ML20206D629

Forwards Responses to DC Scaletti 880912 Request for Addl Info on Ssar for Advanced Bwr.Responses Principally Pertain to Chapters 1,2 & 3.Responses to NRC Also Encl

ML20206D629
Person / Time
Site:	05000605
Issue date:	11/14/1988
From:	Marriott P GENERAL ELECTRIC CO.
To:	Chris Miller NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), Office of Nuclear Reactor Regulation
References
MFN-79-88, NUDOCS 8811170177
Download: ML20206D629 (61)

Text

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GE NucIcar Energy wm: Ecc cax.y ris caec Amr w m cA sst:s l

November 14,1988 MFN No. 79-88 b Docket No. STN 50-605 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Charles L Miller, Director Standardization and Non Power Reactor Project Directorate

Subject:

Submittal of Responses to Additional Information as Requested in NRC Letter from Dino C. Scaletti, Dated September 12,1988

Dear Mr. Miller:

Enclosed are thirty four (34) copies of the Responses to Request for Additional Information (RAI) > , the Standard Safety Analysis Report (SSAR) for the Advanced Boiling Water Reactor (ABWh). These responses principally pertain to Chapters 1,2 and 3. Also included are other committed responses to RAls from Scaletti's letter, dated July 7,1988.

It is intended that GE will amend the SSAR with these responses in December 1988.

Sincerely, t

/

P. W. Marriott, Manager Licensing and Consulting Services ec: D. R. Wilkins (GE)

F. A. Ross (DOE)

J. F. Quirk (GE) ,

g l D. C. Scaletti (NRC) -

8911170177 891114 PDR ADOCK 05000605 I

i N

A PDC e

v ABWR 23AM00AD Standard Plant PEV D TABLE 2.01 ENVELOPE OF ABWR PLANT SITE DESIGN PARAMETERS Maximum Ground Water Icel: Extreme Wind: Speed:

2 feet below grade Basic 110 mph %g/130mphg Maximum Flood (or Tsunami) level:I I Tornado: N 1 foot below grade Maximum tornado wind speed: 260 mph Translational velocity: 57 mph

- Radius: 453 ft Precipitation (for Roof Design): Maximum atm AP: 1 A6 psid Maximum rainfall rate: 10 in/hr Missile Spectra:

Maximum snow load: Per ANSI /ANS 2.3 50 lb/sq. ft.

Design Temperatures: Soll Properties:

Amblent Minimum Bearing Capacity (demand): 15ksf 1% Exceedance Values Minimum Shear Wave Velocity: 1000 fps Maximum: 100"F dry bulb /77 F coincident wet Liquification Potential:

bulb None at plant site r 3 ,

Minimum: 10 F from OBE and SSEgiting  :

A Ocr Exceedance Values (Historicallimiti Maximum: 115"F dry bulb /82"F coincident wet j Seismology bulb Minimum: 40 F OBE gagiround Acceleration (PGA):

0.10g Emergency Cooling Water inlet: 95 F SSE PGA : 0.N)g(5)

SSE Response Spectra: per Reg. Guide 1.60 SSE Time History: Envelope SSE Response Spectra IIE

$0 year recurrence interval; value to be utilised for design of non sofety related structures only.

IIE 100 year recurrence interval; value to be utilised for design for safety related structure s only.

IE Probable maximum flood level (PMF), as defined in ANSI /ANS 2.8, "Determining Design Basis Flooding at Power Reactor Sites."

VE 1,000,000-year tomado recurrence interval, uith associatedparameters based on ANSI /ANS 2.3.

Ib] Free field, at plant grade elevation.

(6)

For conservatism, a value of 0.15g is employed to evaluate structural and compenent responses in Chapter 3.

See item 3 in Section 3A.]for additionalinfonnation. $

A AmenJment 3 2.02 L

ABM 23miooxc Standard Plant nt v n well below the design pressure and temperature of 3.1.2.8.2.2 Evaiu: tion Against Criterion 51 the structures. This provides an adequate margin for uncertainties in potential energy sources. The primary containment vessel (PCV)is a re-inforced concrete structure with ferritic parts .

The design of the containment system thus (ths removable head, personnel locks, equipment meets the requirements of Criterion 50. hatches and penetrations), which are made of ma. A terial that has a nil dugtility transition tem-

' For further discussion, see the following peratve of at least 30 F below the minimum sections: service temperature.

Chapter /

The PCV is enclosed by and is integrated with Section liik the r.;nforced concrete reactor building. The pre operational test program and the quality (1) 3.7 Seismic Design assurance program ensure the integrity of the containment and its ability to meet all normal (2) 3.8 Design of Seismic Category I operating and accident requirements.

Structures The containment design thus meets the (3) 6.2.1 Containment Functional Design requirements of Criterion 51.

(4) 6.2.2 Containment Heat Removal System For further discussion, see the fol'owing sections:

(5) 15 Accident Analyses Chapter /

3.1.2.5.2 Criterion $1 Fracture Presention Section Iilk of Containment Pressure Houndary (1) 3.8 Design of Seismic Category I 3.1.2.5.2.1 Criterion $1 Statement Structures The reactor containment boundary shall be (2) 17 Ouality Ass' rane designed with sufficient margin to assure that under operating, maintenance, testing, and 3.1.2.5.3 Criterion 52 Capabliity for postulated accident conditions: Containment leakage Rate Testing (1) its ferritic materials behave in a nonbrittic 3.1.2.5.3.1 Criterion 52 Statement manner; and The reactor containment and other equipment (2) the probability of rapidly propagating which may be subjected to containment test fracture is minimized, conditions shall be designed so that periodic integrated leakage rate testing can be conducted The design shall reflect consideration of at containment design pressure, service temperatures and other conditions of the containment boundary material during operation, 3.1.2.5.3.2 Nluation Against Criterion 52 maintenance, testing, and postulated accider.t conditions and the uncertainties in deternaning: The containment system is designed and constructed and the necessary equipment is (1) material properties; provided to permit periodic integrated leak rate tests during the plant lifetime. The testing (M residual, steady . tate, and transient program is conducted in accordance with 10CFR$0 stresses; and Appendix J.

(3) site of flaws. The testing provisions provided and the test program meet the requirements of Criterion 52.

Amendment 3 3 t 15

.- D '

ABM 2 m i<m e Standard Plant RN D SECTION 3.5 CONTENTS (Continued) section Htle Page 3.5.2 Stmetures. Systems. and Comnonents to be Protected from Externally Generated Afisslies 3.5 7 3.5J Barrier Deslan Procedures 3.57 3.53.1 Local Damage Prediction 3.5 7 3.53.1.1 Concrete Structures and Barriers 3.5-7 3.53.1.2 Steel Structures and Barriers 3.57 3.53.2 Overall Damage Prediction 3.57 3.5.4 Interfaces 3.5-7 l 3.5.4.1 Protection of Ultimate l{ cat Sink 3.58 3.5.4.2 Missile Generated by Natural Phenomena from Remainder of Plant Structures, Systems and Components 3.58

. 3.5.43 Site Proximity Missiles and Aircraft flazards 3.58 6

3.5.4.4 Secondary Missiles Inside Containment 3.58 3.5.5 References 3.58 1LLUSTRATIONS Figure Htls Pace 3.51 Missile Velocity and Displacement Characteristics Resulting from Saturated Steam and Water Blowdowns (1050 psia Stagnation Pressure) 3.59 3.5 iii AmendmeM 3

i ABWR m amic Standard Plant REV R 3.5 MISSILE PROTECTION Criterion 4 of 10CFR50 Appendix A, General g Design Criteria for Nuclear Power Plants. 7 The missile protection design basis for Seismic Category I structures, systems and Potential missiles that have been identified components is described in this section. A are listed and discussed in later subsections.

tabulatic's of safety related structures, systems, and components (both inside and outside After a potential missile has been containment), their location, seismic category, identified, its statistical significance is and quality group classification is given in determined. A statistically significant missile Table 3.21. General arrangement drawings is defined as a missile which could cause showing locations of the structures, systems, and unacceptable plant consequences or violation of components are presented in Section 1.2. the guidelines of 10CFR100.

1 Missiles considered are those that could The examination of potential missiles and result from a plant related failure or incident their consequences is donc in the following including failures within e.e3 outside of manner to determine statistically significant containment, ensironmental generated missiles and missiles:

site proximity missiles. Tbc structures, shields, and barriers that have been designed to (1) If the probability of occurrence of the withstand missile effects, the possible missile loadings, and the procedures to which each missfle (P3 ) is determincd to be less than 10' per year, the missile is dismissed barrier has been designed to resist missile from further consideration because it is impact are described in detail. considered not to be statistically significant.

3.5.1 Missile Selection and Description (2) If (P 3) is found to be greater than 10 per year, it is examined for its probability Components and equipment are designed to have of impacting a design target (P,).

a low potential for generation of missiles as a

• basic safety precaution. In general, the design (3) If the projuct of (P 3) and (P,) is less that results in reduction of missile generation than 10 per year, the missile is potential promotes the long life and usability of dismissed from further consideration.

a component and is well within permissible limits of accepted codes and standards.

(4) If the produchof (P ) and (P,) is 3 greater than 10 per year, the misille is Seismic Category I structures have been examined for its damage probaHlity (P )'

analyzed and designed to be protected against a If the combined probability (i.c P3 x i

wide spectrum of missiles. For example, failure g 3 P, x P 3 = P4) is less than 10 per i of certain rotating or pressurized components of ye'ar, the missile is dismissed.

equipment is considered to be of sufficiently high probability and to presumably lead to (5) Finally, measures are taken to design generation of missiles. Ilowever, the generation acceptable protection again of missiles from other equipment is considered to (P ) gre at e r t ha n 10'$tper missiles with year io be of low enough probability and is dismissed reduce (P 3

), (P,), and/orj (P ), so from further consideration. Tornado generated 3 that (P4) is less'than 10 per year.

missiles and missiles resulting from activities particular to the site are also discussed in this Protection of essential structures, systems section. The missile protection criteria to and components is afforded by one or more of the i which the plant has been analyzed comply with following practices:  :

I Amendment 3 g,g l

o 6 -

23A6100AE Standard Plant nry n 33.1.1 Internally Generated hfissiles (Outside (1) The equipment design and manufacturing Containment) criteria mentioned previgusly result in (P1 ) being less than 10' per year; or These misslies are considered to be those missiles resulting internally from plant (2) Sufficient physical separation (barriers equipment failures within the Nuclear Island but and/or distance) of safety related and outside containment. redundant equipment exists so that the combing probability (P3 x P2 ) I' I'88 33.1.1.1 Rotating Equipment than 10 per year.

3J.1.1.1. Missile Characterization These conclusions are arrived at by noting that pumps, fans, and the lince have synchronous Equipment within the general categories of motors and thus cannot achieve an overspeed pumps, fans, blowers, diesel generators, compres. condition. At rated speed, if a piece such as a sors, and turbines and, in particular, components fan blade breaks off, it will not penetrate the in systems normally functioning during power re- casing. As an example, a containment high purge actor operation, has been examined for any possi- exhaust fan has been analyzed for a thrown blade ble source of credible and significant missiles. at rated speed conditions using an analytical expression from Reference 1. It is determined, 33.1.1.1J RCIC Steam Turbine based on maximum thickness this blade could penetrate, that the blade would not escape the The RCIC steam turbine driving the pump is not fanfasing and consequently (P3 ) is less than a credible source of missiles. .It is provided 10 per year.

with mechanical overspeed protection as well as automatic governingt very extensiv: industrial 3.5.1.1J Pressurized Components and nuclear experience with this model of turbine has never resulted in a missile which penetrated 3.5.1.1J 1 bilssile Characterization the turbine casing.

Potential missiles which could result from 33.1.1.1.3 Afaln Steam Turbine the failue of pressurized components are analyzed in this subsection. These potential Acceptance criterion 1 of SRP Section 3.5.1.3 missiles may be categorized as contained considers a plant with a favorable turbine gen- fluid energy missiles or stored strain energy t erator placement and orientation and adhering to (clastic) missiles. These potential missiles the guidelines of Regulatory Guide 1.115 ad. have been conservatively evaluated against the equately protected against turbine missile haz- design criteria in Subsection 3.5.1.

. ards. Further, this criterion specifies that y exclusions of safety related structures, systems, Examples of potential contained fluid. energy C

or components from low trajectory turbine missile missiles are valve bonnets, valve stems, and '

strike zones constitutes adequate protection retaining bolts. Valve bonnets are considered against low trajectory turbine missiles. The jet. propelled missiles and have been analyzed as turbine generator placement and orientation of such. Valve stems have been analyzed as the ABWR Standard Plar.t now meets the guidelines piston. type missiles, while retaining bolts are of Regulatory Guide 1.115 as illustrated in examples of stored strain energy missiles.

Figure 3.5 2.

3.5.1.1JJ hilssile Analpes 3J.1.1.1A Other Allssile Analpis Pressurized components outside the contain-No remaining credible missiles meet the ment capable of producing missiles have been significance criteria f having a probability reviewed. Although piping failures could result (P,,) greater than 10 per year for rotating in significant dynamic effects if permitted to ce pressurized equipment, because either: whip, they do not form missiles as such because the whipping section remains attached to the Amendment 3 333

.e. o .

ABM us6icoxe .

Standard Plant arv.n remainder of the pipe. Since Section 3.6 factors against failure, and, coupled with the j addresses the dynamic effects associated with low historical incidence of complete severance pipe breaks, pipes rre not included here as failure, were detemined to not be a potential potential ieternal missiles. misslic source (Ref. 9).

All pressurized equipment and sections of pip. (2) Valve Stems All the isolation valves ing that may periodically become isolated under installed in the reaetor coolant systems pressure are provided with pressure relief vahes have stenn with a back seat which clitnicates acceptable under the ASME Code, $cclion 111. the possibility of ejecting valve stems even if the stem threads fall. Since a double The only remaining pressusfred components failure of highly reliable components would considered to be potentially capable of producing be required to produce a valve stem missile, missiles are:

the overall pgbability of occurrence is less than 10 per year. Hence valve (1) valve bonnets (large and small); stems can be dismissed as a source of missiles.

(2) valve stems; (3) Pressure Vessels Moderate energy vessels (3) pressure vessels; less than 275 psig are not credible missile sources. The pneumatic system air bottles (4) therruowelh; are designed for 2500 psig to the ASME Code, Section 111 requirements. These bottles are ,

(5) retaining boks; and not considered a credible source of missiles l2 for the following qualitative analysis: i (6) blowout panels.

(a) The bottles are fabricated from heavy.

These are analyzed as follows: wall rolled stect; (1) Valse Bonnets Valves of ANSI rating 900 (b) The operating orientation is vertical psig and above and constructed in accordance with the ends facing concrete slabs, with the ASME Code, Section til are The bottles are topped with steel covers pressure scal bonnet type valves. Valve thick enough to preclude penetration by bonnets are prevented from becoming missiles a missile, by limiting stresses in the bolting to those defined by the ASME Code and by designing (c) The fill connection is protected by a flanges in accordance with applicable code permanent steel collar.

requirements. Safety factors involved 2 against failure of these type bonnets are (d) The bottles are strapped in a rack to O sufficiently high that these pressure prevent them from toppling over. The scal type valves are not considered a rack is seismically designed to the ASME potential missile source (Ref. 9). Code, Section !!!, Subsection NF require.

ments.

