LD-94-012, Forwards Listed Matl to Close follow-on Questions to Dser Responses Re Sys 80+

Forwards Listed Matl to Close follow-on Questions to Dser Responses Re Sys 80+

ML20067D336
Person / Time
Site:	05200002
Issue date:	02/22/1994
From:	Brinkman C ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20067D341	List: LD-94-012, Forwards Listed Matl to Close follow-on Questions to Dser Responses Re Sys 80+ ML20067D344
References
LD-94-012, LD-94-12, NUDOCS 9403080173
Download: ML20067D336 (252)

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_ ABB February 22, 1994 LD-94-012 Docket No.52-002 Attn: Document Control Desk U.S. Nuclear Regulatory Comaission Washington, D.C. 20555

Subject:

System 80+" Information for Issue Closure

Dear Sirs:

The attachments to this letter provide material to close follow-on questions to DSER responses. Attachment 1 provides minor revisions to S?ction 1.4 of CESSAR-DC in response to a comment from Mr. T. Boyce.

Attachment 2 transmits a revision to Table 3.9-15 as requested from and fTxed to Mr. C. Li on January 11, 1994.

. Attachment 3 presents a copy of information on the Contani.nert Spray and Shutdown Coolirig Systems which was faxed to Mr. S. Sun on February 2,1994.

6ttachment 4 provides a revision to Appendix SE on the Intersystem LOCA issue which vas previously faxed to Mr. D. Terao.

Attachment 5 transmits a revision to Section 9.5.3 on emergency lighting which was requested by Mr. C. Thomas.

Attachment 6 transmits a revision to Section 12.3.1 to address a comment from the ACRS on cobalt content of matcrials in contact with the reactor coolant. This revision should be given to Mr. C. Hinson.

Attachment 7 provides a revisico to Section 14.2.12 which addresses a confirmatory item on relief valve bench testing.

l Attachment 8 presents a response 1o the issue on Safety Injection pump durability under mini-flow conditicas (Section 6.3). This information was faxed'to NRC on January 14, 1994, and should be given to Mr. T. Collins.

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ABB Combustion Engineering Nuclear Power T

Corntuston Ergrwems ir<: >x m a 3 8&1911 94030B0173 940222 wr.osor. c1 oco95

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U.S. Nuclear Regulatory Commission LD-94-012 l February 22, 1994 Page 2 Attachment 9 transmits a copy of a fax to Mr. N. Saltos on fire protection inside containment.

Attachment 10 transmits a revision to the System 80+ Fire Hazards Assessment to make it consistent with CESSAR-DC, Amendment U re,isions to which the NRC staff has agreed. A corresponding change to Sec' ion 1.6 is also included.

Attachment 11 transmits revisions to Chapters 2 and 3 which have been previously provided as a result of meetings with NRC staff. Attachment 12 provides corresponding changes to Chapters 1 and 19. Changes to the Seismic Margins Assessment of Chapter 19 will be provided within the next few days, along with other Chapter 19 changes.

Attachment 13 provides a revised copy of the Software Program Manual for Nuplex 80+ to address a comment from the ACRS. The revision, in Section <

9 of the manual, has been discussed and agreec' to by NRC staff.

CESSAR-DC changes provided above will be printed in Amendment V. In addition to the above revisions, the Severe Accident Mitigation Design Alternatives analysis in Appendix 19A will be revised to be consistent with data in Amendment U. It is understood that the revision to Appendix 19A will not significantly change any result or NRC staff conclusion.

If you have any questions, please call me or Mr, Stan Ritterbusch at (203) 285-5206.

3 Very truly yours, COMBUSTION ENGINEERING, INC.

a e o- tw;a::= -

3 C. B. Brinkman Acting Director Nuclear Systems Licensing CBB/ser ,

cc: J. Trotter (EPRI)

T. Wambach (NRC)

P. Lang (DOE)

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CESSAR En#ication 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4.1 APPLICANT'S QUALIFICATIONS AND EXPERIENCE (Presented in site-specific SAR.)

1.4.2 ARCHITECT-ENGINEER'S QUALIFICATIONS AND EXPERIENCE (Presented in site-specific SAR.)

1.4.3 COMBUSTION ENGINEERING'S QUALIFICATIONS AND EXPERIENCE Combustion Engineering, Inc. (hereafter referred to as C-E, Combustion, ABB Combustion Engineering Nuclear Power or ABB-CE) nuclear power activities are of three general types: design, development, construction and operation of reactor and auxiliary systems; design and fabrication of nuclear components; and, support of design, development and analytical projects.

A summary of the company's efforts, accomplishments, and operating experience in the light water reactor field is provided below.

1.4.3.1 Pre-Commercial Reactor Programs 1.4.3.1.1 Naval Propulsion Program Du:-ing the period 1955 through 1960, Combustion was a major contributor to the U.S. Naval Reactors program. The Company designed and built, at its Windsor, Connecticut site, the prototype of a small attack submarine power plant. This prototype (SIC) went into operation in 1959 and ic ctill being Wa$

operated as a naval training facility. A second plant of this type was also designed and built by Combustion for installation in the USS Tullibee (SSN-597) which has been operated as a part of the Unite.d States nuclear submarine fleet.

In the design, development, construction and operation of the ,

prototype system and the submarine power plant, Combustion's i responsibilities included all safety aspects of the reactor  ;

systems.

1.4.3.1.2 Boiling Nuclear Superheat (BONUS) Plant Combustion was responsible for the nuclear design and for the direction of startup and initial operation of the BONUS plant in l Puerto Rico.

The design of this reactor system presented a number of unique problems, e.g., control and safety analysis of a two-region core, ,

l Amendment N 1.4-1 April 1, 1993

CESSAR nnincum ..

design of a superheater' fuel element, design of a steam control system to assure adequate cooling of superheater fuel under all credible conditions, and design of a containment building of the

" total containment" type to house the entire power generating  !

installation.

The BONUS plant achieved full power operation in September 1965, and was the first nuclear power plant under USAEC control operating with an integral superheating core.

1.4.3.2 Development and Design of Commercial PWR Systems ,

The development and design by Combustion of a pressurized water reactor for utility service dates back to 1958. At that time, i the Company was selected by the AEC to undertake the design, analysis and economic evaluation of a 250 MWe PWR plant, in conjunction with an architect-engineer. This effort provided initial technical and economic guidelines for combustion's commercial development of the PWR.

With a subsequent decision by the Company to concentrate on the development of the PWR for large nuclear power stations, a program was initiated to guide required design and development work along appropriate lines. The following is representative of the types of PWR-oriented work which have been performed:

A. Evaluation of overall plant and systems to establish optimum physical' arrangement and design criteria from the standpoint of economics and safety. Much of this work has been performed in conjunction with qualified architect-engineering organizations; e.lemen4~ -

B. Design and development of nuclear components such as controlA

• assemblies, control and auxiliary systems equipment. 4 assembly elesnent drive mechanisms, C. Extensive testing of PWR nuclear components, such as fuel assemblies and reactor control components, under actual service pressure, temperature and flow conditions.

Combustion Engineering's Nuclear Laboratories- have been engaged in the development and testing of fuels, fuel elements, control assemblies, reactor components and materials for reactor application. Particular emphasis has been given to UO2 and Zircaloy cladding technology, involving both in-pile and out-of-pile investigations. The initial efforts in the

. , laboratories were associated with submarine reactor programs.

rm,snj 6-inoe- 1960, _the personnel of the nuclear laboratories M'/~ ,

80 actively participated in the joint U.S. AEC - Euratom research and development program for fuels development. In addition to il 1.4-2 .

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J these programs, personnel in the Nuclear Laboratories have been-responsible for materials design activities for the HWOCR study and for pressurized water, boiling water, nuclear superheat, and fast breeder reactor systems.

1.4.3.3 Major Component Design and Fabrication During the period of 1955-1961, Combustion Engineering (C-E) was a major supplier of nuclear cores for naval propulsion service.

C-E has fabricated the boiling and the superheating-fuel for the BONUS reactor. The boiling section of the Bonus core was made up of Zircaloy clad, rod type, UO fuel elements ' fundamentally ,

similar to those being utilized id the C-E Standard fuel design.

The superheater fuel utilizes Inconel-clad, rod type, UO fuel elements. The superheater cladding is designed for an opehating temperature of 1250'F.

forWdPMbf Combustion Engineering has performed 4the design engineering and i i

fabrication of control rod drive mechanisms and fuel rods for all of the commercial power reactors listed in Table 1.4-1.  !

Combustion Engineering has fabricated and shipped many reactor i vessels for utility plant service and for naval service.  !

I Additional vessels for plant sizes up through 1300 Mwe are now in service.

~

Combustion Engineering has fabricated nuclear steam ' generators )

for naval service and for all its commercial PWR plants. In l addition, the company designed and fabricated the 10 ' steam  ;

generators in the Hanford Production Reactor facility.

and reaCW Coolahf funtp domyonenfS Combustion / Engineering manufactures reactor vessel internal structuresVat its Newington, N.H. facility.

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• 4 e,n vn,,4 no-rin.v h :tdd L gauf;3ug ;;peilitj g hf ?hr4 cation *f-reactor ccolant pumps by 4*e antry in iT't 7i nte j oint-c': mech i n of the CE/MSB ?"rp Cc=pesy -

1.4.3.4 Facilities The C-E laboratories at Windsor, Connecticut, and Chattanooga, Tennessee, provide complete facilities for the development, design, analysis and testing of PWR components and systems.

These laboratories include equipment for:

l Amendment J 1.4-3 April 30, 1992

1 CESSARnuhou  !

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i A. Mechanical Testing.

B. X-ray and Radiography.

C. Metallography. ,

D. Ceramics Development.

E. Analytical and Radio-Chemistry F. Fuel Fabrication Development ,

G. Corrosion Testing __ -

H. 2500 psi Component Performance Testing of MSC I o Yhd.

hue,l G.brictibon I. 2500 psi and 5000 psi Steam Generation (Oce55 J. Welding Development.

C-E maintains a fuel manufacturing facility in Hematite, Missouri. The Hematite plant is used to convertSUF, gas to UO.,~j

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• ipped Wi nd nen - rnnnoceicne en ccmp1__ the fuel fchrie=*4nn

.procesc, ar decer-ibed belou - Currently,-tha Ha== tite-fccility ic cnmn1neing 3 mn1*i-mt114nn an!1ar i;provc;cnt preg"= te mcdcrnized manufacen"4ng ^perctionc. In-the aceu f u i.ur e u r. ill

-treusfet ull fucl f*"4"*4^n ="t4"4*4a= *^ the :eestite plenu.-

e.lemPAt The Windsor facilities of Combustion Engineering, Inc. are ,

equipped to fabricate, and rovide the necessary quality control for, fuc-1 assemblies, contro assemblies, control tassembly drive mechanisms, and other specialized nuclear components. ~ e.leme.nt Combustion Engineering's Chattanooga Plant includes a separate facility to design, fabricate, and provide quality control for large reactor pressure components. The facility has such special equipment as heavy duty cranes and large capacity machine tools capable of performing work on large, heavy parta. to close tolerances and fine surface finishes. It is also equipped with-the latest testing and quality control equipment, including a linear accelerator for weld examination. ,

1.4.3.5 Commercial Reactor Operation Table 1.4-1 lists all Combustion Engineering Pressurized Water Reactors designed and built to date.

