ML19333B820
| ML19333B820 | |
| Person / Time | |
|---|---|
| Issue date: | 10/02/2019 |
| From: | Quynh Nguyen Advisory Committee on Reactor Safeguards |
| To: | |
| Nguyen, Q, ACRS | |
| References | |
| NRC-0619 | |
| Download: ML19333B820 (69) | |
Text
Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION
Title:
Advisory Committee on Reactor Safeguards Open Session Docket Number: (n/a)
Location: Rockville, Maryland Date: Wednesday, October 2, 2019 Work Order No.: NRC-0619 Pages 1-37 NEAL R. GROSS AND CO., INC.
Court Reporters and Transcribers 1323 Rhode Island Avenue, N.W.
Washington, D.C. 20005 (202) 234-4433
1 1
2 3
4 DISCLAIMER 5
6 7 UNITED STATES NUCLEAR REGULATORY COMMISSIONS 8 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 9
10 11 The contents of this transcript of the 12 proceeding of the United States Nuclear Regulatory 13 Commission Advisory Committee on Reactor Safeguards, 14 as reported herein, is a record of the discussions 15 recorded at the meeting.
16 17 This transcript has not been reviewed, 18 corrected, and edited, and it may contain 19 inaccuracies.
20 21 22 23 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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1 1 UNITED STATES OF AMERICA 2 NUCLEAR REGULATORY COMMISSION 3 + + + + +
4 667TH MEETING 5 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 6 (ACRS) 7 + + + + +
8 WEDNESDAY 9 OCTOBER 2, 2019 10 + + + + +
11 ROCKVILLE, MARYLAND 12 + + + + +
13 The Advisory Committee met at the Nuclear 14 Regulatory Commission, Two White Flint North, Room 15 T2D10, 11545 Rockville Pike, at 1:00 p.m., Peter 16 Riccardella, Chairman, presiding.
17 COMMITTEE MEMBERS:
18 PETER RICCARDELLA, Chairman 19 DENNIS BLEY, Member 20 CHARLES H. BROWN, JR., Member 21 WALTER L. KIRCHNER, Member 22 JOSE MARCH-LEUBA, Member 23 DAVID PETTI, Member 24 JOY L. REMPE, Member 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2 1 DESIGNATED FEDERAL OFFICIAL:
2 QUYNH NGUYEN 3 ZENA ABDULLAHI 4
5 ALSO PRESENT:
6 JAMES (ALAN) BEARD, GEH 7 YOUSEF FARAWILA, Framatome 8 JANE MARSHALL, NRR 9 JASON PAIGE, NRR 10 WALTER (SKIP) SCHUMITSCH, GEH 11 JAMES SHEA, NRO 12 ASHLEY SMITH, NRR*
13 DANIEL TINKLER, Framatome 14 15 *Present via telephone 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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3 1 CONTENTS 2 Page 3 Opening Remarks by the ACRS 4 4 Advanced Boiling Water Reactor (ABWR) 6 5 Design Certification Renewal 6 Framatome's Topical Report, RAMONA5 for 34 7 Anticipated Transient without SCRAM 8
9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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4 1 P R O C E E D I N G S 2 (1:03 p.m.)
3 CHAIRMAN RICCARDELLA: This meeting will 4 now come to order. This is the first day of the 667th 5 meeting of the Advisory Committee on Reactor 6 Safeguards. I'm Pete Riccardella, Chairman of the 7 ACRS. The ACRS was established by the Atomic Energy 8 Act and is governed by the Federal Advisory Committee 9 Act, FACA.
10 The ACRS section of the U.S. NRC public 11 website provides information about the history of the 12 ACRS and provides FACA-related documents, such as 13 charter, bylaws, Federal Register notices for 14 meetings, letter reports, and transcripts of all full 15 and subcommittee meetings, including all slides 16 presented at the meetings.
17 The Committee provides its advice on 18 safety matters to the Commission through its publicly 19 available letter reports. The Federal Register notice 20 announcing this meeting was published on September 18, 21 2019, and provided an agenda and instructions for 22 interested parties to provide written documents or 23 request opportunities to address the committee, as 24 required by FACA. In accordance with FACA, there is 25 a designated federal official for today's meeting.
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5 1 The DFO for this meeting is Mr. Quyhn Nguyen.
2 During today's meeting, the Committee will 3 consider the following: advanced boiling water reactor 4 design certification renewal and Framatone's topical 5 report, RAMONA5 for anticipated transient without 6 SCRAM, and, three, preparation of ACRS reports.
7 There is a phone bridge line. To preclude 8 interruption of the meeting, the phone will be placed 9 in a listen-only mode during the presentations and 10 committee discussions. We have received no written 11 comments or requests to make oral statements from 12 members of the public regarding today's session.
13 There will be an opportunity for public 14 comment, as we have set aside ten minutes in the 15 agenda for comments from members of the public 16 attending or listening to our meeting. Written 17 comments may be forwarded to Mr. Quyhn Nguyen, the 18 designated federal official.
19 A transcript of open portions of the 20 meeting is being kept, and it is requested that all 21 speakers use one of the microphones, identify 22 themselves, and speak with sufficient clarity and 23 volume so that they can readily be heard. Also, 24 please silence any phones or other devices to avoid 25 interruption of the meeting. We're somewhat NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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6 1 short-handed this week, as we have two members 2 overseas, one sick, and one not attending, but we'll 3 have adequate coverage of the topics based on the 4 members who are in attendance.
5 Also, I'd like to acknowledge the passing 6 of a former ACRS member and chairman, Dr. Mario 7 Bonaca. He served with the Committee until 2014. The 8 first topic at the meeting is ABWR, advanced boiling 9 water reactor design certification renewal.
10 This is the -- we've done many design 11 certifications, but this is the first one to come up 12 for renewal. With that, I will turn the meeting over 13 to Jason Page, who is acting branch chief of Licensing 14 Branch 3 in NRO. Jason.
15 MR. PAGE: Thank you. I'm just going to 16 introduce myself again. My name is Jason Page, acting 17 branch chief in the office of new reactors. I'll turn 18 it over to Jim Shea. He's the PM of this activity.
19 MR. SHEA: Thanks, Jason. Again, I'm Jim 20 Shea. Good afternoon, everyone. I appreciate the 21 opportunity for us to present our staff review of the 22 ABWR D.C. renewal. GEH will go first, and then the 23 staff will follow. It will be an abbreviation of what 24 we did before for the subcommittee. Thanks. I would 25 like to turn it back over the Chairman.
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7 1 MR. BEARD: Good afternoon. My name is 2 Alan Beard. I'm with GE-Hitachi Nuclear Energy. With 3 me is Walter "Skip" Schumitsch. He's the program 4 manager for this effort to renew our certification for 5 the ABWR.
