ML20236B663

SER Re West Valley Demonstration Project Supernatant Treatment Sys.Addl Measures Should Be Taken by DOE to Minimize Radionuclides in Treated Supernatant

ML20236B663
Person / Time
Issue date:	09/30/1987
From:	NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
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ML20236B651	List: ML20236B648 ML20236B663
References
REF-PROJ-M-32 NUDOCS 8710260292
Download: ML20236B663 (29)

Text

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SAFETY EVALUATION REPORT ON THE

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WEST VALLEY DEMONSTRATION PROJECT SUPERNATANT TREATMENT SYSTEM l

PROJECT M-32 U.S. NUCLEAR REGULATORY COMMISSION 1

0FFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS SEPTEMBER 1987 s

8710260292 PROJ 971009 PDR PDR M-32

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p TABLE OF CONTENTS f

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Section- Page.

1. INTRODUCTION. . . . . . . .. . . . . ........... 1

2. THE SUPERNATANT TREATMENT SYSTEM . . . . . . . . . . . . . . 3 2.1 The Physidal System . ................. 3 2.2 The STS Process . . . . . . . . . . . . . . . . . . . . 8 2.3 Administrative Controls . . . . . . . . . . . . . . . . 12 3, PERFORMANCE EVALUATION ............ .. .. 13 e

3.1 Normal Operations . . . . . .............. 13 3.2 Off Normal ard Accident Conditions. . . . . . . . . . . 14 3.2.1 Externally Caused Releases . . . . . . . . . . . 15 3.2.2 Internal Phenomena . ... .......... 20 3.3 ALARA Review. . . . . . . . . . . . . . . . . . . . . . 22 3.4 Independent Safety Reviews. . . . . . . . . . .... 23

4. DECONTAMINATION AND DECOMMISSIONING. . . . . . . . . . . , . 23

5. OPERATIONAL SAFETY REQUIREMENTS. . . . . . . . . . . . . . . 25 APPENDIX A - References . .. ................ . 26 i I

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1. - INTRODUCTION' l I

l The U.S. Department of: Energy (D0E) is. conducting 'a high-level waste

solidification demonstration program at the Western New York Nuclear Service '

j Center. The initial 1 step,of this' demonstration is the processing of radioactive  !

iliquidwastescuNrently. stored'inthehigh-level;wastetanks. The waste'.will ,

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'be' processed through.a series of ion exchange columns.to remove'the dissolved 4 radioactive. constituents, primarily Cs-137, The-ion exchange medium will be a naturally; occurring, mineral: called zeolite. The zeolite particles, to which .

the Cs-137-will adhere,-will be further processed at a later date into a final f

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form.

The system'used to perform this separation is called the Supernatant i Treatment System-(STS). This review addresses the public health and safety aspects of processing of the supernatant. liquid high-level waste through the STS.

The major source of information.used in evaluating this system-is the DOE report ]

entitled:" West Val'f ey De.aonstration Project Safety Analysis Report, Volume III, Supernatant'. Treatment System, Rev.-1." This-Safety Analysis Report describes

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the structures and equipment that will be used.for the process system, the pro-cess itself, the' discharge's that will occur during normal operation, and the j releas'es 'of radioactive material that could occur during off-normal or accident  !

conditions.

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l In-the sections that follow, we present a brief overview of the STS, j focusing on its equipment,' process, and operating procedures as they relate to public health and safety. The SER examines the releases to the environment j that are expected during normal operations and the releases that could occur during potential accidents. The offsite dose consequences of these releases are estimated and compared to applicable standards, i

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g: The conclusions of;this SER can be summarized as follows:

t ID Routine gaseous effluents from the STS-are expected to be low.'. A

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hypothetical member of the public spending'one year at the site. boundary during-continuous STS operation would receive a.50 year whole-body dose' commitment-of. q

about:1E-11 mrem. This dose is.far below al.lowable limits'. .

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2.  : We_ examined potential STS accidents and considerithe most severe i

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credible accident:to-be.a failure of the STS HEPA filters, enabling unfilteredi "

venti.lation air from the valve aisle, pipeway, and shield _ structure to be released'to the environment'. The 50 year whole-body-dose commitment to'a >

. hypothetical individual at the.' nearest site boundary-from a two-hour-long

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Taccident would be less than 0.001 mrem. I

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3. We conclude.that additional. measures should be taken by DOE to l

1 minimize the radionuclides in the' treated supernatant. Specifically, DOE J should make a commitment to remove at least 99.9. percent _of.the cesium in the- l t

supernatant, and-should. establish'~an operational safety requirement related to'  !

removing as much cesium beyond the minimum fraction-as reasonably achievable.

DOE should also consider whether some other radionuclides, particularly Sr-90 and ' i

, transuranic, could be partly removed for the supernatant without-seriously Eaffecting cesium-removal', We recommend that DOE perform a relatively short test

.run (perhaps'two or three full cycles) before they begin continuous operation of the STS, following which decontamination factors for cesium and other radio- l

-nuclides can be accurately measured anc' adjustments can be made in operating practices should the data warrant.

