ML18081A224

Part B of CSF-1, Application (Selected Parts as Related to the Radiation Safety Program - Part 3)

ML18081A224
Person / Time
Site:	West Valley Demonstration Project
Issue date:	10/12/1962
From:	Nuclear Fuel Services
To:	US Atomic Energy Commission (AEC)
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ML18081A225	List: ML18081A116 ML18081A222 ML18081A223 ML18081A224
References
Download: ML18081A224 (130)

Text

AUG g 19~

Before the Washington', o. C.

(@ W::LEAR FUEL SERVICES, IOC.

_For Licens~s for a Spent Fuel Processing Plant

---.. -- - -- ~

Cnder Sections 53, 63, 81, 104 (b), and 185 of the

\\

Atomic Energy Act

(

~

AEC Docket No. 50-201

• 1

..:.J

• I -

r..;

@.su~is~ion ~o__: _~:_:~ _!inal Sa_fety Analysis Rep~~!

• Revision of Sect A2Pfl0V£D F.., "'"'"'J VII.- Protec ~oh "'of-'thJ1$'uiiif-E

\\

0 Revision 2, August 20, 1964

VII PROI'Ecria. CF THE PUBLIC Swmnary 7.1. The plant and process which have been described in detail in preceding sections are designed to operate so that, under all normal operating procedures, any discharge of radioactivity to the' environment will be well within the limits set forth in 10 CFR Part 20.

7.2 Radioactivity can be lost from the process complex at the fo!lowing points,

1.

Stack

2.

Waste storage tanks

3.

Storage lagoon *

4.

Burial ground

~.

Egress of personnel and n,aterial

6.

Product shipment In subsequent paragraph~, each' of the above possibilities as analyzed to show that *the *statement of Paragraph 7.1 is

,,aucr;-.:.--Soaie of the detailed calculations are shown in

.Appendicies as noted

• 7.*3 Further, this plant and its site are shown to be so designed and located that, in the unlikely event of the

- moat sarious accident which could possibly be deemed credible, there will be no discharge to the environment which results in levels of exposure in excess of those set forth in Sections 100.ll(a)(l), (2) and (3) of 10 CFR Part 100; and further that steps can be taken to assure. that, even in the event.

of such an accident, the disch3rges to surface waterways at the site boundary canoe kept within the limits specified in 10 CFR Part 20 th!ough the use of reasonable correction measures after the accident or release has occurred.

7.4 The following abnormal events have been postulateds

1.

The complete rupture of a waste tank releasing 600,000 gallons of high-level waste.

2.

A criticality incident anywhere in the plant involving a total of 1019 fissions in a single buiat or a multiple continuing event totalling 1~ fissions.

Revision 1, Aug. 20, 1964

3

• A criticality incident in the fuel storage pool which sets up a 10-mwt boiiing water reactor which operateJ[Sfor as long as 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> before it can be. shut down.

4.

A chemical explosion in the plant which is assumed to rupture a vessel containing a full day's charge of the maximum fission product content possible.

~.

The complete failure of the iodin~ removal

.equipment so that for a period of up to one day the complete charge of iodine is lost to the atmosphere.

The rationale for the selection of these events has been to select for the pla~t, the stack, and the tank farm1events which represent the upper limit of catastrophe which could occur in each of these areas, even though we believe that the likelihood of occurrence is very small. In subsequent paragraphs, each of the above possibilities is analyzed.

7.5 Throughout this section, a number of. assumptions recur. Values for such recurring assumptions are r~llected in Table 7.5. Assumptions specifically related to a part-icular calculation are included in the calculation

• Normal Operations*

Stack 7.6 As explained in Paragraphs 6.3 through 6.21, the ventilation systems are designed to assure that, under normal operating conditions, flow of air is always from areas of least contamination into those of higher contamination.

There are separate systems for vessels, dissolvers, and the cells themselves. These join together and are filtered before discharge through a 65-meter ~tack. The total volume of air discharged is 32,000 cfm.

Iodine removal facilities are d3igned to collect 99.5% of the incident iodine. It is assumed that all of the noble gases in a daily charge escape during the course of the day. Under normal operating conditions, the amount of solid fission products taken into the gas stream is assumed to be low enough that the filtering of this stream will reduce them to the point where they are negligible in c~mparison to the gaseous activity. Calculationa are based on an average fuel which we may expect to process in this plant represented by the following parameters&

Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964

1.

2.

• Table 7.~

Assumptions Used in Calculations Sections VII & VIII The dispar,ion parameters used are those given in Table 2.14 and in "Nuclear Safety", Volume 2, No. 4, June 1961 *. Figures V-1 and V-2 provide horizontal and vertical dispersion coefficients respectively for distances up to 1<>5 meters and for meteorological conditions ranging from "extremely unstable" to "moderately stable. In all calculations performed in this section, "slightly unstable" coefficients have been assumed to represent average conditions and "moderately stable" coefficients have been assumed ~o represent inversion conditions.

~*,~

~ ~

Wind velocities of 1 meter/second for inversion conditions and 4 meters/second for average conditions have been used.

The foll~ing wind distribution data has been used, Wind D.istribution ( t...

<; i'~"'.)

(Per Cent Per Octant)

Wind Direction Sumner Winter Average N

8%

NE 4

2 E

5 2

SE 17 9

s 23 21 gf 13 25 w

9 12 NW

• 20 21 Fuel is cooled 150 days before processing.

3.

High-level waste is stor ed at 410 g~lloras per ton which is equivalent to1 132 c/gal 166 c/gal

~7 c/gal at the time of storage.

Sr-90 Cs-137 Ru-106 Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964

. I

Ta~l* 7.~ (Cont'd) *

4.. The rate of travel in the surficial till is 1.0 foot/day. The rate of travel in the silty till 1* ~ x 10-~ foot/day.

~.

~

of Sr-90 is associated with sludge in the tank.

6.

99.CJ' of Sr-90 is adsorbed on soil on passage through it.

7.

~

8.

9.

10.

99.9CJ' of Cs-137 is adsorbed on passage through the 700 feet of soil.

No Ru-106 is adsorbed at all.

Tritium is assumed to go 25% to stack, 10% to wa&te tanks, 65% to s*am.

For long-lived isotop~s the fission products are

• taken as 7'11, from u235 - 30% from Pu239.

For short-lived isotop~s they are.taken as 60% from Pu239 -

40% from u235

• Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964

Burnup Specific Power Irradiation Time Load Factor Cooling Time 20,000_ D!'f_d/ton

. 32 IJ'llljt:_on ',

. 2 years.

85 per. cent 150 days Using these parameters the input activity to the plant was

/

calculated. The gas1ous activity* input is1

~/~

Kr-85 I-129 6.3 x 1o3 curies,,J,

~)

1.. 131 Xe-13lm Xe-133 Tritium

. 0.022 curie/......-*t°'* ~ ;,- ~

. 1.8 curies 1.0 curie)

_.,, "" = *.)

3.8 x 10-curie

,,),* \\. r :.,,.,.

50 curies

• Under the conditions stated above, the total daily discharge from the stack using the average activity level fuel contemplated will bes 6.3 x-103 curies 1.1 x 10-4 curie 9.0 x 10-3 curie 1.0 curi!3 Kr-85 I-129 I-131 Xe-13lm Xe-133 *.

Tritium 3.8 x 10 curie 50 curies

-- -~

.,,; t. I *

• 7.7 The concentrations of each of these isotopes at various distances and under various meteorological conditions are ~alculated f~om the following formulae1 for short-term calculations1 X*

Q exp 2a-2 z

(7.7a) for Long-period average concentrations rt..0-0lf Q a-

• U X

z 21T exp 8

2o-

~

z (7.7b)

• At 150 days cooling these are the only signiricant gaseous isotopes.

Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964

//

I I

I

'i

• j

\\

~

~*.

• l 1

Where X

• concentration ln curles/m3 (,c/cc}

Q

• emission rate ln curies/second*

°y* cr'z

• dispersion coefficients. in meters b

• stack height in ~eters u * *wind velocity in meters/second x

• distance downwind in meters f

• wind frequency *in per cent/octant The calculation has been carried out for both inversion and average conditions aver the range 1500 to 51,000 meters (see Appendix 7.7). The results of these calculations are presented in Table 7.7. The maximum concentrations are given for both the average and inversion conditions. For average conditions the maximum concentration occurs at the site boundary; under inversion conditions the maximum concentration ~ccurs over the range of about 4000 to 10,000 mei ers downwind from the stack. It can be seen that all of the concentrations are well within the MPC values with the exception of the Kr-85 concentration under inversion conditions. The inversion concentrations given are centerline concentrations and include no wind diverstty factor; they are not expected to persist for more than a few hours at a time. The yearly average concentration, which is permitted under 10 CFR Part 20, will not be significantly increased by these occurrances *

. 7.8 Although 10 CFR Part 20 contains no provision for limits on the.deposition of radioiodine on pasturage, the plant is designed to release iodine at concentrations lower than the MPC for concentration in air in order to protect those areas surrounding the plant site which are used for dairying. Using the long-period average concentration and a deposition velocity of _0.01 meter per second, ~he deposition rate has been calculated (Appendix 7,8). Since yearly average contentrations are used, it is reasonable to assume that the equilibrium. conditions are reac~edJ i.e. the rate of depos-ition equals the rate of decay. The south, southwest and northwest octants have the highest yearly average wind frequencies, ranging from about 19 to 22 per cent. Therefore, a wind frequency of 25 per cent per octant has been used in these calculations. It was found after the Windscale incident that a grazing area contamination level of l µc per square meter resulted in about 0.1 µc/liter of milk*.

Using this relationship the resultant activity levels in milk ha~e been calculated. The milk activity levels are shown in Table 7.8.

7.9 The Federal Radiation Council has established a Radioactivity Intake Guide for Iodine-131 of 100 µµc per day, b~sed on the uptake by children as the most sensitvie segment of the population. As can be seen from Table 7.8, the con-sumption of about five liters of milk per day from dariy cattle grazing inrnediately adjacent to the site boundary would be required to equal the level of intake as established by

• TI0-8206, Page 56 Revision l, Oct, 29, 1962 Revision 2, Aug. 20, 1964

Isotoeesd Kr-85 I-129 I-131 Xe-13lm Xe-133 Tritium Table 7.7 Maximum Concentration of Gaseous Isotopes Under Inversion and Average Meteorological Conditions x, ~/cc CUrieslSecond Inversiona Avera2e6 MPCC JICLCC 7.3 X 10*2 7.3 X 10*7 1.6 x 10-8 3 X 10*7

-9 1.3 X 10 1.3 X 1o*l4 2.8 X 10*16 6 X l0-11 LO x 10-7 1.0 X 10-12 2.2 X 10-14 3 X lO*lO 1.15. X 10-5 1.15 X 1o*lO 2.5 X 10*12 4 X 10-7 4.4 X 10-8 4.4 X 10*13 9.7 X 10*l5 3 X 10-7 5.8 X 10-4 5.8 X l0-9 1.3 X lO*lO 2 X 10-7 a

Maximum concentration occurs at about 6000 meters from the stack1 concentration within about 10% of the maximum occur from about 4000 to 10,000 meters from the stack.

b Maximum concentrations occur at the site boundary (1500 meters).

c Table II, Appendix B, 10 CFR Part 20.

d At 150 days cooling, these are the only significant gaseous isotopes.

e Based on 1 triton produced per 104 fissions (reported as 1 in 1 to 4 x 104) with 25% lost up the stack, 65% lost in liquid w.-.ste effluent, 10% to storage tanks.

Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964

Table 7.8

• Iodine Deposition and Milk Concentration8

• Ground Concentration Milk Concentration Distance in Meters euc/m2 we/liter 1500 200 22 2000 150 15

,ooo 31 3.1 10000 8.9 0.89 20000 2.6

. 0.26 a

See Table 7.5 -for assumptions and Appendix 7.8 for detailed calculations.

Revision 1, Oct. 29, 1962 Reyision 2, Aug. 20, 1964

this Guide. This rate lf consumption is higher than any that can be expected, probably by a factor of at least four.

In addition, no credit is taken for dilution (during processing) by milk containing lesser (or no) amounts of radioiodine or the fact that cattle are pastured in Western New York State only about half the year.

Waste Storage Tanks 7.10 The design of ~le waste storage tanks has been discussed in detail in Paragraphs 5.50 through 5.56 and in Submission 1 date July 1, 1963. These t~nks are built in a "cup-and-saucer" design. Operating procedures call for monitoring of the annular space between the tank and its saucer and of the water introduced under the tanks.

If there is significant leakage from the tank into the saucer, the entire tank contents Nill be transferred into a spare tank kept for that purpose. Thus, under normal operating conditions there will be no loss of activity from these tanks. c<<~,.-t J.r..,.\\c,.~,lt. o,\\;;;-,1 Storage Lagoon 7.11 The very low-level wastes from this process--

overheads from acid fractionation, solvent wastes, and miscellaneous wastes--can be put through the gene.ral purpose evaporator and the overheads from this can be put through ion exchange ~olurnns if necessary. It is expected that the normal activity content of the overheads from the general purpost!

evaporator -will contain about 10-6 µc/cc.of activity. This can be further reduced by a factor of 30 by the use of simple, non-regenerated cationic ion exchange r~sulting in a concent-ration of 3 x 10-8 µc/cc.

The expected volume of these wa*stes is 40,000 gal/day. The average available flow in

• Buttermilk Creek is 41 cfs which is equal to 2. 7 x!_07 ga.1/day.

Thus, the available on-site dilution factor is 6.8 x 102.

In Cattaraugus Creek an additional dilution factor of about 8.5 is available. The concentration in Cattaraugus Creek would be expected to be about 10-lO µc/cc.

Furthermore, the residual activity in this stream will be largely Ru-106 and I-131 with some ~-Nb-95. The MPC's for these isotopes are 1 x 10*5~ 2 x 10*, 6 x 10-5, and 1 x 10*4 µc/cc rather than 1 x 10-1 for unknown activities when radium is absent.

Therefore, the available factor of safety is about 103 with-out any analyses of the effluent and about 104 if we choose to carry out specific fission product analyses on this effluent stream. This stream will also carry about 130 curies per day of tritium since there is no known way to process it to remove the tritium. The concentration of tritium on-site in

-Buttermilk Creek will average 1.3 x 10-3 µc/cc.

The on-site MPC is 10-l µc/cc.

In Cattaraugus Creek the tritium concentration is expected to average 1.5 x 10-4 µc/ccJ the MPC here is 3 x 10-3 µc/cc.

• Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964

I 0

7.12 This low.level.stream can be discharged directly to Buttermilk Creek and the level of activity at the site boundary will remain well within the MPC levels of 10 CFR Part 20. In addition there will be a series of lagoons available for use as an emergency holdup area. Their use will permit time for the decay of shorter-lived isotopes and will allow the adsorption on the soil of some of the longe~lived isotopes. The low level waste streams from the plant ~ischarge into the interceptor, a concrete pit of 50,000 gallons capacity,.. ich is designed for batching of wastes. A valved interceptor drain lice will permit collecOion of one days output from the plant which will then be sampled for gross alpha, beta, ganma and tritium. The pH of the sample w,ill be checked and the interceptor contents neutralized if necessary ~o pH to 6 to a. A line is available for pumping the interceptor contents back to the plant for further processing. Normally, after sampling, the inter-ceptor drain valve will be opened and the contents allowed to drain by gravity to the first holding pond, a 300,000-gallon settling basin vlith a hlgh level overflow to the second pond *. The second and third ponds each have capacity of about 2.3 million gallons. Between the second and third ponds will be a high level overflow and a valved drain line about 18 inches above the bottom of the pond. A valved drain line from the third pond will discharge to the creek. The capacity of the ponds above the overflows will allow complete holdup of 100 days output from the plant.

7.13 In view of the factors of safety available, no hazard will be presented by the routine handling of this aspect of the operation.

Burial Ground 7.14 Two types of wastes will be buried in the ground in conjunction with the operation of this plant. One is low-

• level solid trash of all sorts coming either from the plant operation itself or shipped in for burial from off -site users of radioactivity. The other is high-levP.l solid trash in the form of leached hulls or equipment discarded from the plant. Activity associated with the former type is considered to be "available" in the sP.nse that it could be leached out of the waste if it were contacted with water.

The radioactivity associated with hulls and discarded ~quipment, Revision 1, Aug. 20, 1964

on the r,ther hand, is not considered to be "available". In the case of the hulls, the radioactivity is induced in the hulls themselves which are either $tainless steel or zirconium. Both of these metals are highly refractory and and would not be expected to corrode in the burial environment to any significant extent. They will have been carefully leached in bolling nitric acid prior to burial, inspected, and an aliquot analyzed to assure that significant quantities of fuel values ar~ not being discarded with them. Equipment to be discarded will have been exhaustively decontaminated in place before bringing it out of the cells and it will then be further decontaminated in the Equipment Decontamination Room before it is buried. Hence, significant quantities of "available" activity is not expected to be associated with this type of waste either.

7.1~ Burial.of both types of solid waste will be done in the silty till described in paragraphs 2.17 through 2.25 and 2.41. We h~ve now had considerable experience in working with this material in various excavations in the course of constructing the plant and in the operation of a low-level waste burial *operation for wastes of the first type described in paragraph 7.14. From this experience it is possible to accept the very low permeabil*i ty figures which were obtained during the subsurface investigations reported in Section II.

Therein a calculated *horizontal flow rate of 5 x 10-5 ft/day was reported.. Since we expect to carry out no burial operations within...100-.feet of any ravine, -this calculates to something over 5000 years for any leached activity reach the ravine. Further this silty till has been shown to have good ion exchange capacity for the longer lived isotopes, Cs-137 and Sr-90. Thus, we expect the natural defenses~-* this material to cont~in completely the activity buried in it.

7.16 Silty till does not, however, act as a natural ion exchange material for ruthenium. This is a relatively short-lived isotope, however.

For the sake of illustration assume that a curie of ruthenium were to escape from the burial site and begin to work its way toward on of the ravines. Further.assume that discontinuities or chemical reaction of the waste with the soil should increase its velocity by a factor of 100. It would still take over 50

.years for the activity to reach the stream.

In this period of time the curie ruthenium would have decayed to 10-l5 curie.

• The yearly flow in Cattarau~us Creek averages 3.5 x 1013 cc.

Thus, for each curie/year which was leached from the burial ground, the concentration in Cattaraugus Creek would be 3 x 10-23 ~c/cc. The MPC is 10-5 pc/cc.

Revision l, Aug. 20, 1964 l

@4Clittl :J

7.17 We expect the release of activity to the environment from the operation of the waste burial ground--either from low-level trash containing "available" activity or from the high-level waste described above--to b_e completely inconse-quential.

Egress of Personnel or Material 7.18 The control of release of activity into the environ-ment by carrying it out on th~ persons or clothing of personnel or on material leaving the plant must be accomplishad.by administrative means.

Personnel working with radioactivity in the plant will be provided with protective clothing which must be changed before they leave the plant. They will also be required to take a 1hower. Hand and foot counters will be provided for monitoring all persons--visitors included--

who leave the working areas.

7.19 Similarly procedures will be set up whereby nothing rnay be sent off the plant without first having been surveyed and smeared by Health-Safety personnel. Guards will be instructed not to pass out any material which does not have Healt~-Safety certification.

7.20 While it is possible that occasionally barely detectable quantities of activity might slip thro~gh these procedure~, it is essentially impossible for significant quantities of activity to get outside the plant in this manner.

No difficulty in contamination of the environment is expected from this operation.

Product Shipment 7.21 Radioactive shipments are covered by AEC r~gulations in 10 CFR Part 71 and 72 primarily.

All regulation! in effect at !h9 time of the shipment pertaining to such shipments are expected to be complied with by the shipper and the carrier.

The only way in which radioactivity could enter the environment by way of product shipments is fo~ the shipment to become involved in a serious accident. The regulations on product shipping containers are designed with that possibility in mind.

The hazard thus involved is not one peculiar to this plant, its design, or its operation. There is a considerable body of experience on this aspect of the business and we expect in no way to increase the degree of risk above that which has already been accepted.

Revision 1, Aug. 20, 1964

7.17 We expect the release of activity to the environment from the operation of the waste burial ground--either from low-level trash containing "available" activity or from the high-level waste described above--to b_e completely inconse-quential.

Egress of Personnel or Material 7.18 The control of release of activity into the environ-ment by carrying it out on th~ persons or clothing of personnel or on material leaving the plant must be accomplishad.by administrative means.

Personnel working with radioactivity in the plant will be provided with protective clothing which must be changed before they leave the plant. They will also be required to take a 1hower. Hand and foot counters will be provided for monitoring all persons--visitors included--

who leave the working areas.

7.19 Similarly procedures will be set up whereby nothing rnay be sent off the plant without first having been surveyed and smeared by Health-Safety personnel. Guards will be instructed not to pass out any material which does not have Healt~-Safety certification.

7.20 While it is possible that occasionally barely detectable quantities of activity might slip thro~gh these procedure~, it is essentially impossible for significant quantities of activity to get outside the plant in this manner.

No difficulty in contamination of the environment is expected from this operation.

Product Shipment 7.21 Radioactive shipments are covered by AEC r~gulations in 10 CFR Part 71 and 72 primarily.

All regulation! in effect at !h9 time of the shipment pertaining to such shipments are expected to be complied with by the shipper and the carrier.

The only way in which radioactivity could enter the environment by way of product shipments is fo~ the shipment to become involved in a serious accident. The regulations on product shipping containers are designed with that possibility in mind.

The hazard thus involved is not one peculiar to this plant, its design, or its operation. There is a considerable body of experience on this aspect of the business and we expect in no way to increase the degree of risk above that which has already been accepted.

Revision 1, Aug. 20, 1964

Conclu,12n,.

7.22 a, the ba1i1 of the data and cafculations presented in Paragraph, 7.6 through 7.21, i'n the normal operation of the ch1111ical processing plant described herein, there will

• be no discharge of radioactivity to the environment in excess of the l imits set forth in 10 CFR Part 20.

Abnormal ~erations 7.23 In Paragraph 7.4 five abnormal events were hypothe-sized. These *events range from the unlikely to the incredible but they delineate, we believe, the upper limit of any cat-astrophe which could occur in this plant and its related facilities. None of these accidents would result in levels of exposure to the general public exceeding the quide limits for gaseous emission suggested in Section 100.11 of 10 CFR Part 100, and further there is reasonable assurance that liqiud discharges at the site bounda11 could be kept within the concentrations for drinking water purpose specified in 10 CFR Part 20.

Loss from High-Level Waste Tanks 7.24 Careful measures have been taken to ensure the reliability of the high-level waste tanks, to provide multiple means of detecting any leakage in the unlikely event that any defects should develop and to minimize the effects on the environment of such leakage.

7.25 There are several methods of detecting leakage from the waste* tanks barriers between the stored waste and the environment. The tanks have been e.quipped with liquid level measurement systems which are accurate to 1/4 inch or about 700 gallons.

• The tanks are located within saucers and each saucer is equipped with a liquid monitoring system.

Each tank and saucer is contained within a reinforced concrete vault; the vault in turn is constructed upon four feet of graded gravel into which water is introduced for the primary purpose of maintaining the moisture content--and thus the bea~ing properties--of the underlying silty till.

There are eight wells located within a foot of the vault which go down into the gravel area and through which the level of the water is measured and from which samples may be drawn to determine if there has been any leakag*e through the first three barriers. If there sho~ld have been any large penetration of the first thr~e barriers, it would be possible to retrieve the activity with relatively little dilution by pumping out of the gravel area thro,*;!l any of

• the eight wells. This 3rea thus represents the forth barrier to the escape of activity.

Revision 1, Aug. 20, 1964

7.26 Th* local environment provides two additional

~nlen to the **cape 6f r1dioactivi ty from the al te. The tanks *r* located in the approximate center of a peninsula with a thick layer of aUty till. rt has been shown ~at

~h* perm.. blllty of this silty till la so low that essentially complete containment would be expected of any waste that did

.-.cape the first four barriers. The till, then, is a fifth and most important barrier. The aeninsula is bounded by Erdman Brook and Quarry Creek.

U&iS geologists who did the survey work on the site assure us that any radioactivity which escaped either onto or into the ground on this peninsula would eventually have to show up in one or the other of these creeks lf lt were not adsorbed on the soil by ion exchange.

At the confluence of these two creeks there is established a sampling *st&tlon to detannine again that activity has not escaped from the site. The average yearly flow at this point is about 2 cfs. While it would be expensive, it

~ould not be impossible to collect the total flow at this point and pump it bac~ up to the plant site for additional processing if this should prove to be necessary. This represents the sixth barrier. There is still a final sampling of th~ ditcharge in Cattaraugus Creek at the point where the effluent leaves the plant property. This will provide the legal record of the plant discharges.

7.27 A spare tank identical. to the working tank is provided so that in case the working tank begins to leak the contents may be transferred to the spare. Initially there will be a lal sparing ratio. It is.contemplated that during the first 1~ years of.operation of the plant two additional working tanks will be built and that the spare will serve all.three. Th~ eveotual sparing ratio will be dictated by plant experience.

7.28 c;;,~~that a waste tank could be ruptured only by sabotage or*by a major earthquake. The former is outs!de the scope of the requirements of this review. The latter has been shown to be highly unlikely (see Paragraphs 2.46 through 2.48). In the event that a tank shbuld rupture, however, the combination of the vault, the gravel area and wells, and the impermeability of the surrounding silty till can be expected to maintain the tank contents within the iamedlate area for a long period of time. There would be more than ample time to arrange a temporary piping system to permit pu:nping the waste solution from the tank, the saucer, or wells into the gravel, into the spare tank.

Revision 1, Aug. 20, 1964

7.29 The aultlpllclty of method* for determining. any leakage fl"OIII th* tanlrmake lt e11entlally impossible that.

auch leakage could remain undetected. There are so many

.batti"lrl between the waste and the environment that sign-ificant escape into the uncontrolled environment ls also conaidered impossible. We even consider it possible to suffer a complete tank %Upture--a most serious hypothetical and unlikely accident--and still maintain Cattaraugus Creek below the IIPC levels of 10 CFR Part 20

• Critic~llty Incident Anywhere in the Plant 7.30 7here have been eleven criticality incidents in solution systems.* Eight of these have resulted in a total number of fissions ranging from 4 x 1016 to 1.3 x 1018.

Oie, that at Idaho Chemical Processing Plant in October, 1~9, resulted in 4 x 1019 fissions. Except in one case in which there wa, some warping of a tank bottom, none of these resulted in any physical damaga. The assumption is made her~ that a criticality incident producing 1019 fissions in a single burst or 1()20 fissions in a repeating incident is experienced anywhere in the plant and that the entire production of noble 9aseous fission products plus 1/3 of tht iodines (from 1020 fissions) are lost. The value of 1019 fissions is chosen to conform to calculations made at Savannah River suggesting this value as the upper l~~it of a single burst. These same calculations suggest 10 fissions as the resultant of a maximum repeated burst. It*will be ohown that the limiting problem with this incident is not a public protection problem but rather the exposure of in-plant personnel to penetrating radiat-ion at the time of the burst. For a repeating incident there would be time to evacuate personnel after the first burst and the exposure to penP.trating radiation can be considered equivalent to that from a 1019 fission burst.

