ML20209F414

Safety Evaluation Re West Valley Demonstration Project, Principal Design Criteria & Mgt Organization

ML20209F414
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Issue date:	04/30/1987
From:	NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
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ML20209F405	List: ML20209F399 ML20209F414
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REF-PROJ-M-32 NUDOCS 8704300283
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SAFETY EVALUATION RELATED TO THE WEST VALLEY DEMONSTRATION PROJECT

.j . PRINCIPAL DESIGN CRITERIA AND MANAGEMENT ORGANIZATION PROJECT M-32 U.S. NUCLEAR REGULATORY COMMISSION.

OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS APRIL 1987 i

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. 8704300283 PROJ MO427 PDR PDR .

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TABLE OF CONTENTS Page l' INTRODUCTION AND PURPOSE ........................................ 1-1 2 PROCESS OVERVIEW ................................................ 2-1 3 SITE CHARACTERISTICS . .......................................... 3-1 3.1 Seismicity ................................................. 3-2 3.2 Meteorology ................................................ 3-4 3.2.1 Tornado and High Winds .............................. 3-4 3.2.2 Tornado Missiles .................................... 3-6 3-7 3.2.3 Meteorological Dispersion ...........................

3.3 Hydrology .................................................. 3-11 4 SAFETY ISSUE IDENTIFICATION AND MANAGEMENT ................. .... 4-1 l 4.1 Hazard Identification ...................................... 4-2 4.2 Hazard Elimination or Midigation............................ 4-6 i 4.3 Independent Safety Review .................................. 4-7 5 RADIATION PROTECTION PROGRAM .................................... 5-1

, 5.1 Source Term Estimation ...... .............................. 5-1 5.2 Dose Assessment Methods .................................... 5-2 5.3 Release Point and Environmental Monitoring ................. 5-4 5.4 ALARA..................................................... 5-5 5.5 ' Documentation .............................................. 5-6 6 TRAINING .. ..... . . . ... .................. ................. 6-1 7 ACCIDENT ANALYSIS ............................................... 7-1 7.1 Engineering Confinement .................................... 7-1 7.2 Characteristic Upper Limit Accident ........................ 7-3 7.3 Emergency Planning ......................................... 7-3 8 QUALITY ASSURANCE ............................................... 8-1 l

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TABLE OF CONTENT (Continued)

LIST OF FIGURES l

Figure Page 2.1 Simplified Processing Scheme for West Valley High Level Waste.... 2-3 2.2 Existing and Future Facilities Associated with the West Valley Demonstration Project ........................................... 2-4 3.1 Estimates of Tornado and High Wind Probabilities (from Fujita 1981)..................................................... 3-5 4.1 WVNS Organizatio'n Chart ......................................... 4-8 LIST OF TABLES i 3.1 Comparison of Annual Average X/Q Values at Site Boundary by j Sector for Ground-Level Releases ................................ 3-9 3.2 Comparison of Annual Average X/Q Values at Site Boundary by

Sector for Elevated Releases .................................... 3-9 3.3 Comparison of Two-Hour X/Q Values at Site Boundary by Ground-Level Releases ................................................... 3-10 3.4 Comparison of Two-Hour X/Q Values at Site Boundary for Elevated Releases ................................................ 3-10 4.1 Identification of Major Activities for Developing, Reviewing, and Approving WVDP Systems and Listing Issues Reviewed by NRC. .... 4-3 i 4.2 Summary of Independent Check Made by WVNS ...... ................. 4-10 l

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1 INTRODUCTION AND PURPOSE The Department of Energy (the " Department or DOE") is authorized by Public Law 96-368, the West Valley Demonstration Project Act (the "Act"), to carry out a demonstration project (the " Project") at the Western New York Nuclear Service Center (the " Center" or " site") in West Valley, New York.

Specifically, the Act calls for DOE to (1) solidify, in a form suitable for transportation and disposal, the high-level radioactive waste at the Center by vitrification or other technology; (2) develop containers suitable for the permanent disposal of the high-level waste solidified at the Center; (3) as soon as feasible, transport, in accordance with applicable provisions of law, the waste solidified at the Center to an appropriate Federal repository for permanent disposal; (4) in accordance with applicable licensing requirements, dispose of low-level radioactive waste and transuranic waste produced by the solidification of the high-level waste; and (5) decontaminate and decommission, in accordance with such requirements as the U.S. Nuclear Regulatory Commission (the " Commission") may prescribe, (a) the tanks and other facilities at the Center in which the high-level radioactive waste solidified under the Project was stored, (b) the facilities used in the solidification of the waste, and (c) any mate-ial and hardware used in connection with the Project.

The Department of Energy assumed control of the West Valley site in February of 1982 with West Valley Nuclear Services Company (WVNS), a subsidiary of Westing-house Electric Corporation, as their operating contractor. The Center was previously used primarily as a licensed facility for the reprocessing of spent nuclear fuel from nuclear power plants. It was those reprocessing activities which resulted in the accumulation of liquid high-level wastes stored in under-ground tanks at the Center.

The Department entered into an agreement with the Commission whereby the Commis-sion would review and consult with the Department on the Project with respect to the health and safety of the public. The agreement, in keeping with the 1-1

Act, specifies that the review and consultation process shall be conducted in an informal manner and shall not involve formal procedures or actions by the Commission pursuant to the Atomic Energy Act, the Energy Reorganization Act or any other law. The agreement or Memorandum of Understanding (MOU) between the Department and the Commission was executed on September 22, 1981 (DOE 1981).

The MOV calls out five specific provisions. These five provisions state that the Department will: (1) prepare a Project Plan which addresses DOE's approach for conduct of the Project, describing what is going to be done and how it is to be accomplished. The Commission is required to review and comment on the Plan, (2) consult with the Commission with respect to waste form and waste container selection, (3) prepare a Safety Analysis Report (SAR) on Project activities and submit this to the Commission for independent review, (4) pre-pare an analysis of different decontamination and decommissioning modes and a Project decommissioning plan which is to be reviewed by the Commission, and, (5) afford the Commission access to the site in order to monitor activities with a potential for impact on public health and safety.

The Department issued the Project Plan described in (1) above on December 12, 1984, following consultation with the Commission (DOE 1984). The Project Plan was prepared to describe how the public health and safety will be assured while complying fully with the Act. Prior to issuing the Project Plan, the Department selected borosilicate glass as an appropriate waste form for solidifying West Valley high-level waste (DOE 1983a). In making this decision, the Department relied, in part, on the fact that borosilicate glass had been previously selec-ted for solidifying the similar high-level wastes at the Savannah River Plant.

On October 28, 1983, the Department further decided to proceed with the slurry-fed ceramic melter as the process technique to produce the borosilicate glass waste form (D0E 1983b). Although adopted to specific wastes at West Valley, the Department again relied to a large extent on the development of the process at Savannah River. ^

The Project Plan discussed a general processing strategy which the Department intends to follow, i.e., to minimize the amount of nonradioactive waste constit-uents being incorporated in the solidified high-level waste. This is to be 1-2

accomplished by ion exchange treatment of the high-level waste supernate and washing of the high-level waste sludge. The highly radioactive streams from these processing operations would be the portions that are homogenized and vitrified in the Joule-heated ceramic melter. The remaining lower radioactive streams would be treated and disposed of in other forms.

The Plan also described the Project's approach to identifying and accommodating processing hazards. The basic approach is to (1) analyze proposed systems to estimate the consequences of failure; (2) based on the estimated consequences, ,

identify the appropriate safety category with its attendant minimum require-ments for codes, standards and quality assurance level; (3) establish actual j design criteria for an individual project and subject this to internal and j external review; (4) prepare a safety analysis report which examines all aspects of the operation.

The Department plans to submit the SAR required by the MOV in modular units I

because different parts of the project will be undergoing design, construction, and operation at different times, and because DOE and their operating subcon-tractor want to avoid a situation where SAR preparation and review become the

limiting factor. (WUNS 1984) Therefore, a series of SAR modules have been identified. These modules are

i SAR MODULES Volume Title Part I Project Overview and General Information A II Existing Plant and Operations B III High-Level Waste Vitrification o Vitrification Facility C o Supernatant Treatment System D o Sludge Mobilization System E IV Waste Management, Storage and Disposal o Vitrified High-Level Waste Storage F o Cement Solidification G o Liquid Waste Treatment System Upgrades H 1-3

o Size Reduction Facility I o Lag Storage Facility J o Disposal Area Operations K V Final Decontamination, Decommissioning and Waste Shipment L VI Operational Safety Requirements M (Technical Specifications)

Since the SARs are being prepared and submitted to the Commission in a modular fashion, and the Commission's staff (the staff or NRC) is reviewing the Project and consulting with DOE on a routine basis, the staff has concluded that the most appropriate way to document its review and evaluation of.the Project's SARs is to prepare several Safety Evaluations (SER) reflecting review of one or more modules. This_ report presents our evaluation of information in the Project Overview SAR module (Volume I) as-well as other information supplied by the Project to NRC, Subsequent reports will evaluate the protection of the health and safety of the public relative to that particular operation.

