ML20196J391
| ML20196J391 | |
| Person / Time | |
|---|---|
| Issue date: | 10/19/1998 |
| From: | Callan L NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | |
| References | |
| SECY-98-239, SECY-98-239-01, SECY-98-239-1, SECY-98-239-R, NUDOCS 9812100038 | |
| Download: ML20196J391 (156) | |
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October 19.1998 SECY-98-239 FOR: The Commissioners FROM: L. Joseph Callan Executive Director for Operations
SUBJECT:
POST-DISPOSAL CRITICALITY RESEARCH PURPOSE:
To inform the Commission of the staff's assessment of whether to continue post-disposal criticality research at low-level waste (LLW) disposal facilities, and to obtain the Commission's direction on additional research.
BACKGROUND:
This paper responds to the April 29,1998, Staff Requirements Memorandum (SRM) concerning SECY-98-010," Petition for Envirocare of Utah, Inc., to Possess SNM in Excess of Current Regulatory Limits" (Attachment 1). This SRM directed the staff to review the Oak Ridge Il j 3
National Laboratory (ORNL) report on post-disposal cr;ticality at the Barnwell, South Carolina, disposal facility and to inform the Commission of its findings and recommendation on whether to continue post-disposal criticality research. The SRM also directed the staff to consult with the Advisory Committee on Nuclear Waste (ACNW) on generic issues. ;
As part of the staff's evaluation of the petition for rulemaking submitted by Envirocare of Utah, Inc. (Envirocare), staff performed a bounding analysis to evaluate the potential for special nuclear material (SNM) to migrate and reconfigure at LLW disposal facilities, resulting in an inadvertent criticality. As a result of this bounding analysis, the staff concluded that CONTACT: Tim Harris, NMSS/DWM (301) 415-6613
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I The Commissioners 2 post-disposal criticality concerns could not be dismissed as a possibility. Given the uncertainties associated with the assumptions and scenarios evaluated, the staff determined i that technical assistance was required to further evaluate the hydro-geochemical processes necessary to result in a post-disposal criticality. ORNL was contracted in 1995 to provide this i technical assistance, using the Envirocare site as a model. Following the study at Envirocare, :
the staff concluded that a LLW disposal site, in a humid climate, using different disposal '
methods, should be evaluated to better determine the likelihood of post-disposal criticality at LLW disposal facilities. The Barnwell disposal facility was selected as a model for this additional study. ,
In 1997, the staff identified the need for research in two areas of LLW criticality. The first i research need was prompted by the Commission's direction in the SRM on SECY-96-268, '
" Final Rule to Amend 10 CFR Part 71 for Fissile Material Shipments and Exemptions" (Attachment 2). The revisions to Part 71 set limits on unusual moderators such as beryllium,
- when shipping fissile material. Babcock and Wilcox (B&W) identified this concern during shipment preparation of a waste product from the down-blending of weapon-usable fissile material that could result in nuclear criticality. The Commission directed staff to consider the criticality issues raised in SECY-96-268 in a broad context and to examine previously unanticipated fissile materials and moderators in other areas of the fuel cycle and waste ,
programs. The second area of research was to develop a generic methodology to quantify the .
risk associated with post-disposal criticality. These two issues were consolidated, and a statement of work (SOW) was sent to ORNL. In response to the SOW, ORNL submitted a cost proposal on March 29,1998. This research was placed on hold pending implementation of the Commission's direction in the SRM on SECY-98-010.
Consistent with the SRM on SECY-98-010, the staff briefed the ACNW on July 20,1998. The presentation described the previous studies (i.e., ORNL's studies of Envirocare and Barnwell and the staff's work in support of the " Draft Environmental Impact Statement for the Shallow Land Disposal Area in Parks Township, PA"), and discussed the limitations of these studies in the context of the research requests. in response to this presentation, ACNW issued a letter dated July 30,1998, concluding that significant research on post-disposal criticality was not warranted. The Committee stated that the studies contained elements of a risk assessment but lacked consistency of application in propagation of realistic uncertainties through the analytical model. It recommended performing a quantitative risk assessment on a specific site and that this assessment be externally peer-reviewed.
DISCUSSION:
The staff's evaluations of the Envirocare, Bamwell, and Parks Township studies are discussed in Attachment 3. This evaluation included performing additional hydrologic and geochemical evaluations for the Barnwell site to determine whether additional research should be conducted.
A copy of the Barnwell study is included as Attachment 4. As discussed in Attachment 3, if the trench cover remains in place, under reasonably credible scenarios, a post-disposal criticality would be unlikely for 10,000 years. Staff also did an analysis conservatively assuming the trench cover is removed after 500 years, and found that criticality was unlikely for 1000 years.
The consequences of a post-disposal criticality were not evaluated for the Barnwell site.
Although the uncertainties associated with this likelihood estimate have not been completely quantified, the studies performed to date indicate that, while theoretically possible, post-disposal criticality is unlikely. Based on the additional work, staff concludes that additional
O e The Commissioners 3 research is not a high priority and that the uncertainty does not need to be precisely quantified.
However, additional work could be performed if the Commission wishes to quantify this uncertainty.
Attachment 5 is a differing professional view (DPV; as allowed under Management Directive (MD) 10.159) on the conclusions in this Commission Paper, submitted by a staff member on October 2,1998. The staff will review the DPV in accordance with the procedures in MD 10.159.
With respect to the unusual moderator issue raised in SECY-96-268, the staff responded to the SRM on SECY-96-268 in a memorandum to the Commission, dated May 21,1997. Staff concluded that the fuel cycle regulatory process is adequate to ensure public safety with respect to criticality issues of unanticipated moderators and fissile material. Staff proposed evaluating whether or not there is a need to restrict or provide constraints on the burial of unusual moderators in LLW facilities. As discussed above, the first research need noted that unusual moderators had not been considered in the previous post-disposal criticality studies.
Staff recommended this research in order to reduce any residual uncertainty concerning the issue of unusual moderators in waste.
Unusual moderators as they pertain to emplacement LLW criticality have been evaluated in NUREG/CR-6284," Criticality Safety Criteria for License Review of Low-Level Waste Facilities."
This report states that mass of beryllium should be limited to five times the mass of uranium-235 (U-235) and the mass of carbon (graphite) should be limited to twenty times the mass of U-235. To gain a general sense of the presence of unusual moderctors in LLW, staff contacted State officials, familiar with past disposal at LLW sites. Staff was informed that, during the past several years, large quantities of unusual moderators (beryllium and graphite) have not been disposed of at the Barnwell, Richland, and Clive sites, although the presence of such unusual moderators is not routinely noted on LLW manifests. This is consistent with staff's knowledge of LLW disposal practices. With the exccption of B&W, staff is not aware of any significant sources of beryllium among facilities licensed by NRC. In addition, large sources of graphite %
would be more economically disposed of by incineration. Further,10 CFR Part 71 limits the quantity of unusual moderators in fissile exempt shipments well below the limits identified in NUREG/CR-6284. Even if significant quantities of unusual moderators were present in LLW, it is unlikely that thay would be commingled with SNM to create a critical array.
Staff concludes that significant sources of unusual moderators are not present at LLW facilities.
In addition, in response to the SRM on SECY-98-010, staff is preparing guidance on emplacement criticality and will emphasize the need to limit significant quantities of unusual moderators in LLW disposal units, which is consistent with existing practices.
Ootions:
The staff has identified the following three options on post-disposal criticality research. If option 1 or 2 is selected, staff would not continue with further evaluation of unusuai moderators. For the reasons cited above, staff considers that research on unusual moderators in LLW disposal independent of research on post-disposal criticality is not warranted. If option 3 is selected, unusual moderators would be included as part of the generic methodology.
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- I The Commissioners 4 1.
Cease further review of oost-disoosal criticality. Existing studies indicate that long-time frames would be required to reconfigure the SNM into a critical mass. Monitoring of wells and sumps at disposal facilities, as suggested by ORNL, could mitigate any remaining concerns by providing early warning of any sign of reconcentration in the trenches.
However, signs of problematic conditions, if they occur, would be expected to take hundreds of years to appear and reliance on long-term monitoring is not consistent with the institutional control provisions in 10 CFR 61.59.
The principal advantage of this option is that no additional resources are required. Option 1 does not provide a comprehensive basis for concluding that post-disposal criticality poses no significant health and safety concern, because the uncertainties have not been fully quantified. Staff's analyses suggest that the likelihood that criticality will occur is very low over thousands of years.
- 2. Conduct a limited-scope study to reasonably cuantify the associated risk. This is the approach recommended by the ACNW. To implement this approach, staff would request technical assistance to evaluate an additional two to three trenches at the Barnwell site, using a two-dimensional or three-dimensional flow and transport computer code to model radionuclide transport under variably saturated conditions. Probability density functions would be established for key model parameters to evaluate the uncertainty. A probabilistic approach would be used to calculate the maximum U-235 accumulation and quantify the likelihood of occurrence. The consequences for any potential criticalities would then be determined. The risk would be obtained by multiplying the likelihood times the estimated consequences. If this option is selected, staff would brief the ACNW on this scope and obtain the Committee's input before proceeding with the technical assistance.
The principal advantage of this option is that it may reasonably quantify the risk of a post-disposal criticality with fewer resources than option 3. The results from this effort might provide the staff with a firmer basis for concluding that post-disposal criticality is not a significant concern. However, this option would not develop a generic methodology that could be used to evaluate other sites.
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! 3. Develoo a aeneric methodoloav. This option would develop a methodology, to quantify risk from a post-disposal criticality, that could be used by existing and future LLW disposal facilities. This option would include consideration of unusual moderators in LLW disposal.
The principal advantage of this option is that it produces a tool which could be used in the j future to quantify the risk associated with post-disposal criticality. However, expending i the resources to complete this option may not be warranted, considering the perceived small risk.
RESOURCES:
l Option 1 would not require any resources. The staff estimates that the cost to complete option 2 ,
is approximately $250K in contractor support and approximately 0.3 full-time equivalent (FTE).
The staff estimates that the costs to complete option 3 is approximately $400K in contractor support and approximately 0.5 FTE. Resources to conduct the activities described in these l
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The Commissioners 5 options are currently not included in the FY 1999-2000 budget for the Office of Nuclear Regulatory Research or the Office of Nuclear Material Safety and Safeguards. If option 2 or option 3 is chosen, resources will be reviewed during the next budget cycle.
RECOMMENDATIONS:
Considering the current staff workload, budget constraints, and the apparent low likelihood of reconcentration and criticality, the staff considers that additional technical assistance and/or research is of low priority, and therefore, recommends option 1, cease further review of post-disposal criticality. The staff considers this option to be consistent with the memorandum from the Chairman to the Executive Director for Operations, dated August 7,1998, which instructed staff to prioritize areas and increase the threshold for commencing new initiatives.
COORDINATION:
The Office of the Gsneral Counsel has reviewed this Commission Paper and has no legal objections. The Office of the Chief Financial Officer has reviewed this paper for resource implications and also has no objections.
N k .d. /L<w L. Joseph Callan 'q Executive Director \
for Operations Attachments: As stated Commissioners' completed vote sheets / comments should be provided directly to the Office of the Secretary by COB Wednedsay, November 4, 1998.
Commission Staff Office comments, if any, should be submitted to the Commissioners NLT October 28, 1998, with an information copy to the Office of the Secretary. If the paper is of such a nature that it requires additional review and comment, the Commissioners and the Secretariat should be apprised of when comments may be expected.
DISTRIBUTION:
Commissioners OGC OCAA OIG OPA OCA ACNW CIO CFO EDO SECY s
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l ATTACHMENT 1 - SRM ON SECY-98-010 l
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UNITED STATES Cys: Callan NUCLEAR REGULATORY COMMISSION Tht.dani
- i j WASHINGTON. D C. 20555-0001 Thompson
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- Norry o April 29, 1998 Blaha
- ...+ Bangart, SP or,,cc o, 7sg Knapp, RES SECRETARY Larkins, ACNW Funches, CF0 MEMORANDUM TO: L. Joseph Callan j a a te. N Executive Director for Operations C Jesse L. Funches Chief Financial Officer Anthony J. Galante Chief Infor ation Officer FROM:
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- ohn .H le, Secretary
SUBJECT:
STAFF REQUIREMENTS - SECY-98-010 - PETITION FOR ENVIROCARE OF UTAH TO POSSESS SPECIAL NUCLEAR MATERIAL IN EXCESS OF CURRENT REGULATORY LIMITS The Commission has not approved the staffs proposal to send a letter to Envirocare requesting additionalinformation regarding Envirocare's 1992 petition and 1997 exemption request at this time. Instead, the staff should focus its limited resources on Envirocare's Part 70 license application and inform Envirocare of the Commission's decision on this matter.
(hMSS) 9800081 The Commission has approved the staffs plans to develop guidance on emplacement criticality safety which could be used by Agreement States for existing and proposed low-level waste (LLW) disposal facilities. The staff should also investigate whether emplacement criticality requirements should be an item of compatibility, in accordance with the Commission's policy on adequacy and compatibility and based on realistic scenarios, and inform the Commission of its findings.
7/26/99 (EOG) (NMSS) (SECY Suspense: develop guidance: W30/99 9800082 compatibility determinations: -1134/99) 1/25/99 A'ter the Oak Ridge report is issued in final, the staff should review it and inform the Commission of its findings and of the staffs recommendations for resolution of whether the NRC research work on post daposal criticality of LLW should continue.
(EOG) (NMSS) (SECY Suspense: WB4/GB) 9800083 7/24/98 Based on the new policy and technical issues, the E <ecutive Council should consider program adjustments in FY 1998 and FY 1999 to commit resources for the LLW program to ensure that this program can meet its current demands. The Executive Council should inform the SECY NOTE: THIS SRM AND SECY-98-010 DISCUSS SENSITIVE INFORMATION AND WILL BE LIMITED TO NRC UNLESS THE COMMISSION DETERMINES OTHERWISE.
Attachment 1
Commission of the impact of this decision on the Strategic Plan, Strategic Goals, and existing programs.
(EDO/CFO/ClO) (SECY Suspense: 7M5/98) 9800084 NMSS 7/8/98 The staff should address any future year requirements in its FY 2000 budget submission.
(EDO) (NMSS) (SECY Suspense: -745/G8) 9800085 7/8/98 The staff should consult with the Advisory Committee on Nuclear Waste on generic issues associated with the Envirocare facility and other LLW sites, and consult with and obtain the Commission's approval on pokcy proposals necessary to resolve these issues.
(NMSS) 9800086 cc: Chairman Jackson Commissioner Dieus Commissioner Diaz Commissioner McGaffigan OGC OCA OlG
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ATTACHMENT 2 - SRM ON SECY-96-268 l 4
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Cys: Thompson UNITED STATES
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NUCLEAR REGULATORY COMMISSION Jordan Norry
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- a IN RESPONSE, PLEASE !
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- ' January 27, 1997 REFER TO: M970122D !
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- Blaha i OFFICE OF THE Meyer, ADM SECRET W Shelton, IR' ,
Lieberman, (
MEMORANDUM FOR: Hugh L. Thompsin, r. Miraglia, Ni Acting Executive Director for Operations Tanious, i l John F. Cordes, Acting Director Of ce of ommission Appellate Adjudication Joh C. oy k
e, Secretary i
FROM:
SUBJECT:
STAFF REQUIREMENTS - AFFIRMATION SESSION, 11:30 A.M., WEDNESDAY, JANUARY 22, 1997, COMMISSIONERS' CONFERENCE ROOM, ONE WHITE i i
FLINT NORTH, ROCKVILLE, MARYLAND (OPEN TO l PUBLIC ATTENDANCE) l 4
I L SECY-96-268 - Final Rule to Amend 10 CFR Part 72 for Fissile l
Material Shioments and Exemotions !
l The Commission approved publication of an immediately effective 2 .
final rule amending 10 CFR Part 71 to correct a recently !
discovered defect. The following changes should be made to the Federal Recister notice.
- 1. On page 2 ,in the first line of the Background section, add
' Babcock & Wilcox, Naval Nuclear Fuel Division (B&W),' after
' licensee.'
- 2. On page 3, in the first line on the page, replace ' Babcock &
Wilcox, Naval Nuclear Fuel Division (B&W),' with 'B&W '
- 3. The following sentences should be added to the " Summary" section and included in the public announcement.
"The regulatory defe.ct is not indicative of unsafe Rather, it was fissile material shipments in the past.
2 Section 201 of the Energy Reorganization Act, 42 U.S.C.
Section 5841, provides that action of the Commission j
shall be detemined by a " majority vote of the members present." Commissioner Diaz was not present when this item was affirmed. Accordingly the formal vote of the decision.
Commission was 4-0 in favor of the i
j Commissioner Diaz, however, had previously indicated that i he would approve this paper and had he been present he
- would have affirmed his prior vote.
t Attachment 2
or an identified by B&U during preparation for shipment ,
unprecedented type of fissile material that could l result in nuclear criticality under current requirements. This unique material is produced as a !
waste product from processing of strategic material l
resulting from operations to commercially downblend weapons-usable fissile material from the former Soviet l Unicn." i
- 3. above should be added after i The addition noted in number to.each of the Congressional letters !
sentence 1, paragraph 2, contained in enclosure 2. In line 1, paragraph 2, of the !
l Congressional letters, add 'on September 11, 1996,' after l
' notified the NRC.' l Following incorporation of these changes, the Federal Recister j
notice should be reviewed by the Rules Review and Directives i Branch in the Office of Administration and forwarded to the '
office of the Secretary for signature and publication. 02/28,'371 9600!! ,
(sDo) (RES) .7 suspense:
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The staff should ensure that codes used for criticality j calculations consider the n42n beryllium reaction. The photo '
neutrons are a minor component and difficult to take into account. The interaction between arrays (cake) could also be j important, ,
The~ staff should consider the criticality issues raised in 'h I i
SECY-96-268 in a broad context and examine previously- ;
unanticipated fissile materials and moderators in other areas o the fuel cycle and waste programs. The staff should consider 7 criticality issues regarding special moderating materials in " ,
processes at licensees' facilities, in storage awaiting !
transportation, and after disposal at waste facilities. 3/31/97) .j 97000:
(seet (NMSS) (SECY Suspense: {
The staff should consider issuing guidance to clarify the f i
application of the tables on pages 22 and 23 to situations where fissile materials with.different hydrogenous moderators may be 96001 shipped in the same container. (RES) j
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. . j 1 II. SECY-9~-004 - Seouovah Fuel Corporation and General Atomics; LBP-96.:24 Aporovina Settlement with General Atomics and DismiFsina Proceeding i
The Commission approved an order granting the petitions filed 2 1 by the State of Oklahoma, Native Americans for a Clean )
Environment, and the Cherokee Nation for Commission review of the l Atomic Safety and Licensing Board's Memorandum and Order, LBP .
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24, dated November 5, 1996, in which a majority of the Board j approved a settlement agreement between the NRC staff and General i l
Atomics. l (Subsequently, on January 22, 1997 the Secretary signed the i Order.) l l i l
l cc: Chairman Jackson l Commissioner Rogers l Commissioner Dicus i Commissioner Diaz I Commissioner McGaffican
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OCAA OCA OIG Office Directors, Regions, ACRS, ACNW, ASLBP (via E-Mail)
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l 2 Section 201 of the Energy Reorganization Act, 42 U.S.C.
Section 5841, provides that action of the Commission shall be determined by a " majority vote of the members present." Commissioner Diaz was not present when this item was affirmed. Accordingly the formal vote of the Commission was 4-0 in favor of the decision.
Commissioner Diaz, however, had previously indicated that he would approve this paper and had he been present he would have affirmed his prior vote.
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I ATTACHMENT 3 - DWM STAFF EVALUATION OF PREVIOUS POST-DISPOSAL CRITICALITY STUDIES l 1
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DWM STAFF EVALUATION OF PREVIOUS POST-DlSPOSAL CRITICALITY STUDIES A. INTRODUCTION The U.S. Nuclear Regulatory Commission has evaluated three low-level waste (LLW) and decommissioning sites for post-disposal criticality: (1) Parks Township, Pennsylvania; (2)
Envirocare. Utah; and (3) Barnwell, South Carolina. NRC staff evaluated the Parks Township site with some assistance from Oak Ridge National Laboratory (ORNL). ORNL evaluated the Envirocare and Bamwell sites with some assistance from the staff. These three sites have several attributes in common:
- 1. They disposed of uranium, including depleted uranium (DU) and special nuclear material (SNM)in the form of enriched uranium.
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- 2. There may be sufficient inventory and enrichments of uranium to lead to a theoretical criticality under ideal circumstances.
- 3. The enriched uranium was widely dispersed with soil or other wastes and often with much-larger quantities of DU or natural uranium.
B.
SUMMARY
OF STAFF ANALYSES FOR PARKS TOWNSHIP The Parks Township Shallow Land Disposal Area received waste, including SNM from the Apollo plant (NRC,1997). The staff analyzed the potential mechanisms of SNM migration from a dispersed state to a concentrated state, with favorable geometry and water content that could lead to criticality. The criticality calculations assumed uranium with an enrichment of 20 percent.
The analyses were limited to 15,000 years, judged to be the maximum time that the trenches would remain intact before erosional forces would disperse their contents. A critical configuration would require the migration of uranium from a highly dispersed state to a concentrated state. Vertical migration of uranium would not lead to criticality. Focused flow toward a hypothetical drain with capture of virtually all uranium would be required to accumulate a potential critical mass. However, the staff concluded that the accumulation of a critical mass within a single trench would require a highly efficient concentration mechanism to operate at the 3 expense of other, more likely redistribution processes. Using the available data, the analyses I indicated that there was reasonable assurance, making allowances for the time period and uncertainties involved, that the potential for a critical mass to form was so unlikely that criticality need not be considered further. It should be noted that the SNM inventory at the Parks Township site is significantly less than at LLW disposal sites.
C.
SUMMARY
OF ORNL ANALYSES OF ENVIROCARE AND BARNWELL The Envirocare study examined mobilization, migration, and concentration of uranium initially l
dispersed in soil, using geochemical transport models (ORNL,1997). This study considered the '
two concentration processes: (1) sorption onto materials such as iron oxide; and (2) precipitation in a chemically reducing zone. The analyses demonstrated that nuclear criticality could occur under certain unlikely conditions by vertical reconcentration of the uranium to a Attachment 3
2 thinner layer at the bottom of the disposal cell, by the process of chemical reduction and precipitation. Sorption processes alone did not lead to any critical conditions for the cases studied. This study assumed that the source was 100 percent enriched uranium at the maximum allowable concentration specified in the State of Utah's license. These conditions do not exist at the Envirocare site, and the actual conditions are much less conducive to nuclear criticality. The estimated average enrichment of the inventory through 1993 is only 0.42 percent, which is less than natural uranium and the theoretical minimum enrichment necessary for criticality in a water-moderated system.
The Barnwell study examined precipitation of uranium in reducing zones postulated to be caused by decaying organic material such as wood and cardboard, and corroding iron such as steel drums. However, the study noted that reducing zones would require saturated conditions, which do not presently exist in the trenches at the site (ORNL,1996). This analysis assumed that the uranium was enriched to either 10 percent or 100 percent and initially in containers.
Unlike the Envirocare evaluation, the Bamwell study considered both vertical migration of uranium to the bottom of the trench, and then horizontal transport, within the trench, to form a critical geometry of reconcentrated material.
The Barnwell study concluded that the minimum concentration of uranium-235 (U-235) to produce a criticality was 1.6 times greater for 10 percent enrichment than for 100 percent enrichment. Below 10 percent enrichment, significantly greater conce:.?ations of U-235 are required. Assuming the formation of reducing zones under saturated conditions (limiting oxygen in the system), reducing zones may be effective in precipitating uranium and appear to be stable even with the influx of oxidizing water. However, to date, the formation of reducing zones is hypothetical and has not been observed at Barnwell. For the conditions assumed in the Bamwell report, approximately 10,000 years would be required to mobilize and reconfigure the uranium. Based on actual uranium areal density of the disposal trenches, horizontal flow would be required to increase the areal density to pose a criticality concem.
ORNL suggests monitoring the sumps at the site for uranium, iron, and organics. If uranium is detected, its enrichment should be determined, and the redox condition should be evaluated by determining the speciation of iron. ORNL notes that placing caps over the trenches will limit infiltration and promote oxidizing conditions. Commingling SNM and source material (natural and depleted uranium) will reduce the average enrichment and thus reduce the concern of post-disposal criticality. Chem-Nuclear LLC, operator of the Barnwell facility, has been constructing caps over the waste trenches as part of the tritium migration mitigation.
- 1. Simplifying assumptions and uncertainties Some of the simplifying assumptions used in the Envirocare and Bamwell analyses included:
(1) uniform uranium distribution in waste; (2) one-dimensional flow; (3) saturated conditions; (4) stable reducing zones; and (5) U-235 and U-238 mobilized uniformly. The reports concluded that criticality might be possible under certain unlikely conditions, and long-time frames would be required. Assuming average enrichment and areal density, post-disposal criticality does not appear to be likely at the Envirocare and Barnwell facilities.
One key uncertainty with the ORNL studies is the assumption that the enrichment and U-235 concentration are homogeneous throughout the trench. For trenches 1 to 36, only the quantities of enriched uranium were reported, so it was conservatively assumed in the criticality
3 evaluations that the inventories were 100 percert enriched in U-235. Trenches 37 and above reported both SNM and source material. These records clearly show that large amounts of source material are commingled with the SNM in most trenches, or, in one case, that the amount of disposed SNM was very small. In the Barnwell study, one of the older trenches, for which only SNM was reported, (trench 23) was identi'ied as being of possible concern because of its high mass of U-235. To resolve this concern, the staff obtained trench loading records from the State of South Carolina. ORNL analyzed these records and concluded that there is three orders of magnitude more source material than SNM, and that the materials are commingled sufficiently ;
that the assumption of uniform enrichment is reasonable. However, other trenches might i contain localized areas of concentrated and enriched uranium that could pose a potential i concem. Localized concentrations of SNM in areas relatively free of source material could lead l
to accumulation of a critical mass in shorter time frames if there is " funneling" or convergence of I flow to a central location; immobilization mechanisms for uranium that worked efficiently; and configuration of the accumulated mass that had favorable neutronic properties. The staff cannot completely rule out this possibility without evaluating detailed disposal records, such as was done for trench 23.
Another area of uncertainty that may be mitigative is that the studies did not consider the effects of kinetics of sorption and redox reactions. Some research suggests that precipitation of uranium does not occur readily at low temperature (Meunier, et. al,1990; Duff, et. al,1997).
Further, the ORNL studies assumed reducing conditions would exist and be stable for time frames necessary to reconcentrate the SNM, which has not been demonstrated, and in fact is cornraindicated in most landfills (e.g., Greenfield, et. al,1990). Other assumptions that were not considered and could affect the time required to reconcentrate SNM include: (1) transient infiltration and concentrated flow conditions; (2) variable container weathering; and (3) combination or competition of various geochemical processes.
- 2. Additional analyses at Bamwell Staff conducted additional analyses for the Bamwell site to determine whether the Agency needs additional research and technical assistance regarding criticality at LLW disposal and decommissioning sites. The staff performed a computer modeling study of flow in the trench '
under the influence of infiltration at the surface, with and without the soil cap in place. To simulate moisture movement in a typical Bamwell trench, the staff used a two-dimensional, variably saturated flow code (Celia,1990). The trench was assumed to be 23 meters wide, six meters deep on one side and sloping to seven meters deep on the other side. A one meter deep French drain was modeled on the bottom of the lower side of the trench. The water table was assumed to be five meters below the trench bottom. Recharge in the native soil around the trench was treated as a steady state, constant flux boundary equal to 1.2 x 104 cm/s along the top of the model.
The condition 4
of a failed soil cap was treated as a constant flux boundary equal to 3.8 x 10 cm/s to conservatively represent no cover over the trench. The staff used a value of 4
hydraulic conductivity in the clayey sand of 10 cm/s. The modeling exercise with the missing cap showed that a moisture front reaches the trench bottom within the first year. There appears to be some lateral flow into the French drain, but no ponding there because water exfiltrates faster than it can accumulate.
Staff attempted to model moisture movement, assuming a lower hydraulic conductivity of
4 1.4 x 10 # cm/s for the clayey sand, but was unable to obtain convergence with the computer code. The staff believes that the higher values are more reasonably applied to the trench-scale model than the lower reported values, which were derived from small core damples. The higher value of hydraulic conductivity is also consistent with a number of field-scale studies at and near the site.
The staff also performed additional analyses using HELP (Schroader et. al,1988 ), a quasi-two dimensional water routing code for cnalyzing water movement in covers. The staff simulated the as-built trench, and concluded that infiltrating water would flow mostly in the vertical direction, with very little diversion (0.3 m /yr) from the trench to the French drain. The staff's analysis also showed that assuming no soil cap results in a small amount c; lateral drainage to the French drain (roughly 5 m / year on average). This analysis also showed that the maximum head on the clayey sand layer would be about 0.15 m, and the peak flow to the French drain would be 2 m8/d, leading to the conclusion that there would be little if any ponding within the trench. These analyses show that with a soil cap on the trench, the time needed to accumulate a sufficient mass of U-235 is on the order of 10,000 years, as predicted by ORNL. Staff also did an analysis in which it conservatively assumed that the trench cap was removed at 500 years and determined that at least another 500 years would be needed to reconfigure the SNM to form a critical mass. These time estimates assume that the reconfigured uranium has an enrichment of 10 percent. However, disposal records indicate that the average enrichment is significantly lower, and in most cases, below the minimum 1 percent enrichment needed to produce a criticality. Therefore, longer timeframes would be required if the enrichment was between one and ten percent. Moreover, there are additional reasons, which are discussed below, why post-disposal criticality is not likely.
Because it had been assumed that reducing conditions would be required to precipitate uranium, the staff also performed additional geochemical calculations, using PHREEOC (Parkhurst, 1995), to determine if uranium could be reconcentrated under oxidizing conditions (i.e., the conditions expected to exist in the Barnwell trenches). The staff modeled the exposure of urananite to water under oxidizing conditions. Such exposure resulted in the conversion of urananite to schoepite (another species of uranium oxide). This solution in a medium containing silica (i.e., sand such as exists in the Barnwell trenches) resulted in the precipitation of soddyite (another species of uranium oxide). The uranium concentration in the solution was significantly decreased, thus immobilizing most of the uranium in solid form. Based on this analysis, the staff concludes that reducing conditions are not required to precipitate uranium. However, the kinetics of these reactions are currently unknown, and so the efficiency of the analyzed mechanism under oxidizing conditions is uncertain.
l In summary, there are severallines of reasoning why the staff believes criticality at Bamwellis
- not likely
! . Actual areal densities of uranium in the 1.*enches requires horizontal flow to accumulate l sufficient mass of uranium to form a criticality. With the cover intact, virtually all flow will l be vertical in the unsaturated zone, with little if any flow to the French drain. With the i cover removed, most flow would still be vertical, but there would be greater diversion, up to 5 m'/ year, to the French drain. There would not likely be any accumulation in the French drain, however, because water would exfiltrate through the clayey sand as fast as it infiltrated.
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5 Long-term ponding of water to form stable reducing conditions does not appear likely.
Because the French drain will be above the water table and generally well-drained, even with the soil cap removed, the environment should be oxidizing, thereby promoting the rapid decay of organic matter. Being near the bottom of the trench, there would be no likely avenues for organic material from the surface (e.g., dead leaves) to accumulate.
