ML20154D809
| ML20154D809 | |
| Person / Time | |
|---|---|
| Site: | 05000605 |
| Issue date: | 09/12/1988 |
| From: | Scaletti D Office of Nuclear Reactor Regulation |
| To: | Marriott P GENERAL ELECTRIC CO. |
| References | |
| NUDOCS 8809160120 | |
| Download: ML20154D809 (13) | |
Text
I September 12, 1988 Docket No. STN 50-605 Patrick W. Harriott, Manager Licensing & Consulting Services feneral Electric Company Nuclear Energy Business Operations e Mail Co& 682 275 Curtner Avenue' San Jose, California 95125
Dear Mr. Marriott:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE GENERAL ELECTRIC COMPANY APPLICATION FOR CERTIFICATION OF THE ABWR DESIGN In our review of your application of your Advanced Boiling Water Reactor Design, we have identified a need for additional information.
Our request for additional information, contained in the enclosura, addresses the areas of SRP Chapters 1, 2, and 3 reviewed by the Plant Systems, Radiation Protection, Structural & Geosciences and Materials Engineering Branches. Questions related to the review being carried out by the Mechanical Engineering Branch e
will be provided in the near future.
In order for us to maintain the ABWR review schedule, we request that you provide your responses to this request by November 15 If you have any concernsregardingthisrequestpleasecallmeon(305)1988.
492-1104.
Sincerely,
/s/
Dino C. Scaletti, Project flanager Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects III, IV, Y and Special Projects Office of Nuclear Reactor Regulation
Enclosure:
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,I w AsMGTON, D. C. 20655 September 12, 1988 Docket No. STN 50-605 Patrick W. Marriott, Manager Licensing & Consulting Services General Electric Company Nuclear Energy Business Operations Mail Code 682 275 Curtner Avenue San Jose, California 95125
Dear Mr. Marriott:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE GENERAL ELECTRIC COMPANY APPLICATION FOR CERTIFICATION OF THE ABWR DESIGN In our review of your application of your Advanced Boiling Water Reactor Design, we have identified a need for additional information. Our request for additional information, contained in the enclesure, addresses the areas of SRP Chapters 1, 2, and 3 reviewed by the Plant Systems, Radiation Protection, Structural & Geosciences and Materials Engineering Branches. Questions related to the review being carried out by the Mechanical Engineering Branch will be provided in the near future.
In order for us to maintain the ABWR review schedule, we request that you If you have any concerns regarding this request please call me on (30115,)1988.
provide your responses to this request by November 492-1104.
Sincerely, P
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QManager Dino calet,Projec Standardization and Non-Power i
Reactor Project Directorate Division of Reactor Projects III, IV, Y and Special Projects Office of Nuclear Reactor Regulation
Enclosure:
As stated l
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L D C W 60RE REQUEST FOR ADDITICNAL I
INFOfMTICH DOCKET HO. S'IN 50-605 4.
STRUCTURAL DUINNIE In section 3.5.3, for local damage prediction of concrete structures 220.'1 and barriers, the concrete wall and roof thicknesses deterinined should not be less than those listed for Region 11 in Table 1 of SRP Section 3.5.3 unless justification is provided.
'Ihe soil-stmeture interaction (SSI) analyses of the mactor building 220.2 (RB) dia:ussed in Section 3.7 of the AIMS SSAR are based on Revinion 2 of SRP Sections 3.7.1 and 3.7.2 as provided for by the Licensing It abould be noted that Revision 2 Review Bases dated August 7,1987.
iu currently in the process of public comments and to this date has Consequently, there may be changes to Revision 2 not been finalized.
which may mquire further discussion of this topic at a later date.
It is indicated that computer programs SASSI and Ct.ASSI/ASD will 220.3 be used to perform SSI analyses.
Indicate hcw these programs are In CLASSI the contribution of radiation damping cannot validated.
be determined on a mode by mode basis and it can have a substantial Provide results of sensitivity studies.
impact on building response.
Since the responses due to SSE are obtained in ratio to the 220.4 response.from the OBE ana' lyes, indicate what is the purpose of establishing response spectra with.07 and 0.10 damping.
In Section 3.7.2.9, a number of conservative assumptions.are listed 220.5 in the calculation of floor response spectra.