-l Amendment N 1.4-4 April 1, 1993 i i

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q-TABLE 1.4-1 (Sheet 1 of 2)

C-E PRESSURIZED WATER REACTOR PLANTS Commercial Nominal Plant Operator Utility Plant Location Operation Hwe Net Non-System 80 Plants Palisades ' Consumers Power Co. Michigan 1972 800

Maine Yankee Maine Yankee Atomic Power Co. Maine 1972 800

Fort Calhoun Omaha Public Power District Nebraska 1973 475

Calvert Cliffs Unit 1 Baltimore Gas &-Electric Co. Maryland 1974 850 Millstone Point Unit 2 Northeast Utilities Connecticut 197S 865 ,

Baltimore Gas & Electric Co. Maryland 850 A Calvert Cliffs Unit 2 1976 St. Lucie Unit 1 Florida Power & Light Co. Florida 1976 810 -

Arkansas tiuclear One Unit 2 Arkansas Power & Light Co. Arkansas 1980 900 St. Lucie Unit 2 Florida Power & Light'Co. Florida 1983 810 San Onofre Unit 2 Southern California Edison Co. California 1983 1100 San Onofre Unit 3 ' Southern California Edison Co. California '1984 1100 Waterford Unit 3 Louisiana Power & Light Co. Louisiana 1985 1100 a

Amendment A September 11, 1987

1 TABLE 1.4-1 (Cont'd)

(Sheet 2 of 2)

C-E PRESSURIZED WATER REACTOR PLANTS Commercial Nominal Plant Operator Utility Plant Location Operation Mwe Het System 80 Plants Palo Verde Nuclear Arizona Public Service Arizona 1300 A Generating Station Company Units 1 1986 2 1986 3 1988 lB l

Washington Nuclear Project Washington Public Power Supply Washington 1300 l A Unit 3 System g Yonggwang Korea Electric Power Company Republic of g Units 3 Korea 19 W 1000lgEB

,g 4 199E* s U lc.hin korea. Dec}(ic feet biparg Repub)ic, o9 UnUS 5 %orea o%sk /@

N /797 4 A

• Anticipated Commercial Operation I

Amendment E December 30 788

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4 EH ICATION Se[: b$6)Q. hw d.:3, $ - 3 TABLE 3. 9-15 (Bhoot 1 of 80)

INSERVICE TESTING BAPETY-RELATED PUMPS AND VALVE 8 i

(h) (i) l Safdy Test Test Test CFSiAR-DC Pump {

rim Paremeter Du Cponfm Hany No, j CCW PUMP 1A 3 DP SP3 ,SPo, t %i93 mo.

2 16 9.2.2-1.1 CCW PUMP IB DP.SP 3 SPo, V 3 3 mo. 16 9.2.2 1.1 CCW PUMP 2A 3 DP,SP3,SPo, 3 mo. 16 9.2.2-1.7 l CCW PUMP 2B DP,SP 3 ,SP o ,V 3 3 mo. 16 9 1 2-1.7 MD EFW PUMP 1 3 DP.SP3 ,SPo,Q,V 3 mo. 21 10.4.9-1.1 TD EFW PUMP 1 3 N,DP,SP3 ,SPo,Q,V 3 mo. 10.4.9-1,1 21  ;

MD EFW PUMP 2 3 DP SP3,SPo,Q.,V 3 mo. 21 10.4.9-1.1 TD EIN PUMP 2 3 N.DP SP3 ,SPo, c 3 mo. 21 10.4.9-1.1 SI PUMP 1 2 DP.SP3 ,SPo,Q, (44 3 mo. 18 6.3.2-1A S1 PUMP 2 2 DP,SP3 ,SPo,Q, (46 3 mo. 18 6.3.2-1B SI PUMP 3 2 DP,SP3 ,SPo,Q, f,,) 3 mo. 18 6.3.2-1A SI PUMP 4 2 DP SP3 ,SPo ,Q, 6) 3 mo. 18 6.3.2-1B SC FUMP 1 DP SP 2 3 ,SPo,Q, ~' 3 mo. 19 6.3.2 1A SC PUMP 2 2 DP,SP3 ,SPo,Q,V 3 mo. 19 6.3.2-1B CS PUMP 1 2 DP,SP3 ,SPo,Q,V 3 mo. 19 6.3.2-1A CS PUMP 2 2 DP,SP3 ,SPo,Q,V 3 mo. 19 6.3.2-1B SSW PUMP 1A 3 DP.SP,Q,V 3 mo. 17 9.2.2 1.1 SSW PUMP 1B 3 DP,SP,Q,V 3m 17 9.2.2 1.1 SSW PUMP 2A 3 DP SP,Q,V 3m 17 9.2.1 1.3 SSW PUMP 2B DP.SP,Q,V 3 mo.

3 17 9 1 1 1.3 ECW PUMP 1A 3 DP,SP3 ,SPo,Q,V 3 mo. 20 9 1 9-1.1 ECW PUMP 1B 3 DP,SP3 ,SPo,Q,V 3 mo. 20 9 1 9-1.1 ECW PUMP 2A 3 DP,SP3 ,SPo,Q,V 3 mo. 20 9.2.9 1.5 ECW PUMP 2B 3 DP,SP3 ,SPo,Q,V 3 m. 20 9 1 9-1.5 DG BUILDING SUMP PUMP 1A 3 DP,SPc,Q,V 3 mo. 17 9.5.9-1 DG BUILDING SUMP PUMP IB 3 DP.SPc,Q,V 1m 17 9.5.9-1 DG BUILDING SUMP PUMP 2A 3 DP,SPc,Q,V 3 mo. 17 9.5.9-1 DG BUILDING SUMP PUMP 2B 3 DP.SPc,Q,V 3 mo. 17 9.5.9-1 RB SUBSPEERE QUAD A SUMP PUMP 1 3 DP.SPc,Q,V 3 mo. 17 9.3.3-2.1 RB SU'BSPHERE QUAD A SUMP PUMP 2 3 DP,SPc,Q,V 3 mo. 17 9.3.3-2.1 RB SUBSPHERE QUAD B SUMP PUMP 1 3 DP SPc,Q,V 3m 17 9.3.3-2.2  ;

RB SUBSPHERE QUAD B SUMP PUMP 2 3 DP,SPc,Q,V 3 mo. 17 9.3.3-2.2 -

RB SUBSPHERE QUAD C SUMP PUMP 1 3 DP,SPciQ.V 3" 17 9 3 3'2 I RB SUBSPHERE QUAD C SUMP PUMP 2 3 DP,SPc,Q,V 3m 17 9.3.3-2.1 RB SUBSPHERE QUAD D SUMP PUMP 1 3 DP,SPc,Q,V 3 mo. 17 9.3.3-2.2 RB SUBSPHERE QUAD D SUMP PUMP 2 3 DP.SPc,Q,V 3 mo. 17 9.3.3-2.2 SPENT FUEL POOL COOLING PUMP 1 3 DP SP3 ,SPo,Q,V 3 mo. 20 9.1.3 i

SPENT FUEL POOL COOLING PUMP 2 3 DP,SP3 ,FPo,Q,V 3 mo. 20 9.1.3 '

Amendment T November 15, 1993 l

CESSARnab a w .os a q - r h o 4-a-TABLE 3.9-15 (Sheet 60 of 80)

INSERVICE TESTING SAFETY-RELATED PUMPS AND VALVES range of the pump, and RPM measurements are talan to demonstrate pu=p speed control settings correspond with the actual pm.p FPM.

Q (40)-4 &I-390r-4I-391r GI-392rSI-393 -

b Se, f These valYe'a'nmet_open in a severe accident to allow-WL or in the IRWST to O flood the reactor cavity -debffs . To toat those valves, manua1 va1ves upneream mo% cover ot 3o_eto o proyone fio, of wator from th.

IRWST to the HoldapJolu(Tank when thoso alvoaJro opened. Closing the j_ onta manual vntluoda not practical during oporatio'nT uinca. they requiro entry Therefore, thooo valvos will be te st2diIringJold (41) Valves: SI-612, SI-622 , SI-632 , SI-642 , SI-619, SI-629, SI-639, SI-649 These air-operated SIT Nitrogen Presouro Control Valvos are atroked in the courso of plant operation as a control of the SITS at a frequency matter of normal operation and prosaure which satisfies tost requiromonts of quarterly testing. Fail-saf e actuation (FS) on a 3 month baulo, however, is impractical during plant operations (quarterly test frequency) or cold ohutdown of the SITSbecause such testing involves entrion to containment to proximity (high radiation dono and airborno contamination aros) to fail air to the air diaphragm valve actuators. Therefore, the FS tent for those valvos will be performod on a refueling outago basis for ALARA purposes.

(42) Valves: SI-322, SI-332, SI-611, SI-618, SI-621, SI-628, SI-631, SI-638, SI-641, SI-648, SI-661, SI-670 Thane air-operated valves are stroked on a quarterly frequency. Fall-safe (FS) actuation testing on a 3 month basio, however ,is impractical during plant operations (quarterly test frequency) or cold shutdown because such tooting involves entrios to containment to proximity of the SITS (high radiation dono and airborne contamination area) to fail air to the air diaphrag:n valvo actuators. Therefore, the FS test for theon valves will be performed on a refueling outage basis for ALARA purposes.

(43) Although those Emergency Diesel Generator support system cceponents are Safety Clans 3, they are procured, tooted and maintained ao part of the Emergency Diesel Generators themselves, and reliability by the plant Technical Specifications.which are tested for operability Therefore, these components are testod by Technical Specifications Surveillanco Requirements of Technical Specification Section 3.8.

i (44) Pronouro Isolation valvoo (PIva) are not roverso flow tested quarterly, since testing of thoso valvos during power operation would require

) containment entries to high radiation and airborno contamination areas.

PIVs are not reverse flow touted ovary Cold Shutdown, because of the extensive teot equipoent setup which could extend the Cold Shutdown. The

{ RF function is verified, however, by leakage testing each valvo in the roverno flow direction during unit startup for the testing frequency outlined in Technical Specification Surveillanco Requirement 3.4.13.1.

This surveillance requiremont states that leakage tenting of those valves la required every 18 months AND prior to entoring Mode 2 whenever the plant has boon in Mode 5 (Cold Shutdown) for 7 days or more, if leakage tunting has not been performed in the previoup 9 months AND within 24 hourn following valve actuation due to automatic or manual action or flow through the valve (s) .

Adk .rd s e. ,- t' 45 Amendment S September 30,1993

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INSERT 46 (To be inserted in CESSAR-DC Table 3.9-15 in Amendment V)

(46) I'er inservice testing of the safety injection pumps during refueling outages, a walkdown visual examination of safety injection system piping and components outside Containment will be conducted to verify the leak tight integrity of the system.

