ML19309C900

Forwards Response to 800124 Comments & Questions Raised by NRC on License Amend Application 18

ML19309C900
Person / Time
Site:	Framatome ANP Richland
Issue date:	02/21/1980
From:	Estey H SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:	Crow W NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
Shared Package
ML19309C901	List: ML19309C900 ML19309C912
References
15618, NUDOCS 8004090370
Download: ML19309C900 (11)

Text

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.D a a Eh NUCLEAR COMPANY,Inc.

2101 Horn Rapids Road P. O. Box 130. Richland, Washington 993S2 Phone:(509)375 8100 Telox: 15 2878 February 21, 1980 lir. W. T. Crow, Section Leader Uranium Process Licensing Branch Division of Fuel Cycle & fiaterial Safety U.S. Nuclear Regulatory Commission Washington, D. C. 20555 License No. SNM-1227 Docket No. 70-1257

SUBJECT:

NRC Comments & Questions on License Amendment Application No.18

Reference:

Letter, Robert L. Stevenson (NRC) to H. Paul Estey (ENC),

dated January 24, 1980 Exxon Nuclear Company, Inc. hereby submits its response to the commcats and questions raised by your staff on the subject license amendmer,t application, as transmitted by your letter of January 24, 1980.

Respective pages of the License Amendment Application have been appro-priately amended in accordance with this response, and seven copies of the amended pages are enclosed. Also, one copy has been sent to Region V IE.

The License Renewal Application has also been appropriately amended to incorporate the commitments reflected in the attached response.

Sincerely,

-2 o

.f H

aul E ey,11anager s

Licensing and Compliance Operating Facilities Enclosures cc:

Mr. W. J. Cooley (USNRC Region V IE) l

'B oon n o37o AN AFFILIATE OF EXXON CORPORATION

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, o Response To Comments and Questions on License Amendment Application No. 18 (Application Dated November 19,1979)

Exxon Nuclear Company, Inc.; License No. SNM-1227; Docket No. 70-1257 1.

Comment Page I-5.33, para. 5.2.2.4.2 - it sitould be confirmed Biat (a), (b) and (c) are :tddLtive requiremen,ts.

Para. (c) sitould add Biat, in tlie absence of,a fLteptoof barriet, special con,ttots teilt be used to pteven,t fitts and.to conttal ase of moderators in fitefigItting in sucIt process areas. Furtlichmore, (c) appears to alloto ase of concenttalion conttot bcsed solelij on administta,tive procedures.

It sltoald be made etcar Bia,t sucIt applications taill be limited to situations teltcre Bie na.ture of Bte ptocess and operations make violatio, of tite concenttation Li.mLt unlikelij even after fa,iture of any s. ingle conttol.

Response

It is agreed that Sections 5.2.2.4.2(a), (b) and (c) constitute additive requirements, and Section ~ '.2.4.2 of the Application has bec.; mcdified to show this.

Section 5.2.2.4.2.(c) of the Application has been expanded to include the statement, "In the absence of a fireproof barrier, special control shall be used to prevent fires Cid to control the use of moderators in firefighting in such process areas".

Section 5.2.2.4 of.the Application has been expanded to include the statement, "... where the nature of the process and operations made violation of the concentration limit unlikely even after failure of any single contol."

Sections 5.2.2.4, 5.2.2.4.1 and 5.2.2.4.2 (as modified above) of this Amendment Application have been incorporated into the License

g i

Renewal Application as Sections 3.2.2.4, 3.2.2.4.1 and 3.2.2.4.2, l

respectively.

2.

Question Page 11 - 2.16b - What is die justifica< tion fo.t not employing Ble usual tempetatute conttols on ute ion exchange column feed stteams and Bic coltunn itsetf t f

Response _

I The normal operating temperature range for the conversion process waste stream is between ambient temperature and 100 F.

The ion exchange columns and a significant amount of the associated piping l

is plastic (PVC, polypropylene and fiberglass) which is used for corrosion resistance where operating temperatures do not exceed 140 F.

The fiberglass ion exchange columns would fail at around 220 F.

All of the above temperature points are less than the 275*F E

(135*C) at which accelerated nitration of the resin could pres-i surize the column.

In addition to the above process limits on the

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temperature, the ion exchange column effluent temperature will be monitored.

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3.

Comment Page 11 - 2.26 - Since, as stated on page 2 of die stansnLttal l

tettet, it is destted Blat Bsc dty end of Btc kUR be licensed a.t this time, addi.tional infotmettion should be ptovided identifging ute basis for nacica1 etiticality safety in Bia.t part of die kUR.

Response

P The Waste Uranium Recovery Facility dry end will receive 55 gallon barrels of uranium-contaminated waste and used HEPA filters which will be assayed for uranium content prior to processing.

If these items contain greater than 150 gU/ item, they will be moved to i

the second floor of the facility for separate processing.

The l

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barrels will be dumped and the contents will be sorted, shredded, and separated into low and high uranium content waste.

HEPA filters will be shredded and the material separated into low and high uranium content waste. The high uranium content material will

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be repacked and stored for subsequent leaching.

The low uranium-contaminated waste will be repacked for burial.

