ML20153B792

Rev 0 to Fermi 1 Sar

ML20153B792
Person / Time
Site:	Fermi
Issue date:	08/31/1998
From:	DETROIT EDISON CO.
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ML20153B787	List: ML20153B792 NRC-98-0129, Forwards Rev 0 to Fermi 1 SAR, in Response to NRC Comments & Updated Info Re 971215 Application for License Amend. Revised Pages Marked Rev 0,but with Updated Date as NRC Has Not Yet Approved SAR
References
NUDOCS 9809230244
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U.S. Nuclear Regulatory Commission Docket No. 50-16 NRC License No. DPR-9 Enrico Fermi Atomic Power Plant, Unit 1 Fermi 1. Safety Analysis Report.

Detroit Edison l 9009230244 980913 PDR ADOCK 050000164 P PDR l

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Fermi 1 Safety Analysis Report Table of Contents PAGE LIST OF TABLES iii LIST OF FIGURES iv 1.0

SUMMARY

1-1

2.0 INTRODUCTION

2-1

3.0 DESCRIPTION

OF FERMI 1 3-1 3.1 Description of Plant 3-1 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 4-2 l 5.0 SURVEILLANCE 5-1 5.1 Surveys 5-1 5.2 Inspections and Testing 5-1 6.0 ADMINISTRATIVE CONTROLS 6-1 6.1 Organization and Responsibilities 6-1 6.2 Quality Assurance Program 6-2 6.3 Review Committee 6-6 6.4 Radiation Protection 6-7 7.0 FIRE PROTECTION PROGRAM 7-1

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8.0 . ENVIRONMENTAL ASSESSMENT ~ 8-1 8.1 Distribution of Estimated Population 8-1 8.2 . Non-RadiologicalImpacts 8-1 8.3 RadiologicalImpacts .

8-3 8.4. Postulated Radiological Accidents 8-4 8.5 Alternatives Considered 8-10 8.6 Conclusions 8-12 l 9,0 References 91 li Rev 0, .18 l

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All the blanket shipments were made using two Philadelphia Electric Company Model PB-1 shipping casks, one of which was owned by General Atomic Company. To utilize these casks to handle Fermi-1 blanket subassemblies, the Columbus Ohio Laboratories of Battelle Memorial Institute designed 14 ,

baskets and basket containers, which were then fabricated by the Central Ohio Welding Company under l Battelle's supervision. The subassembly baskets were constructed of carbon steel gridwork, sized to fit l into a cylinder 25 inches O.D. x 158 inches long, containing 37 individual storage Lies. The 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.

l Five dry runs were made to develop handling 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 inch axial subassembly movement. A 2-inch-diameter, carbon steel mill tube of appropriate length, 5 to 62 inches, weighing approximately 0.735 lb/ft was placed over each subessembly as a spacer. Once the subassembly and spacer were in place, the basket was raised from the pool and allowed to dry overnight. Radiation levels up to 50 R/hr at the basket surface necessitated isolation of the building and rigid health physics controls of personnel while the basket was unshielded.

A helium mass spectrometer leak test of the container weld followed, prior to closing the cask for shipment.

No problems were encountered other than when the bottom supports of one basket failed, resulting in loose pins and basket damage. There were no radiation exposure incidents, and contamination levels were low. Several pins which were not located by the time the last shipment was made to Idaho were shipped to Nuclear Engineering Company's Sheffield, Illinois, burial site on September 23,1975.

3.2.2 Sodium and NaK Removal 3.2.2.1 Secondary Sodium l At the beginning of the retirement program, about 34,600 gallons of secondary sodium was stored in l three systems: the secondary system, service system, and storage tanks. The secondary sodium system had been drained for steam generator maintenance in 1972 before the decision was made to retire the plant. Samples of secondary sodium were analyzed by three laboratories and determined to be non-radioactive since the level of beta activity was no higher than the activity in naturally occurring isotopes found in regular commercial grade clean (radiologically non-contaminated) sodium. The major volume of secondary sodium was drained into the three 12,000-gallon storage tanks via the service system by normal operating procedures. A complete drain, except about 3600 gallons in the IIIX tube bundles, was accomplished by gravity drain supplemented by evacuating the storage tanks and pressurizing the nitrogen cover gas system. In November 1973, the sodium from the IHX tube bundles was drained into the primary sodium system.

The secondary sodium service system was drained directly to a barrel fill station. This was accomplished by removing the old secondary sodium system fill line and pressurizing the system with nitrogen cover gas. A sodium barrel fill station was constructed in the NaK room located in the Sodium Building, adjacent to the former primary and secondary sodium outdoor tank car unloading station. A barrel storage area was built in another adjacent area between the Sodium and Waste Gas Buildings. The sodium barrels used were USA Standard Department of Transportation No.17 E, Type 1, 55-gallon drums. All drums were visually inspected and leak-tested before use.

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!' The transfer of secondary sodium to drums began on May 31,1973, and was completed July 24,1973.

A total of 569 drums (about 30,400 gallons) were filled. All filled drums were sold and shipped to Fike Chemical Company of Nitro, West Virginia, to be processed into sodium methylate.

Based on the above; there is an assumed 600 gallons of sodium remaining in the secondary sodium system, the secondary sodium service system, and the secondary sodium storage tanks. The sodium is a l j~ 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 or less is in the storage tanks' as verified by visual inspection (Spring 1997).

l 3.2.2.2 Primary Sodium _ l l .At the beginning 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 FARB service system. Prior to draining the primary sodium, samples were taken from the primary service system No. I sodium ~ storage tank and the FARB transfer tank for analysis. The level of

. radioactivity in this sodium was 0.022 Ci/g, and was primarily of 22Na,137 C s, and 90Sr.

All nondrainable components;had to be siphoned by opening the components and installing special l siphon pipes connected separately to the service system. It was necessary to drain the primary system in l steps, beginning May 8,1973. First,3800 cubic feet of sodium from the primary system and the IIIX 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, l .except for the plenum, in three individual efforts. The plenum was drained separately via the service system recirculating line. About 400 gallons of sodium from the overflow' tank was removed by drilling into the bottom of the tank and siphoning. The FARB transfer tank contained about 8000 gallons of sodium with another 280 in the adjacent overflow tank. Both tanks were siphon drained with individual siphon pipes. The two storage tanks in the FARB cold trap room contained about 775 gallons of sodium l .which was transferred to the primary storage tanks with the rest of the primary sodium. I 1

There was more primary sodium than could be stored in the three primary storage tanks. The remainder u . was placed in drums. Dremming primary sodium was accomplished using the equipment for barreling l secondary sodium, with some minor modifications. A total of 630 steel drums were filled with about l' 32,000 gallons of primary sodium in November 1973. The filled drums were coded and moved to L temporary storage in the fuel transport trestleway and Reactor Building. Each drum measured about 5

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mR/hr at a distance of two feet. The drums were stored four to a pallet, two and tree tiers high within controlled areas with restricted access for periodic inspection and smear checks.

