ML20246G293
| ML20246G293 | |
| Person / Time | |
|---|---|
| Issue date: | 02/13/1989 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2608, NUDOCS 8905150257 | |
| Download: ML20246G293 (31) | |
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' ft DATE ISSUED:2/13/89
' h (ed., i :, i.: a MINUTES OF SUBCOMMITTEE
-ON ADVANCED BOILING WATEP, REACTORS MEETING NOVEMBER 15-16, 1388 The Subcommittee on Advanced Boiling Water Reactors met on November 15-16, 1988, 7920 Norfolk Avenue, Bethesda, Maryland. Mr. Carlyle Michelson was the Chairman of this subcommittee. Other ACRS Members in attendance were:
C. J. Wylie, D. Ward, P. G. Shewmon, and F. Remick.
D. Okrent was a consultant to the subcommittee.
H. Alderman was the Cognizant ACRS Staff Member for this meeting.
Introductory Remarks by Chairman Michelson. He noted that the meeting is a continuation of the review of the Advanced Boiling Water Reactor.
He noted that the November 15th meeting would be a preview of Chapters 1, 2, and 3 of the Safety Analysis Report. The November 16th meeting would review questions pertaining to Chapters 4, 5, 6 and 15-1 of the Safety Analysis Report. Mr. Michelson mentioned that notice of these nieetings was published in the Federal Register on October 27, 1988.
n8' Dr. Walt Ardenne made the first presentation for the General Electric Company. Dr. Ardenne noted that he would discuss the safety classifica-N N$
tion for the Advanced Boiling Water Reactor. He remarked that the basis 08 g
for classification in the ABWR is the same methodology that has been
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$g used on all previous GE BWR's. He noted the criteria for classification is the American Nuclear Society Standard 52.1 for Boiling Water Reac-M"g)ggg}y$
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. Reactors, November 15-16,.1988 Mr. Michelson remarked that his understanding was that the NRC staff has not accepted ANS 52.1. Mr. Brammer, NRC agreed that the staff had not accepted ANS 52.1.
Mr. Brammer noted that certain areas of ANS 52.1 are not consistent with Regulatory Guide 1.26.
Mr. Ward asked if the staff does not accept ANS 52.1, what do they use as a basis for review? Mr. Brammer replied that the staff used Regu-1 latory Guide 1.26 and past experience.
Dr. Ardenne noted that there are two broad classes of classification, safety class equipment which is divided into safety class 1, 2, and 3 and non-safety class equipment. The non-safety class equipment is divided into two sub groups. One subgroup has specific requirements that are commensurate with the functions performed by that equipment.
The second sub group does not. All of the safety classes are Seismic Category I.
Dr. Okrent asked if the staff has under consideration the consideration of other than SSE based criteria for judging the seismic capability of equipment?
l Mr. Brammer replied that Dr. Okrent was contemplating something beyond the design basis event that the staff uses in their routine review.
Dr. Okrent said that the safety goal policy goes beyond the SSE with regard to the seismic contributions to risk.
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Reactors, November 15-16, 1988 4.
Chairman Michelson noted that application'of safety goals to the ABWR will be a special topic which will be handled at a later date. He said that for the present.we are trying to limit our discussion to the i
design basis, a
There was a discussion of the classification,of the fuel storage pool.
Dr. D. Ardenne stated that it has a Seismic Category I makeup system but I
i the general classification is non-nuclear.
Dr. Okrent asked if either GE or the staff had any thoughts as to whether Fire Protection should have some form of seismic capability?
Mr. Quirk responded that GE is looking at the Fire Protection capability of the nuclear island.
1 Pr. Miller - Staff Trip Report i
Mr. Scaletti noted that the Japanese have provided maintenance space l
around a lot of components inside of containments. They have added a second equipment hatch directly on the other side of the recirculation i
pump so they have direct access from the equipment hatch to the recirculation pump. They have improved the# design from the standpoint of maintenance and maintaining the facility, i
Chairman Michelson asked if Fukashima had guard pipes around the verti-4 cal relief valve pipes in the air space portion of the suppression pool area?
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s; Reactors, November 15-16,1988 j
Mr. Quirk replied that they.didn't.
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i Mr. Miller noted that the turbine is completely dismantled every year.
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Mr. Miller observed that radiation areas are indicated by flashing lights. This provides a visual warning.
