The potential for facility accidents and the magnitudes of their consequences are important factors in evaluating the stockpile stewardship and management alternatives addressed in this programmatic environmental impact statement (PEIS). The health risk issues are twofold:
Guidance for implementing Council on Environmental Quality regulation, 40 Code of Federal Regulations 1502.22, as amended (51 FR 15618), requires the evaluation of impacts which have low probability of occurrence but high consequences if they do occur; thus, facility accidents must be addressed to the extent feasible in this PEIS. Further, public comments received during the scoping process clearly indicated the public's concern with facility safety and consequent health risks and the need to address these concerns in the decision-making process.
For the No Action case, potential accidents are defined in existing facility documentation, such as safety analysis reports, hazards assessment documents, National Environmental Policy Act (NEPA) of 1969 documents, and probabilistic risk assessments. The accidents include radiological and chemical accidents that produce high consequences but have a low likelihood of occurrence, and a spectrum of other accidents that have a higher likelihood of occurrence and lesser consequences than the high consequence accidents. The data in these documents includes accident scenarios, probabilities, materials at risk, source terms (quantities of hazardous materials released to the environment), and consequences.
For new, modified, or upgraded stockpile stewardship and management facilities, the identification of accident scenarios and associated data would normally be a product of safety analysis reports performed on completed facility designs. However, facility designs have not been completed for the alternatives analyzed in the programmatic portion of this PEIS. Accordingly, the accident information developed for this PEIS has been developed based upon existing information for similar facilities. The likelihood and consequences of accidents (which are site dependent) are recomputed for each of the stockpile stewardship and management proposed sites where a facility may be located. This calculation reflects the effects of such site parameters as population size and distribution, meteorology, and distance to the site boundary.
This analysis also acknowledges, semi-quantitatively, the differences in likelihood of accident initiators at specific sites (e.g., aircraft impacts, beyond design basis seismic events, and so forth), as well as qualitatively discussing the opportunities for risk reduction afforded by the potential incorporation of new technologies, processes, or protective features in the stockpile stewardship and management facilities that will enhance public health and safety over the existing facilities.
Subsequent to this PEIS, evaluation of the specific benefits achieved by such measures would be presented in the tiered project-specific NEPA document for each facility. Also, for each new facility, a Hazards Analysis Document that identifies and estimates the effects of all major hazards that have the potential to impact the environment, workers, and the public would be issued in conjunction with the Conceptual Design Package. Additional accident analyses for identified major hazards would be provided in a Preliminary Safety Analysis Report (SAR) to be issued during the period of Definitive Design (Title II) Review. A Final SAR would be prepared during the construction period and issued before testing begins as final documented evidence that the new facility can be operated in a manner that does not present any undue risk to the health and safety of workers and the public.
The accident scenarios chosen to represent the impacts for each alternative were arrived at through a screening process based on a larger set of accidents presented in existing safety documentation for similar facilities. Documents such as those shown in table F.1.1-1 were reviewed for applicable accident scenarios and data. The process sought to identify a bounding accident in each of several classes of events (e.g., fire, explosion, spill, mechanical, criticality, natural phenomena initiators, and external initiators) applicable to the alternative. The process also sought to identify bounding accidents over the spectrum of high to low probability of occurrence in order to include high-consequence/low-probability and low-consequence/high-probability accidents. These accidents are generally referred to as beyond evaluation basis accidents and evaluation basis accidents, respectively. In accordance with Department of Energy (DOE) NEPA Guidelines, beyond evaluation basis accidents are generally in the probability of occurrence range of 10 -7 to 10 -6 per year (yr), and evaluation basis accidents generally have a probability of occurrence greater than 10 -6 /yr. These two designations are used only if formal SARs have not been prepared. In cases where SARs have been prepared, they are the source documents for two equivalent designations "beyond design basis accidents" and "design basis accidents." Based on discussions and meetings with experts, including a workshop, the accident scenarios were modified to reflect expected stockpile management facility conditions. For example, the material at risk identified in a safety report for a similar facility was adjusted to reflect the material at risk applicable to the Stockpile Stewardship and Management Program. A complete description of the development of accident scenarios for the alternatives is provided in a topical report (HNUS 1996a).
