ML20004F744
| ML20004F744 | |
| Person / Time | |
|---|---|
| Issue date: | 05/19/1981 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1851, NUDOCS 8106220265 | |
| Download: ML20004F744 (31) | |
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ISSUE DATE:
5/19/81 MINUTES OF THE k
ACRS SUBCOMMITTEE MEETING ON ADVANCED REACTORS DES PLAINES, IL APRIL 21 & 22, 1981 The ACRS Subcommittee on Advanced Reactors met on April 21 & 22,1981 at the Royal Court inn in Des Plaines, IL.
The purpose of the meeting was to dis-cuss with DOE and some of its contractors the safety design criteria of the Conceptual Design Study (CDS). The CDS is on LMFBR conceptual design that incorporates experience from FFTF, CRBR, AND PLBR and serves as a focus for the U.S. LMFBR base program. The CDS provides a logical next step in the U.S. program.
The Subcommittee did not receive dither written statements or request for time to make oral statements from any members of the public. The list of attendees, schedule, notice, and all the handouts received at the meeting are attached to the Office Copy of these minutes.
Dr. M. Carbon opened the meeting at 8:30 a.m. and made a few organizational remarks to the Subcommittee members and consultants.
He then briefly commented on the agenda and its contents and called on the first speaker.
F. Gavigan, DOE, stated that the purpose of the meeting was to ' rief the ACRS o
regarding the CDS design safety considerations, to assist the ACRS in its plan to develop a "large LMFBR" design criteria and to discuss the criteria for reviewing NRC safety R&D program. Phase I of the CDS program was completed on March 31, 1981. The next phase of the program w
_bg to establish prelim-inary design parameters for a large plant. Th1[p s
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progress.
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MINUTES OF THE MEETING ~ ;
ON ADVANCED REACTORS APRIL 21 & 22, 1981 The contractors of the CDS program under DOE are:
Atomics International, j
Babcock & Wilcox, Combustion Engineering, General Electric Company, and Westinghousa Electric Corporation (reactor manufacturers), Bechtel Cor-poracion, Burns and Roe, and Stone &
'ter (architect-engineers), and Boeing Engineering and Construction (technical integrator). The organi-zational structure and responsibilities of the ' participants were discussed.
The architect-engineers work through Boeing Engineering and Construction who reports to the Department of Energy Project Office. The architect-engineers are concerned, essentially with facility design. Tne NSS does not report to Boeing but reports to the Department of Energy Project Office directly.
The major plant pErameters for the CDS program established in Phase I are:
1000 MWe Plant size Reactor ccnfiguration -
loop 4
Number of loops mixed oxide fuel / heterogeneous core core First core steel containment with concrete confinement Containment concept (double wall) tandem compound Turbine generator The basic NSSS consists essentially of the core within the pressure vessel, the primary heat transport system, the intemediate heat transport system, and the steam turbine-generator system. The primary and intermediary trans-port system contains sodium. There are two decay heat removal systbs, di-verse and redundant which are located in the primary and intermediate loops.
Mr. Gavigan stated that this design is mature enough to support high confidence cost and schedule estimates.
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ON ADVANCED REACTORS APRIL 21 & 22, 1981 A ron-licensed safety review will be perfomed in accordance with applicable DOE regulations and with requirements and procedures of an independent tech-nical review as agreed upon by DOE and NRC.
The approaches to core disruptive accidents (CDAs) include prevention, miti-gation, and risk assessment.
In prevention, they will reduce the probability of CDAs sufficiently to justify their exclusion from plant design basis acci-dent (DBAs) spectrum. 00E has issued a report, CDS 300-3, which addresses CDA prevention through reliability of safety features.
Mitigation will be
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provided by design conservatisms beyond those nomally included in DBAs to ensure an acceptably low risk to the public from accidencs beyond the DBA spectrum. CDS 100-12 addresses CDA mitigation. A probabilistic risk assess-ment shall be performed to demorstrate that public risk from plant operation on its specified site shall not exceed LWR risks described in WASH-1400.
TMI Lessons Learned have been considered in applicable aspects of the design.
John Graham, Westinghouse, discussed the key safety design considerations of the CDS program concerning areas of the core, safe shutdown, NSSS plant configurations, B0P considerations, and containment.
Mr. Graham stated that there are strong licensing implications of homo-geneous versus heterogeneous core configurations. However, it is known that the HCDA energetics associated with sodium voiding and early fuel failure tends to increase and becomes more sensitive to uncertainties
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MINUTES OF THE ON ADVANCED REACTORS APRIL 21 4 22, 1981 as core sodium void worth increases. With this consideration, a hetero-geneous core would have a licensing advantage provided that such a design does not introduce features which unacceptably compromise plant reliability and other operational considerations. A heterogeneous core design was selected.
