ML20205Q166
| ML20205Q166 | |
| Person / Time | |
|---|---|
| Issue date: | 03/03/1986 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2401, NUDOCS 8605280196 | |
| Download: ML20205Q166 (8) | |
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DATE ISSUED: 3/3/86 h
3 f Ti PROPOSED MINUTES /
SUMMARY
OF THE ACRS ADVANCED REACTORS SUBCOMMITTEE MEETING FEBRUARY 25, 1986 WASHINGTON, DC Purpose The purpose of the meeting was to review two liquid metal reactor (LMR) conceptual designs (the PRISM-power reactor inherently safe module, and the SAFR-sodium advanced fast reactor). These two conceptual designs were submitted by DOE and its contractors to the NRC Staff.
Meeting Attendees ACRS NRC Staff M. W. Carbon, Chairman T. King J. C. Mark, Member W. Kerr, Member Others M. El-Zeftawy, Staff N. Brown, GE S. Davies, GE DOE F. Tippets, GE A. Millunzi C. Snyder, Bechtel F. X. Gavigan R. Lancet, RI M. P. Norin J. Brunings, RI J. E. Stader Highlights, Agreements, and Requests 1.
T. King, Safety Program Evaluation Branch /NRR, overviewed the status of NRR's interactions with DOE on the LMR program. The NRC Staff has been committed to review two advanced LMRs conceptual designs over approximately a two year time period (1986-1987) and would conclude with the issuance of a Safety Evaluation Report and a licensability letter based upon the review of a Preliminary 8605280196 860303 DESIONATED GRIGINAL PDR ACRS
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2401 PDR Certified By
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Advanced Reactors Meeting February 25, 1986 Safety Information Document (PSID) to be submitted by DOE. How-ever, as a result of the Gram-Rudman-Hollings Act and the Office of Management and Budget (0MB) review, the NRC resources have been reduced. The current available resources will only apply about two professional staff-years in each fiscal year and minimal technical assistance funds. Future interactions with DOE will be limited to major issues only such as:
Non-safety grade balance of plant (B0P)
Severe accidents Containment issues Mr. King indicated that, at this point, the NRC Staff would still like to solicit the ACRS coments and interactions.
2.
F. Gavigan, Director of the Office of Advanced Reactor Program at DOE, overviewed the LMR program. The GE (PRISM-design) and the RI (SAFR-design) contracts consist of three phases with the option for renewal at the end of each year. DOE is now at the end of the first phase, after which there will be approximately two years of design activities.
In fiscal year 1987, DOE will have to face the issue of the Gram-Rudman cuts, depending on the resolution by the Congress.
DOE have started interacting with the utilities to understand the problems they have on buying large nuclear reactors.
DOE plans to have a cost-competitive design that would give the utilities a cost advantage as compared to other alternate competi-tive sources available in the year 2000.
At the completion of the third phase, DOE expects to achieve through interaction with the NRC a set of licensing criteria along with a licensability letter. DOE also expects to receive a market analysis and comercialization plans.
Advanced Reactors Meeting February 25, 1986 3.
GE Presentation The GE's design called the PRISM concept emphasizes inherent safety characteristics and modularity, reduce owner's risk, and reduce costs. The reactor modules are a single standard design that would be built in a factory and are shippable by rail as a unit.
The plant uses nine PRISM reactors, with each module producing 425 Mwt power. The plant combined power output is 1246 Mwe.
Each module is a pool type LMFBR design with its own intermediate heat transport system and steam generator system.
Each reactor module is housed within its own seismically isolated silo.
Each reactor and silo is housed in a reactor building with outer silo and equipment cells. Common facil-ities are the control building, maintenance building, radwaste building and the fuel handling building.
The guard vessel, reactor vessel, and reactor closure are the major components of the reactor enclosure. The guard vessel assures that the core will not be uncovered if the reactor vessel leaks. The guard vessel and the reactor vessel are made of 316 stainless steel.