6 We're just going to very quickly go over 7 what we've done as part of the renewal process. Just 8 a quick overview of what the ABWR is, for those of you 9 who are not real familiar with it, a brief discussion 10 of the timeline that we've been through through the 11 renewal effort, the scope of what we did during that 12 renewal effort, and then just a real quick talk about 13 some of the major design changes that we made as part 14 of that process.
15 ABWR was first built and operated in 16 Japan, at the Kashiwazaki-Kariwa site. Units 6 and 7 17 are both -- were in operation, are currently suspended 18 in operation because of the events of the tsunami, but 19 there are plans to bring those back online. Japan has 20 three additional ABWRs that were operating prior to 21 the tsunami. As far as we know, there are plans to 22 bring those back online. Additional two are under 23 construction in Japan, and then two under construction 24 in Taiwan, although the Taiwan construction effort is 25 currently suspended. ABWR is licensed in Japan and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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8 1 Taiwan, certified in the U.S., and underwent the 2 approval process, GDA approval process in the United 3 Kingdom, so it's had a lot of review.
4 CHAIRMAN RICCARDELLA: Did the plants at 5 KK, did they operate before the tsunami?
6 MR. BEARD: Oh, yes, they had about 25 7 years of operation between the two of them.
8 CHAIRMAN RICCARDELLA: Twenty-five years.
9 MR. BEARD: Yes. I would also note that 10 the ABWR, at least in our opinion, is the first of the 11 Generation 3 reactors that has been in operation.
12 MEMBER REMPE: Out of curiosity, because 13 I missed your subcommittee meeting in -- educate me.
14 The ones in Japan and other ones that are up and were 15 running, do they operate in the MELLLA+ region or 16 MELLLA region? Do they just use control rods for 17 power changes, or do they use flow, also?
18 CHAIRMAN RICCARDELLA: Do you know?
19 MR. SCHUMITSCH: I do not -- sorry, this 20 is Skip Schumitsch. I'm sorry; I do not know the 21 answer to that, either.
22 MEMBER REMPE: It's probably not relevant.
23 I just was curious. Do you know, Jose, from your work 24 on it?
25 MEMBER MARCH-LEUBA: From the original NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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9 1 reviews of ABWR, they do use flow for power control, 2 yes.
3 MEMBER REMPE: So they never have gone to 4 any sort of expanded operating domain, then?
5 MEMBER MARCH-LEUBA: I suspect it's the 6 same with ABWRs and BWRs -- I don't know -- that they 7 have such high power density that there's no need to 8 go to MELLLA+. They're already a DPU when they were 9 licensed.
10 MEMBER REMPE: Okay, thank you.
11 MR. BEARD: Would also like to note --
12 it's not on this slide, but the ABWR --
13 MR. SCHUMITSCH: I'm sorry; we got a text 14 from somebody that's listening. The answer is not 15 MELLLA+. Thank you, David.
16 MR. BEARD: -- to note that the ABWR was 17 the first plant that underwent the Part 52 process and 18 was issued the No. 1 certification, Appendix A of Part 19 52. That was back in May of 1997. I will note, 20 there's a picture over on the wall there. Our initial 21 meetings with the ACRS during that ABWR certification 22 were actually held down in the green building, off of 23 Norfolk Avenue, in Bethesda, when the ACRS was still 24 meeting down there. There are names on the table 25 there of people that were actually part of the review NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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10 1 initially. Then one other note, just kind of a 2 historical interest.
3 We actually submitted the ABWR under Part 4 50, originally, seeking a standard safety analysis 5 report, then Part 52 became promulgated, became a 6 regulation, and we chose to take advantage of that, so 7 we switched our application over to a Part 52 8 application.
9 Just like to note that the ABWR was a 10 collaborative effort. It was developed in between the 11 efforts of GE, Tokyo Electric Power Company, Hitachi, 12 and Toshiba. As I noted before, the first plants that 13 were built and operated, Kashiwazaki-Kariwa Units 6 14 and 7.
15 We'll note that both of those plants were 16 built on time and ahead of schedule and under budget.
17 It can be done. The Japanese have a very good way of 18 doing that. Hopefully, we can learn some lessons from 19 them. Primary drivers for the ABWR, at least from 20 GEH's perspective, were we wanted to enhance the 21 safety, and we wanted to improve the constructability 22 and maintainability, as well. Some of the major 23 design enhancements we made, we have an improved 24 containment design. We went to pretty much a right 25 cylinder design that's kind of a combination of our NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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11 1 Mark II and our Mark III pressure suppression 2 containments.
3 It also is a reinforced concrete 4 containment vessel with a steel leakage liner. It is 5 a relatively compact reactor building. The emergency 6 diesel generators are actually in the reactor 7 building, so they moved a lot of equipment into a 8 pretty small footprint.
9 We'll note that with the inclusion of the 10 reactor internal pumps and the removal of our external 11 recirculation loops, we were able to show, for all our 12 design basis accidents, that we never have core 13 uncovery.
14 So there's always water over the core, 15 very little heat up when you do go into a transient 16 situation. Although our reference design was based on 17 the Japanese design at K6 and K7, our probabilistic 18 risk assessment people led us to include some 19 additional items in the design. Here's a list of 20 several of those that we did. The reason I'm noting 21 this is we do believe that these are -- if they hadn't 22 been in the design, probably would have been added to 23 address the post-Fukushima tsunami event. I'd like to 24 point out that we were being proactive. We were 25 looking at the design to identify some safety NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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12 1 enhancements. This is a list of those.
2 The first of those is we have an 3 air-cooled combustion turbine generator that serves as 4 an alternate AC power source. For the certification, 5 the ABWR is classified as an alternate AC power plant 6 for the station blackout rule. Having said that, we 7 do have a reactor core isolation cooling pump, which 8 also provides us with an AC independent means of 9 cooling the core in the event of a station blackout.
10 We have what we call the AC independent 11 water addition system, ACIWA. Very fancy way of 12 saying we tied the fire water into several of our 13 safety-related systems so that if we don't have other 14 means of injecting water, we can use the fire water 15 storage tanks and the fire water pumps to pump water 16 into the reactor pressure vessel, into the 17 containment, and into the spent fuel pool.
18 We also had a means of passively 19 addressing cooling of core debris that would have 20 melted through the bottom of the vessel in the event 21 of a severe accident. This used thermally-actuated 22 valves, kind of like what you have in the overhead 23 here. What they did was they opened up and they 24 allowed water from the suppression pool to float over 25 to the debris that would have relocated down to the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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13 1 bottom of the drywall.
2 Then finally, containment overpressure 3 protection system, we did engineer in a leakage path 4 or a vent path. That is from the suppression pool air 5 space, so that we do credit scrubbing of any release.
6 I believe we had a decontamination factor of ten that 7 we credited for gasses and about 100 for particulate 8 matter. That just allows that excess pressure to vent 9 out to the atmosphere.