I The remaining primary sodium, about 45,000 gallons, was temporarily stored frozen in the three storage tanks in the Sodium Building. When the sodium was ready to be shipped offsite, the sodium from the L

. three storage tanks was transferred to steel drums. During this drumming operation, samples were collected from representative drums. The maximum levels of activity were: 1.07E-3 pCi/gm of 22Na, 6.83E-4 pCi/gm of 137Cs and 1.15E-3 Ci/gm gross p. A total of 1347 drums, including the drums in j' temporary storage, were shipped offsite in 1984. The sodium was sent to Idaho to the Department of

Energy.

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3.2.3.3 Sodium Building The 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 overflow tank end the intermediate heat exchangers (IHX) drain line were cut. The nozzles of the tank were capped. The residual sodium on the walls of the piping was converted to Na2 CO3 by purging the piping with CO 2. The remainder of the primary sodium service system including the hot trap economizer, heat exchanger and pump, and all piping except the storage tank line have been removed from the Sodium Building.

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 sodium 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 Boundary The Fermi i Protected Area Boundary was revised to exclude many nonradioac6ve reas such as the Steam Generator Building, the Control and Office Buildings, and Turbine BMam- & Protected Area Boundary as shown in Figure 3.4 is marked by a seven-foot-high chain-lit. nmce nd building walls that enclose the FARB, the Reactor Building, the Sodium Building, and the ,vw.e Gas and Inert Gas Buildings. The Health Physics Building has been dismantled. The Steam Generator Building was detennined to contain a small amount of tritium within the secondary sodium and secondary sodium service system in 1998.

3.3 CURRENT PLANT CONDITIONS 3.3.1 Primary System Retirement of the primary system included draining all the primary sodium and refilling with an inert gas (nitrogen) to which CO 2was added to reduce the residual sodium deposits to inactive solids. The system was then sealed and maintained at slightly positive inert gas pressure to prevent the entrance of water and/cr moisture and to minimize dispersal of any remaining radioactive material. A gastight system was established consisting of the primary 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 caps. This system was filled with carbon dioxide, which reacts with the residual sodium to form an inert solid compo md (Na2CO3). The primary system is presently being maintained under a positive pressure by use of bottled CO2-The machinery dome has been seal-welded in place. The machinery dome glass observation ports, gaskets, and gasket compressors were replaced with carbon steel plates which were seal-welded in place.

Its access door was also seal-welded in place.

3.3.2 Primary Sodium Storage Tanks Nitrogen cover gas is maintained on the primary sodium storage tanks. When N2 Pressure decreases, additional N2 is manually added. The door to the sodium storage tank room remains locked except during entries.

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333 Liquid Waste Disposal System All potentially contaminated drains and sumps collect in the hot sump in the Fuel and Repair Building.

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 these 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 prepared if discharges were to be required in the future.

33.4 Electrical Supply Power is supplied to Fermi 1 through the 120KV switchyard located south of Fermi 1. Three offsite l power lines supply the switchyard. The switchyard provides power through the switchgear room in the Fermi 1 Turbine Building 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 located on the second floor of the Control Building provides DC power to Fermi 1.

33.5 Sodium Remaining in Plant Systems As part of the initial decommissioning of Fermi-1 following shutdown, the sodium (Na) in the primary system, secondary system and supporting systems was drained. Due to the configuration of some of the 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 equipment and systems forming heels. The majority of the equipment and piping which was abandoned in place was blanketed with CO2gas to try to react any residual Na (passivate) to form non-reactive sodium carbonate (Na2 CO3). Experimentation at the time showed that due to the permeability of CO2 gas through Na2 CO3, the potential exists ibr a Na2CO3crust to form less than 1 /4" thick, beyond which no further reaction between the CO2and the Na may occur.

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 historical consistency and consistency with the size of tanks and drums for comparative purposes. Based 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 - 1200 gallons of residual sodium is estimated to be contained in Fermi 1 systems, of which approximately 435 gallons is primary sodium. Some of this sodium has been converted to sodium carbonate or bicarbonate, but some is still sodium metal.

The residual primary sodium is expected to be located in the following buildings:

• Reactor Building and sodium tunnel - approximately 374 gallons

• FARB - approximately 11 gallons e Sodium Building - approximately 50 gallons Approximately 225 - 725 gallons of secondary sodium residuals are estimated to be in the Steam Generator Building. Visual observation of the secondary sodium storage tanks identified 34 - 70 gallons of solid sodium metalin each tank.

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33.6 Taylor instrumentation Taylor Pressure Instrumentatior was used throughout the plant and contained NaK as a transmitting medium which is contained withm armor covered capillary lines. These capillary lines are expected to contain unreacted NaK since they have been sealed. A total of 33 of these instrument lines were identified. The 25 NaK filled instrument lines that were installed in the Steam Generator Building and Waste Gas Building were removed in 1997 and 1998.

33.7 Monitoring and Alarms 33.7.1 Water Intrusion System Monitoring detectors for water intrusion are located in two areas: 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 Fermi 2. The water intrusion alarm in the Reactor Building biological shield 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.

33.7.2 Cover Gas Monitoring The primary system cover gas pressure is monitored with high and low pressure alarms. The monitors and the alarm circuitry are periodically checked and calibrated in accordance with the Technical Specifications and written procedures. l i

3.4 ACCESS CONTROL I The area encompassed by physical barriers and to which access is controlled is the Protected Area. The Protected Area is enclosed by either a chain link fence or building walls which provide equivalent degree l of resistance to penetration. The fence is topped by three or more strands of barbed wire or brackets l angled outward with an overall height of no less than seven feet. Normal entry to the Protected Area is l through a normally locked gate in the fence ad,iacent to the Sodium Building. Other doors that are a part 1 of external walls, which act as a part of the Protected Area boundary, are locked or permanently sealed. l Access to the Protected Area is controlled, limited, and recorded. The access key is at the Fermi 2 l Radiation Protection Control Point. A second key is held in safe keeping by the Custodian for use only  ;

in extenuating circumstances.

Written procedures delineate the requirements associated with entry into the Protected Area and specific areas within the Protected Area to prevent unauthorized entries and to protect the safety and health of authorized personnel.

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SECTION 4: RADIOLOGICAL CONDITIONS 4.1 Total Nuclide Inventory

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' 4.1.1 Radionuclide 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 SAFSTOR period. Reference Numbers 11 and 13 provide the details for Table 4.1, a summary of the results of analyzing the reactor vessel, primary shield tank, secondary shield wall (concrete), reactor vessel l internals, sodium residuals, and liquid waste samples.

The 40 year SAFSTOR will result in the following estimated reductions between 1986 and 2025:

e' 85% reduction in total activity )

• 90% reduction in liquid waste activity j

• 89% to 90% reduction in exposure .