Some of the assumptions listed are not relevant to the generation of the floor response spectra, It is stated that the but to the overall design of the equipment.
floor response spectra obtained from the time-history analysis of the In view of building are broadened plus and minus 10% in frequency.
the fact that response spectra for all site-soil cases are combined to arrive at one set of final response spectra (Section 3.7.2.5),
1 indicate how the i 10% broadening is accompitshed.
l In section 3.7.3.2.2, for fatigue evaluation it is indier.ted that only 220.6 10 peak OBE stress cycles are taken into account which appears to be 4
very low, considering the fact that the reactor building may also be subjected to SRV loadings.
As indicated in the SRP Section 3.7.3 larger number of cycles should be cansidered.
in appendix 3A.6 the following statement is made in the first 220.7 paragraph:
The "The behavior of soil is nonlinear under seismic excitation.
soil nonlinearity can be conveniently separated into primary and secondary nonlinearities. The primary nonlinearity is associated with the state of deformations induced by the free-fielo ground The secondary nonlinearity is attributed to the SSI effects.
motion.
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This secondary effect on structural response is usually not significant and is neglected in the appendix."
Indicate if the secondary effect includes the radiation damping, if it does not, indicate how it is considered in the analysis.
220.8 In appendix 3A.6 the computer program SHAKE is used to perform free-field site response analysis. To staff's knowledge, analyses based on SHAKE under certain site conditions may give unrealistic results and it cannot be used indiscriminate 1y.
In view of this observation, indicate what control or caution has been exerted in your use.
220.9 It is noted that ABWR is designed for 60-year life versus the 40-year life for plant design in current regulation.
Fron the point of view of structures, provide your justification for the longer plant life.
220.10 Since the containment is integral with the reactor building, the following are staff's concerns:
(1) The thermal and pressure effects of the containment on the reactor building, especially under severe accident conditions.
(2) The restraint effects of the reactor building floor slabs on the behavior of the containment, especially on the ultimate capacity of the containment.
(The staff has not received Chapter 19 which is believed to contain the estimate of the ultimatecapacity).
(3) The behavior of small and large penetrations which span between containment and reactor building, especially under severe accident conditions.
Your approaches to resolve these concerns should be provided.
If the resolution is to be accomplished through testing, provide a description of the tests to be performed.
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220.11 In section 3.8.1.1.1 it is noted that the main reinforcement in the l
containment wall consists of inside and outside layers of hoop and l
vertical reinforcement and radial bars for shear reinforcement.
It appears that no diagonal seismic reinforcement is used.
Indicate how the tangential shear due to horizontal earthquake is to be resisted.
220.12 in section 3.8.4.3.1.2, for the same loads considered the first load combination under item (1), if compared with the.first load combination under the (2), should obviously be the governing one.
It appears that a re-examination of the load combinations in this section should be made to weed out load combinations which are obviously not controlling the design unless there are errors in the combinations.
Furthermore since the RB is integial with the containment, effects due to such integration should be reflected in the load combinations of structural elements or components outside the containment unless considered othe rwi se.
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220.13 The terms, G1, Gr, and G all as defined in section 3.8.1.3.1 are not listed in table 3.8.- I while the terms lv and ALL listed in Tab'.e 3.8-1 are not defined. Clarification of the table is requested.
220.14 In table 3.8 - 5 for load combiriation No. 3, it appears the acceptance criterton should be changed to S from U unless justified otherwise.
220.15 Discuss the potentials for severe accidents that can be caused by external initiators such as high wind, tornado, tsunami, and earthquakes, and specifically flood since the reactor building has a standard soil embedment of 85 feet.
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O GEUTECHNICAL DGINEERItu Table 2.01 in the Advanced BWR Standard Safety Analysis Report (SSAR) 241 1 (Table 2'01) gives an envelope of ABWR plant site design parameters. This table gives the minimum bearing capacity and the minimum shear wave velocity of the foundation soil. The table also gives the values of SSE and OBE and indicates (a) that the SSE response spectra will be anchored to Regulatory Guide (RG) 1.60, and (b) that the SSE time history will envelope SSE response spectra. The following additional information/ clarification should be provided in the SSAR:
a.
While the SSE (PGA) of 0.3g anchored to RG 1.60 could, in general, be considered conservative for many sites in the Central and Eastern United States, the SSAR should recognize and reflect the fact thet localized exceedances of this value cannot be ruled out categorically and that adequate provisions will be made in the seismic design to consider site-specific geological and seismological factors, b.