For each alternative, a number of evaluation and beyond evaluation basis accidents have been identified and are generally referred to as the "composite set of accidents." Two subsets of the composite set are also referred to as the "composite set of evaluation basis accidents" and the "composite set of beyond evaluation basis accidents." Impacts are presented for the composite set of accidents to reflect the combined impacts of evaluation basis and beyond evaluation basis accidents. The impacts for the composite set of evaluation basis accidents are also provided to reflect the impacts of high-frequency/low-consequence accidents and impacts for the composite set of beyond evaluation basis accidents are provided to show the impacts of low-frequency/high-consequence accidents. Evaluation basis accidents are generally in a frequency range greater than 10-6/yr, while beyond evaluation basis accidents are generally in a frequency range of 10-7 to 10-6/yr. In some cases, accidents less than 10-7 are included in the composite set of beyond evaluation basis accidents to provide information that is relevant to decisionmaking and that otherwise would not be considered.
For each alternative, each accident is analyzed to estimate its risk (i.e., mathematical product of an accident's probability of occurrence and the accident's consequences) and consequences (e.g., cancer fatalities) to a noninvolved worker, a member of the public at the site boundary and the population out to 80 kilometers (km) (50 miles [mi]) from the accident. The estimated risks for the composite set of accidents analyzed for the alternative are mathematically combined to obtain an average risk (cancer fatalities per year) and consequences (cancer fatalities), given that the accidents occurred. The data on individual accidents used to calculate the composite values are provided in section F.2.
ItemNumber |
Title |
|
Site |
|
Report Number |
|
Date
|
|---|---|---|---|---|---|---|---|
|
01 |
"The Continued Operation of the Pantex Plant & Associated Storage of Nuclear Weapon Components EIS" |
Pantex |
|
Draft Rev. 2 |
|
January 1995 |
|
|
02 |
Stockpile Stewardship and Management/PEIS Expanded Data Call Addendum to the Alternative Report for "Pit Manufacturing at Los Alamos National Laboratory" |
LANL |
|
none |
|
June 1995 |
|
|
03 |
Stockpile Stewardship and Management/PEIS Expanded Data Call Addendum to Alternative Report for "Pit Manufacturing at Los Alamos National Laboratory" |
LANL |
|
LA-UR-95-2670 |
|
Sept. 1995 |
|
|
04 |
Appendix D "Accident Analysis" |
LLNL |
|
Volume II |
|
Feb. 1992 |
|
|
05 |
Stockpile Stewardship and Management PEIS "Canned Secondary Assembly and Case Manufacturing Facility" Data Report |
LLNL |
|
SST 95-07-006 |
|
July 17, 1995 |
|
|
06 |
Draft EIS and EIR for "The Continued Operation of Lawrence Livermore National Laboratory & Sandia National Laboratories, Livermore" Unclassified Controlled Nuclear Information |
Sandia/LLNL |
|
Volume 1 |
|
Feb. 1992 |
|
|
07 |
Preliminary Draft EIS "The Continued Operation of the Pantex Plant & Associated Storage of Weapons Components" Unclassified Controlled Nuclear Information |
Pantex |
|
DOE/EIS 0225 |
|
Sept. 1995 |
|
|
08 |
EA for the "Proposed Interim Storage of Enriched Uranium Above the Maximum Historical Storage Level at the Y-12 Plant, Oak Ridge, Tennessee" |
Y-12 |
|
DOE/EA-0929 |
|
Sept. 1994 |
|
|
09 |
"Basis for Interim Operation for the Pantex Plant, Amarillo, Texas" |
Pantex |
|
none |
|
June 1995 |
|
|
10 |
"Revision 2 of the Basis for Interim Operation for TA-55-4" |
LANL |
|
ESH-3:94-105 |
|
June 1994 |
|
|
11 |
"Submittal of Revised JCO for CMR Facility" Unclassified Controlled Nuclear Information |
LANL |
|
none |
|
Feb. 1995 |
|
|
12 |
"Accident/Event Analysis" (Safety Information Document) |
Pantex |
|
Draft-Rev. 2 |
|
Jan. 1995 |
|
|
13 |
"CMR Facility (SM-29) Final Safety Analysis Report" Unclassified Controlled Nuclear Information |
LANL |
|
CMR-FAC-94-001 |
|
