skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: An outboard shield design for TIBER-II with potential for tritium self-sufficiency: Appendix A-1

Abstract

The outboard breeding shield design for TIBER-II is described. The design allows for tritium self-sufficiency without compromising magnet protection, design simplicity, and the testing mission of the device. The shield consists of a beryllium pebble front zone backed by a steel pebble zone. The shield is cooled by an aqueous solution containing 16 g LiNO/sub 3/ per 100 cm/sup 3/. A double first wall is used to insure uniform cooling and minimize pressure. The design pressure for the outboard shield is 0.19 MPa and the coolant temperature is less than 75/sup 0/C. 6 refs., 5 figs., 3 tabs.

Authors:
;
Publication Date:
Research Org.:
Wisconsin Univ., Madison (USA). Fusion Technology Inst.
OSTI Identifier:
5757365
Report Number(s):
CONF-871007-49
ON: DE88002459
DOE Contract Number:
FG02-87ER52140
Resource Type:
Conference
Resource Relation:
Conference: 12. symposium on fusion engineering, Monterey, CA, USA, 12 Oct 1987; Other Information: Paper copy only, copy does not permit microfiche production
Country of Publication:
United States
Language:
English
Subject:
70 PLASMA PHYSICS AND FUSION TECHNOLOGY; SHIELDING MATERIALS; DESIGN; TIBER-X TOKAMAK; BERYLLIUM; BREEDING BLANKETS; DIVERTORS; FIRST WALL; GRAPHITE; HYDRAULICS; LITHIUM COMPOUNDS; THERMAL ANALYSIS; TRITIUM; ALKALI METAL COMPOUNDS; ALKALINE EARTH METALS; BETA DECAY RADIOISOTOPES; BETA-MINUS DECAY RADIOISOTOPES; CARBON; ELEMENTAL MINERALS; ELEMENTS; FLUID MECHANICS; HYDROGEN ISOTOPES; ISOTOPES; LIGHT NUCLEI; MATERIALS; MECHANICS; METALS; MINERALS; NONMETALS; NUCLEI; ODD-EVEN NUCLEI; RADIOISOTOPES; REACTOR COMPONENTS; THERMONUCLEAR REACTOR WALLS; THERMONUCLEAR REACTORS; TOKAMAK TYPE REACTORS; YEARS LIVING RADIOISOTOPES; 700201* - Fusion Power Plant Technology- Blanket Engineering; 700209 - Fusion Power Plant Technology- Component Development & Materials Testing

Citation Formats

Sawan, M.E., and Sviatoslavsky, I.N. An outboard shield design for TIBER-II with potential for tritium self-sufficiency: Appendix A-1. United States: N. p., 1987. Web.
Sawan, M.E., & Sviatoslavsky, I.N. An outboard shield design for TIBER-II with potential for tritium self-sufficiency: Appendix A-1. United States.
Sawan, M.E., and Sviatoslavsky, I.N. 1987. "An outboard shield design for TIBER-II with potential for tritium self-sufficiency: Appendix A-1". United States. doi:. https://www.osti.gov/servlets/purl/5757365.
@article{osti_5757365,
title = {An outboard shield design for TIBER-II with potential for tritium self-sufficiency: Appendix A-1},
author = {Sawan, M.E. and Sviatoslavsky, I.N.},
abstractNote = {The outboard breeding shield design for TIBER-II is described. The design allows for tritium self-sufficiency without compromising magnet protection, design simplicity, and the testing mission of the device. The shield consists of a beryllium pebble front zone backed by a steel pebble zone. The shield is cooled by an aqueous solution containing 16 g LiNO/sub 3/ per 100 cm/sup 3/. A double first wall is used to insure uniform cooling and minimize pressure. The design pressure for the outboard shield is 0.19 MPa and the coolant temperature is less than 75/sup 0/C. 6 refs., 5 figs., 3 tabs.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = 1987,
month = 1
}

Conference:
Other availability
Please see Document Availability for additional information on obtaining the full-text document. Library patrons may search WorldCat to identify libraries that hold this conference proceeding.