TheSSARgivesanOBE(PGA)valueof0.10gandstatesthat,"for conservat. ism, a value of 0.15g is employed to evaluate structural and component responses in Chapter 3." The staff, however, considers the OBE valve to be 0.15g as per criterion 2 of 10 CFR 50 Appendix A and para" graph V of 10 CFR 100 Appendix A which require, in part, that for seismic design considerations the OBE shall be no less than one-half of the SSE.
c.
The SSAR should indicate the procedures that wruld be adopted to evaluate the liquefaction potential at selected soil sites.
It is not sufficient to say that the liquefaction potential will be "none at plant site resulting from OBE and SSE."
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9 COfPONENT INTEGRITY 251.12 Criterion 51, Fracture Prevention of Containment Pmesum (3.1.2.5.2.1) Boundary, is only applicable for contairmente made of ferritic materials.
Since the AIHR containment is made of conente, this section should clarify the applicability of Criterion 51 to the AEHR containment.
251.13 his section must include a discussion of all potential turbine (3.5.1.1.1.3) missiles and mechanisms of missile ger.eration. W e turbine missile discussion should include failure of turbine discs and blades.
251.14 h is section must include e discussion of a favorable turbine (3.5.4.1) orientation or prwide a discussion on maintenance of the main steam turbine to protect against turbine missiles.
252.15 Leak-Befom-Bmak (LBB) - he staff considers LBB evaluations to (3.6.3) be plant specific because parameters such as potential piping degradation mechanisms, piping geometry, materials, fabrication procedures, loads and leakage detection systems are plant specific. Remfore, the detailed LBB analysis should be provided when an application references the AIMR design.
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ATTYTLTAW SYSTDiS 410 1 Section 3.5.1, "Missile Selection and Description," states:
"The missile (3.5.1) protection criteria to which the plant has been analyzed cogly with the intent of 10 CFR 50 Appendix A. G' aral Design Criteria for Nuclear Power Plants." Provide a list of thost instances where the protection criteria are in strict compliance with 10 CFR 50 Appendix A, and those instances where the protection criteria cogly only with its "intent."
Provide an explanation of and justify the acceptability of those missile prctection criteria which are in ecq11ance only with "intent" of 10 CFR 50, Appendix A.
410.2 Section 3.5.1 states:
"A statistically significant missile is defined >
(3.5.1) as one which could cause unacceptable plant consequences or violation of the guidelines of 10 CFR 100." Provide an explanation of "unacceptable plant consequences."
410.3 Section 3.5.1.1, "Failurerates(Pl)yGeneratedMissiles(OutsideContainment}4" "Internall (3.5.1.1) states for valve bonnets are in the range of 10 to 10"; per year." Prcvide a reference or analysis in support of the above statenent.
410.4 Regarding the physical separation requirements, provide a list of (3.5.1.1) all systems (required for safe shutdown, accident prevention or mitigation of consequences of accidents) whose redundant trains do not have missile-ptcof barriers, and include the minimum separation distances. Provide, for the limiting case of the minimum separation distance, an analysis demonstrating the acceptability of the approach of not calculating P2, and instead relying on the "extremely low" prcbability of a missile strike to both trains, or a missle from one train striking the redundhnt train.
410.5 Explain how safety-related systems or components are protected from (3.5.1.1) missiles generated by non-safety-related conponents.
It is the staff's position that missles garerated from nonsafety related components should not impact safety related components since a single active failure is assumed concurrent with the missile.
410.6 Discuss the means by stich stored spent fuel is protecta6 from damage (3.5.1.1) by internally generated missiles.
410.7 Section 3.5.1.1.1.4, "Other Missile Analysis," discusses the exam >1e (3.5.1.1) of analysis of a containment high purge exhaust fan for a thrown ) lade.
Provide the details of this analysis, such as the max ^mun penetration of the blade and the thickness of the fan casing. Discuss whether this analysis is conservative with respect to other rotating equipment missile sources.
410.8 Regarding Section 3.5.1.1.2.2, "htssile Analysis," provide the details (3.5.1.1) of the rack, strap and cover assemble design for the pneumatic system air bottles, showing the thickness of the steel cover and the distance to the concrete slab.