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2 PROCESS OVERVIEW The West Valley Demonstration Project will process and solidify the high-level waste presently stored in underground tanks to produce a borosilicate glass waste form acceptable for final disposal in a repository. A byproduct of the solidification process will be a low-level waste cement form, also suitable'for final disposal.

The high-level radioactive wastes are contained in two underground tanks at the i Center. By far the largest quantity of high-level wastes (approximately 550,000 gallons) is stored as a basic solution (pHN10) in a carbon steel tank, desig-i nated as Tank 80-2. Although most of the waste volume in Tank 80-2 is in liquid l form, a solid phase (sludge), which rests on the tank bottom, was formed when i

the acid wastes from reprocessing operations were neutralized with excess sodium hydroxide. Tank 80-1 is an identical tank. It was maintained as a spare storage volume to receive the waste if a significant leak had occurred in Tank 8D-2.

Both tanks rest in their individual carbon steel pans and are ccatained within separate, reinforced concrete vaults. There have been no leaks 67 either tank.

The spare', Tank 80-1, receives water condensate from the cooling of water evapo-rated from Tank 80-2. This water is slightly contaminated and is occasionally treated for discharge. The tanks are 70 feet in diameter and 27 feet high.

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The vaults are covered with approximately 8 feet of earth overburden. For a discussion of the NRC staff's safety evaluation related to current storage, refer to the SER for the dormant West Valley Reprocessing Facility (NRC 1982).

A much smaller quantity (12,000 gallons) of high-level waste is contained in acid solution in Tank 8D-4. Since this waste was not neutralized, no signifi-

, cant sludges are present in Tank 80-4. An idec.tical spare, Tank 80-3, serves the same purpose as Tank 80-1, as described above. Both of these tanks are made of austenitic, stainless steel and are contained within a common, reinforced concrete vault.

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The sequence _of-activities which the Department is planning for the project are:

(1) The 8D-2 supernate will be processed through ion exchange columns containing an inorganic zeolite exchange medium.to remove cesium from the supernate. The separation of cesium from the nonradioactive salts in the waste solution will enable a more concentrated high-level waste form. The ion exchange columns will be installed and operated within the existing spare high level waste tank (80-1).

When each column batch of zeolite is nearly saturated with cesium, it will be discharged directly into the-tank. The supernate is to be treated further to meet the requirements for low level waste and will be solidified in a cement matrix for disposal.

(2) After the supernate has been treated for cesium removal, .the sludge in 8D-2 will be washed to remove soluble salts. This wash solution will be processed through the ion exchange columns in 8D-1 for cesium removal. The treated wash solution will also be solidified as low-level waste in a cement waste form.

(3) Af ter sludge washing is complete, (1) the loaded zeolite in tank 80-1, (2) the washed sludge in tank 80-2, and (3) the acid waste in tank 80-4 will be combined and fed to the melter in the vitrification facility.

A simplified schematic of these processing steps is shown in Figure 2.1.

A plan view of the site with its existing and future facilities is_ presented in Figure 2.2. The figure shows the location of existing facilities which will be utilized or modified and new facilities which will be constructed to support Project activities.

l Most gaseous discharges including melter off gas and the vessel vent system in I the vitrification facility will be from the main plant stack, although several areas will have separate discharges for cell or room ventilation systems. The 2-2 l

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major routine source of release of radionuclides to the atmosphere will be the

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vitrification off gas system. The NRC wi.ll be examining the. performance of-this and other off gas 'and ventilation systems. All liquid wastes from various plant operations will be processed through a liquid waste treatment system which will also be reviewed by the NRC.

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4 3 SITE CHARACTERISTICS The Center has been the subject of numerous site investigations, analyses, and

- evaluations since its inception in the early 1960's. Should the adverse effects caused by the extremes of natural phenomena on the facility (such as an earthquake or a tornado) not be . mitigated, the public health and safety could be influenced through surface water, ground water, or air transport of released radioactive matcrials. Site investigations relative to these potential trans-port mechanisms have been conducted by Nuclear Fuel Services, Inc. (NFS), the original site developer; the U.S. Geological Survey; the New York State j Geological Survey and its contractors; and the Department and its contractors.

, The NRC (or its predecessor, the Atomic Energy Commission, AEC) has evaluated the site and its associated design basis natural phenomena. The early site investigations were discussed in safety reports issued prior to construction (AEC 1963) and operation (AEC 1965) of the reprocessing plant.

Later, additional evaluations of site phenomena were performed by the staff and documented in reports on the dormant facility. (NRC 1977, NRC 1982). These documents presented an assessment of the natural. phenomena which must be con-sidered in the design of a nuclear fuel reprocessing facility at the West i

Valley site, as well as the meteorology and hydrology which should be considered for analysis of transport of radionuclides into the off-site environment.

Some of the site phenomena have been further investigated since the-NRC per--

formed its last assessment. In addition, new design basis-phenomena have been

[ proposed by the Department because of differences between the waste solidifi-cation activities of the West Valley project and the previous spent fuel reprocessing activities.

In this section we describe and evaluate current information on pertinent site characteristics. We also discuss the relationship of current activities to evaluations of previous activities.

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N 3.1 Seismicity The Center is located in seismic risk Zone 3, as designated by the U.S. Geolog-ical Survey. The original buildings at the Center were designed in the early 1960's to meet Uniform Building Code Zone 3 seismic requirements. In 1973 NFS, the site operator and a co-licensee with the State of New York for the repro-cessing facilities, applied for a permanent operating license under 10 CFR Part 50,

" Domestic Licensing of Production and Utilization Facilities," that included a construction permit to modify the facility by increasing the reprocessing capacity and improving certain safety aspects. In connection with that appli-cation, the Commission concluded that an appropriate acceleration value for anchoring a seismic design response spectrum (zero period) was 20 percent of the acceleration due to gravity (0.2g). This value was related to the proce-dures set forth in Appendix A to 10 CFR Part 100, " Reactor Site Criteria,"

i.e., the value was related to methods used for siting nuclear power plants.

Prior to the license application discussed above, the NRC staff had neither published nor developed techniques for determining design basis seismic inputs for fuel cycle facilities. ,

NFS did not implement its plans for modification of the reprocessing pla,nt complex, so the 0.2 g design criteria was not specifically identified with any new structures.

The NRC independently analyzed the effect of an earthquake at various accelerations up to and including the 0.2 g value for the existing reprocess-ing facility structures at West Valley. These analyses were discussed in the 1982 safety evaluation of the dormant reprocessing facility. The analyses ,

indicated that in the event of an earthquake:

1. Some degree of cracking in the pool area might be expected and some -

seepage of pool water might occur, but there would be no risk to the public because of the impermeability of the soil.

2. Some failure of the General Purpose Cell might occur at seismic levels as low as 0.1 g, but other portions of the processing plant would not 3-2

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fail _until seismic levels were higher than 0.2 g. If such partial j failures were to occur, the staff concluded it was unlikely for radio-active material to be released,and dispersed.

3. Some cracking and concrete spalling in the high level waste tank vault could be expected for earthquakes up to 0.2 g, but no gross failure is expected of either the tank or vault.

The Department recognizes the historical background of 0.2 g site design basis earthquakes for nuclear fuel reprocessing plants. In Section A.3.6-E of the supplement to'the Project Overview andsGeneral Information SAR module, the DOE s states its belief that waste processing and solidification operations, partic-ularly the ones being planned for the West Valley Demonstration Project, present

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- a lower potential for offsite risks than a nuclear reactor because of differences in operating lifetimes and potential driving forces.

The Department has proposed the un cf a lower design basis seismic input of 0.1 g. The Department selected this value because the operating life of the processing plant is expected to be about one and one-half years as compared to a conventional nuclear power plant lifetime of 40 years. This reduces the prob-ability of occurrence of a natural phenomena _during the operational phase of the facility by more than an order of magnitude. Another factor in DOE's selection was the low energy nature of the solidification process itself with aged nuclear waste in which there is a limited potential for the dispersal of radioactive material.