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! 2. 'THE SUPERNATANT TREATMENT SVSTEM i The STS is being' constructed in the area of.the existing high-level waste tanks at the West Valley site. Figure 2.1 presents an overview of the site and l identifies the location of the existing high-level waste tanks and the STS relative.to the other facilities at the site. The following sections consider the STS structures, the STS process, and the administrative-controls to ensure that the STS is operated safely,  ;

c, 2.1 The Physical System j i

STRUCTURES i

i The STS consists of four major structures. The first is a confinement shield over the top of the existing 8D-1 vault that provides space for piping serving process equipment and beams to support process equipment suspended in the existing 80-1 tank. The second=is a pipeway through which piping passes as it connects the process equipment with the valve aisle. The third is the valve aisle, from which all process flows are controlled. The fourth is the STS support building, from which the process is controlled and which provides process materials (ion exchange zeolites) and utilities. The location of these structures relative to each other'and the existing high-level waste

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tanks is shown in Figure 2.2. j The shield structure is made of concrete and steel, and provides a 10-foot-high shielded space over the northwest third of Tank 80-1. Within this shielded l

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space are steel I-beams used to support the process equipment (ion exchange I columns, filters, and tanks) that will be used for the STS process. The structure provides space for the piping that runs between the various pieces )

of process equipment and the valve aisle. Soil overburden provides radiation I shielding for the pipes. The load of this new structure is primarily carried by pilings positioned around the existing 8D-1 tank vault.

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l 8 SHIELD STRUCTURE 1

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Figure 2.2 STS FACILITIES PLAN (Reference 1)

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< 1 l' l F The pipeway, located between the shield structure and the valve aisle, is l

! also constructed of reinforced concrete. It is underneath part of the STS sup-port building, and has an extra-thick roof so it does not need a soil overburden.

l The load of this structure is primarily carried by pilings like those used for

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the shield structure. l The valve aisle is a box with front, top, and two sides comprised of steel ,

4 plates 1-foot thick. The back wall is a 1-inch steel plate that supports the valves and jumpers necessary to control the flow to the process equipment sus-pended in the 80-1 tank. The valve aisle box rests on a concrete floor. The l

front of the valve aisle box has viewing windows for cbserving the condition of the equipment on the back valve aisle wall and manipulators for operating the equipment to control the process flow or performing maintenance work. All high-level waste flow control will be performed inside this structure. Any leakage from the flow control operations will flow to a valve aisle sump that will be transferred to 80-2 via an air eductor.

The STS support building is also constructed of concrete and steel, and 1

contains the control room for the STS, the cold support facilities, and cooling

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water and ventilation support equipment. The support building is a two-story structure, the lower half of which is a shielded structure that adjoins the steel valve aisle and provides the operating area for the valve aisie. The ,

upper half is a sheet metal stru:ture that houses the control room and the equipment for handling the zeolite prior to its introduction into the STS process.

Another structure associated with the STS facilities is a new steel-lined, concrete pump pit located atop Tank 8D-2. Details are not provided in the SAR but the Dames and Moore report by Gates (Reference 1) mentions the existence of a drain.

VENTILATION SYSTEMS In addition to these structures, there is an STS ventilation and off gas treatment system that conducts air from the STS support building through the 6

L operating areas and into the contaminated areas of the valve aisle and shield structure. The air is exhausted at the rate of about 4100 cfm from the con-taminated areas through a mist eliminator and two HEPA filters before it is discharged to the atmosphere. The equipment for treating the off gas is located in a sheet metal structure adjacent to the existing 80-1 tank. Although not shown in the SAR, the structure is located on the northeast side of Tank 80-1.

The STS off gas will be monitored for radioactivity.

-The NRC staff reviewed and examined the installed STS ventilation system and verified that the STS SAR representation of the system is accurate. The roughing filter and two stagas of HEPA filters that constitute the primary barrier to atmospheric releases have been tested to demonstrate particulate removal efficiencies in excess of 99.95 percent per HEPA filter. Using the FIRAC ventilation system accident analysis computer code, the NRC staff has established that the system can function as designed for normal operating conditions.

The high-level waste tanks will continue to be served by the existing tank farm ventilation system, which maintains the tanks at a negative pressure and treats the off gas from these tanks. The off gas from Tanks 80-1 and 80-2 and the ventilation of the equipment suspended in Tank 80-1 is processed through a condenser and the off gas from Tanks 8[' and 80-4 is processed through a caustic scrubber. The combined off gas (about 150 cfm) is routed through a knockout drum, a demister, a heater, and one of two HEPA filter and blower trains.