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NUCLEAR POWER l FAX TRANSMITTAL

!' To: M.X. Franovich or From: J. E. Robertson T. V. Wambach Project Managers l Date: February 2,1994 FaxRef,:

Company : U.S.NRC Company: ABB C-E Dept: Standardization Project Dept: Fluid Systems Fgineering i Directorate I'

Phone: 301-504-1121/1103 Phone: (203) 285-4688

{ Fax: 301-504-2260 Fax: (203) 285-3267  !

I Thispage with 4 pages tofollow.

I I

i Mike / Tom, 1

Attached are marked-up pages from CESSAR DC Tables 5.4.7-2 and 6.5-3, in response to l I

verbal questions from Mr. Summer Sun of the Reactor Systems Branch. The questions relate to consistency between the Shutdown Cooling System and Containment Spray System Failure

! Modes and Effects Analyses (the tables) and the system P&ID (Figures 6.3.2-1 A, B, and C). l

Please forwant this information to Summer. l I

These changes will be added to Amendment V. Please call me if you have any questions, t

Jhn Robertson

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TABLE 5.4.7-2 (Cont'd)

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SNUTDOWN CoctthC SYSTEM FAlluaE PcDES AND EFFECTS ANALYSTS S wh Cnnss - bd .

syrptces and Local Eff ects inherent Remarks and

& Marne Fallstre Mode Cause includire Dependent Falls.res Method of Detection Coroensating Provision other Effects I

f II) ft f. a) faits Open Elect. Malf.,

eechanical Loss of one ECS train Lev temperature in 5CS; periodic Redundant ECS train Valve is norsaLIV tecked

( lsetati elve binding testing, valve closed st- ,51-342 position indication in the control races b) Fails Elect. Malf., No effect en SC5 speration Pericdic testing, None required closed mechanical valve positien binding indication in tie control recra

12) 50CMX a) false Elect. Matf. Inability to control contdown Comparison alth Redundant SCS train Intet/Dutlet indica- rate in affected train. redundant indicators, Tesperatare tien Possible isolation of with mit other Recorder T 300, functional Scs train process T-301 , instrumentation and

' valve positimes Indications consistent. Pertadie testing Amendment N April. 1, 1993

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(sheet 7 of 73 520TOOtt:: C00t;,0 ST5 FEM FAttuRE MOSES AWD EFFECTS AWAttsis t

symptoms and Local E f fects MA W orse Feiture Mode Cause :nherent Remarks and includine Dependent Feitures Method of Detection Consensat t m, 'P rovis f ori Other Ef fects

16) SCS Test Return a) Falls Cerrosion, Wome Isolation Yatve open migh temperature Series isolstfon mechanical st-314,51-315 birding Indication from valves in InwsT return 51-688, St-693 T-330, T-301; '

periodic testing b) Falls Corrosion, no effect on SCs operation Closed Periodic testing mene required mechanical binding 173 SCS/CS$ A a) Falls Elect. Malf., tone Periodic testing, secovever valve open mechanical CSS is norselly 33-341, sI-volve position fartated hindins. Indication in the s- JJ b) Falls Elect. Metf.,

control reas No effect en SCS operation Periodic testing, None required Closed mechanical valve position binding 1ralication in the

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18) PCPS tse ion a) Fall Mech. bi mooe Per ic testing a valve t series valve V is normelt evides isolation ocked closed vetve St-455, s 0, b) Feits Mech. bindin Mone st-45 Periodic ting t-454 Ct ed A A ^

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Junendment N April 1, 1993 t

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TABLE 4.5-3 (Cerut'd) *

(sheet & ef 6)

CONTAtsMENT SPEAT SYSTEM FAf ttRE RODEE AaB EFFECTS AAALT5f5 Symptamm and Local Effects teereret temarks and g same Fallesre made cassee incitaNrw Rwerufent Feltures method of tetection Caweasating Provision _ other Effects

13) CS Pusp/5CS a) Fa!La Corroeien, Effective toss of one Lou ilow Indication SCs is norantly Pump suction open mechanical containment spray passp F-338, F-34&; high isolated Crasa Connect binding, C5 pump suctica Yelves operator error, pressure Indication 51-340, sf-342 electrical P-310, P-329, fatture periodic testing, valve positten indicator b) Feits Corrosion, ut effect on CSS operation Periodic testing, Bene required closed mechanical valve pcsition binding, indicator electrical failure

14) CE Pump /ECs a) Fails Carrosion. Effective toss of one sigh flew SCS is normstly Pump Discherme open mechanical containment sprey pump indication isalated Cross-Cennect binding, F-338, F-348; Volves operator error, periodic testins, SI,341, St-M3 electrical valve position failure indicator b) Felin Correstem. Wo effect on Css eperetlen Periodic testing, none required closed rechanical valve position binding, frulicator electriest r n f a(L1tre '

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FCPS Isolation a) Falls Reeb. binding home Periodic testing " _ ,2n: ;;;L. ..i.e Valve is normally Valve . Open provides isolation Locked closed at valve

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CESSARnuhou option A design features rely on inherent physical attributes of a system or subsystem which will prevent failure when it is pressurized to normal RCS operating pressure. Option A features do not require any immediate action by equipment or operators to satisfy the ISLOCA acceptance criteria. This approach is intended to provide the optimum protection against ISLOCA challenges and to allow the operator the necessary time to properly assess and restore the system to normal conditions.

Examples of Option A features satisfying the ISLOCA acceptance criteria include e locating the system or subsystem completely within containment; e

designing the system or subsystem to normal RCS design Replace pressure M e . .

designing the system or subsystem to at least 40% of the RCS n,p" normal operating pressure; A

1. Y U N Agning the cycter or subsyst<>m ce that--th e a l t iet-e- w ctr-eng t h M-t-he-mahr i a l comprising the cyc t-emwed e Phn strenc produaed in th mater;al by p unaure equal Lv no c me4 4l Rc9 operating prarrurc; are 4-

• physically separating the system or subsystem from the RCS <

during conditions when the RCS pressure exceeds its design pressure.

Option B design features are design responses to ISLOCA events consisting of specific equipment and instrumentation which perform actions to prevent or mitigate the consequences of an ISLOCA. Option B design responses that have been considered will not require operators eventually require to prevent or mitigate the event, but will operators to perform remedial action, inspection of equipment following the event and returning the plant systems to normal configuration.

Option B design features are intended to be applied to systems for which it is impractical to apply Option A design features.

Examples of Option B design features are

• i the isolation of a system or subsystem in the pressurization l pathway at the interface between the lower pressure system or subsystem and its pressurization source; and pressure relief to limit the pressurization to within the design capabilities of the system.

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Amendment Q  ;

SE-4 June 30, 1993

~S Ac c H As4 E No Auts: -rs- 2 7r RN o f" "O _ _ . _ _

. . .- _. . . - . . - . . - . = . . . - - . - - . .- - . .- .- . ..-. . . . . - . . . . . .

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Insert "A" - Page 5E-4 Modify the third item as follows:

"* designing the system or subsystem to a pressure of at least 40% of the RCS normal pressure. Austentic Stainless Steel piping' will use a minimum wall thickness corresponding to standard weight for sizes less than 16 inch NPS and schedule 40 for 16 inch NPS and larger sizes;"

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SENT BY:DE&S  ; 2-21-94 ; 8:06 ; DUKE ENGR & SRVS* 2032855203;# 2/ 3 CESSA'REHL -

The design of tha plant lighting systems is in accordance with applicable industry standards for illumination fixtures, cables, grounding, penetrations, conduit, and controls.

All lighting fixtures and other components of the lighting system located in normally occupied areas or in areas containing safety earthquake equipment are supported so as to enhance the survivability of these components and to ensure, in particular, that they do not present a personnel or equipment hazard when subjected to a seismic loading of a design basis earthquake.

l Tha normal lighting system is used to provide normal illumination under all plant operation, maintenance and test conditions.

Table 9.5.3-1 summarizes typical illuminance ranges for normal lighting.

The sectaity lighting system provides the illumination required I# to monitor isolation zones and all outdoor areas within the plant protected perimeter. The security lighting system complies with

( the intent of NUREG CR-1327.

E The emergency lighting system is used to provide acceptable q 1evels of illumination throughout the station and particularly in areas where emergency operations are performed, such as control (

]' S rooms, battery rooms, containment, etc., upon loss of the normal lighting systam.

g Lighting circuits which are connected to a class 1E power sourca are treated as associated class 1E circuits.

