ML20214P049

Application for Amend to License SNM-1107,authorizing Release of Industrial Waste Treatment Products,Consisting of Calcium Fluoride & Other Mixtures,For Distribution to Briquette Mfgs.Revised License Condition Encl.Fee Paid

ML20214P049
Person / Time
Site:	Westinghouse
Issue date:	04/24/1987
From:	Nardi A WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Crow W NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
References
28149, LA-87-20, NUDOCS 8706030185
Download: ML20214P049 (14)

Text

{{#Wiki_filter:- (3 76-ns7 o pock m o NN u m m m m-ss f APR 3 0190 ) -{ LA 87-20 o ss {Ag[jd J g345 Westinghouse Water Reactor Electric Corporation Divisions g p m aacass s S w e q April 24, 1987 $EN ? 'U. S. Nuclear Regulatory Commission E g3 019 q ATTN: Mr. W. T. Crow M' Uranium Fuel Licensing Branch q E Office of Nuclear Material Safety and Safeguards y Division of Fuel Cycle and Material Safety cn j Washington, DC 20555 A Gentlemen: REF: SW-1107, Docket 70-1151 Westinghouse hereby requests an amendment to License SW-1107 to authorize free release of industrial waste treatment products (primarily calcium fluoride) and other homogenous mixtures, in which the mean concentration of uranium constituents does not exceed 30 picocuries per gram, without continuing EC controls. We are requesting this amendment to authorize distribution of dried calcium fluoride to briquette manufacturers to be mixed with other steel flux forming materials, briquetted, and further distributed to steel manufacturers in the production of steel. Enclosed are revised Section 4.5 describing the license conditions and new Section 1.11 describing in detail the proposed activities. Westinghouse contends that this use will have no adverse effects on members of the public or the environment since the uranium levels in the calcium fluoride produced at Columbia are the same or lower then those found in naturally occurring calcium fluoride. In addition, the slightly increased levels of U-235 will result in negligible increases in specific activity or calculated radiation dose levels from the calcium fluoride.

Finally, Westinghouse expects a positive environmental impact from this approach (when compared with the existing practice of land burial of the calcium fluoride) since the calcium fluoride will be eventually encapsulated in furnace slag, at uranium concentrations well below 30 pCi per gram, resulting in decreased fluoride leachibility and decreased radiation dose levels to the public.

This proposal would also free up valuable space at the chemical. landfill and conserve natural resources. If you have any questions, please write or telephone me at (803) 776-2610, Extension 3247. -m Very truly yours, q _ h.M. ce 10- ,[... f' T 90 Dat3Ch'#' # ' Received Bl* * * * '

• p.M...... A. J. Nardi, Fanager i

NES License Administration WP1452E: b ' [ 8706030185 870424 Enclosures [@ 00

1.11 Nonradioacti;m Industrial Products In the ADU processing lines, ammonium fluoride, water and ammonia - are the major components of the waste stream from the plant. This waste stream is treated in an advanced wastewater treatment process which removes the residual uranium from the liquid waste to less than 0.3 ppm. Slaked lime is then added to precipitate calcium fluoride. The resultant slurry is then processed through a distillation column where the ammonia is steam-stripped to ammonium hydroxide and returned to the ADU process. The calcium fluoride is then transferred to a holding lagoon and/or a dewatering device to increase the solids content of the calcium fluoride to greater than 60 weight percent. Dewatered calcium fluoride is discharged into storage / transport containers. The calcium fluoride is then transported off-site in closed vehicles or transport containers. A comprehensive sampling and analysis program exists to assure that uranium concentrations in the calcium fluoride meet the conditions specified in paragraph 4.5. For calcium fluoride obtained directly from the holding lagoons, samples of calcium fluoride are obtained from representative portions of the lagoon and analyzed for uranium content. An analysis is then performed to assure that the mean value for the calcium fluoride is less than 30 picocuries per gram of dry material. For calcium fluoride obtained directly from the distillation column, representative samples are obtained from either the dewatered calcium fluoride discharge or from the still feed tank and analyzed for uranium content. Again, uranium concentrations must meet the conditions of paragraph 4.5. Note: Dewatered calcium fluoride discharge sampling may replace lagoon sampling as determined by the Radiation Protection Component. Sample analyses from the still feed tank will be correlated to the dewatered calcium fluoride. Docket No. 70-1151 Initial Submittal Date: 4/24/87 Page No. 1.11-1 License No. SNM-1107 Revision Submittal Date: 4/24/87 Revision No. 10