Chairman Michelson requested layout drawings for the plant.
Mr. Quirk said the layout drawings will be in Mr. Michelson's hands either December 31st or January 30th.
Mr. Rubenstein noted that there is a good writeup of the Japanese regulatory process.
Mr. Ward said he would like to see this.
(Note - This has been provided to D. Ward, C. Wylie and-C. Michelson on December 6, 1988).
Mr. Rubenstein noted that a PRA is scheduled for January 1989.
Dr. Okrent expressed his concern that the containment design pressure will be established prior to the PRA.
Mr. Rubenstein agreed that there is an element of risk if GE and their designers don't anticipate what the final staff requirements on severe accidents will be.
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Reactors, November 15-16, 1988 Dr. Okrent asked if there was a document that would explain what the staff policy was on severe accidents for the future.
Mr. Quirk replied that the licensing review basis document establishes as an acceptable criteriaa core damage frequency for both internal and external events of ten to the minus five per reactor year.
For events having a frequency of ten to the minus 6 per reactor year, GE has committed to a requirement that the offsite exposures would be less than 25 rem.
Progress Report by General Electric Mr. Quirk briefly discussed the Licensing Review Basis Document.
He noted-the Licensing Review Basis Document attempts to identify the traditional areas that have proved to be problems in past reviews and establish aggressive acceptance criteria such that'once the General Electric Company proves that the design meets those requirements, the design will be licensable.
If during the review the staff develops new requirements, those re-quirements would need to be promulgated by staff management before they are addressed by the General Electric. Staf'f management would be defined as the CRGR and the NRR Office Director.
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l Mr. Quirk noted that during the ACRS Subcommittee meeting in June 1988, i
Chapters 4, 5, 6 and 15 were presented to the subcommittee. He remarked that Chapters 1, 2, and 3 had been submitted to the NRC staff in March
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'1988. He saidlthat they had submitted Chapters on instrumentation,
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auxiliary systems, and. quality assurance.to the staff in June' 1988. The outstanding chapters to be submitted are Chapter 19, (Severe Accidents
-- will include an internal and external events' PRA and power. blockout),
Chapter 18 (Human Factors Engineering),' Chapter 16 (Tech. Specs.), and; Chapter 10 (Turbine).
l Mr. Quirk noted that they had received two groups of NRC Review Questions. He said.they had responded to those questions _in April and.
September of this year.
l Chairman Michelson requested that the NRC staff provide ~to the ACRS copies of the GE Responses.
Mr. Scaletti agreed.
Mr. Quirk noted that the GE ABWR Safety Analysis report will have a I
separate volume listing every question, and have a matrix liF$1ng where the answer to that question appears.
Dr. Okrent noted that he had read a speech by one of the Commissioners where the subject of sabotage was raised in terms of advanced reactors.
He asked if this had been addressed?
l Mr. Rubenstein responded that the staff has a special task force on external events and he thought that they would address sabotage.
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Reactors, November 15-16, 1988 3
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Chairman Michelson noted that when module 3 comes up for review, he-intends to discuss sabotage during the subcommittee meeting.
Mr. Quirk noted that they had reviewed the quality requirements as practiced in Japan and concluded that they met the requirements of 10' CFR 50 Part B.
Mr. Wylie asked if General Electric would seek certification on the entire plant.
i Mr. Quirk responded that they would.
Mr. Wylie asked what level of detail is planned for the design certil fication.
Mr. Quirk replied that they will specify procurement information down to the capacities, redundancies, the ratings and functional performance.
They will stop short of nameplate information in order to ensure the competitive bidding practices can be employed for purchase of equipment.
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Mr. Quirk noted that regarding Seismology, they were designing for an OBE of 0.1g and an SSE of 0.3. These are based upon the EPRI ALWR 9
requirements. Mr. Quirk noted that the staff prefers the OBE to be about one half of the SSE.
He noted that this is a point of discussion between the staff and EPRI.
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Reactors, November 15-16, 1988 Regarding tornado frequency, Mr. Quirk noted that the EPRI requirements I
specify a tornado frequency of reoccurrence on the order of ten to the minus six years. The NRC Regulatory guide specifies ten to the minus seven. This is being discussed by the NRC staff and EPRI.
Regarding the tornado missile spectrum, GE references the missile spectrum identified in ANS 2.3.
