ML20236T891
| ML20236T891 | |
| Person / Time | |
|---|---|
| Issue date: | 06/24/1987 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2509, NUDOCS 8712020150 | |
| Download: ML20236T891 (21) | |
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ih j DATE ISSUED: 6/24/87 7/1/f7 PROPOSED MINUTES /
SUMMARY
OF THE ACRS ADVANCED REACTORS SUBCOMMITTEE MEETING JUNE 17, 1987
. WASHINGTON, D.C.
Purpose The purpose of the meeting was to review and discuss the three DOE sponsored advanced reactor conceptual designs. These are: (1) Modular HighTemperatureGas-CooledReactor(MHTGR),(2)PowerReactorInherent-lySafeModule(PRISM),and(3)SodiumAdvancedFastReactor(SAFR)..
Feeting Attendees ACRS NRC Staff D. Ward, Chairman T. King, RES i
J. Ebersole, Member J. Wilson, RES l C. Siess, Member C. Allen, RES C. Wylie, Member R. Landry, RES M. El-Zeftawy, Staff Engineer R. Johnson, RES A. Tabatabai, Fellow Others DOE N. Brown, GE F. Gavigan F. Tippets, GE j A. Millunzi R. Lancet, RI G. Sherwood J. Brunings, RI A. Campise, RI l S. Ball, ORNL V. Gat, ORNL l A. Neylan, GA .l G. VanTuyle, BNL l J. Kendall, GCRA i
P. Kroeger, BNL F. Silady, GA l D. Pedersen, ANL 1 l Y. Boyer, Consultant !
l S. Maia, TEPC0 D. Graf, P0C0 i
Highlights, Agreements, and Requests
- 1. Mr. Ward introduced the ACRS Subcommittee members present and stated that the purpose of the meeting is to familiarize the Subcommittee members with DOE advanced reactor plans and concepts 0712O20150 870624" 9 pg Certified By _ s,
I~ Advanced Reactor Deigns Minutes June 17, 1987 i
and the NRC Staff review plans. Mr. Ward indicated that this j review and discussion will be helpful to prepare the ACRS for the task of reviewing the formal design proposals and the Staff's safety review over the next several months or years for those advanced designs. ) 1
- 2. Mr. F. Gavigan, Director of the Office of Advanced Reactor Program at DOE, overviewed the HTGR and LMR programs. He stated that the demiseoftheClinchRiverBreederReactor(CRBR)wasadepressing )
event which indicated the Congress was not interested in large- l scale breeder reactors. DOE decided to look at the utility's needs .)l and at smaller size (modular) reactors. The modular reactors reduce required utility planning horizons, have easier public acceptance, reduce on-site utility construction, and enhance QC/QA factory production. The new directions will have the following ..
advantages:
l l All safety systems are centered on the nuclear island, Balance of Plant (B0P) is not needed.
Heavy emphasis on passive and inherent safety. 1 Il
- Factory fabrication and short construction time.
- Plant investment is a separate requirement from safety requirement.
Mr. Gavigan indicated that early interactions between DOE and the NRC through the advanced reactor policy statement and NUREG-1226 (implementation plan) have appreciably helped clarifying DOE's needs for the advanced concepts. GE and W have advanced two large evolutionary improved LWRs with the Japanese in a joint funding program. Less than two years ago, DOE and EPRI jointly got in- +
volved in a mid-size LWR (600 MWe) design. The funding level for the mid-size LWR program is $15 million/ year from DOE. The HTGR, y
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-l1 Advanced Reactor Deigns Minutes June 17, 1987 I
DOE requested $5 million for this year, however, the Congress voted out for $28 million in addition. For the LMR, the designs them- )
selves are about $14 million/ year, with the total LMR program
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roughly $200 million/yr. Attachment I is a recent LMR program j schedule. Attachment II presents a general schedule of application' q milestones, i i
- 3. General Electric Presentation Mr. N. Brown, GE, described the design approach for the PRISM.
concept. The PRISM concept emphasizes inherent safety characteris- !
tics 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 j plant combined power output is 1245 MWe. Each module is a pool .
type LMFBR design with its own intermediate heat transport system and steam generator system. Steam from the steam generators is ;
piped to a single turbine' generator to form a power block of about i 415 MWe and there are a total of three power blocks for the plant.