This is considered in Paragraphs 8.26 and 8.27.

Insofar as the general public is concerned there is no ~azard from the iaaadiate radiati.on at the time of the burst. It,.s well

• ~tablished that the limiting condition in an occurrence of this type is the thyroid dose from the iodine isotopes re-l eased. Therefore, this event is analyzed on the basis of thyroid dose to a person on the periphery of the site, at Springville, and at Buffalo. All three are calculated for the average and inversion conditions specified in Table 7.5.

In the case of Springville and Buffalo the total population dose ls calculated and expressed in man-rem.

• Nuclear Safety, Quarte~ly Literature Review, V~l. 3, No. 2, Dec. 1961, Pages 34-37 plus a subsequent Hanford incident and one in Olarlestown, Rholde Island in July, 1964

• Revision 1, Aug. 20, 1964

• 7.31 Table 7.31 lists the peak activity of each of the iodine laotopea 131 through 1~ and the time after the accident when the peak o~~~*! __ These have been calculated uiing NRDL-456,

• c.1culated Actlvltlis-and Abundances of U-235 Fission Products".

Wlth one exception, the p~ak activities have been assumed in calculating the popunticm-dose. This procedure ls conservative but by a relatively small amount over the time periods involved. The one exception is the activity of iodine-134 at the time it reaches Buffalo under inversion conditions.

The tranait time in this case ls so large in relation to the half-life of iodine-134 and its precursors that its activity level was found to b~ ner*igible compared to the remaining iodine isotopes.

7.32 Th* off-site doses have been computed assuming that the iodine 11 released from the stack instantaneously. The to~l inhaled activity has been calculated using Equation 7.7a for short-term center-line concentrations. The talculations have been performed for average (slightly unstable) meteorological conditons and for inversion (moderately stable) conditions. The distances involved area 51 te periphery*

Springville Buffalo 1,500 meters 7,200 meters 51,000 meters The use of Equation 7.7a J valid for the first two distances.

Extrapolation to 50,000 meters is questionable, but gives a fair estimate. The results so obtained are given in Table 7.32a.

Then using the approximations suggested in 10 CFR Part 100 for the thyroid dose from each of these isotopes, the total rem p,r person and the fraction of the 300 rem reference value are calculated 'for the three locations and for.both typee of meteorology. Total man-rem values have also been calculated.

These date are presented in Table 7.32b. Calculations supporting the numbers shown in these three tables are given in -Appendix 7.32. It can be seen that in no case is the reference value used for evaluaton of reactor sites exceeded or even closely approached. The highest value indicated, a 1.95-rem/person dose in Springville under inversion conditions, is not expected to be encountered since it is the opinion of meteorologists (see Paragraph 2.13) that an inversion aimed at Springville would be caught and held in the Buttermilk-Cattaraugus Valley systems. Even t~is value is only about 0.7 per cent of an emergency dose of 300 rem.

Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964

Table 7.31 Quantlti** of Iodine Iiotopes Formed from 1o20 fissions*

\\i_...

p,..-.:..

U~ Nr>tL* IS' Isotope Time of Peak Activity Peak Activity, Curies

!tz :....f ~

>>:!'?'E.

S,II, "'*"

I-131 5.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

, '.,. '/7, <<:

25 f""" 1) 7,",..

I-132 7.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

-i+t-,YJ 80

~. '? N

)... " ho.,,.

I-133 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> I Z. So ~ '11 -:

420

.:z, ~

I-134

~'

~770 1\\-11" 1 46.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.

-" 3 M

l.

~ ;"'

I-13'

/~

,+ 3 '2.o:,,

1470

<,7 H I r ;

Total 7765

/

""",,,... \\ C.f'

...., + L....u,i 0

a Assuming 1/3 of the iodines are lost from the stack

• Table 7.32a ToS,1 po,, PY* to Radloiodlnes, Rem!Person*

Inversion Average Location (lloderatelv Stable)

(Slightly Unstable)

Site Boundary 0.09 0.63 Springville 1.95 0.06

• Buffalo 0.33 2.8 X 10*3 a From instantaneous release of 1/3 the iodines from io20 fissions.

Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964

0 Criticality Incident in Fuel Pool 7.33 The fuel pool is designed to hold 1000 fuel elements 1~ racks of such_ geometry that the establishment of a critical array is impossible even if the elements were all of tha max-lllum reactivity of any fuel before it is irradiated. Allowance is made for ~vlng through the storage array an element of the highest reactivity. This is discussed in Section VI and ttie

• occurrence of a criticality incident here is shown to be extramely_unlikely. Despite the fact that a criticality incident in the fuel pool is extremely unlikely, the following event ls hypothesized*

It ls assumed that an element is janrned into the interstice between four elements and that the

• five elements are involved in a critical event.

that a 10-nwt boiling water reactor will be set.

up, and that it will operate 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> before it is possible to shut it down. It is further assumed that all five elements are defective and, thus, that some gaseous activity can escape from the element.

7.34 Calculations supporting this sect ion are shown ~in Appendix 7.34. The beet released would raise the temperature of the water in the storage pool only about 16F even if the pool water coolers failed to operate. Therefore, there is no danger that the water level in the pool would drop sign-ificantly and consequen~ly the shielding provided Ly the water would prevent any hazar~ from increased radia~ion levels from direct radiation.

• EBWR defect test studies have shown that the fraction of noble gaGaous activity lost per second from a defective fuel element is about 4 x 10-s. This same test showed that the iodine loss was at least an.order of magnitude less than this. The total inventory of gaseous activity ir, the five fuel elements assumed to be involved in this incident and the amounts which may reasonably be lost from the fuel pool water are shown in Table 7.34. These quantiti,s of iodine isotopes are much less than the amounts which ~ave already been *shown to be readily tolerated by this environment (see Paragraphs 7.30 to 7.32). Consequently the iodine releases result in less hazard than has already been shown to be acceptable. The releases of kryptons are also much less than those which have. already been shown to be within MPC.

Similarly, the xenon-133 discharge results in concentrations under the worst *condtions of only 0.01 MPC.

The only aspect of this hypothetical incident which has not already been calculated in the section is the xenon-138 release.

ReviQi~n 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964

I1otoee Kr~

Kr85

.Ki-88 1131 1132 1133 1134 1135 xel33m xel33 Xel35 Xel38 Table 7.34 Gaseous Activities Lost from Fuel Pool Durin9 Assumed Criticality Incident

• Inventroy fraction Lost Curies In 3 Hours 6.1 X 104 4 X 10-4 2.5 X 103 4 X 10-4 2.4 X l@

.4 X 10-4 2.7 X 103 4 X 10-5 3.2 X l@

4 X 10-S 5.4 X 104 4 X 10-5 1.6 X 106 4 X 10-5 1.6 X lo5 4 X 10-5 5.2 X lo2 4 X 10-4 9 X lo3 4 X 10-4 1.5 X 106 4 X 10-4 3.7 X 106

.4 X l0-4 Total Lost in 3 Hours1 Curies 24 l

10 0.1 13 2.2 64 6.4 0.21 3.6 600 1500 Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964

Using the method employed in Paragraph 7.7 (Equation 7.7a)

~h* concentration of xenon-138 at the site boundary under inversion con~itions is f.4 x 10-6 p.c/cc.

No MPC for this isotope is given in either 10 CFR Part 20 or in NBS Hand-book 69 but it would not appear that this would result in any hazard. This event can, therefore, be tolerated without exceeding published feiPC' s_.

Chemical Explosion 7.35 The a*ssumption is next made that a vessel containing one full day's charge of fuel in solution suffers an explosion which ruptures the vessel distributing the contents through-out a cell and putting some fraction of the contained solution into the ventilating system. The ventilating system will 8ithstand the rupturing of a tank. However, there might be some plugs or windows loosened. So long as ventilation is maintained air flow should remain into the cell except for the instant of the explosion. In analyzing this event some assumptions contained in "Radiochemical Facility Hazard Evaluation", by E. o. Arnold, A. r. Gresky, and J.P. Nichols,

~L-CF-61-7-39, July 10, 1961 are used. This is a very similar analysis of a completely analogous situation to that considered here. The assumptions are made therein that aerosols pene-trating high-efficiency filters will contain 0.14 mg/M3 of material with the same concentration as the original dispersed solution and that the MPC for mixed fission products is 6.6 x 10*9 a,.c/cc. The ventilating air passing through the fllt.ers of this plant amounts to 32,000 cfm or 900 M3/min.

Then 0.14 x 900 or 12~ mg/min of the original solution may be assumed to pass through the filter. w~ further assume that the gaseous activity has already been released and that in twenty.minutes the ventilating system will have picked up nearly all of the gross activity that it is going to; Under these conditions about 2.5 grams of solution will be released.

The maximum activity to be expected in the plant is about 700 curies per liter or 0.45 curie per gram for a total discharge of 1.1 curies. Following the methods of Paragraph 7.7 the poorest value of X/Q (at a distance ~f abou~ 5,000 meters) ls 1 x 10-5. Q is equal to 1.1 3600

• 3 x 10-4 curie/sec.

The X

• 3 x 10-9 p.c/cc. This is le s than t e MPC for mixed - :;..

fission products assumed by ORNL in the above report and it would appear possible to accept this particularly untoward accident.

• There would be, of course, a big cleanup job in the cells. This would be undertaken according to methods outlined in Paragraphs 8.8 and 6.54 through 6.56.

Revision 1, Oct. *29, 1962 Revision 2, Aug. 20, 1964

failure of Iodine Removal Equipment

. 7.36 Finally it is assumed that the silver reactors and other iodine removal equipment all fail and that this ls not discovered for a period of one day. That this could remain undetected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is extremely unlikely since the stack monitor would detect the iodine increase at once. If the entire charge of iodine-131, 1.7 curies, were to be lost during a day the value of Q is 2 x 10-5 curie/sec. Under worst conditions (at a distance5of about 5,000 meters) the poorest value of X/Q is 1 x 10-and the concentration of iodine-131 at this point would then be 2 x 10-10 p,t;/cc.

lhis ls less than the MPC for continuous exposure off-site.

,, *"' /

)

Conclusion I I.,o I u,,, **

7.~7 All of the abnormal incidents hypothesized in Paragraph 7.4 have been ana0yzed. It has been shown.that

_in all cases except the 1o2 fission criticality incident

• the limits prescribed in 10 CFR Part 20 for continuous exposur~ are met and that in this one case there is no dose at any point which exceeds, or even closely approaches, the guides suggested in 10 CFR Part 100 for emergency conditions.

~ince we expec~ the probability of these events to ba very low,.to the point of incredibil~ty, and since they can be handled by the environment even if they should occur, we submit that the operation of this plant does not constitute an undue hazard to the general public beyond the site boundary.

Revision 1, OCt. 29, 1962 Revision 2, Aug. 20, 1964

Before the UNITED STATES ATCIIIC* ENERGY Ca.lMISSION Washington, D. c

• 0 In the Matter of the Application of Ntl:LEAR FUEL SERVICES, Itl:.

for Licenses for a Spent fuel Processing Plant Under Sections 53, 63, 81, 104 (b), and 185 of the.

Atomic Energy Act "AEC Doc:ket No. ~-201 Submission No. 16 - final Safety Analysis Report Revision of Section VIII - Protection of Plant Personnel Revision 2, August 20, 1964 0

~""~

I SJ)U44 =* 7 &A JS

ca C

J U ¥ £ It 1

VIII PR0rEC1'I0i Of PUNT PERsmHEL l>e*lf! Criteria 8.1 The design criteria and the operating rules of the N1S plant have been set up so that the plant.will conform to th* rules and regulations specified in 10 CFR Part 20, Standards for Protection Against *Radiation.

8.2 The plant will have an across-the-board industrial safety program (see Section IX) aimed at reducing accidents of all types. It will maintain a con~tant program designed to increase.tM s,afety morale of all of its personnel, both ln the area ~f normal industrial safety and in that of radistion safety.

8.3 The radiation safety program is designed to protect the plant personnel froms

a.

external radiation,

b.

inhalation, *

c.

ingestion.

All three have been taken into consideration in the design of the plant. They also dictate the conditions under which the plant will be operated. In subsequent paragraphs each of these areas is discussed in detail to demonstrate that the plant as designed can be operated in accordance with the provisions of 10 CFR Part 20.

In addition, the accidents which were hypothesized in Section VII are reanalyzed from the standpoint of personnel in the plants and some less serious but more probable events are discussed from the view-point of pers*onnel protection.

Protection from External Radiation 8.4 The primary protection for t he worker from penetrati ng radiation is to interpo1e sufficient shielding between him and the radioactivity at all times. The plant shielding has been described in some detail in Paragraphs 6.~9 through 6.65. The shielding has been designed so that, when the most active unit which ~ould be in any particular section of the plant is present,. the radiation level on contact outside the shielding in a normal access area would be 1 mr/hr.

In many cases the point *of contact will not be readily accessible. to personnel and the percentage of the time that the shielding wall is subjected to the maximum activity level 11 small. The shielding design has been based on a "design" Revision 1, Aug. 20, 1964

fu*l !Mving the following lrradlatlon on history*

Burnup Speclflc Power Cooling Time 30,000 llfD/r

~ */r

~o Daya a.,

Fuel la brought into the plant in shielded casks which have had their design carefully checked to en~ure that adequate protection ls available. A shipment will be surveyed befo:re it is sent out.. It will be survey_ed again upon arrival at the plant. Before the carrier is opened, lt is placed under sufficient water (see Paragraph 3.7) so that, as a fuel element la removed. there will be at least 11 feet of water over the to~ of the longest type of fuel element. Movement of the elements ln the storage pool,and their storage are also conducted under at least this much water. Transfer to the PII: ls done remotely under water and back of concrete ahleldlng. The mechanic1l operations in the PMC and GPC are carried out remotely back of. concrete shielding. The transfer to the CPC ls handled remot4!ly. All operations ln tho CPC are remote. Transfers to the remaining contact-maintained cells

• are fluid transfers carried out remotely. All operations in the enti:re process, therefore, are carried out behind shielding until product ls decontaminated to the point where external radiation ls no longer a problem. Plutonium products containing high concentrations* of Pu-240 will be placed semi-remotely into con'tainers with suffic.ient shielding so that they may be handled safety.

8.6 Sampling ls an operation which can contribute significantly to exposure of personnel. The sampling systems, which were described in detail in Paragraphs 6.22 through 6.36, have been designed to permit most of the sampling to be carried out compl~tely behind shielding and to provide working back-ground of l mr/hr or less. Many dilutions will be made inside the shielding and only the diluted analytical sample will be

• brought out. This will reduce considerably the potential for spillage and also the resultant exposure in the event of spillage.

8.7 In order to maintain the background levels in the plant at design levels, it is necessary not only to have adequate shielding but also to maintain strict controls to prev~nt spillage. -This is done fir.st by keeping the activity back of the shiolding--there are no planned withdrawals of activity except for the samples, many of which, have been already dlluted1 second by a careful and continual radiation survey program to detect areas in which there may have been an inadvertent introduction of activity1 and third by a prompt and imnedlate cleanup of such areas at the same time deter-mining the cause of the event and corr*cting it.

Revision 1, Aug. 20, 1964 J

a.a

• tntenance work, both routine and major, can be expected to contribute aoaawhat to the whole body radiation of the plant ptrtoMtl. It l* the intention of NFS to

. l'llit.. 1ntenance work only under such conditions that no worker wlil be expo,ed ln excess of the limits defined in 10 CPR Part 20. The maintenance procedures, which are deacrlbed 1n detail ln Section IX have been set up to ainlaize the exposure of the persoMel. However, it ls clear that maintenance work will have to be done in high radiation area, (areas in which the background levels excHd 100 ar/hr). Such work will be controlled by a work perait ayat.. as described in the Health-Safety portion of Section IX and be a~thorlzed by the plant manager.

In attacking any maintenance job, the area ls carefully surveyed and the amount of time that may be permitted a worker in the are ls calculated. Work in the*radiatlon field is done under closely 1upervi1ed conditions. Accurate time is kept from outside the field. Recording meters as well as film badges are worn during the operation and a log of the exposure ls kept and this ls added to.each worker's permanent radiation record. The level to which an area will be decontaminated before maintenance is attempted-will vary with the amount of.time needed to carry out the job, but in no case will a worker be all"'.'fed to enter a radiation field exceeding 2 r/hr rdthout special approval of the plant manager.. It will be normal plant practice to limit the exposure of any individual for any single maintenance job to 0.2 rem. Subject to the maximum limitation specified above, the balance between time and activity level will be a decision to be made by plant susfervlslon in each instance.

8.9. In the.normal operation of the plant we expect that an operator will spend no more than two hours per day ln the full 1 mr/hr permitted ln a normal access area. It la expected that most of the normal access area will have a background much less than this. For planning purposes we have assum*d that the additional six hours per day will be ln an average backgrcund of 1/6 mr/~r. The total background radiation for the quarter would then amount to 0.2 rem.

This would leave about l rem per quarter for maintenance operations without exceeding 1.25 rem/quarter. With exposcre limited to 0.2 rem/maintenance jobs, a given individual could perform five such maintenance jobs per quarter. There will be 3bout. sixty men in the plant who can be called upon to carry out such jobs so that the plant can carry ou~ a maximum of 300 such operations per quarter, about five per day.

Revision 1, Aug. 20, 1964

0 Inhfl*tton 8.10 The prlaary protection of the workers from.

lnhalatlon it... in keeping the activity inside the process

• er tpaent ltaelf. As a second line of defense, all of the

• "'11pment ls contained in cells maintained by a separate ventilation system at a pressure negative to the working areas. As third line *of defense, masks and supply-air equipment are available. These ventilation systems have be6n described in detail in Paragraphs 6.3 through 6.21.

Under all normal operating conditions no process activity is expected to escape past the first two barriers and into the operating areas.

8.11 There will be a system of fixed air samplers backed up by a program of air monitoring with portable air monitors to assure that the air in the working area does, indeed, remain free of activity. This monitoring program has already been described (see Paragraphs 6.66 through 6.76). The monitors will have audible and visual alarms set to operate at the lowest practical level so that remedial action may be taken before any consideration of evacuation is nece ssa_ry.

8.12 Consideration has also been given to the mechanism whereby activity could be brought into the plant by recycle into the building air intake of. air discharged

• from the plant stack which ls located on top of the

• building. In Appendix 8.12 there are shown calculations for average and inversion conditons which indicate that the amount of recycle t~ be expected ls completely negligible in either case. There is, however, an infrequent condition whereby the discharge from a stack may come directly down upon the stack. Under these conditions the amQunt of dilution could be small. A calculation is shown in Appendix 8.12 for the normal iodine-131 discharge. This shows the concentration of Iodine-131 at the stack exit with no dilution at all except that afforded by the ventilation air in the stack itself. The concentration of iodine discharged from the stack would be 6.7 x 10-8 ~c/cc which ls only a factor of 7.~ higher than the.occupational MPC of 10 CFR Part 20. This particular meteorological condition is not expected to occur very frequently or to persist for any long period of time. Even with.Q2. dilution, and it would be expected that there wculd be som~~perhaps a factor of ten, the concentration is such that under the provisions of Paragraph 20.103b the iodine-131 present in the building air could be tolerated for five hours. Such a condition would be picked up very quickly by one or more of the monitors.

This iodine concentration would be attained only during the course of. a dissolution1* there would be ample time to shut down the dissolver or evacuate the building or both

• Revision 1, Oct *. 29, 1962 Revision 2, Aug. 20, 1964

8.13 There will be an ample supply of protective equlpnent such aa Scott Alr Packs available for use during emergency conditions or during maintenance work inside cells. The Health-

• Safety program (see Section* IX) will in~lude frequent training aeasiona and drills for all personnel in the use of this equipment so that ln an emergency the equipment should be used promptly and properly.

I Ingestion 8.14 The control of the problem of ingestion of radio-activity ls largely one of developing <<ithin the workers good safety morale and habits of personal hygiene and of providing them.with adequate protective clothing, devices, and monitors to facilitate the execution of the program. It ls a problem which has ~~en dealt with routinely at all of the presently operating chemical processing plants of the AF.C without creating any serious hazards.

a.1~ Protective clothing will be issued to all personnel working in the plant areas and must be worn therein. This clothing will not be worn outside the plant areas. It will be laundered at the plant and returned to service or discarded to solid waste depending on monitoring preceding and following the laundering process. An ample supply of other specialized

~rotectlve clothing such as sur9ical, heavy rubber, and *cotton gloves, caps, boots, and tape (for taping gloves to coveralls, for instan,e) also will be maintained.

8.16 Eating and smoking will be controlled throughout the plant. Eating will be done only in the designated lunch room. Protective clothing will not be worn into the lunch room. There will b~ hanci and foot counters at the door *. The area will be checked frequently by the Health-Safety survey.

Smoking will be done only in designated areas.

8.17 At the conclusion of each shift each individual who works in the plant areas will be required to change clothes, and check his hands and shoes before leaving the premises.

Two hand and foot counters are provided.

Revision 1, Aug. 20, 1964

8.18 The above program of hy,1iene has proved to be aatlsfactory to malntaln the ingestion of activity to neglig-ible levels at other installations. ~ The' coamon practice to back up Jils program with a medical program will be followed at the NfS plant. _The medical plans for the plant are as followsa The medical program will consist of a very thorough pre-employment medical history and physical examination for each prospective employee. The medical history will be aimed at not only past illnesses and injuries but particular attention wlll be pald to history of past radiation exposure, al~ergies, blood dyscrasias, tumors, and any evidence of emotional instability. The laboratory studies*

• .on all applicants will consist of a minimum of complete blood count, serology, urinalysis, chest X-ray, and vital capacity determinations.

~.ach employee will h~ve a complete physical examination yearly. A complete blood count will be done twice yearlyJ clinical urinalysis monthly. Bio-assays will be done on an

• across-the-plant sta~lstical survey" plan and follow-up examination, as this survey may indicate. The pre-employment physical examination and laboratory studies will be repeated on each individual leavirag the employ of the company.

A dispensary will be maintainP.d for care of ordinary minor on-the-job injuries. There will be facilities for intensive first-aid care of severe* injuries such as burns, fractures, and gross contamination with radioactive material. Inrnunization against tetanus will be routine for all employees.

Close liaison with the Health and Safety Department will be maintained.

The medical director will assist in health and gafety training and indoctrinationo He will review with the superintendent of health and safety all industrial radiation exposure recordsJ air, water, and plant radiation survey records. He will cooperate with the superintendent of health and safety in plant inspections.

Revision 1, Aug. 20, 1964 I

I j

Detailed records df all the above will be

..intained by the medical director.

8.19 As explained in the foregoing paragraphs we

~"ticipate no difficulty in conducting the normal operation of this plant within the framework of permissible levels of exposure. It remains in the remainder of this section to analyze the consequence to employees of ac~idents.

• Analysis of Accidents 8.20 In Paragraph 7.4 five highly abnormal hypothetical incidents were proposed and the effect*

of these upon the public was considered in Paragraphs 7.24 through 7.37. These same incidents are now considered with reference to the plant personnel.

Tank Rupture 8.21 It has been shown that the soil in which the tanks will be constructed in quite impervious and that the liquid from a ruptured tank would be held in the iaaediate vicinity for some period of time. It is proposed that the waste be transferred into one of the spare tanks as quickly as possible. The type of action envisioned would involve pumping the solution into a spare tank probably through temporary lines which might be laid overground with only a minimum of shielding.

The laying of this pipe would not require personnel exposure except for the connection into the ruptured tank. This would be done by lowering a flexible hose into the tank through one of the spare nozzles on the tank. If necessary, such an operation could be accomplished from back of a temporary shield constructed outside the radiation *field and pushed into place wi th a payloader or crane. If the earth shield remained intact, no additional shielding would be required. If the condenser system was inoperative, it might well be necessary

. to carry out this operation with the protection of supply-air masks. The transfer system would probably be set up with two pumps in the system in an effort to avoid any maintenance on the pumps during the transfer operation.

The pumps would be operable from outside the radiation field.

8.22 Maintenance of the pumps, if required during this operation, would certainly entail operations in a high radiation. In the transfer set-up a tee would be ln1erted upstream from the pumps so that the lines could be flushed and decontaminated somewhat before such Revision 1, Aug. 20, 1964

e maintenance would be attempted. It would have to be done from behind a portable shield. In the case of so serious a problem as this there would be no question that oper~tions of the plant would be shut down and all available exposure time would be used in solving this probl*em.

Supervisory personnel to the highest levels and individuals from other plants operated by the company--those who receive no radiation in the course of their work--could be brought in if necessaryo No individual, however, need be exp,>sed beyond permissible levels.

8.23 Most of the exposure associated with this incident would be in cleaning up afterward. Dismantling the highly contaminated lines, pumps, and valves and disposing of them would certainly require operations in high radiation fields. However, the need for speed would no longer be present and enough time and people would be used to assure that the task was accomplished within the permissible radiation exposures.

Criticality Incident in Plant 8.24 This incident, discussed in Paragraphs 7.30 through 7032, assumed that the ventilation system remained in working order since that situation results in the most iamediate and complete discharge of the 9aseous isotopes to the environment. In that event the air inside the plant would be completely safe, as evidenced by the calculations shown in Appendix 8.12, in all cases except that of the unlikely recycle. In Appendix 8.25 it is shown that if such a downdraft occurs under average conditions the amount of dilution will amount to a factor of about 5000 If it occurs under inversion conditions the dilution factor will be about

~. For the purpose of this calculation the dilution is taken as 10. It is further assumed that during the course of the entire discharge (assumed to be 10 minutes) the downdraft will be centered precisely on the air intake ten percent of the time. It is also assumed that some personnel are exposed to the resulting concentration* for the entire ten minutes and that they are so preoccupied with rendering assistance to other p~rsonnel or are otherwise so upset that they did not make use of the available supply-air equipment.

Under these circumstances they would be exposed to the concentrations and receive the thyroid doses shown in Table 8.24. These calculations are also shown in Appendix 8.,250 The total thyroid dose is less than the dose suggested as an emergency guide in 10 CFR Part 100.

Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964

Isotos,4t I-131 I-132 I-133 1-134 I-135 Table 8.24 Thyroid Dose During Recycle Coincident With a Crltlcallty Incident X

Cone.,

Q X 1,1,c/cc Frequency 2.6 X 10*3

...L 100 6.3 X 10*3

...L 100 0.04

_L 100 0.6

_L 100 0.16

..L 100 Dose Rate, Rem/

Time,

'1c/(ccHsec)

Seconds 110 600 4

600 31 600 2

600 6

600 Total Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964 Dosef Rem 5

0.7 24 22

~

74.7

a.~ There la no r1a1on to assume that the events which could lead to a criticality incident would al10 lead

.to ahuttlrig down the ventilation 1ystem. There may be, however, a iaall probability that the two events might occur 1laultaneou1ly. The possibility is a difficult o~e to analyze since the amount of li,akage of activity from the cell would be expected to vary considerably depending on the conditions. If th' supply air remains on, the exterior of *the cells.would remaln'at a higher pressure tlijln the cells and little, if any, leakage should occur. If both the supply air and the exhaust go out, there may still be a little negative pressure in the cells due *t., the natural draft of. *the 65-meter stack. If the static pressure in the cell becomto equal to that outside the ce-lls, some leakage wc,uld occur but it should not be large since any leakage path would be tortuo~s. 'There would not seem to be a mechanism whereby th~ cella would reach a higher pressure than the surroundings. Failure of the ventilation system will activate

• an alarm which:will require evacuation of the plant unless counte%111anded by plant supervision. We conclude, therefore, that *tf a *critical.tty incident were to occur coincident with a failure of the ventilating system, the plant would be evac~~ed long before anyone could receive a significant dose from inhalation.

8.26 The most*tmportant aspect of protection of plant

• personnel in connection With a criilcality incident is to assure th~t no one rec~ives a serious dose of penetrating radl~tion at the time of the incident. The first line of de{ense -is, ~f course, to prevent the occurrence of the incident. Gre~t care ls being exercised in -the design of this plant and in the setting up of its operating procedures to ensure tha.t

• a* crl ticall ty incident does not take place.

The whole subject of crlticality control throughout the plant ha* been pr,sented in detail ln the final paragraphs of Section vi; We believe that we have reduced the probability of such*~ incident to an absolute minimum.

However, there have been eleven such accidents in solution systems. Every

• mijor site save one has had one. There have been five incidents in metal-air systems at Los Alamos.

8.27 An Olk Ridge study* has calculated the prompt neutron and gamna dose at the outside of a normal concrete shield from a nuclear reactor of 1018 fissions and these data are shown in Table 8.27. They can be used fqr a 1019 fission event by direct ratio. Th9 concrete shielding walls ORNL-cF-61-7*39, "Radiochemical facility Hazard Evaluation",

E. D. Arnold, A. T. Gresky, and J.P. Nichols, July 10, 1961, Page 6.

Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964

Table 8.27 Th* Proapt Neu_tron and Gaaaa Dose*at the OJtslde

• Of *t Nort11l *eo,,crttf Shleld From a Nuclear a.actloh of 1018 Fifsions*, 6 QrdlnaryC Concnte Shield Ib\\c1to1**, Ft

.Dose at Q.Jtsldt of Shield, rem Metal Nuclear O Nuctear-Reactlon in Rea:tion Aqueous Solution -

1 3

88,000 317 17.0 0.960 o.~9 5,200 23 1.9 0.14 0.012 a

The dose rate may be calculated for any other number of fissions through the use of a direct proporetion.

b ORNL-CF-61-7-39* "Radiochemical Facility Hazard Evaluation", e.o. Arnold. A.l. Gresky. and J.P. Nichols.

July 10, 1961, Page 6.

c F~r high density concrete the ganwa dose is reduced by a.factor of 1.6 for a given concrete thickness

• for the GPC, PIC and a>>c have openings for viewing windows which aze 4t1Ulvllant in shielding valuo to the concrete walls for gaaaa radiation but offer lea, protection than th* concrete for neutron radlatlon. Table 8.27 does not

~uilect thelncreaaed neutron do1e to an employee who might be 1(1 front of one of the viewing window,. In the case of a 1o.r.9 burst from a criticality accident in the dissolver, the total prompt neutron plus 91111111 dose to an employee at the nearest vi1Wing window would be about 300 rem, lf the window were coaipletely transparent to neutrons. The only...,

place ln the plant where ametal-alr incident 11 at all poaaible 11 in the PIii: - GPC. There we have four feet of hlgh denaity concrete shielding and the resulting dose would be negligible. A solution system event could conceivably occur from the di1solver on. In the CPC there are six feet of concrete and the dose would be even less than in the

~. In Cell #1 there are flve feet of concrete shielding.

The dote would still be negllgible. In the remaining four cell* there are three feet. At the lower end of the process there ls no need for this much shielding from the fission product *c~ntent. The minimum of three feet of concrete ahleldlng has been carried down to the end of the proc*ess in order to assure that even if a 1019 fission critlc5l incident should occur, and a worker should be standing

.right opposite the point ln the cell at which the event occurred, he would still not receive a MLD of penetrating radiation.

8.28 When the product must be removed from the piant and put into storage a~d eventually onto a truck for shipment, conta,t with the product ls required. Therefore, particular car*.ha* been exercised ~1th product shipment plans. This 11 di1cu111~.in Paragraph 7.21.

Criticality Incident in the Fuel Poo_!

8.29 The hazard of 3 criticality incident ln the fuel pool to t:,e general publlc has been discussed ln Paragraphs 7.33 and 7.34. It was shown therein that the amount of heat released ls not enough to destroy the integrity of the water shielding which ls enough* to ~eep such an incident from ir-radiating anyone significantly from the prompt neutrons and gallDls. It ls necessary to consider the gaseous activity which ls given off, however. The quantities of gaseous isotopes expected to be released during three hours was 1hown ln T~ble 7.34. In Table 8.29 these quantities are shown as pe/cc and their ~oncentratlons ln the fuel receiving and storage area air are ~hown assuming that it is diluted wlth the 11,000 cfm of ventilating air which ls drawn through Revision 1, Oct. 29, 1962 at.vision 2, Aug. 20, 1964

Table 8.29 Ga1eou1 Activities Loat into Fuel Receiving and Storage Area During Assumed Criticality Incident-Activity Released

Cone, Isotope

,,,.c/sec

,,,.c/cc..

MPC Kr-85aa 2300 4.4 X 10*4 6 X 10-6 Kr-~

93 1.8 X 10-5 1 X 10-5 Kr-88 930 1.8 X 10-4 I-131 9.3 1.8 x* 10-6 9 X 10-9 I-132 1200 2.3 X 10*4 2 X 10-7 I-133 200 3.8 X 10-5 3 x 10-8 I-134 6000 1.1 X 10-3 5 X 10-7 I-1~

600 1.1 X l0-4 1 x 10-7 Xe-133111 20 3.9 X 10*6 Xe-133 330 6.4 X 10-5 1 X 10-5 X~l35m 5.5 X 104 1

x

• 10*2 Xe-138 1.5 X 1<>5 2.7 X 10-2

that area. Th*** concentrations range from twice the 40-hour MFC for Kr-&!\\ to 2000 times the MPC for I-134. -These IIPC'* are for continuous breathing and can be scaled up or down ~1th tllle. Taking the I-134 as controlling, the room would have to be evacuated within 1/2000 of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> or in just undtr one minute in order not to exceed one week's allow-able inhalation. It will certainly be possible to evacuate this room ln less than a minute and an event such as this would be iaaediately obvious to anyone in the room.

Presumab\\y there would b~ a visible flash. Monitors would trip and alarm, and the fuel pool itself.would be visibly agitated. After evacuation, personnel could put pn supply-alr equipment in other parts.of the building before re-entering to take remedial action.

Chemical.Explosion 8.30 In.Paragraph 7.3~ a chemical explosion is assumed which ruptures a tank containing an entire day' s charge.of activity.les~ the gaseous activity (sine~, in orde~ to have the full day's charge in solution, it will have to have been through the dissolution step during which the gaseous ac~ivity ls lost). The cell ventilation system has *been designed to withstand *the effects of sucn an explosion.

So long as the ventilation system is maintained, no activity,hould get out of the cell in which the explosion took place e~cept s~e that would be lost during the period of oyerpressure following the. explosion. This period is estimated to be about one second.

The calculations of Appendix 8.12 show that,' for all conditions except a direct recycle of stack discharge into-the air intake, there will be negligible concentration of ~ctivity in the building air.

In the unlikely event that such a recycle and an explosion should coincide and using o calculative method analogous to that shown in Appendix 8.25, the concentration of unfiltered solid* activity at thr tnroat of the stack would bea 0.12~ g/min x 0.45 curie/gram x 106 pc/cc 5

/

32,000 cfm x 28317 cc/cf

• 602 x io-p.c cc As in Appendix 8.25 a dilution factor of 1/10 from stack to intake and a frequency factor of l/10 were used~ The resulting concentration in the building would be 6.2 x 10-7 µc/cc.

This is about 100 times the MPC assumed by Oak Ridge for mixed fission products as aerosols and it implies that about 25 minutes would be available for evacuating the plant. This ls more than adequate.

Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964

8.31 Although** believe that the ventilation system

• 111 be.. intained in operation if an event such as this ahould occur the possibility that it does not continue to function 111.11t be considered. As in the discussion of Paragnph 8.25 we find this si'tuation diflicult to analyze and for the same reasons. In this case it is certain that the cell would be pressurized for a short time, perhaps several seconds, and that du1ing this time some activity would escape. If the building ventilation system were iOt functioning, the air in the building will be essentially stagnant. In the ianediate vicinity of the cell quite high concentrations can be hypothesized depending on the assump-tioni chosen. However, it seems difficult to hypothesize a mechanism whereby this activity will spread ver, quickly from the immediate area if the building air is not moving~

An explosion ~uld alert anyone close to it and the inmediate area would be evacuated in a matter of seconds. Reconnaissance and remedial action would then be carried out with supply-air masks.

Failure* of Iodine RP1Doval Equipment 8.32 As in all of the situations in which activity is put up the stack, here again there is no hazard inside the plant at all except in the unlikely case of direct*

recycle the concentration of iodine that might be founc in the building air would bes 20 ~c/sec x 0.1 dilution factor x O.l frequency l.~ x 107 cc/se~

= 1.3 x 10-8µ:/cc The 40-hour MPC for iodine-131 is 9 x 10*9 p,c/cc which is lower than *the above calculated concentration by only about 30 percent. Thus, this concentration could be permitted for over two days. It is unlikely that the recycle would be in-just the right position for more than an hour during the day. In any event this concentration would trip all the building air monitors and the incident would be dis-covered more readily and the dissolver shut down.

Minor Accidents 8.33 We do not believe that the accidents which have been discussed in Sections VII and VIII will occur. It.is a good deal more likely, however, that during the course of the operation of this plant there will be a number of much more minor occurrances which pose no hazard 3t all to the general public but which, if they were not handled properly, could lead to additional exposure of the plant personnel. Such Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964

event, alght tp., llluatr1ted bya

a.

Spllllng of activity, particularly analytical aamplea or special samples such as waste tank sample.

b.

Tracking of spilled activity from on& place to another.

c.

Pulling activity into jet steam lines by improper venting.

d.

Leaking waste lines in diversion boxes,

Spilling of product solution.

The problem with all of this type of event ls the same.

In one way or anothe~ they lead to an increase ln the background radiation which the worker may receive. This is undesirable since in nuclear work one wishes to avoid any unneces.sary radiation. It is also undesirable since it is important to keep the "operating background" as low as possible in order to leave a cushion with which to carry out the required maintenance work. There are several lines of defense against this sort of problem and they are the same for all of thems

a.

First, the plant has been designed to eliminate, insofar. as possible, the necessity for handling even small amounts of activity.

b.

Second, the operating rules are designed to eliminate 9very possible exposure.

c.

• Third, the fixed monitoring system (see Paragraphs 6.66 through 6.76) is designed to detect increases in either background radiation or air concentrations.

d.*

The fixed monitoring system is backed up by a formal mobile monitoring and survey program (see Section IX).

e.

Each employee will wear both meters and film badges. He will be trained to chec~ his own e:q,osure rate frequently and to use portable monitors himself so that he need not rely completely upon Health-Safety Coverage.

Revision 1, Aug. 20, 1964

Thu*, the** alnor accidents should not go undetected. tlone of th*** accidents could credibly result in exposure to plant personnel ln excess of the limits set forth ln 10 CFR Put 20.

. 8.34 In-this section we have shown that we are able to operate the NFS plant under all normal conditions within the requirements of 10 CFR Part 20 as they pertain to the protection of the plant personnel. This includes protection froai external radiation, inhalation, and ingestion. Both the operation of the process proper and all the necessary aalntenance operations are included in this stat~ent.

a.~ The same aeries of hypothesized major accidents that were discussed in Section VII have been considered from the standpoint of the protection of the workers in the plant. It i ~~~~9Wn that each of these unlikely events could be sustalntd without undue risk of exposure to plant personnel.

8.36 Finally the problem of minor accidents that could lead to increases in the background ~adiation received by plant personnel ls considered and the multiple series of defenses against these *are shown.

8.37 We conclude that the NFS plant can be operated within the requirements of 10 CFR Part 20 as they apply to the protection of its own personnel'.

Revision 1, Aug. 20, 1964

I 0

0 Before th*

UHITED STATES ATCJlIC ENERGY ca.NISSION Washington, D. c.

I In the Matter of the Application of

~CLEAR FUEL SERVICES, INC.

For Licenses fo~ a Spent Fuel Processing Plant Under Sections ~3, ~3, 81, 104 (b), and 185 of the Atomic Energy Act AEC Docket No. ~0-201

/-_* \\

~-**.

f. ;,;*~-

. \\

I I I I

Submission ~o. 18

• Final Safety Analysis Report Paragraphs 9.0 through 9.96 of Section IX of the Safety Analysis I

u~nn I I

0 IX PLANT OPERATION 9.1 Detailed in this section are the following ftems1 The organizational make up of the Spent Fuels Reprocessing Plant; aspects of administra~:ve control and procedures in various operations of the plant; Training Program, Health and Safety Program; Fire Safe~y Program and Emergency Proce~ures; the uses of the Operating.Procedures and Letters of Authorization;~ discussion of Maloperation; and the use of Maintenance Procedures.

Organization Plant Manager 9.2 The Plant Manager is responsible for all activities at the plant and is, therefore, concerned with all aspects of plant operation.

The more important areas include production, technical services, health and safety, and nuclear safety.

Production Manager The Production Manager is responsible for carrying out production in accordance with approved procedures and accepted Health and Safety standards.

Health and Safety ~irector The Health and Safety Director serves in a police and guidance capacity to assure conformance to approved Standard Operating Procedures and to advise in plant operations from a Health and Safety standpoint.

Technical Services Manager The Technical Services Manager is concerned with the technical soundness of the operations proposed, the surveill~nce of material, and ;particularly, as a membgr of t~~_criticality group, the maintenance of a critically safe system. -He generates**appllcations for license revisions. He reviews proposed SOP's and Letters of Authorization to ~onfirm compliance with the license.

Plant Criticality Committee

• 9.3 The Plant Criticality Committee consists of the Plant Manager, the technical Services Manager, *the Health and Safety Director

• and the Production Manager. This committee sits in individual judgment on all SOP's and Letters of Authorization. Each member satisfies himself that the proposed procedure is in compliance with approved Health and Safety policies and that no criticality problem is involved. Each member gives particular attention to the function that he represents. The usual seq Jence for review is1

1. Production M~nager

2. Technical Services Manager

3. Health and Safety Director and

4. Plant Manager.

HI.Jo&*UtSl9o 9.. 4 The *ln functlon ~*of una,... nt*':t.& t,r:~afa,ly ~

ec~cally ~lnlattr 111 operations r*latt~e t~ ~- ~~t... Th~. _fl*".'~

llaNSger, -~ Ml overall respon1lbillty f~i> Pl&niJ ~r*f~r.l~*~I *rr** *d~l*CJ~ted.,

cert~in reaponslbllltles as 3numtrated ln par*gr:tii't"9.Q~ :.:-.I~ :additl,on to the &bovt, ~.ther delegation of responslblllty t:S, ij_,: ;~!)owtr ~

_tbe-- A~ttis~~~t!vt

~ganlz~tlori Chart, f lgure 9.4.

-,~ -

• Operating Procedures and Letters o*f ~Au$h9~lza~t*on..

,* ~

. 9.~ P;roc:e11ln9 of all spe~ial ;,,wel_e~r aaatiril'l}~~~~~;d/.

under the

• llcenst ls done in accordanc* wtth* th* crl~eri'ilt set-:f~r:th'.'1in

~ Uc-tne&. ' AU opera::tlon$ in th* Spent* F~ct.l Repr-C,C*l>SStng PJ:"(fi~*~afe

~* in accordance wt th*'-appraved operating :pr~ed~~~s.;,~fch*:,det~n,. th**

Ntf\\ods to bo used and incorP.)rate* criteria. contair5,~d : tn tji~ l:l~on"t*e..,

)...

r

~

9.6 It wll'l be th* responslbUlty o°f. eac.h e~plt>Jtt.. to.z:.ad, 1',nder1.tand at'\\d foUow expllcltly the dlrt~t1ons -cont*alned-ln the. Stanaa*rd*

Operating P~ocedures fo~ j~bs which. he \\s ~alltd :upon t~~~; it 1~ tht

.r*1ponslb1Uty of. ~etch s~pervi~or to,know* ~tie. Stanclard. Operat:lng" Proc'edures whl.ch ipply in* his area,* tc,.have copies 9f. these.sop**s~.~vat&a~!'a.for.,_~

employeea to read and to-:be certain that llru:Mviduals und~.r hi~ su~~lsion read anict' und*J'lstand.each procedure. It ls: Manage~nt '-s

• zespon's.lbiltty to r*~lew and re-issu.}SOP':*~ as necessary, to re:ne_et\\ ct>lang_es*',in tp~ procsss and to insure that the inst.ructions contained in SOP' s rep:r,eset"\\t* ~ safa*

and efUclent,!19thod, for ?~compllshing tne work.. *.

g..1 SpeclaI nuclear material ls. recQlV~d into the plant*

and

• pr_oces~ed by Appr:oved Letters of Authorliation whic~ 'state* the ~perat ing proclkt\\µ'e(s) to be tisiKI, specl1l handl \\ng where, r*equired,,* th~ cust9iHr.

fen wf\\Odl tn* -p!'OCi!tslng ls being done, the contaln~r(s). in *which the SJl&terlal will be* found and the material to be.uned... as to t.ypti; enri(:hment 1nd weight. '81,f~rt a Letter ot.Authorlz~tlo~.cin~be used it ls inaependently

.tOView~ by. tac~ mem..,er ~, th~ Plant Critltality Con11J1lttee t <i:assura its confc,~e, with approved license criteria. It h. the 11Spons1b1Uty of each tm~loy,~ to r,idr under.stand l~nd follow exp~~c~tly th, *dlre~iiono co{'taln~**1n th** Le.tt~r{s) of ~u~horlza_tl~n.. It! ls th~--r~sJjonsibil~ty of each s~~rvJso~ to have a copy uf t~e Letter(~) 9f AJtho~lzatiqn av111~~lt for -t~ndl~id~als under his su~~vis~~n to raa~ a~d to be certain that )~h*~ reQd and unders~and the inst~ctlons containe~ ln the Letter(s) of:~u\\~orb*tlon. It.ls Manage~nt's respons!ibillty* to iss~e Letter(s)

{if.., Authorization t.,ereby scheduling work thrataghout the plant$ Normally, p.ioc*es.s engln'lerS "nd~r the supervision of. th*e Pr~uction Manager or tha fechnic*l S.rvlces Manager draft the procedures or authorization. T_hese

~nglneirs, serve aito t~ tec~lc~l liaison and guida~ce in prQduction and thiy conduct and supei:vlse englneet'ing development.

9.8

• the general ad&niftlstrattve philosophy will b, to

~st~bli.h standal'Cl~proc~dures. for at many situations as* possible and to

~ontrol th* t1fectlvtnes* of the,*-procedures by aneins of regularly

.malnt..lned.log* and check-off lists." These procedures,toget~et with th~ir

Se-curl ty Of fleer Industrial Rel*tlona Office Manager C

Pl'OducUon lf.ana9!r I

Assistant Production ~r Plant Engineer I

r Shlft-S rvlson Fl9un:.9.4 C... t>

plant Ot9anl1atlon *Ctwrt.

(

.Plant 1ta~9er I

Aa1l1tant -to :

r~

I Tr,;:! Di~ctor I Mecllcal-Dlrecto~

Technical Director Technlcal-Sirvlces llana9!.:

I Accountjtil_l}ty Of fleer Analvtlcal Services Ma r

Dh*-ctor

avpponlng 1091 lf4 chtck-off llatr.wUl be 1ubJ*ct to re~lar, but randoa.

lns,-ctlon by higher levtll of autho.rltyr. For in~t*nce, c1~ain routine txuln.tlon* and..,,u.r... nt1 wlli bt carrted out dally :*ccordln9*to approved chtck*off li1t1 and duly logged. In th*** ca111 th* next higher level of 111pervt1ion wlll, once a week at a random tJ*, follow through

~

apec,tr tc procedur* and d1tt?!9ln1 that lt 11 being properly ca~rlod out. One* every_ two aontha the *.~ext higher l1v1l of supervision will do llkewi*** On;*..;:* year th11~ suae( pro~edur** will ~ \\observed by the h~gbest level :;of authGrlty. ~* Pers~nal 'ltspc}nsibUity wlll be 111phalized by hwlng *~ch one of theie*.'tnspie tlon1 : recorded by. 1 lgnature and date a the log* wlll be kept.11 f~. pe~nent' r1c~rd.

In adcUtion;* duty ll1t1, lddrt*~** and telt~cm* "J!~*~*.,-~11~*,ba-:~1,.t1lntd, and 11lect1d gr9ups

~e off*du.ty perso~et,... ~"'al.1 *1~1~ of authority and skill, will b.e nqulr.td to keep

• the:pJ~i ;info~ ~rF that.. ; wher,about1 at all times for

~r941ncy c'&11.

Jrtt~l09 of. Plint Pegso~nel_

_

• 9o9 Th* initial. staff cadre will be largtl¥ IRl<lt *~p of

-~Ople wl~.~xtenslve experltn*~* -in ~- handling, pro~~sslng_ and *~nitering.

af.1'4ioac\\ lvo Ntirials. Thh grlNPe. und,r *the Training Director, will

. ~~t ~ - training c°-"i-.tes -f~?> *111

• a~l tlonal. employees.

The curriculum

(--*Appendix 9.J!t.~wtll.. be* dlr*ct~

• towar~ th~. ~ucation *of ce~tain ph~at P!iti'ioMel ~n the.pr~c*e,1e1 ;and r*l*~ed o~~*~lons in. such detail air to

. *tn,~. coaplet* famlll*ritj with th* equipment, tts function and.~oaipet,nce in.. tt, operation.

~

9.10 It ls th*' lntent of the *tral~lng _program to,nable proc*** (iperat~r* to successful~y s3~isfy-AEC ;requlteoent1 for oper~t9r1' llcen*** i,y: te*t and.. e)(amlnat.lon... Approxlmattl_y 7~ operators* wlll ~ so train~* by.,pemn*ent* or 'tn~rary.staft~*~mt,,rs..Initially, three types oJ operator, wUl-be t g11lried *fo*r trork ln thl'ee dlff erent types :of are*ata

1. l:it~~~l1t lv1 Pi oce11lng Oper~tlons

2. Chnlc.al Processing Operations

3. c.on~rol Operitions tu. cadi"e, in ac!iltlo~> to s,rv*l~ 11 the faculty, will.take these and adctttlon~l.. courses ?'e*19ned to satl_sfy AEC requ,lr~nt~. for ~enlor.

Oper&tors' Llcensgs. C.rt~in,employees such *~~w1tchmen, secietarlei, etc.

will ~

.txM~~ fl'.~~~~ of _-th, more tee_hnicil aspects of the currlculuftl.

but~*ll ~loy~s will ~ ;.expo;ed to, a radlelogl~al famlllarlzatlon cunlcul.._; Wt.l tt!n 'fxtstii*~atlona, g:r;ded according to*.leve l of re1pon1.U,l U ty Ind WOfk exooaul'*.; *w,Ul ~be.co~ducted~

• A

• *J._~l th~ c~rl~l~~wlll lnclud1 an. -l!'t::oductlon (comprising detail* ~.;; th~.. b~kg~.nd,nc1 :~eactlptifl' material o()tbe pli~t) detalla of tht proee1101, healih-and *afetv., *1nat~ntat1on, 1qul~nt description and usage; -.aechanlcal:.. ~nlpu*l&tion, *process control, proee11:.maloperatl'on, d~.,nt*lnatlon procedures, *waste treatment., ***rgency *1sur11, *,ccounta-b111tl, **conomlc ~~ ct1t1c~11ty conaldtr~tlon,~,nd lay ch,mlttry and

• pi\\y1lc1 ***oclatf!Cl.wt'llt reactor os-r1tion1 and chealcal reproce11in9

• In f!ddltlon, tl\\t currlcul~ for th* cadre and others preparing for Senior ~

Ope~*t~* llc*n*** wlll include.th* condition* and limitations in the factllty llcen*** ti\\e design and operating limitations in technical speclficatl~~*j ihe aiec~nisa for any changes ln the limitations in the 11*cen~*-oi-lptelflcatlons and more advanced study of che1~istry and

• radioactivity. ' The training program for plant personnel will be a

.contfl'!Ulng one. - Regula~_process operators will.be given, periodically, a rtorfentatlon exposure to radiation safety ar.d to processes and equipnent lnvqlved ln their particular plant.specialty. New employees will be tndoctrtnated by training as are the *1niti*l employee~ and will be requ~red ;to pass the same, m:s examinations in *ddition to AEC license ex.,.1.nations.

Training of Outsid~ Or91niz~tion1 9.12 Pa~tly as *a matter *of public relations but primarily to obtain efftctiv* 1nd. non-p1nicky assistance if an emergency requiring t.b*lr coo~~atl~n ihould develop, local town, county, state police officers, fire departmentt.of.the ar,a,.civil defense organizations and. elected off ~clala wlU be_ invl ted to lectures at the plant. The subjects covered will be,Jialnly those connected with protection of the *pu~lic and will be designed to ett1blish *method1 of Ua1son.and.cooperation 1(_dealrabl& and nee***~ und~r hypothetical e~rg,ncy conditio~s so that assist,nce it IIIO&t*effectlvt a~d radlologlcal haza:1(1s to outsiders are mlnimiied.

Health ~net Safety Program

~.13 The Health&' Safety i:>e'pa~n~ is charged with the re1P9.naiblllty* fot* protecting. plant per,onne.l from all job hazards and the

-public frQill *haza.rdous quan~ities o( ~a<ilation*and* ridloactlve materials.

lflthln th* scope of thit re~pondblll ty 'the Heal th~ & Safety Department wllh.

1. lil~itor for... radlation and contimlnation.au plant

~eis and. op.rati(?r\\ff (see ~ppendlx 9.13 *for equlp111nt.)