9.5.3.2 System Description c

~~~

9.5.3.2.1 Normal Lighting system The Normal Lighting System provides general illumination

---)throughout the plant. in accordance Illuminating with illuminationSociety.

Engineering levels recommended by the Incandescent lighting is used in the containment Building.While incandescent, fluorescent and high intensity discharge lighting is provided in the remainder of the plant and on the plant site.

Power for the Normal Lighting System is provided independently from the Normal Auxiliary Power System via dry-type transformers and lighting panelboards.

Indoor lighting is designed for continuous operation, switching is by individual plant circuit breakers except in office areas.

Outdoor lighting is controlled by photocells.

The normal lighting system is As considered part of the plant such, the normal lighting permanent non-safety systems. I system is energized as long as power from an offsite power source or a standby non-safety souren (combustion Turbine) is available.

Amendment T 9.5-48 November 15, 1993 l

SENT BY:DE&S  ; 2-21-94 ; S:07 ; DUKE ENGR & SRVS4 2032855203;# 3/ 3 CESDAR Teena Gert3ency 'a n Oc moi,3 enMr.k room is ar,.g>prowdel Sud kh

[ d leut -fw inhecJ5 c of' 19;h$ 46 aced b Mferen4 cim3 dE dViskns.

Normal system operation is not affected by the failure or unavailability of a single light;.ng ' transformer.

The circuits to the individual lighting fixtures are staggered as much as possible, with the staggered circuits fed from separate electrical divisions, to ensure some lighting is retained in a room in the event of a circuit failure.

9.5.3.2.2 Security Lighting System The security lighting system is considered part of the permanent non-safety systems and is fed from the Alternate AC (AAC) Source (Combustion Turbine), which is located in a secure vital area for protection. Selected portions of the security lighting system essential to maintaining adequate plant protection are powered from a non-Class 1E battery power source.

The COL Applicant shall provide a security lighting system that l will meet CCTV illumination requirements within ' camera viewing areas to permit prompt assessment of intrusion alarms.

The security lighting system is designed to provide a minimum illumination of 0.2 foot-candles when measured horizontally at ground level.

l 9.5.3.2.3 Basrgency Lighting  ;

I Emergency lighting is located in vital areas throughout the plant as identified in Emergency Procedures and Hazards Analysis for safe-shutdown of the plant following an accident or hazard.

Included in the vital areas will be the Control Room, Technical Support Center, Operations Support Centar, the Remote Shutdown Panel Room, the stairway which provides access from the Control Room to the Remote Shutdown Panel room, Sample Room, Hydrogen Rocombiner Rooms, Electrical System Areas, Main Steam Valve 4 Houses, the Chemistry Labs, routes for personnel passage and 5 !

egress, and other areas where operator access is required post- C accident or hazard. F

.G The emergency lighting system in the main control room is integrated with the normal lighting system.pwl vill M c:nfip;r:d  ;

so-that nc"-21 an: 0;;;c gsacd 8 -~cauf.2: w12 he staggerac and fed  ;

fec: different ::tety - ' - *" """""" *"* 'ighting-4e retain e in the event e f- --; circ.:it failuse. The emergency lighting system in the main control room maintains -ed:g::t: e illumination levels in the e control room during -ekk emer M. condiElons, including station blackout. -The er:rgency 11;;gencyhti.g-syster in th rain-control recr ic powered-4 rom-a-olese-48 batte..y pe"ar ecurre_

The emergency lighting installations which serve the main control room and other areas of the plant where safe shutdown operations may be performed are designed,to remain functional during and after a design basis earthquake.

Amendment T 9.5-49 November 15, 1993

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1 14.2.12.1.21 Shutidown Cooling System Test 1.O OBJECTIVE ,

J 1.1 To demonstrate proper operation of Shutdown Cooling l System and the Shutdown Cooling Pumps.

1 2.0 PREREOUISITES 2.1 Construction activities on the systems to be tested are l complete.

2.2 Plant systems required to support testing are operable and temporary systems are installed and operable.

2.3 Permanently installed instrumentation is operable and l calibrated.

2.4 Test instrumentation is available and calibrated.

2.5 All lines in the Shutdown Cooling System have been filled and vented. .

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j 3.0 TEST METHOD 1 3.1 Verify proper operation of each shutdown cooling pump l with minimum flow established. l 3.2 Verify pump performance including head and flow characteristics for all design flow paths which include the normal decay heat removal flow path and  ;

1

1. Shutdown cooling system flow to the chemical and volume control system for purification.

2. Shutdown cooling system transfer of refueling water to the IRWST.

i

3. Shutdown cooling system to cool the IRWST. ]

Perform a full flow test of the shutdown cooling .

3.3 system.

l 3.4 Verify proper operation, stroking speed, position indication and response to interlock of control and -

l isolation valves.

3.5 Verify the proper operation of the protective devices, controls, interlocks, indications, and alarms using actual or simulated signals.

3.6 Verify isolation valves can be opened against design differential pressure.

Amendment U 14.2-54 December 31, 1993

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SENT BY:DE&S  ; 2-22-94 ; 7:55 : DUKE ENGR & SRYS* 2032855203 # 3/ 3 CESSAR 8!n%cmon O. Electrical Service and Lighting a The System 80+ design provides good lighting and convenient electrical service. This will facilitato maintenance and inspection activitics and reduce the anticipated personnel lamping in high expouuro. Reliablo extended service radiation areas will be used, whenever possible, to minimize the frequency of maintenance required. The lighting fixtures aro located to minimizo personnel exposure during maintenance. These features are in accordance with Regulatory Guide 8.8, Position C.2.1 guidance.

P. Spent Fuel Pool Decontamination system 80+ provides the capability to uso high pressure domineralized water for the decontamination the spent fuel pool. Alternative methods of decontamination, such as use of a strippable coating, may be ovaluated by tho operator, as practical.

12.3.1.3 Bource Term control Source term control is an important aspect of the system 80+"

design. The following design features reduce the overall dose due to operation, maintenance, and inspection activitics.

t A. Fuel Performance The System 80+" design features assure low primary system sources with improved fuel clad leakage performance of less than 0.1% fuel clad failures, as well as an extended fuel cycle.

B. Corronion Product Control System 80+" design includes design features that reduce corrosion product production in the primary system.

1. Primary System Materials 7-C.0 6 The System 80+" esign specifics primary system materials vi

~ corrosion rates and very low cobalt l

impurities (0.07 w/o for equipment in direct contact with the prim coolan ted vjhro No roVeda e n/4e$b.

The presence of antimony in E

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r ngs as prcsonte a problem with hot particles in the current generation of nuclear plants. In the System 80+ design, the reactor coolant pump bearings will be designed to minimize the presence of antimony.

Steam generator tubes are fabricated to roliove straanes to reduce stress corrosion cracking. This will reduce the probability of tube plugging activities and further reduce maintenance exposures.

Amendment U  !

12.3-9 December 31, 1993 1

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DSElt Open item 6.3.3-1 The applicant should provide the design criteria for the Si minimum recirculatio operating data or test results demonstrating operability at low recirculation How.

Ailll-CE's Itesponse in the NRC/ABB-CE/ DOE Senior Management Meeting of December 15, 1993 for the System 80+

Design Certification Program, ADH-CE presented the bases for the minimum now a type testing that is required to be performed by the vendor for the Safety injection P is provided for further clarification and to summari7e the commitments made The Safety injection pumps have been designed to meet all the requirements fo System. They are low-How, high-head, multi-stage, low-suction-specific-speed pump for operation from the specified minimum now to the design runout. This type of pum to operate successfuHy, especially at low now without susceptibility to hydraulic y and wear. instabilit The instability phenomena is a characteristic of a single-stage, high-flow, high s pump. -

The issues in NRC lhdletin 88-04 and NUREG/CR-5706 have beer, addressed e and design of the System 804 SIS. These issues include arrangement items such as th due to the location of mimflow connections .relative Also, system to the pum in the recirculation line and the minimum Gowtate has b The at BEILincrease in the size of the miniGow line allows all inservice testing e for System This eliminates all normal operation at minillow and minimizes wear on the pum provides a better bases for assessing the perfoi nance of the Si ,pump head and as the design D vibration values provided by the vendor are used as references values for ASME Further, NUREG/CR-5706 states that yBWR designs ke+4 pan the safety pumps at reduced now rates. a as~!

they p:"'do notance.

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has been committed to and documented in maximum protection for the SI pumps. Therefore, System 80+ design assures

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in previous responses to DSER Open item 6.3.3-1 and in a separate fax from M Volodzko of ABB-CE to S. Sum of the NRC on 3/l2/93, Al&CE has providcd summaries of utility nd v endor testing that and continuously. This has been summarized from ins j

the hfe of the plants. r '

For the System 80 design at Palo Verde, the llPSIP, which are the same for System 80+ SIS, are tested quarterly at miniflow. These pumps are disassem llPSI pumps. wear every third outage. To date, there has been no reported e wear requir 1 Further, ABB-CE previously committed to provide vendor test requirements lCfd*

pt841tbin CES Section 6.3.4.1.1). -

o acceptance criteria for the vendor's testingf iifkatirqmrementrTo-be- minimum Cow operation ymewt wirIq ,t y in

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exceed)be maximum safe limit fo lubricationHearing e th. its t_emperature wi not .11 equi

1. 1: fo/ bearing oil temperature)[ y endorsed a limit f 180 Pun $ shaft seal leakage temperature: f .

The \ (typically, ven s

egeed the maxifnum safeJIInit for the seals ( typically, 21: for the shaft seal cooling water), imit of 175 vpndors have

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[ Pump frame vibratiok Testing will demonstrate,Aiyit x vibrati established in Fig /ure'ISTB5.2 from the ASME OM Colle.

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/ s pump to with' stand opedtion at minimum s hours or until all ofliie key parameters identified above stabilize i (14e flow.

e,is ther ebht SI The time the Si pum w are required to operate at. Eigh for ieuatural m' tours.

imlhuJ]owrepresents ase dis maximur culationhqldown. cus d in CESSAR-D Appendix SD

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s Ck.bbNN cbd5YlCAYlON 6.3.4 INSPECTION AND TESTING REQUIRTMENTS During fabrication of the SIS components, tests and inspections are performed and documented in accordance with code requirements to assure high quality construction. As necessary, performance tests of components are perf ormed in the vendor f acilities. The SIS is designed and installed to permit in-service inspections and tests in acccrdance with the ASME Code,Section XI.

6.3.4.1 SIG Performance Tests Prior to initial plant startup, a comprehensive series of system flow tests, as detailed in Section 14.2, will be performed to verify that the design performance of the system and individual components is attained.

Preoperational tests and analyses will be performed to confirm that the as-built SIS fulfills operability requiremente ' nd provides a level of performance that satisfies safety analyses.

