l
threction j
from j
site 0-1' l-2 2-3 34 4-5 5 10-20 20-30 30-40 40 0-50 i
N 37 353 225 100 251 12367 80,617 '
404,245 686.014 384.721 1,595,478 NNII' 0
.I30 15 114 103 6.016 77,351 554,465 712.580 443,530 2.029.419 Nii 0-
'329 167 15 0
0 13,198 12.021 943.019 GIS
' 48,661 g
1:Nii 0
0 0
0 0
0 8.215 19,534 22.2S3 19,949 64.518 li o
o 0
0 0
0 720 8,330' 16.820 3.785 33.250 ISI:
0-0' O
G 0
0 0
0 20,415 0
3.363 31i 0
0 0-0 0
0 0
467.
3.356 52,282 60,553 SSE O
O' O
O O
O O
1.225 7,804 23.242 43,758 l.
S-
$2 733 65' O
O O
O 7,036 '
19,291 37,136 62,454 SSW 0
903 271-'
O O
O 2,858 I12,942 17,432 55,081 204,097.
SW 0
265 11 0
149 1,190 14,001 310.191 32,285 28,302 430.800 WSW 0'
31 1,076 2,844 2,263 43,849 8,54G 12.122 76,691 11,975 94,525 W
-0
-74 37 210 763 5.712 7,173 13,934 11,825 35,952 97,255 WNW 0
23 39 66 139 4,841 7,879 21,130 33,400 15.071 67,61r NW 4
97-449 813 404 6,286 5,636 116,337 140,874 92,109 363,009 NNW 0
178 309 81 98-3.263 14.889 92.050 72.099 106.183 289.150
[
TOTAL 93 3.098 2,420 4.243 4,170 83,5N 241,077 1,685,705 2,143,612 1,314,966 5,482,908 x
4 Source: ER-OL
- To convert miles to kilometers, multiply by 1.6093.
Ls 9~
i i:
a _.,.. ~
-..--__..,--.----._.,-.~._,,._..--.-.-.-_..,l
- L)
SECTION 9:
REFERENCES 1.
Enrico Fermi Atomic Power Plant, Unit 1. " Administration Control and Surveillance Procedures Manual" 2.
"Enrico Fermi Atomic Power Plant", APDA 124 (Atomic Power Development Associates, Inc.), January 1959.
3.
" Retirement of the Enrico Fermi Atomic Power Plant", NP-20047, Power Reactor Development Company (PRDC), dated March 1974.
4.
" Retirement of the Enrico Fermi Atomic Power Plant, Supplement 1", NP 20047, Supplement 1, PRDC, dated October 1975.
5.
Enrico Fermi Atomic Power Plant," Technical Information and llazards Summary Report",
Vol.1 10, AEC Docket No. F-16, license No. DPE 9, Power Reactor Development Company.
6.
" Analysis of Fermi i Sodium", letter from R.R. Eberhardt to N.W. Ewbank; Attachment II, October 5,1983; page 3.
7.
" Decommissioned Enrico Fenni Unit 1 Reactor and Associated Building and Equipment, p
Administrative and Surveillance Procedures", Revision 1 July 1980.
v 8.
" Fermi 2 Updated Final Safety Analysis Report", Volume 1 NRC Docket No. 50."$41, License No. NPF 43, Detroit Edison.
9.
" Fermi i New Age for NuclG rower" American Nuclear Society,1979.
10.
Detroit Edison Letter, NE 85-0714, dated May 17,1985.
I 1.
Detroit Edison Let:er. VP-86-0092, dated July 23,1986.
12.
Detroit Edison Letter, VP 86-0118, dated September 15,1986.
13.
Detroit Edison Letter, NRC 87 0174, dated Septernber 25,1987.
14.
Detroit Edison Letter, NRC 88-0226, dated September 15,1988.
15.
Detroit Edison Letter NRC 88-0294, dated December 22,1988.
16.
Detroit Edison Letter, NRC 89-0100, dated May 17,1989.
17.
Bum, Reed Robert," Evaluation of Sodium Storage at the Enrico Fermi 1 Atomic Power Plant", University of Michigan. September 1982.
r3 Rev 0,12/97 9-1
to NRC-97-0115 Attaciunent 3 Safety Evaluation for liypothetical Secondary Sodium Scenario O
O
Safety Ev:.luation 97-006 FERAll 1 SAFETY EVALUATION O
V Safety Evaluation No*:
97-006 Change Request No:
Page No:J of _
A.
Description of Change: This safety evaluation is being written to address the existence of the residual sodium heels contained in the secondarv sc,dium storage tanks located in the bnement of the Fermi 1 Steam Generatorjuilding. Should the sodium ec. ne into contact with water.
hydrogen gas is generated as a bynroduct of the reaction which createe the patential for an explosion.
O Continuation II.
Safety Evaluation Fermi 1 evaluations are based on the desenption of the facility and Technical Speci0 cations at the time of receipt of1.icense Amendment No. 9, Aptil 28,1989, including any subsequent amendments or desi;;n modifications.
1.
Will the Proposed Change:
YES NO
[]
[X) 1.
Increase the probability or corsequences f an accident or malfunction of equipment important to safety previously evaluated for Fermi 1 SAFSTOR7
[]
[X) 2 Cn: ate an accident or malfunction of a different type than any previously evaluated for Fermi 1 SAFSTOR?
[]
[X]
3.
Reduce the margin of safety as defined in the Fermi 1 Technical Speci0 cations?
[]
[X) 4 Affect the environment as presently described for Fermi 1 SAFSTOR?
Document the basis of the responses on continuation sheet.
2.
Unreviewed Safety Question Answ er "YES" if any of the above boxes are checked "YES "
[ } YES
[X) NO,llowever desenption of scenario will be submitte i for NRC review h Date: / 7/// 7 C.
Prepared liy: Danny R. Swindle f S&L)
Lynne Goodman Approved liv: L=,M Weuos F m.tm i _.JL 0 4 Date: /2 - T-9 7 Review Conimittee Chairman:
0D%AN Date: 1 1
- P )
t Technical Review: Explosion llazard'-see attached letid. Attachment 1 - NaOli llazard-4 N 6 M
%Ih6 V Lyfon i2/)!q p
- Request from Fermi i Project fermi 1 Manual, Appendix IL Safety Evaluation, For,n 2, Pill, Rev 14,081897 Page1
Safety Ev:luation G.006 l
B C SAFETY EVALUATION 11.1 DISCUSSION
%e secondary sodium y rage tanks have maintained a sodium heel since the tanks were originally drained epproximately 25 years ago. An accident created by a reaction of this sodium was not specifically addressed in the design basis documents, which is why this safety evaluation is being performed, nis safety evaluation is to postulate a worst case accident due to the existence of the tank sodium heels, and to determine if the consequences are greater than accidents previously analyzed.
1.0 INTRODUCTION AND BACKGROUND
Decommissioning of Fermi 1 was initiated on November 27,1972. At that time, there was about 34,600 gallons of non radicactive sodium stored in three systems: the secondary sy. tem, the secondary service system, and the three secondary storage tanks. Due to safety coni,iderations and post retirement economics, the decision was made to dispose of the secondary sodium. His was accomplished by draining the secondary sodium system into the three storage tanks (performed earlier in 1972 for steam generator maintenance). He sodium was then barreled and shipped to Fike Chemical Company in Nitro, West Virginia to be processed into sodium methylate (Reference 3.3).
To assure a complete drain of the secondary sodium, the storage tanks were evacuated and the nitrogen cover gas system pressurized. Rese actions resulted in approximately 210 gallons of solidified sodium remaining in the tanks as residual heels. In order to prevent any hazard from fire or f
hydrogen formation, the thrt e storage tanks were purged with 7 8 psig CO on January 9,1974 to 2
\\
passiva:e the residual sodium deposits by conversion to sodium carixmate (Reference 3.3). Recent walkdowns indicate that the storage tanks appear to have been sealed after the CO purge.
2 There has been inleakage of water into the sodium storage tank rooms during heavy rain events over the past 25 years since the imtial CO passivation, nis has esulted in degradation of he asbestos 2
t insulation covering the storage tanks, especially the insulation on tank # 1. Pressure Relief Valve (PRV) 11021 was removed on Apnl 16,1997 in order to visually inspect the sodium heel in tank #
1, which appeared m(talhc and chemically unreacted with the CO. The inside surface of the piping 2
at the relief valve connection was covered in a thin layer of white powder which lab analysis identified as sodium carbonate and sodium bicarbonate.1.ater visual inspections of tanks # 2 and # 3 gave the same results. Walkdown of the tanks on 09/02/97 following removal of the asbestos msulation showed only surface rust on the tanks. Analysis of the cover gas in tank # 1 resulted in the following composition:
- O Page 2
~.
Safety Evaluation 97-006
~
i Component Cover Gas Sample Concentration Air Concentration
% hy volume
% by volume (ref. 3.7) liydrogen 0.04 trace Oxygen 13 21 Nitrogen 78 78 Carbon Dsoxide 10
< 0.1 i
farbon Monoxide
<0.05 None Argon 0.2 1
Moisture ambient (40 50"F 1)ew Point) ambient Table 1.1 Note the elevated CO, the depletion of 0, and the 11 level well below the lower explosive level 2
2 2
(LEL) of 4% liydrogen by volume with air (Reference 3.7).
- 2.0 AnhC11mtat of RhkAkkociated With Fermi 1 Secondary Sodlum Storane Tsak(s) Residual Sodluin Het]
This Safety Evaluation is being written to assess the potential risk associated with unreacted sodium heels in the secondary sodium storage tanks on site at Fermi 1. These three (3) tanks are located in the Steam Generator (SG)Iluilding basement, which is outside of the Radiologically Restricted Area (RRA).