10 MEMBER REMPE: Is it a multi-unit 11 application?
12 MR. BEARD: No.
13 MEMBER REMPE: It's just a single --
14 MR. BEARD: Following the EPRI guidance, 15 it was designed as a single-unit standalone plant.
16 MEMBER REMPE: Okay.
17 MR. BEARD: So there's no sharing of 18 systems.
19 MEMBER REMPE: I was going to ask about 20 the standby gas treatment system, but I suppose, then, 21 that doesn't come up in the application.
22 MR. BEARD: Quickly, the reactor core 23 power of 3,926, that is kind of an edifice from Japan.
24 They license on electrical output. The electrical 25 output, when they backed it out, we got 3,926 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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14 1 megawatts of thermal. Eight hundred and seventy-two 2 fuel bundles, very similar to our operating fleet, 3 using the latest and greatest technology, 12 feet in 4 length, 3.7 meters in length.
5 We characterize it as a moderate power 6 density, 51 kilowatts per liter. To control that, we 7 have 205 control blades. We did introduce -- one of 8 the other major design enhancements with the ABWR is 9 what we call our fine motion control rod drive. Prior 10 to that, we actually had what we called a locking 11 piston.
12 It was purely hydraulic insertion and 13 withdrawal of the control blades, six-inch increments.
14 With the fine motion control rod drives, we went to an 15 electric motor that drives the lead screw. We now get 16 five eighths of an inch increments for each notch 17 position.
18 We also maintained the ability to 19 hydraulically SCRAM. We actually have diverse means 20 of inserting control blades. We can do it either 21 hydraulically, which is the preferred safety-related 22 means, or if that should fail for whatever means, we 23 can also drive them in electrically.
24 MEMBER BLEY: How long does that take?
25 MR. BEARD: Hydraulic --
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15 1 MEMBER BLEY: No, the --
2 MR. BEARD: To put this in perspective, 3 the hydraulic SCRAM, less than two seconds, the 4 electric drive in is a little less than two minutes 5 from full out.
6 MEMBER MARCH-LEUBA: Since we are curious, 7 on the control room, do you still use the CO248 8 display for control rod, or do you use inches?
9 MR. BEARD: I actually don't know. I have 10 not been in the control room.
11 MEMBER MARCH-LEUBA: It's likely you used 12 the CO248.
13 MR. BEARD: Yes. I'm sure we're going to 14 get an answer. This is just an overall flow chart of 15 the ABWR. I won't spend a lot of time talking about 16 it, but given the power rating, a single high-pressure 17 turbine, followed by three low-pressure turbines in 18 series, so pretty standard conventional side of the 19 plant. On the nuclear side, you see a pressure 20 suppression containment denoted there, and then I'll 21 talk about a little bit of the safety-related systems 22 on the next slide. The approach for the ABWR was we 23 had three divisions of safety related equipment.
24 They were operated by four divisions of 25 instrumentation and control logic that was making the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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16 1 decisions, but each of those three divisions has both 2 a high-pressure injection capability, as well as a 3 low-pressure injection capability.
4 One of those high-pressure injection 5 capabilities is the reactor core isolation cooling 6 turbine and pump that I spoke of, which provides us 7 with an AC independent means. The three high-pressure 8 systems are sufficient to inject enough cooling water 9 to maintain the core cool in the event of -- should we 10 have an isolation event.
11 Then the low-pressure systems inject 12 plenty of water to handle all the break scenarios. As 13 I said before, design basis accident point, we never 14 have core uncovery. In the event that we go beyond 15 design basis and we only have a single pump injecting, 16 any one of our five motor-driven pumps is sufficient 17 to keep enough water into the core to maintain 18 adequate cooling.
19 CHAIRMAN RICCARDELLA: No recirculation, 20 though, internal pumps.
21 MR. BEARD: Yes, reactor internal pumps.
22 That's these yellow cans hanging down here. There are 23 ten of them. They're about 700 kW each. Excuse me --
24 yes, 700 kW each. The next picture's just that of an 25 artist's rendering of what this plant looks like. You NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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17 1 can see that the turbine structure is orientated 2 perpendicular to the reactor building.
3 Somewhat interesting is the control 4 building is located in between our reactor building 5 and the turbine building here. The vast majority of 6 the control building is actually below grade, 7 including the main control room, itself. Timeline, I 8 don't need to read this to you. It's there for your 9 information.
10 It has been a relatively long and lengthy 11 process. We've gotten through it, and we're glad to 12 be at this final stage and hope to get good report out 13 from the Committee today. In our original submittal 14 to the NRC, as part of our renewal request, these are 15 some of the items we addressed, aircraft impact, 16 obviously a post-9/11 requirement. We did have some 17 containment re-analysis we had to do based on some 18 knowledge we gained from further projects. We did 19 some selected design updates, and we also corrected a 20 couple of errors that had been identified by GEH, 21 again, in some of the construction projects we had 22 going on. In addition to that, the NRC developed a 23 list of 28 topics that they sent to us in a letter.
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18 1 through all those. Anticipating the question, out of 2 those 28 topics, there were several that GE proposed 3 not be addressed. An example of that would be 4 upgrading of the digital instrumentation control 5 system.
6 As we mentioned last time we were here, 7 because of the speed that technology continues to 8 evolve, we felt that wasn't a worthwhile investment at 9 this time because we did not know when our next 10 potential customer would be and that we would upgrade 11 at that time as part of the license application.
12 MEMBER REMPE: I apologize if this was 13 brought up during the subcommittee meeting, but with 14 the errors identified by GEH, did you look to 15 understand why those errors occurred and provide 16 yourself some sort of assurance that there wouldn't be 17 other errors?
18 MR. BEARD: Yes. Following all the GE 19 practices, we go through a corrective action program 20 that we look at what are the root causes for these 21 things. An example of that, that I talk about here 22 later -- I'll go ahead and do it now -- is for the 23 containment overpressure protection system, we did not 24 have exact pipe routing at the time we were submitting 25 this design.
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19 1 We had some design assumptions on what the 2 pressure losses would be in that system. When we 3 finally got around to designing one of those for one 4 of our projects, we determined that we were 5 non-conservative with some of our design assumptions 6 for that, so went back and corrected it, indicated 7 what prior diameters were needed and how many elbows 8 and flow restrictions we had in there in order to 9 maintain the assumptions we had for the safety 10 analysis.
11 MEMBER REMPE: You looked for any other 12 possible situations similar to that and didn't find 13 any.
14 MR. BEARD: Correct.
15 MEMBER REMPE: Sounds good. Thanks.
16 MR. BEARD: I keep forgetting I have to 17 drive myself. Some of the significant design changes 18 that we did incorporate listed here. Post-Fukushima 19 1, we did add two safety-related wide-range spent fuel 20 level monitors. That gives a time to main 21 reflectometry concept. We did enhance our ECCS 22 suction strainers to address continuing concerns about 23 plugging the strainers in the event of a LOCA event.
