ML20206B132
| ML20206B132 | |
| Person / Time | |
|---|---|
| Issue date: | 07/11/1988 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2583, NUDOCS 8811150346 | |
| Download: ML20206B132 (21) | |
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DATE ISSUED: July 11,1988
SUMMARY
/ MINUTES OF THE ACRS SUBCOMMITTEE MEETING ON THE ADVANCED BOILING WATER REACTOR JUNE 1, 1988 WASHINGTON, D. C.
Purpose A meeting was held by the ACRS Subcomittee on the Advanced Boiling Water Reactor on June 1, 1988. The purpose of the meeting was to begin the Final Design Approval review of this standard plant concept. This meeting concentrated on the first review module consisting of the Standard Safety Analysis Report, Chapters 4, 5, 6, and 15-1.
Presen-tations were made by the General Electric Company, and the NRC Staff.
Notice of this meeting was published on May 18, 1988, in the Federal Register. Attachment A is a schedule of presentations.
The meeting was entirely open to the public.
Richard Major was the cognizant Staff Engineer for this meeting.
Attendees ACRS GE Nuclear Energy C. Michel. ion, Chairman D. Wilkins C. Wylie, Member J. Quirk W. Kerr, Member C. Sawyer F. Remick, Member I. Kobsa D. Ward, Member D. Neal J. Ebersole, Consultant G. Wade R. Major, ACf 5 Staff Others NRC Staff S. Sharron, Bechtel D. Scaletti N
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,o Minutes /Aavanced Boiling Water 2
Reactor, June 1, 1988 G. Brown, NUKARC J. Tsao A. Omoto, TEPC0 C. Jan S. Maia, TEPCO L. Neal R. Ketchel L. Rubenstein Meeting Highlights, Agreements and Requests:
Openino Statement - C. Michelson, ACRS Subcomittee Chairman Mr. Michelson discussed ACRS plans for reviewing the Advanced Boiling WaterReactor(ABWR).
Currently the ABWR review is divided into four modules.
For each module the NRC Staff will conduct their review and issue an SER. The ACRS will also review each module one at a time and issue a letter report on a single module. There will be a fifth ACRS report commenting on the overall final design approval review. Prior to each ACRS letter, a particular review module wili be discussed at three subcommittee meetings. At the first subcormiittee meeting, an overview will be given.
Detailed questions will follow at the second meeting and any problems will be resolved during the third (letter due) session.
Mr. Michelson also mentioned three potential problem areas associated with.this review. These areas include:
(1)apotentialproblemifthe EPRI standardization effort becomes out of phase with the ABWR review; (2) how to address severe accident considerations; and (3) handling the collection of unresolved safety issues and generic safety issues.
Overview of the ABWR - Dan Wilkins, GE Dr. Dan Wilkens, General Manager of the ABWR program at General Electric, presented an overview of the ABWR design. He noted an
Minutes / Advanced Boiling Water 3
Reactor., June 1, 1988 international design effort to develop a standard plant has been ongoing for eigh! years. As applied in Japan the ABWR will be a 1350 MWe plant.
Currently, GE intends to meet the EPRI ALWR requirements program. GE is going to pursue design certification on the ABWR program. GE intends that those plants now planned for Japan will be essentially the same as the U.S. standard plant design that receives design certification.
Dr. Wilkens explained that many features on the ABWR have been used in existing BWR plants in the U.S. and around the world. The advanced core and fuel has been used in the U.S.
Internal recirculation pumps are used in Sweden's BWRs. The fine motion control rod drives have been used in Germany. The advanced pressure suppression containment is similar to U.S. designs. Engineered service and maintainability features have corre from Japan. However, the solid state digital control systems is a new technology used for the first time on the ABWR.
Tokyo Electric power Company has announced plans to proceed with the first two ABWR units in Japan. They will be Kashiwazaki 6 & 7.
The license applications have been recently submitted. Corr.ercial operation of the two units is scheduled for 1996 and 1998. The plants will be constructed by a joint venture of GE, Hitachi and Toshiba. GE will supply the nuclear steam supply, the fuel, and the turbine generators for these two units.
The current construction schedule for*the Japanese ABWRs is 48 months from the initial concrete pour to comercial operation. Prior to the first concrete pour, it is estimated that it will take 18 montl.s to 1
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Reactor, June 1, 1988 excavate the site. The Kashiwazaki site is located on a rocky sea coast on the Japan sea. The current U.S. standard plant review has a final design approval scheduled to be issued in late 1990 and a. design certification one year later. The Japnese construction permit for K-6 is expected shortly after the U.S. FDA is issued.