O 2.1 Backgr.nund. AcddenLinalynis When sodium (Na) reacts with water (110), the exothermic process produces caustic sodium 2
hydroxide (NaOll), hydrogen (11 ), and heat according to the following reaction:
2 2Na + 2110 => 2Naoli + 11 + heat (equation 1) 2 2
In this reaction, the hydrogen gas that evoh es inside the tank will tend to blanket the unreacted udium, thereby reducing the cate of the reactin. De heat given off by this highly exothermic reaction may combust the hydrogen as it forms and may not allow enough 11 to accumulate to 2
perpetuate a massive explosion. The calculativn m reference 3.1 (Appendix 1, p.1.1) showed that for a complete sodium water (Na Il 0) teaction,24,652 pounds of11 would be generated for the 70,000 2
2 gallons of primary Na that was stored in the Fermi i RRA in 1982.
ne accident scenario presented m Reference 3.1 tonservatively postulates that a total Na il O 2
reaction occurs instantaneously, releasing the entire inventory of Na as 11 and NaOli concentrated 2
^
vapor. The entire volume of11 is conservatively assumed to subsequently explode. flowever, this 2
assumption is not entirely possible as the 11 must first mix with atmospheric 0 in order to combust.
2 2
De consequences conside;ed in Reference 3.1 are the impact of the 11 explosion, the release of 2
NaOli vapor at the nearest site boundary (3,280 feet) for Fermi 1, and the release of radioactivity due to primary No. The radiological hazards portion of the analysis is not applicable to the secondary Na system since the Na is not radioactively contaminated Page 3
S fety Evaluation 97 006 (l
'lhe results of the hazaid analysis in Reference 3.1 are as follows:
\\ J Fxplosion liarard: If the entire 24,652 pounds of hydrogen produced were instantly exploded, e
the raosi severe damage at the site boundary would be window breakage. No human physiological harm would result at the site boundary or beyond.
Chemical llazard: Total combustion and vaporization of approximately 307 gallons of Na is required to produce a NaOli vapor concentration of 0.01 gm/m' at the site boundary (respiratory protection is recommended for concentrations higher than 0.01 gm/m').
2.2 hnentRay AtchitatStenario At the present time, the three secondary sodium storage tenks are estimated to contain a total of about 210 gallons of solid sodium (approximately 70 gallons per tank) that was assumed passivated with CO at the time the system was drained. There has been intermittent inleakage of water into the 2
storage tank bays over the past 25 years. Recent drain tile installation has reduced this leakage.
Visual inspection of the three tanks indicated the presence of an t nreacted sodium heel, and an air quality monitor indicated the lack of explosive gases overlying the heel in the # 1 tank (only tank in which the atmosphere was sampled).
llyvn'Attical Accidcal The worst case postulated event would be an undeterr..ined breach of all three secondary sodium storage tanks with the presence of sufficient water in the tank bays to react completely with the - 210 gallons of Na. '" The presence of water given a tank bicach is plausible since the water table O) elevation at Fermi is approximately 575 feet (Reference 3.5), and the top of the secondary sodium
(
storage tanks is at approximately 570 feet (Reference 3.6). Assurning a complete and instantaneous Na Il O reaction "'is conservative for estimatmg available hydrogen which can be detonated. Usmg 2
the stoichiometric equation for this reaction (equation 1 of this SE), and the TNT equivalence (keference 3.1 equation 1.1), the hydrogen gas produced per tank is approximately 24.7 pounds, or equivalent to about 190.7 pounds 0' of trimtrotoluene (TNT). The NaOli resultbg from the reaction is about 986 pounds. Table 2.1 summarizes the reactants and products for this Na.110 reaction.
2 Reactant / Product Gallons Pounds Molecular Weight (gm/ mole)
Sodium (Na) 70 0 567.0 23 Water (11:0) 53.2 444.0 18 Sodium Ilydroxide (NaOll) 986.3 40 Ilydrogen (11;)
24.7 2
Tahic 2,1 Notes:
(1) Although the worst easc scenano would be for the Na from all three tanks to react and su'oseguently explode simultaneously. this scenario is considered to be extremely improbable. It is plausible to postulate that the reaction / explosion of one tank will result in the reaction /
explosion of the other tanks. Reference 3.12 provides guidance for hazardous chemical releases, g) of whieb sectmn C.5.a states that multiple container failures need not be postulated for a page 4 i
Safety Evaluatmn 97 0C3 l
l l
C maxim > m concentration accident unless they are connected. The three tanks are essentially isolated from each other, (2) It is highly unlikely that this event will occur since for the entire volume to react instantaneously, l
the surface area available for reaction would theoretically have to be comprised of the entire mass of sodium. In addition, the hydrogen must mix or dilute with atmospheric oxygen to i
combust, and the assumption is made that none of the hydrogen leaks out of the tank bay or I
prematurely ignites.
liased on Table 2.1, only $3.2 gallons of water are required to completely react the 70 gallons of Ns. Over the past, there was easily more than $3.2 gallons in the # 1 tank bay floor at some given times. Due to the layout of the tank bays and equipment, the structural integnty of the tanks, and the mechanism of the Na ll 0 reaction process, it is highly unlikely that Na would completely 2
react to fonn hydrogen which would then collect within the tank and be detonated by the heat of reaction. A more realis'.ic scenario would be where water is introduced into the tanks via a crack or pitting holes in the tank wall with hydrogen formed and contained within the confines of the l
tank. Note that sodium is a solid at room temperature, therefore, if a hole or crack were to fona in the tank bottom, the sodium would not pour or drip ou' of the tank. If the breach is small (e.g.,
a pinhole) the reaction will be selflimiting in water (and moist atmosphere) due to the very small sodium surface area available for reaction. Any hydrogen gas which escapes the tank would rise to the ceiling of the tank bay (bottom of the operating floor), and exit the tank bay via the vent paths in the operating floor (grated opemngs). This hydrogen would then be dispers:d within the much larger area of the upper SG 11uilding, thus eliminating any concerns with this free portion f
of the hydrogen exploding (i.e., additional vent paths out of tne SG Iluilding, nydrogen
\\
concentration being below the minimum explosion limit, etc.). llowever, assuming that the reaction n.te is instantrneous without loss ofil from the tank bay is conservative.
2 (3) Reference 3.4 section C presents recommendations in the design of nuclear power plants to withstand the possible effects of explosions. In particular, it states that for " calculating TNT equivalents, assumptions of 100% TNT (mass) equivalence for solid energetic materials and 240
% TNT (mass) equivalence for substances subject to vapor phase explosions are acceptable upper bounds". This is not applicable to the mass of sodium since the sodium or sodium vapor is not expkiding, but the hydrogen generated as a result of the postulated sodium water reaction is the explosive material. The TNT equivalence for hydrogen, as given in Reference 3.1, was calculated by multiplymg the mass of hydrogen (24.7 lbs) by the ratio of the beat of combustion for hydrogen ($ 1,$71 litu1b) to TNT (2.000 litu1b). and factormg in a hydrogen combustion efficiency (30%). Ther efore, the equivalent TNT value of 19i).7 lbs from Reference 3.1 will be used as the basis for this SE. Ilowever, using an equivalent charge weight of(567 lbs)(240%) or 1,360.8 lbs in this Safety Evaluation would result m the same final conclusion as using 190.7 lbs.
Powlated Scenario l'or the explosion analysis m this Safety Evaluation, it uill be postulated that all of the Na in one tank reacts completely and mstar taneously, and that all of the generated hydrogen is available for explosion. The impact of the possible reaction / explosion on the remaining two tanks will also be addressed.
O v
Page 5
Safet;* Evaluation 97 006 I
),) l'm1 La= timmard 2.3.1 Denttipties of the StCondary Sadl=nn Tanka mmd Tank Hats
%ere are three,12,000 gallon secondary sodium tanks which are located in three separate rooms (tank
- ~
bays) in the Fermi i Sicam Generator Huilding Basement (see Figure 1). These tanks are horizontally mounted on saddle supports, and have ellipsoid heads. %e tanks have an outside diameter of 10'0", n wall thickness of / ", and a 22' 4'/ " overall kngth (Reference 3.6.b). The centerline of the tanks is i
3 4
approximately 6'6" from the basement Door (Reference 3.6.c).
l He following information was obtained from Reference 3.6.d (See Figure 1). The elevation of the top of l
the fmished Door in the steam generator tank bays (50 Huilding basement)is at $$8*6". The basement Door is 2'6" thick and sets on bedrock Rese rooms aiso have a '/s" steel liner on top of the finished Goor. The operating floor is l'0" thick with a 2" finished grout layer (top of operating Door at elevation
$90'2"). He Steam Generator Huilding basemer i door (elevation $$8'6"), operating Door (elevation
$90'2"), and the tank bay outside walls, are cone, rete reinforced with steel bar, ne centerline of the reactor is approximately 45 feet from the north wall, with the Reactor Building grade at elevation $90'0".
%e bottom section of the north wall (up to elevation $74') is 4'0" thick with the top section (up to elevation 590'0") 3'2" thick. %e north wall from elevation 590'0" to 629'l'/ " is 2'6" thick reinforced 2
concrete. The remaining north wall up to the roofis corrugated transite siding. The bottom section of the cast wall (up to elevation $73'6") is 2'9" thick with the top section (up to elevation 590'0") l'4" thick.
