ML20043D935
| ML20043D935 | |
| Person / Time | |
|---|---|
| Issue date: | 03/06/1990 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2674, NUDOCS 9006110290 | |
| Download: ML20043D935 (13) | |
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SUMMARY
/ MINUTES Meeting 1of the Advanced Boiling Water Reactor' Subcommittee,JOctober 31, 1989 7920 Norfolk Avenue, Bethesda, Maryland The ACRS Advanced' Boiling Water Reactor Subcommittee metTat 3
i 8:30 am, October 31, 1989, Room P. 110, 7920 Norfolk Aventra,- Bethesda, Maryland. The ACRS meubers in attendance were C. Michelson, SubcommittSe: Chairman, D. Ward, I. Catton and D. Okrent, Consultant. H. Alderman was the Cognizant ACRS staff member. q] The meeting commenced at 8:30 am. f The purpose of the meeting was.to review the draft _ Staff' i Safety Evaluation report of module one of the General Electric company Ster.da; <'aty Analysis Report (SER). L Mr. Michelse noted in his introductory remarks that-he~was: concerned _about the status of the SER, given its large number of L open items. He pointed out that'the subcommittee should g've-careful thought on how to proceed in' bringing this case to the-full committee at this time, l Mr. Charles Miller, NRh n te the opening statement. He - noted that the staff's intent was to'be ac specific as possible concerning the rtatus of the review and to. 'swer any questions concerning-the chapters or subchaptors of the SER. Mr.~D. Scaletti, NRR, discussed the open-issues in the SER i l Lfor the ABWR. He noted that 14 of these issues are related to ~ I i [pjb ( DESIGNATED ORIGINA1, 9006110290 900306 m hs; ACRS Coctified _By[g)[ b - l ..~
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Minutes / Advanced Boiling 2 Water-Reactor, Oct.31,'89 --information to be presented to the committee in modules 2, 3 and it'of-the Safety Evaluation Reports. Most~of the'open issues: deal i with.information that-that'the. staff has under review. In some ~ cases the staff hasn't decided if they need more information. In some' cases the staff isLwaiting-for! policy guidance from;the Commission. Mr. Scaletti noted that future modules of the Safety Evaluation Report will include: p Module 2 site characteristics and. design criteria for o structures,-systems and components.- i J i Module 4 - Chapter 15-analysis, technical specifications,- ~ o control room design. review and severe: accident design considerations and the Probablistice Risk' Assessment (PRA).- l Mr. Scaletti discussed some of the outstanding'open issues: 1 Containment Overpressure System - GE has proposed a pansive o device with a blowout diaphragm at the end of the line which would rupture just prior to reaching the ultimate. strength of the containment. This issue is still uncer. discussion. .[ Containment leak-testing - GE will provide additional n information in Amendhant 9. The staff will complete its review when it receives the additional information.,
O I [ *;. Advanced Boiling Water-3 1 ) L Reactor Subcom.Mtg. Oct. 31,'89' j R 1 l Control' Room habitability - the information has been E o provided to the staff and is under review.- -) o Atmospheric clean-up system - information to be provided in { a lator SSAR supp,lement. o Main Steam isolation valve leakage ~ system -:this11ssue is under review by the staff. ;The etaff may need' additional' information to compete its review. - Dr. Okrent asked if-any research was;needed regarding thei flooding of the molten core? Mr. Scaletti replied that he thought there was; ongoing o research. He noted that there was a difference.of opinion among. q the staff as to whether or not steam explosions should be ] considered, l l Mr. Michelson noted in many cases'the. kind of information that is provided in the SSAR is not suitable for obtaining specific equipment or construction. He said that the licensing: basis agreement states, The degree d design detail necessary-providing casentially complete design f.s to be that detail that is suitable for obtaining specific equipment, and for construction bids, and'to demonstrate conformance to the safety, the design safety limits and criteria." V Mr. 7oe Quirk, GE noted that there.are-two sources that G" ww-WD' WN w e-2--. --4e m