Dr. Wilkens noted that the l! censing review bases document between NRR and GE outlines key acceptance criteria. These include a core damage probability of less than 10E-5 per year and a freruency of exceeding 25 REM off site at less than 10E-6 per year. The licensing review bases document also suggests the level of detail to be included in this application and the design certification process. Guidelines ware also established for the detail necessary to review the advanced electronics design.
Dr. Wilkens stated that two of the FSAR review modules have been submit-ted to the NRC. The final two review modules will be submitted before the end of the year.
In response to a question from Dr. Kerr, Dr. Sawyer, GE, noted that GE does have internal performance goals for non-safety control systems. GE I
does have a written quantitative goal for the multiplexing system.
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Containment Arrangement Overview - Gentry Wade, GE Mr. Wade discussed the ABWR containment configuration. He explained the building is designed for ease of servicing and maintenance which hs a
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s Reactor, June 1, 1988 positive impact on safety. To aid equipment handling, servicing monorails ar1 incorporated into the design.
The ABWR uses a pressure suppression containment.
It is a cylindrical, lined, reinforced concrete containment. The reactor building and containment structure have been integrated into one structure. The drywell and wetwell are linked with horizontal vents to the suppresion pool. The design pressure of the containment is 45 psig. The seismic capability is greater than 0.3 g.
Plant mechanical equipment is separated into three distinct areas; plant control and instrumentation are separated into fcur quadrants.
In response to questions from Mr. Et,ersole, it was reported that SRV lines pass from the upper drywell to the suppression pool through the floor of the upper drywell.
It was pointed out this could form a i
potential suppression pool bypass path. Mr. Wade stated that these l
lines would not use double walled piping, but would be fabricated to a high quality standard, f
Dr. Remick asked if there were any provisions to vent the space above l,
the suppression pool. GE responded that there is no specific venting path, although inerting and de-inerting connections are in place. There is presently no specific containment venting system incorporated into the ABWR design similar to the European filtered vents. There is a standby gas treatment system that routes containment atmosphere through filters to the plant stack if certain radiation levels are exceeded.
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Minutes / Advanced Boiling Water 6
Reactor, June 1, 1988 In response to a question from Dr. Kerr, it was noted the 45 psig l
t containment design was based on a design basis accident analyses.
ABWR SSAR Expanded Scope - Joe Quirk, GE Mr. Quirk explained that the scope of the ABWR design has expanded from the nuclear island (containment and control building and the equipment containedineach)toalsoincludethebalance-of-plant (turbineisland and radwaste facility).
Items which still remain outside the scope of the ABWR include:
the service building, switchyard, cooling tower and ultimate heat sink.
t tiRC Staff presentation on Their Recent Trip to Japan - Dino Scaletti, NRC Mr. Scaletti explained the purpose of the Staff's trip to Japan was to support the Staff's ABWR review.
The objectives of the visit included establishing centacts with appropriate Japanese organizations, comparing i
U.S. and Japanese regulatory requirements, comparing GE's domestic design to the Japanese design, reviewing Japanese standardization policy, and visiting component testing laboratories.
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The organizations the Staff ret with in Japan included:
the Ministry of International Trade and Industry (MITI). Tokyo Electric Power Co.
(TEPCo).Hitachi,LTD.,andToshibaCorporation. The topics discussed with the Japanese covered severe accident features, PRA, fire 1
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Reactor, June 1, 1988 protection, design and analysis, and the licensing and construction schedule.
The Staff explained that the Japanese concentrate more on accident prevention than on severe accident mitigation, although severe accidents are considered to a limited degree. The Japanese perfonn PRA's and study the results, but severe accident considerations are not fonnalized as in the U.S.
The Japanese have no severe accident requirements. Loss of offsite power is considered in Japan, but station blackout is not.
The Japanese ao not believe their backup diesel generators will fail.
The Staff also had an opportunity to see a working model in 1/5 scale of the reactor vessel and internal pumps.
Full scale testing was being conducted on a single internal recirculation pump and on a fine motion control rod drive. The Japanese had also built and destructively tested a 1/10 scale redel of the ABWR containment. The rodel was pressurized and seismically tested.