7 He bottom section of the west wall (up to elevation $77'2") is 2'9" thick, the middle section (up to elevation 582'2") is l'4" thick, with the top section (up to elevation 590 0") l'0" thick The top sections of the basement east and west walls are above the ground level on those sides (lower grade elevation of
-p) t 582'2"). The inside walls separatmg the tank hays, and the tank bay south wall, are poured concrete for the Orst 3 feet from the Ocor and then l'0" thick brick to the bottom of the operating door. The brick wall served as a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Orc protection barrier. The personnel doorways to the tank bays are 2'6" wide by 1
7'0" high, with the bottom at elevation $61'6". There is a walkway between the bays at an elevation equal to that of the top of the tank,'with a personnel doorway between bays. The entrance to the tank bays from the basement level are via personnel doorways. The door to the # 2 & # 3 tank bays is located L
on the east side of the tank bay south wall, and the door to the # 1 tank bay is located on the west side of the tank bay south wall. The steam generator opening in the operating Door is approximately 10'4"in diameter, with the centerline of the opening offset from the centerline of the Na storage tanks.
2.3.2 Methanitter the rsplullalakhin_thclatbedlitutlut_t The loading from an explosion within a vented or untented structure consists of two almost distinct
- phases, Octonation wave He Orst phase is that of a shock wave which is sustained by the energy of the chemical reaction (i e., detonation) m the highly compressed explosive medium in the wave. It consists of the mitial high pressure shock wave, w hich is of short duration, followed by several reDected pulses.
+
The imtial shock wave and associated re0ceted wave will have the largest amplitude because of.
I irreversible thennodyntmic processes, and due to mterference of the subsequent re0ceted waves inside the complex structure, Due to the complex con 0guration of the room and equipment, and location of the explosion relative to the structure, modeling of the reDected wave response would be extremely difficult with questionable accuracy m the results Detonation waves travel faster than sound.
Page 6 I
Safety Evaluation 97,006 A
/V)
Combustion was The second phase is that of a pressure wave which propagates by the process of heat transfer and diffusion, forming a structural loadmg. As the blast waves re' lect and lose energ/ to the room atmosphere, and as heat is released from the explosion, the atmospheric temperature of the room increases until a quasiotatic pressure is reached. This hydrostatic pressure rise may a:so be augmented by fonnation of additional gasses ia the explosion. The duration of this pressure depends on the energy loss rate to the sanounding structure, the sire of available vent paths from the room, and the cooling effects ofinternal expansion as gases are expelled. Note that due to the magnitt de cf the initial pressure shock wave from the explosion and th*. additional quasi static pressure within the initial second, the contribution of any venting from the room aas little effect in reducing the peak pressure loading on the structural enclosure of the tank room (bay).1he maximum pressure reached depends on; 1) weight of the explosive,2) type of explosive (chemical composition),3) volume of the room, and 4) ambient pressure and temperature of the air mitially in the room. He combustion waves are slow compared to the velocity of sound, therefore, the tima of occunence of peak loading for the detonaton and combustion waves differ. The shock were pulse causes a sigmfuntly greater loading than does the quasi. static pressure.
Note that factors such as residual water m the ro'om, czistence of piping, etc. which would tend to absorb some of the blatt energy and thus reduce the effects of the above scent,rio, will be conservatively ignored.
2.3.3 Calculatlutt.ulthe Prenure Londlus The first step in qualitatively analyzing the effect of the blast on surrounding structures is to estimate the inedent overpressure. The incident overpressure versus time curve is required for quantitative analyses, N(b however, it is considered acceptable io estimate the nagmtude of the overpressure for :his qualitative teview. As stated before, due to the irregular geom:try c.f the room and the presence of piping and equ pment, the prenure wave reactmn is dif0 cult to model and will be conservatively estimated.
2.3.3.1 lilastShocLhn s
I igure 2 57 from iteferenec 3.9 will be used to estimate the s a loads. Note that the actual distribution of the blast loads is highly inegetar because of the multiple reDections and time phasing and results in localued high shear stresses in the structure. The use of the average blao loads is considered acceptable since the scinforced concrete of the SG lluildmg is expected to transfer tnese localized loads to regions oflowo stress.
I' rom lleference 3.9 Figure 2o7, th( iollowmg parameters are required to estimate the magnitude of the renected pressure wavc:
Z = IMV' 3 (equation 2) 3 where:
Ils = Nonnat distance (perpendicular distance fiom the wall of concem), ft Charge weight,lb W
=
Scaled normal distance, ftilh' '
Zs
=
I: rom section 2.2 of this SE, the equivalent charge weight (W) calculated in llefeience 3.1 is 190.7 lbs.
p The closest stru,:ture assummg the blast is centered at the center of the tank is the tank bay Door at a V
page 7 i
Safety Evaluation 97 006 e
6 distance of 6'6". The closest structure assuming the blast is centered at the center of the is the i
north or south wall at a distance of 12'7", This gives a scaled normal distance (Z ) of 1,13 Mb i
4 2.19 Mb" respectively. Using figure 2 57 of Reference 3.9, the estimated peak pressure at a distance of i
6'6"is approximstely 6,800 psi, and approximately 1,700 psi at 12'7" from the blast. As the pressure wave emanates from the blast center, the blast pressure drops by a factor of approximately 4 for a p'
doubling of the distance These values are very conservative since a preliminary calculation using the' methodology of Reference 3.9 problem 2 14.2.1 gave values of 925 psi for 6'6" and 475 psi for 12'7".
1 The simplified pressure loadings and time durations are graphically represented below.
i Where:
- p o
P
= Pressure
/
E' I
P,
= Peak reflected pressure P,. m Peak gas pressure Blast pressure loading P.
= Ambient pressure
)/
T
= Time t,
' m Duration of P, P,
1, m Duration of P, _
N Gas pressure ionding
' t,
= Time of explosion f
i N
j t
t, t,
o 2.3.3.2 Gas.hrnure Loadine iOuaildlatitprnsur.c) l Ilased on the shock wave pressure calculated above, the block fire walls of the tank bays are expected to fail, thereby providing a large vent area from the tank bays. Should these wall (s) fail, the gas pressure
- buildup will be minimal. As stated above, the shock wave pulse causes a significantly greater loading than does the quasi static pressure. Therefore, since this evaluation is qualitative and not quantitative, the magnitude for gas pressure buildup will not be calculated.
1 rangibility A frangible element as defined m Reference 3.9 is an element that exhibits a resistance to internal shock loads equal to or less than 25 pounds per square foot and will undergo significant displacement during the loading time of the shock pressures and, thereby, reduce the effects of the shock pressures acting on
- both the frangible panel itself and re0cetions to other elements of the structure. The fire walls which are made of concrete bhicks may be considered frangible, hoiever, it is conservative at this point to assur.1e that they remain stationary followmg the blast, thus producing the maximum calculated reflected -
pressure wave. If the element fails, then the portion of the shock pressure impulse displacing the element as well as that portion of the impulse being redeeted to oth5r elements will be reduced because of additional venting area produced by the elementchecak up".
O!
Page 8 m
Safety Evaluation 97 006
('
Leakage Prcinurs d
Following the hydrogen explosion within the tank bay, the shock pressures escape to the surrounding areas along with the venting of the gas pressures. Trailing shock waves overrun and coalesce with the lead shock wave to form a single diverging shock wave. This is due to the speed of sound m the heated area behind the initial shock wave being greater than that in the ambient atmosphere. As the shock waves emanate frem the room via the vent paths, the structure itself affects the wave pattern forming highly turbulent vortices at the corners. his results in a shock wave with a higher pressure in the direction of venting, and a reduced pressure at the sides and rear of the wave front. Figure 2187 of Reference 3.9 provides a reasonable estimate of the peak positive pressure at the front of a partially vented four wall cubicle. This method assumes that there is only one vent area, however, the tank bays have several vent atcas (i.e., personnel doors, grated openings, side walls following their collapse, etc.). In addition to direction, these pressures are a function of scaled distance and the vent area divided by the volume to the two. thirds power (A / V").
Tbn vent area from the bays, except for bay # 1 which has the steam generator removed, is small in comparison to the room volume. The free volume is equal to the total volume of the tank bay minus the volume of all the interior equipment (e.g., piping, tanks, steam generators, etc.), structural elements (e.g.,
walkways, columns, etc.), etc. Although the tank bay volumes are approximately the same, except for bay # 1 which has the steam generator removed, the tank bay with the smallest volume is # 3. The actual free volume of this room will be less since the steam generator penetrates the operating floor with a portion of the SG occupying some of the tank bay space, in addition, prior to the asbestos abatement of the SG lluilding..'here were several cu% feet of space taktn up by the large insulated sodium piping
(
w hich is routed through the tank bays. Ilowever, using the total area for calculations is considered L
acceptable for the purposes of this evaluation since we are looking for order of magnitude numbers for quahtative analysis. The total volume of tank bay # 3, assuming all of the piping and equipment were removed,is approximately (refer to figures 1 & 21:
Y ' (30'0" - l'9" - 6")(53'5" 2'l 1" - 23'1")(589'0"- 558'6")
= (27'9")(27'5")(30'6")
= 23,205 ft' The main s ent area from the bay is throurh the two personnel doors. Thcre are also wall / ceiling penetrations which would provide vent paths (e.g., there is at least one 2'x2' grated opening ar d one 2'x5' grated opening in tige ceihng o,f each tank bay). In addition, the # 1 tank bay would have an additional nr* = (n)(5'2")' or 83.9 fr of sent area due to the steam generator removal. Using the vent area,for the two personnel doors of(2)(2'o")tT) or 35 ff.,one 2'x2' grate or 4 ft, and one 2'x5' grate or 2
10 ft*, the immmum vent area from each tank oay is 49 ft'. Usmg equation 2 3 from Reference 3.9, the ratio of the vent area to the 2/3 root of the free solume of the room is:
A / %" = (49 ft*) + (23,205 ft')2 ' = 0.0602 3(V Page 9 i
Safety Evaluation 97 006 4
ep Note that this value would be larger if the actual free volumes of the tank bays were used. From V
Iteference 3.9 Figure 2 187, at a A / V" value of 0.0602 (shown above), the pressure vented to the remainder of the SO 13uilding is:
Scaled Distance 3
4 5
6 7
6 9
10 20 30 (IVib")
Actual Distance (feet) 17.3 23.0 28.8 34.6 40.3 46.1 51.8 57.6 115.2 172.7 Pressure (psi) 70 40 25 18 13 10 8.5 7
1.9 0.9 lhese pressures are considered conservative since the pressure will be vented from the tank bays from separate openings; one personnel door into the south basement room, one personnel door into the adjacent tank bay, a portion through the Door gratings, and the remainder via the gaps in the wall and 41oor penetrations. This does not account for the assumed failure of the block fire walk, nor the pressure energy absorbed by the structure and components.1hc pressure decreases with increasing scaled distance and/or increasing room volume and/or decreasing vent area.