24 We had a new fuel vault in the original 25 design, just a big hole in the operating floor on the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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20 1 refuel floor, where we would put -- initial concept, 2 when you brought the new fuel on site, inspected it, 3 you would put it in there and store it. Most 4 utilities, I think maybe all utilities, have stopped 5 doing that.
6 They pull the fuel out; they inspect it; 7 and they go ahead and put a channel on it and put it 8 in the spent fuel pool, so it is ready for that, which 9 eliminates a handling step and the potential to damage 10 the fuel while doing that.
11 We did address the NRC Bulletin 2012-01 12 dealing with the out-of-phase current issues that were 13 identified at -- I know it was an Exelon site in the 14 Midwest, but I can't remember the exact site. We did 15 some design changes to our electrical distribution 16 system to monitor and to detect an out-of-phase 17 condition, and then to isolate the out-of-phase 18 condition and allow the diesel generators to come on 19 and support the necessary safety functions.
20 MEMBER BLEY: That solution's in the 21 design cert.
22 MR. BEARD: That solution is in the 23 renewal design cert. I would point out that we did 24 have commitments in the initial design that one of our 25 three safety-related divisions was to actually be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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21 1 powered from a different offsite source. We had 2 requirements for two offsite sources.
3 We had a commitment that at least one of 4 the three had to be on the auxiliary site. Fukushima 5 Recommendation 4.2 mitigation strategies. We included 6 several items to do that. We enhanced our fire 7 protection system connections. We originally only had 8 one external connection. We added a second external 9 connection on a different face of the building, just 10 to address the possibility that debris might have 11 blocked access to the original thing. We did that.
12 MEMBER BLEY: A couple years ago, we had 13 a presentation from NEI and the owners' group on the 14 PWR strategies for using FLEX and other systems. Is 15 that part of the design cert, or is that going to be 16 -- the procedures for using all of that going to be 17 done later?
18 MR. BEARD: Anything that was 19 administrative or procedural in nature was deferred 20 until we had an applicant for license.
21 MEMBER BLEY: Makes sense. Okay.
22 MR. BEARD: The change to the containment 23 overprotection system I talked about already. Then we 24 included some other changes to enhance the capability 25 to implement the FLEX strategy, as outlined by NEI.
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22 1 That completes our prepared comments. I'm prepared to 2 answer any other questions.
3 MEMBER KIRCHNER: This is, perhaps, a 4 late-in-the-game detailed question. Where are your 5 FLEX connections to the fire mains and such?
6 MR. BEARD: Let me see.
7 MEMBER KIRCHNER: Are they between the 8 control room and the reactor building? It's a leading 9 question because I'm looking at your layout and I just 10 noticed -- the control room's in the middle.
11 MR. BEARD: The control room's here, yes.
12 MEMBER KIRCHNER: It's in a hardened 13 building, right?
14 MR. BEARD: It's in a reinforced concrete 15 --
16 MEMBER KIRCHNER: But you've got your 17 steam lines running right over, right?
18 MR. BEARD: Yes.
19 MEMBER KIRCHNER: And feedwater lines.
20 MR. BEARD: Mm-hmm.
21 MEMBER KIRCHNER: Are the FLEX connections 22 into between those two buildings, or are they 23 somewhere else?
24 MR. BEARD: We have two sets of FLEX 25 connections. There's one over on what I call the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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23 1 south wall.
2 MEMBER KIRCHNER: Okay, you answered my 3 question.
4 MR. BEARD: Then there's another one over 5 on this wall.
6 PARTICIPANT: Which wall?
7 MR. BEARD: This east wall. North is this 8 side of the building right here, east, south, west.
9 PARTICIPANT: Thank you.
10 MR. BEARD: Mm-hm.
11 MEMBER MARCH-LEUBA: You mentioned there 12 was a connection to add water for a very long-term --
13 MR. BEARD: Yes.
14 MEMBER MARCH-LEUBA: Where would that come 15 from?
16 MR. BEARD: We'd use the fire water 17 system. We have two 500,000-gallon tanks.
18 (Simultaneous speaking.)
19 MEMBER MARCH-LEUBA: So you just refill 20 the fire tanks outside the containment, and then use 21 that piping to come in?
22 MR. BEARD: We use the fire water system 23 to connect to the residual heat removal system, and 24 then the pipes within the residual heat removal system 25 give us the capability to flow water either to the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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24 1 containment, whether that be the dry well or the wet 2 well, into the pressure vessel, itself, or into the 3 spent fuel pool.
4 MEMBER BLEY: Did you build in any special 5 filtering for that water or just --
6 MR. BEARD: No, we figured, at that point, 7 that --
8 PARTICIPANT: Water's water.
9 MR. BEARD: Water's water. It's not 10 salted water. It's clean water, but it's not --
11 MEMBER MARCH-LEUBA: You never know what.
12 MR. BEARD: It's not demineralized or 13 anything like that. The answer back to the question 14 about rod position indication, it's 0 to 200.
15 MEMBER MARCH-LEUBA: Two hundred.
16 PARTICIPANT: Two hundred steps.
17 MEMBER MARCH-LEUBA: You have to have some 18 additional training for operators. I would have gone 19 to 100 percent, maybe.
20 MR. BEARD: Any additional questions? If 21 not, we thank you for your time and interest.
22 CHAIRMAN RICCARDELLA: Okay, with that, 23 we'll bring up the staff for their presentation.
24 MR. SHEA: Good afternoon. My name is 25 James Shea. I'm the staff project manager for the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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25 1 ABWR DC renewal. Today, the staff will present an 2 overview of the GEH design certification renewal 3 review, and we'll review the ABWR DC renewal upcoming 4 schedule activities and rulemaking.
5 As mentioned before, the ABWR is the only 6 Generation 3 nuclear plant in operation today, not in 7 the United States, at least yet, anyway. The ABWR was 8 initially certified in 10 CFR Part 52, Appendix A, on 9 May 12, 1997. The ABWR DC renewal application was 10 submitted on December 7, 2010. In a July 20, 2012 11 letter, the NRC staff identified proposed DCD design 12 changes that the staff believes should be considered 13 for renewal. GEH provided, Revision 6 of the DCD of 14 the ABWR, on February 19, 2016, in response to the 15 staff-requested design changes, and then the staff 16 completed its supplemental SER at the end of June of 17 this year. GEH submitted the ABWR DC renewal 18 application under Subpart B, standard design 19 certifications of 10 CFR Part 52.
20 Scope of the ABWR DC renewal included a 21 total of 39 design items proposed by staff or 22 submitted by GEH. We talked about the 28 -- there was 23 28 specific staff items that were requested. Out of 24 those, 22 were accepted by GEH, and six items were --
25 GEH wrote back to us that they thought they had NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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26 1 already covered. We re-reviewed those. Then in a 2 February 2018 letter, we agreed with that assessment 3 and wrote an additional assessment from the staff side 4 in that letter.
