Text

Part B of CSF-1, Application (Selected Parts as Related to the Radiation Safety Program - Part 1)
	ML18081A222
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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;
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(l / I / Before The I UNITED STATES ATOMIC ENERGY COMMISSION Washington, D. C. In the Matter of the Application of NUCLEAR FUEL SERVICES, INC. For Construction Permit and Licenses for . a Spent Fuel Processing Plant Under Sections 53, 63, 81, 104 (b), and 185 of the Atomic Energy Act AEC Docket No. 50-ZOl Part B --Safety Analysis Amendment No. I October lZ, 1962 . \ [ 97R7 I *1--Part B --Safety Analysis Amendment No, I October 12, 1962 {3n July Z6, 1962, Nuclear Fuel Services filed with the Division of Licensing and Regulation an application to build and operate a fuel reprocessing plant for spent reactor fuel. A two-volume Safety Analysis was submitted as Par_t B of this application.

During the past two months there have been several design changes. These will be reflected in a series of amended pages to the Safety Analysis.

The amended pages will be submitted as soon as they can be prepared and the required copies printed. These changes are summarized as follows:]

[!_> Elimination of Thorex EquipmenfJ The expected load of ThOz-UOz fuel is less than originally templated so that it does not now appear appropriate to include the expensive facilities needed to provide a processing capability for this type of fuel alent to that of the uo 2 fuels. " Therefore, the plant capacity for Th0 2-*.uo 2 fuels will be 500 kg/day in place of the 1000 kg/day described in the Safety Analysis.

Also, the facilities for decontamination of the recovered thorium have been eliminated.

Thorium will be permitted to go into the high level waste stream and will be stored in stainless steel tankage along with the fission pr od*ucts from this particular fuel. Stainless steel tankage will be provided for t his purpose as required.

E Reduced Capacitl £or Stainless Steel-Cermet Fueli) The Safety Analysis* indicates the inclusion of a Darex facility capable of handling 225 kg/day of stainless steel, Th e capacity of this unit has been reduced to 125 kg/day and the capital aU ovtan b~'.t i nc H\i c lE F th c ed sl-.. : of this capabil i ty. If development work on electrolytic dissolution at SRP continues to look favorable prior to freezing of design, NFS may change the design to provide for electrolytic dissolution at a processing rate of 125 kg/ day --------

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

___ _ of s*tainless steel. Allowance has been made in cost estimates to permit the inclusion of either (not both) s eta of equipment.

t Waste Storase Facilities)

Facilities for storing neutralized wastes which will be installed at the outset will include two 750, 000-gallon carbon steel and concrete tanks. One of the tanks to be installed will be used for storage of the neutralized wastes from the entire processing sequence.

The second will be held as a spare. Other waste storage shown in the Safety Analysis will not be built at the outset. Stainless steel tankage will be provided by NFS as required, but funds will not be committed until receipt of firm commitments for the cessing of fuels whose wastes require this type of storage (e.g., stainless steel-uo 2 cermets, Th0z_ uo 2 , depleted uranium-molybdenum, aluminum, or uranium-zirconium alloys). The depreciation schedules for waste storage are sufficient to provide a revolving fund for new mild steel. waste storage as required and contractual commitments will permit tional capital as needed. E Removal of *One Dissolverj As a result of the reduced requirements on the plant for the ThOz-UOz fuels, one of the three dissolvers was removed. The remaining plant facilities including fuel receiving and storage, mechanical cell, extraction equipment, acid recovery, solvent handling, prod--2-__________________

___ ...... ______ _ r --

1l uct packaging and handling, utilities, maintenance, and ana.lytical facilities remain eaeentially unchanged as described ln the Safety Analysis.

These changes have already been communicated to the USAEC in a letter to Mr. R. C. Blair, dated September Zl, l 96Z. lett~r dated *September 5, l 96~, the Division of Licensing and Regulation of the USAEC raised eleven questions formal answers to which are hereby submitted)

1) Question:

The application does not give the types and maximum quantities

/ . of radioactive material that will be present in each processing step and in the storage areas during normal operation of the plant. This information is necessary to determi11e the probable releases that could be anticipated in the event of an accident.

Answer: In Table A-I-1 there is given a tabulation of the quantities of radioactive materials expected in each process stream. Data are given for fissionable materials, fertile materials, total fission products and specific fission products.

The data presented are representative o(. th~ most active fuel which we expect to process in the plant. These data are for a fuel burned to ZO, 000 mwd/ta*n, Z7. 5 mw/ton, irradiated two years at*85o/o load factor and cooled 150 days. The maximum quantity of fuel which can be stored in the storage pool is about 1000 fuel elements.

It is*expected that normally no more than ZS% of the pool will be full. The amount of activity stored therein can only be estimated since it will depend upon the past history of the particular fuels delivered to the plant. An estimate of the amount of fuel normally in storage is given in Table A-I-2. The amount and type of activity in waste storage will change with time, increasing as the tank is filled and at the same time decreasing due to decay. The maximum amount of activity in storage in a given tank will be present just at the completion of the filling period. An estimat,r

of the total quantity of waste and the major specific fission product contributors at the time the first waste tank is filled is given in Table A-I-3. 2) Question:

It does not appear that a "mock-up" shop will be included in the plant. Existing Commission facilities which utilize "remote maintenance" have "mock-up" shops and these shops have reduced exposure considerably during maintenance.

How does NFS plan to minimize employee exposure during maintenance operations without such a &hop? Answer: The NFS plant does not have a full-scale "mock-up" shop complete with crane of the same type as that used at Hanford and Savannah River. However, the plant does have the capability for carrying out the function of the 1'rr1ock-up" shop, viz, to check out the exact dimensions of a piece of equipment to be installed in the chemical processing cell. Jigs are provided in the maintenance shop to set up an equipment piece going into the CPC and to check all of the pertinent dimensions against the known-require-ments. The need for this type of facility is less in the NFS plant than in the AEC production plants since there are only 15 pieces of equipment in the CPC which are handled in this manner and the proposed procedures will adequately allow the checking of such a small number of equipment pieces. Further, it is felt that this is not a safety problem at all ~ut rather an economic one. The installation of equipment in the CPC is done remotely and does not involve the exposure of personnel in excess of the normal plant ( *' background.

The penalty for making an error in the measurement of dimensions (resulting in failure to be able to install the equipment) is not taken in increased personnel radiation exposure but rather in lost production ( C C __ ...... __ ..... ..,. .. ____ .. .,,. time. If it were not possible to install the equipment piece, it would be necessary to move it out into the decontamination area,clcan it up to the point where it could be worked on again, and make the necessary corrections

.... 3) Question:

The chopping technique for the variety of fuels that will be cessed has not been employed before in spent fuel processing plants and the probability of widespread contamination in the cell, as anticipated by NFS,

does necessit..\te "remote maintenance.

11 How is this remote maintenance going to be accomplished on large items such as cell windows and the out ram? Answer: The chopping technique has not been in ro1,tine plant operation, .. but it has been operated on full plant scale at Oak Ridge National Laboratory by the Chemical Technology Division.

The amount of "dusting" is estimated to be about 1 per cent. To confine this and any fines produced, the chopper will be separately sealed and washed down at t he end of a cycle. Nonetheless, the cell is expected to be highly contaminated and the specifications for every equipment piece installed in the cell require that it be capable of remote maintenance for most repairs, and that it be removable either in pieces or in toto if that is required.

The windows are removable from the outside of the cell. Since the push-out r l'\m was questioned specifically, a step-by-step method for its removal is given below: 1) The drive system, control, and power connections are located outside the cell a!ld are not contaminated.

These are first removed. 2) All the supporting connections outside the cell are then removed after which the supporting connections inside the cell are removed or loosened by use of the ,na.nipulator.

-3) A wall section of the operating aisle opposite to the PMC must be removed to allow complete removal of the unit. 4) Around the ram drive mechanism there is some unit sbielding.

This is removed from outside the cell. A partial radiation check can then . . be made of the unit. Decontamination of the unit will undoubtedly be quired. This is done using the remote manipulators inside the cell. tional radiation checks are made during the course of the decontamination.

5) Roller units are set up in the operating aisle to allow the unit to be brought out horizontally.

A special dolly is brought up outside the building to support the unit as it is being removed and to carry it to the maintenance shop. 6) A sling is attached to the crane in9ide the cell and the sling is secured to the end of the ram housing. 'l1le housing is raised sufficiently by the crane to take the weight off the supports.

7) The rollers are also raised until the ram housing no longer con-tacts the supports.

The ram unit is now free to move. 8) A cable is attached to the ram in the operating aisle, threaded through the opening in the outside wall and attached to a winch located outside so that a straight line pull on the ram unit is possible.

9) The winch is actuated and the ram slowly pulled out of. the cell. The crane inside the cell must be moved in parallel with the ram movement.

Coni\tant radiation monitoring is carried out during this operation to tain that the decontamination has been sufficient.

It may be n~cessary to .. stop and do additional decontamination.

Additional decontamination may be done either in the operating aisle, with or without the use of portable shielding, or it may be done inside the cell by partially_ reversing the removal operation.

10) As the unit emerges it is placed on the dolly and then transported to the maintenance shop. The opening in the PMC wall is temporarily shielded, Q 4) Question:

In paragraph

4. 11 the return of the fuel transfer basket from the Process Mechanical Cell (PMC) to the Fuel Stor:\ge Pool Complex is described, The transfer of contamination to the fuel pool is not discussed during this step. What evaluation has been made to determine the quantity of contamination that will be transferred to the pool? How well will the ion exchange equipment (see paragrap l =.. 4) remove the contamination?

Since only a single circuit of cooler, filter and ion exchange equip:a.nent (see Figure 5, 4) is going to be used, how will cooling be accomplished out the pool while decontamination of only one compartment of the pool (see paragraph

5. 5) is in progress?

Answer: Revaluation of the contamination problem has led NFS to decide G on a slightly different approach to the problem of transferring the fuel ments into the PMC from the FRS. The fuel storag~ baskets will be sized L so that the elements will protrude a short distance above the top of the basket. The basket will be affixed to the underwater transfer conveyor as indicated in paragraph

4. 8 and brought into position below the hatch. The crane will then be used to pick up the fuel element from the basket leaving the basket in place in the underwater transfer conveyor.

The basket is returned to the fuel storage pool never having gone into the PMC where it might become contaminated.

Paragraphs

4. 8 and 4. 10 are being rewritten to reflect these changes and new pages will be submitted.

It would appear that the remainder of the questions in this section were prompted by the contamination possibility created by the return to the pool of baskets which had been in the PMC and, therefore, they may be considered to be answered by the above change. There is, of course, some t=_* ------

-__ ___;.;.._...-

_____ _ chance of pool water contamination from other l:3ources such as the storage of the elements.

This is not a problem which is unique to this plant, ever, and the proposed method for handling the cooling and the decontamina-tion of the pool water has proved effective in many previous installations.

There is no justification for duplicating the ion exchange equipment.

No contamination problem in a fuel pool is so acute or so sudden that the cleanup facility has to be in constant operation.

The same is true of the cooling system. The fuel pool is a tremendous heat sink. At the design rate of heat release, 600, 000 Btu/hl"--* a number which we believe to be conservative, equivalent to the heat from the ZOO-day-cooled output of 6000 mwt of reactors--the temperature of the fuel pool would increase only O. 1 F per hour. Thus it does not appear to be necessary to provide a spare cooling loop. 5) Question:

In reference to the proposed ventilation system what evaluations indicate that (1) a manifolded and butterfly valve controlled system on the exhaust of the blowers (see Figure 6. 3) will prevent blow back in the event of a blower failure and (Z) that a "minimum" 100 £pm face velocity thru openings will prevent backmixing from active to less active areas. (see paragraph

6. 9) Answer: 1) A det a iled answer i s being prepared and will be submitted.

Z) The 100 ft/min figure has been generally used throughout AEC installations for design purposes.

6) Question:

According to Table z. Ila and z. llc freezing conditions and snowfall can be expected from October to April. Under these conditions, will operational and contamination problems occur in outdoor areas, such as the washdown area (see paragraph

3. 6 and 4. 3 ), the general purpose evaporator (see paragraph

4. 97 and 5. 43 ), the concrete lined burial bins for high level solid waste (see paragraph

4. 98 ), the low level trash burial area (see graph 4. 98 ), the water seals on the thorium and depleted uranium storage -' . -----I 1 I ,

~-..................

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.. .... tanks (see Figure 4. 96) and the diversion box and outdoor sampling points (aee Figure 6. 23e)? Answer: ( 1) Operational problems can be expected to occur in an outdoor waahdown area, Therefore, this concept has been abandoned.

All washing and decontamination will now be done inside the building.

(2) The general purpose evaporator has also been moved int.o the building.

(3) It is no longer planned to use the concrete-lined burial vaults. These were for the storage of metal hulls which have been leached in boiling nitric acid and thoroughly rinsed with water. Activity which is not removed by this treatment is not likely to be removed by water at a pH of near neutrality.

We propose to bury these directly in the silty till, a tion which is not a aquifer and in which the rate of movement of any water that should get into it is essentially ze~q. (4) Low level trash burial is an operation which can be halted in extremely inclement weather. Burial ditches will be dug somewhat ahead of time and sloped to one end so that rainwater collecting in the open trench will collect away from the point at which burial is taking place*. If water does collect in the end of the trench, it will either be pumped out prior to the use of the end of the trench or the end will be back-filled without making use of it for low level burial. (5) The seals called for are liquid seals. During winter months, the liquid will be of a non-freezing type. (6) The diversion box function has also been taken back into the building.

(7) The only outdoor sampling points are on the waste tanks themselves and into the annular space around the tanks. These will be a -----------------

, ¥ ,. ** .. ii$ simple thief-type sampler which will operate in inclement weather. Appropriate char.gee have been made in the Safety Analysis to re-fleet the above and the revised pages will be submitted.

7} Questlon:

Where and in what quantity will cell penetrations be provided for the future anticipated requirements as indicated by Figures 3. l 9(a), 3. l 9(b), 3. 22, 4. 21 (a), 4. 33 (c), and 4. 39(a)? Answer: There are provided in the contact cells 10% spares for ~.all penetrations and there are 20% spares in the remote cells. These tions take the form of stainless steel pipe or tubing so arranged that there is no leakage of radiation in excess of the design shielding for the wall in which they are installed.

Both ends are sealed by welding. ConsequentJ-y, they represent no safety hazard. Their exact locations will be shown on the final drawings.

8) Question:

Clothing, monitoring and change facilities are discussed in paragraphs

8. 14 to 8. 17 but this discussion does not define the boundaries of the controlled zones in the plant, the traffic pattern in the controlled zones or the type and size of facilities provided at each boundary to prevent carry over of contamination from one zone to the next. A discussion of these point& for the proposed plant layout will be necessary.

Answer: Drawings are being prepared, coded to show the five kinds of plant areas from a contamination control standpoint.

These are: a) unrestricted access b) access when wearing plant clothing and shoes c) access when wearing plant clothing, shoes, and special shoe covers d) r.io access at all except after thorough decontamination, health physics surveys, special clothing, and shoe change l I I I I* .1 I e) a few limited areas in which either (a) or (b) is permitted.

Persons will enter the plant only through the main entrance.

,,. .. They will have free acce11 to the (a) areas without changing clothing 01~ 1boe1 and may al10 go into (e) areas. In the case of (e) areas visitors will not be permitted unle1s accompanied by plant per1onnel.

The workers will change clothes and shoes in the locker rooms after which they will have acce** to the (b) areas. At the interface between all (b) and (c) areas there . will be shoe cover rack* and ~ere will be a change of footgear at every cros aing of the1e interfaces.

The (d) areas will not be entered at all except under full health safety coverage and there will be clothes and shoe change areas set up at the point of entry. 9) Question:

What are the average ar.d maximum discharge concentrations and flow rates of 1129 and 1 131 that will be exhausted through the stack? What C. total quantity of these materials will be released per year? C -i --Answer: In v~ew of the implications of question 11 (see below) the requested data are provided not only for the iodines but also for krypton 85. In Table A-1-4 there are presented data represe11ta.tive of the discharges which are expected from average fuel during the fi'rst few years of operation.

The discharges of these same isotopes from a fuel repl"esentative of the highest burnups which we contemplate processing in the NFS plant were shown as a part of Table* A-1-1 (see answer to question 1) and are repeated here for convenience:

Kr I I Xe Xe 85 1Z9 131 131m 133 -ll .. 8. ZS x 103 curies O. 017 curies 1. 7 curies 1. 1 curies 5 x 10-3 curies

] <<r I It bad been our intention to die cuss this type of fuel when the time came to develop technical 1pecific;atione.

In view of the interest shown by the ACRS at the meeting held on October 5, 1962, we have decided to redo the calculations of Sections VII and vm using the above maximum numbers. Revised pages for these section* will be submitted as soon a* po11ible.

10) Ques~on: It is c.uggested in paragraph

7. 12 that the 1torage lagoon will be uaed a* an emergency holdup area for the overhead*

from the general purpo*e evaporator.

What maximum concentration of activity in theae heads would be diacharged to the a to rage lagoon? Do any other 1tream1 feed thi1 lagoon? U 10, what type and concentration, of activitiee wilt' be in these stream*? How operable will the lagoon be during winter weather condition*

1* Answer: There appears to be some mi1under1tanding concerning the function of the storage lagoon. It i* not our intention to operate thi* as a seepage basin. That is, it is not the intention to routinely percolate wastes out into the stream through the ground using the ion exchange capacity of for additional decontamination.

All the liquid discharges will pass through the storage lagoon. They will be monitored and a record kept of the volume and activity which has been discarded.

We expect to discard liquid at this point sucli that 10 CFR Part 20 will be met in Buttermilk Creek on a gross count basis assuming the absence of radium (1 x 10*7 uc/ cc). The average dilution factor available in Buttermilk Creek has been calculated to be 2. 7 x 103. Using a dilution of 103 would imply that the discharge from the storage lagoon could be as high as 10-4 uc/ cc. If the storage lagoon discharge were to become higher than that required to meet 10 CFR Part 20 in Buttermilk Creek the overflow would be ~--------

C L *topped and the di1charge held up until the condition causing the higher activity level had been corrected or until 1pecific fia 1ion prociuct analy1e1 could *how that the waste could be aatiafactorily di'achared within the limit* ol 10 CFR Part 20. It i* then the function of the lagoon to hold up the activity.

In dry warm weather 1ome aeepage into the ground can be expected and for the amount which doe* 10 aeep advantage would be taken of the ion exchange capacity <<>f the ground. Thia would be an abnormal and not a routine tion, however, In cold weather it ia conceivable that the volume held up would freeze. Since the ice could be expected to ~w a 1:.1 in place the atorage lag(),)n could then be 1aid to be carrying out its function very well. The proce11 streams which go to the storage lagoon and their ex-pected activity levels are a1 follows: 1) Overhead, from acid fractionation----1 O-5 uc/ cc 2) Overhead*

from General Purpose Evap----10-6 uc/cc 3) Floor drain* from non-contaminated area a of the plant-----------------------------

zero to 10-6 uc/ cc Miscellaneous wastes such as laundry waates and laboratory wastes can go to the storage lagoon but they do so by way of the General Purpose Evap Feed Tank. If they prove to be low enough ao that evaporation is not necessary they can be then routed to the storage lagoon. We expect to set an operating limit for any waste discharged to the storage lagoon at about 10-2 uc/ cc.

We expect not to discha t*ge a~ything from the storage lagoon so that Buttermilk Creek could become in excess of 10 CFR Part 20.

This is approximately the limit used at Savannah River also.

-* 11) Queetlon: Since at1nlftcant quantltlea of Kr 85 and 1 131 (aee para1raph

1. 6 and Table 1. 1) will be diachar1ed to the atmoaphere, with what degree of certainty can we be aaaured that the generalised parameter*

uaed are conalatent with the actual alte metereology?. Anawer: Before operation baa atarte~ we expect to have collected on-alte meteorolopcal data for at leaat a year and to have aome evidential aupport for the meteorological parameter*

uaed. In the meantime the data uaed in making all calculation*

have been deduced by Dr. Maynard Smith from a atudy *of the alte. He waa aaked to aupply a "conaervative" aet of data. At a meetln1 of the ACRS aubcommlttee held in Buffalo on September 12, ~962, Dr. Smith defined hia concept of the degree of aervativiam aa followa: The parameter*

were selected to reflect tion* which are per}lapa aa much aa three timea b~ter than the worst po11ible condition and about 1000 timea worae than the beat conditions. Subacribed

.and sworn to before me thi1 /,,k day of~ I 1962.<~~-c.;!!~ My Commi1slon expireaH'/(/f 7 ........... Reapectfully submitted, Nuclear Fuel Service,, Inc. Preaident

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t ' '--.. :..:_.. _.-; ... Tbif vil1 r efer to ~ur l e~te r of Oc~oo e: 15 su~witt1~g A m eodm eo t No. l. Certa l o dr awings referred to in this Ame nd~ent ~ere no t in cl uded since they had not yet ~e*-comple te d. We are u erev ith ~ranamit ting 52 copi es ol' th ese d~awings ~hlc h should oe added to Ameod..me n~ No. l. BGB:dJt Eocl oo ures Very Lruly yourG, ,., .. , / ,J .. ... r* r -..

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I I I ' i I I I (j L VII PROTECT!~

CF THE PUBLIC Surrmary 1*.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 los~ from the process complex at the following points* 1. 2. 3. 4. 5. 6. Stack Waste storage tanks Storage lagoon Burial ground Egress of personnel and material Product shipment In subsequent paragraphs, each of the above possibilities is analyzed to show that the statement of Paragraph 7.1 is valid. Some 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 most-serious 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; anrl further that steps can be taken to assure that, even in the event of such an accident, the discharges to surface waterways at the site boundary can be kept within the limits specified in 10 CFR Part 20 through the use of reasonable correction measures after the accident or release has occurred.

7.4 ThP. following abnormal events have been postulated1

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 siugle bu~at or a multiple continuing event totalling 10 fissions.

Revision 1, Aug. 20, 1964 ..

---* **--.... ----*--*------*--*-*---

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3. A criticality incident in the fuel storage pool which sets up a 10-nMt boiling water reactor which operated for as long as 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months 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.

5. The r.omplete failure of the iodine 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 plant, the stack, and the tank farm events 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~= the above possibilities is analyzed.

7.5 Throughout this section, a number of assumptions recur. Valu~s for such recurring assumptions are collected in Table 7.5. Assumptions specifically related to a icular calculation are included in the calculation.

Normal Oper~tions Stack 7.6 As explained in Paragraphs 6.3 through 6.21, the vent.ilation systems are designed to assure that, under normal operating conditions, flow of air is always from areas of least contam i nation 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 stack. The total volume of air discharged is 32,000 cfm. Iodine removal facilities are dsigned 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 comparison to the gaseous activity.

Calculations are based on an average fuel which we may expect to process in this plant rep~esented by the followin g parameterss Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964 I I l I I l I I j l I l \ I I 1 I C L Table 7.5 Assumptions Used in Calculations Sections VII & VIII 1. The dispersion 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 105 meters and for meteorological conditions ranging from "extremely unstable" to "mccfarately 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 to represent inversion conditions.

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

The following wind distribution data has been useds Wind Distribution (Per Cent Per Octant) Wind Direction Suniner Winter Averag§. N 8% 8.5% NE 4 2 3 E 5 2 3.5 SE 17 9 13 s 23 21 22 SN 13 25 19 w 9 12 10.5 NW 20 21 20.5 Fuel is cooled 150 days before processing.

High-level waste is stored at 410 gallons per ton which is equivalent tos 132 c/gal 166 c/gal 57 c/gal at the time of storage. Sr-90 Cs-137 Ru-106 Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964 I ll 1--------.. .._.,._-______ :...--, _ ........ _ .........

-

' --* LL, *. * *~t * *** -,.,_ .... , ....... *LO ' ** ---------------:,:..*--------~---

Table 7 .5 (Cont* d)

4. 5. 6. 7. a. 9. 10. The rate of travel in the surficial till is 1.0 foot/day.

The rate of travel in the silty till is 5 x 10-5 foot/day.

90% of Sr-90 is associated with sludge in ~he tank. 99.~ of Sr-90 is adsorbed on so°il on passage through it. 99.99% 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 stac~, 10% to waste tanks, 65% to steam. For long-lived isotopes the fission products are taken as 7r:JI, from u235 -3r:JI, from Pu239. For lived isotop,s they are taken as 60% from Pu239 -40% from u235.

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

(_ ... Burnup Specific Power Irradiation Time Load Factor Cooling Time 20,000 mNd/ton 32 mN/ton 2 years 85 per cent 150 days Using these parametex*s the input activity to the plant was calculated.

The gaseous activity*

input iss Kr-85 I-129 I-131 Xe-13lm Xe-133 Tritium 6.3 x 103 curies 0.022 curie 1.8 curies 1.0 curie~ 3.8 x 10-curie 50 curies Under the conditions stated above, the total daily discharge from the stack using the average activity level fuel contemplated will bes Kr-85 I-129 I-131 Xe-13lm Xe-133 Tritium 6.3 x 103 curies 1.1 x 10-4 curie 9.0 x 10-3 curie 1.0 curi!3 3.8 x .10 curie 50 curies 7.7 The concentrations of each of -these isotopes at various distances and under various meteorological conditions are calculated from the following formu l ae& for short-term calculationss X

g exp 2o-2 z (7.7a) For Long-period average concentrations h2 2rr , exp ____ 1 __ 8 2o-2 z (7.7b) Titer U X z

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

Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964 Where X = concentration in curies/m3

(,c/cc) Q = emission rate in curies/second

°y* <rz = dispersion coefficients in meters h = stack height in meters 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 over 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 occurs over the range of about 4000 to 10,000 meters 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 diversity 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, the deposition rate has been calculated (Appendix 7.8). Since yearly average concentrations are used," it is reasonable to assume that the equilibrium conditions are reached; i.e. the rate of 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 1 µc per square meter resulted in about 0.1 µc/liter of milk*. Using this relationship the resultant activity levels in milk have been calculated.

The milk activity levels are ~hown in Table 1.g-. 7.9 The Federal Radiation Council has established a Radioactivity Intake Guide for Iodine-131 of 100 µµc per day, based on the uptake by children a s the most sensitvie segment of the population.

As can be seen from Table 7.8, the sumption of about five liters of milk per day from dariy cattle grazing inrn~diately adjacent to the site boundary would be required to equal the level of intake as established by *TID-8206, Page 56 Revision 1, Oct. 29, 1962 Revision 2, Au~. 20, 1964 ~~,,_, _______ ---*-----------------=-

* ----*-*--*-

... --._,_ ___ __

I i \ 1 l C . --.............. . Table 7.7 Maximum Concentration of Gaseous Isotopes Under Inversion and Average Meteorological Conditions x, pc/cc Isotopesd Curies/Second Inversiona AverageE MPCc pc/cc 7.3 X l0-2 7.3 X 10-7 1.6 X 10-8 3 X 10-7 ::," J Kr-85 I-129 1.3 X 10 -9 1.3 10-14 208 X 10-16 6 X l0-11 I-131 1.0 X l0-7 1.0 X 10-12 2.2 X l0-14 3 X 10-lO Xe-13lm 1.15 X 10-5 1.15 X 10-10 2.5 X 10-l 2 4 X 10-7 Xe-133 4.4 X 10-8 4.4 X 10-13 9.7 X 10-15 3 X 10-7 Tritium 5.8 X 10-4 5.8 X l0-9 1.3 X 10-lO 2 X 10-7 a Maximum concentration occu1*s at about 6000 meters from the stack; concentration within about 10% of the maximum occur from about 4000 to 10,000 meters from the stack. b Maximum concentratio

.ns occur at the site boundary (1500 mete:r-s).

c Table II, Appendix B, 10 CFR Part 20. d At 150 days cooling, these are the only significant gaseous isotopes.

e Based vn 1 triton produced per 104 fissions (reported as 1 in 1 to 4 x 104) wi~h 25% lost up the stack, 65% lost in liquid waste effluent, 10% to storage tanks. Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964 ,.f *\ \ .... ,,.

-l *\' . ----:--. *-. ' . -. -. . . -. -. . . -.. -._ . ---' . --....,..,._.._..,._--...,..._,....._

.. _~ ... ... _..._. -.....--.......---*--~

.. ...,. .........

_. ... __ ................

_______________

_..;; _ ___________

__..:.__,.....,,,,.,~==m::t=!l===-~?

Table 7.8 Iodine Deposition and Milk Concentrationa Ground Concentration Milk Concentration Distance in Meters uuc/m2 uec/liter 1500 200 22 2000 150 15 5000 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 Revision 2, Aug. 20, 1964 -------=-=-=-

-=:-.::.;==

=_;:_-~:::.:::---===

===-=-===-

--=-:-:=-----*-------~-=

I C C this Guide. This rate of consumption is higher than any that can be expected, propably 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 the waste storage tanks has been discussed in detail in Paragraphs 5.50 through 5.56 and in Submission l date July 1, 1963. These tanks are built in a "cup-and-saucer" design. Operating prccedures call for monitoring of t~e 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 will be transferred into a spare tank kept for that purpose. Thus, under normal operating conditions there will be !lQ loss of activity from these tanks. Storage Lagoon *1.11 The very low-level wastes from this process--'. overheads from acid fractionation, solvent wastes, and miscellaneous wastes--can be put through the general purpose evaporator and the overheads from this can be put through ion exchange columns if necessary.

It is expected that the normal activity content of the overheads from the general purpose 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 resulting in a ration of 3 x 10~8 µc/cc. The expected volume of these waste~ is 40,000 gal/day. The average available flow in Buttermilk Creek is 41 cf~ which is equal to 2.7 x107 gal/day. Thus, the available on-site dilution factor is 6.8 x 162. 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 Zr-Nb-95.

The MPC's for these isotopes are 1 x 10-5 2 x 10-6, 6 x 10-5, and l x 10-4 µc/cc rather than l x 10-~ for unknown activities when radium is absent. Therefore, the available factor of safety is about 103 out any analyses of the effluent and about 10 4 if we choose to carry out specific fission product analyses on this effluent stream. This stream will als o carry about 130 curies per day of tritium since there i s 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 ~c/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/cc; the MPC here is 3 x 10-3 µc/c~. Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964

. . * *-* * ..,..'! ) * , * * * .. * .' > ,Rf;. * > * * !.Ji -~..._ * '~. 1Utt ,...,,.._

P J ae *---------~-==

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 longer-lived isotopes.

The low level waste streams from the plant discharge into the interceptor, a concrete pit of 50,000 gallons capacity, which is designed for batching of wastes. A valved interceptor drain line will permit collection of one days output from the plant which will thon be sampled for gross alpha, beta, garruna and tritium. The pH of the sample will be checked and the interceptor contents neutralized if necessary to pH to 6 to 8. A line is available for pumping the interceptor contents back to the plant for further processing.

Normally, after sampling, the 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 with a high 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 thi~ plant. One is 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-level 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 sense that it could be leached out . of the waste if it were contacted with water. The radioactivity associated with hulls and discarded equipment, Revision 1, Aug. 20, 1964

( on the other hand, is not considered to be "available".