6.3.4.1.1 Flow Testing Each installed SI train vil] be tested to reasure SI pump developed dif fer ential pressure at miniflod, measure runout flow through the DVI lines and, for SI pumps 3 and 4, measure runout flow through the hot leg injection lines. Runout flow testing will be conducted with the RCS at atmospheric pressure conditions. Test results will be used to confirm SI pump performance characteristics over the operating range of the pump and to confirm system resistance characteristics. Test conditions, including fluid temperature, suction and discharge side fluid elevations, and potential instrument uncertainties will be taken into account in performing an analysis. The analysis will use test results, with adjustments made for test conditions, to determine system performance during postulated accident conditions. The calculated system performance shall be within the limits used to perform safety analyses in Section 6.3.3 and Chapter 15.

r -v W W Testing will be performed to confirm that the SI pump miniflow p' rate in the installed system meets or exceeds pump vendor's minimum flow requirements. The pump vendor will perform a type test to generate data that verifies the SI pump's ability to gf o

k< operate continuously within the design limits of key parameters at minimum flow for the mission time. These key paraneters are:

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• Ucaring oil equilibrium temperature: The testing will gfM demonstrate that its temperature will not e<ceed the pp i

maximum safe limit for lubrication (typically, vendors have endorsed a limit of 180*F for bearing oil '"

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CESSAR naincuiou v

• Pump shaft seal leakage temperature: The testing will demonstrate that the temperature will not exceed the \

maximum safe limit for the seals (typically, vendors /

have endorsed a limit of 175*F for the shaft seal /

cooling water), and

• Pump frame vibration: Testing will demonstrate that vibrations are within the criteria established in Figure ISTB 5.2 from the ASME OM Code.

The duration of the test, or mission time, is either eight hours or until all of the key parameters stabilize, depending on p whichever is longer. Eight hours represents the maximum time the SI pumps are required to operate at minimum flow as discussed in Appendix SD for a natural circulation cooldown. After completion of the test, the SI pump will be inspected and the condition evaluated.

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Testing will be performed to confirm that the SI pump return line to the IRWST allows each SI pump to be operated at a flow rate equal to or greater than design flow during inservice testing.

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l INSERT TO CFSAR-DC SECTION 6.3.4.1.1 (PAGE 6.3-41) l Testing will be performed to confirm that the Si pump miniflow rate in the installed system meets or exceeds the pump vendor's minimum flow requirements. The pump vendor will perform a type test with Guid conditions representative of those during a natural circulation ,

cooldown to generate data that verifies the SI pumps' ability to operate continuously within the design limits at minimum flow for the mission time. The pump type test will monitor the normal pump qualification test par:uneters and will use the following as acceptance criteria:

Bearing oil equilibrium temperature: The testing will demonstrate that its temperature will not exceed the maximum safe limit for lubrication (typically, vendors have endorsed a limit of 180 'F for bearing oil temperature).

Pump shaft seal leakage temperature: The testing will demonstrate that .the temperature will not exceed the maximum safe limit for the seals (typically, vendors have endorsed a limit of 175 'F for the shaft seal cooling water),

Pump bearing housing vibration: Pump bearing housing vibration will be measured in accordance with ISTB 4.6.4 from the ASME OM Code. Testing will demonstrate that. vibrations are within the criteria established in Figure ISTB 5.2 from the ASME OM Code.

Pump inspection: After completion of the test, the SI pump will be disassembled and the internals inspected with the conditions of the parts being evaluated for excessive wear as a result of degradation during the pump test.

1 The duration of the test, or mission time, is either eight hours or until the bearing oil temperature, the pump shalt seal leakage temperature and the pump bearing housing vibration >

stabilize, whichever is longer. Eight hours represents the maximum time the Si pumps are required to operate at minimum now as discussed in Appendix SD for a natural circulation cooldown.

4 ALWR-FS-268 Rev. 01 Page 4 of 4 l

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FAX  ;

To: Nick Saltos i USNRC - NRR Mail Stop 10E4 Phone (301)504-1072 Fax (301)504-2260 FROM: David Finnicum ABB-CE Phone (203)285-3926 Fax (203)285-5881 XC: J. J. Ilerbst (w/o)

R. E. Jaquith (w/o)

M. Ruben (NRC)

Adel El-Bassioni (NRC)

J. Longo Jr, S. E. Ritterbusch .

9424 Files 9612 files DATE: November 29,1993 NUMBER: OPS-93-0998 SUIMECT: Evaluation of Fire Inside Containment In our conversation of November 22,1993, you requested that ABB-CE provide an assessment of the potential impact of a fire inside containment on equipment needed for a safe shutdown and the attendent potential risk.

ABB-CE believes that a fire inside containment that would damage all equipment inside l containment is not credible because of the limited amount of combustibles inside containment, the spatial separation of equipment, and the physical barriers witin containment. i Thus, the maximum credible fiore would only affected a limited complement of equipment inside containment.

As part of the response to DSER Open Item 9.5..l.2.1-1, ABB-CE has performed an analysis of protection of redundant functions. This analysis concludes that spatial separation and location of equipment inside containment assures that a fire inside containment will not damage redundant functions. (A copy of this analysis is attached.) The primary equipment of concern inside containment includes the motor operated shutdown cooling suction isolation Page 1 of 16 l

l valves for each division, the Rapid Depressurization Valves, and instrumentation associated with the steam generators. 1 The SCS suction isolation valves, called RCS pressure isoiation valves in the attached analysis, are located by division 180 degrees apart. The valves are located near the crane wall on either side of containment and are over 100 feet apart. They are located at elevation i 10l+8 which is 10 feet above the noor elevation of 91+6 feet. This separation is sufficient i to ensure that at least one division of SCS is available to perform the coolown function.

The Rapid Depressurization Valves (RDVs) are located inside containment. The valves for

one division are located inside the pressurizer cavity and the valves for the other division are located outside the pressurizer cavity. Thus, the RDVs for the redundant divisions are separated by a concrete wall.' This ensures that one division of the Rapid Depressurization i System is available to perform its function.  ;

A fire at either steam generator could damage instrumentation associated with that steam generator. However, the other steam generator would not be affected and would be available to achieve safe shutdown. In addition, there are four channels of steam generator level and pressure instrumentation for each generator and each channel is located in a different quadrant around the steam generator, As seen from the discussion above and that in the attachment, the maximum credible fire inside containment would affect at most the equipment associated with one division of one system used for safe shutdown (shutdown cooling system, Rapid Depressurization System or steam generator), this is consisten with the assumption in the Scoping Analysis for Fire and Flood in the System 80+ PRA that the worst possible fire could affect, at most, one division. The fire frequencies used in the scoping analysis were derived from data presented in NUREG/CR-4840, including the frequency for fires in the reactor building. A review of the information presented in table A-2 of NUREG/CR-4840 shows that this category included 7 fires inside containment. (Note: The reported fires inside containment were local in nature.)

Therefore, fires inside containment are covered in the fire frequency used in the scoping analysis t

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DSER Ooen Item 9.5.1.2.1-1 The staff does not accept the concept of radiant heat shield and 20 ft. of separation. Each such deviation inside containment must be fully justified.

Pronosed Onen Item 9.5.1.2.1-1 Resolution CESSAR-DC Soction 9.5.1.1.2.C will be revised to state that for inside containment or the annulus safe shutdown following a fire is ensured by separation of redundant divisions by quadrant to provide sufficient spatial separation, as proven by engineering analysis.

Separation for safe chutdown cables is provided through use of mineral insulated cables which qualify as a three hour rated barrier.

CESSAR-DC Section 9.5.1.3.9 will be revised to provide an analysis of protection of redundant functionc. This analysis is attached and describes arrangements for which location and spatial separation assure that fire inside containment will not damage redundant functions.

i open Item 9.5.1.2.1-1 1 Rev. C DRAFT 05/14/93 i

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9.5.1.3 FIRE PROTECTION SAFE SHUTDOWN ANALYSIS i

9.5.1.3.1 ASSUMPTIONS 1

+

The Fire Protection Safe Shutdown Analysis includes the '

effects of one worst case spurious actuation.

Fire is postulated with or without loss of offsito power  ;

(which ever is the most severe challenge to the ability to l achieve safe shutdown). I

. Inside containment, cables for safe shutdown valve motor 1 operators and instruments are three hour fire rated.

. Fire is not postulated concurrent with simultaneous, coincidental failures of safety systems, other plant accidenta  !

or the most severe natural phenomena.

)

I 9.5.1.3.2 FIRE PROTECTION SAFE SHUTDOWN DESIGN DASIS GOALS

. Achieve and maintain subcritical reactivity conditions in the primary system.

j 1

Maintain reactor coolant inventory, l l

• Achieve primary system temperature and pressure conditions. I Maintain Reactor Coolant System (RCS) process variables within those predicted for loss of AC power.

. Prevent fuel clad damage, failure of the primary system pressure boundary, or rupture of the containment boundary, j 9.5.1.3.3 FIRE PROTECTION SAFE SHUTDOWN DESIGN BASIS OBJECTIVES The following Design Basis Objectives are met in order to assure  !

the Design Basis Goals stated above are satisfied:

• Maintain RCS pressure boundary integrity (i.e., reactor coolant pump seal integrity, CVCS letdown isolation, Safety Depressurization System isolation and RCS sample line isolation).

Assure the reactivity control function maintains the available shutdown margin at greater than 1% Ak/k with the highest worth control element assembly (CEA) fully withdrawn.

+ Assure reactor coolant make up is available to maintain

reactor coolant in the pressurizer within prescribed limits. -

. Maintain RCS decay heat removal function and cool down the RCS to cold shutdown conditions.

Provido direct reading of procesa variables necessary to perform and control reactivity, reactor coolant pressurizer level and decay heat removal.

. Maintain support functions (process cooling, lubrication, etc.) for equipment required for safe shutdown.

9.5.1.3.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN

. The RCS provides reactivity control by control element assembly (CEA) insertion and also removes decay heat from the core through natural circulation.

. Emergency Feedwater System (EFW) provides secondary side decay heat removal capability.

. Atmospheric Dump Valves provides secondary side pressure control capability.

. Shutdown Cooling System (SCS) provides residual heat removal function for cooldown from hot shutdown to cold shutdown conditions.

. Safety Injection System (SIS) provides makeup capability for inventory control and boron addition for reactivity control.

. Safety Depressurization System (SDS) provides primary system pressure control capability.

. Essential Chilled Water System (ECWS) provides chilled water for IWAC heat removal to all safety related room recirculation cooling units.

. Component Cooling Water System (CCWS) provides decay heat removal capability and equipment cooling for

• the Shutdown Cooling System, Safety Injection System, Essential Chillers, Emergency Diesel Generator Coolers, etc. , as well as other non safo chutdown functa.ons.