This is a little different then that which is specified in the staff regulatory guide and is being discussed with the NRC staff.
Dr. Okrent asked about the possibility of liquification being a problem at values above the SSE.
The response was that this will have to be considered when an applicant comes in with a specific site.
Mechanical Equipment, (Charles Dillmann, Manager, Mechanical Equipment Desion ABWR Project)
Mr. Dillmann discussed questions raised at a previous nyeting. The first-question was with regard to whether the backup scram system has been used elsewhere. The backup hydraulic scram is ut,' 3 on all BWR's.
The backup scram allows the scram valves to operate by venting the air header. The plants also have Alternate Rod Insertion (ARI) which is a second mode of backup.
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Reactors, November 15-16, 1988 Another question asked if the nitrogen and water accumulator on the Hydrualic Control Unit (HCU) had been used in the.past. The accumulator is larger.
Basically the HCU is an existing component sized for a new duty.
Another question is what happens on low air pressure? The concern is that the low air pressure would allow the scram valves to creep open and they start to leak.
If they leaked, they would be using water from the accumulators.
The response was that water is continually charged to the accumulators by the Control Rod Drive (CRGD) pump.
If the leakage were large enough, it would cause the pump head to reduce to the point where there would be inadequate scram pressure. This would be prevented by an 31 ann and rod block.
If the pressure continued to decline after the alarm and rod block, the reactor would scram.
Another question was about low voltage to the scram solonods. At low voltage, the scram Solenoids would chatter and may eventually stick.
This happens at below 90 volts. There is quite a bit of margin before this happens. This problem has only occurred during construction.
If it should happen, the backup scram and electric run-in feature wculd allow shutting down the plant.
r Mr. Dillman briefly discussed ongoing failure modes and effects analyses. He noted that GE will be submitting a combined failure modes i
and effects analysis by the end of the year.
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Minutes / Advanced Boiling Water 10 Reactors, November 15-16, 1988 He noted the use of an electro-mechanical brake on the control rods.
The brake is actuated to open when the rods are driven in and out. The rest of the time the break is locked closed.
If the scram hydraulic line is ruptured, the brake keeps the rod from being driven out. The brake fails in a locked position.
Mr. Dillman discussed the reactor's internal pump. The question was, "with a destructive short circuit, what would happen to the pressure boundary?" Mr. Dillman noted that the destructive short circuit would be terminated very rapidly.
The power circuit to the Reactor Integral Pump (RIP) drive has redundant overcurrent protection. There are two circuit breakers in series, either one can open and break the circuit.
In addition, the solid state power circuit uses Thyristors. The thyristors will stop firing when subject to overcurrent.
A second part of the question was what would happen to any debris from a motor chort circuit condition. Mr. Dillman pointed out that because of the geometry involved and water flows in the area, any debris would be contained in the motor housing.
Mr. D111mann presented an overview of Section 3, (mechanical component design criteria).
He noted that all of the mechanical components have had extensive operating experience. Any changes to the components have been thoroughly tested.
In some cases, test data in conjunction with
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Reactors, November 15-16, 1988 1
analysis has been used to prove performance.
He said that appropriate l
pre-operational and start-up testing will be performed.
l Dr. Okrent asked if GE has any problems with the design in providing a considerable margin for scram given an SSE of 0.3.
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Mr. Dillman said that he believes they have a considerable margin.
Mr. Dillman discussed the environmental design and qualification. He noted that the first step was to define the environments in normal, abnormal, ~ test, accident, and post-accident situations. The definition includes time, temperature, radiation, and humidity. All of this information is tabulated. The equipment is then characterized as safety-related or not. The safety related equipment is characterized as operating in a harsh or mild environment. A harsh er.vironment is one where the environment changes as a function of the the event that the equipment is mitigating. Each equipment item will have its performance requirements identified including safety function and function time.
Some equipment operates during a LOCA and after a LOCA for a hundred days. Other equipment may only operate for the first 30 seconds of a LOCA. Acceptable and unacceptable failures are considered. The events involved are considered. Thisincludeswhakeventsaretheequipment exposed to and what events does the equipment have to mitigate.
Chainnan Michelson asked how is GE going to address external events?
Specifically, Mr. Michelson asked how they woulc handle a break analysis which included flooding and water effects.
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Mr. Dillman responded that they would provide a flood analysis. They would corisider leakage of non-seismic pipes and the consequences of a failure.