Feb. 1994 |
|
|
14 |
Executive Summary - "Hazards Analysis of the Los Alamos National Laboratory Plutonium Facility (TA-55)" Unclassified Controlled Nuclear Information |
LANL |
|
TA-55 FSAR |
|
July 13, 1995 |
|
|
15 |
Stockpile Stewardship and Management/PEIS "Alternative Report for Pit Manufacturing at SRS" Unclassified Controlled Nuclear Information |
SRS |
|
NMP-PLS-950176 |
|
Sept. 1, 1995 |
|
|
16 |
Draft Safety Analysis Report for "The Device Assembly Facility at the Nevada Test Site" Unclassified Controlled Nuclear Information |
NTS |
|
DAF SAR- 001-193-5394C |
|
March 1995 |
|
|
17 |
"U.S. Department of Energy Defense Programs Safety Survey Report" |
DOE |
|
DOE/DP/70056-HI |
|
Nov. 1993 |
|
|
18 |
"U.S. Department of Energy Defense Programs Safety Survey Report" |
DOE |
|
DOE/DP/70056-HI |
|
Nov. 1993 |
|
|
19 |
"U.S. Department of Energy Defense Programs Safety Survey Report" |
DOE |
|
DOE/DP/70056-HI |
|
Nov. 1993 |
|
|
20 |
"U.S. Department Of Energy Defense Programs Safety Survey Report" |
DOE |
|
DOE/DP/70056-HI |
|
Nov. 1993 |
|
|
21 |
"TA-55 Final Safety Analysis Report" Volume I Unclassified Controlled Nuclear Information |
LANL |
|
TA-55-PRD-108-01.0 |
|
July 13, 1995 |
|
|
22 |
"TA-55 Final Safety Analysis Report" Volume II Unclassified Controlled Nuclear Information |
LANL |
|
LA-CP-95-169 |
|
July 13, 1995 |
|
|
23 |
"TA-55 Hazard Analysis" Unclassified Controlled Nuclear Information |
LANL |
|
LA-CP-94-0076 |
|
July 13, 1995 |
|
|
24 |
"Nuclear Explosive Facilities Final Safety Analysis Report Nuclear Explosive Cells Module" (Buildings 12-44 Cells 1-6, 12-85, 12-96, and 12-98) Unclassified Controlled Nuclear Information |
Pantex |
|
Volume 1 - Draft B |
|
July 1995 |
|
|
25 |
"Nuclear Explosive Facilities Final Safety Analysis Report Nuclear Explosive Cells Module" (Buildings 12-44 Cells 1-6, 12-85, 12-96, and 12-98) Unclassified Controlled Nuclear Information |
Pantex |
|
Volume 2 - Draft B |
|
July 1995 |
|
|
26 |
"Chemical High Explosives Hazards Assessment for the Pantex Plant, Amarillo, Texas" |
Pantex |
|
none |
|
Oct. 1993 |
|
|
27 |
(Data Call) Tab D: "Facility Operations" Unclassified Controlled Nuclear Information |
Y-12 |
|
OR-9183 |
|
no date |
|
|
28 |
"Nuclear Explosive Facilities Final Safety Analysis Report Nuclear Explosive Bays Module" (Buildings 12-64, 12-84, 12-99, and 12-104) Unclassified Controlled Nuclear Information |
Pantex |
|
Rev. 1 Draft 2 |
|
Dec. 1994 |
|
|
29 |
"Nuclear Explosive Facilities Final Safety Analysis Report Nuclear Explosive Bays Module" (Buildings 12-64, 12-84, 12-99, and 12-104) Unclassified Controlled Nuclear Information |
|
|
Rev. 1 Draft 2 |
|
Dec. 1994 |
|
|
30 |
"Preliminary Safety Analysis Report Special Nuclear Materials Component Staging Facility" Unclassified Controlled Nuclear Information |
Pantex |
|
none |
|
April 1989 |
|
|
31 |
"Safety Analysis Report - On-Site Transportation" Unclassified Controlled Nuclear Information |
|
Pantex |
|
Draft B |
|
Sept. 1995 |
|
32 |
Stockpile Stewardship and Management/PEIS "Assembly/disassembly Nevada Test Site Alternative" |
|
NTS |
|
Volume 1 |
|
Aug. 4, 1995 |
|
33 |
Appendix 11-K - Release Fraction Data, Appendix 11-J - Consequence Equations Used in the Accident Analysis, Appendix 11-F - Seismic Accident Analysis, Appendix 11-E - Derivation of Data Values Used in the Accident Analysis Unclassified Controlled Nuclear Information |
LANL |
|
CMR-FAC-94-001 |
|
Feb., 1994 |
|
|
34 |
Draft "Design Process for Accident Mitigation" Pit Disassembly and Conversion Facility Unclassified Controlled Nuclear Information |
LANL |
|
Section 8 |
|
Aug. 21, 1995 |
|
|
35 |
"U.S. Department of Energy Defense Programs Safety Survey Report"
|
DOE |
|
DOE/DP/70056-HI |
|
Nov. 1993 |