Most valves of ANSI 600 psig rating and below are valves with bolted bonnets. These (4) Thermowells Thermowells are welded to l 'E type valves were analyzed for the safety sockolet connections which in turn are Amendment 3 34-4

MN 23A61ooAn Standard Plant nry n missile consequence mitigation by structural ments.

walls and slabs. These walls and slabs are designed to withstand internal missile effects; 3.5.1.2.4 Evaluation of Potentlal the applicable scismic category and quality group Gradtational hfisslies Inside Containment classification are listed in Section 3.2.

Pene ration of the structural walls by internally Gravitational missiles inside the containmc.21 genrrated misslics is not considered credible. have been considered as follows:

3 's.1J Internally Generated hilssiles (Inside Seismic Category I systems, components, and

< ontainment) structures are not potential gravitational mis-site sources.

Internal missiles are those resulting from plant equipment failures within the contain- Non Seismic Category I items and systems ment. Potential misslic sources from both loside containment are considered as follows:

rotating equipment and pre:,surized components are considered. (1) Cable Tray 3J.1J.1 Rotating Equipment All cable trays for both Class IE and non.

Class 1E circuits are seismically supported By an analysis similar to that in Subsection whether or not a hazard potential is 3.5.1.1.1, it is conclu de d t h a t n o it e m s of evident.

rotating equipment inside the containment have the capability of becoming potential missiles. (2) Conduit and Non. Safety Pipe All reactor internal pumps are incapable of achieving an overspeed condition and the ecolors NorrClass 1E conduit is seismically sup-and impellers are incapable of escaping the ported if it is identified as a potential casing and the reactor vessel wall, respectively, hazard to safety related equipment. All Nurlear Island non safety related piping 33.1.2.2 Pressurized Components that is identified as a potential hazard is seismically analyzed per Subsection Identification of potential missiles and their 3.7.3.13.

consequences outside containment are specified in Subsection 3.5.1.1.2. The same conclusions are (3) Equipment for hiaintenance drawn for pressurized components inside of con.

tainment. For example, the ADS accumulators are All other equipment, such as hoists, that is designed to the AShfE Code, Section Ill, require- required during maintenance will either be ments and are therefore not considered a credible removed during operation, moved to a missile source. One additional item is fine location where it is not a potential hazard ,

motion control rod drives (FhtCRD) under the reac. to safety related equipment, or seismically tor vessel. The FhtCRD mechanisms are not cred. restrained to prevent it from becoming a ible missiles. The FhtCRD housings are designed missile.

(Section 4.6) to prevent any significant nuc! car transient in the event of a drive housing break. 33.1.3 Turbine hilssiles 3J.1.2.3 blisslie Barriers and leadings See Subsection 3.5.1.1.3.

i f

Credit is taken in some cases of rotating and 3.5.1.4 bilssiles Genrrated by Natural pressurized components generating miniles for Phenomena missile consequence mitigation by structural walls and slabs. Penetration for the containment Tornado generated missiles have been deter-walls, floors and slabs by potential missiles is mined to be the limiting natural phenomena not considered credible, llowever, credible sec- hazard in the design of all structures required g ondary missiles, e.g., concrete fragments, may be for safe shutdown of the nuclear power plant.

2: formed following impact of primary missiles. See Since tornado missiles are used in the design Subsection 3.5.4.4 for interface require- basis, it is not necessary to consider missiles Amentment 3 33 6

ABM. 2sx6ioore Standard Plant nrv. n valent static load concentrated at the impact 3. A. Amirikan, Design of Protectin Struc-area is determined. The structural response to tures, Bureau of Yards and Docks, Publica-this load, in conjunction with other appropriate tion No. NAVDOCKS P 51, Department of the design loads, is evaluated using an analysis Navy, Washington, D.C., August 1960.

procedure similar to that in Reference 6 for rigid missiles, and the procedure in Reference 7 4. A. E. Stephenson, Full Scale Tornado Mis-for deformaHs missiles.' site Impact Tests, EPRI NP 440, July 1977, prepared for Electric Power Research 3.5A Interfaces Institute by Sandia Laboratories.

3J.4.1 Protection of Ultimate Heat Sink 5. W. B. Cottrell and A. W. Savolainen, U. S.

y Reactor Containment Technology, ORNL-t Compliance with Regulatory Guide 1.27 as NSIC 5, Vol.1, chapter 6, Oak Ridge Na-j' related to the ultimate heat sink and connecting tional Laboratory, is conduits being capable of withstanding the E effects of externally generated missiles shall be 6. R. A. Williamson and R. R. Alvy, Impact '

demonstrated. Effect of Fragments Striking Structural Ele.

ments, llotmes and Narver, Inc., Revised 3J.4.2 h!!ssiles Generated by Natural Phenomena November 1973.

from Memainder of Plant Structures, Systems and Components 1. J. D. Riera, On the Stress Analysis of Structures Subjected to Aircraft impact The remainder of plant structures, systems, forces, Nuclear Engineering and Design, and components shall be analytically checked to North llolland Publishing Co., Vol. 8,1968, ensure that during a site specific tornado they will not generate missiles exceeding the missiles 8. American National Standard For Estimating considered under Subsection 3.5.1.4 Tomado and Other Ettreme Hind Characteris.

t i cs at Nuclear Power Sites, ANSI /ANS 2.3.

3J.43 Site Prosimity hilssiles and Aircraft liarards 9. River Bend Station Updated Safety Analysis ,

Report, Docket No. 50 458, Volume 6, pgs. n Analyses shall be provided that demonstrate 3.5 4 and 3.5 5, August 1987.

that the probability of site proximity missiles (including aircraft) impacting the Nuclear Island and causing conseqt.ences gresyr than 10CFR Part 100 exposure guidelines is 5.,10 per3 ear.

3.5.4.4 Secondary hilssiles Inside Containment h Protection against the secondary missiles Inside containment described in Subsection 3.5.1.2.3 shall be demonstrated.

3.5.5 References

1. C. V. bloore, The Design of Barricades for Hasardous Pressure Systems, Nuclear Engl-necting and Design, Vol. 5,1967.

2. F. J. hloody, Prediction of Blowdown Thrust and ht Forces, ash!E Publication 6911T 31, August 1969.

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Figure 3.5 2 ABWR Standard Plant Imw-Trajectory Turbine Missile Ejection Zone

, e 2M61(4AE Standard P_lant Rw n 3.6 FROTECTION AGAINST DYNAMIC Subsection 3.6 3 describes the implementation EFFECTS ASSOCIATED WITH THE of the leak before break (LBB) evaluation proce.

POSTUIATED RUPTURE OF PIPING dures as permitted by the broad scupe amendment to General Design Criterion 4 (GDC 4) published This Section deals with the structures, in Reference 1. The piping systems that are systems, components and equipment in the ABWR demonstrated by these procedures to qualify for Standard Nuclear Island. the LBB behasior (See Appendices 3E and 3F) are not postulated to break in the design and evalu-Subsections 3.6.1 and 3.6.2 describe the ation that are required to be performed, la design bases and protective measures which ensure accordance with Subsections 3.6.1 and 3.6.2, for that the containment; essential systems, compo- the potential dynamic effects from postulated nents and equipment; and other essential strue- piping breaks. However, such piping systems are tures are adequately protected from the conse. evaluated for pipe crack effects in accordance quences associated with a postulated rupture of with Subsections 3.6.2.1.5 and 3.6.2.1.6.2.

high energy piping or crack of innderate energy piping both inside and outside the containment. 3.6.1 Postulated Piping Failures In Fluid Systems Inside and Before delineating the criteria and assump. Outside of Containment tions used to evaluate the consequences of pip.

ing failures inside and outside of containment, Th;s subsection sets forth the design bases, it is necessary to define a pipe break event and description, and safety evaluation for determin-a postulated piping failure: ing the effects of postulated piping failures in fluid systems both inside and outside the con-Pipe break event: Any single postulated tainment, and for including necessary protective piping failure occurring during normal plant measures-operation and any subsequent piping failure and/or equipment failure that occurs as a direct 3.6.1.1 Design Bases consequence of the postulated piping failure.

3.6.1.1.1 Criteria Postulated Piping Failure: Longitudinal or circumferential break or rupture postulated in Pipe break event protection conforms to 10CFR50 high energy fluid system piping or throughwall Appendix A, General Design Criterion 4.. Ensiron-leakage crack postulated in moderate energy fluid mental and hiissile Design Bases. The design p, system piping. The terms used in this definition bases for this protection is in compliance with y are explained in Subsection 3.6.2. NRC Branch 'T echnical Positions (BTP) ASB 3-1 and blEB 31 included in Subsections 3.6.1 and 3.6.2, Structures, systems, components and equipment respectively, of NUREG 0800 (Standard Review that are required to shut down the reactor and Plan).

mitigate the consequences of a postulated piping failure, without offsite power, are defined as hf EB 31 describes an acceptable basis for essential and are dest gned to Seismic Category I selecting the design locations and orientations requirements, of postulated creaks and cracks in fluid systems piping. Standard Resiew Plan Sections 3.6.1 and The dynamic effects that may result from a 3.6.2 describe acceptable measures that could be postulated rupture of high energy piping include taken for protection against the breaks and missile generation; pipe whipping; pipe break cracks and for restraint against pipe whip that reaction forces; jet impingement forces; compart. msy result from breaks.

ment, subcompartment and casily pressuritations; decompression waves within the ruptured pipes and The design of the containment structure, coin-seven types of loads identified with loss of ponent arrangement, pipe runs, pipe whip re-coolant accident (LOCA) on Table 3.9 2. straints and compartme. talizatbn are done in Arrendment 3 3 &l

ABM 2miOOxE Standard Plant _. REV. R Table 3.81 LOAD COMBINATIONS, LOAD FACTORS AND ACCEPT C CRITERIA FOR THE REINFORCED CONCRETE CONTAINMENTg'J, {2), (3), H)

Accep-tance ,

ht@fPNmat' %,1 K IA&d CAMdifien g SRV Description No. D .L Pi Po Pa Pi Ps Tt To Ta E E' W W' Ro Ra M l' ADS O LOCA -

SERVICE 4 '

Test 1 1.0 1.0 1.0 1.0 S l Construction 2 1.0 1.0 1.0 1.0 S Normal 3 1.0 1.0 1.0 1.0 1.0 1.0 I 1.0 S FACTORED Sewrv 4 1.0 13 1.0 1.0 13 1.0 1.0 1.0 U Eauronmental 5 1.0 1.5 1.0 1.0 13 1.0 1.0 1.0 U Entreme 6 10 1.0 1.0 1.0 1.0 1.0 1.0 1.0 U l Emironmental 7 1.0 1.0 1.0 10 1.0 1.0 1.0 U Abnormal 8 10 10 13 to 1.0 10 Note s U Sa 1.0 1.0 1J 1.0 1.0 1.0 1.0 Note 5 U 86 1.0 1.0 13 1.0 1.0 1.0 1.0 Nas5 U 9 10 1.0 1.0 1.0 13 1.0 Not: 5 U 9a 10 1.0 10 10 1.25 1.0 1.0 Note 5 U 9b 10 1.0 1.0 1.0 1.25 1.0 1.0 Note 5 U L 10 10 1.0 1.15 1.0 1.0 1.25 Note $ U too 1.0 1.0 1.15 10 10 1.25 1.25 Nas5 U '

10b 1.0 1.0 1.25 1.0 1.0 1.25 1 25 Notc 5 U j Abnormal / 11 1.0 1.0 1.15 1.0 1.25 1.0 1.0 Note $ U Sewre 11: 10 1.0 1.25 1.0 1.15 1.0 1.0 10 Nae 5 U i Emironmental lib 10 10 1.15 10 125 1.0 1.0 1.0 Note $ U 12 1.0 1.0 1 15 1.0 1.25 1.0 1.0 Note 5 U  ;

124 1.0 1.0 1 25 10 1.25 1.0 1.0 1.0 Nue s U 1:b 10 10 1.25 10 1.25 1.0 1.0 1.0 Note $ U t

13 1.0 10 1.0 LO 10 U 14 10 1.0 1.0 1.0 1.0 U Aeriormat/ 15 1.0 1.0 10 10 10 10 10 1.0 Note 5 U I streme 15a 1.0 10 1.0 1.0 10 1.0 10 1.0 10 Nue5 U retronrneatal 15b 1.0 1.0 10 1.0 1.0 10 1.0 1.0 10 Note $ U NOTES:

1. The loads are described in Subsection 3.& L3 and acceptance criteria in Subsection 3.& LS.  ;

2. For any load combination, if the effect of any load component, other than D, reduces the combined load then

• the load component is deletedfrom the load combination.

3. Since Pa, Pi, Ps, Ta, SRVand LOC 4 are time dependent loads; their efTects millbe superimposed accordingly.

4 Load E in load combination 13 is based on the post accident recovery waterflood height in the containment.

(

3. LOCA loads, CO, CHliG, VLCand PS are time dependant loads. The sequence of occurrence is given in Appenda -

38. The loadfactorfor LOC 4 loads shall be the same as the corresponding pressure load Pa, Pi or Ps.

it l' includes )), )*m and )'r. ,

[

7. Qe sequence of occurrence of SR Vloads is given Irk Appendit 3D. ,

1 (G, or G ),yADS and GALL are not concurrent, w hen they are indicated in the load combination.

f1 i Amendment 3 382? '

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ABM 2346i004r REV.A Standard Plant _-

20.2.2 Chapter 2 Questions 241.1 Table 101 in the Advanced BWR Standard Safety Analysis Report (SSAR) 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/clarifica.

tion 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 that 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. The SSAR gives an OBE (PGA) value of 0.10g and states that,'for conservatism, a value of 0.15g is employed to evaluate structural and component responses in Chapter 3.* The staff, however, i considers the OBE valve to be 0.15g as per criterion 2 of 10 CFR 50 Appendix A and paragraph 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 would 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 riant site resulting from OBE and SSE.'

i Amendment 2 20.2

ABWR maar Standard Plant nrw. A 20.2.3 Chapter 3 Questions 220.1

(

In section 3.5.3 for local damage prediction of concrete structures and barriers, the concrete wall and roof thicknesses determined should be less than those listed for Region 11 in Table 1 of SRP Section 3.5.3 unless justification is provided 220.2 The soil structure interaction (SSI) analyses of the reactor building (RB) discussed in Section 3.7 of the ABWR SSAR are based on Resisjon 2 of SRP Sections 3.7.1 and 3.7.2 as prosided for by the Licensing Review Bases dated August 7,1987. It should be noted that Revision 2 is currently in the process o' public coraments and to this date has not been finalized. Cor.sequently, there may be changes to Revisic,n 2 which may require further discussion of this topic at a later date.

220.3 It is indicated that computer programs SASSI and CLASSI/ASD will be used to perform SSI analyses.

Indicate how these programs are validated. In CLASSI the contribution of radiation damping cannot be determined on a mode by mode basis and it can have a substantial impact on building response.

Provide results of sensitivity studies.

220..!

Since the response due to SSE are obtained in rat o to the response from the OBE analysis,indicte what is the purpose of establishing response spectra with .07 and 0.10 damping.

220.5 in Section 3.7.2.9, a number of conservative assumptions are listed in the calculation of floor response spectra. Some of the assumptions listed are not relevant to the generation of the floor response spectra, but to the overall design of the equipment, it is stated that the floor response spectra obtained from the time history analysis of the building are broadened plus and minus 10% in frequency, in view of 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), indicate how the .10% broadening is accomplished.