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. '..p I a -l Advanced Boiling Water 4 Reactor Subcom.Mtg. Oct. 31,'89 describe the information. One source isLin the Standard safety j! Analysis Report, the other_ source is in the bacxup' documentation l which is available:at G.E.- Mr. Michelson asked that if.the design.that would be cresented for' certification had' differences from the detailed j i Japanese design. would.GE provide a conceptual design' showing the-differences in the United States design? a i Mr.:Quith said that GE would show the differences. i Charles Dillman, Mana.ger of Mechanical Equipment design for the ABWR program, GE. discussed Chapter 4 of'the GE SSAR.- He pointed out--that the Fine-Motion Control Rod Drive (FMCRD) has diverse means for-insertion. -It'has; hydraulic scram' and electric run-in. - The FMCRD' eliminates-the scram discharge . volume. The control rod is positively coupled to the drive with a bayonet coupling. i Mr. Dillman pointed out that the control rod drive housing i ~ cannot be ejected.from the core because the guide-tube has a flange-on the end that is larger than the mounting holesin-the core. Dr. Okrent asked what was the maximum reactivity reinsertion rate during startup? Mr. Dillman replied that the didn't know but would.obtain i that information. Mr. John Tsao, Materials and Chemical-Engineering Branch, NRR. f Mr. Tsao.noted that the staff's review of control rod I
bi 1 t g -j 3 Advanced' Boiling Water-5 Reactor Subcom.Mtg. Oct. 31,'89' 'l structural materials is complete and there are_no open agens. He stated the staff had used the Standar6 Review Plan-(SRP) 4.5.1 as, their guideline to_ review the GE submittals.. -He_also noted the-staff-had' relied on ASME Code Section III,-appendix 1, for l material properties and allowable stresses,;ASME Section'II'for-a natorial. specifications and Reg.-Guide:1485. _The welding-technologies were reviewed using Reg. Guide. . 31.- Mr.'Tsao noted the core support structures were_roviewed'uming SRP'4.5.2. Mr. George Thomas, Reactor ~ Systems Branch, NRR. .[ Mr. Thomas noted that the control rod drive' system review is_ j essentially complete. One open issue'is that the electric scram a is not safety grade. There are two control rod. drives:per accumulator. Two adjacent rods would be connected to neparate, accumulators, so.a single-failure would.,cc have an effect on the shutdown capability. Dr. Okrent asked it tnere were any seismic. considerations in-the staff's evaluation process? j Mr. Thomas replied that seismic consideration were not { considered. Dr. Okrent suggested that the staff consider. seismic considerations. i Mr. Scaletti noted that.the staff Jill consider the design bases and accident scenarios in Chapter 19. This,is currently k under review, i Mr. Michelson asked if the staff had done a failure mode i analysis of the Control Rod Drive System. He. asked if the staff o e-- -e. 4
l . 4 Advanced Boiling. Water 6 Reactor Subcom.Mtg.;Oct. 31,'89 has looked at the full' spectrum of air pressure?. Jur. Thoma.s said the staff;would lookLat degraded air-pressure. L Mr.- Anthony J. Jr.mes, General Electric, Newport News Division, presented a brief overview of Chapter-5. Mr. James noted that there were 10; reactor internal pumps with adjustable epeed motors. .Mr.EJames discussed the. provisions that prevent failure:and: blowout of-the-reactor internal pumps. 7 1 He noted the core penetration for reactor internal pumps was about one half square foot.. The. stretch tube supports the i assembly, the motor is retained by straps.- J L Dr-Cotton asked what could-be'-the consequences of a ~ l disintegration of one of the reactor internal pumps. Mr. James said you would have loose. parts in the reactor but it wouldn't jeopardize the reactor pressure boundary.. He noted' that there had been a thorough evaluation lof possible: failure-1 modes. Mr. James discussed the reactor core--isolation cooling p system. -He noted that this is a pump system that.if main feedwater flow is unavailable, this' system will provide reactor '1 coolant. It is.a small turbine driven system.. The turbine uses \\ 1 l' steam from the main steam lines inside-containment and exhausts 1: L the steam to the suppression pool. The turbine runs a pump which l can connect to either the condensate storage tank or-the suppression pool. The condensate storage tank is the1 preferred 1 C