2.3.4 llupact of111 51 FrInnte The high pressure blast wave and the high speed of the ejected fragments are the primary hazards associated with the postulated hydrogen explosion.
I'rimmy / Secondary Fragments O
As defined in Reference 3.9, primary fragments are from the explosive casing, and secondary fragments are objects in the vicinity of the blast, either free standing or restrained. which may be displaced by the blast (e.g., piping, equipment, etc.). For this postulated accident, the soutum storage tank itselfis the explosion casing.1he radioactively contaminated equipment of concern (e.g., primary system, primary Na storage tanks, etc.)is located north of the SG lluildmg.1he reinforced concrete north wall of the SG Iluilding is considered adequate to stop or signincantly reduce the kinetic energy of any primary fragments.1herefore, any etfects associated with fragments generated from the blast are bounded by the blast and ground shock evaluation.
GmundShack The shock wave imparted to the ground around the SG lluilding is only a concern for those structures on the north side of the buildmg (i.e., Fermi i buildings which contain radioactive contamination are on the north side of the SG lluildmg). and with the safety of Fermi 2. Ground shock results from ait induced and direct mduced effects of the energy relea:ed by the exnlosion. The ait induced shock is due to the peak positive incident pressure (P ) from the blast compressing the ground, thus sending a stress pulse mto the g ound which attenuates as it pases through the media. The direct induced ground shock results from the explosive energy being transmitted directly through the ground. Rather than calaulating the ground shock response spectra, a quahtative approach will be used.
Ilased on the relatively high reDected blast pressures, and the fact that the SG Ilmlding is resting on bedrock, tia response spectra due to ground shock is expected to be high. To address the Fermi i response,it is assumed that the ground shock due to the blast is suf0cient to breach the F rmi i primary
' (b systein inside the 1(eactor fluildmg. or the primary Na storage tanks, or other primary Na containing Page 10
Safety Evaluation 97 006 O
tanks or piping, even though in reahty a bt 4ch is not expected based on all the conservative U
assumptions. The concem with breaching the primary system is due to the presence of radioactive contamination and residual Na, The postulated breach in the primary system provides a path for the radioactive contamination within the system to escape into the atmosphere (also assuming that there is a coincident breach of the primary containment vessel). Ilowever, this accident is bounded by the previously evaluated accident which was postulated to occur during SAFSTOR (References 3.10, section 2.1.3 and 3.11, section 3.2). This accident assumed that a lhe or other catastrophic event results in an airbome release to the environment of the residual Na including the entire radionuclide mventory. The second concern is witii the release and subsequent reaction of the residual Na with water, resulting in the fonn-
'nd subscquent release of caustic vapors. The second concern is not considered to be a realists o s Ario for the following reasons;
'lhe residual primary sodium is a solid at room temperature and is located in the low points of the system or trapped between crevices within components (i.e., pumps, valves, tanks or heat e xchanger intemals) or within the reactor. Therefore, any residual Na will not be easily displaced.
The Reactor Iluildin t is surrounded by a building annulus with a water intrusion alarm which is maintained as part of the Fenm 1 Technical Specitication surveillances.
Unlike the plume calculation perfonned for the caustic fumes resulting from the reaction of the secondary Na storage tank residuals with water, there is no explosion associated with the failure of the pnmary Na system which would cause the caustic vapors to become airborne. Therefore, there is no driving force for any caustic vapors generated tr leave the Reactor Iluilding lower level, or the n
Reactor fluilding containment, or other buildings.
U To also address the Fermi 2 response to the ground shock, the evaluation that was performed for the postulated explosion of the Fermi 2 hydrogen tank farm located by the natural draft cooling towers will be compared to the equivalent hydrogen from the postulated Fermi i Na.li 0 seaction and the distance 2
of the SG lluildmg from the I crmi 2 buildmgs. The Fermi i postulated accident consisted of 190.7 lbs of equivalent TNT explodmg in the SG lluilding which is approximately 800 feet from the closest comer of the Fermi 2 Turbine lluildmg and approumately 1,100 feet from the closest comer of the Auxihary lluildmg (Reference 3.6.e). From Reference 3.5 section 2.2.3 A "a 20,000 gallon liquid hydrogen storage tank is located at the llWC pas supply facihty. The gas supply facility is approximately 1100 feet northwest of the RilR Complex. Reference 3.5 states that the result of the postulated explosion has the potential to damage the roof and sidmg of select Fermi 2 buildings but does not impact the safety mieprity of the plant." liased on the similar distances from the Fermi 2 safety related buildings for the tw o postulated explosions, and the orders of magmtude difference in hydcogen weight being detonated,it is concluded that the results of the postuhted explosion (i c., blast fragments, ground shock, etc.) in the Fermi l SG lluildmg will not impact the safety mtegrity of Femii 2.
2.3.5 Imp actoLLilutamLG as. L'Insurnanlittut t ura LElements Eck I trcEalls Since the tank bay dividing walls are formed of concrete blocks starting at 3 feet above the Door, they will not withstand the initial load of the esplosion. It is expected that due to the layout of the supporting p
columns m relation to the blocks. that the block wall will fail grossly, with some damage to the steel V
Page11
Safety Evaluation 97-006 I
(V}
beam column 1he same argument applies to the initial load that will be felt by the south wall. Failure of the wall (s) will dramatically increase the vent area for the room, thus reducing the gas pressure in the tank bay to an insi.rnificant value. In addition, the additional vent path is expected to decrease the
!eakage pressure loading on the remainder of the SG Iluilding.
Duter Walls Since the bottom Door of the buildmg is 2'6" of reinforced concrete resting on bedrock, there is no concern as far as structural integrity for the initial loading.1he north wall at this elevation is 4'0" thick reinforced concrete with a few feet of ground between the north wall and the Reactor !!uilding annulus.
There is a possibility that the initial impulse load on the north wall may buckle and crack the wall; however, the majority of the fot.c will be absorbed by the wall, the ground, and the outer annulus wall.
Opentung Floor (Tank Bay Ceiling)
The integrity of the operating Door is questionaile since it was designed to withstand a downward load l
and not the upward load being applied by the blast. It is also questionable whether the operating Door will give, or if the steam generator will be ejected, or if the load will be shared by the two. Note that the
- 1 steam generator has been previously removed; however since 6e hole is offset from the assumed blast center, some Coor damage may still occur due to the initial load.
2.3.6 Ruuks n
'the analysis assumed that the 70 gallons of sodium in one tank reacts instantaneously and completely V
with water to form hydrogen. The hydrogen is trapped either within the tank or the tank bay and is then assumed to detonate. Detonation of the hydrogen within the sodium tank bay would produce a nonuniform, high intensity blast load on the structure and components within the tank bay. There is no radioactive sodium assnciated with the secordary sodium storage tank residuals, therefore, there will be no radiological releases resulting from the imtiating explosion.
'lhere is no concem with the structural mtegrity of the bottom Door of the building (tank bay Door) due to the blast since it is reinforced concrete restmg on bedrock. The north wall at the elevation of the blast is 4'0" tluck reinforced concrete, with a few feet of ground between the north wall and the Reactor lluildmg annulus. There is a possibihty that the initial impulse load on the north wall may buckle and crack the wall; however, the majority of the force will be absorbed by the wall, the pund, and the outer Reactor fluilding annulus wall. Since the tank bay dividing fire walls and tank bay south wall are formed of conciete blocks, portions of the walls impacted by the blast are expected.. fail grossly, with some damage to the steel beam column. Failure of the wall (s) will dramatically increase the vent area for the room, thus reducing the gas pressure in the tank bay to an msigmacant value. Since the operating Door was designed to withstand a downward load and not the upward load being applied by the blast, its structural integrity is questionable. Due to the size and connection points of the steam generators, and the estimated blast load, the steam generators are not expected to be ejected a great distance. It is expected that the load will be shared by the steam generator and the operating Door, however, failure and collapse of the operating Door and steam generator into the tank bay is not a concem.
Due to the reinforced concrete constructmn of the SG lluilding outer walls and the tank bay ceiling n
(operating Ooor), fragments from the explosion are considered to be either stopped by the structure or the V
Page 12
Safety 11 valuation 97 006 O]
kinetic energy of the fragments reduced to a small amount while penetrating the SG Building.1hese
/
fragments may impact other Fermi i buildings; however, the efTects are bounded by the ground shock evaluation. The conservatively calculated leakage pressure into the remainder of the SG lluilding is considered sufficient to blow out the transite siding on the upper portions of the building wads. Should the superstructure of the SG lluilding fail outwards, the result would also be bounded by this evaluation.