2. Monitor, for radiation and cont'amination, areas

. ex\\ernal to th* plantp

3. Approve ~oeedt;Jrts for work wl th radioactive, aaterial* 1.

4. Esta~U*h e1JMtrgancy,-procedures1 f

~ *. :&tabllsh.Ua11on with all other departments and advli* tttm lrrmatter1 p,,rtaining to health and 1afety1

, *.S~perv1** 't)rje rec*lpt atid 1hipnent of all h81ardou1 aaterlat*,

7. Provide currlculua and teach 11pec:t1 of the health and 1a,fety progr*1

a. !1t1bll1h and maintain plant fire brigades trained to cope with radiation area firesa

9. Conduct a continuing safety training program for all 111ploye111

10. Conduct inspections of all areas for fire and safet'l hazards and institute corrective action when necessary,

11. Viaintain complete, accurate records of personnel exposure, radiation-contamination con~itions in and around the plant, and perform radiation instrument calibration.

Health and Safety Organization 9.14 Specific responsibiliti*s for members of the Health and Safety group are as follow**

Health and Safety Director *

• Plan, organize and supervise the work of the department.,

MalntMin close liaison with the Medical Director advising -* and seeking advice cancernlng the employees* health and welfare. Maintain cl~se liai,on with

• other departments and advise them in matters pertaining to health :and'*safety.

Maintain complete, accurate records of plant, personnel and *environmental radlation-contamin~tion conditions. Administer health and s*af1tty aspects of training programs. Organize and train plant fire brigad.~s.

Inapect and llliintain fire fighting and emergency equipment~ ~repafe material for use in safety meetings. Conduct fire and safety inspection,.

Lead Ttchnlcia,i

• Perform routine. and non routine mon!toring. task~ at /directed.

Writ* complete, accurate reports of condittons observed. C&Jib.tite and check monitoring instru,11er1ts. Obtain and count air samplll'~'-*. JYerfol'i!i safety ~nspections. Participate in shift safety *traini~g 'prcgrams a~d

• ~fety meetings. Maintain exposure and survey. re~ord.fil, i.. Check 91111111 dosimeters and record results. Prepare film badges for distribution a~ _processing and record results.

T9chn1cian

• Medical Perform routine and non rou*tine medical tests on employees inc_luding processing bio-assay specimen,. Receive environmental and aonf~oring samples and 'prepare them for counting.

I*shntc\\an - Shtft ParfOl'II routine and non routine monitoring and inspection t11k1 a, directed. Participate in shift safety training programs and saffty

..,tinga. Wrtte complete, accurate reports of 2:tivities.

Radiation Area Work Procedures

. 9.1~ All radiation area work ls governed by procedures approved by responslb~e persons in Productio~ Plant Engineering and Health and Safety.

It is the intent of these procedures, in accordar1ce *with NFS policy, to incorporate sound ind~strial ~afety practice and to maintain exposure of employees to ionizing aadiat!on and radioactive,.contamio*ation at a level below the limits stated in 10 CFR 20.101 and appendix B, through the use of monitoring,decontamination.and shielding techniques and through the use of protective clothing, respiratory protective devices an~ other safety equipment as requit ed.

9.16 For the purpose of defining radiation areas, the following

~es are establisheda Zone t All areas beyond the site perimeter boundary, Zone II All areas within the site perimeter boundary which are normally free of radiation--contamina-tion in *~cess of 500 d/m alpha and 0.05 mrad/hr beta-ganma 1 Zone III All areas wi~hln the cite perimeter boundary

~hich may have detectable radiation-contamina-tion but !n which the radiation level is normally less than 100 mrem/hr and the contam-ination level is not signlflcant1 Zone IV All areas within the site in which the radiation level e>:ceeds 100 mrem/hr or in which signifi*

cant contamination exists.

Zoning of the plant and site will be the responsibility of He~lth and Safety.

9.17 The General Regulations for Radiation Area Work will apply to all work protedures. See Appendix 9.17 for listing of equipment.

General Regulations The minimum requirements for protect-ive clothing area

.a. For entry to Zones I and II, no protective clothing is required J

b. for entry to Zone III areas, for inspection only, the alnlaua protective clothing required shall bes Laboratory coat, shoe covers and glovesl

c. For entry to Zone III areas to perform work, the minimum pr~tectlve clothing required shall bea Coveralls, shoe covers, gloves and cloth hatl

d. Protective clothing required for entry to Zone IV will bo specified on a Special Work Procedure.

No one will be peX'llitted to enter a Zone IV area until L Special Work Procedure has been completed and signed and all provisions of that procedure have been implemented~

• • Respiratory protection requirements will be posted in the "hot" lobby.

The minimum requirements for personnel monitoring area

a. For entry to the Plant, the minimum requirement for personnel monitoring shall beJ *-Badge

b. For entry to Zone III areas, the minimum requirement for personnel monitoring shall.beJ*-Sadge, dosimeters and dose rate type radiation survey meter.

c. For entry to*zene IV areas, the personnel monitoring requirements will be specified on the Special Work Procedure.

The exiting,rece4ure 111

a.

When leaving Ze~e IV areas the minimum requirement for personnel survey shall be A complete clothing and body survey by Health and Safety Personnel,

b. When leaving Zone III areas the minimum requirement for personnel survey shall be A complet& self survey at the station monitors located at the Zone III - Z~e II boundaryf

c. When leaving Plant Zone II the minimum requirement for personnel survey shall be A hand and shoe check using the ftand and shoe counters and station monitors in the building lobby. This survey shall also be made before entering the building lun:h room.

The rules for radiation area conduct area

a. 1o smoking, eating, drinking, or chewing shall be permitted in Zones III and IV.

Zone II Plant areas in which smoking 1* not permitted will be so designatedJ

I 0

b. l!ffl'Y *~(ace and every piece of equipment in Zones III and IV and every tool or article taken into th*** Zone,

,hall be regarded 11 being contaminated until surveyed and released by a representative of H*alth and SafetyJ

c. All the provisions of applicable work procedures shall be read. understood and followed explicitly by the per1onnel performing the work;

d. Each employee is responsible for the care and treatment of equipment issued to hiffl and for his conduct in the performance of assigned work. Careless or willful mis-handling of equipnent or misconduct on the job will not be tolerated and will constitute gr~unds for dismissal.

9.18 for work of a routine nature in areas normally free of significant radiation and/ or contamination and where conditions are known and the work to be performed will not cause any significant change in these conditions, work is governed by Extended Work Procedures which may be modified or terminated at any time by Health and Safety personnel. Such Extended Work Procedures are given a date of termination not exceeding twelve months from the date of issue.

On, or in adv_ance of, the date of termination, the procedure is reviewed by responsible persons in Plant Engineering, Production, and Health and Safety, changed as necessary to reflect current working conditions, amd re-issued with a new termination date.

9.19 For work of a special or unusual nature or work in areas or on equipment which does involve significant radiation-contamination,.a Special Work Procedure* is issued. Eaeh Special Work Procedure is valid* for or.e shift only. Approval of responsible persons in Plant Engineering, Production and Health and Safety is required prior to the start of any work and befor* work can continue on succeeding shifts.

Job Planning and Scheduling 9.20 Each day responsible representatives of Plant Engineering, Production and Health and Safety meet to plan and schedule work for the following day. *A Work Schedule is prepared and distributed and Special Work Procedures are prepared and approved in ad~ance of the work.

The

• Work Schedule lists the personnel assigned to each task, the time and place to 1111et for each job, the estimated duration of each. job, the applicable

, proced~res governing the work and other information of general interest.

9.21 The Plant Engineer is reiponsible fora

a. Estimating the time _and manpower required to accomplish each maintenanc* jobl

b.

• Assigning maintenance personnel to **ch scheduled main-tenance Jo~

c. Assuring that all maintenance personnel read and under-stand applicabl3 work procedures and are thoroughly trained ln radiation-contamination workl

ct. A11urln9 that scheduled ulntenance persoMel und1r-1tand what work ls to be accoapllshed and *that the proper tool* and ~lpment, ln good condltlon, are avallablt ln advance of the job 1

• • A11urlng that assigned maintenance personnel are available at th-place and time indicated on the schedule.

9.22 The Production Manager ls responsible fora

a.

Establishi.19 priority of mainLe,,ance ir, the plant1

b. Determining what the 9ffect will be of scheduled ulntenan~e work on plant operations,

c. Arranging for equipment or area shutdown as necessary to accomplish the scheduled work1

d. Arranging for pre-maintenance decontamination and/or shielding 11 requiredJ

e. Assigning operating pel'sonnel to scheduled jobs as required'f'.

f. Assuring that all operating personnel read and under-stand applicable work procedures and are thoroughly

• trained in radiation-contamination work1

g. Assuring.that operating personnel understand what their duties will be for each scheduled job and that the necessary equipment, in good* condition, is available in advance of the jobs

h. Assuring that assigned operating personnel are available at the place and time indicated on the schedul~1

1. Issuing the work schedule following each planning and tchedullng meeting.

9.23 The Health and Safety Di~ector ls responsible for,

a. Determining what radiation-contamination conditions arwJ/or other special hazards will be encountered in performing the scheduled work1

b. !>etermlning whether or not a Special Work Procedure will be required for each scheduled job and if not, which Extended Work Procedure will apply1

c. Deteralnln9 requlr... nt1 for protective clothing ard/or other 11fety equipment for scheduled work and a11urlng that such equipaent, in good condition, 11 avallU>le ln advance of tht work1

d. Scheduling and leading a pre-job confer,nc:e lf requlrtd1

• • As1lgnlng Health Physics personnel to scheduled jobs as requlred1

f. A11uring that all Health Physlc1 personnel read and understand applicable work procedures, are thoroughly trained ln all phases of radlatlon-contamlnatlon work and are tr*lned and equipped to respond to unusual or

... rg,ncy condltlon11

g. A11urln9 that assigned Health Physics personnel are available at the place and time indicated on the achedul11

h. Inltlating Special Xork Procedures following each plannl~g and scheduling meeting.

Unco"\\dltlo~al Release 9.24 Rtltast surveys of equipment are the responslblllty of H11lth and Safety. Any item leaving Zone IV or Zone III to go to Zone II

~

or Zone I or any ltea leaving the plant site from any Zone, must be accompanied by a coapleted Unconditional Release. The original of the release accompanies tht equipment, and on, copy (in the case of an item leaving the plant site) 11 presented to the Pl~nt.Seeurlty Guard who is responsible for enforcing this procedure. Thia procedure also applies to cOC11Dercial vehicles and rallw*y cars. The Unconditional Release states the radiation-contamination level* on the lttms described, and releases them with no conditions or restrictions 11 to th*tr use.

eonc:11t12n11 B*\\****

9.25 Th* us* of. a Conditional Release la normally restricted to equipment which 11 not to leave Zone III. for exampli, a process pump which ls to be taken to ~e Equipment Decontamination Room or the Main-tenance Shop for repair will r~ulre a Conditional Release. The Conditional Relta1* describes the item releised, lists the radiation-contamination status of the ittm and lists any special precautions which must be taken for handling, dismantling, and ~*pairing the item.

Loct and Tag Procedure 9.26 The Lock and Tag Procedure is used to lock out valves, control*, and switch**, the unautho~lzed or inadvertent use of which could c1u11 process upset, damage to facilities and equipment or personal injury.

. Each department will have ita own locks and will be re1pon1lble for applying

lock* to equipment** required for.. ploy** protection even if this practice re1ult1 ln aeveral locks on the,... switch. The responsibility for removing locks will rest with the department head (or his delegated assistant) of the departaent *~1pon1ibl* for applying the lock. Non :ompliance with this provision will not be tolerated. Maintenance locks are normally applied only during aalntenance work on equipaent and are removed when the work ls completed.

The tags are used to i~dicate the reason for the lock and to warn all personnel of the possible consequences of violating this procedure.

~fety Hazard Tao Procedure 9.27 Any !*S empl:,yee ls responsible for tagging or posting any equipment.or condition which represents a safety hazard and/or unsafe working condition. After taking.such action he should notify his foreman or supervisor so that the condition may be corrected promptly. The Supervisor or Foreman shall notify the Director of Health and Safety.

Radiation and Contamination Protection 9.28 In this paragraph there a:e dltcussed a r.umber ~f adainistratlve limits of radiation exposure for t~o NFS Plant. It ls expected that these limits may be modified as pl,nt experi4nce dictates.

NFS tmployees may be exposed to radiation up to the limits stated in the following table with the approval of the employee's lnmediate supervisors Table 9.28 Rems Per calendar Quarter

a. Whole body1 head and trunk1 active bloc( fornlng organ11 ltns of 1y~11 or gon~ds------------------------------------------- 1-1/4

b. Hands and forearms feet and ankles-------------------------------18-3/4

c. Skln of whole body***--------------------~------------------------ 7-1/2 Whole body exposure to penetrating radiation in any 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period shall be llaited to 0.1 rem or, if approved in advance by the Health and Safety Director, 0.2 rem, Planned single exposures in excess of 0.2 rem must be approved in advance by the Plant Manager.

In emergencies involving the life of personnel, it shall be the resoonsl-billty of the NFS Senior representative present to determine and authorize, if such be his decislo~, e~try into higher fields of radiation.

9.29 Th* whol* body dose and skin dose ls available from badge raadlngs. TI°'e dose to extremities ls controlled in the field. If the dose rate to the hands and forearms or feet and ankles ls more than 1, times th* dose rate to the whole body, the time limit for the work ls bas i on th8 dose rate to the extremities. With p?'ior approval of the Plant Manager and the individual concerned, an employee of NFS may be

pendttld to recelv* 1 dose to the whole body greater than that permitted under paragraph 9.28 provided thats

a. Dur, g any calendar quarter the total whole body dose shall not exceed 3 rems, and

b. The dose to the whole body, when added to the accumulated occupational dose to the whole body, shall not exceed~ (N-18) rems where "N" equals the employee's age in years at his last birthday; and

c. The employees accumulated occupational dose. to the whole body has been determined using Form AEC-4, in accordance with the instructions in paragraph 20.102 of 10 CFR-20.

9.30 The consequence for intentionally causing erroneous film badge or dosimeter readings is dismissal.

9.31 ~S employees, who have been certified in the use of radiation monitoring instruments by the Health and Safety 01:ector, may in the course of their normal duties, self monitor in areas where the dose rate does not exceed 100 mr/hr, except Zone IV areas. In areas in which the dose rate exceeds 100 mr/hr or in Mll Zone IV areas monitoring for any entry shall be by Health and Safety Technicians. In no case shall employ.. s enter an area in which the dose rate exceeds 2 r/hr unless prior approval of the *?lant Manager has been obtained.

(See 9.28)

Maximum Permissible Levels of Radioactivity 9.32 *The maximum ~llowable surface contaminations for the West Valley Plant are shown in Tacle 9.32a. The Maximum Permissible Concentrations in alr of some radionuclldes expected to be encountered in the West Valley Plant are shown in Table 9.32b.

The Maximum Permissible Concentration in on-site. nonpotable water in Buttermilk Cre11k of some radionuclldes expected to be encountered at the West Valley Plant are shown in Tabl* 9.32c.

Air*sampling 9.33 The air sampling program provides for the,valuation of alpha and beta-gamma air contamination in all building areas, the plant site and the site perimeter. Included in the program are ~4 in-plant area particulate samplers, 19 remote in-cell particulate samplers, 7 in-plant ccntinuous air monitors, 1 plant site sampler and 3 site perimeter air monitors. This equipment ls described in Appendix 9.33, is located as per Figure 6.67, and discussed in Paragraphs 6.66 to 6.67.

9.34 ll'ie -filter paper used for particulate sampling is Whatman #41 or equal, two inches in diameter. Whatman #41 filter paper has a collection efficier,ey of 98 per cent fvr 0.18 micron particulate or larger at a flow velocity of ~ centimet~rs per second. To obtain thla *tow velocity a minimum flow rate of~ liters per alnute ls used for in-plant air sampl,rs. Self absorption in Whitman #41 paper ls zero for ~t* and. about 0.3 for alpha.

Table 9.32a Maxiaann Allowable Surface Contamination for West Valley Plant Smearable Non-Smearable Surface Alfha d!(m}Z:100 cm2}

Beta-Gamma As ShownZ100 cm2 Altha dL{mlZ:100 cm2}

Beta-Ganrna As ShownZ100 cm2 Skin No Detectable 500 100 c/m Personal Clothing No Detectable 500 100 c/m Plant Clothing 500 100 c/m 1)000 2,000 c/m Plant Vehicles 500 100 c/m

.

• 1,000 5,000 c/m Coanercial Vehicles 500 100 c/m 500 0.4 mrad/hr *

• Zone I Zone I limits are per 10 CFR - 20, Appendix B, Table.:1 Zone II 500 100 c/m 500 100 c/m Zone III 5,000 10 mrad/hr 5,000 100 mrad/hr Zone IV*

50,00C 2 r/hr*.

50,000

~ r/hr

• For personnel entry Conditional Release 1,000 5,000 c/m 5,000 10 mrad/hr Unconditional Release 500 100 c/m 500 100 c/m

Table 9.32b Maximum Permissible Concentration Cuc/ml)

Mixed fission Products No respirato:-y protection full face filter mask Supp} 4.fd air mask Strontium-90 No respiratory protection full f~ce filter mask Supplied air mask Cesium-137 No respiratory prote~tion Full face filter mask Supplied air mask Plutonium 239 No respiratory protection full face filter mask Supplied air mask Natu+al Uranium No respiratory protection full face filter mask Supplied air mask High Enri~hed Uranium No. respiratory. protection full face filter mask Supplied-air mask Iodine-131 N~ respiratory protect i on Supplied *air mask Krypton-85 No respiratory protection Supplied air mask Footnotes to Table 9.32b l X 10-9 2 X 10-8

• Above 2 x 10-8 3 x-10-lO 6 X,10-9 Abovo 6 x 10-9 1 X 10-8 2 X l0-7 Above 2 x 10-1 2 X l0-12 4 X 10-ll Above 4 x 10-11 6 X 10-ll 1 X 10-9 Above 1 x 10-9 l X 10-10 2 X *10-9.

Above 2 x 10-9 g *x io-9 10-9 Above 9 x 1 X l0-5 Above 1 x 10-5 I

Maximum Permissible Concentrations for other radio-nuclides is as indicated in 10 CFR-20, Appendix B, Table I. When a mixture of radionuclides is encountered and the identity and concentration of each radionucli~e in the mixture are known, the Maximum Concentration is derived as followsa If radionuclides A, B, V, are present in concentrations Ca, Cb, Cc and the applicable MPC~ s. are MPCa, MPCb, and MPCc re*spectively, than the concentrations shall be limited so that the following relationship exists a Ca + Cb i

Cc MPCa MPCb MPCc.$ l

Table 9.32c Maximum Permissible Concentration (µg/ml}

On-Site-Buttermilk Creek Off**Site

~

Cesium-137 4 X 10*4 2 X 10*5 Cobalt-60 1 X 10*3 3 X 10*5 Tritium 1 x.10*1 3 X 10*3 Iodine-131 6 X 10*5 2 X 10*6

. Plutonium-239 1 X 10*4 5 X 10-6 Ruthenium-103 2 X 10*3 8 X 10*5 Ruthenlum-106 3 X 10*4 1 X 10*5 Strontium-90 4 X 10*6 l X 10*7 Natural Uranium 5 X 10*4 2 X l0-5 High Enriched Uranium 8 X.10*4 3 X 10*5 Footnotes to Table 9.32c Maximum Permissible Concentrations for other radionuclides are as stated in 10 CFR-20,Appendix.B. When mixtures of radionuclides are encountered and the identity and concentrations of each is known, the procedure stated in the footnote to Table 9.32b is used to determine the MPC

• I 9.~ Air samples are collected and analyr.ed for.r-adi-oactive material according to the schedule shgwri in Table 9.35. This ochedule is subject to revisi~n as experience is gained in operating the plant.

Continuous a}r monitora are used in some occupied areas to provide an iamediate alal'f'l should high air contamination exist. The other remote samplers will be used occas lonally to obtain very short, ~pct samph's of air contamination co~ditions. in the c~lls. These remote samplers are:

Miniature Cell General Purpose Cell Chemical Process Cell Mechanical Cell X-Cell 2 X-Cell 3 Product Purification and Concentration Chemical Process Cell Crane Decontamination Area Process Mechanical Cell Cr~ne Decontamination Area 9.36 As each sample is removed from the sample head it is placed in an envelope which is marked with the sampler location, date-time started, date-time changed and the flow rate. When all samples have been changed, accoxding to the schedule, they are brought into the Health Physics Lab, removed from the envelopes, placed in planchets and -surveyed with portable beta-gamma and alpha detection instrllfflents (Appendix 9.36}. Any samples which show unusually high activity are segregated for special handling and prompt attention in the counting room.

9.37 Alpha, Beta proportional counters (Appendix 9.37} are used to analyze in-plant air "samples. All samples.receive a one minute alpha, beta count as soon as possible after being delivered to the counting room.

The beta/alpha ratio is determined based on this count. Since the beta/alpha ratio is constant for natural activity, it may be possible at this time to make a preliminary estimate of the amount of long-lived emitters on the sample. The concentration of beta emitters on the sample will be determined based on the initial count. This is accomplished as shown in Appendix 9.37a. All-samples receive a five minute alpha count five to seven hours after sampling and a second five minute. alpha-beta count 23 to 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> after sampling. These counts are used to calculate the alpha counts due to long-lived alpha activity (product} on the sample. This is accomplished as shown in Appendix 9.37b. Any samples which, on the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> c,unt, show le~s than 1 c/m alpha and less than 1800 c/m beta are

1

'r:

Shiftwise Hot Lobby Mechanical Operating Aisle-west Ram Equipment Room Chemical Viewing Ai:le-north Ventilation Wash Room Process Sample Enclosure-1 Process Sample Enclosure-2 Analytical Aisle Extraction Sample Aisle-west Extraction Sample Aisle-east Ventilation Exhaust Cell Pulser Aisle

~ :---..;.~- ~-=--

Table 9.35.

Start Up Schedule for Air Sampling Daily

. GFC Operating ~isle**west GFC Operating*Aisl~-east Lower Warm Aisle-west Lower Warm Aisle-east Acid Recovery Pump Aisle Scrap Removal Mechanical Operating Aisle*east X-Cell entrance air lock U-Product cell Product purification cell Product Packaging l, 2 and 3 Fuel Storage 1 and 2 Chemical Viewing Aisle-south Equipment Decontamination Room Chemical Operating ~isle-north Chemical Operating Aisle-south Lower Extraction Aisle-ea.st Upper Warm Aisle-west Upper Warm Aisle-east Off Gas Cell-2 OGC-ARC Aisle Chem Lab-east and west Product Lab Emission Spec. Lab Mass Spec. *tab Alpha Lab Stack Sampler Upper Extraction Aisle-west Upper Extraction Aisle-east Extraction Chemical Room-east Laundry Weekly n,trd Floor Offic@

Se~ond Floor Office Main Lobby Maintenance Shop Utility Room Manipulator Repair Area Product Packaging CAM Fuel Storage CAM EDR Viewing Station t.cR Air Lock Ventilation Supply Room Ho.t Lobby CAM Mechanical Operating Aisle CAM VSR Access Aisle Off Gas Cell-3 Analytical Cell Decon. ~.rea Alpha Lab CAM Lab Access lisle

. Control Room Extraction Chemical Room CAY Plant Area Perimeter-1 Perimeter-2 Perimeter-3 GPC Crane Room Mechanical Crane Room Chemical Crane Room

I 1*

I discarded. These counts at maximum counting error, represent a~ut 1%

of MAC for plutonium-239 and strontium-90 respectively. Samples which exce,d either or both of the counting limits will be held for a final count. The ftnal 30-minute count on in-pl~nt air samples is tak.en a minimum of four days after sampling to allow the natural activity to decay essentially to zero. All of the alpha counts are assumed to be counts due to product and the c~ncentrations are calculated as follows1 Alpha

~c/ml = c/m (1.31 x 10-12)

M3 Since the counting err~r for a 30-minute count at 95 percent confidence level is+/- 10% at 10 c/m, the minimum detectable alpha concentration on a 24-hour sample is1 10 (1.31 x 10-12> = 1.5 x 10-13 with~ 10% accuracy 86.4 Beta-Gamma

~c/ml = c/m (9.19 x 10-13)

M3 9.38 Some in-plant air monitors are ~oving-filter type and the filter tape is not normally analyzed in the counting room. Portions of the tape may be counted and/or gamma scanned if this information is needed.

9.39 The perimeter samples are changed weekly and are analyzed once as soon as practical after sampling and again four days after sampling. The samples are analyzed by counting for one hour in a low background aipha, beta proportional system (see ~ppendix 9.39a). The geometry of this system is 50% for beta and 35% for alpha. The background is about l c/m. The concentration of beta emitters is determined as shown in Appendix 9.39b and the concentration of alpha emitters is determined as shown in Appendix 9.39c. A log is kept of air sample results. These results become part of the permanent record of radiation-contamination conditions in and around the plant.

9.40 Radioiodine activated charcoal filters from the stack and p~rimeter ~tations are analyzed as follows:

The filters are gamma.

scanned to dete~ine if there are other gamma emitting isotopes present and in what proportion. Since the radioiodine filter is preceded by a particulate filter, there will normally be no interference from other isotopes. The radioiodine filter is then counted in one of the propor-tional counting systems and the concentration is calculated as shown in Appendix 9.40

• 9.41 The numbers used in this section for geometry of counters and efficiency of and self absorption in filter paper are numbers furnish~d by the manufacturers. The method used to determine actual counter *geometry is described in the calibration section {Appendix 9.37). The collect!on efficiency of n~tman #41 filter paper can be checked by using a membrane type filter behind the Whatman #41 filter to test the penetration under various conditions of use in the plant. The self absorption of alpha in

e::~tf~! ~~~ !~1~:!~~~~e: :ra~~~~!i~~dac~~~:r~; :~:s~~:~~~ei~e i~!ter, absorption correction then becomes; filter count/planchet count. These tests will be run on each batch of filter paper received.

Radiation - Contamination Survey Program 9.42 Beta-gamma film badges are supplied to each employee and all visitors to the plant through an arrangement with a conunercial film badge processor. Badges are exchanged and read weekly for most personnel; monthly for administrative personnel. This schedule is subject to change as operatin~ experience is gained.

Immediate notification by phone or

_wire ls given for badges which show a dose in excess of 100 mrem.