220.6 in section 3.7.3.2.2, for fatigue evaluation it is indicated that only 10 peak OBE stress cycles are taken into account which appears to be 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 considered.

220.7 In appendix 3A.6 the following statement is made in the first paragraph:

'The behavior of soil is nonlinear under seismic excitation. The soit nonlinearity can be conveniently separated into primary and secondary nonlinearities. The primary nonlinearity is associated with the state of deformations induced by the free field ground motion. The secondary nonlinearity is attributed (c. the SSI effects. The secondary effect on structural response is usually not significant and is neglected in the appendix.*

20.2-Amndant 2

ABM zu6iooar Standard Plant RIV A Indicate if the secondary effect includes the radiation damping. If it does not, indleate 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, analysis based on SHAKE under certain site conditions may give unrealistic results and it canno: b used indiscriminately. In view of this observation, indicate what control or cause 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. . From 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 staffs concerns:

(1) The thermal and pressure effects of the containment on the reactor building, especially t.nder severe accident conditions.

(2) The restraint effects ot' 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 ultimate capacity).

(3) The behavior of small and large penetrations which span between containment and reactor building, especially under severe accident conditions.

Your approaches to resolve tliese concerns should be provided. In the resolution is to be accomplished through testing, provide a description of the tests to be performed.

220.11 in section 3.8.4.3.1.2 it is noted that the main reinforcement in the containment wall consists of inside and outside layers of hoop and 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 Soverning 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 RH is integral 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 otherwise.

220.13 The terms, Gl. Or and G all as defined in section 3.8.1.3.1 are not listed in table 3.8.1 while the terms lv and ALL listed in Table 3.8.1 are not defined. Clarificatbn of the table is requested.

Amndmnt 2 20 2

MN 23A6100AT Standard Plant REV.A 220.14 In table 3.8 5 for load combination No. 3, it appears the acceptance criterion should be changed to S from U unless justified otherwise.

t 220.15 Piscuss the potentials for severe accident that can be caused by externalinitiators such as high wind, tornado, tsunami, and earthquakes, and specifically flood since the reactor building has a standard soil etabedment of 85 feet.

251.12 u

Criterion 51, Fracture Prevention of Containment Pressure Houndasy, is only applicable for containments made of ferritic materials. Since the ABWR containment is made of concrete, this section should clarify the applicability of Criterion $1 to the ABWR containment. (3.1.2.5.2.1) 251.13 This section must include a discussion of all potential turbine missiles and mechanisms of missile generation. The turbine missile discussion should include failure of turbine discs and blades.  ;

(3.5.1.1.1.3) 251.14 This section must include a discussion of a favorable turbine orientation or provide a discussion on maintenance of the main steam turbine to protect against turbine missiles. (3.5.4.2)  :

251.15 ,

l Leak Before Break (LBB) The staff considers LBB cvaluations to be plant specific bacause parameters such as potential piping degradation mechtalsms, piping geometry, materials, fabrication y procedures, loads and leakage detection systems are plant specific. Therefore, the detailed LBB ,

analysis should be provided when an application references the ABWR design (3.6.3) ,

410.1 l Section 3.5.1, ' Alissile Selection and Description,' states: 'The missile protection criteria to which the plant has been analyzed comply with the intent of 10CFR 50 Appendix A, General Design l Criteria for Nuclear Power Plants.* Provide a list of those instances where the protection criteria  ;

are in strict compliance with 10 CFR 50 Appendix A, and those instances where the protection criteria  !

comply only with its ' intent.' Provive an explanation of and justify the acceptability of those missile protection criteria which are in compliance only with

• intent' of 10 CFR 50, Appendix A.

(3.5.1) l i 410.2 Section 3.5.1 states: "A statistically significant nissile is defised as one which could cause l unacceptable plant consequences or violation of the guidelines of 10 CFR 100.' Provide an explanation of "unacceptable plant consequences.' (3.5.1) i l

i 1

l

' m2 Anw M a ni2 I

?h_

fj M 23A6100AT I Standard Plant W^ l 410J Section 3.5.1.1,"Internally Generated Missiles jputsidc Containment)* states: 'Fallure rates (P1) for value bonnets are in the range of 10 to 10'5 per year." Provide a reference or  ;

analysis in support of the above statement. (3.5.1.1) >

410.4 Regarding the physical separation requirements, provide a list of all systems (required for safe  !

shutdown, accident prevention or mitigation of consequences of accidents) whose redundant trains do  :

not have missile proof barricts, and include the rsinimum separation distances. Provide, for the limiting case of the minimum separation distance, an analysis demonstrating the acceptability of the e approach of not calculating P2, and instead relying on the 'entremely low' probability of a missile strike to both trains, or a missile from one train uriking the redundant train. (3.5.1.1) 410J Explain how safety related systems or componcats are protected from missiles generated by non. safety related components. it is the staff's position that missiles generated from nonsafety related components should not impact safety related components since a single active failure is assumed concurrent with the missile. (3.5.1.1) 410.6 Discuss the means by which stored spent fuelis protected from damage by internally generated missiles. (3.5.1.1) 410.7 Section 3.5.1.1.1.4,'Other Missile Analysis,' discusses the example of analysis of a containment high purge exhaust fan for a thrown blade. Provide the details of this analysis, such as the maximum 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. (3.5.1.1) 410J Regarding Section 3.5.1.1.2.2, ' Missile Analysis.

• provide the details of the rack, strap and cover assemble design for the pheumatic system air bottles, showing the thickness of the steel cover and the distance to the concrete slab. (3.5.1.1) 410.9 Regarding Section 3.5.1.1.3, ' Missile Barriers and Loadings,* provide a list of all local shields and barriers outside latended to mitigate missile effects, giving their specific locations and design data. Provide an example of an analysis showing that the design of the shleid or barrier will withstand the mest energetic missile which could credible impact it. (3.5.1.1) 410.10 Section 3.5.1.2.1,' Rotating Equipment * (which can contribute to internally generated missiles inside the containment, 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 potential missiles.* Pro %ide the details of this analysis. (3.5.1.2) 20 2 AmeMineat 2

!- Standard Plant. .. usa .

1 1

410.11 l

Regarding Reactor Internal Pump (RIP) motors and impellers which can contslbute to icternally generated missiles inside the containment, explain the bases for concluding that the RIPS are incapable of achieving an overspeed condition and that the motors and impeliers 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 j backward rotation of the RIP will prevent its overspeed during the course of a LOCA or during normal i plant operation when one RIP is stopped and the other RIPS are operating (see SSAR Section 5.4.1.5).  !

(3.5.1.2) .

l j

410.12 l Regarding pressurized co ponents, provide justification for the Autrment,'FMCRD mechanisms are not credible missile sources, ade in Section 3.5.1.2.2.

t 410.13 l Regarding Ser, tion 3.5.1.2.3, ' Missile Barriers and 1.oadings.' pro,*A . . same data for internally generated missiles inside the containment, as that requested under Question No. 410.8 l j above. (3.5.1.2) i i I 410.14 l

Clarify whether secondary miniles generated as a result of the impact of primary missiles have been considered. Explain how protection against credible secondary raisilles is provided. (3.5.1.2)  ;

i 410.15 a

Regarding Section 3.5.1.2.3, "Evaluation of Potential Gravitational Missiles inside Containment

!

• Item 3,' Equipment for Maintenance,* describe any interface requirements imposed by this item on i applicants referencing the ABWR. (3.5.1.2) l

! 410.16 l

Regarding missiles generated by natural phenomena, provide the details of the tornado. missile analysis performed, identifying flic tornado region (as defined in RG 1.76) and the missile spectrum.

Discuss the compliance of the analysis with NUREG.080'), Section 3.5.1.4 accepts nce criteria, Regulatory Guide 1.76, Positions C.1 and C.2; and Regulatory Guide 1.117, Positions C.1 through C.3 j (3.5.1.4) f i

410.17 i

Provide specific descriptions of all provisions made to protect the charcoal delay tanks against ,

caternally generated tornado missiles. Discuss any interface requirement imposed by these design [

provisions.

410.18 .

Regarding SSC to be protected from externally generated missiles, discuss ccmpilance with NUREG 0800, Section 3.5.2 acceptance criteria; Regulatory Guide 1.13, Position C2; Regulatory Guide i 1.27, Positions C2 and C); and Regulatory Guide 1.117, Positions C1 through C3. (3.5.2) t

[

20 2 AmtmJment 2 i

t

. - _ - _ . . --- - . . - - . _ - - _ _ . _ . - _2

ABWR m-r Standard Plant RIX A 410.19 Clarify whether all nonsafety related SSC, that may adversely impact (as a result of their failure due to an external missile) the intended safety fnnction (i.e. achieving and maintaining safe shutdown, mitigating the consequences of an accident or prever 'og an accident) of a safety related SSC, are protected from external missiles. Describe how such SSL are protected. (3.5.2) 410.19a SSAR .cction 3.5.1.3.2.2, "Separation,' relies on physical separation between redundant essential systee" 'uding their related auxiliary syrtems as the basic protective measure against the dynamic effecta + mstulated pipe failures. The general arrangement drawings (e.g., Figure 1.2 2) are schedulo, be submitted in December 1988. Note that additional information on Section 3.0.1 miy be requested a a result of the resiew of the above drawings. (3.6.1) 420.20 Section 3.6.1.1.1, "Criteria,* stetes that the overall design generally complies with BTP ASil 3 1. Specify those criteria which are in strict compliance, and those which are not in strict compliance with the BTP. Also, provide justificati ca for the items that are not in strict compliance. (3.6.1) 410.21 Provide a listing of all the moderate energy piping outside the containment, but within the scope of ABWR, Also, describe how safety related systems are protected from jets, flooding and other adverse environmental effects that may result from ti re failures in moderate energy piping systems.

(3.6.1) 410.22 Justify the non inclusion of pipe failure analyses for the Process Sampling System, Fire Prc::ction System,ilVAC Emergen:y Cooling Water System and the Reactor Building Cooling Water Splem as related to the Ultimate IIcat Sink. Provide a summary table li' ting the protective measures provided against the effects of postulated pipe failures in each of the abose systems and the systems listed in SSAR Tables 3.6 :. and 3.6 4. (3.6.1) 410.23 Give details for the worst case flooding arising from a postulated pipe failure and include the mitigation features provided. Note that for flooding analysis purpo .cs, the complete failure of non-seismic Category I moderate energy piping systems should 1e considered in lieu of cracks in determining the worst case flooding condition. (3.6.1) 410.14 Identify a;l the high energy piping lines outside the containment (but within the ABWR scope), the adveric effects that may result from failures of applicable lines ameng there, and the protection provided against such effects for each of such lines (e.g., barriers and restraints). (3.6.1) 410.25 Clarify whether the reactor building steam tunner is part of the break exclusion boundary. Also, proside a subcompartment analysis for the steam tunnel. Discuss how the structural integrity of the tunnti and the equipment in the tunnel are protected gainst failures in the tunnel. (3.6.1)

AmenJrwat 2 N 2-i

0 9 ABWR mmr i Standard Plant _

RFV A  !

410.26 l i

State how the M51V functional capability is protected. (3.6.1) 410.27 Provide a summary table of the findings of an analysis of a postunted worst eau DBA rupture of a ,

high or moderate energy line for each of the following areas: 1) MCIC compartment,2) equipment and valve room,3) other applicable areas outside the containment (e.g., housing RilR piping) (3.6.1) l 410.28 '

Clarify whether protection for safety related systems and components against the dynamic effects of pipe failures include their enclosures in suitable design structures or compartments, drainage systems and equipment environmental qualification as required. Il so, give typical examples for the ,

above type of protection. (3.6.1) 410.29 Regarding interfaces (Section 3.6.4.1), incli:de results of analyses of moderate energy piping failures (currently, the interface requirements address only the high energy piping failures <

analysts). (3.6.1) ,

410.29a ,

Appen,liz 31,' Equipment Qualification Envirr.nmental Design Criteria," is scheduled to be submitted  !

la December 1988. Note that additional information may be requested based on redew of the above l appendix. (3.11) 410.30 Although there are no detailed equipment qualification requirements for safety related mechanical equipment in a harsh environment, GDC 1,'Ouality Standards and Records,' GDC 4,

• Environmental Missile Design Vases,' and Appendix B to 10 CFR 50,'Ouality Assurance Criteria for Nuclear Power -

Plants and Fuel Processing Plants" (Sections 111, ' Design Control,' and XVil, 'Ouality Assurance Records") contain the following requirements related to equipment qualification: .

a) Components shall be de igned to be compatible with the postulated environmental conditions, i including those asto< lated with LOCAs. [

b) Measure shall be established for tht selection and review for suitability of applicatic.n of I materials, parts und equipment that are essential to safety related functions, j c) Design control mesures shall be established for wrifying the adequxy of design, d) Equipment qualitication records shall be maintained and sbu include the results of tests '

and material analyses.

Clarify whether the design complies with all the above requirements for safety related mechanical l equipment in a harsh emironment within the ABWR scope. Provide jr'ification for the non compliance items above and identify any interface requirements needed to comply wh t e abow. (3.11) <

Amendmen 2 20.2-

ABWR m6imr Standard Plant nry A 2C.2.5 Chapter 5 Questions 451.1 What are the bases (including references) for the site envelope of the AllWR design meteorological parameters listed in Table 2.017 Are these values intended in reflect the indicated maximum historical values for the contiguous USA? What is the combined winter precipitation load from the addition of the 100 year snow pael and the 48 hout probable maximum precipitation? What is the duration of the design temperature and wind speed values? What gust factors are associated with the extreme winds? Arc any other meteorolor,i:al factors (e.g., blowe 'ust) considered in the ABWR design?

451.2 Short term dispersion estimates for accidental atmospheric releares are not provided explicitly in Section 2.4.3. If you X/O values which are listed in Chapter 15 reprer,ent an vrper bound for which the ABWR is designed; what is the bases for their selection?

Amendment 2 N 2-

a <

23A6t00AT Standard Plant sum n 20.3.2 Response to Second RAI - Referune 2-QUESTION 430.2 Regarding Reactor Coolant Pressure Boundary (RCPB) leakage detection systems, provide information on the following: (5.2.5) i (a) Describe how the leakage through both the inner and outer vessel head flange seal, will be detected and quantified.

(b) 1.ist the sources that may contribe to the identified leakage collected in the Reactor Building Equipment Drain Sump.. l (c) Describe how potential intersptem leakages will be monitored for the 1) Low Pressuie Coolant Injection System,2) liigh Pressure Core Spray System,3) Reactor Core Isolation Cooling System (RCIC) - Water side and 4) Residual licat Removal System Inlet and discharge sides. Your response should include all the applicable (for the ABWR design) systems and components [

connected to the Reactor Coolant System that are listed in Table 1 of SRP Section 5.2.5 and other systems that are unique to ABWP,(except thote that you have already discussed in SSAR {

Subsection .t2.5.2.2, item 11).  !

FISPONSE 430.2a Subsection 5.2.5.2.1(7), which describes the reactor vessel head flange seal monitoring, indicates tuat leakage through both inner and outer seals will be detected by *other* dry well leak  ;

detection instrumentation. Such leakage, through both the inner and outer head Dange 0. ring seals would be detected principally by increases in drywell floor drain sump. All up/ pump out rates, changes in water level of the floor drain sump, increases in the drywell air cooler condensate flow rates and increasing temperatures of the drywell ambient temperature moni: ors, s

llowever, without any further data, none of these *other' drywell leak detection methods can specifically identify the leakage as coming from the failure of the O. rings at the vessel head flange seal. That is, these other methods would only indicate an increase of unidentified leakage into the drywell.