Even with the cover in place, the sampled groundwater in the trenches is moderately oxidizing.
Some scientific evidence in the literature indicates that urananite does not form quickly or easily at low temperatures, even in highly reducing zones (e.g., Meunier,1990; Duff, 1997). However, precipitation of uranium in oxidizing conditions is possible (e.g.,
Hemingway 1982, Moll et. al,1996, and Nguyen et. al,1992). Even if it can be assumed that precipitation of uranium occurs in oxidizing conditions, the uranium would become essentially immobile in the drainage sand layer, and increased concentration of uranium due to horizontal flow would not occur.
The average enrichment of the uranium in the trenches is likely to be low Trenches for which the inventories of SNM and source material were reported generah Itad low average enrichments or very small inventories of SNM. There remains uncertainty about the average enrichment in the older trenches, for which only SNM quantities were reported. However, one of those trenches, trench 23, had less than 0.3 percent average enrichment when monthly disposal records were inspected. Although there might be local pockets of higher enrichment in any of the trenches, the scenario that depends on migration of uranium from a large area to the French drain would also homogenize the enrichments to a uniformly lower level.
D. CONCLUSIONS The NRC staff and its contractors have completed three evaluations of potentialin-ground criticality at LLW disposal and decommissioning sites where enriched uranium was disposed.
Although the uncertainties have not been fully quantified, these studies conclude that post-disposal criticality, vhile theoretically possible, is remote or unlikely.
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References l
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Celia, M. Personal communication,1990.
I j Duff, M.C., C. Amrhein, P. Bertsch and D. Hunter,1997,"The chemistry of uranium in evaporation pond sediment in the San Joaquin Valley, California, USA using X-ray fluorescence l
and XANES techt,.nes", Geochimica et Cosmochimica Acta, Vol 61, no 1, pp 73-81,1997.
Greenfield, B.F., A. Fosevear, and S.J. Williams, " Review of the microbiological, chemical and radiolytic degradation of organic material likely to be present in intermediate level and ;ow-level radioactive wastes", DOE /HMIP/RR/91/002, Department of Environment, HMIP, England, June 31,1990.
Hemingway, B.S.,1982, Thermodynamic properties of selected uranium compounds at 298.15 K and 1 bar and at higher temperatures--Preliminary models for the origin of coffinite deposits, USGS Open-File Report 82-619.
Meunier, J.D., P. Landais, and M. Pagel,1990, " Experimental evidence of uraninite formation from diagenesis of uranium-rich organic matter", Geochimica et Cosmochimica Acta. Vol 54, pp 809-817,1990.
Moll, H., Geipel, G., Matz, W., Bemhard, G., and H. Nitsche,1996, Solubility and speciation of (UO2)2SiOg2H2 O in aqueous systems, Radiochimica Acta, Vol. 74, pp 3-7.
Nguyen, S.N., Silva, R.J., Weed, H.C., and J.E. Andrews Jr.,1992, Standard Gibbs free energies of formation at the temperature 303.15 K of four uranyl silicates: soddyite, uranophane, sodium boltwoodite, and sodium weeksite, J. Chem. Thermodynamics, Vol. 24, pp 359-376.
NRC,1997," Draft Environmental Impact Statement Decommissioning of the Babcock & Wilcox Shallow Land Disposal Area in Parks Township SLDA," NUREG-1613 (withdrawn 1997).
ORNL 1997, "The potential for criticality following disposal of uranium at low-level waste facilities
- Uranium blended in soil", NUREG/CR-6505, Vol.1, June 1997.
ORNL 1998, "The potential for criticality following disposal of uranium at low-level waste facilities
- Vol. 2: Containerized waste", NUREG/CR-6505, Vol. 2, September 1998.
Parkhurst, D.L.,1995, User's guide to PHREEOC A computer program for speciation, reaction-path, advective-transport, and inverse geochemical calculations, USGS Water Resources investigation Report 95-4227.
Schroader, P.R., R.L. Peyton, B.M. McEnroe, and J.W. Sjostrom,1988,"The Hydrologic Evaluation of Landfill Performance (HELP) Model- User's Guide for Version 2," Vol. 3, U.S.
Army Engineer Waterway Experiment Station, October 1988.
4 ATTACHMENT 4 - NUREG\CR-6505, VOL. 2.,
"THE POTENTIAL FOR CRITICALITY FOLLOWING DISPOSAL OF URANIUM AT LOW-LEVEL WASTE FACILITIES"
NUREG CR 6505. Vol. 2 ORNL/TM-13323/V2 The Potential for Criticality Following Disposal of Uranium at Low-Level-Waste Facilities Volume 2: Containerized Disposal Prepared by L. E. Toran, C. M. Hopper, C. V. Parks, ORNL V. A. Colten-Bradley, NRC Oak Ridge National Laboratory Prepared for U.S. Nuclear Regulatory Commission l
ATTACHMENT 4 i
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' Die purpose of this study was to evaluate whether or not fissile uranium in low-level-waste (LLW) be concentrated by hydrogeochemical processes to permit nuclear criticality. A team of exp geology, geochemistry, soil chemistry, and criticality safety was formed to develop and test some reaso scenarios for hydrogeochemical increases in concentration of special nuclear material (SNM) and to us scenarios to aid in evaluating the potential for nuclear criticality. The team's approach was to performl simultaneous hydrogeochemical and nuclear criticality studies to (1) identify some possible scenarios i;f uranium migration and concentration increase at LLW disposal facilities, (2) model groundwate subsequent concentration increase via precipitation of uranium, and (3) evaluate the potential for nuclea 1
criticality resulting 2 from potential increases in uranium concentration over disposal limits. The al was restricted to "U in the present scope of work. The work documented in this report indicates that the l
potential for a criticality safety concern to arise in an LLW facility is extremely remote, but not Theoretically, conditions that lead to a potential criticality safety concem might arise. HoweverI1 hydrogeochemical mechanisms, the associated time frames, and the factors required for an indicate that proper entplacement of the SNM at the site can eliminate practical concems relative to the!
occurrence and possible consequences of a criticality event. ;
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Eage ABSTRACT........................................................................... iii LIST OF FIGURES
...................................................................vii LIST OF TAB LES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
EXECUTIVE S UMMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .l 1
ACKNOWLEDGMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .:
l 1 PURPOSE.............................................................................1 I 2 PREVIOUS WORK . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . '
i 3 S ITE DESCRIPTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . :
4 APPRO ACH . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ,
4.1 NUCLEAR CRITICALITY EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 4.1.1 Code Description and Validation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
4.1.2 Analytical Approach . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... . . . . . . . . . . . . . . . . . 13 4.1.3 Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ,
4.2 HYDROGEOCHEMICAL MODELING . . . . . . . . . . . . . . . . .. .. . .. .. 15 ........
4.2.1 Conceptual Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
4.2.2 M odels Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . )
4.2.3 Parameters and Model Grid . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
4.2.4 Sensitivity Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... .. .. .. .. 19 .......... '
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1 5 AS SUMPTIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
F 6 CRITICALITY SAFETY EVALUATION RESULTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .23. . . . . . . . .
6.1 2"U ENRICHMENT INFLUENCE ON CRmCAL MASS OF URANIUM .....
............23 6.2 TRENDS FOR U(10}.-H 0-SIO: MIXTURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 2
. 6.3 COUPLING OF NUCLEAR CRITICALITY AND HYDROGEOCHEMICAL MODELING . . . 28 P 7 HYDROGEOCHEMICAL MODELING RESULTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 7.1 REDUCING ZONES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 7.2 SENSITIVITY ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 7.3 PRECIPITATION OF SILICATE MINERALS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 4 l
7.4 3-D HYDROGEOCHEMICAL MODEL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 8 DI S CU S S I O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
8.1 ENRICHMENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 8.2 S OURCE TERM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
8.3 GEOCHEMICAL PROCESSES . . . . . . . . . . . . . . . . . . . . . . .. . .. .. 39 ........ i NUREG/CR-6505, y
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. 8.4 HORIZONTAL-VS-VERTICAL FLOWPATHS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 8.5 OTHER MITIGATING FACTORS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 45 9 CONCLUS IONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . 4 7 10 REFERENC ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1 APPENDIX A: Criticality Study Results . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55 APPENDIX B: Suberiticality Evaluation for Chem-Nuclear Systems, Inc., Trench 23 . . . . . . . . . . . . . . . . . 79 t
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t LIST OF FIGURES i Figure page 3.1 Typical construction of waste trenches . . . . . . . . . . . . ............... ... ..... ... ...... 6 3.2 Histograms of source material in trenches
.. .... ...... .. ............ . ............8 3.3 Example ofredox zonation in a landfill cross section
............................ .. ...... 11 4.1 Conceptual configuration for nuclear criticality evaluations ... .............. . . . . . . . . . . . . . . . 14 4.2 Schematic of mode; grids showing oxidized injection into reducing zone . . . . . . . . . . . . . . . . . . . . . 17 2
6.1 Critical mass "U vs H/X for UO F,-H 2 O 2 in spherical H 20-reflected systems . . . . . . . . . . . . . . . . . . . . 24 6.2 Critical mass U vs H/X for UO2 F,-H 2O in spherical H2 0-reflected systems . . . . . . . . . . . . . . . . . . . . . . 25 6.3 2 2 Infinite slab areal density (kg "U/m ) vs g 2H 0/g SiO 2 and log scale of2 g "U/g SiO 2
. . . . . . . . . . . . 27 7.1 Reaction progress for hydrogeochemical model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 8.1 Relationship between enrichment and source term . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 8.2 Critical (k,,. 2 0.95) masses of spheres and " pseudo-infinite" slabs and cylinders 2
as a function of water content for different densities of "U in waste matrix . . . . . . . . . . . . . . . . . . .
8.3 Histograms of calculated concentrations of"'U in old and new disposal trenches 2
using reported SNM or "U mass and reported disposal volumes . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 8.4 Histograms of calculated areal densities of 2"U in old and new disposal trenches 2
using reported SNM or "U mass and reported disposal volumes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 8.5 Cross sections showing horizontal flow (a) within the waste zone due to preferential flow paths and (b) at the base of the trench due to less-permeable sediments at the bottom botmdary and trench drainage
....................................................... .............43 A.1 Infinite media neutron multiplication factor (k.) vs g H2 0/g SiO2 and g2 "U/g SiO2 . . . . . . . . . . . . . . . . 57 A.2 Infinite media neutron multiplication factor (k.) vs g H2 0/g SiO2 and log scale of g2 "U/g SiO2 ......58 A.3 Infinite slab thickness (cm) vs g H2 0/g SiO2 and g 2"U/g SiO 2 ................................59 2 2 A.4 Infmite slab areal density (kg "U/m ) vs g H2Olg SiO2 and g2 "U/g SiO2 .......................60 2 2 A.5 Infinite slab areal density (kg "U/m ) vs g H 20/g SiO2and log scale of g2 "U/g SiO2 . . . . . . . . . . . . . . 61 A.6 Infinite cylinder diameter (cm) vs g H 20/g SiO and 2 g 2nU/g SiO2 .............................62 2
A.7 Infinite cylinder linear density (kg "U/m) vs g H2 0/g SiO2 and g2"U/g SiO2 ....................63 NUREG/CR-6505, vii Vol. 2
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A.8 Infinite cylinder linear density (kg "U/m) vs g H2 O/g SiO2 and log scale of 3g "U/g SiO 2
. . . . . . . . . . 64 A.9 Sphere diameter (cm) vs g H2 0/g SiO, and g 2"U/g SiO 2. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 65 A.10 Sphere mass (kg2 "U) vs g H2 0/g SiO2 and g2"U/g SiO 2 .................. ....... ... ... . 66 2
A.11 Sphere mass (kg "U) vs g H2 0/g SiO2 and log scale of g2 "U/g SiO2 . . . . . . . . . . . . . . . . . . . . . . . . . . 67 1
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LIST OF TABLES ,
h Tahle hge 3.1 Concentration of dissolved nonmetals and metals in trench water samples taken at the LLW !
burial site near Bamwell, S.C. j
..........................................................9 j 4.1 l Components and reactants (minerals and aqueous complexes) used in ParSSim . . . . . . ............18 l
6.1 U(10) plus H2 O plus SiO 2 -soil (S-S) results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 7.1 Parameter variation and results of hydrogeochemical modeling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 8.1 Disposal records from Barnwell, S.C., and calculated enrichments and density . . . . . . . . . . . . . . . . . . . 35 !
A.1 U(10) plus H2 O plus SiO2 -soil (S-S) results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 68 i
I B.1 Suberiticality evaluation assumptions and ramifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 83 I B.2 Raw and transformed data from Autry, 1998 .............................................. 84 '
B.3 Concentration factars for criticality concern . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 85
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. 4 EXECUTIVE
SUMMARY
This work is Volume 2 of a two-volume study to address the potential for nuclear criticality resulting increasing2 uranium concentrations in low-level-waste (LLW) disposal facilities. In contrast to Volume 1 focused on "U blended with soil, this report focuses on containerized waste with 10 weight perce in uranium, U(10). Hydrogeochemical modeling assumed precipitation of uranium in reducing zones in concentration on sorption sites as a means of forming a critical mass. As in the earlier study, several assumptions were made in developing the hydrogeochemical and criticality models. These are discussed Section 5.
2 The criticality safety calculations showed that higher concentrations of "U were needed for U(1 as expected. Using the minimum 2 concentration values necessary for a potential criticality, 2 the mass of "U' 2
i U(10) was 1.6 times greater than the "U as U(100). Differences in "U concentration have not be i a point-by-point basis and could be larger and smaller than the 1.6 value observed at the minimum con posing criticality concems. l l
The mechanism of precipitation for increasing the concentration of uranium in assumed reducing zone saturated conditions has been evaluated. These reducing zones formed very efficient barriers to uranium transport, precipitating nearly 100 wt % of the uranium in solution. The r sults of the geochemical mode indicated that the reducing zone did not become oxidizing despite the influx of oxidized water. The source reducing agents is postulated to be steel dmms or wooden crates, thus serving as plausible '
of uranium. Further study of the geometry of these reducing zones would be needed to evaluate the ,
! concentrating relatively small critical masses (e.g., spherical masses). Other limitations may be identified through the evaluation of reducing zones.
I Nonetheless, disposal practices at the Chem-Nuclear Systems, Inc., disposal facilities at Bamwell, S.C., res the possibility of criticality safety concerns in several ways. Very low average 2nU enrichments have been reported for most trenches, below the2 I wt % limit to produce a criticality concem under typical disposal conditions. For most trenches with higher "U enrichments, the source term (e.g., mass) for uranium is too low 2
to produce a slab of sufficient size with the required increase in concentration of "U needed for criticality concern. One exception is Trench 23, which has a high enrichment (greater than 80 wt %) and a large source (175 kg) of2 "U.
2 Even for the limited examples that potentially have sufficient "U, very long times are needed to accumulate a critical mass. For the most conservative travel time, assuming one-dimensional (1-D) flow and no dispersion tens to hundreds of thousands ofyears are needed. Bree-dimensional (3 D) modeling indicates even longe times are needed when dispersion is incorporated. The flow paths would need to funnel the uranium from trench volumes to relatively small reducing zones in order to increase the concentration of 2"U to the level that 2 ;
would pose a criticality safety concern. Iflarger reducing zones form either the "U will be too diffuse to pose a criticality safety concem, or larger sources of 2"U than are reported to be present in the trench are required.
Uranium travel times are long enough to allow monitoring and possible mitigation of conditions that could pos criticality safety concems.
His study results in the following recommendations for consideration oflicense reviews of LLW facilities:
- 1. Minimize those factors that enhance SNM accumulation.
Reduce groundwater infiltration Reduce enrichment i NUREG/CR 6505, xi Vol. 2 l
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Executive Summary .
Minimize opponunities to create isolated zones of reducing conditions. Avoid organic matter in waste cells Design trench to minimize focused flow ,
- 2. Limit the areal density of the fissile materials.
- 3. Model trench performance using site-specific conditions on a scale that addresses the potential for criticality.
Consequently, the observation that the average enrichment of a trench is less than 1 wt % "'U in the uranium does not necessarily eliminate a criticality concern for the trench. Burial reports may suggest that localizr/
regions of a trench contain quantities of fissile material that greatly exceed the average enrichment.
- 4. Continue to use sumps in disposal trenches to monitor for the presence ofiron, organics, and uranium as I indicators of mobility in the trenches. If uranium is observed in the sumps, determine its enrichment. !
Changes in redox conditions may be monitored by changes in different iron species. Even though it may take many years for sufficient buildup of uranium, early detection of mobile iron and uranium would indicate changes in the trench water chemistry.
t Disposal trenches at the Bamwell, S.C., LLW facility have waste materials containing uranium with average "'U !
enrichments less than I wt %, insufficient masses of"'U at enrichments larger than I wt %, or distributions and mass proponions of"'U and "U such that criticality safety concerns are not a realistic issue. For the single i disposal trench, Trench 23, having a large mass of waste material containing highly enriched uranium, suberiticality is ensured by the physical distribution and commingling of the material with substantial quantities of" source material," which is typically nonnal or depleted uranium (e.g.,0.7 or 0.2 wt % n5U in uranium ,
respectively). Uranium concentration factors (i.e., hydrogeochemical relocations and densification of uranium) ;
larger than 10 are required to pose a potential criticality safety concern. As demonstrated by the evaluation of the reaction for hydrogeochemical cumulative uraninite precipitation for long time frames, it requires a minimum of 7,000 years to increase the "'U density from about 0.002 g/cm' to 0.02 g/cm'. It isjudged that the same hydrogeochemical processes will redistribute and concentrate the commingled source material that is present, thereby further ensuring suberiticality through isotopic dilution of the SNM with normal or depleted uranium to ?
"'U enrichments less than I wt %. <
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- ACKNOWLEDGMENTS This work was supported by the NRC under Task 14, " Reconcentration of SNM in Low-Level Disposal Facilities," of JCN L1376, Technical Supportfor Design. Construction, Operation. and Performance R Low-Level Waste. V. Colten-Bradley of the NRC provided technical direction and contributed substantially the work throughout the duration of this project. He cooperation received from the staff at the Chem-Nuclear Systems, Inc., disposal facilities at Barnwell, S.C., and the State of South Carolina, Department of Health and Environmental Control, was greatly appreciated by the authors.
The authors acknowledge the helpful assistance provided by Patricia B. Fox and Lester M. Petrie, Jr., for performing numerous neutronic calculations and producing representations of the computational results. We als acknowledge thoughtful reviews of this manuscript by Gary Jacobs and John McCarthy.
P NUREG/CR-6505, xiii Vol. 2 h
1 PURPOSE ne purpose of this study was to evaluate the potential for hydrogeochemical processes to nuclear material (SNM) in containerized low-level-waste (LLW) disposal facilities such that there is concentration increase and geometry reconfiguration to permit nuclear criticality. This particular eval restricted to criticality safety concerns and geochemical processes associated with uranium. T (1) to identify some reasonable scenarios for uranium migration and increase in concent facilities, (2) to model coupled groundwater transport and geochemical speciation of uranium the potential for nuclear criticality in terms of passive geometry configurations and increases in uran concentration.
the evaluation. A combination of hydrogeochemistry and criticality safety experts worked togethe This study extends the previous work reported in The Potentialfor Criticality Following Dispos Low-Level Waste Facilities, Volume 1: Uranium Blended With Soil (Toran et al.,1997). The pres\ i emphasizes the disposal of containerized uranium instead of disposal of uranium blended in soil. So scenarios and concentrations are evaluated. In particular, the emphasis in this report is on the mobiliza uranium under oxidizing conditions, with immobilization under reducing cor..htions. In the previl uranium was desorbed from the soil and resorbed in a zone of high-sorption sites. This process did nol the uranium concentration enough to cause nuclear criticality safety concems, and the scenario will not reexamined 2 here. In addition, the previous criticality safety analysis assumed the uranium was enriched to) 100 wt % "U [ referred to as U(100)). In this report, additional calculations are presented 2 for 10 wt uranium [ referred to as U(10)], and some comparisons are drawn between these U(10) calculations and ti previous U(100) computational results reported in Vol.1.
The Chem-Nuclear Systems, Inc. (CNSI), LLW disposal facility at Bamwell, S.C., was used as a site conditions for containerized disposal. Specific disposal practices at the site were evaluated as the potential for the "U concentration to increase.
However, the models were process-oriented rather than site-specific. That is, the models emphasiz that could occur in disposal settings, rather than being a detailed construction of site conditions su used in a performance assessment. Some details could not be addressed without a site-specific model that incorporates transient soil moisture conditions, or without additional data such as packing configurations a weathering rates. Assumptions were selected based on judgment regarding the potential conditions that would increase the possibility for criticality.
The questions addressed in this study are:
Is there sufficient inventory for the available geometries requisite for criticality?
e 2 How does the concentration of "U needed for criticality compare with systems containing U(10) vs U(1 e
What chemical conditions and physical aspects of trenches are conducive to increasing uranium concentration?
e Can reducing zones, which precipitate uranium, be sustained to enable critical masses to accumulate?
How could disposal pra.Tes, in particular at Bamwell, S.C., enhance or mitigate the development of entical masses?
I NUREG/CR-6505, 1
Vol. 2
s t Purpose . Section 1 ne questions addressed in this study reflect the imponant processes that could be evaluated with the available data, hydrogeochemical models and criticality safety analyses. The results provide bounds on conditions that could raise nuclear criticality safety concems. Insights gained from the hydrogeochemical modeling and criticality safety analyses will be used as a basis for recammendations concerning future disposal practices for Sh%f.
NUREG/CR-6505, Vol. 2 2 t
2 PREVIOUS WORK h
In the previous report, Vol.1 (Toran et al.,1997a), nuclear criticality evaluations and hydrogeochemic were based on licensed soil-contamination limits specified for Envirocare of Utah,Inc. The maxl ,
concentration of"'U permitted 2in disposed waste under the State of Utah license 2 (UT 23002 i
soil, which equates to about 0.0006 g of "U per em' of soil given a soil bulk density of about 1.6 g If disposal occurred at this maximum concentration, there is the theoretical possibility of a accident, given assumptions about hydrogeochemical influences on reconfiguration of the ura .
narrow range of conditions resulted in sufficient increase in uranium concentration, and the length of time required to increase the concentration of uranium is expected to be many thousand of yearsl to criticality will further mitigate consequences that occur with rapid approaches to critical conditions.
However, it is important 2 to note that reviews of disposal records from Envirocare of Utah, Inc., indica concentrations 2 of "U in the waste material are more than a factor of 10 less than allowed by the licens the average site "U enrichment is below the minimum I wt % (Pruvost and Paxton,1996) required t nuclear criticality. Bus the likelihood of a criticality accident is vanishingly small. i i
Because of the numerous combinations ofparameters that could be considered in nuclear criticality and hydrogeochemical modeling, bounding and simplifying assumptions were used in the analysis. N criticality evaluations were performed for simple geometries using two generic2 soil types: SiO soil (the m conservative medium because pure 2 SiO is the least likely soil composition to absorb neutrons, thereby i enhancing the potential for criticality) and a " nominal soil" composed of minerals and secondary phase representative of a world-average soil composition.
Potential, direct radiation exposures 2 were estimated for two postulated types of criticalities: one wi concentration factor (large increase in "U concentration) and one with a low concentration factor. The locations l
of the determined radiation exposures were for two positions 1 m above grade. One position was dir !
the deposit, and the other position was 90 m away from the deposit. The assumed fission yields fr(
systems were based upon the fission energy release necessary to remove the quantity of water that is i
moderate neutrons to sustain nuclear criticality throughout an over-moderated condition. These assump j were predicated upon a geologically slow approach to a non-idealized critical geometry, thereby permitting '
localized steam generation and self regulation and shutdown of the fission-chain reaction. Alternative idealized assumptions have been pt 'ulated by others (Bowman and Venneri,1994; Greenspan, Armel, Ahn and Vujlj 1997) that present more severe consequences. '
ne criticality evaluation showed that the SiO 2 -soil results are similar to the nominal soil results. In terms of the I
hydrogeochemical processes that can increase uranium concentration, the critical slab configurations are more i readily achieved than cylindrical or spherical configurations (i.e., lower concentration factors are required).
De criticality evaluation also provided a minimum concentration needed to achieve criticality safety concems, which was the target concentration for hydrogeochemical modeling.
Simplifying assumptions in the hydrogeochemical modeling included one-dimensional (1-D) transport and saturated conditions. The hydrogeochemical scenario was postulated based on the geometry of the Envirocare site with disposal in soil containing sorbed uranium, then mobilization to a zone of higher-sorption capacity below the disposal trench. This sorption zone could potentially contain uranit'm in a zone of higher concentration. A reducing zone to capture uranium was also hypothesized, but not explicitly modeled,in the previous study. A reducing zone was difficult to define given the limited supply of reducing agents in the trenches and the unsaturated conditions in soil, which would keep the system oxygenated. A sensitivity analysis <
Vol. 2 l l
Previous Work Section 2 was performed to evaluate various factors, such es concentration of complexing agent, quantity ofinitial uranium source term, and groundwater velocity on the r stential to increase uranium concentration.
The previous work noted that the concentration of complexing agent and the size of the source term were limiting factors in the reconfiguration of uranium. For most scenarios, once sufficient uranium was mobilized, the concentration of a complexing agent was important because it outcompeted sorption sites in the high-sorption zone and prevented increases in uranium concentration. The possibility ofimmobilizing uranium in reducing sones was presented as a more-likely scenario (to be evaluated in the present work). Furthermore, if the initial concentration of uranium could be limited during disposal (e.g., by limiting disposal thickness), it would not be possible to increase the uranium concentration sufficiently along a 1 D flow path to pose a criticality safety concern. Much uncertainty exists in the estimates of the time frame for the increase in uranium concentration, but analogs from soil-forming processes suggest that these processes can require thousands ofyears.
Volume 1 of this report provided the following recommendations for consideration during a license review of LLW facilities having uranium blended with soil:
- 1. Minimize the factors that enhance the increase in the concentration of uranium. For example, reduce water infiltration, dilute the *U by reducing the enrichment, and minimize opportunities to create zones of reducing potential that precipitate uranium readily (e.g., by maintaining unsaturated conditions, and avoiding organic matter in waste cells to prevent methanogenesis).
- 2. Limit the area; density of uranium to a safe value by limiting the licensed depth of the disposal cell and the licensed disposal concentration. Results suggest that criticality safety concerns can be reduced or eliminated even under worst-case hydrogeochemical transport by reducing the disposal cell depth.
NUREG/CR-6505, Vol. 2 4
.. _ _ _ _ _ _ _ . __ _ . . _ _ ~ . _ _ - _ . _ _ _ _ _ _ _ _ . _ _
3 SITE DESCRIPTION The parameters used in the models are based on site conditions at the CNSI LLW disposal facility in ,
S.C., although not all physical and chemical conditions were explicitly modeled. A variety of geologic hydrologic information on the site is available from previous studies (e.g., Weiss and Colombo,1979; ;C 1982; Dennehy and McMahon,1987; and data provided by CNSI). '
He Barnwell facility was opened by CNSIin 1971. The facility receives approximately 8490 m' (300, to 11320 m'(400,000 ft') of LLW per year. Barnwell receives Class A, B, and C waste. The majority SNM is contained in Class A waste. These waste classifications are defined in 10 CFR 61.55.
About two-thirds of the weste disposed at the Bamwell facility comes from nuclear power plants, with the othe j
third of *U.coming The S from other industry and govemment sources (such as the U.S. Army). The SNM con U enrichment is typically less than 10 wt % for trenches with enrichment data available.
Approximately 90 wt % of the SNM is dry active waste. Other waste includes resins, which have been from the dry waste since the mid-1980s. Scintillation vials and other organic liquids were banned in 1979.
Originally, waste was received and buried at the Barnwell facility in containers, such as cardboard boxes and drums. During tl e 1980s, shipments more commonly came in plywood boxes. Then disposal requireme '
became more stringent, and steel drums (then high-density polyethylene containers) became the standard in mid to late 1980s. These containers are now overpacked in concrete vaults or cylinders. The containers are ,
disposed in trenches that today are typically 30,480 cm (1000 ft) long,6,096 cm (200 ft) wide, and 762 deep. Smaller trenches were used in the past and are used today for high-concentration radioactive waste (Class C).
He trenches are excavated through the surficial sand into clayey sand. Prior to waste placement, buffer sand is placed on the bottom of the trench. The bottom of the trench is sloped, and a French drain is ur M to move water i
to sumps, where it can be sampled for detection of contaminant movement (Fig. 3.1). In the past, the French i drains were constructed on one side of the trenches and consisted of gravel. In the late 1980s to early 199I slotted plastic pipes within high-permeability material were used to collect water on both sides of the trench.
When the trenches are full, they are ba:kfilled with sand, then covered with earthen caps. New high-dens polyethylene caps have been emplaced on some older trenches to further inhibit water from infiltrating the trenches.
From January ~ 1970 to 1984, there were 12 amendments to the original SNM disposal specifications that '
influenced nuclear criticality safety. The original license included a 200-g possession limit for all SNM with no package limit. The license was amended to increase the possession limit and included package limits of 15 g 50 g. Later amendments included mass and spacing limitations for accumulations ofpackages based on 2
disposals from specific generators. In 1981, an areal density limit of 200 g/ft was imposed. In addition, the waste form requirements and practices varied over time until 1984 when the 10 CFR Part 61 criteria were implemented.
l NUREG/CR 6505, 5
Vol. 2
cT E'
w a o e a
h 3.
a sp* Trench standpipe (number varies with trench) i Sump Sump 1 -
4e 4p Surface sand N
1 m French drain '
4 25 ft Buffer sand "
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- r Compacted clayey sand
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h.f$$)fbM' Undisturbed clayey sand # IjjM' i Figure 3.1 Typical construction of waste trenches. Graded bottom is lined with capture water for French drain. l.ocation of drains varies. Some typical dimensions are shown. Taken from Dennehy and McMahon (1987) { ,
g.
b .!
1 l
o a Section 3 Site Description CNSI's last Radioactive Material License (No. 12-13536-01) with the NRC allowed 350 g of"'U per package.
The 350-g limit pertains to "'U, which represents the vast majonty of the SNM waste. Other SNM, such as "'U and isotopes of plutonium, make up less than I wt % of the total grams of SNM. Shipments commonly contain less than 100 g per package. Any shipment with I g or more SNM is reportable. Shipments of SNM are presently placed only at the bottom of the trench if they contain at least 30 g of SNM. In the past, configuration of the SNM in trenches was determined by " operational randomness"; that is, packages of waste containing were not placed on top of packages already emplaced with SNM. These blocks were typically 304.8 cm (10 ft) =
304.8 cm (10 ft) x 365.76 cm (12 ft) in size, then later 762 cm (25 ft) x 762 cm (25 ft) x 1524 (50 ft).