Neutron monitoring is accomplished on an area basis. Neutron badges are placed

.in the product storage and product packaging and handling areas to establish and check the neutron dose rate in these areas. The neutron badges are changed monthly* during startup but this may be changed to quarterly at a later date.*

9.43 Each prociuctio.1 employee and each visitor is -issued a 0-200 mr g~a dosimeter which is read and the dose recorded during the shift following the shift on which it is.used. The dose is recorded on th'l "Dosimeter Re~dings" form, and is transferred later to the "Exposure Record" card which is also used to record padge readings. The 5 x.a* inch card, designed to be used in a "Victor Visible" type file, contains all of the information required by AEC Form 5. Each card represents 13 weeks exposure data. See Appendix 9.43 for the "Exposure Record" card refer~ed to above.

9.44 A limited number of self reading dosimeters are available for use during "hot" area decontamination and maintenance work*.

These dosimeters will be used as the second line of defense *against overexposure. The primary control will be monitoring, by Health and Safety or by the individual performing the work, and timekeeping, by the individual or by a timekeeper assigned t o the job.

• 9.45 Health and Safety responsibility for product shipments entails checking the shipping papers for Production signature approval, for product specifications, accountability certification and surveying the shipping containers t~ insure conformance with all applicable federal, state, and local regulations. The signature of an authorized Health and Safety representative on the ship.ping i,apers will constitute approval to.

ship

• 9.46 With-the exception of Zone IV egress, personnel surveys a?e the responsibility of each employee. Health and.Safety will aud.it the frequency and adequacy of such surveys. Personnel found in a Zone II or Zone I area with contaminated clothing may be subject to dismissal.

9.47 A regular schedule of routine surveys will be performed by Health and Safety. The routine survey program is designed to supplement the reports of radiation contamination conditions which are encountered during maintenance and other work, and to insure that all plant areas are surveyed on a regular basis. Each routine survey is described in consider-able detail on the "routine survey" form \\See Appendix 9.47) which will serve as a guide for the Heal th and Safety person11el performing the survey.

A list of routine surveys is shown in. Table 9.47. A written record is made of every survP.y performed by Health and Safety personnel. This record which is executed on pre-numbered survey log sheets becomes part of the permanent record of radiation-contamination conditions in and around the plant.

Environmental Survey Program 9.48 The environmental survey program, pre-operational and post-operational is divided into three categories:

1. Atmospheric monitoring including air particulate monitoring;

2. Water monitoring including surface and -ground water sampling;

3. Earth and biota monitoring including samples of silt, mud, -plankton, fish an.ci shellfish from Buttermilk Creek and Cattaraugus Creek; soil, vegetation and milk samples from the site and surrounding area *and small game from the site.

~

\\,,._ V

\\ *. *,. 7

"~

..... #. 4(.

9.49 The pre-operational progre~ is divided into two phases; the first phase, started in the spring of 1963, to establish on site gross activity background with a few analyses for specific isotopes and the second phase, starting in the fall.of 1964, to include more analyses for specific isotopes. Phase II will continue into the post-operational period: Both Phases are detailed in Appendix 9.49. A summary of the Environmental Monitoring Program is presented in Tables 9.49a and 9.49b.

-Waste Disposal Control Program--Gaseous Waste 9.50 Gaseous waste control is accomplished by treatment of waste gases be(ore release, continuous monitoring at the point of release and environmental monitoring to determine the effect, if any, of released activity in the environment. Waste gas treatment is discussed in some detail in Section V~, Paragraphs 6.66 to 6.70. Prefilters, air sc: ubbers, silver reactors and high efficiency filters are used to minimize the amount of radioactive gases and particulates released routinely from the plant. It is anticipated that the routine releases will be well below the maximum ~llowable under applicable

Survey No.

S-1 S-2 5-3 S-4 0-1 D-2 0-3 0-4 D-5 0-6 0-7 0-8 0-9 0-10 0-11 W-1 W-2 W-3 W-4 W-5 W-6 Table 9.47 Routine ~urveys Title Check Dosimeters and Record Results Pick up Air S~mples Check Charts on Gamma Alarm, Sample

.System and Weather Monitoring SY.stem Count Samples.

Check Station Monitors and Hand Countets Calibrate Instruments Spot Chack Laboratories Surve~ Hot Lobby

• Transfer Dosimeter Readings to Exposure Record Cards Spot Check Sample Aisle, Pulser Aisle and Warm ~quipment Aisle Survey Lunch Room Survey Step-off Pads Prepare Control Samples for Counting Prepare Environmental Samples for Counting Spot Check Product, Packaging and Handling Survey Alpha L_ab Survey Chem Labs Survey Spec Labs Su~ey Product Lab Survey Zone III offices S urvey Mens Locker Room Shift Assigned 1,2,3 1,2,3 1,2,3 1,2,3 3

3 2

3 1 *

. 3 2

3 1

1 2

1 1

1 1

3 2

Table 9.47 con't Shift Su.rvay No.

Title Assigned W-7 Survey Five Personnel 1,2,3 W-8.

Survey Ventilation Penthouse 3

W-9 Survey Upper Warm Equipment Aisle 3

w-10 Survey Access Aisle 2

W-11 Survey C~erating Aisles 2

W-12 Survey Sample* Aisle 3

\\'t-13 Survey Fuel Receiving & Storage 3

W-14 Survey Product Packaging & Handling 2

W-15 Survey Decontamination Area 3

W-16 Survey Scrap Transfer Area 3

0 W-17 Survey Health Physics Lab 1

W-18 Survey Mechanical Cell Viewing Area 2

W-19 Survey Laundry 3

W-20 Obtain Environmental Samples 1

W-21 Survey Womens Locker Room l

W-22 Survey Warm Equipment Aisle 3

W-23 Survey Mobile Equipment 1

W-24 Survey.Ventilation Equipment Rooms 2

M-1 Survey Analytical Viewing Ar~a 2

M-2 Survey Instrumept Shop 2

M-3 Survey Main Lobby 1

M-4 Survey Cold Chemical Penthouse 2

M-5 Survey Chemical Process Cell Viewing Area 2

• o Survex Ho.

Jltl*

M*6 M-7 M-8

• M-9 M-10 M-11 M-12 M-13 Q-1.

Q-2 Q-3 Q-4 Q-~

Q-6 Survey Maintenance Shop Survey Guard House Survey Tank Farm Survey Burial Ground Survey Remote Operating Station Survey Fi~st Aid Obtain Environmental Samples Autoradiogr..1ph Env!ronmental Air Samples Survey Ut111 ty Building Survey Roads, Walks, Farking lot and R.R. Spur Survey Storage Lagoon and Hardstand Areas Survey Ory Wells Survey Zone II offices Obtain Environmental Samples S = Shiftwise 0 = Daily W = Weekly M = Monthly Q = Quarter 1 y Shift Assigned 2

1 1

1 2

1 1

3 2

1 1

3 1

Air Sampling 3 Perimeter 1 Plant Site Rain & Snow

!Plant Site

" I I'

Su3:face Water 1 Erdman Brook 1 Buttermilk Creek 1 Cattaraugus Creek Mud ~Si_lt 1 Erdman Brook 1 Buttermilk,~~ek 1 Cattaraugus C1*eek f

Well Water 1 Plant Site Vegetation 3 Perimeter Milk

~eighboring Farm Small Game 1 Plant Site Table 9.49a Environmental Monitoring Phase I - Type of Analysis Weekly Monthly Gross Alpha Gross Beta-Gamma Gross Alpha Gross Beta-Gallllla, Trltium Gross Alpha Gross Beta:..

Gama, 1*r1t1um Gross Alpha Gross Beta-Gama Gross Alpha Gross Beta-Galmla, Tritium Gross Beta Gamma I-131 Sr-90 Semi-Annually Gross Alpha Gross Beta-Garrma I-*131 Sr-90 Gross Alpha Gross Beta-Gama I-131 Sr-90

I

!'d>le ?-49b Environmental Monitoring Phase II - Type of Analysis Weekly Monthly Semi-Ann..Y,1Ally Air Sampling Gross alpha 3 PeriHter Gross Beta-Ganma Scan 1 Plant Site Gama Rain and Snow Gross Alpha l Pl~nt Site Gross Beta-Sr-90 Ganma Tritium Surface Water Gross Alpha -

l Erdman Brook Gross Beta-1 Buttermilk Creek Ganma, Tritium l Cattaraugus Creek Mud and Silt Gross Alpha l Erdman Brook Gross Beta-Sr-90 l Buttermilk Creek Ganma 1 Cattaraugus Creek Well Water Gross Alpha 1 Plant Site Gross Beta-Garrma, Tritium Vegetation Gross Alpha 3 Perimeter Gross Be'ta-Sr-90 Gamma I-131 Milk Gross Alpha LJ1iant Site Gross Beta-Sr-90 Ganma I-131 Fish and Shellfish 1 Cattaraugus Creek Gross Alpha Gross Beta-Gaaaa I-131 Small Game Gross Alpha l Plant Site Gross Beta-Ganma I-131 Sr-90

.. \\

federal*ind state regulations. Spare units and aut01r,atic controls are used

~* necessary to prevent the escape of high level bursts of activity caused by major equipment failure.

9.51 A continuous stack gas m~~tor, described in Appendix 9.51, is used to d~~ect concentrations of 3 x 10-

~c/ml or less of gross beta-gaama particulate activity and about the same~concentr~tion of I-131. A significant increase in concentration of either particulates or radioiodine will cause an alarm in the plant control room. The exact alarm positions will be field selected based on operating experience1 they will be kept at the lowest practical level to provide the earliest possible warning* of off-standard condttions.

9.52 Environmental monitoring to determine the effects on the environment of waste gas disposal is c.oncentl'ated in air sampling and sampling of soil, vegetation, milk and rainout. Three site perimeter continuous air monitoring stations are established to determine concentrations of ~adioactive particulates and radioiodine at these stations. One station is located 3,lrO meters south-east of the plant, the second station is located 2,100 meters north-east of the plant and the third station is located 4,000 meters north-north-west of the plant. This places a monitoring station at either. end of and adjacent to Buttermilk Valley and, accord!ng to prevailing wind.patterns, will place one of the three monitors down wind of the stack n,arly 60 per cent of the time.

9.53 Th6 routi~e. soil, vegetation, milk and rainout sampling program is defined in Table 9.49 b. The entire sampling program is subject to change as operating experience is gained but it is expected that any changes will be minor in nature. Special samples wilJ b~ analyzed lf the stack monito*t indicates an alarm cond~tion. The weather monitoring itations, (see* Appendix 9;53} will supply data which may be used to determine the direction of travel of stack fumes and the distance at which the maximum ground level concentration occurs. A mobile motor-generator sampler set will allow sampling down wi nd of the stack regardless of wind conditions.

Waste Disposal Control Program -- Liquid Waste 9.54 The primary control of high level liquid was.te is in the facilities provided. The waste tank itself, the concrete saucer for secondary containment, *the impervious "silty till" formation and the spare tank all contribute to a high degree of confidence in the system. See Paragraphs 5.~o - 5.56, 7.10, 7.14 -7.18, 7.25 - 7.37. Facilities are provided for monitoring or sampling in the annular space between the tank and the vault. Routine surveys will be performed in the wells located adjacent to the waste tanks. A continuous water sampler located ngar the confluence or Erdman Brook and ~uarry Creek will serve as a third monitoring point of control of liquid waste.

9.55 Low level liquid waste will be discharged to Cattaraugus Creek via Erdman Brook and Buttermilk Creek. Waste water at a volume of about 40,000 gallons per operating day is received in the interceptor, batch neutralized if necessary, and discharged to a series of holding

~

oonds. The interceptor volume is about 50,000 gallons and the ponds

providt holdup for 4,000,000 gallons or 100 operating days above the minimum overflow ~int,. Overflow points between ponds are a valved line at two feet aboveth* bottom to ~ovide for solids collection, and an open overflow at one foot from the top.* The discharge line to the creek is valved so the amount of waste discharged may be regulated.

9.~

Stream gauging and sampling stations are provided near the confluence of Quarry Creek and Erdman Brook ~nd on Cattaraugus Creek. Gauging is performed in order to determine the rate at which waste solutions may be metered into Erdman Brook. Samples from these Gtations will be collected and.!!'alyzed_weekly~ Analyses will include gross alpha, beta and gamma, tritium and specific isotope analyses as required for control. (Appendix 9.56)

Waste Disposal Control Program--High Level Solid Waste 9.57 A burial area for waste generated in the plant will be mai~tained in an area north of the plant between the waste tank farm and the confluence of ~uarry Creek and Erdman Brook. This area will be reserved for process scrap and discarded process equipment. Process scrap, fuel eiement end pieces and leached hulls, will be packaged in 30 gallon drums, loaded into a shielding cask on a carry-all type trailer and transported to the burial area.

(See Paragraph 7.14.) At the burial area a truck mounted crane with remote controls, 100 feet away, will b~

used to lift *the lid of the cask, remove the scrap drum and place it in the trench. At the end of each burial operation, which may require several trips, the crane clam attachment or front end loader, will be used to backfill where necessary to maintain an exposure rate at the security fence of 2 mrem/hr.

The drums will be covered with sufficient dirt to r*duce*the exposure rate.at ~he edge of the trench to 200 mr/hr. Final backfilling when the trench, or a portion of the trench, is full will be to a radiation level of 1 mr/hr or less. The minimum dirt covering will be four feet thick.

9.~8 A similar procedure will be followed for burial of process equipment.

The *equipment, after decontamination, will be suitably packaged and loaded on the truck in the Equipment Decontamination Room, transported to the burial area, loaded into the trench with the crane and bac~filled.

Packaging techniques will vary depending on the equipment itself and the radlailon-contaminatlon co,,di tions. Generally a sprayed-on coating or a covering of plastic film wiU be used..

Medical Program 9.~9 The medical program, under the direction of the Medical Director, will consist of a very thorough pre-employment medical history and physical examination for each prospective employee.

The medical history will be aimed at not only past illnesses and injuries but particular

\\ ~, ~ attention wlll be paid to history of past radiation~exposure, allergies,

~e *~crasias, tumors and any evidence of emotional instability. The

~1:'

1,r~

laboratory studies on all applicants will consist of a minimum of complete 1-t"*

blood count, serology, urinalysis, chest x-ray end vital capacity deter-minations. Each employee will hav9 a complete physical examination yearly

• 4£4 S

04 t:::::te J P A.

A complete blood count will be done twi~e yearlys clinical urinalysis monthly.

The pre-employment physical* examination and laboratory studies will be repeated on each individual leaving the fi4ploy of th~ company.

9.60 B10-1111y1 will be scheduled for employees using an "across-the-plant statistical survey" plan. The number of times each employee is sampled ear~ year and the type analyses performed will depend on his Nork location. Office employees annually, for totai alpha and gross fission products; mechanical head end, extraction operators, Health & Safety techni-cians, mai,.tenance and utility operatori semi-annually for total *alph*a and gToss fission products; product purification and packaging operators quarterly for plutonium and total uranium. Additional samples will be obtained to confirlil any positive result and special samples will be obtained when inhala-tion or ingestion is suspected for any employee.

9.61 Thyroid monitoring of employees will be performed at le*ast once eac~ year in conjuction with the annual physical examination. Special monitoring will be performed as indicated b~ air sample counting results.

9.62 A dispensary will be maintained for care of ordinary minor on-the-job injuries. There will be facilities for intensive first-aid care of severe injuries such as burns, fractures and gross contamination with radioactive materials. Irmnunizati~n against tetanus will be routine for all employees.

9.63 Close liaison with the Health and Safety Department will be maintained. The Medical Director will assist in health and safety traihing and indoctrination. He will review with the Health and Safety Director, all industrial radiation exposure records; air, water and plant

. radiation survey records. He will cooperate with the Health and Safety Director. in plant inspections.

Radiation exposure data for each employee sh~ll be kept on form AEC-5 as

~rt of the permanent record of each employ~e.

A permanent chec~off list shall be attached to each employee's permanent record covering all of the plants* requir~ments regardi;;g physical examinations and pe~ onal radiation exposur~ recording and control as well as all requirements uf 10CFR-20.

Emergency Procedure

• Fire Protection Organization 9.64 The Health and Safety department has ~he primary respon*sibili ty for training personnel and auditing procedures and activities for fire prevention as well as for fire fighting. The fire fighting function will be carried out through shift fire brigades organized as indicated in Appendix 9.64.

Organization for Radiation Emergencies 9.6~ There are a very large number of combinations of conditions which might constitute or cause an emergency. It is, therefore, not possible to prescribe inflexible procedures* for emergency action. However, there are m

kEI egg [ C'

~ad categories of emergencies for which general procedures may be stated and certain general rules which apply in nearly ~11 cises. In any radiation

... rgency, the Health and Safety Department has.the primary responsibility to define the magnitude and extent of the problem and to recommen~ a course of action wh' ch will restore the affected areas promptly and safely.

9.66 In*any radiation emergency the responstble group (Production or Analytical} in the area in which the emerge~cl condition exists must take iaaediate steps to accomplish the followings

a. Protect plant,.>ersonnel by evacuating affocted areas and take action to confine the condition and eliminate or moderate the cause.

b. Notify the Health and Safety Director (or Technician on off shift) giving all possible details about the nature and location of the emergency.

c. If the emergency involves property damage, personal injury, significant radiation levels, production interruption, or possible off-site contamination, the following must be notifieds He~lth and Safety Director Medical Director Laboratory or Production Manager Plant Manager & Assistant Plant Manager Assistant to the Plant Manager Security Officer Plant Engineer

d. following the survey by Health and Safety~ barricade and post the affected area to prevent inadvertent*entry.

e. Devi** **a plan for restoring the area and assemble the required men and materials.

9.67 Generally, the following rules apply in handling an emergency conditions

a. If incident involves wreckage and a person is believed to be alive and trapped, make every possible effort to rescue him. The usual radiation -rule$ may be abrogated upon the authority of the senior person present.

b. Segregate and detain for further examination those persons who have had possible contact with the radioactive material. Parform complete contamination surveys of such personnel and institute decontamination at once if significant exposure could result from a delay. ~ormally, it is best to leave skin decontamination to those persons with specific training in this function.

o.

Remove injured persons from the scene with as little direct personal contact as *possible. Limit first aid and medical procedures to

.1

tho1e that must be done promptly until the doctor is present.

d. Do only what is necessary to preserve life and property prior to the arrival of Health and Safety specialists.

e. Wor within the framework of any applicable SOP*s covering a specific type of emergency.

Plant Maintenance Program 9.68 The Nuclear Fuel Services maintenance program has been planned to insure continued safe operation of the plan~ co111nensurate with Paragraphs 9.13 to 9.41, with a minimum of downtime consistent with economic considerations.

9.69 The routine inspection and maintenance program is similar to that for a noi,nal chemical plant, except where modified to reflect more stringent requirements for the nuclear aspects of the plant. The maintenance program is based upon utilizing conventional methods and _procedures for performing contact maintenance work.

Special controls are incorporated to cover work within contamination and radiation zones. Work on contaminated equipment or syste~s is done under the surveillance of the Health & Safety

-Department which recoamends required control measures. C~reful planning, prewritten job procedures, and close coordination with Production and Technical Services Departments assure safe and efficient plant operation.

Normal inspection contemplates periodic shutdowns to permit inspection and maintenance of those portions of the plant not readily accessible during routine operation

• 9.70 Certain equipment is deemed vital to the safe and continuous operation of the plant. This equipment is defined as 1. equip-ment that could become critically unsafe from a nuclear* standpoint, and

2. any malfunctioning piece of equipment which could reasonably require the shutdown of the plant.

A list, referred to as the Vital Equipment List initiated bi the Production and Technical Services Departments and approved by the Plant Manager, is compiled and issued to the Production Department.

(See 9.82) The list states the requirements to be met before the equipment is thken out of service, and what tests and requirements are to be met bef9re the equip-ment is returned to service. All equipment not specifica1ly de1signated on the Vital Equipment* List is considered as non critical and may be taken out of service, repaired, and returned to service according to normal standard maintenance practice.

Q!' ganization 9.71 Maintenance work on Nuclear Fuel Services equipment and systems ls performed by ?lant Engineering. Plant Engineering is responsible for all mechanical, instrument and electrical maintenance work.

Each of these categories is und,r the direction of a group leader. Close cooperation between these groups is maintained to facilitate scheduling, conserve man-pow, and minimize downtime.

Under normal conditions, mechanical and electrical maintenance is accomplished on a day schedule, five days per weik~

Much of the routine instrument main-tenance ls carried out on a similar schedule; however, instrument. technicians are normally on shift with operations personnel.

Plant Engineering Section Personnel 9.72. Plant Engineering is composed of a plant engineer, mechanical engineers, maintenance mechanics, instrument technicians, and stenographer.

The Plant Engineer is responsible for,

1. Planning, scheduling, and controlling personnel, materials, equipment and tools.

2. Initiating training and educational programs for maintenance personnel.

3. Establishing and supervising the maintenance of a readily accessible file of design.and vendor information, parts dat2,.preventive main-tenance records, and historical records.

4. Supervision of all maintenance assignments, including instructions to cover *safe working practices, radiation protection measures and as>proved maintenance repair procedures *

5. Making technical studies on maintenance of mechanical, instrument and electrical equipment, and making reconimendations on design changes.

6. Preparing labor and mate~ial costs estimates for non routine work.

The Plant Engineer is primarily assisted by two mechanical engineers to whom any of the above responsibilities may be delegated. Technical

~upport is available from the Technical Services and the Health and Safety Departments which will provide specialists as required.

Facilities 9.73 The.Plant Engineering Section and shop facilities are organized primarily to perform field maintenance work.

On site shop work consists basically of minor repairs, replacement of defective components and checkout of equipment. The bulk of the work is of *short duration and minor complexity, and the shops are equipped accordingly. Machine, electric, instr\\.!fflent, pipe, carpentry and welding shops are provided.

• In cases where maintenance functions require facilities not provided at the site, privately operated*

shops in nearby Buffalo, New York will b~ utilized where possible

• o*

Instrwpent Maintenance Personnel 9.74 The maintenance of instrumentation and control sy~tems is the responsibility of the Plan~ Engineer assisted by the Instrument Engineer.

These responsibilities are as follows,

1. Adequacy of the maintenance facilities and the training of personnel to meet all requirements, both routine and emergency;

2. Planning and scheduling of all instrument maintenance in cooperation with mechanical maintenance personnel;

3. Establishment of a preventive maintenance program for all control systems and components, with particular emphasis on those involving the safety of the.plant.

4. Planning and maintenance of a file system th~t contains the infor-mation necessary to analyze, design, order spare parts and components, apply preventive maintenance procedures and provide history of repairs on all equipment. This will be done in conjunction with mechanical maintena~ce.

Instrument Shop Facilities 9.75 The instrument shop is equipped with services, (water, air, electricity, tools and test equipment) necessary for the calibration and maintenance of either pneumatic or electronic instruments.

Maintenance Categories 9.76 Plant Engineering performs three categories of work; preventive maintenance and inspection, routine maintenance and non routine maintenance *. Any of these categories of work niay involve hazardous conditions due to radiation or contamination. The procedures used in performing this work depend on both the category of work and the degree of hazard inv~lved due to direct radiation or contamination. These procedures will be subject to approval by the Health and Safety Director in those cases involving radiation hazards.

Preventive Maintenance and Inspection 9.77* The preventive maintenance program minimizes shutdowns and breakdowns by systematically inspecting equipment, making calibrations or adjustment, and scheduling repairs and overhauls before failure occurs.

Each piece of equipment ls studied thoroughly, and a schedule of routine inspections is determined and established under the following c~assificationss

a. A-Class, Major inspection (complete check of equipment,)

b. B*Classa A "middle-of-the-road" inspection. Usually made quarterly to semi-annually and, on occasions, monthly~

c. C-Classa. A minor inspection (ordinarily visual and frequent.) Usually made monthly to quarterly and, on occasions, weekly.

r

As each piece of equipment ls studied, a complete list of items to be checked on each inspection is made.

A contral control system indicates when inspec-tions are due. If inspections do not interfere with normal plant operation, th~ inspections are scheduled and carried out ~n accordance with work loads in the section. Inspections that require shutdown of equipm~nt or interfere with normal plant operations are coordinated with the Production Department.

After an inspection is completed, information is transferred from the inspection sheet to a card as a continuing record. If any repairs are necessary, such repairs fall in1'o the category of routine maintenance and

~e scheduled ~ccording to the urgency required.

o Routine Maintenance 0

9.78 Routine maintenance includes all maintenance work on equipnent or systems which is' directed toward restoring the equipment or system to its normal functioning capability, without altering its basic design function. Routine maintenance is conducted during normal plant operation, as well as during scheduled shutdowns.

Normal routine maintenance work is either requested by the Production Depart-ment or results from the preventive maintenance program. Because there is generally a backlog of work, all work is given a level of priority to facilitate effective scheduling. Priority is based on safeguards considerations, production loss resulting from the equipment being shut.down, or the probability of a breakdown if a repair is not made, with consequent damage to equipment.

!,on Routine Maintenance 9.79 Non routine maintenance includes modifications or additions to systems or processes ~s differentiated from repair or replacement of faulty equipment. Depending ~pon the nature and extent of tne work, main-tenance or construction forces are used. In the latter case, Plant Engineering is responsible for maintaining close contact with the.work to see that it is performed in accordance with.specifications, within the cost estimate, and reporting on the progress of the job during the construction period.

Administrative Procedures for Carrying Out Program I

9.80 All work performed in the various categories of the main-tenance program, including those of the* Plant Engineering Section both during normal plant operation and during plant shutdown, ~re in accordance with established administrative procedures described below. These administrative procedures deal with the conditions or requirements that must be satisfied to initiate and complete a maintenance operation rather than to exercise control over the actual repa~r work.

Non Vital Components 9.81 Administratively controlled maintenance procedures are not required on non vital components for safe operation of the facility.

Therefore, preventive maintenance or routine maintenance operations on non vital components is carried out by the maintenance sections in accordance

with normal standard maintenance practice, except as noted in Section 9.83.

The maintenance work on non vital components is coordinated with the Production Department to minimize downtime. Detailed maintenanc~ procedures f~r most pieces of equipment are provided by the vendor or are written by maintenance personnel; for hazardous conditions the operation may be altered and is

..administratively controlled as described in Section 9.83. Non routine maintenance of a non vital component is discusse~ in Section 9.84.

Vital Components 9.82 Administratively controlled maintenance procedures are required on vital components for safe operation of the facility. Therefore, prior to performing preventive maintenance or routine maintenance, it is necessary to evaiuate the effect of performing the maintenance work.

Such an evaluation is made on all items listed as vital equipment.

The Vital Equipment List is prepared by the Production Department and the Technical Services Department and approved by the Plant.Manager. If the maintenance work does not involve a radiation or contamination hazard, the work is initiated after approval by the Production Manager. If a radiation or contamination hazard is associated with the maintenance job, it is necessary to alter the operation as described in Section 9.83.

Non routine maintenance of vital components is discussed in Section 9.84.