Criticality safety for the dry end of the Waste Uranium Recovery Facility will be assured on a safe batch control basis.

The uranium content of the barrels and filters being dumped will be recorded, and when the accumulated uranium equals one safe batch the equipment will be shut down and cleaned out.

The equipment included in the batch limit area will include the barrel dumper, shredder, separator and the barrel filling station. All of the equipment included in the safe batch control area is designed for -

convenient inspection and cleanout. Any storage areas associated with the dry end of the Waste Uranium Recovery Facility will also be limited to one safe batch.

The use of safe batch control as a method of assuring criticality safety is used in several locations throughout the fuel manufac-turing building and is demonstrated in Part II of Document No. JN-2.

4.

Comment Pagc 11 - 2.27, second and third Lines - Ptcase shtte specificalbj e

witat codes will be met bt).the building design.

Response

A copy of a letter from the building designer, which specifies the codes and standards followed in the design, is attached. A copy of this letter has also been included in Attachment A, Section 7, of the License Renewal Application.

_a_

ARCHITECTS ENGINEERS CONSTRUCTORS FACILITIES SYSTEMS ENGINEERING CORPORATION 23C5 UTAH AVENUE EL SEGUNDO. CAllFORNIA 90245 (2131 776-7300 9170/ TBS 723 THE PARrWAf, P O. 447. RICHLAND, WASHINGTON 99352 (509) 943-6771 22 January 1980 NECen f y,-

Exxon Nuclear Company 2955 George Washington Way g8,M.8f.jpOU 7

Richland, Wa 99352 Attn:

Jack Fastabend "8 ECD Gentlemen:

The Waste Uranium Recovery Building, to be constructed at Exxon Nuclear Company's Horn Rapids Site, will be used to store and handle special nuclear materials.

This building was designed to the following codes and standards:

Washington State Highway Standard Specification for e

Road and Bridge Construction - Roadway & Parking e

U.B.C.

Uniform Building Code, 1979 Edition e

National Electrical Code ANSI-C1, 1978 Edition

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e Uniform Fire Code, 1976 Edition Uniform Plumbing Code, 1976 Edition e

e Uniform Mechanical Code, 1976 Edition e

ASHRAE Standards, 1979 Edition City of Richland Ordinance Numbered:

3777 (Adopt. Building Code) e 3877 (Adopt. Plumbing Code) 3977 (Adopt. Mechanical Code)

Very truly yours, I

~

Thomas B.

Swearingen, Resident Manager Facilities Systems Engineering l

Corporation TBS :sj f I

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Comment Page 11 - 4.76 - There appear to be exponotts omLtted in some of the descriptions of manimum resin loading (e.g., 3rd line up on l

pag e 4. 76 ).

Picase review this page and 11-4.77 and corteet if nece6aary.

i

Response

l paragraph 4.6.T3.2(a) and paragraph 4.6.13.3(a) of the application have been corrected with the proper exponents.

In addition, the i

minimum critical concentration for a twenty inch diameter cylinder 3

is stated both in kgs U/ft. of column length and kgs U/ft. This data is from ARH-600, Table III.B.7-6.

6.

Comment Page 11 - 4.77 - Picase relate Die maximum resin loading va. lues quoted by.the manufactwtch to.the values 9iven based on Bie Exxon experiments, and show Biat ble latter ate cons.iStent teint Bic i

manafaeturcr's values.

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Response

The maximum resin loading values based on Exxon experiments are lower than the resin loading values quoted by the manufacturer because the manufacturer loading factors represent total ion exchange or equilibrium capacity where as the Exxon numbers repre-sent a breakthrough capacity.

Since the chelating resins were developed specifically for copper, nickel, zinc and iron recovery, and the uranyl ions are competing with metal cations for resin sites, it would be difficult to reach the manufacturers quoted i

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-9.,-.w-v r--

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resin loading values. The total exchange capacity for the resins f

quoted by the manufacturer is equivalent to N 45 gU/1, or 1.27 3

kgU/ft, as shown below.

At Wt (I)

I (1) grams / liter = equivalent / liter x Vaience i

(2) Q vol. = (1-8) p Q weight (2)

N Where Q vol = volume capacity of packed bed in equiv./ liter; Q weight = theoretical weight capacity in meg /g(dry); s = fractiorici void volume; w'= water content and; p = density 'Sp.G.) of swollen Resin.

Equation (1) yields a larger exchange capacity as follows:

IRC-718 Chelating Resin Total chelating / cation exchange capacity:

5.3 meg /g (dry)

Moisture Content:

68 percent (nominal)

Shipping Weight (sodium form): 42 lbs/ft3 = 0.67 kg/t l

Resin dry weight basis: 0.67 kg/t x 0.32 = 214 g/t 5.3 meg /g x 214 g/t = 1134.2 meg /t or 1.134 eg/t 1.134 eg/t x 2j8 = 45 g/t Paragraph 4.6.13.3(a) of the Application has been rewritten using the manufacturer's maximum resin loading values (instead of the l

values based on experiments), since these can be calculated.

The use of the manufacturers maximum resin loading values instead of i

l the experimental values has no significant effect on the basic assumptions for the analysis.