T The inventory of radionuclides in the waste wa vas based on measurements of samples from the waste water sump, tank dose rates and activity measurunents for the final total discharge afler the 1973 to 1975 decommissioning work. The 1986 total activity for 60Co ar.J 137Cs was estimated at 6.0E+3 Ci each. 1 This activity will decrease to 3.1E+1 Ci for 60Co and 2.4E+3 Ci for 137 Cs in forty years. j

- A review ofliquid waste system water samples and smear analyses performed from 1989 through 1998 identified additional radionuclides in the liquid waste system. The combination of radionuclides varied

~ in the different analyses, as did the ratio of radionuclides. The radionuclide inventory is not spread i evenly and the data suggests very localized deposits may be present in areas of the liquid waste system. l The following radionuclides have been identified in one or more analysis from the sump and waste tanks:  ;

3H 137Cs 14C 238Pu 55Fe. 239/240Pu 60Co 241Am 63Ni 242Cm 90S r ,

242/243Cm During the 1983 sodium drumming 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 22Na and 4.8E+3 pCi for the 137Cs respectively. After 40 years, the activity would i reduce to 2.5E-2 Ci and 2.0E+3 pCi, respectively. This amounts to a 65% reduction in sodium activity l In 1997, a sample was taken from a drum of sodium previously shipped offsite. The tritium content was l 2.2E-4 Ci/g. Based on the 435 gallons of primary sodium residues estimated to remain, the primary 4

sodium would contain 3.5E+2 Ci of tritium. A more conservative estimate based on interpolating from the higher tritium concentrations found in one sample of secondary sodium would be 4.4E-2 Ci/g, or 7.0E+4 pCi tritium in the primary sodium.

In 1998, samples of secondary sodium were analyzed for tritium. Tritium concentrations varied from lE-2 pCi/g to less than minimum detectable. The secondary sodium cold traps were not sampled, so conservative estimates were made of the possible tritium content in the secondary sodium systems, 4-1 Rev 0,8/98 l l

- ,.wy = , - - ,- y ew--.v .w.,

including the cold traps. NaH would concentrate in the cold traps, which removed impurities from the system. Tritium would be contained in the NaH. A conservative estimate of total secondary sodium tritium content is 0.62 Ci.

Amendment 12 to the Fermi 1 license permits additional byproduct material onsite for sample analysis, instrument calibration, or associated with radioactive apparatus, hardware, tools, and equipment. The cumulative quantity of the byproduct material must not exceed the criteria contained in 10 CFR 30.72, Schedule C.

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 Building (FARB), and the Sodium Building were surveyed as part of the scoping effort performed in 1997. The observed dose rates were typically < 0.1 mrem /hr with dose rates in many .

areas on the R/hr order of magnitude. Table 4.2 summarizes the results of the 1997 scoping st.rveys.

The highest dose rate observed in the Reactor Building basement was located on the No.1 IHX (3 mrem /hr on contact and 1.5 mrem /hr at 30 cm). The general area dose rates in the remainder of the basement were on the order of 0.1 mrem /hr. Other areas where general area dose rates exceeded 0.1 mrem /hr include the primary sodium 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 unit was used to identify the isotopic mix of the crud causing the hotspot on the IHX. The only isotope detected was 137 C s (30.07-year half-life).

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 observed.

4.2.2 Contaminated Areas The scoping surveys performed in 1997 focused primarily on areas assumed to be radiologically 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 in the 1972 Fermi 1 Retirement Report (140,000 dpm/100 cm2),

The highest contarnination levels were on the fuel transfer ports to the cut-up pool and to the transfer tank. Gamma spectroscopy analysis of an air sample from the chamber identified 137Cs as the only radionuclide present.

A room located in the basement of the FARB was used as a temporary storage area for depleted uranium.

The uranium was removed before or during the 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.

Surveys of the decay pool and cutup pool were last performed in 1986. Radiation and surface contamination levels measured are shown in Table 4.3.

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TABLE 4.1 CALCULATED ACTIVITY IN THE REACTOR VESSEL, ITS INTERNALS, THE PRIMARY SHIELD TANK, AND TliE SECONDARY SHIELD

• JULY 1986

• REACTOR VESSEL TOTAL ACTIVITY. CURIES ISOTOPE JULY 1986 JULY 2026 JULY 2086 94 Nb 1.87 E-06 1.87 E-06 1.87 E-06 60 Co 1.29 E-00 6.73 E-03 2.53 E-06 59 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 14 C 9.50 E-10 9.50 E-10 9.50 E-10 55 Fe 5.09 E-01 1.77 E-05 3.61 E-12 l TOTAL 2.05 1.96 E-01 1.23 E-01 l

• - REACTOR VESSEL INTERNALS TOTAL ACTIVITY. CURIES 94 Nb 4.06 E-02 4.06 E-02 4.06 E-02 60 Co 274. 1.41 5.2 E-04 59 Ni 3.92 3.92 3.92 63 Ni 87 64 41 55 Fe 110 2.5 E-03 2.92 E-10 TOTAL 475 69 45 l e PRIMARY SHIELD TANK TOTAL ACTIVITY. CURIES 14 C 3.92 E-10 3.92 E-10 3.92 E-10 55 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 e CONCRETE SECONDARY SHIELD TOTAL ACTIVITY. CURIES 152 Eu 2.75 E-03 3.44 E-04 1.52 E-05

• Reference 11 4-5 Rev 0,8/98

l SECTION 6: ADMINISTRATIVE CONTROLS The Detroit Edison Company (Edison), as the licensee for Fermi 1, has the responsibility for maintaining the License and Technical Specifications. Administrative controls have been established to ensure that management and administration of Fermi 1 is performed in a consistent manner that complies with l regulatory requirements.

I Responsibility for Fermi 1 is delegated through the line organization of the Senior Vice President,  !

Nuclear Generation. A Fermi 1 Review Committee functions to advise the Fermi 1 Custodian on all l matters relating to nuclear safety and to review and approve procedures, design changes, Licensee Event ,

Reports, and other activities.

6.1 Organization and Responsibilities l

6.1.1 Senior Vice President, Nuclear Generation l l

The Senior Vice President, Nuclear Generation is responsible for overall plant safety of Edison nuclear l power plants, including Fermi 1.

6.1.2 . Assistant Vice President, Nuclear Assessment l

l The Assistant Vice President, Nuclear Assessment reports to the Senior Vice President, Nuclear l Generation, and is responsible for the Fermi 1 facility. Reporting to the Assistant Vice President, Nuclear Assessment is the Fermi 1 Custodian.

6.1.3 Other Support l Fermi 2 organizations provide support for the Fermi 1 facility as needed.

6.1.4 Fermi 1 Custodian l The Fermi 1 Custodian or Custodial Delegates shall be responsible for directing all Fami 1 activities, l seeing that the activities are done in a safe manner and in compliance with the Technical Specifications, and reporting these activi:ies to the NRC. Key responsibilities include:

• Coordinate, approve, and assign work done at the facility.

