ML20003E294
| ML20003E294 | |
| Person / Time | |
|---|---|
| Site: | 07001308 |
| Issue date: | 01/31/1981 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20003E291 | List: |
| References | |
| NEDO-21326C-RC3, NUDOCS 8104020788 | |
| Download: ML20003E294 (82) | |
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g ATTACHMENT G PROPOSED REVISICN C3, NEDO-21326C CONSOLIDATED SAFETY ANALYSIS REPORT FOR MORRIS OPERATION The material in this attachment consists of proposed revisions and additions to NEDO-21326 made in compliance with requirements of 10 CFR 72.
Revised and new =aterial is indicated by double-bar vertical lines in the right margins. Some incidental editorial changes and new material is included when they occurred en the same page as Part 72 revisions, or were otherwise related. These changes are indicated by a single bar and the letters E (Editorial) or N (New Material).
1 See the attached letter for other commento applicable to this material.
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NEDO-21326C3 Jtnurry 1981 In Dece=ber 1975, General Electric received a license a:end=ent to incre the o
fuel storage capacity of the facility frem about 100 TeU3 to 750 TeU,.ne V
installation of fuel s crage syste= of a new design and through appropriate changes in fuel handling and suppcrt systems.
~his project, designated by GE as Morris C;eration-?roject I, ccnverted the for:er high level waste storage basin to a fuel storage basin.
Se capacity expansion project us co=pleted in 1976 1.1.1 Corporate Entities, Business, and Excerience l
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The facilities described in this report are owned and operated by General Elec-tric Company, a ccrpcration under the laws of the State of New York, with its principal place of business at Schenectady, New Ycrk.
2e facility is operated through General Electric's Nuclear Fuel and Services Division (NFiSD), with head-
{E quarters at San Jose, California and operatiens at Morris, Illinois.
General Electric is a broadly diversified ccepcration involved in research, design,
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=anufacturing, and marketing products and services in several fields including industrial products, technical syste=s and =aterials, consu=er products, and O) pcwer syste=s.
2.e latter activity includes nuclear systems, equip =ent, V
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l' and services.
t General Electric's experience in nuclear activities includes research and develop-
=ent of prototype reacters for nuclear sub=arines, operation of the govern =ent's Hanford facilities for =cre than 17 years and the develop =ent, design, manufacture, and erecticn of 41 boiling water reactors currently operating at electric pcwer statiens in the l'nited States and throughcut the world.
2e staff of the C7:pany's Nuclear Energy Group includes literally thousands of scientists, engineers, and techniciars, representing cne of the largest pools of nuclear knowledge and experi-DQ ence in the world.
1.1.2 Plant Location Morris Operation faciities are located on the northern end of a rectangular tract of about 815 acres cwned by General Electric Co= par.f. in Gocaelake Township, Grundy County, Illinois, near the ecnfluence of the Kankakee ard Des Plaines rivers (Figure 1-1).4 v
1-2
NEDO-21326C3 January 1981 The tract (Figure 1-2) is about 15 air miles southwest of Joliet and about 50 miles southwest of the Chicago, Illinois - Gary, Indiana area. Morris, Illinois, the county seat of Grundy County, is about 7 miles to the west of the tract.
The Illinois 'daterway and Kankakee River are separated from the tract to the north and east by lands cwned by Com=onwealth Edison Co., the site of the Dresden Nuclear Power Station (DNPS) and related facilities, and a privately owned plot of about 50 acres. The developing Gooselake Prairie State Park is to the west and a refractory mining operation borders the tract to the south.
O The terms used in the text of this document to describe the General Electric property are as follows:
Tract / Controlled Area a - all land holdings of General Electric as defined I
in Section 3 Site - the developed area of the General Electric tract, including the protected area, sanitary lagoons, and evaporation pond.
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NEDO-21326c3 Jcnuary 1981 1.1 3 Existing Facilities i
O The existing facilities occupy about 52 acres at the north edge of the tract I
(Figure 1-3).
The principal plant structures, including the ventilation stack, are within a 15-acre fenced protected area, while the sanitary waste treatment facilities and the industrial waste evaporation pond are located i= mediately south of the protected area.
The sanitary waste facilities are fenced, also, l
but not as part of the protected area.
The evaporation pond is not fenced.
O 1.1.4 Fuel T70e and Exposure The design basis fuel to be stored is UO 2 fuel having had an initial enrichment of 55 U-235 or less, with stainless steel, ::irconium or Zirealoy cladding, and i
in a " bundle of rods" geometry. The design basis fuel may have been irradiated
.sc apecific power levels of up to 40 kW/kgU, with exposure to 44,000 mwd /TeU (reactor discharge batch average), arxi must be cooled for at least 1 year after reactor shutdewn and prior to receipt at Morris Operation.
O iia 2 ru 1 1= ster==-
Irradiated fuel from PWR's and BWR's has been received and stored at the Morris Operation facilities since 1972.5 These activities have reaffir=ed experience elsewhere that fuel can be handled and stored safely with no impact on the envi-ronment.
There has been no significant fuel leakage (as determined by measurement of basin water activitr), indicating that the fuel is a stable, inert material while in the storage bcsin environment.
Effective control of water quality, radioactive material concentration in the water, cask contamination, and airborne radioactive material has been demonstrated.
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1.2 GENERAL PLANT DESCRIPTION The following descriptions are of those aspects of the Morris Operation facili-I ties that are related to irradiated fuel storage or shipment.
Facilities origi-nally intended for reprocessing are mentioned only as related to fuel stortge operations.
1-5
NEDO-21326 C3 January 1981 1.2.1.5 Envirens Summary The distances from the plant stack to the tract (controlled area) boundaries are l!
2265 ft to the east, 6512 ft to the south and 3100 ft to the west.
The tract boundary to the north is about 950 ft fec= the stack; however, tne DNPS site provides an effective exclusion distance 7 of about 5950 ft.
Studies of population and land usage in surrounding areas were made and reported in the course of DNPS development, as well as during the MFRP prcgram and Morris Operation Project g
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Factors of specific interest are su==arized below.
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Industrial:
On the DNPS site there are three nuclear power reacters situated about 0.7 =11e northeast of the Morris Operation stack loca-tien. A large fossil-fired power plant is located about t miles west-southwest of the stack. A chemical plant is located about 1.5 ziles from the stack to the northwest. Adjacent to the scuth boundary of the Morris Operation tract there are clay mining and clay products =anu-facturing activities about 1.t miles frc= the stack.
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Residential:
The residences nearest to the tract are on about 50 acres directly east of the facilities, between General Electric pecperty and the Kankakee River. The tract owner, who =aintains his permanent resi-dence there (about 0.5 mile fec= the stack), has leased individual river fron sites on which approxi=ately 30 cottages have been b' tilt, largely for recreaticr.al purposes. There are other residences acrces the Kankakee River, the nearest about 0.7 =ile from the stack.
The total population within a 5-mile radius is esti=ated to be about 5000 including summer visitors, increasing to about 8830 by the year
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2000. A population of about 32,400 resides within a 10-m11e radius of the plant, and should increase to about 68,000 by the year 2000.
N-Population in the 5-to 20-mile radius zone, wnich includes the cities of Aurora and Joliet, is about 252,900. This populatien should increase to about 432,500 by the year 2000.
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NEDO-21326C3 Jcnucry 1981 In general, population projections for the State of Illinois have been
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lowered in recent years.
Current projr.ctions indicate a relatively N
slow growth rate as compared to the over-all U.S. rate.
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Recreational:
In addition to fishing, hunting, and boating accivities near the confluence of the Kankakee and Des Plaines Rivers 1 to 2 miles east of the plant, the Goose Lake Prairie State Park has been established adjacent to the Morris Operational tract.
This natural prairie preserve
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of about 1800 acres is west of the tract, with the nearest point being about 0.6 tr'.le from the stack.
1.2.1.6 Tract Ownership The tract is wholly owned by General Electric Company. Since purchase of the original tract, which then totalled 1380 acres, approximately 70 acres located at the southwest corner and approximately 50 acres in a 400-ft-wide strip along the south edge of the tract have been sold to
- a. P. Green Refractory Cc=pany, Illinois Products Division, to be used in connection with their clay mining and
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clay products manufacturing activities.
A parcel to the north and east was sold to Commonwealth Edison Company for construction of canals to a cooling lake for the DNPS reacters.
1.2.2 Facility Descriptions The largest building on the site (the =ain building) was originally constructed i
i to house the fuel reprocessing chemical facilities, as well as waste manage-r f
ment, fuel handling, and fuel storage facilities (Figure 1-4).
I OV 1.2.2.1 Main Building The main building is a massive structure of reinforced concrete, about 204 ft by l
78 ft in plan, and about 88 ft high above ground.
The western end of the build-l l
ing houses most of the fuel storage facilities.
This portien of the building is of steel frame and insulated metal siding construction, and is attached to
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the concrete main building.
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January 1981 1.3.7 Emergency Provisions l
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The structures and systems at Morris t'are designed to more rigorous standards than would be required for spent fuel s'.arage.
Emergency plans are in effect, and assistance agreements exist with appropriate local agencies. Structures
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provide access for law enforcement, medical, fire, or other emergency services.
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NEDO-21326C3 Jcnuary 1981
1.4 REFERENCES
O 1.
License and docket information and a list of applicable documents are con-tained in Appendix A.1 and A.2.
2.
The Morris Operation does not encompass 3WRTC activities, although both are General Electric operations.
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3.
Storage capacity expressed in terms of metric tens of uranium (TeU) as contained in LWR fuel rods. A metric ton equals 1,000,000 grams, or one megagram (Mg). Abbreviation for met.*ic :enne (Te) used as recomrended by American Institute of Physics, American Chemical Society, and others.
t Throughout this report, TeU = MIU.
4 The BWRTC is also on this tract. Hereafter BWRTC is referred to only when germane to the purpose of this report.
4a.
" Controlled area" as defined in 10 CFR 72 3 (h).
O 5.
K. J. Eger, Operating Experience - Irradiated Fuel Storage - Morris Operatien, Morris, Illinois, General Electric Company, May 1978 (NEDO-209693).
6.
See Chapter S.
7.
See 10 CFR 100.3 (a) definition.
8.
Previously, the ficor could be drained to the site runoff drain system or the cladding vault. The runoff drain has been disconnected and mpped.
9.
Also, see Figures 1-12 and 1-13.
In various documents, the cask unloading ba * *
- as been called fuel unloading pit, cask unloading pit, etc.
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NEDO-21326C3 January 1981 3.2.2.4 Boundaries for Establishing Effluent Release Limits
~he controlled area boundary (the tract boundary shown in Figure 1-2) is the bouncary for establishing dose equivalents as defined in 10 CFR /2.67 and 72.68.
No credible acts tf nature, san-induced events er accidents have been identified that would result in a biologically significant release of radioactive material or a direct radiation dose in excess of the li=its of 10 CFR 72.68 cutside of the controlled area boundary.
Therefore, the E=ergency Planning Zone (EPZ)
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for Morris Operation coincides with the controlled area boundary.
3.2.3 Population, Distribution and Trends The data base for the following sections is fcunded on infor=ation developed by agencies of the States of Illinois and Indiana, as well as information developed by General Electric and Ccc=cnwealth Edison. la,1b,1c 3.2.3.1 Population Between 0 and 5 Miles (Figures 3 4 through 3-5A)
The pspulation in the i==ediate vicinity of the Morris Operation is very low.
Within a radius of 5 miles the population is about 5,000, including 1,500 in the village of Channahon, about 4 =iles to the northeast.
Included in this accounting are several residences at the Dresden Lock and Dam. The 1970 pop-ulation figures within a 5-mile radius are based on a 197t actual house count assuming three persons per house and are not intended to represent 1970 U.S.
census data. The 1980 projections for the 0- to 5-mile radius are based on an assumed 5% annual growth in all areas except those in which the tract is located.2 f)
The population within 5 miles of the site is projected to increase to 8830 a
by the year 2000. Within 10 miles, the existing population is about 32,400 and is projected to reach 68,000 by the year 2000.2a i
I 3.2 3.2 Population Within 50 Miles (?igures 3-4 through 3-7A) i s
The total population within the 50-mile radius was found to be about 6,051,500
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in 1970 and is projected to reach 7,500,600 by 2000 with about 91% of the total l
beyond the 30-mile radius.3 3-7 t
1 NEDO-21326C3 Jtnu,ry 1981 The 1978 population projections prepared by the State of Illinois and the State-of Indiana for counties within a 50-mile radius surrounding the facility antici-pate relatively miner population growth between 1978 and the year 2005.3" Studies by Co=menwealth Edison's Industrial Development Department indicate that since 1946, 82% of the new industries locating within the Com=cnwealth system are located within 25 miles of downtown Chicago.
In 1965, 80% of the new industries also located according to this pattern.
Current indications are that this industrial growth pattern is slowing but continuing within the 25-mile belt. Thus, the growth adjacent to the GE-DNPS sites (which are cut-side of the 25-mile belt) should continue but at relatively low rates.
The Joliet and Aurora areas are the closest areas that are likely to see signifi-cant population increases.
3.2 3 3 Transient Population There are small seasonal variations in population in the farm lands of the area because of harvest manpower requirements.
Unlike some rarm areas, harvest O
ceivittee ere a18h17 =ecae=1=ee e=e retetive17 rew e.de1tione1 werxere ere needed.
Almost all manufacturing and other industrial activity is nonseasonal and draws upon a population base that resides in the same general area.
For example, with the largest part of Chicago's industrial and residential areas within tae 50-mile radius, th" mily movements of people within Chicago and environs result in a relative ly insignificant statistical change from the viewpoint of ccrisiderations applicable to the Morris Operation site.
As discussed elsewhere in this chapter, recreational uses of lands and water in the area result in small seasonal changes in population in cottages, etc.
These changes have been estimated by observation and incorporated in Figures 3-4 and 3-5.
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3 2.4 Users of Nearby Land and Waters
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/d' The immediate neighbors of Morris Operation (Figure 3-3) are the DNPS site l
on the north, the A. P. Green Refractory Company, Illinois Products Division, 3-12
NEDO-21326C3 Jcnuary 1981 on the south and the Goose Lake Prairie State Park to the west.
To the east p
is the Dresden cooling lake and a privately owned property of about 50 acres, O'
divided into about 30 cottage sites.
Co==enwealth Edison's Collins Station (a fossil-fired plant) is to the west-southwest of the Morris site.
The present land use patterns in the area seem likely to centinue for so=e ti=e to come.
The Northeastern Illinois Planning Cc= mission does not expect a change in the pattern in the southwestern corner of adjacent Will County, either s (The county line is approximately 1-1/2 =iles east of the GE tract.)