The staff has reviewed the icwer design basis criteria and finds it to be rea-sonable for this purpose. Staff analysis shows that the consequences associ-ated with potential accidents during the processing of West Valley waste are less than the consequences associated with potential accidents in a nuclear

. fuel re'ph ocessing plant. The NRC staff does believe that DOE should, prior to startup'of any new operations, provide analyses which shows that structural and confinement system failure will not occur for a 0.1 g seismic event. In addi-tion, DOE should document that there are adequate safety margins beyond a 0.1 g earthquake before significant failure occurs.which may lead to release of material. Further analyses should be undertyen by the Department for selected l

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structures and/or equipment related to confining radioactive materials in order to demonstrate a structural capacity to function at acceleration inputs up to and-including 0.2 g.

3.2 Meteorology- 3 3.2.1 Tornado and High Winds During its 1974 review of the NFS application to modify the facility, the NRC indicated that an appropriate design basis tornado for the West Valley site would include a maximum wind speed of 325 mph. (NRC 1977). A 1981 evaluation based on analyses by Fujita (Fujita 1979) indicated that the recurrence inter-val for a 325 mph tornado at West Valley might be one billion years or greater.

Based on this updated information, the NRC established a new design basis tor-nado which had maximum wind speeds of 200 mph and a recurrence interval of about two million years. Fujita conducted further analyses using a larger, more accurate data base which again revised the recurrence interval for the various winds (Fujita 1981). Fujita's new estimate of the recurrence interval for a tornado with maximum wind speeds of 200 mph is five million years.

The Department has proposed a design basis tornado with a maximum wind speed of 160 mph. In the Project Overview and General Information SAR module, the I

Department cites the latest data of Fujita, but does not present a logic for selecting the 160 mph tornado from spectrum of tornados presented in Fujita's report. The Department does not discuss design basis straight line winds. The NRC has reviewed and accepts the new data and analysis presented by Fujita in his latest report. The design basis tornado established by the NRC in 1981 had a recurrence interval of two million years. Figure 3.1 (the summary figure from Fujita's report) shows that the new estimate for a tornado with a two million year recurrence interval has a maximum wind speed of 178 mph. A tornado with a maximum wind speed 158 mph has a recurrence interval of about one million years. The 160 mph tornado is therefore expected to have a recur-i rence interval of slightly over one million years. Fujita's latest analysis also shows that wind speeds of less than 165 mph have a higher probability.of being associated with straight line winds than with a tornado.

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Figure 3.1 Estimates of Tornado and High Wind Probabilities (from Fujita 1981) j 10-1 -

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~The staff concludes that the 160 mph tornado is acceptable as the design basis tornado. While we agree with this value for a tornado, the analysis indicates

.that straight line (non-tornadic) winds of the same frequency would be 165 mph.

The staff recommends that wind speeds greater than design basis value be used l

.to. analyze safety margins for structures and systems having potential impor-tance to public health and safety.

Associated with its 160 mph tornado, DOE has proposed a design basis maximum

, pressure drop of 0.35 psi and a maximum pressure drop rate of 0.15 psi /sec.

The staff developed its own estimates of maximum pressure drop and maximum pressure drop rate for a 160 mph tornado using correlations presented by 2

MacDonald (MacDonald 1981) and Fujita (Fujita 1978). The MacDonald corre-i lation predicted a maximum pressure drop of 0.35 psi and'a maximum pressure-

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drop rate of 0.08 psi /sec. These values are similar to the values proposed by DOE for the project. The Fujita correlation, however, projected a maximum 4 pressure drop of 0.41 psi and a maximum rate of over 0.2 psi /sec, which are 1 more severe than the values proposed by DOE for the project.

The staff recommends that the structures and systems which confine high-level j

waste should be analyzed for the more severe pressure drop and pressure drop rate developed from the Fujita correlation.

3.2.2 Tornado Missiles The Project proposes to use two design basis missiles discussed by MacDonald.

(MacDonald 1981) These missiles are:

(1) Timber plank, 10 cm x .3 m x 3.7 m, 63 kg, at a velocity of 144 kph (89.5 mph).

(2) Steel pipe, 8 cm diameter and 3m long, 34.4 kg, at a velocity of 104 kph (64.6 mph).

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-In considering potential tornado missiles at the West Valley site, the NRC staff reviewed the kinds of material expected to be in the' vicinity of the site that could become tornado missiles. Because of outgoing site construction, there are planks and pipes, steel rods, wooden poles and automobiles which are potential missiles of the type described in the NRC Standard Review Plan (NRC 1981). While many missiles types are considered to be credible, it is expected that one or two of the types will be the most severe, and therefore, more apt to influence the design of specific project structures. As the specific pro-ject structures are designed, the staff will evaluate the missile types used in the design analysis for those structures that provide a confinement for radio-active materials which could endanger the health and safety of the public.

3.2.3 Meteorological Dispersion As airborne radionuclides are transported to offsite areas, the concentration will decrease because of atmospheric dispersion. Atmospheric dispersion fac-tors (X/Q) were calculated by the Department and'used in predicting consequences of offsite dose. The dispersion factors are based on one year of onsite data collected during October 1983 through September 1984. (A discussion of the data collected and its analysis is presented in Volume I (Section A.3.3.3), in the Supplemental Information Volume (Sections A.3.3-A, A.3.3-B, and A.3.3-C),

and in a report by WVNS (WVNS 1985). The staff reviewed the Department's esti-mates of atmospheric dispersion.

This review addressed two topics: the placement or location of modeled recep-tor points with respect to the site exclusion zone boundary, and the suitability of the calculated X/Q values. The report by Yuan presents 16 points (receptor points) for which X/Q values are calculated. These points are located at the sixteen compass directions and various distances which represent the site bound-ary. The receptor point locations were plotted and their locations found to be very near the actual site boundary.

The calculated X/Q values were reviewed utilizing a different methodology than that used by the Department. It utilized a line sector averaged Gaussian plume 3-7

4 model, COMPLEX I, which accepts one year of hourly meteorological data and in-corporates terrain adjustment factors which are necessary to adequately repre-sent the wind flow in this region of fairly complex terrain. The Department utilized EPM 3, a variable-trajectory Gaussian puff dispersion model, to cal-culate annual average X/Q values along_with PAVAN, a straight-line Gaussian plume model, which accepts onsite meteorological data in the form of a joint frequency distribution of wind speed, direction and atmospheric stability in its analysis of short-term impacts.

Annual average X/Q values predicted for the site boundary are presented in

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Tables 3.1 and 3.2 for postulated ground-level and elevated release cases,

respectively. The NRC's values in the tables were calculated using COMPLEX I and then adjusted to account for depletion according to the methodology des-cribed in Regulatory Guide 1.111. In Table 3.1 the comparison of X/Q values calculated for ground-level releases shows very similar results. The X/Q l values we calculate are, on the average, 55% higher than those calculated by the Department. Table 3.2 presents the comparison of X/Q values for elevated (60 meter) releases. Again the NRC-calculated X/Q values are slightly higher l (an average of 47%) than those calculated by the Department. A comparison of

) NRC-calculated X/Q values for other distances supports the conclusions drawn from the information presented in Tables 3.1 and 3.2. NRC-calculated X/Q are slightly higher than the Department values. At short distances (850 meters)

NRC-calculated X/Q values average 12% higher for elevated releases and 29*4 higher for ground-level releases. At greater distances (2400 meters), our X/Q values average 1.7 times those presented by the Department for elevated releases, and 2.6 times those presented for ground level releases, i

The hourly X/Q values predicted by COMPLEX I were statistically analyzed to determine the 0.5*4 maximum sector and 5*; overall X/Q values according to Regu-latory Guide 1.145. The greater of these, the 0.5% maximum sector X/Q values which are expected to be exceeded 0.5's of the total time, were selected as  !