.~. ~ Dewatered (greater _ than 60 'edo. solids). calcium fluoride is. transported to -off-site briquettors in closed vehicles or containers to prepare the material for use as a fluxing agent. The uranium concentration in calcium _ fluoride generated at the Westinghouse Columbia Plant is in the same range as natural calcium fluoride (fluorspar) used as a fluxing agent in ~the manufacture of steel as shown below: Source Uranium Level (ppm)- Seaforth Mineral and Ore (Natural) 14.0 - 14.2 Mercier (Natural) 2.53 - 2.78 Westinghouse Electric (Columbia, SC) 1.41 -

4. 01 30 pC1/gm, 5 w/o U-235 12.6 Consequently, the environmental impact of calcium fluoride generated at the Columbia Plant will not differ significantly.from natural sources. To demonstrate this, we have calculated the dose 4

to a worker involved in briquetting of this material using the following assumptions. t 1. 250 days per year, 8 hours per day exposure. 2. 30 pCi per gram of total uranium in the. calcium fluoride (12.6 ppm uranium based upon. a specific Ractivity of 2.38 uC1/gm). 3. 5 w/o U-235 enrichment (SA = 2.38 uCi per gram). 4. Dust loading during briquetting 'is _ a maximum for a nuisance (non-toxic) dust of 10 milligrams per cubic meter. 5. Dose conversion factor (U-234) for-inhalation 6 1.3x10 rem /Ci from NUREG-0518. Note: The doses-from ingestion and immersion - are significantly below that from inhalation. The U-234 dose conversion factor is the highest of all the uranium isotopes. Docket No. 70-1151 Initial Submittal Date: 4/24/87 Page No. 1.11 ' License No. SPN-1107 Revision' Submittal Date: 4/24/87 _ Revision No. 10 ~

6. Breathing rate - 20 liters per minute. Therefore, the whole body dose to a briquettor due to inhalation would be given as: 10x10-3gm dust 3x10-11Ci 2.4x10 m 1.3x10 rem 33 6 Dose (mrem /yr) = x x x 3 m gm dust year Ci Dose (mrem /yr) = 0.94 mrem / year *

• 40CFR190 limit is 25 mrem / year This dose can be considered conservative for the following reasons:

1. This calculation assumes that the worker handles only Westinghouse-generated calcium fluoride. 2. The calcium fluoride contains the maximum allowed uranium values (30 pCi/gm) at a maximum allowed enrichment (5 w/o U-235). 3. The worker is continually breathing dust at the rate of 10 milligrams per cubic meter which would be extremely uncomfortable. Consequently, the worker would not likely be exposed to such a level for any extended period of time. 4. All other persons would be exposed to levels far below the assumed 10 milligrams per cubic meter. 5. Briquettors will be routinely exposed to natural uranium in the calcium fluoride. Assume that a worker is exposed to 8.4 ppm natural uranium in calcium fluoride, with a specific activity of 3.6x10-7 C1/gm. 10x10-3gm dust 3x10-12Ci 2.4x10 m 1.3x10 rem 33 6 Dose (mrem /yr) = x x x 3 m gm dust year Ci = 0.094 mrem / year Docket No. 70-1151 Initial Submittal Date: 4/24/87 Page No. 1.11-3 License No. SNM-1107 Revision Submittal Date: 4/24/87 Revision No. 10

Not3: This dosa does not include the daughter products which are in equilibrium with natural calcium fluoride. Consequently, the incremental dose due to our calcium fluoride is: 0.94 - 0.094 = 0.85 mrem / year Once the material is briquetted, it is transported to a steel manufacturer for use as a fluxing agent. A survey. conducted for Westinghouse indicated that the annual usage of " natural" fluorspar is approximately 170,000 tons per year. Therefore, the Westinghouse Columbia Plant rate of approximately 1,000 tons per year would not significantly impact the current supply system, and the uranium it contains would be diluted before use in the steel making process. Uranium in the briquette becomes part of the slag during the steel making process (> 99%). This is based upon a considerable amount of work performed by the National Lead Co. at the Feed Materials Plant in Fernald, Ohio and by Union Carbide at the Oak Ridge National Laboratory.1 Based upon this evaluation, the total ) uranium level in the slag will increase to approximately 2.66 ppm (for 15 ppm, 3% U-235 in the calcium fluoride, in an electric furnace) from approximately 1.98 ppm from calcium fluoride from " natural sources." The increase in activity is to approximately 2.52 pCi per gram of 1 H. Cavendish, " Treatment of Metallic Wastes by Smelting," J NLCO-ll57, September 25, 1978. 8. Heshmatpour and G. L.