Disposal records provided for this project by CNSI vary in detail. Specifically, the records for Trenches 1-35 (herein refen ed to as the "old" trenches) provide only total SNM mass, whereas the records for "new" trenches (38-87), specific isotopes of uranium have been identified, so grams of"'U are available and average trench enrichment can be calculated. No data on disposal amounts were provided for Trenches 36 and 37. The dispo record data indicate that, in general, smaller quantities of SNM were disposed ofin the "old" trenches. Most trenches have less than 40 kg, with masses ranging between 0.5 kg and 175 kg (Fig. 3.2). In the new trenches, disposed quantities of"'U range fromjust a few grams to 1600 kg, with most trenches containing less than 300 kg. (
r The Barnwell facility is located on the Atlantic Coastal Plain. The surficial deposit at the site is known as the Tobacco had Formation (formerly the Hawthorn Formation). The deposit is approximately 1828.8 (60 ft) to 2438.4 cna 180 ft) thick, and contains dominantly a sandy clay [e.g.,85 wt % quartz (Pietrzak et al.,1982)] with !
coarse sand occasionally present near the base. The present-day water table is within this deposit, typically around 1066.8 cm (35 ft) below the land surface. Beneath the Tobacco Road Formation is the Dry Branch Formation (formerly the Barnwell Formation), which 4 is a massive medium-grained sand. The permeability of the surficial deposit varies from 3 x 10 to 2 x 10 cm/s (Cahill,1982, p. 38) based on laboratory tests on core collected in the region from a variety of depths. Dennehy and McMahon (1987, p. 28) analyzed shallow cores near experimental trenches and found a permeability range from 7 x 10 to 7 x 10' em/s. Field measurements i using slug tests tend to the upper end of the hydraulic conductivity (Cahill,1982, p. 38). Porosity of the deposit is estimated to be around 40%. Effective porosity typically is somewhat lower, around 30%, but the porosity of the waste matrix could be higher, up to 40 to 50%, as reported at other waste sites (Spalding,1987).
Soil moisture above the water table is typically high due to the humid, wet climate. Cahill (1982) repons soil moistvre measurements made over 1.5 years, with values typically greater than 90%. The annual rainfall is about 114.3 cm (45 in.) per year, but only 35.56 cm (14 in.) to 43.18 cm (17 in.) per year is expected to infiltrate the regional flow system (Dennehy and McMahon,1987). Estimates ofinfiltration in the disturbed area around the trenches with earthen caps have not been reported but may be higher due to runoff from the caps.
Groundwater velocity in the trenches has been estimated through tracer tests and groundwater modeling.
Dennehy and McMahon (1987) constructed experimental trenches similar to waste disposal trenches and monitored water levels, soil moisture, and a salt tracer to estimate groundwater travel times from and within the trenches. Salt granules (Nacl) were placed at the bottom of the experimental trenches and at or near the land surface. Detection of the tracer in monitoring points was used to estimate the velocity of 3 x 104 cm/s in the cap and around 4
6 x 10-' cm/s in the backfill material. Cahill (1982) estimated a similar lower vertical velocity of !
2.5 x 10 cm/s in a regional groundwater flow model in the area. Both the upper and lower ranges of values were used in modeling here.
NUREG/CR-6505, 7 Vol. 2
f m r
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h U-235 Inventories in the Barnwell LLW Facility E
8;
{
8 3-
- - 8-S- Trenches 1-35
& m - g@-
c Trenches 38-87 E $
=
a li 3e - t y _ o m -
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m 0 200 400 600 800 1000 0 200 400 600 800 1000 kg SNM kg U-235 Figure 3.2 IIistograms of source material in trenches. Grams ofSNM or 2"U per trench are given for (a) pre-198 trenches, SNM reported, and (b) post-1981 trenches, '"U reported. Older trenches have lower source terms m
2 C.
8 -
w
Section 3 Site Description f
. Monitoring of the sumps and other wells surrounding the waste facilities has detected a tr 914.4 m (3000 ft) in length. Tritium is likely to travel at the velocity of groundwater, being .
and provides an early waming system ofleakage in the trenches. Tritium was first detected o!
in 1978 and measured in an off-site well in 1990 along the base of the Tobacco Rol were installed over the oldest trench area first to help minimize future migration of tritium. Ichimur ,
(1994) report that the travel time of the tritium is similar to estimates of velocity in the fast ho!
of 1982).
I x 10" cm/s beneath the trenches. Cobalt-60 and organics have also been detected be Trench water chemistry reported by Weiss and Colombo (1979) is presented in Table 3.1. Alt i chemistry is variable, it indicates moderately oxidizing conditions. However, this does no zones will not form. Zonation of redox species in landfills has been reported elsewhere, and anal other sites are useful to consider here.
Table 3.1 Concentration of dissolved nonmetals and metals in trench water samples taken a!
the LLW burial site near Bamwell, S.C. (Weiss and Colombo,1979) i Dissolved Trenches component (mg/L) 3 5 6 8 25/21' Total alkalinity (as CACO3 ) 100 200 40 600 80 Inorganic carbon 24 -
11 130 38 Dissolved organic carbon (DOC) 7 -
2 170 12 Chloride 7 10 90 85 42 Nitrogen (N)(ammonia) 0.3 -' l.4 59 25 Nitrogen (N) (NO[ + No[) <0.04 <0.1 23 8.0 15 Silica 4.3 7.6 5.8 6.0 5.0 Sulfate <5 7 18 34 56 Total anions (meq/L) 2.3 4.4 4 16 5.7 Calcium 4.0 3.2 16 34 21 fron 0.15 1.5 0.4 1.2 0.2 Magnesium 2.5 3.3 1.0 18 3.3 Manganese 0.24 0.34 0.45 0.72 0.32 Potassium 1.0 4.6 1.4 12 3.5 Sodium 2.3 20 29 87 37 Total cations'(meq/L) 0.55 2.4 2.3 12 4.8
' Trenches 25 and 21 are reponed together.
- Insufficient sample for analysis.
' Includes nitrogen as .NH[
NUREG/CR-6505, 9 Vol.2
i Site Description . Section 3 i
Relatively few studies detail the redox conditions in landfills because of the problematic nature of obtaining reliable measurements. One of the main difficulties is that not all redox couples are in equihbrium, so a given !
measurement may not be relevant to all redox couples (Lindberg and Runnells,1984). A series of redox l reactions may occur in zones around landfill leachate (Baedecker and Back,1979; Christensen et al.,1994). The !
zones can be identified by pattems in water chemistry, such as concentration of redox couples, evolving from the f aerobic zone, to nitrate-reducing, iron-reducing, sulfate-reducing, then methanogenic (Fig. 3.3). The range of !
redox variation in landfills is quite large and controls the mobilization and concentration of many redox sensitive - i i
species. The observed values range from -200 to +600 mV over distances of hundreds of meters. Similar redox variation has also been observed in natural systems (Champ et al.,1979) on the scale of hundreds of meters to !
I kilometers.
Although the zones may be extensive horizontally, they are often quite thin vertically, limited by the lack of }
vertical mixing in the landfill plume. These zones develop over decades, but may initiate on the order of 5 years !
[e.g., the Bemidji spill (Baedecker et al.,1993)]. Some of the key factors in the development of zones are the !
mineralogy of the sediments [in panicular, availability ofiron minerals, according to Heron et al. (1994)], the i organic carbon content, microbial degradation rates, and the moisture content. The size of the different zones !
also depends on the landfill size, venical mixing, and rates of groundwater flow.
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Vol. 2 10 l l
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Section 3 l Site Description i
Distance from landhil(m) 0 100 200 300 400 f f
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l331 AEROBIC 22 Figure 3.3 Example of redox zonation in a landfill eross section. Five zones from methanogenic to aerobic were mapped in multilevel piezometers, B1 through B9. Reducing zones extend over 350 m from landfill source. Adapted from Christensen et al.,1994 NUREG/CR-6505, 11 Vol. 2
\
4 APPROACH 4.1 NUCLEAR CRITICALITY EVALUATION 4.1.1 Code Description and Validation :
i The SCALE (1995) code system was used to calculate the k,,of the designated system factor of a system. Problem-dependent processing of the cr resonance self-shielding is performed using the NITAWL and BONAMI codes. For this study,;I code was executed by the CSAS module to provide the k,, values. XSDRNPM is a deter the Boltzmann equation for neutron transport in a 1 D mathematical system using a!
SCALE was used because ofits historic and recognized success in the performance of benchm !
applications analyses for licensing activities. i The stationary system of the SCALE codes used for this study and validation, CSAS, BONA XSDRNPM, and KENO V.a, were created May 30,1995. ,
The Brookhaven Evaluated Nuclear Data File B Version V(ENDF/B-V) point cross-section library, which was collapsed to a 238-ne (Greene,1994), named REF01.XN238, was created May 26,1995, and resided on the s the SCALE suite of codes during the period of this study. The 238-energy-group library w currency of evaluation, testing, and benchmarking. He hardware platform, the SCALE comp and the 238-energy-group library used were validated through the computation ofverification i benchmarks involving '"U systems before and after the evaluations performed for this stud ;
and validation benchmark calculations provided identical results for calculations peri those performed after the study, thereby demonstrating the stability of the software and d study. The bias and uncertainties of the benchmark calculations were within -0.5% of !t I that is, the calculated k,, values of the 14 critical experiment benchmarks were between 0.9 Vol. I for more details. . .
4.1.2 Analytical Approach !
i The analytical approach taken for the nuclear critical evaluation was performed in two segments <
segment was to evaluate the infinite-media multiplication constant, k.,2 of a fixed-density SiO so differing degrees of 2"U and water contents or densities within the soil. These results provided combinations of zuU, soil, and water that could support self-sustaining nuclear fission chain reactio i essentially infinite sea of material (i.e., k 2 0.95). He second segment involved examining threel that have relevance to the evaluation: spheres, cylinders ofinfinite length, and slabs ofinfin . .
i In Fig. 4.1, the dimension r + 4 m refers to the determined critical radius plus 4 m of uncontaminal water, and the dimension h + 8 m refers to the thickness of the determined critical slab 2plus 8 m of SiO al 2 2 water. He evaluations of the infinite slabs approxiniate the effects of the "U, contaminating the soil-like settling vertically performed for reviewing onto LLWafacilities.
waste-cell floor and are consistent with previous evaluations (Hopper e 4.1.3 Parameters Consultations among Oak Ridge National Laboratory (ORNL) and U.S. Nuclear Regulatory C staff, evaluating the CNSI disposal records, permitted the inference of a representative uranium enrich i
NUREG/CR-6505, I 13 Vol. 2
El k "e n Infinite Media Model Spherical Model
!.l 6
8 r+4 w =
w i w
i a f i
Infinite Cylinder Model Infinite Slab Model _
/
q ) ._ n+8m r+4m y m i Figure 4.1 Conceptual configuration for nuclear criticality evaluations .
1J
Section 4 Approach
{
the nuclear criticality computational evaluations. Though CNSI has been licensed to enrichment of uranium, the assumed arbitrary representative uranium enrichment for the compu evaluations was 10 wt % 2"U in uranium.
He nuclear criticality computational studies that are reported in Vol. I were performed for 100 2 uranium, U(100), and water in one of two hypothetical waste matrixes. Both waste matrixes with or water had identical bulk densities (i.e.,1.6 g/cm') and equivalent 0.4 void fractions. The basic m
" Nominal Soil (N-S)" consisted of" average" weight fractions of earthen elements and the matrix "SiO:-Soil, (S-S)" consisted of 2only SiO . The nuclear criticality computational results in Vol.i that the ratio of critical uranium areal densities for the S-S matrix, divided by the critical uran for the N-S matrix, was generally on the order of 0.7. The S-S matrix was chosen for t 2
conservative may be more likeestimates S-S than N-S. for the lesser uranium enrichment of 10 wt % "U. Furthermore, the sand a The reported results from this study are for various oncentrations 2 of 10 wt % "U in uranium\
in a 1.6-g SiO/cm' hypothetical waste matrix as it may relate to the LLW facility operated by C Bamwell, S.C. The results are presented in tabular and graphic form, followed by discussions abou relevance of the results to practical initial disposal conditions. Some compansons are drawn betwe U(10)-H 0-SiO 2 and previous U(100)-H 2 0-SiO 2 computational results that were reported in Vol. .
The lowest 2"U concentration in the nuclear criticality evaluations is the concentration of 2"U used in Vol. I of 2
this study, which was the permissible State of Utah license limit for "U. Although this concentratio1 relevant to disposal as containerized waste, it is below the level of concem for nuclear critica be Vol.1. used as a staning point. However, it is no longer a reference point for a concentration factor, as 4.2 HYDROGEOCHEMICAL MODELING 1 4.2.1 Conceptual Model As stated previously, a process-oriented model was developed to evaluate hydrogeochemical mechan would increase the concentration of uranium in disposal settings. The model was not intended to p site-specific predictions, but data (where available and applicaPC from the Bamwell site were used in th model. The' specific process modeled was the reduction and precipitation of oxidized uranium in The interplay between reducing zones and oxidizing water that infiltrates requires a coupled trans geochemistry model because, instead ofjust a single component (uranium), the transpon of multi (e.g., uranium, oxygen, complexing agents, competitive electron acceptors such as Fe'*) is involved.
His study focused on an increase in concentration of uranium at a hypothetical boundary between oxidize reduced zones, rather than the development of zones. The reducing agent was assumed to be either elemen iron, which represented the 55-gal drums, or methane (CH.), which represents the organics contained in the waste and the cardboard, plastic, or wood containers. This study did not evaluate kinetic aspects or time varia infiltration.
One-dimensional (1-D) transport of uranium through the trench was assumed. Even though three-dimensionaj (3-D) transpon is more realistic, dispersion will reduce the concentration of uranium transported fromi to another, as shown in selected runs. By using 1-D flow, the results of the modeling will be conservative; the i
travel times will be shortest, and the concentrations will be maximized. Transport pathways that would mimic l NUREG/CR-6505, 15 Vol. 2
Approach Section 4 !
vertical flow through the trench and horizontal flow along the drainage systems can be modeled with two different 1 D legs.
In summary, conditions that tend to enhance the potential for increasing uranium concentration were modeled, but only ifjudged to be within reasonable bounds, or the limitations could be specified. A detailed description of model assumptions and further discussion are provided in Sect. 5.
4.2.2 Models Used Preliminary hydrogeochemical modeling was conducted using two codes: PHREEQC and ParSSim. PHREEQC (Parkhurst,1995) is a chemical speciation code that models 1-D transport using mixing cells. It has a fairly complete geochemical database, but neglects transport effects such as dispersion, which can reduce concentrations. ParSSim (Wheeler et al.,1997) runs on a supercomputer and incorporates full transport and user-dermed geochemical reactions in a multidimensional, multispecies transport code. This code permits more realistic simulations, but is also less stable numerically and more time-consuming to use (in terms of run time and output analysis). Not all cases that run for PHREEQC were run successfully with ParSSim. However, a representative 3-D problem has been run successfully with ParSSim, which provided some error bounds on the simplified PHREEQC modeling.
Both codes were tested by comparison with a field and modeling problem involving nitrate removal by oxidation of pyrite, which creates a sharp redox front (Engesgaard and Kipp,1992). Numerical methods in the codes were selected to help code stability over the large concentration ranges resulting from redox problems (Toran et al.,
1997b).
4.2.3 Parameters and Model Grid !
The model grid represented a 1-D flow field. A continuous input of oxidized water containing dissolved uranium was introduced into a reduced zone and allowed to flow through the reduced zone at velocities of I x 104 and '
1 x 104cm/s. The model grid was 5 m long and represented a reduced zone (Fig. 4.2). The 5-m length was selected arbitrarily to monitor movement of the reduction boundary during the injection. In most cases, the i boundary stayed within the model grid. The grid began at the redox boundary because this is the location where '
most of the geochemical alterations are presumed to occur. This approach neglects some interactions within the oxidized zone in the natural flow field, but these interactions are likely to be less important than the reactions ;
taking place at the redox boundary. ,
The oxidized water assumed in this modeling exercise was geochem;cally similar to trench water at the CNSI site .
(Trench 25/21. Table 11) with an assumed (rather than measured) eoncentration of uranium. The uranium was <
input as a dissolved species. The uranium concentration was varied in the different runs, with values ranging I from 1 to 20 mg/L. The initial uranium concentration is limited by mineral solubility. For example, in this water, dissolution of schoepite results in an equilibrium concentration of about 20 mg/L (based on PHREEQC modeling). Furthermore, tens of mg/L of dissolved uranium has been observed in water running off of uranium mill tailing piles in oxidizing environments. A lower limit of I mg/L uranium was selected as the input concentration for most runs, a reference concentration used in previous studies (Sims et al.,1993). Infiltrate '
water in the model column is also represented by the geochemistry of Trench 25/21, with a low uranium concentration of 0.01 mg/L. ,
~
NUREG/CR-6505, Vol. 2 16 d
Section 4 Approach 1 i
A
, Redudng Zone i Oxidized j \
U Bearing H20 / /
4
- 8. C. /
Redudng Zone Redudng /
Zone 0
/ /
0xidized 0xidized !
U.8eadng U-8dadng M0 !
20 ,
Figure 4.2 Schematic of model grids showing oxidized injection into reducing zone.
A. One-dimensional model used in PHREEQC and 1-D ParSSim evaluation. B. Three- i dimensional ParSSim model with small source term, allowing dispersion in three dimensions.
C. Three-dimensional ParSSim model with larger source term, limiting 3-D dispersion along centerline of plume NUREG/CR-6505, 17 Vol. 2
The oxygen content of the trench water is not well known; Eh and dissolved oxygen were not reported in CNSI monitonng data or the hydrologic summary of Cahill (1982). A few (5) dissolved oxygen measurements were reported by Weiss and Colombo (1979), ranging from 0.1 to 1.5 mg/L. Calculations of the redox potential of the i Trench 25/21 data also indicate an oxidizing Eh (on the order of 400 mV). The undersaturated conditions of the l trench also suggest dissolved oxygen is present in the trench water. A base-case value of 2 mg/L dissolved l oxygen was selected, and the value was varied in the sensitivity analysis (see below).
l Ten components from this background water were selected to form the basic components for the geochemical l modeling (Table 4.1). From these components, equilibrium species form and mmerals can be selected for equilibration. For PHREEQC, the database contains an extensive list of complexed species. For ParSSim, the user identifies key complexes for equilibration, and a more limited number of complexes is favorable for faster convergence. The complexes selected for equilibration in ParSSim (Table 4.1) were based on PHREEQC modeling, which indicated those complexes formed in significant concentrations (at least one order of magnitude greater concentration than other complexes of the same component).
Table 4.1 Components and reactants (minerals and aqueous complexes) used in ParSSim" Components mg/L Products log (K,) Minerals / Phases log (K,,)
H' 5.9 (pH) OH- -14.00 FEMETAL -84.0 C O3 80 H 2% -44.67 CH, -130.9 Ca 21 HCO-3 10.35 UO2 -27.72 H 2CO 3 16.68 Fe2O3 -'
Na* 37 Mg 3.3 ~ U(OH),% -41.018 Fe(OH)3m Fe** ' O.2 UO 3(CO 3)3 17.00 FeS 3 -'
0 2% 2 (varied) UO3(CO3 )% 9.63 S0l' 9.0 Fe -7.76 U O3 '* 1.0 (varied) FeHCO3 ' -5.76 Cl- 42 Fe(OH),' -5.67
- For PHREEQC components and complexes, see PHREEQC database.
- Used for PHREEQC only. Ksp not calculated for ParSSim database.
' Iron input as Fe in ParSSim equilib ates to Fe
In the models PHREEQC and ParSim, minerals will not dissolve or precipitate unless they are specifically selected for equilibration with the model solutions. Minerals selected for equilibration were uraninite (the reduced uranium mineral) for the case of a carbon reducing agent (CH,), and uraninite, hematite, amorphous iron hydroxide, and pyrite for the case of an iron-reducing agent (FEMETAL). These iron minerals resulted in a solution that was not supersaturated with respect to any remaining minerals; however, other combinations of iron-mineral equilibration could have been used.
NUREG/CR-6505, Vol. 2 18
Section 4 Approach The reducing agents, FEMETAL (Fe') and CH., were selected to represent waste contain FEMETAL represents SS gal drums and other steel containers commonly used for wa for organic matter, such as wood crates used in disposal. The amounts of these reducing agen the model, but were typically less than the molar content of a single banel or wooden crate The FEMETAL was introduced by setting an equilibration constant with respect to elemental oxidation state). The simulations were conducted with FEMETAL undersaturated in the system. W saturation with respect to FEMETAL was assumed, the model predicted a large release of redu system, along with very low pe, and high pH (> 10). More likely, a slower release ofiron occurs, a represented in the model by setting the FEMETAL equilibrium to subsaturated conditions. A sa
-15 produced a rate ofrelease ofreduced iron in solution that was reducing at near-neutra source concentration it isofreduction not known of FEMETAL in the model is well below the total iron conte wouldifbe.
the iron in the trenches would become localized in a zone to form a redo rate e At Bamwell, there are fewer sources of organic matter than at municipal or industrial la practices included wood and cardboard boxes for containers. In addition, small amounts ofscint were included in the waste prior to 1976. Thus there are potential sources of organic matter in the w trenches that could serve as reducing agents or sorbing surfaces to increase the concentration of Organic matter is very important in the concentration and reduction of sedimentary uraniu Landais,1996; Wood,1996). In particular, sources (including tree trunks) are believed to b 1996). Kinetic inhibition of reduction may be overcome by heat or microbial activity. Organic ma common in landfills and is associated with methanogenic reducing zones.
Thermodynamic data are not available for wood or cardboard, so a sunogate species must be u CH, as a reducing agent represents an end-member composition for these materials, alt reducing than wood or cardboard, and is readily dissolved in water. A partial pressure of CH, ga maintained such that there was 2 mg/L in solution.
Precipitation in the reducing zone was controlled by the mineral uraninite, a commonly observed ur mineral in reducing zor,es ofnatural ore deposits. Model runs were typically 500 L ofpore fluid, which also conducted. The longer times were run only with PHREEQC, represe 4.2.4 Sensitivity Analysis A series of runs were conducted to help provide bounds on conditions that could limit or enhan for nuclear criticality safety concems. The parameters selected for evaluation were the initial con dissolved uranium, the concentration of oxygen, the amount and size of reducing zones, additional u analyses were conducted with PHREEQC, A more limited numb and size were done with ParSSim (Fig. 4.2). The range in parameters was selected to span likely fie rather than worst-case scenarios. Thus the uranium concentration was varied from 1 to 20 mg of observed values around waste sites); the oxygen concentration varied from 1 to 8 mg/L the solubility limit for oxygen dissolved in water); the leachate complexes were chosen based on ob for the anions nitrate and fluoride at the Bamwell site.
\
NUREG/CR-6505, 19 Vol.2
5 ASSUMPTIONS All models are simplifications ofreality. Some of the processes not explicitly modeled in the simulations are evaluated here. The scenarios modeled typically represent bounding calculations.
- Saturated, steady state flow The models assumed saturated, steady-state flow to approximate conditions over the long time frame. These conditions represent worst-case scenarios because unsaturated, transient flow likely involves longer trave
- Geochemical complexes and solid phases The geochemical complexes and solid phases that were selected for equilibration were based on a important uranium species and what is known about site chen,istry. The decision was made to focus on the reduced form of the uranium (i.e., uraninite). Reactions among species were assumed to be in equilibrium Although additional complexes and phases could change the mobility of reactants, and kinetic considerations could change the rates, the calculations represent likely bounds and best available data.
e Simplistic deposit geometries Simplistic deposit geometries, having no density gradients, were used in the criticality assessment. That is th
"'U and source material were assumed to be unifonnly distributed over the volume of the trenches. Sm quantities of fissile material in equivalent volumes may be required to reach criticality for certain density gradients. An extreme, but actual, critical experiment performed by Morfitt (1953) was the assembly of five concentric cylindrical uranyl fluoride solution regions having variable densities of 93 wt % enriched uranium.
Solution uranium densities were selected to produce a nearly uniform thermal neutror, core flux. Doing produced equal volume.a critical system with 1061 g "'U as compared with a homogeneous core mass of 1162 g "'U in an
- Enrichment For the nuclear criticality calculations, all "'U enrichment of 10 wt % is assumed. Pased on disposal record (see Section 8 and Table 8.1), this represents an arbitrary upward bound on likely enrichments in the trench.
Because of incomplete records and a lack of regulatory limits on enrichment, a theoretical upper bound of i
100 wt % enrichment is conceivable, but not likely. The case of 100 wt % enrichment was modeled p in Vol.1, and the calculations presented there are applicable to issues of this study. In evaluating individual trenches, the average ennchment was assumed to be appropriate.
Enrichment does not affect hydrogeochemical mobility of uranium. Relative enrichments of"'U are assumed to i
remain constant between the source and the precipitated secondary phases. The geochemical model accounts '
for total uranium, irrespective of the "'U isotopic enrichment. However, because the amount of dissolved uranium associated with the waste can span an order of magnitude from I to 10 mg/L, one esn interpret the 1 mg/L as
"'U/L). The geochemical models deal only with chemical species and not NUREG/CR-6505, 21 Vol.2
Assumptions . Section 5 ;
o Reducing zones It was assumed in this analysis that reducing zones exist within the disposal environment. The development and !
longevity of reducing zones was not specifically modeled in this analysis because of a lack of site-specific ,
information on weathering rates of containers, amount of organics in specific trenches, and 3-D data on chemical l zonation within and around the trenches. Reducing zones also require a water saturated environment, which is !
not currently present in the trenches. Thus the reducing zones represent a worst-case scenario of future !
conditions, which would enhance precipitation of uranium. l
- Containerdegradation ,
For the reasons given above, the degradation of the waste containers was not modeled explicitly in this analysis; all of the source term was assumed to be exposed to migrating fluids. Realistically, degradation is fast relative to !
the time required for significant chemical migration. Steel can degrade in tens ofyears, concrete can crack, and i high-density polyethylene (HDPE) degrades in hundreds of years. !
i
- Horizontal and preferential flow paths !
An important assumption regarding the mobilization of uranium in trenches is that multiple vertical flow paths {
can be funneled into a demobilization zone. This funneling occurs through a horizontal component of flow. !
! Although the areal density disposal limits and the package limits mitigate development of a critical mass along a ;
J venical path, horizontal flow along the base of the trench may result in the concentration of material within the '
{ drainage system or the sumps. In this exercise, two-dimensional (2 D) flow was modeled as a combination of
[
l D segments.
i i
i i
Vol. 2 22 f t
6 CRITICALITY SAFETY EVALUATION RESULTS 6.1 23sU ENRICHMENT INFLUENCE ON CIUTICAL MASS OF 2
2 .
The critical mass of "U is inversely proportional to the "U enrichment of the urani equal (e.g., degree of neutron water moderation, volume, chemical composition, te 2
as the weight percent "U enrichment in uranium decreases, the critical mass of "U 2 figures are taken from an illustrative study (Jordan and Tumer,1992) of this effect rel 2
neutron water moderation, expressed as the ratio of hydrogen atoms to "U atoms (H/X), for fu spheres of homogeneous 2 2 2 UO F -H O solutions. The figures are provided to illustrate t 2 2
mass response to moderation and uranium mass response to moderation at various "U en 2
demonstrates that for a given H/X of 500 the critical "U mass is estimated to be a 2
2 whereas the critical 2"U mass is estimated to be about 1.2 kg "U for U(10). Figure 6.2 d given H/X of 500 the critical uranium mass is also estimated to be about 0.76 kg uranium the critical uranium mass is estimated to be about 10.2 kg uranium for U(10). Thus for a ,
l of the minimum critical masses of 2"U for U(100) and U(10) As can beis about observed in 0.63 (0.7i Figs. 6.1 and 6.2, for water moderation less than an H/X = 500, the 2 optimum for minimum significantly smaller ratios of U(100):U(10) critical masses. For instance, at a poorly moderat 2
U(100), the critical 2 2 mass is estimated to be about 6.2 kg "U, whereas for U(10) the critical '
be about 31.0 kg "U, a "U mass ratio of 0.2 or uranium mass ratio of 0.02. These effects ar the slowing-down power of water for high fission-energy neutrons escaping the neutron re characteristic of 2"U in the U(10). With less 2 water present in the mixture, neutrons cannot be th rapidly as2 when present in optimum ratios of water to "U, about H/X = 500, and therefore the ne captured by "U resonance capture.
6.2 TRENDS FOR U(10)-H2 0-SIO: MIXTURES 2 :
Similar trends for increased uranium and "U mass are present for the mixtures of U(10}-H 0-S 2 2 this report. The obvious differences between the above-referenced illustrative study and this assumption of elemental uranium (i.e., no chemical compounds were considered) and the presence o dioxide at a fixed density of 1.6 2 g SiO /cm'. The presence of 90 wt % "U in the uranium in these ca 2
increased the importance of SiO as a poor moderator of neutrons and a poor capturer ofn 2
in the studies with U(10) having no water, the resonance capture 2ofneutrons by the "U was import suppressed the infinite2 media neutron multiplication factor, k., from 0.955 for U(100) to2 0.713 for U(l concentration of 0.000886 g "U/g 2 SiO . A slight introduction of water to the matrix (i.e.,0.01813 g H 2 2 for both systems decreased the k. for the U(100) to 0.867, whereas the k. was increased for the U This response is the direct result of the SiO affording excess neutron slowing down for the U(100) 2 thereby allowing hydrogen neutron capture to suppress k..
In the case of the U(10), the SiO2provides excess neutron slowing down for the U(10) matrix, but these neutrons are forced to slow in energy 2 into the "U resonance capture region and be captured. The slight addition of water to the U(10) matrix can slow neu large amounts and thereby perrrit neutrons to be moderated or slowed to energies that 2are less than the "U l
l resonance capture energies, the'eby offsetting some of the negative effects of the SiO2 over moderation.
Table 6.1 provides extracted results of the computations to highlight the extremes of this study. Th listing of the results of this study are provided in Table A.1 of the Appendix to this report. The results are provided in the same format that was used in Vol.1, Appendix C. Figure 6.3 provides an interpolated and 2
1 smoothed surface plot for the entical infinite slab areal density vs H:0 and nU concentrations relative to SiO .
2
\
NUREG/CR-6505, 23 Vol. 2 1
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a s Figure 6.2 Critical mass U vs II/X for UO,F,-II,0 in spherical II,0-reHected systems.
n O ' E-A. 4 O< Ch.
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PC v.
M U. c m
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a
, Table 6.1 U(10) plus H2 O plus SiO2 -soil (S-S) results
'"U content Wseer consent critical mfinite slab Criticalinfinite cylinder Critical sphere Line k. or entry 's '"lFem' g'"U/g5 S g H 0/cm' g H,0fgS-S k-inf Dickness Areal density Diameter Linear density Diameter Mass i
(cm) (kg "'U/m') (cm) (kg "'U/m) (cm) (h g '"U) 31 0.00163 0.0010188 0324 0324 0.445 69 0.0024761 0.0015476 0.029 0.01813 0.982 464.22 11.4946 75834 111.84 1020.78 1378.9967 f Ch 110 0.0032722 0.0020451 0.0885 0.05531 1.031 174.48 '5.7093 328 62 27.7534 439.7 145.64911 156 0.0057k12 0.0035714 0.324 0.2025 0.%5 234.28 133872 401.80 72.4544 484 339.22677 162 0.0063 0.0039375 0.819 0.07438 1.222 66.02 4.1593 123.50 7.54681 17432 17.473536 208 0.2817 0.1760625 0.029 0.01813 1.110 13.83 38.9591 90.94 182.% 74 143.62 436.9495 212 0.2817 0.1760625 0.251 0.15688 1334 11.24 31.6631 33.84 253352 53.06 22.0337 213 0.2817 0.1760625 0324 0.2025 1379 9.82 27.6629 28.80 183510 44.% 13.4049 .
a D
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d
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Section 6
, Cnticality Results 6
g H20/g SiO2 1
15
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Siab Area: Density 10 (kg U2.35/m^2) 0 0.003 t 8.01 M scale of(g U239g $102) s.43 1
2 2 l Figure 6.3 Infmite slab areal density (kg "U/m ) vs g H 0/g SiO and for scale of g "'U/
2 2 2
i n
, NUREG/CR-6505, 27 Vol. 2
)
. -.- - ~ . _ . .- . - - _ - - . - . . _ - . . - . .-__ . - . _ _
~_ _ - .