Mr. Tinkler of the NRC Staff mentioned that the Japanese have procedures and a pressure hardened vent path for containment venting. The vent path is through the standby gas treatment system to the plant stack.
This would be usrd when the etntainment pressure reached 80 psig.
Details of the design were not available to the Staff, GE comented that they were not aware of this feature. GE is aware that TEPCo is implementing emergency procedure guidelines which bring into play existing vent paths in certain instances, but GE did not believe TEPCo has engintered a specifically hardened containment vent path.
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Reactor, June 1, 1988 GE was asked to clarify the Japanese use of containment venting at the next subcommittee meeting.
ABWR SSAR Chapter 4 Reactor Desion - I. Kobsa, GE The ABWR vessel and internals are intended to contribute to the design objectives of the ABWR. These include design simplification, additional margins, reduced maintenance costs, and lower occupational exposures.
All BWR st.eterials concerns have been addressed. The design will use either technology that has been established through experience, or a new design feature will be tested before use.
The ABWR reactor pressure vessel (RPV) is larger than the largest BWR-6 vessel.
It is 27 inches larger which allows room for the reactor internal pumps (RIP).
The RPV has a reduced radius closure head and fl a r.ge s. This reduces the number of closure bolts and size of the head reducing manufacturing costs and the time necessary to bolt and unbolt the vessel head.
The RIP penetrations are machined in the bottom head knuckle forging.
There are steam flow restricters which are integral to the vessel nozzle. Located in the RPV top head is the vessel vent and head spray nozzle.
Regarding the reactor internals, the ABWR uses a BWR/6 type stean dryer and steam separators. The core support structure is also similar (3 the BWR/6. The ABWR will use a boron carbide control rod without a velocity limiter. Thereareseparatelowpressurefloodspargers(2)inthe
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' Minutes / Advanced Boiling Water 9
Reactor, June 1, 1988 vessel located below the feedwater sparger which serves as a third low pressure flood train. There are two high pressure flood spargers that are permanently installed below the shroud head.
The ABWR will use the latest approved GE standard 8 by B fuel design.
As new fuel designs are approved, it is expected they will be used in the ABWR, Almost all reactor internal components are austenitic stainless steel type 304L or 316L per ASME or ASTM material specifications depending on product form.
The exception to this is the shroud support structure which makes the transition between the low alloy steel vessel and the stainless steel core support structure which is nickel-chrome-iron (inconel 600) as are the shroud head bolts.
Fabrication and processing of austenitic stainless steel conforms to the guidance given in NRC Regulatory Guides 1.31, 1.37 and 1.44 It was noted that the reactor vessel will have no welds in the beltline region. Thebeltline(areaof vessel that parallels the reactor core) will be formed by a single ring forging.
Mr. Michelson requested detailed presentations on ABWR material perfor-mance at the next subcomittee meeting.
The reactivity control system was discussed. The new fine motion control rod drives (FMCRO) have the sa'me scram response time as the piston drives used on the BWR/6. The drives position and individually support each control rod.
The drive prevents a rod drop through
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separation detection. The FMCRD can also prevent or limit the rate of rod ejection in a postulated pressure boundry failure through use of an electro-mechanical brake and a check valve used to equalize pressure between the FMCRD and vessel preventing a rapid blowout of a control rod. The control rod is positioned by an electric motor; positioning steps are now rnuch smaller than previous hydraulic drives.
The FMCRD features a diverse means of rod insertion. Both a hydraulic scram and an electric run in can be used. The scram discharge volume has been eliminated. The new control rod drives minimize rod drop consequences.
If a control rod becomes separated from the control drives during withdrawal the loss of weight is sensed, signaled, and rod withdrawal is blocked.
During a hydraulic scram latches engage in the guide tube and prevent the rod from backing down.
Mr. Michelson is interested in studying a failure modes and effects analysis performed for the FMCRDs. He wished to see a study where potential trouble source have been evaluated and dismissed or corrected, j
The control rod hydraulic scran system is power by a control rod drive pump (charging purop). The systen has a pressurized nitrogen backup i
i source.
i General Electric believes the FMCRD represents a proven technology based on operating experience. The FNCRDs have been used in Europe for fifteen years with very good results.
There has been no observed
' Minutes / Advanced Boiling Water 11 Reagtor, June 1, 1988 performance degradation. There is an extensive testing base developed in Europe over the years.
i Mr. Michelson requested further details on the experience obtained on the total drive units not just the drives themselves.