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l 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 facilities 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 guards vessel and the reactor vessel are made of 316 stainless steel.
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The PRISM reactor core is a homogeneous design which metallic fuel j (uses oxide backup) with an average temperature rise of 265 F. The core structural material was chosen for its low irradiation swell-ing characteristics. The core lattice is being selected to be j capable of breeding, ultimately' with on-site fuel reprocessing, i
Mr. F. Tippets, GE, indicated that the small size of each reactor I module facilitates the use of passive inherent self-shutdown and shutdown heat removal' features. The PRISM plant is equipped with I three methods of removirg shutdown heat from the reactor; condenser cooling, auxiliary cooling system (ACS), and the reactor vessel .)
auxiliarycoolingsystem(RVACS). The normal shutdown heat removal is by condenser cooling. Failing that, shutdown heat is removed j 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),
RVACS 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 tempera- i ture, the higher the RVACS heat removal rate. 1 The reactivity control and shutdown consist of the primary control l rod system and the secondary control rod system. The primary control rod system consists of six assemblies for power control, burnup compensation as well as shutdown capecility in response to demands from the plant control system. The secondary control rod system is comprised of three assemblies and provides redundant >
shutdown capability. It deploys in response to loss of electrical power, signals from the plant protection system or sustained over-temperature coolant. Failure.to scram probability is less
l4 Advanced Reactor Deigns Minutes June 17, 1987 than 10-6 per demand. The balance of_ plant (B0P) is completely disconnected from the primary loop safety considerations. l 1
Mr. Tippets indicated that the PRISM individual and societal risks are well below the NRC safety goals guidelines. The PRISM design
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complies with the advanced reactor policy statement.. Attachment !
III describes the PRISM project plan.
- 4. Mr. A. Millunzi, DOE, presented the overview of the MHTGR program.
DOE developed a plan of technical interactions with the NRC Staff I l
regarding the MHTGR's design and licensability. The technical l interaction approach includes the submittal of top-level criteria.
The top-level criteria are defined as the standards for judging l licensability that directly specify acceptable limits for pro-- I l
tection of the public health and safety. The philosophy is to produce a safe, economical design which meets NRC and user require- i l
ments by providing defense-in-depth through pursuit of four goals:
Maintain safe plant operation Maintain plant protection !
- Maintain control of radionuclides release-Maintain emergency preparedness i The sources used for development of the tcp-level regulatory L criteria are:
NUREG-0880 - Safety Goals for Nuclear Power Plant operation.
10 CFR 20 - Permissible dose levels and activity concen-trations in unrestricted areas.
10 CFR 50, Appendix I - Numerical. dose. guidelines for meeting the criterion, "As Low As Is Reasonably Achievable," for power reactor effluents.
Advanced Reactor Deigns Minutes June 17, 1987
" 10 CFR 100 - Numerical dose guidelines for determining the exclusion area boundary, low population zone, and population ;
center distance. j EPA-520/1-75-001 - Protection Action Guide Doses for Protec- ,
tive Actions for. Nuclear Incidents. I i
DOE started out with over 16 HTGR design concepts and after careful evaluation of the economic and technical feasibility through interactions with the NRC Staff, it decided on one smaller sized HTGR design concept. In March 1985, the side-by-side steel concept !
l was selected as the reference design to be further developed. The i concept was studied with two types of fuel, pebble bed and pris-matic. In August 1985, the concept using prismatic fuel was selected for further development in FY 1986. Attachment IV l 1
outlines the schedule for the preapplication with NRC interactions. I Mr. V. Boyer, Chairman - Gas Cooled Reactor Associates Management Committee, summarized the utility / user requirements for MHTGR. The criteria are as follows: l
- Total outage of less than 20% over lifetime of the plant. !
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- Probability of exceeding safety-related design limit of less than 10-5/ plant year.
No sheltering or evacuation planning is required.