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_m y 4 o i - Advanced: Boiling - Water . 31,'89 7 ~ Reactor Subcom.Mtg. Oct. alignment.' l l Mr. James discussed the Residual' Heat Removal System (RHR). There are three completely independent loops. They are cooled by the intermediate cooling loop, i Mr. James discussed the integrity of the' reactor coolant-pressure boundary. He noted they;are meeting the key code requirements. It is~a section III vessel. They have overpressure protection that conforms to all of the requirements. Mr. James discusned the reactor vessel internals.- The' -l - recirculation flow piping below the core has been eliminated by i using internal pumps. Mr. James noted that by using a larger diameter vessel, the gap between the shroud and the vessel is larger. This water filled gap results in lower vessel fluence levels than current BWR's. Mr. Michelson noted that the GE - SAR stated that the vessel will not be designed to be' annealed. He asked if that'was.still GE's position? Mr. James replied that it was. He noted that it would be i possible to be annealed if you had enough money-but it would be 1 impractical. i 1 Mr. Chandra, Plant Systems Branch discussed the Reactor Coolant Pressure Boundary Leakage Detection System. He noted this system complies with SRP 5.2.5, GDC 2 and:10, R.G. 1.29, GDC 30 and Reg. Guide 1.45. Mr. John Tsao discussed reactor vessel integrity. Mr. Tsao Y
i Advanced Boiling Water-8' 4 l Rt. actor Subcom.Mtg. Oct. 31,'89 discussed reactor vessel annealing. 'He said that if the ' l ductility' transition temperature exceeds 200 Fahrenheit:and the upper shelf enerry is:below 50 foot pound, then the reactor vessel will have to be annealed. Mr. Tsao said-that based upon.the staff's analysis, tho . i staff did not believe the GE ABWR vessel will have to~be
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t {_ Several of the committee members discussed intersystem LOCA's-with regard to.the GE ABWR. ] GE and the staff noted there is a possibility of~small-l leakage but the probability of pipe rupture is low. Mr. Scaletti i noted that the staff thought the design criteria-thatfGE used was- { satisfactory to solvr the problem.- Mr. Michelson auked if the interfacing' system LOCA;was an-open or closed issue? Mr. Scaletti'said that as far as.the staff r was concerned, the issue was closed. L 1 Mr. James discussed Chapter 6 Engineered. Safeguards. He note the criteria for containment *was that it hadLto.be economic, t it had to be a pressure suppression containment it had to be sized for the ABWR and it had to be designed to enhance maintainability of equipment. I o He noted that containment-heat removal has three completely f separated divisions. The ccntainment heatfremoval system takes-1 l-i suction from the suppression pool,-rejects the heat load to a heat exchanger and discharges back to the suppression pool. Mr. James noted that emergency core cooling has three i
2 4 Advanced Boiling Water 9, ReactorLSubcom.Mtg. Oct. 31,'89 L L - separate mechanical and electrica1' divisions and:three functions: core-cooling, containment cooling, and shutdown cooling. He P noted=that there wasn't any core uncoverage for any postulated-pipe break. Mr. Ehlert, General Electric Co., discussed the primary and t - secondary containment-functional design. He noted-the primary. containment was a reinforced concrete cylinder. It is designed i i for 0.3G SSE and the design pressure is 45 psig.' He~noted the: containment is designed to both the SSE postulated LOCA worsti case along with loss of offsite power. He noted that-the secondary containment operates.at a-negative pressure with respect to the atmosphere anC the.cleanc zone of primary containment. It operates at about alquartereof-i an inch water-negative pressure. Everything in secondary containment is. considered to be contaminated'or"could be 'i contaminated due to pipe breaks. Mr Ehlert noted'that the only systems that pass from within the direct pressure boundary through the secondary containment wall are main steam lines and J main water lines. Mr Michelson asked about the use of inflatable seals. He asked what provisions are used to assure continuity of air pressure for a long period of time? He also asked what is the plan-in case the seals fail? Mr. Ehlert responded that he'couldn't answer at this time and he would have to check and get back to him. In response to a question regarding the location of the
c 3 -w 'Advanced. Boiling' Water 10 Reactor-Subcom.Mtg. Oct. 31,'89 sceam tunnel,1Mr. Ehlert noted that the steam tunnel goes over the control room. Mr. Micholson<noted that it is non-seismic. piping. Mr. L; 1ert pointed out' that it was a seismic one-4 building. Mr. Ehlert pointed out'that this topic will be covered t in Chapter 3. Mr. George Thomas, NRC,' discussed the ECCS system. He noted i that the'LOCA analysis will be by approved current methods. Several of the members question the way the analysis will beidone using appendix K instead of "best estimate" codes. Mr. Thomas i said that they have used both methods. He said-that the appendix K values were higher than~the "best estimate": values'.- Dr. Catton questioned the results and noted that this will i be discussed further in a Thermal Hydraulics subcommittee --I meeting. Mr. Chandra, Plant Systems' Branch,.NRC discussed control room habitability. He noted.thet NRC wtaff is concerned-about-