I being applicable for accident evaluations and are presented in Tables 3.3 and 3.4. Tables 3.3 and 3.4 show that the assumptions used with the PAVAN model yield more conservative results (i.e., the Department predicted X/Q values are 1

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Table 3.1 Comparison of Annual Average X/Q Values at Site Boundary by Sector for Ground-Level Releases DIR DISTANCE (M) NRC/ COMPLEX I X/Q DOE / EPM 3 X/Q (sec/m') (sec/m')

S 2200 1.65E-07 1.55E-07 SSW 2000 1.04E-07 1.34E-07 SW 1850 1.76E-07 1.25E-07 WSW 1400 1.69E-07 1.34E-07 W 1300 2.85E-07 3.43E-07 WNW 1050 7.02E-07 8.47E-07 NW 1050 1.47E-06 1.40E-06 NNW 1600 2.17E-06 8.33E-07 N 2350 1.33E-06 5.14E-07 NNE 1850 1.08E-06 -5.75E-07 NE 1700 8.54E-07 5.62E-07 ENE 1900 4.75E-07. 3.04E-07 E 2550 3.48E-07 1.68E-07 ESE 2700 3.70E-07 2.22E-07 SE 2850 4.37E-07 2.11E-07 SSE 2500 2.97E-07 1.85E-07 7

Table 3.2 Comparison of Annual Average X/Q Values at Site Boundary by Sector for Elevated Releases DIR DISTANCE (M) NRC/ COMPLEX I X/Q DOE / EPM 3 X/Q (sec/m') (sec/m')

1 S 2200 5.88E-08 1.09E-07 SSW 2000 6.86E-08 1.38E-07 SW 1850 1.18E-07 2.25E-07.

WSW 1400 1.48E-07 6.45E-08 W 1300 5.94E-08 3.05E-08 WNW 1050 2.00E-08 1.73E-08 NW 1050 2.00E-08 1.52E-08 NNW 1600 3.92E-08 1.17E-08 N 2350 2.94E-08 1.99E-08 )

NNE 1850 1.96E-08 1.63E-08 NE 1700 3.92E-08 2.97E-08 ENE 1900 4.90E-08 4.57E-08 E 2550 6.86E-08 5.60E-08 ESE 2700 8.82E-08 6.29E-08 SE 2850 5.82E-08 4.10E-08 SSE 2500 1.47E-07 5.18E-08 3-9

Table 3.3 Comparison of Two-Hour X/Q Values at' Site Boundary for Ground-Level Releases DIR DISTANCE (M) NRC/ COMPLEX I X/Q DOE / PAVAN X/Q (sec/m') (sec/m')

S 2200 1.22E-05 1.93E-05 SSW 2000 1.08E-05 1.52E-05

. SW 1850 1.45E-05 -2.24E-05 4

WSW 1400 1.38E-05 2.21E-05 '

il 1300 2.19E-05 3.22E-05

, WNW 1050 5.46E-05 6.10E-05 NW 1050 6.31E-05 1.28E-04 NNW 1600 3.08E-05 6.48E-04 N 2350 2.28E-05 7.07E-04 NNE 1850 2.93E-05 -9.77E-05 NE 1700 3.32E-05 7.53E-05 ENE 1900 3.63E-05 5.20E-05 E 2550 2.17E-05 -3.37E-05 ESE 2700 1.70E-05 2.87E-05 SE 2850 1.38E-05 2.46E-05 SSE 2500 1.46E-05 2.11E-05 Table 3.4 Comparison of Two-Hour X/Q Values at Site Boundary for Elevated Releases DIR DISTANCE (M) NRC/ COMPLEX I X/Q DOE / PAVAN X/Q (sec/m') (sec/m')

S 2200 8.34E-06 3.41E-05 SSW 2000 4.05E-06 2.47E-05 SW 1850 4.70E-06 4.07E-05 WSW 1400 4.25E-06 5.61E-05 i

W 1300 1.02E-05 5.68E-05 WNW 1050 1.64E-06 4.00E-05 NW 1050 1.19E-06 3.67E-05

NNW 1600 8.83E-07 1.85E-05  ;

N 2350 9.30E-07 1.19E-05 l NNE 1850 1.02E-06 1.26E-05 I NE 1700 1.79E-06 6.72E-05 l ENE 1900 1.41E-06 5.02E-05 E' 2550 2.56E-06 3.63E-05

ESE 2700 1.17E-06 2.00E-05 SE 2850 1.08E-06 1.48E-05 SSE 2500 8.28E-06 2.91E-05 4,

3-10 l

consistently greater than those calculated by NRC). For these short-term acci-dent scenarios, Department values are consistently greater and therefore more conservative than NRC values by a factor of 1.5 to more than an order of magni-tude.

Aside from examination of the magnitude of predicted X/Q values, the locations of maximum X/Q values were also compared. For the annual case, both analyses predict maximum X/Q values to occur in -the same general direction (NW or NNW) for the ground-level release, and in the WSW or SW for the elevated release.

During a 2-hour release, both models predict maximum X/Q values to occur in the NNW-NW sectors for the ground-level release. For the short-term elevated re-lease, PAVAN predicts the maximum to occur in the NE with a secondary maximum in the W to WSW. COMPLEX I did not predict the NE maximum for this case, but did predict the maximum to occur in the W sector.

While the differences in magnitude predicted by these models may at times seem substantial, such variation is not uncommon in modeling analyses and is within the uncertainty of the analytical method. Likewise, minor differences in maxi-mum locations can be attributed to the slight differences in receptor point locations and in the models' differing treatment of terrain. The Department-calculated X/Q values are found to be suitable for use in evaluations of bounding impacts. Dose calculations based on these dispersion factors will be acceptable.

3.3 Hydrology The geohydrological characteristics of the West Valley site have been studied for several years. Some of the studies have been related to plant construction, but many more have been related to the disposal of low-level radioactive wastes at the site. A particularly useful report was prepared by the New York State Geological Survey for the NRC (NYSGS 1983).

The site's surface hydrology was first characterized by Nuclear Fuel Services, Inc., in the early 1960s. The site is drained by Quarry Creek and Frank's Creek 3-11 i

I l

1 to Buttermilk Creek to Cattaraugus Creek and then to Lake Erie. The characteri-zation of surface runoff is considered acceptable. There is no evidence that

, runoff from current operations is causing problems. Surface _ discharges are projected to decrease in the future.

The subsurface hydrology of the site is strongly influenced by the glacial deposits which rest on shale bedrock. The most prominent feature is t.'e

, clay-silt deposit identified as Lavery Till. The Lavery formation is a very hydraulically tight clay and considered by some to be virtually impermeable to water flow, if undisturbed. In the disposal area, the stratigraphic horizon between the weathered and unweathered Lavery Till lies at a depth of about ten feet from the surface. The permeability of the weathered clay is significantly greater than the underlying Till. In other site areas, alluvial deposits of sands and gravels also result in higher permeabilities than the Lavery. Direct-ly beneath the Lavery is a Lacustrine unit, which appears to be a more permeable

formation, and below the Lacustrine is another tight clay, the Kent Till.

I In the low-level waste disposal area, the surface strata (the top 3 meters) is a brownish, weathered till which can usually be distinguished from the Lavery formation which is a dark grey. The hydrologic characteristics of these mate-l rials has been recognized for its importance to the safe confinement of the

radioactive wastes already disposed of in holes or trenches. The migration experience in the disposal areas shows that some subsurface flow can occur and that the flow is nonuniform. The hydraulic behavior of these flows is currently being studied.

l

(

i 3-12 4

e

I 4 SAFETY ISSUE IDENTIFICATION AND MANAGEMENT As discussed in Chapter 1, the physical systems for waste processing are in various stages of design and construction, therefore an integrated review of 4

design details of the systems is not feasible at this time. The staff has re-viewed the project management systems to. determine the methods whereby the

. protection of public health and safety are considered. Specifically, we have determined if the management system is likely to identify potential public health and safety hazards, eliminate or reduce the hazards, and provide inde-pendent checks of the identification and mitigating efforts. This chapter presents our conclusions in this matter.

The public health and safety aspects of the Project are conducted according to the requirements of DOE orders. The staff is aware of these orders and the fact that DOE-Idaho periodically inspects the operations at West Valley to assess Project compliance with DOE orders. The staff has and expects to con-tinue review of the independent assessment reports from these DOE inspections to help carry out the monitoring of public health and safety at the site.

The Project Overview and General Information SAR Module (Volume I) does not provide a detailed discussion of the management system for identifying and then eliminating or mitigating potential hazards to the public health. The SAR module does, however, identify the WVNS Policies and Procedures Manual as a document which describes the management system addressing these and other issues.

, The staff has reviewed this Policies and Procedures document in our evaluation of the Project's program.for protection of public health and safety.

The WVNS Policies and Procedures manual describes how WVNS performs its various tasks at the West Valley site. There are procedures which address administra-tion, engineering, personnel relations, purchasing, services, accounting, and safety. Because this specific review focuses on public health and safety, only an applicable group of WVNS policies and procedures were examined. The policies f

2 4-1

l l

1 and procedures which were selected are based on an understanding of the major WVNS steps associated with developing, reviewing, and approving a design, pre-paring for startup; and controlling operations.