INTERCONNECTING PIPING The raw supernatant will be transferred from the 80-2 tank to the shield structure over the 80-1 vault through a buried, double-walled stainless steel pipe within a stainless steel conduit. The top of the conduit is nowhere less than three feet below the surface. The conduit extends from the pump pit above l Tank 80-2 to the STS pipeway adjacent to Tank 80-1. The double walled pipe i inside the conduit is sloped so that any liquid leaked into the annular space will flow back toward Tank 8D-2.

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T LAt'the' structural.' interface b'etween the-conduit'and the STS pipeway,:the iconduitLis encased;in an' outer; larger-diameter pipe.

. Inside the STS.'pipeway;and shield structure there are double-walled pipes:

'for the lines /which will. handle raw-supernate. The annular space ~for these double-walled pipesiis neither drained'nor monitored. DOE may wish to consider some modifications:to permit' detection'for drainage of any leakage of supernate'into the annularsspace, 2.2 The STS Process t

The STS process can be viewed as five operation's: . feed transfer and preparation,-ion exchange processing, liquid effluent handling, resin removal, '(

and: fresh resin loading, .The following paragraphs discuss each of the'se  ;

operations in more detail.

FEED TRANSFER AND PREPARATION' The feed for the STS will be decanted supernatant from'th'e 80-2. tank. The,

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' solution will_be pumped from 8D-2 using a submerged air-driven; pump through a.

filter. The. filter is of the crossflow type where the majority of the flow is  !

across the~ filter face and only a small portion of the flow will be through the

-filter. 'The pump capacity is 40 gpm at-80 psig. The material that passes  ;

through the filter (3-6 gpm) will be stored in a feed tank, The material'in the feed tank can be diluted or pH-adjusted by:the addition of caustic soda (NaOH) 'f or nitric acid (HNO3 )'

l The supernatant will be pumped from the feed tank through a cooler to the

' ion exchange columns. The cooler reduces the temperature of the supernatant to l

about 6 C before it is processed through the ion exchange columns. The pump )

'that moves the material through the cooler and the ion exchange columns has a l

capacity ~ of 26 gpm at 95 psig and a net forward flow rate of 1-3 gpm, ION EXCHANGE SYSTEM

'The ion exchange system consists of four columns and a post-column filter.

A schematic of the system showing the major elements present during continuous 8

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operation is presented in Figure 2.3. A schematic depicting the proce:s equip-ment and flow directions for ion exchange column refill /changeout is presented' in Figure 2.4. Three of the columns will be online during processing while the.

fourth column is having loaded resin removed and fresh resin added. The ion  !

1 l exchange process consists of three columns in series. Each column is 3.4 feet  ;

in diameter, 14,5 feet high, and has a packed length of 6 feet. The solution  !

is pumped through the columns downward with a superficial velocity of 83 cm per hour. The ion exchange resin is zeolite (IE-96), which is highly selective for cesium over other cations, including sodium, the major cation in the feed.  !

This. selectivity is decreased by higher pH (up to 13) and increased by lower temperature (down to 6 C). The minimum performance goal is expected to be 99.9 percent removal of the cesium from the feed supernatant (i.e., decontam-ination factor = 1000). The effluent from the columns will be passed through a sand filter to remove zeolite fines. Each column has a vent line and a pressure relief device that are connected to the supernatant feed tank. After filtration, the ion exchange column effluent will be transferred to an under-ground storage tank (80-3) where the liquid will be accumulated and sampled before transfer to the Liquid Waste Treatment System.

When the concentration of cesium in the effluent from the ion exchange column reaches a predetermined value, the column will be taken off line, pressurized with air to force contaminated supernatant back to the high-level waste tank (80-2), and then rinsed with water. The rinse water will also be sent to tank 8D-2. The spent zeolite will then be removed from the columns by '

backwashing with water to expand the bed, and by opening an outlet valve on the '

bottom of the column so the zeolite is dumped to the bottom of tank 80-1. The spent zeolite will be kept under water so that the heat associated with radio-active decay can be dissipated without overheating the zeolite. New zeolite will be added to the column by using a slurry transfer system to transport zeolite from the storage area of the STS support building to the zeolite columns in Tank 8D-1.

INSTRUMENTATION AND CONTROL t

The STS system description in the SAR discusses the general requirements for instrumentation and controls. The process instrumentation for the STS l

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1 process is presented in Table 6.5-1 of the SAR. The table describes the nature of the p'rocess instrumentation and the type of actuation for the various valves.

} The table aT30; identifies part of the control logic for the specific instruments 1 orvalves,,',IIf ,

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Administrate.tve costrols also play a role in orotecting public health and 1 .,

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'm safety. ' T, N,xst' specific of the WVDP administrative controls are the WVNS t procedures that implement the requirements of DOE Orders. These WVNS procedures M} address the requiremd)ts of, projects as'they move from the design stage through ds the operational'9 bass. The' general WVNS policies and procedures were previously 1 tl reviewed by, the 11RC staff and were determineo to be a satisfactory system for

,orbtectd9 thd' hvblic health and safety (Reference O).