Copeland, "The Effects of Slag Composition and Process Variables on Decontamination of Metallic Wastes by Melt Refining," ORNL/TM-7501, January 1981.

E. W. Mautz, et al., " Uranium Decontamination of Common Metals by Smelting - A Review," NLCO-lll3, February 5, 1975. Docket No. 70-1151 Initial Submittal Date: 4/24/87 Page No. 1.11-4 License No. SPN-1107 Revision Submittal Date: 4/24/87 Revision No. 10

sitg from 1.29 pCi per gram of slag. This is the worst case activity increase and is well below the present licensed free release limit of 30 pCi per gram. The environmental evaluation performed for the briquetting operation will be conservative with respect to the steel making process for the following reasons: 1. Westinghouse-generated calcium fluoride will be mixed with natural calcium fluoride. 2. The uranium in the calcium fluoride transported to a steel mills will be encapsulated in the briquettes. 3. Uranium will be "encapsulateda in the slag at concentrations less than 30 pCi per gram. Consequently, doses to the public will be well below 40CFR190 limits. Docket No. 70-1151 Initial Submittal Date: 4/24/87 Page No. 1.11-5 License No. StN-1107 Revision Submittal Date: 4/24/87 Revision No. 10

4.5 Nonradioactiva Industrial Wastes Under the auspices of 10 CFR 20.302, industrial waste treatment products (primarily calcium fluoride) and other homogeneous mixtures in which the mean concentration of uranium constituents does not exceed 30 picoeuries per gram shall be released for disposition in a chemical disposal

site, industrial

landfill, or to off-site briquette manufacturers (calcium fluoride only) without continuing USNRC licensing controls.

4.5.1 A sampling plan shall be implemented which will characterize the industrial wastes in accordance with NUREG/CR-2082 as follows: 4.5.1.1 The estimation of the population mean for the uranium concentration shall be representative of the industrial wastes located onsite and any subsequent shipment of this material.

4. 5.1.2 The sample size, used to calculate the mean value, shall be determined such that the 95 per cent confidence limit of the mean value shall be less than 25 per cent of this value (e.g., + ts/ [li<.25 O where "t" is from students t tables, "s" is the standard deviation, "x" is the mean, and "n"

is the population size. In addition to these requirements, the sampling plan shall provide a minimum confidence level of 95 per cent that the true mean value determined for these industrial waste products is less than the maximum allowable limit of 30 picocuries per gram of dry material. Records pertaining to the release of these materials shall be maintained along with the names of the individuals receiving these materials. Reasonable efforts shall be made to reduce the level of contamination in line with the ALARA concept. 4.5.2 Calcium fluoride released to off-site briquette manufacturers shall contain a minimum of 60% solids. Docket No. 70-1151 Initial Submittal Date: 4/30/83 Page No. 4.5-1 License No. StN-1107 Revision Submittal Date: 4/24/87 Revision No. 10

TABLE OF CONTENTS Section Pg Table of Contents i Schedule of Revisions 111 Revision Record vili 1. Safety Demonstration Table of Contents 1.0-1 1.1 Introduction and Corporate Information 1.1-1 1.2 General Description of Operations 1.2-1 1.3 Site Description 1.3-1 1.4 Organization 1.4-1 1.5 Personnel 1.5-1 1.6 Emergency Procedures

1. 6-1 1.7 Radiological Safety Procedures 1.7-1 1.8 Nucleo." Criticality Safety 1.8-1 1.9 Processing Operations 1.9-1 1.10 Off-Site Release Evaluations 1.10-1 1.11 Nonradioactive Industrial Products 1.11-1 2.

Minimum Specifications and Capabilities Table of Contents 2.0-1 2.1 Scope of Licensed Activities 2.1-1 2.2 Facility and Equipment Design Features 2.2-1 2.3 Nuclear Criticality Safety Limits 2.3-1 2.4 Minimum Conditions of Operation 2.4-1 2.5 Emergency Plan

2. 5-1 2.6 Operating Procedures 2.6-1 2.7 Environmental Monitoring 2.7-1 2.8 Decommissioning 2.7-1 3.