These major steps associated with the development, review and approval of' pro-ject facilities are identified in Table 4.1. The table lists the activities from development of functional and operating requirements through operational readiness review. It identifies the issues which are associated with each of the activities that were reviewed by the staff. It also identifies the speci-fic WVNS Policies and Procedures which address these activities and were exam-ined in this review. In general, the staff reviewed the WVNS policies to verify that they establish a management system which is likely to identify hazards, properly evaluate actions for the elimination or mitigation of hazards, and provides for independent checks of hazard identification and elimination or mitigation.

It is important that potential hazards associated with the operational phase of the WV0P be identified and then eliminated or mitigated as well. In addition to the Policies and Procedures identified in Table 4.1, two others were identified for review in order to evaluate the management system during the operational phase.

These were WV-195, Rev 2 " Configuration Control" WV-222, and Rev 0 " Trouble Record Reporting System."

The following sections present the staff's review of WVNS's Procedures and

, Policies concerning hazard identification, hazard elimination or reduction, and

'. independent safety review.

4 4

4.1 Hazard Identification In the WVNS system, hazard identification can occur at several stages. Pre-design hazard identification occurs when the inherent hazards associated with process materials are identified. The predesign hazard identification should 4-2 J

s

' Table 4.1 Identification of Major Activities for Developing, Reviewing, and Approving WVDP Systems and Listing of Issues Reviewed by NRC Issues Being WVNS Policy WVNS Activity Reviewed by NRC~ and Procedure Development of o identification of public health and WV-362, rev 2, Functional and safety hazards for the system 104, rev 2

! Operational o independent check of hazard Requirements identification Development of o identification of hazards which the WV-104, rev 2 Design Criteria design must address

. o independent check of hazard i

identification

Approval of o independent check that the design WV-363, rev 1, j Designs and addresses identified hazards. 104, rev 2 Equipment o independent check that no new hazards

.l- Purchases are introduced by the design i Review and o independent check of hazards WV-906, rev 3 i Approval of identification and mitigatior.

]

Safety Analyses Preparation of o identification of necessary operational WV-365, rev 1 Operational controls ,

Safety o independent check of identification Requirements of operational controls i Conduct o independent review of hazards WV-368, rev 0 t Operational identification and management for i

Readiness the entire system

, Review l

1 1

1 l

)

J 4

4-3 i

occur in the preparation of the WVNS Functional and Operating Requirements and the Design Criteria Documents mentioned in Table 4.1. Postdesign hazard identi-fication is the identification of hazards which could be associated with the design. This should occur during the design review and approval stage. Next, there should be a preoperation identification of hazards which checks the results of the previous two stages of hazard identifications. This should occur during the operational readiness review. Finally, the operational phase hazard identification notes the need for corrective action of any hazards not previously identified. These paragraphs summarize the findings of the staff's review of selected WVNS Policies and Procedures.

WV-362, Rev 2 discusses how WVNS establishes Functional and Operating Require-ments (F/0R) for selected systems, major equipment acquisitions, and signifi-cant construction or operational tasks. The F/OR identifies the need and requirements for the project and establishes an order of magnitude cost esti-mate. The procedure presents a guideline for preparing the F/ ors which contain a safety requirements section. On this basis, the staff concludes that the procedure does support the identification of public health and safety hazards during the preparation of Functional and Operating Requirements.

WV-363, Rev 1 is the WVNS procedure which specifies the contents of design cri-teria documents and the responsibilities of WVNS personnel for preparing, reviewing, and approving these documents. The procedure specifically calls for the design criteria to have a safety requirements section which addresses the protection of the health and welfare of both site personnel and the public.

The procedure only briefly describes the contents of this section, but the staff concludes that the procedure does support the identification of public health and safety hazard during the preparation of design criteria documents.

Design Reviews are discussed in WV-360, Rev 3 " Design Review Procedure." The procedure details responsibility for various phases of the design review and provides sample design review checklists to follow, but no individual or organi-zation is identified as being responsible for addressing safety during the 1

l 4-4

design reviews. The sample checklists address a multitude of topics but only j appear to address safety in passing. There is no WV procedure which addresses design approval, but there are procedures for approving drawings and equipment specifications which are the products of the design effort. WV-305, Rev 1,

" Preparation of Drawings" states, by reference to WV-104, Rev 2, that drawings are reviewed and approved by the Radiation and Safety Department. WV-306, Rev 1,

" Equipment Specifications" also states by reference to procedure WV-104 that the equipment specifications are reviewed and approved by the Radiation and Safety Department. While the design review procedure does not specifically call for a safety review, the staff concludes that the system promotes identi-fication of hazards at the design review stage because the Radiation and Safety Department reviews and approves the results of the design effort, specifically the drawings and equipment specifications.

WV-906, Rev 3 describes the WVNS Safety Review Program. The procedure states that new activities of the type that the NRC is concerned about at West Valley will be analyzed and reviewed prior to startup. The cognizant engineer is responsible for preparation of the safety analysis and the appropriate Opera-tional Safety Requirements. The fact that the procedure calls for the prepara-tion of safety analyses and operational safety requirements means that some examination of safety hazards will be required. The staff concludes that the WVNS procedure on safety review programs promotes the identification of hazards prior to startup of new or modified operations.

If potential safety problems are identified during the operational phase, WVNS has mechanisms for focusing on and resolving the safety issue. Procedure WV-222, Rev 0 " Trouble Record Reporting System" presents the steps for identi-fying and reporting system deficiencies. The procedure states that facility changes can be initiated using an Engineering Change Notice but details such as how this occurs and what public health and safety reviews are conducted are not presented. The workings of the Engineering Change Notice System will have to be examined by the NRC. Procedure WV-195, Rev 2 " Configuration Control" de-scribes the WVNS management system for controlling the design baseline. Ac-cording to this procedure the cognizant engineer is responsible for preparing 4-5

the Design / Field Change (DFC) package and identifying the type of review re-quired. He or she must specifically decide whether the Radiation and Safety Committee and the Radiation and Safety Department must concur on the DFC pack-age. The NRC believes that this DFC procedure provides a mechanism for

' identifying hazards associated with facility modification after the initial design and possibly construction is complete.

4.2 Hazard Elimination or Mitigation After hazards are identified, they should be managed to either eliminate or reduce the identified hazards. Hazard management will occur in the design phase as physical controls are incorporated into the design. Hazards are also

managed administratively as procedures are written to control operations. WVNS j procedures WV-305, Rev 1, " Preparation of Drawings" and WV-306, Rev 1, " Equip-ment Specifications" were reviewed to determine if the WVNS management system promoted hazard management in the design stage. Both of the procedures, by j reference to WV-104, Rev 2, call for the WVNS Radiation and Safety Department i to review and approve the results of the design efforts, the drawings and the equipment specifications. The staff concludes that requiring the Radiation and Safety Department to review and approve drawings and specifications promotes the management of hazards during the design phase.

WV-365, Rev 1 discusses the preparation, review and approval of operational safety requirements (OSRs). The OSRs are based on results of the Safety Analysis. They are incorporated into procedures and constitute major adminis-trative controls. The staff concludes that WV-365, Rev 1 promotes the manage-ment of public health and safety hazards through the preparation of these administrative controls.

Facility modifications made after the initial baseline design has been established are covered by either an Engineering Change Notice (WV-222, Rev 0) or a Design / Field Change (WV-195, Rev 2). The procedure for handling Engineer-ing Change Notices is unknown and will have to be examined. The procedure for l handling the Design / Field Change, as discussed in WV-195, Rev. 2, requires the 4-6 i

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

i cognizant engineer to decide if the Radiation and Safety Committee or the Radiation and Safety Department must concur with the proposed action. The staff finds that either of these organizations are capable of reviewing the package for its impact on public health and safety.

4.3 Independent Safety Review While the WVNS Policy and Procedures system does provide opportunities to-identify and then eliminate or reduce hazards, the staff also reviewed the policies and procedures to see if there were independent reviews of the hazards identification and subsequent hazard elimination or mitigation.

The engineering efforts to develop and build the West Valley processing facil-ities is performed by the WVNS Technical Division. Therefore any independent review of work performed by this division should be performed by personnel from another division. The future operations of the project waste processing facil-ities will be conducted by either the Facilities Division or the D&D Operations Division. An independent review of work by these divisions would have to be performed by personnel from other divisions.