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a l Withthda[bmittalofsrMcificinformationontheSTS,theNRCstaff

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reytw ed the specific STS administrative procedures that provide for public

,, shealth and protection during' operation. These specific administrative proce-

! ,1 e dures are identified in Chapter 10 of the STS SAR and include preoperational ,

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,f 3k test procedures, training programs, normal operating procedures, and emergency l h,' procedures.

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' j f ', ' I 4 Th,e' prep 6Mational tests are intendet. to test the functioning of various A com)nenu, first in a static situation and ther with a simulated supernatant.

Perhrhirig these checks will provide greater assurances of the ability of the

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., I,n iodition to the preoperational tests, the NRC staff recommends that

,/ the, performance of the STS ion exchange system be carefully evaluated after two

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' tot.hreeepmpleteionexchan0{: cycles (8 to 12 column dumps) to help establish

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tktinum operatint.;,s procedures for obtaining high decontamination factors under j reasonable operathne, conditions (Spe Sectibn 3.3).

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I i The discussion of.the training program in the SAR ident ti [the topics

.that/wi'Mbe.coveredinoreratortraining. It addresses' formal classroom  !

train (ng,on-the-jobtraining,andtesting.

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The. staff has reviewed the outline

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oppgv con 3ractor's STS-related operator qualifi ation currit:ulum and course conterd This information covers the topics important to public health and safety,

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3. PERFORMANCE EVALUATION ,

q The NRC Staff has reviewed the STS and made independent assessments of the ' potential for release of radioactivity to the., environment. For those situations where reledses were' judged to be possible, independent estimates were made of release quantities add off-site dosei. This chapter of the Safety Evaluation Report preepnts.the results of our independent evaluation.

The releases that could result from normal operation are presented in Section 3.1 and the releases that could occur under accident conditions are presented in Sectica 3.2. j 3.1 Normal Operations k

i The STS, under normal operating cor.ditions, will involve the transfer of 1

high-level waste solutions from Tank 8D-2 to Tank 8D-1 where the STS equipment  !

is located. 'There will be no liquid discharges to the environment directly from the STS. All liquid effluents , um the STS will be transferred to the )

Liquid Waste Treatement System which will be +ha :ut iect of a future Safety Evaluation Report. The only routine releases of radioactive material expected to occur from STS operations are those associated with the ventilation systems.  !

These releases can be divided into two categories: 1) gaseous releases from trx S qh-level waste tanks discharged through the existing tank farm ventilation systam aim 2) gaseous releases from leaks and spills in the shield structure, pipewg , and alve aisle, discharged through the STS off gas system.

The off ga n s generated by the high-level liquid waste stored in Tanks 80-2 and 80-4 are processed through the tank farm ventilation system. The present 1

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l 13

release of off gases from the tank farm ventilation system has been reviewed by the NRC staff and found to present no danger to public health and safety (Reference 3). The 50 year dose commitment at the site boundary due to the tank farm off gas is approximately 7E-3 mrem per year of exposure.

The site-boundary doses from the STS ventilation system have also been estimated. There will be no large volumes of free liquid that could release material to the ventilation system. There is a possibility of small leaks and small-scale accumulation of radioactive material in the STS sumps. With a con-servative assumption that the sump is always full of untreated supernatant, an upper bound estimate on the site-boundary dose has been made. The results of I' this calculation .w presented in Table 3.1. The analysis assumes that there '

are 10 liters of <upernatant in the sump. The dose factors for unit releases are from previots independent AIRD05-EPA runs.

The dose from the STS off gas system (IE-11 mrem / year) will be insignificant compared to that from the tank farm ventilation system (7E-03 mrem / year), which is itself far below standards norma'1y applied to nuclear fuel cycle and waste management operations. The EPA regulations for fuel cycle facilities (40 CFR 190), for example, limit exposure to the maximally exposed off-site individual to 25 mrem / year whole body.

3.2 Off-Normal and Accident Conditions The SAR recognizes that off-normal or accident conditions could result in off-site releases greater than those associated with normal operation. The doses for normal operation are primarily associated with the existing tank farm ventilation system where radionuclides are released to the ventilation air from r the free surface of the high-level waste and passed through a condenser, a knockout dru,n, a demister, and a HEPA filter. In order for a larger release to occur, there would have to be damage to the off gas system or a breach of containment that would bring liquid wastes into direct contact with the environment.

The most serious credible accident identified by the staff was the failure of the HEPA filters for the STS off gas system. If this were to occur, : mall 14

amounts of' radioactive material resulting from spills in the valve aisle, pipe-way, and shield structure would be released directly to the environment. The NRC staff develooed independent estimates of the off-site dose that would result ,

from such an accident, The staff estimated a 50 year whole-body dose commitment of less than 0.001 mrem to an individual at the site boundary during a two-hour exposure period.