Dr. Okrent' asked if each of the safety trains had its own heating and
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venti 1'ation system and no interconnections?
Mr. Quirk responded in the affirmative.
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Dr. Okrent asked if there was any problem in getting insulation with a 60 year life.
Mr. Dillman replied that the detailed evaluation has not been done. He said if the detailed evaluation shows this to be a problem, then alter-native solutions will be proposed.
Dr. Shewmon asked if any regulations would have to be changed to license a plant for 60 years.
Mr. Miller said a change would be necessary. Under the current legis-lation, NRC cannot license a plant for more than 40 years.
Mr. Dillman noted that testing is performed for the qualified life of a component.
It is aged to the end of that life, and this may include vibration if necessary. The component is then run through dynamic events which include the OBEand SSE in combination with hydrodynamic 4
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Minutes / Advanced Boiling Water 13 Reactors, November 15-16, 1988 loads. Whatever event or events it must mitigate, and exposure to that environment are (,nsidered.
As part of the program, maintenance is addressed. A plan is proposed and documented to maintain the component qualification status throughout the plant life.
Mr. Dillman discussed materials.
He said that all the materials that i
are specified for the ABWR are based on proven successful test and l
experience and include the resolution of all known problems that have occurred.
He said the design is based on understanding and using all the materials properly.
He noted the environment is controlled to enhance the natural performance.
Dr. Shewmon asked if the water chemistry requirements in Japan are i
different than the EPRI guidelines?
Mr. Dillman said that for current piants they are different, but on the ABWR they will follow the EPRI guidelines.
Dr. Shewmon asked what is GE doing about design for inspectability?
Mr. Dillman replied that they make sure there is access to all the welds that require in-service inspection (ISI).
Dr. Shewmon asked if there would be welds in the central part of the pressure vessel?
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Mr. Dillman replied that the welds will be above and below the core.
j Dr. Shewmon asked if there were any welds with tracks for inspection.
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Mr. Dillman replied that they have provided tracks for automatic in-spection from the outside.
Mr. Ward asked about the wall thickness of the main ring section, i
Mr. Dillman replied about 71 inches.
Mr. Dillman noted that materials used include 304, 316, and nuclear grade stainless steel.
He said they also used the L grade with low carbon. To compensate for the low carbon, nitrogen control maintains strength.
Solution heat treatment is required wherever the material has been exposed to a sensitizing operation. He said in the case of cast-ings, they require a ferrite number or ferrite percentages of 8 to avoid IGSCC.
Dr. Shewmon noted that the 8 percent was a lower limit.
He asked if there was a upper limit. Mr. Dillman replied that he thought it was 11.
1 Mr. Dillman pointed out that they use Inconel 600 at certain stress areas in the reactor internals.
He noted that there have been problems with cracking of Inconel in crevices over the last 8 or 9 years.
He said that to prevent cracking, Inconel should iiot be used in a crevice application unless it has been stabilized.
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Minutes / Advanced Boiling Water 15 Reactors, November 15-16, 1988-Mr. Dillman said they use carbon steel primarily in piping. They require tough grades of material. They apply special fatigue design rules'above and beyond the code.
Dr. Shewmon asked if they use carbon steel in contact with the coolant.
Mr. Dillman replied they have carbon steel lines. They use carbon steel feedwater lines.
He noted that carbon steel was pretty good from a stress corrosion standpoint.
Mr. Dillman stated that the choice of materials and low fluence means that NDT shift is not_a problem on the ABWR even in the 60 year case.
Chairman Michelson asked, "do you use cladding for other than the reactor vessel?"
Mr. Dillman said he couldn't think of any use off hand. He said there may be some cladding in the heat exchangers.
Chairman Michelson asked what cladding materials are used for the reactor vessel.
Mr. Dillman replied they use 308L; or if it is two layers, 309 for the l
first layer and 308 for the second layer. He noted that the Japanese are using a single layer with special qualification requirements.
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Chairman Michelson asked what kind of heat treatment is given to the cladding?
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I Mr. Dillman replied that the cladding gets a heat treatment with the vessel.
Chairman Michelson asked if the 308 L cladding was adequate, even through it may be elevated to 1150F for long periods.
Mr. Dillman replied that as long as you have the ferrite level up, sensitization is below 1150.
He noted that you pass through the sensi-
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tization range getting up there and coming down.