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the steam tunnel running-through the control building.- He noted the staff is concerned about the lack of a compartmentLanalysis i for the steam tunnel including the effects of~high energy. The staff is concerned about the effects of piping failures:in.the 1 tunnel, on safety-related structure and components. 'He noted the if GE would provide an analysis which would indicate that the. t failure of high energy piping in the tunnel does not impair safety-related structures and components, then the staff might find the location for this tunnel to be acceptable. 1 Mr. Scaletti noted that this will be discussed in Chapter 3, i
]' q -- m; .4 q y b. Advanced Boiling Water 11 o -Reacto~r Subcom.Mtg. Oct. 31,'89 Mr. Phillip Novak, General' Electric presented an. overview of-Chapter 17 - Quality' Assurance. Mr. Novak noted that three companies, General Electric, Hitachi, and Toshiba ce'all 1 involved in common engineering for the'GE ABWR. The three companies are each-responsible and are all responsible to produce about 1300 common ' engineering ' documents. Each company must formally review'andEapprove each' document. Each company.had lead' l responsibility for about one third-of the documents or about 450 documents. Document review is based upon internalureview memorandums. Each and every comment must be, formally addressed and resolved to t the satisfaction of the.commenter. All three organizations must agree to.the resolution of every comment. i Mr. Michelson asked what kind of. assurance is provided t. hat ] the translations between Japanese and English and'vice. versa are accurate? I Mr. Novak noted they use independent translators and the documents are reviewed by a responsible engineer. Mr. Michelson asked the staff if they believed that this process provided adequate control? Mr. Scaletti said that he thought it did. Mr. Novak noted that GE, Hitachi and Toshiba are committed to 10 CFR part 50 appendix B and JEAG-4101 -:1981, wnich is a Japanese standard for quality assurance. He noted that JAEG-4101 follows the essence of 10 CFR part 50 appendix,B.-
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+- L1 o. Advanced Boiling Water-12- ~ Reactor Subcom.Mtg.-Oct.-31,'89 ler. - Novak-noted that ' Genaral Electric: isl responsible for : the L content-of'all.of the common en;h,eering documents. He remarkedz thht GE has an annual review of the partners of-the ABWR. Nr. Kennoth Hooks, NRC, discussed QA' review. He noted'that he had been involved with the review-of the GE work. He noted the Jack Spraul had reviewed Chapter-17 and Mr. Spraul' believed J that QA as. described in Chapter 17.and the GE Standard QA Plan-met the review requirements as stated;in Chapter 17. E Mr. Hooks remarked that-he'and others had visited GE fnd had-looked at systems designed by Hitachi and Toshiba. He1said-that l it was their opinion that'the three party review was very l-extensive and was equivalent to design verification processes y that are done in the United 1 States. ) L Mr. Hooks remarked that it is NRC's understanding'that GE i L . holds the design that is.being licensed in the United States. GE s accepts total and full responsibility.for the design.- Mr. Quirk GE, noted the-GE has a'QA process that: tracks any difference between the ABWR certification design and the Japanese dGo4gn. He noted that at the point where GE has a U.S. I r.pplication, then det ailed engineering will be done for the 4 j differences between the Japanese design and the U.S. design. l Mr. Quirk noted the GE position on containment venting. Ile said they would provide a rupture disc set to failure ~pressureiot. i the containment. The containment venting can be subsequently closed and containment can be reestablished. He noted that they ) meet EPRI, requirements and in this case excsed the requirements. 1 o i .J 1
[. V $. E - s 4 c, -Advanced Boiling Water 13 Reactor'Subcom.Mtg. Oct. 31,'89 Mr. Wardiasked how GE defines-containment failure? -Mr. Quirk said the design pressure _for:the' containment is_45 PSI.- He noted ultimate containment failure-w'ould=be about 100 PSI. He said the rupture disc would be designed,for 80;to_90, ~ PSI. The meeting was adjoined at 6:45 pm.- ' NOTE :- A transcript of the meeting;is available at the NRC. Public Document Room,_Gelman Bldg. 2120-"L" Street, NW., Washington, D.C.' Telephone-(202)_634 3383 or can'- be purchased from Ann. Riley & Associates, - LTD.,-'1612 K. 'St. NW., Suite 300,-Washington, D.C. 20006,-((202)_293-3950. "I- ....}}