3.2.1 Externally Caused Releases 1 Af ter a-_ site review and evaluation, the staff has identified the natural I disasters that are credible and should be recognized in the design of structures

-and systems (Reference 2). Three natural phenomena were identified as being  ;

potentially significant at the site and therefore requiring attcntion in the design of structures and systems that will handle high-level waste. The phenomena are seismic disturbances, high winds, and tornado missiles.

SEISMIC-In the Overview SER on the WVDP (Reference 2), the NRC agreed that the design basis earthquake at the site would be one causing a horizontal surface acceleration of 0.1 g. The STS design criteria and the STS SAR identify seismic design criteria for the 80.1 and SD-2 structural additions, the pipe chase between 80-1 and the valve aisle, and the STS building below grade to be the Uniform Building Code.for Zone 3 with an importance factor of 1.0. (Equipment ,

suspended in the tanks, pipe chase, or valve aisle will also comply with ANSI A 58.1.) There were no seismic requirements for the STS ventilation system or the above grade portion of the STS support building. (SAR Table D.5.2-2).

To determine the performance of the STS structures under design basis earthquakes and situations beyond the design basis, Dames and Moore performed a simplified Seismic analysis of the STS barriers and reported the results in a document titled "STS Confinement Barrier Integrity Review, West Valley demonstration Project" (Reference 1). The report described in general terms the nature of the seismic analysis or indicated that engineering judgment was l-l l

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___j

h '.

TABLE 3.1 CALCULATED 50-YEAR WHOLE-BODY DOSE' COMMITMENT FROM NORMAL OPERATION OF THE STS OFF-GAS SYSTEM DURING STS OPERATION i

ACTIVITY ANNUAL ,

IN THE OVERALL RELEASE CALCULATE LISOTOPE. STS SUMP SYSTEM ' AMOUNT DOSE (C1) DF (C1) (REM /YR)

H-3 5.15E-04 1.00E+05 5.15E-09 1.06E-16 C-14 6.85E-04 1.00E+05 6.85E-09 1.39E-14

-NI-63 4 23E-03 1. 0')E+13 4.23E-16 7.26E-21

.SR-90 1 45E-02 1.00E+13 1.45E-15 8.93E-19 TC '8.00E-03 1.00E+13 .8.00E-16 4.99E-19  :

RU-106 1.40E-05 1.00E+13 1.40E-18 4.18E-22 SB-125 3.60E-04 1.00E+13 3.60E-17 2.25E-20

.TE-125m 8.30E-05 1.00E+13 8.30E-18 1.22E-23 I-129 1.05E-06 1.00E+05 1.05E-11 1.15E-14 CS-134 9.70E-02 1.00E+13 9.70E-15 5.29E-18

'CS-137 3.72E+01 1.00E+13 3.72E-12 1.87E-15

.PM-147 1.09E-03 1.00E+13 1.09E-16 1.45E-22 EU-154 7.45E-05 1.00E+13 7.45E-18 4.65E-21 EU-155 1.37E-05 1.00E+13 1.37E-18 3.36E PU-238 6.50E-04 1.00E+13 6.50E-17 7.04E-19 PU-239 1.25E-04 1.00E+13 1.25E-17 1.51E-19 PU-240 9.50E-05 1.00E+13 9.50E-18 1.15E-19 PU-241 7.50E-03 1.00E+13 7.50E-16 1.87E-19 4

TOTAL 7.22E+01 2.25E-08 2.73E-14 16

used to predict performance of STS structures during' seismic events. The con-clusions of the report are that the system will remain intact for the design

. basis earthquake of_0.1 g free field lateral acceleration. These conclusions were incorporat7d into the DOE SAR. Seismically weak areas were identified as:

o the interface between.the high-level waste tank and perlite insulat!on blocks where the blocks could become crushed resulting in possible tank rupture at 2xDBEQ o the: vault wall which could crack at about 2xDBEQ o the steel column under the Tank 80-2 pump pit which could fail at 2xDBEQ.

o the steel walled pump pit which could rupture at 2xDBEQ (It is noted that the analysis-by Gates of Dames and Moore which is cited as the reference for the WVNS SAR, claims that the pump pit is stable to greater than 4xDBEQ) o the pile foundation for the pipeway associated with the STS building which could fail at 2xDBEQ o the steel and reinforced concrete structure around the valve aisle which could slide at about 1.5xDBEQ o the shield structure over the 80-1 tank which could crack at about 2xDBEQ (It is noted that the report by Gates of Dames and Moore which is cited as the source of information by WVNS states that the structure could crack at 1xDBEQ)

The NRC reviewed the calculations which were the basis for determining the seismic capability of the high-level waste confinement system. The review focused on the STS transfer line, the pump pit over 80-2, and the shield structure over t

17 ,

l

_= _ -_

l.

L 80-1. The NRC staff concluded that the analyses performed by DOE was conserva-tive and that.these important aspects of the STS confinement system will be capable of withstanding the site design basis earthquake with adequate safety margins.