Leakage through only the inner reactor vessel head flange seal is detected and can be identified by monitoring the increase in pressure measured between the two O ring seals. The Icakage amount cannot, however, be quantified, as long as the outer O. ring seal maintains its integrity and the manual valve in the drain line to the drywell equipment drain sump remains closed.

Assuming that leakage through the inner O ring seal has occurred prior to any subsequent leakage through both the inner and outer vessel head flange seals (a reasonable assumption), then the

, potential exists to both identify and quantify leakage through both seals of the reactor vessel head i flange. The leakage through the inner seal would have been previously identified by the increase in pressure being monitored between the two O ring seals. Both an alarm aad measurement indication of l pressure are prosided in the control room.

Any subsequent loss of integrity of the outer O. ring seal (assuming that loss of integrity of the inner O. ring seal had previously occurred) should at this time result in additional alarms indicating an increase of unidentified leakage into the drywell. This new unidentified leakage can be t

5

• This is a continuation of response to the second RAI submitted September 14, 1988. All responses not included herein will be provided by Februaq 28,19S9.

20 3 Amendment 3

{

4 ,

ABWR nA6ioo^r Standard Plant arm n quantified, Also, at the same time, these should be a marked, noticeable drop in the value of the pressure being monitored between the inner and outer O ring seals and this pressure drop would be indicated in the control room.

Thus, by correlating past versus current drywell leakage information, leakage through both the inner and outer vessel head flange seals could be both detected, identified and quantified.

QUESTION 430.2b For the above ABWR, the reactor building equipment drain sumps are also designated as reactor building low conductivity waste (LCW) sumps. Reactor building Door drain sumps are also designed as reactor building high conducthity waste (llCW) sumps.

Sources that may contribute to the identified leakage collected in the reactor building equipment drain sumps include the following examples:

(1) RCIC, RiiRA, B&C and liPCF B&C pump shaft seal drains (2) RCIC barometric condcaser vacuum tank relief vahr discharge (3) RCIC turbine and pump lube oil cooling line relief valve dicharge (4) RilR A, B & C and HPCF B&C pump seal vent and pump s.;ction vent discharges (upon opening valves to allow discharge)

(5) RilR heat exchange; drains (6) Vahr stem packing drains sia return from valve gland leakage treatment system drain lines (7) Process piping sent, leak tightness test and sample drain lines (8) CRD sptem pumps drain and vent lit.es (9) CRD System high point vent and low point drain lines (10) CRD sptem filter drains and vents (11) CRD system pump suction a.nd purge water heater pressure relief valve drains (12) 11CU equipment drains RESPONSE 430.2e items 5,6, and 7 of Table 1.11 (BWR) of SRP Section 5.2.5 were already discussed in Subsection 5.2.5.2.2(11), item 4 of Table 1.11 is not applicable to the ABWR, For all other items (i.e.,1,2 and 3) of Table 1.11 of the SRP, any potential intersystem from the RCPB that might be applicable to the ABWR would have to be postulated to occur as leakage through closed check vahrs and/or closed containment isolation vahes.

For the water side of the reictor core isolation cooling (RCIC) system of the ABWR,i.e , item 3 of Table 1.Il of the SRP, RCIC inlet suction now is drawn from the condensate storage pool (or from the suppression pool) and not from the reactor coolant system pressure boundary. during RCIC operations, the RCIC discharge flow would be through the low point check valve in the RCIC piping, through the RCIC injection valve, and then through another RCIC check valve before flowing intn feedwater line B upstream of the outboard isolation check valve of feedwater line B.

When the RCIC turbine driven pump is not in the RCIC piping, from the condensate ;,torage pool down to the lowpoint RCIC check valve,is filled witi water from the condensate storage pool. Between the outlet of the low point check valve and the closed RCIC injection valve, the RCIC piping is maintained full of water from the condensate makeup water system through two check vahrs interfacing this section of RCIC piping. Any potentialintersystem leakage into the RCIC piping system resulting from leakage through closed RCIC check valves or the closed RCIC injection valve would be from the feedwater system o condensate makeup water system and not from the RCPB.

Amendment 3 20.3

e e ABWR ummr Standard Plant nry n Thus item 3 ofiable 1.11 of the SRP is also not considered as being applicable to the ABWR, as the RCIC water side disch rge is through feedwater line B and is not considered as being directly connected to the reactor system pressure boundary.

It should be noted, however, that there are test points id test valves that are provided in the RCIC system piping that are located both between the inner (Iow point) RCIC check valve and the RCIC injection valve and between the injection valve and the outer RCIC check valve, that are intended to be used to specifically test for the leak tightness of these three RCIC vahes.

Item 2 of Tcble 1.11 of the SRP addresses the inlet and discharge line components of the residual heat removal (RilR) system that are connected to the reactor coolant system.

For the ABWR, the inlet RilR pump suction line of each of the three RilR subsystems, i.e., RilR subsystems A, B and C,is connected to the reactor coolant system by piping, through an RilR shutdown moling line suction valve and through both an outboard and an inboard containment isolation valve, Use suction lines, for each RilR subsystem, are used only the shutdown cooling mode of operation.

Duing most reactor operations, the shutdown cooling mode suction line suction valves are ke) locked

(* he closed position and the outboard and inboard containment isolation valves in the three shutdown cooling mode suction lines are all closed by high reactor pressure isolation signals.

For normallineup of RilR suction, for all three RilR subsystems, the RilR suction lines are also connected by piping through normal RilR pump suction valves to the suppression pool. The three normal RilR pump suction valves are keylocked in the open position such that the normal suction for the RilR pumps is from the suppression pool and not from the reactor coolant system pressure boundary. This normal RilR lineup supports the low pressure core flood (LPCF) mode of operation of the thr(e RilR subsystems.

Potential intersystem leakage from the reactor coolant system into the inlet or suction side of the three RilR subsystems can thus only occur through the shutdown cooling mode suction piping. Such intersystem leakage can only be postulated to result from leakage through three closed valves in each of the three RilR subsystems shutdown cooling mode suction lines. That is, for each RilR subsystem, leakage through both closed inboard and closed outboard containment isolation valves and then tiuough the keylocked closed shutdown cooling mode suction vahr.

For each RilR subsystem, a test point and test vahes are located between the inboard and outboard containment isolation valves of the shutdown cooling mode RilR pump suction lines to be used to specifically test for the leak tightness of the inboard containment isolation valves. Leakage through both containment isolation valves would be detected by pressure sensors located between the outboard containment isolation vahc and the keylocked closed RilR shutdown cooling mode suction valve. liigh pressure in this section of piping would result in a control roam alarm. Significant pressurization at this section of piping, resulting from postulated intersystem leakage through the containment isolation vahes, would be discharged to the suppression pool via pressure relief valves.

For the ABWR, the RilR system discharge or injection lines are used by all three RilR subsystems to return flow to the reactor for both the shutdown cooling mode of operation or the LPCF mode of operation. The discharge lines of the three RilR subsystems are normally best fuli of water up to the injection valves by RilR discharge line fill pumps A, B and C. Suction flow for filling the discharge lines is drawn from the suppression pool and not from the reactor coolant system.

Only the RilR subsystem B and subsystem C discharge lines communicate directly with the. Reactor Coolant System pressure boundary. The RilR subsystem A discharge is not directly to the reactor coolant system pressure oounda.ry but rather such discharge (whether for shutdown cooling mode of operation or for the LPCF mode of operation) is through RilR subsystem A injection valve (normally Amendment 3 20.3

l ,

ABWR mim Standard Plant arv n closed) and through R}iR subsystem A check valve (normally closed) into feedwater line A.

Thus potential intersystem leakage from the reactor coolant system is only postulated to occur into RilR subsystems B and C discharge lines resulting from leakage through normally closed discharge check valves and normally closed discharge check valves and normally closed injection valves of RilR subsystems B and C. Test points and test valves are located between the discharge check valves and the discharge injection valves of RilR subsystems B and C to be used to specifically test for the leak tightness of the discharge check valves and injection valves (both normally closed). Substantial leakage through the discharge check valve and closed injection valve of either RilR subsystem B or C would result in pressuritation of the discharge lines which would lead to a control room alarm.

Significant pressurization of the discharge piping of RilR subsystems B or C, resulting from postulated intersystem leakage, would be discharged to the suppression pool via pressure relief valves.

Item 1 of Table 1.11 of the SPR addresses the components of the safety injection systems that are connected to the Reactor Coolant System.

For the ABWR, the low pressure safety injection system is the low pressure core flood (LPCF) mode of the RilR system. The connections between the reactoi coolant system and the discharge lines at the RilR subsystems were presiously discussed abose, in the LPCF mode of operation of the RilR subsystems, inlet suction flow is drawn from the suppression pool and not from the reactor coolant system.

For the ABWR, the high pressure safety injection system consists of the two high pressure core flood (llPCF) systems B and C (and also RCIC which was presiously discuued). There is no connection between the reactor coolant system and the inlet suction of IlPCF systems B and C. Both systems draw their suction flow from the condensate storage pool (or the suppression pool) and not from the reactor coolant system.

The discharge lines of IIPCF sy-tems B and C connect to the reactor coolant system through discharge check valves and injection valves. Potential intersystem leakage from the reactor coolant system is only postulated to occur into llPCF systems B and C discharge lines resulting from leakage through normally closed check valves and norinally closed injection valves. Test points and test valves are located between the discharge check valves and the injection valves of HPCF systems B and C to be used to specifically test for the leak tightness of the discharge check valves and the normally closed injection valves.

Normal lineup for both IIPCF systems B and C is through normally open suction valves connected to the condensate storage pool. These suction lines will fill with water down to the condensate storage pool suction check valves. The discharge lines for llPCF systems B and C are maintained full of water with water sourced from the makeup water system (condensed). Substantial (potential) leakage from the reactor coolant system through closed discharge check valves and closed injection valves into either llPCF B or C discharge lines would result in pressuriration of both the discharge line and the llPCF pump i.uction line which would lead to a control room s. arm indicating high ilPCF B (or C) pump suction pressure. Significant pressuritation of the suction piping for either llPCF pump B or C, resulting from postulated intersystem leakage, would be discharged to the suppression pool sia a pressure relief vahe.

QUESTION 430.3 Discuss compliance of reactor coolant leak detection systems with Regulatory Guide (RG) 1.45,

' Reactor Coolant Preuure Boundary Leakage Detection Systems *, Positions C4, CS, C6, C8, and C9 with respect to the following items: (5.15)

Amendment 3 M

ABWR umm Standard Plant nry n (a) Indicators for abnormal water !cvels er flows in all the affected areas in the event of intersystem leakages.

(b) Sensitivity and response time of leak detection systems used for unidentified leakages outside the drywell.

(c) Qualification relating to seismic events for dr>well equipment drain sump monitoring system and leak detection systems outside the drywell.

(d) Testing Procedures hionitoring sump levels and comparing them with applicable flow rates of fluids in the sumps.

(c) Inclusion of reactor building and other areas floor and equipment drain sumps in ABWR Technical Specifications for leak detection systems.

Note that a few of the questions above arise because in Subsection 5.2.5.4.1 you state that the totalleakage rate includes leakages collected in drywell, reactor building and other area floor drain an equipment drain sumps.

RESPONSE 430.3 As noted above, several questions arose because in Subsection 5.2.5.4.1 it was stated that:

'The total . . . leakage rate consists of all leakage, identified and unidentified, that flows to the drywell, reactor building and other area floor drean and equipment drain suraps.'

The italicized wording was incorrectly included with the text of Subsection 5.2.5.4.1. Subsection 5.2.5.4.1 has bec resised accordingly, liistorically, total leakage rate limit, as established by Plant Technical Specifications, have been associated only with the potential leakage into the reactor primary containment (drywell) as collected by the drywell floor and equipment drain sumps and as monitored by different drywell leakage detection systems, e.g., drywell atmosphere (gaseous and/or particulate) radioactivity monitoring, drywell sump / level monitoring and drywell air coolers condensate flow monitoring. Also, the recommendatiens and regulatory positions of Regulatory Guide (RG) 1.45 have been interpreted in the past as applying only to reactor coolant leakage into the primary containment. RG 1.45 Positions C1, C2 and C3 specifically address leakage to the primary reactor containment and indication of leakage to the containment.

This questions addresses compliance with RG 1.45 Position C4 which recommends that provisions should be made to monitor systems connected to the RCPB for signs of intersystem leakage and also suggests that monitoring and indicators to show abnormal water level or flow in the affected areas.

Specifically, this questions requests discussion of compliance with RG 1.45 Position C4 with respect to the ' indicators for abnormal water level or flows in oli the affected areas in the event of intersystem leakages.'

As indicated in the Subsection 5.2.5.7 discussion that is related to RG 1.45: compliance, radiation monitoring of the reactor building cooling water coolant return lines from the RilR, RIP, CON and SPCV heat exchangers is the monitoring method used for determining potential intersystem leakage from the RCPB within these heat exchangers.

Also, i the discussion related to the response to part C of Question 430.2, it is indicated that Amendment 3 20 3

ABM 2346ionar Standard Plant RrN. n pressure monitoring (l.c., alarming of abnormally high pressure) was the monitoring method utilized for determining potentint intersystem leakage into the other systems connected to the RCPB that are listed in Table 1.II of the SRP.

Since all potential intersystem leakage from the RCPB for the ABWR would be into closed systems, '

normally filled with water (whether flowing or not),' indicators for abnormal water levels or flows

• are not utilized in the ABWR leak detection systems. i RESPONSE 4M3b This question apparently expands upon the RG 1.45 Position C5, which recommends that the sensitivity and response time of allleak detection systems addressed by RG 1.45 Position C3,i.e.,

those that are employed for unidentified leakage into the priniary reactor containment, should be adequate to detect a leakage rate, or its equivalent, of one gpm in less than one hour. This question requests discussion of the sensitivity and response time of leak detection systems used for unidentified leakage outside the dr>well.

For the ABWR, where the monitoring methods of the leak detection systems that are used for defecting unidentified leakages outside the drywell are the same as or similar to the methods used for detecting unidentified leakages inside the drywell (primary reactor containment), it is intended that leak detection systems outside the drywell will be designed for the same sensitivity and response time requirements as recommended by RG 1.45 Position CS,i.e. adequate to detect a leakage ,

rate, or its equivalent, of one gpm within one hour. The siring requirements for the reactor i building floor drain sumps and for the number and capacity of the associated sump pumps are exactly the same as those required for the dr>well floor drain sump and equipment drain sump and pumps.

i RESPONSE 4303c This question is related to RG 1.45 Position C6 compliance and specifically addresses the  !

qualification related to seismic events for the drywell equipment drain sump monitoring system and the leak detection system outside the drywell. .

Note first that, in the response to Question 430.5, it was indicated that all elements of the leak j detection systems (both inside and outside the drywell), which must accomplish a safety function or  !

whose failure could prewnt accomplishment accomodate a SSE and remain functional, and such elements  ;

will be designated as Seismic Category I equipment.

For the ABWR, the dr)well equipment drain sump monitoring system will be qualified for OBE and this meets RG 1.45 Position C6 requirements.