In the case of the hulls, the radioactivity is induced in the hulls themselves which are either stainless 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 boiling nitric acid prior to burial, inspected, and an aliquot analyzed to assure that significant quantities of fuel values are not being discarded with them. Equipment to be discarded w111 have been exhaustively decontaminated in place before bringing it out of the cells and it will then be further decontami~ated 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. 7.15 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 have 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 permeability 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 of this material to contain 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 ~eriod of time the curie ruthenium would have decayed to 10-5 curie. The yearly flow in Cattaraugus 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 pc/cc. The MPC is 10-5 ~c/cc. Revision l, Aug. 20, 1964 7.17 We expect the release of activity to the environ~~nt from the operation of the.wa1te burial ground--either from low-level trash containing "available" activity or from the high-level waste described above--to be completely quential.

Egress of Personnel or Material 7.18 The control of release of activity into the ment by carrying it 9ut on the persons or clothing of personnel or on material leaving the plant must be accomplished 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 shower. Hand and foot counters will be provided for monitoring all persons--visitors who leave the working areas. 7.19 Similarly procedures will be set up whereby nothing may 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 Health-Safety certification.

7.20 While it is possible that occasionally barely detectable quantities of activity might slip through these procedures, 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.

!roduct Shipme.nt 7.21 Radioactive shipments are covered by AEC regulations in 10 CFR Part 71 and 72 primarily.

All regulations in effect at the 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 for 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 I C Conclusion . 7.22 Qa the basis of the data and calculations presented in Paragraphs 7.6 through 7.21, in the normal operation of the chemical processing plant described herein, there will be no discharge of radioactivity to the environment in excess of the limits set forth in 10 CFR Part 20. Abnormal ~erations 7.23 In Paragraph 7.4 five abnormal ev8'lts were sized. These events range from the unlikely to the incredible but they delineate, we believe, the upper limit of any 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 1001 and further there is reasonable assurance that liqiud discharges at the site boundary could be kept within the concentrations for drinking water purpose spec~.fied 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 froni the waste tanks barriers between the stored waste and the environment.

The tanks have been equipped 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 bearing 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 ls measured and from which samples may be drawn to determine if there has been any leakage through the first three barriers.

If there should have been any large penetration of the first three barriers, it would be possible to retrieve the activity with relatively little dilution by pumping out of the gravel area through any of the eight wells. This area thus represents the forth barrier to the escape of activity.

Revision 1, Aug. 20, 1964 ,. -*.* ..... * *

Y.-W .

7.26 Th* local environment provides two additional barriers to the escape if radioactivity from the site. The tanks are located in the approximate center of a peninsula with a thick layer of silty till. It has been shown that the permeability of this silty till is so low that essentially complete containment woula be expec~ed of any waste that did escape the first four barriers.

The till, then, is a fifth and most important barrier. The peninsula is bounded by Erdman Brook and Quarry Creek. U&;S geologists who did the survey work on the site a*ssure 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 if it were not adsorbed on the soil by ion exchange.

At the confluence of these two creeks there is established a sampling station to determine 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 would not be impossible to collect the total flow at this point and pump it back up to the. plant site for additional processing if this should prove to be necessary.

Thi& represents the sixth barrier. There ls still a final sampling of the discharge in Cattaraugus Creek at the point where the effluent leaves the plant propet'ty.

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 111 sparing ratio. It is contemplated that during the first 15 y*ears of operation of the plant two oddttional working tanks will be built and that the spare will se1"Ve all three. The eventual sparing ratio will be dictated by plant experience.

7.28 We believe that a waste tank could be ruptured only by sabotage or by a major earthquake.

The former is outside 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 should rupture, howev~r, the combination of the vault, the gravel area and wells, and the impermeability of the surrounding silty till can be expected to mainta,.I'\ the tank contento within the inmediate area for a long period of time. There would be more than ampie time to arrange a temporary piping system to permit pumping the waste solution from the tank, the saucer, or wells into the gravel, into the spare tank. Revision l, Aug. 20, 1964 C G 7.29 The multiplicity of methods for determining any leakage from the tank make it essentially impossible that such leakage could remain undetected.

There are so many barriers between the waste and the* environment that ificant escape into the uncontrolled environment is also considered impossible, We even consider it possible to suffer a complete tank rupture--*

most serious hypothetical and unlikely accident--and still maintain Cattaraugus Creek below the IIPC levels of 10 CFR Part 20. Criticality Incident Anywhere in the Plant 7.30 There 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. Ole, that at Idaho Chemical Processing Plant in October, 1959, resulted in 4 x 1019 fissions *. Except in one case in'which there was some warping of a tank bottom, none of these resulted in any physical da"88ge. Theassumptio~

is made here that a criticality incident producing 1019 fissions in a single burst 9r 1o20 fissions in a repeating incident is experienced anywhere in the plant and that the entire production of noble gaseous fission products plus 1/3 of the iodines (from 1020 fissions) are lost. The value of 10 1 9 fissions is chosen to conform to calculations made at Savannah River suggesting this value as the upper 1~5it of a single.burst. These same calculations suggest 10 fissions as the resultant of a maximum repeated burst. It will be shown that the limit1og problem with this incident is not a public protection problem but rather the exposure of in-plant personnel to penetrating radiation 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 penetrating 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 hazard from the inrnediate radiation at the time of the burst. It is well established that the limiting condition in an occurrence of this type is the thyroid dose from the iodine is~topes re-1*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 Springvill~

and Buffalo the total population dose is calculated and expressed in man-rem.

Nuclear Safety, Quarterly Literature Review, Vol. 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 * --... ..__.... .... _____________

_ --------*

l I i .. Isotope Sr-90 Cs*137 Ru-106 . Table 7.30 uantltles and Concentrations of Sr-Various Points In the Event~of !e!c Quantities, Curles In Monltorln~

Leakage Stat Im_ . 2 X 106 20 2 X 106 20 1 X 106 100 Nonltorlng Station 1 X Jo*5 1 X 10*5 5 X 10*5

100 gallons per day for 100 clays. Concentratlonsd, ec/cc Butter'llnk Cattaraugus Creek Creek 5 X 10*7 5 X 10-8 5 X 10*7 5 X 10-8 1.5x10*' 2.5x10*7 b Tanks located* shc!NI on Figure J.2 so that* 11ln1 ... of 700 f*t of travel to n .. rest trlb~tary Is avalllble.

c For off*slte, Appendix I, Table II, 10 CFR Part 20. d Appllcele during the eleventh ,aar follawlng the accident; concentrations during foll~lng years wl 11 be lower. Decay has been neglected for cesh* a11d:*strontl1111

. J X 10-7 2 X 10*5 1 X 10*5 l 0 G b Oft 7.31 Table 7.31 ll1t1 the peak activity of each of the lodlne l1otope1 131 through 135 and the time after the accident when the peak.occurs. The1e have been calculated using NRDL-456, "Calculated Actlvltlea and Abundance, of U-235 flaaion Products".

11th.one exception, the peak activltlea have been assumed ln calculating the population dose. This procedure is conservative but by a relatively small amount over the time periods involved.

The one exception ls the activity of iodlne-134

at the time it reaches Buffalo under inversion conditions.

The transit time ln this case ls so large in relation to the half-life of lodlne-134 and lts precursors that its activity level was found to be negligible compared to the remaining iodine l1otopes.

7.32 The off-site doses have been computed assuming that the iodine ls released from the stack instantar.eously.

The total inhaled activity has been calculated using Equation 7.7a for short-term center-line concentrations.

The calculations have been performed for average (slightly unstable) meteorological conditons and for inversion (moderately stable) conditions.

The distances involved area Site periphery Springville Buffalo 1,500 meters 7,200 meters 51,000 meters The use of Equation 7. 7a is valid for the first two disi.nces*. 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 per person and the fraction of the 300 rem reference value are calculated for the three locations and for both types 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 do** in Springvi~le under inversion conditions, is not expected to be encountered since 1t la the opinion of meteorologists (aee Paragraph 2.13) .that an inversion aimed at Springville would be caught and held in the Buttermilk-Cattaraugus Valley systems. Even this 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

-~ ... Tabl*e 7.31 Quantitie1 of Iodine I1otope1 formed from 1o20 Fi11ion1 1 I1otope Tille of Peak Activity Peak Activity, Curie* 1-131 ~.2 houri 1-132 7.6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , 80 I-133 3.~ hour, 420 1-134 46.0 houri ~770 1-1~ 2.2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , 1470 Total 77~ a A11uaing 1/3 of the iodine, are lo1t froaa the -..tack. ___ ..,._, _____ _

0 G f, *lo -n J I SW Table 7.321 ftM/P*r*eo*

Inver1lon Average L2ai120 'llod*rtitl!

S~bltl ,s11sbil1 uo1tab11l , Site Boundary 0.09 0~63 ' . *Springville l.~ 0.06 Buffalo 0.33 2.8 X 10-3 a From instantaneous release of 1/3 the iodines from 1o20 fissions.

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

_ .. _._ -------...... ----*--Table 7.32b Individual and Population D01es at Several Points fn Event of a Criticality Incident Site Periphery Inversion*

Ayeraq,! . SDrinqville Inver11on8 Average* Buffalo Inversion*

Average* Individual Dose, RIii/ 0.3~ 2.8 x 10-3 Personc 0.09 0.63 1.95b 0.06 Fraction of 300 Rem/ 3.1 X 10-4 2.6 X 10*3 6.3 X 10.3b 2.1 X 10*4 10*3 b 9.3 X 10-6 Person Dose 1.1 X Exposed Population

.*-. Dose, Rem 24b 593b Total ------0.8 5 a See Table 7.5 for conditions assumed. b It is not expected that, under inversion conditions, any activity will reach Springville or Buffalo1 but rather.will be trapped in Cattaraugus Valley. c See Table 7.32a. 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* in racks of such geometry that the establishment of a critical array is impossible even if the elements were all of the imum reactivity of any fuel before it is irradiated.

Allowance is made for moving through the storage array an element of the highest reactivity.

This is discussed in Section VI and the occurrence of a criticality incident here is shown to be extremely unlikely.

Despite the fact that a criticality incident in the fuel pool is extremely unlikely, the following event is hypothesized*

It is assumed that an element i~ janrned into the interstice between four element~ and that the five elements are involved in a crltical event, that a 10-IIINt boiling water reactor will be set up, and that it will operate 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months 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 thit section are shown i in Appendix 7.34. The heat released would raise the temperature of th~ 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 ificantly and consequently the*shielding provided by the water would prevent ~ny hazard from increased radiation levels from direct radiation.

EBWR defect test studies have

shown that the fraction of noble gaseous activity lost per second from a defective fuel element is about 4 x 10-B. This same test showed that the iodine loss was at least an order of magnitude less than this. The total inventory of gaseous activity in 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 quantities of iodine isotopes are much less than the amounts which have already been shown to be readily tolerated by this environment (see ~aragraphs 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. Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964 Isotoee Kr85m Kr85 Kr88 1131 1132 1 133 1 134 1135 x 8 133m x 8 133 Xel35 Xel38 .. Table 7.34 Gaseous Activities Lost from Fuel Pool During Assumed Criticality Incident Inventroy Curies 6.1 X 10 4 2.5 X 103 *,, 2.7 X 103 3.2 X 105 5.4 X 104 1.6 X 106 1.6 X 105 5.2 X 102 9 X 103 1.5 X 106 3.7 X 10 6 Fraction Lost Total Lost in In 3 Hours 3 Hours I Curies 4 X 10-4 24 4 X 10-4 1 4 X 10-4 10 4 X ------. 0.1 ..._ ' 4 X 10-5 ' 13 4 X 4 X 4 X 4 X 4 X 4 X 4 X 10-5 2.2 10-5 64 10-5 6.4 10-4 0.21 10-4 3.6 10-4 600 10-4 1500 Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964 Using the method employed in Paragraph 7.7 (Equation 7.7a) the concentration of xenon-138 at the site boundary under inversion conditions is 1.4 x 10-6 p.c/cc. No MPC for this isotope is given in either 10 CFR Part 20 or in NBS book 69 but it would not appear that this would result in any hazard. l h is event can, therefore, be tolerated without exceeding published MPC 1 s. Chemical Explosion 7.35 The assumption 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 withstand the rupturing of a tank. However, there might be some plugs or windows loosened.

So long as ventilation is maintained air fl<A~ should remain into the cell.except for the instant of the explosion.

In analyzing ihis event some assumptions contained in "Radiochemical Facility Hazard Evaluation", by i. D. Arnold, A. T. 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 trating high-efficiency filters will contain 0.14 mg/M of material with the same concentration as the original dispersed solution ijnd that the MPC for mixed fission products is 6.6 x 10-~c/cc. The ventilating air passing through the filters of this plant amounts to 32,000 cfm or 900 M3/min. Then 0.14 x 900 or 125 mg/min of the original solution may be ~ssumed to pass through the filter. We 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.

Th~ 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. Follow i ng the metho~s of Paragraph 7.7 the poorest value of X/Q (at a distance of about 5,000 meters) is 1 x 10-5. Q is equal to 1.1/3600 = 3 x 10-4 curie/sec.

The X = 3 x 10-9 ~c/cc. This is less than the MPC for mixed fission products assumed by ORNL in the above report and it would appear possibl~ to accept this particularly untoward accident.

There would b_e, of course, a big cleanup job in the cells. This would be undertaken according to methods outlined in Paragraphs a.a 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 ls assumed that the silver reactors and other iodine removal equipment all fail and that this is not discovered for a period of one day. That this could remain undetected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is extremely unlikely since the stack monitor would *dete~t the iodine increase at once. If the entire charge of iodlne-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 distance 5 of about 5,000 meters) the poorest value of X/Q isl x 10-and the c c nce~tration of iodine-131 at this point would then be 2 x 10-0 pc/cc. 'nlis is less than the MPC for continuous e>eposure off-site.

Conclusion 7.37 All of the abn9rmal incidents hypothesized in Paragraph 7.4 have been anaAyzed.

It has been shown that in all cases except the 1o2 fission criticality incident the limits prescribed in 10 CFR Part 20 for continuous exposure 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.

Since we expect the probability of these events to be very low, to the point of incredibility, and since th~y 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, 0:t. 29, 1962 Revis i on 2, Aug. 20, 1964 I rrr I . ---------------*---.. _,_ ... _____ ....... --J t ' (j (_ ---*--*--.:,.---

Criticality Incident Anywhere In the Plant 7.38 There hive been ten criticality Incidents In solution systems.* Eight of these have r1sulted *tn: a totar number of fJsslons ranging from 4 x 1016 to*l.3 x 1016. One, that at Idaho Chemical Process i ng Plant In October, 19S9,. resulted In 4 x 1019_flsslons.

Except In one case In wh t ch there was some warp Ing of a. tank bottom*, none* of thest! *resu 1 ted In any physical damage. The assumption Is made here that a criticality Incident producing 1019 fissions In a single burst or 1020 fissions In a repeating Incident ts experienced anywhere In the plant and that the entire production of gaseous fission products .(from 1020 fissions) 1$ lost. The value of 1019 fissions Is chosen to conform to calculations made at Savannah RI ver suggesting th Is va 1 u.e as the upper 1 Im It of a single burst. These same calculations suggest 1020 fissions as the resultant of a maximum repeated burst. It will be shown that the limiting pro~lem with this incident Is not a public protection problem but rather the exposure of In-plant personnel to penetrating radiation at the time of the burst. For a repeating Incident there would be time to evacuate personnel af~er the first burst and the exposure to 19 penetrating radiation can be considered equivalent to that from a 10 fission burst. This ts considered In Paragraphs 8.27 and 8.28. Insofar as the general public ts concerned there ts no hazard from the lmm~dlate radiation at the time of the burst. It ts well established that the 1 lmltlng condition In an occurrence of this type ts the thyroid dose from the Iodine Isotopes released.

Therefore, this event ts 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.S. In the case of Springville and Buffalo *the total population dose ts calculated and expressed In man-rem. 7.39 Table 7.39 lists the peak activity of each of the Iodine Isotopes 131 through 13S and the time after the accident when the peak*occurs. These have been calculated using NRDL-4S6, "Calculated Act I v i t I es and Abundances of U*23S Flss Ion Products*J: WI th one exception , the peak activities have been assumed In calculating the populat i on dose. This procedure Is conserv*tlve but by a relatively smal 1 amount over the time periods Involved.

The on'e exception Is the activ i ty of lodlne~134 at the time It reaches Buffalo under Inversion conditions. The transit time In this case ts so large In relation to the half-life of lodlne-134 and Its precursors that Its activity level was found to be negligible compared to the remaining Iodine Isotopes.

-.'t *, .... Nuclear Safety, Quarterly Literat u re Review, Vol. 3, No. 2, Dec. 1961, Pages 34-37 plus a recent Hanford i ncident not yet* reported. J__..__________,_..........._._

_______ ...-...,. . '"" .. 7' P Ffl,:,.TM1iW U F L Y &*B ...... .,._...-.....-..

It

~"'-"--...,_ . ...._ ________ _ -Table 7.39 Quantities o(.lodlne Isotopes Formed from 1020 Fissions Isotope Time of Peak Activity Peak Activity,* Ci.fries 1-131 .5 .2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 73 1-132 J.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 24o 1-133 3.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 1,250

1-134 46.o minutes 17,300 1-135 2.2 minutes 4,4oo Total 23,300 t l f I ----*".': ---------

C 0 7 , 40 The off-site doses have been computed.assuming that all the Iodine Is released from the stack Instantaneously.

The total Inhaled activity has been calculated using Equation 7.7a for short-term center* line concentrations.

The cilculatlons have been performed for average (sllghtly unstable) meteorologlcal conditions and for Inversion

(moderately stable) conditions.

The distances Involved are: Site periphery Springville Buffalo 1,SOO meters 7,200 meters Sl,000 meters The use of Equation 7.7a Is valid for the first two distances.

Extrapolation to S0,000 meters Is questionable, but gives a fair estimate. The results so obtained are given In Table 7.40a. Then using the approximations suggested in 10 CFR Part 100 for the thyroid dose from each of these isotopes, the total rem per person and the fraction of the 300 rem reference value are calculated for the three locations and for both types of meteorology.

Total man-rem values have also been calculated , These data are presented In Table 7.40b. / . Calculations supporting the numbers shown In these three tables are given In Appendix 7.40. 1t can be seen that In no case Is the reference value used for evaluatlo~

of reactor sites exceeded or even closely approached.

The highest value Indicated, a S.8S-rem/person dose In Sprlngvl*lle under Inversion conditions, Is not e>tpected to be encountered since It ts the opinion of meteorologists (see Paragraph 2.13) that an Inversion aimed at Springville would be caught and held In the Cattaraugus Valley systems. Even this value ts only about 2 per cent of an emergency dose of 300 rem.

Criticality Incident In Fuel Pool 7.41 The: fuel pool Is designed to hold 1000 fuel elements In racks of such geometry that the establ tshment of a crl ti cal array ts Impossible even If the elements were all of the maximum reactivity of any fuel before It Is *irradiated.

Allowance Is made for moving through the storage array an element-of the highest reactivity. Thls Is discussed In Section JlI and the occurrence of a criticality Incident here Is shown to be extremely unlikely.

Despite the fact that a criticality incident In the fuel pool ts extremely unlikely, the following event Is hypothesized

It Is assumed that an element Is Jammed Into the Interstice between four elements and that the five elements are Involved In a critical event, th~t a 10-mwt bolling water reactor will be set up, and that It will operate 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months before It ts 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. " * *f Table 7,4oa Total Dose Due to Radlolodlnes, Rem/Person*

Inversion Aver*ge Location (Moderately Stable) (Sllahtly Unstable)

Site Boundary 0.28 2,5 Sprlngvl I le s.as 0.19 Buffalo I, 0 8.4 X 10*3 ..

From Instantaneous release of all the Iodines from 1020 fissions.

---.,.. ~' t: l i ..... l , 1

lndtvtdual Dose, R8'1Personc Fraction of 300 Rem/Person Dose Exposed Popul*tlon Dose, Rem Total { Table 7.ltob Individual and Population Doses at Several Po i nts In Event of a Cr i tical i ty Incident S. i te Per lphery Sprln9v 11 le Inversion*

Average* Inver s ion* Averap* 0.28 2.s s.8sb 0.19 9.3 X 10-4 8.3 X 10-3 1.9 X 10-2~ 6.3 X 10-4 ------73b 2.4 a See Table 7.S for condit i ons assumed. Gd ' ' '-~-,-luff*lo Invers i on* Average* 1b . 8.4 X 10-3 3.3 X 10-~ 2.8 X 10-S 1750b IS b It ts not expected that, under inversion conditions, any activity will reach Springv i lle or Buffalo; but rather will be trapped in Cattaraugus Valley. c See Table 7.4oa. -I j i t* i I i ' ' ' l ' I

t . '-f . __ ....

I i I i I I i I i : i ' *-,c* 7.42 Calculetlons supporting this section are shown In Appendix 7.42. The ~eat released would raise the temperature of the water In the storage pool only about *J6F even If the pool water coolers-failed to operate. Therefore, there Is no danger that the water*level In the pool would drop significantly and consequently the shielding provlded by the water would prevent any hazard from Increased radiation levels from direct radiation.

EBWR defect test studies have'shown that the fraction of noble gaseous activity lost per second from a defective . fuel element Is about 4 x 10-8. This same test showed that the lo~lne loss was at least an order of magnitude Jess than thl.s. The tota: Inventory of gaseous activity In the five fuel elements assumed to be Involved In this Incident and the amounts ..tllch may reasonably

.be Jost from the fuel pool water are shown In Table 7.42. The quantities of Iodine Isotopes are much less than the amounts ..tllch have already been shown to be readily tolerated by this environment (see *Paragraphs 7.38 to 7.40)~ Consequently the Iodine releases result In less hazard than has already been shown to be acceptable.

The releases of kryptons are also much Jess than those which have already been shown to be within MPC. Similarly, the xenon-133 discharge results In concentrations

.. from: IO~to 109.: MPC's. The only aspect of thls hypothetical Incident ~lch has not already been calculated In this section Is the xenon-138 release. Using the method employed In Paragraph 7.7 (Equation 7.7a) ' the concentration of xegon-138 at the site boundary under Inversion conditions Is 1.4 x 10-~c/cc. No MPC for this Isotope Is given In either 10 CFR Part 20 or In NBS Handbook 69 but It would npt *appear that this would result In any hazard. This event can, therefore, be tolerated without exceeding published MPC's. ! Chemical Explosion 7.43 The assumption Is next made that a vessel containing one full day's charge of fuel In solution suffers an explosion

~lch rupt~r.es the vessel distributing the contents throughout a cell and putting some fraction of the contained solutl_on Into the ventilating system. The ventllat*lng system wtll withstand 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 I nstant of the explosion.

In analyzing this event some assumptions contained In "Radiochemical Facl 1 lty Hazard .Eval\la: t')on, 11~ by: E: .. o*. *~rnold, A.T. Gresky, and J.P.Nlchols, ORNL-CF-61-7-39, July 10, 19'1 are used. This Is a very slml lar analysis of a completely analogous s*ltuatlon to that considered here. The assumptions are made therein .that aerosols penetrating high-efficiency filters wlll contain 0.14 mg/M 3 of material with the same concentratlori as the original dispersed solution and that the MPC for mixed fission products Is 6.6 x 10*9'-'c/cc. The ventlJatlng air pasting through the filters of this plant amounts to 46,000 cfm or 1300 Ml/min. Then 0.14 x 1300 or 180 mg/min of the original solution may be assumed to pass through the filter. We further assume that the gaseous activity has already been released and that In one hour .the ventilating sys t em will have.picked up nearly all of the gross activity that Is going to. Under these conditions about 11 grams of solution wlll be released.

The maximum activity to be expected In the plant Is about 200 curies per liter or 0.13 curie I , J I: i __ ,1...t---------------------------------------r-1

11 . ......:-----* '°""--................. -.... -.---.---------

-*iii;--tt-

_,.. __ ----*-------

---_ _.

-f Isotope Kr8Sm Kr85 Kr88 1131 1132 :t133 1134 7135 xe133m l 0 xe133 1 xe13Sm * '

l xe138 l. , I I . r I C ' J J I I l I If -llt i;;:Q 8 Table 7.42 G**~* Activities Lost from Fuel Pool During Assumed Criticality Incident Inventory Fr*ctlon Lost Tot*I Lost 1 , , Curles In J Hours J Hours 1 Curles 6.1 X 104 4 X 10*4 24 2.5 x 1o3 4 X 10*4 1 2.4 X 105 4 X 10*4* 10 2.7 X 1o3 4 x 1o*S ~-.1 3.2 X 105 4 X 10*5 13 5.4 X 104 4,: 10*5 2.2 1.6 X 106 4 X 10*5 64 1.6 X 105 4 X 10*5 6.4 5.2 X 10 2 I 4 X 10*4 0.21 9 X 1o3 4 X 10*4 3.6 1.5 x.:10 6 4 X 10*4 600 3.7 X 106 4 X 10*4 1500 t*.---~i---------

.........

1 ______ .. ____ --*"'------~--:-----*--------=--, .. -*-*"""' Ulft--------~

-:--------.

-.* -'. ~ea* gram for a total discharge of 1.4 curies. Fol lowtng : the methods of Paragraph 7.7 the poorest value of X/Q at ihe*slte boundary*I~ 1 x 10*5. Q ts equal to 1.4/3600 * *3.9 x :10-~ curie/sec. Then X

3.9 x J0*9~c/cc.

This 1.s less than the *"PC for mixed fission.products assumed by ORNL In the above report and It would appear possible to accept this particularly untoward accident.

Ther~ 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. Failure of Iodine Reppval Equipment 7.44 Finally It Is assumed that the silver reactors and other : Iodine removal equipment all falls and that this _Is not discovered for , a period of one day. That .this could remain .un~tect~d for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Is extremely

unlikely since the stack monitor w:,uld detect the Iodine Increase at once. If the entire charge of ... Jodlne-*131.,: J.6.cur i lei,. were to*be*lost during a *~Y.the value of Q ls: 1.9 x : Jo-s curle,'sec.

At the site boundary under worst conditions the value of X/Q Is Ix 10 10-S. The concentration at the site boundary would then*be*l.9 x .10* )Jc/cc. This Is less than the MPC: for continuous exposure off-site.

[conclusion 7 .45 Al 1 of the abnormal. Incidents hypoteslzed In -P*ragtaph

,7 .4 have been analyzed.

It has *been shown that .ln*all cases except the 1020 f.l ss I on .cri t I ca 11 ty Incl dent the llmlts prescr I bed In : 10 CFR Part 20 for continuous exposure are met and that *tn .~his on'e case*there ls.no dose at any p9lnt which exceeds, or even closely approaches, .the guides suggested In 10 CFR Part 100 for emergency. conditions.

Since we except the prr.>babl 1 lty of these eve_nts to be *very _low, .to the point of Incredibility, and since they can .be handled by.the environment even If they should occur, we *submit that the operation

.of this plan_t does not constitute an undue hazard to*the general publl, beyond the s I te boundary :.J .. ' , , I C I I I I I I ' ., -.&....+---_,.

_____________________

....., ___________ -a,..-i.

-.* -'. ~ea* gram for a total discharge of 1.4 curies. Fol lowtng : the methods of Paragraph 7.7 the poorest value of X/Q at ihe*slte boundary*I~ 1 x 10*5. Q ts equal to 1.4/3600 * *3.9 x :10-~ curie/sec. Then X

3.9 x J0*9~c/cc.

This 1.s less than the *"PC for mixed fission.products assumed by ORNL In the above report and It would appear possible to accept this particularly untoward accident.

Ther~ 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. Failure of Iodine Reppval Equipment 7.44 Finally It Is assumed that the silver reactors and other : Iodine removal equipment all falls and that this _Is not discovered for , a period of one day. That .this could remain .un~tect~d for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Is extremely

unlikely since the stack monitor w:,uld detect the Iodine Increase at once. If the entire charge of ... Jodlne-*131.,: J.6.cur i lei,. were to*be*lost during a *~Y.the value of Q ls: 1.9 x : Jo-s curle,'sec.

At the site boundary under worst conditions the value of X/Q Is Ix 10 10-S. The concentration at the site boundary would then*be*l.9 x .10* )Jc/cc. This Is less than the MPC: for continuous exposure off-site.

[conclusion 7 .45 Al 1 of the abnormal. Incidents hypoteslzed In -P*ragtaph

,7 .4 have been analyzed.

It has *been shown that .ln*all cases except the 1020 f.l ss I on .cri t I ca 11 ty Incl dent the llmlts prescr I bed In : 10 CFR Part 20 for continuous exposure are met and that *tn .~his on'e case*there ls.no dose at any p9lnt which exceeds, or even closely approaches, .the guides suggested In 10 CFR Part 100 for emergency. conditions.

Since we except the prr.>babl 1 lty of these eve_nts to be *very _low, .to the point of Incredibility, and since they can .be handled by.the environment even If they should occur, we *submit that the operation

.of this plan_t does not constitute an undue hazard to*the general publl, beyond the s I te boundary :.J .. ' , , I C I I I I I I ' ., -.&....+---_,.

_____________________

....., ___________ -a,..-i.

.,._ ti r r a 11 Ii

n 1 .... ,

I (l it t I I I j I G I G r::._:_ __ . *---* -----VIII PROTECTIOO OF PLANT PERSOONEL Design Criteria 8.1 The design criteria and the operating rules ~f the NFS plant have been set up so that the plant will conform to the 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 constant program designed to increase the safety morale of all of its personnel, both in the area of normal industrial safety and in that of radiation safety. 8.3 The radiation safety program ls designed to protect the plant personnel from* 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 th& standpoint of personnel in the plant1 and some less serious but more probable events are discussed from the view-. point of personnel protection.

P~otection from External Radiation 8.4 The primary protection for the worker from penetrating radiation is to interpose sufficient shielding between him and the radioactivitY.

at all times. The plant '4J shielding has been described in some detail in Paragraphs 6.59 through 6.65. The shielding has been designed so that, when the most active unit which could 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 is small. The shielding design has been based on a "design" Revision 1, Aug. 20, 1964 ~-::-------*-----------,----~-

-

fuel having the following irradiation on historys Burnup Specific Power Cooling Time 30,000 MWD/T 35 MN/T 150 Days 8.5 Fuel is brought into the plant in shielded casks which have had their design carefully checked to ensure tha t. adequate protection is available.

A shipment will be survey~d before it is sent out. It will be surveyed again upon arrival at the plant. Before the carrier is opened, it is placed under sufficient water (see Paragraph 3 .. 7) so that, as a fuel element is removed, there will be at least 11 feet of water over the top of the longest type of fuel element. Movement of the elements in the storage pool and their storage are also conduct2d under at least this much w~ter. Transfer to the PMC is done remotely under water and back of concrete shielding.

The ~echanical operations in the PMC and GPC are carried out remotely back of concrete shielding.

The transfer to the CPC is handled remotely.

All operations in the CPC are remote. Transfers to the remaining contact-maintained cells are fluid transfers carried out remotely.

All operations in the entire process, therefore, are carried out behind shielding until product is decontaminated to the point where external radiation is no longer a problem. Plutonium products containing high concentrations of Pu-240 will be placed semi-remotely into containers with sufficient shielding so that they may be handled safety. 8.6 Sampling is 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 completely behind shielding and to provide working back-* ground of 1 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 tho plant at design levels, it is necessary not only to have adequate shielding but also to maintain str i ct controls to prev~nt spillage.