. Station Service Water System (SSWS) takes suction from the ultimate heat sink and provides cooling water flow to the CCWS heat exchangers for cooling and decay heat removal.

. The Control Building, Nuclear Annex, Subsphere and Diesel Generator ' Building Ventilation Systems provide ambient temperature control within parameters required to assure componenta function as intended to achieve safe shutdown conditions.

w< r a

. Reactor coolant pump seal cooling is provided by-either soal injection from the CVCS charging pumps or direct cooling from the CCWS.

- The Pool Cooling and Purification System provides decay heat removal from the spent fuci pool.

• The onsite Emergency Diesel Generators provide power for lE busses for equipment power, control and instrumentation required to achieve safe shutdown conditions.

. The Combustion Turbine (AAC) provides onsite power to the permanent non-safety busses which provide power to the CVCS Charging Pumps and associated valves and controls.

Two shatdown paths are provided by the above systems. These are Division 1 and Division 2. For fires outside of the control room, one of these divisions is ensured to be available to bring the plant to safe cold shutdown.

For a Control Room fire, the Remote Shutdown Panel will be utilized as alternative shutdown capability. A fire in the Control Room is the only fire scenario which requires the Remote Shutdown Panel to be utilized. Shutdown from the Control Room can be accomplished for fires originating in all other fire areas. For the Control Room fire, both shutdown paths (i.e. Division 1 and Division 2) are available to safely shut the plant down to cold shutdown from the Remote Shutdown Panel.

Each of these systems includes adequate controls and instnunentation in the Control Room and at the Remote Shutdown Panel to assure safe shutdown can be achieved.

CESSAR-DC Section 7.4.2.5 describes the instrumentation and controls that are on the Remote Shutdown Panel that are required to bring the plant to safe cold shutdown conditions.

CESSAR-DC Section 18.3.2 describes the personnel requirements for the Control Room and Remote Shutdown Panel. Shutdown procedures following a fire will be the same as described in the plant Emergency Procedures for achieving safe cold shutdown and will not require additional personnel for the fire scenario. Safe cold shutdown can be. achieved with one shutdown division within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after reactor trip.

9.5.1.3.5 SYSTEMS WHICH REQUIRE ISOLATION

= SCS pressure isolation valves until RCS is cooled and depressurized to SCS entry conditions.

• SDS to prevent uncontrolled blowdown of the RCS.

• CVCS letdown to prevent uncontrolled letdown of the RCS.

RCS sample lines to prevent uncontrolled letdown of the RCS.

• Main Steam System to prevent uncontrolled blowdown of the steam generators.

• Atmospheric Dump Valves to prevent uncontrolled blowdown of the steam generators.

Main Feedwater System to prevent uncontrolled blowdown of the steam generators and steam generator over fill.

Steam Generator Blowdown System and steam generator sample lines to prevent uncontrolled blowdown of the steam generators.

• EFW to prevent steam generator over fill.

Each of these systems includes adequate controls and instrumentation in the Control Room and at the Remote Shutdown Panel.

9.5.1.3.6 ASSOCIATED CIRCUITS The potential for electrical interaction duo to fire mandates that a study be conducted to assure that redundant safe shutdown systems are not damaged by a single fire. Generic Letter 81-12 Rev. 1 defines Associated Circuits and provides guidance for documenting the Associated Circuits Study.

Outside of containment the System 80+ plant configuration provides complete separation of redundant safety related divisions by three hour fire rated barriers. Division 1 is located (plan) north of column line 17. Division 2 is located (plan) south of column line

17. An exception is the Control Room and the Remote Shutdown Panel room which are physically separated and electrically isolated and provide redundant shutdown capability. Transfer switches which i transfer control from the Control Room to the Remote Shutdown Panel are located in the Control Room. Transfer switches are arranged i such that when power is transferred from the Control Rodm to the l Remote Shutdown Panel, manual operations in all four Vital i Switchgear rooms are required to return control capability to the  ;

control Room. Thus associated circuit interaction in the Control Room will not affect the ability to achieve safe shutdown from the Remote Shutdown Panel.

9.5.1.3.7 SAFE SHUTDOWN FOLLOWING FIRE OUTSIDE OF CONTAINMENT j As discussed in section 9.5.1.3.6, " Associated Circuits", redundant safe shutdown divisions are separated by column line 17. Each fire area is enclosed in three hour fire rated barriers. Three hour fire rated barrier walls are located along Column Line 17, except at elevations 115+6 and 130+6 where the Control Room is

located. The exception to complete divisional separation is the Control Room and the Remote Shutdown Panel room which have redundant control function capability. They are physically separated and electrically isolated from each other. CESSAR-DC Figure 9.5.1 depicts the separation of redundant electrical divisions outside of containment.

Thus a fire in any fire area outside of containment will not affect redundant safe shutdown systems, equipment, or components.

9.5.1.3.8 SAFE SHUTDOWN FOLLOWING FIRE INSIDE CONTAINMENT The Containment and Annulus are a single fire area. The only components inside the Containment and Annulus which are required for safe shutdown are motor operated valves and instruments associated with safo shutdown systems.

Inside the Annulus and Containment, three hour fire rated cable protective systems (i.e., mineral insulated cablos) are used for cables associated with safe shutdown functions. An exception to the three hour fire resistance rating may be containment penetrations which are currently commercially available with a one hour fire resistance rating. Three hour fire rated containment penetrations will be purchased if available.

The only in situ combustible material inside containment that may be exposed to a fire is insulation of cables that are not associated with safe shutdown functions. Redundant trains of valves and instruments analyzed as an assured method of achieving safe shutdown are physically separated such that a potential fire will not affect redundant equipment as stated in section 9.5.1.3.9.

In situ combustible material inside containment is limited to those materials which are essential for unit operation (i.e., cable insulation, lubricants, etc. ) . The largest quantity of combustible materials is RCP motor lubrication oil. All potential leak points are enclosed in a seismically designed oil collection system which drains to a seismically designed oil collection tank. If oil were to escape from any reactor coolant pump, it would drain into the containment holdup volume. There are no safe shutdown components located in the containment holdup volume which may be damaged due to a fire at this location.

Transient combustible material will be administrative 1y controlled to avoid unacceptable fire hazards.

9.5.1.3 9 PROTECTION OF REDUNDANT FUNCTIONS

1. OBJECTIVE: Maintain primary system pressure boundary integrity (i.e., reactor coolant pump seal integrity, CVCS letdown isolation , SCS isolation, SDS isolation and RCS sample line isolation).

ANALYSIS:

A. RCP seal integrity is maintained by either seal injection from the CVCS charging pumps or direct cooling from the i CCWS. The CVCS is discussed in CESSAR-DC Section 9.3.4, and is shown in Figure 9.3.4-1. The CCNS is discussed in CESSAR-DC Section 9.2.2 and is shown in Figure 9.2.2-1.

The RCP seals are discussed in CESSAR-DC Section 5.4.1 and are shown in Figure 5.1.2-2.

Outside containment the two divisions of CCWS are separated by a three hour fire rated barrier. In addition, the redundant CVCS charging pumps are separated by a three hour fire rated barrier. However, each division of CCWS provides seal cooling for two of the four RCPs. Should one CCWS division be lost from a fire outside of containment, RCP seal integrity of the two RCPs cooled by the CCWS division is maintained through seal injection from the CVCS charging pump in the unaffected division. The seal injection lino penetrating containment is located 90 degrees apart from each containment penetration for the CCWS supply and return line to the RCPs. Each of the CCWS supply and return lines to the RCPs has two isolation valves, (For division 1, the isolation valve located inside containment has control power supplied from channel B and the isolation valve located outside of containment has control power supplied from channel A. For division 2, the isolation valve located inside containment has control power supplied from channel A and the isolation valve located outside of containment has control power supplied from channel B.). There is an isolation valve in the seal injection line located outside of containment. This valve han control power supplied from Channel C. Thus a fire outside containment cannot simultaneously isolate both seal cooling means.

Inside containment isolation and control valves on the CVCS seal injection, RCP controlled seal bleedoff and CCWS supply and return lines for the RCP seal coolers are protected such that spurious signals from a fire inside i containment can not simultaneously isolate both RCP seal cooling means. Seal injection isolation valves on each side of the high pressure seal coolers are normally open with the breakers racked out. The CCW supply and return line isclation valves to each RCP are powered from the i permanent non-safety electrical power busses and are  !

normally deenergined (e.g. MCC breaker is open). The seal injection ficw control and controlled seal bleedoff line valves are located near each associated RCP inside the Reactor Building crane wall. These valves are powered from the permanent non-safety electrical power busses. The containment isolation valves for the RCP seal cooler CCWS supply and return lines are powered from

1 l

l Class 1E busses and are normally deenergized (e.g. MCC l breaker is open). The RCP controlled seal bleedoff line containment isolation valves are also powered from Class i 1E busses. These valves are normally energized (e.g. the j solenoid actuators are energized to maintain the valves 1 in the open position). However, should the RCP controlled seal bleedoff valves spuriously close, a relief valve located inside containment opens and allows continued RCP controlled seal bleedoff to the reactor drain tank. Therefore, a fire inside containment cannot simultaneously isolate both means of seal cooling.

B. The CVCS letdown line is discussed in CESSAR-DC Section 9.3.4 and is shown in Figure 9.3.4-1. The letdown line has two power operated valves in series. Each isolation valve is powered from..a different division of Class lE power and is separated and protected such that a fire incide containment can not prevent both isolation valves from closing.

C. Each division of SCS has two RCS pressure isolation valves in series. These pressure isolation valves are shown in CESSAR-DC Figure 6.3.2-1C and are discussed in Section 5. 4.7. Each valve has power supplied from a different Class 1E channel and is normally deenergized (e.g. MCC breaker is open). The MCC for each valve is located outside containment in separate fire areas. In l addition, the valves are " status controlled" locked closed from the Control Room via administrative procedures which require the component's discrete control device (e.g. switch or soft touch screen) to be enabled I prior to valve operation. The valves are also  ;

interlocked such that they ennnot be opened until RCS pressure reaches SCS entry conditions. Thus, neither a fire inside or outside containment is capable of l spuriously opening both valves in a division.

1 D. Each division of SDS from the pressurizer to the In l Containment Refueling Water Storage Tank (IRWST) has two 1 power operated valves in series. Each valve has power supplied from a different Class lE channel and lu' i normally deenergized (e.g. MCC breaker is open) . The MCC '