Save / Share:
  • The compactness of the TIBER-II reactor has placed a premium on the design of a high performance inboard shield to protect the inner legs of the toroidal field (TF) coils. The available space for shield is constrained to 48 cm and the use of tungsten is mandatory to protect the magnet against the 1.53 MW/m/sup 2/ neutron wall loading. The primary requirement for the shield is to limit the fast neutron fluence to 10/sup 19/ n/cm/sup 2/. In an optimization study, the performance of various candidate materials for protecting the magnet was examined. The optimum shield consists of a 40more » cm thick W layer, followed by an 8 cm thick H/sub 2/O/LiNO/sub 3/ layer. The mechanical design of the shield calls for tungsten blocks within SS stiffened panels. All the coolant channels are vertical with more of them in the front where there is a high heat load. The coolant pressure is 0.2 MPa and the maximum structural surface temperature is <95/sup 0/C. The effects of the detailed mechanical design of the shield and the assembly gaps between the shield sectors on the damage in the magnet were analyzed and peaking factors of approx.2 were found at the hot spots. 2 refs., 6 figs., 2 tabs.« less
  • In the initial design of TIBER-II inboard (I/B) shield, multilayers of tungsten shield and coolant were deployed with a total thickness of 48 cm. It was thought during the design process to replace W by PCA. The motivations are: (1) accumulated activation level in the I/B shield at shutdown is larger in the W-shield in comparison to the PCA-shield, and (2) concerns regarding cost/fabrication. This design change required an I/B shield thickness of --58 cm to reach the same performance level of the 48 cm W-shield. In this paper a detailed comparison between the two types of shield is givenmore » regarding the accumulated radioactivity, biological hazard potential (BHP), and afterheat levels at shutdown and various times thereafter. In addition, a substantial part of the present work is devoted to studying the impact of the present neutron cross-section uncertainties in the prediction of the radiation damage parameters in the S/C magnet. In this regard, an extensive cross-section sensitivity/uncertainty analysis was performed to assess the required increase in the I/B shield thickness in both cases to account for these uncertainties. It was shown that the economic penalty of such an increase is 13 - 17 M$ in the W-shield case as opposed to 10 - 14 M$ in the case of the PCA-shield.« less
  • This paper describes the preliminary analysis of LOFA (loss of flow accident) and LOCA (loss of coolant accident) in the TIBER-II engineering test reactor breeding shield. TIBER-II is a compact reactor with a major radius of 3 m and thus requires a thin, high efficiency shield on the inboard side. The use of tungsten in the inboard shield implies a rather high rate of afterheat upon plasma shutdown, which must be dissipated in a controlled manner to avoid the possibility of radioactivity release or threatening the investment. Because the shield is cooled with an aqueous solution, LOFA does not posemore » a problem as long as natural convection can be established. LOCA, however, has more serious consequences, particularly on the inboard side. Circulation of air by natural convection is proposed as a means for dissipating the inboard shield decay heat. The safety and environmental implications of such a scheme are evaluated. It is shown that the inboard shield temperature never exceeds 510/sup 0/C following LOCA posing no hazard to reactor personnel and not threatening the investment. 7 refs., 6 figs.« less
  • TIBER-II is an engineering test reactor designed to establish the technical feasibility for fusion, and is a US option for the prospective International Thermonuclear Test Reactor (ITER). The TIBER-II baseline design has 3 m major radius, 3.6 aspect ratio, and 1.1 MW/m/sup 2/ average neutron wall loading. The inboard shield is about .5 m thick and structurally consists of tungsten alloy and PCA alloy. The outboard is 1.52 m thick and utilizes PCA as structure and beryllium as a neutron multiplier. An aqueous solution of 160 g LiNO/sub 3//liter is used throughout as a coolant and breeder. A one-dimensional cylindricalmore » model for TIBER is used to calculate the neutron flux and the radioactivities. Activities are calculated during and after 2.5 full power years (FPY) of operation. TIBER total activity is approx.2 MCi/cm at the end of operation and is dominated by the inboard activity. The high volumetric fraction of the PCA/W alloys in the inboard shield, used to provide magnet protection at the limited inboard space, makes the inboard specific activity two orders of magnitude higher than that of the outboard and dominant all the time. The decay heat due to ..beta.. and ..gamma.. decay produces about .05 W/cc in the inboard at shutdown and for a few weeks. Under adiabatic conditions, this heat would raise the inboard shield temperature up to 1150/sup 0/C. A considerable part of this heat is generated by the ..gamma.. decay which might help, through the ..gamma.. transport, to smooth the heat concentration. Waste disposal ratings of the TIBER structures have been calculated, and it is found that both the inboard and the outboard shield are classified as Class C radwaste. 4 refs., 4 figs., 3 tabs.« less