This is done first by kaeping the activity back of the shielding--there are no planned withdrawals of activity except for the samples, many of which , have been already diluted; second by a careful and continual radiation survey program to detect areas in which there may have been an inadvertent introduction of activity; and third by a prompt and immediate cleanup of such areas at the same time mining the cause of the event and correcting it. Revision 1, Aug. 20, 1964 8.8 Maintenance work, both routine and major, can be expected to contribute somewhat to the whole body radiation of the plant personnel.

It is the intention of NFS to permit maintenance work only under such conditions that no ~orker will be expos~d in excess of the limits defined in 10 CfR Part 20. Tho maintenance procedures, which are described in*'detail in Section IX have been set up to minimize .the exposure of the pe~sonnel.

However, it is clear that maintenance work will have to be 'done in high radiation areas (areas in which the background levels ~xceed 100 mr/hr). Such work will be controlled by a work permii' sys.t;em as describe~

in the Health-Safety portion of Section IX and be authorized by the plant manager. In attacking any maintenance job, the area is carefully surveyed and the amount of time ~hat may be permitted a worker in the are is calculated.

Work in the radiation field is done under closely supervised conditions.

Accurate time is kept from outside the. field. Record~ng meters* as well as film badges are worn during the ope~ation and a log of the e~posure is kept and this is 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 wor~er be allowed to enter a radiation field exceeding 2 r/hr without special approval of the plant , manager. It will be normal plant practice to limit the exposure of any individual for any single maintenanc~

job to 0.2 rem. Subject to the maximum limitation specified above, the balance be~een ti.Me and activity level .will be a decision to be made by plant supervision 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 in the full 1 mr/hr permitted in a normal access area. It is expected that most of the normal access area will have a background much less than thls. for planning purposes we have assumed that the additional six hours per day will be in an average*background of 1/6 rrµ:'/hr.

The total background radiation for the quarter.would then amount to 0.2 rem. This would leave about 1 rem per quarter for maintenance operations without exceeding 1.25 rem/quarter.

With exposure limited to 0.2 rem/maintenance jobs, a given individual could perform five such maintenance jobs per quarter. There will be about sixty men in the plant who can be called upon to carry out such jobs s~ that the plant can carry out a maximum of 300 such operations per quarter, about five per day. Revision 1, Aug. 20, 1964 Inhalation 8.10 The primary protection of the workers from inhalation lies in keeping the activity inside the P-ror.ees equipment itself. As a second line of defense, all 'of the equipment ls contained in cells maintained by a separate ventilation system at a pressure negative to the work,.ng areas. As third line of defense, masks ~.nd supply-*atr equipment are available.

These ventilation

systems have been described in detail in Paragraphs 6.3 through 6.21. Under all normal operating conditions no process activity ls 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 portabl* air monitors to assure that the air in the working area does, indeed, remain free of a*ctivity. 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 ls necessary.

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 to be expected is completely negligible, in either case. There ls, however, an infrequent condition whereby.the discharge fiom a stack may come directly down upon the stack. Under these conditions the amount of dilution could be small. A calculation ls shown in Appendix 8.12 for the normal lodlne-131 discharge.

This 1hows 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-B .,.c/cc which is only a factor of 7.5 hlgher;than the occupational IFC of 10 CFR Part 20. Thia p~rt.i~ular meteorological condition ls not expected to occur very frequently or to persist for any long perlo,t of time. Even with !!2 dil_ution, and it would be expected that there would be some--perhaps a factor of ten, the conr.entration is such that under the provisions of Paragraph 20.l03b l 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 mor, of the monitors.

This iodine concentration would be attained only during the course of a dissolutlon1 th,re would be ample time to shut down the dissolver or evacuate the building or both. ..... d

Revision 1 1 Oct. 29, 1962 Revision 2, Aug. 20, 1964 I Q Q 8.13 There will be an ample aupply of protective equipment auch a, Scott Air Packs available for use during emergency condition, or during maintenance work inaide cells. The Safety program (see Section IX) will include frequent training 1e11ion1 and drill* for all personnel in the use of this equipment 10 that in an emergency the equipment should be used promptly and properly.

Ingestion 8.14 The control of the problem of ingestion of activity 11 largely one of developing within the workers good safety morale and habits of personal hygiene and of providing thell with adequate protective clothing, devices, and monitors to facilitate the execution of , the program. It is a problem which has been dealt with routinely at all of the presently operating chemical processing plants of the AEC without creating any serious hazards. 8.15 Protective clothing will be issued to all personnel working in the plant areas and aust 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 protective clothing such as SU1"9ical, heavy rubber, and cotton gloves, caps, boots, and tape (for taping gloves to coveralls, for instance}

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 ~he lunch room. There will be hand and foot counte~s at the door. The area will be checked frequently by 'the Hea.lth--Safety survey. Smoking will be done only in designated areas. 8.17 At the conclusion of e,ch shift each individual who works in the plant areas will be required to change clothes, and check his hands and shoe*, before leaving the premises.

Two hand and foot counters are provided.

kevision 1, Aug. 20, 1964 -------------~

8.18 The abov* program of hygiene has-proved to be ,atisfactory to maintein the ingestion of activity to neglig-1blt levels at other installations.

The common practice to back up this program with a medical .program will be followed at the NFS plant. The medical plans for the plant are as follows* The medical program wi'll consist of a very thorough.pre-employment medical history and physicai examination tor each prospective employee.

The medica*~ , history will be aimed at not OJllY past illne~.s.~s and injuries but particular attention wilt be paid to history of past radiation exposure, allergies, blood dyscrasias, tumors, and any eviaence 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.

Each employee will have a complete physical examination yearly. A complete blood count will be done twice yearly1 clinical urinalysis monthly. Bio-assays will be done on an "across-the-plant

statistical 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 leaving the employ of the company. A dispensary will be maintained for care of ordinary minor on-the-job injuries.

There will be far-ilities for intensive first-aid

~*re of severe injuries such as burns, fractures, and gross conura~.nation with radioactive material.

IfflJ'ii\lnization against tetanus will be routine for all employee~.

Close liaison with the H~alth and Safety Department will be maintained.

The medical director will assist in health and safety training and indoctrination.

He will review with the superintendent of health and ~afety all industrial radiation exposure records1 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 acwaaa. . I Q C Deteiled records of all t h e above will be maintained by the medical dj , recto r. 8.19 Ae explained in the foregoing paragraphs we anticipate no difficulty in conducting the normal operation of this plant withi~ the fr~m.-ork of permissible levels of exposure.

It remains in the ~emainder*of this section to analyze the consequence to employees of accidents. Analysis of Accidents 8.20 In Paragraph 7.4 five highly abnormal hypothetical' incidents were proposed and the effect of these upon the public wa, considered ln Paragraphs 7.24 through 7.37. These same incidents are now considered wl th reference t~ 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 ln the innediate vicinity for some period*of time. It ls proposed that the waste be transferred in~o one , of the spare tanks* *as quickly as possible.

-The type of* action envisioned Would involve*puq>ing the"&Olut~c;>n into a spare tank probably through iemporary* lines which might be la'i.d *overground with only a minimum ~f shie.lding. The laying of .thfs pipe would not require personnel exposure exc.ept for th~ *connection i~to 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 cauld*be accomplished from* back of a temporary*s~lel~ constructed outside the radiation field and pushed into place with 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 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 ins~rted upstream from the pump s so that the lines could be flushed and decontaminated somewhat before such Revision 1, Aug. 20, 1964 ____ _...,. .... _..... .. _., ...... _-,-___ -.. b -~""""-----

-*-* --fJa:'4U

___ _

maintenance would be attempted.

It would have to be done from behind a portable thie!.d. In the ca1e of 10 1eriou1 a problem 11 thi1 there would be no que1tion that operations of the plant would be 1hut down and all available expo1ure time would be u_sed in solving this problem. Supervi1ory personnel to the highest levels and individuals from other plants operated by the company--thoae who receive no radiation in the course of their work--could be brought in if necessary. No individual, however, ne*d be exposed beyond permissible levels. 8.23 lloat of the exposure a11ociated with this incident would be in cleaning up afteiward.

Dismantling the highly contaalnated line,, puaps, and valves and disposing of them would certainly require ope ra tions in high radiation fields. However, the need for speed would no longer be present and enough time and peopl e would be used to a11ure that the task was accoapliahed within the permissible radiation exposures.

Criticality Incident in Plant 8.24 This incident, di1cu11ed in Paragraphs 7.30 through 7.32, assumed that the ventilation syst.lfr remained in working order since that situation results in the moat ianediate and complete discharge of the gaseous isotopes to the environment.

In that ev~nt the air inside the plant would be completely safe, as evidenced by the calculations shown in Appendix e.12, in all cases except that of the unlikely recycle. In Appendix 8.25 it i1 shown that if such a downdraft occurs under averago conditions the amount of dilution will amount to a factor of about ~00. If it occurs under inversion conditions the dilution factor will be about 25. For the purpose of this calculation the dilution is taken as 10. It ls furthe r assumed that during the course of the entire diicharge (asa u m ed 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 10 preoccupied with rendering a ssistance to other personnel or are otheiwise 10 upset that thoy 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 8.25. 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 1 ! I l I I I t f_

Isotope I-131 1 ,.132 I-133 I-134 I-~35 Table 8.24 Thyroid Do1e During Recycle Coincident With a Criticality Incident X -X Cone., Q .,.c/cc Frequen~y 2.6 X 10-3 ..J.. 100 6.3 X 10-3 ....L 100 0.04 ..J.. 100 0.6 ..J.. 100 0.16 ..J.. 100 Dose Rate, Rem/ Time, ,c/(cc)(sec)

Seconds 110 600 4 600 31 600 2 600 6 600 Total Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964 Dose, Rem 0.7 24 22 74.7 8.2~ There is no reason to assume that the events which could l*ad to a criticality incident would also lead to shutting down the venti~"tion system. There may be, however, a small probability that the two events might occur simultaneously.

The possibility ls a difficult one to analyze since the allh)unt of leakagi of activity from the cell would be.expected to vary considerab l y depending on th~ conditions.

If the supply air remains on, the exterior of the cells would remain at a higher pressure than 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 to the natural draft of the 65-meter stack. If the static pressure in th&, cell becomes equal to that outside the cells, some leakage would occur but it should not be large since any leakage path would be tortuous.

There would not seem to be a mechanism whereby the cells 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 countermanded by plant supervision.

We conclude, therefore, that if a criticality incident were to occur coincident with a failure of the ventilating system, the plant would be evacuated long before anyone could receive a significant dose from inhalation.

8.26 The mo~t important asper.t of protection of plant personnel in connection with a cri ticali*~y incident is to assure that no one x-eceives a seri'ous dose of penetrating radiation at the time of the incident.

The first line of defense is, of course, to prevent the occurrence of the incident.

Great care is being exercised in the design of this plant and in the setting up of its operating procedures to ensure that a criticality incident does not take place. The whole subject of criticality control throughout the plant has been presented in detail in the final paragraphs of Section VI. We believe that we ha~e reduced the probability of such an incident to an absolute minimum. However, there have been eleven such accidents in solution systems. Every major site save one has had one. There have been five incidents in metal-air systems at Los Alamos. 8.27 An Oak Ridge study* has calculated the prompt neutron and gamma 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 for a 1019 fission event by direct ratio. The concrete shielding walls

ORNL-CF-61

.. 7-39, "Radiochemical facility Hazard Evaluation", E. o. Arnold, A. T. Greiky, and.J. P. Nichols, July 10, 1961, Page 6. Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964 l1otoe, 1-131 1-132 t...:1 33 1-134 1-135 -l I l ~----Table 8.25 Thyroid Dose Durlna Recycle Coincident With a Crltlca11ty ln~ldent !x ~01e Rate, Cone, Q Ra/ ,vc/cc Freguen~ ,+1c/(cc).(1ec) 5.6 x 10-3 ' 330 -100 1 .9 X 10*2 1 12 -100 o. 1 ' 92 loo 1.3 ' 6 -100 0.34 ' 25 100 Time, Dose, Seconds Ran --600 . 11 600 : 1 600 55 600 47 600 51 -Total 165 I : I* ; ------* lf *" ' Table 8.27 The Prompt Neutron and Gamma Dose at the QJtside Of a Normal Concrete Shield From a Nuclear Reaction of 1018 Fis*iionsa, 6 Ordinaryc Do1e at 0Jt1id1 of Shield 1 rem Concrete Shield Metal Nuclear Nuclear Reaction in Th1ckneas 1 Ft Reaction Aqueous Solution 1 88,000 , 5,200 3 317 23 4 17.0 1.9 5 0.960 0.14 6 0.059 0.012 a The dose rate may be calculated for any other number of fissions through the use of a direct proporation.

b ORNL i CF-61-7-39, "Radiochemical Facility Hazard Evaluation", E.D. Arno l d, A.T. Gresky, and J.P. Nichols, July 10, 1961, Page 6. c For high density concrete the ganaa dose is reduced by a factor of 1.6 for a given concrete thickness. -"""'--...........

---..... --*-1"1:l!.ii-~-=-"'--*---J: C C ' ' L C -----for the GPC, PMC and CPC have openings for viewing windows which are equivilant in shielding value to the concrete walls for ganma radiation but offer less protection than the concrete for neutron radiation.

Table 8.27 does not reflect the increased neutron dcse to an employee who might be ir front of one of the viewing windows. In* the case of a 10 9 burst from a criticality accident in the dissolver, the total prompt neutron plus ganma dose to an employee at the nearest viewing window would be about 300 rem, if the window were completely transparent to neutrons.

The only place in the plant where a metal-air incident is at all possible is in the PMC -GPC. There we have four feet of high density concrete shielding and the resulting dose would be negligible.

A solution system event could conceivably occur from the dissolver on. In the CPC there are six feet of concrete and the dose would be even less than in the *PMC. In Cell #1 there are five feet of concrete shielding.

The dose would still be negligible.

In the remaining four cells there are three feet. At the lower end of the process there is no.need for this much shielding from the fission product content. The minimum of three feet of concrete shielding has been carried down to the end of the process in order to assure that even if a 1019 fission critical incident should occur, and a worker should be standing right opposite the point in 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 plant and.put into storage and eventually onto a truck for shipment, contact with the product is required.

Therefore, particular care has been exercised with product shipment plans. This is discussed in Paragraph 7 .. 21. Criticality Incident in the Fuel Pool *a.29 -The hazard of a criticality incident in.the fuel pool to the general public has been discussed in Paragraphs 7.33 and 7.34. It was shown therein that the amount of heat released is not enough to destroy the integrity of the water

shielding which is enough to keep such an incident from radiating anyone significantly from the prompt neutrons and garrmas. It is necessary to consider the gaseous activity which is given off, however. The quantities of gaseous isotopes expected to be released during three hours was shown in Table 7.34. In Table 8.29 these quantities are shown as pc/cc and their concentrations in the fuel receiving and storage area air are shown assuming that it is diluted with the 11,000 cfm of ventilating air which is drawn through --.......--

..........

Revision 1, Oct. 29, 1962 Revision 2, Aug. 20, 1964 l a b .,_ .. -----.... _ ... Table 8.28 The Prompt Neutron and Gamma Dose at the Outside Of a .Normal Concrete Shield From a Nuclear Reaction of io*~ Flsslonsa, 6 Concrete Shield Thickness, Ft Dose at Outside of Shield, rem Metal Nuclear Nuclear Reaction in Reaction Aqueous Solution 3 4 5 6 88,000 317 17.0 0.960 0.059 5,200 23 1.9 0.14 0.012 The dose rate may be calculated for any other number of fiss i ons through the use of a direct proportion.

ORNL-CF-61-7-39, "Rad iochem t ca 1 Fae fl i ty *Hazard Eva 1 uat i on 11 , E.D. Arnold, A.T. Gresky, and J.P. Nichols, July 10, 1961, Page 6. :J!*** .,: _ .. ,

... T~.b le 8.'29 Gaoeous Aettv1tl~~-Lo~t into Fuel Receiving and :, (Pie

_ Storage .Area, 'Durlna,., Assumed Cr.iticalitr

£~c~a-ent ... A*t , ti'vi"ty Re_le.ased Cone, Iso;t ope . . v.c/sec µc/cc Kr-85m 2300 4.4 X 10-4 Kt--85

93 1.8 X 10-5 Kr-88 930 1.8 X 10-4 <I-131 9,.3 1.8 X 10-6 I-132 1200 2.3 X 10-4 I-l3~ 200 3:, 8 X 10-5 I-134 6000 1.1 X 10-3 I-135 600 1.1 X l0-4 Xc-l:33m 20 3.9 X l0-6 Xe.-133 330 6.4 X 10-5 Xe-13Sm 5.5 X 104 1 X 10-2 Xe-138 1.5 X 105 2.7 X 10-2 MPC 6 X 10-6 l X 10-5 9 X 10-9 2 X 10-1 3 X 10-S 5 X 10-7 1 X 10-7 1 X 10-5 th*t area. These concentrations r*nge from twict the 40-hour MPC for Kr-8~ to 2000 times the IF(: for I-134. These IIPC'a are for continuous breathing and. ~*n be scaled up or down with time. Taking the I-134 as controlling, the room would have to be evacuated within 1/2000 of 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months or in just under one minute in order nQ~ to exceed one week's able inhalation.

It w.ill certatnly be possible to evacuate this room in leas than a ainute~and an event such as this would be imnediately obvious\~

anyone in the r~om~ Presumably there would be a visible fJaah. lbni*tors would trip and alarm, and the f~e r*pool itself would .be visibly agitated *. Aftet' evacuation, personnel could put on supply;. air equi~nt in other parts of the bulldlng*before enterlng.t~ take remedial action. Chemical_ E5>"loslon 8.3~ . In Paragraph

7 .. 35 a ~hemical e~piosion* is a~sumed which rupture* a*tank.con~i~ing an eht ftr~ day's tl: \arge of activ'ity leis. *l ne:.'ga*s,o~s );~tivity*:.(slnee, in arder to h~ve the full day-'*s

charge .-i fi sotution,* 1.t wlll ."I * * , # -~' A * ,* -, have to. hl*ve been thr.ough the:* dissolution r step ,*during which ttl*e geieous* .activity *11 I'ost*l *. ~~The**cell* ventflation system ha, , '.bti~n designed. to ~w.!thGtand : the* Eif~e~ts : of *such an* explos.iop. ;g'.~g ai 1 the -Veriti*J;atio*'

system i& maintained, no acti-.:1'.'ty should~ ge.:t OU~ of ~th~ cell. in which the explo~Jon took place_ except !.ome th~t wo_uld ~* lost during .the per'i od of oyerpias~ur

.&*,: f~ll~lng. : th.e. : exp , losion. Thi, period* -is e;tl-.ted to* .bii about bne -second. The calcu l ations of

App'in , 4ix *.8.;i~

sliow.'.that*-; °f'or *all conditions e>,ecept a di;ect* recy~le -'of ~~<:ic'* disc~.a:r .. ~e. into the. air int~ke,: th,~.* w~*11 be '1it.gllg.lble; conce_nt1ation of activl t , y in ~he .~uildlng .at.r., I~ thf u~~ikely t j\'iot: that such a recycle and an explosion* should:* _co.ltipide *and using a. calculative method analogeys to t~.t shown -1'1 App~ndix 8.~5,**, the. con*centration. of , . , unfilttred. a d lt~ act1vl~¥ at thr th~oat ,-of the .. it.ck would be* * * -* *Q~!~ g/min x 0.45 curie/gram x 106*11e/cc *-32,000 cfm x 28317 .. cc/cf * * * * <'0112 x .l0'."" 5 1-c/~c As in Appendfx 8: 25. a dilution: fa~to:r of 1/10 from stack to 1ntak6l and a frlfque~cy

.Jactor of: 1/10 were used. :!'he rt'.sultb1g .e 4?n*entratlon l,n the buildiri_g would be f!~2. x 19":'7 , p.;.J~c. This is about 100 times *the *MPo asiUSMd by Olk Ridge ,. for ml~l_ed fission products a.s** ~ero(e,ls an'd 1 t 1mp.~it% ~hat **about 25 mi~utes would be av<<~la61, for evacuating the 1 ,~1nt; Thi& is more than ad~quate.' Revl1lon 1, Oet. 29, 1962 Revision a, Aug. 20, 1964

I I I I ' * . I . t I Table 8.JO Ga1eou1 Activities Lost Into F~~I Receiving and e *Storm Aru purl~ Al1*a

l1pt0N Kr*8Sa Kr*8S Kr*N 1-131 ,~112 , ... ,, 1-134 1-135 Xe* 133nt Xe-, 133 X**13Sm Xe-138 Crtl*cal I ty Inc dent Activity Refused* Cone, ~c,,ec JJ.Clcc 2300 4.4 X Jo*4 ,,. 1 .a x 10*5 *930 I .8 X 10*4 ,., 1.8 X 10*6 1200 2.J X 10 ,*4 200 J.8x 10*5 6000 I.IX 10*3 600 I.Ix 10*4 20 ].9 X *~-, 330 6.4 X Jo*S 5.5 X. 104 1 X 10*2 J.4 X 105 . 2.7 X 10*2 . ,. MPC 9 X Jo*9 2 X 10*7 ] X 10*8 5 X 10*7 1 X 10~7 -

f f

r t e.31 Although we believe that the ventilation ayetem will be maintained 1n opera*tton lf an* event such as this thould oceur the possibility th4t it does not continue to function muet be considered.

Aa*ln the dl1cu11lon of *Paragraph a.~ we find th11 situation

.difficult to analyze and for.the same rea1on1.* In this case it 11 certain that the cell would: be pressurized for a sho~t time, perhaps several seconds, an~ that during this. time &Olbe activity would escape~ If the buildi'ng ventila , tion system were not functioning, the air *in the building wil~ be essentially stagnent.

In the ininediate vicinity. of the cell quite hJ 1 gh conc.ntrations can be hypothesized depending on the tions chosen. Howev*r,*it seems difficult to hypothesize a mechanilll whereby*t~is actlvity will *spread , very quickly from the iaaedlate area if th'e building air is n&>t moving. An explosion~would alert anyone close to it and the lnnediate area. would.be evacuated in a matter-of secondi. Reconnaissance

~nd remedial action would then be carried out with supply-air masks. Failure of Iodine Ri!llOval Equipment* 8.32 As in all*of the situations in which activity is p~t*up the stack, here ag a in there ls no hazard , inaide the.plant at all except* in the unlikely case ~i direct recycle the concentration of iodine that might be found in *the building air wo,...~d *be, . *" ' I * *20 .. Jl e/s e e x-0.1 _ dilu~b}JI factor x 0.1 frequency

_8 . -* ,:~ .r.5\~ .. =10.7" cc/sec = 1.3 x 10 JJ.C/cc 't h& 4!9-houi* Mf?C fG i i:-iodine-l~I is 9 X 10-9 ~/cc which is lowe~~:~han the_ab,ove 4 calculated concenttatfon by only about 30 pe rc ent 9 Tn.P._s-.-*th i.~ co*ncentration

-could. be permitted

~or over two days. I t* is*11 nlikely that the recycle would be in ju s t*the r~ght p9sit lQ n for more than an hour during t he day. In *any 'e v~nt .;t his *.'e9.ncentration would trip all the buildin g °ii ir .;jp}to 11i s_an i1. 'the inc i dent would be c ove:red tnote , readi l-y*an d ,*t ti e dissolver shut down. '" .... .. '-! .,. Mipor i\ccident*r;,: * .. ' .. <-*. , .. t.. . "' . . . .,. _. -' ' 8.33-1 We do n~~~**bel ti:&ve .. t Q at the accidents which have been discus s ed-i n se ei ions VII' a m"d VIII will occur. It is a . 9ood , deal mo~~ li~e~y~ ~owev tf , ttjat ~~~i ng *the c~urse ~f the opt2:a~f"on of.-,-this l;\lant th~r e ~111 , ~e. a ~u~er of .much more minor *occurr!nces wJ)l9h*~~!t no hazard.a.~ all .to the general publio but.w~t~~. if J t.hey ,er* not handl*d*properly, could lead to addi t~onal exposure~*ef th~ p l ant personnel.

Such * .. Revi1iqn 1, Oct. 29, 1962 Revision 2, Aug. 20 1 1964 event, might be illu1trated bya a. Spil l ing of activity, particularly analyti~al saap'te* or special samples such as waste tank 1amp~e. b. Tracking of 1pllled activity from one place to another_. c. Pulling activity into_jet ,team lines by improper venting. d. Leaklng w11te lines in diversion boxes. e. Spilling of product solution * . The probl* with all of *this iype o*f event is the same. In one way or.another they lead to an inc~ease in the background radiation wt,.t ch the worker may receive. Thir, is undesirable since in n~clea;.work one wishes to avoid a~y unnecessary radiation.

It is also undesirable

'since it is important to keep 1 'the "'?~*rating

bac'kgrounu" **~low as possibltt in order to leave a cu~hion with which to carry out *the required maintenance work. Thero are s,veral lines of defense against this sort of probiem and they a1e the same for all of themt a. First, the plant~has been designed to eliminate, insofar as possible, the rtecessity for handling eve~ small amounts of activity.

b. Second, the operating rules are designed to eliminate every possible exposure.

c. Third, the fixed monitoring system (see Paragraphs 6.66 thr.ough .6.76.} is d~signed to det@ct increases in either.background radiation or air concentrations.

d. Tfie fixea monitoring system is.backed up by a formal mobile monitoring and survey pr.ogram (see Section fx). e. Each employee ~ill wear ~oth meters and film badges._ He will be trained to check his own expoiure rate frequently and to use portable monitoro himself to that 'he* need not rely completely upon Health-Safety Coverage.

Revi,ion 1, Aug. 20, 1964 J Thus, th*** minor accident, thould not go undetected.

None of the1e accident, could credibly re1ult in expoaure to plant personnel in exce11 of the limit* set forth in lv CPR Part 20. Suniury . 8.34 In .*this section we have shown that we are able to operate the NFS plant under ~11 normal conditions withln the requirements of 10 CFR Part 20 as they pertain to the protectlon

of the.-plant personnel.

This includes protection from*external radiation, inhalation, and ingestion.

Both the operation of the proceis proper and all the necessary maintenance operations are included in this statement.

8~-3~

T~e same aerie-. of hypothesized major accidents that were di1cu11ed in Section VII have been considered from the standpoint of ~he protection of the workers in the plant. It is shown that each of th-se unlikely event, could be sustained without undue risk of exposure to plant personnel.

8.36 Final.ly the problem of minor accident, that could lead to increases in the bactground radiation received by plant personnel is considered and the multiple aeries 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.

Revl1lon 1, Aug. 20 1 1964 IX PLANT OPERATION 9.1 Detailed in thia aection are the following itemsa The org,nizational make up of the Spent Fuel* Reproce11ing Plant1 11pect1 of act.iniatratlvc control and procedure*

in various operations of the plant1 Training Prograa, Health and Safety Prograa1 Fire Safe~y PrograrA and Eatergency Procedure11 the u1e1 of the Ol)9rating Procedures and Lotter& of Authorization1

-discussion of Maloperation1 and the use of Maintenance Procedures.

Organization Plant Manager 9.2 The Plant Manager is responsible for all activities at the plant and 11, 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 Director 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 ~echnical Services Manager is concerned with the technical soundness of the operations proposed, the surveillance of material, and particularly, as a member of the criticality group, the maintenance of a critically safe system. He generates applications for license revisions.

He reviews proposed SOP's and Letters of Authorization to confirm compliance with the license. Plant Criticality Coaaittee 9.3 The Plant Criticality Connittee consists of the Plant Manager, the Technical Services Manager, the Health and Safety Director and the Production Manager. This conaittee 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 crit1cality problem is involved.

Each member give, particular attention to the function that he represents.

The usual sequence for review 111 1. Production Manager 2. Technical Services Man*ger 3. Health and Safety Director and 4. Plant Manager. pt st; ....

1 ' J I Admini1tratioo 9.4 The main function of man1gement 11 to safely and economically

dadniater all operation*

relative to the plant. The Plant Manager, who h11 overall re1pon1ibility for Plant Operation*, haa delegated certain re1pon1ibilitie1 a, enumerated in paragraph 9.2. In addition to the above, other delegation of responsibility is a, shown on the Adlliniatrative Organi:zation Chart, Figure 9.4. Operating Procedureb ,nd ~etters of Authorization

9.5 Processing of all special nuclear material handled under the license is done in accordance with the criteria set forth in the license. All operations in the Spent Fuel Reprocessing Plant are done in accordance with approved operating procedures which define the methods to be used and incorporate criteria contained in the license. . . 9.6 lt will be the responsibility of each employee to read, understand and follow explicitly the directions contained in the Standard Operating Procedures for jobs which he is called upon to do. It is the responsibility of each supervisor to know the Standard Operating Procedures which apply in his area, to have copies of these SOP's available for employee$

to read and to be certain that individuals under his supervision read and understand each procedure.

It is Management's re&P9nsibility to review and re-issue SOP's as necessary t9 reflect changes in the process and*to insure that the instructions contained in SOP's represent a safe and efficient method for accomplishing the work.

9.7 Special nuclear material is received into the plant and processed by Approved Letters of Authorization which state the operating procedure(s) to be used, special handling where required, the customer for whom the processing is being done, the container(s) in which the material will be found and the material to be used as to type, enrichment and weight. Before a Letter of Authorization can be used it is independently review&d by each member of the Plant Criticality Conaittee to assure its conformance with approved license criteria.

It is the responsibility of each employee to read., understand and follow explicitly the directions contained in the Letter(&) of Authorization.

It is the responsibility of each supervisor to have a c opy of the Letter(s) of Authorization available for individuals under his supervision to read and to be certain that they read and understand the instructions contained in the Letter(s) of Authorizatio

n. It is Management's responsibility to issue Letter(s) of Authorization thereby scheduling work throughout the plant. Normally, process eng i neers under the supervision of the Production Manager or the Te c hn ic al Services Manager-draft the procedures or authorization.

These engineers serve also in technical

, liaison and guidance in production and they conduct and supervise eng i neering development.

9.8 The general administrative philosophy will be to establish standard procedures for as ns..,ny situations as possible and to control the effectiveness of these protedures by means of regularly maintained logs and check-off lists. These procedures,together with their

,; r-, ,.~: Assistant Plant Manager I . Security Officer ......-t 'Industrial Relations Manager Office llanaaer Production Manager I Assistant Production Man*ger ineer Mechanical Enaineer ... ..... ----*-------

....... -* \; Figure 9.4 Plant Organization Chart Plant Manager I Assistant to the Plant Manager Director lledicar D1nctor Technical Director Technical Services llanaaer Accountability Officer Analytical Services Manager Shift Supervisors Health and Safety Director I supporting 1091 and check-off liat\will be 1ubject to regular, but random, inspection by higher levels of autho~ity.

For instance, certain routine exaainatlona and mea1ur ... nt1 will be carried out dally according to approved check-off ll1t1 and duly logged. In th*** ca*** the next higher level of aupervialon wtll, once a week at a random tille, follow through the apecific procedure and deteraine that it 11 being properly carried out *. Once every two months the next higher level of 1upervi1ion will do likewise.