Due to the equipment contained within the SG Building (i.e., non-safety), and the contents of the systems (i,c., non-contaminated), failure of the SG Building itselfis not a concem. Should failure result to the remaming secondary sodium storage tanks subsequent to the explosion in the first tank (e.g., the blocks from the separatmg walls rupture the adjacent tank), the resulting accident will be no worce than that evaluated above.
The effect of the resultant ground motion caused by the blast on surrounding buildings and equipment was not quantitatively evaluated due to the expected high values for the response spectra. The possible consequences of the expected high values were evaluated qualitatively and were considered to be bounded by existing analyses for Fermi 1 and 2. The concem at Fermi l is with the primary system or primary Na storage tanks, or other primary Na c9ntaining tanks or piping being ruptured by the blast or ground shock, leading to the release of radioactively contaminated Na. A previous analysis evalu;.ted the effect of a total release of the primary Na inventory and found the results acceptable. The concem at fermi 2 is with the safety integrity of the plant being jeopardized by the blast or ground shock, leading to an unreviewed safety question. A previous analysis which evaluated the results of an explosion of the 20,000 gallon liquid hydrogen tanks located near the natural drafl cooling towers is considered to bound the effects due to the smaller hydrogen explosion postulated in the Fermi i SO lluilding, and similar distances from Fermi 2 safety related structures.
'the quantity of hydrogen available for explosion from the postulated scenario in this Safety Evaluation is small compared to that addressed in section 2.1 of this Safety Evaluation. Ilased on the order of magnitude difference, the hazards at the Fermi site boundary are minimal to nonexistent by comparison to the identified explosion hazard in section 2.1 of this Safety Evaluation. Ilowever,it should be noted that there may be industnal hazards associated with an explosion of one or more of the secondary sodium storage tanks at Fermi 1. These harards may include the chemical hazard for personnel in close proximity to said explosion and the hkelihood of serious mjury and'or fatality for persons in the Steam Generator fluilding at the time of explosion. Ilowever, there is no requirement for continuous staffing at Fenni 1, so this is an industrial safety concern vs. a plant safety concern.
2.4 Chetultaillarant Approximately 986 lbs, of Naoll (4 47 x 10' g) are produced by totally reacting 70 gallons of sodium with water. The calculation performed m Reference 3.1 gave the site boundary concentration of concentrated Naoll vapor produced from reacting 70.000 gallons of sodium with water, No credit was taken for any deposition on buildmg surfaces for the analysis in Reference 3.1.The calculated concentration was above the respiratory protectmn hmit. The evaluation also concluded that if 307 gallons of sodium wcre totally combusted and dispersed, the concentration at the site boundary for the few seconds that the vapor cloud passed would just be at the respiratary protection limit of five times the Threshold 1.imit Value (0.002 g / m') or 0.01 g i ml Smce in this postulated scenano only 70 gallons are assumed reacted vs. 307 gallons, the concentration at the site boundary would be less than previously calculated. Since this scenario occurs inside a
'v' Page 13
Safety Evaluation 97 006 I
4 (7
building, if credit was also taken for deposition on the Steam Generator 13uilding suifaces as taken in the
(..)
Attachment I analysis, the site boundary concentration would be below the threshold limit value. Since the site boundary calculation was previously performed in the University of Michigan study (Reference 3.1), it will not be repeated in this safety evaluation, and the study conclusions will be used.
The potential impact of the chemical release on the Fermi 2 control room was not previously evaluated. to this Safety Evaluation contains the evaluation of the potential impact of the postulated
+
Naoll aerosol on the Fermi 2 control room. Attachment i provides detailed information on the Naoll calculaHons and associated assumptions. 'the conclusion reached is that the maximum Naoli 3
conten stion in the Ferm. 2 control room is less than 0.002 g / m,
'!he Naoll release could affect people at Fermi l or elsewhere onsite, but this is an industrial safety concern vs. a nuclear safety impact.
A calculation of the radiation dose to the Control Room Operators was not performed because a comparative evaluation shows that the dose would be negligible. Per reference 3.11, the whole body dose at the Fermi ! exclusion area boundary (EAll) fr'om the release of all of the radioactive material in the sodium would be 1.31E 3 mrem. Since the Fermi 2 Control Room intake is about a third of the distance from the Feimi 1 So lluildmg as the EAll, the maximum dose at the control room intake from the release of all of the radioactive sodium at Fermi 1 would be on the order of 0.0: mrem.
nV
,A
(/
Page 14 l
Salety Evaluation 97-006
,Q 3.0 IEEERENCES U
3.1 Bum, Reed Robert," Evaluation of Sodium Storage at the Enrico Fermi 1 Atomic Power Plant".
September,1982.
3.2 Detroit Edison, Fermi i Manual. Revision 14 3.3 Power Reactor Development Company. " Retirement of the Enrico Fermi Atomic Power Plant".
hiarch,1974.
I 3.4 U.S. Nuclect Regulatory Commission. Regulatory Guide 1.91; " Evaluations of explosions postulated to occur on transportation routes near nuclear power plants". Revision 1, February 1978.
J 3.5 Fermi 2 UFSAR sections:
2.2.3.4 Onsite Storag of Fuels and Explosives 2.4.1 "I.2 59urces 3.3.2.3.2
' Reactor / Auxiliary Building Above the fif h Floor (Blow Away Sidind 3.3.2.3.7.3 Turbine Building 3.5.1.1.2 Design Evaluation 10.4.10.3 Safety Evaluation 3.6
- a. Fermi I drawings:
'l p
1, 6P721-1049 3 Rev. G d
2.
h1160
- 3. hil61 4.
P !01-G1630 series
- b. Fermi 2 drawings:
1.
6 hts 721-108, '09,-112 9
d 3.7 Ilandbook of Chemistry and Physics,73 Edition 3.8 Crane Technical Paper No. 410 3.9 Army Th15-1300 / Na.y NAVFAC P-397 / Air Force AFR 80-22,"Sv ctures to tesist the u
Effects of Accidental Explosions", November 1990
-3,10 The Office of Nuclear Reactor Regulation, Safety Evaluation Supporting Am:odment No. 9 to Possession-Only License No. DRP-9. Fermi Unit No.1 Docket No. 50-16, dated April 28,1989 3.l ?
DECO letter NRC-87-0174, dated Sep:mber 25,1987,"Supr emental Environmental information Enrico Fermt Atomic Pow,:r Plant. Umt 1" 3.12 Regulatory Guide 1.78, Dated.'une l74," Assumptions for Evaluating the Habitability of a iiuclear Power Plant Control Room During a Postulated Hazardous Chemical Release"
/^
Page 15
q Safety Evaluation 97-006 9
11.2 EVALUATION OFIDE PROPOSED CHANGE
- 1. Will the proposed change increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated for Fermi 1 SAFSTOR?
A. The residual secondary sodium has been in the tanks since the tanks were drained approximately 25 years ago and an accident created by a reaction of this sodium was not speci6cally addressed. He Retirement Repor' did mention there were about 50 gallons left in the secondary sodium storage E
tanks (and additional residues elsecere). The sodium residues were thought to be passi.vated, but do not appear to be. Sin e the sodium has been there for the past 25 years, the actual probability of an ac:ident is not increased, however, the secondary sodium was not previously addressed in the licensing basis as to whether it could cause an accidct.
What is being evaluated here is whether the probability of an accident has increased compared to what was previously discussed in the licensing basis accident analysis. The existence of the secondary sodium does not in italf cause an, accident, but rather, its existence means a secondary Na accident can occur, his safety evaluation addresses the worst case postulated accident that could t
occur, not that it is likely to occur. The previous safety analysis docketed already recognized that a Dre or other catastrophic event could occur that could result in release of all primary sodium remaining. The sc,urce of the fire or catastrophic event was not identified. It is very unlikely that an 3,
4' event involving the secondary sodiu n could lead to release of the primary sodium, but based on available information whether or not a primary system breach could c,ccur during the event could not
~
be calculated. Therefore, the assumption is made.n this safety evaluation that a worst case postulated secondary sodiura accident could cause the release of pnmary sodium and so be one of the unidentified causes of that already analyzed event. Since it wouk' ake a catastropluc type event to cause this postulated secondarv sodium scenario, there is no discernible in-rease in probability of a pnmarv sodium release when comparing tne probability of a pnmary sodium release event taking this secondarv sodium scenario into account than when it wasn't considered. Additionally, the postulated secondary sodium scenano could result in the analyzed liquid waste tank release event occurring. Again, since it wculd take a catas'rophi: type event to esuse the postulated secondary sodium scenario, there is no discemible increase m the probability of the liquid waste accident compared to the existing licensmg basis.
Storage of residual sodium heels in the itcmdary sodium storege tanks does not increase the onsequences of a previously analyzed accident or malfunction of equipment important to safety because the storage has no impact on the consequences of the analyzed liquid waste release accident, nor the worst case pestulated sodium radiological accident (i.e., pnmary sodium release). Since that accident alread, postulates the release of all the pnmary sodium and all radioactivity contained in th' primary sodium, the potential worst case offsite radiological consequences of an accident involving the non-radioactive secondary sodium m the storage tanks can be no worse than the previously analyzed pnmary Na and hquid waste release evcats. The evaluation looked at whether a postulated accident involvmg the secondary sodium could impact safety re:ated systems or structures at Fermi 2 and so increase consequences by mvo'vmg Fermi 2. The evaluation concluded that Fermi 2 safety related structures and systems, and the safe operation of the plant, would not be affected by the postulated accident since this postulated m:enario is bounded by the analyzed explosion of the 20.000 gallon Fermi 2 hydrogen water chemistry torage tank. Also, the chemical release is not expected to result in concen:ratioa levels m the Ferm ). control room above allowable limits for Page 16 i
(
Safety Ev aluati:n 97-006
(]
NaOll. There is no requirement or credit taken in the Fermi 1 safety analyses for personnel to be U
continuously located at Fermi 1 (in the Control Room or elsewhere in the facility), or w quickly respond to any accident, so the potential harm to personnel located there from a postulated accident involving the secondary sodium is an industrial safety concem, not a concem with increasing the consequences of an accident. Based on the University of Michigan study, a reaction of 70 gallons of sodium will result in acceptable consequences offsite.