1 Dr. Shewmon asked about radiation enhanced stress corrosion cracking?
Mr. Dillman replied that this is under investigation. The only indica-tions so f.ar have been in the cores and control rods that receive high fluence in a few years compared to the rest of the intervals.
Mr. Dillman pointed out that they are using low cobalt materials in most of the internals to reduce radiation from crud buildup.
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Chairman Michelson asked about Zinc injection to minimize radiation buildup.
l Mr. Dillman replied that an observation was made that plants with high zine levels in the reactor coolant had lower radiation levels in the
Minutes / Advanced Boiling Water.
17 Reactors, November 15-16, 1988 piping. -He said that in a zine rich environment the crud picks up'zine rather than cobalt. 'This led to zinc being injected into the reactor coolant to lower the cobalt levels in the crud.
Dr. Okrent asked how do' you get from.a 5000 hour0.0579 days <br />1.389 hours <br />0.00827 weeks <br />0.0019 months <br /> test to a hundred year life?
Mr. Dillman replied that the test results are extrapolated.
Dr. Okrent asked how good was the correlation.
Mr. Dillman replied that it wasn't totally an impirical correlation.
It was grounded in theory and empirically confirmed.
Chairman Michelson asked if GE was requiring hydrogen water chemistry.
Mr. Dillman replied that the ABWR is designed to use hydrogen water chemistry but is is not presently planned to use it in Japan.
He noted that the' Japanese are planning to have GE run a mini test in one of their plants to measure the increased radiation levels in the steam lines. He said they expect there will be a move toward hydrogen water
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chemistry in the future.
Mr. Dillman noted that in addition to the selection of proper materials, they watch the stresses.
He said the stress limits have been developed i
by a combination of theory and fact. They are specific to the material and other factors. The concern is to avoid stress corrosion cracking.
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Dr. Shewmon asked what is done in water chemistry to avoid crud buildup?
Mr. Dillman replied that they don't avoid crud buildup as much as to avoid what is in the crud. By keeping things like copper out, they can avoid crud buildup that is deleterious to materials. By keeping out things like cobalt, they avoid crud buildup with a high radiation level.
l Chairman Michelson asked about corrosion problems with zirconium?
Mr. Dillman noted that he was not knowledgeable about this subject and suggested that a special discussion on this subject would be order.
The recorded portion of the meeting was adjourned at 4:50 p.m. to reconvene the next morning, f
Thursday, November 16, 1988 I
John Chambers, General Electric Co.
Mr. Chambers said he would discuss conformance to the general design l
criteria (GDC). Mr. Chambers noted that the fine motion control rod l
drives met GDC 12, 22, 23, 24 and 28.
Mr. Chambers noted that with respect to GDG 12 (suppression of reactor power oscillation), they had some new features. These include select control rod run-in, ganging of many of the rods, the ability to move them in and out in finer increments, and anticipatory actuation of
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select control rod run-in on such events as loss of. feedwater heating or a trip of two or more reactor internal pumps.
l Chairman Michelson asked what was the power supply for the control rods.
1 Mr. James said it was the Class IE bus.
Mr. Chambers noted with respect to GDC 22 (protection system indepen-dence), with the scram backup being a motor drive there is a diverse way to get the rods in.
For GDC 23 (protection system failure mode), the scram discharge volume has been eliminated. This eliminates the potential fer.ommon mode failure.
Dr. Shewmon asked how ATWS was addressed? Mr. Chambers replied by (1) enhanced rod drop / ejection performance via electro-mechanical break, (2) class IE rod separation detection and (3) a standby liquid control system.
Mr. Quirk pointed out that in Section 3.1 of Chapter 3, all the general design criteria are addressed, and the manner in which the ABWR complies with each is listed, and references are made to other parts of the SAR where details may be found.
I Mr. Chambers noted that elimination of the external loops helps reduce l
the potential for pipe breaks.
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tiinutes/ Advanced Boiling Water 20 Reactors, November 15-16, 1988 For GDC-28 (Reactivity limits with an increased number of pumps), there is a less severe reactivity transient with a failure of any single pump.
For GDC 32 (inspection of reactor coolant pressure boundary), with the elimination of the external piping there is better access. This allows for better automation for ISI equipment and results in less overall exposure of personnel for maintenance and ISI.