Once a year these .saae procedures will be observed by the highest level of authority.

Personal responsibility will be empha1lzed by having each one of these inspection*

recorded by signature and date1 the 1091 will be kept as a permanent record. In addition, duty lists, addresses and telephone nuabers will be maintained, and selected group, of off-drjty per,onnel, at all levels of authority and skill, will be required l to keep the plant infol'll8d on their whereabouts at all times for emergency call. Training of Plant Personnel 9.9 The initial staff cadre will be largely aade up of people with extensive experience in the handling, processing and 110nitoring of radioactive materials.

This. group, under the Training Director, will conduct the itraining courses for all additional employees.

The curriculum (see Appendix 9.9) will be directed toward the education of certain plant personnel in the processes and related operations in such detail as to* ensure complete familiarity with the equipment, its function and competence in its operation.

9.10 It is the intent of the training program to enable process operators to successfully satisfy AEC requirements for operators' licenses by test and examination.

Approximately 75 operators will be so trained by permanent or temporary staff membe~s. Initially, three types of operators will be trained for work in three different types of areasa 1. Manipulative Processing Operations

2. Chemical Processing Operatio~s

3. Control Operations The cadre, in addition to serving as the faculty, will take these and additional cou~~e* designed to satisfy AEC requirements for Senior Operators' Licenses.

Certain employees such as watchmen, secretaries, etc. will be exempted from most of the more technical aspects of the curriculum but all employees wiil be exposed to*a radiological familiarization curriculum.

Written examinations, graded according to level of responsibility and work exposure.will be conducted. . 9.11 The curriculum will include an introduction (c011pri1ing details of the background and descriptive material of the plant) details of the processes, health and safety, instrumentation, equipment description and usage, .. chanical manipulation, process control, process maloperation, decontamination procedures, waste treatment, emergency measures, bility, economic and criticality conaiderations and lay chemistry and . i f i I *, J t I O* . I

l -----*~-~----~-~-~

:-:-** --**-physic* 111ocl1ted with reactor operations and chemical reprocessing.

In addition, the curriculua for the cadre and others preparing for Senior Operator,*

lice~*** will include the conditions and liaitations in the facility license, the de1ign and operating 11.IDitations in technical 1peolfic1tion1, the *chani111 for any change, in the llaitations in the llc*n** or specifications and more advanced 1tudy of chemistry and radioactivity.

The training prograa for plant personnel will be a continuing one. Regular process operators will be given, periodically, a reorientation exposure to radiation safety and to processes and equiplftnt involved in their particular plant specialty.

New eaployees will be indoctrinated by training as are the initial employees and will be required to pass the saae NFS exaainations in addition to AEC license examinations.

Training of Outside Organizations 9el2 Partly as a aatter of public relations but primarily to obtain effective and non-panicky assistance if an ell8rgency requiring their cooperation should develop, local town, county, state police officers, fire departaents of the area, civil defense organizations and elected officials will be invited to lectures at the plant. The subjects covered will be mainly those connected with protection of the public and will be designed to establish methods of liaison and cooperation if desirable and necessary under hypothetical ell8rgency conditions so that assistance la most effective and radiological hazards to outsiders are miniadzed.

Health and.Safety Program 9.13 The Health & Safety Department ls charged with the responsibility for protecting plant personnel from all job hazards and the public from hazardous quQntitles of radiation and radioactive materials.

' Within the scope of this responsibility the Health & Safety Department wills 1. Monitor for radiation and contamination all plant areas and operations, (see Appendix 9.13 for equiP_Mnt)

2. Monitor for radiation and contamination, areas external to the plant1 3. Approve procedures , for work with radioactive materials1

4. Establish emergency procedures, 5. Establish liaison with all other departments and advise them in matters pertaining to health and safety1 6. Supervise the receipt and shipment of all hazardous materials1

7. Provide curriculum and teach aspects of the health and safety program, a. Establish and maintain plant fire brigades trained to cope with radiation area fires, 9. Conduct a continuing safety training program for all employees, 10. Conduct inspections of all areas for fire and safety hazards and institute corrective action when necessary, 11. Maintain complete, accurate records of personnel exposure, radiation-contamination conditions in and around the plant, and perform radiation instrument calibration.

Health and Safety Organization 9.14 Specific responsibilities for members of the Health and Safety group are as follows1 Health and Safety Director Plan, organize and supervise the work of the department.

Maintain close liaison with the Medical Director advising and seeking advice concerning the employ9es' health and welfare. Maintain close liaison with other departments and advise them in matters pertaining to health and safety. Maintain complete, accurate records of plant, personnel and environmental radiation-contamination conditions.

Administer health and safety aspects of training programs.

Organize and train plant fire brigades.

Inspect and maintain fire fighting and emergency equipment.

Prepare material for use in safety meetings.

Conduct fire and safety inspections.

Lead Techniciaq Perform routine and non routine monitoring tasks as directed.

Write complete, accurate reports of conditions observed.

Calibrate and check monitoring instruments.

Obtain and count air samples. Perform safety inspections.

Participate in shift safety training programs and safety metttings.

Maintain exposure and survey record files.. Check gamma dosimeters and record results. Prepare film badges for distribution and processing and record results. Technician

-Medical Perform routine and non routine m&dical tests on employees including processing bio-assay specimens.

Receive environmental and moni"toring samples and prepare them for counting.

t I I i . , I I l 1 C C j Technician

-Shift Perform routine and non routine monitoring and inspection tasks as directed.

Participate in shift safety training programs and safety meetings.

Write complete, accurate reports of activities.

Radiation Area Work Procedures 9.15 All radiation area work is governed by procedures approved by responsible persons in Productio~

Plant Engineering and Health and Safety. It is the intent of these procedures, in accordance with NFS policy, to incorporate sound industrial safety practice and to maintain exposure of* employees to ionizing radiation and radioactive contamination 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 and other safety equipment as required.

9.16 For the purpose of defining radiation areas, the followi~g zones are establisheds Zone I All areas beyond the site perimeter boundary; Zon~ II All areas within the site perimeter boundary which are normally free of tion in excess of 500 d/m alpha and 0.05 mrad/hr beta-gamma; . Zone 1II All areas within the site perimeter boundary which may have detecta9le tion but in which the radiation level is normally less than 100 mrem/hr and the ination level is not significant; Zone IV All areas within the site in which the radiation level exceeds 100 mrem/hr or in which cant contamination exists. Zoning of the plant and site will be the responsibility of Health and Safety. 9.17 The General Regulations for Radiation Area Work will apply to all work procedures.

See Appendix 9.17 for listing of equipment.

General Regulations The minimum requirements for protective clothing ares a. For entry to Zones I and II, no protective clothing is required J ,_

I Ji@W b. For entry to Zone III areas, for inspection only, the minimum protective clothing required shall bea Laboratory coat, shoe covers and glovesJ c. For entry to Zone III areas to perform*work, the minimum protective

~lothing required shall bea Coveralls, shoe covers, gloves and cloth hatJ d. Protective clothing required for entry to Zone IV will be specified on a Special Work Procedure, No one will be permitted to enter a Zone IV area until a Special Work Procedure has been.completed and signed and all provisions of that procedure have been implementedJ

e. Respiratory protection requirements will.be posted in t~e "hot" lobby. The minimum requirements for personnel monit~ring area a. ~or entry to the plant, the minimum requirement fo~ personnel monitoring shall bel --Badge b. For entry to Zone IiI areas, the minimum requirement fo~ personnel monitoring shall be1--Badge, dosimeters and dose rate type radiation survey meter. c. For entry to Zone IV areas, the personnel monitoring requirements will be specified on the Spe~ial Work Procedure.

The exlting proce~ure isa a. When leaving Zone-IV areas the minimum requirement for personnel survey shall be A complete clothing and body surv~y by Health and Safety PersonnelJ

b. When leaving Zone Ill areas the minimum requirement for personnel survey shall be

A ~omplete self survey at the station monitors located at the Zone III -Zone II boundary, c. When leaving Plant Zone II the minimum requirement for personnel survey shall be A hand and shoe check using the hand and shoe counters and station mor,: tors in the building lobby. This survey shall also be made before I entering the building lunch*room. The rules for radiation area conduct area a. No smoking, eating, drinking, or chewing shall be permitted in Zones III and IV. Zone II Plant areas in which smoking is not permitted will be so designatedf

..

0 G b. Every surface and every piece of *Jquipment in Zones III and IV and every tool or artic l e taken into these Zones shall be regarded as being contaminated until surveyed and released by a representative of Health and Safetyf c. All the provisions of applicable work procedures shall be read, understood and followed explicitly by the personnel performing the workJ d. Each employee is respons.w~e f or the care and treatment of equipment issued to him and for .his conduct in the performance of assigned work. Careless or willful handling of equipnent or misconduct on the job will not be tolerated and will constitute grounds for difmissal.

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 tweive months from the date of issue.* On, or in advance of, the date of termination, the procedure is revJewed by responsible persons in Plant Engineering, Production, and Health and Safety, changed as necessary to reflect current working conditions, afid re-issued with a new termination date. 9.19 For work of a special or unusual nature or work in areas or on equipnent which does involve significant radiation-contamination, a Special Work Procedure is issued. Each Special Work Procedure is valid for one shift only. Approval of responsible persons in Plant Engineering, Production and Health and Safety is required prior to the start of any work and before work can continue on succeeding shifts.

Job Planning arid Scheduling 9.20 Each day responsible representatives of Plant Engineering, Produ c tion 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 advance ~f the work. The Work Schedule lists the personnel assigned to each task, the time and* place to meet for each job, the estimated duration of each job, the applicable procedures govern i ng the work and other information of general interest.

9.21 The Plant Engineer is responsible fors a. Estimating the time and manpower required to accomplish each maintenance Jobi b. Assigning maintenance personnel to each scheduled tenance jobl,. c. Assuring that all maintenance personnel read and stand applicable work pro c edures and are thoroughly trained in radiation-cont a mination works

d. Assuring that schedulad maintenance personnel stand what work is to be accomplished and that the proper tools and equipment, in good condition, are available in advance of the jobf e. Assuring that assigned maintenance personnel are available at th~ place and time indicated on the schedule.

9.22 The Production

.Manager is responsible fors a. Establishing priority of m~inten 9 nce in the pl~nt1 b. Determining what the effect will be ot scheduled maintenance work on 'pl&nt operations1

c. Arranging for equipment or area shutdown as necessa~y to accomplish the scheduled work1 d. Arranging for pre-maintenance decontamination and/or shielding as requireds

e. Assigning operating personnel to scheduled jobs as required,;*. f. Assuring that all operating personnel read and 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 job1 h. Assuring that assigned operating personnel are available at the place and time indicated on the -schedule1

i. Issuing the work schedule following each planning and scheduling meeting. 9.23 The Health and Safety Director is responsible fora a. Determining what radiation-contamination conditions and/or other spacial hazards will be encountered in performing the scheduled work1 b. Determining whether or not a Special Work Procedure will be required for each scheduled job and if not, which Extended Work Procedure will apply1 G c. Determining requirements for protective clothing and/or other safety equipment for scheduled work and assuring that such equipment, in. good condition, is available in advance of the work*~ d. Scheduling and leading a pre-job conference if required*

e. A-i. *ning Heal th Physics personne 1 to scheduled jobs as l.dquired1

f. Assuring that all Health Physics personnel read and understand

~pplicable work procedures, are thoroughly trained in all pha , se~ of radiation-contamination work and ~re trained and equipped to respond to unusual or emergency conditions, g. Assuring that assigned Health Physics personnel are available at the place and time indicated on the schedule, h. Initiating Special Work Procedures following each planning and scheduling meeting. Unconditional Release 9.24 Release surveys of equipment are the responsibility of Health and Safety. Any item leaving Zone IV or Zone III to go to Zone II or Zone I or any item leaving the plant site from any Zone, must be accompanied by a completed Unconditional Release. The original of the release accompanies the equipment, ~nd one copy (in the case of an item leaving the plant site) is presented to the Plant Security Guard who is responsible for enforcing this procedure.

This procedure also applies to commercial vehicles and railway cars. The Unconditional Re~ease states the radiation-contamination levels on the items described, and releases them with no conditions or restrictions as to their use. Conditional Release 9.25 The use of a Conditional Release is normally restricted to equipment which is not to leave ~one III. For example, a process pump which is to be taken to the Equipment Decontamination Room or the tenance Shop for repair will require a Conditional Release. The Conditional Release describes the item released, lists the radiation-contamination status of the item and lists any special precautions which must be taken* for handling, dismantling, and repairing the item. Lock and Tag Procedure I 9.26 The Lock and Tag Procedure is used to lock out valves, controls, and switches, the unauthorized or inadvertent use of which could cause process upset, damage to facilities and equipment or personal injury. Each department will have its own locks and will be responsible for applying

/ ........ ~-*w,,.*"""""~""" ...............

,.......~

..................... .............

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

____ __ -* -l locks to equipment

~s required for employee protection even if this practice rasul t§ ln several locks* on the same switch. The responsibility for removing r ocks will rest with the department head (or his delegated assistant) of the department responsible for applying the lock. Non :ompliance with this provision will not be tole;ated.

Maintenance locks are normally applied only during maintenance

~ork on equipment and are removed when the work is completed.

The tags are used to indicate the reason for the lock and to warn all personnel of the possible cons~qUences of violating this procedure.

Safety Hazard Tag Procedure

.' 9.27 Any NFS employee is responsible for tagging or posting any equipment or con~ition which represents a safety hazard and/or unsafe wo~kifig 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 are discussed a number of administrative limits of radiation exposure for the NFS Blant. It is expected that these limits may be modified as plant experience dictates.

NFS employees may be exposed to radiation up to the limits stated in the following table with the approval of the employee's inunediate supervisor, Table 9.28 Rems Per Calendar Quarter a. Whole body; head and trunk; active blooc forming organs; lens of eyes; or gonads-------------------------------------------

1-1/4 b. Hands and forearm; feet and ankles-------------------------------18-3/4

c. Skin of whole body---~--------~-~---------------------------------

7-1/2 Whole body exposure to penetrating radiation in any 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period .shall be limited to 0.1 rem or, if approved in advance by the Health and Safety Director, 0.2 rem. Planned single e~posures in excess of 0.2 rem mµst be approved in advance by the Plant Manager. In emergencies involving the life of personnel, it shall be the bility of the NFS S~nior representative present to determine and authorize, if such be his decis i on, entry into higher fields of radiation.

9.29 The whole body dose and skin dose is evailable from badge readings.

The dose to extremities is controlled in the field. If the dose rate to the hands and forearms or feet and ankles is more than 15 times the dose rate to the whole body, the time limit for the work is based on the dose rate to the extremities.

Vli th prior approval*of the Plant Manager and the individual concerned, a n employee of NFS may be 0 G permitted to receive a dose to the whole body greater than that permitted under paragraph 9.28 provided that. a. During any calenda~ quarter the total whole body dose shall not exceed 3 rems, and b. The dose to the whole body, when added to the accumuleted occupati_onal dose to the whole body, shall not exceed 5 (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 ior intentionally causing erroneous film badge or dosimeter rea~ings is dismissal.

9.31 NFS e~ployees, who have been certified in the use of radiation monitoring instruments by the Health and Safety Director, may in the*course of their normal duties, self mbnitor 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 all Zone IV areas monitoring for any entry shall be by Health and Safety Technicians.

In no case shall employees enter an area in which the dose rate exceeds 2 r/hr unless prior approval of the Plant Manager has been obtained. (See 9.28) Maximum Permissible Levels of Radioactivity 9.32 The.maximum allowable surface contaminations for the West Valley Plant are shown in Table 9.32a. The Maximum Permissible Concentrations in air of some radionuclides expected to be encountered in the West Valley Plaot are shown in Table 9.32b. The Maximum Permissible Concentration in on-site, nonpotable wate1 in Buttermilk Creek of some radionuclides exp~cted to be encountered at the West Valley Plant are shown in Table 9.32c. Air Sampling . 9.33 The air sampling program provides for the evaluation of alpha and beta-gamm*a air contamination. in all building areas, the plant site and the site perimeter.

Included in the program are 54 plant area particulate samplers, 19 remote in-cell particulate samplers, 7 in-plant continuous air monitors, 1 plant site sampler and 3 site perimeter air monitors.

This equipment is described in Appendix 9.33, l s located as per Figure 6.67, and discussed in Faragraphs 6.66 to 6.67. 9.34 The filter paper u~ed for particulate sampling is *Whatman #41 or equal, two i.nches in diamC!ter.

Whatman #41 filter paper has a collection efficiency of 98 per cent for 0.18 micron particulate or larger at a flow velocity of 50 centimeters per second. To obtain this flow v~locity a minimum flow rate of 60 liters per minute is used for in-plant air samplers.

Self absorption in Whatman #41 paper is zero for beta and about 0.3 for alpha.

Table 9. 32&

Maximum Allowable Surface Contamination for"West Valley Plant Smea'rable Non-Smearable Alfha Beta-Gamma Alfha Beta-Ganma Surface As Shown/100 cm2 As Shown/100 cm2 d/(m)r100 cm2} d/(m)r100 cm 2) 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 Commercial Vehicles 500 100 c/m 500 0.4 mrad/hr Zone I Zone I limits are per 10 CFR -*20, Appendix B, Table II Zone II' 500 100 c/m 500 100 c/m Zone III 5,000 10 mrad/hr 5,000 100 mraci/hr Zone IV* 50,000 2 r/hr 50,000 2 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

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

___ ...,....,._..._.__

---.. ---....-_____ ..,.._., ___ _._ .... ._. __ .. ,,.. ---**--'.,.._.,..NI;.

............. .........

***----------1 ---------Table 9.32b G Maximum Permissible Concentration Cuc/ml) C Mixed Fission Products No respiratory protection Full face filter mask Supplied air mask Strontium-90 No respiratory protection Full face filter mask Supplied air mask Cesium-137 No respiratory protection Full face filter mask Supplied air mask Plutonium 239 No respiratory protection Full face filter mask Supplied air mask Natural Uranium No respiratory protection Full face filter mask Supplied air mask High Enriched Uranium No respiratory protection Full face filter mask Supplied air mask Iodine-131 No respiratory protection Supplied air mask . Krypton-85 No respiratory protection Supplied air masi ootnotes to Table 9.32b l X 10-9 2 X 10*8 Above 2 x 10*8 3 X 10*10 6 X 10-9 Above 6 x 10-9 l X 10-8 2 X 10*7 Above 2 x 10-1 2 X 10-12 4 X 10-ll Above 4 x 10-11 6 X 10*11 1 X 10-9 Above 1 x 10-9 l X 10-10 2 X 10-9 Above 2 x 10-9 9 X 10-9 Above 9 x 10-9 1 X l0-5 Above 1 x 10-5 Maximum Permissible Concentrations for other radio-nuclides is as indicated i~ 10 CFR-20, Appendix B, Table I. When a mixture of radionuclide$

is encountered and the identity and concentration of each radionuclide in the mixture are known, the Maximum Concentration is derived as followsz If radionuclides A, B, V, are present in concentrations ca, Cb, Cc and the applicable MPC's are MPCa, MPCb, and MPCc respectively, than the concentrations shall be limited so that the following relationship existsz ~+_9L-t...f£_

~l. MPCa MPCb MPCc I ms I L p .,._ ............

_ ......................

~.-.,._._.

____ . .___ .........

*~-*-*----* -... . ______________

____ ..,,,, __ _ Table 9.32c Maximum Permissible Co~centration (ric/ml) On-Site-Buttermilk Creek Off-Site Cesium-137 4 X 10-4 2 X 10-5 Cobalt-60 1 X 1 o-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 Ruthenium-106 3 X 10-4 1 X 10-~ Strontium-90 4 X 10-6 ... 1 X 10-1 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.

C 9.35 Air *samples are collected and analyzed for radioactive material accor d ing to the schedule shown in Table 9.35. This schedule is subject to revision as experience is gained in operating the plant. Continuous air monitors are used in some occupied areas to provide an invnediate alarm should high ai~ contamination exist. The other remote samplers will be used occasionally to obtain very short, spot sample& of air contamination conditions in the cells. These remote samplers ares Miniature Cell General Purpose Cell Chemical Process Cell Mechanical Cell X-Cell l X-Cell 2 X-Coll J Product Purification and Concentration Chemical Process Cell Crane Decontamination Area Process Mechanical Cell Crane Decontamination Area 9.36 As each sa1'Dple 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, according to the schedule, they are brought into the Health Physics Lab, removed from the envelopes, placed in planchets and surveyed with portable beta-gamma a~d alpha detection instruments (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 *.1atural activity, it may be possible at this time to make a preliminary

~stimate of the amount of long-lived emitters on the sample. The concentrat~~n of beta emitters on the sample will be determined basecl 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 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months 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 i 1 n Appendix 9.37b. Any samples which, on the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months count, show less than 1 c/m alpha and less than 1800 c/m beta are -Jr ** ._.. ______ _ ,,

Shiftwise Hot Lobby Mechanical Operating.Aisle-west Ram Equipment Room -Chemical Viewing Aisle-north Ventilation Wash Room Process Sample Enclosu..re-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 GPC Operating Aisle-west GPC Operating Aisle-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 Produc~ purification cell Product Packaging 1, 2 and 3 Fuei *storage 1 and 2 Chemical Viewing Aisle-south Equipment Decontamination Room Chemical Operating Aisle-north Chemical Operating Aisle-south Lower Ex~raction Aisle-ea~t Upper Warm Aisl~-west

Upper Warm Aisle~east Off Gas Cell-2 OOC-NC Aisle Chem Lab*east and west Product Lab Emission Spec. Lab Mass Spec. Lab _ ~lpha Lab Stack Sampler Upper Extraction Aisle.,.st Upper Extraction Aisle-east Extraction CheMical Room-east Laundry Weekly Third Floor Office Second Floor Office Main Lobby -Maintenance Shop Utility Room Manipulator Repair Area Product ~ackaging CAM Fuel Storage CAM EDR Viewing Station ICR Air Lock Ventilation Supply Room Hot Lobby CAM . Mechanical Operating Aisle CAM VSR Access Aisle Off Gas Cell-3 Analytical Cell Decon. Area Alpha Lab CAM Lab ~ccess l i sle *control Room Extraction Chemical Room CAM Plant Area Perimeter-1 Perimeter-2 Perimeter-3 GPC Crane Room Mechanical Crane Room Chemical Crane Room _,_ e:n. ::---=-:-~: eL.._!~ .., .:-,-I + 71 I I

I I . .-,:*!.D ,> f':,:c'[.;~.

-~..:~~::,"':

~-'!!.-:.. -.. it(£$ ... -II.) -.. -' ~~~--*

G I I Q discarded.

These counts at maximum counting error, represent about 1% of MAC for plutonium-239 and ,trontium-90 respectively.

Samples which exceed either or both of the counting l\mits will be held for a final count. The final 30-minute count on in-plant air samples is taken a minimum of four days after saq:,ling to allow the natural activity to decay essentially to zero. All of the alpha counts are a~sumed to be .counts due to product and the concentratio

ns are calculated as followsa Alpha ~c/ml = c/m (l~~l x 10*12) Since the counting error for a 30-minut, count at 95 percent confidence level is i 10% at 10 c/m, the minimum detectable alpha concentration on a 24-hour ~ample is1 10 (1.31 x 10*12> 3 1.5 x 10*13 with! 10% acc~racy 86.4 Beta-G,.ama

-~~/ml= c/m (9.19 x 10*13} M3 9.38 Some in-plant air monitors are moving-filter type and the filt~r tape is not normally analyzed in the counting room. Portions of the tape may be counted and/or g8Jlllla 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 hou; !n low background alphat beta proportional system (see Appendix 9.39: a). The geometry of this system is 50%* for beta and 35% for alpha. The t i ackground is about 1 c/m. The concentration of beta emitters is determinft

~ as shown in Appendix 9.39b and the concentration of alpha emitters* is de tt ermined as shown in Appendix 9.39c. A log is kept of air sample results. These results become. part of the permanent record of radiation-contam

~.nation conditions in and around the plant. 9.40 Radioiodine activated charcoal filters from the stack and perimeter stations are analyzed*as follows* The filters are gamn1 scanned to determine if there are other 91111111 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 tt_len counted in one of the 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 furnished by the manufacturers.

The method used to determine actual counter geometry is described in the calibration section (Appendix 9.3~. The collection efficiency of Whatman #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 ~se in the plan t. The self absorption of alpha in w;aatman #41 can be determined by co~nting a filter, dissolving the filter, evaporating the solution on a planchet and counting the planchet.

The 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-gaama film badges are supplied to .each employee and all visitors to the plant through an arrangement with a conmercial film badge processor.

Badges are exchanged and read weekly for ~ost personnelJ monthly for administrative personnel.

This schedule is subject to change as operating experience is gained. Immediate notification by phone or wire is given for badges which show a dose in excess of 100 mrem. Neutron monitoring is accomplished on an area basis. N&utron badges are placed in the product storage and product packaging and handling areas to establisli and check the neutron dose. rate in these areas. The n,utron badges are changed monthly during startup but this may be changed to quarterly at a later date.

9.43 Each productio.1 employee and each visitor is issued a 0-200 mr ga11111a dosimeter which is reaij and the dose recorded during the shift following the shift on which it is usea. !he dose is recorded on the "Dosimeter Readings" form, and 1:s transferred later to the "Exposure ,. Record" card which is *also used to record badge readings.

The 5. *x.a*.inch card, designed to be used in a "Victor Visible" type file, contains ali of the information required by AEC Form 5. 'Each card r,presents

13 weeks exposure data. See Appendix 9.43 for the "Exposure ijecord" card referred 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 to the job. 9.45 Health and Saf~ty responsibil

, ity for product shipaents entails checking the shipping papers for Production dignature approval, for product specifications, accountability certification and surveying the shipping containers to insure conformance with all applicable federal, state, and local regulations.

The*signature of an authorized Health and Safety representative on the shipping papers will constitute approval to ship.

o 0 -.,-r------

--* **-~--*-______ ,, _ _,...,......

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

.. ....,._., .... ..,. .. __ .,. ......... --~ .. . , " 9.46 With*the exception of Zone IV egress, personnel survey* are the reapc;naibility of each employee.

Health and Safety will audit the frequency and adequacy of such aurveya. Peraonnel fou~d in a Z9ne II or Zone I area with contaminated clothing may be aubject to di1mi11a1.

9.47 A regular schedule of routine aurveya will be performed by Health and Safety *. The routine survey program ia designed to aupplement the reports of radiation contamination conditions which are encountered during maintenance and ot11er work, and to insure that all plant areaa are surveyed on a regular basis. Each routine survey 18 described in able detail on the "r.outine survey" form (Sc* Appendix 9.47) which will serve as a guide for the Heal th and Safety peraon,,el performing the survey. A list Qf routine surveys 11 a~own in. Table 9.47~ A written record is. made 9f every survey performed by Health and Safety peraonnel.

This record which 11 executed on pre-numbered aurvey log 1heet1 become* part of the permanent record of radiation-contamination condition, in and around the plant. Environmental Survey P~ogram 9.48 The environmental survey program, pre-operational and post-operational is divided into three ~ategoriesa

1. Atmospheric monitoring including air particulate monitoring, 2. Water monitoring including surface and ground water samplings

3 *. Earth and biota monitoring including samples of silt, mud, plankton*

fish and shellfiah from Buttermilk Creek and Cattaraugua Creeks soil, vegetation and milk sample* from the site and surrounding area and small game from the site. 9~49 The pre-operational

program is divided into two phaaes1 the first phase, started in the spring of 1963, to establish on aite gt~** activity background w i th 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 ~ill continue into the poet-operational period. Both Phases aro detailed in Appendix 9.49. A sunaary 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 11 accomplished by treatment of was.ta gases before release, continuous monitoring at the point of releaae and environmental monitoring to determine th~ effect, if any, of released activity in the environment.

Waste gas treatment is discussed in some detail in Section VI, Paragraphs 6.66 to 6.70. Prefilters, air scrubbers, silver reactors-and high efficiency filtere are used to minimize the amount of radioactive ga1e1 and. particulates released routinely from the plant. It is anticipated that the routine releases will be well below the maximum allowable under applicable Survey No. s-1 S-2 S-3 S-4 D-1 D-2 D-3 D-4 D-5 0-6 D-7 0-8 D-9 D-10 0-11 W-1 W-2 W-3 W-4 W-5 W-6 -------------~----~-----

Table 9.47 Routine Survey* Titlf Check Dosimeters and Record Re1ult1 Pick up Air Samples Check Chart* on Gaaa Alan, S111ple System and Weather Monitoring System. Count ~lea Check Station Monitors and Hand Counters Ca~ibrate Instruments Spot Check Laboratories Survey Hot Lobby Transfer Dosimeter Read1ngi* to Exposure Record cards Spot Check Sample Aisle, Pulser Aisle

and Warm Equipaent Aisle Survey Lunch Room Survey Step-off Pads . Prepare Control Samples for Coun~lng Prepare Environmental Samples for Counting Spot Check Product, Packaging and Handling Survey Alpha Lab Survey Chem Laba Survey Spec Laba Survey Product Lab Survey Zone III offices S *urvey Mena Locker Room Shift A11lgned 1,2,3 1,2,3 1,2,3 1,2,3 3* 3 2 3 1 3 2 3 l *l 2 1 1 1 l 3 2 Q l .... ,. .......................

--.-......*

--..,........