For instance, has the backup hydraulic scram system proposed for the ABWR been employed elsewhere? What are the internal details of the nitrogen accumulators?
l Mr. Quirk promised to supply a failure modes and effects report for the CRD system, included in this report would be a study of air supply contaminants, a gradual reduction of air pressure, over and under voltage problems, and seal material contamination.
ABWR SSAR Chapter 5, Reactor Coolant System and Connected Systems - I, Kobsa, GE Mr. Kobsa noted the reactor coolant system components include the reactor vessel and internals, the ten reactor internal pumps, the main steam, safety relief and feedwater systems to the isolation valves. The system does have an automatic depressurization feature. Connected to
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the reactor. Coolant systen are one reactor water cleanup system, two high pressure core flooder systems, and three residual heat removal l
systems, with a low pressure core flooding feature.
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Mr. Michelson requested that GE be prepared to discuss the details of i
i the safety / relief valves at a later meeting.
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Minutes / Advanced Boiling Water 12 Reactor, June 1, 1988 A major advantage associated with internal reactor coolant pumps is the elimination of the external recirculation piping and flow nozzles.
These nozzles had been located on the reactor pressure vessel below the core. Flow restricters h. ave been placed in the RPV steam outlet nozzles. The fact that the ABWR uses flow restricters in steam nozzles and has. eliminated external recirculations loops reduces the loss-of-coolant accident loads on reactor internals and the containment.
The internal pumps have variable speed motors which can be used to adjust power without rod moyen,ent. The internal pump inertia also provides a slow flow coastdown to keep the fuel within thermal limits.
The large RPV diameter places the pressure boundry material well away from the core. This reduces the neutron fluence and allows long RPV life with better protection against brittle fracture.
Mr. Michelson asked what percent of reactor power could be handled by i
the tafety relief valves.
Dr. Sawyer explained the SRVs could handle 100% reactor power at around 1200 psi. Mr. Michelson also asked if the SRV's "pop" open on the spring action.
GE agreed to check on the operation of the SRVs. The concern is simering effects that could erode the valve seat.
GE noted that hydrogen water chemistry will be offered as an option to provide additional margin against stress corrosion cracking. The Japanese have hydrogen water chemistry under study, but have not adopted a position on it yet.
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' Minutes / Advanced Boiling Water 13 Reactor, June 1, 1988 Regarding in-service examination of the reactor vessel, it was noted that there are no welds in the core beltline. The vessel itself will be made of forged rings. There will be no welds in the bottom head section containing the control rod drive penetrations. The circle seam between the CRDs and the reactor internal pumps will be inspectable with a remotely operated ultransonic positioner. With the elimination of jet pumps, there is ipproved access to the inside of the RPV.
Mr. Kobsa discussed key features of the reactor internal pumps (RIPS),
These features include:
wet motor, seal-less design with casing welded to the RPV, solid state adjustable frequency speed control.
continuous purge with clean water from the condensate storage
- tank, motor cooled by reactor building closed cooling water system via closed cooling loop on the RIP (one heat exchanger per pump),
impellers and motors removab:a without reactor draining, and back seating the shaft and blow out restraint hangers provide redundant LOCA prevention.
The Subcomittee questioned GE over the effects of a destructive short circuit in an reactor internal purp motor. Of concern was the path debris from the motor could take, possibly entering the reactor core and the effects of a cracked rotor casing.
The largest leak GE thought was possible from such a scenario was the leakage down the pump shaft.
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Minutes / Advanced Boiling Water 14 Reactor, June 1, 1988 Further discussion of these questions will be rescheduled at the next subcomittee meeting.
Mr. Kobsa discussed operating experience associated with the reactor internal pumps. There are nearly 100 pumps in service in European BWRs.
The usage is approaching 600 pump-years of experience. There has been only one incident of a major forced outage due to a pump-related prob-lem.
There has been extensive European testing of RIPS by vendors and NSSS suppliers.
In Japan testing is being conducted by Hitachi and Toshika.
There has also been Japanese government and utility testing.
The RIP is believed to be a proven technology.