One of the major design goals for stockpile stewardship and management facilities is to achieve a reduced risk to workers and the public relative to that associated with similar facilities in the existing Nuclear Weapons Complex. Significant changes exist between stockpile stewardship and management facilities and the current facilities design criteria and safety standards, which will reduce total risk to the public. These changes include design to current DOE structural and safety criteria; smaller throughput, batch size and inventories of certain hazardous materials; and elimination of some hazardous materials. This will reduce potential offsite health effects if an accidental release were to occur.
Stockpile stewardship and management facilities will be designed to comply with current Federal, state, and local laws; DOE orders; and industrial codes and standards. As a result, a facility will be provided that is highly resistant to the effects of natural phenomena, including earthquake, flood, tornado, high wind, as well as credible events appropriate to the site, such as fire and explosions, and manmade threats to its continuing structural integrity for containing hazardous materials. The facilities will be designed to maintain their continuing structural integrity in the event of any credible accident or event, including an aircraft crash, if credible at these sites.
The design process for new and modified stockpile stewardship and management facilities will comply with the requirements for safety analysis and evaluation in DOE O 430.1, Life-Cycle Asset Management and DOE Order 5480.23, Nuclear Safety Analysis Reports. Safety assessment is required to be an integral part of the design process to ensure compliance with all DOE safety criteria by the time that the facilities are constructed and in operation.
For new facilities, the safety analysis process begins early in conceptual design by identifying hazards with the potential to produce unacceptable safety consequences to workers or the public. As the design develops, failure mode and effects analyses are performed to identify events that have the potential to release hazardous material. The kinds of events considered include equipment failure, spills, human error, fire and explosions, criticality, earthquake, electrical storms, tornado, flood, and aircraft crash. These postulated events become focal points for design changes or improvements to prevent unacceptable accidents. These analyses continue as the design progresses to assess the need for safety equipment and to assess the performance of this equipment in accident mitigation. Eventually, the safety analyses are formally documented in an SAR and/or in a probabilistic risk assessment. The probabilistic risk assessment documents the estimated frequency and consequence for an entire spectrum of accidents and helps to identify design improvements that could make meaningful safety improvements.
The first SAR is completed at the conclusion of conceptual design and includes identification of hazards and some limited assessment of a few enveloping design basis accidents. This analysis includes deterministic safety analysis and failure modes and effects analysis of major systems. A detailed, comprehensive Preliminary SAR is completed by the completion of preliminary design and provides a broad assessment of the range of design basis accident scenarios and the performance of equipment provided in the facility specifically for accident consequence mitigation. A limited probability risk assessment may be included in that analysis.
The SAR continues to be developed during detailed design. The safety review of this report and any supporting probabilistic risk assessment is completed and safety issues resolved before the facility construction is initiated. There is also a Final SAR produced that documents safety-related design changes during construction and the impact of those changes on the safety assessment. It also includes the results of any safety-related research and development that has been performed to support the safety assessment of the facility. Final approval of the Final SAR is required before the facility is allowed to commence operation.