.. _ ... _ _...... .._.._.. ........ ....... ------*---***.,,..'11 ...... ,'"'-**r---.....--

--~ ... --... --...... Table 9.47 con*t Shift Survey No. Title A11iqned 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 Acce1~ Aisle 2 w-11 Survey Operating Aisles 2 w-12 Survey Sample Aisle 3 W-13 Survey fuel Receiving

& Storage 3 W-14 Survey Product Packaging

& Handling 2 W-15 Survey Decontaadnatlon Area 3 W-16 Survey Scrap Transfer Area 3 0 W-17 Survey Health Phy1ic1 Lab 1 W-18 Survey Mechanical Cell Viewing Area 2 W-19 Survey Laundry 3 w-20 Obtain Environmental Sample* 1 W-21. Sul'Vey Womens Locker Room 1 W-22 Surv~y Warm Equipment Aisle 3 W-23 Survey Mobile Equipment 1 W-24 Survey Ventilation Equipment Rooa1 2 M-1 Survey Analytical Viewing Area 2 M-2 Survey Instrument Shop 2 M-3. Survey Main Lobby 1 M-4 Survey Cold Chemical Penthouse 2 M-5 Survey Che11ical Proce11 Cell.Viewing D Area 2 Survey No. Title 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-5 Q-6 Survey Maintenance Shop Survey Guard Houae Survey Tank Fara Survey Burial Ground Survey Rellote Operating Station Survey First Aid Obtain Environaental SaJ1Pl** Autoradiogra~

En"ironaental Air. Saaplea Survey Utility Building Survey Roads, Walka, Parking lot and R. R. Spur * . . >> * .. Survey Storage Lagoon and Hardstand Areas Survey Dry Wells Survey*.zone II offices Ob~ain Environmental Samples. S

Shiftwise

.D*** Daily W

Weekly M

Monthly Q

Quarterly

' ( Shift Assigned 2 1 l 1 2 1 . -.1 l 1 3 re :

Air Sampling 3 Perimeter l Plant Site l!!_in & Snow l Plant Site Surface Water 1 Erdman Brook 1 Buttermilk Creek l Cattaraugus Creek ~and Silt l Erdman Brook 1 Buttermilk Creek 1 Cattaraugus Creek Well Water 1 Plant Site Table 9.49a Environmental Monitoring Phase I -Type of Analysis Weekly Gross Alpha Gross Gaaa Monthly Gross Alpha Gross Gaaa, Tritium Gross Alpha Gross Gaana, Tritium ~oss Alpha Gross Gama Gross Alpha Gross Gama, Tritium Semi-Annually

~------------------------------------------------------------------~.i _;-------Vegetation 3 Perimeter Neighboring Farm Small Game 1 Plant Site ** Gross Beta Gamma I-131 Sr-90 Gross Alph:l Gross Gama I-131 Sr-90 Gross Alpha Gross Gama I-131 Sr-90

-* Air Sampling 3 Perimeter l Plant Site Ra\n and Snow 1 Plant Site Surface Water* l Erdman Brook 1 Buttermilk Creek 1 Cattaraugus Creek Mud and Silt l Erdman Brook 1 Buttermilk Creek 1 Cattaraugus Creek Well Water 1 Plant Site Vegetation 3 Perimeter Milk 1 Plant Site Fish and Shellfish 1 Cattaraugus Creek Small Game 1 Plant Site Table 9.49b Environmental Monitoring Phase II -Type of Analysis Weekly Gross Alpha Gross Gama Gross Alpha {",ross Gama, Tritium Gross Alpha Gross Gamma I-131 Monthly Gama Scan Gross Alpha Gross Ga'l'lla Tritium Gross Alpha Gross Gama Gross Alpha Gross Gama, Tritium Gross Alpha Gross Gamna I-131 Semi-Annually Sr-90 Sr-90 Sr-90 Sr-90 Gross Alpha . Gross Gaama I-131 Gross Alpha Gross Ganma I-131 Sr-90 0 federal and state regulations.

Spare units and automatic controls are used ~s necessary to prevent the escape of high level bursts of activity caused by major equipment failure. 9.~l A continuous stack gas mo~~tor, described in Appendix 9.51, is used to detect concentrations of 3 x 10-~c/ml or less of gross beta-gamma particulate activity and about the same concentration of 1-131. 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 experience; they will be kept at the lowest practical level to provide the earliest possible warning of off~ standard conditions.

9.52 Environmental monitoring to determine the effects on the environment of waste gas disposal is concentrated in air sampling and sampling of soil, vegetation, milk and rainout. Three site perimeter continuous air monitoring stations are established to determine concentrations of radioactive particulates and r!dioiodine at these stations.

One station is located 3,100 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, according to prevailing wind patterns, will place one of the three monitors down wind of the stack nearly 60 per cent of the time. 9.53 The routine 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 be analyzed lf.the -sta*ck monlto*t indicates an* alarm cond~.tion.

The weather monttoring

stations, (see ippendi,c': 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 wind of the stack regardless of wind condi~ions.

Waste Disposal Control Program --Liquid Waste 9~54 The primary contro.i' of high level liquid waste 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 5.50 -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 near the confluence or Erdman Brook and ~uarry* Creek will serve as a third monitorir t9 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 necessa~y, and discharged to a series of holding oonds. The interceptor volume is about 50,000 gallons and the ponds I._ ... ~*--*--*--__ ........_..

........ ....,.,,,.ltr.rlol

.. "" ..................... 9'fl ..... *---,... ............... ~--*-........ -, ______ ol,.,.,._..__......_, _________

....,,. .. ....,,. -~----------v-----. ** **-------* provide holdup for 4,000,000 gallons or 100 operating days above the minimum overflow points. Overflow points between ponds are a valved line at two feet above the bottom to provide 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.

9.56 St~eam gauging and sampling stations are provided near the confluence of Quarry Creek and Erdman Brook and 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 stations will be collected* and analyzed weekly. Analyses will include gross alpha, beta and gamma, tritium and specific isotope analyses as required for control. (Appendix 9.56) Waste Disposal Control Pro9ram--Hi9h 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 element 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 be used to lift the lid of the cask, remove the scrap drum and place it in the trench. At the end of each burial operatiop, 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 reduce the exposure rate at the 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 l mr/hr or less. The minimum dirt covering will be four feet thick. 9.58 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 backfilled.

Packaging techniques will vary depending on the equipment itself and* the radiation-contamination conditions.

Generally a sprayed-on coating or a covering of plastic film will be used. Medical Program 9.59 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 will be paid to history of past radiation exposure, allergies, blood dyscrasias, tumors and any evidence of emJtional instability.

The laboratory studies on all applicants will consist of a minimum of complete blood count, serology, urinalysis, chest x-ray and vital capacity minations.

Each employee will have a complete physical examination yearly.

t_ -----' -A co~plete blood count will be done twice yearly; ~linical urinalysis monthly. The pre-empl , oyment physi c al examination and laboratory studies will be repeated on each individual leaving the employ of the company. 9.60 Bio-assays will be scheduled for employees using an "across-the-plant statistical survey" plan. The number of times each employee is sampled each year and the type analyses performed.will depend on his work location.

Office employees annually, for total alpha and gross fission products; mechanical head end, extraction operators, Health & Safety cians, maintenance and utility operators semi-annually for total alpha and. gross fission products; product purification and packaging operators quarterly for plutonium and total uranium. Additional samples will be obtained to confirm any positive result and special samples will be obtained when tion or ingestion is suspected for any employee.

9.61 Thyroid monitoring

_of employees will be performed at least once each year in conjuction with the annual physical examination.

Special monitoring will be performed as indicated by 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.

I111nunization 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 training 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 shall be kept on form AEC-5 as ~rt of the permanent record of each employee.

A permanent checlt-off list shall be attached to each employee's permanent record covering all of the plants' requirements regarding physical examinations and personal radiation exposure recording and control as well as all requirements of lOCFR-20.

Emergency Procedure Fire Protection Organization 9.64 The Ht!a H , h and Safety department has the primary responsibility 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 fer Radiation Emergencies 9.65 There are a very large number of combinations of conditions which might constitute or cause an emergency.

I t is, therefore, not possible to prescribe inflexible procedures for emergency action. However, there are broad categories of emergencies for which general procedureb may be stated and certain general rule& which apply in nearly all cases. In any radiation emergency, the Health and Safety Department ha& the primary responsibility to define the magnitude and extent of the problem and to recommen~

a course .of action which will restore the affected areas promptly and safely. 9.66 In any radiation emergency the responsible group (Production or Analytical}

in the area in which the emergency condition exists must take inmediate steps to accomplish the followings

a. Protect plant personnel by evacuating affected 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 notifieda Health 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. Devise 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 rad~ation rules 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.

Perform complete contamination surveys of such personnel and institute decontamination at once if significant exposure could result from a delay. Normally, it is best to leave skin decontamination to those persons with specific training in this function.

c. Remove injured persons from the scene with as little direct personal contact as possible.

Limit first aid and medical procedures to : ,, ,, ... t.

C ........ A those that must be done promptly until the doctor is present. d. Do only what 11 necessary to preserve life and property prior to the arrival of Health and Safety specialists.

e. Work*within the framework of any applicable SOP'& covering a specific type of emergency.

Plant Maintenance P.rogram 9~~a The Nuclear Fuel Services maintenance program has been planned to insure continued*safe operation of the plan\ conaensurate 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 normal 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 control& are incorporated to cover work within contamination and radiation zones. Work on contaminated equipment or systems is done under the surveill~nce of the H.eal th & Safety Department which recoaaends required control measures.

Careful planning, prewritten job procedures, and close coo~ination with Production and Tec h nical 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. ment that could become critically unsafe from a nuclear standpoint, and 2. any ma*lfunctioning piece of equipment which could reasonably require the shutdown of the plant. ' A list, referred to as the Vital Equipment List initiated by 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 taken out of service, and what tests and requirements are to be met before the ment is returned to servf c e. All equipment not specifically designated on the V i tal Equipment List is considered as non critical and may be taken out of serv ic e, repaired, and returned to service according to normal standard mainte n a nc e practice.

Organization 9.71 Maintenance work on Nuclear Fuel Services equipment and systems is performed by Plant Engineering.

Plant Engineering is responsible for all mechanical, instrument and electrical maintenance work. Each of these categories is under the direction of a group leader. Close cooperation between these groups is maintained to facilitate scheduling, conserve power, and minimize downtime.

Under normal conditions, mechanical and electrical maintenance is accomplished on a day schedule, five days per week. Much of the routine instrument tenance is carried out on a similar schedule, *however, instrument technicians are normally on shift with operations personnel.

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

The Plant Engineer is responsible fora 1. Planning, scheduling, and controlling personnel, materials, equipment and tools. 2. Initiating training and edu~ational p1*ograms for maintenance personnel.

3. Establishing and supervising the maint~nance of a readily accessible file of design and vendor information, parts data, preventive tenance records, and historical records. 4. Supervision of all maintenance assignments, including instructions to cover*safe working practices, radiation protection measures and approved maintenance repair procedures.

5. Making technical studies on maintenance of mechanical, instrument and electrical equipnent, and making recomnendations on design changes. 6. Prepdring labor and material 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 dele~ated.

Technical support is available from the Technical Services and th~ Health and Safety Departments which will provide specialists as r.equired.

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, instrument, pipe, carpentry and welding shops are provided.

In cases where maintenance functions require facilities not provided at the site, privately operated shops 1~ _nearby Buffalo, New York will be utilized where possible.

' l 0 C Instrument Maintenance Peraonnel 9.74 The maintenance of instrumentation and control systems ia the responsibility of the Plant Engineer assisted by the Instrument Engineer.

Theae responsibilities are as followaa 1. Adequacy of the maintenance facilities and the training of personnel to meet all requirements, both roijtine and emergency, 2. Planning and scheduling of all instrument maintenance in cooperation with mechanical maintenance personnelf

3. Establishment of a preventive maintenance progrQD 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 that contairaa

'the mation necessary to analyze, design, order spare parts and components, apply preventive maintenance procedures and provide history of repairs on all equipnent.

This will be done in conjunction with mechanical maintenance.

Instrument Shop Facilities 9.75 The instrument shop is equipped with services, (water, air, electricity, tools and test equipnent) necessary for the calibration and maintenance of either pneumati~

or electronic instruments.

Maintenance Categories 9.76 Plant Engineering performs three categories of work1 preventive maintenance and inspection, routine maintenance and non routine maintenance~

Any of these categories of work may 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 involved 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 is studied thoroughly, and a schedule of routine inspections is determined and established under the following classificationsa

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

b. B-Classa A "middle-of-the-road" inspection.

Usually made quarterly to semi-annually and, on occasions, monthlYJ, c. C-Class I A minor inspection ( ordinarily visual and .frequent.)

Usuall y made monthly to quarterly and, on occasions, weekly.

As each piece of equipment le studied, a complete list of items to be checked on each inspection is made. A central control system indicates when inepec~ tiona aro due. If inspection, do not interfere with normal plant operation, the 1nepec t 1ons are scheduled and carried out in accordance with work loads in the section. Inspections that require shutdown of equipment or interfere with normal plant operations are coordinated with the Production Department.

After an.inspection la completed, information is transferred from the inspection sheet to a card as a continuing record. If any repair& are necessary, such repairs fall into the category of routine maintenance and are scheduled according to the urgency required.

Routine Maintenance 9.78 Routine maintenance includes all maintenance work on equipment 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 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.

!fon Routine Maintenance 9.79 Non routine maintenance includes modificati~ns or additions to systems or processes as differentiated from repair or replacement of faulty equipment.

Depending upon the nature and extent of tne work, 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 per~od. Administrative Procedures for Carrying Out Program 9.80 All work performed in the various categories of the tenance program, including those of the Plant Engineering Section both during normal plant operation and during plant shutdown, are 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 I

with normal 1tandard 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 maintenance procedures for most pl*c** of equipment are provided by the vendor or are written by maintenance per1onnel1 for hazardou1 conditions the operation may be altered and 11 administratively controlled as described in Section 9.83. Non routine maintenance of a non vital component is discussed 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 evaluate the effect of performing the maintenance work. Such an evaluation is made on all it,ms listed as vital equipment.

The Vital Equipment List 11 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

-9.83 When hazardous conditions exist, it is necessary to alter normal maintenanc*

procedures before maintenance is initiated.

In all cases, a Special Work Procedure is required.

This work procedure is obtained and is administered as described in Paragraph 9.17. The use of this permit provides maximum assurance that both the worker and ma~agement take adequate steps to minimize the consequences of radiation or contamination associated with the job. In all cases involving hazardous maintenance, it is necessary to fulfill the requirements set forth in the Special Work Procedure.

After this is done, the maintenance operation is performed in acccrdance 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 maintanance, 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 Criticality Connittee.

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 C maintenance.

Production Department 9.86 The Production Department is responsible for the operation. and maintenance of the proce11ing plant and its related process services.

The organization 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 follows* a. Fuel and Mechanical Handling1

b. Chemical Processingl

c. Plant Engineering1

d. Utilities and Process Services.

9.87 The Fuel and Mechanical Handling group is responsiblb for the Fuel Receiving and Storage area including cask transport, handling fuel assemblies, transfer and storage1 operation of the FRS water treatment facilities1 Process Mechanical Cell operation including fuel assembly transfer, handling, disassembly by saw or mechanical means, fuel shearing, handling of scrap, utility services to the area and hot equipment repair or ment1 General Purpose Cell including the loading, handling, storage and transfer operations of fuel baskets, scrap material and equipment utility services1 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 facilities.

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 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 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 ofa all utility systems within the utility room, plant area and off-plant facilities, non radioactive systems for both solids and liquids1 operation of the conventional low level burial and scrap removal from the plant1 material handling including the transport, handling, warehouse and distribution of equipment and supplies as required for plant operations1 decontamination of ~reas and facilities not included under other groups1 material control including records of input, output and inprocess material necessary to effect control, and accountability of source material and (j C Q ..,. M I ..... ._ ____ ,:_..,, * ......,. ________ .._......_~

...,.._, ... --:---:~ special nuclear materiat as necessary for Production Department requirements.

9.91 r.le basic plant operation and control is carried out physically by the process operators and shift iupervisors1 however, in a lar9e processing c~plex such aa the NFS plant, additional support inclu~ing technical and analytical services, monitoring, accountability, maintenance and c~ntrol 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 suppo~t. 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, aaministrative control is maintained by 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 together with a general 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, trative controls required during the operation, references to related SOP or other procedures, detailed instructions f o r functional operation of the equipment and,insofar as possible, the mechanical limitations of the ment. This last item may, in some instances, more appropriately be included in Run Sheets. 9.94 Run Sheets are another set of pr~cadures 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 co l ullll 1 as well as a desired operating flow. Separate Run Sheets are used for each flowsheet authorized under the operating license. The pub l'ished 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 strict enforcement and control of the plant operation.

The shift supervisor cannot operate outside the specified limits of the Run Sheet. However, exte n sion 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 operation within the lim i t s pe cif ied by the Run Sheft 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 l are reviewed periodically and amended ~s deemed necessary.

Under no conditions is th e , plant operated outside the te c hnical specifications included in the operating license.

l 9.95 Letters of Authorization are an ijdministrative procedure directing actual plrnt operation as described

!n Paragraph 9.7. They are used to authorize a specific Run Sheet and/or &uxiliary procedures for a particular processing campaign and in addition, are used to modify any of the restrictive procedures established for plant control. All Letters of Authorization are approved as discussed in Paragraph 9.7. 9.96 The actual operation of the complete processing plant is performed by personnel licensed as described in Paragraph 9.9 through 9.11. The basic areas of operator responsibflity are broken down into

specific catagories or areas of the plant consistent with production plant operating techniques.

The specific are~s are manned by operators consistent with their group license. The specifically-assigned areas for each shift are as followsa 1. Central Control Room1 2. Process Mechanical Celli 3. General Purpose Celli 4. Fuel Receiving and Storage-Chemical Processing Celli 5. Sampling1

6. Chemical Makeup1 7. Product Packaging and Handling1

a. Waste Handling.

In ad~ition, non licensed personnel are assigned to the following areasa 1. Utility Room1 2. Yards and ground, etc *

A brief description of each of these areas outlining the basic operator responsibilities for the respective areas is as follows*z l. 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 continuing through feed adjustment, solyent extraction, product puri(ication, concentration and storage are operated from this location.

Complet~ ~ontrol of the process is exercised from the control room with the exception of non routine operations such as manual block valves 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 paoel is a semi-graphic type for-ease of identification

and efficient operation.

In addition to posting the Run Sheets 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 using 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 dec~ntamination and in-cell remote maintenance.

* 1 I

0

C* , . . *

The run sheets for the PMC are somewhat different than those.for the rest of*the processing complex. They are made up of detailed fuel handling procedures which, in effect, are similar to an.SOP. Each different category of fuel requires specific instructions for handling throughout the PMC. The fuel handling procedures indicate the adapter& and fixtures required for

handling different fuels within the cell, detailed instructions for handling the fuel on the saw table, the disassembly, inspection and push out table, shear feed magazine, sequence of shearing and special precautions to be taken. Included also are throughput quantities so that they can be coordinated

- 1th chemical proce~sing.

3. General Purpose Cell a. Chopped fuel loading, handling, storage and transfer operations.

b. "Leached hull sampling, handling and transfer operations.

c. Receiving and transferring scrap and other ~z terials to the 11\C. d. Fuel basket handling including liners, capping and material control. e. In-cell remote maintenance

a. b. c. d. e. a

b. c. d. 4. Fuel Receiving and Storage Accountability as applicable to Production Department responsibilities.

Cask receiving, unloading and.transport.

Fuel assembly handling, transfer and storage. Coordination with mechanical head end operation.

Operation of pool water treatment systems. 5. Chemical Process Cell--Mechanical Handling Charging dissolver with fuel. Dissolver discharge of leached hulls. Equipment replacement by remote*mechanical means. other remote operations within the CPC requiring use of the remote mechanical handling facilities.

6. Process Sampling Sampling of the various process streams and vessels is conducted for ability, procass control, and waste loss determination throughout the chemical processing portion of the plant. The samples are taken at times prescribed by the Run SheetJ auxiliary samples may be taken as determined by the shift supervisor.

Laboratory analytical data from the samples are transmitted to the accountability officer and also to the control room where the results are logged. The shift supervisor then makes process adjustments or transfers within the limits prescribed by the Run Sheets. 7. Chemical Makeup Areas The chemical makeup area includes process solution makeup for the chemical cessing portion of the plant. These include all cold chemical influent streams for the solvent extraction columns and other cold process solutions such as dissolution and regeneration solutions.

Each process solution is made up from _.._ __ _ .. . . -~

a prescribed detailed form listing the constituent concentrations and total amounts of each solution.

The solution is then sampled and held for fication.

Following certification, and upon process demand, the solution is then transferred to a run tank for subsequent introdu:~i on to the process or, in some cases, directly into the process vessels. 8. Product Packaging and Handling a. Load out of plutonium 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 lo~ enriched uranium produr.t to transport vessel. All operations are conducted on a batch basis following specific instructions by the shift supervisor.

9. Supporting Areas a. Scrap Removal Areas, including the receiving and transfer to the burial area of*leacheo hulls ana other*head-end scrap generated.during processing, and transfer of new materials to the General Purpose Cell for head-end processing.

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 used in the facility.

10. Utility Room Operation--All Plant Services Contained Within the Utility Room Complex a. Water--raw, filtered, process, demineralized, and potable. b. Air--process, instrumentation.

c. Steam--equipment, process .and heating. d. Electrical--normal and emergency.

( vrs:r -.. ---------Process Maloperation 9.97 M.Jch 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 thoroughly tested process and one which is 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 made. Obviously, such ideal conditions will not always exist. This section discusses possible maloperation in various areas of the plant, the results of such maloperation, the method of determining a 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 (Head End) Proc~ssing 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 fuel elements or baskets and constitute little or no hazard, as such, but rather an venience and time loss during processing.

Maloperations in the FRS, P~C, GPC, and CPC are itemized ln table 9.98a, b, c, and d. Dissolution

-Feed Adjustment 9.99 Dissolution-Feed Adjustment steps under normal tions are discussed in Sections 4.21 through 4.34. The nature of a operation can vary with the fuel being processed.

For discussion purposes, the fuels will be divided into (1) ceramic U0:2 or Th9'2 fuels, cylindrical in form, clad in *stainless steel 1Jr Zircaloy tubing (as represented 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 i~ all types of fuels processed by NFS. A ~ummary of dissolution maloperation for these three fuel types, is given in table 9.99. Contents of this table are discussed in the following paragraphss Ceramic Fuels, Stainless-or Zircaloy-Clad Yankee, Commonwealth-Edison, and Northern States Power 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 concentration or too little or too much dissolvent added to the dissolver are the maloperations most likely to occur. The results of, and corrective actions*required by, such a maloperation are given in table 9.99.

. Measured Variable Column Pulse (Frequency-Pressure)

Density Flow Interface Position _,{Column)

~J~id Level Pressure -Pressure Differential Temperature f --,:* .~.-... U *.. --..;; .. a...,,,-. ' '. .. Table 9.97 INSTRt.lENT FUNCTIONS (Nomenclature used in Maloperation Discussion) 1 Display 7 Controlling Devices ,/ ;;; Devices * .::.* If 1/: -,-.ob f . l'\~ glf l# 9~ . <t""' .~ l9. * .... § ! ~.; ~<t c1 i:,; c1 i:,; .,., I (J,,, (ij~ . j j ;:,4 0~ (j 'b (j (ij 'I 'I 1

cl cl t., T_tw JH nloi C CIC CAL CAH D DI DR DIC DAL DAH F FI FR FIC FG FAL IRC IAH/L IAH/L I (LIC-(hi ah-low) ~, . L LI LR LIC LIC LAL LAH ' p Pr PR PIC PG PAH Pd PdR .. PcCL PdAL PdAH T* TI TR TIC TR:: TG TAL TAH ------*------

__________ .;_ __ _ -.---. ..... -, 'J 1 I I

(' (-.. Table 9.98a MALOPERAnON IN FtEL RECEIVIt<<;

AND ~GE AREA (FRS) Maloperation Result Indication

---Corrective Action . -1. Grapple failure during Fuel element drops to floor Visual. Special underwate1

ex-fuel transfer.

of cask unloading pool. tension wrenches and i !

tongs used to retrieve : fuel. ' 2. Fuel element cannot be re-Inconvenience and time loss, Poor visibility or binding Cask cover replaced and triaved from floor of cask. between cask and pool wall cask removed to decon-r unloading pool. or canisters.

tamination area to pro-I ' vide working space. t 3. Fuel element stuck in in-Inconvenience and time loss, Visual. Use underwater working I I dividual shipping slot. tools or fabricate equipment for removal. I I i . I I I I l . I I ---._... -*-**---

I ______ _. I *:'* Table 9~98b IIALOPERATION IN PROCE!!S IIEC~mCAL CELL AREA (PIC) ---:-'t' .. -----** ............

....... .A. ... -...........

~n~ct~ .... w:::-~~--*-** 1. Major breakdown of shear Inoperative shear Visual. Hold fuel element or during operation.

mechanism.

segment in PIC as long as necessary, monitor-ing temperature oc-casionally.

If temper-ature is too high, water cool with spray nozzles I in shear magatine. . Repair shear

2. Saw cut through fuel Saw cuttings containing Visual. Coolant will be proces-element (wet cuttin~).

fuel.particles drawn into sed if desirable.

saw coolant. 3. Saw cut through fuel Visual. Remove filter to fuel element (dry cutting}.

basket for subsequent I proce~sing ira dissolvers if ecoi'581Dics dictates , . . recovery of fuel * ., ;; i1

('1 (* -0 Table 9.98c MAI.OPERATION IN GEtl:RAL PURPOSE CELL AREA (GPC) . Mal operation Result Indication Corrective Action 1. Grapple failure during Part of fuel may be spilled Visual. *Power Manipulator transfer of loaded fuel onto floor of GPC. (2V-73) brought in and basket. pi~ces retrieved one at a time to fuel basket. *Notes Same procedure is ,sed to retrieve *spilled leac ~d hulls. 2. fines dropped on cell Lost fines1 inconvenience Visual. Wash down cell floor to floor from dropped fuel and time loss. crlticality safe suap. basket. Reaove auap pan and place fines in fuel basket. -3. Sheared fuel lodged in Stuck fuel piece in reducer

G-1-aonitor (2-LAH-l) at Reaove fuel basket and reducer during descent. 9-Jnch di ... ter point in fuel chute tip1 naow reducer. fuel pieces with 11Hter . slave *nipulatora or power aanipulator.

i I t i t 7---..al.~fei:iiW

-*

1. llaloi>eration Grapple failUflt during tranafer of fuel basket or leached hulls to or

from Dissolvers

1:-1 and :1:-2. t----_J_ --,u J --Table 9.98~

IIALOPERAnON IN THE CtEIIICAL PR<aSS CEU. AREA ( CJlC) Fuel or leached hulls spilled or i floor of cell. I Viaual. Recover pieces Iii th poar aanlpul,tor.

i I I

.) . } I L, Conaolidated-Edison fuel i s 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 .M hydrofluoric acid in the dissolvent to dissolve the thorium oxide. Possible maloperations in this case include all the above cases plus insufficient, or excess, or omission of, the fluoric acid. Boric acid is also added to the dissolvent on the Consolidated-Edison fuels. Omission or change in concentration of this component is a criticality consideration and is discussed in Section VI. Aluminwp-Uraniym 411oy Fuels These fuels, primarily of the MTR type, are charged into the baskets without chopping.

They are dissolved in 5.4 .M nli ric acid in which .005 .M mercuric nitrate serves as a dissolution catalyst. tions would include the use of acid with too low or to~ high a centration, too little or too much acid of the correct concentration, the addition of too little, too much, or the omission, of the catalyst.

These maloperations too, are shown in table 9.99. The addition of too much or too little acid to the dissolver is likely since the dissolvent is made up batchwise as required and is certified prior to transfer in total to the dissolvers.

The ing maloperations, their consequences and corrective action required are given in table 9.99

Zircooium-Uranium Alloy Fuels Dissolution of fuels of this type is accomplished by the appropriate addition of 1.0 .M nitric acid to the dissolver with baskets of chopped fuel in place, and to which is added 27.6 .M hydrofluoric acid at a sufficiently slow rate that little free fluoride ion is present (fluoride is complexed by the dissolving zirconium).

After coq>letion of the dissolution, aluminum nitrate-chromic acid solution is added to further c omplex the fluoride and to oxidize tin (from the Zircaloy) and uranium to their h i gher, and more soluble, valence states. Maloperation can occur from incorrect quantities and concentrations of the three solutions added to the dissolver.

Consequences of, and corrective measures to be taken after, such rnaloperations, are given in table 9.99. Stability regions for the dissolver product-feed solutions are given in TID-10089.

Other Fuels The SCRUP fuel is clad in aluminum and this cladding is removed with an NaOH-NaN0 3 solution before dissolut i on of the f uel. Excess cladding solution or high component co nc entrations i n this solution will result in higher-than-normal wast e solution volumes but will have no serious process consequences.

Insufficient decladding

.!Ir 11:liiil3l*r~

Table 9.99 IIALOPERATION AND CORRECTIVE ACnON DlllIM1 DISSOLUTION (3::-1 or 3::-2) .FueLiwe llalonAration Riruuilt.

I ic+iva &~+ton 1. JCeramic1 Stainless Steel1 I Low acid concentration in Zirconium Clad. dissolvent.

Low concentration of fertile material in solver productf all fuel not dissolved1 low acid. Recharge dissolver with additional dissolvent and complete dissolution.

Acid adjustment in 30-1. 2. 3. 4. 5. Ceramic1 Stainless Steel1 Zirconium Clad. ICeramic1 Stainless Steel1 Zirconium Clad. ICeramic1 Stainless SteelJ Zirconium Clad. Too high an acid concentra-1 High acid in extraction tion in dissolvent.

feed1 low decontamination factors if run. ,..._ I Insufficient dissolvent.

Excess solvent. -LAL-21 LCL-21 high tration of fertile and fissile materials in solver product. Low acid concentration.

High acid and low *tals content of dissolver product. Dilute high acid in 30-11 run w1 th low tions of uranium, thorium, end plutonium through the partition

cycle. Adjust concentrations in 30-1 into HAF tionsf add additional solvent to dissolver.

Dilute to proper acid concentration

in .3D-11 zun*to extractien syatea with low *tal ... content. *If .filled too high, could loverflorto the other dissol\ler or tCJ 30-1. In this event~ a batchwiM adjustaent of tha solution of all thilN vesHla' contents in 30-1 laould be neceasary-.

Uranium-aluminum alloy. Low acid concentration in I Low* alUllin\llll concentration disaolvent.

in dlssol"9nt product. *

Acid deficient product.

Saa fuel aay be solved. tl>TEa The indication

in alll of the *above maloperationlf is by sample analysis.

Add additional dissol"9nt t.r 3C-l and 3:-2 tf necessary to complete dissolution.

If sary, add acid to 30-1*. Boil down*to correct aluminum concentration.

I. -*---------... "'**-* ........... '4_ ... ,_w;* ilii1a1 .. *r.*----.* :..;-. .. .. .-.. ; lilaiiia __ ........ ..1! -* --~---**--* J ' ! 'J i? I I I r I i I I ., 6. F -Table 9.99 Continued MALOPERAnON AND CCMECTIVE ACTION DtRING DISSOLUTION

{3C-l or 3C-2) .EueLTwe Uranium-aluminum alloy. Malooeration Reault High acid concentration in I High acid in dissolvent di.ssolvent.

product1 low salti~ strength for extraction.

r.nrrective Action Add Al(~)3 to 30-1 and adjust to extraction feed specifications.

7. IUranium-aluminum alloy Low Hg{N03)2** Dissolution incomplete; dissolver solution low in aluminum, high in JR>3* Retain solution in solver1 add Hg{NO.:a)2 and continue dissolutron.

a. 9. Uranium-aluminum alloy. High Hg{N03)2*

Zirconium-uranium alloy.* High~ concentration.

No process consequences1 however, may limit tration of waste. Possible precipitation from feed upon long standing.

Adjust waste1 boil off if necessary.

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

Nitric acid is added to dilssolvers batchwise.

Hydrofld:>ric acid is metered s~parat~ly into the nitric acid. Stability ranges a~ giveqin 1I1RI& for zirconium-uradium dissolution.

---***-------------------

.. ----------------------+---------------------

... ~-------------------

10. 11. Zirconium-uranium alloy. Low~ concentration.

Zirconium-uranium alloy. Excess HF. -Lowered solubility of zirconium in dissolver solu* tion. Aluminum precipation in feed adjustment.

High equipment corrosion.

OOTEa The indication in alll of the above maloperationsl is by sample analysis.

Add concentrated HN~ to dissolver or to feed adjustment tank-if solution has already taken place. Dilute dissolver product with water or additional HN0 3 as stability region periiits.