- 7. Will the proposed change create an accident or malftmetion of a different type than any previously evaluated for Fermi 1 S AFSTO" A. Storage of residual sodiu-eels in the secondary sodium storage tanks does not create an accident or malfunction of a different type than any previously evaluated for Fermi i SAFSTOR since the analysis used in this safety evaluation has a worst case postulated accident scenario sirr.aa. to that used in Faference 3.1, though with considerably less sodium. Any damage to the SG Building, or fragments generated as a result of the event, will not result in primary system damage with consequenct.s greater than that assumed in tlyis SE. Also, any radiological hazard as a result of the explosion potentially breaching the primary sodium containing piping or tanks and Reactor Building was previously evaluated and found ac:eptable. The ground shock may result in rupture of the liquid waste tanks, however, this event has previously been analyzed. Smce the postulated secondary sodium accident scenario was determined not to affect fermi 2 safety related structures or systems, nor cause a concentration of NaOli in the Fermi 2 control room above the limit, no different type of accident has been created at Fermi 2 based on Fermi 1.
p
- 3. Will the proposed change reduce the margin of safety as defined in the Fcrmi i Technical d
Speci6 cations?
A. Storage of residual sodium heels m the secondary sodium stcrage tanks does not reduce. Se margir, of safety as defined in the Fermi 1 Technical Specifications since the Fermi i Technica Specifications and Reference 3.10 do not rddress secondary sodium. The only sodium addressed was the primary sedium since it contains residual radionuclides. The analyses (Reference 3.10 section 2.1.3 & Reference 3.1I section 3.2) concluded that any accidental release of 100% of the activity in the primary residual sodiun would result in less than 0.003 mrem whole body exposure at the site boundary. Also, the analyzed hquid waste accident resulted in concentrations well within alloweble hmits. Therefore, the NRC agreed in their SER that the residual sodium and liquid waste do nc :
present a signi6 cant radiological hazard to the health and safety of the public.
While the letters submitted to the NRC which mentioned the residual secondary sodiurs assumed it was passivated, no radio!agical hazard m addition to that already assumed in Referen' : 3.10 is posed by its existence in a non passivated form. since the greatest radiological hazard a postulated accident involving the secondary sodmm can result m has already L.een determined by the NRC to not present a signi6 cant hazard to the health and safety of the pubhc. Therefore the margin of safety has not been decreased. Also, the Fermi i Techmcal Speci6 cation do not require or assume continuous staf6ng at Fermi 1, thus the NaOli haza J to personnel at Fermt I for the duration of the postulated event will not reduce the margm of safety as de6ned in the Fermi 1 Technical Speci6 cations.
V Page 17
Safety Evtluati:n 97 006 o
,-p.
Additionally, the tanks appear to have adequate structural integrity and drain tiles were installed
'd along the north side of the SG Duilding, which have prevented further substantial water in.rusion
. (3.g., approaching the bottom of the tank)into the tank bays.
4.
Will the proposed change affect the enviromnent as presently described for Fermi 1 SAFSTOR7 A, Storage of residual sodium heels in the secondary sodium storage tanks does not change the effect on the environment as presently described for Fermi i SAFSTOR, since previous analyses bound the worst case postulated event covered by this safety evaluation from a radiological standpoint. Also, the offsite chemical release will result in concentrations at the site boundary less than the respiratory protection level based on the previously performed University of Michigan evaluation.
While the answers to these questions lead to a conclusion that the residual secondary sodium does not constitute an unreviewed safety question, the scenario will be submitted to the NRC for review. Since a Fermi 1 Safety Analysis Report (FISAR) is being created, this scenario will be specifically addressed in the FISAR as a separate scenario.
NRC review will be requested by the submittal letter. This decision is based on judgment that this scenario which hadn't been addressed during the licensing basis for Q AFSTOR should be provided to the NRC for review.
O V
(G Page 18
Safety Evduation 97-006 FIGURE 1
N, VEHICLE ENTRY MT t.OWER LEVEL) o-4 g
r 24
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OPERATING F1.00R Note: This figure to illustrate buildirs wall layoJt only,
. STEAM GENERATOR BUILDING p,y, g and does not depict actual PLAN VIEW equipmnt layout.
MEMBOANE,
Safety Evaluation 97-006 p
f O
S0010M SEPARATOR UNIT g,, 7 g 30-TON CRANE DRIDGE L
I I
a9't$ '
. s;-9001UM WATER v
REACTION VENT NITROGEN
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SECONDARY b E STEAM OENERATOR SODluM E
p
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1, =
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CYLINDERS
' CHEMICAL TANKS
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- H"*
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g FW LDESp AREA f
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4 SODIUM y
c INLET
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b
=.N
- n k-
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FRSE --
i CEILING
's FALSE i
W VENTILATION n'
,g CEILING S e
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- OUTLET OPENINGS
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- O OU rify wef i
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'T i'
--a 6'
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a4 M 8' STORAGE welkwy ga,,h kys TANK
- FIREWAl.,L e.
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M
, gMETAL LINEN I if---
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e SAND BAGS' D
Note: This figure to illustrate building wall layout only,.
STEAM GENERATOR BUILDING and mes i.ct epict actual ELEVATION Page 20 equipmnt layout.
Safety Evaluation 97-006 1.
l O
Attachment i N) l'01tallaumpact of Sodium Ilydroxide Aerosol on Fermi 2 Control Room This calculation will conservatively estimate the Fermi 2 cocol room maximum concentration of NaOli in the t-vent of the postulated worst case scenano. The scenarh mvolves breach of a secondary rodium tank, instantaneous reaction of the solid sodium (70 gallons) with water, and explosion of the 11 gas 2
produced 'uy the reaction. The reaction produces NaOli as well as 11 gas. Approximately 986 lbs. of NaOll 2
(4.5 x 10'g) are produced by reacting 70 gallons o sodium. When the 11 instantaneously explodes (rather r
2 than some or all leaking out of the room), the NaOli is postulated to become airbome and then is released from the room in the fonn of a puff when the tank room is breached by the explosion.
Other Assumntions:
The Na; Ji is not diluted by any water in the roorn.
Even though it is unlikely the NaOli would travel as far as Fer.ni 2 due to its heavier than air density and due to the likehhood that it will react with CO and moisture in the air, losses due to settling and 2
reactions are neglected.
The puff travels towards the Fermi 2 south coritrol room air intake, though this is not the prevailing wind direction. Wind from SSW, SW, WSW and W is more predominant than SSE or S, which would be nteessary for the puff to reach the intake.
bs 3
Control center envelope contams 252,731 ft per Fermi 2 Design Calculatios. (DC) 5252.
e Normal ventilation inlet Oow is 2933 cfm of outside air per Fermi :' DC-5252.
The highest concentration at the control room intake is reached when x = 0, in the formula from Appendix B of RG 1.78, where x is the distance from the puff center in the horizorital alongwind direction and X = D - pt, where D is the source-reception distance, p is the wind speed and t is the time after release. In other words, the highest concentration at the intake is when the cloud has traveled the distance to the intake.
Since the release is instantaneous. the puffis only briefly at the intake.
The Fermi 2 control room does not have a NaOli detutor.
The puff does diffuse per the equation m Appendix B of RG 1.78, June 1074.
Ground level release is assumed. The tank room is below grade, but the explosion would elevate the e
puff. The transite walls will fail, allowing the puft to leave the Steam Generator Building.
Even though NaOli is heavier than air, this evaluation allows some of the NaOli to climb during dispersion to reach the control room intakes.
Page 21
..J
Safety Evaluation 97-006 The disunce of 338 meters from the Steam Generator Building to the closest Fermi 2 Control Room intake
(
e V
on the south side of the Fermi 2 Auxiliary Building was obtained from converting 1114 ft, scaled fro.n Fenni 2 drawings 6MS721 109 Rev. A and 6MS721 ll2 Rev. A to meters.
For this calculation, it is assumed that the contrcl room normal ventilation remains in service, and the control room is not manually isolated.
Informption on NaOll; 3
ne OS11A occupation limit (8 hrs / day,40 hr work week)is 0.002 g / m. This ie also the ceiling limit. A separate Short Term Exposure Limit (STEL) has not been established. Tlie immediately Dangerous to Life and licalth (IDLII) concentration is 0.250 g / m'. Per a University of Michigan study (" Evaluation of Sodium Storage at the Enrico Fermi 1 Atomic Power Plant", Reed Robert Bum, University of Michigan, Sep'tember 1982), respiratoiy protection is recommended at 5 times the Threshold Limit Value of 0.002 g / m, which is 0.01 g / m'. (This study concluded that if 307 gallons of sodium were totally combusted and dispersed, the concentration at th,e site bc.undary for the few seconds the vapor cloud passed would be just at the respiratory protection limit.)