Administrative and Procedural Requirements Table of Contents 3.0-1 3.1 Administrative Controls 3.1-1 3.2 Surveillance Requirements 3.2-1 3.3 Technical Qualifications 3.3 1 Docket No. 70-1151 Initial Submittal Date: 4/30/83 Page No, i License No. StN-1107 Revision Submittal Date: 4/24/87 Revision No. 10

TAELE OF CONTENTS Section Page 4. Authorizations and Exemptions Table of Contents 4.0-1 4.1 Release for Unrestricted Use 4.1-1 4.2 Authorization for Use of Materials at Offsite 4.2-1 Locations 4.3 Record Storage 4.3-1 4.4 Exemptions from the Requirements of 10CFR70.24 4.4-1 4.5 Nonradioactive Industrial Waste 4. 5-1 4.6 Possession of Licensed Material at Reactor 4.6-1 Sites 4.7 Leak Testing Scaled Plutonium Sources 4.7-1 4.8 Disposal of Aqueous Products 4.8-1 l Docket No. 70-1151 Initial Submittal Date: 4/30/83 Page No. 11 License No. StN-1107 Revision Submittal Date: 4/24/87 Revision No. 10

SCHEDULE OF REVISIONS Page Revision Page Revision Number Number Number Number 1.9-49 2 1.9-80 2 1.9-50 2 1.9-81 2 1.9-51 2 1.9-82 2 1.9-52 2 1.9-83 2 1.9-53 2 1.9-84 2 1.9-54 2 1.9-85 2 1.9-55 2 1.9-86 2 1.9-56 2 1.9-87 2 1.9-57 2 1.9-88 2 1.9-58 2 1.9-89 2 1.9-59 2 1.9-90 2 1.9-60 2 1.9-91 2 1.9-61 2 1.9-92 2 1.9-62 2 1.9-93 2 1.9-63 2 1.9-94 2 1.9-64 2 1.9-95 2 l.9-65 2 1.9-96 2 1.9-66 2 1.9-97 2 1.9-67 2 1.9-98 2 1.9-(9 2 1.9-99 2 1.9-69 2 1.9-100 2 1.9-70 2 1.9-1C1 2 1.9-71 2 1.9-102 2 1.9-72 2 1.10-1 2 1.9-73 2 1.10-2 2 1.9-74 2 1.10-3 2 l.9-75 2 1.11-1 10 1.9-76 2 1.11-2 10 1.9-77 2 1.11-3 10 1.9-78 2 1.11-4 10 1.9-79 2 1.11-5 10 Docket No. 70-1151 Initial Submittal Date: 4/30/83 Page No, v f License No. SFN-1107 Revision Submittal Date: 4/24/87 Revision No. 10

SCHEDULE 0F -REVISIONS Page Revision Page ' Revision Nunber Number Number Number 2.0-1 4 2.3-11 '2 2.1-1 2 2.3-12 2 2.1-2 2 2.3-13 2 2.1-3 2 2.3-14 2 2.1-4 2 2.3-15 3 2.1-5 2 2.3-16 deleted 2.2-1 2 2.3-17 2 2.2-2 2 2.3-18 2 2.2-3 6 2.3-19 2 2.2-4 3 2.3-20 2 2.2-5 3 2.4-1 3 2.2-6 2 2.5-1 7 2.2-7 2 2.6 5 2.2-8 2 2.7-1 3 2.2-9 2 2.7-2 3 2.2-10 2 2.7-3 2 2.2-11 2 2.7-4 2 2.2-12 4 2.7-5 2 2.2-13 4 2.7-6 2 l 2.3-1 4 3.0-1 2 2.3-2 4 3.1-1 2 2.3-3 3 3.1 2 2.3-4 2 3.1-3 2 2.3-5 2 3.1-4 2 2.3-6 2 3.1-5 2 2.3-7 2 3.1-6 2

2. 3-8 2

3.1-7 2 2.3-9 2 3.1-8 2 2.3-10 3 3.1-9 2 Docket No. 70-1151 Initial Submittal Date: 4/30/83 Page No. v1 License No. SNM-1107 Revision Submittal Date: 4/24/87 Revision No. 10

.~ _. SCHEDULE OF REVISIONS -Page-Revision' Page Revision! Nunber ' Number Number Number 3.1-10 2 3.1-11 2 3.1-12 3 3.2-1 2 3.2-2 2 2 3.2-3 2 3.2-4 3 3.2-5 2 3.2-6 2 3.2-7 2 3.2-8 3 1 3.2-9 3 i 3.2-10 2 3.2-11 3 3.2-12 3 3.3-1 2 1 3.3-2 2 4 3.3-3 2 4.0-1 3 4.1-1 3 4.2-1 3 4.3-1 2 4.4-1 3

4. 5-1 10 4.6-1 2

4. 6-2 2

4.7-1 8 4.8-1 9 Docket No. 70-1151 Initial Submittal Date:' 4/30/83' Page No.= vil-License No. SNM-1107 Revision Submittal Date: '4/24/87 Revision No. 10 1 -.. ~ .----.-n