• Maintain the physical facility as defined by Technical Specification B.1, Figure B-1.

• Comply with the Technical Specifications; administrative controls; and local, state, and Federal Regulations.

• Plan, control, and monitor decommissioning activities.

• Assign duties to the Custodial Delegates and Custodial Agents as required.

• 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.

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• . Review Fermi 1 Manual and design changes.

• Maintain records in accordance with the type and retention period stated in the Technical Specifications, Section 1.9.

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I The Fermi 1 Custodian shall be appointed in writing by the Assistant Vice President, Nuclear j Assessment, and approved by the Senior Vice President, Nuclear Generation. The Fermi 1 Custodian  !

shall, as a minimum, have a basic understanding of surveillance and Health Physics procedures and the l

Fermi 1 Manual. 1 I

6.1.4.1 Custodial Delegates Custodial Delegates shall act in the absence or on behalf of the Fermi 1 Custodian and shall assume such duties and responsibilities listed in Section 6.1.4 as required ar assigned by the Fermi 1 Custodian.  !

Custodial Delegates shall be appointed in writing. l 6.1.4.2 Fermi 1 Health Physicist l The Fermi 1 Health Physicist shall review all procedures and limits involving the handling of radioactive l materials. This individual shall be responsible for ensuring that all plant discharges and shipments are l within the limitations set forth in the Code of Federal Regulations. i The Fermi 1 Health Physicist shall be appointed in writing by the Fermi 1 Custodian and shall have two l years of specialized training in health physics or equivalent and three years work experience related to l radiological health and safety.

6.1.4 '8 Justodial Agents Custodial Agents shall have unescorted access to the Protected Area to perform activities at Fermi 1.  !

Custodial Agents shall be authorized in writing by the Fermi 1 Custodian or Custodial Delegates.

l 6.1.4.4 Health 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. Qualifications are addressed in Section 6.2.2.

6.1.5 Title Changes If personnel titles change without a change in Fermi i responsibilities, the Fermi 1 Safety Analysis Report revision may await the next planned review and update.

6.2 Quality Assurance Program 6.2.1 Introduction The purpose of the Fermi 1 Quality Assurance Program is to provide assurance that work is performed at the Fermi 1 nuclear facility in a quality manner. The Quality Assurance program provides such assurance through compliance with implementing documents and assurance that such documents are adequate, reviewed, and used. Adherence is required by all personnel working on the nuclear portion of the facility. Audits are performed ofimplementation to assure adherence to the program.

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i The program meets the requirement of 10 CFR 50.54(a). It has been established for the SAFSTOR l condition, based on the condition, status, and history of the facility. Prior to the program's inception, l there were administrative controls and review requirements established, but not a Quality Assurance Program. The Quality Assurance Program applies to work in the nuclear portion of the Fermi 1 facility, )

which is the portion inside the Protected Area. The area curesponds to the description of the facility in l Fermi 1 Technical Specifications. Note that this Quality Assurance Program does not implement 1 Appendix B of 10 CFR 50. There are no systems or components at Fermi i that are used to prevent or I mitigate the consequences of postulated accidents that could cause undue risk to the nealth and safety of I the public. Special attention is focused on assuring radiological surveys are performed by qualified j personnel using calibrated instrumentation and that radiological calculations performed for decision j making are reviewed. l l

The Fermi 1 Quality Assurance program will be revised in accordance with 10 CFR 50.54(a) Changes that reduce commitments in the program description previously accepted by the NRC will be submitted to the NRC and receive NRC approval prior to implementation.

If personnel titles change in the Quality Assurance program without a change in Fermi 1 responsibilities,  ;

the Quality Assurance program revision may await the next planned review and update. I 6.2.2 Organization and Qualifications The Senior Vice President, Nuclear Generation is responsible for Detroit Edison's nuclear power facilities. The Fermi 1 Custodian is responsible for the Fermi 1 facility. Detail o,. organization structure, responsibilities, and qualifications are discussed in Section 6.1. Appointed Custodial Delegates can fulfill the responsibilities of the Fermi 1 Custodian. All 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.

The Fermi 1 Health Physicist is responsible for reviewing procedures and limits involving the handling of radioactive materials at Fermi 1.

Radiation Protection (Health Physics) technicians at Fermi 1 are qualified in accordance with the Fermi 2 Radiation Protection Program for the tasks they perform. Such training and qualification is documented.

Personnel shipping radioactive waste and/or hazardous waste receive appropriate training. Personnel performing activities which per regulations require certification, will be qualified for such activities and such training and qualification documented.

The Fermi 1 Review Committee functions to advise the Fermi 1 Custodian on matters relating to nuclear safety. The appointment of the Chairman of the Fermi 1 Review Committee is approved by the Senior Vice President, Nuclear Generation. This provides independence from the Fermi 1 Custodian. The Fermi 1 Review Committee is addressed in Section 6.3. Audit team leaders perform audits under the

. auspices of the Fermi 1 Review Committee.

6.2.3 Procedures Procedures ensure that the requirements of the Technical Specifications are carried out in a proper and j timely manner. They also serve as training and reference units for future Fermi 1 CmuJians. Custodial l Delegates and Custodial Agents. Administrative procedures include Custodial ecporMoilities and authority, Procedure Manual control, Custodial Delegate and Custodial Agent seheum. and function, 6-3 Rev 0,8/98 l

l l l 4 reporting procedures and Fermi 1 Review Committee functions. In addition, there are appropriate technical procedures for details ofinspections, surveillances and operation.

l Per Fermi 1 Technical Specifications, procedures are required to be prepared and utilized for radiation ,

control, maintaining cover gas supply, and facility inspections. Fermi i procedures state that the Fermi 2 Radiation Protection Program is applicable to Fermi 1.

i

- Fermi i procedures are reviewed and approved by the Fermi 1 Review Committee. The Fermi 1

[ l Custodian may temporarily change a procedure by written document following a determination that the change does not constitute a significant increase in hazards associated with the facility.

The Fermi 2 Radiation Protection Program and Environmental Prograin apply to the entire site, including activities at Fermi 1. Procedures to implement these programs are typically controlled via the Fermi 2 procedure program, Some items specifically applicable to Fermi 1 are included in the Fermi I procedure i manual.

6.2.4 Design Control Design change documents are used to modify installed Fermi I systems in the Protected Area. l Modifications 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 information on the configuration of the plant.

l Additionally, the Fermi 1 Safety Analysis Report will be updated at least once per 24 months to provide l a current integrated description of the facility.

l l Design control and level of detail in design change documents are commensurate with the potential I

impact on quality and health and safety of the public. Design change documents are reviewed by the I Fermi 1 Custodian, an engineer, and Fermi 1 Health Physicist, or their delegates, and approved by the Fermi 1 Review Committee.

l 6.2.5 Work Control Repair, maintenance and modification of installed systems, structures and components shall be performed using a work control document. Approval to start work shall be noted on the document by the

~

Fermi 1 Custodian or delegate. Any post-maintenance testing, inspection, and Technical Specification l L out-of-service requirements applicable to the work activity shall be specified on the work control L

document.