for each valve is located outsido containment in separate fire areas such that a fire inside containment cannot i result in spuriously opening both valves in a division. l Each division of SDS from the pressurizer to the reactor drain tank and from the top of the reactor vessel to the reactor drain tank has two power operated valves in series. Each valve has power supplied from a different Class lE channel and is normally deenergized. The power source for each valve is located 'outside containment in separate fire areas such that a fire inside containment cannot re'ault in spuriously opening both valves in a

I 1

division. A fire outside containment could only affect a single valve. The SDS is discussed in CESSAR-DC Section 6.7 and is shown in Figure 5.1.2-3.

E. Primary nampling lines have a flow reducing orifice which .

restricts the flow to less than the normal makeup  !

capacity. In addition, each sample lino has a normally closed isolation valve inside containment and a normally closed isolation valve outside containment. Each containment isolation valve associated with a sample line penetration is powered from a different division of Class lE power. Thus, a fire inside containment can only affect the operation of one of these valven. The sample system is discussed in CESSAR-DC Section 9.3.2, i Containment Isolation is discussed in CESSAR-DC Section 1 6.2.4, and the containment isolation valves are shown on  !

Figure 6.2.4-1.

2. OBJECTIVE: Assure the reactivity control function maintains the available shutdown margin at greater than 1% Ak/k with the highest worth CEA fully withdrawn. i l

ANALYSIS: Reactivity control is maintained by the CEAs and by j boration. The Safety Injection System (SIS) is the primary method of injecting boron into the primary system. The SIS is discussed in CESSAR-DC Section 6.3 and is shown in Figure 6.3.2-1. The majority of components in the SIS are located outside containment where each division is separated by a three hour fire rated barrier. The Safety Injection Tanks ,

(SITS) are located inside containment. To ensure an available i flowpath from each SIT, the discharge isolation valves are I normally open with the breaker racked out. To prevent spurious opening of the single isolation vent valves on the SITS, the solenoid valve power supply fuses are normally removed. Thus a fire inside containment will not affect the ability to maintain reactivity control.

3. OBJECTIVE: Assure reactor coolant make up is available to maintain reactor coolant in the pressurizer within prescribed limits.

ANALYSIS: The Safety Injection System (SIS) is used for make up to the RCS. See item 2, " reactivity control", above for description and protection.

4. OBJECTIVE:

~

Maintain reactor coolant decay heat removal function and cool down the RCS to cold shutdown conditions.

ANALYSIS:

, A. Emergency Feedwater System (EFWS) provides decay heat removal from hot standby to hot shutdown conditions by supplying feedwater to each steam generator. The EPWS is discussed in CESSAR-DC Section 10.4.9 and is shown in l

~

I

Figure 10.4.9-1. Each division has a motor driven ancLa steam driven EFW pump. Each of these pumps is sized for full capacity so that only one pump per division is neceoonry to achieve safe shutdown. Each pump discharge line has two motor operated valvea in series. The motor driven and steam driven EFW pump of each division feed into a common supply header. All pumpa and power operated valvos are located outside of the containment.

Thus a fire incide of containment will not affect the EFW function. Outside of containment each EFW train is caparated by a three hour fire rated barrier. In order to prevent steam generator over fill, the motor operated

. control valve at the discharge of each pump has power supplied from a different Class 1E channel compared to the associated pump controls. The valve and pump along with associated cables are located and routed through different fire areas. to prevent losing both pump and valve control due to spurious signals.

D. Steam Generator pressure control la maintained by the atmospheric dump valves which are part of the Main Steam Supply System. These valves are discussed in CESSAR-DC section 10.3 and are shown on Figure 10.3.2-1. Each of the four main oteam lines has an atmospheric dump valve (ADV) and its associated block valve located upstream of the main steam isolation valven. These valves are located outside containment in the main steam valve houses (MSVH). Thus a fire inside containment cannot af fect their operation. Each MSVH, which contains two of the four ADVs, is located on opposite sides of the Reactor Building and is separated by a three hour fire rated barrier. Only one steam generator and one of the ADvo associated with that steam generator are required for decay heat removal and cooldown. Each ADV in a division has power supplied from a different Class 1E channel in its respective division. Therefore, a fire outside containment can only affect the operation of the ADVs located in the division in which the fire occurs.

Thus, the ADVs in the unaffected division will be available to control pressure in the steam generator performing the cooldown function.

C. In order to prevent uncontrolled blowdown of the steam generator and steam generator overfeed, the main steam, main feedwater, and steam generator blowdown systems and the steam generator cample lines require isolation.

The Main Steam System is discussed in CESSAR-DC Section 10.3 and is shown in Figure 10.3.2-1. Each steam generator has two main steam linea. Each main steam line has a main steam isolation valve. Each main steam isolation valve has redundant solenoida powered from different Clano 1E channels. In addition, these valves I fail closed on loss-of-power. The main steam isolation i 1

1

valves are located outside of containment in their associated main steam valve house. Thus, a fire inside containment does not affect the operation of these valves. In addition, main steam can be isolated at the turbine ntop valves.

The Main Feedwater System is discussed in CESSAR-DC Section 10.4.7 and is shown in Figure 10.4.7-1 Each steam generator has a economizer feedwater line and a downcomer feedwater lino. Uncontrolled blowdown of a steam generator is prevented by two check valves in series on each of these lines. Steam generator over feed is prevented by closing the two feedwater isolation valvea located in aeries on each of these lines. Each feedwater isolation valve in serien has power supplied from a dif ferent Class lE channel. In addition, these valves fail closed on loss of power. The feed water isolation valves are located outside of containment in their asnociated main steam valve house. Thua, a fire inside containment does not affect the operation of these valves. Main feedwater can also be isolated by stopping the main feedwater pumps.

The Steam Generator Blowdown System and the Process Sample System are discuased in CESSAR-DC Sections 10.4.8 and 9.3.2 respectively. The Steam Generator Blowdown System is shown in Figure 10.4.8-1. Each steam generator blowdown line and each steam generator sample line can be isolated by their associated containment isolation valves. Each containment penetration han a containment isolation valve located inside containment and a containment isolation valve. located outside of containment. Each valve associated with a containment penetration is powered from a different division of Class lE power. Thus, a fire inside containment can only affect the operation of one valve.

D. The Shutdown Cooling System (SCS) provides decay heat removal and cooldown after the primary system is cooled and depressurized to the point that allows opening of the RCS pressure isolation valves. The SCS cools ' the RCS from hot shutdown to cold shutdown conditions. The SCS in described in CESSAR-DC Section 5.4.7 and is shown in Figure 6.3.2-1 The SCS has redundant divisions. Each division taken suction from a different RCS hot leg and returns the RCS after it is cooled directly to the reactor vessel. The majority of the SCS system is located outside of containment and the redundant divisions outside of containment are separated by a three hour fire rated barrier. Only the motor operated RCS pressure isolation valves are locatt.M inside containment.

There are two valves in series in each of the redundant flow patha which are located by division 180 degrees  ;

apart. These valves are located near the crane wall on l

either side of containment such that they are over 100 feet apart. They are located at elevation 101+8 which is 10 feet above the floor elevation of 91+6. This distance is sufficient to ensure that one division of SCS is available to perform the cooldown function.

E. The Component Cooling Water System (CCWS) and the Station Service Water System (SSWS) transfer decay heat from the SCS to the ultimata heat sink. In addition, they provide process cooling to equipment and components required for safe shutdown. These systems are discussed in CESSAR-DC Sections 9.2.2 and 9.2.1 respectively and are shown in Figures 9.2.2-1 and 9.2.1-1 respectively. Each of these systems are located outside of containment and would not be affected by a fire inside containment. Outside of containment each division is separated by a three hour fire rated barrier. An exception is the CCWS cooling to the RCP seals which has valves located inside containment. See item 1A above for analysis of this item.

5. CBJECTIVE: Provide depressurization of the RCS to allow Shutdown Cooling System to be placed in service to obtain cold shutdown conditions.

ANALYSTS: RCS depressurization is accomplished utilizing the Safety Depressurization System (SDS) . The SDS is described in CESSAR-DC Section 6.7 and is shown in Figure 5.1.2-3.

Depressurization is accomplished by opening the valves and controlling flow from the pressurizer to the reactor drain tank. These valves are located inside containment. Two divisions of valves located in parallel are provided. Each division of SDS from the pressurizer to the reactor drain tank has two power operated valves in series. Each valve has power supplied from a different Class lE channel. The valves and cables are adequately separated and protected (i.e. one division is inside the pressurizer cavity and one division is outside of the pressurizer cavity) to ensure one division of ,

SDS is available for RCS depressurization in the event of a '

fire inside of containment.

6. ODJECTIVE: Provide direct reading of process variables necessary to perform and control reactivity, reactor coolant l pressurizer level and decay heat removal.

ANALYSIS: Instrumentation (Incore instrumentation, T-Hot, T-Cold, S\G Pressure, S\G Level, Pressurizer Pressure, Pressurizer Level, Neutron Flux): Cables for all of these instruments are three hour fire rated. l A. Neutron Flux instrumentation, T-Hot and T-Cold are l located inside the primary system and are not susceptible I to fire damage.

1 B. Pressurizer Pressure and Level instruments are located at i the pressurizer. There are four channels of pressurizer pressure and level instrumentation. Each channel is j located in a different quadrant around the pressurizer.  !

C. S\G Pressure and Level instruments : Fire at either steam generator may damage instruments associated with that steam generator. However the other steam generator would not be affected and would be available to achieve safe shutdown. In addition, there are four channels of steam generator pressure and level instrumentation and each channel is located in a different quadrant around the steam generator.

7. OBJECTIVE: Maintain support functions (process ' cooling, lubrication, etc.) for equipment required for safe shutdown.

ANALYSIS:

A. Component Cooling Water System (CCWS) and Station Service Water System (SSWS) are discussed in , item 4E above.

D. Lubrication: There is no equipment inside containment which requires lubrication for safe shutdown.

Lubrication requirements outside of containment are divisionalized and separated by a three hour fire rated barrier.

C. Ambient cooling:

The Essential Chilled Water System (ECWS) provides cooling water to area room coolers located outside containment. These coolers are contained in the Control Complex, Reactor Building Subsphere, and Nuclear Annex Ventilation Systems. These systems are discussed in CESSAR-DC Sections 9.2.9, 9.4.1, 9.4.5, and 9.4.9 respectively, and are shown in Figures 9.2'.9-1, 9.4-2, 9.4-5,. and 9.4-8 respectively. Each of these sy0 tem has two divisions which are entirely located outside of containment and are separated by a three hour fire rated barrier.