\

I I Table 9.99.Continued MALOPERAnON AND CCJUlECTIVE ACTION IXJRING DISSOLUTION (X-1 or X:-2) Fuel T llaloperation

12. IZirconium-uranium alloy. Insufficient HF. 13. IZirconium-uranium alloy. Excess Al(N0 3)3* 14. IZirconium-uranium alloy. Insufficient Al(N0 3)3*

Result Zirconium precipitation in dissolver or feed adjust~ ment. All alloy may*not dissolve.

Aluminum precipates.

Zirconium precipitates1 high corrosive rates on stainless steel equipment.

1 NOTE1 The indication in ail .of the above maloperations fs by sample analysis.

Co~ctive Action If solution is still in dissolver, add additional HF, heat, and sparge. Add HF at controlled rate, using dilution air. Add additional HF to bring composition into stable range. Add additional Al(~)3 to dissolver or feed adjustment accountability tank as required

0 solution or low component concentrations in this solution will result in only partial jacket removal. As a result, dissolution may be complete or perhaps may not even start. A repeat of the decladding step will be necessary in this case. Decladding should not be attempted 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 fe~d 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.* . Solvent Extraction

-Partition Cycle; Extraction Column (4C-1}; Feed PYmp Pots (4C-13a and 13h}; and Meter Head Pot l4Y-14} 9.100 The most likely maloperations that can occur on the column, the feed pump pots, and the meter head pot, will be discussed in this section. ~' Many of these maloperations will be conrnon to .all pulse columns and tabulations made for 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 downstream columns. The plant operating and surveillance procedures are designed to detect maloperations before consequences can significantly upset any column. The maloperations that will be tabulated and which are applicable to other columns follows 1. Incorrect pulse amplitude-frequency settings;

2. Incorrect flow ratios between the various streams; 3. Incorrect stream compositions;

4. Interface crud and organic quality; 5. Column flooding;

6. Loss of column jacket cooling water; 7. Nozzle plate fouling. Pulse amplitude-frequency settings, if incorrect, can resul t 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 a Theoretical Stage) with varying pulse amplitude settings.

If the combination of the column design and flowsheet requirements are met by conditions at point A (which represen ts minimum HA column pulse amplitude requirements for the Yankee fuel flowsheet), then operation to Table 9.99 Continued MALOPERATION OF FEED ADJUSTMENT AND ACCOUNTABILITY TANK (30-1) Mal operation

1. Low steam in coils. J?p~111+ Poor temperature control. Cannot evaporate solution for concentration ion. 2. I Low cooling water in coils~ Poor temperature control for jetting. Indication TI-2-5; LR-6; DR-4; Sample 3C. TI-25; DR-4. Corr~c~ive Action Hold solution in tank until sample shows it is of correct tration; add cold chemicals to proper concentration.

Check cooling water valve; hold so l ution in tank until proper tions are achieved.

3. I Condenser water not on. Loss of solution (vapors) to vent system. TI-2-4 in condensate; LR-5 I Check cooling water valve. in 3D-2. 4. I Incorrect cold chemical addition.

5. I Air sparger off. Incorrect concentration of I Sample 3C; LR-6; DR-4. solution.

Poor mixing; incorrect I Sample 3C; l-[;-24. sample. Hold solution in tank; add proper concentration of cold chemicals to bring the solution to proper concentration.

Turn on air; hold solu-, , tion in Tank H:-24 until concentration is correct. I

1. l 2. 3. Maloperation Low cooling water. No air sparging.

Incorrect cold chemical addition.

Table 9.99 Continued MALOPERATION OF FEED TANK 10 PARnTION CYCLE (40-1) Result Indication Poor temperature controls I Tl-2-14. high feed temperature, re-sulting in poor column efficiency if HAF is too hot. Poor mixing, which results . in a bad sample. Wrong solution composition, which can cause a loss to product to the HAW stream or poor df. IC-29 off; manual check of hand control valve. Sample 4C. i Corrective Action Increase cooling water flow. Tum on Hand Control vaive from air supply line (1-C-29) 1/2 hour prior to taking sample. Certification and istrative check of proper amounts of cold chemicals to bring the solution to the proper concentration.

r: I l;,

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 maxilll.lm amplitude-frequency situation would be represented by point B. Amplitudes above B would result in increased stage heights due to adequate phase disengagement or other effects. Hence, for a given pulse frequency, the amplitude operating range would fall between points A and B. Operating outside these raQges would be considered maloperation.

Figure 9.100 HEYS VARIATION WITH PULSE AIFLITlllE

-HA OOLUIN At a particular frequency (cycles per minute).(See note) Maximum acceptable HEYS ,.,----~-Operating Range~~~~~__,;;:*

Pulse Aq,litude NOTEs Different frequencies would generate a family of curves with similar characteristics.

For satisfactory operating conditions, low frequencies are necessarily bined ~1th high amplitudes and high frequencies with low amplitudes.

Workable operating combinations will be established for each column and flowsheet.

Mlloperation would be due to hwun error or equipment failure and the correction needed would be obvious in each case.

f I I I I I . I i

I

. ' j i (_* flow ratios of the organic-aqueous streams (0/A ratio) have direct bearing on waste losses and decont,mination factors. Losses from the HA column, as well as the lA, IO, and !IA columns, will decrease with increasing 0/A ratios. The magnit~de of 0/A increase is limited, however, by the incr v ased flow or ratio change*.possible in the downstream column, or by flooding the column in question.

For 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 rate), to approximately 0.80, the waste losses in the~HAW stream will start exceeding

.the economically permissible value* of Q.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 limit$ on the*O/A ratios will be set on each of the columns in the syste~*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 puq:,. In each of the above, pulse or flow-ratio maloperation, the most probable result is high fissile losses in 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 * ._!operations in the Partition Cycle Feed Puq> Pots and Meter Head Pots are mainly concerned with flow rates to and from the column. These are also discussed in table 9.100. Plutgnit.p cycle feed Conditioott Jank C4Q-6) 9.101 Maloperations of the Plutonium Cycle Feed Conditioner Tank (4D-6) *are shown i~ table 9.101. The maloperations are as described in paragraph 9.99. feed Cgnditioner to first Uranium Cycle C4P:9) 9.102 Maloperations of the Feed Conditioner Tank for the ~irst Uranium Cycle (4D-9) are shown in table 9.102. The maloperations are as described in paragraph 9.99. Second Uranium cycle feed Cgnditioott (40-12) 9.103 Maloperations of the Second Uranium Cycle Feed Conditioner Tank (~0-12) are shown in table 9.103. The maloperations are as described in*paragraph 9.99. Plutgnit.p Purification c,11 9.104 The operation of the Plutonium Purification Celi is discussed in paragraphu 4.73 to 4.75. Malo~rations thit have been con* sidered consist in general of human errors. These include the flowing of tanks, improper s~lution adjustment, impro~~ v~lving, and failure to operate Plutonium Ion Exchangers 5C-1A, 5C-1B, and 5C-1C properly.

These maloperatlons are tabulated in table 9.104.

1 l !* C * ,**,,_, --Table 9.100 Continued MALOPERAnON OF HA COLtllN ( C-1) Malo ration Result Indication o IV Q. la. Flooding caused by high Same as item 1. Same as item 1. I Reduce flow rates. lb. le. 2. flow rates. -Flooding caused by poor disengaging time of organic. Same as item l. Flooding.caused by-pulserl Same as.item 1. amplitude-frequency.

Cyclic flooding.

Probable larger amounts of entrained organic in HAW stree. Excessive product loss to HAW and poor contamination in column. Indication will be same as item 1 in column. If , organic disengaging time is the cause, analytical l8s will indicate poor engaging time or organic . degradation.

Same es item l for column instrumentation.

Pulser frequency can be counted on oscillations of PR-8 or by counting RPM of pulser f poppet cam shaft. Pulser amplitude determined by maximum pressure on pulser surge tank (pulse stroke). Interface controller LR:-29 will be erratic, caused by a repeated build-up and breaking of an aqueous layer in the extraction sec* tion of the colUllll.

This will reflect an erratic havior of DR-23 also. PR-8 will increase slowly as the aqueous layer builds.up, then drops sharply to norma: as it breaks. Stop.flooding as in item l. Correct organic problem by solvent wasting or replace solvent. Adjust RPII of pulser motor or adjust pulser surge tank pressure.

Reduce flow rates or re C:,Uce pulser amplitude.

Check analy tical data for organic disengaging time.

--(i 1. Flooding.

l ' (" Table ~.lex> Continuftd MAI.OPERATION OF HA COLlllN (C-1) Organic leaves the 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 pr9duct to the product stl'.9am.

Both will require rework of material.

Interface controller LIC-29 will fi1at become erratic then will decrease loadi'1g pressure to pot 4Y-15 nificantly allowing organic to flow out HAW line. DR-23 (~xtraction section density) will increase dicating layer build-up of aqueous in the column. The layer of aqueous may break occasionally showing a sharp retum , to near normal of DR-23 reading with a following gradual build-up.

PR-8 (column static pressu will be erratic brat will show a substantially higher reading, indicating sed aqueous in ttie colUlll. lll-22 (~nsity in top engaging head) will crea .. significantly when aqueous replaces the organic in this section. When aqueous flows to surge pot 4Y-5 and thence to HBX* colUllln C-2, the interface controller, LIC-33 will be unstable due to increase in f aqueous flow to that column I I a Reduce colmn put rates and/or pu l ser frequency and upl i tude. Rework material as sary. Organic which left through the HAif line will be separated in the decanter 4Y-l and returned to the No. 1 solvent system.

Malooeration

3. Pulser stops. 4. Loss of HAF. 5. Loss of HAX. . Table 9.100 Continued MALOPERAnON OF HA COL\IM (C-1) Result Column will iaaediately go into a total flooded tion as in item 1. Loss of product in nol'llal , stream. Possible over of excess fission product to subsequent cycles. Loss of product to waste system. I XA-19 pulser alarm will. sound. Oscillation effect of pulser on iristr\lNnts PR-8, UC-29 will stop. Erratic behavior of *nts described in item 1 will not occ~r. PR-8 will show increase as aqueous builds up in column displac ing organic. DR-23 will in crease, finally showing total aqueous. Organic flow to SUrge Pot 4Y-5 will decrease to approximately 1/2 since flow will be equal to aqueous influent only. LR:-29 loading pressure will decrease nificantly to allow organic to flow out HAW line. Zero reading on LR-281 terface Controller LIC-29 will decrease sharply until controller adjusts to bring level back to desired point DR-23 will decrease1 DR-22 will decrease1 PR:.a will decrease.

FAL-9, FR:-22 on HAX1 flow to Surge Pot 4Y-5 stops1 LIC-301 DR-23 and PR-8 will 1 increase.

I I I I Shut down partition cycle columns immediately.

Restart after pulser trouble j.s corrected.

Rework material as *necessary.

If not restored within five minutes, shut down partition cycle. Shut down partition cycle innediately.

RellOrk as necessary.

' <. l f c , llalos:,e_ration

6. Loss of HAS. 7. High HAF flow rate. a. Low HAF feed rate. 9. I High HAX feed rate. 10. I Low HAX flow rate. 11. I Higl HAS flow rate. 12. I Low HAS flow rate. ' r Table 9.100 Continued MALOPE8AnON OF HA COLlllN ( C-1) Rewlt_ Decontamination will reduce by a factor of 10 to 100. Fission products will be carried to downstream cycles. L~ss of product to HAW. Indication FAL-7, FIC-1 on HAS1 LRC-29 will decrease sharply until controller adjuats for duced aqueous flow. DR-23 and PR-8 will decrease.

LR-28 increase (HAF ing pot J. PR-8, DR-23, and DR-22 ir.crease I I.R:-29 lower loading pressure.

Possible stripping of exces,LR-28 decrease (HAF fission product and carry-ing pot). PR-8, DR-23, and over to subsequent cycles. DR-22 decrease1 LJC-29 Same as ltem 8 Same as item 7. hi~r loading pressure.

FRC-22 (HAX) increase1 PR-8, DR-23, DR-22 will decrease.

FIC-22 (HAX) decrease1 PR-8, DR-23, DR-22 will increase.

~"'T"Ntcti ve Acilon If not restored within 5 minutes, shut down partition cycle. Lower HAF feed rate (FRC-16).

Rework material as necessary.

Increase HAF feed rate (FRC-16).

Decrease HAX rate with FIC-22. Increase HAX rate with FIC-22. Increase waste volume of HAlftFIC-1 on HAS1 PR-8, DR-23

1 Reduce HAS flow with FIC-1. Possible loss of product to will increase1 LRC-29 load-Rework material as neces-lwaste. ing pressure will decrease.

sary. Fission products will not be I FIC-1 on HAS1 PR-8, DR-23 I Increase HAS flow with scrubbed from organic in will decrease1 LIC-29 load-FIC-1. scrub section of column. ing pressure will increase.

_,: ..

Table 9.100 Continued IW.OPERAnON OF HA COLl.111 (C-1) llalooeration Result Indication

13. ILoss of salting agent from I Same as item 12. Will also I DI-1 and lab analysis on HAS. cause refluxing in column 140-171 PR-8, al-23, and and loss of product to waste DR-22 will decrease.

stream. Corrective Action Certification of HAS make-up solution before use. 14. INo cooling water on HA colUJlll c-1 jacket.

Depending on teq:,erature of !Cooling water valves close4,. Open cooling water influent. streams, may cause Sample analysis of effluentlvalves.

less efficient decontamina-streams. tion factor and waste loss. 15. IExcess interface crud. 16. IFouling of column nozzle !plates. Increase disengaging time. Possible poor decontaminat of Zr and Nb. Sample analysis of effluen streams. HAW and uranium and plutonium. streams from subsequent columns. Reduce contact of organic Can only be determined by aqueous and n\Dber of colUJlll critical analysis of all stages with resulting los~ factors affecting colUllll of column efficiency.

operation.

f .. If severe enough to affect product tions, increase solvent clean-up to improve engaging time. If trouble still prevails, displace interface to HAif syst* by Standard Operating Pl'ocat,re.

Clean out system and try to remove fouling ial by chemical flushes. If unsuccessful, replace column. I I C i 0 Table 9.100 MALOPERATI~

OF PARnTION CYCLE FEED PlJIP POTS (<<:-13A -138) llaloa:-~tinn

1. I Air supply is high. I Cause high flow of HAF LR-28 on *ter Head Pot stream. 4Y*l4 and FR::-16 on air supply. 2. I Low air supply. Cause low flow of HAF 1trea1 LR-28 on Meter Head Pot 4Y-14 and FR::-16 on air supply. 3. I Loss of pot due to crud I Low or loss of HAF stream. I LR-281 FR::-16. in check valves. Q ' -*-"" Correct settings on LR-28 and FJC-16 instrwnents.

Co~ct setting on LR-28 and FR::-16 instl'Ullants.

.. Switch to air lifts for HAF feed. l

-*-...-.-.-L

...... 1. High level in pot. 2. Low level in pot. . . . < . . -.. . .. -' ... '* . I Table 9, 100

IMLOPERAn~

OF METER JEAD POT (4Y-14) -~-------**--..... *--** The ~olution will go throuj, I.1C-2S1F1C-l6.

the overflow line to Tanlc 41).l until maloperation is corrected.

Low flow to HA col\1111\.

LJC-281 FIC-16. '* " \ . . . . . . l .. . . -. . . ~* .. -,. ---' *' ~---....... -................

-.. LIC-28 will adjust air flow to PUii> pots C-13A and 1381 reset instnJ-ments if necessary.

LIC-28 will adjust air flow to Pllllp Pots lC-13A and 1381 reset instrum if necessary

" -. . -; -,l ' : I I' 'f I J ! 1 0 n -) Table 9.101 MAI.OPERATION OF PLUTCJUlll-cYCLE FEED C~DinONER TANK (40-6) ___ .., ____ -*-**------**---.... .&.v1. """..,.&-........ .,........

.--_ ..... _, ** 1. High level in tank. Cause overflow to 60-3. La=-11 sample in Vent Correct flows in and System Catch Tank 60-3. out of tank and recyc l e . product from vent*system to rework. 2. Low oxident concent~ation Results in insufficient FR1 FAL on 'cold chemical Add oxident to correct of cold chemicals.

oxidation of plutonium, feed tank1 sample 9C1 DR-1. concentration and recycle which causes the plutonium IIAW from 40-8 to rework. *product to go with the IIAW stream. 3. No air 1parging1 poor Poor tank mixing1 poor Air Supply Valve 11:-11 thin Open.Jl:-11/2 hour pr i or mixing. sample. line on UC-1. to sampling; check and l o g . opening of spare valve * . ' .. .

Table 9.102 MALOPERATION OF FEED COODITIONER TO FIRST URANIUM CYCLE (40-9) Maloperation Result Indication

1. Wrong cold chemical Cause wrong feed concentra-DRC-3; sample 22:. addition.

tion to Uranium Cycle and loss of product to IAW or poor df. 2. Low cooling water on coolel Poor temperature control ma1 Valves closed on cooler; 4E-4. cause poor column efficicnC)I column effluent samples. 3. Air sparger off. Poor mixing, resulting in a OC-4 off; sample 22:; DRC-3 bad sample and poor concen-tration control on IAF.

Corrective Action Correct concentration an d rk recycle IIAW to the rewo system. Open valves on cooler. Insure sparger on for 1/2 hour prior to sampling. .

Table 9.103 MALOPERATIOO OF SECCND t.RANIUM CYCLE FEED ~DITIONER (40-12) 1. Malo~rati.on Incorrect cold chemical addition to 4P-I2. Result Wrong feed concent~ation, 7esulting in poor extraction and loss of pr:oduct to IIM or poor df. --2. High level in tank 40-12&, ,*Overflow of t~nk1 loss of . product to 60-3. 3. Lo* level in Tank 40-12. 4. JL~w cooling water to Cooler 4E-5~ -5. INo air sparging.

I Loss of IDF flmc. Poor IDF te~*ratu::-e contl'ol Poor mixi.ny of lOF. lndicat§on

= r. -":iv~ A~tinn # FR on Cold Ch~i es l Feed Tan:tJ sample 2i>B~ DR-7 , .Co2>rect concen~rai;Jon

'in tank" :and recycfe IC. fl"OtD.~13 to re~rk; 40-12. * " !Ji-2.?1>

i.AH-6 ii) :n~nk 40-12. . ,., .: -----"I. * ¥ **.,:, ~:;:. -LR~'i.2 in 'i'~~'*.i!U)-12 .. * , . -* .. _ .... , ...... Va!v_es*. cloie~~i' ,~r(coclerJ

~oiurfil effluent $aD!ple~-. R:-8 offj saq>le 26B. '~ ~* ~di~!~~5!9W~" into ~nd out ,.?}** of , tank afld recycie fros off-gas system to nl-"°rlc*

Adj~st flo~ into ~Rd out of Taruc 40..:12 .. ; Open VJ}VeS on Col}Oitt-.

Tum on .c-a. J

1. 2~ table 9.104 MALOPERAnON IN PLUTCIUlll PlRIFICATI<*

CELL (PFC) llalol>eration Plu~onf\lin Ion.E~ange Feea Conditioner 5>-1 has been allowed to overfl~~ . Rnult Overflow goes to Ion change Recy~l-. Wilste Tank ~2 and then to the Plutonium Feed Conditioner

't~ &>-6 for rework * . -*

Indication LAH*l2 and LR-14 on 50-lt LR-1 on 5>-2. Improper Plutonium Ion '11ax1~ plutonium absorptionlTIC-3 on 5>-11 TR:-4 on Exchange Feed. Condi ti~ner by !ni.on resin will not 5)-20. 50-1 temperature and/or .. accompli~c:l if 1'~he 60 Ct temperature at Pluto~ium 10 , C plut~niua t" Ion E~changer l!)n Exchangers

--~*lA,~. . . , ~mperatun

i's not. mtlnt&rl ~-).B, and ~-lC>:'dUe'to At ~l~er and tower tempera-. pump ~7 not*operit1~ ~~.tures, resin*e):Cchange improP,er he~~i~g cS tioi *

fieoien_cy is ~'i~ii. :

water tpt 51)-20~ I ..: ' 3 *. I Improper Plutoniua**Ion . ith~ut \JP.9rglng, 138~ ~148;-l~*~ and 168 Exchange. feecf conc*:a~ra:.

.. . prqper: -~justmen~

of feed ln tiil)ier, gi Ye J Jf flu,ent tlon ( 5>-1) .due to~ paoi t. * ~l canri~t,**, be* obta , inect. If c,ntration1 1parging lndl-sparging or 'improper f,ee~.,\ ~3 i f f. not; ;_~_ciiu~~.a*. . . ~ted .by varlatiori' in , level ** adjustaent of and .,, ~-r 7.*2*! ~,-/U,le .. , pluton~ 1, co~1 fa\lun to t~ NlfDi.

0ot .ojd:dizec!

i,*.~\~ 'indl~ate~

by* FR*91 .ni ~l,~ . 1~ -'p).~ont ... ; wll}:*

by LR-14. *

11 ., .. ... .:.~th8 *Ion. Ix-,* pa war ,:-':~*~1~~ \ * * ... changers (s;.:u, 5:-18*, or. S: ,-lC) dur.ing:'ihe -.l<*Ji'ng 1 . *~ ****"'* *-" ; .. cycle, t9 , tbe. {qn ~c!Mffige

.*. iate *Tanlc. S>-2 .. end "111 .

t ' .. "') ,, ""' **

1W .to be returned ., to *tha *1Plutoni\ll Feed Condlt'io;..:r

'ank ( c-6) for nwo~~c.' orrect1ve A~tion Start fl°" thx'outh one of the Plutonium Ion E::~hanger1

~-lA, S:*lB, Ol' ~-lC. Readju1t temperature and refer to Standard Operating Procedur*s lf nect11ary1 check ~20 for nece11ary watex-J*velt adjust proper teaperatur.

with TIC-4. Adjust N~ and HNO addltlon to tank 50-f if product does not .. t 1peclflcatlon1.

l

4. '~--6. "1. a. 9. lla , lopera~1on Improper;valve.open.durlng *elutr1at1on

-~ycle of Ion . Exchanges

~~lA, !C-lB, and ~-le.*

I~r valv.e.open ~ing loading or wash cycle of Ion Exchanger

(~-lA, ~-1B, ~-lC). Ollit wash cycle o.f Plutonium Ion Exchangers

~-lA, ~-1B, ~~1c. No-~ or' l~ss than 0~2~ elutriant used in'*.elutriation cycle Plutoni1.m Ion Exchange Recycle Waste Tank. Overflow of Plutonium Ion Exchange .... 9ecycle Waste * . TanJc (5>-2). Improper use of Steam Jet 511:.32 (Pump-Out Eductor on 5D-2). ....... -\ . Tabl* 9.lO\ Contf.1aued . , . IW.OJIERAnON

~N PLtJTOfilll PlllIFICAtIClf QLL (PPG) Result* ~. -_Imil.catton

Ex~e~s p~utonium will be* J* $pple ~-14t~*-.SV-16, or carrii,d to the *Recycle *** . 5Y*l8. Tani 50-2. ' *~~ritles an~ fission ducti will_~ carried t~ . the PbJtonha Product Feed . Eva~r.ltot 1: ank '5)-4.. . Thi~ 1 **terial ~n be sent barut tc .. . i the PluiOS)ium Ion Exch~nge *Fe~*Conditiol)er

.Tank '( 50-l) for'* rework 0 O(f product specification.

-. . " J Fo~tion of pluioniunl' polymets which inlght plug resin bed. ..... LAH-13, LAH-14, LAH-15 on Ion Exchange Col\lll\S11:R~l9 on 9>-4

Sample 148, 151, or 168 on ~-.lA, ~-:1B, or !C-lC and Sampl~r l'1B'on tank 50-4. Tank 140-46 analy1e11 no flow by LI-27 and LR-28. Overflow will be carried ~hro~gh.vent *line to the Vessel-Off-Ga~C~tch Tank ( 6D-3) 'where 1 t .. can be wo~~-d. or sent to waste. **i' LAH-1 or LR-1 ori Ttnk 50-2. May result in transferrl~g

plutonium prod\,Jct: into Plutonium Waste.Catch Tank 40-8. LR~22 on Plutonium Wa*te Catch Tank (41>-8) and ple analysis of .40-8. *-r.. .... t ua, &...+t ftft Double check.cornet val"' upon start of 1 cycle. \ . Correct valve po1ltlon1.

Rework product to tanlt 51)-1. . . Certlfic1tion of acld concentration before UM. Steari Jet 51-32 to the Plutonl\11 Waste Catch Tank*4D-8 or by gravity flow to C>-6. Shut off 91-32 and work <<>-81 send.to tank 7D-8.

I I ,--------...... ",i#* -..... --,-----------:-----

.... -----.......-

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

-..... ---.-~ Tahle9.104 Continued IW.OPEJ,ATION IN PLtJrOUtll PlRIFICATION C'EU. (PPC) -~ .. . .* .... ' 10. MalooeEation lutonium Product vaporat~r Feed Tank 50-4 1verflow.

Result -. Plutoni um*.'can:ied to PFC * ,/.! S\DJ> and*to Solvent Waste Tanlt (1 30=-8) .. by i51-8 Steam Edµ~r. *. SUlllp and floor are sized to. be , safe for la~st vessel !n*PPC *. Tanlt 130-8. has boron Raschig rings and is' critica l ly safe * . ft. iiiuton i un Prod\lct :Liquid. is returned to Tanlc l&>-41 some l iquid may carry !over through condenser to 15)-6. ~porator ~-2 is allowed o*QwnlQW. 1a o lProduct boil-down to a _high_COIJCentration in *, Evaporator

~-~* IPlutonium concentration is 00 high~ 13. I Plutoni~ Evaporatoi;. Con-* ~ondensate is c~led densate*ranlt 5>-6 overf1ow~through over f l"ow vent t o ' . *essel Off-Gas' Ga~ch' Tanlc . --,D-3. It can be. handl*d at t point as re<l'}.ihd lol . . *-__ ,.. -... . ,; . l'4: : I Insufficie n t HN03* in ~1'-**Y form p J ,~t on,ia pol~r . '* -~ 5:-i. ' Excess fission p rodu c t~ i n '\19~: \--~t product 5>-4. : ~*pe~iiica~ions. , .. r-* Indication. LR-19 on 5D-4J LAH-16 on iump e DR-17 and LJC-20 on S:::-2 and 53-10 respectively. .JC-20, *n-2-31, and DR-17 on ~-2. I-2-32, LAH-16, and LR-21 n S>-61 LR-2 on 6~3. i5>-4 suple analysis.

4 sample. ... Start Pullp Si-10 to . evaporator or stop flow into S>-4. Stop pump Si-lOJ return 5>-6 solution to 9)-4 with IC-41. ;Watch DR-17 cloMlYJ dilute with dilute acid *

in !C-2~ Route to rework 7D-8 or ste 70-2. iAdd HfO:a if polymer has formed In !C-2 and digest ntil*in solution.

Return to 5>-1 with pump !IG-10 for rework e l,.. ' ** ~* . * : r _ . =-Iii' -s>ta:-arn

rt *sr*rs; ftiP*~' i iJ * '"" *. !:~*!#t;-tt:sm:d\

' -~ I..~.:!,. ..,, ,--.: Jo; . .... . -* .-. . ... ...... ~-~* -

~-... -CJ ration 16. I Resin degfadation

~-n, 5:-lB, ~-1c. . Ta~l* 9.1>4 Continued IW.OPERAnON iN.* PLUTtliilll PlllIFICATION CEU. (PPC) . . Exce~sive product in .. 2. I ~2 SUl)le. 1-7*; I lllproper discharge of . I So t>J.t,t!_f~

n ~rected* to.~~ I !-.AH-16 on Tank S>-6. *contents of Plutonium

tank, J

Evaporator Condensate Tanlc (5>-6). ,) /?':\ v' Replace naln ** 1peciflecf

ln standard Opentlng covering lon **change colulin opentlon.

Enauze that t.-contents an known prior to discharge.

-Uranium Produ,t Pyrification 9.1~ Uranium Product Purification is discussed in paragraphs 4.67 to 4.72. This discussion includes the operation of the Uranium Product Evaporatori the.Silica-Gel Beds, and the storage of uranium product. Again, the maloperations

discussed concern primarily those associated with human errors. These include high and low levels of solutions in tanks, improper valving, improper evaporatpr operat'ion, and improper sampling.

These maloperations are listed in table 9.105. frUduct Pi~kaging and Shipping Area . . 9.106 In the Product Packaging and Shipping Area, product plutonium

... solut-ion is placed in smal._l bottles for shipment. riched urinium product is also packaged in this area_.and low-enriched uranium is transferred to tank trailers from this*area. Due to the possibility of hazard from alpha contamination in the *Product Packaging and*Shfpping area, much of the equipment is enclosed in glove !?oxes and electrical interlock controls are installed to insure proper operator procedure and ventilation control. Maloperations in the Product Packaging*and Shipping Area include any spillage of mo~e than a minute quantity of plut~nium so~ion, cross contamination of the products with any foreiijn material, an *any act that could cause a crttJcal array. All equipment

~sed for hi~h-enriched uranium and plutonium is geometrically safe, including the sumps in the handling area. Here again~ tions involve human errors and are ~iscussed in table 9.106. *Bewotk Evaporator System 9.107 The rework system consists of a~ evaporator, $er, feed tank, and associated piping. The rework system is used to boil down and adjust any one of *the *aqueous waste **streams that are sampled and found to be too htgh in product to dis.card *before rerunning through the solvent extraction system. Organic streams are not run through the rework system. The streams reaching the rework system will vary from very dilute . solutigns approaching water, to stronger solutions containing up to 6.014 HN03. S~me streams will have only a few grams per liter uranium in themiothers will be of much higher concentration, and plutonium may also be present. When high-enriched fuel is processed, these streams may contain some of the structural alloys*of the fuels, such as aluminum and zirconium.

Thorium may also be present during thorium fuel processing.

Since the feed material ~ntering the rework system will vary widely in composition, each batch will require special consideration and operating inttructions.

Any serious deviation from these instructions would stitute a maloperation and.would result in a product stream u~suitable or less than l optil'lllm in composition for return to the extr~ction syst~m. Off-specification feed to the extraction system will result in l~~* of product and/or further rework operations.

If .solutU,n~

should t,e: conctntrated in the rewor~ system, a nuclear criticality incident could result. Prtvention of such incidents, however, 1, assured by certai~

-1. 2. llaloperation High level in Uranium Product Evaporator Feed Tarilc 50-7. ' , Low enriched uranium transferred from 50-7 to 50-9 by ...istake.

3. +High. enriched uranium . product misdirected to Lo. Enriched Ur-nium Product 4 ~vaporator

!ll'.:-4.

gh level in product .t evaporators.

-~-6. : I High density in product I evaporator.

Low density product in evaporator.

Table 9.105 .IW.OPERAnON OF \MNIUM PRODlCT PlRIFICATIClf Result Indication Tanlc overflows to 6D-31 I LAH-191 LJC-26. tank may boil1 no adverse effect unless ignored for long period. This tank is not normally heated. Possible degradation of high enriched uranium. produc~ solution.