Hold points may be inserted into a work control document which require an independent check of data or performance of an inspection or activity by individuals specifically qualified for the activity or ,

inspection. Work cannot proceed beyond a hold point unless the hold point activity is performed or the individual responsible for performing the hold point waives the hold point.

6.2.6 Document Control The Fermi 1 Manual has been issued as a controlled document to a specified users list. Procedures for use are obtained by copying from one of the controlled manuals. When a procedure is revised, work in progress will be evaluated for impact and workers provided with an updated procedure, if needed.

6-4 Rev 0,8/98 l

- - - - - - . - - . - -. - . ... - - - = _ _.- - - .- . . -

t

. Copies of records are obtained from the Fermi 2 records management organization.

A change control process has been established which requires specific reviews for adequacy and approvals of change 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. Procedures are reviewed by the Fermi 1 Health Physicist, Fermi 1 Custodian, and the Fermi 1 Review Committee. Typically, the Fermi 1 Review Committee approves the change documents.

[

. In the future, as design changes are implemented, an appropriate drawing (s), ~as applicable and available, will be marked up and maintained in the Fermi 1 records files. Sketches may be used as an alternate.

6.2.7 Special Processes Welding, heat treating, and non-destructive testing ofinservice systems will be controlled and performed by qualified personnel in accordance with applicable standards. Non-destructive testing performed for l information only purposes does not fall within the controls of this Quality Assurance Program, nor does cutting using welding equipment.

e 6.2.8 Measuring and Test Equipment Measuring and test equipment (M & TE) used to perform Technical Specification surveillances shall be calibrated periodically. Instruments used to obtain radiation survey data or laboratory analysis for decision making or surveillance purposes shall be calibrated in accordance with manufacturer's recommended frequency or evaluation of past performance, but at least annually. If an instrument is found out of calibration, an evaluation shall be performed to determine the validity of previously made measurements. The M & TE found out of calibration will be clearly identified with a tag or other appropriate means and segregated from the calibrated equipment until recalibrated and properly tagged.

Computer programs used to process radiological data for final site survey purposes shall be verified and validated. Hand calculations for such purposes shall be reviewed.

6.2.9 Records Records required by Fermi 1 Technical Specifications are maintained as Quality Assurance records for at least five years, or duration of the license, as specified in the Technical Specifications. These records 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 concerning areas and locations in which radioactive materials were stored and used were substituted. This alternate information currently includes a special list of where radioactive material was used, some drawings, and current design changes after implementation. Audit reports are also maintained as Quality Assurance records for at least five years.

The Quality Assurance records are maintained in Fermi 2 facilities for records and are retrievable. l l ':

lL U

L l

6-5 Rev 0,8/98 l

I l

l 6.2.10 Corrective Action l

Significant conditions adverse to quality shall be documented, a cause analysis performed, and corrective action to prevent recurrence implemented. Conditions meeting the Technical Specification reportability criteria shall be considered significant conditions adverse to quality.

6.2.11 Review The Fermi 1 Review Committee reviews performance at the Fermi 1 facility. Their reviews include l procedures, design changes, license amendments, and 10 CFR 50.59 safety evaluations. See Section 6.3 l l for further discussion on the Fermi 1 Review Committee.

Audits of facility activities and Quality Assurance Program adherence will be performed at least two times per year. The purpose of the audits is to assess implementation of and verify compliance with the Quality Assurance Program and the Technical Specifications. Implementation of radiation protection control activities shall be covered at least once per year and will include radiation surveys and instrument calibration.

The individual or team performing the audit will be independent of line responsibility for Fermi 1.

Review responsibilities (e.g., a Fermi 1 Review Committee member) are not considered line responsibilities for determination ofindependence. The audit team leader will be a certified audit team leader / lead auditor (per ANSI N45.2.23, NQA-1, or ISO-9000), or have at least two years Quality Assurance audit or management experience.

The audit team leader possesses stop work authority that can be used if the audit team observes an unsafe act or unsatisfactory work. The Fermi 1 Custodian is responsible for addressing discrepancies identified by audits and reporting to the Fermi 1 Review Committee actions taken to resolve discrepancies. If any discrepancies are disputed and cannot be resolved between the Fermi 1 Custodian and auditor, the issue and proposed solution shall be presented to the Fermi 1 Review Committee for resolution.

Significant conditions adverse to quality identified during an audit shall be processed via the corrective action program.

The audit scope and results shall be documented and presented to the Fermi 1 Review Committee and Fermi ! Custodian.

Vendors providing 10 CFR 61 analysis services for Fermi i shall be evaluated at least triennially if such services have been provided. The evaluation may be by review of an audit covering similar services for another licensee.

Self-assessments may be used to assess adherence to regulations and management expectations for selected activities. Personnel performing self-assessments will be knowledgeable in the activities they are assessing.

,- 6.3 Fermi 1 Review Committee l The Fermi 1 Review Committee shall function to advise the Fermi 1 Custodian on all matters relating to l nuclear safety.

I 6-6 Rev 0,8/98 l

l The Fenni 1 Review Committee shall be responsible for review and approval of the following: l

• Procedures for activities at the facility l

'* Annual report to the NRC e Licensee Event Reports .

• Design changes

. . 10 CFR 50.59 Safety Evaluations e License Amendments, including Technical Specification changes e~ Facility monitoring results The Fermi 1 Review Committee meets et intervals not exceeding 13 months and prepares and distributes formal minutes ofits meetings. Special meetings may be called by the Chairman, Fermi 1 Custodian, or i one of the Custodial Delegates, as required.

The Fermi 1 Review Committee is composed of five or more personnel from within the Detroit Edison l l ' organization or consultants, at least three of whom have had two years or more of experience in a responsible position at an operating nuclear power facility and have had basic health physics training.

6.4 Radiation Protection Program The Fermi 1 radiation protection practices are based on the Fermi 2 Radiation Protection Program which contains the procedures, practices, and training needed for an operating reactor. Radiation protection activities at Fermi 1 are covered by Fermi i specific procedures or applicable Fermi 2 procedures. l l Radiation protection technicians used at Fermi 1 meet the Fermi 2 qualification and training program requirements for tasks they perform. l-l i

e l

6-7 Rev 0,8/98 l

Personnel dose resulting from decontamination and decommissioning activities will depend on future factors not presently available; i.e., availability oflow level burial sites, adoption of de minimus levels, development of new dismantlement and decontamination techniques and acceptance in burial sites of material that has been contaminated with sodium. All of these factors, if realized, along with normal decay will provide for lower personnel exposure 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 ifimmediate dismantlement was performed.