Hydrogen Detonation and Damage +,o Safety Relief Valves - J. Quirk, GE i
The Subcomittee asked for assurance from GE, that an event which damaged several safety relief valves in an oversea BWR could not occur in the ABWR. The cause of the detonation was the fast compression of hydrogen and oxygen (generated by radiolytic decomposition of water which migrated from the reactor core to the main steam lines) by activt. tion of the pilot valve line lead to the SRV internals. To correct the problem either prevention of high concentrations of hydrogen and oxygen, or actuation of the SRY without compressing tile gases would be effective. The ABWR will use direct acting SRVs which activate with decompression. This should prevent any such problem, i
gdRSSARChaptor6EngineeredSafetyFeatures-C. Sawyer,GE 1
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"Minutes / Advanced Boiling Water 13 P.eac. tor, June 1,1988 Dr. Sawyer discussed the Standby Gas Treatment System. This system uses an advanced filter train design. The configuration of the filter l
housing and flow patterns should eliminate any untreated bypass of the filters. The filter itself is a single train. The capacity of this system is reduced from that used on previous BWRs. Although there is a single filter train, active components such as the dryer trains and exhaust valves are redundant to meet the single failure criterion. This system is only used when high radiation levels are detected in the containment atmosphere.
For normal operation a separate ventilation system is used.
Mr. Michelson requested design details for the Standby Gas Treatment System. Of concern was the maximum containment pressure that could be handled by this system. Loads associated with accidents were of inter-est.
Dr. Remick also question 6d the use of space heaters in the filter train.
Are these devices passive? Can they cause fires in the charcoal fil-ters? He was s1so interested in the possibility of de:ay heat from fission products trapped in the filter creating a fire. How would such a fire be extinguished?
Containment System (Section 6.2) - G. Wade, GE Mr. Wade described the ABWR containmen't. The primary containment structure is a reinforced concrete cylinder 97 feet high and 95 ft. at the inside diameter. The containment is steel lined. The containment
' Minutes / Advanced Boiling Water 16 Reaqtor, June 1, 1988 is structurally integrated with the surrounding reactor building and upper pools. The seismic design is for 0.3 g.
The design pressure is 45 psig positive pressure with a design negative pressure of 2 psi.
The volumes of the containment chambers were reported.
The upper drywell. volume is 223,000 cu. ft. and the lowtr drywell volume is 36,000 cu. ft.
The suppression chamber contains 210,000 cu. ft. of air space and 126,000 cu. ft. of water.
Mr. Michelson raised a concern over the timing of de-inerting the containment. What is the likelihood of venting the containment to de-inert and having a fairly significant accident? Could the action of containment sprays create a negative containment pressure (now that the non-condensibles have been exhausted) beyond the design? Can the containment be isolated again given an accident? These questions will be pursued at a future meeting.
There are eight vacuum breakers that connect the suppression chamber airspace with the lower drywell. They remain closed during chugging and pool swell. They would open on drywell or wetwell spray actuation.
The ABWR secondary containment surrou,1ds the primary contairrent.
It is entered through air locks and has sealed penetrations.
It operates at a negative pressure relative to the primary containment. The air flow in the secondary containment is toward those areas which could be poten-tially contaminated. Exhaust is monitored and can be routed through the standby gas treatment system.
' Minutes / Advanced Boiling Water 17 Reactor, June 1,1988 Emergency Core Cooling Systems (Section 6.3) - C. Sawyer, GE The ABWR has three divisions of ECCS.
Each division has a high and low pressure ECCS function. One branch of the Low Pressure Core irdectier, (LPCI) system can serve as a heat removal syctem which meant ;!ere are now three residual heat removal (RHR) divisions. One divi +,5cn of the high pressure ECCS (RCIC) u,.loys a steam driven pump. Wi+sn the elimination of the external recirculation piping, the isrgest single LOCA pipe break was eliminated.
Smaller capacity ECC5 systems prevM.
enough water flow to keep the core covered for all design basis accidents.
The core spray function has been eliminated. The core spray sparger has been replaced by a flooder sparger.
The change was made to be consis-tent with EPRI ALWR requireirents. The rationale for the change was to simplify maintenance and surveillance. This change has had basically no effect on the DBA analyses.
3 ABWR SSAR Chapter 15 Accide_nt Analysis - C. Sawyer, GE Chapter 15 covers transients and accidents.
Events considered are those which produce a decrease in coolant temperature, increase in pressure, decrease in flow rate, reactivity anomalies, increase in coolant inven-tory, decrease in coolant inventory, and anticipated transients without scram. GE has also produced a nuclear safety operational analysis.
This is similar to a failure modes and affects analysis for 56 indiviA -
al events.