A secondary shutdown system is required to meet design guls of shutdown reliability with system diversity and redundancy.
In addition, the design must be able to accomodate a self-actuated shutdown system (SASS) function which may be combined or independent of the secondary system.
Natural circulation cooling tests were perfomed on the FFTS in 1980 and 1981 at power level s of 5%, 35%, 75%, and 100%.
Results showed good agreement with calculations, though the calculations were generally conservative.
A discussion of NSSS plant configuration, pool versus loop design, was not discussed in depth.
The basis for the selection of the loop design was not explored fully by DOE and its contractors et this presentation. This topic will be discussed in detati at a future meeting, although DOE referred to a report CDS 500-1 for further infomation. This report will be distributed to the Subcommittee when received.
In-service inspection has a strong safety implication, but the best choices with regard to trade-offs between continuous monitoring, periodic testing, and periodic examinations and alternatives (e.g., inherent design features, plant protection features, and back-up and redundant design features) are not yet known.
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MINUTES OF THE ON ADVANCEh REACTORS APRIL 21 8 22, 1981 BOP considerations sucn as steam cycle, fuel storage, ventilation, cell lintrs and catch pans, hardening of steam generator buildings (seismic Category 1) and sodium smoke were discussed.
Cell liners are provided for all cells containing radioactive sodium in order to prevent sodium-concrete reaction and increase the risk of release of radioactivity following a sodium spill. A catch pan will be provided for all cells containing non-radioactive sodium.
Design descriptions and compliance with design requirements for the design basis event (dBE) were discussed f,or the reactor shutdown system, shutdown heat removal, and containment.
The power coefficient is expected to be negative over the complete power range of operations.
The design basis events were discussed. There are 27 events listed, many of them are similar to the LWR events.
The failure probability of the RSS is predicted to be 4 X 10-8/yr, primary, and 5 X 10-6/yr, secondary; the required design failure probability is < 10-0/yr.
The unavailability of the RSS is predicted to be 22 hrs /yr; the required design unavailabiity is <25 hrs /yr.
The shutdown heat removal system (SHRS) of the LMFBR dissipates the residual and decay heat of the operations during reactor shutdown to keep fuel and 4
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MINUTES OF THE MEETING
-6 ON ADVANCED REACTORS APRIL 21 & 22,1981 reactor systems temperatures below acceptable design limits. A reliable decay haat removal system is essential to protecting the health and safety of the public and the structural integrity of the plant.
This is 3chieved by:
(1) a direct reactor auxiliary cooling system (DRACS) which operates to remove heat directly from the sodium of the reactor vessel, and is independent of the main sodium loops, and (2) an intermediate reactor auxillary cooling syste.n (IRACS) which' ren; oves heat from the sodium in the main intermediate sodium system.
The CDS containment building is a so-called containm6nt/ confinement design.
The NSSS is confined by an ir
'e hemispherical-cylindrical steel shell of about 1.5 inches thick.
Surrounding this steel shell is a concrete con-finement of similar shape to the steel shell and has a thickness of 2.5 feet. An air space separates the inner and outer structure. The floor is concrete with a 0.25 inch thick liner.
Radiological dose analyses were performed using source terms from the CRBRP and ANS.
Hanford site meteorology (X/Q) was used. Doses were witnin NRC limit.
The containment / confinement assembly was designed to Region III tornado (240 mph and 1.5 psi), OBE of 0.15g and SSE of 0.30, is ternal missiles 9
l Large sodium spills are defined as a spill size of 620,000 lbs. caused by a pipe hreak area of 4 inches square.
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Cd ADVANCED REACTORS APRIL 21 & 22, 1901 The containment / confinement structure was designed for events beyond DBA or HCDA. The goals for HDCA accomn. dation are:
(1) minimum 24 hr. delay for venting, and (2) passive features inside RCB and active systems outside RCB, if needed.
R. Lancat, Atomics International, discussed safety system reliability. Pre-liminary conclusions of this analysis are listed below:
Reliability pr3 gram instituted to help assure safety.
Reliability goals established as design targets.
Reliability evaluation perfonned to assist in design tradeoffs.
Accident prevention reliability goals appear achievcble.
Accident mitigation features assessed in risk assessment.
P. Greebler, General Electric, discussed core disruptive accident ICDA) ener-getics considerations. The initiating accident scenario involves failure 4
to scram following 1 css of flow (e.g., loss of main pumping power for all primary loops) and transient overpower (e.g., uninterrupted control rod wi thdrawal ). The design approach to this concern is the use of a hetero-geneous core (low sodium void worth). This approach renders the calculated CDA energy releases very much lower than SMBDB (700 MJ fuel vapor isent?opic expansion to 1 atm).