83.3 Releases of Radioactive Effluents All releases of radioactive effluents will be made in accordance with the Fermi 1 Technical Specifications. The amount of such releases is inherently limited since no addition.il radioactive material is being produced at Fermi 1 and the amount which can be received onsite is limited by the Fermi 1 license.

While the liquid and gaseous release systems used during plant operation are no longer functional, provisions will be implemented in accordance with the Technical Specifications prior to initiating radioactive effluent releases. Releases will not exceed the limits in 10 CFR 20, Appendix B, Table 2 on an instantaneous basis. Monitoring or sampling of radioactive effluents will be conducted at a location that will enable determination of an actual or conservative radioactive effluent release rate.

83.4 Shipment of Radioactive Materials Shipments of radioactive materials will be made in accordance with applicable regulations and facility procedures. Waste shipments may be made to intermediate processors or directly to a disposal site.

8.4 Postulated Radiological Accidents There are three main postulated radiological accidents that could occur during SAFSTOR. These are described in the following sections. Additionally, a hypothetical secondary sodium scenario is addressed which could release tritium or potentially lead to the liquid waste or sodium t: leases, though it is highly unlikely.

8.4.1 Liquid Releases It is assumed that two liquid waste tanks in the Fuel and Repair Building rupture. For the analysis, it is assumed that the tanks contain a total of 7550 gallons of radioactive radwaste containing 6 mci of 60Co and 6 mci of 137 Cs. This assumption is based on 1986 activity.

Scenario A: Airborne Release Assumptions:

Tanks rupture / malfunction and radioactive inventory is spilled on floor.

. 25% of the inventory is assumed to be released through a vent to the environment.

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 an individual is exposed for the entire time at the exclusion area boundary (EAB).

• Dose factors from Regulatory Guide 1.109 and ICRP Publication 30 are used.

8-4 Rev 0, 8/98 l I. .

l 1

Assumptions (Continued):

l

l. e X/Q = 1.55 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 f

! percentile value at Fermi 2 Exclusion Area Boundary (EAB) of 915 meters NW. This l' I - distance is conservative since the Fermi 1 EAB is approximately 1211 meters NW. l j e Maximum Permissible Concentration (MPC) from 10 CFR 20, Appendix B, Table 11. l

.(Values used are based on regulations in existence in 1986)

]

l Results: 1 l Liouid Water Tank Source - Airbome Release l Concentration Ci/ml  !

Nuclide In Tank At EAB MPC (air) C/MPC*  !

60Co 2.10 E-4 3.23 E-12 1 E-8 3.23 E-4 137Cs 2.10 E-4 3.23 E-12 2 E-9 L62 E-3 l

. *C/MPC = ratio of EAB concentration to MPC l 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Dose at EAB. mrem I Adult Child l Whole Body 3.27 E-4 1.11 E-4 l L Lung

• 4.46 E-3 5.29 E-3 i l

-

• Lung is most critical organ. l ]

L 'A conservative evaluation was performed of the potential maximum offsite' dose due to this event )

L considering the additional radionuclides identified in one or more smear or sample analysis from the l

liquid waste system from 1989 - 1998. These radionuclides are addressed in Section 4.1.1. The conservatively estimated maximum dose was 0.17 mrem committed effective dose equivalent (CEDE) at l- . the EAB. The methodology used differed in that values from the current 10 CFR 20 were used.

i Scenario B: Liouid Release to Lake Erie l

~ Assumptions:

L e .Since the liquid radwaste tanks are located in the basement of the FARB, minor cracking of

- the structure could occur in the event of an earthquake. The tanks could undergo stress ' -

cracking 'and leaking to allow fluid flow between the interior of the structure and the ]'

surrounding earth. 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 ,

! ground water and the water level inside the structure would rise until it reached the elevation )

of the piezometric level..of the ground water. At that time the radioactive liquid may be i diluted by as much as 10:1; however, no credit is taken for dilution via the influx of water.

• Tanks are approximately 450 ft. from the Lake Erie shoreline.

l.

• Flow rate within the aquifer is 0.24 ft/ day.-

l

• Delay time in traveling from the tank to Lake Erie is 1875 days plus 40 days to move upward -

L through till and lake bottom sedunents, (Fermi 2 UFSAR, Section 15.7.3.2). l h e Dilution factor of 77 at Monroe City Water intake (Fermi 2 UFSAR, Appendix 11 A). l L

>e Radioactivity decay with delay time is assumed,

i. . 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.

- * - Dose factors from Regulatory Guide 1.109.

.* MPC from 10 CFR 20, Appendix B, Table II.  ;

i c

8-5 Rev 0,8/98 l l~ gp w.

g .+-w9 .y y g ye pa.s-- w -w--y 7 - ge.- , --' _ .q. --+- ,- *

• Jm'1-

. - _. - . - - -. . .- -. - .~ .- . , - . . . -, .. . - ~ _ - .

l i'

L

. Results:

Liouid Waste Tank Source - Monroe City Water Intake  ;

_ Concentration, pCi/ml '

Nuchde In Tank Entering 1&ht At Intake MPC (water) C/MPC*

60Co 2.10 E-4 1.04 E-4 1.35 E-6 5.00 E-5 0.03 '

137C s~- 2.10 E-4 1.86 E-4 2.41 E 2.00 E-5 0.12 i 4

l *C/MPC = ratio of concentration at intake to MPC.

L Innestion Dose. mrer- t Water Fish Invertebrate Iotal

. Adult Whole Body 0.36 29.90 2.38 32.64

. Adult, Liver 0.53 30.30 3.62 34.45 Child, Whole Body 0.19 4.24 0.54 4.97 i Child, Bone 1.10 29.90 3.68 34.68 A conservative evaluation was performed of the potential maximum water ingestion whole body dose due to this event considering the additional radionuclides identified in one or more smear or sample analysis from the liquid waste system from 1989 - 1998. These radionuclides are addressed in Section j 4.1.1. The conservatively estimated maximum dose was 93 mrem CEDE. The methodology used differed in that values from the current 10 CFR 20 were used.

8.4.2 ^ Airborne Releases from Sodium It i.s assumed that a fire or other catastrophic event results in the release to the environment of the l residual primary sodium including the entire radionuclide inventory which contains a total of 0.98 mci 22Na and 4.84 mci 137Cs and possibly 70 mci of tritium (Section 4.1). l .;

l . Assumptions:

e 100% ofinventory becomes airbome.

)

• 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.

• ' Dose factors from Regulatory Guide 1.109 and ICRP I'ublication 30.

e X/Q = 1.55 E-5 sec/m3, .

l

• MPC from 10 CFR 20, Appendix B, Table II.

• For tritium,~ 10 CFR 20, Appendix B, Table 2, Col. I was used for maximum effluent concentration.