Criticality Resuits Section 6 Figures A.1 through A.]1 provide interpolated and smoothed surface plots of the data provided in Table A.I.
These plots are provided in the same format that was used in Vol.1, Appendix D. In future comparisons, care i
' should be exercised to recognize that even though there are similarities between the general appearances of the .
surface plots of the U(100) and the U(10), the significantly larger "'U values are for uranium enriched to only 10 wt % "'U in the uranium.
6.3 COUPLING OF NUCLEAR CRITICALITY AND HYDROGEOCHEMICAL MODELING ne nuclear criticality safety calculations were used to establish the configurations and associated uranium concentration increases that are required to reach a level of concem. Then, using the hydrogeochemical modeling approach and assumptions discussed in Sections 4 and 5, the time required to increase the uranium concentration from the reported disposal values to one of the minimum concentration levels of concern was detennined. Several minimum values of the critical mass were selected to provide a conservative scenario as benchmarks for evaluating the potential of developing a critical mass of"'U under the conditions of SiO matrix, water, and 10 wt % carichment "'U. 2 De lowest uranium concentration that could support a nuclear reaction was 8
in a nearly dry system: 0.0024761 g "'U/cm' and 0.029 g 2H 0/cm (Table A.1, line entry 69). This concentration requires a slab 464.2 cm thick, or 1379 kg, ifin a spherical configuration. By contrast, concentrations of 0.0224 g "5U/cm' and 0.2817 g "'U/cm' under almost fully saturated conditions would require slabs of 19.7 cm and 9 cm thick, respectively. The hydrogeochemical model assumes a water-filled porosity 40 vol %. If uranium precipitates under wet conditions, then the system would need to dry out in order to ,
achieve the minimal configuration necessary to support criticality. In order to develop the higher uranium '
concentrations, funneling of source material by horizontal transpo:t is necessary.
NUREG/CR-6505, Vol. 2 28 r
+
7 HYDROGEOCHEMICAL MODELING RESULTS Hydrogeochemical modeling results are presented as a comparison of simulations with differen The evolution of uranium precipitation over time is used as the basis for comparnon (Table 7.
modeling is conducted in terms of pore volumes, which can be translated into yi its by assuming velocity.
4 A pore volume in the model was assumed to be 1 L of water. Using an assumed g of 10 crrvs (1 m/ year), most simulations were conducted for 80 years (500 pore volumes) with selected simulations up to thousands ofyears. Relevance of the modeling to site conditions at Bamwell is d Section 8. !
7.1 REDUCING ZONES ,
1 The reducing zones maintained a low redox state over the duration of modeled influx of oxidii 30,000 L for the longest simulation; typically 500 L). The CH. created a reducing zone that had a with a pH of 6, whereas FEMETAL (Fe') affected a reducing zone with a pe of -1.9 and a pH of 6. Th release of the reducing agents was on the order of 25 mg/L per year for methane (Fig. 7.1) for FEMETAL (a small fraction of a barrel). The values are derived from the modeI meant to represent the actual release rates in a disposal setting. Rese modeled release rates are not like' consume the supply of wood from dispo< l crates or metal from weathering dmms. However, the kinetics o weathering of these disposal containers it not well known. Both types ofreactions could be media bacteria, providing accelerated weathering rates and formation ofreducing zones.
The reducing zones as modeled capture very close to 100% of the uranium input (e.g.,99.998% fo producing a linear increase in the amount of uranium precipitated over time (Fig. 7.1). One rea zones are efficient at capturing uranium is that reduced uranium minerals, such as uraninite, have Another important factor is that there were modest amounts of complexing agents present in the observed water.
7.2 SENSITIVITY ANALYSIS For the conditions simulated, the most significant parameter in predicting the amount of uranium pr waste.the initial source of uranium or the dissolved concentration associated with wastewater filtering was The uraninite precipitation was directly related to the initial concentration of uranium, and complete pr occurred for initial concentrations up to 20 mg/L (maximum modeled). The uraninite precipitation was not sensiti've to the oxygen concentration in the infiltrating water, which has the potential to compete with uran for electron acceptor sites. The concentration of oxygen dissolved in water can vary only over a small ra it is not a sensitive parameter. The model predicted essentially the same amount of precipitation for cases of half 1
the concentration of the reducing agent. Small, but realistic, amounts of complexing agents were added with no i significant effect. In particular, increasing the bicarbonate concentration from 80 to 400 mg/L did not inhibit uraninite precipitation. De uranyl-carbonate complex is not dominant under reducing conditions. Sensitivity the precipitation of silicate minerals and hydrodynamic dispersion is discussed in the following sections.
l NUREGICR-6505, l 29 Vol. 2 !
Hydrogeochemical Modeling Results . Section 7 Table 7.1 Parameter variation and results of hydrogeochemical modeling Run No. Amendment Value U PPT CH4 U PPT FE O Base case (one set of runs for each of two See below 2=10' 2=10' reducing agents: FeMetal and CH,)*
I Vary uranium in leachate 2 mg/L 4x10" 4 x 10' 2 10 mg/L 2 = 10 2 x 10
3 20 mg/L 4 x 10 4=10
4 Vary oxygen in leachate 2 mg/L 2=10' 2 x 10' 5 8 mg/L 2=10" 2 x 10' 6 Vary reducing zoce 1 mg/L CH, or 1 M 1.4 x 10' 2 x 10' FeMetal (1.4 x 10')
7 1 cell reducing 2 x 10" 2 x 10' 8 Vary carbonate concentration 240 mg/L HCOj 2 x 10" 2 x 10' 400 mg/L HCO; 2 x 10' 2=10' 9a Vary leachate complexes Add 2 mg/L F 2 x 10' 2=10' 9b Add 20 mg/L NO3 2=10' 2=10' 10 Vary precipitate (ppt in zone of higher Si, ppt. coffmite 2 x 10" 2x10' equilibrated with quartz) 1I ppt. soddyite 1.0 x 10" 12a incorporate transport (dispersion) 1-D transport in ParSSim 1.8 x 10' 12b 1-D, slower velocity 3.8 x 10-8 12c l-D reducing zone halfway 1 x 10' down grid 13 3 D transport with small source 2.6 x 10
area 14 3-D transport v.ith large source 1.3 x 10
area 4
15 Larger dispersivity 2.9 x 10 (1.4 x 10)
' Base Case: 1 mg/L U in teachate Reducing Agent 1: 10 M FeMetal (equilibrated, but undersaturated to create a slow release)
Reducing Agent 2: 2 mg/L CH , constant source 2 mg/L 0 3 80 mg/L HCOj 5 cells reducing Equilibranon minerals: uraninite, also for FeMetal: pyrite, hematite, Fe(OH)3 U PPT = g U-mineral precipitated per em' soil using CH, or Fe' as reducing agents.
For runs with number shown in parentheses, the run did not go to completion. Value should be compared with value shown in parentheses for equivalent timestep.
NUREG/CR-6505, Vol. 2 30
e Section 7 Hydrogeochemical Modeling Results f ou l
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(b) l Figure 7.1 Reaction progress for hydrogeochemical model: (a) Cumulative j CH release vs time for short time frame; (b) Cumulative uraninite precipitation for long ;
d time frame. Times shown are based on a velocity of 3 x 10 em/s and a porosity of 0.4 i NUREG/CR-6505, 31 Vol. 2 I I
Hydrogeochemical Modeling Results Section 7 7.3 PRECIPITATION OF SILICATE MINERALS Some alternative uranium mineral precipiutes were considered by including silica in the model. The silica in the infiltrating water is expected to be relatively low, given that the source area contains waste rather than soil.
However, when the leachate reaches the drains at the base of the trench they will encounter a zone of backfill containing sand. This is likely to result in water in equilibrium with an amorphous silicon dioxide such as chalcedony. The scenario modeled was that of water infiltrating into a zone in which chalcedony dissolution occurs, increasing the silica concentration up to about 6 mg/L. Precipitation of uranyl silicates was modeled with and without reducing agents present. When the uranium infiltrates this zone, uranium silicates precipitate. The minerals considered in this analysis were soddyite, an oxidized uranium silicate, and coffinite, a partially reduced uranium silicate. The soddyite does not completely precipitate all of the available uranium because it has a higher solubility than uraninite. The coffinite captured the uranium as efficiently as the uraninite.
7.4 3-D HYDROGEOCHEMICAL MODEL The main error induced by using a mixing cell model, such as PHREFQC,instead of a coupled geochemistry and transport code, such as ParSSim, is that dispersion is neglected in the mixing cell model. Dispersion is a function of the dispersivity and the velocity. It can occur in three dimensions, although the two transverse directions dispersivity tends to be smaller in the transverse directions rather than the longitudinal direction (in the direction of flow). The effects of dispersion on transport and reconcentration were evaluated and compared in a variety of ways: (1) 1-D vs 3-D dispersion, representing an infinite source and a point source; (2) dispersion along the flow path before the reducing zone is encountered; (3) longitudinal and transverse dispersion (ratio 3:1) after release from a 1-m by 1-m source area; (4) longitudinal and transverse dispersion after release from a larger source,3 m by 1 m; and (5) fast vs slow velocities. These runs were conducted with ParSSim, using the same background solutions and the same equilibration minerals as in the CH, reducing zone case.
The results (Table 7.1) indicated that 1-D dispersion reduces the amount of uraninite precipitated, but that the amounts are reduced less than an order ofmagnitude, assuming typical scaled values for dispersivity (0.2 m).
The errors due to neglecting dispersion in one dimension are relatively small compared with other uncertainties, such as the kinetics of weathedng, mineral dissolution, and precipitation, and the size and distribution of source term. Bree-dimensional dispersion results in a decrease in the amount of uranium precipitation of approximately an order of magnitude. His decline in concentration ofprecipitated uranium indicates that isolated sources of uranium that can disperse in three dimensions (Fig. 4.2) will require (1) much longer times to reach levels of concern for nuclear criticality safety, (2) larger source terms, or (3) may never develop concentrations sufficient to pose a criticality safety concern. Table 7.1 indicates that the model results are not highly sensitive to certain parameters [e.g., velocity (Run Nos.12a and 12b), source term (Run Nos.13 and 14) and flow path length (Run Nos.12a and 12c)].
NUREG/CR-6505, Vol. 2 32
. = _ ~. - . - _ - - . . .. _ _ _ . . _ - . . -_ . -
. 8 DISCUSSION Several conditions restrict the potential for a criticality safety concem in specific disposal trenches Bamwell LLW disposal facility. Among these are the enrichment, the source term, the chem the trenches, and the potential flow paths. Because of uncertainties associated with the disp ,
uncertainties associated with the results of this analysis need to be underscored. The basi precipitation in reducing zones, does not appear to be a limiting factor based on the curre However, long times may be needed. Uncertainties and other mitigating factors are also discussed 8.1 ENRICHMENT The average enrichment of homogenized material within each trench is available only for "r ,
38 through 87. For the trenches with enrichment data (Fig. 8.1), the average "'U enrichmeul only six trenches have more than the minimum I wt % "'U enrichment needed to reach a :c Trench 66 has a reported "'U enrichment of 100 wt %. It is judged from historic and currel and data (see Section 3 and Appendin Fg that SNM and source material disposals are comingle Because of this comingling and dispersion of SNM, very few of these trenches need to be conside criticality safety concems. Again, this is based upon the reasonable assumption that "'U and nonfis behave in a similar geochemical fashion.
For the older trenches (trench numbers less than 38), only grams of SNM (Table 8.1) were repo uranium is the most common SNM, to simplify the calculations, it was assumed that 2 all of the SNM though minor amounts of plutonium may have been disposed ofin the trenches. No data were avail amounts of nonfissile uranium, so again the conservative assumption is that the uranium is all "'U and the enrichment is 100 wt % (Fig. 8.1). Two of these trenches have no reported SNM, which leaves 35 t uncertain enrichment.
Most of these trenches are unlikely to have high enrichments, although Trench 23 is an exception. The response to a specific NRC staffinquiry resulted in the identification of relative qu proportions of SNM, source material (SM) and byproduct material activity (BPM) that were placed in The reference to and summary of that response are provided in Appendix B, Suberiticality Evaluatio Chem-Nuclear Systems, Inc., Trench 23.
8.2 SOURCE TERM The minimum mass of"'U to chieve criticality was calculated and compared with the total mass each trench. The minimum mass for a sphere provides a lower bound on source calculations, and these ar shown in Table A.I. For slabs, the minimum mass for the source can be calculated assuming the l side of an " infinite" slab is ten times the calculated critical thickness. These finite volumes are critical uranium concentration to give total mass required for a pctential criticality concem. These uranium values were used to determine the time frames needed to achieve threethe selected minimum uranium d concentration to achieve criticality (0.0024 g/cm'), approximately 10 times that density (0.022 g/cm'), and ;
approximately 100 times that density (the maximum calculated in Table A.1,0.28 g/cm').
For spheres, the critical masses range from 3 kg to 2000 kg over these density ranges (Fig. 8.2). For sla critical masses vary from 20 kg to 6000 kg (Fig. 8.2). With the exception of Trench 66, the "new" trenches that contain SNM with "'U enrichments greater than 1 wt % have less than 30 g of"'U (Table 8.1), which is well below the minimum critical masses calculated. Trench 66 has a reported source term of 2300 g of 100 wt %
NUREG/CR-6505, 33 Vol.2
Discussion , Section 8 !
t o '
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I NUREG/CR-6505, '
Vol. 2 34 l
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Discussion l Table 8.1 Disposal records from Barnwell. S.C., and calculated enrichments and density
SNM mass "5 !
U "'U "5 U Enrichment Disposal No. (e) ~ (e) Density' (e) (wt %) volume (ftY 1 1.32e+04' -
e/cm' 2
4.2e+04 1.11e-05 i 3.73e+04 1.1e+05 1.20e-05 3 1.0le+04 l 4 2.3e+04 1.52e-05 4.66e+04 !
1.3e+05 1.26e-05 5 1.71e+04 (
1.8e+05 3.36e-06
- 6. 2.53e+04 . l 2.4e+05 3.66e-06 7 3.35e+04 f 8
2.4e+05 5.00e-06 2.45e+04 j 9
2.0e+05 4.29e-06 4.10e+04 l 1.9e+05 7.47e-06 10 2.91e+04 l 2.0e+05 5.18e-06 11 2.91e+04 1.5e+05
)
12 1.73e+04 6.8ie-06 2.0e+05 3.09e-06 13 1.49e+04 14 2.le+05 2.50e-06 2.46e+04 {
- 2.2e+05 3.95e 06 1
15 3.43e+04 l L 2.8e+05 4.32e-06' )
16 1.62e+04 2.6e+05 2.20e-06 17 1.30e+04=
1.8e+05 2.57e-06 18- 1.78e+04 -l 1.7e+05 3.63e-06 19 2.51e+04 l 2.4e+05 3.69e-06 20- 1.07e+04 !
2.4e+05 1.61e-06 21 1.42e+04 .!
. I 22 8.0de+04 9.2e+05 3.10e-06 23 1.75e+05 9.Be+05 4.91e-06 24 5.01e+02 2.6e+05 6.73e-08 l 25 1.42e+04 2.7e+05 1.83e-06 26 7.31e+04 !
8.9e+05 2.90e-06 !
27 1.44e+04 2.le+05 2.39e-06 j L 28 8.30e+04 8.9e+05 3.29e-06 !
l: 29 7.31e+04 1.2e+06 2.24e-06 f 30 6.86e+04 ' ,
4.8e+05 5.07e-06 1 .- 31 1.39e+04 2.5e+05 1.93c-06 32 5.56e+04 7.le+05 2.76e-06 '
33 7.60e+04 6.6e+05 4.07e-06 34 3.78e+04 ,
L 5.4e+05 2.47e-06 !
35 7.04e+04 7.2e+05 . 3.47e-06 i 36 '
r NUREGICR-6505, !
-35 Vol. 2 !
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Discussion Section 8 Table 8.1 (continued)
SNM mass 22'U 2U 2"U Enrichment Disposal Density' No. (c) (c) (c) (wt %) volume (ft')' c/c m '
37 38 9.18e+07 2.59e+05 9.16e+07 0.28 5.le+05 1.79e-05 39 1.10e+08 4.16e+05 1.09e+08 0.38 5.5e+05 2.66e-05 40 9.16e+04 9.16e+04 1.8e+04 41 1.68e+08 4.17e+05 1.68e+08 0.25 5.Se+05 2.55e-05 42 2.93e+08 2.87e+05 2.92e+08 0.10 6.l e+05 1.67e-05 43 3.88e+08 3.02e+05 3.88e+08 0.08 7.8e+05 1.37e-05 44 2.81e+08 3.79e+05 2.81e+08 0.13 7.2e+05 1.85e-05 45 3.42e+03 4.63e+00 3.42e+03 0.14 1.0e+05 1.61e-09 46 1.14e+08 3.66e+05 1.14e+08 0.32 5.7e+05 2.28e-05 47 5.65e+02 5.65e+02 4.le+04 48 1.96e+03 2.78e+01 1.93e+03 1.42 6.4e+04 1.54e-08 49 6.7e+04 50 9.4e+04 51 9.93et07 4.70e+05 9.88e+07 0.47 6.0e+05 2.74e-05 52 7.95e+06 7.95e+06 8.2e+04 53 2.61 e+08 1.78e+05 2.61e+08 0.07 5.3e+05 1.19e-05 54 6.02e+08 1.90e+05 6.02e+08 0.03 6.4e+05 1.05e-05 55 2.13e+02 4.63e+00 2.08e+02 2.17 8.6e+04 1.91e-09 i 56 3.87e+02 3.87e+02 9.9e+04 57 6.30e+03 6.30e+03 8.3e+04
- 58 5.68e+08 3.94e+05 5.68e+08 0.07 7.3e+05 1.90e-05 '
59 60 l 1.le+03 61 9.90e+08 1.76e+05 9.90e+08 0.02 7.7e+05 8.05e-06 62 5.14e+08 7.27e+05 5.13e+08 0.14 7.5e+05 3.40e-05 63 5.11e+08 1.07e+06 5.10e+08 0.21- 7.7e+05 4.90e-05 64 2.-42e+08 7.57e+05 2.41e+08 0.31 5.le+05 5.22e-05 65 2.2e+04 66 2.32e+03 2.32e+03 0.00e+00 100.0 2.0e+04 4.12e-06 67 1.24e+02 4.63e+00 1.19e+02 3.74 1.4e+04 1.15e-08 68 4.30e+02 1.39e+01 4.16e+02 3.23 1.9e+04 2.61e-08 69 5.7e+03 70 1.22e+08 2.61e+04 1.22e+08 0.02 2.2e+05 4.23e-06
'71 3.00e+0S 3.30e+05 3.00e+08 0.11 6.le+05 1.90e-05 72 3.05e+08 1.83e+05 3.05e+08 0.06 1.0e+06 6.30e-06 i
NUREG/CR-6505, Vol. 2 36
Section 8 Discussion Table 8.1 (continued) i" 235 SNM mass U 2"U No. 2"U Enrichment Disposal Density' (c) (u) (c) (wt %) volume (ft')* e/cm 2
. 73 2.51 e+07 3.08e+04 2.51e+07 0.12 3.7e+05 74 4.00e+07 8.59e+04 2.91e-06 4.00e+07 0.21 3.8e+05 75 8.06e-06 i 76 5.4e+03 77 1.3e+04 6.41e+01 4.63e+00 5.95e+01 -!
7.22 1.2e+04 78 1.36e-08 l
- 79. 6.0e+03 1.95e+03 1.85e+01 1.93e+03 !
0.95 3.3e+04 80 2.00e-08 !
81 2.13e+04 !
1.06e+02 2.12e+04 0.50 82 3.le+04 1.20e-07 3.68e+05 6. Ole +01 3.68e+05 I 0.02 3.0e+04 83 2.21e+05 7.16e-08 1.14e+03 2.20e+05 0.52 l 84 2.6e+04 1.54e-06 85 4.52e+06 2.01e+03 4.52e+06 0.04 2.1e+05 86 9.88e+05 3.46e-07 3.16e+03 9.84e+05 0.32 5.7e+04 1.96e-06 87 1.95e+04 1.95e+04 2.5e+04
" SNM masses for Trench 38 and greater are determined by summing o opeuranium masses. is t Trench numbers less than 38 are reported SNM. Blank indicates data not reported or could no calculated (enrichment or density) from available data.
i
- Values reported in units of cubic feet. Cubic meters can be determined by mi 0.0283.
2
' Density for Trench 38 and greater is the density of 35U in g/cm'. For trench number the density not reporte . is the g/cm' of SNM. Blank values for the density indicate the volume or require Read as 1.32 x 10t NUREG/CR-6505, 37 Vol. 2
t Discussion Section 8 1
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Water contents (g H2 0/cm )
t l
l- Figure 8.2 Critical (k,g. 2 0.95) masses of spheres and " pseudo-infinite" slabs and cylinders as a 1 function of water content for different densities of 2"U in waste matrix l
l l \
1 v
i NUREG/CR-6505, Vol. 2 38 I L 1
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Section 8 Discussion 2
ennched "U, within the range for critical spheres. Because it is difficult to form a sphere under the transp conditions existing in a disposal trench, this small mass does not pose a criticality safety concern. The "old" trenches with unreported enrichments contained 10 to 175 kg of SNM (Trench 23), with a mean value of 30 k These masses are within the range of critical masses for spheres and for some slab configurations, assu 2 enrichments of 10 wt %. Note that larger masses will be needed for enrichments between I and 10 wt %.
Distribution of the source term within a given trench is an important component in the potential to deve critical mass. License conditions in effect between 1970 and 1981 limited the mass of "'U in a package, as as gave some constraints on disposal. Prior to 1977, package limits varied between 15 and 50 g. The areal 2
density of the material in the 2"old" trenches is less than 0.05 kg/m , less than one-third of the minimum "new" trench areal density of 0.16 kg/m (see Fig. 8.3). Although the enrichment of the early disposed material is not known, the smaller masses of SNM, the lower areal densities, and lower package limits indicate that the SNM i more dispersed in the "old" trenches (1-37) than in the "new" trenches (38-87). Therefore, while the older trenches have an inventory of SNM that theoretically would be sufficient to form a critical mass at 100 wt %
enrichment, it is2 likely that the SNM is dispersed enough that significant funneling would be required to reconcentrate the "U into a critical mass.
Trench 23 is an exception. Trench 23 contains 175 kg of highly enriched material and sufficient quantity either a sphere or slab with a critical mass. However, a subsequent evaluation of Trench 23 was performed Appendix B, "Suberiticality Evaluation for Chem-Nuclear Systems, Inc. Trench 23") that demonstrates that criticality would not be achieved because of:
(1) the limited thickness of the disposal burials (i.e., approximately 4 to 1I ft),
(2) 2 the conservatively evaluated low "U densities (e.g., between a maximum of about 2 0.00168 g U/cm' for 350 g SNM in a single 55-gal container to a generally expected global density ofless than about g 2.5 x 10-5 2"U/cm' within the trench), and (3) the significant quantity of source material (i.e.,210 metric tons of normal or depleted uranium or thorium),
that is, codisposed with the 175 kg SNM that results in an homogenized trench 2 averaged "U enrichment of about 0.08 wt %.
8.3 GEOCHEMICAL PROCESSES To increase the concentration of uranium, a geochemical process is needed that will capture uranium in a solid phase / Evidence from ore bodies indicates that uranium concentrations can increase through sorption onto a subs' rate or by precipitation as reduced uranium minerals. Because sorption was evaluated under a variety o conditions in Volume 1 of this study, this analysis assumed precipitation of uranium minerals.
One important result of the geochemical modeling is that given reasonable amounts of reducing agents, red zones are stable in a water ; sturated environment (with limited oxygen). A more realistic approach to modeling the redox conditions in the trench would be to model transient wet / dry cycles, and weathering of drums and wood crates. It is known that such containers weather fast enough to be breeched in recent disposal history.
However, the weathered containers are still present, not completely weathered away. The remnants of such containers will serve as surfaces for sorption or redox-driven precipitation of transported uranium.
NUREG/CR-6505, 39 Vol. 2
l
< o E-5~
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9 E-9 Concentrations of U-235 in Waste / Bamwell LLW Facility 8
o m _ n ~ --
1 i
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S- Trenches 1-35 @- Trenches 38-87 E 8 O
- "E E
.a S -
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i Ilili i i i i i i i i i i i i i 0.0 0.00002 0.00004 0.00006 0.0 0.00002 0.00004 0.00006 SNM (U-235) g/cm' U-235 g/cm' 2
Figure 8.3 IIistograms of calculated densities of "U in old and new disposal trenches using reported SNM or 2"U mass and reported disposal volumes (not total trench volume) m 3
d 5'
3 9 y s e
Section 8 Discussion The longevity of the transport processes needed to increase the concentration of uranium ca amount of water transporting dissolved uranium. Because of solubihty limitations, the concentration of dissolved in pore water is significantly less than the initial concentration on the waste. Thus the amou required to flush through the waste matrix in order to dissolve sufficient uranium for developing a cr quite large.
1 To mobilize the 4.4-kg requisite to form a slab with a entical density of 0.0224 g/cm' (line of 10 mg U/L (or 1 mg *U/L, assuming 10 ut % enrichment),3.5 2 10' L is needed. For a flow rate of 1 m/ year, the volume of water to pass through I m' (1-m 2 footprint) of waste with 40% porosity is 400 L per y The critical slab has a footprint of approximately 4 m and a thickness of approxunately 20 cm. As rate of I m/ year, ahuost i 1,000 years are needed to mobilize the 4.4 kg. Unsaturated flow, or intermittent flo through the trench will also lengthen this estimated mobilization time.
The amount of reducing agent available was not a limitation in this analysis. Only a few grams are cons 1 uranium reduction per pore volume, much less than the mass ofiron- or organic rich containers presen trenches. Because the reductant was not a limiting reactant, the modeled geochemical reactions were not ab provide a constraint on the conditions for mobilizing and immobilizing *U. However, other changes in hydrogeochemical conditions may occur over the time period of disposal.
8.4 HORIZONTAL-VS-VERTICAL FLOWPATHS i The initial areal densities within the waste trenches were estimated from the disposal records. The areal densiti were calculated by assuming that the waste concentration is projected on a 1-m footprint (Fig. 8.4) The .
areal densities for the Barnwell facility trenches are significantly less than the minimal areal density require enticality in a slab configuration (Table A.1, Appendix A), indicating that simple 1-D vertical flow paths will b insufficient to produce a critical areal density. Therefore, funneling of vertical flow paths or channeling of flo paths into horizontal flow (drainage system) must be considered.
He potential for horizontal flow in unsaturated soils has been evaluated by Kung (1990a,b). Kung (1990a,b) delineates three types of preferential flow in the unsaturated zone: short circuiting, fmgering, and funneling Short circuiting is the concentration of flow in macropores or fractures; fingering is the splittmg of flow pat to instabilities; and funneling is the horizontal movement and combining of flow paths caused by heterogenei Kung gives field evidence for funneling using dye tracers in sandy soils. Typically, a coarse unit will form a barrier to movement because of the large suction required to penetrate the larger pores; the penetration of this barrier is related to the hydraulic conductivity, the slope of the unit, and the height of the available water column. If funneling occurs, there is no need for saturated flow to develop. If flow occurs in 50% of the matrix, this amount is !
equivalent to increasing the capture zone for potential source material by a factor of 2, and if flow occurs in 1% of the matrix, the capture zone could be nearly a factor of 100 larger than assumed for strictly 1-D flow.
In general, the flow through the trenches at Barnwell is vertical. However, due to local heterogeneity and the transient nature of vertical groundwater flux, some horizontal flow may occur. Horizontal flow could occur within the waste zone or at the base of the trenches (Fig. 8.5) Horizontal flow is anticipated in the drainage system the base of the trench.
NUREG/CR-6505, 41 Vol. 2
i EE a r, -
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9 o
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Areal Density of SNM (U-235) kg/m2 Areal Density of U-235 kg/m 2 Figure 8.4 liistograms of calculated areal densities of 2"U in old and new disposal trenches using reported SNM or 2"U mass and reported disposal volumes 2
b 3
! Oc *
, . i Section 8 Discussion I
I i
i VV i
[a] [b]
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Figure 8.5 Cross sections showing horizontal flow (a) within the waste zone due to preferential flow paths drainage and (b) at the base of the trench due to less-permeable sediments at the bottom boundary and trenc l
i 4
4 NUREG/CR-6505, 43 vol 2
Section 8 Discussion Direct evidence indicates that horizontal Dow occurs in the Barnwell trenches. Tracer tests were conducted by the U.S. Geological Survey in experimental trenches (constructed like the actual trenches) at Barnwell.
and McMahon (1987) report a bromide tracer experiment in which "the data from the probes . . indicated the response . . was due to ponded water in the trench moving laterally into the undisturbed sediments." The attribute lateral flow to the impermeable nature of the " clayey" sand barrier, and also repon evidence for temporary perched water, which has potential for saturated lateral Cow.
A second piece of evidence for lateral flow is the appearance of tritium-contaminated water in the sumps monitor trench discharge (Ichimura et al.,1994; see their Fig. 5). These sumps collect water at the base of the trenches from lateral movement along a sand buffer, which is part of the trench design (Fig. 3.1). Quantitative estimates of How in the sumps were not available for this analysis. Much of the sump data are suspect because (1) video logs indicate that sump monitoring wells have filled in, and (2) discontinuities in water levels wi same trench indicate that monitoring wells act as conduits for surface water infiltration.' in either case, once the
! sump monitoring well is damaged, neither wet nor dry monitoring data can be used to understand water movement at the location. Thus, accumulation of wastewater in the sumps cannot be ruled out.
An estimate of the potential amount of horizontal flow at the base of the trenches was conducted by Mark Thaggard of the NRC using the HELP code, a 1-D water-routing code developed to analyze percolation 2
waste facilities (Schroeder et al.,1994). Thaggard used three layers to represent a waste trench and input regional precipitation data from Cahill (1982). The three layers represented waste backfill, a drainage zo base of the trench (used to collect water at sumps) and undisturbed soils beneath the trench. The hydraulic conductivity of the basal unit was an important variable in determining the amount of horizontal Dow that wo occur. The hydraulic conductivities measured in four samples of undisturbed 4 clayty sand vary from 6.7 x 1 j
4.8 x 10'(Dennehy and McMahon,1987). Using hydreulic conductivity of I x 10 cm/s, Thaggard estimated 5
90% of recharge moves horizontally at the base of the trench and using 10 cm/s,0.03% of the flow is horizont These values might be considered bounding estimates, although Thaggard points out that the higher conduct value affects horizontal Dow, which corresponds to recharge estimates that better me.tch regional recharge, it is l
not known whether recharge beneath the trenches should match regional recharge, however. The calculation leaves a large range in uncertainty for quantifying horizontal flow. However, the sensitivity of the calculation to hydraulic conductivity suggests that more in situ characterization of hydraulic conductivity in the clayey sa j
would help to better define the nature of Dow in the bottom of the Bamwell facility trenches.