410.9 Regarding Section 3.5.1.1.3, "Missile Barriers and Loadings," provide (3.5.1.1) a list of all local shields and barriers outside intended to mitigate missile effects, giving their specific locations and design data.
Provide an example of an analysis showing that the design of the shield or barrier will withstand the most energetic missile which could credibly impact it.
410.10 Section 3.5.1.2.1, "Rotating Equipment" (which can contribute to inter-(3.5.1.2) nally cenerated missiles inside tie containrent), states:
'By an analysis similar to that in 3.5.1.1.1, it is concluded that no items of rotating equipment inside the containment have the capability of becoming petential missiles." Provide the details of this analysis.
410.11 Regarding Reactor Internal Pump (RIP) motors and impe11ers which can (3.5.1.2) contribute to internally generated missiles inside the containment, ex) lain the bases for concluding that the RIPS are incapable of ac11eving an overspeed condition and that the moters and impe11ers are incapable of escaping the casing and the reactor vessel wall (SSAR Section 3.5.1.2.1).
Your response should explain how the provision of an anti-rotation device at the bottom of the RIP motor which 'prevents backward rotation of the RIP will prevent its overs)eed during tie course of a LOCA or during nornal plant op(eration w1en one RIP issee SSAR stopped and the other RIPS are operating l
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1 410.12 Regarding pressurized components, provide justification for the (3.5.1.2) stat ment, "FMCRD mechanisms are not credible missile sources,"
imade in Section 3.5.1.2.2.
410.13
'Regarding Section 3.5.1.2.3, "Missile Barriers and 1.cadings," provide (3.5.1.2) the same data for internally generated missiles inside the contain-iment, as that requested under Question No. 430.67 above.
410.14 Clarify whether secondary missiles generated as a result of the impact (3.5.1.2) of primary rdssiles have been considered. Explain how protection against credible secondary missiles is provided.
410. M Regarding Section 3.5.1.2.4, "Evaluation of Potential Gravitationti (3.5.1.2)
Missiles Inside Containment "Item 3 "Equipment for Maintenance,"
describe any interfC rec,utror.ents imposed by this item on applicants referencing the ABWR.
410.16 Regarding missiles generated by natural phenomena, provide the details (3.5.1.4) of the tornado-missile analysis performed, identifying the tornado region (as defined in RG 1.76) and the missile spectrum. Discuss the compliance of the analysis with NUREG-0800, Section 3.5.1.4 acceptance criteria; Regulatory Guide 1.76, Positions C.1 and C.2; and Regulatory Guide 1.117, Positions C.1 through C.3 410.17 Provide specific descriptions of all provisions made to protect the charcoal delay tanks against externally generated tornado missiles.
(3.5.2)
Discuss any Interface requirerent imposed by these design provisions.
Regarding SSC to be protected from externally generated missiles, 410.18 discuss compliance with NUREG-0800 Section 3.5.2 acceptance criteria; i
(3.5.2)
Regulatory Guide 1.13, Position C2; Regulatory Guide 1.27, Positions C2 and C3; and Regulatory Guide 1.117, Positions C.1 through C.3.
Clarify whether all nonsafety-related 550, that may adversely impact 410 19 (as a result of Weir failure due to an external missile) the intended (3*b*2) safety function (1.i. achieving and maintaining safe shutdown, mitigating the consequences of an accident or preventing an accident)
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of a safety related SSC, are protected from external missiles. Describe how such SSC are protected.
(3.6.1)
$$AR Section 3.6.1.3.2.2, "Separation," relies on physical separation between redundant essential systems including their related auxiliary systems as the basic protective measure against the dynamic effe Ms of postulated pipe failures. The general arrangement drawings (e.g.,
Figure 1.2-2) are scheduled to be submitted in December 1988. Note l
that additional information on Section 3.6.1 may be requested as a result of the review of the above drawings.
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Section 3.6.1.1.1, "Criteria," states that the overall design generally 410.20 conplies with BTP ASB 3-1.
Specify those criteria which are in strict (3.6.1) ccepliance, and those which are not in strict comsliance with the B1P. Also, provide justification for the items t1at are not in strict compliance.
Provide a li$ ting of all the moderate-ener g piping outside the con-410.21 tainment, but within the scope of ABWR. Also, describe how safety-related (3.6.1) systems are protected from jets, flooding and other adverse environnental effects that may result from pipe failures in moderate energy piping systems.