Results: 1 Airborne Release Concentration pCi/ml Nuclide - At EAB MPC (air) C/MPC* )

l 22Na 2.11 E-12 6 E-9 3.52 E-4 137Cs 1.04 E-11 2 E-9 5.21 E-3 l 3 1-1 1.51 E-10 Max. Effluent Cone. C/ Max. Effluent l 1 E-7 Conc.

1.5 E-3

, *C/MPC = ratio of EAB concentration to MPC

[ Adult Dose. mrem i Whole Body 1.3 E-3 .

Liver

• 1.5 E-3 l

!->-

• Liver is most critical organ.

8-6 Rev 0,8/98 l ,

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. ~ - , _. - -. - .

8.4.3 Discussion i Both Scenario A and B and the releases from the residual sodium result in concentration levels that are well below the MPC values in 10 CFR 20, Appendix B, Table II for releases to unrestricted areas.

The doses associated with Scenario B are below the limits at which precautionary measures would be taken for an accident-type release. The doses associated with the fish and invertebrates are the result of the concentration factors and the models in Regulatory Guide 1.109. In this Scenario, the radioactive liquid is released to the aquifer and groundwater. The results are extremely conservative since no credit

. was assumed for:

e Dilution from the initial influx of water into the F ARB.

. Removal of suspended particulates by filtering action of the soil and clay. l'

• Removal ofionic forms through adsorption by the soil and clay.

The 1875 day travel time to the shoreline of Lake Erie provides ample time to sink wells, follow the

- progress to the lake, and take remedial action should it become necessary.

8.4.4 Hypothetical Secondary Sodium Scenario ,

l There are three 12,000 gallon secondary sodium storage tanks located in three separate rooms in the l Fermi 1 Steam Generator Building Basement. The Steam Generator Building is located outside of the  ;

Fermi 1 Protected Area. Due to the siphoning method used for draining the tanks, there is a maximum l estimated 210 gallons of solid sodium remaining in the three tanks (approximately 34 - 70 gallons per l 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 CO2 on January 9,1974 in an attempt to passivate the residual sodium deposits by converting the sodium to sodium carbonate. Based on visual inspection of j the tanks during 1997, the residual sodium appears to be unreacted.

. 8.4.4.1 Sodium-Water Reaction / Hydrogen Reaction I When sodium (Na) reacts with water (H 2O), the exothermic process produces caustic sodium hydroxide (NaOH), hydrogen (H2), and heat according to the following reaction:

2Na + 2H2 O :o 2NaOH + H2 + heat The presence of water given a tank breach is plausible since the water table elevation at Fermi is approximately 575 feet, and the top of the secondary sodium storage tanks is at approxi_mately 570 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 reduced inleakage. l This postulated scenario is highly unlikely due to the structural integrity of the tanks, layout of the tank bays and equipment, and the mechanism of the reaction. Sodium is a solid at room temperature and would therefore not drip out of a hole or crack in a tank. A more realistic scenario would involve water being introduced 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 and the small hole area. The hydrogen which escapes the tank would l dissipate in the larger areas of the building. The worst case hypothetical scenario is evaluated here involving instantaneous reaction of all the sodium in a secondary sodium storage tank.

8-7 Rev 0,8/98 l

l L Assumptions:

.* Sufficient water is present te completely react the 70 gallons of Na estimated to be in the tank.

• Total and complete Na-H O2 reaction (i.e., maximum amount of hydrogen and caustic L generated).

• - The reaction (i.e., hydrogen and caustic formation) occurs instantaneously.

• The generated hydrogen concentrates within the specifi,c tank bay.

• The air to hydrogen ratio is conducive for complete combustion. i e The center of combustion is at the center of the respective Na storage tank.

• All of the hydrogen combusts instantaneously (i.e., no partial burn during hydrogen generation).

• Hydrogen combustion efficiency of 30%.

I 8.4.4.2 Structural Hazard i The 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 would result in failure of some Steam Generator Building structural '

members. _ However, since the building is outside of the Protected Area, the only radiological release i from building failure itself would be from the tritium in the secondary sodium. The Steam Generator-Building rests on bedrock, with the outer walls made of reinforced concrete. 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 Building grade (582'). A larger radiological release might l l potentially occur if the ground shock were of sufficient amplitude and frequency to cause failure of equipment and structures outside of the Steam Generator Building which contain radioactive

! contamination, such as the primary system in the Reactor Building or primary sodium storage tanks, but i this is highly unlikely.

Assumptions:

Detonation Waves -

• Any leakage pressure from the tank bay into'the upper levels of the Steam Generator Building will result in blowout of the transite siding prior to failure of the concrete walls, e 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 Release Hazard The generation of hydrogen and NaOH, with the potential subsequent explosion of the hydrogen, could provide a means for sending the NaOH airbome. With the postulated Steam Generator Building

structural failure or transite siding blowout, a vent path of the caustic NaOH from the Steam Generator ,

l Building to the atmosphere is available. The caustic plume was evaluated for the onsite (i.e., Fermi 2 I operations) and offsite impact.

i l

=

s 8-8 Rev 0,8/98 l l' _ .. . - - - . . . - - .-- -. - -- - -- - - - - - - - - --

Assumptions:

. . The NaOli is not diluted by any water in the room, e Much of the NaOH remains in the Steam Generator Building, with 40% escaping to the environment, which is a conservatively high percentage based on technical reports. ,

e The release is at ground level.-

• The NaOH reacts rapidly with the CO2 in the air. By the time the cloud reaches the vicinity of I

the Fermi 2 control room intake, some or altsost all the NaOH has been converted to sodium carbonate, depending on the assumed wind speed. For the evaluation, the fraction not converted to sodium carbonate by the time the leading portion of the plume reached the intake was used.

. The puff travels directly towards the Fermi 2 south control room air intake, though this is not the prevailing wind direction.

• The puff does diffuse.

e The NaOH climbs during dispersion to reach the control room intake, even though NaOH is heavier than air.

. . The highest concentration at the control room intake is when the cloud has traveled the distance to the intake.

. Since the release is instantaneous, the puffis only briefly at the intake.

• - It is assumed that the control room normal ventilation remains in service, and that the control room is not manually isolated based on an explosion at Fermi 1.

. NaOH is dispersed by and deposited on buildings in the pathway between Fermi 1 and the Fermi l 2 control room, decreasing the amount of NaOH in the plume.

8.4.4.4 Results Radiological Hazard Conseauences The radiological release resulting from the initiating explosion could consist of the tritium contained in ,

the secondary sodium residuals in the tanks. The maximum dose to an offsite member of the public would be 'l.4 E-6 mrem, assuming the tritium content of the sodium residuals in the tank was IE-2 pCi/g. Any subsequent failure of the Steam Generator Building or the ' production of fragments due to the explosion could release tritium remaining in other portions of the secondary sodium system, including the cold traps. The maxim.im dose to an offsite member of the public could be 3E-4 mrem.