5 MEMBER MARCH-LEUBA: How many of those 6 were related to the aircraft impact analysis, just the 7 main major analysis?
8 MR. SHEA: None of those. You mean of the 9 28 design items?
10 MEMBER MARCH-LEUBA: Mm-hm.
11 MR. SHEA: Actually, aircraft impact was 12 submitted originally. That's one of the -- if you 13 look at that 11 additional design items, it was 14 actually submitted with the initial renewal.
15 MEMBER MARCH-LEUBA: Okay, so it wasn't 16 additional.
17 MR. SHEA: As part of the rule of Part 52, 18 it required AIA, if it wasn't already completed, 19 which, in this case, prior to its original 20 certification, the AIA rule had not been promulgated; 21 therefore, as part of this rule, it is required to 22 submit an AIA.
23 MEMBER MARCH-LEUBA: I understand from the 24 subcommittee that was one of the main efforts.
25 MR. SHEA: That was, yes. It was one of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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27 1 the main efforts. That's why I wanted -- my one 2 particular staff member who worked on that, that was 3 probably the most significant item that was addressed 4 as part of the renewal.
5 Some of the other key significant renewal 6 design changes included ECCS suction strainers, 7 Fukushima design enhancements that GEH talked about.
8 That included -- they talked about the AC independent 9 water addition, and also connections to enable offsite 10 sources to come in and connect to a fire truck or 11 other water sources, in order to -- for a COL 12 applicant, essentially, to meet the mitty bitty rule 13 (phonetic), included AC connections offsite -- again, 14 offsite, non-safety-related electrical generator could 15 be brought onsite and connected to safety-related 1E 16 electrical components, in order to mitigate a beyond 17 design basis event, and EP enhancements mostly related 18 to staffing, and also fuel pool instrumentation.
19 We just talked about two sets of fuel pool 20 safety-related instrumentation, redundant. I think we 21 mentioned it at the subcommittee meeting. They're 22 backed by the AC and DC batteries that are backup, 23 those two safety-related instruments. The ABWR AIA, 24 mentioned that. That was submitted with the original 25 renewal.
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28 1 The PCT modification, that came up very 2 late in the game, actually, probably toward the end of 3 our review, when the staff noticed that PCT should 4 have been updated for the renewal. Then GEH went back 5 and did some changes based on 50.46 reporting 6 requirements and made some adjustments to their 7 evaluation model, and then resubmitted. Finally, a 8 containment overpressure protection system, which they 9 talked about, also, as being -- that came in, again, 10 from GEH, originally, because they found an error in 11 their analysis. Just as an example, we used the AIA 12 as an example of the 39 items that we addressed, in 13 this particular case, the SER Supplement, Chapter 19, 14 Section 19.5, Aircraft Impact Assessment.
15 GEH submitted its assessment, again, 16 initially, with the renewal. Changes included 17 enhanced fire protection design features and ITAAC 18 that ensures penetrations are not installed on the 19 control building roof without an AIA cognizant 20 engineer review.
21 In short, GE didn't change any parameters, 22 as far as the design, the walls, the locations, the 23 buildings, essentially took the NEI template for doing 24 AIA and applied it and verified that the aircraft 25 impact would not adversely impact the plant or meet NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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29 1 the parameters for ensuring safety of the plant.
2 Next, what we have is scheduled -- after 3 this meeting, we do have our rulemaking that we'll --
4 we already had some preliminary rulemaking meetings.
5 We'll kick that off, essentially, after this Phase 3 6 of the ACRS review. That schedule would include about 7 a year to 14 months for rulemaking. That will go in 8 parallel with the final SER that the staff will work 9 on and have published for the rule effort, which would 10 be Phase 4, FSER, with no open items. With that, I 11 just want to summarize everything. The staff 12 evaluated the GEH-proposed design updates to the ABWR 13 and validated the findings in NUREG-1503 and 14 NUREG-1503, Supplement 1.
15 This ABWR DC renewal safety evaluation 16 report, Supplement 2 to the NUREG-1503, documents the 17 NRC staff's review of GEH's application to renew the 18 ABWR DC. Except as modified the supplement, the 19 findings made in NUREG-1503 and its Supplement 1 20 remain in full effect.
21 The NRC staff made its safety 22 determinations on specific modifications and 23 amendments proposed by GEH as part of the DC renewal 24 application. These modifications and amendments were 25 found to meet the applicable regulatory requirements NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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30 1 and are therefore accepted. That ends the staff 2 presentation, unless there's any questions.
3 CHAIRMAN RICCARDELLA: Thank you. Are 4 there any --
5 MEMBER KIRCHNER: I'm sorry, go ahead.
6 CHAIRMAN RICCARDELLA: Are there any 7 further questions from the Committee?
8 MEMBER KIRCHNER: Yes, Jim, just for the 9 record, what is the expected time for the staff to 10 complete a rulemaking on something like a DC renewal?
11 MR. SHEA: I think the goal that we have 12 -- currently, we are planning to have a direct final 13 rule, so it won't go out for the normal rulemaking 14 process. That should shorten the process. What we 15 have in planning phases, right now, is expected about 16 12 to 14 months.
17 MEMBER KIRCHNER: Why does it take so 18 long? It's not a safety question; it's a process 19 question.
20 MR. SHEA: Part of the fact that we still 21 have Phase 4, the FSER. Step 1 is GEH will submit us 22 their Revision 7 to the DCD, which they haven't done 23 yet. They're still going through their final 24 validation process on that. Once they submit that to 25 us, we use that as the basis to go back and look at NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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31 1 all the SERs that we completed that were based on 2 markups of Rev. 6.
3 For example, AIA, we had some RAIs 4 following the Revision 6 to the DCD that were then 5 addressed and included markups, which will now be 6 incorporated into Revision 7. At that point, the 7 staff -- we will verify all those changes were made in 8 Revision 7. That's going to take some time, probably 9 until the end of the year. Right now, the schedule 10 has us completing the FSER in -- I think we have it in 11 March 2020.
12 MEMBER KIRCHNER: Yes, that's what you 13 have on Slide 9.
14 MR. SHEA: That all will depend on when we 15 get the DCD. It should only take us -- in this 16 particular case, it should only take us a few months 17 to complete the FSER process. You need the FSER and 18 the final DCD to actually submit it as a rule. We've 19 got to have those two main administrative pieces done 20 before we can actually go to final rule.
21 MEMBER KIRCHNER: I should have framed my 22 question differently. When you get done with the 23 final SER with no open items in March of next year, 24 which is your target, how much time, then, is 25 allocated for the rulemaking, itself?
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32 1 MR. SHEA: Like I said, that can be done 2 -- we should be kicking the rule off within a month or 3 so. That would take us, from that point, 12 to 14 4 months. That includes, in parallel, us completing the 5 FSER. That's part of that process. It wouldn't be in 6 addition.