The NRC review of the WVNS Policies and Procedures identified two organizations which play a role in performing independent safety reviews. The first is the Radiation and Safety Department which is in the Radiation and Environmental -

Safety Division. This group is functionally and organizationally separate-from the Facilities Division, the D&D Operations Division, and the Technical Divi-sion (Figure 4.1). This determination of organizational independence is based on an organization chart received from Project personnel during a site visit the week of February 17, 1986. The staff notes that this is an improvement over the organization chart presented in the Project Overview and General Information SAR module (Figure A.10.1-2). In this previous organization chart the personnel that make up the current Radiation and Environmental Safety Division were under one of the operations divisions of the Technical Division.

i 4-7 1

1 ,

Figure 4.1 WVNS Organization Chart i

T President t

1 1 Radiation & D&D r Facilities Environmental Operations Technical Safety 2

4

? Responsible for Responsible for rad- Responsible for Responsible for

'* upkeep and operation iological support decontamination developing and l of shutdown repro- for operation and operations, fuel constructing new j cessing plant. environmental and shipsients, and waste processing safety analyses.

disposal operations. facilities.

l t

I-Senior Tech Project ouclity Personnel Controller Administrative coneuttant Assurance Relations Services Control

• Managers of these divisions are merrbers of WVNS Radiation and Safety Committee.

** Manager of this division is Vice Chairman of the WVNS Radiation and Safety Committee.

l ***Chainnan of the WVNS Radiation and Safety Comittee.

l The second group which functions as an independent review group is the Radia-tion and Safety Committee. This group reports to the President of WVNS. It is made up of major divisional managers and is chaired by the WVNS senior techni-cal consultant (Figure 4.1). The staff concludes that these groups are capable of performing an independent review of work performed by the WVNS divisions.

NRC, after reviewing the policies and procedures which define the major role of the Radiation and Safety Committee (WV-906), discussed the makeup of the com-mittee with the Department and its prime contractor. The NRC stated its view that the Radiation and Safety Committee membership should reflect broad tech-4 nical interests and capabilities. The contractor has not identified specific minimum requirements for each position en the Radiation and Safety Committee in order to maintain some flexibility with personnel management. The Depart-ment and contractor have provided informati6n on the current makeup of the committee and the background of the indiv' duals (DGE 1986). In addition, both the Department and the contractor have agreed to provide similar information for any new members to the Committee.

A third group which functions in an independent review capacity is the Config-uration Control Board which reviews and either approves or disapproves all de-sign / field changes. This committee has as one of its members the Chairman of the Radiation and Safety Committee (WV-195, Rev 2). The staff finds that this independent committee, with the Chairman of the Radiation and Safety Committee as a member, is capable of performing an independent review of the public health and safety consequences of any proposed changes prepared under the DFC system.

The WVNS Policies and Procedures were reviewed to determine if they had require-ments for independent review and approval. The results of this evaluation of independent review and approval are summarized in Table 4.2. The review showed that independent reviews do occur at all the phases of hazard identification and management. There are, however, some inconsistencies between procedures WV-362 and WV-104, and between WV-906 and WV-104.

4-9

Table 4.2 Summary of Independent Check Made by WVNS Activity Organization Performing Independent Check Hazard identification in the Chairman of the Radiation and Safety-preparation of Functional Committee according to WV-362, Rev 2 and and Operating Requirements Radiation and Safety Department according to WV-104, Rev 2 Hazard identification in the Chairman of the Radiation and Safety preparation of design criteria Committee according to WV-363, Rev 1 and documents both Radiation and Safety Committee and Radiation and Safety Department according to WV-104, Rev 2 Hazard identification during Radiation and Safety Department reviews design reviews and approves drawings and equipment specifications according to WV-305, Rev 1 and WV-306, Rev 1 Hazard identification during The Radiation and Safety Committee re-Safety Review views and approves safety analysis according to WV-906, Rev 3 and both Radia-tion and Safety Committee and Radiation and Safety Department review and approve safety analysis according to WV-104, Rev 2 Hazard management at the The Radiation and Safety Department design stage reviews and approves drawings and equipment specifications according to WV-305, Rev 1 and WV-306, Rev 1 Hazard management during the Both the Radiation and Safety Department operational phase and the Radiation and Safety Committee review and approve Operational Safety Requirements according to WV-365, Rev 1.

The Configuration Control Board review and approve Design / Field Changes according to WV-195, Rev 2 i

1 4-10

. . = .

l l

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i T

5 RADIATION PROTECTION' PROGRAM

The NRC has reviewed the radiation protection progran, for the West Valley Demonstration Project, as related to public health and safety. The current

review of the Radiation Protection Program addresses five topics. These are:

source term estimation, dose assessment methods, release point and environ-mental monitoring, ALARA, and environmental reporting.

l 5.1 Source Term Estimation

{ Since Project activities are currently in the design and construction stage, some source terms or release quantities for the process systems may not yet be estimated. The staff has examined the methods used by project personnel to estimate release quantities for both normal operations and potential accident situations and assessed the completeness and the degree of conservatism l associated with the methods.

The solidification and other processes planned by the Department will involve both atmospheric and liquid releases. The atmospheric releases will come from

ventilation systems for rooms, cells, and vaults; from vessel vent systems; and-from process off gas systems such as that for the melter. There will be liquid i releases from the existing low-level waste treatment facility which is intended

! to be replaced soon by a new liquid waste treatment system. The new system will have some liquid discharges, but quantities are expected to be signifi-cantly less than those associated with the current low-level waste treatment l

facility.

I

! Section A.8.6.3.1 of the Project Overview and General Information SAR module

! discusses in general terms how source terms will be estimated. The SAR does-

not discuss specific data sources which might be used to develop source terms i because the presentation is intended to be more generic. Therefore, the staff examined the kinds of assumptions made for specific releases associated with 4

i h 8 j 5-1 i

i

the cement solidification facility and other current operations provided in other SAR volumes currently being reviewed. The review included both atmo-spheric and liquid releases. The staff finds that the general approach being taken by the project in these specific instances appears to be conservative.

Conservative release estimates are based on physical phenomena, measured data or limiting case assumptions. A detailed review of release estimates for each of the project operations will still be conducted when the individual SAR modules are reviewed.

5.2 Dose Assessment Methods The offsite doses resulting from plant operations are being projected prior to plant operation. The staff has reviewed the method for developing these pro-jections since they are also used to develop an appropriate monitoring program and to assess the impacts of plant operation.

Section A.8.6.3.4 of the Project Overview and General Information SAR module identified the codes being used for offsite dose evaluation. Additional details on the analytical methods are described in a report by WVNS (WVNS 1985).

For the evaluation of routine atmospheric releases, the Project is using the Dames and Moore code EPM 3 to calculate meteorological dispersion and a modified version of AIRD05-EPA to calculate doses. The LPGS code is used for routine liquid releases and the LADTAP-II code is used for accidental liquid releases.

The most recent ICRP recommendations (ICRP 26/30) are used for calculating doses to offsite individuals.

The staff has reviewed these codes and dose conversion factors and found them to be acceptable for analysis of the offsite radiological consequences at the West Valley Demonstration Project. 1 1

The staff reviewed the input used for these dose calculations and assessed the degree of conservatism associated with the individual input parameters. The staff made independent LADTAP-II computer runs as part of a determination of the liquid release unit dose factors which are being used by the Department. l l

5-2

LADTAP input parameters used by the Department were reviewed against parameters l identified in Regulatory Guide 1.109 and against site-specific information when such information was'available in the SAR. No discrepancies in the input data

-were identified, but a few parameters such as the dilution factors and service

populations for water supplies along Lake Erie could not be verified.

The runs were made using the standard LADTAP-II dose factor library (which is based on ICRP-2 methodology) and the input parameters identified in the WVNS 4

report. The NRC LADTAP-II outputs were adjusted using the ratio of the dose i conversion factors for ICRP 2 and ICRP 26/30. Unit dose factors for both the ._

maximum exposed individual and the population were compared. The unit dose factors for 21 isotopes were independently calculated by the staff. The unit dose factor which we generated was within a few percent (less than 9 percent)

, for all isotopes except for Ni-63 which was 37 percent higher. This difference is due to an old Ni-63 dose conversion factor being used by the Project.

j Project use of the current Ni-63 dose conversion factor would produce identical i unit dose factor. The NRC concludes that the unit dose factors which the Department will be using for continuous liquid releases are acceptable.

{

l The staff also made its own set of AIRDOS-EPA runs as part of an independent  !

j check of the maximum unit dose factors for atmospheric discharges. The Depart-

{ ment's AIRDOS inputs were checked against standard pathway parameters identi-

fled in the AIRDOS-EPA manual and NRC Regulatory Guide 1.109. In all cases examined, the Department's parameters were at least as conservative as the EPA

{

! or NRC values.