For GDC 33 (reactor coolant makeup) and for GDC 35 (emergency core cooling), elimination of the large nozzles below the top of the fuel decredses the load on the ECCS system.
Mr. Chambers said the three division ECCS increases the overall indepen-dence, redundance and separation with regardito several GDC's.
Mr. Chambers mentioned the reinforced concrete containment vessel.
It is an integral design such that the containment and the reactor building are all integrally designed and constructed to add to the overall strength. This relates to GDC's 2, 4, 16 and 50.
Mr. Chambers discussed control and instrumentation (C&I).
He noted that they were using a digital measurement and control approach to C&I. He mentioned the increased redundancy, two out of four logic at both sensor and division level, and greater testability of the individual sensor and division of sensor bypass capabilities.
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' The subcommittee discussed the potential for failure of the I&C' systems L
through heat, fire and other causes.
It was agreed to postpone a failure modes discussion until the next meeting when this topic will. be again discussed..
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Mr. Quirk noted that one of the agenda items dealt with containment venting by the Japanese. He said that the Japanese plants are not designed for containment' venting.
- Chairman Michelson asked what was the ABWR position on venting?
Mr. Quirk replied that they don't have a position on venting as yet.
He said that their. goal was to demonstrate that the ABWR capabilities under severe accidents was acceptable without venti,ng.
Hal Townsend of the General Electric Company discussed pipe whip. He noted that the ABWR is designed to GDC-4 of 10 CFR Part 50, Appendix A for postulated pipe ruptures.
He said they are addressing postulated pipe ruptures of the high energy piping both inside and outside of containment. He remarked that they are considering all the protective measures specified in the standard review plan, Sections 3.6.1 and 3.6.2.
Mr. Townsend said they are also considering failures in moderate energy fluid systems and those are limited to cracking only. He said as that for those lines it is unnecessary to consider pipe whip protection.
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- The basic approach will be to follow current practices and place the necessary jet shields and pipe whip restraints on all the high energy
. piping per the current procedures.
p Mr. Townsend discussed leak-before-break.
He noted there is an option to consider leak-before-break in the current GDC-4. He said they intend to take credit for that, or at least provide the option for taking credit for that.
He said the SAR provided the methodology for taking credit for leak-before-break. The methodology addresses the fracture methodology for determining the critical crack size.
He noted that the lines selected as being mort applicable for leak-before-break (including the stream and feedwater lines) are in the drywell and steam tunnel.
He said they had not completed their study regarding which lines are' applicable and they have not selected all of the lines.
Mr. Townsend said the basic approach to leak-before-break is to provide any' applicant who is referencing the ABWR design the optfor, to apply (leak-before-break. The base design is being designed without that l-capability.
t Mr. Townsend pointed out that the current design is based on postulated pipe ruptures and not leak-before-break.
He noted that they would request the NRC review and accept the procedure for applying leak-before-break to the larger piping with the objective that the
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applicant could come' back and remove pipe whip restraints in its final application by applying an agreed to procedure.
iDr. Okrent. asked why Japan has not accepted leak-before-break?
Mr. Townsend said it was a matter of timing.
GE plans to discuss this with the Japanese.
..Dr. Okrent asked the NRC staff if leak-before-break has been accepted for BWR's?
Mr. Tsao, NRC Staff, replied that they have not accepted leak-before-break for any BWR's because of IGSCC problems. This has not been resolved.
Mr. R. Townsend discussed the seismic design of the plant. He noted that the peak ground acceleration for safety shutdown earthquakes being considered is 0.39 and the 0.B.E. design is 0.19 For the soil struc-ture interaction, he said they have evaluated and used 0.15g or half the SSE.
Mr. Townsend pointed out that two alternative analytical methods have been used. The basic method is a computer program known as SASSI.
Its a finite element modeling of the soil. As a check, an alternate computer program called CLASSI/ASK is used.
It treats the soil situation differently by using a linear continium mechanics analysis instead of finite elements.
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.. '., Minutes /Ady'anced Boiling Water 24 Reactors Noven.ber 15-16, 1988 The soil damping is obtained using a computer program called SHAKE.
This treats >the non-linear characteristics of the soil.
The SSE was not directly calculated. The SSE response was assumed to be double.the response from the OBE analysis. This is considered conserva-tive for the SSE.
i Mr. Townsend noted that the conclusion reached from the SASSI analysis was that the dominant influence on the response spectrum'within the reactor building is the stiffness of the soil around the site.