The extent of lateral Dow needed to increase the areal density of the current configuration of 2"U in the Barnwell trenches is unknown and is difficult to estimate given the uncertainties concerning distribution of the source term in the trench, weathering of containers and vault materials, and How through the vaults. In the new trenches (38-87) if one vault contains SNM, then the adjacent vaults do not contain SNM. The areal densities of 2"U in the trenches are significantly less than the areal densities necessary to sustain a nuclear reaction. In order to achieve a minimum areal density of 4 kg/rd (Appendix A; Table A.1), significant lateral flow and/or funneling must occur. For new trenches, the lateral flow would have to be directed along the bottom of the trench in the drainage system, directly below the SNM. It is not known to what extent this or. curs. The areal densities for the ,
older trenche : suggest that the distribution of SNM in the trenches is probably more diffuse than in the newer
'Vemon Ichimura, Chem Nuclear Systems, Inc., personal communication to Laura Toran, September 10,1996.
2 Mark Thaggard, NRC, personal communication to Laura Toran, October 25,1996.
NUREG/CR-6505, 44 Vol. 2
O
- Section 8 Discussion water in increase significant the older trenches is in concentration the French drain system, which runs along the side of S .
of the tre necessary. U to occur, a combination oflateral and funneled flow will be In summary, it will be important to continue to monitor the reliable sumps draining the wa evidence of waste mobilization. Although the sumps are meant to provide a collection may enhance horizontal flow, they also can serve as a means of detecting uranium tra the sumps becomes water-saturated, there is the pote .
zones By keeping the trenches unsaturated and oxidized,it is p thicknesses that would be of concem for nuclear criticality safety.
8.5 OTHER MITIGATING FACTORS Other factors in the disposal setting may mitigate the reconfiguration of fissile uranium int Because these factors cannot be quantified with the present information and scope of wo they limit criticality is not presently known. Nonetheless, these considerations can be and used to identify any concerns that need to be addressed.
In all trenches at the Barnwell facility, there are multiple barriers specifically included t SNM. These barriers, which include disposal containers, vaults, and fill material, will inh uranium.
t The reducing zone. Y hey form, will require a water-saturated environment without disso Presently, the trenchu are not saturated with respect to water and the water chemistry oxidizing. The effects of changing rainfall rates and trench capping could be modeled to for saturating all or part of the trenches. Several trenches have high-density polyethylene ca capping is planned. Another effect of assuming saturated conditions is faster estimates of travel tim The potential distribution (amount and location) of reducing zones in this setting is a la in this analysis. As stated previously, if the reducing zones are extensive, the precipitated uraniu disperse and the concentration will be lower than needed to provide a criticality safety c zones do not form, the mechanism for increasing the concentration of uranium needs to chan con oded waste packages was not modeled in this analysis. Sogtion on mineral surfaces was consi Volume 1 of this study, and did not result in sufficient increase in uranium concentration.
l Small masses of "U may be considered a criticality safety concem (Figure 8.2); however, they m sphere, which is difficult to achieve given the generally dispersive nature of flow. Nonspherical g the same uranium density, require considerably greater masses of "U. Even though it may be easie slab or cylinder by flow paths in a trench, the masses required to form such geometnes are, in m greater than the inventory of an entire trench.
l i
l l
NUREG/CR-6505, 45 Vol. 2
i 9 CONCLUSIONS This study presented calculations and assumptions to evaluate the potential for a nuclear criticali LLW site where containerized waste is buried. Many operational conditions were considered. Data ne further evaluation are suggested. Even though conditions specific to Barnwell were used in the ca work was also intended to provide an example for types of analyses for detennining the potenti other disposal sites. The answers to questions posed in this study are as follows:
e is there sufficient inventory for a criticality safety concern?
The inventory of 23sU in the trenches is sufficient to form a critical geometry. 2 '5 However.*U U inventory, the enrichment and the spatial distribution of the uranium must be considered in combination tojudge whe criticality safety concern in a LLW trench. The critical mass of 235 U for pseudo-infmite slabs (length and width at least 10 times the thickness) comprised of silica 2 soil (1.6 g SiO /cm with variable densities of enriched 3
and water) ranged from a minimum of about i 1.3 kg to well over 2000 kg of SU, depending upon the concentration of water and uranium contamination in the available spaces of the soil porosity. This b critical masses coincides with a broad range of uranium density and system volumes or sizes. The smal masses require a greater concentration (i.e., densification) of enriched umnium than the larger critical masses example, a 60 x 60 x 6 m pseudo-infmite slab ofsilica soil contaminated with 0.00155 g *U/g of silica soil requires about 53,000 kg SU as 10 wt % enriched uranium to be critical whereas , .
a 131 x 1.31 x 0.131 m pseudo-infmite slab of silica soil contaminated with 0.032 g "U/g of silica soil requires about i1.3 kg *U 10 wt % enriched uranium to be critical. This differential is a nearly 21-fold increase in uranium den nearly 5,000-fold decrease in uranium mass.
The available mass of *U in the Barnwell trenches ranges from less than 1 kg to about 1070 kg with assu homogenized enrichments of 100 and 0.21 wt %. Taken together with the available information on distnbution of the uranium in the trenches (including average *U density), these combinations of mass and enrichment were not judged to present a criticality safety concern for the Bamwell LLW site. How Barnwell disposal trench, Trench 23, the 175 kg of highly enriched uranium contained in the trench caused sufYicient concem that a trench-specific investigation was performed (see Appendix B). The result indicated suberitMty was ensured by the physical distribution of the fissile material and the comm material with sutantial quantities of SM. Without site-specific information regarding the spatial distribution mass of disposals, such as for the Barnwell Trench 23 disposal, there could be concem for disposa sufficient inventory to pose a potential criticality safety concern.
o How decs U(10) compare to U(100)? How much more concentration of uranium is needed for U(10) th U(100)?
The criticality safety calculations showed that higher concentrations of S U were needed for U(10) than U(100), as expected. At the minimum concentration that achieved safety concem, the concentration ratio of mU f that for U(100) was 1.6. Variances in this ratio have not been analyzed on a point-by-point basis and could larger and potential smaller critical than mass of S the value at the minimum concentration. Furthermore, the differences in the minimu U for U(10) and U(100) are small. Only when enrichments drop below 10 wt % do minimum masses increase significantly.
NUREG/CR-6505, 47 Vol. 2 I
l
)
Conclusions Section 9 What chemical conditions and physical aspects of trenches are conducise to increasing uranium concentration?
Some trenches have suf6cient disposal masses to pose a criticality safety concern, and the conditions are potentially available to mobilize and precipitate uranium. At the Barnwell LLW disposal facility, the water is l slightly oxidizing, which enhances the potential for the transpon of uranium. The presence of wood and iron '
containers within the trenches increases the potential for reducing zones to form. Transport could occur as water flushes through the waste and forms uranyl carbonate complexes. Degradation of waste containers can produce a -
reducing zone in the waste trench, if a water-saturated environment is present to limit oxygen concentrations.
These hypothetical reducing zones can precipitate uranium, as discussed below. The zones are hypothetical because they have not actually been observed at the site.
The hypothetical nature of the reducing zones is an important limitation to the current work because the mass required to cause a criticality safety concern is dependent upon the geometry of the fissile material, which in turn l is dependent on specine geometries of the reducing zones. Multiple flow paths would need to funnel the uranium L from large trench volumes (if the SNM is distributed across the trench volume) to very small reducing zones in I order to increase the concentration of '"U to a criticality safety concern. Iflarger reducing zones form, either l the 2"U will be too diffuse, or larger sources of 2"U than are reported to be present in the trench are recuired.
It is difficult to calculate travel time for the accumulation of uranium. Since kilogram quantities of 2"U are needed, and transporting waters have only mg/L of 2"U, millions ofliters of water need to flush through a reducing zone. Speculation about flow rates and conditions is needed to make estimates, but travel times are long. Tens of thousands of years are needed, assuming a 1-m/ year velocity, a 40-vol % porosity, and no dispersion.
- Can reducing zones, which precipitate uranium, be sustained to enable critical masses to accumulate?
Yes. In the analysis, reducing zones formed very efficient barriers to uranium transport, precipitating nearly 100% of the uranium in solution. The reducing zones did not become oxidizing despite the influx of oxidized water over the ran;;e of pore volumes modeled: 500 to 30,000 L. The source of the reducing agent was postulated to be steel drums or wooden crates.
- How could disposal practices, in particular at the Barnwell LLW disposal facility, enhance or mitigate the development of critical masses of SNM?
Some disposal practices and trench designs can increase the potential for criticality and others can reduce it.
Trenches dug in impermeable material tend to retain the fissile material within their confines, increasing the potential for criticality. Trenches that allow for contaminant release reduce the potential for criticality by increasing the volume of the geologic medhim into which the uranium can spread. At Barnwell, tritium release has occurred, but uranium might or might not be released from these trenches, depending on the geochemical conditions. Trenches whose floors are sloped can enhance focused flow, increasing the potential for accumulation of uranium. Consequently, French drains and sumps can be sites for SNM accumulation, indicating the desirability to restrict (when possible) their geometries and size to avoid potential critical geometries.
Maintaining relatively impermeable caps on the trenches should tend to reduce the uranium migration by limiting the quantity of water available and the potential for locally reducing conditions where 2"U could accumulate by promoting aeration of the trench due to its unsaturated conditions. Commingling of SNM and SM may reduce the enrichment of the mobilized uranium if the two are leached similarly and well blended. To assume similar NUREG/CR-6505, Vol. 2 48
Section 9 Conclusions To assume similar leaching and thorough blending, it would be necessary to consider th containing SNM and SM and their locations in the trench to estimate the time when packa the rate ofleaching. The Trench 23 records that were examined in detail indicate that homog and SM would likely occur because of the relatively uniform intermingling of SNM and SM e .
that sorb uranium can slow the transfer process and disperse . e reducing the needed density, but increasing the needed mass, o Although conditions that permit criticality safety concems are not impossible, disposal prac potential LLW facilities: concern. This study results in the following recommendations for consideration oflic 1.
-Minimize those factors that enhance SNM accumulation.
Reduce groundwater infiltration Reduce enrichment Minimize opportunities to create cells isolated zones ofreducing conditions. Avoid organic matte Design trench to minimize focused flow
- 2. Limit the areal density of the fissile materials.
- 3. Model trench performance using site-specific conditions on a scale that addresses Consequently, the observation that the average enrichment of a trench is less than I wt %
2 does not necessarily eliminate a criticality concem for the trench. Burial reports may sugges regions of a trench contain quantities of fissile material that greatly exceed the average ennchme 4.
Continue to use sumps in disposal trenches to monitor for the presence ofiron, organics, and u indicators of mobility in the trenches. If uranium is observed in the sumps, determine its enric Changes in redox conditions may be monitored by changes in different iron species. E take many years for sufficient buildup of uranium, early detection ofmobile iron and uranium changes in the trench water chemistry.
In summary, the concentration of dissolved uranium in reducing zones is a possible mechanism for reconcentrating SNM within LLW disposal facilities. Further study of the geometry of these reduc needed to evaluate the potential for concentrating relatively small critical masses (e.g., spher ,
the associated predicted consequences. Further evaluation of flow through the trenches would effect of the source term and its distribution within the trench on the development of a critical m of mitigating factors, such as multiple-barrier disposal in the new trenches and the effect and iron presented oxyhydroxides here. to inhibit transport out of the waste containers, will also better define th NUREG/CR-6505, 49 Vol. 2
10 REFERENCES Autry, V. R., Director, Division of Radioactive Waste Management, Bureau of Land and Waste Management, S Department of Health and Environmental Control, letter to T. E. Harris, Project Manager, Low-leve Issues Section, Low-Level Waste and Decommissioning Projects Branch, Division of Waste Managemen Material 1998. Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-000 Baedecker, M. J., and W. Back,"Hydrogeological Processes and Chemical Reactions at a Landfill," Groun 429-37(1979). ,
Baedecker, M. J., L M. Cozzarelli, R. P. Eganhouse, D. I. Siegel, and P. C. Bennett, " Crude Oil in a Shallo Gravel Aquifer-IIL 8,569-86(1993). Biogeochemical Reactions and Mass Balance Modeling in Anoxic Groundw Bowman, C. D., and F. Venneri, Underground Autocatalytic Criticalityfrom Plutonium and Other Fissile Mate LA-UR-94-4022, Los Alamos National Laboratory,1994 ,
Cahill, J., Hydrology ofthe Low LevelRadioactive Solid-Waste Burial Site and Vicinity Near Barnwell.
U.S. Geological Survey Open Fi.e Report 82-863,101 pp.,1982.
Champ, Earth D. R., (1979).
Sci. 16,12-23 J. Gulens, and R. E. Jackson, " Oxidation-Reduction Sequences in Ground Water Flow Sys Christensen, T. H., P. Kjeldsen, H.-J. Albrechtsen, G. Heron, P. H. Nielsen, P. L. Bjerg, and P. E. Holm, " A Landfill Leachate Pollutants in Aquifers," CriticalReviews in EnvironmentalScience and Technolog Dennehy, K. F., and P. B. McMahon, Water Movement in the Unsaturated Zone at a Low-LevelRadioactive Site Near Barnwell. South Carolina, U.S. Geological Survey Open File Report 87-46,66 pp.,1987.
Engesgaard, P., and K. L. Kipp, "A Geochemical Transport Model for Redox-Controlled Movement ofMiner Groundwater Flow Systems: A Case of Nitrate Removal by Oxidation of Pyrite," Water Resour. Res Greene, N. M. et nl., The LA WLibrary-A Multigroup Cross-Section Libraryfor use in Radioactive Waste Calculations, ORNI/IM-12370, Martin Marietta Energy Systems, Inc., Oak Ridge Natl. Lab., August 1994.
Greenspan, E., S. Armel, J. Ahn and J. Vujic," Critical Depositions of Uranium in Rock Having Positive Feedback,"
Chelan, WA. American Nuclear Society September 7-11,1997 Topical Meeting of the Nuclear Criticality S ,
Heron, 153-58(1994).
G., T. H. Christensen, and J. C. Tjell, " Oxidation Capacity of Aquifer Sediments," Environ. Sci. a Hopper, C. M., R. H. Odegaarden, C. V, Parks, P. B. Fox, Criticality Safety Criteriafor License Review Facilities, NUREG/CR-6284 (ORNLffM-12845),1995. i ofLo House, W. B., Corporate Director, Regulatory Affairs, Chem-Nuclear Systems, Inc., (RA-0372-96), letter to Prrject Manager, Low-level Waste and Regulatory Issues S:,. tion, Low-Level Waste and Decommissioni Division of Waste Management, Office ofNuclear Material Sdty and Safeguards, U.S. Nuclear Regulatory Co Washington, D.C. 20555-0001, dated August 27,1996. ,
Ichimura, V. T., B. S. Smith, and M. T. Ryan, " Cover Design for the Low Level Radioactive Waste Disposal Site n Barnwell, (1994).
South Carolina," Toxic Substances and the Hydrologic Sciences, American Institute of Hydrology,478 NUREG/CR-6505, 51 Vol. 2
References Section 10 {
t Jordan. W. C., and J. C. Tumet. Estimaled Critical Conditionsfor UO F;-H O Systems in Fully Water-Reflected Spherical i Geo,netry. ORN1/Ihl2292, Martin Marietta Energy Systems, Inc., Oak Ridge Natl. Lab., December 1992. I i
Kung, K-J. S.," Preferential Flow in a Sandy Vadose Zone: 1. Field Observation," Geoderma 46,51-58 (1990a). I Kung, K J. S., " Preferential Flow in a Sandy Vadose Zone: 2. Mechanism and Implications," Geoderma 46,59-71 (1990b).'
Landais, P., " Organic Geochemistry of Sedimentary Uranium Ore Deposits," Ore Geology Reviews 11,33-51 (1996).
i Lindberg, R. D., and D. D. Runnells. " Ground Water Redox Reactions: An Analysis of Equilibrium State Applied to Eh l Measurements and Geochemical Modeling,"Sc ence 225,925-27 (l984).
j r
Morfitt, J. W., Minimum CriticalMass and Umform ThermalNeutron Core Flux in an ExperimentalReactor, Y-1023, Union ,
Carbide Corporation, Nucl. Div., Oak Ridge Y-12 Plant,1953.
I Parkhurst, D. C., User's Guide to PHREEQC-A Computer Programfor Speciation. Reaction-Path. Advective-Transport. and Innese Geochemical Calculations USGS WRI 95-4227,1995.
I Pietrzak, R., K. S. Cryseinski, and A. J. Weiss, Evaluation ofIsotope Migration-Land Burial Water Chemistry at Co'mmercially Operated Low-Level Radioactive Waste Disposal Sites Status Report October 198Meptember 1981, NUREG/CR-2616,1981.
Pruvost, N. L., and H. C. Paxton, Nuclear Criticality Safety Guide, LA 12808, Los Alamos National Laboratory, September 1996.
SCALE: A Modular Code Systemfor Performing Standardi:ed Computer Analysesfor Licensing Evaluation. Vols.1-111, NUREG/CR-0200, Rev. 4 (ORNL/NUREG/CSD-2/R4)(April 1995). Available from Radiation Shielding Information Center as CCC-545. !
Schoeder, P. R., C. M. Lloyd, P. A. Zappi, and N. M. Aziz, The Hydrologic Evaluation ofLandfillPerformance (HELP) {
Model: User's Guidefor Version 3, EPAl600-R 94/l68a,1994.
}i Sims, R., T. A. Lawless, J. L. Alexander, D. G. Bennet, and D. Read, " Uranium Migration Brough Intact Sand-stone:
Effect of Pollutant Concentration and the Reversibility of Uptake," Migration 93, Charleston, S.C.,1993.
f j
Spalding, B. P., "Insitu Grouting of Buried Transuranic Waste with Polyacrylamide," Oak Ridge Model Conference !
Proceedings, CONF-871075, 39-76,1987.
Spirakis, C. S., "De Roles of Organic Matter in the Formation of Uranium Deposits in Sedimentary Rocks," Ore Geology f Reviews 11,53-49 (1996). t Toran, L. E., C. M. Hopper, M. T. Naney, C. V. Parks, J. F. McCarthy, B. L. Broadhead, and V. Colten-Bradley, The Potentialfor Criticality Following Disposal of Uranium at Low-Level Waste Facilities, Volume I; Uranium Blended With Soil, NUREG/CR-6505, Vol.1 (ORNI/TM-13323/V1), U.S. Nuclear Regulatory Commission,1997a. !
Toran, L. E., S. Bryant, J. Eaton, and M. F. Wheeler, " Coupled Geochemistry and Bioreactions for Remediation Modeling,"
Submitted to the American Geophysical Union Spring Meeting, '3altimore, MD, May 27-30,1997b. {
j Weiss, A. J., and P. Colombo, Evaluation ofIsotope Migration-LandBurial Water Chemistry at Commercially Operated Low-LevelRadioactive Waste DisposalSites Status Report Through September 30.1979, NUREG/CR 1289, l979.
i i
i NUREG/CR-6505, Vol. 2 52 '
I r
i l
. e-1 Section 10 l References I i
. Wheeler, M. F., T. Arbogast, S. Bryant. C. N. Dawson, F. Saaf, and C. Wang, "New Computation Chemically Reactive Transport in Porous Media," Next Generation EnvironmentalModels and Col!
Proceedings ofthe U.S. EnvironmentalProtection Agency Workshop (NGEMCOM). G Delic Philadelphia, pp. 217-26,1997, ,
Wood, S. A.,"The Role of Humic Substances in the Transport and Fixation of,Metals V)," Ore Geology Reviews II, 1-31 (1996). , ,,
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NUREG/CR-6505, 55 Vol. 2
Appendix A Criticality Study Results
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Criticality Study Results Appendix"A i
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Appendix A Criticality Study Results l l
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Criticality Study Results Appendix A 1'
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Appendix A Criticality Study Results U
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Cnticality Snidy Results Appendix A r U:i< = sus 2
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Criticality Study Results Appendix A 1
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NUREGICR-6505, 65 Vol. 2
Cnticality Study Results Appendix A ti et
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235 l Figure A.10 Sphere mass (kg U) vs g H2 0/g SiO2 and g '"U/g SiO2 l
i l
l l
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l NUREG/CR-6505, Vol. 2 66 l
l l --
Appendix A Criticality Study Results e
.05 g ,gyg,,g yg;
.1 l
15 k
i
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i 1 50 l1tW)
J .%phe.t W ssIA t D!
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3 Figure A.11 Sphere mass (kg "5U) vs g2 H 0/g SiO 2 and log scale of g "'U/g 2SiO 4
i NUREG/CR-6505, 67 Vol. 2
P Tr.ble A.1 U(10) plus II,0 plus SiO,-soil (S-S) results t.a Q
"'U content Water content Criticalinfinite slah 5 Q
Line Criticalinfinite cylinder l Critical sphere k, or d
W entry g '"U/cm' s "'U/gS-S g II,O/cm' g II,O/gS-S k-int 1hickness Diameter Q
Linear density Diameter Mass 4 m (cm) (kg "t/mdensit7)
Areal (cm) (kg "'U/m) (cm) (kg '"U) M "o I 0.0005 0.0003125 0 E
." 0. 0.430 2 0.0005 0.000313 0.058 0.03625 0.356 x 3 0.0005 3
0.000313 0.119 0.07438 0.282 E.
G 4 0.0005 0.0003I3 0.183 0.11438 0.230 5 0.0005 0.000313 0.251 0.15688 0.193 6 0.0005 0.000313 0.324 0.2025 0.164 7 0.0005 0.000313 0.4 .
0.25 0.142 8 0.0014179 0.000886 0 0 0.713 9 0.0014179 0.000886 0.029 0.018I3 0.779 10 0.0014179 0.000886 0.058 0.03625 0.73 g II 0.0014179 0.000886 0.0885 0.05531 0.672 12 0.0014179 0.000886 0.119 0.07438 0.62 13 0.0014179 0.000886 0.151 0.09435 0.572 I4 0.0014179 0.000886 0.I83 0.11438 0.53 15 0.0014179 0.000886 0.217 0.13563 0.492 16 0.0014179 0.000886 0.251 0.15688 0.458 17 0.0014179 0.000886 0.2875 0.17 % 9 0.427 18 0.0014179 0.000886 0.324 0.2025 0.4 19 0.0014179 0.000886 0.362 0.22625 0.375 20 0.0014179 0.000886 0.4 0.25 0.353 21 0.00163 0.0010188 0. O. 0.744 22 0.00163 0.0010188 0 029 0 029 0.832 23 0.00163 0.0010l88 0.058 0.058- 0.787 >
E 2
ee x
i t
.i
'"U content Water content Table A.I (continued) du
- Critical inGnete slah Criticalinnnite cylinder Line k,or Critica; sphere O entry a '"U/cm' 8 "UtgS-S 8
g II,O/cm' g II,O/gS-S k int Thickness Areal density 9-Diameter Linear density Diameter Mass *
(cm) (ng "U/m') (cm) (kg '"U/m) (cm) (kg '"U) >
24 0.00163 0.0010188 0.0885 0.0885 0.73 25 0.00163 0.0010188 0.119 0.119 0.677 26 0.00163 0.0010188 0.151 0.151 0.627 27 0.00163 0.0010188 0.183 0.I83 0.583 28 0.00163 0.0010188 0.217 0.217 0.543 29 0 00163 0.0010188 0.251 0.251 0.507 30 0.00163 0.0010l88 0.2875 0.2875 0.474 31 0.00163 0.0010l88 0.324 0.324 0.445 I
32 0.00163 0.0010188 0.362 0.362 0.418 33 0.00163 0.0010l88 0.4 0.4 0.394 m
W 34 0.0018 0.001125 0. O. 0.763 35 0.0018 0.001125 0.029 0.029 0.870 36 0.0018 0 00ll25 0.058 0.058 0.828 37 0 0018 0.001125 0.119 0.119 0.719 1 38 0.0018 0.001125 0.I83 0.183 0.623 39 0.0018 0.001125 0.251 0.25I 0.544 40 0.0018 0.00ll25 0.324 0.324 0.479 41 0.0018 0 00ll25 0.4 0.4 0.426 42 0.0018738 0.00I1711 0. O. 0.771 h, I 43 0.0018738 0.0011711 0.029 0.01813 0.884
=.
O I 44 0 0018738 0.0011711 0.058 0.03625 0.845 k
O 45 0.0018738 0.0011711 0.0885 0.05531 0.789 en E
46 0.0018738
<@a t 0.0011711 0.119 0.07438 0.736 y
28 u .*
47 0.0018738 0.0011711 0.15 0.09438 0.684 a
E i
Bi ,
_ . . _ . _ _ .__ -.-.m....___._.m _..__.._..m.. _ . _ _ _ _ _ _ _ _ _ ____.____________e____ _______s_ _ _ _ _ _ _ . _ _ _ _ _- _ - . _ _ . _ - ._ __
- 2. Table A.1 (continued) n
. w 3 M-o 8"U content Water content Criticalinfinite stah Critical infimte cyliraler Critical sphere E.
8 ps 1.ine g '"U/cm' g '"U/gS-S g it,O/cm' k_or q entry g II,O/gS-S k-inf nickness Areal density Diameter Linear density Diametei Mass
& (cm) (kg 8"U/m') (cm) (kg '"U/m) (cm) (kg '"U) t.n E
o u 48 0 0018738 0.0011711 0.183 0.11438 0 639 k 49 0.0018738 0 0011711 0.217 0.13563 0.597 k
1%
50 0.0018738 0.00I1711 0.251 0.15688 0.560 51 0.0018738 0.00I1711 0.2875 0.I7%9 0.525 52 0.0018738 0.001171I 0.324 0.2025 0.494 53 0.0018738 0.0011711 0.362 0.22625 0.465 54 0.0018738 0.0011711 0.4 0.25 0.439 55 0.002154 0.0013463 0. O. 0.794 l
56 0.002154 0.0013463 0.029 0.01813 0.935 57 0.002154 0 0013463 0.058 0.03625 0.902
, 4 L O 58 0.002154 0.0013463 0 0885 0 05531 0.848 59 0.002154 0.0013463 0.119 0.07438 0.795 i 60 0.002154 0.0013463 0.151 0.09438 0.744 61 0.002154 0.0013463 0.183 0.11438 0.697 62 0.002154 0.0013463 0.217 0.I3563 0.654 63 0.002154 0.0013463 0.251 0.15688 0.615 -
64 0.002I54 0.0013463 0.2875 0.17969 0.578 65 0.002154 0 0013463 0.324 0.2025 0.546 66 0.002154 0 0013463 0.362 0.22625 0.515 i i
67 0 002154 0.0013463 0.4 0.25 0.487 ,
68 0.0024761 0.00I5476 0. O. 0.811 69 0.0024761 0.0015476 0.029 0 01813 0.982 464.22 11.4946 758.34 III.84 1020 78 1378.9 % 7 y 70 0.0024761 0.0015476 0.058 0.03625 0.957 966.8 23.9389 1
1684.72 551.97 2867.82 30578.9410 0 O.
71 0 0024761 0.0015476 0.0885 0 05531 0.907 k *i I
e l 1
Table A.I (continued)
'"U content Water content d.
t Line Critreal in6nise slah Cntical in6 nite cyhnder k,or Critical sphere O entry g '"U/cm' g '"U/gS-S g it,O/cm' g II,O/gS-S k-inf 9" lhickness Are. Jensity Diameter Linear density
- Diameter Mass (cm) (kg "'U/m') (cm) (kg '"U/m) Icm) (k g '"Up >
72 0 0024761 0.0015476 0.119 0 07438 0 856 73 0 0024761 0 0015476 0.151 0 09438 0.804 74 0 0024761 0.0015476 0.183 0 11438 0.757 75 0.0024761 0.0015476 0.217 0.13563 0.713 76 0.0024761 0 0015476 0.251 0.15688 0 673 77 0.0024761 0.0015476 0.2875 0.17% 9 0.635 78 0.0024761 0.0015476 0.324 0.2025 0.600 79 0.0024761 0.0015476 0.362 0.22625 0.568 80 0.0024761 0.0015476 0.4 0.25 0.539 81 0 0028465 0.0017791 0. O. 0827 4
~
82 0.0028465 0.0017791 0 029 0 01813 1.027 275.92 7.8541 461.7 47 6563 629.I6 371.18729 83 0 0028465 0.0017791 0.058 0.03625 I.011 258.9 7.3696 429.26 41.1947 576 8 286 01284 84 0.0028465 0 0017791 0.0885 0 05531 0.%5 486.26 13.8414 829.7 153 901 1188.78 2503.8843 85 0.0028465 0 0017791 0.l19 0 07438 0.916 86 0.0028465 0.0017791 0.151 0.09438 0.865 87 0.0028465 0.0017791 0.183 0.I1438 0.818 88 0.0028465 0 0017791 0.217 0.13563 0.774 89 0.0028465 0 0017791 0.251 0.15688 0.732 90 0.0028465 0.0017791 0.2875 0.17969 0.693 O c.
91 0.0028465 0.0017791 0.324 0.2025 0.658 C.
92 0.0028465 0.0017791 0.362
{
=-
0 22625 0 624 h 93 0.0028465 0 0017791 0.4 tn
<m o
W 94 0.0028639 0 0017899 e 0.25 0.
0.593 0.828
(
y 2$
wy 95 0.0028639 0 0017899 0.029 0.01813 1.019 270 62 7.7503 456 82 o
46.9394 621 22 350 49490 $
a
o
.. Table A.I (continued) n 3.
o '"U content Water content Critical infinite slab Cnticalinfinite cylinder Critical where ---
3 Line k.or ps entry g '"U/cm' g '"U/gS-S g II,O/cm' g II,O/gS-S k-inf 7hickness Areal density Diameter Q
Linest density Diameter
& (cm) (kg '"U/m') (cm) (kg '"U/m) (cm)
Mass (kg '"U) t/2 E
y % 0 0028639 0.0017899 0.058 0.03J25 1.013 253.3 7.2543 420 06 39 6889 488.7 175 01830 k
97 Oh028639 0.0017899 0.0885 0 05531 0.967 452.94 12.9717 741.02 123.511 1045.96 1715.9383 $
98 0.0028639 0.0017899 0.119 G 0.07438 0.918 99 0.0028639 0.0017899 0.151 0.09438 0.868 100 0.0028639 0.0017899 0.183 0.11438 0.821 101 0.0028639 0.0017899 0 217 0.13563 0.776 102 0.0028639 0.0017899 0.251 0.15688 0.735 103 0.0028639 0.0017899 0.2875 0.17 % 9 0.696 104 0.0028639 0.0017899 0.324 0.2025 0 66 105 0.On28639 0 0017899 0.362 0.22625 0.626 w
N 106 0.0028639 0.0017899 0.4 0.25 0.5%
107 0.0032722 0.0020451 0 0. 0.839 108 0.0032722 0.0020451 0.029 0 01813 1.076 206.12 6.7447 356 8 32.7174 488.7 199 97028 109 0.0032722 0 0020451 0.058 0.03625 1.071 205.9 6.7375 299.36 23.0312 410.46 118.48161 110 0.0032722 0 0020451 0.0885 0.05531 1.031 174 48 5.7093 328 62 27.7534 439.7 145 64911 Ill 0.0032722 0.0020451 0.119 0.07438 0.975 314.6 10.2943 513.12 67 6654 696 78 579 59612 112 0 0032722 0 0020451 0.151 0 09438 0.926 113 0.0032722 0.0020451 0.183 0.11438 0.880 114 0.0032722 0 0020451 0.217 0.13563 0.835 IIS 0.0032722 0 0020451 0.251 0.15688 0.793 116 0.0032722 0.0020451 0.2875 0.17969 0.754 l17 0.0032722 0 0020451 0.324 0.2025 0.717 g
IIB 0.0032722 0 0020451 0.362 0.22625 0.682 119 0.0032722 0.0020451 0.4 0.25 0.651 e-X -
1._-____.________-- - - -
_ - - - - - - -- _ -W
Table A.1 (continued) 8"U content Water content Cnticalinfinite slab d~
m Line Criticalinfinite cylimler Cntical sphere entry g '"U/cm' k.or O g '"U/gS-S g li,O/cm' g it,OrgS-S k.inf &
Thickness Areal density Diameter I_incar density Diameter Mass M '
(cm) (kg '"U/m')
120 0.0036054 0.0022534 (cm) (1g '"U/m) (emi (Lg '"U) >
0 0. 0.844 125 0.0036054 0.0022534 0.029 0.01813 1.094 178.4 6.432 314 58 28 0223 435.52 155.94676 122 0 0036054 0.0022534 !