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- Seismic (groundacceleration)of0.3gSSE. I
- Fuel cycle with enrichment level of less than 20%. I
- Installed capital cost of less than $2000/KW (1986 Dollars). ,
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Advanced Reactor Deigns Minutes June 17, 1987 i
Mr. A. Neylan, Director - MHTGR Project Division /GA Technologies, summarized the MHTGR general design description. The standard i MHTGR design is based on relevant experience derived from the design and operation of such plants as Peach Bottom and Fort St.
Vrain. It uses ceramic coated fuel particles and pressurized i helium as a coolant. The fuel (LEU /Th-Uranium TRISO coated parti-cles) is arranged in an annular pattern with control rods located in both the central and side reflector regions. )i The prismatic fuel option uses four modular, steel vessel reactors in the side-by-side configuration..each operating at a power level of 350 MWt and supplying steam to two turbine generators. The net plant electrical output is 558 MWe.
i The reactor is housed in an uninsulated steel vessel about the size of a BWR reactor vessel. The reactor vessel is contained in an underground concrete silo or cavity which is designed as a heat removal system of last resort. This configuration also reduces l seismic amplification. The nuclear island portion consists of four reactor enclosures and adjacent structures that house fuel handl- j ing, helium processing, and other essential reactor service sys-tems. A common control room is used to operate all four reactors l and the turbine plant. A specific site has not been selected yet.
Normal decay heat removal is accomplished by forced circulation of helium and removal of heat through the steam generator or through a small auxiliary system cooler. Passive means are provided as a backup. The design has no containment. A confinement system is used in the design. ,
DOE claims that radiation releases, including the release that could come from a core heatup during the passive decay heat re-jection mode, result in public doses that are well below statutory 1
Advanced Reactor Deigns Minutes June 17, 1987 limits. In addition, doses will be below the EPA Protective Action Guidelines (PAG) dose of less than 1 rem whole body and 5 rem thyroid, indicating that a public evacuation plan should not be '
required.
Mr. F. Silady, GA Technologies, summarized the MHTGR licensing approach as follows:
Safety design approach emphasizes radionuclides retention within fuel with passive design features.
Licensing basis events systematically selected using PRA.
- 10 CFR 100 design criteria developed from top-level regulatory criteria.
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- 5. Rockwell International's Presentation Mr. J. Brunings, Rockwell International (RI), presented a brief description of the SAFR design. The SAFR design is a modular liquid metal reactor designed to produce 350 MWe net per module for a utility power grid. The plant has been designed by a project team headed by RI with Bechtel group, Inc., and Combustion Engi-neering as subcontractors under the sponsorship of D0E.
The overall objectives of the SAFR program is to develop a concep-tual plant design that is economically competitive with other power j sources such as coal and LWRs. RI envisions four 350 MWe SAFR modules per site. Each SAFR unit is a pool-type design with passive decay heat removal. The reactor outlet temperature is 950 F, .using superheated steam cycle. The reactor core uses metal-fuel, and can accommodate derated oxide as a backup. The personnel radiation exposure shall be no higher than 20% of current LWRs.
Advanced Reactor Deigns Minutes June 17, 1987 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 (currently set at900MWt). All major components will be designed to be shop D bricated and shipped as complete units. No site has been select-ed for the plant. The layout is considered generic and is intended for c wide range of site conditions.
The nuclear island (NI) consists of three buildings, the reactor servicebu!1 ding (RSB),thesteamgeneratorbuilding(SGB),andthe reactorconteinmentbuilding(RCB). These buildings are placed on a common mat at grade level. The NI buildings are compactly l arranged around the containment building and share common walls.
The reactor assembly is located close to the center of the NI.
This arrangement minimizes the seismic effects on components and piping, and reduces cost and complexity. The SGBs are closely coupled to the RCB, thus minimizing the length of the large diame-ter sodium piping in the intermediate heat transport system.
A number of considerations important to plant design and operation were factored into the overall plant arrangement. For example, all potentially contaminated areas of the plant are grouped together so that personnel access and radioactivity can be tightly controlled.
Safety-related facilities are separated from non-safety-related facilities. This allows the use of non-nuclear-grade standards and procedures for design and construction of the non-nuclear side of the plant, with significant capital cost advantages.
The features of site arrangement allow a SAFR unit to be construct-ed in 28 months from start of structural concrete to initial fuel load.