Hazardous Maintenance

. 9.S3 When hazardous conditions exist, it is necessary to alter normal maintenance procedures before maintenance is initiat~d. In all cases, a Special Work Procedure is required. This work procedure is obtained and is administered as described in Parag~aph 9.17. The use. of this permit provides maximum assurance that both the worker and management take adequate steps to mini,nize the *consequences of radiation or contamination associated with the job.

In all cases involving haza_rdous maintenance, it is neces~ary to fulfill the requirements set forth in the Special Work Procedure. After this is done, the maintenance operation is performed in accordance with Sections 9.81 and 9.82 *.

Non routine Maintenance 9.84 Non routine maintenance involves changes in basic design or additions to equipment.

When it is necessary to perform this type of maintenance, on either.vital or non vital components, such maintenance is not carried out until a complete evaluation of such a change is conducted and approved by the CritJcality Committee. After the procedure is approved, the maintenance operations are performed in* accordance with Sections 9.81 9.82 or 9.83.

Work Completion 9.85 Representatives of Plant Engineering, Production and Technical Services (if involved) and Health and Safety Departments (if involved) observe the testing and return to operation of the components or system involved in maintenance

• Production Department 9.86 The Production Department ls responsible for the operation and maintenance of the processing plant and its rel~ted process services.

The organ.zation and administration of the department has been planned to provide safety to the public and plant personnel and to effect operation and maintenance of the facility within the operating license limitations.

In order to effectively operate the plant within the prescribed limitations, the Production Department has been broken down into groups to achieve effective control of the necessary operations. The group breakdowns are as followsz

a. Fuel and Mechanical Handling;

"* "a - rz. ' ' s

b. Chemical Processing; S"., -!-

, ~ i 1 ~ 2 I c~

Plant Engineering;

., 1 F, -.1 ~ er

d. Utilities and Process Services. ~ ¥~ * : I I - 1:3 9.87 The Fuel and Mechanical Handling group is *resporisible for the Fuel Receiving and Storage area including cask.transport, handling fuel assemblies, transfer and storage; operation of *the FRS water treatment facilities; Proc.ess Mechanical Cell operation including fuel assembly transfer, handling, disassembly by saw or *mechanical means, fuel shearing, handling of scrap~ utility se~vices to the area and hot equipment repair or replace-ment; General Purpose Cell including the loading, handling, storage and transfer operations of fuel baskets, scrap material and equipment utility services; Chemical Process Cell-Equipment Decontamination Room including the charge of fuel into and discharge of leached hulls out of the dissolvers, replacement of equipment, and remote handling operations within the CPC and the EDR; Scrap removal including the handling and transfer operations of waste and materials into and out of the mechanical head end iacilities.

Accountability and material control coordination consistent with Production Department requirements.

9.88 Chemical Processing group responsibilities include feed dissolution, solvent extraction, solvent recovery systems, product purifi-cation and concentration, acid recovery, sampling, cold chemical make up, waste concentration and rework operations, process off gas systems, building ventilation and accountability in these areas consistent with Production Department requirements.

9.89 The Plant Engineering group is responsible for the main-tenance of the facility as necessary to maintain continuity of operation as described in detail in Paragraphs 9.68 through 9.85.

9.90 The Utilities and Process Services group includes the operations of: all utility systems within the utility room, plant area and off-plant facilities; non radioactive systems for both solids and liquids; operation of the conventional low level burial and scrap removal from the plant; material handling including the transport, handling, warehouse and distribution of equipment and supplies as required for plant operations; decontamination of areas and faci) ities not included under other groups; i:aterial control including records of input, output and inprocess material nece~$ary to effect control; and accountability of source material an~

Production Department 9.86 The Production Department ls responsible for the operation and maintenance of the processing plant and its rel~ted process services.

The organ.zation and administration of the department has been planned to provide safety to the public and plant personnel and to effect operation and maintenance of the facility within the operating license limitations.

In order to effectively operate the plant within the prescribed limitations, the Production Department has been broken down into groups to achieve effective control of the necessary operations. The group breakdowns are as followsz

a. Fuel and Mechanical Handling;

"* "a - rz. ' ' s

b. Chemical Processing; S"., -!-

, ~ i 1 ~ 2 I c~

Plant Engineering;

., 1 F, -.1 ~ er

d. Utilities and Process Services. ~ ¥~ * : I I - 1:3 9.87 The Fuel and Mechanical Handling group is *resporisible for the Fuel Receiving and Storage area including cask.transport, handling fuel assemblies, transfer and storage; operation of *the FRS water treatment facilities; Proc.ess Mechanical Cell operation including fuel assembly transfer, handling, disassembly by saw or *mechanical means, fuel shearing, handling of scrap~ utility se~vices to the area and hot equipment repair or replace-ment; General Purpose Cell including the loading, handling, storage and transfer operations of fuel baskets, scrap material and equipment utility services; Chemical Process Cell-Equipment Decontamination Room including the charge of fuel into and discharge of leached hulls out of the dissolvers, replacement of equipment, and remote handling operations within the CPC and the EDR; Scrap removal including the handling and transfer operations of waste and materials into and out of the mechanical head end iacilities.

Accountability and material control coordination consistent with Production Department requirements.

9.88 Chemical Processing group responsibilities include feed dissolution, solvent extraction, solvent recovery systems, product purifi-cation and concentration, acid recovery, sampling, cold chemical make up, waste concentration and rework operations, process off gas systems, building ventilation and accountability in these areas consistent with Production Department requirements.

9.89 The Plant Engineering group is responsible for the main-tenance of the facility as necessary to maintain continuity of operation as described in detail in Paragraphs 9.68 through 9.85.

9.90 The Utilities and Process Services group includes the operations of: all utility systems within the utility room, plant area and off-plant facilities; non radioactive systems for both solids and liquids; operation of the conventional low level burial and scrap removal from the plant; material handling including the transport, handling, warehouse and distribution of equipment and supplies as required for plant operations; decontamination of areas and faci) ities not included under other groups; i:aterial control including records of input, output and inprocess material nece~$ary to effect control; and accountability of source material an~

• special nuclear material as necessary for Production Department requirements.

9.91 The basic plant operation and control is ~arried out physically by the process operators and shift supervisors; however, in a large proce, ing complex such as the NFS plant, additional support including technical. and analytical services, monitoring, accountability, maintenance and control is necessary to assure proper operations. The groups, listed in Paragraph 9.86 and staffed by production supervisors, have been established within the Production Department to provide the defined portions of this

~

support. Their primary function is to maintain an up-to-date intimate knowledge of their respective areas of responsibility. These staff functions have functional responsibility for their areas, however, administrative control is maintained ~y the Production Manager or an Assistant Production Manager.

This type of organization provides a decentralized type of functional responsibility, yet maintains centralized control over operations.

9.92 To the maximum practical extent all details of plant operation are controlled by written procedures. These include Standard Operating Procedures, Run Sheets (including administrative controls) and Letters of Authorization. These procedures are maintained in a current status as described in Paragraphs 9.5 through 9.7 and 9.94.

9.93 The Standard Operating Procedures include a detailed step-by-step.procedure for functional operation of each piece of equipment and/or process function in the plant. The format for SOP toyether with a gen,ral listing of the major systems covered by SOP are shown in Appendix 9.93. Included in each SOP is the scope encompassed, a general description of the operation involved, cautions to be observed in operations; adminis-trative controls required during the operation, references to related SOP or other procedures, detailed instructions for functional operation of the equipment and,insofar as possible, the mechanical limitations of the equip-ment. This last item may, in some instances, more appropriately be included in Run Sheets.

• 9.94 Run Sheets are another set of procedures used to maintain control of the plant operation. They list the operating conditions for the campaign of a particular fuel beginning with mechanical processing and continuing through the process to product storage. They include the upper and lower limits for each flow of plant processing. For example, maximum and minimum flow rates are list~d for each influent stream to each solvent extraction column as well as a desired operating flow.

Separate Run Sheets are used for each flowsheet authorized under the operating license. The published Run Sheets available to the shift supervisor and his operators are generally more restrictive than those permissible under the operating license. This practice allows more st~ict-enforcement and control of the plant operation. The shift supervisor cannot operate outside the specified limits of the R;Jn Sheet. HaNever, extension of these limits may be made, within the limits of the operating license, by an approved Letter of Authorization. If the supervisor cannot maintain the ope~~tion within the limit specified by the Run Sheet the affected portion of the operation must be shut down until the condition is corrected or approval to modify the run sheet is received. Run Sheets are reviewed periodically and amended as deemed necessary. Under no conditions is the.plant operated outside the technical specifications included in the opera~ing license.

0 9.~ Letter* of Authorization are an administrative procedure dlrecting actual plant operation as described in Paragraph 9.7. They are used to authorize

• apeclfic Run Sheet and/or auxiliary pr_ocedures for a

~tlcular proc111ln9 campaign and in iddition, are used to modify any of the re1trlctlve procedures established for plant control. All Letters of Authorization ar* approved as discussed in Paragraph 9.7.

9.96 The actual operation of the complete processing plant ls performed by personnel licensed as described in Paragraph 9.9 through 9.11. The basic areas of operator responsibility are broken down into specific catagories or areas of the plant consistent with production plant operating techniques. The specific ar~~s are manned by operators consistent with their group license. The specifically-assigned areas for each shift are as follows,

1. Central Control Room;

2. Process Mechanical Cell; 3; General Purpos.e Cell;

4. Fuel Receiving and Storage-Chemical Processing Cell;

!>. Sampling;

6. Chemical Makeup;

7. Product Packaging and Handling;

8. Waste Handling.

In addition, non licensed personnel are assigned to the following areas,

1. Utility Room;

2. Yards and ground, et~.

A brief -description of each of these areas outlining the basic operator responsibilities for the respective areas is as follows,

1. Central Control Room The Chemical processing portion of the plant is controlled from a Central Control Room located on the fourth floor of the process building. Processing beginning with dissolver operations and continuJ.ng through feed adjustment, solvent extraction, product purification, concentration* and storage are

_qperated from this location. Complete control of the process is exercised from the control room with the exception of non routine operatio~s such as J!!l_ny_al_block_y_alyes for the process service requirements which are located in the Upper and Lower Extraction Aisles. Manual valving in the Upper and Lower Extraction Aisles is performed by other individuals at the request of the control room operators or shift supervisors. The control room panel is a semi-graphic type for ease of identification and efficient operation.

In addition to posting the Run S~eets in the control room, many of the instruments are individually posted showing the limits of operation.

2. Process Mechanical Cell

a. Fuel assembly transfer and handling.

b. Fuel assembly disassembly usi~g saw or mechanical means.

c. Removal of extraneous hardware.

d. Make up of fuel modules and shearing.

e. Handling individual fuel elements.

f. Scrap handling, cell decontamination and in-cell remote maintenance.

a pre1cribed detailed form listing the constituent concentration, and total amounts of each tolution. The aolutton la then sampled and held for certi-fication. Following certification, and upon process demand, the solution la then transferred to a tun tank for subsequent introduction to the process or, in acme cases, directly into the process vesselJ.

8. Product i>ackaging and Handling

a. Load out of plui~nium product into bird cages and interim storage in the process building.

b. Load out of high enriched uranium into bird cages and interim storage.

c. Load out of low enriched uranium product to transport v,lssel.

All operations are conducted on a batch basis following specific instructions by the shift supervisor.

9. Supporting Areas I

a. Scrap Removal Areas, including the receiving and transfer to t.~e burial area of leached.hulls and other head-end scrap generated during processing,

. and transfer of new materials to the General Purpose Cell for head-end p~ocessing.

b. Equipment Decontamination Room, including the mechanical handling to and from the chemical process cell.

c. Process Laundry for decontamination of the anti-contamination clothing

.use~ in the facility.

10 *.!!1!,!ity Room Operation--All Plant Services Contained Within the UtJ..11 ty Room Complex

a. Water-**raw, filtered, process, demineralized, and potable.

b. AiT--process, instrumentation.

c. Steam--eq-,Jipment, process and heating.

d. Electrical--normal and emergency, Ct C

.... SU

Before the UNITED STATES ATOMIC ENERGY COi.MISSION Washington, o. C.

In the Matter of the Application of NUCLEAR FUEL SERVICES, INC.

For Licenses for a Spent Fuel Prpcessing Plant Under Sections 53, 63, 81, 104{b), and 185 of the Atomic Energy Act AEC Docket No. 50-201 Submission No. 21 - Final Safety Analysis Report Paragraphs 9.~7 through 9.117 of Section IX of the Safety Analysis, Revision of Table of Contents for Chapter IX, and Paragraph 1.91 October 26, 1964 5067.

.COjTENJS continued Section Paragraph Number IX PLANT OPERATION Organization 9.1 Administration 9.4 7raining of Plant Personnel 9.9 Training of Outside Organizations 9.12 Health and Safety Program 9.13 Emergency Procedures 9.64 Plant Maintenance Program 9.68 Production Oepartr.~nt 9.86 Process Maloperation 9.97 D

Revision l, October 26, 1964

JX PLANT Af'EBAJJ<J,I 1.91 Included in thls section are details relating to the foll~ tng subjects* Plant Organization, Plant Admini~tration, the training of plant personnel, the Plant Health and Safety Program, the Plant Emergency Program, the _Plant Maintenaryce Program, the operation a.nd function of the Production Department, and a discussion of Process.Maloperations.

1.92 Plant Organization discusses the duties of the Plant Manager, Production Manager, Health and Safety Director.

Technical Service Manager and Plant Criticality committee.

1.93 Operating Procedures, Letters of Authorization, the procedures involved in initiating and changing same, and the methods used. to insure that these procedures and Letters of Authori-zation are being followed are discussed in the Administration Section.

1.94 The Plant Training Program at various levels throughout the plant is covered under this topic. The Appendix to this section lists the curriculum to be employed in performing this training.

• 1.9~ The Health and Safety section defines the responsi-bilities of the Health and Safety Department, the general regulations and procedures for performing radioactive work, the maximum -allowable levels of radiation jn the plant (by zones), the monitoring sampling procedures, and the medical program.

1.96 Details of the fire protection organization and general procedures to be followed in the event of radiation emergencies are defined in this se*ctlon.

1.97 The operation of the Plant Maintenaoce Department, its organization, duties and functions are defined in this section.

1.98 Prcduction Department operations, including administrative controls within the Production Department, a break-down and a description of the operations in the plant by major areas are described in this section.

1.99 Process maloperations; results, the method determining the maloperation and the corrective action to be taken is presented, largely in tabular form, in this section.

Revision 1. October 26, 1964

Proc;;a,, Miloperattoo 9.97 lllch design, operating information, and experience exists on the chemical processing of spent reactor fuel, particularly on processing by solvent extraction. *The Purex-type of extraction process, which will be used in the NFS plant, is a thoro49hly tested process and one which.is ex-pected to operate without unusual difficulty. Plant operation is predicated upon the "norm" or usual condition where the equipment operates as designed and no *human mistakes are {Jlade. Obviously, such ideal conditions will no~

always exist. This section discusses possible maloperation in various areas of the plant, the results of such rnaloperation, the method of determining c maloperation, and the corrective action to be taken in the case of the particular maloperation. This information is largely presented in tabular form for ease in review and *assimilation. A list of abbreviations used in the discussion is found in Table 9.97.

Mechanical Cttead End} Processing.

9.98 Maloperations that may occur in the head-end processing are mechanical in nature.

• M>st of the maloperations envisioned involve a failure of a manipulator grapple during the transfer of f~el elem*nts or baskets and constitute little or no hazard, as such, but rather an incon-venience and time loss during processingr Maloperations in the FRS, PM:,

GPC, and CPC are itemized in table,.,sa, b, c, and d.

Dissolution - Feed

• Adjustment.

9.99 Dissolution-Feed Adjustment steps Jnder normal condi-tions are discussed in Sections 4.21 through 4.34. The nature of a mal-operation can vary with the f ue-1 be lng processed.

For dis cuss ion purposes, the fuels will be divided into (1) ceramic UC>i or Th<?:2 fuels, cylindrical in form,. clad in stainless steel or Zircaloy tubing (as repre-eented by Consolidated-Edison, Yankee, Commonwealth-Edison, and Northern States Power Fuels), (2) uranium-aluminum alloy fuels (MI'R type), and (3) Zr-U alloy* fuels (STR type). Maloperations in dissolving these types of fuels will be fairly representative of difficulties that can be encountered in all types of fuels processed. by NFS.

A summary of dissolution maloperation for these three fuel types, is given in tabl~ 9.99. Contents of this table are discussed in the following paragraphs a Ceramic Fuels, Stainless-or Zirc;aloy-Clad.

• Yankee, Commonwealth-Edison, and Northern States PaNer fuels are practically identical in processing. These fuels are chopped, loaded into baskets, and the baskets placed in the dissolvers, as described in Section IV. Dissolvent acid of the appropriate strength is added to the dissolvers, the solution heated, and the dissolution of the fuel proceeds.

Too low or too high a dissolvent concent~ation or too little or too ~uch dissolvent added to the dissolver are the maloperations most likely to occur. The r~sults of, and corrective actions* required by, such a maloperation are given in table 9.99.

f I

Measured Variable Column Pulse (Frequency-Pressure)

Density Flow Interface Position (Column)

Liquid Level Pressure Table *9.97 INSTRlllENT FtK:TIONS (Nomenclature used in Maloperation Dlscusalon)

~

°,3 i:

~.::.*

~'b

_p'.;

~O) g"'

~

"J-1\\.~,f'

~*

I A..,~

C D

DI F

FI I

L LI P.

PI Display Devices

~

. § l' 8

~

Z Cont:lling Device*

,..~

~~

l1 '1

% ~

f.

J'~

~

~o.

G~

~*.#

s

~.

• 'fi

~

l!

.:f 1 ~ o" -.."',

I

~

a

~~ -v

-v I I

A$

~

C?~ow

• NI nh

~li

~~

CIC.

CAL CAH DR FR LR

• LIC DR:

FFC IRC (LFC-T~\\

LFC FG DAL DAIi.

FAL IAH/L I IAH/L (hihh-low)

LAL LAH Dl:r

-~ -** * -

I I

I -- *1

. I...... I I -

I I

I Pli PG PAH Pressure Differential Pd PdR Pcl:L PdAL PdAH I Te-rature

'T TI TR TIC TIC 1G

. TAL TAH

Table 9.98a MALOPERATION IN FlEL RECEIVING AND smt\\GE AREA (FRS)

Maloperation

. 1. I Grapple f allure during fuel transfer.

2.

Fuel element cannot be re-trieved from floor of cask unloading pool.

Result Fuel element drops to floorl Visual.

of cask unloading pool.

R " \\ e ci ~ ~......., "0 J j"'* ~,

C

<,o.t r, ~,~ \\/

Indication Inconvenience and time

'/:

t.. **

. () ((Cf..'" (/

r e,c..

41.-

fl l..

, l U* t. -

lossJ Poor visibility or binding between cask and pool wall or canisters.

3. I Fuel element stuck in in-I Inconvenience and time lossJ Visual.

dividual shipping *slot.

- 1,,..,"

JA-~,-~,,4' A /corrective* Ac.tion

'" Special underwater ex*

tenaio, wrenchee and tongs used to retrieve fue 1. W.... ~ w *.,/ ~ ~

-,o Cask cover replaced and cask removed to ~-,, **

tamination area to pro-vide working space.

Use underwat~r working tools or fabricate equipment for removal.

Table 9.98b MALOPERATION IN ~C:X::ESS MECHANICAL CELL AREA (AC).

Maloperation Result Indication r.o==- -~1ve Ar-t'.'"'"

1.

Major breakdown of shear Inoperative shear Visual.

HQld fu~l element or during operation.

mechanism.

segment in PIC as l!.)ng as necessary, mon1°r-ing temperature oc-casionally. If temper*

ature is too high, water cool with spray nozzle*

in shear magazine.

Repair shear.

2.

Saw cut through fuel Saw cuttings containing Visual. krw~

Coolant will be proces*

element (wet cutting).

fuel particles drawn into sed if desirable.

saw coolant.

3.

Saw cut through fuel Saw cuttings containing Visual. ~7,.

Remove filter to fuel element (dry. cutting).

fuel p~rticles are drawn basket for subsequent i nto blower system filter.

processing in dissolvers if economics dictates recovery of fuel

• I

Table 9.98c MAI.OPERATION IN GENERAL PURPOSE CELL AREA (GPC)

Maloperation Result Indication Corrective Action

~rapple failure during Part of fuel may be spilled Visual.

• Power Manipulator transfer of loaded fuel onto floor of G~.

(2V-73) brought in and basket.

pieces retrieved one at a time

• o fuel basket *

• Notes Same procedure is, sed to retrieve spille~ leac ried hulls.

2.

Fines dropped.on cell Lost fines I inconvenience.

Visual.

Wash down cell floor to floor from dropped fuel*

and time loss.

criticality safe sump *

. basket!

Remove s~p pan and place fines in fuel basket.

3.

Sheared fuel lodged in Stuck fuel piece in reducer ~ Ganma monitor (2-LAH-1) at Remove fuel basket and reducer during descent.

9-inch diameter point in fuel chute tip1 remove reducer.

fuel pieces with master slave manipulator& or power manipulator

• l I

I f

I I

I I

~

i I

I

1.

Table 9.98d MALOPERATION IN THE CHEMICAL PROCESS CELL AREA (CR:)

Malo ration Result 1na1ca~1on Lorrccsixc OE~IBQ 1

I Grapple failure during Fuel or leached hulls transfer of fuel basket spilled on floor of cell.

or leached hulls to or from Dissolvers 3:-1 and 3:-2.

I Visual.

Recover pi~ces with power manipula~or.

Consolidated-Edison fuel is the exception in this group. This fuel contains mixed thorium-uranium oxides in the fuel and requires a much higher nitric acid concentration than other fuels of clad-oxide-fuel type, plus.04 Ai hydrofluoric acid in the dissolvent to dissolve the thorium oxide.

Possible malopfrations in this case include all the above cases plus insufficient, or excess, or omission of, the hydro-fluoric acid.

Boric acid is also added to the dissolvent on the Conso ~idated-Edison fuels. Omission or change in concentration of this component is a criticality consideration and is discussed in Section VI.

Aluminym-uraniuo Alloy Fuels These fuels, primarily of the MrR type, are charged into the baskets without chopping. They *are dissolved in 5.4 1:1 nli ric acid in which

.00~ M mercur,ic nitrate serves as a dissolution catalyst. Malopera-tions would include the use of acid with too low or too high a con-centration,,too little or too much acid of the. correct concentration, the addition of too l ittle, too much, or the omission, of the catalyst.

These maloperations too, are show~ in table 9.99.

• The addition -of too much or too little acid to the dissolver is un-likely since the dissolvent is made op batchwise as required and is certified prior to transfer in total to the dissolvers. The remain-ing maloperations, their consequences and corrective action required are given in table 9.99.

Zirconium-Ura~i*~ Alloy Fuels Dissolution of fue l s of this type is accomplished by the appropriate addition of 1.01 *nitric acid to the dissolver wfth ha~kets of un-chopped fuel in place, and to which is added 27.6 M hydroflu0ric acid at a sufficiently slo'N rate that little free fluo'ride ion ls present (fluoride is complexed by the dissolving zirconiu~). After completion of the dissolution, aluminum nitrate-chromic acid solution is added to further complex the fluoride and to oxidize tin (from the Zircaloy) and uranium to their higher, and more soluble, valence states

• Moloperation can occur from incorrect quantities and concentrations bf the three soluti ons ~dded to the dissolver. Conseq~ences of, and corrective measures to be taken after, such maloperations, are given in table 9.99. Stability regions for the dissolver product-feed solutions are give~ in TID-10089.

Other Fuels The SCRUP fuel is clad in aluminum and this cladding is removed with an NaOH-NaN03 solutior before dissolution of the fuel.

Excess de*

cladding solution or high component concentrations in this solution will result in higher-than-normal waste solution volumes but will have no serious process consequences. Insufficient decladding

~ -.*- *-

u Table 9.99 MAI.OPERATION AND CORRECTIVE ACTION OlMlt<<i DISSOLUTIOk (X:*l or X:*2)

~ ** 1 TVD8 Ualo~-ration Dae.111 +

I

--*4ua 4 r +4""

1. ICeramic1 Stainless Steel1 I Low acid concentration in Zirconium Clad.

dissolvent.

i Low concentration of fertile material in dis-solver product, all fuel

~vt dissolved1 low acid.

Recti.rge dlaaolver with additional dlasolvent,and.

complete i la1olutlon.

Acid adjustaent in 30-1 *.

2.

Ceramlc1 Stainless Steel1 Zirconium Clad.

Too high an.acid concentra-1 High acid in extraction tion in dissolvent,.

feed1 low decontamination factors if run.

Dilute high acid 1n 30-11 run w1 th low concentra-tions of uraniua, thoriua, end plutonium through tha partitign cycle.

-*,------------+-------------+-------------+----------

3.

4.

~.

Ceramic1 Stainless Steel1 Zirconium Clad.

• Ceramic1 Stainless Steel1 Zirconium Clad*~

Insufficient dissolvent.

• Excess solvent.

LAL-21 LCL-i1 high concen-tration of fertile and fissile materials in dis-solver product. Low acid concentration.

High acid and low metals content of dissolver product.

Adjust concentrations 1n 30-1 into HAf specifica-tions1 add additional solvent to dissolver.

Dilute to proper acJd concentration in 30-11 run to extraction syatea with low *tali content.

• If filled too high, could loverllow to the other diaaol~r or to 3D-l. In this event, a batchwise adjustment of the solution of all thilee vessels' contents in 30-1 lwould be necessary.

Uranium-aluminum alloy.

Low acid concentration in dissolvent.

Low aluminum concentration in dissolvent productr Acid deficient product.

Some fuel may be undis-.

solved.