RESPONSE 4303d This question requests discussion of the compliance of the reactor coolant leak detection systems with Position C8 of RG 1.45 with respect to

• Test Procedures . Monitoring sump levels and comparing them with applicable flow rates of fluids in the sumps.'

The recommendations of RG 1.45 Position C8 have usually been interpreted as the

• Testability

• requirements placed upon the leak detection requirements placed upon the leak detection systems, i.e., the requirements related to Paragraph 4.10 of IEEE279; Capability for Test and Calibration as per the RG 1.45B discussion of Signal Correlation and Calibration.

' Testability *, i.e., the ability to test for operability and the ability to calibrate, differs Amendment 3 20.3

,I 1, _ -

ABWR mean Standard Plant RiiV. H from "Testing Procedures". As was indicated in Subsystem 5.2.5.7, the Position C8 requirements of RG 1.45 are satisfied, as, per the requirements, the leak detection systems of the ABWR are ' equipped with provisions to readily permit testing for operability and calibration during plant operations.'

The SSAR text provides example testing methods to show how pronsions had been made to permit testing for operability and calibration during plant operations.

In the context of this c,uestion, "Testing Procedures

• are those viable methods which can be used during reactor operations to confirm the operability of specific leak detection systems, or are the methods which, because of design features or provisions, can be used to confirm that adequate calibration has been maintained, e.g., by the cross comparing or correlation of tht, signal outputs from two or more leak detection systems.

As an example of provisions in the design, the sump desiga for the ABWR requires that the sumps be configuicd such that the sump volume increases as a function of water level in the ratio of 16 gallons per vertical inch. The sump level monitoring is compatible with this sump configuration. By using surap pump timers, the rate at which a pump fills with refeience to sump pump operations can determine the degree of abnormalleakage collected in the sump. Also, the rate of actual sump lesel change, which is also being monitored can determine the degree of abnormalleakage. Because of the required sump configuration, these two measures of the degree of abnormal leakage have a known correlation. As another example, the measurement of drywell air cool:rs condensate flow can be checked against sump level rate of change.

Similar examples o such

• Testing Procedures

• ur methods as provided in Subsection 5.2.5.7 to show satisfaction or compliance with Position C8 requirements.

RESPONSE 430.3e Part e of Ouestion 430 apparently requests discussion related to compliance with RG 1.45 Position C8 with respect to the possible inclusion of new limiting conditions in the ABWR Technical Specifications for the leakage collected outside the drywell, i.e., unidentified and identified leakage collected in the reactor building and othcr at :, (e.g., main steam tunnel area) floor drain (llCW) sumps and equipment drain (LCW) samps.

Such inclusion for the AllWR Technical Specifications is not being proposed. As indicated at the outset of this response, the statement in Subsectior. 5.2.5.4.1 has been revised.

QUESTION 430.18 Describe the manner in which suppression pool dynamic loads resulting from postulated loss of coolant accidents, transients (e.g., relief valve actuation), and seismic events have been integrated into the affected containment structures. Provide plan and section drawings of the containment illustrating all equipment and structural surfaces that could be subjected to pool dynamic loads. For each structure or group of structures, specify the dynamic loads as a function of time, and specify the relative magnitude of the pool dynamic load compared to the design basis load for each structure. Provide justification for each of the dynamic load histories by the use of appropriate experimental data and/or analyses.

Describe the manner by which potential asymmetric loads were considered in the containment design. Characterire the type and magnitude of possible asymmetric loads and the capabilities of the affected structures to withstand such a loading profile. (6.2) 20.3 Amendment 3

gg 23A6tooAT j

Standard Plant "" '

RESPONSE 430.18 .

The pool dynamic loads such as vent clearing, pool swell, condensation oscillation, chugging, etc., resulting from postulated loss of coolant accidente and the pool dynamic loads associated with safety relief valve actuation during transients are to be found in Appendix 3B, The containment vessel, basemat, pedestal, access tunnels and quencher supports are subjected to these loads. In Appendix 3B, the time histories including magnitude and duration as well as spatial variation and distribution of these loads are specified. The seismic loads on the structures such as pedestal, containment vessel structure, etc., at:io be found in Appendix 30. The combinations of seismic loads OBE/SEE with pool dynamic loads for the containment design are in accordance with Tables 3.81, 3.8 5 and 3.8-6.

QUESTION 430.19 Provide information to demonstrate that the ABWR design is not vulnerable to a safety relief valve discharge line break within the air space of the wetwell, coupled with a stuck open relief valve after its actuation as a result of the transient. (6.2)

RESPONSE 430.19 f There is a high degree of assurance that the ABWR.SRYDL wetwell piping is not vulnerable to a safety relief valve discharge line break within the air space of the wetwell due to a stuck open relief valve. Pressure / temperature conditions in the discharge line during valve actuation have been analyzed and the calculated stresses and fatigue usage factors for all the load combinations which 1 include the stuck open valve loads were within code allowables. This provides confidence that the corubination of discharge line failure and valve actuation is a very low probability event and is not a significant issue for the containment. The design and analysis of the wetwell piping is briefly described below.

i Materials 7

(1) Corrosion Resistance l The material selected for the air space piping is 316 stainless steel with .05% max carbon.

This materialit resistant to corrosion in the high humidity of the wetwell and also has high resistance to intergranular stress corrosion, i

(2) Non Ductile Fracture l

Stainless steel is not subject to non ductile failure, a concern with some grades of carbon stccl.

Desian and Anahsis 4

The wetwell piping was designed and analyred to ASME class 3 requirements. All known (postulated) l loading conditions were evaluated in the stress analpis.

i j Protection against some specific potential failure modes is discussed below.

(1) Fatigue Failure by Thermal Expansion j Class 3 piping is safe for 7000 cycles of thermal expansion when the stress ratio is 1.0 or l

XL3-Amendment 3 i

ABWR m6iar nry n Standard Plant below. The maximum expansion cycles for any valve is only 1790 cycles and the stress ratio is below 0.8 which assures no fatigue failure.

(2; System Collapse Due to Weight Stress

'lhe weight of the upper wetwell 10 inch piping is supported by the heavier 12 lach piping extcoding vertically down to the quencher. The stresses are far below the code allowable.

(3) System Collapse due to Dynamic Loads The anchors and the lateral supports control the stresses due to dynamic loads at every location in the piping location within the code allowable stress limits.

Fabrication and Installation (1) Welds and Welding To guard against fabrication defects, all welding and welding nondestructive examination is done to ASME class 3 requirements. This prosides 100% radiography of he welds in the welwell air space.

The number of welds in the wetwell air space was minimited by using bent pipe wherever possible. There are no longitudinal pipe welds due to the use of seamless pipe.

(2) Hydrotest Following installation, the piping is hydrostatic tested to assure no leaks and demonstrate structural integrity.

QUESTION 0036 For isolation valve design in systems not within the ABWR scope, identify the systems and the relevant interface requirements. Include a discussion on essential and non essential systems per Regulatory Guide 1.141 and the means or criteria provided to automatically isolate the nonessential systems by a containment isolation signal. Also, include a discussion on the requirement that the setpoint pressure which initiates containment isolation for nonessential penetrations be reduced to the minimum value compatible with normal operations. (6.2)

RESPONSE 43036 Allisolation vahrs are within the scope of the ABWR Standard Plant.

QUESTION 43037 Specify all plant protection signals that initiate closure of the containment isolation valves.

(6.2)

RESPONSE 43037 A summary of the plant protection signals that initiate closure of the containment isolation valves is provided below:

Amendment 3 20.3

ABM uraioort am n Standard Plant (1) Reactor vessel water level (1 1/2,2 or 3) I (2) Reactor pressure high (3) MSL turbine arca high amb temp (4) MS tunnel ambient temperature high i (5) MS line high 11ow rate (6) MS line low pressure (reac mode sw in 'run' pos)

(7) MS line high radiation (8) Drywell high pressure (9) St.C switch in run start position r

(10) RCIC steam press low (11) Main condenserlowvacuum (13) RHR equipment area high amb ten.p (19) RCIC equipment area high amb temp ,

(15) RCIC exhaust diaphragm pressure high (16) RCIC steam supply diff pressure (hlgh flow)

(17) CUW process piping high differential flow (18) CUW equipment area high amb temp QUr3 TION 430.38 1.'escribe the leakage detection means provided to identify leakage for the outside containment remo;e manual isolation valves on the following influent lines: Feedwater. R}{R injection,11PCS, standby liquid control, RWCU connecting to feedwater line, RWCU reactor vessel head spray. (6.2)

RESPONSE [

t These valves will be tested periodically in accordance with the procedures of Appendix J of 10CFR50- type C tests to determine the leakage tightness across the valve seat. Gross valve leakage out of the system will be dete-ted (depending on location) by radiation monitors, temperature sensors (main steam tunnel) or drain sumps.

i QUESTION 430.40 With respcct to Figure 5.2.Ta:

(a) include the holation valve arrangement of the standby liquid control system line.  ;

(b) Identify the line labeled in the figure as 'WDCS A' (it joins the RWCU li'ie prior to its connection to the feedwater line), and discuss the isolation provisions for that line.

l RESPONSE 4M.40a The standby liquid control system will be added to the next revision of Figure 6.2 3ba. The valve '

configuration is shown in Figure 20.313. ,

RESPONSE 430.40b l

'WDCS' is now 'RiiR* I,tt.is correction will be made to the next revision of Figure 6.2 38a). The f isolation for this line is p; raided by the feedwater system outboard motor operated valve shown in Figure 6.2 38a.

I AmMment 3 20 }

ABM ux630aar Standard Plant nnv. n QUESTION 430.41 Provide a diagram or reference to figure (s) showing the isolation valve arrangement for the lines ,

identified below. For the isolation valve design of each of these lines, provide justification for not meeting the explicit requirements of GDC 56, and demonstrate that the guidelines for acceptable alternate containment isolation provisions contained in SRP 6.2.4 are satisfied. The lines in i

question are: l HPCS and RHR test and pump mininow bypass lines RCIC pump miniflow bypass line RCIC turbine exhaust and pump miniflow bypass lines SPCU suction and discharge lines (6.2)

RESPONSE ,

4 GE is applying the same alternate criteria to the ABWR that has been reviewed and approved by the f NRC on past licensing applications. Specifically, the ABWR design applies GE Safety Standard 20 No. [

8 te No. 9 to the lines in question as shown in Table 20.3 7.

I i QUESTION 430.44 identify the system lines whose containment isolation requirements are covered by GDC 57 and discuss conformance of the design to the GDC requirements. (6.2) ,

1

RESPONSE 430,44  ;

GDC 57 addresses closed loop systems which penetrate the containment but do not communicate with

! the containment interior. The system lines shown in Table 20.3 8 have been identified and are l j considered to ronform to GDC57 with the valve configuration as shown. The heavy lines denote an extension of the containment boundry, l

! i

]

t i i l

,! I I

5  !

I i

I I

I d

I I

20 S Amendment 3

ABWR mmr myn StatidarAflant ,

2033 Response to Third RAI Reference 3 QUESTION 220.1 i

in section 3.5.3 for local damage prediction of concrete structures and barriers, the concrete wall and roof thicknesses determined should be less than those listed for Region 11 in Table 1 of SRP Section 3.53 unless justification is provided RESPONSE 220.1 The ABWR meets the acceptance criteria defined in Subsections .s.53 and 3.5.1.4 cxcept for Reg.

Guide 1.76. The Design asis Tornado and the Design Basis Tornado Missiles are defined using ANSI /ANS 2.3. The 10) tornado has been used for this design.

L QIJESTION 220.2 The soil. structure interaction (SSI) analyses of the reactor building (RB) discussed in Section 3.7 of the ABWR SSAR are based on Revision 2 of SRP Sections 3.7.1 and 3.7.2 as provided for by the Licensing Review Bases dated August 7,1987, it should be noted that Revision 2 is currently in the process of public comments atid to this date has not been finalized. Consequently, there may be changes to llevision 2 which may require further discussion of this topic at a later date.

RESPONSE 220.2 GE agrees that significant differences between Draft Revision 2 of SRP Section 3,7.1 and 3.7.2 as delineated in the ABWR Licensing Review Bucs sad their finalissued form may require further discussions.  ;

QUESTION 220J It is indicated that computer programs SASSI and CLASSI/ASD will be used to perform SSI analyses.

Indicate how these programs are validated. In CLASSI the contribution of radiation damping cannot be i determined on a mode by mode basis and it can have a subs:antial impact on building response.

Provide results of sensitisity studies.  ;

RESPONSE 2203 The SASSI computer program is validated by running 14 test problems. These problems are designed to verify the four major capabilities of the program, finite element library, impedance analysis, i scattering analysis, and SSI analysis. The verification matrix is shown in attached Table 203 5. l Depending on the complexity of the problem, the SASSI results are compared to hand calculations, i solutions from other computer programs, or solutions in published technicalliterature. The result comparisons are good for all test problems. j l

The ClASSI/ASD computer program is validated by running 12 test problems. Of these 12 problems, l 10 problems are analyzed for complete SSI solutions. The verification summary is prosided in l attached Table 203 6. The CLASSI/ASD solutions are in good agreement with published results l calculated using different methodologics.  ;

The statement that in 'CLASSI the contribution of radiation damping cannot be determined on a r mode by. mode basis

• is not correct. The CLASSI family of programs computes the soil structure 20 1 I Amendment )  !

i 4

MM 23A6100AT Standard Plant nry n interaction response of the foundation in the frequency domain. The governing equations of motion are given in Equation 3AA 11 in Attachment A to Appendix 3A, Seismic Soil. Structure Interaction  !

4 - Analysis. The [K(w)] term in the equation includes bvth the real (stiffness) and imaginary (damping) l 4

frequency. dependent impedance terms. For each discrete frequency, the appropriate radiation terms  !

are used to calculate the SSI foundation motions. Hence, the contribution of radiation damphig is  !

explicitly considered on a mode.by mode basis.

QUESTION 220.4 Since the response due to SSE are obtained in ratio to the response from the OBE analysis,indicte  ;

t what is the purpase of establishing response spectra with .01 and 0.10 damping.

RESPONSE 220.4  ;

The purpose of establishing response spectra with 0 07 and 0.10 damping is to demonstrate the degree of conservatism inherent in the SSE response which is defined to be twice the calculated OBE t response using lower OBE damping values. For reinforced concrete structures, a damping value of 0.07 ,

is allowed for SSE design according to Regulatory Guide 1,61, and a damping vnlue of 0.10 is a more realistic value according to NUREG/CR-0098. Therefore, seismic design margins are ensured in view of response spectra with 0.07 and 0.10 damping.

4 QUESTION 2203 l l In Section 3.7.2.9, a number of conservative assumptions are listed in the calculation of floor i response spectra. Some of the assumptions listed are not relevant to the generation of the floor I response spectra, but to the overall design of the equipment, it is stated that the floor response ,

i spectra obtained from the time. history analysis of the building are broadened plus and minus 10% in i

{ frequency. In view of the fact that response spectra for all site. soil cases are combined to arrive l 4 at one set of final response spectra (Section 3.7.2.5), indicate how the .10% broadening is ,

j accomplished. l RESPONSE 2203 i

j The response spectra broadening process is accomplished in three steps. The first step of the process is to envelope three steps. The first step of the process is to envelope all of the SASSI cases in frequency to create a SASSI envelope response spectra. The next step is to envelope the ,

CLASSI/ASD cases in frequency to create a CLASSI/ASD envelope response spectra. The last step is to  !

peak broaden and scale the SASSI en elope until it bounds both the SASSI and CIASSI/ASD envelopes. ,

The minimum peak broadening applied is .2.10%.  ;

i QUESTION 220.6 [

t in section 3.7.3.2.2, for fatigue evaluation it is indicated that only 10 peak OBE stress cycles are taken into account which appears to be 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 considered.