0.058 0.03625 1.095 144.80 52206 254.54 183465 351.28 81.829957 123 0.0036054 0.0022534 0.0885 0 05531 1.059 159.32 5.7441 255.18 18.4389 348.92 80.191741 124 0.0036054 0.0022534 0.119 0.07438 1.015 179.48 6.471 30034 25.5428 405.2 125.59153 125 0.0036054 0.0022534 0.151 0.09438 0.%8 338.54 12.2057 559.48 88.6363 770 861.83520 126 0.0036054 0 0022534 0.183 0.11438 0.923 127 0.0036054 0 0022534 0.217 0.13563 0.878 128 0.0036054 0 0022534 0.251 0.15688 0 836 129 0.0036054 0 0022534 0.2875 0.17969 0.796 {
4 W
130 0.0036054 0.0022534 0324 0.2025 0.760 131 0.0036054 0 0022534 0362 0.22626 0.724 '
132 0.0036054 0 0022534 0.4 0.25 0.692 133 0.004539 0.0028369 0. O. 0.848 134 0.004539 0.0028369 0.029 0.01813 1.147 137.54 6.2429 252.60 22.7465 354.52 135 105.89635 0.004539 0 0028369 0.058 0.03625 3
1.168 106.26 4 823I 194.66 13.5083 272 8 136 48.249396 0.004539 0 0028369 0.0885 0 05531 1.143 99.46 4.5145 17830 113331 248.16 137 36.320692 0.004539 0.0028369 0.119 0.07438 1.107 101.48 4.6062 178 04 11.3001 244.74 138 34 83 % 39 0 004539 0.0028369 0.151 0.09438 1.064 111.42 5.0574 191.08 13.0160 O 261.02 42.264928 139 0.004539 0.0028369 0.183 Q-0.11438 1.022 134 88 6.1222 22438 17.9480 304.96 67.404271 w
140 0.004539 0.0028369 0.217 Ed.
0.13563 0.98 204 24 9.2705 33330 39 6022 444.22 Q
o 141 0.004539 0 0028369 0.251 0.15688 0.939 208 33062 i
W
<" 6 I42 0.004539 0.0028369 0.2875 0.17969 0.899 x k r S.$ 143 0.004539 0.0028369 0324 0 2025 0.862 W
84 h
t"n_
.o' 7 0 Table A.I (continued) to R o "'U content Water content Criticalin6 nite slab Criticet in6 nite cylinder
$ line k.or Cntical sphere -:.
pc g '"U/cm' g "'U/gs.S entry g II,O/cm' g fl aO'gS-S k-int 7hickness Areal density Q
& (cm) (kg '"U/m')
Diameter 1.inear density Diameter Mass tt)
(cm) (kg '"U/m) (cm) (kg '"U) E yi 144 0.004539 1
__... . M28369 0.362 0.22625 0.826 ky I45 0 004 ' e - 0 0028369 0.4 0 25 0.793 g E.
146 0.0057142 0.0035714 0. 0 0 845 G I47 0.0057142 0.0035714 0.029 0 01813 I.187 113.84 6.505 217.88 21.3049 309.16 88.410517 148 0.0057'42 0.0035714 0 058 0 0.1625 1.229 85.06 4.8605 162.50 11.8509 230.72 36. W 974 I49 0.0057142 0.0035714 0.0885 0 05531 1.218 76 08 4.3474 142.68 9.I3631 201.26 24.390822 150 0.0057142 0.0035714 0.119 0 07438 1.190 73.42 4.1954 134.86 8.16227 189 08 20.225106 151 0.0057142 0 0035714 0.I51 0.09438 I.I54 74 4.2285 133.02 7.94106 184.84 is.894783 152 0.0057142 0.0035714 0 183 0.11438 1.117 77.36 4 4205 135.94 8.29353 187.6 19.753885 153 0.0057142 0 0035714 0.217 0.13563 1.078 84.16 4 8091 I45.14 9.45407 198.28 23.323341 N 154 0.0057I42 0.0035714 0.251 0.15688 1.039 97.18 5.5531 163.42 11.99 222.I4 32.797016 155 0.0057142 0.0035714 0.2875 0.17969 1.001 125.58 7.1759 206.36 19.11 277.76 64.115516 156 0 0037142 0 0035714 0.324 0.2025 0.%5 234.28 13.3872 401.80 72.4544 484 339 22677 157 0.0057142 0.0035714 0.362 0.22625 0.929 158 0.0057142 0.0035714 0.4 0.25 08%
159 0 0063 0.0039375 0. 0 0 841 160 0.0063 0.0039375 0.029 0.01813 1.200 106.76 6.7259 207.36 21.2749 296 06 85.6011 161 0.0063 0.0039375 0.058 003625 1.251 78 62 4 9531 152.84 11.5585 218.36 34 344595 162 0 0063 0.0039375 0.119 0 07438 I.222 66 02 4.1593 123.50 7.54681 174.32 17.473536 163 0.0063 0.0039375 0.183 0.11438 1.154 66.58 4.1945 119.22 7.03280 165 E8 15 056403 164 0.0063 0 0039375 0.251 0.15688 1.08I 75 82 4.7767 130 56 8.43432 178 8 18 855658 165 0.0063 0 0039375 0.324 0.2025 1.008 108.14 6 8128 166 0.0063 0 0039375 0.4 0.25 0.940 177.52 15.5928 240 68 45.989549 do 167 0.01167 0 0072938 0. O. 0.809 e-M -
,_. ,. y-, .- - .. . ,w . . - - . , - - . - . e :~s.. - . ., ,. - - . . , - . - . . - __ms
I O
Table A.1 (continued) 8"U content Water con'ent d*
a I.mc Cnticalinimite slah Criticalinimite c)linder k.or Ontscal spliere D entry g '"U/cm' g '"U/gS-S g II,O/cm' g II,O/gS-S k-int 7hid ness Arealdensity thameter I.incar density Diameter Mass *
(cm) og 8"U/m')
168 0 01167 (cm) og '"ti/m) (cm) o g "'U) >
0 0072938 0.029 0 01813 1.232 80 08 9.3694 170.30 26.6497 249.76 95 4452 169 0.01167 0 0072938 0 058 0 03625 1.336 54.32 6 3391 117.66 12 6887 173 48 31.901984 170 0 01167 0.0072938 0.119 0.07438 1.376 40 88 4.7707 85.8 6.74737 125.52 12 083931 171 0 01167 0.0072938 0.183 0.I1438 1.351 36.38 4.2455 73.4 4 93801 106.14 7.3064548 172 0.01167 0.0072938 0.251 0.15688 I.307 34 48 4.0238 67.14 4.I3164 96. 5.4060885 t 173 0 01167 0.0072938 0.324 0.2025 1.255 34.02 3.9701 64 08 3.76361 86 6 3.% 84702 174 0.01167 0 0072938 OA 0.25 1.201 34 68 4.0472 63.54 3.70045 88 82 4.2815573 175 0.01704 0.0I06500 0. 0 0.798 176 0 01704 0 01065 0.029 -
0 0l813 1.216 72.62 12.3454 161.24 34 7114 238.58 120 8789 177 0.01704 0 01065 0.058 0.03625 1.346 46.56 7.9338 4 107.44 15.4486 160 9 u 37.165187 178 0 01704 0 01065 0.I19 0.07438 I.425 33 44 5 6982 75.14 7.55616 111.94 12.514805 179 0 01704 0.01065 0.183 0.11438 f.428 28.69 4 8888 62.06 5.15445 91.56 6 8483436 180 0.01704 0 01065 0.251 0.15688 1.406 26 2 4 4645 54 68 4 00143 79 92 IBI 4.554436l 0.01704 0 01065 0.324 0.2025 1.372 24.76 4 2191 50.02 3.34847 72.4 3.3859761 182 0 01704 0.01065 0.4 0.25 1.332 23.94 4 0794 4706 2. % 389 67.5 2.7439709 183 0.0224 0 014 0. O. 0.801 184 0 0224 0 014 0.029 0.01813 1.192 69.72 15.6173 158 62 44 2630 235.40 152.9914 185 0.0224 0.014 0 058 0 03625 1.335 42.74 9.5738 102.72 18.5630 154.88 186 43.574522 0.0224 0 014 0.119 0 07438 IA4 29.72 6 6573 70.08 P 8.64024 105 68 187 0 0224 0.014 a !!2 dii424 1.463 25.02 13 842834 3 5.6045 56.78 5.15445 84 86 7.1673019 C2 188 0 0224 0 014 0.251 0 15688 1.457 22.44 5.0266 _
49.10 4 00I44 h
(")
189 0.0224 0 014 0.324 0 2025 1.436 20.78 4 6547 72.76 4 5177759 44.08 N 190 0 0224 0 014 0.399 024938 1.407 19 66 4 4038 3.34847 64.74 3.1824728 k
40.58 2.9639 2.$ 19: 00509 0 0318125 0 0. 0873 59 16 2.4284r 08 N t o .* [J, 5:
a
- - - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ ___ m___ _ _ _ .- _ _ _ _ . _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ - ___.__-+-___m__ _ _ _ _ _ _ _ . _ _ _ _ - - - -___._____._______._______.___J
l l
l n
o Table A.I (continued) E.
[ '"U content Water content Cntical infinite slab Criticalinfinite cylinder Critical sphere
$ Line k.or 4 W entry g '"U/cm' g '"U/gs.S g II,O/cm' g II,O/gS.S k. int Thickness Areal density Diameter Linear density Diameter Mass t.n
& (cm) (kg '"U/m') (cm) (kg '"U/m) (cm) Ikg '"U) E o
vi 192 0.0509 0 0318I25 0 029 0 01813 1.107 68 04 34 6324 160 24 102.6478 241.14 373.7009 y 3
193 0.0509 0 0318125 0.058 0 03625 1 253 36 34 18.4978 95.92 36.7912 I47 06 84.761779 c Gi 194 0.0509 0 0318125 0.119 067438 1.404 23.1 11.7579 61.74 15.2384 94.96 22.821201 195 0.0509 0 0318125 0.183 0.11438 1.474 18.56 9.447 48.16 9 27214 73.86 10.7385I6 l
1% 0 0509 0 0318125 0.251 0.15688 1.51 16 00 8.I44 40.22 6.46683 61.46 6.1871985 197 0.0509 0.0318I25 0.324 0.2025 1.525 I4.28 7.2685 34 86 4.85805 53 02 3.97224 %
l 198 0 0509 0.0318125 0.397 0.24813 1.53 13 06 6 6475 30.98 3.83681 46 82 2.7353355 199 0.0794 0.049625 O. 0 0 929 200 0.0794 0.04 % 25 0 029 00l813 1.078 65.12 51.7053 157.32 154.3354 237.98 560 3270 201 0.0794 0 049625 0.058 0 03625 1.2 34.34 27.266 94.38 55.5482 145 82 128 905I8 Ch 202 0 0794 0 04 % 25 0.I19 0 07438 1.353 21.14 I6.7852 59 50 22.0772 92.28 32 669452 203 0 0794 0.049625 0.183 0.11438 l.437 16 66 13 22 45.76 13.0582 70.74 14.716829 204 0 0794 0.04 % 25 0 251 0.I5688 I.487 I4.18 11.2589 37.8 8.91033 58.24 8 2126578 205 0 0794 0 04 % 25 0.324 0.2025 1.518 12.48 9.9091 32.44 6.56254 49 8 51346060 206 0.0794 0 049625 0.3% 0.2475 1.536 11.3 8.9722 28.68 5.I29 43.88 3 5I25212 207 0.28I7 0.1760625 0 0 1044 53 61 151 019 I49.I6 492.2307 230.28 1801.I676 208 0.28I7 0.1760625 0.029 0.018I3 1.1 I0 27.66 77.918 90.94 182.9674 143 62 436 9495 209 0.2817 0.I760625 0.058 0.03625 1.I46 20 06 56.509 72.38 115.9047 114.88 223 6242 2IO 0.2817 0.1760625 0.119 0.07438 I.217 I6 45 072 52.16 60.1920 82.54 82.9428 211 0.2817 0.1760625 0.I83 0.11438 1.28I 13.13 36.9872 4I 00 37.1904 64.58 39.7264 212 02817 0.1760625 0.251 0 I5688 1.334 II.24 31.6631 33.84 25.3352 53 06 22 0337 213 0:2817 0.I760625 0.324 0.2025 1.379 9 82 27 6629 28.80 18.3510 44.96 13.4049 d 214 0.2817 0.1760625 0.385 024063 I .4 I 9 25.353 25 82 14.7499 40 20 95821 n.
215 1. 0 625 0. O l.ISI E*
p Table A.I (continued) >
- .a
..2"U canent Line ,
Weser content .
k.or Oiticalinfinise sish Oiticalinrmise cylinder Critical .p.;a 3n-enery g '"L#cm' g '"U/gS.S 3H/Wem' .i g H,OrgS-S k.inf Thickness Arealdensety Diameter 9: I Lineer density Diameter - Mass *
(cm) (kg '"U/m') !
216 1. 0.625 0.058 (cm) (kg '"U/m) (cm) -- (k g 2"U) .> ,
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APPENDIX B Suberiticality Evaluation for Chem-Nuclear Systems, Inc., Trench 23 NUREG/CR-6505, 79 Vol.2
APPENDIX B 1
Suberiticality Evaluation for Chem-Nuclear Systems, Inc., Trench 23 Introduction A criticality safety review of the Bamw ell, S.C., Chem-Nuclear Systems, Inc. (CNSI) Trench 2 information (Autry,1998) was perfonned to evaluate the suberiticality of theentburial trench as nuclear criticality safety studies (Toran et al.,1997). The information summarized burial information f Trench 23 (100 feet wide, by 20 feet deep, by approximately 992 feet long) for the months o through April 1978 The monthly " Burial Activity Report (s)" included:
Information Abbreviation used Byproduct Material, millicuries" in this evaluation Source Material, pour:ds' BPM Special Nuclear Material, grams 3M(Ibs.)
Total Volume Buried, cubic feet SNM(g)
- SNM packages Tot Vol.(ft')
Location ofmaterials in trench by #SNM Pkgs.
I SNM Shipments (i.e., no. shipment groupings)
- SNM Grps 11 Source Material Shipments (i.e., no. shipment groupings) #SM Grps Burial position (ft) of Shipments along the length ofTrench 23 from beginning of month to end ofmonth Start ft End ft
! No mformation was provided regarding specific individual burial volumes of SNM, Source M Material. Additionally, no infonnation was provided regarding the mass or placement of Byproduct Material within the trench. It was necessary to apply some assumptions to evaluate the subcritic Those evaluation assumptions and their effects on the evaluation are prosided below.
1 Title 10 of the U.S. Code of Federal Regulations Part 30, {30.4 Defmitions, " Byproduct
[x su to ti ra tio e de r s p gr g pec nuc e December 31,1997. ,
' Title 10 of the U.S. Code of Federal Regulations Part 40, 640.4 Definitions, " Byproduct M means the tailings or wastes produced by the extraction or concentration of uranium or thorium f ore processed primarily for its source material content, includmg discrete surface wastes resulting fro uranium solution extraction processes," December 31,1997.
i Title 10 of the U.S. Code of Federal Regulations Part 70, (70.4 Definitions, " Source Materia means: (1) Uranium or thorium, or any combination thereof, in any physical or chemical form, or (2) o which contain by weight one-twentieth of one percent (0.05%) or more or: (1) Uranium, (ii) thorium (iii) any combination thereof," December 31,1997 NUREG/CR-6505, 81 Vol. 2
Subcriticality Evaluation Trench 23 Appendis B Assumptions in order to perform the comparative evaluation of Trench 23, relative to information provided in NUREG-6505, Vol. I or the Nuclear Criticality Safety Guide (Pruvost and Paxton,1996), it was necessary to assume that the containerized waste matrix, contaminated with SNM, was either a hydrogenous material like plastic, water, wood, and paper or was a relatively ineffective neutron-absorbing material like SiO.. Other assumptions, and their effects on the subcriticality evaluations, are provided in Table B.1 below.
Burial Information The Trench 23 information used for the comparative evaluation was derived from Autry,1998, and is summarized in Table B.2. Footnotes to the table provide explanations as to how the reference data were used to derive the values used in the comparative evaluation. The derived values were then compared with information published in Toran et al.,1997 (Vol.1) and Hopper et al.,1995 to demonstrate suberiticality of Trench 23. The 2
primary values ofinterest for the comparative evaluation were2 g "U/cm'(or g SNM/cm') and kg "U/m2 (or kg 2
SNM/m ).
Results of Comparative Evaluation Information from Table C-2 SiO-soil (S-S) results provided in Toran et al.,1997 (Vol.1); and guidance from Hopper et al.,1995, was compared with information extracted from Tab!c B.2 above (highlighted cells) to determine the magnitude of SNM density increase (i.e., concentration factor) to approach a criticality concern.
The following information was extracted from Table B.2 for the comparative evaluation:
+ maximum density is 2.6 x 10' g SNM/cm',
maximum " infinite" slab areal density is 5.0 x 102 kg SNM/m 2,
+
maximum single package SNM mass is less than 350 g2"U (by license) but calculated to be less than 45 g SNM/Pkg,
+ minimum package volume calculated is 60.6 ft',
maximum single "SNM Shipments" burial is 52369.5 SNM(g), and
- total SNM mass in Trench 23 is 174.93 kg SNM.
The information presented in Table B.3 below provides comparisons between the maximum extracted values from Table B.2 above and information from Toran et al.,1997 (Table C-2) and Hopper et al.,1995 (Table 1).
The concentration factor necessary to alter the Trench 23 values to the reference values is also provided.
As determined i om information in Table B.2, the ratio o '#SNM Grps to #SM Grps is about 1.8 to 1. However, the Eff wt % of 2"U is about 0.08%. Given the reported maximum calculated value of 45 g2 "U per package (typically larger than 4 ft x 4 ft x 4 ft), significant concentration of 2"U (via hydrogeochemical processes after breach of package walls) will necessarily involve the very large masses of SM (nearly 1200 g SM per g SNM) in the trench. The concentration and migration of SNM and SM will be effected through repeated dissolution and reduction of the uranium, thereby significantly reducing the effective enrichment of the uranium.
NUREG/CR/-6505, Vol. 2 82 j
a e
to Table B.1 Suberiticality evaluation assumptions and ramifications Suberiticality evaluation assumotion Ramification (1) ne agaified mass of SNM is assumed to be grams of 2"U without 2"U. Likely an excessively conservative assumption since most NRC fuel cycle facility waste uranium enrichments are on the order ofless than 5 wt % '"U; however, many burials in Trench 23 are reporied2 to be in excess of 90 wt % '"U. (conservative - actual enrichment is typically less than 90 wt % "U)
(2) De volume of byproduct material was ig.wd as being inconsequential. Based upon infonnation for Trenches 38 and later (after October 1981)(l!ouse,1996) the
" Total Fuel Cycle Volume Percent" was commonly 78 - 98% of the total burial volume with the balance burial volume being "Non Fuel Cycle"(Byproduct Material). (competing 2
conservative / nonconservative - ignoring volume reduces density of "U, thereby reducing reactivity ab iion) of'"U but including the volume increases density of""U but provides neutron 1 l
(3) ne number ofSNM packages per"SNM 1 Sh:yments* groups (Grps) placement on the Because of the US NRC license limitation of 350 g '"U per package 1977 during
- 1978 this O trend. plat n ap is equivalent to the number of pened and because of the essentially unrestricted mass of'"U or thorium as Source Material SM packages per " Source Material Shipments ** (SM) per package, it is likely that the actual volume per gram of SM would be less than the Grps placements u *.he trench plat map. %is actual volume reducing the a;,.... per gram of SNM thereby providing greater volumes for the SNM, thereby permits the determination ofSM packages and . d SNM densities in the assumptions. (conservative in that the volume trench volume occutned by the SM packases. per gram of SM is hkely smaller per " Source Material Shipments")
(4) One evaluation asa... J that all SNM Grps were placed together at the determined average ne CNSI burial records indicate that the SNM Shipments Grps were generally intermingled depth of waste for the trench. with Source ab WienMaterial Shipments Grps, thereby reducing the density of "U and providing 2
neutron with (5) One evaluation assumed that all SNM Grps and '"U. (conservative in that the SNM was not dispersed with SM) all SM Grps were intermingled at the he CNSI burial records indicate that the SNM Shipments Grps were generally intermingled determined average depth of waste for the with Source Material Shipments Grps, thereby reducing the density of '"U andg. gproviding neutron absorption with trench. '"U. (competing conservative / nonconservative - providing the diluting volume of SM Grps is nonconservative; however, the inclusion of the SM does g provide unaccounted-for neutron abe,sion) g:
(6) ne average burial thickness of about 10.2 ft m was assumed for the determination of kg he CNSI burial records only show volume of burial and approximate length of trench used 2
for a specific burial month. (competing conservative / nonconservative in that some T 2"U/m .
will be more iu, or more shallow) y burials E
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Suberiticality Evaleation Trench 23 Appendix B i
Table B.2 Raw and transformed data frorn Autry,1998.
Month l Sep-771 Oct-771 Nov 771 Dec 771 Jan781 Feb-78l Mar-781 Apr-78l SumWAseraces
' SM(Ib) l 14 01 79807.2I 9746 5l 18716.6I 56251.4l 196346 91 34121.3l 67625 11 462628 9
' SNM(g) I 1102.01 19646.7l 29131 4l 28986 4l 27127.8 l 52369.5l 7228.7l 9337 7l 174930.2l i Eff wt % (1) l 14.779%I 0 054%l 0.653 %I 0 340%l 0.106%l 0.059 %I 0 047%I 0.030%l 0 083 % !
Tot Vol. (ft') l 17101.5I 125224 Bl 180667.71 159599.2l 110198.8l 129574 El 104820.7I 158087.7l 985275.Il ;
!#SNM Pkgs. l 64l 1273l 1026l 1339l 1153l 1176l 366l 886l 7283' l r SNM/Pkg 17.2l 15 4l 28.4I 21.6l 23.5 44.5l 19 8l 10.5l 24 Oi l0SNM Grps 3l 29l 491 53I 36 33 201 231 246:
iOSM Grps 3l 16l 17l 19I 20 27 11 18l 13ll l0SM Pkgs. (2) 64l 702l 356l 4801 641l 962I 201 693i 4100I
!0Pk es. (3) 128l 1975 1382l 1819l 1794l 2138l 567 1579 113831 jft'/Pke (4) 133 6l 63.4 130.7l D.1 61.4 60.6 184.8 100.1 86 6l
!r SNM/cm' f 5) 4 6E-06I 8.6 E-06' 7.7E-Od 8.7E-06 1.4E-05 2.6E-05 38E-06 3.7E-06 9.8 E-06 !
ig SM/cm'(6) 2 6E-05l 2.9E-02 3 4E-03 7.lE-03 2.3E-02 54E-02 1.5E-02 1.6E-02 2.lE-02i ig SNM!cm'-T (7) 2.3 E-06l 5.5E-06 5.7E-06 6 4E-06 8.7E-06 f .4 E-05 2.4E-06 2.lE-06 6 3E-06; Ig SM/cm'-T (8) 13E-05l 1.0E-02l 8.7E-04l 1.9E-03l 8.2 E-03 2 4E-02 5 2E-03 6 9E-03! 7.5E-03i
!Stan ft (9) 0 0l 105l 280l 300 550 756 750I O 01 lEndil(10) 30 105l 280l 456I 525l 754; 850 964l 964!
Depth ft (II) 5.7 11.9l 10 3l 9.1 4 9l 64l i 1.21 74 10.2, !
Sc SNM/m'(12) 7.9E-03l 3. ! E-02 2 4E-02 2.4E-02 2.0E-02 5.0E-02 1.)E-02 8.4 E-03 3 1E-02' (kg SNM/m8 (13) 3.7E-04l 1.9E-03 1.7E-03 1.6E-03! 1.2E-03 2.6E-03 7.7E-04 4.4E-04l 1.8E-03!
Notes:
(I) One hundred times the mass of SNM divided by the sum of the SNM plus SM masses in grams.
(2) The #SM Grps times the #SNM Pkgs. divided by the #SNM Grps assuming equivalent number of packages per group irrespective of
(
(3) The sum of #SNM Pkgs. plus #SM Pkgs.
(4) The Total Volume Buned, cubic feet, divided by the #Pkgs. ,
(5) The mass of SNM divided by the product of #SNM Pkgs. times ft'/Pkg, (6) The mass of SM divided by the product of #SM Pkgs. times it'/Pkg.
(7) The mass of SNM divided by the trench volume (i.e., the trench width times the burial depth times the difTerence of the End minus the Start), expressed in g SNM/cm'.
(8) The mass of SM divided by the trench volume (i.e., the trench width times the bunal depth times the difference of the End minus the Start), expressed in g SM/ cm'. '
(9) Startmg position within the trench for a burial,in feet.
(10) Endmg position within the trench for a burial,in feet.
(tl} Effective depth of a burial determined from Total Volume Buned divided by the product of the trench width (100 ft) times the difference of the Start minus the End of the bunal, expressed in feet of depth.
(12) The mass of SNM divided by the volume of SNM Pkgs. (i.e., #SNM Pkgs. times it'/Pkg) times the burial Depth, expressed in kg
'"U/m'.
(13) The mass of SNM divioed by the Total Volume Buned, expressed in kg8 U/m'.
l t
SVREG/CR/-6505, Vol. 2 84
. . _ _ ~ . _ _ . __ __ _ ._ . _ _ _ .
a O Appendix B Suberitical Evaluation Trench 23 Table B.3 Concentration factors for criticality concem Suberitical reference values and concentration factors for Trench i information extracted from Table B.2 above Table C-2 !
Concentration Table B.1 (Trench 23 study) (Toran et al,1997) Concentration factor (Hopper et al.,1995) i factor '
2.6 x 10 g SNM/cm' ~1.4 x 10 g2 "U/cm' ~54 1.16 x 10 2 g2 "U/cm' 446 5.0 x 10 2 kg SNM/m2 3.1 x 10' kg2 "U/m 2 62 4.0 x 10' kg2 "U/m 2 80 45' g SNM/Pkg (unit) 2.02 x 10' g2 "U 44 7.6 x 102 g2 "U 17.3
- 350 g SNM/Pkg 2.02 x 10' g2 "U I (unit) by License 5.7 7.6 x 102 g2 "U 2.2 {
i l
Hydrogeochemical Potential for' Uranium Concentration i
The results of this study for determining the potential for criticality following the disposal of u low-level-waste facilities as containerized waste provide information regarding the cumulatil '
precipitation for long time frames in an environment consistent with the CNSI Bamwell, S Figure 7.1). The results, as shown in Figure 7.1, show that increasing cumulative uran about 0.002 g/cm' to about 0.02 g/cm' (a tenfold increase) would require about 7000 though somewhat preferential, the concenuation of the SNM within Trench 23 will beco by the same mechanisms and in the same uranium proportions placed in the trench.
j Conclusions We conclude that the areal density of the buried SNM (disregarding the commingling of SM sufficiently small (i.e., one-eightieth or one-sixty-second of the areal density of concem criticality cannot be achieved as placed in the trench. Even though SNM concentration fac concem could develop over approximately 7,000 years, the same hydrogeochemical mechanisms 3 cause vertical and horizontal 2 migration of SNM will also migrate SM, thereby further reducing t criticality by2 reducing the effective "U enrichment in the blended materials to well below I wt % 2"U
(-0.08 wt % "U).
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i NUREG/CR-6505, 85 Vol. 2 l
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ATTACHMENT 5 - DIFFERING PROFESSIONAL VIEW ,
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n e NOTE TO: David Brooks October 2, 1998 FROM: John Bradbury EUBJECT: CONCERNS ABOUT POST-DISPOSAL CRITICALITY at LLW SITES This is a differing professional view that takes exception to 1) the basis for asserting that the probability of criticality at LLW sites is sufficiently low that the NRC does not consider this event a significant public healt.h and safety issue, and 2) that, accordingly, the analysis recommended by the ACNW need not be pursued. Based on the work to date, the probability for reconcentration and criticality at a LLW site has not been determined quantitatively. Consequently, the assertion that reconcentration and criticality is of ". . apparent low likelihood. . " (Recommendation Section of Commission Paper) cannot be adequately supported.
As a geochemist who has studied the formation of ore deposits, I am aware of processes that mobilize and reconcentrate metals into ores in the natural environment. Generally, the formation of ore deposits requires the mobilization of trace quantities of metals dispersed over vast regions into local areas with high concentrations of the metals. Although the formation of ore deposits requires special conditions, and thus, may be rare (spatially and temporally) in nature, it can be suggested that LLW sites and decommisyioning sites are small man-made features in which all the correct ingredients have been placed that may promote reconcentration of metals. When the metal is fissile, evaluations of the probability of assembling a critical mass are suggested.
On reviewing the ORNL report on The Potential for Criticality Following Disposal of Uranium at Low-Level Waste Facilities Volume 2; Containerized Disposal, NUREG/CR 6505 Vol.2, it was determined that some important aspects to the Barnwell site were not considered that might weaken the conclusicus of the report. The review involved other documents to be used as background and supporting information. These other documents included NUREG/CR-6284 Criticality Safety Criteria for License Review of Low-Level Waste Facilities,1994 NUREG/CR-6505 Vol.1 The Potential for Criticality Following Disposal of Uranium at Low-Level Waste Facilities: Uranium Blended with Soil, 1997 Burial Procedures,1996 in letter f rom House, W.B CNSI to T. Harris, April 9, 1996.
Dennehy, K.F. and P.B. McMahon,1987 Water Movement in the Unsaturated Zone at a Low-Level Radioactive Burial Site Near Barnwell, South Carolina, U.S. Geological Survey, Open-File Report 87-46.
Although the Barnwell report was intended to study post-disposal criticality at a site that buried containerized waste, as suggested by its title, the approach taken was strikingly similar to that taken in the Envirocare report.
In both cases, uranium migrates in 1-D through a soil-like material to a location where it drops out of solution, due to local geochemical conditions that are favorable to precipitation or sorption.