Advanced React s P?igns Minutes June 17, 1987 Mr. R. Lancet, RI, presented the safety characteristics of the SAFR design. He stated that the plant design incorporates a high degree of inherent safety to protect the public and the plant safety. It will be a design goal to ensure that anticipated design basis events will not require active safety' systems to prevent signifi-cant damage to the core or primary coolant boundary, and that time margins in excess of 8 hr are.available for corrective action.
A passive, natural circulation means will be provided to remove decay heat from the reactor in the event of unavail-ability of the normal power generation train with sufficient reliability to meet NRC goals. (Currently 10~4 core melt l accidents per reactor year and < 5 x 10~7 prompt or 2 x 10-6 latent fatalities per reactor year.)
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- Two independent reactor shutdown systems will be provided and will be actuated by trip signals from the plant protective system. One system shall also be self-activated upon sodium overtemperature.
- The design will be insensitive to plant transients originating outside the primary reactor vessel.
System arrangement and plant layout will be such that sodium spills, leaks, or fires will not preclude reactor shutdown or shutdown heat removal.
The shutdown heat removal system (SHRS) for SAFR consists of three heat removal paths: the normal path through the steam generator and condenser; the reactor air cooling system. (RACS) 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
Advanced Reactor Deigns Minutes -11 . June 17, 1987 sodium to air heat exchanger. The RACS system operates continuous-ly and is considered the only safety grade system designed to accommodate an SSE and a design basis. tornado. It is considered inherently safe and reliable, because it has no active components and it does not require operator action, and is ' independent of electrical power. DRACS 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 l for economic protection, i l
l 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 V). The SAFR plant conforms with the NRC advanced reactor policy statement.
- 6. Mr. T. King, Acting Chief / Advanced Reactors and Generic Issues Branch /NRC/RES, briefed the Subconcittee regarding the NRC activ-ities related to the advanced reactors programs. The objective of I
Mr. King's presentation was to familiarize the members of the Subcommittee with the DOE advanced reactor plans and concepts and to highlight the NRC Staff review plans. No ACRS letter is re- !
quested at this time. The purpose of the Staff reviews is to:
- Provide guidance early in the design process on the licensing-requ,rements for each concept.
l Assess the potential of the concepts to meet these require-ments.
- Provide a preliminary assessment of the adequacy of the D0E proposed R&D programs supporting each concept.
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- Provide insights as to where NRC sponsored research should be ]
undertaken to support NRC review of future licensing applica- .I tions of advanced HTGRs and LMRs.
For each of the three concepts, the NRC Staff is reviewing three j documents submitted by DOE, namely: (1)PreliminarySafetyInfor- ]
mationDocument(PSID),(2)PRA,and(3)R&Dprogramdescription.
I The Staff reviews for each concept will be documented via an SER, including (NRR concurrence, ACRS review, and Commission review).
SERs are intended to form the basis for Staff review of an actual application, if and when such an application is filed. The Staff
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is also planning to perform independent reviews and analysis on key safety issues of the plants through contracts with BNL and ORNL.
l The advanced designs are expected to have enhanced margin of safety over current generation of LWRs, Mr. King pointed out several factors that will be considered in the review of plant safety with emphasis on the capability and margin included in the design to prevent and mitigate severe accidents.in compliance with the Commission's severe accident and safety goal policies. Previous operating experiences, existing technology, proposed R&D supporting
, the design, and the compliance with the licensing criteria devel-4 l oped for the design are also important factors in the review consideration.
l Mr. King stated that there are two major issues which have to be resolved: (1)treatmentofsevereaccidents,and(2) containment plans. The advanced concepts are being designed to accommodate low l probability accidents (ATWS and Station Elackout). For the con-tainment issue, the aovanced concepts proposes:
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- MHTGR - fuel to perform containment function (no conventional containment building).
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Advanced Reactor Deigns Minutes June 17, 1987 i
PRISM - low pressure / low volume containment.
- SAFR - low pressure / low volume primary containment.
Other issues in the advanced concepts that the NRC Staff is study-i ing are the design and reliability of shutdown and decay heat removal systems. For the MHTGR - graphite structure, air / water ingress. For LMRs - use of metal fuel and sodium. fires. l
- 7. As a result of the Subcommittee's discussion, the Subcommittee !
members raised some concerns regarding the following:
- Dr. Siess commented that the off-site evacuation plans should be the FEMA's responsibility. The Staff replied.that is true, J however, it is the NRC's decision to determine the radius of 1 the emergency evacuation zone.