^

HIGH WINDS The previous NRC staff SER (Reference 2) agreed that the design basis winds for.the site would be a tornado with a wind velocity of 160 mph and straight-line winds with a velocity of 165 mph. The performance of the STS structures and systems under high wind loads was not analyzed in either the SAR or the. Gates report. The NRC staff believes nevertheless that the structures which confine the radioac+.ive material will be able to function under these winds because the structures are massive (18-36" of reinforced concrete) and most of  ;

~the structures are below grade. The most vulnerable of the STS structures is the upper level of the STS support building and the STS off gass structure both of which have a sheet metal outer wall. There is no analysis of this portion of the structure in Reference 1, but the NRC staff has assumed that a high wind j could knock down the metal walls and render equipment inoperative in either the STS structure or the STS off gas building. No credible control system failure j or ventilation system failure could be postulated which resulted in a release greater than the. filter failure accident discussed previously.

MISSILES The previous review of the site by NRC staff identified tornado generated missiles as generally credible phenomena but deferred the identification of tornado missile analysis until specific facilities were considered.

The STS design criteria document states that no tornado generated missile loading is required for the STS design, but the ability of the structures to withstand tornado generated missiles was analyzed by Dames and Moore in the report "STS Confinement Barrier Integrity Review West Valley Demonstration Project." The analysis predicted the consequences of two missiles (a timber r plank 4. inches by 12 inches by 12 feet weighing 139 pounds traveling at 87 miles per hour; and a pipe 3 inches in diameter, 10-feet long weighing 76 pounds 18

i l

1 traveling at 50 miles per hour) impacting on concrete, steel plate, and soil.

Five analytical methods were used to analyze concrete' penetration, one method i

L was used to analyze steel penetration, and two methods were used to analyze I soil penetration. The analysis concluded that the below grade structures had safety margins of 3 to 6 for the DOE-defined design basis missiles.

The staff supports the use of modified Petry formula and the Ballistic Research Laboratory formula as analytical tools for predicting the ability of missiles to penetrate structures. The staff performed its own independent analysis of the missile penetration resistance for the STS. Missile impact i analysis can be complex because of the geometry, relative orientation, mass, tnd energy considerations. For the STS structures in this review, the staff chose to perform a simplified bounding missile analysis by using the guidelines provided in NUREG-0800, Section 3.5.3, which define the minimum acceptable barrier thickness for nuclear power plants to resist a variety of missiles that are more severe than the DOE spectrum of missiles. The table in NUREG-0800 j shows that the West Valley STS structures must have a roof thickness of at least 16 inches and a wall u.ickness of at least 20 inches to withstand the j more stringent missile spectrum for a concrete strength of 4000 psi.

Those parts of the STS structure that provide confinement for high-level waste (the shield structure, the pipeway, and the lower portion of the STS support building) have concrete roofs at least 2-feet and in some instances, 3-feet thick. The structures with 2-foot concrete roofs also have 8 feet of l soil cover, which is more than required according to NUREG-0800. The wall thickness for those structures that confine high-level waste range from 3-feet

-thick for the shield structure, to 18 inches for the operating aisle, to none for the transfer pipe routed between 80-2 and 8D-1. The wall thickness for the shield structure is adequate without taking any credit for the soil over- I burden. The 18-inch wall thickness of the operating aisle is slightly less than the 20 inches called for in NUREG-0800, but there is also a 12-inch steel  ;

wall associated with the valve aisle that must also be penetrated before high-level waste confinement would be breached.

19

s The NRC staff. reviewed the calculations which were the basis for the estimates of the ability of the STS transfer line and the 80-2 pump pit to withst'and tornado generated missiles. The staff' concluded that the DOE calcu- )

lations were conservative and the STS transfer line and the tank 8D-2 pump pit j would withstand DOE design basis missiles.  !

l l

The upper portion of the' STS support building as well as the STS ventila- 1 tion system structure are metal structures and are not considered capable of withstanding design. basis tornadoes. The upper portion of the STS support building.contains the contral room for the ion-exchange process. The staff review did not identify any failure mechanism of the control panel that would result in the release of untreated supernatant to areas which have not been designed to handle such material. Missile penetration of the STS off gas system would result-in greater releases due to the loss of HEPA filters as discussed previously (p. 15).

3.2.2 Internai Phenomena The NRC staff also examined the STS to determine credible internally initiated events that could result in the release of radionuclides to the environment in quantities larger than associated with normal operation. Four '

different types of events were examined - criticality, fires, explosions, and spills.

CRITICALITY The SAR says that the STS feed will be monitored so that fissile material concentrations do not exceed the values of the proposed operational safety requirement 0.11.4.4, which establishes fissile plutonium concentration limits as a function of fissile uranium concentration. The SAR also makes the state-ment that there is no fissile material concentration mechanism down stream from the point where fissile material concentration is measured.