Failure to consider important ef f ects of the burial of containerized waste in the Barnwell report lead to gaps in evaluating probable situations that could pose a criticality concern. For example, current burial practice at Barnwell involves the use of concrete vaults whose dimensions are approximately 10'x 11'x 10'. These vaults are designed to contain thirty-six 55-gallon drums.
The vaults are stacked three high. Burial Procedure S20-OP-030 requires that packages containing more than 30 g "t shall be placed in the lower level vault. This lower vault can contain up to 7.2 Kg t, consistent with license amendment limits. The top and middle vaults pref erentially contain SM. The vaults have a single drainhole in the center of their floors. Dennehy and i McMahon,1987 demonstrated with experimental trenches designed to simulate Barnwell LLW trenches that transient perched water can occur in the bottom of Attachment 5
_ - .7 l
i the trench if the cap is poorly designed. Their trenches were filled with a 1 soil-like material . However, when concrete vaults are buried, pref erential paths along the sides of the vaults are anticipated. It is likely that preferential flow could contribute to transient perching. Sump water level data may suggest localized transient perched water as the result of preferential flow past the well seal. If transient perching occurs, the bottom vault could flood through its drainhole, whereas the top and middle vault would tend to remain high and dry, as long as their lids last. j i
This scenario would suggest that the commingling of SNM and SM to reduce j enrichments as assumed in the Barnwell report could be unusual. Instead !
modeling should assume SNM reconcentration f rom the lower vault at the i enrichment contained in that vault. The drain can act as a focussing j mechanism. Consequently, spherical configurations could be considered. The Envirocare report calculates that only 1.5 Kg 22t in a sphere 51 cm in diameter is needed for criticality.
Besides failing to consider important aspects of containerized waste, the Barnwell report did not use license amendment limits as initial conditions as was done in the Envirocare report. If it had, application of analyses in NUREG/CR-6284 would have suggested that Barnwell license amendment limits alone could not ensure suberiticality on emplacement. For example, license amendments have allowed burial of packages containing up to 350 g 8b U per package whose base was a minimum of 2 f t .2 If the licensee had placed cardboard boxes next to each other in a planar array extending over an area 2m x 2m, with the conditions described above, and covered it with 2.4 m of sand, on resaturation (i.e. , optimally moderated) , NUREG/CR-6284 suggests suberiticality could not be ensured. If hydrologic conditions change such that the reflector sand dries out, NUREG/CR suggests criticality could occur.
If suberiticality on emplacement can not be ensured, how can post-disposal criticality be so unlikely that it leads to a recommendatien that work in this area should be stopped?
Based on the omissions in the Barnwell report described above, and the limited or uncertain technical information used to support the conclusion that post-disposal criticality is of " apparent low likelihood", it is recommended that the staf f support Option 2, and implement the ACNW recommendation. My concern is that a decision to cease work on LLW criticality may send a message to the LLW community as a whole that the NRC thinks LLW criticality is no problem. 3 If the current disposal practices at Barnwell are any measure of practices at other sites, it is suggested that licensees may also have little concern for LLW criticality. Burying potentially critical masses of 8't in vaults that can be flooded and that contain a single drain for focussing should indicate the need for heighten awareness of LLW criticality concerns. Assignment of low probabilities for post-disposal criticality requires stronger evidence )
than that currently provided. J The ACNW recommended that the staf f conduct a limited-scope study to reasonably quantify the associated risk. Current analysis supports this ;
recommendation. is a draf t report written by R. Codell describing a risk assessment methodology that could be applied to LLW sites. It is suggested l that this type of approach could reduce uncertainty and provide better support j for NRC staff conclusions on LLW criticality. Furthermore, this approach is consistent with the ACNW recommendation.
i is a table that lists categories or properties of conditions at [
Barnwell along with the associated pros (lower probability of criticality) and !
cons (higher probability of criticality) . ,
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lh. A m t Risk Assessment for a Nuclear Criticality Accident at the Parks SLDA Site by Richard Codell l
- 1. Background on criticality issues at the Parks Sl_DA site !
The Parks SLDA site received waste material from the Apollo Fuels plant in the 1960's and j
1970's. Since'some of these wastes contained enriched uranium, it was necessary for the staff to consider the likelihood of a criticality event. The staff evaluated the inventories and !
4 enrichments of uranium disposed at the site based on manifests, regulatory limits and onsite measurements. The staff concluded that a suf6cient inventory of Hssile U-235 exists on site.
but that it would have to be concentrated greatly to lead to a criticality event The staff l evaluated a number of scenarios that could potentially lead to criticality, ultimately reaching a j conclusion, stated in its Safety Assessment, that criticality it is not credible at this site.
l We make criticality cr'ulations in Part 11 based on (1) conservatism, (2) identincation of sub-i processes necessary for criticality, and (3) estimates of the ranges of parameters of these sub- !
processes to place the staff's conclusion into a numerical range of probabilities and consequence. In part 111. we evaluate the consequences of a postulated criticality event at the j site based on an estimate of the likely power range of such an in-ground reactor. We i calculated the dose to a person residing on site from direct radiation and consumption of i radionuclides produced in the reactor. '
1
- 11. Probability of criticality at the Parks SLDA site i A. Introduction We have taken an event-tree approach to estimating the probability of criticality at the Parks site because it has been suggested in some other studies of criticality (Kastenberg, et al, 1997), and is similar to the approach taken in performance assessment in general. It is a four -i step process: t
- 1. List the scenarios that express the range of possible futures for the site.
- 2. Estimate the probabilities of the scenarios that could lead to criticality eliminating the I others. -
- 3. For the scenarios identified in (2) above, estimate the ranges of parameters for phenomena that are necessary to calculate whether criticality could occur.
- 4. Calculate the conditional probability of criticality in the identined scenarios.
The overall probability of criticality is then the sum of the products of the scenario probabilities and their respective conditional probabilities. Figure 1 is a general flow chart showing the steps in the procedure.
1
- 4 B. Identitiine Scenarios that could i ead to Criticalits The tbliowing are a list of the minimum requirements for criticality to occur in a lanJ!ill with initially suberitical concentrations of l'-235:
- 1. Sufficient inventory of l'-235
- 3. A mechanism to concentrate dispersed uranium into a smaller volume
- 4. Neutron poisons limited to low levels
- 5. Good neutronics There are probably sufficient quantities of enriched uranium in several of the trenches at the ;
site (we did not consider migration between physically distinct trenches). Under optimal conditions of light-water moderation it takes less than one kilogram of 100% U-235 to have a critical mass (Knief,1985). This quantity increases for lower enrichments and for conditions where there are other constraints such as low porosity and presence of materials other than >
water. Based on neutronics calculations using SCALE (ORNL,1993), nominal soils at the Parks site, ideal spherical geometry and 20% enrichment, the minimum quantity of U-235 likely to lead to criticality is about 3 Kg. Estimates of U-235 content at the site range from 27 Kg based on burial manifests and operating guidelines to 650 Kg based on water and sediment samples.
The staff evaluated severa! mutually exclusive scenarios looking at concentration mechanisms 4
necessary to focus the uranium to a critical configuration. Staff considered the following scenarios for further concentration of uranium:
- 1. Areal loading - The staff first used the recommendations of NUREG/CR-6284 (Hopper,1995) as a screening tool for maximum areal loading (i.e., about 1 kg 100%
2 enriched U-235/m surface area). This model assumes implicitly that the contents of the trench has compressed vertically to an optimal infinite-slab geometry with water moderation, and also has a built-in factor of safety of about 20%. Estimates of uranium content from direct measurements were significantly below the recommended areal loadings at any measurement site, so we concluded that there ..ould have to se a fur *er focusing mechanis-- to attain a :
potentially critical mass.
- 2. Horizontal migration along a trench - This model considers that there is a horizontal flow along the long axis of any of the 9 trenches in the upper site area, with non-mechanistic precipitation of uranium to an optimal slab geometry at the downgradient end. Water flows along the length of the trench, but also out the bottom. Trenches 2 and 6 were the most likely trenches where this mechanism could be effective in producing a criticality event, but were rejected on the basis of a conservative analysis of potential accumulation at the trench end, which is described later.
2
- 3. I Concretions and sorption processes - Staff considered and reieeted as potential mechanisms for the formation of concretions by molecular diffusion. and sorption onto hig sorptive iron ox3 hydroxides from rusting metal.
4.
Bathtub scenario - Staff considered that there could be a portion of a trench sealed except for a drain in the center. and that dissolved uranium would precipitate non-mechanistically to an optimal spherical geometry in the drain. Fdr trench 2. the trench with the greatest estimated quantity of U-235. the " bathtub" is approximately 1/7th of the trench :
volume. approximately 10 x 10 meters length and width, and 3 meters thick (Figure 2). Under the stated conditions. the uranium would accumulate in a small area of the convergent flow held. The dimensions in trench 6 are slightly larger. I10 x 110 meters.
Staff considers that the bathtub scenario in trenches 2 and 6 are the only possibilities that could lead to criticality at the Parks site. The staff then went on to consider the subjective probability. of factors necessary for criticality in each of these scenarios, and the conditional l
probabilities of criticality.
C. Bathtub Scenario in Trench 2 i 1.
Scenario nrobabilitt for bathtub in trench 2 I J
For criticality to be possible. a large number of processes must work in unison to collect the dispersed uranium into an optimal configuration. Unlike the formation of an ore body. the supply of uranium at the Parks site is more limited, so these processes must work with high economy and efficiency so that most of the uranium available collects in one place. The probability of the bathtub scenario can be estimated by considering the probability of the '
coincident factors necessary:
- a. Factor - Impermeable bottom and sides for the bathtub.
Probability range - less than 0.001 - 0.01 1
i
' o :tionale for choice of probability - Elicitation from the involved staff c i... ep with the following rationale for the choice of the range of probabilities:
l (1) The trenches are mainly soil with waste interspersed. There was no deliberate attempt to create an impermeable liner for the trenches.
(2) Low permeability of the weathered bedrock beneath the soil trenches or a tar roof l
' disposed and burned in the trenches have been suggested as possible mechanisms to form an impermeable basin.
(3) The trench bottom and sides does not have to be impermeable, only have higher !
permeability at places which would preferentially allow water to be funneled to a single place. l i
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i
with relatively small losses elsewhere.
b, Factor - A single drain in the center of the bathtub through which all infiltrating water must pass.
Probability range - 0.01 to 0.1 !
l Rationale for choice of probability i
(1) Hydrographs of water levels in the various Temporary Well Sampling Points (TWSPs) in i the trenches shows that some of them apparently drain much faster than others. There are j several possible explanations. ranging from permeability differences in the trench filling l materials, leakage around the collars of the TWSPs. draining through the bottom of the trench ,
and errors in measurement. j (2) There would have to be a single drain rather than multiple drains in the bathtub since the ,
latter would split up the potential accumulation into smaller parts. ;
(3) Lineaments from areal photographs ~suggest there are fractures through the site. However -l recent field reports ( ARCO.1996) indicate that the lineaments are not associated with i I
fractures. :
l
- (4) Fractures through the rock, if plausible. would lead to a " slot" of high permeability. Such l
l a shape would be far from ideal from a criticality standpoint. and in this instance would not i l
be considered a single circular drain. Criticality from accumulation along a vertical fracture l has not been considered here, but does not appear to be as likely as the cases considered.
l
, i j (5) For maximum efficiency, the drain would have to be close to the center of the bathtub. .An j
- . off-center drain would allow faster penetration of clean water from the surface, therebv ;
l capturing less of the total material present.
- c. Factor - The ex' .nce of a reductant or sorbant in the single drain (and only ir '.e l I
single drain) for accumulating uranium by chem: al reduction to tetravalent form or sorption.
Probability range - 0.001 to 0.01 Rationale for choice of probability: ;
(1) Uranium would have to be collected at a single location because the inventory of U-235 in any trench or portion of a trench is limited. The diameter of a spherical critical mass l containing the minimum amount of 20%-enriched uranium predicted by neutronics l calculations, about 3 Kg U-235. is 0.62 meters.
. i
)
4 l t
f l
i -
4 (2) Material suggested capable of concentrating the uranium were a wood log. large piece of base metal. or a barrel of Kimuipes. The applicant inJicates that there was no etidence or large logs that would have been buried in the trenches ipersonal communication. Tom Potter.
1997). but large pieces of steel machinery may be burieJ.
(3) It is not likely that a barrel of paper or other organic waste could survive the many thousands of years that it would take to accumulate sufficient uranium. since most such material below the water table would decay in a shallow landfill (Greentield.1990).
(4) The coincidence of the oorbing or reducing agent with the location of the drain would be unlikely. The log would have to occupy most of the drain so that the uranium-laden water would filter through it. or. else part of the uranium would flow-by without being captured.
(5) Even if there was such an effective mechanism to collect the uranium, it is just as likely that there would be m :;iple places as a single one, thus effectively spreading out the accumulation among several locations. The large affinity of uranium for iron oxides and oxyhydroxides (Hsi.1985. Waite.1994) from rusting iron and steel waste widely dispersed in the trenches would also serve to tie-up uranium in place.
- d. Factor - The ability of the reductant or sorbant in the trench to reduce or immobilize most uranium flowing through it with high efficiency.
Probability - 0.01 to 0.1 Rationale for choice of probability - Uranium is known to be accumulated on organic materials (Wood.1996). and may be concentrated by certain organisms like iron-reducing bacteria (Duff.1997). These bacteria must remain in an anaerobic environment. however, and the predicted invasion of dissolved oxygen in the bathtub would eventually destroy this environment and cease the concentration process. Observed concentration factors (i.e.. Ka) for uranium in evaporation ponds where these bacteria are present range from about 7000 to 70.000. Doi (1975) demonstrated a Ka on peat of about 15.000. Still higher concentrations of uranium have been found in natural lignified wood samples. Meunier (1990) reports concentrations of up to 19% l y weight for sma!: agments and up to 6.2% for samples up te 30 cm in diameter. apparently sorbed or complexed whh the carbonaceous material but not reduced. Projecting the latter figure to the 62 cm diameter sphere (about 9 times larger if the sample was a sphere) determined for the minimum critical dimensions and 2000 Kg/m) and an enrichment of 20% at the Parks site would yield about 3.1 kg U-235, about equal to that required for an ideal spherical reactor. These are unusually high cre concentrations for most natural bodies, and cannot be called typical of most ore bodies or of conditions that exist at the site.
T ae concentration factors observed in these experiments may be more typical of adsorption of aosorption of uranium into living and dead organic material. Enrichment factors in the range 10.000 to 100.000 on organic material are widely reported. and are the result of uranyl cation 5
l .
l l
l exchance rather than reJuetion (Wood.1996L .\leasurements on ponJ sediment indicateJ that most of the uranium was in the -6 oxidation state. esen though the sediments themselves l were highly reducing (Duff.1997L This observation may indicate that reaction kineties for reducing uranium on organic material are slow (.\leunier.1990). It is likely therefore that the uranium captured in humie or other organic material at low temperature is by sorption or l complexation rather than reduction. The importance of this observation is that the l
concentration factor for precipitauon of reduced. uranium is essentially unlimited, but sorption
! processes would be limited by the strength of the mass action. determined by equilibrium coefficients. Accumulation of sufficient densities of uranium by sorption alone would be unlikely, unless there was the coincidence of both high dissolved concentrations and high K, Observations of uranium concentrations in TWSPs shows that concentrations and Kfs are negatively correlated; i.e.. high concentrations are generally found with low Kfs and vice-versa.
l I
l Although the free enerries of formation for tha uranium species would suggest potential l reduction of hexavalent uranium on metallic iron available in the waste. there is no compelling evidence that this would occur with high efficiency at low temperatures. While l iron could serve as a reductant. it is also an effective neutron poison, so the mass of uranium accumulated would have to be greater than the stated minimum. Furthermore. tetravalent i uranium may also be mobile under high pli conditions or the presence of organic ligands.
I particularly fultic and humic acids (Duff.1997L which are likely to exist at a temperate site j covered with live and decaying segetation.
- e. Probability of bathtub scenario Overall probabilities would be multiplicatise if they are uncorrelated (the most conservative case). None of the above factors appear to be correlated, so the estimated probability range of the bathtub scenario occurring in 15.000 years of postulated trench stability is therefore:
l P(bathtub) = (0.001 to 0.01) x (0.01 to 0.1) x (0.001 to 0.01) * (0.01 to 0.1) = 10" to 10*
Furthermore. it could also be interpreted that this probability should be spread over the estimated 15.000 years of postulated trench s.arility to calculate a probability per war.
- 2. Conditional probability of criticality under bathtub scenario The conditional probability for criticality under the bathtub scenario can be estimated from the manipulation of several other factors pertaining to this scenario:
There was a wide range of estimates of inventory of U-235 in trenches 1-9, ranging from about 27 to 650 Kg. The inventory from downhole gamma measurements were estimated with a bootstrap statistical procedure (Efron.1982). Downhole gamma measurement at Trench 2 indicated that it had the highest inventory of any trench. The bootstrap method gave a range 6
- - _ - . = - - _ _ . . - =.
of 11.4 to 62.1 Kg l?235. with a median of 27.6 kg and a 95th percentile of 45 Kg. Ihj3 estimate employed both positise and negative measurements because they are all statisneally valid. but conservatively ignored spatial correlations. which would tend to reduce the 3pread of the distribution ifit had been included. and would therefore leaJ to a lower estimate of the probability of criticality. The inventory in the bathtub is taken to be .17th part of the trench l
(because there are 7 TWSPs in this trench). The bootstrap distribution of trench 2 is shown in Figure 3.
The inventory estimate for the bathtub scenario was based directly on the bootstrap results.
This is a rationally conservative-but not extreme--approach to estimating the inventory.
Trench 2 is estimated to contain about 33.7% of the total U-235 in trenches 1 through 9. i Extrapolating the range of the suggested bootstrap estimate to all trenches would predict a minimum inventory of 33.8 Kg and a maximum inventory of 184 Kg U-235 in trenches 1-9
- b. Amount of uranium needed for criticality - Assuming that uranium would accumulate at the location of the accumulating medium is only part of the requirements for a potential criticality event. The reactivity of the resultant mass of uranium would depend on a number of factors such as geometric contiguration and water content. Under nominal soil conditions at the Parks site and assuming a spherical mass of homogeneously dispersed uranium. the minimum mass of U-235 necessary for criticality is about 3.1 kg for 20% enrichment. Any deviation from the optimal contiguration of course will lead to a greater mass required for criticality. Factors that would tend to increase the needed mass are:
- (1) A geometry other than a sphere i (2) A sphere larger or smaller than the optimum size (3) Heterogeneity of uranium in the accumulated mass (4) Lower water content
]
(5) High content of neutron poisons such as iron l (6) Lower enrichment l Conversely, factors that would tend to decrease the needed mass from 3.1 Kg are:
(7) IQher water ~ntent (8) The presence of a high-efficiency moderator such as carbon, or possibly dry silica (9) Higher enrichment j
Concerning point (1), it is hard to imagine a process that would allow the uranium to accumulate to a perfect sphere. Concretions are often spherical, but their slow rate of growth l
by diffusion would rule out the acciimulation to sufficient size to form a critical mass. The l permeability of the wood may be higher than or less than that of the surrounding soil, l depending on such things as its species, alignment in the flow field, and state of decay. If there was a perfectly spherical mass of reductant in the drain that was porous to the flow, and the rate of accumulation of the uranium was fast, the uranium would accumulate at the
! leading edge of the mass. Thus. the accumulated mass would be a crescent shape rather than a
- i 7
l i
i
sphere. A slower rate of uranium accumulation could lead to greater penetration into the maw.
but lower collection efficiency oserall because part of the uranium woulJ leas e the .9 stem before it could react. Furthermore. as uranium accumulated. it uould begin to till the pores of the reductant mass. changing the flow properties so that water would likely be diserted from the tilled-in areas. Desiations from a spherical shape lead to lower reactivity systems ihr an equivalent mass of U-235. Mathematical model studies of consergent flow in the bathtub j scenario indicate that the shape of the precipitated mass would be far from an ideal sphere. l Concerning point (2L the reductant mass or the drain size will not likely be exactly right to create a critical mass of minimum size. Just changing the radius of the sphere will have a big effect on the mass needed for criticality. Figure 4 shows approximately the dependence of l necessary U-235 mass versus sphere radius for a set of SCALE runs that were within 5% plus or minus of criticality fbr nominal Parks soil. For this particular series of runs. the minimum quantity of U-235 required was about 3.1 Kg. with a radius of about 29 cm. The necessary mass increases for smaller and lareer radii. A sphere twice the diameter would require a critical mass of at least 4 times the minimum. The minimum U-235 critical mass should therefore be just a point on the probability distribution of necessary critical masses.
Also concerning point (2). consider that the optimally spherical mass of 3.1 Kg U-235 were l compressed into a short cylinder of the same density and a thickness 1/10 the original diameter of the sphere. lo hase the same lesel of criticality. the cylinder would require about 2.4 times the mass. or 7.4 Kg (1235 according to Figure 4 in LA10860MS (Paxton,1986).
The areal density of this e>linder would be 3.7 Kg U-235'm2 . significantly higher than the l 2
screening estimate of 1 Kg m recommended in llopper (1995) used previously. j Concerning point (3). natural materials tend to be heterogeneous, and it is not likely that there i would be homogeneity in the uranium concentration or the reductant to allow a homogeneous .
accumulation. Furthermore. the relatise speed of the reduction reaction would dictate the degree of homogeneity of the precipitated uranium mass. despite the homogeneity of the materials present.
I liopper (1995) determined that critical loadings for dry silica were in some cases lower than I water-moderated cases. acerning points (4.. (7) and (8L there is likely to be a watcr l content characteristic of the surrounding soils. Di conditions are improbable because of the l
' temperate climate of the site. and high capillary suction of the fine grained materials. While there may be some variance in the porosity in the present-day trench contents because of the disposed materials, most of the contents of the trenches is soil. We can expect that most of the added organic material and base metal will degrade and the deformable clay and silt soil will compact itself over the many thousands of years that would be necessary to attain criticality. Actual porosity measurements in nearby soils are less than the 40% conservatively assumed in the neutronics calculations (ARCO 1996).
Concerning points (5) (poisons) and (8) (better moderators). there are a number of materials disposed in the trench. including a large amount of base metal and carbonaceous materials 1
8 i
such as paper. rags. wooJ. rooting tar and coal (naturally occurring as stringers in rock and soil of the regiont Iron is a neutron poison and carbon may be a beuer moderator We hat e done some neutronies calculations on the etTect of iron. but consider only a n.,minal carbon content of the soil of 4.3% by weight in the criticality calculations. The stof does not expect excess carbon to have a significant effect on criticality, honeser. NUREG CR-6284 (llopper. !
et al.1995) recommends that the carbon content of the disposed waste not exceed the U-235 content by a factor of greater than 20 on a neight basis. Neutronics calculations presented indicate a small decrease tabout 6"o) in the required mass of U-235 in a particular system tiir a carbomU-235 mass ratio of about 80. For the minimal critical contiguration of 3.1 Kg U-235 in a sphere 0.62 meters in diameter consisting entirely of wood with a dry density of 800 ,
Kg/m*. the carbon content would be approximately 44 Kg. for a carbon /U-235 ratio of about -
14.
i Concerning factors (6) and (9). enrichments vary from place to place in the trenches. The average enrichment of aoout 23 o' occurred in Trench 7. but this trench has a projected !
inventory of only 39 Kg U-235. For trenches 2 and 6. the average enrichments are about 19 !
and 10%. respectively.
Each of the factors discussed above may increase or decrease the minimum amount of U-235 required for criticality. The >talT has chosen the distribution of U-235 mass necessary for
! criticality to range from somewhat less than 3 Kg. to considerably more than 3 Kg U-235. A i triangular distribution was used to take into account these modifying factors. The minimum amount of 2.5 Kg accounts thr carbon moderation. higher enrichment and greater water l content. The apex of the triangle would be the 3 Kg from our neutronics calculations with l
nominal soil. The upper end of the range would be 15 Kg to take into account the likely non- '
ideal shape. heterogeneity. lower enrichment. lower water content, and neutron poisons.
l c. Fraction of uranium collected - There are a number of factors that would increase or reduce the rate or time span for uranium collection. Factors that would tend to reduce the rate, efficiency or operational period of uranium accumulation are:
l (1) Diversion of flow away from the reductant i
(2) Decreased infiltration (3) Destruction of the reductant by decay or from invading dissolved oxygen 1
l (4) Destruction of the trenches by erosional or man-made forces
(
Factors that would tend to increase the rate, efficiency or operational period of uranium i accumulation are:
c e
(5) Increased infiltration s
9 2
(6) Increased dis 3ohed concentration of uranium l
(7) Longer trench stability Concerning point (1). even for a drain with all the material being diverted through the reductant. the accumulation of uranium in the pores of the reductant might change flow patterns, forcing flow around the reductant mass.
Concerning points (2) and (5), infiltration could be different from the 0.12 m'yr estimated by the applicant. Earlier estimates ofinfiltration were as high as 0.25 m/yr based on regional data. Engineered improvements to the site would likely reduce the infiltration even under l partially failed conditions. Climate change over the many thousands of years that would be necessary to accumulate the uranium could increase or decrease intiltration.
l Concerning points (4) and (7), trench stability would be increased by engineered improvements to the site. Erosion of the unimproved site would depend on the site cover (bare soil to dense forest). downcutting from Dry Run, and frost actions.
Concerning point (3). modeling studies with the stylized bathtub scenario suggest that flow from the center part of the basin will reach the drain in less than 6000 years under nominal l conditions and even less time under higher infiltration conditions. Additionally. the water l table would be depressed there. Therefore. the fraction of uranium collection will be reduced I because the higher flows at the center will speed the invasion of dissolved oxygen, possibly i
destroying the reducing environment and possibly re-dissolving uranium previously l precipitated.
l l Concerning point (6). the accumulation of uranium was based on low concentrations of l dissolved uranium currently observed in the trenches. It is possible that this concentration I
could increase with time because more uranium would be mobilized by oxidation in the upper layers of the trench. However concentrations used in the models are consistent with 1( values representative of soils in general (Sheppard et al.1990).
The staff used the curves of accumulation versu time generated from geochemieN *ansport and hydraulics analyses to estimate the fraction accumulated in 15.000 years under nominal conditions. The estimates then were adjusted to account for the increased or decreased collection efficiency from some of the above factors. This factor is represented by a triangular distribution, with a lower collection fraction of 0.28 in 15.000 years based on results from a hydraulic model for oxygen invasion at less than 6000 years with an infiltration of 0.12 m/yr, an apex of 0.54 based on accumulation in convergent flow model without taking account of oxygen invasion, and an upper limit of 0.9 based on the 0.25 meter / year infiltration, no redissolution of uranium. and total capture. Applying this fractional accumulation would tend to conservatively overestimate accumulation of U-235 if the uranium concentration were solubility limited.
10
.._ _._ _ __ _ _.. _ _. _ _ ._ _ _ __..~ _ _ _
. i- ;
i
- 3. Combinine Factors to nredict criticality :
A thousand samples were taken from the proposed distributions tbr the inventory in the !
bathtub and collection efficiency in 15.000 years. SampleJ values from these distributions i were then multiplied by each other, and compared to a sample from the distribution fbr i
inventory necessary for criticality, if the quantity of accumulated U-235 exceeds the amount necessary. then a count is added to the criticality fraction. The total number of criticalities j divided by 1000 is then the conditional probability of criticality given the bathtub scenario.
- 4. Resuhs For the stated conditions, the conditional probability of criticality for the bathtub scenario in !
trench 2 was estimated to be 3.3'6 A histogram of the ratio of the accumulated saass to the i
needed mass for the 1000 samples is shown in Figure 5. The staff believes this is a robust :
estimate of criticality becau3e it conservatively ignores the influence of freshwater I breakthrough on the accumulation of uranium in the drain over most of the range. !
i The total probability thr criticality is the product of the scenario and conditional probabilities. t i.e.. the probability that there would be a criticality in the next 15.000 years is: I Pteriticality) = t ltr" to 10*
- 0.033 = 3.3 x 1042 to 3.3 x 10'
, or on an annual basis:
p(criticality) = 2.2
- 10* to 2.2 x 1042/ year.
D. Bathtub scenario in trench 6 Based on corrected TWSP downhole gamma measurements, the area surrounding TWSP 6.1 could contains approximately 55.5 Kg of uranium (estimated by the bootstrap method) enriched up to 24.2*L. Nieasured liquid concentrations in this TWSP are considerably lower than those in trench 2. indicating that a higher K, of 4810 ml/g should be used in the model.
- For a surface area v.' 114 m' representing the region around TWSP 6.1, the esti..nted mass of U-235 which could accumulate in 15.000 years would be 1.03 Kg for 0.12 m/yr infiltration and 2.15 Kg for 0.25 m/yr infiltration. These numbers are considerably below those for the calculations in Treach 2 and therefore this scenario is considered to be not credible.
E. Bathtub scenario in uench 6 with mobili7ation of uranium The relatively low rate of uranium accumulation in the bathtub model for trench 6 is a reflection of the low measured concentrations and high K,. If the Ko in the trench were to I
decrease, more uranium would be available to migrate in the dissolved phase. The high K, may be caused by sorption onto organic components 'in the soil, and that if this material were 11 1
i
to be degraded, then the K; could Jecrease a nell. Ihmeser. tb mesurcJ natural carbon content of the soil is approximately 4.3"o ( ARO L l'%. l he total Jegradation of the carbon in the soil is unlikely because the infusion of oxygen bs intihratine water is 31ou. anJ'at ieast some of the carbon (e.g.. coalj is not reaJily oxidized. Alaterial3 such a3 iron oxide.s froni rusting metal also dispo. sed in the trench coulJ sers e as powerful sorbants. even in the absence of organic materials. Finally. it would be .somewhat incon3istent to allow the degradation of carbon responsible for the high K, of uranium. but not allou the total degradation of the carbon leading to its accumulation into a concentrated mass.
If K, suddenly is reduceJ from 4810 to 200 ml g. and all other mechanisms in the bathtub scenario work as specified. then there would be a high probability of criticality in 15.000 years. Ilouever this higher probability would be offset by the low probability of this additional mechanism to mobilize the uranium. To quantify the probability of this scenario in a conservatis e manner. the analysis for Trench 2 was modified in the following way:
- 1. The area of the trench was increased to 114 m to reflect the larger volume of soil surrounding TWSP 6.1.
- 2. The shape of the probability distribution of inventory in the portion of trench 6 surrounding TWSP 6-1 was assumed to be the same as the inventory in trench 2. but aJ.iusted for the median salue.s of trenches 6 and 2.
- 3. The shape of the distribution of fractional accumulation was based on the hydraulie bathtub model lbr each salue of K, used C00.1000 and 4810 ml/g;.