(1) Water Supply and Treatment, National Lime Association (2)

Ion Exchange, McGraw-Hill, Helfferich, 1962.

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7.

Comments S Question Page 11 - 4.18, pata. 4.6. I3.3(b) - Tlte calculation fo.1 nomtt c

conditions of etution is based on dissotation of ammonium diatanate l

(ADU). Witat is Ble expetimental basis for concluding Btat Bnc etution reaction is alauys concetty reytesented by Bic IWO -AOU 3

reaction?

It is necessang for a valid safety conetusion to be able i

to confirm tlutt Ble etution opetation cannot resul,t in a deep plug of sotation above tite minimum esitical concentnation.in Ble bed.

Response

i Since the IX feed is alkaline (pH>9.0) from the upstream precipitaion l

j of ADU, it is assumed that the form of uranium loading on the resin is ADU of extremely small particle size that has not been removed by l

the upstream centrifuges and filters. Since the exact resin loading mechanism and form of uranium loading is unknown, it was decided to i

experimentally determine the maximum concentrations of uranium in 2N nitric acid, by dissolving excess amounts of ADU, U0 and U 0 2

38 to saturate the 2N nitric acid.

These uranium concentrations are compared to the concentrations derived theoretically, from the following generally accepted equlibrium dissolution reactions.

t 4

U02 7 + 6NH03 + 2002 (NO )2 + 2NH N03 + 3H 0 (Nil)2 l

ADU:

3 4

2 4

I.

2N nitric yields 0.67 Mole of UNH or 159 gm U/t 2 (NO )2 + 2N02 + 2H 0 U0 :

UO2 + 4HNO3 > U0 2

3 2

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I 2N nitric yields 0.5 Mole of UNH or 119 gm U/t

-* 3002 (NO )2 + 2N0 + 4H 0 3U02 + 8HNO3 3

2

'2N nitric yields 0.75 mole of UNH or 178.5 gm U/t

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8-2(NO)2+2N02 + 4N 0 U03 8 + 8 HNO3 + 300 U0:

3 2

3g l

I 2N nitric yields 0.75 mole of UNH l

1 2 (NO )2 + 2NO + 10 H 0 30 03 g + 20 HNO3 + 900 3

2 2N nitric yields 0.9 mole of UNH or 214 gm U/t I

As can be seen, the latter equation gives the highest uranium l

concentration..This compares with experimental uranium saturated 2N nitric acid concentrations of 97 gm U/t for ADU, 53 gm U/t for i

U0, and 124.6 gm U/t for U 0.

It should be noted that in 2N 2

38 nitric acid. ADU dissolves quickly and completely. U 0 slower but 33 complete, and U02 probably does not reach saturation in 2N nitric.

i It is very doubtful that U 0 w uld load on the column.

The only 38 source of U 0 w uld come from a scrap uranium UNH stream.

UNH is 38

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prefiltered, precipitated, centrifuged (3 centrifuges in series) and filtered again prior to entry to the IX columns.

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Page II-4. 78, ptaa. (c)

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Question (a)

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Wita.t is tite ba. sis for asswning a ::cro resin loading (dien the l

in.tchstices teche asswned flited toltlt AQU for tite calculittion i

reported Incret i

Response-Ion exchange is typically a surface phenomenon where the resin contains bound groups carrying an ionic charge in conjunction with i

free ions of an opposite charge that can be displaced or " exchanged."

-,,n-

,y e

.-,---,,-,,.--,,r

-w

-,,,e

-w, nr.,-

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,w, e

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r The total surface area is comprised of the external surface of the resin particles plus the areal surface of macropores, therefore, l

resin loading and bed packing occur in the same space.

i Question (b) l The values quoted for normal resin unloading appear to be small compared to the uncertainty in, and possible. impact of, resin bed void volume. Are there no measurements of resin bed void volume l

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that would be dircetty applicable to the sLtua. tion analy:cd?

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Response

1 Three alternate methods are available for determining the resin bed i

void volume:

s

1) Manufacturers report void volumes of 30-40 percent;

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2) Manufacturers give a bulk density and particle density with the j

relationship, ob = pp(1-c) where ob = bulk density, pp = particle i

density and c = void fraction;

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For the resins to be utilized, the bulk density is 42 lbs/ft3 (0.67 kg/t.), and the particle density (or Sp. G.) is approximately 1.09, representing a 38 percent void volume.

3) Volumetric fleasurement Laboratory measurements yield an average of 39.3 percent void I

volume. There appears to be good correlation for a void volume of 40 percent. The situation analyzed assumes a 50 percent void volume.

Ques ti on le)

Why tus it assumed that AQU would not f Lll the unpacked scetion?

s t

v

-- =

,--,y,

--,.---.-r-,,.-,.v-_,

-ew.,-e

-,y

- r

.-c-

Response

If a surge of ADU in a concentration considerably greater than the 300 ppm U limit were to enter the column the ADU would not accumu-late on top of the resin bed, but would pack into the void spaces, as analyzed, causing the column to blind.

Differential pressure measurecents and controls preclude packing of the bed to a point where the upper section of the column would accumulate ADU.

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