' Minutes / Advanced Boiling Water 18 Reactor, June 1, 1988 The ABWR will use the standard BWR fuel design currently in use. This f';e1 provides substantial core design margins between steady state operation and limiting conditions for operation, i
New features o' he ABWR to eliminate rod withdrawal errors include a wide range net,
monitoring system for period based protection during i
startup and an advanced rod block monitor. The ABWR incorporated a three channel voting logic digital control system to make the control system single failure proof.
1 Transients in the ABWR are ext.ted to be milder due to design features such as increased water in the vessel and a reduction in the void coefficient due to the new control rods. The severity of overpressure j
transients has been reduced primarily due to the reduction in void T
coefficient.
(Neutron flux will increase at a slower rate than in previous EWRs due to the void collapse resulting from the pressure i
increase.)
I The ABWR has the necessary features to deal with ATWS. The control rods have diverse insertion modes (hydraulic scram and electric run-no). The j
ABWR has a recire. pump trip. A standby liquid control system is not e
i necessary for ATWS mitigation, but manual capability is av611able.
!.nalyses by GE of the effects of an ATWS indicate adequate margin in the f
ABWR design.
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'5e Nuclear Safety Operational Analysis (NSOA) was performed to identify safety functions and to demonstrate compliance with the single failure
' Minutes /Advanctd Boiling Water 19 Reactor, June 1, 1988 criterion.
Fifty six individual events were studied. The process r
involved studying each event to define the equipment necessary to achieve success given a single failure.
closing Remarks - Chairman Michelson Mr. Michelson closed the meeting by reminding the Staff and GE that the next Subcommittee meeting will be in November (15-16), 1988. This meeting will introduce the second ABWR review module and provide a opportunity for detailed questions on the first review module. A detailed schedule for the meeting will be developed by early October.
The meeting was ad,iourned at 5:00 p.m.
h0TE:
A transcript of the meeting is available at the NRC Public Document Room,1717 H Street, NW., Washington, D.C. or can be purchased from Heritage Reporting Corporation, 1220 L Street.
NW., Washington. 0.C. 20005 Telephone (202) 628-4888.
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F-
- s ACRS SUBCOMMITTEE ON THE 1
ADVANCED BOILING WATER REACTOR JUNE 1, 1988 WASHINGTON, D.C.
ABWR - FDA REVIEW MODULE 1 8t30 a.m.
1.
Opening Statement - C. Michelson
(.
Min.)
a.
Purpose, objectives b.
Future ACRS actions c.
k3 view of days schedule 8:45 a.m.
- 11. Overview of ABWR - G.E.
(45 Min.)
Overall description Special errphasis on a description t
of arrangertents inside containment Discussion of new balance-of-plant scope 9:30 a.m.
III. NRC Staff Discussion o' Recent Trip (60 Min.)
to Japan to Observe .e Japanese ABWR Program (15 Min.)
BREAK 10:00 a.m.
10:45 a.m.
IV. Reactor Design (75 Min.)
a.
Overview of Chapter 4 - G.E.
b.
Additional discussion - more detail - G.E.
o reactormaterials(Section4.5) o design of reactor control systems (Section4.6)
Review Status and Coments - NRC Staff c.
(60 Min.)
LUNCH
- 2:00 NOON 1:00 p.m.
V.
Reactor Coolant System'and Connected Systems (75 Min.)
a.
Overview of Chapter 5 - G.E.
b.
Additional Discussion: G.E.
o integrity of reactor coolant pressure boundary (Section 5.2) reactorvessel(Section5.3) o
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Schedule / June 1. 1988 l >
- Advanced Boiling Water Reactor i
o component and. subsystem design (Section 5.4)
Review Status and Comment - NRC Staff
.c.
2:15 p.m.
.VI.
Engineered Safety Features (75 Min.)
Overview of Chapter 6 - G.E.
a.
b.
Additional Discussion - G.E.
Containmentsys',as(Section6.2) o o
Emergency Core Cooling Systems (Section 6.3)
Review Status and Comtents - NRC Staff c.
(15 Min.)
BREAK 3:30 p.m.
3:45 p.m.
Vll. Accident Analysis (75 Min.)
Overview of Chapter 15 - G.E.
a.
b.
Additional Discussion - G.E.
decrease in reactor coolant o
inventory (Section 15.6)
ATWS(Section15.8) o Plant Nuclear Safety Operational o
Analysis (Appendix 15A)
Review Status and Comments - NRC Staff c.
5:00 p.m.
Closing Remarks - Adjeurn -
ACRS t
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