3 2.4.1 Industrial In addition to the A. P. Green Refractory Company operations south of the tract and the Co sonwealth Edison holdings to the easti north, and northwest, another industrial area is located along Highway I-55 This highway runs north and south, about 4-1/2 miles directly east of the tract (Figure 1-1).
Two =iles east of I-55 is the inactive Joliet Ar=y A==unition Plant. A large Mobil Oil petroleum refinery is located where I-55 crosses the Des t
Plaines River.
Industrial sites are also located on the north bank of the Illinois River.
3 2.4.2 Residential Use and Population Centers Residential occupancy in the i= mediate vicinity of Morris Cperation is low.
There is a cluster of about 30 cottages en the west shore of the Kankakee River, about 0.5 mile fn:m the Morris Operation stack.
Ihese are located between Dresden Road and the Kankakee River on a tract of about 50 acres adjacent to j
Any residential development in the immediate vicinity or Morris Operaticn would be limited to this tract which is now nearing saturation.
There is also a similar group of cottages en the east b nl af the Kankakee River at a distance greater than 1 mile from the Morris Operation stack.
Some of the homes in this area are permanent residences, although most have A
been developed for part-time recreational purposes. Surveys by Commonwealth
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Edison indicate that within 2-1/2 miles of the DNPS site there are a total of 129 permanent homes and 191 part-time recreational cottages along the 3-13
NEDO-21326C3 January 1981 Kankakee River. Cther residences in the area include several at the Oresden D
Dam about 1.2 =iles to the north.
There are no =ajor residential centers G
developing south of the Kankakee and Illinois Rivers in the vicinity of the General Electric tract.
Cities and towns having populations greater than 1,000 located within 30 miles of the Morris Operation are listed in Table 3-1.
Population centers of less than 1,000 within about 5 miles of the tract are as follows:
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e Village - Minocka, 5.2 miles N 758 people ::,f 1970.
Population is compact.
Subdivision - Dresden teres, 3 5 miles NW e
Approximately 200 people as of 1975.
Population is ec= pact.
Shady Oaks Trailer Fark - 5.1 miles NNE e
Approximately 40C perople (150 trailers) as af 1975.
Population is compact.
Goose Lake Subdivision - 3 to 5 =iles SW and SS'd e
Approxi=ately li']O people as of 1975 Population is diffuse.
e Feather *4 cods Subdivision - 1 =ile E and ESE Approximately 650 people as of 1975.
Population is compact.
l Other areas and sites involvirag intermittent and temporary congregations of
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persons within 5 miles of the Morris Operation are as follows (data as of 1974-1975):
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Schools - Enroll:ent4 Minooka High School 587 Minooka Junior High School 777 Minooka Grade School (3
Channahan School 772 Illinois Youth Centers 30 l
3-14
N!30-21326C3 Jcauery 1981 No access is allowed by Cocconwealth Edison to the Dresden ecoling lakes for
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recreational uses. The Illinois Waterway, one of the major inland waterways, C'
is adjacent to the DNPS site.
An agreement between GE and Commonwealth Edison provides for access to the Illinois Waterway through the DNPS site so that facilities for boat docking and access reads to the waterway could be developed at some future ti=e if required.
There are two small " finger lakes" about 2-1/2 miles south of the GE tract
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where homes have been built, while othee lakes on which houses are being built are located about 3-1/2 miles southwest. Some of these houses are solely for recreational purposes.
3.3 NEARBY INDUSTRIAL, 'RANSPORTATION AND MILITARY FACILITIES Ncne of the industrial, military, or transportation activities in the area present a credible hazard to the fuel storage facility nor to the transport of irradiated nuclear fuel.
Fuel in storage is located well belcw groand level in a steel-lined, reinforced concrete water basin, and held in stainless steel (m) baskets lat:hed in a supporting grid. Explosions or fires
$" "nearbf' industrial N
~j facilitice would be much too far away to have any influence on fuel in stcrage.
Even the explosion of a passing tank truck would not affect the safety of stored fuel. Likewise, the structural characteristics of fuel casks and the nature I
is of nearby activities result in minimum hazard to transportation of spent fuel.~a
.3.1 Nearby Nuclear Facilities The location and identification of nuclear facilities within 50 miles of the Morris Operation site are shown in Table 3-2.
The closest facilities are the s_-)
DNPS Units 1, 2 and 3, located about 0.7 mile north of the Morris operation l
stack.
The ecmbined radiological impacts from Morris Operation and the DNPS are within the requirements of 10 CFR 72.67 as indicated by calculations and i
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environmental monitoring results. The calculated dose eczmitments from Morris Operation are a small fraction of the dose ecmmitments from DNPS, even consider-ing the design basis accidents evaluated in Chapter 8.
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NKDO-21326C3 January 1981 332 Industrial and Military O
2e GE tract is near several industrial sites in an industrial area alcng the Illinois River (Figures 1-1 and 1-2).
Most of the develop =ent is north of the Illinois River at a distance of over 1 =ile from the Morris Cperation.
The rapid develop =ent of the last few years is slowing as = cst of the suitable industrial sites are already occupied and the Goose Lake Prairie State Park now occupies meet of the remaining land south of the river.
Table 3-2 NUCLEAF. REACTORSa WITHIN 5014TTJR OF MORRIS OPERATICN Capacity Airline Miles to
- ypu (MWe)
On Line Latitude Longitude Morris Operation Name 3WR 200 1960 41c22' 380 4' 1
Dresden 1 1
0 SWR 309 1970 31 22' 38014' 1
Dresden 2 BWR 309 1971 41022' 380 14' 1
Cresden 3 EWR 1,078 1978 41 21' 380 6' 20 LaSalle 1b 0
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SWR i.078 1980 alo,'
e8o36' 20 teSe11e 2D a
SWR 1,100 1981 410 16' 38013' 10 3raidwood 1D SWR 1,100 1982 41016' 380 13
10 3raidwood 2D In addition to DNPS 4-ediately to the north and the A. P. Green Refractory Co=pany's clay products plant i==ediately to the south, other industry in a 6-cile radius of the Morris Operation is listed in Table 3-3 333 Transportation O
v One of the principal factors in the original selection of the Morris site was the ready availability of excellent rail and highway ac:ess to all parts of the United States and water transportation that could be developed if required in the future.
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aAll plants owled by Cocunonk - 'th Edison.
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bUnder construction.
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k s REDO-21326C3 January 1981 CJ T
Highway access to the tract is via a paved county road, known as Dresden Rcad, extending south from the DNPS site parallel to the Morris tract and intersect-ing Pine Bluff Road (Figure 1-2).
Pine Bluff Road (na=ed Lorenzo Road in Will County) runs in an east-west direction approximately 1 mile south of the GE tract boundary and provides access to Interstate 55 approximately 4 miles east of the site, and Illinois 47 to the west.
Interstate 55/U.S. 66 is a limited-access highway between Chicago and St. Louis.
Another limited-access highway,
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Interstate 80, which traverses the State from east to west, is approximately 5 miles north of the GE lands and is accessible either from Interstate 55 or from a State highway, Illinois 47, at a point approximately 2 miles north.
Table 3-3 INDUSTRIAL, TRANSPORTATION, AND MILITARY ACTIVITIESa (6-m11e radius)
Installation Function Proximity Reichold Chemical Plant Chemical plant 1.5 mi NW Amax Aluminum mill products 3 mi NW Northern Illinois Gas Co.
Natural gas afg 3 mi NW i
Rexene Poly =ers Co.
Chemical plant 2 mi ENE Mobil Oil Co.
Oil refinery 4 =i ENE Collins Power Station Electricity generation 4 mi WSW (fossil-fired)
ARMAK Co.
Mfg of fatty acid 4 mi WNW derivatives Northern Petro Chemical Mfg of polyethylene and 4 mi NW Co.
ethylene glycol Joliet Arsenal Munitions plant (inactive) 6 mi ENE Demert Lad Dougherty Filling aerosol cans 6 mi S
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aSee Table 3-2 for nearby nuclear power stations.
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NEDo-21326C3 Jcnucry 1981 The seismic risk map (Figure 3-12) of the conterminous United States was U(s prepared by a group of research geophysicists headed by Dr. S. F. Algermissen of the United States Coast and Geodetic Survey and issued in January 1969.
The site area lies well within zone 1 where minor earthquake damage can be expected. According to this map, :ene 1 corresponds to intensities V and VI en the modified Mercalli scale.
Modified Mercalli intensity VI seems to be the greatest intei sity experienced historically in the site area. This intensity was the result of the 1912 earthquake which was centered approximately 15 miles from the site, and may also have been the result of the 1811-1812 New Madrid, Missouri, earthquakes.
Intensity VI, with its corresponding acceleration (according to Newmann's curve) of 0.01g may be reasonably expected to occur again within the life-time of the facility.
3.7.5 Earthquake Design Basis The design earthquake basis for the basin was a horizontal ground motion of
(~/D 0.1g.
Ihe basin structure and fuel storage system are designed to withstand N_
the design basis earthquake without da= age to structures or components essen-tial to the integrity of stored fuel or fuel being moved in the normal process of-storing or shipping fuel. The design earthquake is defined as a seismic event that has a reasonable probability of occurrence during the life of the facility, based on studies of seismic history and geology. A maximum earth-quake with ground accelerations of 0.2g is also considered in the seis=ic analyses. The design bases are discussed in Chapter 4.
3.8 TRANSPORTATION OF IRRADIATED FUEL Irradiated fuel is received by truck or rail at Morris Operation in casks cer-tified to comply with applicable U.S. Nuclear Regulatory Commission regulations.37 Typical shipping casks are discussed in Section 1.3 As of the end of 1980, 510 shipments of fuel had been completed, moving about rg 316 tonnes - heavy metal in 1220 fuel bundles. Shipments to Morris Operation
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have been canpleted without highway or rail accidents. Rail shipments were all from DNPS.
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NEDO-21326C3 Jcnucry 1981 The environ = ental impact of these transportation operations has been negligible, ll p) thus supporting the conclusions of various studies and analyses.38,39 Projected transportation activity required to store a full ecmplement of fuel
- about 700 TeU, total, or about 1300 additional bundles assuming a 60:40 ratio between PWR and BWR fuel - is shown below. This esti= ate is based on the assu=p-tion that all shipments are by truck; ship =ents by rail would result in even lower environmental impacts.
('s a.
1000 shipments (varies with PWR/3WR =ix) b.
1500-mile trip (one way) c.
Elapsed time per trip - about 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> d.
Turnaround per cask - 18 to 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> e.
Maxi =u= shioments per year - 300 to 400 The nonradiological and radiological impacts of transportation are analyzed in the literature.40 Environ = ental i= pact assessments of the Morris Operation facility by the Nuclear Regulatory Cc==ission staff have also founc no sigrificant environmental impact frem spent fuel transport.41' *2 3.9 SU MARY CF CONDITIONS AFT?ECTING FACILITY OPERATING REQUIREMENTS lE Irradiated fuel storage operations have been unde"way at the Morris Operation since January 1972 when the first shipment of irradiated fuel was received
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under Materials License No. SNM-1265, Docket 70-1308, issued December 1971.
Ihroughout this period of operating experience and during the on-going environmental studies and monitoring programs, no condition has been found to detract from the desirability of this site as a fuel storage location.
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NEDO-21326c3 January 1981 E
391 Significant Factors
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The following factors were significant in selecting the design bases for the existing Morris Operation facility.
3 9.1.1 Meteorology lE p
The climate at the site offers no severe extremes except tcrnadoes. Analysis of tornado activity, including official and unofficial records, indicates a frequency close to the average for all states east of the Rocky Mountains.
I The topography of the site introduces little perturbation in diffusion calcu-lations; only the 630-ft elevation of the Dresden Heights, about 1-1/2 miles north of the Morris Operation stack is of concern in selecting stack design bases. Local fog conditions are involved in dispersion considerations. Diffusion climatology and characteristics have been firmly established and confirmed by the meteorological measurement program.
O lE 391.2 Hydrology Surface hydrology of the site offers no characteristics significant to the selection of design bases (except for the usual consideration of natural drainage pathways, etc.).
Subsurface hydrology shows excellent separation between the upper strata and the deeper aquifers that provide the water supply - almost exclusively - for municipal and industrial use.
The intrusion of groundwater is of concern during construction, based on experience during MFRP work. These flows indicate a complex near-surface V
groundwater system that becomes significant because of localized fracturing induced during construction.
3913 Geology and Seismology lE The site is located in a stable area which has experienced historically low seismic activity. The existing construction is founded'on bedrock of Ordovician
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(Paleozoic) age. Design of the facility and its fuel storage equipment for horizontal ground motion of 0.10g is ' considered conservative.
3-57
h 21326C3 January 1981 3.10 REFERENCES O
1.
See Appendix A.1 for document list.
la. State of Illinois, Bureau of the Budget, Illinois Population Projections (Revised 1977), Springfield, Septe=ber 1977 1b. State of Indiana, State Board of Health, Indiana County Population Projec-g
()
tions, Indianapolis, 1978.
1c. Northeastern Illinois Planning Commission, Regional Data Report, Chicago, June 1978.
2.
The 5". growth in the 0-5 mile area was developed from the assumption that far= land will not experience growth (urbanization) except in a few selected areas. This growth was esti=ated and the overall araa growth integrated.
Most people working in local industries live in the Western Joliet and Morris areas; there has been little growth in s= aller co== unities.
i 2a. Beyond the 5-mile area, population data' totals on charts have been rounded off to the nearest 100.
3 The USNRC staff reported an adjusted estimated 1980 population for the area within the 50-mile radius of about 9,169,337 (Environmental I= pact Appraisal, Docket 70-1308, NR-FM-002).
3a. During research for these data, differences were noted between (for example) the Northeastern Illinois Planning Commission data and Federal census figures.
In general, however, the data appear mutually supportive, particularly at
~
the county levels.
l l
i 4
Within 5 miles of the site the total school population is 800, but at j
slightly more than 5 miles it increases to about 2,140; the larger number is shown.
O s -
4a. See Reference 39: WASH-1238,Section II, E.
ll l
3-58 t
1
NEDO-21326C3 Jcnusry 1981 5.
Correctional institutien (juvenile) at 01annahon, 3 mi$es WNW.
i O
6.
Climatography of the United States, No. 60-11, revised and reprinted June 1969.
7 H. E. Landsberg, " Climates of North A= erica," World Survey of Climatology, Vol.11, edited by Bryson, et al., Elsevier Scientific Publication Co.
(1974).
O 8.
S. S. Visher, Climatic Atlas of the United States, Harvard University Press,
Cambridge (1966).
9.
U.S. Department of Commerce, Climatography of the United States No. 86-9,
" Decennial Census of United States Climate," for Illinois, Washington, D.C. (1964).
- 10. " Final Environmental Statement related to operation of the Midwest Fuel Recovery Plant by the General Electric Co.," Doc. 50-268, USAEC O
(oeoemeer 1972).