Two separate phases of CDA have been studied. The initiating phase (core subassembly geometry retained) and transition phese (meltout molten rool, secordary criticality potential). R sults of CDS initiating phasa ener-getics assessments indicate that the CDA energetics is expected to be bound by the initiating phase.
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MINUTES OF THE MEETING 0-ON ADVANCED REACTORS APRIL 21 & 22, 1981 J. Graham, Westinghouse, discussed the considerations beyond the design basis for the CDS program. The basis for protection is the so-called line of assur-ance (LOA) approach of prevention and mitigation. For prevention, there are two diverse and redundant heat removal systems, and self-actuated shutdown SASS and IHRS.
For' mitigation, there are a neterogeneous low void worth core design, adequate mechcnical strengths (low risk margins) and containment thermal margins (low risk margins).
The mechanical strength of the primary system for the beyond the design basis can adequately resist loadings from a core disruptive accident of 700 MJ.
This loading value is bas ~ed on trends from improved understanding of prior history, new methods, and foreign positions.
Provisions may be made to relieve pressure in containment structures. if required, where those pressures may violate structurai integrity.
The containment thermal margins for the beyond the design basis is based on little holdup of debris in-vessel. The thennal loadings are calculated using the CACECO code which calculater pressure transients, temperature, and hydrogen concentra tions.
Mr. Temme, General Electric, discussed the role of risk analysis in the conceptusi design ttudy. He stated th'at risk analysis can be viewed as an aid to safety dt. cision.
At this stage of the project, there are ',1Ritations of risk analysis because ~ design details are lacking and definition of operation and maintenanc procedures is lacking. Therefore, best estimates and ranges of uncertainty are estimated by judgment.
Tbe meeting was recessed at 6:05 p.m.
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MINUTES OF THE MEETINb 3-O',4 ADVANCED REACTORS
. APRIL 21 &.22,1981 The meeting reconvened at 8:30 a.m. on April 22, 1981.
D. Ferguson, Argonne National Lab., discussed a summary of DOE LMFBR safety program activities in support of CDS.
The safety program includes reactor system reliability, self-actuated shutdown system development, shutdown heat removal under faulted conditions, CDA accommodation, and risk assessment and allocation.
F. Gavigan, DOE, discussed DOE's LMFBR safety R&D program. The purpose of the program is to provide data base to assess the risk to the public of LMFBRs. He stated that a safety " issue" is when a major area of concern and/or contention between regulators and applicants arises which must be resolved before the regulatory process can proceed.
The major safety issues are reliability of safety features, safety margins under degraded plant conditions, accommodation of core disruptive accidents, and plant siting and accident health effects.
An executive session was held to schedule future meetings and to discuss issues that designers of LMFBRs needed to answer before the program could go forward.
Concerning the next Subcommittee meeting, a tentative date has been set for May 14 and 15,1981 at the Royal Court Inn in Des Plaines, kl.
F. Gavigan will check the availability of the people doing the presenta-tion and the possible items to be covered.
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MINUTES 0F THE MEET!%
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- 0N ADVANCED REACTORS
' APRIL 21 &~22, 1981 A preliminary list of the issues that the Subcommittee will address are listed below:
Overall Safety Philosophy and Goals Shutdown-Systems Shutdown Heat Removal
- Preventing Energetics Subassembly Propagation Fuel Failure Propagation Accommodation Energetics Reaction and Accident Debris Accommodation Spent Fuel Cooling Siting Pool vs. Pot Design Global Core Instrumentation Containment (venting, integrity)
Sabotage Equipment Qualification (environment)
Sodium Fires Advance Fuel Design Automatic vs. Manual Operation Seismic (piping, reactor vessel)
Standardization and Long-Term Perspective to Safety Solution The meeting adjourned at 4:15 p.m.
NOTE: For additional details, t con;plete transcript of the meeting is avail-able in the NRC Public Document Room,1717 H St., NW, Washington, DC 20555 or from Alderson Rep 6rters, 300 7th St., SW, Washington, DC, (202)554-2345).
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Federal Register / Vrl. G RS 8C / MG8Y* Aprii7,1981TNitices ~
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En.ployment cnd Tr-Ining section 552b of Title $.Unbed St:tes ronaldered during the balanca,cf the Administration. 801 D Street NW.,
Code.
meeting.
Wa hington.D.C.20213.
Further formation with reference to
%e Subcommittee will then hear Cageed at Wnhington, Dt. this sist day this meeting can be obtained from Mr.
presentations by and hold d;scussions e4 March 196t.
John H. Clark. Advisory Committee with representatives of the NRC Staff.
Im.a Sepulveda, Management Officer. Nations!
their consultants, and other interested k
Acting Director. Office offrogmm Services.