NaOli is normally a solid at mom temperature, having a melting point of 604 F (318'C) and boiling point of 2534*F (1390 C). It has a negligibla vapor pressure. NaOli is soluble in water and solutions are viscous. The reportable quantity (per the EPA) for NaOli, if there is a release, is 1000 lbs. Irritations and bums can result if Naoll is inhaled, ingested, or comes in contact with eyes or skin.
p\\
During a sodium water reaction, the sodium decomposes the water, producing sodium hydroxide (NaOll) and hydrogen (11 ). The sodium water reaction takes place at a very high rate causing the hydrogen to 2
spontaneously ignite. The primary reaction is Na + 11:04 NaOli + 1/211. An intermediate step that 2
produces sodium hydride (Nall)is possible if the sodium is in a liquid state,2Na + 110* NaOli + Nall.
Nail decomposes t vaporization or sublimation and depends on the total pressure above the melting point of Nall. At temperatures below 385
- C, Nall is in equilibrium, Nall = Na + 1/211. Nall teacts 2
violently in moist att and possibly hydrolyzes in excess 11:0 to form NaOli and 11, thus, eliminating 2
intemlediate products (Nail + 1104 NaOli + 11 ). (
Reference:
Sodium - NaK Engineering llandbook, 2
2 Gordon and Breach fo: U.S. AEC, O.J. Faust, Editor 1972)
The NaOli absorbs the carbon dioxide and water in the tank room, creating carbonates. NaOli particles also rapidly a. >rb carbon dioxide and water from air, so NaOli in an aerosol is likely to be reactir.g with the air as it travels.
Path from Steam Generator fluildme to Fermi 2 Control Room Intake The amount of NaOli vapor m the puff decreases from the imtial quantit / in the room as follows per enginectmg judgment. The NaOli is m the form of droplets,,r particles in the air, it is not a gas.
Therefore, when the vapor impacts a sohd surface, much of the NaOli will be deposited on the surface or fall to the ground rather than stay m the puff and move around the obstacle.
The first obstacles encountered are the Steam Generator Building intemal surfaces, including the walls of the tank room. When the tank room is breached, the vapor exits to the rest of the Building. For this p
calculation, it is assumed the transite walls and other non-concrete wall barriers are opened up by the U
I age 22
Safety Evtluation 97-006 l
{
~
Q explosion. These openings provide the release pathway. Part of the north wall, facing Fermi 2, above V
elevation 629'l'/ " is transite. Some of the aerosol most likely escapes the tank room through openings 2
prior to the explosion, but its release pathway is the same. Also, such aerosol would have less NaOli in it. Some of the NaOli in the aerosol will be deposited on the tank room walls and ceiling. Steam Generator Building walls, and other surfaces. Also, a small amount of NaOli will probably react with the CO, which is not a hazardous material. A conservative assumption of only CO in the tanks, forming Na2 3
2 25% deposition and other losses is assumed. This assumption is believed conservative, since only the concrete block walls of the tank room are expected to fail.
The Fenni i Reactor Building is directly north of the Steam Generator Building. Most of the puff emitting from the %m Generator Building heading towards the Fermi 2 control room intake will directly impact the iemti i Reactor Building and deposit the entrained droplets on its surface. Some may rise above and some of the puff may go around the Fermi 1 Reactor Building. The only path ofinterest for this calculation is that towards the Fermi 2 control room intake. A building wake factor of 2.9 is used for the Fermi i Reactor Building (Regulatory Guide 1.3 " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant, Accident for Boiling Water Reactors", Rev. 2, dated June 1974, Figure 2).
Some of the NaOli that makes it past the Reactor Building will then impact the Fermi 1 FARB or the Sodium Building complex. Conservatively,10% is assumed deposited on these surfaces.
The Fenni 2 Tm bine Building further blocks the straight line pathway cf '.he aerosol aerosol making it past the Fermi I Reactor Bcilding and prevents it from reaching the con:rol room intake. A building wake p
factor of 3 is used for the Fermi 2 Turbine Building. Some NaOli will make it around or over the Fermi 2 V
Turbme Building and the aerosol traveling to the west of the Turbine Building does have a straight line path to the south wall of the Fermi 2 Reactor Building which is approximately 62 ft from the western edge of the control room intake (Fermi 2 drawmgs 6MS721-109 Rev. A and 6MS721-112 Rev. A). Based on the con 0guration of the air mtake, air enters from the bottom of the intake through a set of screens.
Additionally, any aerosol makmg it just past the southwest edge of the Fermi 2 Turbine Building, will likely impact th walls protectiri the transformers. There are 3 transfonner walls in series perpendicular to the west wall of the Fermi 2 furbine Building. Another 50% of the NaOli is conservatively assumed to be deposited on or trapped by the transformer walls or the intake screens for the control roem ventilation.
His is conservati,e compared to using a normal buildmg wake factor of 3 for these walls, but appropriate since the height of the walls is less than the height of the control room intake and the height of the Turbine Buildmg. Additionally, a wmd tunnel effect exists between the AIB and the Fermi 2 plant. Wmd tends to blow cast or west in that area and stronger than prevaihng wmds. Any NaOli arriving in the area south of the Turbme Building and north of the AIB will likely be blown east or west and prevented frem reaching the area of the control room intake. No specific reduction factor is assigned for this wind tunnel effect. It serves as another justification for why less than 50% of the NaOli which travels past the Fermi 2 Turbme Building will reach the control room mtake. The amount of NaOli reac'.ing the control room mtake vicmity is assumed then tc be:
NaOH = (4.5x10' g)(1 - 0.25)(1 - 0.10)(1 - 0 5)
= 1.7x10' g (2.9)(3.0) od Fage 23 l
Safety Evdu: tion 97 006 4r
"}
This calculation neglects the removal by reaction to sodium carbonate discussed earlier and any u>
deposition or dropping of NaOli particles or dropicts during free transit.
Dispersion Calculation The following calculations are used to determine the dispersion of the NaOli plume between the Fermi 1 Steam Generator Building and the Fermi 2 Control Room intake, and the NaOli uptake by the Fermi 2 Control Room ventilation system. The dispersion equation used is from Regulatory Guide (RO) 1.78 Appendix B.
f
~
.i 2
2 2
Gi
-7.87(o,/ + o,a y,2, y,2 )W-
\\
xexp -1/2 G,, + G, 4-0,,+ 0,+ 0,,+ 0,> _
- - - =
i 1
1 where:
o,, o, and o, are standard deviations of gas concentrations in horizontal alongwind, horizontal y
crosswind and vertical crosswind directions. o, is assumed equal to o per RG 1.78.
y ci s the iattial standard deviation of the puffin cubic meters.
i
-io
- 01 ai =
-7.8 7, -
7 Where Qi s the puff release quantity, x, is the density of gas.t standard conditions g / m'.
i x, = p = Mol. Weight x 10' x 1.8/82.06 (T + 460)
= 40 x 10' x 1.8/82.06 (70 + 460) 3
= 1655 g / m 7,
.n
,. G, 4.5x10'g c,=
=
_7.87,,
,(7.87)(1655g / m ),
2 I
o, = 3.2 x, y, and z are the distances in meters from the puff center in the horizontal alongwind, horizontal crosswind, and vertical crosswind directions respectively. x in exponential term = D-t, where D is source - receptor distance, p is wind speed and t is time after release.
For t = D, x = 0. This occurs when the centerline of the puff reaches the control room intake. The unit concentration at the Control Room ir.take at that time will be:
1 = 7.87(12 + 3.2 )(6 + 3.2 )'
2 2
2 2
exp -l / 2(0 +
2 + 6, + 3.2 _
2 2
G, 12 + 3.2 Page 24
Safety Evaluation 97-006
=
Let y = 62 ft., since thete is no direct path for the plume to reach the control room intake without running into the Fermi 2 Turbine Building. However, the plume could reach about 62 ft. horizontally away from the control room intake while missing the Fermi 2 Turbine Building. As conservatism, this calculation will assume centerline is 62ft. away from intake rather thsn edge of plume being about 62ft. away from edge of control room intake.
62ft. x 12 in/ft x 2.14.cm x _lm. = 19 meters in.
100cm z - 10 meters based on height ofintake and relea<e points of Steam Generator Building I-- = 13xlo^ lm' Gi (Since the NaOH in this situation is not really a gas, but particles suspended in air, this calculation was repeated using the density of air as a sensitivity case.
p = ll61g / m' = y,
. i,3
- ")
. G, 4.5x10'g cr, =
=
7.87,,
(7.87{l161g / m')_
7 a, = 3.6 7.87(12 + 3.6 f6 + 3.6 )
1=
-.+
= 13x10" 2
2 2
2
-l / 2k0 +
exp 2
2 2
12 + 3.6 6, + 3.6 Q,
Since both the case using the density of NaOH and this sensitivity case using the den.;ity of air result in the same I when the puff arrives at the control room intake (peak concentration outside the control I
toom intaky, the calculation has been shown to be insensitive to which density is used.)
In an iterative process this calculation next determines the time / distance at which the concentration at the control room intake becomes significant. The followmg table summanzes the results.