- 11. Fluor Cooling Products Company, " Evaluated Weather Data for Cooling Equi;=ent Design," Addendum No.
1, Winter and Su=mer Data, Santa Rosa, CA (1964).
- 12. D. W. Phillips, et al., "The Climate of the Great Lakes Basin," Climatological Studies Number 20, Environment Canada, Toronto (1972).
- 13. J. L. Vogel, et al., " Fog Effects Resulting from Power Plant Cooling Lakes,"
i O
Journal of Applied Meteorology, Vol.14 ( August 1975).
G'
- 14. Final Environmental Statement related to the operation of Dresden Nuclear Power Station Unita 2 and 3 by the Commonwealth Edison ca., Docket No.
50-237 and 50-249, AEC (November 1973).
i l
- 15. Applicants Environmental Statement, Dresden Nuclear Power Station Unit 3, Commonwealth Edison Co., Docket No. 50-249 (July 1970).
3-59
NEDO-21326C3 January 1981 p
- 16. Thom suggests an annual extreme-mile (fastest mile) wind speed of 82 ph for 30 ft above ground and for a 100-yr mean recurrence interval.
- Thom, H.C.S.,
"New Distributions of Extreme Winds in the United States," Journal of the Structural Division, Proc. ASCE, Vol. 94 No. St. 7 (1968) Applicants Environmental Report, Midwest Fuel Recovery Plant Morris, Illinois, June 1971.
- 17. liarry and Trettel, Inc. Consulting Meteorologists, Chicago, IL. Letter, Literski (M&T) to Eger (GE), September 23, 1976.
- 18. An increase of 5% for a reprocessing facility; less for a storage facility.
- 19. From Braidwood Station Environmental Report, Commonwealth Edison Co.,
Chicago, IL.
Year of record: July 1971 - June 1972.
- 20. The application of these methods to the Dresden reactors and the description of the techniques used there can be found in Appendix A of the Final V
Safety Analysis Report for Dresden 2 and 3, Docket 50-237.
- 21. The description of the first year's data taken at the site can be found in Amendment No. 13, Question B-11, to the Dresden Unit No. 2 Final Safety Analysis Report, Docket 50-237
- 22. E. C. Watson and C. C. Guertsfelder, "Em ironmental Radioactive Contamina-tion as a Factor in Nuclear Plant Siting Criteria," February 14, 19 6 3, HW-S A-2809
()-
- 23. Dames & Moore report dated January, 1971 (Appendix B).
' 24. Dames & Moore,1550 Northwest Highway, Park Ridge, Illinois 60068.
- 25. Payne, 1940, page 7; and Eardley, 1962, page 45.
- 26. Willman and Templeton,1951, page 123 7.sb 3-60
NEDO-21326C3 Jcnunry 1981
- 27. Bristol and Buschbach, 1973, Plate 1.
- 28. Willman and Templeton,1952; also Bristol an' Buschbach, 1971, Figure 3
- 29. Ekblau, 1956; Da=es & Moore, 1965.
- 30. Kempton,1975.
- 31. See Table 3-13 for studies referenced in this section.
Ov
- 32. Payne, 1940; Willman and Templeton,1951.
- 33. Willman and Templeton,1951.
- 34. Dames & Mccre, report dated December, 1967 (Appendix 3).
- 35. J. A. Udden prepared a report describing observations of this earthquake.
He presents an isoseismal map for this earthquake and, according to his
=ap, the site -as in the area which experienced Rossi-Forel intensity 7I (about V-VI sn the modified Mercalli scale).
- 36. This intensity is based on an isoseis=al =ap prepared by O. W. Nuttli and presented in the Bull. Seis. Soc. Am., Vol. 63, No.
1, 1973
- 37. K. Eger, Operating Experience Report - Irradiated Fuel Storage at Morris Operation - January 1972 to December 1979, General Electric Company, September 1980 (NEDO-209693).
38.10 CFR 51, mmmary Table S-4, " Environmental Impact of Transportation of Fuel and Waste To and From (he Light-Water Cooled Nuclear Power Reactor,"
U.S. Nuclear Regulatory Commission, especially Note 4, "Although the environ-mental risk of radiological effects stemming from transportation accidents is curently incapable of being numerically quantified, the risk remains small regardless of whether it is being applied to a single reactor or
,~s a multireactor site."
()
l l'
I 3-61 r
NEDO-21326C3 January 1981
- 39. Environmental Survey of Transportation of Radioactive Materials to and from i
Nuclear Power Plants, U.S. Atomic Energy Commission, December 1972 (WASH-1238); and U.S. Nuclear Regulatory Cocsission, April 1975 (Supple =ent 1, NUREG-75/038).
- 40. Final Environmental Statement on the Transportation of Radioactive Material by Air and Other Modes, U.S. Nuclear Regulatory Cc= mission, December 1977 (NUREG-0170 ).
O
- 41. Environmental Impact Appraisal by the Division of Fuel Cycle and Material Safety Related to License Amendment for Materials License Amendment 'for Materials License No. SNM-1265 Morris Operation Facility - Grundy County, Illinois for General Electric Company - 1)ocket No. 70-1308, Nuclear Regula-tory Cc= mission, December 1975 (NR-FM-002), especially Section 6.
- 42. Environmental Impact Appraisal related to the Renewal of Materials License No. SNM-1265 for the Receipt, Storage and Transfer of Spent Fuel at Morris Operation - General Electric Company - Docket No. 70-1308, U.S. Nuclear O
aesu1 tor 7 cc==1==1 =. 3==e '98o. e=aeo1>117 sec=1o== 7 5 =a 3 2-l l
l l
l I
O 3-62
NEDO-21326C3 Jr.nucry 1981 4
DESIGN CRITERIA AND COMPLIANCE n/
4.1 INTRODUCTION
A general description of the Morris Cperation and a smw y of its operational functions are contained in Chapter 1.'
Be original design criteria for Morris Operation facilities were developed and established as part of the design for a fuel reprocessing plant - the Midwest Fuel Recovery Plant.
The criteria herein p) are those applicable to the use of those facilities for spent fuel storage.
w 4.1.1 Material To Be Stored The Morris Operation is licensed to store irradiated light water reactor fuel from nuclear power stations. The design basis fuel is UO2 fuel with an initial enrich-ment of 5% U-235 or less, with stainless steel, :1rconium, er Zircaloy cladding, in a " bundle of rods" geometry.
The design basis fuel may have been irradiated at specific power levels of up to 40 kW/kgU, with exposure to 44,000 mwd /TeU (batch average), and =ust have cooled for at least 1 year after reactor shutdown before OV
, entering storage at Morris Operation.
Se accident analyses in this report were originally prepared for fuel cooled 90 and 160 days, and these analyses have not been changed.
The calculated fissi.:n product act:vity contents of fuel irradiated at 40 kW/kgU, exposed at 44,000 mwd /TeU, and cooled 90 days and 160 days are pre-sented in Table 3-1.
Fuel to be received and stored typically will have exposures of 33,000 mwd /Teu or less, with cooling periods much longer than 1 year.
As of the first of Feb-ll ruary 1977, the average exposure of BWR fuel in storage was about 3,100 mwd /Teu and that of PWR fuel about 23,700 mwd /TeU.
The average exposure was less than AC 15,000 mwd /TeU, far less than 44,000 mwd /TeU.
As of the first of February 1977, the average cooling time of BWR fuel in storage was about 61 months and that of PWR was about 52 months. Overall, the average cooling time was about 56 months (4.7 yr).1 Realistic exposures and cooling times based en the fuel in storage have been used in some analyses, as appropriate. Table 4-2 contains a list of analyses and fuel p
exposures and cooling times on which each is based.
d eSee Section 4.7 for references.
lE 4-1
NEDO-21326C3 Jcnucry 1981 including the forces that might be imposed by natural phenomena such as earth-quakes, tornadoes, and flooding conditions.
Standards for assuring that systems, structures and equipment will perform safety functions for their intended service life with a low probability of failure have been based on upper limit temperatures, corrosion rates and other stress condi-tions derived from comprehensive analyses, including consideration of:
accessibility for in-service surveillance, monitoring and repair a.
(or replacement);
b.
potential for short-term exposure to abnormal operating or accident conditions; and c.
ccnsequences of component failure; no single component failure or multi-ple failures caused by a single initiating event shall result in signif-icant radiation exposure to the public.
O 4-ecce eidi11e7 rce e=erseac7 eerv1=ee. 1ac1ue1== =eu1 =ce eteae at=. cire and police services, and other emergency activity.
4.2.1 Wind and Tornado Loadings i
4.2.1.1 Criteria Plant structures and components essential for safety shall be designed to with-stand the effects of short-term wind velocities of 300 mph with pressure differen-tials of up to 3 psi without damage to fuel in storage to an extent endangering
\\
public health and safety. The site is located in USNRC Tornado Intensity Region I, as defined in Regulatory Guide 1.76.
l 4.2.1.2 Compliance i
The fuel basin structure (enclosure) was analyzed with wind loads applied as uni-p form static loads on the vertical or horizontal projected areas of the walls and roof. Only dead load was considered as resisting uplift.
Horizontal wind loads 4-6
NEDO-21326C3 Jtnuary 1981 are distributed by the walls to the floor and roof systa=s, which transfer loads to the lateral load-carrying ele =ents of the structures.
Plant structures and components were designed to withstand sustained wind veloc-ities of 110 =ph without loss of functions.
At higher velocities, the enclosure covering =ay fail or blow away.
These analyses included consideration of a drop in atmospheric pressure of 3 psi
(
in 3 seconds. This condition would damage the basin enclosure, probably da= age or even remove much of the roof and wall sheathing from the basin enclosure, but would *cause no off-site radiological effect.
4.2.2 Tornado Missile Protection 4.2.2.1 Criteria Plant structures and components essential for safety shall be designed to withstand the effects of windborne =1ssiles witnout dwage to fuel in storage to an extent
/3 V
endangering public health and safety.
4.2.2.2 Compliance The folicwing su= nary of analyses indicate that the public health and safety would not be endangered as a result of tornado =1ssiles i=pacting the fuel storage struc-tures or components.
Only those windborne objects which could have a significant downward velocity on entry into the water-filled basin have the potential for causing damage to basin U
contents. Such objects must have been at a significant elevation above ground
)
level, prior to entry, to develop the required vertical velocity component to result l
in damage.
Potential missiles can be classified in regard to their relative eleva-tion, as follows:
Y i
i 1.
Objects in the insediate area which, when the tornado strikes, are at elevations above the level of the basin surface (operating equip-l ment and auxiliaries, components of the enclosing structure, etc.).
l l
4-7 i
l l
!EDO-2132603 I
i j
January 1981 t
j 4.2.7 Basic Water Cooling i
i l
4.2.7.1 Criteria 1
1 4'
Means shall be provided to =aintain basin water te=perature less than 200 F 0
t (93 3 c).
1 I
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4.2 7 2 Compliance i
{
3asin water is cooled by a system described in Section 5.5 3 i
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NEDO-21326C3 Jcnuary 1981 4.3 SAFETT PROTECTION SYSTEMS O
4.3.1 General There are no site-related factors that are sufficiently unusual to require pro-tection systems or special design considerations beyond these normally required for a facility of this type. Operations shall take into account the proximity of DNPS to assure that the cumulative effects of these operations do not constitute an unreasonable risk to public health and safety.
4.3.2 Protection by Multiple Confinement Barriers and Systems The total confinement system consists of one er more individual confinement barri-ers and systems that successively minimize the potential for release of radioactive material to the environment. These features also ;rotect the fuel in storage by E
protecting the fuel frta damage and providing a favorable environ =ent.
4.3.2.1 Criter ia Equipment and systems containing radioactive or potentially contaminated = ate-rials shall provide a continuous boundary against escape of such material and be desiped to have a icw probability of gross failure or significant uncontrolled leakage during the design lifetime.
Secondary confinement barriers such as vaults, ventilation system, etc., shall be designed and constructed to contain the results of primary system failure, under conditions that may have initiated such failure, without less of required integ-rity and to continue operation for the maximum anticipated period of stress.
O Storage vaults and basins shall be designed and constructed for a lcw probability of gross failure or uncontrolled leakage, with =eans provided to monitor leakage and preclude transport of radioactive materials to underlying aquifers. For lined ll structures containing radioactiv,e or potentially contaminated liquids, leak detec-tion and empty-out_ means shall be provided between the liner and the structure so that release of radioactivity to the environs can be avoided by pumping leakage back into storage, effecting repairs where leaks can be located and are accessible, or installing additional facilities in the event repair is not feasible. Water 4-42
NEDO-21326c3 Jcnucry 1981 systems shall be designed to prevent accidental re= oval of water from the basins l!
by any =eans to less than a safe level. Basin water level shall ta indicated and alarmed (1cw water alarm) in the control room.
4.3.2.2 compliance All criteria described above have been satisf!ed; refer to Chapter 5.
4.3.3 suilding ventilation 4.3.3.1 criteria Radioactive material in the building ventilation exhaust shall be reduced to levels that are as icw as reasonably achievable before being released to the environs.
Special venting lines and special enclosurss shall be employed when necessary, such as during cask venting operations, to confine airborne radioactive particulate materials.
I i
4.3.3.2 compliance
%J The principal =ethods used to =eet these criteria include the follcwing:
a.
Generation:
Airborne radioactive =aterial may originate fec= cask decen-tamination and venting operations; icw activity solid waste compactor operation; preparation of contaminated equip =ent for disposal; and from operation of the low-activity liquid waste treatment systems. Other than these principal sources and the minor leakage from fuel in storage, no other significant source exists.Sa These activities can be suspended Dd on short notice whenever higher than prescribed levels of radioactive materials are detected in the ventilation air exhaust stream.
The waste evaporator system is designed to limit radioactive material in its effluent.
b, confinement:
The building ventilation system utilices pressure differ-p entials to maintain air flow paths to exhaust all ventilation air through the filter system and the discharge stack. Special venting systems I
and special enclosures may be employed to confine airborne particulates l
l 4-43
^
NEDO-21326c3 Jcnuary 1981 from cask venting, decontamination activities, or similar sources to the filter - discharge stack system. The ventilation system is designed for all credible normal or anticipated off-nor=al conditions.
c.
Release:
2.e building ventilation system is designed
- ,o collect all ventilation air and exhaust it through a final sand filter of demonstrated capability for removing particulate matter, and a 300-foot-high discharge stack.
O 4.3.4 Protection by Equipment and Instrumentation 4.3.4.1 criteria Equipment and instrumentation shall be provided to monitor radioactivity and other parameters of operation, and to perform related control functions in accordance with the following:
Equipment and systems shall be set and adjusted to alarm and/or initiate a.