Table 1 shows the breakthrough time. collection efficiencies at time of breakthrough.15.000 years at 0.12 m'yr infiltration.15.000 years at 0.25 m yr infiltration and probability of criticality. This table reflects the bathtub accumulation scenario adjusted for trench 6. The second column. " breakthrough time for clean water". is the time at which the clean infiltrating water from the center of the bathtub would reach the drain. signaling the probable end of the accumulation process as oxygenated water encounters the sorbant or reductant. The third column. " Fractional Accumulation at breakthrough" is the amount of total uranium in the bathtub accumu'ated at time of breakthrraph. The tourth and fifth columns. "Fracti al Accumulation at 15.000 Years", is the accumula: in in the drain, conservatively ignoring the potential effects of clean-water breakthrough. The last column, " Probability of criticality". is the result of statistically sampling the accumulation of uranium in the drain, and quantity of uranium needed for criticality, assuming the same dimensions and enrichment as the trench 2 analyses. Assuming that each final K, was equally probable. the conditional probability of criticality under this scenario is 36%.
i 12 l
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Table 1 - Probability of criticality in trench 6
.\1obilization scenario K , - m !!g breakthrough Fractional Fractional Fractional Probability time for Accum. at Accum. at Accum. at of criticality clean water, breakthrough 15.000 yr. 15.000 y r. in next years 0.12 m yr 0.25 m'yr 15.000 years infiltration infiltration 4810 never 0.076 0.076 0.162 0.004 1000 9500 0.22 0.311 0.503 0.346 200 <2000 0.22 0.758 0.925 0.731 If we assume the scenario probability was the same as that for the trench 2 bathtub scenario.
then the overall probability of criticality for this case would be:
l P(criticality) = (10+' to 10 ^ )
- 0.36 = 0.36 x 10" to 0.36 x 10
or on an annularized basis p(criticality) = 2.4
- 10'" to 2.4x 10~"' year.
F. Hori7ontal transport in trench 6 The applicant used a water budget for trench 6 to suggest approximately equivalent amounts l of water leave the bottom of trench 6 as lease the downgradient end. The staff estimates that approximately equal quantities of uranium would be transported past the edge of the trench as through the bottom. The integrated quantity of U-235 leaving the trench through the side j could be no more than about 7.5 Kg even for the unlikely complete mobilization of uranium -
in the trench. If spread over the side area of the trench, the mass loading would be no more 2
than about 0.25 Kg U-235/m . far short of the estimated areal density for an optimally moderated slab geometry without a substantial focussing mechanism. Consideration of this scenario coupled with an additional focusing mechanism would be captured in the convergent ,
bathtub scenario for Trench 6 presented above. Therefore. horizontal transport in trench 6 did not contribute to the overall probability in this exercise. Furthermore, consideration of horizontal trausport past the edge of Trench 6 would diminish the likelihood of the bathtub ,
scenario being effective, and reduce the overall probability of criticality in that trench. "
13
G. Probability and consequences of muhiple potential criticality event.s The conceptual model for the bathtub scenario in trench 2 con.siders that there is 17 of the trench volume and uranium imentory per bathtub. Ihe probability of criticality in that bathtub
- was then determined from the probability of the sub-processes and sampled ranges of the physical parameters.
The contribution to the probability of a criticality from the 6 other sub-areas could range from a multiplication factor 1 to 7 for the following reasons:
- 1. If all bathtubs were identical and all external forces acting on them were identical.
then they would either reach criticality or remain suberitical identically. Therefore. the probability of a criticality would not increase (this assumes. reasonably. that there is no effect of criticality in one bathtub on an adjacent bathtub).
- 2. If each bathtub nad properties and external forces uncorreleated to those in the other 6.
then the probability of a criticality in trench 2 would increase by a factor of 7.
However, the 7 bathtubs share many properties and external forces in common (e.g..
soil type. rainfall).
To account for the possible increase in probability of criticality from multiple potem;al reactors in trench 2. the staff multiplied the estimated probability for a criticality in trench 2 by 4. Consequence analy sis is presented in Section 111. The dose to an individual living on i
the site would not increase sieniticanth
' because of the simultaneous occurrence of two criticality events in adjacent bathtubs. Dose from ionizing radiation was for a person living I
directly above the critical reactor and falls off sharply with distance. There would be an additional 10-11 meters of soil from the drain of an adjacent bathtub. effectively eliminating ionizing radiation as a contribution. Additional radionuclides would be generated by the other reactors. but since they are physically displaced. one could easily assume that they would not find their way into the same source of drinking water.
l H. Conclusions for Probability of Criticalitt t The Staff expects that the criticality analyses presented above covers the range of scenarios that could occur at the site Summing the probabilities of criticality from the two scenarios leading to criticality results in a range of 4.9 x 10'" to 4.9 x 10" over the assumed 15.000 year lifetime, a range much lower than traditionally considered in risk analyses for nuclear accidents. Because there are a number of conservative assumptions that have gone into these estimates, the staff believes that the calculated probability for a potential criticality event is robust, and unlikely to be higher.
14
l l
111.
Radiation dose consecuences of a crincalits accidem at Parks Town hin -
1 A. Introduction 1
Although the probability of a criticality esent at the Parks site appears to be very low. the consequence in terms of health effects to a person in the critical group affected bv the esent l have been calculated in order to complete the risk analysis. The analysis considers only the i bathtub scenario, assuming a minimum quantity of U-235 of 3.1 Kg. concentrated in a sphere !
62 cm in diameter. The spherical reactor is assumed to be near the bottom of the trench.
approximately 3 rneters from the surface. The dose is from gamma rays produced in the nuclear fission. gamma rays produced as a result of radioactive decay. attenuated fast neutrons from the fission. gamma rays from neutron capture and inelastic collisions. and ingested l
radionuclides derived from fission and activation of material in the trench. i l
B. I. imitations of nuclear reacticn In each kilogram of U-235 there are about 2.6x10" nuclei. The energy from nuclear 11ssion of this much uranium is equivalent to about 4 megawatt-years thermal. The energy content is potentially great, but the amount and rate of energy production will have to do meinly with physical factors of the reactor construction, particularly the density of uranium, the w"ter content, the size of the reactor area and the stability of the reactor.
The hypothetical reactor would be composed of a homogeneous mix of uranium in a matrix of soil and water. Water will play a key part in the reaction, so the resp (nse of the water to the heat produced in the reaction will be extremely important. Neutronics calculations performed for the minimal uranium content necessary indicate that the system is undermoderated; that is, the water content (which is restrained in this system to 40% by volume) is too low for maximum efficiency. Therefore, decreasing the water content of the mass by thermal expansion or drying-out of the mass would reduce its criticality and power.
There are several possible scenarios for the reactor:
A steady sate reactor. in which the heat p;oduced u equal to or less thar . beat necessary to boil off the water required to austain the nuclear chain reaction.
A pulsing reactor in which the rate of heat production is high enough that water will expand or evaporate quickly, causing the nuclear reaction to cease. The reactor could go critical again once the area has reverted to its original state.
An exploding reactor, in which the rate of heat production generates steam at such a high rate that the reactor is physically disassembled, and cannot go critical at a later time.
15 1
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, , i l
l
- 1. The explodine reactor case Case 3 described abose sounds like the worst case. but may actually be relatisel,s benign because once the reactor has disassembled. it can no longer po,e a criticality hazard. There have been a small number of criticality accidents around the world. most of which have released the energy equivalent of 10" to 10* lissions (Knief.1989. Only a few of these accidents have invohed energetic disassembly or damage. Many of these accidents have been in vessels in which there was considerably more tissile material than we are dealing with in the bathtub. and also consisted of high-enriched uranium or plutonium. Radiation from such a reactor would occur in a burst, and some of the radiation would result from the decay of l fission and activation products of the reaction that could be released in the accident. Direct i radiation would be relatively minor on a year-long basis, as compared to the continuous reactor cases discussed below. It uould be difficult to predict the extent of damage to the reactor, and whether it would cause an ejection of material into the air or simply create enough local damage to prevent further nuclear chain reactions, yet still keep the material buried. The staff believes that the likelihood for this event is low because if the reactor could I form at all, it would approach criticality from the under-moderated and under-inventoried ,
direction. Further analysis is beyond the present scope. I
- 2. The pulsine reactor case l In this case, the reactor woulJ release energy at a rate fast enough to boil enough water that would subsequently escape through the ground or to the air as steam. Alternatively the l thermal expansion of the reactor could reduce the density of the water, and stop the nuclear i chain reaction. A reactor that was pulsating rapidly because of thermal expansion, but without l boiling, should have an average power level similar to the steady reactor case. A reactor that I
creates a loss of water by evaporation would have a longer cycle because water would have to re-saturate the reactor zone by capillary suction or infiltration of meteoric water. The annual rate of radiation to someone occupying the site over the long term would depend on how quickly the reactor could cycle and the energy production per cycle. There are no simple ways to put bounds on these estimates. Arguments that the radiation released per year is similar to that of the steady reactor case will be based on considerations of the time for resaturation of j the soil and wiC be dis. A later.
l
- 3. The steadv reactor case In this case, the temperature of the reactor would remain below the boiling point, so that the generation of steam would not cause the reactor to dry out, thereby causing the reaction to cease. The maximum steady power would be determined by the temperature produced, since any drying would reduce the power. In this way the reaction is self-limiting.
16
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, o l
C. Analvses of the steadv reactor case i The power produced by the reactor uses the follouing conceptual moJel.s:
- 1. Oklo Natural analog The Oklo natural analog site has been well studied. The estimated parameters of the site are that the reactive zones had very rich uranium ores. 50 to 70 percent uranium by weight, enriched to greater than 3% in U-235. and up to a meter thick (Brookins.1979L The reactor zones were much deeper underground, and could therefore sustain considerably greater heat generation because of the higher boiling point of water under pressures of up to 40 atmospheres. Nevertheless, the steady power pioduction of the Oklo reactors has been !
estimated to be only about 100 watts per cubic meter (Cowan.1978). Projecting this energv i generation rate to the spherical reactor of 0.62 m diameter gives an energy production of 12.5 watts thermal. l i
- 2. Heat-limited case The maximum heat load for a continuous nuclear reactor can be estimated using the following assumptions
- a. :
For the purposes of heat transfer. the reactor is considered to be a thin disk of radius 1 a=0.31 meters and depth z* = 3 meters underground.
- b. The heat generated per unit area is uniform across the disk.
- c. The temperature at the center of the disk is 100 "C at 15.000 yers.
{
- d. The initial temperature of the soil and of the earth's surface is 20 "C.
- e. The heat capacity C. density p and thermal diffusivity k of the soil are constant.
i The temperature at the center of a thin disk in an infinhe medium of zero initial temperature 2
for an instantaneous heat input of q (joules /m ) would be (Carslaw and Jaeger,1939);
7, CI 3 g .a'i m g -z' int 2pc/nk:-
y where z is the downward distance from the surface. The boundary condition at the surface is !
taken into account by including an image at z = -z' 17 l ,
i
. o q ,
c .. . .
EY
~ ~
2 p Cg For a continuous source of heat. the temperature increase above background of the disk will be:
q g - :-:- u: .g - :-:- w: g 3 _ g . n u:
, 2 p cpkC ' '
The heat input q can be estimated by evaluating the integral numerically with an average soil (Carslaw and Jaeger 1959). heat capacity C = 0.2 cal /(gm C). density 2.5 gnt/ cm'. and thermal diffusivity k = 0 0044 cm2 /see and T = 80 C (i.e. 100 - 20 'C). For the stated conditions the heat load would be 150 watts to raise the center temperature to the boiling point in about one year. The heat load to maintain boiling would diminish with time as the isotherms penetrated further from the reactor. but not signi6cantly; i.e., the necessary heat load for 10.000 years is only about 4 b less than the one-year estimate. We will therefore adopt a power level in the subsequent calculations of 150 watts.
D. Dose assessment from in-eround reactor The dose to a person residing on the site directly aoove the in-ground reactor has been calculated under the following assumptions:
The dose is from irradiation by gamma rays, fast neutrons, and ingestion of radioactive Sssion and activation products.
There is three meters of soil shielding.
The reactor is considered to be a point murce of radiation.
Radiation shielding codes were unavailable for this analysis. so the staff used approximate textbook methods for all dose calculations.
- 1. Dose from fast neutrons An approximate formula for the flux of neutrons from a nuclear reactor represented as a point source is (Bonilla.1957):
h = 2.6x10'K x' I8
_ _ __A
o e l
where R is distance in feet and K is the power of the reactor in kihmatts. The attenuauon ot !
prompt neutrons can be estimated from the formula t Bonilla.1957r l
(water)
= n 95e - C . 02e (concrete) I l
l I
r
=
0 . 9 6 e "" - 0 . 0 4 e "
- I "3 !
The attenuation factor in water is much greater. A typical water content for concrete would be i
6%. The porosity used for the soil is 40%. and it is assumed the soil is saturated with water i over at least part ofits thickness. and close to saturu :sn elsewhere. The composition of the '
nominal soils used in this study is given in Table 2. Conservatively using the concrete values.
the neutron density at the surface is 0.0025 neutrons (cm sec). This translates into an approximate dose of about 3.2 mrem year directly above the reactor at the surface for the 150 watt case and 0.3 mrem year for the 12.5 watt case thr 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day occupancy. An occupant ten meters away from the center would receive a negligible dose from prompt neutrons (therefore radiation from adjacent bathtub reactors would not have to be taken into account). For a thick soil shield. the dose from gamma rays produced by neutron capture could exceed dose from the neutrons themselves. This was not evaluated explicitly because such calculation generally require computer programs, but will be taken into account by a factor of safety. i Table 2 - nominal soil at Parks site Constituent Mass Fraction Carbon 0.0429 i Oxygen 0.49 Sodium 0.0068 Magnesium 0.006 Alumir.am 0.071 Silicon 0.33 Potassium 0.0136 Calcium 0.0137 Iron 0.026 l Porosity = 40%
~
\
19 I
i I
i
9 s.
- 2. Dose from prompt camma radiation The number of prompt tission gamma rays proJueed from nuclear fission can be approximated in four energy groups to be 5.2.1.8. 0.22 and 0.025 per tission ihr 0-1.1-3. 3-5 and 5-7 MEV energies. respectively (Lamarsh.1976). For a 150 watt reactor. the tission rate S is about 4.7 x 10" lissions'second. The flux of gamma rays at distance R from a point source through the effective shielding thickness can be expressed:
S 4# ,
4nR S# e
- l where R is the shielding thickness. B, is the buil dup factor and p is the attenuation coefficient. Considering gamma rays only in the 3rd and 4th group because of the thickness of the shield. and assuming the shielding properties of the nominal soil with 40% water by volume, the total dose rate from prompt gamma radiation is about 6 mrlyear for 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> / day exposure directly above the 150 watt reactor. The dose rate for the 12.5 watt reactor would be 0.5 mrent'yr. These doses could increase from the long-term buildup of fission and activation products, but probably not more than a factor of 2 (Lamarsh.1975). This possibility will be taken into account by the factor of safety.
- 3. Doses from radionuclide incestion This calculation projects the dose from the radionuclides produced in the nuclear fission. The dose is based on screening dose factors for ingestion of three nuclides. Np-237.1-129 and Tc-99 as representative of long lived radionuclides produced in a reactor that also have low retardation. The user's well is assumed conservatively to capture the entire quantity of these radionuclides generated in the reactor. and to pump at a rate of 100,000 liters per year (taken as an average per capita usage for rural non farming use [ Miller.1980]). The well is also considered to be far enough from the reactor that only the non-retarded long-lived radionuclides w.,uld rea ;t before significai. decay The rate at which the radionuclides would be generated by the reactor is based on the average radioactive inventory of spent nuclear fuel projected for burial at the Yucca Mountain site (Wilson.1993). For an average burnup of 3.7 x 10' Megawatt-days / metric ton uranium, there would be approximately 0.378. 0.0339, and 14.3 Ci/MTU for Np-237.1-129 and Tc-99.
respectively. Screening dose factors for ingestion from NCRP report 1231 " Screening models 3
for release of radionuclides (1996)" would be 6.38 x 10 .1.24 x 10-8. and 6.64 x 10* Sv/Bq.
for Np-237,1-129 and Tc 99. respectively.
Assuming all of the produced radionuclides from the 150 watt reactor were mixed into the user's well and consumed at a rate of 2 liters per day containing the long-lived mobile 20
. 9 .
radionuclides from the 150 watt reactor, pise3 a Jose from ingestion of 9.56. U.!' ana 03s millirem ~ year from Np-237. :-C9 and Ie-99. re>pectisely. for a total of 10.1 millirem y ear The do3e from ingesting those radionuelides for the 12.5 watt reactor uould be o.84 millirem yr. N! ore rea3onable assumptions about the u3e of well water i.e. multiple users per well or wells tapping deeper, cleaner aquifers. would reduces these doses significantly. Even with these conservatism. the doses from ingestion of the long-lised. mobile radionuclides would be below the regulatory standard of 15 mitem 3r. Doses would be higher only in the unlikely event that the user nell were located so close to the reactor that the short-lived. Je33 mobile radionuclides were taken into the nell E. Estimate of doses from nulsed release scenario The annual dose calculated above for the continuous reactor would probably encompass the dose stemming from the " pulse" reactor scenario, where the reaction is assumed to proceed at a high rate for a limite ' amount of time and then cease until resaturation occurred. This assessment is based on the following considerations:
1.
The reaction rate for the continuous case is equivalent to 2.2 x 10 ' tissions per year. 2 Reported criticality accidents toutside of nuclear reactors) have been in the range of 10" to 4.4 x 10 lissions t Knief.1985).
2.
The release of a large quantity of heat oser a short period of time would lead to vaporization of water and dryout of the soil. Consider for example that the yearly heat in the case of the 150 watt continuous reactor were concentrated into an instan release.
way:
4.7 x 10' joules. The heat would interact with the wet soil in a complicated Heat would be transferred by thermal conduction:
Sensible heat would be absorbed and transformed into latent heat by the vaporization of water:
The heat capacit. of the soil will J..rease as water evaporates:
The increase pressure would force some water out of pores, replacing it with water vapor.
Steam would migrate through the soil under the increased pressure, transferring heat to the soil by conduction and condensation of water vapor.
Calculation of the size of the dried-out region would be complicated, and also require information not available from this site. However it is possible to bound the size of the regi and also estimate the time to resaturation:
21 I
- a. For thermal conJuetion alone. the temperature around an instantaneous point source in an infinite medium wou!J be iCarslau and .laeger.1959t
.1 . = -r Ep: :n' For the conditions ;ho3en. the maximum penetration of the 100 "C isotherm. assuming an initial temperature of 20 "C. would be about 1.3 meters. Temperatures in the reactor l
sphere would not fall below the boiling point until approximately 0.05 years, for I
conduction heat transfer alone.
- b. The heat generated in 10" lissions would vaporize 1 liter of water (Toran.1997).
Iherefore the instantaneous heat source could vaporize 2200 liters equiv-! nt to a sphere 1.1 meters in radius for a soil porosity of 40%. If only half the heat went to vaporize water the sphere would have a radius of 0.87 meters.
- c. Infiltration of meteoric water alone would resaturate the 0.87 m radius sphere in about 1.85 to 3.85 years for an intiltration rate of 25 or 12 cm/yr. respectively. Water would also be expected to imbibe from all directions because of c 'pillary forces, reducing the time for resaturation.
From these approximate calculations therefore. the pulse reactor could take on the order of a year to resaturate. The radiation produced therefore would be expected to be approximately 2
the same as the continuous reactor on a .searly basis. about 10 " lissions/ year for the 150 watt reactor.
IV Risk from reactor A. fonditional probabilits of health erYects eiven that there is a criticality The doses from irradiation and ingestion of radionuclides are combined into a risk for the exposed individual under the following conservative as.sumptions:
- 1. The individual lises their entire 70 year life on site. spending 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per day directly l
above the reactor.
- 2. The individual gets all water from an onsite well located downgradient of the reactor.
The well captures all Np-237.1-129 and Tc 99 generated by the reactor, but none of the other radioisotopes produced by the reactor.
- 3. The lifetime risk of developing a fatal cancer from irradiation is 5 x 10" per rem.
- 4. The irradiation dose from prompt neutrons and gamma rays is increased by a safety factor of 10 to account for uncertainties in the approximate analyses. especially (a) 22
1 . ,
l I
production of gamma rays in the shielding soil by neutron capture and ab) ingrouth of l tission and activation products.
From the above calculations and assumptions. the conditional probability of the lifetime ri3k j of the maximally exposed individual in the case of the 150 watt reactor would be 0.00105 for radiation and 0.00035 from ingestion for a total of 0.0014. The corresponding conditional probability for lifetime risk for the 12.5 watt reactor uould be 0.00012. On an annualized !
basis the conditional probabilities of a health effect would be 2 x 10" year and 1.7 x 10'
/ year for the 150 watt and 12.5 watt reactors. respectisely.
B. Total probability of health effects The total probability of health effects for the site is evaluated by multiplying the conditional probabilities calculated above by the probability of the criticality. The probability of a criticality at this site somedme within the next 15.000, sears has been estimated conservatively to be in the range 4.9 x 10-" to 4.9x 10' The probability of developing a fatal cancer to the most exposed individual residing on the site for 70 years over the next 15.000 years would therefore range from 6.9 x 10'" to 6.9 x 10"" . or 9.8 x 10* to 9.8 x 10* year over the individual's 70 year lifetime.
V Conclusions These calculations bound the probability and consequences of a potential nuclear chain reaction at the Parks site. The potential reactor would have a power output of about 150 watts thermal under the conditions stated for a minimum critical mass of about 3 Kg U-235 in the bathtub scenario. Using the estimated long-term power production rate from the Oklo natural reactor gase a lower power of 12.5 watts thermal for the same reactor. Using simplified dose calculations for prompt tission neutrons and gamma rays. the maximum exposure at the surface would be about 9 mrem / year directly overhead for 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day exposure in the 150 watt reactor case and 0.75 mr' year in the 12.5 watt reactor case. The potential dose to an individual residing on the site took into account direct irradiation and ingestion. considering 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> / day residency to the highest irraJiation rate at the surface and total reliance on an onsite well cap ring all .t '37.1-129 and Tc-99 prodused by tne nuclear reactor. The . .afation .
dose also included a safety factor of 10 to account for uncertainties in the calculations.
primarily because of gamma ray production by r.eutron capture in the shield and buildup of fission and activation products. Despite the large factors of conservatism employed in almost all phases of these analyses. the probabilities. doses and risk all far below any reasonable regulatory concern at this site.
References ARCO " Field Work Report.1995. Parks Township Shallow Land Disposal Area". Atlantic Richfie:Id. Babcock and Wilcox. February 1996. i
\
23 l
t
Bonilla. C.. Nuclear Fneineerine. .\leGraw liill Book Company.195~
l Brookins. D.G.. Thermodynamic considerations unJerlying the migration of radionuelide m geomedia: Oklo and other examples". in GJ. NicCarths. eJitor. Scientific Basis for Nuclear l i Waste N1anagement, v 1. pp 355-366.1979 l l
Carslaw. C. and J. Jaeger. Conduction of Ileat in Solids. Oxford at the Clarendon Pless.1959 Cowan. G.A.. and A.E Norris (eds), lmestication of the Natural Fiuion Reactor Procram.
Report no 7530. Los Alamos National Laboiatory.1978 Duff. 51artine C.. C. Amrhein. P. Bertsch and D. Ilunter. "The chemistry of uranium in evaporation pond sediment in the San Joaquin Valley. California. USA, using X-uy fluorescence and XANES techniques". Geochemica et Cosmochimica Acta. Vol 61. no 1. pp 7 l 73-68. 1997 l
l Efron. B.. "The jackknife. the bootstrap and other resampling plans". Society of Industrial and Applied Afathematics. Philadelphia.1982.
Greentield. B.F. . A. Fosevear and S.J. Williams. " Review of the microbiological, chemical and radiolytic degradation of organic material likely to be present in intermediate level and low-level radioactive wastes". DOE ilNilP RR 91002. Department of Environment. HN11P.
England. June 31.1990 Hopper. C.hl.. R.H. Odegaarden. CX. Parks. and P.B. Fox. " Criticality Safety Criteria for License Review of Low-Level Waste Facilities". NUREG CR-6284,1995 Hsi. C.K. and D. Langmuir. " Adsorption of uranyl onto ferric oxyhydroxides: application of ,
the surface complexation site-binding model". Geochimica et Cosmochimica Acta, vol 49. pp l 1931 - 1941. 1985.
Kastenberg W.E.. P.F. Peterson. J. Ahn. J. Burch. G. Casher. P. Chambre, E.Greenspan. D.R.
Olander. J. Vuji " Con cra' ions of autocat l.stic criticality of thsile materials in geol 3.c ;
repositories". Draft paper to appear in Nuclear T. hnology.1997. '
Knief. R.A., Nuclear Criticalitv Safetv. American Nuclear Society. Lagrange Park 111. 1986 -
i Lamarsh. J., introduction to Nuclear Fncineerine. Addison Wesley Publishing co. Reading .
hiass 1975 24 f
1 i
l Paxton. ILC. and N.L Prosost. " Critical Dimensions of Sptems Containing l'-235 Pu-239 i and l'-233" Los Abmos National Laborator,s. I os Al.uno . N NI. July lux 7 Nieunier. J.D.. Landais. P. anJ Pagel. NL " Experimental esiJence of uraninite fornution trom diagenesis of uranium-rich organic matter" (ieochimica et Co.3mochimica Acta. sol 54. pp 1541-1556 1
N1 iller. D.W.. editor. Waste disposal effects on erounduater: A comprehensive sunev of the occurrence and control of eroundwater contamination resultine from waste disposal practices Premier Press. Berkeley. CA (1980)
Sheppard. NLI. and Thibault. D.H. " Default soil solid liquid partition coefficients. Kgs. for four maior soil types: a compendium". liealth Physics. vol 59.1990.
Toran. L.E. . C.NL llopper. C.V. Parks. Brian Broadhead and V. Colten-Bradley. "The potential for criticality silowing disposal of uranium at low-level waste facilities. Volume 1".
NUREG/CR-6505.1997 !
Waite. T.D.11. A. Davis. T.E. Payne. G.A. Way chinas, and N. Zu. " Uranium (VI) adsorption to ferrihydrite: Application of a surface complexation model". Geochemica et Cosmochemica Acta. Vol 58. no 24. pp 5465-5478.1994.
Wilson 1993. " Total sprem performance assessment.1993" Sandia National Laboratories.
Albuquerque NNI.1993. !
I Wood. S.A.. "The role of humic substances in the transport and fixation of metals of i economic interest ( Au. Pt. Pd. l'. V)" Ore Geology Reviews. Elsevier, vol 11. pp 1-31.1996 l
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i Attachment 2 i A Comparison of information related to LLW post disposal criticality i Category or oronerty _P_ fos or no oroblem Cons or oroblem i i
Inventory Eleven trenches Fif ty-five trenches I contain less than a contain multiple l critical mass critical masses (a critical mass a 1.5 Kg !
23D) . Trench 63 contains 1070 Kg 23 9 l t
License amendment . Assuming.10% enriched Assuming enrichment l limits 200 g 23t/ft8 U homogeneously mixed greater than 10%, i and 350 g 22t/ package with quartz and water NUREG 6284 recommends with minimum base in infinite slab with that in no case may j area of 2 ft: quartz / water reflector the areal density :
requires 409g 2t/fta exceed 94 g 8't/ft8 l to go critical . The For enrichments of lot operational limit is and less the :
294 g 23t/ft2 operational limit is !
(Barnwell report 174 g 88t/ft8 !
6505v2) i Use of NUREG 6284 as Assumes infinite This existing study j screening tool planar array of pure provides an example in uranium metal' which criticality .
optimally moderated would occur if the !
units. waste were buried at .'
the Barnwell license !
amendment limit. !
Recommended areal !
density limits come j f rom criticality ,
experts. No evidence ;
supplied to support i less restrictive ;
limit. The limits in l' NUREG 6284 were to be used with"LLW !
consisting of l contaminated !
hydrocarbons (e.g. j paper, plastic) 1 contaminated i metal / alloys and i inorganics." i Commingling of SNM Assuming perfect Commingling requires and SM mixing of SNM and SM equal mobilization of mostly reduces all forms of uranium enrichment to less (on resins, as metal, than 10%. Trench 66 on paper etc. ) .
is reported to have Depleted uranium is a 100% enriched U. better reflector than water. Consequently, heterogeneous distributions of SNM and SM could promote criticality, instead of inhibiting it.
, i
Burial procedures Attempts to place SM Map of trench'23 shows next to SNM. SIM is to concentration of 4 and be placed at bottom of 5 SNM packages trench. covering areas of 40 ft on a side.
Procedure S20-OP-030 requires packages containing more than 30 g "'U be placed in the bottom concrete vault. Bottom vault is most likely to flood. SM in vaults placed on top of SIM vault likely would remain high and dry.
Therefore, SNM would be preferentially mobilized and reconcentrated.
Reliance on commingling is optimistic.
Ponding in bottom of NRC staf f analysis Ponding was observed trench indicates minimal in one of two ponding when trench is experimental trenches filled with soil like at site (Dennehy and material with McMahon, 19 87 ) .
hydraulic conductivity Saturated conditions of 4E-3cm/s . Sumps lasted half of the containing up to 21 f t year. Preferential of water drain in 4 pathways as would be months. expected when concrete vaults and large boxes are emplaced are likely to lead to transient perching Chemical condition Barnwell report states NRC staf f calculations needed for uranium that ORNL calculations using Barnwell immobilization suggest reducing groundwater data conditions are suggests that required for ef ficient oxidizing conditions immobilization of can lead to ef ficient uranium. Reducing immobilization of conditions are uranium. Saturaten unlikely if a perched conditions need not be water zone is not maintained. Transient maintained at the saturation as seen at bottom of the trench Barnwell is all that is required.
l Holes in concrete The NRC staff cannot Current practice vaults lead to envision an advantage involves emplacement focussing to this design feature of waste in concrete with regard to vaults that contain criticality safety. one drain hole. This design leads to focussing of fissile material.
j i,
_ _ _ .-_....w-_- .
, .. . . ~ . - _. -
- u. s.
\
Enrichment Average enrichments High enriched uranium are low. These {
has been buried. i averages are computed Trench 23 contains over the whole trench, uranium enriched to I or in the isolated greater than 80 wtt l case of Trench 23, (Barnwell Report) .
over the monthly {
Current burial inventories. Maximum procedures require average enrichment of that for each-monthly inventory of individual SNM i Trench 23 is 14.8%. package, a complete ;
Most average isotopic analysis j enrichments are less printout be part of j than 1%. the shipper manif est.
1 Mass in single vault Recent burials have Based on Burial relatively minor 2"U procedure S20-OP-030, 1 inventories 7.2 Kg 2"U is the maximum mass allowed in a single concrete vault. 1.5 Kg 2"U is sufficient for criticality in {
spherical geometry (NUREG/CR 6505 V1) ,
l Actual burial Recent burials have In 1996, licensee practices relative to relatively minor 2"U requested a relaxation license amendment inventories of the areal density limits limit from 200 g 2"U/ft2 to 342 g 2nU/ft2 Such a request suggests past i burials may include areal densities approaching the license limit.
Waste volume is money to the licensee. The main driving force is to bury as much waste as possible in the '
smallest volume, i
s +w
]
Kinetics In extreme cases, There are examples of reconcentration will uranium ore deposits not occur. Natural that have formed at analogues suggest that low temperatures and conversion of oxidized pressures (Meunier et uranium species to al.,1990).
reduced uranium solids Immobilization occurs may be kinetically in stages with the inhibited. in,itial sorption onto organic material and slow conversion to uraninite.
Assumption of equilibrium is standard practice for geochemists. This practice is comparable to the assumption of Darcy flow conditions by hydrologists.
Using kinetics to discredit the equilibrium-calculations, would then draw into question the common practice of using the solubility limit to constrain radionuclide concentrations. This practice is used for high-level waste. (GTP on Solubility).
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