A REVISION RECORD Revision Date of Number Revision Pages Revised Revision Reason 1 12/12/83 1.9-52 through 1.9-76 Respond to NRC questions inclusive, 2.2-10, in NRC letter dated 2.~2-11, 2.3-1, 4. 8-1 April 6, 1983. 2 3/26/84 All pages resubmitted Respond to NRC questions as Revision 2. See in NRC letters dated submittal letter January 26, 1984 and attachment for de-February 23, 1984. scription of changes. 3 1/4/85 All pages resubmitted Respond to NRC as Revision 3. See re-questions. vision lines in right hand columns of each page for specific changes. 4 6/25/85 New pages 2.2-12 and Licensing of uranyl 2.2-13 and revised nitrate storage tanks. pages 2.3.1 and 2.3.2. See revision lines in right hand columns of each page for specific changes. 5 5/14/86 2.5-1 and 2.6-1. Update of Emergency Plan Section. 6 8/8/86 2.2-3 Added criteria for positive pressure gloveboxes. 7 9/22/86 2.5-1 Update of Emergency Plan Section. 8 4/1/87 4.7-1 Added criteria for leak testing sealed plutonium sources. 9 4/20/87 4.8-1 Added criteria for disposal of aqueous products. 10 4/24/87 New pages 1.11-1 through Added criteria for 1.11-5 and revised page disposal of dewatered 4.5-1. calcium fluoride. Docket No. 70-1151 Initial Submittal Date: 4/30/83 Page No. vili License No. SNM-1107 Revision Submittal Date: 4/24/87 Revision No. 10

p. ;

maan.s.e - a.~ n, ~ u-aa.n ~ -a.p-- .an..~w .~.,w- ~.n. .--.,-e p- "N 4_ 3 n~ x 6 ~ w, A.. 9 ,4 w a 6 i k 9 1, g aW 1 4. 9 4 r t i t '.' h y Y.: 1 t -y 4 e d D A .k e 4-j' ? t P - t A l'. g y i t .j .i.r 't. 1 f I . ' ' / n y a-a we S h 's [4 i 1 5 1' l' i b 4 +;. 6 'h e 4'e i t i i I. E -i a v e I.. a a f i. l f i 0 l t k I r I s. ts (T j i \\ a l' $ s l:} ~ .a j .) s s ., = 5 y s p 4 A 5 T-*-wSte' e - w"W4N? MAN w 9-wst'$.w We'-gGN' Tem Dr MT9'K2 T*'D @M 9'kk> art * @ *'FT Me-k'bWv'"Ayp-+ w~wr r~qag*,h e.e mN W w m - m -- t 2-1p r, Mg-t-*-que e ges er-get-r,*yw e ttp qrse.r-w'h ts gr -eid*PWh Wag'*e 14 Pgwpew$wf 1

u Page . l'. 9 Processing Operations-1.9-1 1.9.1 Conversion-Operations l.9-2 i- - 1.9.2' Fabrication Operations 1.9-25 ~ 1. 9.3 Laboratory. Analysis 1.9-36 1.9.4 Scrap Recovery Operations 1.9-36 - 1.9.5 Moderation-Controlled Processing.0perations 1.9-42 ' l.9.6' ' Incinerator 1.9-81

1.9-88 1;9.7'

- Chemical Manufacturing Development Laboratory '1.9.8. Purification of Contaminated Scrap through Solvent- 'l.9-91 Extraction 1.9.9 ~ lF6 Cylinder Washing and Recertification l'.9-101' 1.9.10 Improved Fuel' Processing. Operations 1.9-102 -1.10 Off-site Release Evaluations 1.10 1.11 Nonradioactive Industrial Products 1.11 -1 f .i t j i Docket No. 70-1151 Initial Submittal Date: 4/30/83 Page No. -1.0-2 I-License No. SfN-1107 Revision Submittal Date: 4/24/87 Revision No. 10 4 e, e -ae-. ~ ---m-- ,n, v. ,. - en, w e. -, e+ .-g+p,, p.. --y+, t wd,, rw --ow-g 14~ <, -ww'=wy---* t v 4 *- ~ s'-*m-wv=r

~ 1 I l DCGHET NO. ~//(( CONTRDL llo,

hhff, DATE OF DCC.

_ N__ f I a m w:m. 39 7 _j/_ FCUF PDR FCAF LPDR. l & E REF. _/ S 7ECUNDS _ FCTC. _ 0;fa _ DATE. b _ [ 1;;;r,2,t, - _ - - - -}}