NJTE1 The indication in cill of the above maloperat1on~ is by sample analysis.

~~~

. i'?' __

Add additional diaaolvent to X:-1 and 3C-2 if necessary to complete.

d1ssolut1on. If *neces-sai'f,* add acid to 30-1.

Boil down to correct aluminum concentration.

---

*"~

Table 9.99 Continued MALOPERATION AND cmRECTIVE ACTION DllUNG DISSOLUTION (3C*l or 3C-2)

J:uel Tvoe

6.

Uranium-aluminum alloy.

7. IUranium-al\\JIDlnum alloy a.* IUranium-aluminum alloy.

9.*

Zirconiwn-uranium alloy.*

Malooeration High acid concentration in dissolvent.

Low Hg(N03)2*

High-Hg(~)2*

High HNOJ concentration.

Aoc.11\\t.

High acid in dissolvent product1 low saltine.,

strength for extractiop.

Dissolution incomplete1 dissolver solution low in aluminum, high in HN03*

No process consequences?

however, may limit concen-tration of waste.

Possible precipitation from feed upon long standing.

r.n~rectiv* A~t.fftn Add Al(N0:3)3 to 30-1 and adjust t 1 extraction feed specificati~ns.

Retain solution in dl1-solver1 add Hg(N0..:.)2 and,

continue dissolutron.

Adjust wastea boil off if necessary.

Dilute solution in dis-solver or in feed adju1t-ment tank if dissolution has already occurred.

• Nitric acid fs added to d~ssolvers batchwi~e. Hydrofl*Lr1c*~cid is metered separatJly into the nitric acid.

Stability ranges are givef\\ in Uo:Dl9J for zirconium-uradtum dissolution.

10. Zirconium-uranium alloy.

11. IZirconium-uranium alloy.

Low ffN0:3 concentration.

Excess HF.

Lowered solubility of zirconium in dissolver solu*

tion.

Aluminum precipation in feed adjustment. High equipment corrosion.

t<<>l'Ea The indication in ab.l of the above maloperationsf is by sample analysis

• Add concentrated~ to dissolver or to feed adjustment tank if dis-solution has already taken place.

.Dilute diasolveT product with water or-additional HN03 as stability region permits.

Table 9.99 Continued MALOP£RATI~ AND CCJUlECTIVE ACTI'* WUNG DISSOLUTION (~-1 or ~-2)

.. -w.-........,....

• a.6Vl,IIIIIJ.& Cl 1,,.A.VII

..............,...-.......... -~*---.,.

12 *. Zirconium-uranium alloy.

Insufficient HF.

Zirconium precipitation 1n If solution 11 1tlll in dissolver or feed adjust~

di11ol.ar, ~

additional ment

• All alloy may not HF, heat, and apuge.

dissolve.

Add HF at controlled rate, using dilution air.

13. Zirconium-uranium alloy.

Excess A_l(N03)3*

Aluminum precipates.

Add additional Hf to bring composition into stable range.

14. Zirconium-uranium alloy.

Insufficient Al(N03)3*

Zirconium precipitate&J Ad~additional Al(~)3 high corrosive rates on to dissolver or feed stainless steel equipment.

adjustment accountability tank as required

• NOJ'S*

The indication in a* l of the above maloperations ls by sample analysis.

f

solution or low component concentrations in this solution will result in only partial jacket r~moval.

As a result, dissolution may be in-complete or perhaps may not even start. A rep,eat of the decladding step,ill be necessary in this case. Oecladding should not be atteq,ted until all previous dissolver solution is removed from the dissolver so that precipitates of sodium diuranate cannot be formed.

Maloperation during the actual fuel dissolution will be very similar to that for the oxide fuels and table 9.99 should be referred to for possible consequences and corrective steps. Maloperations in the feed adjustment and accountability tanks are concerned primarily with human error. The most important of these errors involve the failure to add the proper cold chemicals or to thoroughly mix the solution prior to sampling. These and other maloperations are detailed in table 9.99 for tanks 30-1 and 40-1. These maloperations will also apply to other feed adjustment and accountability tanks and to other downstream feed adjustment and neutralizer tanks.

Sglyent Extraction - Partition Cyc1°; Extraction Column (4C-l);

Feed PvwP Pots C'lC-lJa and 1Jb); and Meter H~ad Pot (4y-1a}

9.100 The most likely maloperations that can occur on the column, the feed pump ~ts, and the meter head pot, will be discussed in this section.

~~s Many of these maloperations will be co1Trnon to.a.l.l. pulse columns and tabulatio:1s made (or this column will be applicable to downstream flowsheet columns as well. Further, it is.recognized that maloperation of this column, in a cascade system such as is used here, will affect some of the downstream columns. A discussion of sequential difficulties

• will not be made here; this discussion will be used as being typical of dcwnstream columns.

The plant operating and surveillance procedures are deuigned to ~etect ma1operations before consequences*can significantly upset any colu~n.

The maloperations that will be tabulatad and which are applicable to other columns follows

1. Incorrect pulse amplitude-frequency settir.gs;

2. Incorrect flow ratios between the various streams;

3. Incorrect stream compositions;

4. Int~rface crud and organic quality;

5. Column flooding;

6. Loss of column jacket cooling water;

7. Nozzle plate fouling.

Pulse amplitude-frequency settings, if intorrect, can result in high column waste losses, or column instability and flooding combined with high waste losses.* Figure 9.100 illustrates the variation of HETS (Height Equivalent to~ Theoretical Stage) with. varying pulse amplitude settings.

l: the combination of the column design and flowsheet requirements are met by conditions at point A (which represents minimum HA column pulse amplitude requirements for the Yankee fuel flowsheet), then operation to

Table 9.99 Continued MALOPERATION OF fc'ED ADJUS1MENT AND ACX:OtMTABILITY TANK (30-1)

Malo;-.;;.. ac.inn Dacul+

Tn,H.caf'.ion r.-

1.

Low steam in coils.

Poor temperature control.

TI-2-51 LR-61 DR-41 Sample Hold solution in tank Cannot evaporate solution 3C.

until sample *hows it for concentration correct-is of correct concen-ion.

trationa add cold chemicals to proper concentration.

~

2.

Low cooling water in coils Poor temperature control TI-251 DR-4 *.

Check cooling water for jetting.

valvei hold solution in tank until proper condl*

tions are achieved.

i

3.

Condense.r water not on.

Loss of solution (vapors)

TI-2-4 in.condensates LR-5 Check cooling water valve.

to vent system.

in 30-~.

4.

Incorrect cold chemical Incorrect concentration of Sample 3Ca LR-61 DR-4.

Hold solution in tank1 addition.

solution.

add proper concentration I

of cold chemicals to bring the solution to proper concentration.

5.

Air sparger off.*

Poor mixin~1 incorrect Sample ~I K:-24.

I Turn on air1 hold solu-I I

sample.

tion in T~nk IC-"24 until

• concentration is correct *

~

~

i L I #I 4 L h _ JI &3. 3 4 A ;a

• c, Table 9.99 Continued MAI.OPERATION OF FEED TANK 10 PARTITION CiCLE ( 40-1) aa.a.u~.i-01,.1...,11 1,vaw.a...,

......... ~... __,.,..

1.

Low cooling water.

Poor temperature control1 Tl-2-14.

high feed temperature, re-sulting in poor column efficiency if HAF is too hot *

2.

No air sparging.

Poor mixing, which results tc-29 off1 manual check of in a bad sample.

hand control valve.

3.

Incorrect cold chemical Wrong solution composition, Sample.4C.

addition.

which can cause a loss to product to thf HAW stream or poor~ ~.

111!!11'1111 Increase cooli-~ water flow.

Tum on Hand Control v~lve fTom air supply 1 (1£-29) 1/2 hour prior "'

to taking sample.

Certification and adlain-

. istrative check of prope amounts of cold chemical to bring the solution ta the proper concentration I

r

• the left of this point would result in poor contacting, higher HETS values, fewer stages, and consequently high fissile material losses to the waste stream and.poor decontamination of product. Point A in figure 9.100 represents minimum pulse amplitude-frequency conditions. The maxil1'Alm amplitude-frequency situation would be represented by point B.

Amplitudes above B would result in increased stage heights due to in-

.adequate phase disengagement or other effects. Hence, for a given pulse frequency~ the amplitude operating range would fall between points A aQd B. Op~rating outside ~hese ranges would be considered maloperatio~

• Figure 9.100 HETS VARIATION WITH PULSE AMPLITUDE - HA O)LUIM At a particular frequency (cycles per minute).(See note)

Maximuo acceptable HETS

~~~~Operating Range-~~~~~~

Pulse Amplitude NOTEs Different frequencies would generate a family of curves with similar characteristics.

For satisfactory operating conditions, low frequencies are necessarily com-bined with high amplitudes and high frequencies with lo~ amplitudes.

W~rkable. operating combinations.will be established for each column and flowsheet.

Maloperation would be due to human error or equipment failure and the correction needed would be obvious in each case

• I Flow ratios of the organic-aqueous streams (0/A ratio) have direct bearing on waste losses and decontamination factors. Losses from the HA column, as well as thP IA, IO, and IIA columns, will decrease with increasing 0/A ratios. The magnitude of 0/A increase is limited, however, by the increased flow or ratio change possible in the downstream column, or by flooding the column in question. TFor example, the 0/A ratio in the HA column for the Yankee flowsheet is normally about 1.7. If this ratio is decreased by increasing A (or the feed rat~, to approximately 0.80, the waste losses in the HAW stream will start exceeding the economically permissible value of 0.1% of the uranium in the fuel.

Likewise, increasing the 0/A ratio (by increasing organic - the HAX stream) by a factor of 2.0 would result in a flooding condition in the column and again high uranium in the HAW and, additionally, organic phase in this stream.

Operating limits on the 0/A ratios will be set on each of the columns in the system as discussed for the HA column above. Operation outside these limits would be considered maloperation and would be the result of human error, failed or *erratic flowmeters, or metering pump.

In each of the above, pulse or flow-ratia maloperation, the most probable result is high fissile losses i~ the raffinate stream. Recovery of fissile *material will be discussed under the operation of the rework system. Column maloperations are tabulated in table 9.100.

Maloperati-0ns in the Partition Cycle Feed Pump Pots and 1,,eter.Head Pots are mainly concerned with flow -rates to *and from the column.

These are also discussed in table 9.100.-

Plytooium Cycle Feed Conditioner Tack {4D-6}

9.101 Maloperations of the ?lutonium Cycle Feed Conditioner Tank {40-6) are shown in table 9.101. The maloperations are as described

• in paragraph 9.99.

Feed Conditioner to First Uranium Cy; 1~ (40-9}

9.102 Maloperations o*f the :=eed Conditioner Tank for the First Uranium Cycle (40-9) are shown in table 9.102. The malop~rations are as described in para9raph 9.99.

Second Uranium Cycle Fe 0 d Conditione; (40-12}

9.103 Maloperations of the Se.cond _Uranium Cycle Feed Conditioner Tank (40-12) are shown in table 9.103. The maloperations are as described in paragraph 9.99.

Plutonium Purification c0 11 9.104 The operation of the ?lutonium Purification Cell is disc~ssed in paragraphs 4.73 to 4.75. ~4loperations that have been con-sidered consist in general of human errors. These include th~ over-flowing of tanks, improper solution adjustment, improper valving, and failure to operate Plutonium Ion Exchangers 5C-1A, 5C-1B, and 5C-1C properly. These maloperations are tabulated in table 9.104.

1.

Table 9.100 Continued MALOPERATION OF HA COLll&N ( C-1)

MaJoperatlon 1

Result Jndicat\\on

. corre;tiye Actieo Flooding.

Organic leaves ttae column through the normal *aqueous effluent line *and aqueous thr~ugh the normal organic effluent line if allowed -to continue. This allows product to the waste stream and fission product to the product stream. Both will require rework of material.

Interface controller LR:-29 will fiist become erratic then will decrease loading pressure tu pot 4Y-15 sig*

nificantly allowing organic to flow out HAW line.

DR-23 (Pxtraction section density) will increase in*

dicating layer build-up of aqueous in the column. The layer of aqueous may break occasionally showing a sharp return to near normal of DR-23 reading with a following gradual build-up.

PR-8 (column static pressu*

will be erratic but will show a substantially higher reading, indicating increa-sed aqueous in the column.

DR-22 (densiiy in top dis-engaging head) will in-crease significantly when aqueous replaces the organic in this section.

When aqueous flows to surge pot 4Y-5 and thence to HBX column <<:-2, the interface controller, LRC-33 will be unstable due to increase in aqueous flow to that column Reduce column through*

put rates and/or pulNr frequency and amplitude.

Rework material as nece,-

sary. Organic which left through the HAW line will be separated in the decanter 4Y-l and returned to the No. 1 solvent system.

la.

lb.

le.

2.

u Table 9.100 Continued MALOPERATION OF HA COLUMN (C-1)

Maloperation Fl~oding caused by high flow rates.

Flooding caused by poor disengaging time of organic.

Result Same as item 1.

Same as item 1.

Flooding caused by pulser! Same as' item 1.

amplitude-frequency.

Cyclic flooding.

Probable larger amounts of entrained organic in HAW stream. Exc_essive product loss to HAW and poor de-contaminatic~ in column.

Indication

.Sam& as item 1.

Indication will be same as item l in column. If organic disengaging time is the cause, analytical samp-les will indicate poor dis-engaging time or*organic degr~dation:

Same as item 1 for column instrumentation. Pulser frequen_cy can be counted on oscillations of PR-8 or by counting RPM of pulser poppet cam shaft. Pulser amplitude determined by maximum pressure on pulser surge tank (pulse stroke).

Interface controller LRC-29 will be erratic, caused by a repeated build-up and breaking of an aqueous layer in the extraction sec*

tion of the column. This will reflect an erratic be-havior of DR-23 also. PR-8 will increase slowly as the aqueous layer builds up, then orops sharply to normal as it breaks.

Corre_~tive Action Reduce flow rates.

Stop flooding as in item 1. Correct organic problem by solvent wasting or replace solvent.

Adjust RPM of pulser motor or adjust pulser surge tank pressure.

Reduce flow rates or re-duce pulser frequency-amplitude. Check analy-tical data for organic disengaging time.

0 Table 9.100 Continued MAI.OPERATION OF HA COL~ (C-1)

Pula::

1::::~i.on - - *---- l Col~ wi~~si~di-at~-ly ~;*

into a total flooded condi-tion as in item l.

3.

_ _Jn~J.a1l2n XA-19 pulser alarm will sound. Oscillation effect of pulser on instruments PR-8, LR:-29 will stop.

Erratic behavior of instru*

ments described in item l will not occur~ PR-8 will show increas& as aqueous builds up in column displac*

ing organic. DR-23 will in crease, finally showing total aqueous *. O,rganic flow to-Surge.Pot 4¥-5 will decrease to, approximately 1/2 since flow will be equal to aqueous influent only. LJC-29 loading pressure ~will oecrease sig-nificantly ~o allow organic to flow.out HAW line.

cgrrectiva ActJoo Shut down n*rtltlon.

cycle colUAnnt b111tditt.ly.

Restart 1fter pul.Mr trouble ia corrected.

Rework material**

necessary.

4.

Loss of HAF.

5.

Loss of HAX.

I Loss of product in normal stream. Pos$ible carry-over of excess fission product to subsequent.

cycles.

Loss of product to waste system.

Zero readirig on LR-28; In-terface Controller LR:-29 will decrease sharply until controller adju~ts to bring

• level back to desir,d point-DR-23 will decreas~; DR-22 I will decrease; PR-8 will

! decrease

• I If not re,tored within five minutes, shut down partition ~ycle.

~I FAL-9, FIC-22 on HAX; flow f Shut down partition i to Surge Pot 4Y-5 stops;

, cycle iamediately.

• i LIC-30; DR-23 and PR-8 will f Rework as necessary.

1 increase.

Malor:>eration 6~

Losa of HAS *

1.

High HAF flow rate.

e.

Low HAF feed rate~

9.

High HAX feed rate.

10. I Low HAX flow rate.

11. I HSg\\ HAS* flow rate.

12. I Low HAS flow rate.

0 Table 9.100 Continued MALOPERATION OF HA COLlllN (C-1)

Result Decontamination will reduce by a factor of 10 to 100.

Fission produ_cts will be carried to dQwnstream cycles.

Loss of product to HAW.

Possible stripping of excess fission produ~t and carry-over to subsequent cycles.

Same as item 8 Same as item 7.

Indic._ation FAL-7, FR:*l on HAS1 LAC-29 will decrease sharply until controller adjusts for re-duced aqueous flow. DR-23 and PR-8 will decrease.

LR-28 increase (HAF measur-

.ing pot). PR-8, DR-23, and DR-22 increase, LR:-29 lower loading pressure.

LR-28 decrea$e (HAF measur-ing pot). PR-8, DR-23, and DR-22 decrease, LR:-29 hlcjler loading pressure.

FRC-22 (HAX) increase, PR-8, DR-23, DR-22 will decrease.

FRC-22 (HAX) decrease, PR-8, DR-23, DR-22 will increase.

Cn i+fva lr.t'.f.ftn If not restored within 5 minutes, shut down partition cycle.

Lower HAF feed rate (FRC*l~). Rework material as necessary.

Increase HAF feed rate (FRC-16).

Decrease HAX rate with FIC-22.

Increase HAX rate with FIC-22.

Increase wast~ volume of HAWtFIC-1 on HAS1 PR-8, DR-23 I Reduce HAS flow with FR:-1.

P~ssible loss of product to _ will increase1 LRC-29 load-Rework material as necea-iwaste.

ing pressure will decrease. sary.

Fission rrCtducts w:J 1 not be I FR::-1 on HAS1 PR-8, DR-23 I Increase HAS flow with

..scrubbed from organic in will decrease, LR:-29 load-FIC-1.

scrub section of column.

,ing pressure will increase.

I I

13.

Malooeratlon Lo11 of salting agent from HAS.

14. INo cooling water on HA column 4C*l jacket.

1~. Excess interface crud.

16. IFouling of column nozzle plates.

Table 9.100 Continued MALOPERATION OF HA COL\\lti ( C-1)

Result Indlcatlon Corrective Action Same as 1t~m 12. Wlll al&o 101-1 and lab analyst, on cause refluxing in colunn 140-171 PR-8, DR-23, and and loss of product to waste DR-22 will decrease.

stream.

Certiflcatlon of HAS 2 make-up 1olutlon befo~

u, *.

Depending on temperature of !Cooling water valvee close~ Ope~ cooling water influent streams, may cause Sample analysis of effluen~ valves.

less efficient decontamina-streams.

tion factor and waste loss.

Increase disengaging time.

Possible poor decontaminatkr of Zr and Nb

• Sample analysis of effluent streams.

HAW anc' uranium and plutonium streams from subsequent columns.

Reduce ~ontact of organic Can only be determined by aqueous and number o.f column critical analysis of all stages with resulting loss factors affecting column of column efficiency.

operation.

j I

I I If severe enough to affect product apeclfica*

tion3, increaae aolvent clea~up to laprove dl1-engaging time. If trouble still prevail*, displace interface to HAW syat.. by Standard Operating Proc.lJn.

Clean out system and try

• to remove fouling mater-ial by chemical flushes.

If unsuccessful, replace column.

Table 9.100 MAI.OPERATION OF PARTITION.CYCLE FEED Pll&P POTS (C*l3A

• 138)

Maloperation R11u}t Indlc1tion Cou,ctiyt Action

1. -1 Air aupply ls high.

f I

I

2. I Low air supply.

3. I Loss of pot due to crud in check valves.

Cause high flow of HAP I LR-28 on Meter Head Pot stream.

4Y-14 and FJC-16 on air aupply.

Cause low flow of HAF strea1' LR-28 on Meter Head. Pot 4Y-14 and rJC-16 on air supply.

Low or l.oss of HAF stream. I LR-281.fJC-16.

Correct Mtt1nga on Lft.. 28 and Fll:-16 inatn..,.nts.

Correct 5'.tting on LR-28 and *FtC-lt in&truMnts.

Switch to ai-r lifts for HAF *feed.

Table 9,100 IW.OPERAnOO OF METER iEAO POT ( 4Y.. }4)

1.

High level in pot.

~

solution will go throuj, L~*281FIC*l6.

the overflow line to Tank 40-1 until malo~ration ls corrected.

2.

Low level in pot.

Low flow to HA column.

LR:*281 FR:-16.

I

__,.. _____.. _..-.w.........

Lll:*28 will adJuat au flow to puap,.ota c-m and 1381 naet lnatiu-aa.nt, if nece11ary.

3A..

Table 9.101.

/

).

MAI.OPERATION OF PLUTQUlll-c'a'CLE FEED Ca.DITIMR TANK... { 40-6).

---

~

• • ....--~v......,. **

1.

High level in tank.

Cause overflow to 6D-3.

Lll:-ls sample in Vent Correct flowa in and Syst~m Catch Tanlc 60-3.

out of tank, ld recycle product from vent 1yata to rework.

2.

Low oxidant concentration Results in insuffi'cient FR1 FAL on cold chemical Add oxidant to correct of *cold chemicals.

oxidation of plutonium, feed tank1,sample 9Ca~DR-l. concentration and recycle which causes the plutonium

(

IIAW from 40-8 to rework.

product to go with the. IIAW stream.

3.

No air spargin91 poor Poor tank mixin91 poor

.Air Supply Valve te-11 thin Open te-1 1/2 hour prioz mixing.

sample.

line on uc-1.

to sampling1 check and 1 og opening of spare valve

• Table 9.102 MAI.OPERATIC* Of FEED C~DITI<RR ':0 FIRST lRANllll CYCLE (4D-CJ)

Maloperation Result Indication

1.

Wrong cold chemical Cause wrong feed concentra-

~~31 saq>le ~.

addition.

tion to Uranium Cycle and loss Qf product to IAW or poor df *

2.

Low cooling water on cooleJ Poor teq>erature control ma) Valves closed on ~oolera 4E-4.

cause poor col\\.l'lln efficieOC)I column effluent saq>les.

3.

Air spargar off.

Poor mixing, resulting in a tc-4 offs sample 22C1 DIC-3 bad sample and poor concen-tration control on IAF.

Corrective Action Correct concentration recycle IIAW to the N eystem.

Open valves on cooler.

Insure aparger on for hour prior to saapllng and 1ft 2

Table 9.103 IIALOPERATIO. OF ~D lRANILIA C'tCLE FEED CCWDITIC*ER (40-12) llalo ration Result

1. I Incorrect cold chemical Wrong feed concentration, FR*on Cold Chemical Feed Correct concentration addition to 40-12.

resulting in poor extraction Tanlca sample 26B1 DR-'7 on in tenk a'td recycle IDI and loss of product to ~ar 40-12.

f J'OI\\. 41).13 to nwort.

or* poor df.

2. I High level in tank 40-12. l0verflow of tank1 loss of

• LR-221 LAH-6 in Tank Adjust flows into and out product to 60-3.

40-12.

of tank and recycle frCII I

I off-gas system to rework.

3. I Low level in Tank 40-12. I Loss of iDF -'Ilow.

LR-12 in Tank 40-12.

Adjust flow into ar.d out of Tank 40-12.

4. I Low cooling water to IPoor IDF temperature control Valves closed on coolers I Open valves on cooler.

Cooler 4E-5.

column effluent samples.

5. I No air sparging.

)Poor mixing of IDF.

tc-8 off1 saarple 26B.

I Tum on tc-8.

1.

Maloperation Plutonium Ion Exchange Feed Conditioner !ID-1 has been allowed to overflow.

Table 9.104 IW.OPERATION IN PLUTOUllA P~lFICATION CEU. (PPC)

Result Overflow goes to Ion Ex-change Recycle Waste Tank 50-2 and then to the Plutonium Feed Conditioner Taruc 40-6 for rework.

Indication LAH-12 and LR-14 on 5>-h LR-1 on 50-2.

2. I Improper Plutonium Ion MaxilllWll plutonium absorption,TIC-3 on 5D-i1 TIC-4 on Exchange Feed.Conditioner by the anion resin will not 50-20.

3.

~l temperature and/or be accomplished if the 60 Ct temperature at Plutonium 10 C plutonium Ion Exchanger Ion Exchangers !C-lA*

temperature is not malntai

!C-18, and !C-lC due to At higher and lower tempera-pump ~-7 not operating or tures. resin exchange ef-improper heating of hot ficiency is lost.

water tank 50-20.

Improper Plutonium Ion Exchange feed concentra-tion (50-1) due to poor sparging or improper feed adjustment of H~ and N~.

Without.proper sparglng.

1381 l4B* 158, and 168 proper adjustment of feed in samplers give effluent con-50-1 car.not be obtained. If centration1 sparging indi-HNO is not adjusted cated by variation in level to ~.2 !! or the plutonium is recorder1 failure to add not oxidized by sodium indicated by FR-91 nitrite, some plutonium will by LR-14.

pass through the Ion Ex-changers (~-lA. ~-10. or

!C-lC) during the loading

!cycle. to the Ion Exchange aste Tank 50-2 and will have to be returned to the Plutonium Feed Conditioner ank (40-6) for rework.

Corrective Action Start flow through one of the Plutonlua.Ion Exchanger* !C*lA, !C*lB, or ~-le.

Readjust temperature and refer to *standard Operating Procedures if necessary1 check 5>-20 for necessary -water levela

• adjust proper temperature with TIC-4.

Adjust NaN0:2 and HNO addition to tank s,-f if product does not meet specifications.

I I

4.

~.

6.

1.

a.

9.

llaloperation Improper valve open during elutriation cycle of Ion Exchange11,::-1A,,::-18, and,::-1c.

Improper valve open during loading or wash cycle of Ion Exchanger (,::-1A,

,::-18,,::-1c).

Qait wash cycle of Plutonium Ion Exchangers

,::-1A,,::-18, ~-1c.

No~ or less than 0.2511 ffNO:J elutriant used in elutriatlon-cycle PlutoniUJ Ion Exchanqe Recycle Waste Tank.

Overflow of Plutoni\\.lD Ion Exchange Recycle Waste Tank (50-2).

Improper use of Steam Jet 5H-32 (Pump-Out Ecluctor on

&>-2).

Table 9.lOI Continued 1W.OP£RATI0N IN. PLlm>>nllA PlRIFlCATION CELL (PPC)

Result lndlcatlon Excess plutonium*will be J Sample 5Y*l4, 5Y-16, or ca~led to the Recycle Waa 5Y-18.

Tank ~2.

Impurities and f1111on pro-ducts w111 be carried to the Plutonium Product Feed Evaporator Tank 5D-4. This material can be sent back tc the Plutonium Ion Exchange Feed Conditioner Tank (50-1:

for rework.

Off product specification.

formation of plutonium polymers which might plug resin bed.

Overflow will be carried through vent line to the Vessel Off-Gas Catch Tank (60-3) where it can be re-worked or sent to waste.

May result in transferring pluton!um product into Plutonium Waste Catch Tank 40-8.

LAH-13, LAH-14, LAH*l5 on Ion Exchange Column11LR*l9 on 50-4.

Sample 148, 158, or 168 on

~-lA,,::-18, or,::-1c and Sampler 178 on tank 50-4.

Tank 141)-46.analyses, no flow by LI-27 and LRG28.

LAH-l or LR-1 on Tank 50-2.

LR-22 on Plutonium Waste Catch Tank (40-8) and sam-ple analysis of 40-8.

r~n~,..., u. a,-+f ftft Double ctwck cornet valve upon *t*rt of cycle.

Corre~t valve poaltlon,.

Rework product to tank 50-1.

Certification* of acid concentration before UHe Steam Jet 51-32 to the Plutonium Waste catch Tank 41>-8 or by grav1 ty flow to 40-6.

Shut off !IH-32 and re-work 40-81. send to tank 70-8.

10.

llalooeratlon Plutonium Product Evaporator Feed Taruc 50-4 pverflow.

11. l>lutoni\\lD Product evaporator ~-2 ls allowed

~o overflow.

