scispace - formally typeset
Open AccessReportDOI

Options for treating high-temperature gas-cooled reactor fuel for repository disposal

TLDR
In this paper, the authors describe the options that can reasonably be considered for disposal of high-temperature gas-cooled Reactor (HTGR) fuel in a repository, including whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products.
Abstract
This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options Fort St Vrain Reactor fuel, which utilizes highly-enriched {sup 235}U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched {sup 235}U (without thorium) or for the German pebble-bed fuel Future US HTGRs will be based on the Fort St Vrain (FSV) fuel form The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel From the standpoint of process cost and schedule (not considering repositorymore » cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing« less

read more

Content maybe subject to copyright    Report

ORNI./TM-12027
U0-522, -810, -811
. OAK RIDGE
NATIONAL, .
LABORATORY Options for Treating
High-Temperature Gas-Cooled
Reactor Fuel for Repository Disposal
A. L. Lotts
W. D. Bond
C. W. Forsberg
R. W. Glass
F. E. Harrington
G. E. Michaels
K. J. Notz
R. G. Wymer
r
MANAGEDBY
MARTINMARIETTAENERGYSYSTEMS,INC.
FORTHEUN;TEDSTATES DISFR;L_L,;. _ " 'Z.............. ' ,,.,
, ,_,,..,_.,ML,qT f<-'UNLiMI ]ED
DEPARTMENTOFENERGY

This report has been reproduced directly from the best available copy.
Available to DOE and DOE contractors from the Office of Scientific and Technl-
cal Information,P.O. Box 62, Oak Ridge, TN 37831; prices availablefrom (615)
576-8401, FTS 626-8401.
Available to the public from the National Technical InformationService, U.S.
Department of Commerce, 5285 Port Royal Rd.. Springfield,VA 22161.
This report was prepared as an account of work sponsoredby an agency of
the United States Government. Neither the United States Governmentnor any
agency thereof, nor any of their employees, makes any warranty, express or
implied, or assumes any legal liability or responsibilityfor ti'e accuracy, com-
pletenees, or usefulnessof any information,apparatus, produc_.,or process dis-
closed, or represents that Its use would not infringe privately owned rights.
Reference herein to any specific commercial product, process, or servic_ by
trade name, trademark, manufacturer,or otherwise,does not necessarily consti-
tute or implyIts endorsement,recommendation,or favoring by the United States
Government or any agency thereof. The vie"..;; and opinions of authors
expressed herein do not necessarily state or reflect those of the United States
Governmentor any agar;oy thereof.

ORNL/TM--12077
DE92 014909
OPTIONS FOR TREATING HIGH-TEMPERATURE GAS-COOLED
REACTOR FUEL FOR REPOSITORY DISPOSAL
A. L. Lotts**
W. D. Bond
C. W. Forsberg
R. W. Glass*
F. E. Harrington**
G. E. Michaels
K. J. Notz
R. G. Wymer**
Chemical Technology Division
*Engineering Division
**Consultant
Date Published: February 1992
M ,SIE
Prepared by the
OAK RIDGE NATIONAL LABORATORY
Oak Ridgc, Tennessee 37831-2008
managed by
MARTIN MARIETTA ENERGY SYSTEMS, INC.
for tile
U.S. DEPARTMENT OF ENERGY
under contract r3.E-AC05-84OR21400

CONTENTS
FIGURES ..................................................... vii
TABLES ix
ABSTRACT ................................................... xi
1. INTRODUCTION 1
1.1 OBJECTIVES .......................................... 1
1.2 SCOPE ............................................ 1
1.3 BASIS OF THE ASSESSMENT ............................ 2
1.4 ASSUMPTIONS ......................................... 2
1.5 UNIQUE ASPECTS OF HTGR FUEL ....................... 3
2. INSTITUTIONAL ISSUES ..................................... 5
2.1 REPOSITORY WASTE ACCEPTANCE CRITERIA ............ 5
2.2 CARBON-14 ........................................... 7
2.3 APPLICABLE RADIATION PROTECTION STANDARDS ...... 7
2.4 DISPOSAL OF RADIOACTIVE PROCESSING WASTE ........ 7
2.5 SAFEGUARDS: ISSUES RELATED TO NON-WEAPONS
STATES ............................................ 8
2.6 REFERENCES ......................... . ............... 9
3. DESCRIPTION OF HTGR FUELS ............................. 11
3.1 INTRODUCTION ...................................... 11
3.2 FORT ST. VRAIN FUEL ................................ 12
3.2.1 Physical and Chemical Description of Fort St. Vrain Fuel .... 12
3.2.2 Quantities of Fort St. Vrain Fuel ...................... 17
3.2.3 Radiological Properties of Fort St. Vrain Fuel ............ 19
3.3 PEACH BOTFOM-1 REACTOR FUEL ..................... 22
3.4 COMPARISON TO OTHER NON-STANDARD FUELS ........ 24
3.5 FUTIJRE HTGRS ...................................... 25
3.6 REFERENCES ........................................ 27
4. OVERVIEW OF OPTIONS ................................... 29
4.1 WHOLE-BLOCK DISPOSAL ............................. 29
4.2 DISPOSAL WITH PRIOR REMOVAL OF GRAPHITE ........ 33
4.3 DISPOSAL WITH DISSOLUTION ,.)F SPENT FUEL .......... 34
5. WHOLE BLOCK DISPOSAL .................................. 35
5.1 INTRODUCTION ...................................... 35
5.2 ACCEPTABILITY OF WHOLE BLOCK DISPOSAL 35
5.2.1 Previous Studies and Experiments ...................... 35
5.2.2 Comparison of the Characteristics of HTGR Spent
Fuel with Repository Acceptar, ce Requirements ........... 37
'8*
ill