Tanlc inventory1 LAH-24 in 51>-9. Possible to approach criti-'LJC-27 on ~-4. cal concentration (See section on Nuclear Safety). Poor operating behavior of evaporators.

Possible freeze-up of evaporator.

Low product concentration.

High readings on LJC-27 (low enriched uranium) and LJC-28 (high enriched uranium).

High DIC-20 reading1 high TR-1-3 reading (low.enrichlct DR::-21 and TR-1-1 (high enriched).

Low DIC-Z> reading1 low TR-1-3 reading (low enrichtd high product volume1 DR::-21 and TR-1-4 (high enriched).

Co~ctive Action Shut down extraction until *xcess solution in 5>-7 can be processed.

Lock out high enriched uranium transfer airli~s when processing low enriched uranium. When processing high en~ rlched uranium in ~-4, steam will be locked out of servlce1 air lift ply will be blanked. Reset control point to lower value on UC. Reset DIC to lower control point. DIC automatically dilutes high density product in evaporator.

Reset DIC to higher control point. '* J,

-7. e. -* llaloperation Too 1111ch cooling water to evaporator discharge line 1 jacket. Too little cooling water to product evaporator condensers , 9. ** I High level in Uranium Product Evaporator densate Taruc 50-8. 10. I Low level in &>-8. , ~----, 1;1.. I High level in Uranium PrQduct Surge Tanlc 50-9. . , 12. 1 1 Low level in ~9. I ' Table Cll05 Continued . IIALOPERAnON OF URANIUM PRODtCT PURIFICAnON Result Plugging of evaporator product discharge line if over concentration of uranium. Vapor to vent system; possible uranium loss to vent system. Overflow to vent system; loss of uranium. Indication High evaporator level1 product volume1 LJC-27 (low enriched)1 LJC-28 (high enriched).

High water temperature on Condenser TRC-5 ( low enriched) and TIC-9 (high enriched);

high temperature in 50-8 ( TI-2-36 ~. LAH-20 on 50-81 LJC-29. Lack of strip feed for firstlLAL-211 FAL-28 on ICX feed. uranium cycle. ..; ..........

Corrective Action Warm product line with steam or hot water via temporary connection.

Operate evaporator at lower concentration.

Reset TIC to lower tempe , rature. Stop evaporator1 sample analysis will determine disposition.

Supply solution from cold make-up area. Overflow to vent system to I LAH-24 in 50-91 LR-2 on 6D-~.Increase rate setting of 6D-3. . I pumps ~-2 and ~-2A ard/ or stop influent flows

Taruc runs dry; no adverse I LAL-25 in 50-9. I Adjust pump *nte on ~2 results. or 5r2A at lower setting or stop pump. 13. .i High teq:,erature in 50-9. I Improper feed temperature tofTRC-11 high on 50-9. silica gel units. Reset nc-111 use cooling water if necessary1 check cooling water to 5E-10. 14. : 'l""High feed rate to e"' li~a I Excess solution flows back I High_ ~ading on PR-9 or 1 Decrease Feed Pump 5;-2, gel beds ~-6A and ~-6B. *. to 50-9; no adverse effect. PR-10 on Head Pots 5Y-29 or s;-2A rate setting. 5Y-30. i I . : , , i I" "' J lliL::. * -~_!*.'~-~ ---,, _-z* .,*7 C::* ...-.-~* ** !" P Fl ORGI.F pp* , ..,~__.,..............._~~~

_. __ ~~~~-----.__.

,;; A V llalfunctio<<

15. I Line blinds and valves se wrong on silica gel unit. , I' , I ~~.,-*16' ' High level in High En-riched Evaporator Product I SUrge Tanlc 5D-13A. 17., I Off-Specification product in 5D-13A. 18. I Tra~sfer of 5D-13A to 50-1:11 or 50-13: by error before sampling.

I ' Table 9.105 Continued MALOPl:RAnc.

OF tRANilll PROWCT PURIFICAnc.

0 . Result Indication -tJ4'..-,._ Corrective Action Possible diversion of Valves are interlocked to' product to rework. insure proper setting11 visual observation of blind I Overflow to Tank ~-138 or LAH-4 on 5D-13A. ~-13:1 possible rework of product. Product must be reworked.

Saq,le analysis.

Possible rework of product. I Level inventory on tanks1 LAH-5 or 5D*l38 or LA-6 or

I 5>-1~. Adlllinistrative procedure will insure correct positioning of blinds before either loading or

regenerating silica gel beds. Transfer solution to 50-1381 shut down Product Evaporator

~-5. Transfer to ~-13:. Resuple and rework if required.

19. l High level in High En-** 1 Qverflow to 5D-13A and 1'iched Uranium Storage 50-1~1 possible rework. LAH-4 in 5D-13A1 LAH-5 on I Resample1 rework if &>-138 or LAH-6 on 50-13:. necessary.

Tank 50-138. 20. -I High level in &>-13:. 21. High te~rature in 5>-13A or 138. 22. I High levels in Low riched Uranium Product Sample Tanks 5>-12A or I 128. I Overflow to &>-13A or 1381 I LAH-6. possible rework. Resuple all three tanks (5>-13A, 8, and C) and rework if required.

No adverse effect1 eductor transfer may be difficult.

High(TI-2-22 on 5D-13A or I Check cooling water to TI-2-23 on 5D-13B)reading11 5E-111 allow tanks to inoperable eductors or ):ADIi)* *cool by air sparging.

Overflow to 60-3 V9ssel Off-I LAH-7 on &>-l2A1 LAH-8 on Gas Condensate Catch Tank1 I 50-128.

possible rework of s*olution. Transfer flow to empty tank1 transfer full tank to 50-158 after sampling1 resample tanks involved in overflow1 rework if necessary.

1 l I ~** ,* Table 9 .105 Continued IIAi.OPERAna.

OF mANitll PR<DJCT PlllIFICAna.

__ ._,...,. __ --*---------------1 -------**-t1 -** 23.

High temperatures. in No adverse results1 possible High reading on TI-2-25 on Turn on cooling water. i 5D-12A or. 128. boiling in tank. 5D-12A1 TI-2-26 on 50-128. 24. Failure to sparge tanks Poor sample1 incorrect LR-6 on Tank 5D-12A shows Operator must follow pro-5D-12A or 128 before analysis.

tank was or was not spaiged per sampling procedure1 sampling (See note). by linetrace1 LR-7 on Tank resample tank1 turn on '"} 50-128. sparge air 1/2 hour prior ' Note1This maloparation pl'CI cedure applies to all product tanks requiring sa8')ling.

to sa..,ling.

~-Solution put into w£ong Possible rework of product Level inventory on tanks1 Resample tanks as required tank (e.g.50-12A instead and resampling.

LR-6 on S0-l2A1 LR-7 on * ' of 5>-128. S0-12B;if too full, LAH-7 or 5D-12A or LAH-8 on 50-128. ' 26. High level in Low-Enrichec Overflow to off-specificatior.

LAH-9

Resample and transfer . Uranium Storage Tank compartment in 50-15". solutions as required.

5>-15A. --'ZT. High temperature in No adverse results1 possible High TI-2-27 reading. Turn on cooling water to 50-lM. boiling of solution.

coils. 28. High level in Low-Enrichec Overflow to 5D-15A off-LAH-11. Resa8')le and take invento Uranium Storage Tank . ' specification compartment1 rework as required

5>-1~. resampling required1 possil>b: " ".l rework * . ~r1 " ry1 29. Low level in 50-153. . No adverse results1 auto-When Pump ~-6 stops, low Wait for more solution to matic shutdown of product level is recorded.

accumulate or transfer mo ire : ' *~ -. -J?UDlp ~-6. solution from :50-12A or B . ' *-;, 30. H\gh level in Low-Enriched More solution than required LI-141 scale weight. Transfer to next tank tru ck. uranium Product Weight to fill tank trailer. Tank 5V-l. ., a u -1 .. ~. Degradation of silica gel. Failure to remove zirconium Sample analysis of S0-12A Rework if necessary1 re-"1c and niobium*from uranium and 50-128. generate or replace silic ' I , product. gel beds. 'l .. 1 ,*, , . ' . I l '"': [\ 1***r**1 , -,I *! :r . ), I * .. j I..'* -a .. ,I *, . 1'\ b ; -* .. .. , . I ,' ~i:'!r..~

d ,; -s"'.I . ~* ., ,,;!)' Cl 'lH ,: ".J

1} ,. ...i* QfCl i.iit ,..,.......R

,~; 1 ... ~, ,

,:..;;.. -Table9.106 MAI.OPERATION OF THE PR<>IXCT PACKAGit<<;

AND SHIPP It<<; AREA (PPS) . ....... Vti""'if'

... 8 ..... Vla. &lilliJVV

... 111fiii' ***v*v*1i1*v11

1. Plutonium product bottle Plutonium solution will

Visual observation of spil-' overfilled or filr line spill in bottle filling lage and/or measuring pot mlsconnected.

station sump. to a volume > 10 liters. . 3 . u . 2. Enriched uranium product Spillage of enriched uranium Visual observation of bottle overfilled or im-in fill area. spillage.

proper connection of fill -head. *"":::-*-3. Ll Two filled product bottles May approach a critical Radiation alarm. (plutonium or enriched array (See section on , . uranium) brought together.

Nuclear Safety.) .,(1 *"' 1*. *t. I ' '-* > ' ' :i~ ;-.:, ' *' ' 4. 'cross-over valves opened Cross contamination of Radiation alarms product between plutonium and products.

Plutonium alpha analysis.

,, uranium spill transfer contamination of uranium ' . .lines. product fill area * -' 5. Glove rupture. -* Personnel hand contaminatial Visual. ' ' ' J ~-l . . r ....... vw**--~ -* .,... r* Close valves on filling heads move to drip con-tainera transfer spill to* W-4 with eductors clean.up area with de-contamination washes. -Close fill valves and trans.fer spill to En-riched UraniUID Product Storage Tank S>-138. All equipment and opex ating procedures have been designed to preve more than one uncaged product bottle from be ,t. ing in the area at one tia

If the radiation alarl! sounds, area will be evacuated at once.

These valves will be locked at all times un* le11 a special situatic , n *work. requires their use_s material retumed to re Negative pressure on g box.will prevent leaka of contamination from Hand decontamination a prescribed by Health a Safety. Replace glove .ove J* box. d j

6. 1 Table 9.106 Continued MALOPERATION OF THE PROOU::T PACKAGIOO AND SHIPPING AREA (PPS) Improperly stoppered product bottle.* , *1 Gross co1'ltamination of the birdcage.

Monitoring prior to shipment.

Inmediate evacuation of area1 decontamination of birdcage under Health and Safety supervision1 contamination procedures as prescribed by Health and Safety. I

) procedures and administrative controls, which are discussed in the chapter on Nuclear S~f e ty. The rework system receives solutions into the Rework Evaporation Feed Tank (7D-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 compositions and extraction feed requirements, a procedure is followed for reworking each batch. If soluble neutron poison is required, as determined by ana*lysis, it is added from Tank 14D-32 in the cold tion area. Acid, or other reagents, could be added from the same tank, if required.

Solution in Tank 7D-8 is air-sparged and transferred by -steam-jet eductors to the R~work Evaporator 7D-4. Becaus~ of criticality considerations, Evaporator 70-4 is operated on a batch basis only. A low-level alarm and control prevent overconcentration in the evaporator.

Condensate from the rework evaporator flows to the Low-Level Waste Evaporator Feed T3nk. 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 System, together with possible consequences, alarms, tions, and corrective measures,are listed in table 9.107. High-Level Waste Evaporator System 9.108 A schematic representation of the High-Level Waste Evaporator System is shown in figure 4.81. The system consists of Tank 70-1 (the High-Level Waste Evaporator Feed Tanij, 7C-l (the Level Waste Evaporate~, 7E-5 (the High-Level Waste Evaporator dense~, and 7D-4 (the High-Level Waste Accountability and Neutralizer Tank). The system is fabricated of stainless steel except for the heat transf e r tube bundle, which is made of titanium.

Waste from the tion cyc l e or from the rework evapor a tor can be transferred to the Level Waste Evaporator Feed Tank. Solution may be transferred to the evaporator by air lift or by jet. It is intended that the evaporator bottoms be operated at an acid c oncentration no greater than BM HN0 3* M~loperations in this system are detailed in table 9.108. The evaporator bottoms, after analysis, are either neutralized with caustic and pumped to the tank farm for storage or are recycled, if necessary.

Low-Level Waste Evaporator System 9.!09 A schematic representation of the Low-Level Wast e Evapora t or System is shown in figure 4.83. The system consists of Tank 70-2 (th~ Low-Level Waste Evaporator Feed Tanij, 1:-2 (the Low-Level Waste Evaporate~, ,~-7 (the Low-Level Waste Evap o rator Condense~, and 7D-10 (the Low-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.

The Low-Level Evaporator System evaporates the overheads fr.am the High-Level.

Evaporator System, the aqueous waste streams from all of the solvent extraction steps except l 1

v> Table 9.107 MALOPERATION

SUMMARY

OF REWCRK EVAPORATOR SYSTB~ Maloperation 1.1 Transfer of too much solution from any one of seven feeds to 70-8. 2. Transfer of insufficient solution to 70-8. 3.1 Transfer of organic to 70-8 (see note). Result Indication Overflow of 70-8 to Tank I LAH-7 on 70-8; PAH-3. 40-10 and 40-13, one of the seven feeds to 70-8. 70-8 will run dry when evaporator is running. LAL-8 on 70-8. Organic will flow to 7C-4;. I DI; low densities in seven some solvent degradation ma~ feeds (70-8 and 7C-4). occur. Corrective Action Turn off steam to fer eductor, causing excess transfer.

Shut down*of Evaporator 7C-4 is automatic; supply more solution to 70-8 if available.

Jet out organic to waste tank after using special wash solution to strip out any product from the organic. Note: 8. Hydraulics of equipJnent and piping is designed tp prevent organic reaching anv one of the seven feed to Concentration of aclid used in 7C-4 will not ~ausb serious solvent nitration.

4. 5. 6. Transfer of wrong tank to I Boil down of wrong solution; Sample analysis; tank level 70-8. incorrect rework product inventory will show wrong Insufficient air sparge tc 70-8. Failure to add neutron poison to 70-8 when required.

composition.

tank has been transferred.

Poor mixing in 70-8; this isl HC-5 off; LR on 70-8 will only s3rious when widely djf.f draw thin, even line. fering compositions from two or more tanks are mixed or neutron poison is required.

LRC-4 and LAL-9 on 7C-4 will provide control and alarm tQ prevent possible critical condition.

Level inventory on 70-8 and 140-32; also administrative check and data sheet. Revise boil-down and rework procedure to compensate for different solution.

Turn on HC-5. Administrative procedure designed to provide two or more independent checks in such cases to insure poison has been added. See 1 section on Nuclear Safety. I

! ( ) ( ( ' Table 9.107 Continued MALOPERATION

SUMMARY

OF REWORK EVAPORATOR SYSTEM' Maloperation Result 7. I High level in 7C-4. Solution overflows to 30-1 before it is properly adjusted.

8. I Level too low in 7C-4. Concentration too high or steam coil not covered, resulting in low capacity.

9. 10. Air sparger on 7C-4 turned I No adverse results; bo1ling off. will supply ample mixing. Cooling water to Condenser I Water and nitric acid vapor 7E-8 too low or turned off in Vessel Vent System. entirely.

Indication LAH in 30-1. LAL-9. Corrective Action Level control in 7C-4 set too high or steam supply too low --reset. Dilute as necessary.

None other than position ofl None necessary.

air valve. High temperature on TR-1-8; high temperature in Vessel Vent System; excess sate in Vessel Vent System. Increase water flow to 7E-8.

1. 2. 3. 4. 5. 6. J I I J Table 9.108 MALOPERATION OF HIGH-LEVEL WASTE EVAPORATOR FEED TANK (7D-l) Malooeration Feed lift to 7C-l working too fast. Feed lift to 7C-l working too slowly. RP~tJli: Tndi t'!;::ii:inn 7C-l level rises due to in-I LRC-6 in 7C-l. creased flow. 7C-l level drops due to I LRC-6 in 7C-l. decreased flow. r.n~~ot'!tiuP

-A_ction Automatic cutback of air flow. Automatic increase of air flow. Liquid level too low in Liquid level will drop in LAL-5. Shut off air lift. Shut 7D-l. evaporator 7C-l. down steam to evaporator 7C-l to avoid overconcen tration. Liquid level too high in 70-1. Rate jet fails or is plugged. Transfer of organic when using rate jet 7H-4. ' ~~~~.......-c.;..-~--

t f j J l If alarm is lgnoTed and no I LAH-4. action taken, will overflow back to 40-2, the Partition Cycle Waste Hold Tank. 7C-l level drops. I LRC-6. Organic will decompose in I Sample of tank 7D-4. 7C-l with subsequent trans-fer to underground tank. * . * ! --I t I I *Shut off steam jet from 4D-2; increase steam and air jets in 7D-l, moving liquid as much as necessary (both ally feed 7C-l). Use air lift as na*te transfer mechanism.

No explosion will occ u r because steam is limited to 25 psig; the ature is then below fl*ash point organic; when using 7H-~tank 7D-l will not be ely emptied.

J f 1. 2. 3. ....! I , c , Malooeration Not enough s t~am in ing coils. Too much steam in heating coils. Liquid level too high in 7C-l. (', Table 9.108 Continued MALOPERATION OF HIGH-LEVEL WASTE EVAPORATOR (7C-1} Result Insufficient amount of liquid boiled off; liquid level rises, which matically shuts off air lift from high-level waste orator feed tank (70-1}; level in high-level waste accountability and izer tank (70-4} rises by solution traveling through overflow pipeline from 7C-1 to 70-4 if on continuous processing.

Too much liquid boiled off; liquid level drops; liquid temperature

rises; density rises; level recorder trol will automatically draw more feed from ttigh-Level Waste Evaporator Feed Tank (7D-1}. I Level recorder control will automatically reduce flow from High-Level Waste orator Feed Tank (70-1}. I I f r I Indication TR,;,1-6 in 7C-1; LRC-6 in 7D-) LR-9 in 70-4; OR-5 decrease.

LRC .. 6, ~1-6 in 7C-l; DR-5. LRC-6; check flow from 70-6. 0 Corrective Action Increase steam flow through coils with FIC-3. Stop steam jet from 70-1. .Turn down steam to coils

with FIC-3; use batch transfer jet 7H-4 if necessary.

If automatic control not enough, shut off steam jet from high-lev~l waste evaporator feed tank (70-6}; increase heating steam flow to 7C-l with FIC-3.

I I .,...,. *~ Table 9.108 Continued MALOPERATION OF HIGH-LEVEL WASTE EVAPORATOR (7C-l) ---,~----M_a_l_o.,_e_r_a_t_i_o_n

_____ _,_ ____ .~R.e-s-u_l;.;;t;.....

________ .....;;I;;.;.;n;.;d.i.;.c,;;,at_1;;;.;*

o;.;n..,..

_____ ...;:;C;.;o.;;.r,;;.r.;.e.;.ct~i;;;.;v;.;;e;,,..,;.;A;.;c;.;t_i_o;.;;.n_

..r 4. Liquid level too low in Level recorder control wil LRC-6. If automatic control 7C-l. automatically increase flo not enough, use batch from High-Level W*aste jet 7H-4 i ncrease flow Evaporator Feed Tank (7D-l) 1 from 7D-6; reduce heating steam flow to 7C-l with FIC-3. -1 . IF u ao-<<CA";w: " ....... .....__._*<<

s:41=4tt::

r,z~a...~i-~!3: Q1Y:wzt;r

-zem=: ; to ....... != ~s: e::rn---w : ,. x7i "' --. ,*, ~-.u-..

1. 2. J I 0 (, ---Table 9.108 Continued MALOPERATION OF HIGH-LEVEL WASTE EVAPOOATOR CONDENSER (7E-5) Not enough cooling water to 7E-5. Condenser fouled. Insufficient condensation; will send additional vapor to low-level waste feed tan (7D-2) and then to vent header in liquid waste cell increase in temperature of condensate stream; outlet water temperature will rise abnormally.

Insufficient cooling capacity.

TR-1-9 on bottoms stream. ondensate temperature (TR-1-9) high. 0 Increase water flow. . Reverse flush with dilute HN0 3*

I _____ _ I I ,, f . Table 9.10 2 Continued . MALOPERATION OF HIGH-LEVEL WASTE ACCOUNTABILI1Y AND NEtITRALIZER

°TANK (70-4) 1. 2. Malooeration Too much water in cooling coils of 70-4. Too little water in ing coils of 70-4. 3. IAir sparger not on 70-4. 4. 5. 6. Liquid level too high in 70-4. Not enough caustic added to 70-4. Teo much caustic added to 70-4. ::tc :MT* Na R~~ult. Solution too cold; may caust settling of particles to bottom of tank; will give inaccurate sample. Solution too hot if adding caustic; may cause some vapor to enter off-gas and vent system (see off-gas and vent *system). TIC-8. TIC-8. Tndication Corrective Action Autom~tic temperature control of water; administrative check pri e r to sampling; TIC-8 set too low -reset. Automatic temperature control of water; TIC-8 set too high -reset. Will not provide adequate cooling while adding caustic. IC-15 in off position; LR-91Turn on air sparger with will draw a thinner line IC-15. than if sparger were on. If high-level alarm ignoredJ LAH-14 on 70-4. will overflow into Low-Level Waste Accountability and Neutraliz~r Tank (70-10~ not enough room to complete neutralization if unneutra-lized. Solution sent to 70-10; I Sample taken in 70-4 and more acidic than desired. 70-10. Increase waste volume. I Samples taken in 70-4 and 70-10. Turn off steam jet from high-level waste rator 7C-1; jet* solution into Low-Level Waste Accountabili~y Neutralizer Tank (70-10) and complete neutralization in 70-10. Add more caustic in 70-4 or 70-10. Add additional acidic !waste. Ii.--.. -.,,...

j i. *I I 1. 2. 3. 4. n 0 Table 9.109 MALOPERATION OF LOR-LEVEL WASTE EVAPORATOR FEED TANK (7D-2) ______ . .., ___ ---** .. ,_. ...... ~11u.1,.l"".:..IT...1,.Ufl Organic solution in 70-2. Organic fed into Low-Level DAL-3 on 70-2. Waste Evaporator if not detected by operator.

Liquid level too high in If LAH*ignoreds will over-LAH-2 on 70-2. 70-2. flow into 40-10 or 40-13. Liquid level too. low in Low-Level Waste Evaporator DAL-2; LR-1. 70-2. 7C-2 will drop in level. Steam jets not on. Evaporator 7C-2 liquid LRC-7 in 7C-2. level will fall due to decreased flow. , '* 0 Corr~ .. *~ .. , *,11 Shut off steam jets to 7H-l and 7H-2 to Low-Level Waste Evaporator 7C-2; open steam jet to Solvent Waste Hold Taruc 130-8. .. Pump as much as possible to Low-Level Waste Evaporator 7C-2; stop influent flows. Increase feed to tank where possible; shut do iwn steam in Evaporator 7C-to avoid damage there. Start steam jets.

1. 2. MalopeTation Not enough steam in heatin~ coils. Too much steam in heating coils. Table 9.109 Continued MALOPERATION OF LON-LEVEL WASTE EVAPORATOR (7C-2) Result Liquid level will rise; level recorder control will automatically reduce flow from Low-Level Waste ator Feed Tank (70-2); tempera t ure will drop; sity will drop. Liquid level will drop; level recorder control will automatically increase.flow from Low-Level Waste orator Feed Tank (70-2) to limit of jet in operation; temperature will rise; sity will rise. Indication LRC-7 in 7C-2; TR-1-7 in 7C-2; DR-6. TR-1-7; LRC-7; DR-6 in 7C-2. Corrective Action Increase steam flow to coils with FIC-4. Decrease steam flow to coils with FIC-4. ---+-----------

-=---------+--------------------+---------------------.,-----------------

3. Liquid level too high in 7C-2. 4. IOverconcentration.

Level recorder control will automatically reduce flow from Low-Level Waste orator Feed Tank (70-2). Density will rise; liquid level will drop. LRC-7 in 7C-2; check flow from 70-3. DR-6; LRC-7. Increase heating steam flow with FIC-4 until desired height obtained; stop 70-3 flow if necessary.

LRC-7 will automatically increase flow from 70-2; if no solution available, 7C-2 contents must be diluted.

-( ( '* Table 9.1or Continued MALOPERATION OF LCNI-LEVEL WASTE EVAPORATOR CONDENSER (7E-7) ... _,.. ___ ---** n.c.:,u.i.

t.. i.na1.cai:;1.on corrPr.tivP Ar.t,nn l I 1. Not enough cooling water Insufficient condensation TR-1-10 on condensate stream Increase water flow. to 7E-7. with excess vapor entering vessel.vent system (see . vessel vent system); con-densate will have higher than normal temp~ratur~.

2. Condenser fouled. Insufficient cooling Condensate temperature Reverse flush with capacity. (TR-1-10) high. dilute HN0 3* . " I I I f I I I I

',. -----------------

---Table 9.109 Continued MALOPERATION OF LON-LEVEL WASTE ACCOUNTABILITY AND NEUTRALIZER TANK (7D-10) Malooeration Result 1. Too much water in cooling Solution will be too cold; coils in 70-10. may cause settling of par-ticles out of solution and poor sample. 2. Too little water in coolirg Solution will get to~ hot ii coils in 70-10. adding caustic; may cause some vapor to enter vent system (see off-gas and ven1 system). 3. Air sparger not on in 7D-10. Will not provide adequate mixing for cooling during caustic addition.

4. Liquid level too high in If alarm is ignored, solutia 7D-10. will overflow into High-Lerel waste Accountability and Neutralizer tank (70-4); not enough room to neutralize. . . .. . . *. . .-,:; Tnrli r.::a+i nn U.C-7 TIC-7. HC-12 off; level recorder will draw a thinner line* than if sparger were on. LAH-11. -r----~1 ...... 6-'*11--Automatic temperature control of water; adjust nc-7. Automatic temperature control of water; adjus t TIC-7. .. Turn on air sparger wit h HC-12. Install temporary jet (if not neutralized) an d transfer some of the solution to 70-4; shut off steam jet from low-level waste evaporator (7C-2); shut off steam jet from Hl.gh-Ievel was Accountability and N:!u-tralizer Tclnk (70-4); 1 ready, send to radioact waste storage tank (80--' ~, I . te f v~ ) .