Dr. Remick asked if the groundwater had an effect on the stiffness of the soil? Mr. Townsend said he couldn't answer that question.
Dr. Remick said he would appreciate clarification as to what effect the groundwater has on the stiffness of the soil at some later date.
Mr. Townsend noted that they are complying with regulatory guide 1.12 for a 0.3a safe shutdown earthquake.
He noted the design calls for a complement of accelerometers and recorders to record earthquake response.
Mr. Townsend addressed a question regarding accidents during deinerting.
Mr. Townsend said that both the wetwell and drywell are inerted by the atmospheric control system to oxygen levels below four percent. He noted that the system is sized with the capability of inerting or deinerting the containment within a four hour time period from the start
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Minutes / Advanced Goiling Water 25 Reactors, November 15-16, 1988
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of operating the system, and injecting either nitrogen or air.
He said the criteria that is used is that if the oxygen level rises above four percent while at pows', there is a 21 hour2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> period to reduce the power r
below 15 percent of'the rated power. The concept is that there is a low' probability of operating for a brief period of time with the containment deliherted but above 15' percent.
Chairman Michelson expressed his concern regarding closure of the containment isolation valves upon an isolation signal.
I Mr. James said the isolation valves will be subject to all the require-ments:for isolation valves.
k Mr. Townsend noted that there was a question about inadvertent actuation of the containment sprays and the possibility of negative pressure in the containment.
l Mr. Townsend said he had looked at several cases of actuation of con-tainment sprays under different conditions.
He said the limiting condition was a stuck open relief valve followed by actuation of the containment sprays. Analyses of this event'shows the containment pressure will drop to 1.7 psi negative pressure. This has been used to set the containment liner capability at minus 2 psi.
Mr. Tony James, General Electric Co. discussed the ABWR safety relief valves. Mr. Jones pointed out that there are two types of valves. One I
1 i
e Minutes / Advanced Boiling Water 26 Reactors, November 15-16, 1988 type is the safety valves. These are values that open against springs.
They open automatically on increasing system pressure. They open at f
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pre-established pressure set points and will automatically close on j
spring force at a lower pressure set point.
He said the key point was that no actuating power was required for the safety valves.
The second type is the relief valves.
Some actuating power is needed to open relief valves.
For the ABWR the basic motive force to open the
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relief valves at any system pressure is pneumatic. Thye also open automatically on high system pressure without pneumatic assistance.
Mr. James noted that for BWR applications both functions are combined in Safety / relief valves.
He also noted that for BWR applications, steam is relieved.
He said there wasn't any requirements for high pressure liquid or two-phase flow relief from the vessel. Mr. James pointed out that when the safety mode and relief mode are combined in the same valve, the safety mode must be completely independent of, separated from, and completely unaffected by the relief function.
1 1
The code requirements on safety valves is to provide pressure protection l
during overpressure transients such as load rejection. This has to be done by the spring actuated mode.
In the spring actuated mode there must be sufficient capacity to maintain the vessel pressure below design pressure.
To provide depressurization during loss of coolant accidents, the relief valve function is used. The requirement is to provide sufficient
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Minutes / Advanced Boiling Water 27 l-Reactors, November 15-16, 1988 i
automatic depressurization-in the power operated relief mode so that ECCS can come on and flood the vessel.
The power actuated relief mode provides sufficient capacity to maintain-l l.
vessel pressure below some emergency limits during ATWS.
The safety / relief valves provide for remote manual vessel pressure control during. operations such as plant control.
Mr. James. pointed out that GE is not prescribing the details of the S/RV design for the ABWR. The performance requirements will be specified.
He noted that GE is requiring position monitoring devices on the valves and a tight leakage requirements.
Mr. James noted that 18 S/RV's will be required for the ABWR. These 18 meet the code required over-pressure protection limits. Eight of the 18 are required for automatic depressurization.
l Chairman Michelson asked what type of check valves would be used on the accumulators? Mr. James replied that they had not selected that level of detail as yet.
Dr. Okrent asked if GE had considered vessel overfill?
Chairman Michelson noted that with the vessel overfilling with the steam line, you would now have two-phase flow through the valve instead of signal phase.
He mentioned that certain valves are more susceptible to
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. Reactors, November 15-16, 1988 stem binding than others.