I The NRC staff reviewed the criticality safety limits presented in Table D.11.1-1 of the SAR. The values in the WVNS table were checked using the KENO Monte Carlo code to calculate the infinite multiplication factor, k-infinity, l

l 20 L_ _- _ _ _ _ _

'for the entire range of concentration in a 7 molar sodium nitrate solution.

In all instances, the calculated values of k-infinity were less than 0.95, indicating that the DOE safety limits are conservative. As a further check, we'. considered the consequences of plutonium concentration on the zeolite, j A maximum of 22 grams of fissile material dissolved in the supernatant would j be available in the volume fed to any column. Even in the unlikely event that all of the plutonium adhered to the zeolite, criticality would be impossible.

The staff, therefore, considers criticality to be'an incredible event.

FIRES The discussion of fires'in the SAR focuses on the low potential'for fire due to the limited amount of combustible material normally present in the STS structures. The SAR states that the administrative control measure of limiting combustible materials will be supported by fire suppression systems in the operating aisle and the off gas treatment system building. The SAR also states that fire detection equipment, alarm systems, and suppression systems will be installed as required by the Radiation and Safety Control groups.

The NRC staff reviewed the STS to make an independent assessment of the potential for fires end.the associated consequences. Based on the design details provided in the SAR, we expect that there will be low combustible material loadings under normal situations, particularly in those areas where radioactive material is being processed. We note that maintenance operations could present additional potential for fires because these activities can include both increased combustible material loadings and increased potential for ignition sources. The procedures developed for STS maintenance operations will be reviewed by NRC fire protection personnel to ensure that appropriate fire protection measures are taken.

EXPLOSIONS The NRC staf f examined the STS to determine independently if there is the potential for an explosion at the STS.

21

I i

1 A potential mechanism for an explosion in the STS would be the inadvertent  !

introduction of some improper material into the process, such as an organic ion exchange resin instead of the inorganic zeolite. The SAR and supporting docu-mentation do not discuss the administrative controls which are used to prevent the introduction of improper materials into the process. The NRC staff discussed the issue with site personnel who stated that the STS operating procedures would-

'have a data sheet which would call for recording of zeolite lot number and pro-

]

perties. -The site personnel believed that this would provide adequate protection against inadvertent. addition of undesirable process chemicals. The actual operating procedures and the use of the data sheet will be verified by NRC. i per'sonnel at a later site inspection visit af ter operations have begun. l SPILLS l The potential of spills was reviewed by the staff. Following this review, the staff concluded the potential for spills due to failure of pipes and valves does exist but that the off site consequences of these spills would be very

~

minimal because of the effectiveness of the STS off gas ventilation system.

3.3 ALARA Review 4

The NRC' staff reviewed the SAR to determine if the radiological health and safety of the public is being protected during STS operations. A central aspect of radiological protection is the concept of maintaining doses and con-centrations of radioactive material in unrestricted areas as low as is reasonably achievable (ALARA). ,

While the DOE and WVNS have orders and policies calling for the application  ;

of the ALARA principle to their operations, the SAR and associated material did not demonstrate that the WVDP was applying the principle to the question of I what constitutes adequate decontamination of the supernatant. The only specific requirement for performance of the STS is the operational safety require-ment D.11.1.4, which limits the concentration of Cs-137 transferred to the LWTS to 50 pCi/mi, Allowing for no dilution, this corresponds to a cesium removal I

fraction of 0.987. Laboratory data indicate that a cesium removal fraction of at least 0.999 can be attained. In order to min Mize the amount of radioactive f

22

[. l u 1

{'

I m a '. i . tal routed to the low-level waste' streams'to the extent practicable, the q NRC recommends that administrative controls be established that clearly call for the recycle of supernatant which has not had at least 99.9 percent of the

. incoming cesium removed. The performance of the STS process with respect to rem.oving Cs, Sr, and alpha emitters should be closely monitored during the startup test recommended in Section 2.3.. These data should be evaluated in order to deterniine the potential for improvements in overall radionuclides removal efficiency.

3.4 Independent Safety Reviews The NRC staff had previously reviewed the WVNS procedures which implement DOE orders and found that there are two organizations at the site which provide independent reviews of safety issues of new systems prior to start-up. These two organizations are the Radiation and Safety Committee and the Operation Readiness Review Panel. The staff believes that the proper functioning of these organizations is important to assuring that public health and safety is  ;

protected in an operation such as the West Valley Demonstration Project. The

taff will audit the reviews of these organizations.

4. DECONTAMINATION AND DECOMMISSIONING After the operations of the STS are complete, the facility will be decontaminated and ultimately decommissioned. Since the decontamination and decommissioning activities have the potential for impacting public health and safety, the NRC staff reviewed the status and content of the plans. The purpose of the review was to determine if an effort was being made by DOE to minimize the impact of decontamination and decommissioning on public health and safety.