]

g The runs were made using the standard version of AIRDOS-EPA and meteorological j data identified in Supplement A.3.3.B to the West Valley Project Overview l Safety Analysis Report. The unit dose factors were calculated by the staff for i

1 28 isotopes and compared with the unit dose factors being used by the Project. l i For elevated releases, the NRC-calculated unit dose factors were within 30 percent of the values being used by the project. The differences are due to j the meteorology differences between AIRD05-EPA and EPM 3. The differences were 1

l l

4 5-3 I

i

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

i

slightly greater for ground level releases, but this is also due to differences in meteorological models. The NRC finds that the unit dose factors being utilized by the Project are reasonable.

t i

The staff also verified the unit dose factors being used by the Project for short-term atmospheric releases. When the differences between NRC and Depart-

! ment meteorological dispersion parameters are accounted for, the unit doses

] calculated by the staff were within a few percent of those being used by the

! Project.

i i

i The SAR did not present any unit dose factors for short-term liquid releases l for the NRC to verify.

2 l

5.3 Release Point and Environmental Monitoring i

j The West Valley Project has an environmental monitoring program which measures l both releases of radionuclides from the plant and the concentration of radio- i nuclides in the surrounding environment. Section A.8.6 1 of the Project Over-view and General Information SAR module discusses this monitoring program. '

i l The Project currently has an air sampling system for the main building stack which discharges all gaseous effluent from the process building and the waste l tank farm. The effluent is sampled and monitored for gross alpha concentra-

tion, gross beta concentration, and several specific radionuclides such as i Sr-90, Ru-106, 1-129, Cs-134, and Cs-137.

a ,

l  !

l The project also has a monitoring program for liquid releases. The major  ;

j source of liquid discharges from the site is the existing low-level waste '

) treatment system. This is monitored whenever a discharge is being made. In  :

} addition, other onsite surface waters are monitored in order to help monitor j radioactivity associated with site drainage. In previous evaluations, NRC i

reviewed this liquid discharge sampling program and determined its accepta-i bility. As discussed in Section 5.1, a new liquid waste treatment system will

) be installed soon.

l

5-4 l

It is not expected that changes to the environmental monitoring program will have to be made to accommodate future operations. However, adequacy of moni-toring systems will be verified before individual new operations are initiated.

T Offsite monitoring of the air, water, and food chain is also performed by the Project. Section A.8.6.1.1 of the Project Overview and General Information SAR module describes the offsite monitoring that is conducted. There are three offsite air monitors; one located up-valley, one down-valley, and one across the valley from the plant. Surface waters are sampled in Buttermilk Creek up-stream and downstream of the point where the site drainage enters the Creek and in Cattaraugus Creek upstream and downstream of the confluence of Buttermilk Creek with Cattaraugus Creek. Samples of the fish and game animals are taken in order to monitor the levels of radioactivity in the food chain. In addi-tion, dosimeters are situated around the site to monitor radiation levels.

The staff has reviewed the environmental monitoring program. The review shows that substantial improvements have been made over the program which was in place when the Department assumed control of the site. The changes involve elimination of some iodine monitorin; which is only relevant during fuel re-processing, more automation of the water sampling system and the addition of several new air monitoring stations.

In the Fall of 1985 the NRC conducted a field evaluation of the Project's radio-logical measurement program. The evaluation included an onsite review of laboratory analytical procedures, sampling techniques, measuring equipment, and results of internal and external audits of the programs. In general, the NRC

, concluded that the program was adequate, with some recommendations for improve-ment (NRC 1986).

I 5.4 ALARA A universal concept of radiation protection programs is that of keeping radia-tion doses as low as reasonably achievable (ALARA). We reviewed the West Valley Demonstration Project ALARA program for public exposure.

5-5

WVNS Policy and Procedure WV-984 describes the ALARA program for the West Valley Project. Tne general discussion states that the West Valley ALARA program incorporates measures for controlling population exposure. The de-tailed procedure, however, appears to be focused only toward occupational exposure.

We believe that the ALARA program should be extended to address more clearly the public health and safety. Also, ALARA analysis should be presented as part of system design and operation.

5.5 Environmental Reporting An important part of a complete public health and safety program is the report-ing of program performance. Reports documenting the releases from the facili-ties, the measured environmental concentrations, measured or calculated doses, and any planned improvements to the program should be issued.

The staff receives copies of the environmental reports which DOE issues for the West Valley Project. These reports present estimates of the maximum dose to individuals.

5-6

4 6 TRAINING 1

Operating personnel will be responsible for controlling.those systems and components important to public health and safety under both normal and abnormal circumstances. The emphasis of our review of the Department's contractor's j training program is to ensure that operators understand the safety significance of the operational controls and are properly prepared to respond-to abnormal

! conditions.

1

. Section 10.3 of the Project Overview and General Information SAR Module discuss the training program for the project. It describes how training is given to a-1 wide range of personnel. This evaluation concentrates on the nature of the a

. training given to radiation workers and plant operators because they are the ,

personnel most intimately involved with radioactive materials which'could l threaten public health and safety.

I l The training program is formal, continuing and includes testing. However, it i does not specifically address protection of public health and safety. Two WVNS Policies and Procedures, WV-220 " Review of Training Materials Produced" and i WV-538 " Personnel Indoctrination and Training," were reviewed to see if there I were any specific discussions of public health and safety concerns incorporated a

into training programs. There is no specific discussion in either document of i

training relative to public health and safety. The WVNS Policy and Procedure

! on Operational Readiness Review (WV-368) does state that the Operational

Readiness Review for a particular operation will examine operator training and qualification. The personnel on the Operational Readiness Review Board will have the opportunity to examine the details of operator training for the indi-vidual West Valley waste processing operations.

1 The Department has indicated that training does address detection of abnormal

. conditions and corrective actions. Also, in discussions between the staff and j WVNS training personnel, WVNS personnel stated that the results of engineering

safety evaluations were incorporated into the training program.

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The staff believes that the contractor'.s training program has.the potential to cover all the necessary elements for protecting public health and safety. .The staff will review the specific training program for selected project facilities prior to initiation of processing. In particular, the NRC will determine that each operator having responsibility for the operation of controls of systems

and components important to the health and safety of the public is fully familiar with all pertinent operating requirements. ,

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t 7 ACCIDENT ANALYSIS 1

As part of its analysis of the potential for affecting the health and safety of the public from activities related to the West Valley Demonstration Project, the Department incicdes consideration of credible accidents. A list of these 4

types of accidents is presented in Volume I in Table A.2.4-1. These accidents are related to ongoing operations, which is not the direct subject of our evalu-ation. However, the techniques used in the analysis are useful for understanding the methods to be used for future SAR modules.

Both natural phenoc,dna and " man-induced" causes of accidents are considered by the Department. Chapter 3.0 of this evaluation discusses the considerations given to the development of criteria for designing against effects of natural phenomena. However, the actual design of facilities important-to-safety should be analyzed to determine the effectiveness of the design to suppress or miti-gate the' release of radioactive material or loss of shielding against direct radiation in the event of a natural phenomena induced event. Chapter 3.0 also provides~a basis for calculating estimates of radiological impacts from releases to the environment. The majority of these releases will be t.o the a tmoso,h gre.

7.1 Engineering Confinement In Chapter.4.0 we discuss the general management approach taken by WVNS concern-

[ ing the protection of pubife health and safety. As part of this approach, the i Department's contractor Fas developed a safety classification system described

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in Section A.4.4.2 of Volume I of the SAR. The safety classification system is useado{cgicallyidentifythesafety-relatedsignificanceof. structures, systems and compongr.ts used for the solidification of high-level waste. Four classes are used, viz., A, B, C, and N. To be classified in the A category, the failure of the structure, system or component could cause an off-site effective dose equivalen[tiofgreaterthan25 rem. Class B items could result in a dose greater than 0.5 rcm. Class C items could result in an on-site effective equivalent dose of greater than 3 rem. In no instance has any structure, system, or

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component been classified as "A." The analytical reason for this is that, under normal atmospheric conditions, no motive force is available to spread contamin-ation from the site to obtain such a large dose.