..:.--------------

~~~--,~ --* ----:..:...~: ....... -J l , I . i 5. 6. 7. C i 0 0 Table 9.109 Continued MALOPERATION OF LCM-LEVEL WASTE ACX:OUNTABILI1Y AND NEUlRALIZER TANK (70-10) --.... _ --*---------**--~"Ila..,_....,.

.. ~~rrA~f:1 V~ -* , L_.l._**~I Wrong steam jet turned on If ne**tralized waste storage HC-13 on instead of HC-141 Shut off incorrect.jet to 70-10. solution goes to Rework LR-3 in 7D-81 will increase and turn on correct jet Evaporator Feed Tank (7D-8), on 7D-10J jets are kept it may have to be reworked.

locked1 unlocked tr/ Shift Supervisor only. Not enough caustic added Solution going to waste Sample taken from 7D-10. Add more caustic to to 70-10. storage (SD-1) will be more 7D-10. acidic than desired. Too much caustic added to Solut i on going to waste Sample taken from 7D-10. Let more acid enter 7D-10. storage (SD-1) will be more 70-10. basic than desired. . . . I l I l . I J i the partition cycle, and the aqueous solvent washes. The solution is either jet or air lifted from the feed tank to the evaporator~.

It is intended that the evaporator bottoms operate at an acid concentration no greater than 8 HN0 3* The ~ottoms from the evaporator are collected in the Level Waste Accountability and Neutralizer Tank (70-10) and are sampled. After sampling, the waste is either n~utralized with caustic and pumped to the tank farm or is recycled for further processing.

Various possible maloperations of the Rework Evaporator System are found in table 9.109~ General Purpose Evaporator System 9.110 The General Purpose Evaporator System concentrates wastes from a number of sources, as shown in figure 4.89. The condensable overhead is directed to the interceptor and the evaporator bottoms, after neutralization, to the Waste Storage Tank. Maloperations include incorrect steam.loading, undesirable iiquid levels, incorrect caustic addition, failure to air sparge and improper use of transfer jets. Tabulation of maloperations may be found in table 9.110. Acid Re~overv System 9.111 The Acid Recovery System is described in par~graph 4.85 and is shown schematically in figure 4.85. It consists of an Acid Fractionator Feecl Tank (70-3), a Feed Vaporizer (7E-1), a Hot Acid Storag~ Tank (70-11), , a Hot Acid Batch Tank (70-12), the Acid tor (,'C-3), the Weak Acid Catch Tank (70-6), the Recovered Acid* Storage Tank (70-5), and appurtanant equipment.

The tabulation of maloperations that may be encountered in the operation of this system may be found in table 9.111. Dissolver Off-Gas System 9.112 The Dissolver Off-Gas System is shown in figure 4.26 and is discussed in paragraph 4.26. It consists of a Scrubber (6C-6), Silver Reactors (6C-l'and 6C-la), a Heater (6E-1), a Rocirculation Pump (6G-3), a Cooler (6E-2), Filters (6T-l or 6T-1A), Blowers (6K-l or 6K-1A), and appurtanant equipment.

The Dissolver Off-Gas System has been stalled in order to insure against the release of radioactivity to the atmosphere.

A listing of ma~operations may be found in table 9.112. Vessel Off-Gas System 9.113 A number of process vessels are vented through the

Vessel Off-Gas System. The Vessel Off-Gas System serves also, in the case of some tanks, as the solution overflow line. The system is shown schematica!lly in figure 6.3c. In general, the maloperations considered in the discussion of the Vessel Off-Gas System concern the overflow of process vessels to Catch Tank 60-3 and the treatment of the sables leaving Tank 60-3. A discussion of maloperations in the Vessel Off-Gas System may be found in table 9.113. ( ,-

l I " .. (} C) *o Table 9.110 MALOPERATION OF GENERAL PURPOSE EVAPORATCR (7C-5) Maloperation Result Tndir.ation r.n~r.+ivo

&r.+ion 1. INot enough steam in heating Low evaporation rate or coils. l rate will stop. LAH-19 if solution is being,Correct setting on added; possibly TR~l~ll. automatic Pie of steam flow. 2. I Too much steam in heating I Liquid level will drop; coils. temperature may rise if overconcentration occurs. LR-13; TR~l-11. 3. Liquid level too high. 4. ILiquid level too low. 5. Air sparger not on. s. Too much caustic added *. ' .. If alarm ignored, will over~ LAH-19. flow into laundry and i analytical drain catch tank1 {7 0-13) or interceptor.

Possible exposure of heatirgl LR-13; TR~l-11. coils; temperature will rise.

While tank is cold, contents may tend to form homogeneous solution and give inaccurate sample. Solution more basic than desired. LR-13 will draw thinner line than if sparge~ were on. S,.mple of tank contents.

Correct setting on automatic PIC of steam flow or shut-off evaporator

Increase steam heating or shut off inputs. Shut down steam in heating coils. Tum on sparger for set ti* (1/2 hour) before ta~ing sample. !Penni t addition of more acid. ---,t------------t-------------.....;*-------------1-----------

7. INot enough caustic added. I Solution more acidic than desired'.

Sample of tank ccntents.

~dd more caustic. l *'

I i I I i i t f t f l 1 i ! l 1. *, .... Too little water in cooling coils. .. ... ... ----:, Table 9.110 Continued MALOPERATION OF THE GENERAL PlJRI><& EVAP<RATOR CCIIDENSER (7E-13) Temperature of condensate ITR-1-12.

I Increase water flow. will be higher than desired1 additional steam may enter vessel off-gas header. :'

l l l I i 1. 2. 3. 4. n Malor:,eration Air sparger not on i n 70-3. 0 \-~I Table 9.111 MALOPERATION OF ACID FRACTIONATal FEED TAtlC (70-3) Result Tank contents may layer and give inaccurate sample. Indication Corrective Action* IC-27 off1 LR-15 will draw Turn on Air Sparger a thinner line than if IC-27 and leave on 1/2 sparger were on1 operator hour before taking will initial LR-15 chart sample to permit each time sample is taken solution to mix. to insure thick line has appeared for prior 1/2 hour. Hot vapor entering inste1 Level will not increase of liquid.

properly in 70-3. LR-151 TR-1-10 on 7E-7. Increase cooling water flow to condenser 7E-7 (Low-Level Waste rator Condenser).

Liquid level too high inf I If high-level alarm ignored~ LAH-221 LR-15. 70-3. will back up to 7E-7. Liquid l evel too low in 70-3. Feed to acid fractionator i FAL-24; FR::-5 1 LAL-23. will be reduced. Pump as much as possible into 7E-11 if necessary, stop the Low-Level Waste Evaporator (7C-2). Decrease flow to 7E-l if possibleJ increase flow from 7C-2 if possible.

Table 9.111 Continued MALOPERAnON OF ACID FRACTIONATIJl FEED VAPORIZER (7E-l) *-----~l~e!r~tion Result Indication Corrective Action 1. I Too much steam in heating Liquid level in fractionato1 LIC-18; TR-1-13. Automa~ic control of coils in 7E-l. will rise. ' steam by LIC-18. 2. I Too little steam in Liquid level in fractionatoJ LIC-18; TR-1-13. Automatic control of heating coils in 7E-l. will drop. steam by ur::-1a 3. I Liquid level too high in Po s sible flooding of Acid ur::-1s ur::-1a will increase 7E-l. Fractionator

~1th liquid. steam flow to vaporize exces*s liquid. 4. I Liquid level too low in Possibility of vapor enter-I.R;-18. L1C-1a: will*decrease 7E-l. ing Vaporizer Bottoms steam flow to enable Cooler* (7E-11). liquid to build up. ' -~ ff * .,.. .*.. ,,. "-' .* ,~.,* . P I I~ JIA~i'll'Q9:laSla

'DD-..-.**--

I , I 1. 2. 0 0 Table 9.~ 111 Continued MAI.OPERATION OF VAP<RIZER BOn<<.S COOLER (7E-11) Maloperation Too much cooling water to 7E-ll.

Result No adverse effect. Too 11 ttle cooling* water I Output too hot. to 7E-ll. Indication Corrective Action TG-2 will *show below-normal I Decrease water flow rate. rise in temperature

.of cooling _water. Temperature gauge will showl I~crease water flow rate. above-normal rise in tempe ature of cooling water; TI-2-48 in 70-11 * . '

,-* Maloperation Table 9~111 Continued MALOPERATION OF Har ACID ST~AGE TANK (70-11) Result Indication

1. 2. 3. *. Air sparger not on 70-11.f Solution may layer and givel IC-28 off1 LR-16 will draw inaccu~ate sample. . a thinner line than if sparger were on. Liquid level too high in I If alarm is ignored, solu~: LAH-251 LR-16. 70-11. tion*will overflow to R6~ covered Acid Storage Tank 70-5. . Liquid level too low in I Possible shortage of acid I LR-16. 70-11. -t to dissolvers.

Corrective Action Turn on Air Sparger IC-28 and leave on 1/2 hour fore taking sample to mix solution to mix. It is best to have sparger on at all times. Pump as much as possible into Hot Acid Batch Tanlc 70-121 shut off valve from 70-5. Switch acid source to Concentrated ffN0:3 Day Tanlc 140-3, having.acid enter through 1:40-22.

I i .. 1. 2. 3. 0 Malooeration Air sparger not. on.10~12. Liquid level too high in 10-12. . Liquid level too low in 70-12. n ..._,.. ~* ~J Table 9~111 Continued . MALOPERATION OF HOT ACID B~TCH TANK (7D-12)* Result Tank c ontents may layer and give inaccurate sample W i ll overflow into Recover ed Acid Storage*Tank 70-5. Shortage of acid for Dissolvers 3C-l and 3C-2. }l;-29 off; LR-17 will draw Turn on Air Sparger . a thinner line than if. lC-29* and leave on 1/2 sparger were on. hour before tak i ng sample to permit solu-* tion to mix. LR-17. I If Dissolvers (3C-l, 3C-2) cannot accept a flow, send excess t o Low-Level Waste Tank (7D-2). LR-17. I Use solution from iie-cover&d Acid Storage Tank 70-5, Hot Acid Storage Tank 70-11, and/ or Dissolver Solids Mix Tank 140-22

Table 9.111 Continued MALOPERATION OF ACID FRACTIONATCR (7C-3) Indication Maloperation Result ;:-~nough :i:m in hea;-1 Tempe:.alu:r,:

will drop in ing tubes. Fractionator Reboiler (7E-2). TRC-13; TR-1-14. 2. 3. 4. 5. 6. 7. Too much steam in heating tubes. Liquid entering ator instead of vapor. Liquid level too high in 7C-3. Liquid level too low in 7C-3. Too much weak acid enterfng.

Too little weak acid entering.

Temperature will increase in Fractionator Reboiler (7E-2). Temperature will drop in Fractionator.

THC-13; TR-1-14. TRC-13; TR-1-14. Level recorder control will LRC-19. start pump to evacuate bottoms to Recovered Acid ~torage Tank 10-5. Temperature will rise in I TRC-13;LRC-19.

Fractionator; temperature recorder control will decrease steam flow to heating coils; output pump will be automatically closed 6y level recorder control. Increased h~~t requirement!

TRC-13; DR-15. on Fractionator (7E-1) . will lower temperature and density. Decreased heat requirement I TRC-13; DR-15. on fractionator will raise temperature.

Corrective Actio.lL.__

Automatic temperature control of steam input to reboiler. (THC-13) . I Automatic temperAture

control of steam input to reboiler TRC-i3. -Automatic temperature con trol of steam input THC-13; Correct flow or steam to 7E-l. Check flow rate control from Acid Fractionator Fee1 Tank 70-3; check temperature and level recorder controls on 7C-3 to make sure all three are set correctly.

Check flow rate control irom Acid Fractionator Feed Tank 70-3; check temperature and level recorder contr~ls on 7C-3" to make sure all three are set co~rectly.

Automatic temperature control of steam input with TRC-13. Automatic temperature con trol of steam input with TRC-13. l . I I f

1. ,-Maloperation Too little cooling water in 7E-3. Table 9.111 Continued MALOPERATION OF ACID FRACTIONATOR CONDENSER (7E-3) Result Output stream will have higher than normal ature; insufficient densation, which will over load 7E-9 an~ 7E-10. Ind TRC-14 on cooling water output. ' 1 1 TRC-14 on water flow. I f

1. 2. 3. 4. 5. 6. Maloi:>eration Liquid level too high in 7D-6 *. Vapor,instead of liquid, entering tank 70-6. Table 9.111 Continued MALOPERATION OF WEAK ACID CATCH TANK (70-6) Result If LAH ignored, will flow into Vessel.Off-Gas Header. Indication LAH-26; LR::-20. Vapor may enter. Vessel Off-I TI-2-22. Gas Header ( see Vent Sys1Eoi Too much caustic added tof Solution going to intercepti Sample taken. 70-6. or more basic than desired. Not enough caustic ad~ed io 70-6. Air sparger not on 70-6. Liquid sent to wrong area from*7D-6. Solution going to intercept-I Sample taken. or more acidic than desired. Tank contents may layer and I IC-35 off; LR-20 will draw give inaccurate sample. a thinner line than if sparger were on. If interceptor flow is sent. j LI ta any of the other tanks, it will increase heat dutiet *due to nonproductive fluid being present. If flow for another tank is sent to the interceptor and radioactiv-i~y is in solution, RAH in interceptor flow , will sound. ' in tan k s; RAH-28. Corrective Action Increase flow with LR::-20. Increase cooling water to 7E-3 (overhead denser for acid ator). Permit more acid to enter 70-6. Add more caustic to 70-6 or to interceptor.

Turn on Air Sparger IC-35 and leave on 1/2 hour before-taking sample to* permit solution to mix. Shut off open valve and open the correct one. . f I -** * , ,...,--~-::

... _ ... auCif_'f~swn 1 1 ttz,IJJ; .-..,~" *

.r , ' r , Table 9:111 Continued MAI.OPERATION OF RECOVERED ACID STORAGE TAtlC (7D-5) Maloar1t1on

1. I Liquid level too high in If alarm is ignored, it LAH-201 LR-14. Pump as auch as possible 7D-5. *w111 overflow into Hot Acid into Hot Acid Storage Storage Tanlc. Tanlc 70-11. 2. I Liquid level too low in I Not enough to supply tanks LAL-211 LR-14. Shut off Removal Pump 7G-2 7D-5. that draw on it (Dissolver to penlit backup of acidJ Solids Mix Tanlc 140-221

use alternate supply from : t Partition*Cycle Feed Add 140-3 Concentrated~

Day Tanlc 140-20; Partition Tanlc. i i Scrub Mix Tanlc 140-161 Hot I Acid Storage Tank 7D-111 Hot Acid Batch Tanlc 7D-12). 3. I Liquid sent to wrong tank May overflow 140-16, lJ LAH-25 70-11. Level indi-Shut off\va l ve to incorrec t from 7D-S. 14D-22, 7D-12.or 7D-3 will cators or recorders on all tank and open correct valve1 all overflow back into 7D-tanks should be watched Solution that went into while tank is filling to wrong tank should not harm make sure correct tank is tank. being fed. i*

1. 2. 3. 4. Mal operation Recirculation Pump 6G-3 not operating or no caustic in scrubber 6C-6. High liquid level in scrubber 6C-6. No water spray in scrubber 6C-6. Low heater 6E-l ture. Table 9.112 MALOPERATION OF DISSOLVER OFF-GAS SYSTEM Result Loss of scrubbing action will increase the carryover of fission products to the Silver Reactor (6C-l or 6C-la) and possibly out of the stack. The Silver Reactor will normally still remove 99.5%.of the iodine. There would be less 1 2 than a complete off-gas system failure postulated in 7~44 and 8.32.-Liquid carries over to heater (6E-1) and possibly some moisture to Silver Reactor ~6C-l or 6C-la). Possibly plugged scrubber off-gas flow or loss of

pump;' increased scrubber pressure1 condenser presS\119 may damage equipment and cause leaks. Below 11cc no 1 2 is removed in Silver Reactor (6C-l or 6C-la)1 release of 12 would bt less*than in 7.44 and 8.32. Indication FAL-7; FI-1; PdAL-6; increased radioactivity in stack monitors.

PdAH-5; PdR-3; increased radioactivity in stack monitors.

PdAH-5; PdR-3; TI-2-39 low; increased radioactivity in stack monitors.

n-2-40 lows TAL-9 on 6C-l and TAL-12 on 6C-la (Silver Reactor);

increased activity in stack monitors.

Corrective Action Restart motor 6G-3 or put in caustic. Reduce liquid level by using pump 6G-3 and pumping to Low-Level Feed Evaporator 70-2. Increase scrubber temperature; add caustic; flush out scrubber 6C-6; increase water spray. Increase superheater 6E-6 steam temperature and/or flow, using 111:-1 or TIE-2. j

1 I i \ I 5. 6. 1. a. 9. ' -~ . ..._ Table9.112 Continued MALOPERATIOO OF DISSOLVER OFF-GAS SYSTEM Malooeration High heater temperature 6E-l. Result Above 212: AgN0 3 will_ melt, allowing I 2 to pass through Silver Reactor (6C-l or 6C-la) This case would be less than in paragraphs 7.44 and 8.32. Silver Reactor (6C-l or f I 2 released to atmosphere 6C-la) becomes saturated.

but would be less than that postulated in 7.44 and 8.32, No cooling flow through I 400 F gas from Silver cooler 6E-2. Reactor (6C-l or 6C-la) wilJ damage glass filters. Low cooling flow (6E-2). I Operating at a temperature higher than 150 F will damage filters over an tended period of time. Tndir.saHnn TI-2-40 high and/or high FR:-21 TAH-8 on Silver Rea~tor 6C-l or TAH-ll on 6C-1a; increased activ*tty in stack monitors.

Radioactivity gradually. increasing in stack monitors over a period of time. TR:-3-will indicate high gas outlet temperature from cooler. PR-6 will indicate no coolant flow. TJr;-3 and PR-6. Fouled cooler; valves closed (6E-2). Increased cooler, Sil~.c I PdAH-10 at outlet of Reactor (6C-l or 6C-la) and cooler. scrubber fc-d pressures causing *possible damage as discussed in the previous malfunction discussion, , number 3. f , ---,,_..,-r. --+iv* &~inn Reduce flow and/or superheater 6E-6 top-* erat~ before reaching 212 C if possible&

if AgN0 3 has melted switch to aitemate Silver Reactor (6C-l or 6C-la)J regenerate initial Silver Reactor. Switch to alternate reactor (6C-l or 6C-la) and recharge initial reactor. Open coolant valves. Incr~ase cooling flow to cooler 6E-2. Clean cooler or open proper valves.

I Malfunction

10. Plugged filter or the valves are closed (6T-l or 6T-1A). l 11. Blower not operating (6K-l or 6K-1A). . -12. Filter fails.(blown). ' ' Table 9.112 Continued IIALOPERATION OF DISSOLVER OFF-GAS SYSTEM Result Indication Decrease flow through PR-6 and PAH-14 at the system. inlet to filter1 also shows on differential pressure (FdR-7 on 6T-l and PdR-9 on 6T-1A) and on dissolver pressure.

Altemate system blower will PdR-9, Pc:EL-10, PdAL-16 on automatically switch on. I 6K-l; PdR~7, Pcl:L-8, If it does not, then, if PdAL-15 on 6K-lA1 increase the stack system pressure in radioactivity shown by is greater than the dis-stack gas monitors.

solver off-gas system pnssure, the dissolver check valve.will close until dissolver system pressure builds up and then the check valve would open

Not having a* negative pressure in system would allow soae off-gas leakage into theprocass cells. This gas would I>> carried out the stack through vessel off-gas system. . I Activity to stack. \ PdAL-1~1 PdAL-161 atack monitors will show in-creased radioactivity.

' ,; ? Corrective Action -Switch to alternate fi t :..'r or open proper val .ves1 valves are locked open at all timesJ as dif-ferential pressure in-creases, filters will no1'111ally be switched and changed~ Start blower 6K-l or 6K-1A. Filters will switch automatically through Pcl::L-8 and Pcl::L-10.

I I I \ *~ .. ' 1. 2. 3. 4. t Malooeration Overflow of process vessels to 6D-3. Liquid overflow of catch tank 6D-3. . Discharge of Catch Tanlc 6D-3 fluid into wrong system. . No cooling water or low flow in the Condenser

-6E-3. t-*, Tabl~ 9.113 IIALOPERAnON OF VESSEL OFF-GAS SYSTEM Result Indication This tank is protected by OAL-221 DR-31 saq,le boron-glass Raschig rings1 analysis1 LR-2 increase.

a determination must be made as to when the catch-tank material must be reworked.

Overflow to 7D-8. LAH-171 LR-2 in catch tank1 increase in LR-3 in 7D-8. If product if found in Sample analysis of solutiail Catch TanJc 6D-3 and dis-DAL-221 DR-3 on 6D-31 jets charged into Low-Level Feed are kept locked and trans-TanJc 7D-2, expensive produc1 fer approved only by Shift is lost. On the other hand, Supervisor.

if discharge is made into Rework Evaporator Feed Tank 7D-4, expensive plant pro-cessing time may be lost. When exit water temperature TIC-41 TI-2-42 in Scrubber is greater *than lOOoF, the 6C-3. gas teq,erature in the Scrubber 6C-3 will be high, reducing scrubber efficiency.

I ' r*-. -:.iv. *~+*"" Send to Rework Evapora-tor Feed Tanlc 7D-8 or t , ed Low-Level Evaporator Fe Tanlc 7D-2 as necessary.

Process from Rework Eva :re-1k d. rator Feed.Tank 7D-8 as quired1 also process Ta 6D-3 contents as requi:r If product is put into Low-Level Feed Evapora-tor 7D-2, transfer it to *the Rework Evaporator Feed Tanlt 70-81 if con-densate with no product is transferred to Rework Evaporator 70-8, it can be boiled down and transferred to Low-Level Waste Neutralizer Tanlt 7D-10. Start cool~ng water or increase water flow (6E-3).

I ,----1 ' Malo ration Table 9.113 Continued IIALOPERATION OF VESSEL <FF-GAS SYSTEM Result . In i 5. I Incorrect pressure differ-Indicates fouling1 this PdAH-18 or PdAL-191 PdR-11 ential across the wi ll cause trouble in* on 6E-3. Condenser 6E-3. columns ~t the gas-liquid . . interface.

6. I No caustic and/or deminer-1 Small quantities of radio-LI-3 low1 TI-2-42 (6C-3)and alized water spray in acti v e material will be PI-2 readings will be in-Scrubber 6C-3. pasF.ed through system. correct. The PdR and PdAL However, this will be a should give further indi-lllllC'.h sma ll e~ amount than in cations of a maloperation.

I tha dissolver off-gas sys Increased-radioactivity in stack monitors.

7. I The Scrubber Recirculating!

Same as 6 above. I FI-3 low1 FAL-20 would be Pump 6G-2 is not operating activated in addition to the indication&

listed in a. The Scrubber Tank 6C-3* is allowed to overflow.

If tank is allowed to fill up to the Heater 66-4(a head of 17 feet) the heater woulc vaporize this*solution and . release it to the atmosphe~

through tlie stack. Some radioactive material could be released to the phere, but not of a larger magnitude than in 6 above. However, tank level would remain below this height and feed back to Condenser Catch Tank 6D-3 for a period of time. . . -6 above. ~1-a, PI-2, and~FI~3 would indicate tiigh'level on Scrubber 6C-3. The alarm systems on the Catch Tank 6D-3 would be activated as the overflow from the scrubber tank filled the catch tank. Increased radioactivity in stack monitors. , _____ ._ ......... ,.... .... _ -* -+. D 6,tion Adjust TIC-4 for maxialml cooling water (6C-3. Fill Scrubber TanJc 6C-3 with caustic and/or start. water flow on Scrubber 6C-3. I Start Pump 6G-2. Drain scrubber o y Pump 6G-2*to Low-level Evapo rator Feed Tank 7D-2 to proper height1 regulate water and caustic flow.

9. 10. r; Malooeration Low flow or no heating in Vessel Off-Gas Heater 6E-4. Blower not operating. ( 6K-2 :r 6K-2A) 0 Table 9.113 Continued MALOPERATION CF VESSEL CFF-GAS SYSTEM Resul:t Off-gas will pass through final filters at a than-optimum teq:,erature.

The filters will collect excess i ve moisture and teriorate more rapidly. The whole vessle off-gas system-will lose its tive pressure which will cause some off-gas leakage into the vessel cells. No gas can c~me back from the stack system because of the off-gas check valve. TIC-5 on Heater 6E-41 PdR-13 and PdR-14 will increase.

PdR-131 PdR-141 electrical controller on blowers PIC-1 and PIC-21 PAH*l and PAH-31 PAH-2 and PAH-4. "' \.;! Increase steam flow or start steam flow through Heater 6E-~. . Start blower or'switch to alternate systea (6K-l or 6K-1A) if automatic switch does not work.

Off-Gas High-Leyel Waste Storage System 9.114 Off-gas from the High-Level Waste Off-Gas Storage System is treated prior to its releas~ to the atmosphere.

The system is shown in figure 6.3d. Table 9.114 defines maloperations in this system. Waste Iaok Farm 9.115 Underground tankage is provided for the storage of all process wastes containing nondiscardable quantities of radioactivity.

These will be neutralized wastes largely from the bottoms of the and low-level evaporators.

The system is described in paragraphs 4.88 and 4.89. A discussion of the possible maloperations that may be curred in this system is shown in table 9.114. Also, included in this table is a discussion of the maloperations associated with the storage of Consolidated-Edison liquid wastes. . ' t I r

' 1. Maloceration No caustic in Scrubber Acid Waste Tank ac-1. r-, Table 9.114 MALOPERATION OF OFF-GAS'IN HIGH~LEVEL WASTE SfCRAGE SYSTEM R~sult This allows acidic off-gas t o reach main off-gas waste tank system. Indir.~+inn Corrective

~ction LI-111 low level in tank I Fill Scrubber SC-1 with ec-11 FAL-18 will be set caustic. off and LAL-12 and LAL-13* alarms are set off in control room and at tank farm1 low pH of condensate.

2. I Recirculating Pump SG-3 not operating.

Same as 1 above. FAL-181 PG-71 and as in 1 above. Start Pump 8G-3. Notes no alternative.

3. 4. 5. 6. Scrubber ~-1 is allowed I Caustic will be carried to to overflow.

knock-out drum and to the waste tank~ Cooling fan is not opera-~Off-gas temperature will not ting on Waste Tank Off-Ga be reduced so that most of Condenser (8E-l or 8E-1A}. the waste tank condensate Steam not flowing on denser (SE-1 or 8E-1A). Condenser inlet or exit valves closed (SE-1 or 8E-1A). heat will be-returned to waste tank via knock-out drum. During cold we~ther, if steam line is not operating, coils could freeze and rupture. Pressure will be relieved by waste tank off-gas relief knock-out drum through a 7-foot water leg; will blow seal pot only. (See malopera+

tion of Waste Storage Tank.) LI-111 LAH-12 alarm set I Reduce scrubber level to off in control room and at norma~ by pumping (SG-3) tank alarm. to Knock-OUt Drum 80-6. TR-1-1 and TI-3-1 on Waste I Start fan. Notes fans Tank 80-i wtll increase.

are automatically started. PdAH-4 and PdAH-5. No flow out of steam trap. I Open steam valves. PAH-8 on tank ao-2; PAH-4 on tank 80-1; PR-1 on 80-21 PR-4 on 80-1. *, Valves are locked open. l-_____________

,__....,_, _____________

_ _

Table 9.114 Continued MALOPERATION OF OFF-GAS. IN HIGH-LEVEL WASTE STORAGE SYSTEM Mal operation Indication Corrective Action Result ---. 7. Knock-Out Drum 8D-6 drain All liquid will flow LI-4; LAH-3 both in Open valves to* proper valves closed. through overflow to 8D-l. control room and tank farm, tank or turn on conden-sate pump SG-1 to reduc ; liquid level; open up o : adjust normal flow valv r s. a. Waste Tank Off-Gas The system's negative PRC-3; PdR-2. Start Blower SK-1 or Blower SK-1 or 8K-1A is pressure would not be main-switch to alternate syst not running. tained, allowing seal pot (8K-1A). to blow (see maloperation of Waste Storage Tank). em 9. Filter valves are closed Loss of*negative pressure PRC .. 3; PdR-2. Valves are locked open. or the filter is not in system. operating. (ST-1 or ST-lA: . .. I I I Malo_p_eration

1. !Solution in tank becomes over-concentrated.

Table 9.115 MALOPERATION OF WASTE TAN< FARM Result Indication High boiling temperature in IHigh temperatures recorded; tank; solids precipitation.

low level; sample analysis.

2. !Waste put into wrong tank !Loss of spare tank. via general purpose evapor-Level indication in spare tank. ator discharge.

3. IToo little sparge air. 4. !Acidic waste sent to stor** age tank. ~. !Liquid blown from seal loops. Poor tank agitation; local overheating.

Possible corrosion rate crease in condensate system. Temperature recorders show wide temperature tials; low air-flow readings uneven off-gas flow. If batch of acidic waste is transferred, will not affect pH of tank1 indication will be low p of condensate.

Tank vapors will bypass off-lPAH-2 followed by possible gas condensers; loss of activity alarm. water; possible activity to atmosphere.

a Corrective Action Add water to tank if required.

Lock-out on spare tank. Increase air flow. Route condensate to GPE; analyze corrosive effect (extent);

verify proper amount of free caustic in Ui tank. Add water if required from plant water through knock-out drum 80-6; check valves to insure proper vapor flow through condensers.

........a Table <l.115 Continued MALOPERATION OF WASTE TANK FARM Maloperation

6. ILeak in transfer lines f%om plant to waste age tank. *1 Result Hot waste will flow into line jacket and then into encasement sumpJ high volume leak will then flow to vault sump. 7. IWaste tank develops a leak.lWaste flows to waste tank sump. If pan fills, waste flows to concrete vault. Some release of activity through vault vent. Indication Sample of encasement sump1 LAH-1 LAH-1 on waste tank sump1 activity may be found coming from sump vent. .. ,. . ...... _ -----..........___.......r .* , _, _#..,.._..__.---

* ----------Obtain sample from tank* sump and small sump which first catches spillage before it flows into tank sump. If solution is radioactive, pump back into waste tank. if solution is active, pump to lagoon. Install jumper so that alternate transfer line can be used. Obtain sample of liquid in sump. If active, transfer to lagoonJ if radioactive, return to waste tank. Install able waste transfer pump and line and transfer entire contents of tank to spare tank. I

1. f ! 0 Table 9.116 MALOPERATION OF CONSOLil1'TED-EDISON WASTE STORAGE SYSTEM Malo ration Result Indication Corrective Action . TanJc coil fails. FAH-14 and PCV-9 admits air TAH-15 on tank1 FAH-14. Switch to spare coils. to coi l. High temperature in waste tank and eventual boiling. Acid and*activity to waste tanJc off-gas scrubber.

2. I Tank leak. Tank lea k s to tank vault sump. LI-9 and LAH-10. Transfer solution to spare tank, cease further waste transfer and fuel processing of this fuel. 3. Incorrect solution sition reaching tanks. Possible corrosive damage tolAnalysis of waste before tank or prec i pitate forma-transfer from 70-4 solution tion1 storing of unnecessary inventory.

solution volume. Analysis of all waste before transfer and adjustment of off. specification material1 blank influent lines in CPC. 4. I Vent valve closed at out-* 1Waste tanks will vent into !Higher pressure on tank PR-741 Lock valve open .* let of scrubber (SC-1). tank vault and eventually to possible activity release1 atmosphere.

probable pick-up on galllnj monitor in area. s. High activity in cold Indicates leak of solution Radiation~larm h i gh; RAH-16'Switch to spare coil. water return. into coils. 6. ILoss of cold water. Tank temperature will increase.

TR-2 through 6J TR-2-1 through 6. " Switch to plant process water.

MALOPERATION SOLVENT TREATIENT SYSTEI§ 9.117 Solvent treatment facilities in the plant consist of three 'identical treatment systems which service certai11 sections of the plant. Ea c~ system is identical in process and equipment make up and operation.

l he three systems are tntended to provide organic to, and receive organic from, the following areasa No. !-Partition cycle and alternate for first cycle plutonium; No. 2-First cycle uranium, first cycle plutonium; No. 3-Second cycle uranium. Each solvent system is designed to operate continuously in conjunction with its applicable solvent extraction battery. The processing equipment consists of a packed carbonate wash column, two-stage solvent washer and an acid wash column. The fresh carbona t e solution is introduced to the second stage of the carbonate washer, continues to the washer first stage; then to the carbonate wash column and finally to aq9eous waste. The influent organic stream from the solvent extraction battery is first contacted in the carbonate wash column, then proceeds to number land m umber 2 stages of the solvent washer. Mixing in the solvent washers is accomplished by combining an aqueous and organic flow to the suction side of the recircul&tion pump thereby performing pump mixing. The recirculating pump then dischargel the mixtt J re back into the solvent washer for phase separation and continued mixing and circulation.

The aqueous waste solution from the carbonate and acid wash column is collected in Catch Tan~ 130-7, then transferred to Hold Tan~ 130-8, for sampling and analysis.

These tanks are poisoned with boron glass Raschig rings to prevent a criticality incident.

The solution will be ferred to the Low Level Waste Evaporator Feed Tan~ 70-2, for subsequent processing in the waste system or to rework*if sufficient product loss has been encountered.

> Solvent adjustments can be made to each system during the processing by making up the required TBP and/or diluent in Tank 14D-48 and ferring to Tank 140-18. The solvent is washed in 130-18 and is then transferred to the appropriate solvent system storage tank. Possible maloperations and.their results are indicated ancl described in Table 9.117. The Table has been prepared for one system only but is applica~le to any of the three; thu~ the. identifying instrument numbers have been omitted. l l I I I ! i l 1. 2. 3. 4. 5. Mal operation

(' Table 9.117 SOLVENT TREATMENT SYSTEMS Result Indication Washer circulation

~s---Colvent and aqueous-:111

--, Fl~-::-orde:-on puaps turned off. . rcontinue to flow through the will show no flow. ~ystem, but contacting will e inadequate

.with an effect n the quality of solvents.

0 Corrective Action l 'Turn on puq,s. Valves i~roperly set around circulation pumps. :Either no sol vent flow or solvent is pumped to other parts of the system. Depletion o_f solvent levelsfAdjuit vahes properly.

in storage tariks. Increase in 130-18 or storage tanks 130-15; 16 and 17. Air lifts turned off around solvent washers. Carbonate solution level wil~LIC's in washers. rise in washers until it ~ihally overflows to acid ash column. No carbonate ill flow to carbonate wash olumn. !~roper steam supply to solvent washers. IToo low a temperature in I TI too low in washers. ~olvent washers leading to !Poor washing. Carbonate solution sup-*tq:,roper washing; uranium. plied in wrong strength or land plutonium carry over quantity or turned off. pay not be removed. FI on carbonate solution inventory, solution ysis. Turn on air lifts. Reset te8')erature c&tor controllers to proper temperature.

Set correct flows on controller, analyze acid wash before use.

I I I i ' 1 6. 7. I Table 9.117 Continued Mal operation Result Indication Acid solution &UPPlled in t::vent will not be propnlv I FAL on supply llne1 sol. u.;. wrong strength or quantity onditioned for reuse. . 1. tion inventory1 solution or turned off. lsions and product 1011 analysis.

High uraniua or plutoniua t*ibonate scrub will reaove loss to waste solvent ranium and plutonium ana streaas. ollect them in 13>-7 which ontains fixed neutron ,oison to prevent ty. High uraniua or plutoniua in 130-8 analysis.

Corrective Action 1 Set correct f lOIIS on ontrollen, analyze acid wash before use. rt high product 'aste, check ion coluan operating onditions.

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ML18081A222

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SUMMARY

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