I While several parts of the NRC regulations call for the preparation of decommissioning plans, the requirements are not in all the sections of the regulations and there are no specific requirements for the content of these plans. The NRC has published proposed changes to the regulations that would directly address decontamination and decommissioning. These proposed rules

[ (Reference 5) were used to help establish criteria for this particular review.

l Three criteria were developed for and applied to this review. The criteria 23

i were: (1) did DOE take measures during the design and construction of the STS to facilitate eventual decontamination and decommissioning of the facility, "

(2) does DOE plan to take measures during the operation of the STS to facilitate "

the eventual decontamination and decommissioning of the facility, and (3) has -

DOE prepared a preliminary decontamination and decommissioning plan?

The SAk identifies some measures that have been taken during the design -

L phase for STS that should facilitate the decontamination of the system. These l design features are: the ability to remotely remove components from the HLW tank; the ability to remotely decontaminate some pieces of equipment by flushing; l

the minimization of areas or " pockets" which would collect solution or solids; and the use of materials whic.h can handle a wide range of decontamination solutions and also not incorporate contamination into vessel and piping surfaces.

Other design features which are incorporated are (1) the use of epoxy paint or liners to reduce concrete contamination, and (2) the use of sumps on either side of the back wall of the valve aisle to reduce the extent of contamination.

(3) the use of a drain for the floor of the shield structure (possible pluggage of this drain should be examined during preoperational startup testing). These features are expected to be useful to reduce the amount of decontamination that will be required. An aspect of the STS design which does not seem to facilitate decontamination and decommissioning is the unmonitored, undrained annulus asso-ciated with the double walled pipes in the STS pipeway and shield structure.

This presents the potential for developing an undetected accumulation of high-level waste if a leak occurs and could make decontamination and decommissioning more difficult. The staff recommends that DOE reduce the potential for such -

accumulations.

The specific procedures that will be used during STS operations to facilitate future decontamination and decommissioning are identified in a general manner in the SAR. The NRC staff is aware that WVNS personnel have extensive experience with decontamination. The NRC staff expects that DOE will include in the opera-ting procedures the requirement to promptly clean up spills, maintain records of contamination levels in the facility, and maintain accurate as-built drawings.

24 m _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - - - _ _ _ _ _ _ _ _ _ _ _ _ - - - _ __ - - _ - - - -

t i

5. OPERATIONAL SAFETY REQUIREMENTS  !

The Department and its contractor has provided operational safety i

requirements (OSR).for the purpose' of defining limits to design and operation

'for protection of the health and safety of the public and employees. These OSRs are comparable to those technical specifications the Commission invokes in the licensing process. The staff's review of the OSRs considered this comparability.

Chapter.11 of the'SAR identifies five OSRs that DOE will apply to.STS operations. The fi, y specifies limits for the ventilation system (pressure differentials,. equipment status, and surveillance requirements). The second j covers operation of the off gass treatment system, including air monitoring.

It requires testing and changeout of HEPA filters. The third OSR places ,

requirements on the backup and emergency power systems. The fourth OSR places limits on the fissile concentration of the STS feed supernatant and on the Cs-137 content of'the treated supernatant transferred from the STS to the .

liquid waste treatment system. The last OSR identifies the previous Nuclear Fuel Services technical specifications concerning (1) caustic concentration in i the HLW storage tanks, (2) spare tank capacity, and (3) pan specification.

The staff has reviewed these'0SRs and believes that, with the exception j of the OSR which limits Cs-137 concentration in the treated supernatant, these limits are appropriate administrative controls for the protection of the public health and safety. The staff does not believe that the Cs-137 concentration limit is as restrictive as it should be. This is discussed further-in Section 3.3 of.this SER, the staff's review of ALARA for the STS. I 1

l I

25

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, , j "_

' re

, APr'Df1!X ~ A REFEREsCE

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1 -, Dames and Moore. STS Confinement Barrier Integrity Review--West Valley  ;

Demonstration Project:. August 15, 1986.  !

2. 'USNRC. ' Safety Evaluation Related to the West Valley Demonstration

~

' Project--Principal Design Criteria and Management. Organization. April

-1987.

3. USNRC.~ .. Nuclear Regulatory Stdff Safety Evaluation Report-on 'the Dormant .j

,W est Villey Reprocessing Facility; January 1982. 1

'4 Bray, L~. A.,.et a1. Experimental Data Developed to Support the Selection of a Treatment ' Process for West Valley Alkaline Supernatant.

l

.PNL 4969. January.1984.

i

5. Federal _ Register, February 11, 1985; pages 5600-5625.

- 6. DDE, West Valley Demonstration Project Safety Analysis Report Volume III Supernatant Treatment System Rev. 1, July.1986.

7. USNRC. '"FIRAC User's Manual: A Computer Code to Simulate Fire Accidents in Nuclear Facilities," NUREG/CR-4561, LA-10678-M, prepared by D. B.

Nichols and W. S. Gregory, Los Alamos National. Laboratory, April 1986.

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