O Page 25
Safety Evaluation 97-006 l
Summary Table Time NaOli Concentrati3n at Amount of NaOH NaOli Concentration in (Seconds after release)
Control Room Intake entering Control Room Control Rooin 3
3 (g/m )
during interval (g)
(g/m )
s300 insigmficant 310 0.017 0.24 3.4x 10
320 0.076 1.0 0.0002 330 0.17 2.4
'.0005 338 0.22 24 0.0008 346 0.22 2.4 0.0012 356
- 0. I 7 2.4 0.0015 366 0.076 1.0 0.0017 376 0.017 0.24 0.0017 Calculation Assume = 1 meter /second ifi = 100 sec, x = 338 m - I m/sec (100 sec) = 238 m I= 7.87(12 + 3.2")(6' + 3.2') '
2 exp -l / 2 ( 12, + 3.22 + 3 3 + 3.2 2 + 6 + 3.2 s,
- 2 f
G 2
2 E - 2.3 x 10'83 which is insigni6 cant G
At t = 201).u, x = 338 m - I m/sec (200 sec) = 138 m I
Ie 7.87(12 + 3.2 )(6 + 3.2 )
2 2
2 2
exp -l / 2
( 12 + 3.22 + 12 + 3.22 + 6 + 3.2 s 2
2 G
I ~ 2.0 x 10'32 which is insigm6 cant G
At t = 300 sec, x = 338 m - I m'sec (300 see) = 38 m I = 1.2x10 / m' G
Al_L= 310 sec, x = 338 m - I m see (310 see) = 28 m I = 1.0x10" / m' O
OG Page 26 I
['
- Safety Evaluation 97 006 :-
a
/.
os" "
1[\\;. f At t = 320'sec, x = 18 m fI. ='4.5x10 / m'--
4
.p-At t = 330 sec, x = 8 m.
- d. '= 1.0x10-8 / m'
.Q
' At t = 338 sec. x = 0 m 5 = 13x10-3 / m'
.g From the above calculntions, at t = 310 sec, the unit concentration becomes significant. Therefore, tlic -
calculation of amount of NaOH entering the controi room will start with the period from 300 - 310 sec, using the highest value for that period, which is 1.0 x 10%t 310 sec.
p The concentration of NaOltst the control room irtake at 310 see will be:
a 3
. x (y. /Q)(Q) = (1 x 10) (1.7 x 10' g) = 0.017 g / m j.
The amount of NaOH entering during the period from 300 - 310 see will be:
3 3
3 (2933 ft / min)(1 min /60 sec)(10 sec)( 0 017 g / m )(0.0283m'/ ft ) = 0.24g
- The concentration in the control room at the end of this period will be:
l O.24g i ft' 3.4x10'$ g / m'
=
25?,731ft' x -. 0.0283m' From 310 - 320 sec, use:
I = 4.5x10~*
G
.y = O.076 g / in' at intake 3
3 3 3
.. (2933 ft / min)(1 min /60 s:c)(10 sec)(0.076 g / m )(0.0283 m /f1 ) = 1.0 g Page 27 k
y
.m w
-6 er, a
w.*
Safety Evaluation 97-006
' [
The concentration in the control room at 320 see is:
1.0g + 0.24g ift'
= 0.0002g / m' 252,73Ift' O.0283m' From 320 - 330 seq, use:
E=1.0x10" Q
ysop = 0.17 g / m' (2933 R'/ min)(1 min /60 sec)(.10 sec)( 0.17 g / m )( 0.0283 m /ft') = 2.4 g 3
3 The concentration in the control room at 330 3ec is:
2.4g + 1.0g + 0.24g Ift '
= 0.0005g / m' 252,731ft' O.0283m' Etom 330 - 338 sec. use:
1 = 1.3x10" 0
Xmut, = 0.22 g i m' (2933 ft'/ min)(1 min /60 t,w>(8 sec)(0.22 g / m')(0.0283 m'/ft') = 2.4 g The concentration in ihe control room at 338 see is:
2 2.4g + 2.4g + 1.0g + 0.24g if 3
x
= 0.0008g / n 252,73Ift 0.0283m' 2
Assuming the NaOli concentration in the puff decreases in the same proportion as increased, when the trailing part of the puff reaches the Fermi 2 cone ' room intake, additional NaOH enters the control room envelope as follows.
From 338 - 346 sec, aaother 2.4 g enters y" = (2.4 + 2.4 + 2.4 + 1.0 + 0.24)g ift 5
= 0.0012g / m'
-x 252,73 Ift '
O.0283;n' O
=
Page 28 A
Safety livduation 97-006 l.
l O
From 346 - 356 sec, another 2.4 g enters -
V~
y" = (2.4 4 2.4 + 2.4 + 2.4 + 1.0 + 0.24)g Ift '
3 x
= 0.0015g / m 252,73Ift' O.0283m' From 356 - 366 sec. another 1.0 g enters y" = (1.0 + 2.4 + 2.4 + 2.4 + 2.4 + 1.0 + 0.24)g Ift'
= 0.0017g / m' x
252,73Ift' O.0283m' From 366 376 sec, another 0.24 g enters y" = (0.24 + 1.0 + 2.4 + 2.4 + 2.4 + 2.4 + 1.0 + 0.24)g 1p' 3
x
= 0.0017g / m 252,731ft' 0.0283m' This calculation neglects the removal of NaOli in the control room sentilation exhaust. After 376 sec, insignificant additional NaOli enters the control room and slowly the concentration in the control room decreases from a peak of 0.0017 g / m' due to NaOH contaminated air being exhausted and fresh air being admitted to the control room. At no time does the peak concentration in the control room exceed 3
the 0.002 g / m limit.
A more accurate calculation of maximum concentration would be calculus based rather than at multiple discrete time steps. It would take into account both the input rate and removal rate of NaOH. The change of NaOli would be:
dN
-- = P - kN,where P is the production factor, or inpt.t rate; k is removal rate by ventilation exhaust, di and N is the amount of NaOli in the control room. The highest amount of NaOli is:
(1-e-")
N=
k = 2933 ft' / min 252,731ft' 4
k = 1.16x10 / min x 60sec A = 1.93x10" / see P=
= 1.59x10-' g / sec 76see Page 29
S:fety Evaluation 97 006 o
.p N=
1-cM'#"D* = 12g 1.93:1C / see U
The highest concentration is 12g / 252,731ft.= 4.7x10~5g / ft'x
= 0.0017g / m' 5
3 O.0283rr The peek concentration result is the same for both calculations within the accuracy of the calcult. tion.
A sensitivity check was performed to determine if the peak Fermi 2 Control Room concemration of NaOli would be greater if the release took place over a longer period of time rather than being a p'aff type release. With the decreased release rate over an extended period, the concentration in the Control Room did not increase, if for some reason all 3 secondary sodium storaSe tanks at Fermi 1 were to rupture, with sodium reacting, hydrogen exploding, and NaOli becoming airbome, the concentration in the Fermi 2 control room would increase at most by a factor of three. The amount ofincrease would depend on the time between releates from each tank and how much NaOli was exhausted from the Fermi 2 Control Room before the next puff reached the intake. Note that per NRC Regulatory Guide 1.78 section C.S.a, multiple container failures need not be postulated for a maximum concentration accident unless they are connected. During this 3
highly unlikely scenario, the Naoll concentration would be above the NaOli limit of 0.002 g / m, but less than the 0.010 g / m' that the University of Michigan study described as the respiratoty protection 3
needed concentration. If conservatisms, such as assuming no NaOli reacts with the air to form carbonate (d
and assuming no settling out of the heavier than air particles occurs, are removed, the control room concentration of NaOli fo' lowing even a 3 tank event would likely be within the limit.
Per the Fermi 2 licensing basis (UFSAR Sections 9.4.1.3 and 6.4.1.3), if a fire occurs outside the plant, the operators will manually isolate the contiol center to prevent admission of smoke or noxious fumes.
The control center intake t noke detectors would detect smoke within 1 minute of the smoke entering the intake and the operator can manually isolate in less than 10 seconds. There is immediate access to self-contained breathing apparatus. UFSAR Section 6.4.2.5 describes the mask hose apparatus as providing 30 man-hours supply plus SCB As are available for work outside the control room.
The chemical release fron, this postulated event is very similar to the smoke or fumes created by a fire as postulated in the Fermi 2 UFSAR. with the exception that the smoke detectors would not necessarily detect the NaOll. The operators would likely be informed by site personnel and so may isolate tne 3
control room imake. Since the NaOli concentration is below the 0.002 g / m limit, isolation is not necessary, but would not cause a problem.
V l
Page 30
l.. CMM AN NOV 261997 Sargent & Lundy"*
November 25,1997 Project No. 09471-023 Letter No. SL-97-081 The Detroit Edison Company Enrico Fermi Atomic Plant Independent Review of Fermi 1 Safety Evaluation 97-006 for Blast Loading and Structural Considerations
References:
1.
Draft Fermi 1 Safety Evaluation 97-006, Pages 1 through 27, Prepared Date 09-24-97.
2.
Draft Fermi 1 Safety Evaluation 97-006, Pages 1 through 18 and Figures 1 and 2, Prepared Date 10-24-97. Also, modified O
Paragraph B.2.2.A faxed on 11-25-97.
Ms. Lynne Goodman Enrico Fermi-Unit 1 The Detroit Edison Company 6400 North Dixie Highway Newport, Michigsn 48166
Dear Ms. Goodman:
The Sargent & Lundy (S&L) letter SL-97-056 provided our independent review comments on the draft of subject safety evaluation (SE)in Reference 1. That review provided our comments from a structural view point only (i.e., pressure loading, ground motion estimation, and structural resistance considerations).
We have reviewed the revised SE draft in Reference 2 from the same structural perspective. We conclude that all of our previous comments have been addressed, and that Reference 2 does not require any further modification, provided indicated structural failures in Unit 1 Steam Generator Building and Reactor Building
]
Containment are acceptable to Detroit Edison Company, i
55 East Morroe Stree Chicago. IL 60603 5780 USA + 312 269-2000
y O
M$. ' vane ooodman "ovember 23.,e97 The Detroit Edison Company _
Page 2 Thank you for giving us the opporton'ty to provido review comments to your SE97-006.
If you beve any questions or if I may be of further assistance, please contact me at 312-269-7203.
Yours very truly, M'
Mohamad Amin Consultant, Nuclear Plant Division MA:ljr Copies:
D.L. Leone F. M. Berry M. E. Orth O
S.a.aeupp g%meVermful9706l wp O
_