The PRISM reactor core is a homogeneous, oxide-fuel design (metallic fuel can also be used) with an average temperature rise of 265 F.
The core structural material was chosen for its low irradiation swelling characteristics. The core lattice is being selected to be capable of breeding.
The small size of each reactor module facilitates the use of passive inherent self-shutdown and shutdown heat removal features. The PRISM plant is equipped with three methods of
Advanced Reactors Meeting February 25, 1986 removing shutdown heat from the reactor; condenser cooling, auxiliary cooling system (ACS), and the reactor vessel auxiliary cooling system (RVACS). The normal shutdown heat removal is by condenser cooling.
Failing that, shutdown heat is removed from the steam generators by the ACS augmented by some steam venting. The estimated use of the ACS is less than 10 times per module life time and it is a non-safety grade system.
If sodium has been lost from the intermediate heat transport system (IHTS), R~vACS will remove heat directly from the reactor vessel by natural air circulation flow. The RVACS is a safety related system and its estimated use is less than once per module life time.
It is also self-regulating such that the higher the reactor vessel temperature, the higher the RVACS heat removal rate.
GE is placing emphasis on full scale tests to demonstrate the plant's safety. The balance of plant (B0P) is completely disconnected from the primary loop safety considerations.
The PRISM project schedule that includes safety test and licensing is attached (Attachment A).
- PRISM individual and societal risks of (1.4x10-8,5x10-12) are well below the corresponding NRC safety goals of (5x10-7,
1.9x10-6),respectively.
4.
RI Presentation The RI's design called the SAFR is a liquid metal reactor designed to produce approximately 350 Mwe net per module.
RI envisions four 350 Mwe SAFR modules per site. Each SAFR unit is a pool-type design with passive decay heat removal. The personnel radiation exposure shall be no higher than 20% of I
Advanced Reactors Meeting February 25, 1986 current LWRs. The core design can accommodate either an oxide or metal fuel.
- Each SAFR module will use a building block approach with discrete increments of power generation called power paks.
The power paks are of an identical size optimized to user's needs.
The shutdown heat remoral system (SHRS) for SAFR consists of three heat removal paths:
the normal path through the steam generator and condenser; the reactor air cooling system (9.ACS) which is a natural convection of air past the reactor guard vessel; and the direct reactor auxiliary cooling system (DRACS) which uses a single in-vessel heat exchanger and a natural convection flow path to a sodium to air heat ex-changer. The RACS system operates continuously and is considered the only safety grade system designed to accommo-date an SSE and a design basis tornado.
It is considered inherently safe and reliable, t.ecause it has no active compo-nents and it does not require operator action, and is indepen-dent of electrical power. ORACS requires operation of dampers on the ex-vessel natural draft heat exchanger (DHRX) and an internal automatic gas valve to redirect internal flow under decay heat-low flow conditions. The valve is passively operated by differential pressure. DRACS is a non-safety grade system and it is a backup for economic protection.
- The SAFR plant is designed to incorporate a high degree of inherent safety to protect the public. RI's top level safety goal is to minimize the potential for severe accidents, to assure inherent response to all credible accidents and to eliminate the need for offsite evacuation of the public, and
i Advanced Reactors Meeting 6-February 25, 1986 that time margins of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or more are available for correc-tive actions.
The SAFR plant will be designed to be commercially competitive with coal and LWR plants by the year 2000 and beyond. SAFR standard plant licensing schedule is attached (Attachment B).
The SAFR plant conforms with the NRC advanced reactor policy statement.
1 Future Actions Dr. Carbon will brief the full ACRS in March 1986 regarding the HTGR and LMRs conceptual designs. DOE and the NRC Staff will also present a brief overview for both the HTGR and LMRs programs. DOE then will give a complete presentation regarding the HTGR design.
i NOTE:
Additional meeting details can be obtained from a transcript of this meeting available in the NRC Public Document Room, 1717 H Street, NW, Washington, CC, or can be purchased from ACE-Federal Reporters, 444 North Capitol Street, Washington, DC 20001, (202) 347-3700.
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