He noted that if you l.ad an overfill and popped open the valves and bent the stems, you could end up with several stuck open valves.
He said he would pursue this further when the subcommittee got into a discussion of vessel overfill.
Mr. Michelson suggested a discussion of vescel overfill for the next meeting.
Mr. James discussed questions from a previous meeting.
Q.
What power source is used for the depressurization system?
A.
The eight valves for the depressurization system are opened by.
nitrogen from accumulators in the drywell. Each valve has two solenoids from a different division of DC power.
Q.
The depressurization system may be damaged by contaminants in the pneumatic system.
Is this a concern?
A.
The answer is no. The pneumatic system uses instrument l-quality nitrogen which has tight filtering requirements.
I Mr. James discussed the standby gas treatment system (SG&S). He said there were two types of functions. The first function was safety related. The system must establish and maintain a viegative pressure in i
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Minutes / Advanced Boiling Water 29 ab 1,
Reactors, November 15-16, 1988
'i the secon'dary containment. relative to the environment, following a LOCA.. The reason for this is to have the standby gas treatment system process any radioactive leakage before it is released.
This system is also used under non-LOCA conditions to process any effluent from the primary or secondary containment, when required to limit the discharge of the activity to the environs'during plant opera-tion.
Mr. James pointed out some changes in the source terms used as a basis fo'r the.SGTS. The percentage of iodines released is 50 percent. Of the
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50 percent, half will plate out leaving 25 percent available for leakage into the containment. These assumptions are standard.
7 Changes are made in the assumptions on the chemical species of iodine of elemental and particulate..This is being discussed with the Staff.
q Another change is that credit is being taken for decontamination factors in the suppression pool.
He also noted that it is assumed that the SGTS filter is 99 percent efficient in terms of removing elemental, organic, and particulate iodine.
Mr. James pointed out that the if filters are loaded with fission products, they would be kept cool by operation of the fans. He noted that the charcoal beds are temperature monitored. He pointed out that f
there.is a large margin between the charcoal ignition point and the alarm setpoint.
He noted that there is a flange connection to provide I
Minutes / Advanced Boiling Water 30 l
Reactors, November 15-16, 1988 1,
l water to spray over the charcoal in case of a fire. The use of a spray connection prevents inadvertent spraying of the charcoal. The charcoal beds have capped connection that can be connected to a water supply if there is.a charcoal fire.
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Chairman Michelson asked if GE was postulating that there wouldn't be a j
fire at the time of a LOCA and and that air flow was keeping the char-coal beds cooled down enough not to create a fire?
Mr. James agreed.
Chairman Michelson asked how quickly would you reach some kind of limits if the fans were lost?
Mr. James responded that it would be of the order of tens of minutes if the radioactivity was there in the first plac'e.
He added that realis-tically, there's probably not much activity fn the c%rcoal beds.
Dr. Shewmon asked why you feel there is no hydrogen going to get into that gas treatment system.
Mr. James responded that the hydrogen will be generated by a metal-water reactor and by radiolytic decomposition of the coolant. The amount of hydrogen produced will be very low. This will be diluted by containment air and resultant hydrogen concentration will be very low, well below detonation limits.
Mr. James pointed out that the ABWR secondary containment will have a 50 l
percent per day leakage rather than 100 percent per day of previous
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'I Minutes / Advanced Boiling Water 31
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.;,y Reactors,. November. 15-16, 1988 designs. The analyses of the.SGTS shows that the pressure response of-the system after a LOCA will have a rapid drawdown of pressure before.
~ fission products can' escape to the environment.
Dr. Remick asked how frequently has the SGRTS been called into service.
o Mr. James said he couldn't answer that question.
It will~have-to be b
answered at the next meeting.
Mr. Townsend briefly noted the Japanese seismic terminology. He said
.the.OBE'in Japan'is called an S-1 seismic event and is 0.3g. The unit they use is 300 gals. The SSE is called an S-2 and that is 450 gals or about 0.45 g.
The meeting was' adjourned at 2:23 p.m.
NOTE:
A transcript of the meeting is available at the NRC Public Document, Gelman Bldg. 2120 "L" Street, NW., Washington, D.C.,
Telephone (202) 634-3383 or can be purchased from' Heritage Reporting Corporation, 11220 L Street, NW., Washington, D.C.
-20005, Telephone (202)628-4888.
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