(
action such that specified limits are not exceeded as a result of normal or abnormal occurrences.
b.
Redundancy and independence shall be provided to a degree sufficient to assure that no single failure of an instrument or equipment item can re-sult in loss of protection functions.
Equipment shall be designed to permit inspection, testing, and maintenance.
c.
i d.
The control room shall permit occupancy and allow monitoring of important systems and functions during normal operations and under anticipated off-normal or accident conditions.
4.3.4.2 Equipment Compliance Equipment is designed to permit inspection, maintenance, and periodic testihg of p
functions to specified parameters. Temporary removal of single items of equipment frem service has no safety significance.
4-44
NEDO-21326C3 Jcsuary 1981 4 3.4.3 Instrumentation Compliance GV Instrumentation is provided to assure proper operation or notification of the fail-ure of systems.
Instrumentation is designed or specified to standards of known relia bility.
To assure instrument reliability, periodic testing and calibration checking are perfor=ed in accordance with Operatiens Specification 10.4.4.1.
Alams indicating a set point has been exceeded are annunciated in the control room, and where there may be an immediate effect on personnel such as radiation exposure they are alarmed locally as well.
4.3.4.4 Control Room The control room is described in Section 5.5.5.4 4 3.5 Nuclear Criticality Safety 4.3 5.1 Criteria O
Every reasonable precaution shall be taken to preclude a criticality within the Morris Operation.
Both design and administrative controls shall be utilized.
4 3 5.2 Design Control Compliance The design of the spent fuel storage system includes the following controls to pre-clude a criticality incident:
a.
Initial analyses were made in sufficient detail to demonstrate that the criticality control concepts considered (e.g., favorable geometry) ll were feasible under all credible conditions. Additional detailed nuclear criticality safety evaluations of the final design were made by qualifiec experts in the field to assure that final dimensions and other factors affecting safety margins were adequate to prevent a criticality incident.
The additional detaile'd analyses required to confirm the final design p
are included in this document in Appendices A.10 and B.S.
lE V
4-45
NED0-21326C3 Anunry 1981 b.
In the derivation of subcritical limits, the k.cr calculated for the most reactive credible conditions was specified as 0 95 at a 95 per-cent confidence level.9 4 3 5.3 Administrative Control Compliance The operation of the spent fuel storage facility includes the following adminis-trative controls to preclude a criticality incident.
Safety evaluation, review and approval of operating procedures related a.
to design control parameters, b.
Verification of nuclear fuel parameters for fuel scheduled to be stored at Morris Operation.
Verirication of fuel identity for fuel received at Morris Operation c.
for storage.
(Q d.
Maintenance of fuel storage location records.
e.
Specific fuel and cask handling procedures.
f.
Personnel training.
Independent review and audit procedures are utilized to determine the adequacy of nuclear safety control provisions and the effectiveness of implementing activities.
O 4 3.6 nadio1o.1catereeece1on 4 3 6.1 Criteria Radiation and radioactive contamination conditions at the Morris Operation shall be controlled to provide protection of personnel health and safety at all times.
Emphasis shall be placed on minimizing both individual exposures and total exposure (man-Rem) to as low as reasonably achievable (ALARA).
i l
4-46 L
NEDO-21325C3 Jcnu ry 1981 During normal operations, including anticipated occurrences, the annual dose equivalent to any person located beyond the centrolled area boundary shall not exceed 25 mrem to the whole body, 75 mrem to the thyroid ccd 25 mrem to any other organ as a result of either planned discharges or direc* o.diation from the facility.
Any person located at or beyond the nearest boundary of the controlled area shall not receive a dose greater than 5 Rem to the whole body or any crgan from a design basis accident.
4 3 6.2 Compliance Criteria are satisfied through the follcwing design features and operational pactices :
Confining radioactive materials to prescribed locations.
a.
b.
Clearly defining areas in which significant radiation or con-tamination levels exist.
Applying special provisions and appropriate procedures to c.
assure personnel safety.
d.
Applying rigorous surveillance, housekeeping, and clean-up practices.
Providing comprehensive personnel training in radiological safety.
e.
Dosimeters are provided for assuring accurate detection and assessment of personnel exposure to ionizing radiation, in accordance with applicable procedures. Thermoluminescent devices (TLD's) are positioned throughout the site to assest trends in background dose rates so that increases may be detected and corrective plans initiated.
4.3.6.2.1. Access Control (Restricted Areas)
O V
Provisions have been established for controlling personnel access to areas in which radioactive material is present and are maintained to keep the potential for 4-47
NEDO-21326C3 Jcnunry 1981 contamination spr:ad and exposure to radiation as icw as reasonably achievable.
This is accomplished by maintaining a series of access control barriers with s
increasingly restrictive occupancy constraints and access authorization require-ments.
These access controls were designed as follows:
a.
General Electric Tract:
Agricultural fencing with appropriate posting encloses the tract.
Routine surveillance by operating and security personnel is utilized to assure that unauthorized occupancy for signif-icant periods of time is prevented.
b.
Protected Area:
An 8-ft-high chain link fence topped with barbed wire surrounds the protected area in which the Morris Operation storage facili-ties are located.
Personnel and vehicle access gates are locked or manned by security personnel at all times.
'4hile in the protected area, personnel are required to wear personal identification and dosimeters.
All vehicles, materials and equipment are checked into and out of the area following pro-cedures that require potentially contaminated or radioactive items to be monitored and cleared before entry or exit is authorized.
s
' c.
Ooerating Area:
Personnel access to the operating areas in which radioactive material is stored is controlled by limiting entrance such that occupancy authorization require =ents can be strictly en-forced.
Access to the various areas is controlled by the structural compartmentalization and by authorization procedures commensurate with the conditiens existing in the particular areas.
Access to all potentially contaminated areas requiring personnel occupancy is limited to specific routes that have been provided and is in accordance with prescribed procedures, clothing and monitoring requirements, which are varied according to the particular conditions.
Exit from the operat-ing areas, except under emergency conditions, is by the same controlled routes, through necessary clothing change stations and monitoring facili-ties. Routine radiation surveys of the area are performed, and TLD's are posted.
Equipment requiring access (e.g., basin coolers) can be decon-taminated to permit maintenance.
O V
Materials and equipment required for operation and maintenance will i
be checked into the areas and will be monitored before leaving the 4-48
NEDO-21326 C3 Jcnucry 1981 areas in accordance with prescribed control procedures.
Access for transfer of such ite=s is li=ited to specific points which are pro-vided with =eans for precluding unauthorizeo usage, d.
Controlled Access treas:
Areas with the potential of high dose rates are locked, with access controlled from the Control Room.
4.3.6.2.2 shielding Radiation shielding is provided to restrict personnel exposure to levels that are as icw as reasonably achievable.
4.3.6.2 3 %diation Alarm Systems Sampling and detection systens are provided that have sufficient sensitivity and scope of coverage to assure that any radiation or contamination condition of potential safety significance is accurately and promptly assessed.
()
Area radiation monitors meet the following require =ents:
Monitors will detect gacma radiation within the range of 0.1 to 1000 mR/hr.
a.
/
b.
The high trip alarm is audible locally and also annunciates in the control room.
The criticality accident alarm system meets the following requ!.rements:
c.
(1) The system has gamma-sensitive =enitors that meet the sensitivity
()
requirements of 10 CFR 70.24(a)(1).
(2)
The system produces an audLLle alarm that is unique and cannot be shut off even if the exposure rate decreases.
(3)
Two detectors are provided in the basin.
(4) Tha system is continuously functional.
4-49
NEDO-21326C3 Jtnu ry 1981 (5) Detectors are located in the storage basin area but not underwater.
(6) The upper alar n trip circuits for the system are arranged in parallel so that either alarm will energize all criticality alarms.
(7) The alarm circuit that energizes the criticality horns is designed to stay on once it has been initiated and a manual reset in the con-trol room must be employed to silence the horns.
4.3.6.2.4 Effluent Monitoring Sampling and monitoring systems and associated procedures are provided to measure radionuclides in ventilation effluent and in sample wells. Meteorological data and off-site radioactive materials monitoring are provided by a joint program with DNPS, 4.3.7 Fire and Exolosion Protection
]
4.3.7.1 criteria Structures, systems and ccmponents directly involved in the storate of fuel shall be protected so that performance of their functions are not impaired when exposed to credible fire and explosion conditions.
4.3.7.2 compliance This criterion is met by using noncanbustible and heat-resistant -Jaterials when-ever practical throughout the facility, particularly in locations vital to the OQ functioning of confinement barriers and systems such as the basin areas and the pump room. Fire detection, alarm, and suppression systems are installed in ware-house areas, and certain areas of the main building. Fire extinguishers and other equipment are strategically located throughout the facility. Fire brigade training is furnished to operational personnel. Fire alarms are audible in the control room.
I 4-50
.WDO-21326c3 Jcnunry 1981 4-3.8 Fuel and Radioactive Waste Handling ar! Storage O
4.3.8.1 Spent Fuel Receiving and Storage Criteria The cask and fuel handling syste=s shall provide for the safe, reliable and effi-cient handling of casks and fuel.
4 3 8.2 compliance mU The cask and fuel handling system is capt.ble of receiving irradiated fuel bundles in shielded casks :scunted on trucks or railroad cars.
All major equipment such as cranes located above basin areas containing fuel are designed to ensure that co=po-nents will not fall into the basins.
The cask handling system has been designed to preclude a cask from being moved over the fuel storage basins. Means are pro-vided to preclude lifting a fuel bundle or a fuel storage basket to an elevation within a basin such that the shield provided by the basin water is reduced suffi-ciently to cause excessive exposure to personnel.
Cask drop analyses have determined that energy absceptien provisiens in the fuel unloading basin are adequate.
Treat =ent of the stcrage basin water is adequate to 1nimi::e corrosien and prevent undue exposure of personnel.
4 3.8 3 Radioactive Waste Treatment criteria Radioactive waste shall be stored on-site in a manner that does not preclude re-trieval and transfer off-site.
Provisiens sLall be made for inspection and sam-pling of the waste material.
No liquid radioactive waste shall be discharged from the site.
Solid radicactive waste shall be disposed of in accordance with current regulations.
Criteria for storage facilities are given in Section 4 3 2.1.
(v) 4-51
NEDO-21326C3 Janutry 1981 4.3.8.4 Compliance O
Radioactive liquid waste is stored in the icw activity waste vault, and periodi-cally reduced in volume by evaporation. The vault can be emptied for retrieval.
No radioactive liquid waste is discharged fem the site. A solid waste ecmpactor is providad to reduce the volume of solid waste before disposal via a licensed con-tractor.
4.3.9 Utility Systems ll 4.3.9.1 Criteria Utility systems important to safety shall maintain the capability to perform functions important to safety, assuming a single failure.
i 4.3.9.2 Compliance See Section 5.7.1.
O 4.4 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYSTEMS Th3 primary quality objective of General Electric Ccepany is to provide a nuclear fuel storage facility in which strue:ures, systems ud components contributing to the prevention (and/or mitigation of consequences) of conditions that could re-sult in undue risk to public health and safety will perfora their required func-tions in a predictable manner during their intended service life.
The degree of reliability that must be provided for various structures, compo-nents, and systems is determined primarily by the consequences of failure of 1
that unit. Failure of scxne structures, systems, or components could - if uncor-rected - expose people to ionizing radiation. However, in a passive facility such as a fuel storage basin, repair or replacement of the failed structure, system or component can usually be accomplished long before the consequences could pose undue risk to public or employee health and safety.
Failure of still other structvres, systems or components could result in an unacceptable loss of operating efficiency, but would pose no significant long or short-range danger to employeis or the public.
4-52
NEDO-21326C3 Jcnuary 1981 Ouality assurance history and a list of structures, systess, and cc=Anents i!
i=portant Oc safety are centad-ad Chapter 11 The quality assurance plan
'{
da is contained in Appendix 3.3.
t, 9
t.t.1
- ntensity of Natural Phenc=ena 3.'
l.
Provisions have been =ade to monitor r.atural pnenc=ena in the region related i,
to Merris Operation. Meteorciogical data is provided througn a joint progra=
with CNPS {Section 3.3.3). Likewise, provisiens for seisnic =easure=ents are f;
in place at the adjacent DNPS.
j t-4.5 DECOMMISSICNING 4.5.1 Criteria The Merris Cperation facility shall per=1t effective decenta=inatica and decc=-
- issioning to an extent per=itting return of the site to unrestricted use.
I t.5.2 Ccapliance t
t i
The Morris facility desigr. provides a stainless-steel-lined basin that facili-tates cleaning, volu=e-reducing waste manage =ent facilities, and a ventilatien sand-filter that will facilitate decentamination and decc==1ssioning operations.
i.
Other features - criginally designed for a repeccessing facility - facilitate ft re=cval of compenents and contamination centrol (e.g., the canycn area and LA*4 evaporator). See Appendix A.7.
t 4.5 CODES, GUIDES, AND STANDARDS
(",'
\\
s l,
Codes, guides, and standards applicable to the Merris Operation facility, as noted in this report, are listed in Table t 4 I
C_/
\\
l a-53
_= -
NED0-213I6C3 Jcnucry 1981 Table 4-4 CODES, GUIDES, AND STANDARDS
{
t l
Section i
Item Where Referenced j
Uniform Building Code 531 i
ASIM C150 (Cement) 5.5.1.2 ASTM A15-(Rebar) 5 5.1.2
[)
ASTM 262 (Stainless Steel Liner) 5.5.1 3 Regulatory Guide 1.76 4.2.1.1 Regulatory Guide 1.60 4.2.4.2 Regulatory Guide 1.61 4.2.4.2 AISC Steel Construction Manual 4.2.4.2.ga 7th Edition, Appendix A ACI 318 4.2.5.2.1 ANSI-N18.2A 1975 4.3 5.2 API-650, Appendix D 5.6.1.2 ASTM A514 (Stainless Steel)
Appendix A.3
()f ASTM A235 (Stainless Steel)
Appendix A.13 ASTM A240 (Stainless Steel)
Appendix A.13 AWS-ASTM ' 4elding rod)
Appendix A.13 O
. aOther references, also.
4-54 m
4D 4
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NEDO-21326C3 Jcnusry 1981
4.7 REFERENCES
id O
1.
K. J. Eger, Operating Experience - Irradiated Fuel Storage - Morris Operation, Morris, Illinois, General Electric Company, May 1978 (NEDO-209693).
2.
Other postulated missiles (pipe, wood planks, steel rod, etc.) have less dam-age potential than those missiles considered.
Q3 D. R. Miller and W. A. Williams, Tornado Protection for the Spent Fuel Storage Pool, General Electric Cc=pany, November 1968 (APED-5696).
4.
F. C. Bates and A. E. Swanson, Tornado Design Considerations for Nuclear Power Plants, Black & Veatch, Engineers.
5.