RESPONSE 2204  !

l The adequacy of 10 peak OBE stress cycles is demonstrated from a generic study as stated in l Subsection 3.7.3.2.2. This study established the technical basis for previous GE BWR plants and has l been accepted by the NRC for generic applications. [ Letter, Robert J. Bosnak (NRC) to R. Artigas  ;

(GE) dated February 18,19S2].  :

C Amendment 3 20 1 h

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MM 23A6100AT Stands.rd Plant _ uv. s c

in addition to 10 peak OBE stress cycles, cycile loadings due to SRV actuation with much larger number of cycles are also considered in the fatigue evaluation. The total numt,cr of cycles includes hundreds of scram events with thousands of SRV actuations and each actuation results in several load cycles. l

. 1 QUESTION 220.7 In appendix 3A.6 the following statement is made in the f.rst paragraph:

'The behavior of soil is nor.l!near under seismic excitation. The soll nonlinearity can be conveniently separated into primary and secondary nonlinearities. The primary nonlinearity is associated with the state of deformations induced by the free field ground motion. The secondary nonlinearity is attributed to the SSI effects. The 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,indleate how it is considered in the analysis.

RESPONSE 220.7 Both the primary and secondary soil nonlinearities affect the soll material damping only. The ,

effect of radiation damping is considered in computing frequency dependent foundation impedances in the SSI analysis. In computing these foundation impedances, the strain compatible soil properties which account for the primary soit nonlir.carity, are used.

QUESTION 220J In Appendix 3A.6 the computer program SilAKE is used to perform free field site response analysis.  !

To staffs knowledge, analysis based on SilAKE under certain site conditions may givt unrealistic [

results and it cannot be used indiscriminately, la view of this observation, indicate what control

~

or cause has been exerted in your use.

1 RESPONSE 220J In performing free field site response analysis using computer program SilAKE, the strain dependent shear modulus and damping values corresponding to each site condition was used, in these analyses, controls in the form of limiting the soil modulus degradation and soil matcrhl damping ratios have  :

been made to ensure that results were reasonable. Furthermore, for soft site conditions, the cut off frequency for the SilAKE analyses was also contiolled such that the results would be reasonable. As shown in Tables 3A.3 3 and 3A.3-4 of Appendix 3A the reduction factor used for strain dependent shear modulus was limited to 0.40. The increase in the strain dependent materirl damping was limited to i 15% Free field site responses of the softest sitts (UB and VP2) in terms of variations of maximum acceleration and iterated shear wave velocity with depth are shown in Figures 20.311 and 20.312. l These results show that with the limits considered for the strain dependent soil properties and the control of cut off frequency for analysis, the free field responses obtained as functions of depth were reasonable.

QUESTION 220.9 It is noted that ABWR is designed for 60 year life versus the 40 year life for plant design in current regulation. From the point of view of structures, provide your justification for the longer L I

plant life. L I

E i

Amnammi 3 20 5 f i

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ABM nasimar Standard Plant am n 1

RESPONU,E 220.9

! The limiting factors on a structures life are external events that exceed the plant design basis.

For external events, the ABWR design uses a minimum of a 100 year return period for all safety related structures and a 50 year return period for all non safety related structures. This will provide adequate coverage for a 60 year life.

QUESTION 220.10 j

Since the containment is integral with the reactor building, the follo*ing are staffs concerns:

~i (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 con-tainment, especially on the ultimate capacity of the containment. (The staff has not received Chapter 19 which is believed to contain the estimate of the ultimate capacity).

4 (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, proside a description of the tests to be performed.

RESPONSE 220.10 4

This response will be prosided by April 30,1989.

QUESTION 220.11

? In section 3.8.4.3.1.2 it is noted that the main reinforcement in the suntainment wall consists of

, inside and outside layers of hoop and vertical reinforcement and radial bars for shear

! reinforcement. It appears that eo diagonal seismic reinfore: ment is used. Indicate how the i tangential shear due to horizontal carthquake is to be resisted.

J RESPONSE 220.11 The tangential shear will be resisted by the hoop and vertical reinforcement bars as per the draft ASMF..r%Ie Section til Div II, Subsection CC. Reference to this code will be provided in Tab!c 1.8-21 (this table will be prosided by March 31,1989).

i

QUESTION 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 integral 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 otherwise.

A#nendenent 3 20 3-

ABWR maar Standard Plant nrv n RESPONSE 220.12 (1) GE agrees that not all the load combinations under (1) and (2) of 3.8.4.3.1.2 need to be analyred and examined if a simple comparison can weed out the load combinations which are obsiously not controlling.

(2) lsince the reactor building (RB) is integral with the containment, the finite element model as shown in Figure 3.813 for structural analysis has integrated with RB structure, pedestal, basemst and the containment vessel as a complete model. The load combinations for the containment design have been applied for this integral model. The critical sections for various parts of the RB and the containment vessel are shown in Figure 3.814. The effects due to structural integration are included in the finite element model as well as in the load combinations. The detailed structural evaluation will be given in Appendix 311 which will be provided by January 30,1989 QUESTION 220.13 The terms, 01, Gr and G all as defined in section 3.8.1.3.1 are not listed in table 3.8..I while the terms Iv and ALL listed in Table 3.8.1 are not defined. Clarification of the table is requested.

RESPONSE 220.13 Tables 3.81,5 and 6 have been revised to make the notations consistent with those defined in Subsection 3.8.1.3.1.

QUESTION 220.14 in table 3.fs 5 for load combination No. 3, it appears the acceptance criterion should be changed to S from U unless justified otherwise.

RESPONSE 220.14 The S for load combination No. 3 was a typographical error and it has been changed to S in resised Table 3.8 5.

QUESTION 22015 Discuss the potentials for sesere accident that can be caused by externalinitiators such as high wind, tornado,tsuriami, and carthquakes, and specifically flood since the reactor building has a standard soil embedment of 85 feet.

RES PONSE 220.15 This respome will be prosided by April 30,1989 QUESTION 241.1 Table 2.01 in the Advanced BWR Standard Safety Analpis Report (SSAR) gives an envelope of ABWR plant site design parameters. This table gives the minimum bearing capacity and the minimum shear wave selocity of the foundation soil. The table also gises the values of SSE and OEE and indicates (a) that the SSE response spectra will be anchored to Regulatory Guide (RG) 1.60, and (b) that the 20 3 Amendment 3

ABWR mama Standard Plant RFV II SSE time history will envelope SSE response spectra. The following additional information/clarifica-tion 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 that 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. The SSAR gives an OBE (PGA) value of 0.10g and states that,'for conservatism, 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 $0 Appendix A and paragraph V of 10 CFR 100 Appendix A which require,in part, that for scismic design considerations the OBE shall be no less than one half of the SSE.

c. The SSAR should indicate the procedures that would 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.*

RESPONSE 241.la in conjunction with a SSE of 0.3g anchored to Regulatory Guide 1.60 response spectra, the ABWR scismic design bases includes eight conditions (see Section 3A.1 of Appendix 3A) to be satisfied by utility applicants referencing the ABWR design. In addition to satisfying these eight conditions, at any site where the ABWR Standard Plant is to be used, site specific geotechnical data will be developed by the applicant referencing the ABWR design and submitted for review by the NRC staff to demonstrate comparability with the design analyses assumptions. Satisfying the eight conditions and the review and acceptance of the site specific geotechnical data by the NRC staff will provide assurance that the site specific geological and seismological factors are consistent for the ABWR seismic design bases.

RESPONSE 241.lb The relationship between the magnitude of the OBE and the SSE established in paragraph V of 10 CFR 100 Appendix A is inconsistent with their definitions. The OBE is defined in 10 CFR 100 as the carthquake which could reasonable be expected to affect the plant site during the operating life of the plant, for which those features necessary for continued safe operation of the plant are designed to remain functional. The SSE is based upon the maximum earthquake potential which produces the maximum sibratory ground motion for which certain structures, systems, and components are designed to remain functinnal. In coupling the events, as implied by the current regulatory requirement, the intent of the OBE as a reasonably likely event is lost. The use of a 100 year recurrence level of the OBE is appropriate compared to the plant life and is also appropriately conservative relative to the Uniform Building Code requirements for non safety related structures.

Decouplin;the OBE from the SSE has been an issue in the technical con,munity for quite sorae time.

Both industry and regulatory have recognized the inconsistency in the definitions and some of the undesirable results such as greatly stiffened structures and systems to ircet the more restrictisc OBE stress levels.

Generic issue 119.3, Decoupling of the OBE from the SSE, was introduced into the regulatory process by recommenhtion A 3 of the Piping Resiew Committee. It. the historical background of this generic issue, it is noted that in developing the current regulations,it was assumed that the OBE would serve as a separate check of those systems where continued operation was desired at a lower level of ground motion. Ilowever, in practice, the assumed load factors, damping, stress levels, and Amadmot3 20A

2M6100AT Standard Plant m.g service limits have caused the OBE, rather than the SSE, to control the design for many systems including concrete and steel structives and nuc! car piping, in addition, seismic design for OBE accounts for ct.rtain safety related factors such as fatigue and seismic anchor movement that ste not considered in the design for the SSE. As a further consequence, structures and systems have been greatly stiffened to meet restrictive OBE allowable stress levels. This stiffening is detrimental to actual plant conditions, The actions required to resolve this issue conclude: (1) rulemaking to amend and revise Appendix A to 10 CFR 100 to permit decoupling of the OBE and SSE and to incorporate the use of probabilistic methodology in earthquake design; (2) revising and developing Regulatory Guides;(3) updating pertinent sections of the SRP; and (4) advising various code committees to revise appropriate codes and guides to tellect changes in the regulations.

The following is the NUREG 0933 safety significance statement for this generic issue: 'There is no technical basis for coupling the OBE with the SSE. Designing the pipings systems to the SSE is the primary means of casuring safety. Additional margin is provided by specifying the OBE and thus the level at which inspections will be required before continued operation would be permitted The enore realistic apptcach of using specific probabilities (return periods) for OBE and the decoupling of the OBE levels and frequencies from those of the SSE will allow assurance of public safety to be placed on a more rational basis."

Since the ABWR is a plant of the future and there is sufficient evidence that more flexible designs can exhibit reliability and' safety levels equal to or greater than original stiffer designs (such as piping designs studied by (NUREG/CR 4263), GE intends to continue to pursue this lower I magnitude OBE.

RESPONSE 241.le One of the eight conditions of Section 3A.1 of Appendix 3A requires that there is no potential for l liquefaction at the plant site due to OBE and SSE as redewed and concurred with by the NRC staff.

! It further requires that the liquefaction potential of the foundation and site soils will be investigated and reported for a long duration, New Madrid. type carthquake. A footnote to this effect has been added to Table 2.0-1.

QUESTION 251.12 Criterion $1, Fracture Prevention of Containment Pressure Boundary, is only applicable for containments made of ferritic materials, Since the ABWR containment is made of concrete, this j section should clarify the applicability of Criterion 51 to the ABWR containment. (3.1.2.5.2.1)

RESPONSE 251.12 The primary containment vessel (PCV) for the ABWR plant is a reinforced concrete structure with fertitic parts. Criterion $1, Fracture Prevention of Containment Pressure Boundary, is applicable for containment made of ferritic materials. Thus Criterion 51 is applicable to the removable drywell head, personnellocks, equipment hatches and penetrations which are made of ferritic materials.

Subsection 3.1.2.5.2.2 has been clarified accordingly.

QUE.STION 251.13 This section must include a discussion of all potential turbine missiles and mechanisms of missile generation. The turbine missile discussion should include failure of turbine discs and blades.

(3.5.1.1.1.3) 20 S Amadant 3

MN 23A6100AT arv.n staandard Plant RESPGNSE 251.13 Response to this question is provided in revised Subsections 3.5.1.1.3, 3.5.1.4 and 3.5.4 and new i Figure 3.5 2.

QUESTION 251.14 This section must include a discussion of a favorable turbine orientation or provide a discussion on maintenance of the main steam turbine to protect against turbine missiles. (3.5.4.2)

RESPONSE 251.14 I

As discussed in the response to Question 251.13, the turbine generator tilacement and orier.tation of the ABWR Standard Plant now meets the guidelines of Regulatory Guide 1.115.

QUESTION 251.15 Leak Before. Break (LBB) . The staff considers LBB evaluations to be plant specific because parameters such as potential piping degradation mechanisms, piping geometry, materials, fabrication procedures, loads and leakage detection systems are plant specific. Therefore, the detailed LBB J

analysis should be provided when an application references the ABWR design (3.6.3)

RESPONSE 251.15 i A detailed piping design is not part of the ABWR Standard Plant Scope. The overall layout of the plant is based on the postulated rupture approach which maximlres the piping related spacing requirements. GE is seeking NRC approval of Appendices 3E and 3F which encompass the leak before break (LBB) methodology that an applicant referencing the ABWR design will utillic in his detailed piping design. The NRC staff will review the applicants piping design to ensure that complies with the pre approved LD3 methodology.

QL'ESTION 410.1 Section 3.5.1, ' Missile Selection and Description,' states: 'The missile protection criteria to which the plant has been analped comply with the intent of 10CFR 50 Appendix A. General Design Criteria for Nuclear Power Plants.' Provide a list of those instances where the protection criteria are in strict compliance with 10 CFR 50 Appendix A, and those instances where the protection criteria comply only with its ' intent.' Provivt v. explanation of and justify the acceptability of those

) missile protection criteria which are in compliance only with

• intent' of 10 CFR 50, Appendix A.

J (3.5.1)

RESPONSE 410.1 As stated Subsection 3.1.2.1.4.2, Evaluation Against Crlieria 4, the design of essential structures, systems, and components are meets the requirements of Criterion 4 of the General Design Criteria (GDC). The phrase 'the intent of' included in the second paragraph of Subsection 3.5.1 was included in recognition that the GDC are subject to a variety of interpretations and conformance to a particular criterion is not directly measurable (See Subsection 3.1.1 for further discussion). The ABWR design does in fact meet the requirements of Criterion 4 and Subsection 3.5.1 has been revised accordingly. The resised subsection also appropriately limits it compliance statement to Criterion 4, Environmental and Missile Design Bases.

20 3 Amendment 3

1 . ,

ABWR m-r Standard Plant ium n QUESTION 410.2 Section 3.5.1 states: 'A statistically significant missile is defined as one which could cause unacceptable plant consequences or violation of the guidelines of 10 CFR 100.* Provide an explanation of ' unacceptable plant consequences.' (3.5.1)

RESPONSE 410.2 Unacceptable plant consequences are those consequences that could lead to one of the following:

(1) loss of coatalnment function.

(2) Interfere with achicsing and maintaining safe plant shutdown conditions.

(3) leading to offsite exposures exceeding 10 CFR 100 guidelines.

These consequences are implicit in the criteria given in Subsection 3.5.1 adopted to provide an acceptable design basis for the plant's capability to withstand the statistically significant missiles postulated inside the reactor building.