Table9.l04 Continued MALOPERAnON IN PLUTOUI.M Pl.RIFICATION CELL (PPC)

Result _

Plutonium carried to PR:

&\\lftS> and to Solvent Waste Taruc (130-8) by l~-8 Steam Eductor. ~

and floor are sized to be safe for largest vessel in PR:. Taruc 130-8 has boron Raschig rings and is cri tically safe.

Liquid ls returned to Tank 5D-4; some liquid may carry over through condenser to SD-6.

Indication LR-19 on 50-41 LAH-16 on S\\1Dp*

DR-17 and LIC-20 on ~-2 and_~-10 respectively.

Corr.ct!,,. Action Start Pullp eo-10 to evaporator or atc,p flow into 50-4.

Stop pump ~-101 return SD-6 solution to 5>-4 with tc-41.

12. I Product boil-down to a high concentration in Evaporator ~-2.

Plutonium concentration is too high.

LIC-20, TI-2-31, and DR-17 1watch DR-17 clo1ely1 on ~-2.

dilute with dilute acid *

• in ~-2.

13. I Plutonium Evaporator Con- ~ondensate is carri~d ITI-2-32, LAH-16, and LR-21 densate TanJc &>-6 overflow through overflow vent to on 5D-61 LR-2 on 60-3.

Vessel Off-Gas Catch Tank

~D-3. It can be handled at 1that point as *required.

14. I Insufficient HN03 in SD-4. If.lay form plutonium polymer l5D-4 sample analysis.

in ~-2.

15. IExcess fission products int#on't meet product 150-4 sample.

&>-4.

lspeclf~cations.

Route to rework 70-8 or waste 7D-2.

Add HOO.:.' if polymer ha*

formed In ~-2 and digest until in solution.

Return to 50-1 with P',1111)

~-10 for rework.

0 e

• Table 9.104 Continued MALOPERATION IN PLUTONIUM Pt.lUFICATION CELL (PPC)

Ma1ooera-c1.on ttesu.1~

1no1ca't.1nn 1:n.....-.a... T,v* *.,...,..,,,.,,.

16. Resin degradation ~-lA, Excessive product in 50-2.

50-2 sample.

Replace.resin as

~-lB, ~-1c.

specified ins~.

Operatin~ ~ror.~du~

cove'ring ion -exchange column oper~tion.

17. Improper discharge of Solution directed to wrong LAH-16 on TanJc 50-6.

Ensure that tanJc conten contents of Plutonium tanJc.

are known prior to

,ti Evaporator Condensate Tank discharge.

( 50-6).

I

~

Uranlwn Product BJrlflcattao

9. l~ Uranium Product Purification is discussed in paragraphs 4.67 to 4."72. This discussion in~ludes the operation of the Uranium Product E **~orator, the Silica-Gel Beds, and the storage of ~ranium product. Again*~ ihe malopera,tions discussed concern ptrimarily those associated with.. human errors. These include high and low levels of solutions in tanks, improp~r valving, imp~oper f vaporator operation, and improper sampling.

ln~s9 ~~loperatioos are listed in table 9.105.

9.106 In* the iroduct 1>icka~i.ng.an"l Shipping Area, product pluionium solutfo:1 if. plac~d in small bottles f~r shipment. High-en-richad ur~lum prpduct is also packagPd in this lrea and low-enriched ui;&nium 1s era-sf erred to tank trailers from this area. Due to the pos,ibtlLty, of t:tazard f rora alpha contamlnation in the *Product Packaging antfShippine area; much of the qqulpment is enclc,sed in glove boxes a-nd el-!ctrlc:al 1!\\terlock* controls are in*:i.talled t'a, l:1sure proper operator p~~csdure a~d ventilation cont~\\-t. Mala~~rations in ~he Product F-~ckaging ar1d Shipp,ing Ar~a include any spillaqe of more' tran a minute

~V)tity of plutonlu~ solutl~nt cross conta~inatton of the product3 with

~ny foraign matarial, 1nd any act that could cause a critical arrBy.

All equipment used for hig~-~nricned uranium and plut~nium is geometrically safe~ i~cluding the sumps in the handling area.

Here again, malopera-tions invo.t.*,e-human *errors and are discusgeci in taole 9.106

• 9.107 The rework ~ystcm con~lsts of an evaporator, conden-ser, feed tank, and asso~iated oiping.

The rew~rk system is used to boil down and adjust any o~e of the *a~eous waste streams that ar~ sampl'!d afid found t~ be tQ.Q hi9.h in product ta discard 'before *rerunning tnrough the eolvent extrac;ion syst &m.

OrgJ~ic streams arP. not r~n through the te..,,or!c system.

Tho stre&ms reactilnq tMo rework system will vary f rom very dilute soiutlons approacC1!ng water, to strongec solutions containing up to 6.0 !1 HN03.

Some streams will have only a few grams ~r liter uranium in themt oth~rs will be of o.,c~ higher concentr~tion, and plutonium may also be present, Wh~"' ni9h-enrlched fuel is pro*cessed, these streams may cont~in some of *~he stru~tural alloys of the fuels, such as aluminum and z\\rconium. Thort~m-may al~~ be present during thorium fuel processing.

Since the fee:-J material entering the rework syst~m will vary widel y in compositlol'\\~ each batch w.ll1l rEquir~ special considerat:ion and operating instructioi,v.

Any serious d0vlation from these instructions would con-stitute a maloperation and WQuld result in a product stream unsuitable or less than optimum in ccmpo1!tion for return to the extraction system.

Off-~pecific*atlon feed to the extraction systfim will result in loss of prtiuct a~d/or further rework operations. If so~~tions should be over-cort~entaate(~tn the l'&Wor~ sy~tem, a nuclear criticality incident could resQ1t ~ Prevention of su~n incidents, however, is assured by certain

1.

2.

3.

4

~.

-~~~A~~~~~""t~

Malos,eration High l*vel in Ura.ilium Product Evaporator Feed

• tank ~7.

Table 9,1~

MALOPeRATIOH Of ~IUM PROOLCT PUUflCATION Result

=

Tanlt ov~rflowa to 6D-31 tanlt uy boil,: no. adverse effect unless ignored, for

• long period. *This 'tank 1&

not no~il.Y. ~e~ted.

Indication -

.LAH*l91 UC-26.

Low *~iched uraniU!i'

. -,

• Possible ~gradation of transferred* from !)[)..7 w* !.high enriched uranium Tank lnventory1LAH-241n 5Q-9.

~

-9 *~ mistake.

I* produ~t solytlora.

Ht,gh enriched u~~~ium

~oduc~ misdirected to Lo.

Enriched U~anium Product Evapor~tor !C-4.

High level in product evapol'ators.

Hlgh density ln product

, evapoi:ator.

I Possible to appr~~ch criti-'LJC-27 on ~-4.

cal *concentration* (see iection *@n ~cl$ar Safety).

Poor operatill9, beh&v1_9r of evaporators.

Poss,ible free:e-up of evapor~t.9:r.

Htgh readings on LJC-27 (low enriched uranium) and ur::-28 ( high enriched uranluaa).

High D1r::-20* readingi high TR-1-3 readlng ( low entlct.4

])IC-21 a~d TR-1-1 (high enriched).

6.

Lo. den~1ty product in Low product concentr.stion.

• ev~rator.

-'Low DR:-20 readin91 low TR-1-3.reading ( low enrlclwd high product volume1 DIC-21 and TR-1-4 (hlgh enriched). t J

I I I

I '

I I

l

~

Corractlve Action Shut down *xtr-actloo until exe*** §Qlutlon in ~7 c~n be procaaaed.

~

-, - --~---~~~ -

Lock out hi9h enrlc:btd uranium t~nsfe~airllfta when proceaal~ low enriched uraniua.

When procesalng hlg'n en-riched uranium in_~.._,,

steam will be loclced out of service; alr 0llft sup*

ply will be blank~~*

Re&et control point to loar val~ on UC.

Reset Dir: to lower control point. DR: autoaatically dilutes high density product in evaporator.

Reset Dir: to hl9her control point.

7.

a.

9.

10.

11.

12.

13.

llaloperat*lon.

Too auch*,eoof l~g *ter to

~vaporator dl,charge *11ne*

j1dcet.*

Table <110~ Contlm&ed

~ERATlON OF \\.llANllll PRODtCT P<<lllFICATION Raault Pll.l99lng of evaporator, prod.uet discharge line* lf over concentration of uranlua.

,ndlcatlan High *va~rator levela pl'Oduct volUMI UC-27 (low enrlchacl)a UC-28 (high enriched).

Too little cooling water *1V*.por to vent systems to product evaporator possible urani\\.111 loss to coodens,rs.

vent syst..,.

High -.ter ~rature on Condenser TIC-~ ( low enriched) and TIC*9 (hlgh enrlched)a *high teaperature ln ~~ * ( Tl-2-36 ).

Hlgh level ln Uranium Product Evaporator c~-

densate -Tank ~8.

Low level in &>-8~

Overflow to vent systema loss of uranium.

LAH-20 on.&>-8; UC-29.

Lack of strip feed for firstlLAL-211 F~L-28 on ICX feed.

u?anium cycle.

Co~ctlve Action Wara.pl'Odµc.t

• 1lne wl th 1teu or hot *ter vi*

teaporuy ~ttlon.

Operate eqporator at la.er e,ncentntlon.

Reset TIC to l0wtr temperature.

Stop evaporator, saapl~

• analy,is will detendne dlsposltlon.

Supply.solution fraa c;old make-up area *..

.High leve~ in* Uranium Product Surge TanJc.~9.

Overflow to *~ent system to 60-3.

• LAH-24 in 50-91 LR-2 on '6D-~.Increase rate settl119 of pumps ~-2 and !:G-2A aaJ/

or stop influent flows

• Low level *in ~9.

• ~

Tank runs dryJ no ~ve*r&e results~

LAL-25 in 50-9.

High temperature in S>-9. 1Imp~

/ f~ed temperatun, tolJ'll:>!!ll high on 50-9.

silica.~t Uiltlts..

-*~

._.;./.

Adjust pump rate on 5i-2 or ~-2A at lower Mttlng or stop pump.

Reset nc-111-. ~.. cooling

'water if neces1ary1 chtct I cooling water. to 5':-10.

14. I High feed~&~~ to silica gel beds ~-6A and ~--68..

Excess sole1ti6~ "*t!~

back to 51>-~g *no advert?e effect. I-Hig h reading on PR-9 or 1 Decrease Feed Pullp !G-2, P-R~lO on Head Pots* 5y.29 or ~-2A rate setting.

~-30~

I I

I

~~"~~~..

---'.. **oa,,..~._---~esnttr-~,,. -

_.,.._ -.:..--_~1r~e:-;i:.:

,.*.,.,,.;**.~;ca,;,,:;:*~,........ ~,.-

... a:~)lc-------

~

l~.

I Table 9.10~ Continued, IW.OPERATl<* OF l.RAiUlM PROQJCT Ffji(lfl~Tlc>>f llalfunctlon Result Ltne bl1nd1 ~and,valve1 **1. Poaslble dlver~lon of WJ'On9 on elllca gel unit. produ~t to rework.

Indication Vilvea.are interlocked to insure proper settlng11 visual observation of blind 16,I, Hlgh level* 1n )i,lgh E~- I Ov~rfl<>lfi to Tank !C-138 or I LAH-4 on &>-13A.

rlched Evaporato;- Pl'Oduct,;-13Ca possible rework of SUrge* Tank ~13A.

product.

p

17. (, Off*Ss,.elfication pi-<;duct I Product must be reworke~. I Sample analysis.

in ~13A.

18.

Transfer of 51?-i3A. to I POS$lb1e rework 'Of pro.duct."! Level inventory on tank&a

!D-138 or 50-13C by error*

LAH-5 or ~-138 or LA-6 or before sa_~llng.

50-lX.

19. I High level in High En-r.1ched Uranium Storagt Tank**50-1~.

20. I High level in 5D-13C.

Overflow to 5D-13A and

• I LAH-4 in 5D-13A1 LAH-5, on

• so-1x; possible rework.

50-138 -or UH:..6 on 5D-13C.

Overflow to 5D-13A or l3B; I U.H-6.

possible.rework.

COrrectlv* Action.

Adlainl1tratl ve*. *procedure will 1',l~f'e 'iOJffCt po*1Jlonl"9 :of blincla befoq_elther loading or regenerating alllc* gel beds.

Transfer solution to 50-1381 &hut down Product Evas>9rator ~~~.

Transfer to !C*l3C.

Resample and -rework lf required.

Rttsample 1 'rework 1 f necessary.

Resample all* three tanks

($13A, B, and C) and rework if required.

21.

High temperature in 5D-13A or 138.

• No adverse effect; eductor transfer may be difficult.

High(TI-2-22 on 5D-13A or 1*Check cooling water to T1~2-23 on 50-13B)readlng~a 5E~ll; all~ tanks to inoperable eductors or purip. _cool by ~ir aparglng.

22. l' High levels in Low En-riched Uranium Product S~a.,le Tanks 5D-12A or 128.

Overflow to 60-3 V!ssel Off-I LAH-7 *on 50-12A; *LAH-8 on Gas Condensate Catch Tank; I 50-128.

t possible rework of solution Transfer flcn11rto eapty tank; transfer full tank to 51}-158 after ~sampling1 re sample tanks 'invol ve.d in overflow; rework if neces11

0 Table 9.10!> Continued MAL.OPERATIC* OF lJIANlt.11 PR<D.CT PllUFlCATIC*

---

~*.. *--**

23. High teaperatuns in No adverse results1 posiible High reading on TI-2-2~ on Tum on cooling *ter.

5D-12A 01' 128.

bolling in tank.

!ID-ol2A1 Tl-2-26 on 5D-12B.

24. Failure to sparge tanks Poor sample1 *1ncorrect LR-6 on Tank 5D-12A ahow1 Operator aust follo. pro-

&>-12A or 128 before analysis.

tank was or was not sparged per sapling proceJun1 sampling (See note).

by linetrace1 LR-7 on Tank reaample tanks tum on 50-128.

sparge air 1/2 hour prior NoteaThis maloperation pro cedure applies to all produc~ tanks requiring,ampllng.

to sampling.

2~. Solution put into wrong Possible rework 9f product Level inventory on tanks1 Re&ample tank& as required tank (e.g. 5D-12A instead and resampling.

LR-6 on ~l2A1 LR-7 on of 50-128.

5D-12Bslf too full, LAH-7 or 5D-12A e>r LAH-8 on 50-128.

26. High level in Low-Enriched Overflow to off-specification LAH-9.

Resample and transfer Uranium Storage Tank compartment in 50-1~.

solutions as required.

~l~A.

27. H~gh temperature in No adverse results, possible High TI-2-27 reading.

Tum on cooling water to 5D-15A.

boiling of solution.

coils.

28. High level in Low-Enriched Overflow to 5D-15A off-LAH-11.

Resample and take lnvento J')'l Uranium Storage Tank specification compartment, rework as required.

50-1~.

resampllng requlredJ posslble rework.

29. Low level in 50-158.

No adverse results, auto-When Pump ~-6 stops, low Walt for more solution to matlc shutdown of product lev&l is recorded.

accumulate or transfer ao 1ft pump $"."6.

solution from 50-12A or B

30. High level in Low-Enriched More solution than required LI-141 scale weight.

Transfer to next tank true :k.

Uranium Product Weight to f111i tank trailer.

Tank 5V-l.

31. Degradation of silica gel. Failure to remove zirconium Sample analysis of 50-12A Rework if necessary a re-and niobium from uranium and 50-128.

generate or replace silica product.

gel beds.

I

Table9.106 MAI.OPERATION OF Tl£ PRODU::T PACKAGIOO AA.O SHIPPING AREA (PPS) llaloperatlon Results Indication

1.

Plutonium product bottle overfilled or fill line miaconnected.

Plutonium solution will spill in bottle filling station sump *.

Visual observation* of spil*

lage and/or measuring pot to a volume > 10 11 ters.

2.

Enriched uranium product bottle overfilled or im-proper connection of fill head.

Spillage of enriched uranlu; Visual observation of in fill area.

spillage.

3. t Two filled product bottles I May approach a critic al I.Radiation alarm.

(plutonium or enriched array {See section on uranium) brought together. Nuclear Safety.)

4. ICross-over valves opened between plutonium and uraniun spill transfer lines.

Cross contamination of products. Plutonium alpha 1 contamination of uranium product fill area.

Radiation alarm1 product analysis.

5. I Glove rupture.

Personnel hand contaminat~ Visual.

Corrective Action Close valves on filling heads move to drip con*

tainer1 transfer spill to 50-4 with eductor1 clean up area w1 th de*

contamination washe1.

Close fill valves and transfer spill to En-riched Uranium Product Stor~e Tank 50-138.

All equipment and oper-ating procedures have been designed to prevent more than one uncaged product bottle from being in the area at one tl**

If the radiation alarm sounds, area will be evacuated at once

• These valves will be lo~ed at all times un-less a $pecial situation requires their use1 material returned to rework.

Negative pressure on glove

• box will prevent leakage of contamination from box.

Hand decontaminatio.n as prescribed by Health and Safety. Replace glove *

6.

Table 9.106 Continued IW.OPERATION Of TIE PROOOOT PACKAGit<<; AND SHIPPING AREA (PPS)

~

Improperly stoppered product bottle.

Gross contamination of the birdcage.

Monitoring prior to shipment.

Immediate evacuation of area1 decontae1nation of birdcage \\Wier Health and Safety auperviaion1 de-contam.irwtion procedure, as prescribed by Health and Safety.

procedur&s and administrative controls, which are discussed in the chapter on Nuclear Safety. The rework system receives solutions into the Rework Evaporation Feed Tank (70-8) from seven other collection tanks (40-2, 40-8; 40-10,.40-13, 60-3, 70-10, and 130-8). The solutions in these seven tanks have been sampled; hence, they are of known composition. These analyses determine which solutions shall be routed to Tank 70-8.

On the basis of these known co, positions and extraction feed requirements, a procedure is followed for reworking 9ach batch. If soluble neutron poison is required, as determined by analysis, it is added from Tank 140-32 in the cold solu-tion area. Acid, or other reagents, could be added from the same tank, if required. Solution in Tank 70-8 is air-sparged and transferred by steam-jet eductors to the R!work Evaporator 70-4; Because of criticality considerations, Evaporator 70-4 is operated on a batch basis only. A

.low-ievel alarm anJ control prevent,overconcentration in the evaporator.

Condensate from the rework e*:aporator flows to the Low* Level Waste Evaporator feed Tank. This material will be very dilute acid with some activity in it. the bottoms

• from the rework evaporator are transferred by steam transfer eductor to the Partition Cycle feed Tank (30-1).

Solutions may also be transferred to the Low-Level Waste Accountability and Neutralizer Tank.

Various possible maloperations of the Rework Evaporator sy,tem, togeiher with possible *consequences, alarms, indica-tions, and corrective measures,are listed in table 9.107.

Higb-Levei wa,te Evaporator' System 9.108 A schematic representation of the High-Level Waste Evaporator Syst~m is shown in figure 4.81. the system consists of Tank 7D-l (th~ High-Level Waste Evaporator feed* Tanij, 7C-l (the Low-Level Waste Evaporate~, 7E-5 (the High-Level Waste Evaporator Con-dense~,. and 70-4 (the High-Level Waste Accountability and Neutralizer Tank). the system is fabricated of stainless steel except for the heat transfer tube bundle, which is made of titanium. Wast& from the parti-tion cycle or from the rework evaporator can be transferred to the High-Level Waste Evaporator Feed ~ank. Solution may be t~ansferred to the

• evaporator by air lift or by 1et. It is intended that the ovaporator bottoms be operated at an acid concentration no grea~er than 8.M HN03*

Maloperutlons in this system are detdlled in table 9.108. The evaporator* bottoms, after analys,is, are either neutralized with caustic and pumped to the tank f~rm for storage or are recycled, if necessary.

Jaer-u:vei Jaste Evaporator Sys\\\\e'D 9.109 A schematic representation of the Low-Level W~&te Evaporator System is shown in figure 4.83. The system consists of Tank 7D-2 (the L<lw-Level Waste Evaporator \\Feed ranij, -X:-2 (the Low-Level Waste Evaporato~, 7E-7 (the Low-Level Waste Evaporator Condense~; and 70-10 (the Low-Level Waste Accountability' an*d Neutralizer Tank.

the system ls fabricated of stainless stee~ except for the heat transfer tube bundle which is made of titanium. The lo#*Level Evaporator System evaporates the overheads from the High-Level Evaporator System, ~he aqueour waste streams from all of the solvent extraction steps except

Table 9.107 IIALOPERATION SlMIARY OF REWmK EVAPCIIA'l'm SYSTEM llalo~ration 1.1 Tranafer of too mch solution from any one of Nven feed* to 7D-8.

Result O\\;erflow of 70-8 to Tank 40-10 and 40-13. one of the seven fffds _to 70-8.

Indication LAH-7 on 70-81 PAH-3.

2.1 Transfer of insufficient 170-8 will run dry when solution to 70-8.

evaporator ls running.

LAL-8 on 70-8.

3.1 Transfer of 0?1Janic to 70-8 (*~ note).

Organic will flow to 7C-41 J DI1 low densities in aeven

• some solvent degradation ma feeds (70-8 and 7C-4).

occur.

Cornctiw A~tlon Tum off ate* to tranl*

fer educt "'ft causing ** ce..

transfer.

Shut down of Evapontor 7C*4 11 autoaailca supply more solution t.o 70-8 lf available.

Jet out organic t.o waste tank aft.er using apecul wash solution to 1trip out any product froa the organic.

!!2Y,1 Hydraulics of equiPlnent and piping is designed 1 prevent -O%ganic reaching anL one of tlae Nven feed to 7D-8. Concentration of adl.d used !n 7C-4 will not cau:C serious solvent nitration.

4.1 Trander of wrong tank to I Boll cl-. of wro_ng 1olut1on1 Sample analyslsJ tank level I Revhe boU..i-n and.-zit 7D-8.

incorrect rework product inventory will show wrong procedure to compenaat.e composition.

tank has been transferred.

for different solution.

5. I Insufficient air sparge t1 Pcx;r mixing in 70-81 this i tr::-5 off1 LR on 70-8 will I Tum on tr::-5.

70-8.

only serious when widely d draw thin. even line.

I fering compositions from twci or more tanks are mixed or neutron poison is required.

6. I Failure to add neutron I LIC-4 and LAL-9 on 1C-4 will Level inventory on 7D-B and IAGllnbtratlve procadun poison to 70-8 when provide control and alarm t 140-321 also administrative designed to provide two required.

prevent possible critical checlc and data sheet.

or more independent chtck*

condition.

I in such caae1 to insure poison has been added. SN section on Nuclear Safety.

~

Taille 9.107 Continued IIALOFERATIC:>>t StlUMRY OF REW<llK EVAPCRATCR SYSTEM luloperation Real!lt Indlcatlon

&_.;::th,a Actlftll 7~. **ugh level in 7C-4.

Solution overflows to 31>-l LAH tn 30-1.

Level control ln ?C**

before it ta properly Ht too hlgb or ate*

adjusted.

1upply too low -

n.. t.

~

a.

Level too low in 7C-4.

Concentration too high or LAL-9.

Dilute a... ne~aaary.

steaa coll _not covered, resulting in low capacity.

9.

Air 1parger on 7C*4 turned Ne adverse resul tsJ boiling None other ~n position of None nece11ary.

off.

will supply ample mixing.

air valve.

10. Cooling water to Condenser Water and nltrlc acid vapor Hlgh temperature on TR-l-81 Increa&e water.flow to.

7E*8 too low or turned off in Vessel Vent System.

high temperature in Ve1&el 7E*8.

entirely.

Vent Systems excess conden-sate in Vessel Vent System

• I

~ *~----....... ~.... ~jlillaal------~--

Table 9.108 IW.OPERATION OF HIGH-LEVEL WASTE EVAPalATtll FEED TANK (71).1) 111111-

- ~ 1ftft u....,,11:

TrvH -"*+1 Aft

1.

Feed lift to 7C*l working 1C-l level rises due to in-LIC-6 in 7C-l.

too fast.

creased now.

2.

Feed lift to 7C*l working 1C-l level c:uops due to LR:-6 in 7C-l.

too slowly.

decreased flow.

3.

Liquid level too* low in Liquid level will drop in LAL-!>.

70-1.

evaporator 1C-1.*

4.

Liquid level too high in If alarm is ignored and no LAH*4.

70-1.

action taken, will overflow back to 40-2, the Partition Cycl~ Waste Hold TanJc.

~-

Rate jet falls or is 7C-l level drops.

uc-6.

plugged.

6.

Transfer of organi~ when Organic will d~composv in Sample of tank 7~4.

using rate jet 7H-4.

7C-l with subsequent trans-fer to underground tanJc.

1 l

I l '

r.n

- * *- a....., """

Automatic cutbtck of air flow.

Automatic increa.. ot air flow.

Shftt off alr lift. saw,t

,r down steam to evaporate 7C*l to avoid overcocce n-.

tration.

Shut off steam jet fr 40-21 increase steaa and air jets in 70-1, re-moving liquid as much necessary (both event ally feed 7C-l).

s Use air lift as alter-nate transfer mechani...

No explosion will occ because steam is 11111 ted to 25 psig1 the tempe ature ls then below flash point organlc1 when using 7H-4t tank 70-1 will not be coari,t*

ely emptied.

Table 9.108 Continued MAI.OPERATION OF HIGH-LEVEL WASTE EVA~TCE (7C-l)

-...w*-- 6 V

...I. V'* *

• • v-11...-*.,

......... '"'...... v,,

1.

Not enough steam in heat-Insufficient amount of TR-1-6 in 7C-l I UC-6 in 70- l ing coils.

liquid boiled off1 liquid LR-9 in 70-41 DR-5 decrease.

level rises, which auto-snatically shuts off air lift from high-level waste evap-orator feed tank (70-1)1 level in high-leve~ waste accountability and neutral-izer tank (70-4) rises by solution traveling through cverflow pipeline from 7C-l to 70-4 if on continuous processing.

2 *. Too mch,team in heating Too much liquid boiled off1 LRC-6, TR-1-6 in 7C-11 m-5.

coila.

liquid level drops1 liquid temperature rises1 density rises; level recorder con-trol will automatically draw more feed from High-Level Waste Evaporator Feed Tank

\\70-1) *

3.

Liquid level too high in Level recorder control will IJC-61 check flow from 70-6.

7C-l.

automatically reduce flow from High-level Waste Evap-orator Feed Tank (7D-l).

I I

r f

f I

I i I

I

~V.&.A'liJ,-* * *-

Increase 1teaa flow through coils with FIC-3. Stop steam jet from 70-1.

Turn down steam to coil

• with FIC-31 use batch transfer jet 7H-4 if necessary.

t te ng