The MELCOR Accident Consequence Code System (MACCS) was used to estimate the radiological consequences of all stockpile stewardship and management facilities for all accidents. The CHEMS-PLUS (CHEMS-PLUS, Enhanced Chemical Hazard Evaluation Methodologies, Arthur D. Little, Inc., July 1988) computer code was used to estimate the consequences of nonradiological accidents. A discussion of the MACCS code is provided in section F.1.3.2. A detailed description of the MACCS model is available in a three volume report: MELCOR Accident Consequence Code System (MACCS), NUREG/CR-4691, SAND 86-1562, February 1990.
MACCS models the offsite consequences of an accident that releases a plume of radioactive materials to the atmosphere. Should such an accidental release occur, the radioactive gases and aerosols in the plume would be transported by the prevailing wind while dispersing in the atmosphere. The environment would be contaminated by radioactive materials deposited from the plume, and the population would be exposed to radiation. The objectives of a MACCS calculation are to estimate the range and probability of the health effects induced by the radiation exposures not avoided by protective actions.
In order to understand MACCS, one must understand its two essential elements: the time scale after an accident is divided into various "phases" and the region surrounding the facility is divided into a polar-coordinate grid.
The time scale after the accident is divided into three phases: emergency phase, intermediate phase, and long-term phase. The emergency phase begins immediately after the accident and could last up to seven days following the accident. In this period, the exposure of population to both radioactive clouds and contaminated ground is modeled. Various protective measures can be specified for this phase, including evacuation, sheltering, and dose-dependent relocation.
The intermediate phase can be used to represent a period in which evaluations are performed and decisions are made regarding the type of protective measure actions that need to be taken. In this period, the radioactive clouds are assumed to be gone, and the only exposure pathways are those from the contaminated ground. The only protective measure that can be taken during this period is temporary relocation.
The long-term phase represents all time subsequent to the intermediate phase. The only exposure pathways considered here are those resulting from the contaminated ground. A variety of protective measures can be taken in the long-term phase in order to reduce doses to acceptable levels: decontamination, interdiction, and condemnation of property.
The spatial grid used to represent the region is centered on the facility itself. The user specifies the number of radial divisions as well as their endpoint distances. Up to 35 of these divisions may be defined, extending out to a maximum distance of 9,999 km (6,213 mi). The angular divisions used to define the spatial grid correspond to the sixteen directions of the compass.
Since the emergency phase calculations use highly nonlinear dose-response models for early fatality and early injury, it is necessary for those calculations to be performed on a finer grid than the calculations of the intermediate and long-term phases. For this reason, the 16 compass sectors are divided into 3, 5, or 7 user-specified subdivisions in the calculations of the emergency phase.
The increased likelihood (probability) of cancer fatality to a member of the public is taken as 5.0x10 -4 times the dose in person-rem for values of dose less than 20 rem. For larger doses, when the rate of exposure is greater than 10 rads per hour, the increased likelihood of cancer fatality is doubled. The MACCS code was applied in a probabilistic manner using a weather bin sampling technique. Centerline doses as a function of distance were calculated for each of 150 meteorological sequence samples; the mean value of these doses and increased likelihoods of cancer fatality for the distance corresponding to the location of the maximum offsite individual at each site were reported for that individual. Doses to noninvolved workers were calculated similarly, except that these workers will experience an increased likelihood of cancer fatality of 4.0x10 -4 times the dose in person-rem for doses less than 20 rem or exposure rates less than 10 rads per hour. For larger doses, when the rate of exposure is greater than 10 rads per hour, the increased likelihood of cancer fatality is doubled.
The hypothetical worker was placed at 1,000 meters (m) (3,281 feet [ft]) or at the site boundary, whichever is less. It should be noted that since the doses and cancer fatalities for the maximum offsite individual and the workers reported in the high-consequence/low-probability accident tables are mean values based on approximately 100 meteorological sequence samples, there is no direct correlation between the mean value of dose and the mean value of cancer fatalities.
Offsite population doses and latent cancer fatalities are calculated by MACCS using a methodology similar to that described for the maximum offsite individual. In the case of the population, each of the sampled meteorological sequences was applied to each of the 16 sectors (accounting for the frequency of occurrence of the wind blowing in that direction). Population doses are the sum of the individual doses in each sector. Once again, the mean value of the calculated population doses and latent cancer fatalities for each of the trials are reported.