Endowment for the Arts, Washington, persons regarding this review.
D.C. 20506. or ccll(202) 634-6070.
Further informa tion regardin ; topics arch John H. Clark, to be discussed, whether the meeting Name ofApplicant and1.ocation of Director. Office ofCounci/endPonel l'es been cancelled or rescheduled, the Enterprise and PrincipolProduct or Acuvity Operations. NationalEndo wmentfor the Arts Chairman's ruling on request for the arch 30.1981.
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and the time allotted therefor can be Wynla a cooti m i-u obtained by a prepaid telephon; call to Production of cast semia:rssta'.line siliccn
. bricks and wafers for photovoltaic cells the cognizant Designated Federa!
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Retsil sales of football basebe!!. hockey COMMISSION between 8:15 a.m. and 5-00 pn EST.
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Advisory Committee on Reactor Ohio Safeguardt, Subcornmittee on Subsection 10(d)of the Federal at $Justries. Inc., Siren & Fs un.
v n ed Reactws;Meeung Advisory Committee Act. that it may be necessary to close some portions of this Esconsin.Nevis. Minnesota The ACRS Subcommittee on meeting to protect proprietary Manufacture of wood juvenile fumiture.
Advanct < Reactors will hold a meeting information. %e autaority for such plastic lawn and garden fumiture and at 8:30 a.m. on April:!1 and 22.1981 closure is Exemption (4) to the Sunshine bird feeders
. Royal Court Inn.1750 S. Elmhurst Road.
Act. 5 U.S.C. 552b(c)(4).
TigerTeilDistdlery.Irr Dyersbury.
Des Plaines. IL. De Subcommittee will
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T'""' 8 8" Manufacture of ethanol discuss matters relating to the Dated: April t. test.
development of1.MFBR safety design John C. Hoyle.
MnNa obi f n ure d criteria. Notice of this meeting was AdvisoryCommitte ManagementW/icer.
s ladders published March 27.
pu ow. m-smei ra.4 ++et an..
Regency Health Care Center.LaFo!!ette.
In armrdance with the procedures anwo cooe Foss.es-se 1
Tennessee outlined in the Federal Register on Intermediate care nursing services October 7,1980 (45 FR 66535), oral or
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j Media Arts Panel (Programming in the
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g d Mfy j
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sessions may be necessary to discuss October 7,1980 (45 FR 66535), oral or g
Washington. D.C. 20506.
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%e entire meetin subsecticas (c) (4) and (6) and 9(b) of vi-ws regarding matters to be public attendanca. g willbe open to
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TENTATIVE SCHEDULE FOR THE APRIL ~21 & 22, 1981 ACRS SUBCOMMITTEE MEETING ON ADVANCE REACTORS DES PLAINES, IL g
APRIL 21, 1981 APPROXIMATE TIME 8:30_a.m.
I.
Chairman's Opening Statement II.
Presentation by DOE 8:45 a - 9:45 a A.
Introduction Conceptual Design Study (CDS) documents addressing design safety Summary of CDS positions on design safety (philosophical statements, overall plant layout and design) 9:45 a - 12:00 n B.
General Design Criteria (GDC) and CDS Positions on Design Choices with Respect to Key Safety Issues... the back and forth discussion, etc., between the designers and the safety people.
10:30 a - 10:45 a Break 12:00 n - 4:00 p C.
Description of Systems for Key Safety Issues Shutdown Decay heat Removal
- Containment 1:00 p - - 2:00 p LUNCH 4:00 p - 4:15 p BREAK 4:15 p - 4:45 p D.
Safety System Reliability Estimates 4:45 p - 5:45 p
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CDA Considerations and Evaluation 5:45 p - 6:00 p BP.EAK 5:00 p - 7:00 p F.
Risk Assessment 7:00 p - 7:15 p III.
Chairman's Comments j
7:15 p IV.
Recess I
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~ TENTATIVE-SCHEDULE FOR
^ 2' THE APRIL;21.8 22, 1981' ACRS4 SUBCOMMITTEE MEETING ON ADVANCE REACTORS-DES PLAINES, Il f
J APRIL 22,~ 1981-APPROXIMATE TIMi--
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I.
Chairman's Opening Statement II.
Presentation by DOE. (Cont'd)
~8:45 a - 19:45 a
~A.
Current R&D Approaches and What-is
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B.
Unresolved Issues and Summary -- What are the Contention Items ?
fl0:45 a - 11 45 a.
BRE AK -
11:45 a
=12:45 p III.
Executive Session
~12:45 p 11:45 p-LUNCH 1:45 p_ - 5:30 p-Continue Executive Session (break of 15 mins. as appropriate) 5:30 p Adjournment 4
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