7 MEMBER KIRCHNER: Thank you.
8 MR. SHEA: Like I said, if we can get the 9 DCD back to us even before -- I think we're scheduled 10 to get it before the end of this year, the final DCD, 11 we can then validate. Then this process could take 12 shorter than our goal. That's what we would strive to 13 do.
14 CHAIRMAN RICCARDELLA: Any other questions 15 from members? Let's check for members of the public.
16 Is there anybody in the room from the public who would 17 like to make a comment on this?
18 Not seeing any, we will open the phone 19 line and accept questions from anybody who happens to 20 be on the line. Are there any members of the public 21 out there who would like to make a comment on the ABWR 22 design certification renewal? If so, please state 23 your name and make your comments.
24 Not hearing any, we'll proceed. I think 25 -- we're finished with the presentations. We do have NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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33 1 a draft letter, which --
2 MR. NGUYEN: Chairman, I guess I recommend 3 a break, so I can make copies while we have the 4 licensee here.
5 CHAIRMAN RICCARDELLA: Okay, we'll take a 6 break until 2:00 p.m., 15 minutes, and then we will --
7 okay, we're off the record until the next topic, which 8 is at 2:30.
9 (Whereupon, the above-entitled matter went 10 off the record at 1:43 p.m. and resumed at 2:30 p.m.)
11 CHAIRMAN RICCARDELLA: Next topic on the 12 agenda is Framatome's topical report on RAMONA5 for 13 anticipated transient without SCRAM. A portion of 14 this meeting will be open, and then we will close the 15 meeting for a closed session. With that, I'll ask, 16 Jane, do you want to make a comment?
17 MS. MARSHALL: Yes, I'll make a couple of 18 quick opening remarks. I know the ACRS subcommittee 19 had an opportunity to review this a month or so ago.
20 Framatome has generalized this methodology, so it can 21 now be used at -- they generalized it to a form that 22 can be used at any BWR currently operating in the U.S.
23 The methodology is directly applicable to Brunswick 24 ATRIUM-11.
25 That's scheduled to be presented to the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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34 1 ACRS subcommittee in November, since Brunswick 2 proposed to use an identical analysis. The NRC staff, 3 our contractor from Oak Ridge National Lab, and 4 Framatome staff have demonstrated a -- have been very 5 responsive to each other's needs. We've had an 6 efficient and, we think, satisfactory completion of 7 this complex review, without any challenges or delays.
8 Thanks.
9 MEMBER MARCH-LEUBA: Framatome, you want 10 to do some introductory remarks on the open session?
11 Don't use any proprietary slides until we close.
12 MR. TINKLER: I guess we'll start with 13 some introductions, here. My name is Dan Tinkler.
14 I've been with Framatome, now, for 17 years. Fourteen 15 of that has been working various stability methodology 16 development projects, starting with our long-term 17 stability solution methodology, RAMONA5 based, going 18 up through various plant specifics, and now here, 19 today, moving into the generic ATWS-I.
20 With me is Dr. Farawila. He has many 21 decades of stability experience working various items, 22 such as testing, methods development, pretty much 23 spanning the gamut when it comes to the stability 24 field.
25 MEMBER MARCH-LEUBA: Do you want to say NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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35 1 something on the topic of representation in open 2 session?
3 MR. TINKLER: The generic ATWS-I 4 methodology we're really going to present today is 5 kind of -- it's not a new methodology. It's kind of 6 the culmination of some plant-specific work and 7 previous work we've done on the long-term stability 8 solution. That's really kind of what we're going to 9 present to you today.
10 MEMBER MARCH-LEUBA: Thank you. Mr.
11 Chairman, at this point, I propose that we close the 12 open session.
13 CHAIRMAN RICCARDELLA: Okay, the open 14 session is closed. We'll now go into closed session.
15 MEMBER KIRCHNER: Jose, are we going to go 16 closed, and then open again?
17 MEMBER MARCH-LEUBA: No, this is --
18 MEMBER KIRCHNER: Then something's 19 missing, I would suggest. The public needs to know 20 what's the answer. What I mean by that is has the 21 staff review approved the topical? That should be 22 part of the record.
23 MEMBER MARCH-LEUBA: It should. Let me 24 give a summary. The staff has issued an SER.
25 MEMBER REMPE: Just slow down. First of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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36 1 all, are we still in the open session, or did we close 2 it? I thought I heard the gavel bang. Are we in the 3 open session -- the transcriber's going to have to 4 answer this -- or are we in the closed session?
5 COURT REPORTER: We're still in open.
6 MEMBER REMPE: In addition to this, we 7 need to allow for public comments. That's good.
8 Sorry to interrupt.
9 MEMBER KIRCHNER: The public can't really 10 comment until they know what the answer is. We need 11 a summary of what was presented and what was approved.
12 MEMBER MARCH-LEUBA: The staff has 13 reviewed the proposed methodology to calculate ATWS-I 14 transients in BWRs, on a generic basis, and found it 15 acceptable. That's what their SER says.
16 CHAIRMAN RICCARDELLA: With that, before 17 we go into the closed session, we'll ask are there any 18 members of the public in the room that would like to 19 make a comment?
20 (No response.)
21 CHAIRMAN RICCARDELLA: Are there any 22 members of the public on the phone line? Which I 23 guess is open now, because Ashley's on the line.
24 Ashley, are you there?
25 MS. SMITH: Yes, I'm still here.
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37 1 CHAIRMAN RICCARDELLA: Is there anybody 2 else on the line, besides Ashley?
3 Hearing none, I assume there's nobody from 4 the public who would like to make a comment. Is it 5 okay for Ashley to be on the open line? What if 6 someone else --
7 MEMBER MARCH-LEUBA: This is the closed 8 line, right?
9 MS. MARSHALL: She'll use the closed line.
10 She has that number now, but she's not on it yet.
11 MS. SMITH: Yes, thank you. I'll call 12 back in.
13 CHAIRMAN RICCARDELLA: All right, very 14 good, thank you. With that, we can close the open 15 phone line, the public line.
16 (Whereupon, the above-entitled matter went 17 off the record at 2:37 p.m.)
18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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Presentation to the ACRS Full Committee Staff Safety Review of ABWR DC Renewal October 2, 2019
Agenda Overview of the General Electric Hitachi (GEH) Advanced Boiling Water Reactor (ABWR) Design, Certification and Renewal.