P. L. Dean, Tornados and Tornado Effect Considerations for Nuclear Power Plant Structures Including the Spent Fuel, United Engineers and Constructors.
6.
C. V. Moore, Design of Barricades for Hazardous Pressure Systems, Nuclear O
Ene1=eer1=
e=d Deeien 1967.
7.
Design of Structures to Resist the Effects sf Atomic Weapons, U. S. Army Corps of Engineers.
8.
Ammann and Whitney, Industrial Engineering Study to Establish Safety Design Criteria for Use in Engineering of Explosive Facilities and Operations, April 1953 Sa. See Reference 1: NEDO-209699, Section 4.
E Abl 9.
See ANSI N18.2A-1975, Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants.
1 i
O 1
l 4-55/4-56
NEDO-21326C3 Jcnucry 1981 5.
FACILI ! DESIGN AND DESC2!?TICN OV 5.1 INTRODUC ICN This chapter contains descriptive infer:ation en the buildings and other features of the plant involved in the receipt, storage er ship =ent of *.rradiated fue).
Facilities associated with fuel reprocessing ara discussed only as they relace to irradiated fuel storage activities.
O This infor.tatien has been censolidated frc= docu=ents previously sub=itted and part of' the publi: record.
- he =ajority of descriptive =aterial is based en the MFRP FSAR (NEDC-10178) with amend =ents and supplements, and The Safety Evaluation Report For Morris Operation Fuel Storage Expansion (NEDO-20825).
5.2
SUMMARY
DESCRIPTION Reproductions of maps and other illustratiens in Chapters 1 and 3 (especially Figures 1-1, 1-2, 3-1, and 3-2) provide seographical infor:atien about the d
Merris Operatien tract and show the bot ndaries af the General Electric prop-erty and the general arrangement of buildings and other site features.
(See Chapter 1 for use of the te:=s " tract" and " site.")
A.:cre detailed layout and centour =ap of the site and environs is shown in Figure 5-1.
All radioactive =aterial luindling related to the fuel storage activity at Merris Operation is in facilities located w:. thin the protected area.
There are no radicactive liquid effluents released to the envirens and no burial of radioactive ce centaninated material on the tract.
Ihe enly radicactive or contaminated waste =aterials leaving the site are effluents vented through k
the ventilation stack or solid icw-level radicactive wastes that are shipped off-site. Off-site shipnents are =ade in accordance with applicable United States Nuclear Regulatory Commission (USNRC). United States Department of Transportation (USDOT), and other State and Federal r Bulations.
L 5.2.1 Controlled, Restricted, and Protec ed Areas n
(v)
The entire tract owned by General Electric (Figure 1-3) is enclosed by agri-cultural fencing with appropriate posting and for=s the controlled area 5-1
NEDO-21326C3 Jcuuary 1981 (oxclusion area) as described in Chapter 3 In additien, the DNPS site to I
the north is si=ilarly fenced and postec and these areas to the eas*. Of the Morris Operation site occupied by the Oresden cooling lake and its inlet and cutlet canals are also cor. trolled as exclusien areas by Commonwealth Edison.
?.e cc bination of these areas, supplemented cy county read right-of-ways, provides adequate distances in all directions from che Morris Cperation stack location tn which occe.pancy can be centrolled as required
- ,o assure protecticn of public health and safety.
O 5.2.1.1 Restricted and Protected Areas The restricted areas, as defined in Section 20 3,10 CFR Part 20 are within a 1f-acre protected area en the north 1rn side of the tract (Figure 5-1), en-closed by a chain link fence topped with multiple strands of barbed wire for 7.
total fence height of S f t.
The fence is provided with fence lighti%g and alarm systems for surveillance during hours of darkness and foms a protected area, as defined in Section 73 2(g), 10 CFR Part 73
'a shown in Figure 5-1, facilities located within the protected area include the =ain building, the n[ T adjacent ventilation sand filter and equipment building, three underground v
vaults, the ventilation exhaust stack, the cask service facility, the utilities and service building, the shop warehouse building, the ad=inistration building, the general warehouse, and the water system well and elevated water tank.
Liquid (ncnradicactive) waste discharge lines are routed fr0= the protected area to the sanitr.cy waste treat =ent lagoons and the industrial waste evaporation pond located south of the protected area.
The sanitary waste facilities are fenced, but are not a part of the protected area.
The evaporation pend is not fcaced.
O 5.2 i.2 cates Entrance to the protected area is from the east-west county read (Collins Road),
which bounds the tract on the north side.
Entrances for personnel, road and rail traffic are at the not thwest corner of the protected area Entry is e
controlled from a guard station in the foyer of the administration building p
which includes the personnel entrance and is adjacent to the road and rail gates. Two unmanned gates are lccated on the south and northeast sides of the protected area.
The soutt gr te is for construction equipment access and 5-3
NEDO-21326C3 Jcuucry 1981 The fuel storage basins and the cask unloading basin are constmeted of rein-forced concrete poured on bedrock with a welded stainless steel liner.
The fuel storage basins are filled with demineralized water to a nominal depth of 28.5 ft. The water level may be lowered no more than 2 ft for maintenance or other purposes but at least 9 ft of water is normally maintained above the top of stored fuel.
If the water level falls =cre than 2 ft, pump suction inlets will be exposed. There is no means of accidentally draining the basin, nor can any of the basin water systems inadvertently drain the basin (i.e., the water systems are designed with nonreversible pumps, no drainage system, etc.).
Basin water level is indicated in the control room. The system includes an mudible low-water alarm.
g The cask handling, cask unloading, and fuel storage areas are constructed of con-crete, steel, and other materials that are either nonflammable or fire-retardant.
No significant amount of flammable materials is used in these areas, and other potential fire dangers (bottled gases, etc.) are seldom introduced, and then only under stringent administrative controls. No fire detection or autcmatic fire suppression systems are required in these areas, er in the basin pump
(]
rocm and its extension. Fire extinguishers are strategically located, and.
operation personnel are assigned and trained as a fire brigade. Further pro-tection is provided by surveillance patrols.
Reinforced concrete in the b sin walls and floors has been estimated to have a useful life of more than 100 years, and the stainless steel liner can be expected to have a useful life of more than 100 years because of the nonagressive service environment.
5.5.1.1 Foundation and Excavation The basins are founded on shale be.1 rock (Figure 5-11). Samples of the shale have been tested at ultimate compressive strengths ranging from 6000 to 11,000 pai. Appendix B contains a site survey and foundation report prepared for MFRP construction.I3 The excavation was, overexcavated and backfilled to the south of' Basin 2 to fauilitate expansion of basin storage capacity at some em later date. All 1Nse and disturbed rock was removed prior to concrete con-struction. Backfill consisted of controlled and compacted granular soils.
Concrete mud mats were poured to fill any area excavated more than 4 inches 5-30
NEDO-21326C3 knuary 1981 deeper than required (except for the south wall of Basin 2).
The basin
- all structure is designed to resist pressures frem backfill and soil water where overexcavations were made (south of fuel basin and waste vaults, Figure 5-12).
5.5.1.2 Concrete Structure The flocrs of the storage basins were poured en bedrock and are at least 3 f t 10 in. thick, but are thicker in some areas because of the
-0, 4 in.
tolerance in excavatten requirements.
The basin ficers are designed for live 2
loads in excess of 1000 lb/ft.
Basin walls extend 3 5 ft above grade.
Materials used fer concrete ccnstruction of the basins are typical fer other concrete construction on the Morris Operation site.
Materials used for rein-forced concrete structures were cement conforming to ASTM C150 typ 2;
e o
washed sand; OO e
washed and graded aggregate; and reinforcing steel per ASTM A15, intermediate grade.
e Concrete pours had slump tests and laboratory samples taken, usually at the discharge from the truck, but at times at the point of placenent - partic-ularly en canycn containment walls.
Concrete samples were taken for every pour of 100 yards or less, whenever a pour composition changed for any reason, and cne per 100 yards for pours greater than 100 yards.
A full-time concrete inspection program was in effect during construction.
Reinforcing steel used in the basins is intermediate strength with minimum yield strength 'of 40,000 psi.
Structural welds that carry loads frem one element or reinforping bar to another were not used. Where required, loads were transferred fnsa bar to bar by conventional reinforcement bar laps secured O
in assemblies by steel tie wires. In special cases, U-bolts were used. The k./
only welding permitted was tack-welding the reinforcing steel to brace assemblies away from the forms or to secure imbedded items in position during the concrete 5-33
NEDO-21326C3 Jcnucry 1901 pour.
In :ost cases, asse=bly bracing er i= bed securing was done by the use of additional reinforcing steel or structural steel tack welded to the reinfeccing s
steel assembly.
Isbeds were either welded or cla= ped to this additional steel.
Tack welds were =ade no larger than necessary to produce sound, crack-free welds.
5.5.1 3 Basin Liner The unloading and storage basin co= plex is co=pis*ely lined with 304L stainless steel sheets placed flush against the concrete walls v.d ficers and welded to a gridwork of stainless steel back-up =e=bers e= bedded in the concrete (Figure 5-13).
For the unicading pit floor area, the liner is 1/4-in. thick and is placed over a 1-3/t>in. thick steel plate provided for distributing i= pact leads over the underlying concrete structure.
Additional energy absorbing
=eans as =ay be required by cask drop accident censiderations fer receipt of larger-sized casks will be installed.
The unicading pit shelf liner, also 1/4-in. thick, is placed directly en the ccncrete structure with an -energy absorbing asse=bly placed on top of the liner (seen in Figure 1-13).
For the remainder of the storage basin ec= plex, the ficer liner is 3/16 in.
thick.
~he walls of the unicading pit, including the shelf area, and of the transfer tunnel are lined with 11 gauge sheet.
or the fuel storage basin walls, the liner is 11 gauge sheet from floor level to approximately 16 ft up the wall and is 16 gauge sheet from there to the top of the basin.
The large linr:r sheets (generally on the order of 6 x 16 f t) were welded con-tinuously along each edge to the gridwork of back-up bars and also were slot rh V
welded to e= bedded plates at inter =ediate locations so that the liner is held against the cencrete wall
- J reduce the pCtential for puncture damage.
To faciliate fit-up and to assure high integrity, liner sheets were welded to embeddett stainless steel angles at wall-to-wall and floor-to-wall joints.
Also, tee liner terminates on a stainless steel angle at the top of the basin.
Specifications for liner installation include approved joint designs, welding ry procedures and welder qualification requirements.
All welds were visually inspected and vacuum box tested to assure leaktightness.14 Final verification of liner Jntegrity was provided during basin filling.
5-34
1 NEDO-21325C3 Jcnuary 1981 Through May 1976, the natural heat-dissipating capacity of the at=csphere and O
pool structure adequately =aintained the basin water ta=peratures at less than 340C (111.2 ?).2 Decay heat generated by fuel in the basin is re=oved primarily 0
.by evaporation, with the remainder conducted through the basin walls.
5.5 3 2 safety Evaluation Failure of the basin cooling system is not critical to the safety of the fuel storage system.
~he cooling system has three independent units.
Fach of the two larger units has adequate capacity to dissipate the total expected heat load (6.5 x 106 stu/hr).
In the event of failure of the operational unit, the basin water could be continucusly cooled by the other units while the system is being repaired.
- n the event that both of the larger heat exchanger units should fail, cr it was decided not to activate a carbcn-steel unit, there is enough ti=e to supply sake-up water to the basin while the cooling system is repaired or replaced.
If heat exchanger units were inoperative and the stcra6e basins were full OV of fuel, the temperature of the basin water would 310wly rise (<2 F/hr) and 0
approach boiling in no less than 3 days and possibly longer, as deter =ined by natural conduction and evaporation rates within the building.
( As of January 1979, water tenperature had always been less than 120 'F.)
Meanwhile, verk to 0
repair or replace the cooling system would be initiated.
- n addition, preparations to add : sake-up water to the basin would be sadt: if that should be dee=ed necessary.
If the superstructure covering were removed or blown away, water te=perature would stabiline at abcut 183cy (34cC) because of increased evaporation rate from the open pool.
O rotantia1 1eaus i= the coo 11=. s7ste that cou1d oc=ur as a res=1t of an ae=1 dent have been analyzed and th? resultn given in Chapter 8.
It was ccncluded that the consequence of a leak in the system is insignificant.19 The coolers are 4
periodically inspected for leaks (Table 10-2).
Accumulation of radicactive contaminants in the cooling system components is scnitored, and the system decontaminated when required (Section 7 3 2 3).
{a T
5-w
NEDo-21326C3 Jcuuary 1981 5.5.5.4 Control Room bd The central control room is located in the south gallery area intermediate level (65-ft floor elevation).
The rocm is about 75 x 21 ft in plan, with direct stairway access to the building lobby and secondary access to the unused co=-
puter rocm. Principal items of control room equipment include the main process l
control panel across one side of the room, a control console, and various moni-toring equipment. Fuel storage functions monitored in the control room are listed in Table 5-2.
O Although some functions are normally controlled only from the control rocm (e.g., basin cooler pump and fan controls and well-water pump control), the noncritical nature of all control systems permits establishing local control in case the control room becomes disabled.
The control room is continuously manned, and its location in the main building permits its continued occupancy during all operations, including emergency conditions.
5.5.5.5 Laboratory Area i
The intermediate level of the north gallery area houses the laboratory facili-(]
ties required for fuel storage operation. Personnel access is fran the cor-v ridor which spans the east end of the main building at 70-ft reference eleva-tion.
Equipment is arranged in individual areas as described below:
Speciali::ed counting equipment is housed in a 130 ft2 a.
room located against the canyon wall. The counting room is provided with heavy concrete shielding walls and a labyrinth shielded entry door opening to the accessway described above.
b.
The 630 ft2 laboratory houses a series of fume hoods which provide i
for the ventilation and contamination control required for laboratory operations. The exit door leading to the outside stairway is located in the laboratory.
5.5.5.6 Process Steam Generator Roou The process steam generator, condensate cooler, surge tank, water treatment conden-V sate pumps and other equipment are housed in a separate room. There is a service platfonn at the 60-ft elevation in the room for access to upper equipment levels.
5-50
NEDO-21326C3 Janurry 1981 which is ccnnected to a single leak collection su=p.
The su=p consists of a 6-inch-diameter vertical stainless steel pipe e= bedded in the OV vault wall which extends from the top of the vault to approximately 1 foot below the vault floor level.
It contains a liquid level detector line and necessary piping fer a 5-sps (nceinal) jet-out system.
Auxiliaries for the level detection and jet-out systems, including a scnitoring sample station, are located in the hydraulic equipment room of the =ain building. '4ater frem the jet-out system is routed back to the cladding vault.