QUESTION 410.3 Section 3.5.1.1,' Internally Generated hilssiles putside5 Containment)* states: "Failure rates (P1) for value bonnets are in the range of 10 to 10 per year.' Provide a reference or analysis in support of the above statement. (3.5.1.1)

RESPONSE 410.3 Response to this question is prosided in revised Subsection 3.5.1.1(1), and Subsection 3.5.5.

QUESTION 410.4 Regarding the physical separation requirements, provide a list of all systems (required for safe l shutdown, accident prevention or mitigation of consequences of accidents) whose redundant trains do not have missile. proof barriers, and include the minimum separation distances. Provide, for the I limiting case of the minimum separation distance, an analysis demonstrating the acceptability of the approach of not calculating I'2, and instead relying on the 'estremely low' probability of a missile strike to both trales, or a missile from one irain striking the redundant train. (3.5.1.1)

RESPONSE 410.4 The response will be prcnided by April % 1989 QUESTION 410J Explain how safety related systems or components are protected from missiles generated by son. safety related components. It is the staff's position that missiles generated from nonsafety related components should not impact safety related components since a single active failure is assumed concurrent with the missile. (3.5.1.1)

RESPONSE 4103 This response will be prosided by April .% 1989.

20.3 AmendmeM 3

o .

ABM 2suiour Standard Plant nry n QUESTION 410.6 Discuss the means by which stored spent fuelis protected from damage by internally generated

, missiles. (3.5.1.1) [

RESPONSE 410.6 This response will be prosided by April 30,1989.

QUESTION 410.7 Section 3.5.1.1.1.4, 'Other Missile Analysis,' discusses the example of analysis of a containment ,

high purge exhaust fan for a thrown blade. Provide the details of this analysis, such as the maximum l 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. (3.5.1.1)

I RESPONSE 410.7 t

Because the ABWR is a standard plant, the specific details will only be available following procurement of equipment on a specific application referencing the ABWR design. Therefore, when  !

performing potential missile assessment analysis, representative equipment is selected with judgement applied for worse case analysis. The containment high purge exhaust fan is judged to represent a worst case analysis. The containment high purge exhaust fan is judged to represent a worst case  !

j analysis.

The containment high purge exhaust fan from the standard BWR 6 plant was analyred for a thrown blade a'. rated speed conditions as stated in Subsection 3.5.1.1.1.4 It was determined that the t maximum thickness this blade could penetrate was conservatisely 0.12 inches. Since the fan casipg is l 10 gage (0.134 inches), the blade would not escape and consequently 3P is less than 10' per year.

QUESTION 410.A l

Regarding Section 3.5.1.1.2.2, ' Missile Analysis, ' provide the details of the rack, strap and 1 cover assemble design for the pheumatic system air bottles, showing the thickness of the sicel cover e and the distance to the concrete slab. (3.5.1.1)

RESPONSE 410J i

Because the ABWR is a standard plant most of the specific details requested will only be available

following procurement of equipment on a specific application referencing the ABWR design. The statement under item (3), Pressure l'essels, of Subsection 3.5.1.1.2.2 indicating that the bottles are not considered a credible source of missiles is based on a qualitative rather than a quantitative analysis, item (3) of Subsection 3.5.1.1.2.2 has been resised accordingly.

QUESTION 410.9 Regarding Section 3.5.1.1.3, Missile Barriers and Loadings, provide a list of all local shields and barriers outside intended to mitigate miselle effects, giving their specific locations and design data. Provide an example of an analysis showing that the design of the shield or barrier will l withstand the most energetic missile which could credible impact it. (3.5.1.1)

RESPONSE 410.9 This resportse will be provided by April 30,1989 i

! Amendment 3 N.3 r

e

F ABWR wim I

Standard Plant g,y ,,

i QUESTION 410.10 Section 3.5.1.2.1, Rotating Equipment (which can contribute to internally generated missiles inside the containment, 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 potential missiles.' Provide the details of this analysis. (3.5.1.2)

RESPONSE 410.10 The similar analysis referred to in Section 3.5.1.1.1 is the breaking apart of a fan blade driven at constant speed by a synchronous motor in the containment high purge exhaust fan of a standard BWR 6 plant Refer to the response for Question 410.7 for the details of this analysis.

QUESTION 410.11 Regarding Reactor Internal Pump (RIP) motors and impellers which can contribute to Internally generated missiles inside the containment, explain the bases for concluding that the RIPS are locapable of achieving an overspeed condition and that the motors and impellers are incapable of escaping the easing and the reactor vessel wall (SSAR Section 3.5.1.2.1). Your response should explain how the provision of an anti rotation desice at the bottom of the RIP motor which prevents backward rotation of the RIP will present its overspeed during the course of a LOCA or during normal plant operation when one RIP is stopped and the other RIPS are operating (see SSAR Section 5.4.1.5).

(3.5.1.2)

RESPONSE 410.11 The potential for overspeed and generation of missiles by the ABWR reactor internal pumps (RIPS) has been studied and concluded that oserspeed does not impact the safety of the plants as explained below.

The rescrse overspeed is prevented by a mechnical backstop. antireverse rotation device, (ARD).

This backstop functions like a one.way clutch. Should these devices not function, the maximum achievable speed is limited by the differential premiure (llEAD) across the RIPS and the hydraulic design of the pumps.

In the event of the normal operation, with one pump out of service, the idle pump is exposed to a differential pressure of 1.0 times the pumps rated head. This differential pressure would try to rotate the idle pump backward. The corresponding runaway speed of the idle pump (in absence of the ARD)is limited to 100% of the rated speed. Since the rotational stresses vary as square of the speed, one can estimate the stresses in the rotating part would increase by 225%. This value is well within the safety factor for pumps and motors which are designed for deflection and high cycle fatigue. Consequently, fracture and bursting of the pump impeller and motor rotor and missile generation are unlikely, in the event of a large LOCA (pipe rupture and blowdon) the RIP will be exposed to a momentary pressure differential which is 1.38 times the RIP normal operating head. This peak pressure differential decays in 200 seconds. The aserage differential pressure which would force the RIPS to rotate backward is 81% of the pumps rated head. The backward drising head is less than the head for the condition of the one. pump idle, described above.

Despite the above justifications an *1mpeller hiissile* study has been performed and it has been demonstrated that missiles from the impeller do not have the capability to penetrate the reactor pressure vessel or the shroud walls.

20 3-Amndant 3 e

ABM 2w6:=r mya Samadard Plant  !

i la summary, the RIPS are equipped with ARDs which mitigate backward overspeed. Even if ths de. l vices were ignored the maximum achievable overspeeds do not impose a safety problem, because of the duration of the events, inherent strength of the pump, RPV and shroud, and the hydraulic characteris. i tics of the pumps.

QUESTION 410.12 Regarding pressurized components, proside justification for the statement,'FMCRD mechanisms are  !

not credible missile sources,' made in Section 3.5.1.2.2.

l ItEsPONSE 410,12 An explained in Subsection 3.5.1.1.2.1 the pressure boundary containing the FMCRD me:hanisms has  !

been evaluated against the design criteria in Subsection 3.5.1. The pressure boundary including the .

bolted flange connections are stressed below the ASME Code limits and meet allits requirements.

Furthermore, for conservativeness and to prevent the control rod drop accident, internal restraints  !

are provided to support the FMCRD housing in the hypothetical event that the housing.to the.norile ,

weld falls or the housing fails. Therefore, it is concluded that the failure of the pressure boundary containing the FMCRD mechanisms is incredible and the FMCRD mechanisms are not credible  ;

missile sources.

QUESTION 410.13 Regarding Section 3.5.1.2.3, Misrile Barriers and Loadings, provide the same data for I internally generated missiles inside the containment, as that requested utider Question No. 410.8 [

above. (3.3.1.2) r RESPONSE 410.13 [

F As indicated in the responses to Question 410.9, because the ABWR is a standard plant most of the [

specific details requested will only be available fellowing procurement of equipment on a specific i application referencing the ABWR design. The statement that penetration of the containment walls, ,

floors and slabs by potential mistiles is not credible is based on previous BWR licensing applications.  ;

QUESTION 410.14 1

Clarify whether secondary missiles generated as a result of the impact of primary missiles have I been considered. Explain how protection against credible secondary missiles is provided. (3.5.1.2)

RESPONSE 410.14 The only credible source of secondary missiles is from the formation of concrete fragments on the l Impact of primary missiles with structural walls ard slabs. This consideration has been added to l Subsection 3.5.1.2.3. In addition, an interface requirement (referenced in Subsection 3.5.1.2.3) has i been added as Subsection 3.5.4.4 requiring that protection against such missiles be demonstrated. [

QUESTION 410.15 t

P egarding Section 3.5.1.2.3, ' Evaluation of Potential Grasitational Missiles inside Containment' f item 3,' Equipment for Maintenance,' describe any interface requirements imposed by this item on f applicants referencing the ABWR. (3.5.1.2)

Assiestameent 3 M h

_ - _ . _ __ __ _ _ _ _ . - - . _ - -_ o

ABM iminaar myn i Ssandard Plant i RESPONSE 410.15 There are no interfaces requirements imposed by item 3 of Subicetion 3.5.1.2 on applicants referencing the ABWR design since all of the equipment for maintenance is within the scope of the j ABWR Standard Plant. Refer to Subsections 9.1.4 and 9.1.5 for additionalinform:. tion.

1 QUESTION 410.16 j Regarding missiles generated by natural phenomena, provide the details of the tornad e. missile  ;

analysis performed, identifying the tornado region (as defined in RG 1.76) and the missile nectrum. )

Discuss the compliance of the analysis with NUREG.0800, Section 3.5.1.4 acceptance e.riteria;  !

Regulatory Guide 1.76, Positions C,1 and C.2; and Regulatory Guide 1.117, Positions C.1 through C.3 l 3

1 (3.5.1.4) l RESPONSE 410.16

?

This response will be prosided by April M 1989  ;

i j QUESTION 410.17 I

j Provide specific descriptions of all provisions made to protect the chatcoal delay tanks against ,

externally generated tornado missiles. Discuss any interface requirement imposed by these design l t

provisions.

2 l I

i RESPONSE 410.17 t

! This response wilhe prosided by April M 1959 [

1. I l

) QUESTION 410.14 f

Regarding SSC to be protected from externally generated missiles, discuss compliance with l NUREG 0800, Section 3.5.2 acceptance criteria; Regulatory Guide 1.13, Position C2; Regulatory Guide i 27, Positions C2 and C3; and Regulatory Guide 1.117, Positions C1 through C3, (3.5.2) f

] {

l q

i RESPONSE 410.lg [

1 Compliance with Regulatory Guide 1.13, Position C2; Regulatory Guide 1.1.1, Positions C1 through  !

C3 and the corresponding portions of NUREG.0800, Section 3.5.2 acceptance criteria are assund by i

I

! housing all AF.WR Standard Plant safety.related systems and components in buildings or H.uctuies which are designed as tornado resistant. It should be noted that ele view and spent fuel .winge system and the diesel generators are located in the reactor buildtag. Since the ultimate heat sink  !

t

! is not within the scope of the ABWR Standard Plant, compliance with Regulatory Guide 1.27, Positions C2 and C3 and the corresponding portion of NUREG 0800, Section 3.5.2 acceptance criteria will be [

demonstrated by the applicant referencing the ABWR design. This has been added as a specific [

interface requirement la Subsection 3.5.4.1. l 1 QtTSTION 410.19 Clarify whether all nonsafet).related SSC, that may adversely impact (as a risult of their failure due to an etternal missile) the intended safety function (i.e. achieving and maintaining safe {

shutdown, mitigating the consequences of an accident or preventing an accident) of a safet) related i l

i SSC, are protected from esternal missiles. Describe how such SSC are protected. (3.5.2) {

s i

1 i

20.3- f i Amadmat 3 i

- - - - - - - . _ - _ _ _ _ . __. _--_____--___-m

l .. . j MN 2W100AT  !

mandard Plant nry n RESPONSE 410.19 nis response will be prosided by April 30,1989.

QUESTION 410.19e SSAR Section 3.5.1.3.2.2,' Separation,' relies on physical separation between redundant essential systems incluriing their related auxiliary systems as the basic protective measure against the dyna.nic effects of postulated pipe failures. The general arrangement drawings (e.g., Figure 1.2 2) are scheduled to be submitted in December 1988. Note that additional information on Section 3.6.1 may be requested as a result of the resiew of the above drawin6s. (3.6.1)

RESPONSE 4tt.19e ne general arrangement drawings are now scheduled to be subst.itted by January 30,1989, it is anticipated that additlowl information on Section 3.6.1 regarding physical separation may be requested by the NRC staff following review of these drawings.

QUESTION 420.20 l Section 3.6.1.1.1, ' Criteria,' states that the overall design generally complies with BTP ASB l 31. Specify those criteria whic'4 are in strict compliance, and those which a'e not in st ict compli.

ance with the BTP. Also, provide justification for the items that are not in strict et mpliance.

(3.6.1) j RESPONSE 420.20 The design for pipe break event protection is now in compiiance with NRC Branch "echnical Positions ASB 31 and MEB 3-1 with the commitment to the non mandatory Appendix B of ANSI /ANS 58.2, De first paragraph of Subsection 3A1.1.1 has been resised accordingly.

QUESTION 410.21 Provide a listing of all the moderate.cnergy piping outside the containment, but within the scope of ABWR Also, describe how safet) related sptems are protected from jets, flooding and other adverse cavironmtntal effects that may result from pipe failures in moderate energy piping sptems.

j (3.6.1)

) RESPONSE 410.2I l

nis response will be prosided by April 30,1989 s

QUESTION 410.22

]

Justify the non inclusion of pipe f ailure analyses for the Process Sampling System. T!re Prctection Splem,ilVAC Emerger.cy avalirg Water Sptem and the Reactor Building Coolitig Water Sptem as related to the Ultimate liest Sink. Provide a summary table listing the protective measures 2

provided against the eft u ' postulated pipe failures in each of the above sptems and the sptems listed la SSAR Tab!cs 3.t 2 , 16 4. (3.6.1)

RESPONSE 410.22 Dis response will be prosided by April 30,1989 20 1 Amadmeat 3

4

1. . '

ABM u4simar Standard Plant nev. n QUESTION 410.23 ,

Give details for the worst case flooding arising from a postulated pipe failure and ioclude the mitigation features provided.' No'e that for flooding analysis purposes, the complete failure of

~

non seistnie Category I moderate-energy piping systems should bc considered in lieu of cracks in determining the worst case flooding conditir;n. (3.6.1)

- RESPONSE 410.23

'iltis response will be provided by April 30,1989.

QtASTION 410.24 Identify all the high en;rgy piping lines outside the containment (but within the ABWR scope), the adverse effects that may result from falls 's of applicable lines among them, and the protection provided against such effccts for each of such lines (e.g., barricts and restraints). (3.6.1)

RESPONSE 410.34 This response will be prosided by April 30,1989.

QUESTION 410.25 Clarify whether the reactor building steam tunner is part of the break exclusion boundary. Also, provide a subcompartment analysis for the steam tunnel. Discuss how the structuralintegrity of the tunnel and the equipment in the tunnel are protected against failures in the tunnel. (3.6.1)

RESPONSE 410.25 This response will be provided by April 30,1989.

QUESTION 410.26 State how the MSIV functional capability is protected. (3.6.1)

RESPONSE 410.26 This response will be prosided by April 30,1989.