ABWR Design Certification (DC) Renewal Application Regulatory Basis for DC Renewal Design Change Items Proposed and Reviewed Key Significant Design Changes Staff Conclusions Schedule for the ABWR DC Renewal Activities 2
Overview of the ABWR Design Generation III Reactor with enhanced safety features ABWR is a single-cycle, forced-circulation, boiling-water reactor (BWR),
with a rated power of 3926 MWt Reactor recirculation system applying internal pumps Advanced Fine Motion Control Rod Drive (CRD) System Main Control Room (MCR) with full digital system Reinforced concrete containment vessel Source- GEH ABWR 3
ABWR DC Renewal Application Summary May 1997: Staff FSER NUREG-1503 Supplement 1 based on ABWR design control document (DCD) Revision 4.
May 12, 1997: Initial ABWR DC Rule (Appendix A to Title 10, Part 52)
December 7, 2010: GEH ABWR DC Renewal Application DCD Revision 5 July 20, 2012: NRC staff Identified proposed changes including Fukushima Near Term Task Force Recommendations (NTTF) from SECY-12-0025 February 19, 2016: GEH provided ABWR DCD Revision 6 in response to staff requested changes with GEH responses to those requests June 28, 2019: NRC staff completed Advanced Supplemental SER with no open items 4
DC Renewal Regulatory Basis Regulatory Requirements for DC Renewal Applications 10 CFR 52.57, Application for renewal 10 CFR 52.59, Criteria for renewal GEH submitted the ABWR DC renewal application under Subpart B, "Standard Design Certifications," of 10 CFR Part 52 Application included the ABWR DCD and an environmental report (ER).
5
ABWR DC Renewal Design Items 28 Design Items Proposed by the staff for Consideration:
GEH accepted the changes proposed by the staff for 22 items and included the changes in the February 2016 DCD Revision 6.
6 items not incorporated in revised ABWR DCD.
11 additional design items identified at time of Renewal or during the review of the application.
39 Total Design Items Reviewed and Approved in Supplemental SERs to NUREG-1503 or closed by letter.
6
ABWR DC RENEWAL Key Significant Renewal Design Changes
- ECCS Suction Strainers
- Fukushima Design Enhancements
- ABWR Aircraft Impact Assessment
- PCT Modification
ABWR DC Renewal Issue 29 -AIA Aircraft Impact Assessment:
Design Change Type - Modification SER Supplement Chapter 19 Section 19.5 Aircraft Impact Assessment:
Submitted as part of the DC Renewal (DCD Revision 5) - ABWR DCD Tier 2, Section 19G,Revision 6, GEH Aircraft Impact Assessment," and proposed changes to Revision 6 of the ABWR DCD.
Enhanced Fire Protection Design Features.
Control Building (C/B) penetrations are not installed on the C/B roof without an AIA cognizant engineer review.
The NRC staff also finds that the applicant adequately described the key design features and functional capabilities identified and credited to meet 10 CFR 50.150(b), including how the key design features meet the acceptance criteria in 10 CFR 50.150(a)(1).
8
ABWR DC Renewal Schedule - Letter Dated 5/31/19 Completion Date Key Milestones Actual - A Target - T Application Received Design Certification Renewal Application 12/07/10 - A Acceptance Review NRC to issue Acceptance Review Determination Letter 02/14/11 - A Safety Review Phase 1 - Preliminary Supplemental Safety Evaluation Report (SER) and Requests for Additional 01/21/19 - A Information Phase 2 - Advanced Supplemental SER with No Open Items 06/28/19 - A Phase 3 - ACRS Review of SER with No Open Items 10/19 - T Phase 4 - Final SER with No Open Items 03/20 - T Rulemaking Issue final rule TBD 9
ABWR DC Renewal ABWR DC Renewal NRC Staff Conclusions The NRC staff evaluated the GEH proposed design updates to the ABWR and validated the findings in NUREG-1503 and NUREG-1503 supplement 1.
This ABWR DC Renewal Safety Evaluation report, Supplement 2 to NUREG-1503, documents the NRC staff's review of GEHs application to renew the ABWR DC. Except as modified by this Supplement, the findings made in NUREG-1503 and its Supplement 1 remain in full effect.
The NRC staff made safety determinations on the specific Modifications and Amendments proposed by GEH as part of its DC Renewal Application.
These Modifications and Amendments were found to meet the applicable regulatory requirements and are therefore acceptable.
Thank You!
10
Backup Slides 11
ABWR DC Renewal List of Abbreviations Used ABWR - Advanced Boiling Water Reactor IEEE - Institute of Electrical and Electronics Engineers ac - Alternating Current ITAAC - Inspections, Tests, Analyses, and Acceptance ACS - Atmospheric Control System Criteria ACRS - Advisory Committee on Reactor Safeguards MBDBE- Mitigation of Beyond Design Basis Events ACIWA - Alternating Current (ac) Independent Water MCR - Main Control Room Addition System NPSH - Net Positive Suction Head AIA - Aircraft Impact Assessment NTTF - Fukushima Near Term Task Force ATWS - Anticipated Transient Without Scram Recommendations BWR - Boiling Water Reactor NRC - US Nuclear Regulatory Commission C/B - Control Building RAI - Request for Additional Information COL - Combined License RB - Reactor Building COPS- Containment Overpressure Protection System RG - Regulatory Guide CRD- Control Rod Drive RHR - Residual Heat Removal System DBA - Design Basis Accident RSP - Remote Shutdown Panel DC - Design Certification SER - Safety Evaluation Report DCD - Design Control Document SFP - Spent Fuel Pool ECCS - Emergency Core Cooling Systems SR - Safety Related EP - Emergency Planning SRP - Standard Review Plan ER - Environmental Report SSC - Structure, Systems, and Components GEH- General Electric Hitachi TS - Technical Specifications I&C - Instrument and Control TSC - Technical Support Center 12
Item No. Description Type SER Supplement Chapter 2.0 Section 2.5 1 Geological, Seismological and Geotechnical Modification Engineering SER Supplement Chapter 2.3 Section 2.3.1, 2 Regional climatology Modification SER Supplement Chapter 3 Section 3.3, Wind 2 and Tornado Loadings Modification SER Supplement Chapter 3 Section 3.5.1.4.1 2 Missiles Generated by Natural Phenomena Modification SER Supplement Chapter 2.0 Section 2.6.8 3 ABWR Site Acceptability Modification SER Supplement Chapter 2.0 Section 2.6.2 4 Water Level (Flood) Design Site Parameters Modification SER Supplement Chapter 12 Section 12.3 5 Radiation Protection Design Features Amendment SER Supplement Chapter 12 Section 12.2 6 Radiation Sources (SER covers Issues 6&7) Modification SER Supplement Chapter 12 Section 12.2 7 Radiation Sources (SER covers Issues 6&7) Modification SER Supplement Chapter 11 Section 11.4 Solid 8 Waste Management System Modification SER Supplement Chapter 6 Section 6.2.1.9 9 Containment Debris Protection for ECCS Amendment Strainers SER Supplement Chapter 5.0 Section 5.4.8 10 Reactor Water Cleanup System. Amendment SER Supplement Chapter 9 Section 9.5.1 Fire 11 Protection System Modification 13
Item No. Description Type SER Supplement Chapter 5.0 Section 5.2.5 12 Reactor Coolant Pressure Boundary Leakage Amendment Detection.