5.7 SUPPORTING FACILITIF.S Supporting facilities are descr'5ed in the following sections.
as in previcus sections, these functicns relata1 exclusively to fuel reprocessing are emitted or discussed only briefly.
5.7.1 Utility and service Building q
On the north side of the =ain building is located the single-story high-bay utility and service building (Figure 1-4).
It is 71 x 50 ft in plan and is of conventional steel frame, insulated siding and roof ecnstruction en a grade level concrete foundation. The building is divided into a utility section which houses the plant utility steam system (gas-fired boiler), the decineral-1:ed water system; the primary electrical switchgear; and a personnel section containing change room, lunch room and office areas. The arrangement takes into account the nor nal industrial safety requirements for gas-fired steam generation facilities and for major electrical equipment.
Consideration also is given to isolation of nor lal industrial functions and equipment frem all v;
potential sources of radioactive contamination.
Utility services are not cri-tical to the safety of fuel storage operations.
Interruption of these services' for short periods of time, up to several months, would have no off-site impacts as lang as basin water level is maintained. Principal features are described in the following paragraphs.
OV 5-62 L
NEDO-21326C3 Jcaucry 1981 5.7.1.1 Utility Section nU The 1700 ft2 utility section of the building is divided into two rooms, the larger of which houses the water demineralization and utility boiler systems.
The demineralizer system begins with a carbon filter to remove organic material from water entering the demineralizers.
The de=ineralizer censists of two parallel banks of series cation-anien units with degasification provisions between beds.
It is capable of treating 25 spm continuously or 50 gpm instan-g taneously from the plant utility water supply.
Pumps required for operation, U
distributien and regeneration are located nearby and a 1000-gal demineralized water surge tank is mounted on an overhead platform in the room.
The primary unit of the utility steam system is a 25,000-lb/hr package boiler fired by natural gas which is designed to operate at 270 psig, but is nor= ally limited to 125 psig.
Auxiliaries include a platform-counted 1200-gal condensate return tank and a 300-gal deaerator, as well as condensate return and boiler feed pumps, phosphate and hydrazine makeup and injection facilities, etc.
All normal safety provisions required to assure safe operation and personnel pJ protection and to meet all requirements of the State of Illinois boiler code are included.
A separate 300 ft2 room in the utility section houses the pri=ary electrical distribution switchgear for the plant.
Inccming power from the Commonwealth Edisen distribution system is reduced to 480V prior to entry into the utility building.
5.7.1.2 Service Section The 1800 ft2 service section of the building contains:
l Change room facilities with showers, lavatory and stora6e lockers a.
l to accommodate approximately 100 operating people.
j b.
Lunch roca facilities (stove, sink, refrigerator, etc.) for about j
25 people.
c.
About 450 ft2 I
of office space.
/7
'd 5-63
NEDO-21326C3 knu ry 1981 7 3.2.1 History of Radioactive Material Concentration O
The history of radioactivity in the basin water is shown graphically in Figure 7-1.2 The general trend is a gradual increase in concentration with increasing fuel loading and time, culminating in plateaus and abrupt decreases.
The plateaus may be caused by a reduction in the source, or establishment of a steady-state condition between radioactive material addition and removal.
- he decreases are due to accelerated removal of radiocesium and radiocobalt by the use of filtration and special ion exchange material in the hsin water filter.
,7 3 2.2 Contaminants The principal dissolved radioactive contaminant in the basin water has been fission product cesium with concentrations ranging up to 10-2 Ci/ml.
A means of cesium removal has been found that makes reduction and control of this contaminant relatively simple.
For example, over a 10-week period in 1974, the radiocesium concentration was reduced to one-third of that at the beginning of the period.
The basin water inventory was correspondingly reduced from about 29 to 11 C1.
O in 197s. eurins a a-week period. the raetocestu= concentration was reeucee bv a factor of six and the basin water inventory reduced from 14 to 2 3 C1.
At the end of the latter period, the radiocesium concentration was 0.0009 microcurie per milliliter.2 The MPCw for Cs-134 is 3 x 10-4 Ci/ml.
The ability to dramatically reduce the amount of activity in the basin was the result of extensive studies and tests in which an inorganic =olecular sieve medium, Zeolon,3 was used to selectively remove cesium.
These tests demonstrated that Zeolon-100 could successfully be added to the Powdex syatem and remove about two-thirds of l
the radiocesium per Powdex charge.
By routinely asing Zeolon and adjusting Powdex d
replacement frequency, concentrations are effectively centrolled.
In addition to radiocesium, the radionuclide contributing most significantly to basin water contamination is cobalt-60.
Concentrations of this nuclide in the basin water (typically,1 x 10-4 Ci/ml) are attributed to corrosion products on the surfaces of the fuel bundle which are released to the water, orincipally during p
fuel handling.
Normal filtration and ion exchange reduces the cobalt concentrations l
without special effort.
7-6
NEDO-21326C3 Jcuunry 1981 As the fuel in the basin is increased from about 300 to 750 TeU, the radioactive contaminants, principally radiocesium, will tend to increase. However, with the 7_
k-demonstrated effectiveness of Zeolon in the filter, it is not expected that any increase will tax the existing system.
7.3.2.3 Basin cooler Decontamination After a period of operation, depending on the amount of new fuel received, contami-
)
nants accumulate on the inner surfaces of the cooler piping, tubes, and headers.
~~/
In 1978, a chemical decontamination system was introduced which is available to reduce exposure rates under the coolers to acceptable levels.
7.3.3 Airborne Radioactive Material Sources There are five potential sources that could release radioactive material to ventilation air, where it would be passed through the sand filter and sc=e frac-tion exhausted to outside air via the stack:
fN a.
effluent fecm the LAW evaporator; V
b.
vented gasses frem shipping casks; offgas fecm defective fuel rods in the basin; c.
d.
decontamination activities; and e.
uranium used in MFRP testing.
(])
Although there could be radioative material in the ckmisted effluent from the evaporator, the occurrence would be rare and the amount would be very small.
Vented gasses frem casks are exhausted to the LAW vault, and frem there to the air tunnel and sand filter.
During over 6 years of fuel storage experience, there has been no evidence of gasses leaking fran stored fuel. Incidental airborne contamination fran decon-U tamination activities (and fuel storage area;) could occur, although the use of special enclosures (" greenhouses") and other techniques limit such releases 4
7-8
NEDO-21326C3 January 1981 For comparison, the guideline value for compliance to Appendix I of 10 CFR Part 50 is 15 = Rem /yr to any organ (Regulatory ~ Guide 1.109).
7.7.2.3 Man-Rem Calculations Man-Rem calculations were originally done only for the estimated annual thyroid exposure because it was the maximum dose.
Averages of thyroid exposures were calculated for concentric circles with radii of multiples of 10 miles. These average values were multiplied by the population Within each area which gives an average annual men-thyroid-Rem. The sta of these values for each area out to a radius of 50 miles gives a total of less than four m2.n-thyroid-Rem /yr for y
the period from 1970 to the year 2000.
(An evaluation based on 1-year-old fuel would show an even icwer impact.)
For ccmparison, the population exposure fecm normal background radiation (taken at 100 mR/yr) in the same area is about 665,000 =an-Rem for 1970, to 750,000 man-Rem for the year 2000. Therefore, the radiological impact fecn the Morris Operation is insignificant.
O 7.7.3 Liquid Releases There are no planned releases of liquid wastes from the site. Furthermore, there is no mechanism under normal operating conditions for injection of contaminated water into the waste water treatment system.
7.8 REFERENCES
1.
RESSAR-41 Reference Safety Analysis Report, Vol. 6, Westinghouse, December 1973 and Amendments.
2.
K. J. Eger, Operating Experience - Irradiated Fuel Storage - Morris Operation, Morris, Illinois, General Electric ' $)nny, May 1978 (NEDO-20969B).
2a. State of Illinois, Departmert af *..f.o Health, Monitoring and Rey,ulation of Nuclear Facilities in Illinoi:, Sprin gield, Illinois (1977). The report shows slightly higher levels of radioactivity in the control counties.
7-34 I
NEDO-21326C3 Anucry 1981 3
A pr oprietary product of the Norton Co.
O 4.
The average of 111 Ci gross beta is based on data from 1974 through mid-1976.
This "sverage" has decreased as 1977-1978 data has been incorporated; see NED0-209698, May 1978.
The 111 Ci gross beta value is used in the off-site analysis.
5.
Based on Morris operation experience over = ore than 6 years.
O 6.
T. Rockwell, Reactor Shielding Design Manual, VanNostrand, 1956.
7 R. O. Gumprecht, Mathematical Basis of Computer Code RIED, June 1968 (DUN 4136).
8.
Regulatory Guide 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, March 1976.
O
!lO l
l i
i l
O l
7-35/7-36 L
NEDO-21326c3 Janucry 1981 8.
ACCIDENT SAFETY ANALYSIS (a~)
i
8.1 INTRODUCTION
This chapter centains an analysis of postulated accidents in terms of the cause of such events, the consequences, and the ability of the Morris Operation organiza-tien to cope with each situation.'
()
The function of the Morris Operation is to receive, store and ship irradiated nuclear fuel.
A primary requirement of these operations is to protect the public and employees from excessive exposure to ionizing radiation, as determined by the requirements of 10 CFR 72.68.
Specifically, any individual at er beyond the con-trolled area boundary shall not receive a dose greater than 5 Rem to the whole cody or any organ from any design basis accident (i.e., those accidents described in this chapter).
8.1.1 Release Pathways Exposure of the public and employees might result frem postulated accidents, by direct radiation from the fuel, by airborne release of radicactive raterial, er
.by release of radicactive material to groundwater.
These postulated events are discussed in this chapter.
None of these potential releases have off-site impacts above the requirements of 10 CFR 72.68.
lh 8.1.1.1 Direct Radiation Exposure of tue public and employees could be postulated to result from direct q
radiation from the fuel in storage er by release of radioactive material to the I
b environs. Direct radiation from the fuel would occur only if the water level in the storage basin became too Icw to provide adequate shie'lding.
This would pose a hazard to persons only if they were in relatively close proximity to the basin.
l Loss of water could result from postulated drainage or evaporation of the basin 1
water, but only when a basin make-up water supply quantity or rate is not sufficient t0 keep up with the water loss. Sudden draining of water frem the basin is not
(~')
credible.
i x-l
'See Section 8.11 for references.
l 8-1
NEDO-21326C3 January 1981 8.1.1.2 Airborne Release OV Airborne release of radioactive material could be postulated to result from fbel becoming mechanically damaged sufficiently to release fission gases from 'fithin the plenum of fuel rods. Of the gases released, only Kr-85, I-129, and I-131 would be of concern.
No
'chanism exists in the fuel storage environment to cause an airborne release of particulate radioactive material in quantities sufficient to cause exposures approaching the limits specified in 10 CFR 72.68.
During cask operations in air (e.g., decontamination and venting) particulate releases might occur but in very small quantities, even under the most severe conditions that can be postulated.
These quantities would be Luch too small for an off-site impact.
A criticality can be postulated to occur cy dropping a basket in such a way that all the fuel falls out of the basket and comes to rest in a critical array, or by fuel baskets being deformed into a critical array by a tornado-generated missile.
In reality, however, the above events are cnly marginally possible and the results of either would be substantially less than the criteria of Part 72.68. Since the fuel assemblies are designed to operate in a light water moderated critical array, such a criticality would not cause vaporization of fuel.
8.1.1 3 Waterborne Release Release of water containing radioactive materials can be postulated to occur from the LAW vault intrusion water sump if an inner container leak is assumed (Section 5.6.1.2).
Water from this sump is normally disposed of in the on-site evaporation pond, so that an off-site release would not be likely.
Water from the storage basins can be postulated to be released due to a leak in the basin structure, permitting water to escape to the surrounding rock.
A small amount of water could be released to the ground in the event of a basin water cooler leak.
(Such a leak might also cause a small air release of contaminated water vapor.)
L>
8-2
NEDO-21326C3 Janurry 1981 8.1.2 Accident Description / Discussion U
The following sections contain discussions of various pcstulated acetcents and estimates of the quantity of radioactive =aterial release and projected conse-quences. A sn-"ry of events and the sequence of events involved in pcstulated direct radir. tion and radicactive aterial releases that could result in radiation exposure to the public is shown in Figure 3-1.
No ec=bination of normal and credible accident events has been developed that would result in an off-site release er p
direct radiation that would exceed the regulatory limits for an accident.
\\
Based en several analyses of the accident at Three Mile Island No. 2 (TMI-2),
a similar accident at DNPS would have no un=anageable effects en activities at Morris Operation. A mutual aid agreement exists between Morris Operation and DNPS; this agree =ent and other e=ergency plan i=plementation (10 CFR 50, Appendix E),
assures that operators at Morris would receive early notification of any significant ecer-gency at DNPS.
A release of noble gases and halogens fres DNPS, similar to or greater than at TMI-2 O
would not affect fuel storage safety at Morris.
~he location and constructten of the v
Morris Operation centrol rec =, the availability of respiratory protective = asks and syste=s, the availability of protective clothing, and other radiological emergency preparations at Morris would minimize the impact en Morris Operation of any release frem CNPS.Al Even if it should beccce necessary to temporarily evacuate Morris Operation, the slew less of basin water by evaporation and the ease of replacement negates possible detrimental effects, and protects the public health and safety.
A simultaneous accident at Morris Operation, such as a fuel drop accident (Section S.7), would contribute an extremely Icw additional dese to that from a reactor accident release, less than 0.019 mrem whole body and less than 0.096 sRem thy cid.
8.1.3 Exposure Paths Of the possible exposure paths, enly whole body exposure frem external radiation and exposure through inhalation are consid'ered credible at any off-site location.
p No releases have been postulated that would result in the release of material (such O
as I-131) to farC ands, cattle feed lots, or other sensitive areas that could result in an ingestion dose that would be more than a small fraction of the regulatory limits.
8-3
NEDO-21326C3 Janucry 1981 8.2 LOSS OF FUEL BASIN COOLING O
The basin cooling system is not critical to safety.
'4 hen the cooling system is not in service, the water =ake-up system can be used to replace water lost by evaporation.
Even if the water =ake-up system is out of service, there is adequate time to repair or replace both cooling and make-up syste=s er to provide make-up water from alternate on-site or off-sit e sources.
(The water make-up system includes the water well and all equipmeu related to the normal make-up O
water supply to the basin. )
V A conservative approximation of the time available to provide make-up water if both the cooling system and the water =ake-up system were out of service has been calculated to be at least 9 days. The calculations were based on a constant heat load of 6.4 x 106 Stu/hr, which is the approximate heat load if both basins were full of fuel like that new in storage and that projected to be received.
Other assumpticr s were as follows:
a.
uniform water temperature throughout the basins; O
b.
ambient air at 700F and 70% relative humidity in contact with the basin water surface; and c.
basin enclosure removed, with ::ero air velocity act oss the basin water surface (worst case).