QUESTION 410.27 Provide a summary table of the findings of an analysis of a postulated worst case DBA rupture of a high or moderate energy line for each of the following areas: 1) RCIC compartment,2) equipment and valve room,3) other applicable areas outside the containraent (e.g., housing R11R piping). (3.6.1)

RESPONSE 410.27 This response will be provided by April 30,1989.

QUESTION 410.28 Clarify whether protection foi safety related systems and components against the dynamic effects

, of pipe failures include their enclosures in suitable desigt atructures or compartments, drainage systems and equipment environmental qualification as required. If so, give typical examples for the above type of protection. (3.6.1)

Amendm'nt 3 Mk

9 F-MM 23A6100AT Rrw. ri Standard Plant RESPONSE 410.28 This respone will be provided by April 30,1989.

QUESTION 410.2?

Regarding interfaces (Section 3.6.4.1), include results of analyses of moderate energy piping fail-ures (currently, the interface requirements .iddress only the high-energy piping failures analyses).

.(3.6.1)

RESPONSE 410.29 The information required by Subsection 3.6.2.5 of Regulatory Guide 1.70 pertains to dynamic analyses applicable to high and moderate energy piping systems resulting from pipe breaks and cracks. However, Branch Technical Positions ASB 31 and MEB 3-1 only require dynamic arealyses of high energy piping system postulated pipe breaks. The leakage cracks postulated for moderate energy piping systems are included to evaluate concerns associated with such events such as flooding and compartment pressurization. Therefore, there are no applicable dynamic analyses applicable to moderate energy piping systems.

QUESTION 410.29a Appendix 31,' Equipment Qualification Environmental Design Criteria," is scheduled to be submitted in December 1988. Note that additional information may be requested based on review of the above appendix. (3.11)

RESPONSE 410.29a Appendix 31 is now scheduled for submittal by March 31,1989. It is anticipated that additional information may be requested by the NRC staff following review of this appetriix.

QUESTION 410.30 Al'S ; ugh there are no detailed equipment qualitication requirements for safety related r.iechanical egrq.nent in a harsh environment, GDC 1,'Ouality Standards and Records," GDC 4, ' Environmental Missile Design Vases," and Appendix B to 10 CFR 50,"Quality Assurance Criteria for Nuclear Power Plants and Fuel Processing Plants" (Sections Ill, "Design Control,* and XVII, "Quality Assurance Records') contain the following requirements related to equipment qualification:

l a) Components shall be designed to be compatible with the postulated environmental conditions, including those associated with LOCAs.

! b) Measure shall be established for the selection and review for suitability of application of 1 materials, parts, and equipment that are essential to snfety related functions.

c) Design control measures shall be established for verifying the admvacy of design.

.) Equipment qualification records shall be maintained ara er.H irwtude the results of tests

and material analyses.

Clarify whether the design complies with all the above requirements for safety n.leted mechanical equipment in a harsh ensironment within the ABWR scope. Providu justification for the noi compliance items above and identify any interface requirements needed to comply with the above. (111) l 20.3 Amendment 3 I

ABWR mmour Standard Plant arm a RESPONSE 410.30 The ABWR design complies with all the above requirements for safety related mechamcal equipment in a harsh emisonment within the ABWR scope.

While there are no formalized NRC equipment qualification requirements for safety related mechanical equipment in a harsh emironment GE has succeAifu'4 wpjuted nuou: NTOL epplic=!: !=

developing a mechanical equipment qualification program which is consistent with the NRC's position on mechanical equipment as stated in SRP 3.11. The mechanical equipment qualification program to be applied to the ABWR will use applicable portions of the NRC approved Licensing Topical Report NEDE 243261 P and Regulatory Guide 1.89, Revision 1; and will be consistent with the program for qualification of mechanical equipment in a harsh emironment described in the NRC approved GESSAR 11 design. The ABWR program . cope looks not only at the metallic components of the equipment but also the normetallic components. Metallic components which form a pressure boundary are considered to be qualified by the nature of their pressure retention capability as demonstrated by the application of an ASME Boiler and Pressure Vessel Stamp. Nonmetallie, such as greases, gaskets, lubricants, etc.,

i wiil be shown to be capable of performing their intended functions under accident environments. The design of safety related mechanical equipment associated with the ABWR will be performed under the same internal procedural controls as that used for the design of mechanical components associated tith th( GESSAR 11 design. These controls assure that components are designed to be compatible with their postulated operating environments, that measures are established for the selection and review of the suitability of application of the rna': rial, parts, and equipment that are essential to safety related functions, and that there are design control measures for verifying the adequacy of the design. As stated in NEDE 243261 P a complete set of qualification records are developed for each safety related component.

Since the ABWR Standard Plant comprises the entire scope of safety-related equipment there are no interface icquirements in this area.

QUESTION 451.1 What are the bases (including references) for the site envelope of the ABWR design meteorological parameters listed in Table 2.017 Are these values intended to reflect the indicated maximum historical values for the contiguous USA? What is the combined winter precipitation load from the addition of the 100 year snow pack and the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> probable maximum precipitation? What is the duration of the design temperature and wind speed values? What gust factors are associated with the extreme winds? Are any other meteorological factors (e.g., blowing dust) considered in the ABWR design?

GESPONSE 451.1 A detailed description of all site characteristics is not practical for a standard design which is not based on a specific site location. Ilowever, it is possible to define an envelope of selected site related parameters which will blanket the majority of potential nuclear sites in the conterminous United States. This envelope of site related parameters establishes the conditions of phenomena which the ABWR Standard Plant is designed to accommodate. These characteristics and the specific bounding values were chosen after reviewing the corresponding parameters used in recently licensed plants and potential nuclear sites, provide the bases for design of the ABWR Standard Plant. There are no other meteorological factors considered in the ABWR design except for the determination of atmospheric dispersion factors which follow the guidance of Regulatory Guides 1.3 ,

and 1.145 as described in the response to Question 451.2.

Amendment 3 20 S

23A6100AT Standard T'lant anv. n QUESTION 451.2 Short. term dispersion estimates for accidental atmospheric releases are not provided explicitly in Section 2.4.3. If you X/O values which are listed in Chapter 15 represent an upper bound for which the ABWR is designed; what is the bases for thAt selection?

RESPONSE 451.2 The X/O values found in Chapter 15 do not necessarily form an upper basis for evaluation of design basis accident conditions. Since there is no detailed regulatory guidance on standard plant site parameters, recourse was made to existiag regulatory guides. The X/O values found in Chapter 15 for rifsite dose analysis were calculated using a subroutine of the CONAC03 code (Reference 2, Subsection 15.6.7) using the methodology found in the Murphy /Campe paper (Reference 4, Subsection 15.6.7).

Amendment 3 M 3-

)'

MM 23A6100AT Standard Plant _ RIN. B TABLE 20.3 5 VERIFICATION MATRIX FOR SASSIO1S (Response to Question 220.3)

Verification Problems .

Areas of Verification 1 2 3 f .5 6 2 8 2 10 11 12 13 If A) Finite Element Library:

1. 3D spring element x x

2. 3D beam element x x

3. 3D solid element x x

4. Plate /shell element x x .

5. Stiffness / mass elemen* x

6. Response x B) Impedance Anahsis:

1. Rigid Surface Foundation on Uniform flalfspace 1.1 Circular Foundation x 1.2 Strip Foundation x 1 Rigid Surface Foundation on Layer Soil System 2.1 Circular Foundation x 2.2 Strip Foundation x

3. Flexible Surface Foundation on Uniform IIalfspace 3.1 Rigid-Flexible Fdn. x 3.2 Totally Flexible Fdn. x C) Seattering Anahsis:

1. Vertically Propagating Body Waves 1.1 Free Field x 1.2 Embedded Foundation x

2. Inclined Body Waves 2.1 Surface Foundation x 2.2 Embedded Foundation x D) SSI Anahsis:

1.1 Sing!c Foundation x 1.2 Multiple Foundation 1.2.13 D Analysis x 1.2.2 2 D Analysis x 20.3-Amendthent 3

ABM 2346ioorr uv. n Standard Plant TABLE 203 6 CLASSI/ASD VERIFICATION PROGRAM

SUMMARY

(Response to Qnestion 2203)

ElR{ MA.IOR OBJECTIVE MAJOR RESULT CONTROL T1 Establish a benchmark run to Frequency dependent Pul;lished impedances for a calculate impedances for a impedances similar to corresponding problem single circular foundation on published results viscoelastic half space full model T2 Check symmetric foundation Frequency dependent Verification problem T1 option using quarter model of impedances identical to single circular foundation those in Tl T3 Check eensitivity of imped- Frequency dependent Verification problem T1 ances to foundation mesh size impedances similar to those in T1 T4 Check accuracy of Green's func- Frequeacy dependent Verification problem T1 tion algorithm by simulating a impedances similar to half space using 2 layer half those in T1 space with homogeneous proper-ties T5 Check impedances for a single Frequency dependcat Published impedances for a circular foundation on a shal- impedances similar to corresponding problem low 2 layer inhomogeneous half published results space T6 Check impedances for a single Frequency dependent Published impedances for a circular foundation on a deep impedances similar to corresponding problem 2 layer inhomogeneous half published results space T71 Check multiple foundation Frequency-dependent im- Run T7-3 T7 2 Irnpedances using a parametric pedances which approach T7 3 study of effects of separation single foundation case distance when r/a T8 Check calculation of soil. Maximum response similar Published responses for a structure interaction response to published results corresponding 2 foundation of 2 foundations problem 19 Check calculation of soil- Masimum response similar Published responses for a structure interaction respocse to published results corresponding single founda-of a single foundation tion problem Amendment 3 20 1

l ABWR msuur Standard Plant nry n i

TABLE 203 6 CLASSI/ASD VERIFICATION PROGRAM

SUMMARY

(Response to Question 2203) (Continued)

T10 Check impedance for a single Frequency dependent Published impedances for a suiface square foundation impedances similar to corresponding problem published results T11 Check impedance for a single Frequency dependent Publish impedances for a embedded square foundation impedances similar to corresponding problem published results T12-1 Check multiple foundation Frequency dependent Run Til impedances using a parametric impedances which ap-study of effects of separation proach single foundation distance case when r/a Amendment 3 20 }

l

ABWR uAn=Ar Standard Plant REV B I l

Table 20,3-7 ISOLATION VALVE ARRANGEMENTS NOT MEETING THE EXPLICIT REQUIREMENTS OF GDC56 (RESPONSE TO QUESTION 430.41)

1. PP ABLE LINES CONTAINMENT ISOLATION VAL /E5 g RHR TEST LINE AND NO. 8 MINIMUM FLOW LINE > M N F LOW S/P g G TEST Vym V

V HPCF/RCIC TEST LINE M NO. 8 AND MINIMUM FLOW gj MIN F LOn LINE y3 V

SP y g ; TEST rm V

V RCIC TURBINE EXHAUST y NO. 8 LINE AND RCIC VACUUY $ fP I

PUYP DISCHARGE LINE g g i

y V

SPCU SUCTION / RETURN y NO. 9 LINE 7 (SUCTCN)

SP l l No.8 (RETURN)

U NA L 2 .

VN

V3 asneu i

AmenJment 3 'D f

l

TABLE 20.3-7 ISOLATION VALVE ARRANGEMENTS NOT MEETING THE EXPLICIT REQUIREMENTS OF GDC56 (Response to Question 430.41)

Continued GE Safety Standard 20 No. 8 This criterion applies to a line with the following characteristics.

a. penetrates containment;

b. communicates with containment interior;

c. is not an instrument line; and

d. is not a suppression pool effluent line.

Each of these lines shall be provided with two isolation valves. At least one valve shall be located outside of the containment, the other valve may be located either inside or outside the containment. Alternatively, one isolation valve outside the containment which is normally closed (or a blind flange) and which does not receive a signal to open subsequent to an accident may be used.

On influent lines having two valves, one may be a check valve, and the valve outside the containment must be capable of automatic, or remote manual closure, or should be normally locked closed. On effluent lines or where a second valve is not provided on an influent line, these valves shall be capable of automatic and remote manual closure, or should be normally locked closed. The valves shall be located as close as practicable to the containment.

GE Safety Standard 20 No. 9 This criterion applies to a line with the following characteristics:

a. penetrates containment;

b. communicates with wppression pool;

c. is . Sol an instrument line; and

d. is an effluent line.

Effluent lines shall be provided with one remote. manual valve outside the containment. The valve shall be located as close as practicable to the containment.

ABWR 23A61ooAr Standard Plant REV D Table 20.3-8 LINE WHOSE CONTAINMENT ISOLATION REQUIREMENTS Acte COVERED BY GDC57 (RESPONSE TO QUESTION 430.44)

LINES CONTAINMENT ISOLATION VALVE -

V TIP SEAL PURGE LINE $I '

=

>C A

. 7 TIP GUIDE LINE g;ll*

Dd ><0 RE ACTOR BUILDING COOLING WATER AND HV AC NORYAL COOLING WATE R SUPPLY LINE IA  :

PCV V V RE ACTOR BUILDING COOLING W ATER AND HVAC NORY AL COOLING W ATER RETURN LINE PCV V

SUPPRESSION POOL VACVUY S.P $> >

BRE AKER TEST AIR SUDPLY f* '

LINE -1 V V SERVICE AIR SUPPLY LINE I LC V V INSTRUMENT AIR SUPPLY LINE HIGH PRESSURE NITROGEN SUPPLY se 549 c1 Amendment 3 M

ABWR u^am^r Standard Plant _ RfW D SHEAR WAVE VELOCITY (m/sec) 0 250 500 750 1000 I I l l l MAXIMUM SHEAR WAVE VELOCITY (ft/ sect ACCE LE R ATION (g) 0 1000 2000 3000 0.1 02 03 0- 0 l l l l l

10 -

~]

l 50 -

k 20 -

1

_ l 5 E I 30 -

joo _

k E

l 40 -

150 -

50 -

- INITI AL OBE 60 -

83 589 0J INPUT TIME HISTORY. GE.H1 M AXIMUM ACCELER ATION. 0.15g (OBE)

Figure 20.3-11 SHAKE ANALYSIS RESULTS -SOIL PROFILE UB1D150 (RESPONSE TO QUESTION 220.8)

Amendment 3

i

.. J MN a3A6100A1 any n Standard Plant SHEAR WAVE VELOCITY (miseci 0 250 500 750 1000 l I I I I MAXIMUM SHE AR WAVE VELOCITY (f t/sec) ACCELE R ATION (g) 1000 2000 3000 0.1 0.2 0.3 0

0- 0 l l l l l l

I 1

10 I

50 -

20 -

E E l 6 30 -

joo _ k S i o o 1

40 -

150 -

50 -

- = === I N IT I A L

- OBE 60 -

88 589 C2 INPUT TIME HISTORY. GE H1 MAXIMUM ACCELE R ATION. 0.159 (OBEi Figure 20.3-12 SHAKE ANALYSIS RESULTS -SOIL PROFILE VP20150 (RESPONSE TO QUESTION 220.8) 20 %

Amendment 3

r 2O N 23A6100AT Standard Plant m. n s ,

t e

PCV_

G g j j V3 i 1

L A Vm ,

88 b894A i

Figure 20.3-13 ISOLATION VALVE ARRANGEMENT OF STANDBY LIQUID CONTROL SYSTEM LINE (RESPONSE TO QUESTION 430.40) i A

i I

[

[

4 I

1 I  !

i i

j AmeMment 3 20.3 e

- - - , - <.-e., .

_. _ . , , . -,,. . . . - .