However, the implication of such analyses is that there is a loss of control over the confinement of the radioactive material. Therefore, in addition to analyses performed for the safety classification, analyses are required to

i. demonstrate that, under all credible conditions, control, i.e., the safe con-finement of the radioactive materials, will be assured. The Department has begun to perform these analyses. The technique used is to (1) identify the 4

barriers which form an envelope of containment, (2) define the credible stresses which could be imposed upon the barriers, either as a result of natural I phenomena or from other causes such as fires or explosions, and (3) determine

!- the capability of the barriers to resist the imposed stress and provide con-tinuous confinement.

In a typical design, several barriers to release are present. Even if one or more of these barriers fail from the imposition of a severe stress, other barriers may maintain the necessary confinement. Conversely, if analysis indi-cates a design weakness such that safe confinement cannot be adequately demon-strated, changes may be necessary in the design. For each major system, the staff will review the Department's methods used to demonstrate engineered confine-ment of high-level waste during all processing and storage steps.

Although a major and severe equipment failure is not likely, such an event is considered credible. Since high-level waste liquids will be pumped under pressure, a pipe rupture is censidered by the NRC to be an upper limit to ,

the most severe type of accident which could occur during processing. The slurry feed materials have viscosities ten to one hundred times water and are not axpected to yield source terms comparable to the transferred supernatant.

Melted glass cools rapidly upon exposure to air at room temperature, especially if dispersed. Nevertheless, each of these sources, as well as possible dispersal through fire or explosion, will be examined to determine if the event

, could significantly impact public health and safety.

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7.2 Characteristic Upper Limit Accident Although the staff will examine the potential for accident within each major process step during the review of that step, a hypothetical upper limit acci-dent is postulated in this section under the restriction of continuous con-finement. This upper limit accident is assumed to be the release of super-natant solution from a pipe break during a pressurized transfer from Tank 8D-2 to Tank 80-1. Based on research performed at the Battelle Pacific Northwest Laboratory (PNL 1983), a maximum weight fraction of about 5x10 ' might become airborne. We estimate that a maximum of one-thousand gallons of high-level waste containing 26,000 curies of radioactive material could be released to the pipeway. The airborne material would be swept through the ventilation treat-ment system (discounting settling) and be filtered by two stages of high-efficiency particulate air (HEPA) filters. Each of these filters would remove all but 0.001 of the material. Again, discounting other treatment equipment, such as a mist eliminator, the quantity released from the stack would be about 0.013 milliceries of radioactive material. The staff estimates that the resul-tant effective whole body dose equivalent to an individual at the nearest posi-tion outside the exclusion boundary would be approximately 4x10 ' millirem.

This is only a preliminary estimate that may be modified after additional analysis based on the completed designs.

7.3 Emergency Planning One of the initial activities undertaken by DOE and its contractor upon assum-ing control of the West Valley site was to review, assess, and revise, as i appropriate, the emergency plan which had been established by the previous site 1 operator and evaluated and approved by the NRC staff. Various improvements were made to reflect DOE operating philosophy and new site activities and procedures. The staff did not conduct a formal evaluation of the revised emergency plan, but has received copies for review and participated in discus-sions with DOE and its contractor on the provisions of the plan. .The present j emergency plan does not cover the solidification activities which will be undertaken in the future and are the principal subject of this report.

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Prior to initia' tion of the first processing step toward the solidification'of I the high-level waste, i.e., the supernatant treatment _ system, it is expected that the current emergency plan will'be revised to include the new operation.

The staff will review and evaluate that plan insofar as it' applies to the.

health and safety of the public.

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Document Name:

WEST VALLEY SEC 8 SER Requestor's ID:

JOHNSONJ Author's Name:

CLARK T Document Comments:

4/22/87 Revised ETPB

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l 8- QUALITY ASSURANCE The NRC has reviewed the Contractor's Quality Assurance Program Plan for the 4

Project, including revisions. In the plan WVNS has committed to the basic quality requirements contained in ANSI /ASME NQA-I-1979, " Quality Assurance Program Requirements for Nuclear Power Plants," for the design, procurement, l construction, installation, modification, operation, and testing activities related to the Project. Such an approach is consistent with practices the NRC has concluded is acceptable for licensees. Although ANSI /ASME NQA-1-1979 has specific application to nuclear power plants, it also has been widely applied to fuel cycle facilities, such as the Project.

i l In addition to the staff's review of the Plan, the NRC has monitored the actual i implementation of the Plan. In March 1983 the NRC participated as observers during a Department of Energy audit of the contractor's quality assurance pro-gram. The observations of the NRC were provided to the Department. On July 29

, to August 2, 1985, NRC visited the West Valley site to monitor the implementa-tion of the quality assurance program as it related directly to ongoing con-struction' activities for the Project. The principal int'erest of the NRC for the quality assurance program is the identification and assessment of those design provisions related directly to the health and safety of the public.

Since the structures, systems, and components important to public health and safety had not yet been clearly identified, the assessment of the program was limited to observations of more general construction and design activities.

Several recommendations were provided to the Department (NRC 1985). Since that time considerable progress has occurred in the identification of the design-provisions related to the health and safety of the public. The NRC will con-tinue to monitor the application of the. quality assurance program to these specifically identified features.

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REFERENCES 4

AEC 1963 Safety Analysis of the Spent Fuel Processing Plant Division of Licensing and Regulation in the Matter of Nuclear Fuel Services, Inc.,- and New York State Atomic Research and Development Authority, Docket 50-201, February 15, 1963.

AEC 1965 Safety Evaluation by the Irradiated Fuels Branch Division of Materials Licensing in the Matter of Nuclear Fuel Services, Inc. ,

and New York State Atomic and Space Development Authority, Docket j 50-201, August 20, 1965.

DOE 1981 Federal Register, Vol. 46, No. 223, November 19, 1981,

p. 56960-56962.

L l DOE 1983a Letter, W. H. Hannum (DOE) to A. T. Clark, Jr., (NRC)

West Valley Waste Form, June 17, 1983.

DOE 1983b Letter, James A. Turi (DOE) to A. Thomas Clark, Jr. , (NRC) i October 28, 1983.

DOE 1984 West Valley Project Plan, December 12, 1984.

DOE 1986 Letter from W. H. Hannum (DOE) to A. T. Clark, Jr., (NRC)

" West Valley Demonstration Project (WVDP) Responses to NRC Comments and Questions," February 4,1986.

Fujita 1978 T. Theodore Fujita, Workbook of Tornadoes and High Winds for- ,

Engineering Applications, SMRP Research Paper No. 165, September 1978.

Fujita 1979 T. Theodore Fujita, et al., A Site Specific Evaluation of Tornado r and High-Wind Risks at West Valley Site, New York, SMRP Research Paper No. 178, November 1979.

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Fujita 1981 T. Theodore Fujita, Tornado and High-Winds Hazards at Western New York State Nuclear Service Center, West Valley, New-York, 1981.

Mcdonald 1981 James R. Mcdonald, Assessment of Tornado and Straight Wind Hazard Probabilities at the Western New York State Nuclear Service Center, West Valley, New York, July 1981.

NRC 1977 Nuclear Regulatory Staff Interim Safety Evaluation I, Docket No. 50-201, August 1977.

NRC 1981 Standard Review Plan for the Review of Safety Analysis Reoorts for Nuclear Power Plants Section 3.5.1.4, Missiles Generated by Natural Phenomena,' July 1981.

NRC 1982 Nuclear Regulatory Staff Safety Evaluation Report on the Dormant West Valley Reprocessing Facility, Docket No. 50-201, January 1982.

NRC 1986 Letter, A. Thomas Clark, Jr. , (NRC) to W. H. Hannum (00E), dated April 3, 1986, enclosing NRC report entitled, Monitoring Report on_ Measurement Systems and the Measurement Control and Environ-mental Monitoring Programs for the West Valley Demonstration Project - October, 1985.

NRC 1985 Letter, A. Thomas Clark, Jr. , (NRC) to W. H. Hannum (DOE), dated October 8,1985, enclosing report entitled, Monitoring Report on Quality Assurance Program for the West Valley Demonstration Project - August 1985.

NYSGS 1983 J. R. Albanese et al., Geologic and Hydrologic Research at the Western New York Nuclear Service Center, West Valley, New York, NUREG/CR-3782, November 1983.

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l PNL 1983 Sutter, S. L., Aerosols Generated by Releases of Pressurized Powders and Solutions in Static Air, NUREG/CR-3093, PNL-4566,-

Pacific Northwest Laboratory, Battelle Memorial Institute,

' August 1983.

WVNS 1984 Technical and Administrative Approach'for the West Valley Demon-stration Project Safety Program, August-1984.

WVNS 1985 Radiological Parameters for Assessment of West Valley Demon-stration Project Activities,--August 1985.

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