1 Dr. Siess commented that the Staff must be able to place
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enough confidence in the analysis when considering low l probability events (in the order of 10-0) in the policy statement for advanced reactors to treat severe accidents.
Dr. Siess also mentioned that the safety goal is set so low ;
that nobody will believe that it has been achieved.
- Mr. Ebersole commented that the severe accident policy is conditioned to the existing LWRs. The Staff replied that the ,
severe accident policy should apply in principles to any type of plant.
Dr. Siess commented that the decision to evacuate personnel-is going to be a political decision rather than depending on whether the protective action guidelines (PAG) are to be exceeded or not.
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- Advanced Reactor Deigns Minutes . June 17, 1987 I
'I Dr. Siess requested the Staff to check their number of 10-6 probability for catastrophic failure of pressure vessel. The Staff agreed.
- Dr. Siess asked if there is a utility that expected to buy and build one of the MHTGR or SAFR plants or DOE is planning to-pay for it. DOE's representatives' replied that the government at this time does not intend to pay for the first licensed commercial plant and whatever gets done as part of the con-struction would be in connection with the private sector ;
initiative. However,'the plan for the prototype module (and .
testing) of the PRISM design is expected to be funded by DOE,
- Mr. Ward expressed some concern regarding how the industry would maintain balance-between prevention and mitigation for severe accidents and specifically how to apply it to the '
advanced designs.
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- Dr. Siess questioned the difference between a safety grade control room and a non-safety grade control room. The Staff replied that their current position is a non-safety grade control room would be acceptable provided that the remote shutdown area is in place that can protect the operator to initiate shutdown.
Mr. Ebersole expressed some concern for the PRISM design in a case of a major steam blowdown, the liquid metal systems are ,
subject to extraordinary,v thermal transients, due to the fact that a large steam gener stor would be connected to a smaller
. sodium system.
- Mr. Ward commented that Gd in their design of the PRISM ,
concept has made changes in the direction of simplification,
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i Advanced Reactor Deigns Minutes June 17, 1987 l
l inherent safety and maintainability with emphasis on breeding, which is quite different from what DOE claims to emphasize. 1 GE's representatives agreed.
- Mr. Ebersole expressed some concern for the PRISM design in
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regard to the positive sodium void coefficient and the means j and ways'to get any of that positive void into the reactor.
Mr. Ward auestioned the use of saturated steam rather than superheated steam for the LMR design. GE's representatives replied that simplification was the motive.
Dr. Siess commented that if the ACRS would be interested'in the near future regarding the spent fuel disposal issue for j advanced reactors. I Mr. Ward commented that one of the concerns in LWRs regarding l seismic design is whether, for beyond the the SSE specifica-tion, the whole system falls over a cliff. FortheLMR(e.g.,
PRISM) design, GE is introducing isolators for the lateral acceleration only for better site flexibility and margin. It is not clear by introducing these isolators, may be philosoph-ically, that the cliff issue might be enlarged.
Mr. Ebersole expressed some concern regarding the fire pro-tection issue for advanced reactors. RI's representatives !
replied that the advanced LMR design is very. insensitive to fires.
- Mr. Ward commented that the inherent favo.rable response characteristics depend on the detailed design of the core, l internals, and surrounding str';ctures. -The detailed design has to be constructed and fab;1cated to the exact
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. Advanced Reactor Deigns Minutes June 17, 1987' specification to maintain these characteristics without modifications over 60 years (lifetime of the plant). Yet, the ]
applicant does not seem concerned about the effect of aging on I materials, creeping, modifications, etc.
- 8. Future Actions Mr. Ward will brief the full Committee (ACRS) in July 9-11, 1987 regarding the three DOE sponsored advanced reactor concepts'(MHTGR, PRISM and SAFR). 'No ACRS letter is requested at this time. !
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l 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, D.C., or can be purchased from ACE-Federal Reporters, 444 North Capitol Street, Washington, ,
DC 20001, (202) 347-3700.
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