The event is not expected to impact other systems and/or equipment which may contain radioactive contamination. Ifit does, the effects would be bounded by the following analysis.

The effect of the detonation 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 building r.nd system response. The assumption was made, though considered unlikely, that the ground shock could lead to damage of the primary system, primary sodium storage tanks, other primary sodium containing tanks or piping, and/or liquid 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 concern at Fermi 1 would be with the primary system or primary Na containing tanks or tiiping or

. liquid waste tanks being ruptured by the blast or ground shock, leading to the release of radioactively

. contaminated Na and/or waste water. Section 8.4 already addresses the radiological effects of a total

' release of the primary Na inventory and ofliquid waste tanks 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 evaluated the results of an explosion of the 20,000 gallon liquid hydrogen tanks located near the natural draft cooling towers is considered to bound the effects due to the smaller hydrogen explosion postulated in the Fermi 1 Steam Generator Building and similar distance to safety related equipment. The same conclusion applies if all three secondary sodium storage tanks are affected by the event.

8-9 Rev 0,8/98 l

1

. Chemical llazard Conseauences i

A previously prepared University of Michigan study (Reference 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 seconds that the NaOII cloud passed would just be at the respiratory l 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 in 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 maximum NaOII concentration in the Fermi 2 control room due to the reaction of 70 gallons of Na and the subsequent release of the generated NaOII is less than 0.002 g/m3, the total sodium inventory of all three secondary sodium storage tanks at Fermi I (~ 210 gallons) were w react and form airborne NaOII, the concentration in 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 NaOII concentration could be slightly above the 0.002 g/m3limit. If credit were taken for losses and dispersion due to additional buildings and surfaces in the plume pathway, the control room concentration would be expected to be less than the 0.002 g/m3 Threshold Limit Value even following a three-tank scenario.

Conclusion l

The worst case radiological consequences of the postulated hypothetical secondary sodium scenario are similar to the liquid waste tank release and sodium release events analyzed in Section 8.4.1 through 8.4.3. The possible tritium release from the postulated secondary sodium scenario insignificantly increases the total maximum offsite dose to a member of the public. The postulated secondary sodium scenario does not cause damage at Fermi 2 that could lead to greater radiological consequences.

8.5 Alternatives Considered Once a nuclear facility has reached the end ofits 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 alternatives are available: DECON, ENTOMB, and SAFSTOR. The no action alternative is not viable for Fermi 1 since it is already in a decommissioned state. The three alternatives are discussed below.

8.5.1 DECON DECON is defined as immediately 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 decommissioning alternatives which leads to termination of the facility license and release of the facility and site for unrestricted use shortly after cessation of facility operations. DECON would involve the removal or decontamination of all equipment, structures, and those portions of the facility containing radioactivity. Although the fuel has be removed from the Fermi i site, the reactor vessel, its internals and most of the sodium gng remain.

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Clearing the Fcrmi 1 site for unrestricted use is oflittle environmental value since Fermi i lies within the site baundary of Fermi 2 and could not be used for other purposes. A major effort would be involved in the complete removal 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

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various piping and equipment and shipment in commercially available licensed shipping casks to an offsite licensed burial site. This was considered undesirable due to personnel exposure to additional

radioactivity.

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' This alternative is ~not considered preferable since little or no improvement would be realized in )

personnel exposure, land use, aesthetics, or value.

1 8.5.2 ENTOMB

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' ENTOMB means to encase and maintain property in a strong and structurally long-lived material (e.g.,

3 concrete) to assure retention until radioactivity decays to a level acceptable for releasing the facility for I unrestricted use. ENTOMB is intended for use where the residual radioactivity will decay to levels ]

permitting unrestricted release of the facility within a reasonable time period of continued structural integrity of.the entombing structure; approximately 100 years is considered to be consistent with reconunended EPA policy on institutional control reliance for radioactivity containment .

Primary considerations for retiring the reactor and primary system were: (1) removal of all core and 1 blanket fuel, (2) removal of sodium, (3) gastight seal of the primary system, and (4) passivation and maintenance of the entire primary sodiur'n system with carbon dioxide. The reactor vessel was sealed within the primary shield tank and the outlying components were scaled directly, using the Reactor Building as an isolation structure against personnel access to the primary system.

The primary system was filled with~ nitrogen to which CO2was added to reduce the residual sodium deposits to inactive solids. The system was then sealed and maintained at slightly positive inert gas pressure to prevent the entrance of water or moisture and to minimize dispersal of any remaining radioactive material.

To ENTOMB the Fermi 1 facility 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 SAFSTOR SAFSTOR is defined 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 ,

- subsequently decontaminated to levels which permit release of the facility for unrestricted use.

SAFSTOR consists of a short period of preparation for safe storage, a variable safe storage period of

- continuing care consisting of security, suiveillance, and maintenance and a short period of final decontamination (DECON).

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' Fermi 1 was decommissioned according to the NRC (AEC) rules and directives in effect at the time and l was considered as being left in a decommissioned state. In accordance with present definitions, it is in a SAFSTOR condition and a Preliminary Final Decommissioning Plan for Fermi 1 was prepared and i- submitted to the NRC in Reference 16.

The purpose of the Preliminary Final Decommissioning Plan for Fermi 1 was to state Detroit Edison's intentions regarding the ultimate decommissioning of Fermi 1. Detroit Edison is planning on a

!' SAFSTOR period to be followed by a final decommissioning to achieve a radiologically releasable site l and termination of the Nuclear Regulatory Commission (NRC) license, it is expected that final decommissioning for the Fermi I plant will be performed concurrently with that for Fermi 2.

The Preliminary Final Decommissioning Plan is strictly an overview of the final decommissioning effort.

Prior to commencement of the final decommissioning effort, an updated Post Shutdown .

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- Decommissioning Activities Report (or equivalent) will be prepared and submitted to the NRC in accordance with 10 CFR 50.82.

l- 8.6 Conclusions l'

'SAFSTOR'was determined to be the most viable decommissioning alternative for Fermi 1. DECON l would result in little improvement over SAFSTOR, and ENTOMB is not a viable choice because of the i presence oflong lived radioisotopes.

Retaining Fermi 1 in a SAFSTOR status for a 40 year period (1985-2025) will result in the following:

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.- Reduction in dose rate of more than 90%.

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• Reduction in personnel exposure at the time of final action.

L e Reduction in volume of radioactive wastes at time of final action.

L e Potential increased a_vailability of repository sites for radioactive materials.

j. *- Continued compatibility with the long-term use of the Fermi 2 site since Fermi 1 buildings are being

< used for Fermi 2 activities.

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• Nominal expense and impact on the community because use of Fermi 2 personnel' for Fermi 1 l- surveillance activities provides readily available manpower resources.

l-e Integration of Fermi 1 into the Fermi 2 decommissioning program.

• Continued minimization of the risk to the health and safety of the public.

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