Jurtify the non-inclusion of pipe failure analyses for the Process 410.22 Sampling System Fire Protection System, HVAC Emergency Ccoling Water (3.6.1)
System and the Reactor Building Cooling Water System as related to the Ultimate Heat Sink. Provide a summary table listing ths protective measures provided against the effects cf postulated pipe failures in each of the above systems and the systems listed in SSAR Tables 3.6-2 and 3.6-<.
Give details for the worst case flooding arising from a postulated pipe 410.23 failure and include the mitigation features provided. Note that for (3.6.1) flooding analysis purpoats, the complete failure of non-seismic Cate-gory I moderate-energy piping systems should be considered in lieu of cracks in determinier the worst case flooding condition.
Identify all the hi;h-energy piping lines cutside the containment 410.24 (but within the ABWR scope), the adverse effects that may result from (3.6.1) failures of applicable lines anong them, and the protection provided against such effects for each of such lines (e.g., barriers and res-traints).
t Clarify whethtr the reactor building steam tunnel is part of the break 410.25 (3.6.1) exclusion bound.ry. Also, provide a subcorpartment analysis for the steam tunnel. Discuss how the structural integrity of the tunnsi and the equipment in the tunnel are protected against piping failures in the tunnel.
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410.26 State how the MSIV functional capability 9 ~
1 (3.6.1)
Provide a summary table of the findings of an analysis of a postu-410.27 (3.6.1) latte worst-case DBA rupture of a high or rederate-energy line for each of the following areas: 1) KCIC compartrent, 2) RWCU equiprent land valve room, 3) other applicable areas cutside the containmsrt f(e.g., housing RFR piping).
Clarify whether protection for safety-related systems and components 410.28 against the dynamic effects of pipe failures include their enclosures (3.6.1) in suitably designed structures or compartments, drainagt systems and If so, give typical equiprent environmental qualification as required.
examples for the above type of protection.
Regarding interfaces (Section 3.6.4.1), ine'iude rest 1ts cf analyces of moderate-encrpy piping f ailures (currantly, the intet face require-410.29 ments address on;y the high-energy piping failures analyses).
(3.6.1)
Appendix 31, "Equipment Qualification Environmental Design Criteria,"
is scheduled to be submitted in Dececter 1988.
Note that additional (3.11) information may be requested based en review of the above appendix.
H 71caticn requirements
. Although there arc no detailed equipment q:for safety-related mechanical equ 410.30
' Quality Standards and Records," GDC 4 'Envirm.reatal Missila Design (3.11)
Bases," and Appenoix B to 10 CFR 50, "Quality Lsurance Criteria for Nuclear Power Plants and Fuel Processitig P1n,ts" (Sections III, "Design Control," ar.d XVII, "Quality Assurance Re,ords") contain the following requireiaents related to equiprent qual'.+ 1 cation:
Components shall be designed to be % ;. lble with the prist91at.9J a) envirowntal conditions, i.,cludit., t ter - associated with LOCAS.
b)
Measures shall be established for ti s 'ection niid review for suitabllity of application of materic.), parts, and equisent that are assential to safety-related functions.
c)
Desiga control r.wasures shall be established for verifying the adequacy of design, Equipment qualification records shall be mainteined and shc11 d) include the results of tests and materials analyses.
Clarify whether the design cceplies with all the above requirerents for safety-related mechanical equipment in a harsh environment within the Provide justificatice for the non-corp 11ance items abe'..
ABVR scope.
and identify any interface requirerents needed to comply with ths above.
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N 451.1 ht am the bases (including afemnoes) for the site envelope of the AIMR design meteorological paneeters listed in Table 2.0-17 Am these values intended to aflect the indicated maximum historical values for the contiguous USA?
ht is the combined winter precipi-tation load free the add! tion of the 100-ysar snow pack and the 48-hour probable maxim a pmcipitation?
ht is the duration of the design temperatum and wind speed valusa?
What pst factors are associated with the extreme winds?
Are any other meteorological factors (e.g., blowing dust) mnaidemd in the AIHR design?
451.2 Short-ters dispersion estimatte for accidental atmospheric releases am not provided explicitly in Section 2.4.3.
If the X/9 values which am listed in Giapter 15 repmeent an umr bound for which the ABRR is designed; what is the bases for their selectior?
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