Under these assumptions, the temperature of the basin water would slowly rise 0
(<2 F/hr) for about 3 days and even slower thereaf tse (a nenlinear function of time). The =aximum temperature would be about 0
185 F, and more than 39,000 d
ft3 of water would have to evaporate before the tops of the fuel bundles would be exposed.
This would require scre than 9 days.
The probability of excessively high radiation dose rates resulting from loss of fuel basin cooling is clearly quite small in view of the ample time for repairs to be made and for ' water to be added from any of several sources.
O 8-3a
NEDO-21326C3 January 1981 8.3 DRAriAGE OF FUEL BASEIS O
There are no piping penetrations which, if open, could drain the fuel stor-age basins and there are no potential paths fer siphoning water fecm the basin.
Therefore, to inadvertently drain water fec= the basin the basin liner sust
' be penetrated.
Because the basin structure is below grade and because of the low permeability of the surrounding rock (except for the overburden) and high level of upper strata groundwater, leakage (even if it were undetected) would not uncover the fuel ( Appendix A.13).
s 8 3.1 Basin Liner Rupture Experience An accident occurred in June 1972 that resulted in the rupture of the basin liner and demonstrated the ability of the Morris Operation to withstand and recover from such an incident.
No excessive exposure to ionizing radia:icn was experienced by site personnel or the general public as a result of the incident and no groundwater contaminatien above background levels was detected.
The i= pact en the environment was so slight as to be un=easurable.
~
Difficulties encountered during handling of an empty IF-100 shipping cask resulted in puncturing the liner of the cask unicading pit.
Basin water entered the space between the liner and the structural concrete wall with some seepage into portions of the canyon in the main building.
No special action was necessary to assure the safety of site personnel.
~he timing and sequence of events were as follows.
June 12, 1972 -
bv) 1120 hr Model IF-100 irradiated fuel shipping cen-tainer Serial No. 470033 was tipped while attempting to disengage a jammed lifting yoke and came to rest against the south wall of the' unicading pit.
s I
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i
/
8-5
NEDO-21326 C3 Jcnucry 1981 1130 hr Alarms in the basin leak detection sump indi-(}
cated an inflow of water and it was deter-mined inat the metal lining of the cask unicad-
.g pit wall had been punctured.
~
Approximately 1200 hr Progra=s were started to sample wells adjacent to the main building and plans implemented to remove the cask from the pit and provide a tem-(}
porary repair.
1230 hr Region III Compliance Office of the US AEC, State and GE authorities were notified of the incident.
June 135 1972 -
0700 hr The cask was removed from the unicading basin.
()
1430 hr A temporary patch was positioned over the point of puncture and the outflow of water from the basin to the liner was reduced to a ficw that was handled by the jet-out system (approximately 15 gph compared to a normal rate of 4 gph).
June 16, 1972 -
0900 hr Fabrication of an access chamber for permanent b,/
repair of the liner was started.
s 1200 hr An additional well was drilled northeast of the building in the direction of the nearest cottages.
s I
v 8-6
Jtnucry 1981 June 19, 1972 -
0 1500 hr The temporary patch apparently failed and the leakage rate to the leak detection sys-tes increased rapidly, exceeding the jet-out capacity of the system.
\\
1900 hr The te=porary patch was successfully replaced.
()
The leak' ge rate was deter =ined to be approxi-a mately 15 gph.
June 21,'1972 -
1200 hr Decontamination, inspection, repair and =cdi-fication of the tipped cask was cc=pleted.
June 24, 1972 -
x_/
2230 hr The te=porary patch was re=oved and the access chamber lowered into the unicading basin.
The liner was successfully repaired by welding a stain-less steel plate over the da: aged area.
The regicns of the =ain building where basin water =igrated were the extrac-l tien cell, =echanical cell, and equipment transfer area. Moisture was also found on the wall of the hydraulic equipment rcom but no other leaks were located.
The air tunnel was inspected and found to be dry.
The amount of water that seeped into the process cells was not recorded.
('
Water was periodically pumped from the cells.
Pathways for the leakage were observed to be around pipes that penetrate cell walls i
and through construction joints.
Water also leaked arsund the seams of tne cell wall cladding into the mechanical cell.
(')h l
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I 8-7
i NEDO-21326 C3 Jcaucry 1981 8.3.1.1 Leakage Disposition O
Water loss was carefully monitored from the ti=e the breach occurred at approx-imately 1120 on June 12 until the temporary patch was in place at about 1430 on June 13 Overall loss of water during this perted was about 2700 gal. Of this amount, approximately 400 gal were accounted fer through evaporation and 500 gal through a normal basin filter =edium change.
Another approximately 900 gal were jetted to the LAW vault from the leak detection system.
The re=aining 900 gal were
(}
unaccounted for.
With the basin water activity at about 6 x 10-4 uCi/al, the unac-counted for water contained about 2 mci cotal activity.
The average of samples taken from the basin during this period showed Cs-134 to be 2'x 10 4 uCi/ml, Cs-137 3.4 x 10-4 uCi/ml and Co-60 2 x 10-3 uCi/ml.
During the 4-hour period on June 19 whan the first temporary patch failed and was replaced, approxi=ately 200 gal of basin water containing less than 1.0 mci was unaccounted for.
During the 6-hour period when the second temporary patch was recoved and the permanent repair was accomplished, approxi=ately 1400 gal of water with cen-tamination of approx 1=ately 3 sci was unaccounted for.
mb The total unaccounted-for leakage during the repair period was approxicately 2500 gal.
The disposition of this water, containing an esti=ated 6 =Ci (primarily Cs-134 and Cs-137), is not known with certainty.
It is assu=ed that = cst, if net all of this water retained within the confines of the structure and is contained in minute fissures er flaws in the concrete of storage basins and process cells.
If this water did leak cut of the structure, its probable disposition can be explained as follows:
/]
During construction of the facility, explosives were used to excavate the rock V
formation for the deep structures. This blasting fractured rock formations i==e-diately adjacent to the deep structures.
Water accumulates to some degree in the fracbured rock.
Perched water also collects from precipitation at various levels in these formations.
The fraction of precipitation that enters the perched water zones is not known.
However, a small fraction is sufficient to cause a large amount of dilution of any leakage from the basins.
For example, approx-O imately 6.3 million gallons of precipitation falls on the protected area in an V
8-8
NEDO-21326C3 Jcnuary 1981 average year (e33 in. annual average rainfall over an approximate 550-ft x 550-ft area).
If 10% of this rain entered the perched water reservoirs and if the 2500 gallens of uraccounted-for water is assumed to have leaked from the struc-ture and uniformly mixed with the perched water, tne dilution factor would be approximately 250.
Testa indicate that there is no significant connection be-tween the perched water formation and the aquifers supplying water for domestic or agricultural purposes.
These aquifers are located far below the perched water zones and if contaminated water did leak from the structure, it is unlikely tLat it would ever migrate to the lower aquifers.
It is most probable that the basin water would be captured in the perched water ::enes, becoming more diluted and gradually dispersing.
The.ce are no known water wells in the area that tap perched water zones (other than Morris Operation and DNPS sample wells).
8.3 1.2 Monitoring Program Results I= mediately after the incident, three wells were monitored for radicactivity.
An additional well was drilled 100 f t ~ northeast of the fluorine building to determine if any radioactive material was migrating toward the inhabited cot-tage area.
l The wells were not pumped before taking a sample.
Consequently, samples were taken from stagnant pools. The wells were sampled once a day from June 12 to June 21, 1972.
Afterwards, through August 14, the wells were sampled twice a week and, currently, sampling is done once every 2 weeks.
In addition, the main water well for the plant has been sampled several times since the incident.
l The analyses of all samples taken indicate no activity above background levels I
(<0.5 cpm /ml).
8.3 1 3 conclusions l
t Recovery from the incident was rapid and successful.
The liner was repaired i
l by welding a stainless steel plate over the damage vea.
Corrective actions I
to avoid similar problems were promptly initiated and included:
I l
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8-9
NEDO-21326C3 Jcnuary 1981
+
8.11 REFERENCES O
A1.
According to recent studies in the U.S. and abroad, significant evidence has accumulated to indicate that the consequences of a hypothetical fuel =elting accident may be less than currently predicted by at least one or two orders of magnitude; see staff reports, Appendices E, F, and G, Report of the Presi-dent's Commission on the Accident at Three Mile Island.
1.
C. V. Moore, Design of Barricades for Hazardous Pressure Systems, Nuclear Engineering and Design (1967).
2.
Sandia Laboratories, Full-Scale Tornado-Missile Impact Tests, July 1977 Electric Power Research Institute Report No. EPRI NP 440.
3 See Subsection 5.6.3, Design and Analysis Report of the IF-300 Shipping Cask, GE Document NEDO-10084-1, Docket 70-1220.
4 N. R. Horton, W. A. Williams, and J. W. Holt::elaw, Analytical Methods for Evaluating the Radiological Aspects of the General Electric Boiling Water Reactor, March 1969 (APED-5756).
5.
RESSAR 41, April 1974.
6.
Attenuation in Water of Radiation from the Bulk Shielding Reactor:
Measure-ments of the Gamma-Ray Doso Rate, Fast-Neutron Dose Rate and Thermal Neutron Flux, July 8,1958 (ORNL-2518).
O o) u 8-40
NEDo-21326C3 Jcnutry 1981 9.2.3.6 Senior F.ngineer - Licensing and Radiological Safety O
The Senior Engineer - Licensing and Radiological Safety repor:s to the Manager -
Morris Operation and is responsible for coordinating site regulatory =atters with local, State, and Federal regulatory agencies, and directing the site environmental program activities.
The incumbent reviews facility and operating procedure changes to determine the need for a nuclear safety review and reviews fuel data to assure conformance with criteria for fuel storage.
O 9.2 3 7 Plant Safety Committee In addition to the organi::ation shown in Figure 9-2, a plant safety committee is established within the Morris Operation.
Plant Safety Committee members include: Manager - Plant Operations; Manager - Plant Engineering and Maintenance; Manager - Quality Assurance and Safeguards; Plant Sarety Supervisor and Senice Engineer - Licensing and Radiological Safety, who serves as the committee secretary.
The Manager - Morris Operation normally will be a member of the committee.
How-ever, at his discretion, when items of particular significance are being considered (e.g., in the evaluation of a major operational safety matters and development t
of recommended changes in facilities or procedures affecting safety margins),
he serves as chair an of the committee.
The Plant Safety Committee exercises jurisdiction over those matters having radiological or nuclear safety implications, with review and approval authority.
9 2.3.8 Trained and Certified Personnel General Electric has, and will maintain at its Morris Operation, an adequate complement of traind and certified personnel to operate the facility.
93 TRAINING PROGRAMS To provide and maintain a flexible, well-qualified work force for safe and efficient operation, a comprehensive training program has beca implemented.
p Training includes:
U 9-7
NEDO-21326 C3 Janunry 1981 a.
Crientation and Indoctrination b.
Radiation and Industrial Safety c.
Se,cunity/ Safeguards d.
Emergency Brigade Training n
e.
Quality Assurance V
f.
Basic Plant Facilities and Organization g.
Fuel Receiving and Storage Operations 4.
Utilities and Operating Systems These training programs are adapted rec = the progrs=s originally prepared for fuet_ reprocessing operations and are believed to be nere ec=prehensive than Aculd normally be required for fuel storage functions, only.
The a cunt of training and retraining each individual receives is directly related to his function. All personnel are provic;d general Orientation courses which include description of the Merris Operation and its functions, plant safety censiderations, secu,
require =ents, and e=ergency plans and general E
procedures.
9 3.1 Operator Qualification, Training, and certification All personnel assigned duties involving operntion of systems and equipment directly related to movement of casks, leading or unloading of casks, =ovement of fuel, operation of basin water cooling or cleanup systems, radioactive waste scnagement operations, and other activities in the cask receiving and fuel storage areas, including operations supervisory persennel, shall be trained, tested, and certified by General Electric as qualifi d to perfor= specified duties under e
a program approved by the USNRC.
(n) v 9-8
NEDo-21326C3 Anucry 1981 9.4 NORMAL OPERATIONS O
9.4.1 Plant Procedures P. ant procedures are discussed by category in the fol10 wing paragraphs.
Systems and equipment requiring personnel certified for specific functions =ay be oper-ated by noncertified personnel only if under the direct visual supervision of-an individual trained and certified for the specific operation.
O 9.4.1.1 Morris operation Instructions (MOI's)
A system of specific written instructicns provides guidance and direction for performance of Morris Operation activities.
The instructions provide for proper safety, quality, and functional considerations in the planning and implementation of plant activities, including administration, licensing, plant engineering and maintenance, =aterials, operations, quality assurance, safeguards, safety, field services and transportation.
O 9*i2 st =e re c er t1== rroceeure- (soe r
Cperation of Morris Operatien facilities are in accordance with a system of Standard Operating Procedures designed to provide detailed guidance and con-trol for all anticipated conditions.
Individual procedures are prepared by the Plant Operations Unit and approved by the Plant Safety Committee before being implemented. The Plant Operations Unit is authorized to codify standard procedures on an interim basis as required to cover specific conditions arising during operations. Standard Operating Procedures are modified only after due censideration of the safety implicaticris of the change.
Cperating ;ctivities are monitored on a shift-by-shift basis by the supervisory staff for compliance with Standard Operating Procedures.
9.4.1 3 Safety Manual To provide the necessary control of work involving ionizing radiation and p
radioactive materials a system of radiation protection standards has been developed and documented in the Safety Manual.
The Manager - Quality Assur-ance and Safqsuards is respcnsible for the overall administration of the 9-9
NEDO-21326C3 Janusry 1981 requirements set fcrth in the Safety Manual.
Deviation fecm t!!e established C,)'%
requirements say be required from time to ti=e.
~hese may be en a planned basis, under special operating conditions, er there =ay be deviations required by emergencies.
Planned deviations must have prior approval of the Manager -
Quality Assurance and Safeguards or his delegated representative.
Emergency deviations must be reported promptly to the Operation Supervisor en duty who, in turn, notifies the Manager - Quality Assurance and Safeguards.
-O 9.4.1.4 Special Work Procedures Li Special work procedurcs for cases involving nonstandard operations include
=odifications to standard operating procedures and supplemental operating instructions, prepared for interim use on a controlled basis and based on specific evaluation of safety implications.
There are definite time limits on such special authorizations during which off-standard conditions are to be corrected or established requirements revised.
Special work procedures are approved by Quality Assurance and Safeguards, Plant Operations, and the unit performing the wrk.