SER Supplement Chapter 9.0 Section 9.1.1 New 13 Fuel Storage Amendment SER Supplement Chapter 9.0 Section 9.1.4 Light 13 Load Handling System (Related to Refueling) Amendment SER Supplement Chapter 9.0 Section 9.1.5 13 Overhead Heavy Load Handling Systems Amendment Update the Level 1 and 2 full-power probabilistic risk assessment (PRA) for the ABWR, including 14 its description and results in Chapter 19 of the Issue Closed DCD.
Complete a Level 1 and 2 shutdown PRA for the 15 ABWR, including its description and results in Issue Closed Chapter 19 of the DCD.
Update Appendix 19K to develop a 16 comprehensive list of risk-significant SSCs. Issue Closed SER Supplement Chapter 13 Section 13.5 Plant 17 Procedures Amendment SER Supplement Chapter 4 Section 4.2 Fuel 18a System Design Modification SER Supplement Chapter 9 Section 9.1.2.1 Fuel 18b Racks Modification SER Supplement Chapter 9 Section 9.1.2 New 19 and Spent Fuel Storage Modification (SER covers Issues 19&20)
SER Supplement Chapter 9 Section 9.1.2 New 20 and Spent Fuel Storage Modification (SER covers Issues 19&20) 14
Item No. Description Type Replace obsolete (I&C) and data communication technology. The replacement design should 21 conform to current instrumentation and control Issue Closed related regulations, industry standards, and regulatory guidance.
SER Supplement Chapter 7.0 Section 7.7.1.2.1 22 Control Rod Ganged Withdrawal Sequence Modification Restrictions SER Supplement Chapter 3.0 Section 3.7.3, 23 Seismic Subsystem Analysis Modification Apply the guidance from Regulatory Issue 24 Summary 2008-05, Revision 1, to the existing Issue Closed ITAAC and submit revised ITAAC.
Provide a control room design that reflects state-25 of-the-art human factor principles in accordance Issue Closed with 10 CFR 50.34(f)(2)(iii).
SER Supplement Chapter 22 Sections 5.4.7 26 RHR, 5.4.7.1.1.10 ACIWA, 7.4.1.4.4 RSP, 8.3.4.4 Amendment 1E Buses Chapter 16 TS SER Supplement Chapter 22 Sections 3.2.3 27 Safety Classifications, 7.5.2.1 Post Accident Amendment Monitoring System, 9.1.3 Fuel Pool Cooling SER Supplement Chapter 13 Section 13.3 28 Emergency Planning (SER Covers Issue 28&31) Modification SER Supplement Chapter 19 Section 19.5 29 Aircraft Impact Assessment Modification SER Supplement Chapter 6 Section 6.2.1.3 30 Short-Term Pressure Response Amendment 15
Item No. Description Type SER Supplement Chapter 13 Section 13.3 31 Emergency Planning (SER Covers Issue 28 & 31) Modification SER Supplement Chapter 19 Section 19.2.3.3.4 32 ABWR Containment Vent Design Modification SER Supplement Chapter 8 Section 8.2.5 NRC 33 Bulletin 2012-01 Design Vulnerability Modification SER Supplement Chapter 6 Section 6.2.1.6 34 Suppression Pool Dynamic Loads Modification SER Supplement Chapter 14 Section 14.3.2.3.6 35 Structural Task Group Review Modification SER Supplement Chapter 1 Operating 36 Experience Review (Chapter 1 SER Covers N/A Issues 36 &37)
SER Supplement Chapter 1 Alternate 37 Vendor/Changes to Chapter 1 SE (Chapter 1 N/A SER Covers Issues 36 &37)
SER Supplement Chapter 6 Section 6.3 38 Emergency Core Cooling Systems Modification Supplement Chapter 19 PRA to discuss effect of 39 design changes on PRA. N/A 16
GE Hitachi Nuclear Energy ABWR Design Certification Renewal ACRS Committee meeting 2 October 2019
GEH Presentation
- ABWR Overview
- U.S. Design Certification Renewal Timeline
- Renewal Scope
- Significant Design Changes 2
ABWR Overview
- Three additional ABWRs operational in Japan
- Two more under construction in Japan, and two in Taiwan.
- The ABWR is licensed in Japan and Taiwan, certified in the U.S.,
and approved in the UK (GDA) 3
ABWR Overview (cont.)
The ABWR was developed as a collaborative effort between GE, TEPCO, Hitachi and Toshiba
- First Plants were built at the K/K site as units 6 and 7 Primary Drivers were enhanced safety and improved constructability and maintainability
- Improved Primary Containment design
- Combines features of the Mark II and III containments
- Reinforced Concrete Containment Vessel (RCCV) with steel leakage liner
- Compact Reactor Building of primarily reinforced concrete
- No Core Uncovery during a Design Basis Accident (DBA)
- Reactor Internal Pumps (RIPs) 4
ABWR Overview (cont.)
The U.S. NRC certified design incorporated additional features:
- Combustion Turbine Generator as an Alternate AC power source (aircooled)
- AC Independent Water Addition (ACIWA) System using Fire Protection as diverse water source
- Lower Drywell Flooder utilizing passive thermally activated valves to flood the Lower Drywell in the event of an exvessel core melt
- Containment Overpressure Protection System (COPS)
Passive rupture disc venting from Suppression Pool Airspace 5
ABWR Overview (cont.)
Reactor Specification 3926 Rated MWt 872 Fuel Bundles N Lattice (symmetric water gap)
Active Fuel Length (3.66 m; 12 ft)
Moderate Power Density (51 kw/liter) 205 Control Blades Fine Motion Control Rod Drives (FMCRDs)
- Reduced Fuel Duty
- Fast Hydraulic Scram 6
ABWR Overview (cont.)
Overall Flow Chart 7
ABWR Overview (cont.)
Emergency Core Cooling Systems HPCF LPCF/RHR ADS HPCF RCIC LPCF/RHR LPCF/RHR 8
ABWR Overview (cont.)
9
U.S. Design Certification Renewal Timeline
- Application Docketed by NRC Feb 2011
- Initial Application Review Meeting Mar 2011
- NRC Letter - Proposed Changes Jul 2012 (28 items)
- GEH response to NRC Letter Sep 2012
Renewal Scope Original Submittal Aircraft Impact Assessment Containment Reanalysis Selected design updates Corrected errors identified by GEH NRC identified NRC originally identified 28 topics Final list was 39 items 11
Significant Design Changes
- ABWR added two safetyrelated wide range spent fuel pool level
- Enhanced ECCS Suction Strainers
- Deletion of new fuel vault
- Addressed NRC bulletin 201201
- Included changes needed to enhance FLEX 12
Questions?