5.2.2.1 Allowable Release Rates for Radionuclides
from the Repository ........................ 37
5.2.2.2 Allowable Organics in a Repository ............ 39
5.2.2.3 Combustibility ............................ 40
5.2.3 Comparison of HTGR and LWR Spent Fuel
Under Repository Conditions ....................... 41
5.2.3.1 Physical Effects ........................... 41
5.2.3.2 Chemical Effects .... ....................... 42
5.2.3.3 Combined Physical Form and Chemical Effects ... 44
5.2.4 Options for Improved Whole Block Disposal ............ 44
5.3 REPOSITORY ENGINEERING AND COST
CONSIDERATIONS . ........... 45
5.3.1 Repository Engineering Limits ........... ........ .... 45
5.3.2 Heat Limits ..................................... 46
5.3.3 Volume Limits ................................... 47
5.3.4 Waste Form .................................... 48
5.3.5 Relative HTGR and LWR Spent Fuel Disposal Costs ..... 49
5.4 REFERENCES ..................... ................. 50
6. DISPOSAL WITH REMOVAL OF GRAPHITE .................. 53
6.1 OPTIONS AVAILABLE ................................ 53
6.2 PHYSICAL SEPARATION OF GRAPHITE ................ 53
6.3 CHEMICAL SEPARATION OF GRAPHITE (BURNING) ..... 54
6.4 POSSIBLE IMPROVEMENTS TO FUTURE HTGR FUEL .... 55
6.5 STATUS OF TECHNOLOGY ............................ 56
6.6 REFERENCES ....................................... 57
7. DISPOSAL WITH DISSOLUTION OF FUEL .................... 59
7.1 OVERALL FLOW SHEET .............................. 59
7.2 HEAD-END OPERATIONS ............................. 61
7.3 SOLVENT EXTRACTIONS ............................. 63
7.4 OFF-GAS TREATMENT ............................... 64
'7.5 LIQUID AND SOLID WASTE PROCESSING ............... 64
7.6 STATUS OF TECHNOLOGY ............................ 65
7.7 REFERENCES ....................................... 66
8. SCHEDULES AND COSTS .............. .................... 69
8.1 DEVELOPMENT COSTS ............................... 69
8.2 CAPITAL COSTS ..................................... 71
8.3 SCHEDULES ....................................... 71
8.4 OPERATING COSTS .................................. 72
8.5 SUITABILITY OF WASTE FOR THE REPOSITORY ........ 72
9. CONCLUSIONS ....................................... 73
9.1 CONCLUSIONS ON WHOLE BLOCK HTGR SPENT
FUEL DISPOSAL ..................................... 73
iv

Citations
More filters
Journal ArticleDOI

Recent advances in the treatment of irradiated graphite: A review

TL;DR: A review of the treatment and disposal of irradiated graphite from nuclear reactors can be found in this paper, where the main features of Wigner treatment, thermal treatment, chemical treatment, conditioning, coating and impregnation, gasification were addressed.

Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

TL;DR: A review of the nuclear fuel cycles supporting early and present day gas reactors, and identifying challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program is presented in this paper.
ReportDOI

Update to Assessment of Direct Disposal in Unsaturated Tuff of Spent Nuclear Fuel and High-Level Waste Owned by U.S. Department of Energy

TL;DR: In this article, the authors provide information and guidance to the Office of Environmental Management of the US Department of Energy (DOE) about the level of characterization necessary to dispose of DOE-owned spent nuclear fuel (SNF).
Journal ArticleDOI

Spent Nuclear Fuel and Graphite Management for Salt-Cooled Reactors: Storage, Safeguards, and Repository Disposal

TL;DR: The fluoride salt-cooled high-temperature (FHR) as discussed by the authors is a new type of nuclear power station that combines the graphite-matrix coated-particle fuel and graphite moderator from high temperature gas cooled reactors.
Journal ArticleDOI

Performance Assessment for Geological Disposal of Graphite Waste Containing TRISO Particles

TL;DR: In this paper, a deterministic performance assessment for spent fuel from deep-burn modular high-temperature reactors (DBMHRs) in the proposed Yucca Mountain repository is presented.
Related Papers (5)