9.4.1.5 Regulated Work Procedures An essential element of the systems for control of plant safety is the require-
=ent that fermal authorisation be provided for all operating, =aintenance or repair activities which incolve potentially hazardous conditions, i.e., work in radiation or contaminated areas. The Regulated Work Procedure system is designed to assure that such work is accomplished in a safe and efficient manner in accordance with the standards and requirements set forth in the Safety Manual.
O V
Regulated Work Procedures dccument prescribed requirements and limits for special work to be observed prior to beginning each task.
Responsibility for the procedural system is assigned to the Manager - Quality Assurance and Safeguards, including provisions for shift-by-shift monitoring of activities for compliance with centrol requirements, and maintenance of necessary records of such activities.
Regulated Work Procedures are approved by the Plant Safety Committee.
O l
O t
9-10
NEDO-21326C3 Jcnunry 1981 9.4.1.6 Equipment Maintenance Programs O'
i A Work Request System is employed at the Morris Joeration for initiating requests for maintenance, preventive maintenance repairs, modifications, altera-tiens and new installations. Work Requests are reviewed by Plant Engineering and Maintenance, Plant Operations, and Quality Assurance and Safeguards for conformance to plant procedures and instructions.
Equipment maintenance is performed in accordance with manufacturer's recommended practices and operating experience.
Overall responsibility for equipment maintenance is assigned to
)
the Manager - Plant Engineering and Maintenance.
Assistance is provided by s
other plant operating components, as required, to assure that safety and oper-ability criteria are correctly interpreted and performance capability maintained.
9.4.2 Records and Reports Complete files of activitien relating to plant safety are accumulated to demonstrate the adequacy of design and construction safety censiderations and to assure consistent application of safety principles and objectives to plant i
operation and maintenarce.
9.4.2.1 Record Retention Documented records of plant safety assurance activities are maintained to demonstrate that control requirements have been set, including the procedural system documentation and compliance records notri in the preceding paragraphs; environmental monitoring program : eports; perronnel exposure data and regu-latory activity files.
2 O9a3 r e111er a=41ric *1o==
Major modifications of Morris Operation facilities (those related to nue:aar safety) are subjected to a comprehensive evaluation and analysis in accordance with SFSO procedures, which provide a formal progrpn for design review and quality assurance. Minor modifications and tests and experiments are performed under provisions of Section 9.4.4.
9-11
NEDO-21326 C3 Januerf 1981 9.4.3.1 Safety Evaluation and Project Planning
'4 hen a =ajor =odification or project is proposed, a study of the concept develops technical criteria and preliminary specifications, as 'well as other data necessary for a pre 14-inary safety evaluation (1, Figure 9-3).
his evaluatien is performed by a function within SFSO (Licensing and Transportation) that is separated f~ =
organizatienal ec=ponents directly involved in the proposed project activity.
Engineering data and reco==endations fec= other SFSO cc=penents are censidered in this evaluation, including recc==endations f.~:= the Plant Safety Cc==ittee.
(
The evaluttic" determines the need fer licensing action, as well as special studies er other evaluation of the proposed activity.
'he technical criteria, safety evaluation, and other data (such as inco=ing fuel scheduling, =anpower availability, etc.) fer the basis of a project plan developed by Fuel Storage Projects and coordinated with the Morris Opera 1cn.
In sc=e cases, the project will be executed at Morris Operation without further participation by Fuel Storage Projects.
The plan is presented to =anage=ent IN]
of SFSO and NEPD for approval.
'ihen all administrative and technical require =ents have been satisfied, a project authorization is issued by Manager - SFSO.
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9-11a i-l L'
NEDO-21326C3 Jcaucrf 1981 9.4.3.2 Project Design Activity O
Nuclear safety related and major design projects are conducted by Fuel Storage Projects, with support furnished ~ by Morris Operation for those requirements that can best be satisfied at that location.
Design activity (2, Figure 9-3) results in established functional classifications, specifications, drawings, and other documentation, all subject to an intensive review.
Each document is reviewed by all appropriate organizations within SFSO, including Morris Operation, with requirements that each organ ation approve the document prior to issue.
The various features of the design are also subject to engineering reviews, including design verification reviews.
Throughout the design activities, Quality Assurance Programs personnel =cnitor and check compliance with the Quality Assurance Plan, especially the inspection and monitoring of vendor and contractor activities.
9.4 1 3 Licensing Activity Depending upon the centent and nature of the project, Licensing and Trans-portation may provide an environmental report, final safety analysis report, i
and special safety studies (3, Figure 9-3).
Special safety studies may be requested by Fuel Storage Projects, by Quality Aasurance Programs or by other
=anagement including Manager - Morris Operation. Manage =ent and personnel at Morris Operation provide contributions to licensing activities, especially in health physics and environmental fields.
Licensing activities continue as necessary to obtain regulatory approvas of changes or modifications where required.
9.4 3 4 Project completion In the case of a major project, a Fuel Projects Engineer will be assigned the project responsibility for construction, installation, testing, startup, and related activities (4, Fir.ce 9-3).
The Manager - Morris Operation retains full responsibility for the safety of all other activities involving receipt, transfer, or storage of nuclear fuel or other radioactive materials, including operation of the facility during modification.
9-13 i
NEDO-21326C3 Jcaucry 1981 The Morris Operation will furnish supporting services, liaison with local government agencies, etc., as may be required.
The project and site manage-ment teams coordinate activities during project execution to achieve mutual goals in accomplishing both project and operaticnal activities.
Plant Proce-dures for the new facility or function are developed and implecented as described in Section 9.4.
Upon completion of startup and turnover operations, all project docu=entation is completed and filed (both Morris and San Jose sites), and responsibility for the new facility or function assumed by Morais Operation.
9.4 3 5 Audits and Reviews Policies and resulting requirements established for Morris Operation require periodic audit and review of the varicus aspects of h.11 storage activity.
General topics for audit include:
Nuclear criticality safety e
a e Radiation protection e Physical security e Emergency plans e Environmental protection e Quality i
Internal aucits are conducted by Morris Operation management in safeguards, criticality, and radiation safety.
Formal audits and reviews are conducted I
by teams from other Nuclear Energy Group and SFSO components in accordance l
with established Group and SFSO Policies and Procedures.
I i
9-14
NEDO-21326C3 Jcnucry 1981 9.4.4 Changes, Tests, and Experiments O
Changes in the facilities described in this report or procedures described
{
in this report, and tests or experi=ents not described in this report related
[
to receipt, storage, and transfer of spent fuel, may be perfe ced without prior approval of the Nuclear Regulatory Coc=ission provided that such changes, tests, and experiments do not involve significant unreviewed nuclear safety or environ-t mental issues, nor cequire a change in Technical Specifications or other license l
I ccnditiens.
These activities are conducted under provisions of 10 CFR 72 35 l
V I=plementation of such changes, tests, and experiments is acce=plished as directed by applicable procedures.
In general, the procedures require an appropriate analysis and evaluation, with ccccurrence in proposed activity by appropriate Morris Operation and SFSO staff functions, and license a=end=ent activity when appropriate.
9.5 EMERGENCT PI.MS
[]
9.
5.1 Purpose and Scope
v.
E=ergency plans are established and persorr.el are trained in e=ergenc. procedures so that effective actions can be taken under the stress of emergency ocnditions.
2.e interrelated e=ergency plans for Morris Operation are dias. sed in Figure 94 n:e plans and procedures related to radiological emergencies tre enciesed withi= the dashed line.
(3.e Physical Security Plan and related provisions are not discussed i= this document.)
O v
4 O
9-15
NEDO-?1326C3 Jant.cr 1981 Emergency planning at Morris Operation is related to the overall emergency O
planning of Isaeral Electric's Nuclear Energy Group, and radiological assis-tance plans cr the State of Illinois and the Department of Energy.
An arrange-ment has been established between Morris Operation and Commonwealth Edison (Dresden Nuclear Power Station) for mutual assistance in emergency situations.
Likewise, emergency assistance arrangements have been =ade with law enforce-ment, :nedical, arsi other local agencies and services. l*
O O
i O
O.
'See Section 9 7 for references.
l 9-17
... -_-_= - - _,,,-_..-
. _ -.... - _. _ _ _ =
2 NEDO-21326C3 J
Jcnuary 1981 10.
OPERANON SPECIFICATIONS i
O In accordance with requirements of 10 CFR 72, f
proposed technical specifications for Morris Operation have been sutcaitted to the USNRC.
i Therefore, Chapter 10 his been deleted.
O-I f
4' l
i O
O i
9 O
10-1/10-2 i< - -,,, _. ~ -,, _ _,. - -.,.,. _. ~,, _ _. _, - _.... - _. _... _ - _. _. _... _. - _ _ _ - _ _. - - - - - _.
.C O-21326C3 Jr.nuary 1981 11.
OUAU T ASSURANCE (m*) 11.1 IrRCDUCH ON 2e activities at Merris Operatien are cenducted in acccedance with a qualitf assurance plan reviewed and accepted by the USNRC and i:ple ented by instructiens t
and pmcedures at the Morris facility.
Se quality assurance plan is docu=ented as j ;
- e Spent yuel Services Operation CumMty Assurance Plan (NECO-20776).
A nicrofiche 5 -
copy of this plan is included in this repert.
tL
^
11.2 QUACH ASSURANCE ELF.
i i
t The initial design and constructicn of de Morris facility as a fuel reprocessing l
i plant ca=e under a quality assurance program developed by General Electric.
t During the ecnstructicn peried, the USAEC - then the regulatory agency --
f increased its e=pnasis en the specific rethods of acnievia.g quality assurance, i
proposing a=end=ent of 10 Cy3 50 to include Appendix 3, "Ouality Assurance Criter:.a for Nuclear Pcwer Plants."
m r
s 4
P 3efere Appendix 3 was published, General Electric had inccrpcrated quality assumee i-provimiens into the ever-all safety assurance ;regra: fer the reprecessing plant.
I; ii Except for specific equi-e=ents related to decu=ented recced ace"-"'ation, the i
key ele =ents called for in the then-propcsed a:end:ent (as applicable to fuel i,
l,i reprocessing facilities) had been included in the General Electric pregras, which was docu=ented in Supple =ent 3 to the " resign and Analysis Reper - Midwest Fuel j
i' Recovery Plant."
Construction of the facility was cc=pleted under this progra=.
Af ter the decision not to operate the facility as a eprocessing plant, but i
p) te continue fuel stcrage operations, General Electric proposed the i-stallation
(
3,
(
cf a new fuel stcrage system.
This system was licensed by USNRC in Dece ber l;
1975. 2e design, fabrication, and installatien of this syste= were perfer=ed under the currect quality assurance plan, wnich is in acccedance with applicable requirements of Appendix 3,10 Cya 50.
)
v 11-1
NEDO-21326C3 Jcnucry 1981 11 3 STRUCTURES, SYSTEMS, AND CCMPONENTS IMPORTANT TO SAFE"!
n The structures, systecs, and cc=ponents ~.~portant to safety are listed below, with a basis for designation, t
Fuel storage basin - concrete walls, floors, and expansion gate.
The l
a.
basin's concrete structure is a principal elecent in protection of stored fuel, and in the isolation of basin water from the enviren=ent.
(mj b.
Fuel storage basin - stainless steel liner.
Be liner for:s a second element in fuel protection and basin water isolation, facilitating decontamination.
Fuel storage system, including baskets and supporting grids.
~he c.
storage system is a principal element in protection of stored fuel.
d.
Unloading pit docrway guard.
~his device is designed to prevent a leaded fuel basket frem being tipped so that fuel bundles cculd fall into (q) the cask unloading basin.
The unicading pit doorway guard is an element in protection of fuel during =cvement of loaded basket.
I e.
Filter cell structure. Be concrete cell, part of the basin pu= proc =
area, provides radiation shielding to reduce occupational exposure.
n v
11-2
i NEDO-21326C3 Jtnunry 1981 A.7.5.2 Shipping and Disposal Costs OO Shipping and burial cost estimates include the 1978 costs of shipping containers (ncnreusable), transportation fees, and burial charges at a low-level waste dis-posal site. The cost estimate includes weights and volumes of materials based on past experience of the Morris Cperation.
The transpectation costs assume that the waste will be transported to the Hanford Reservation near Richland, Washington.
/%U Disposal of " clean" materials is not included in the costs shown in Table A.7-3 since noncontaminated items are not addressed in this plan.
(See Section A.7.2.2. )
A contingency of 25% of the decommissioning cost (Table A.7-3, Tasks 1 through 4) was included in the total cost shown.
A.7.5 3 Financial Assurance The decommissioning costs for General Electric's irradiated nuclear fuel storage facilities near Morris, Illinois, estimated to be $6,033,000, are s=all co= pared to the total assets of the General Electric Co=pany.
There fore, it is unlikely that General Electric would be unable to meet the financial co=mitments generally associated with the decom=issioning activities as outlined and esti=ated.
On April 15, 1980, Dr. Bertram Wolfe, Vice President and General Manager, Nuclear Fuels and Services Division, General Electric Co=pany, submitted a letter to the Nuclear Regulatory Commission concerning financial arrange =enes for deco =missioning the Morris Operation.
This letter is reproduced in Figure A.7-1.
By action of the Board of Directors in meeting on April 27, 1979 (Minute #9640, April 27, 1979), a Vice President of General Electric Company may axecute su'ch an obligation on behalf of the Company.
A copy of this action of the Board was attached to Dr. Wolfe's letter of April 15.
A. 7-114 s
NEDO-21326C3 Jcnucry 1981 GENER ALh ELECTRIC otscnw ctrcmc come
- 75 CumTseCR.S EMU C
...c.c4ure.
....m.
Ca..ER?m.as WOLrt
...m............
April 15,1980 Office of Nuclear Material Safety and Safeguards U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Attention:
R. E. Cunningham, Director Fuel Cycle & Material Safety
SUBJECT:
FUNDS FOR DECOMMISSIONING MORRIS OPERA 7 ION Occket No. 70-1308 Gentlemen:
(m General Electric's general revenues and retained earnings, as snown d
by the 1979 annual report, are sufficiently large that, at the time of decommissioning, General Electric will have available the resources deemed necessary to satisfy its obligation to dacommission its Morris Operation near Morris, Illinois used for the interim storage of spent fuel.
The decommissioning of the Morris Operation will be carried out by General Electric in accordance with then applicable federal laws and regulations.
Attached is a copy of General Electric's Board Resolution 59540 dated April 27, 1979 concerning the execution of contracts and other instruments which authorizes a Vice President of General Electric Company to sign _this letter.
Sincerely, J
i
- f)p.,k m G'J.Wi,_
l Attachments Figure A.7-1.
Letter frem Dr. Bertram Wolfe, Vice President and General Manager, Nuclear Fuel & Services Division, l
Regarding Financial Arrangements for Decommissioning Morris Operation
(
A.7-15/A.7-16 I
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