The Diesel Generator Building Ventilation System maintains the ambient conditions within the diesel generator rooms to ensure operation of the diesel generators and controls. This system is discussed in CESSAR-DC Section 9.4.4 and is shown in Figure 9.4-7. This system is located outside of containment and each division is separated by a three hour fire rated barrier.

Equipment located inside containment .is qualified for high post accident temperatures., Therefore, i

1

containment cooling is not required to ensure operation of safe shutdown equipment following a fire.

8. OBJECTIVE: Remove decay heat from the spent fuel pool.

ANALYSIS: Decay heat is removed from the spent fuel pool by the Pool Cooling and Purification System. The Pool Cooling and Purification System is discussed in CESSAR-DC Section 9.1.3 and is shown in Figure 9.1-3. All components associated with the spent fuel pool cooling function are located outside of containment and each division is separated by a three hour fire rated barrier.

9. OBJECTIVE: Provide an assured source of on-site electrical power to equipment and components required for safe shutdown.

ANALYSIS: The assured source of electrical power is either of the emergency Diesel Generators for equipment and components powered from the Class lE busses. The electrical distribution system is discussed in CESSAR-DC Section 8.3 and is shown in Figures 8.3.1-1 and 8.3.1-2. The emergency Diesel Generators and associated Class 1E busses are located outside containment and each division is separated by a three hour fire rated barrier. The Class lE busses are separated from the non-lE busses by two isolation breakers in series. The CVCS charging pumps are powered from the permanent rion-safety busses.

Emergency on-site power is supplied to those busses by the combustion turbine. The permanent non-safety busses are located in the turbine building. The Turbine Building is separated from the Nuclear Minex by a three hour fire rated barrier. The combustion turbine is located in its own structure which is separated from the Turbine Building and Nucloar Annex. Cables from the permanent non-safoty busses are separated by the divisional three hour fire rated barrier after they enter the Nuclear Annex.

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• CESSAR !anneuion Report No. Title Date Issued CESSAR-DC Chapter l NPX-IC-DR-791-02 Human Factors Engineering May 1992 18 Standards, Guidelines, and Bases for System 80+ LD 069 tB-93=110- System 80+ Fire Hazards R0 March 13, 9 m _qy -of;t. Assessment 1992 R1 April 5,

% bgg ~x NPX80-IC-RR790- Human Factors Evaluation and March 1993 18 02, Rev. 1 Allocation of System 80+

Functions LD-93-056 NPX80-IC-DB-790- Nuplex 80+ Advanced Control September 18 01 Complex Design Bases LD 1992 102 NPX-TE-790-01 Nuplex 80+ Verification May 1992 18 Analysis Report, LD-92-065 NPX80-IC-DP790- System 80+ Function & Task May 1989 18 02, Rev. 01 Analysis Report, LD-92-065 NPX80-SQP-0101.0 Software Program Manual for January 1993 7 NUPLEX 80+, LD 009 NPX80-IC-QP790-2 Nuplex 80+ Software Safety January 1993 7 Plan Description, LD 009 NPX-IC-QG790-00 Qualification Guidelines for November 1992 7 Instrumentation and Controls Equipment for NUPLEX 80+, LD-92-113 NPX80-QPS-0401.1 Requirements for The Supply May 1992 7 of Commercial Digital Hardware and Software Components to be used in NUPLEX 80+ Safety Systems, LD-92-Il4 Amendment R ly 30, 1H3 1.6-5

l l

l ABB COMBUSTION ENGINEERING O NUCLEAR POWER COMBUSTION ENGINEERING,INC.

l DEPARTMENT OF ENERGY ADVANCED LIGHT WATER REACTOR ,

CERTIFICATION PROGRAM SYSTEM 80+~ DESIGN CERTIFICATION FIRE HAZARDS ASSESSMENT O

PREPARED BY W.L. INGLES and M.L. EDWARDS f

J- 0- 0; m T.D. CROM ADVANCED NUCLEAR PROGRAMS JANUARY 5,1994 O

l

()

ABB COMBUSTION ENGINEERING NUCLEAR POWER COMBUSTION ENGINEERING. INC.

DEPARTMENT OF ENERGY ADVANCED LIGHT WATER REACTOR CERTIFICATION PROGRAM SYSTEM 80+~ DESIGN CERTIFICATION O FIRE HAZARDS ASSESSMENT REVISION NO. ISSUE DATE DESCRIPTION OF REVISION 0 March 13,1992 Original Issue, Volumes 15 Includes Fire Areas 1175 1 April 5,1993 Revised entire contents of Volume 1 2 August 5,1993 Revised Fire Areas 9, 10, 11, 12, 2-4,44,45,101 and 102. Eliminated q Fire Area 49. Added Areas 48,116

(V and 120. Revised pages 28,29 and 30 of Volume 1

In the separation of equipment and components inside containment.

Reactor Coolant System, motor operated valves which serve e r, 2

pressure bondaries for interconnection to low pressum ~;

(i.e., high-low pressure interfaces) and are required to be closed during normal power operation.will have one of the valve motors in each division deenergized during power operation.

L 7.7 REDUNDANT FIRE AREAS CONTAINING SAFE SHUTDOWN EOUIPMENT The following identifies fire areas that contain equipment required for Safe Shutdown following a fire, and the redundant areas for the opposite division.

REDUNDANT AREA FIRE ' DEA EQUIPMENT 3

1 Div. 1, Channel A Vital Instrumentation 4

2 Div. 1, Channel C Vital Instrumentation Div. 2, Channel B Vital 1 3

Instrumentation 2

4 Div. 2, Charmel D Vital Instrumentation Div. lA, CCW Pump 12 i 9

Div. 1B, CCW Pump 11 10 Div. 2A, CCW Pump 10 11 9 i 12 Div. 2B, CCN Pump 16 15 Div. 1, Control Room HVAC Div. 2, Control Room HVAC 15 ,

16 Channel A Cable 22 21 Div. 1, Div. 2, Channel B Cable 21 22 48 l 24 Div. 2, CCW Piping Div. 2, Cable 33 32 Div. 1, Cable 32 33 Div. 1, Motor Driven Emergency 35 34 28 Rev. 2 l

I.

i I

REDUNDANT AREA FIRE AREA RUIPMENT Feedw?.ter Pump Div. 2, Motor Driven Emergency 34 35 Feedwater Pump ,

Div. 1, Turbine Driven Emergency 37 f 36 Feedwater Pump Div. 2, Turbine Driven Emergency 36 37 Feedwater Pump Div. lA, SI Pump 39 38 Div. 2A, SI Pump 38 39 Div. 2B, SI Pump and 41 40 Div. 2, Shutdown Cooling Pump Div-. lA, SI Pump and 40 41 Div. ' . , Shutdown Cooling Pump Div. 2, Emergency Diesel Generator 43 42 Div. 1, Emergency Diesel Generator 42 43 Cearging Pump 45 44 Div. 1, Div. 2, Charging Pump 44 45

24. l 48 Div. 2, CCW Piping Div. 2, Cable 54 53 Div. 1, Cable 53 54 Div. 2, Cable 56 55 Div. 1, Cable 55 56 Div. Essential Chilled Water 58 57 1, Div. 2, Essentill Chilled Water 57 58 Div. 1, Channel C Cable 64 63 ,

Div. 2, Channel D Cable 63 64 Channel A Equipment 66 65 Div. 1, Div. 2, Channel B Equipment 65 66 Div. 1, Channel C Switchgear 71 70 Div. 2, Channel D Switchgear 70 71 .

Div. 1, Channel C Equipment 74 73 74 Div. 2, Channel D Equipment 73  !

Fuel Pool Cooling Equipm;ent 80 ,

77 Div. 2, O

V 29 Rev. 2

i l

1 REDUNDANT AREA FIRE AREA EOUIPMENT O

78 Div. 2, Valve Gallery 79 78 79 Div. 1, Valve Gallery l Fuel Pool Cooling Equipment 77 80 Div. 1, 1, Emergency Feedwater Tank 83 82 Div.

Emergency Feedwater Tank 82 83 Div. 2, Div. 1, Main Steam Valve House 85 84 Div. 2, Main Steam Valve House 84 85 96 95 Div. 1, Channel A Penetration Room Div. 2, Channel B Penetration Room 95 96 98 97 Div. 1, Channel C Penetration Room 97 98 Div. 2, Channel D Penetration Room 120 116 Div. 1, Channel C Multiplexer Room Div. 2, Channel D Multiplexer Room 116 120 167 166 Div. 1, CCW Surge Tank Div. 2, CCW Surge Tank 166 167 0 .

t t

O 30 Rev. 2

/

i -- ,--a

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

1 4

FIRE AREA 24 I. GENERAL

\ .)

A. DESCRIPTION l i

1. Pipe Chase Located on EL. 50+0 through 170+0 at column N-O and column 11-12.

(See CESSAR-DC Figure 9.5.1-2 through 9.5.1-9)

2. Construction Features

a. Walls Walls are constructed of reinf:rced concrete.

b. Ceiling / Floor Ceiling and floor are constructed of reinforced concrete.

c. Interior Finish There are no combustible inter:.or finish materials in this area.

3. Occupancy

a. This area contains piping.

(1) Safety Division / Channel

( Equipment in the area is associated with safety Division II.

(2) Major Equipment (a) Piping (3) Function for Safe Shutdcwn This area will contain ::mponent Cooling Water and other safety related piptng.

. (4) Importance to plant operation This area contains normal operating system piping.

(5) Location of Redundant Systems / Equipment Equipment which provides redundant safe shutdown functions is located in Fire Area 48 (Division I). l (6) High Energy Equipment / Voltages None (7) Heat Sensitive Equipment None (8) HVAC

--sem,as mm ~*~re-em . ee

(8) HVAC.

~

None (9) Acceptable Level of Risk Category 2 B. OPERATOR ACTIONS

1. Normal Procedures TBD

2. Non-Tire Emergency Procedures TBD C. MAINTENANCE ACTIVITIES _

1. Equipment Disassembly & Laydown TBD

2. Use of Solvents TBD D. OTHER ACTIVITIES _ ,

1. Health Physics TBD

2. Chemistry TBD

3. Testing TBD E. RADIOLOGICAL / TOXIC MATERIAL Radiological or toxic material are not present in the area.

F. POTENTIAL IGNITION SOURCE None G. CURBS, DRAINS. EQUIPMENT PEDESTALS

. None H.

SUMMARY

OF COMBUSTIBLE MATERIALS None II. FIRE PROTECTION FEATURES A. FIXED AUTOMATIC SUPPRESSION SYSTEMS None u

a. Location of Device

) TBD (In accordance with NFPA 101)

b. Annunciation Location TBD

6. Fire Barrier / Insulating Material

a. Walls / Floors / Ceiling (1) Location As shown on CESSAR-DC Figure 9.5.1-6.

(2) Rating Three hour fire rating for walls, floor and ceiling.

(3) Method of Qualification Laboratory test or engineering analysis

b. Doors (1) Location As shown on CESSAR-DC Figure 9 . 5.1 -6 (2) Rating fs Three hour fire rating (w,- )

(3) Method of Qualification Laboratory test or engineering analysis

c. Dampers (1) Location TBD (2) Rating TBD (3) Method of Qualification Laboratory test or engineering analysis

d. Penetration Seals (1) Location TBD (2) Rating Wall, floor and ceiling penetrations are sealed to maintain a three hour rating.

f o

t )

'~~' (3) Method of Qualification i

l

Q

b. Cloths Janitorial Supplies

( ,

(_j}

c.

d. Documents i

II. FIRE PROTECTION FF.ATURES A. FIXED AUTOMATIC SUPPRESSION SYSTEMS None B. MANUAL FIRE SUPPRESSION SYSTEMS

1. Hose Stations  !

i

a. Location i TBD (In accordance with NFPA 14)

b. Length 1 Nominal 75 feet or 100 feet, TBD

c. Nozzle Nonadjustable spray nozzle with fixed angle of spray for use on electrical fire to avoid the possibility of applying a straight stream on the electrical equipment.

2. Fire Extinguisher

a. Location rm

( ) TBD (In accordance with NFPA 10)

%./

b. Type TBD

3. Fire Suppression System Valves

a. Location TBD

b. Control Function TBD

c. Supervision TBD

4. Detection I

a. Type The equipment used is a UL listed or FM approved smoke detection type.

b. ' Selection f or Hazard r^g Ionization detector

/

'~

S. Alarms / Pull Station i

O l 4 l (8) HVAC

,O TeD l (9) Acceptable Level of Risk Category 2 B. OPERATOR ACTIONS

1. Normal Procedures TBD

2. Non-Fire Emergency Procedures TBD C. MLINTENANCE ACTIVITIES

1. Equipment Disassembly & Laydown TBD

2. Use of Solvents TBD D. OTHER ACTIVITIES

1. Health Physics TBD

2. Chemistry TBD

3. Testing TBD E. RADIOLOGICAL / TOXIC MATERIAL TBD F. POTENTIAL IGNITION SOURCE High voltage electrical equipment is a potential ignition source.

G. CURBS, DRAINS, EQUIPMENT PEDESTALS Electrical equipment is mounted on 6-inch pedestals to prevent water infiltration in the event of uncontrolled water release into the area.

H.

SUMMARY

OF COMBUSTIBLE MATERIALS

1. Insitu combustible Materials

a. Cable Insulation 2 -. Transient Combustible Materials

)

a. Cleaning Solvents

4 FIRE AREA 116 f')

• Q I. GENERAL A. DESCRIPTION

1. Division I channel C Multiplexer Located on EL.115 + 6 at columns Q-0 and rows 19-21 (See CESSAR-DC Figure 9.5.1-6).

2. Construction Features

a. W' alls Walls are constructed of reinforced concrete.

b. Ceiling / Floor Ceiling and floor are constructed of reinforced concrete.

c. Interior Finish There are no combustible interior finish materials in this area.

3. Occupancy

a. The area contains electrical equipment.

- (1) Safety Division / Channel Equipment in the area is associated with safety Division I Channel C.

(2) Major Equipment (a) Multiplexer (b) MCC (3) Function for Safe Shutdown TBD (4) Importance to plant operation TBD (5) Location of Redundant Systems / Equipment Equipment which provides redundant safe shutdown functions is located in Fire Area 120 (Division II channel D).

(6) High Energy Equipment / Voltages (a) MCC 480 VAC (7) Heat Sensitive Equipment

/ Multiplexer

( MCC

3

]

'(1) Primary Smoke detection (2) Secondary Fire extinguisher and hose lines 3

b. Passive Systems Fire barriers l

c. Defense-in-Depth The defense-in-depth philosophy involven a' combination of fire barriers and manual suppression systems (i.e. hose lines and extinguishers).

14. Consequences of Fire

a. With Detection & Alarm Systems Functioning The smoke detection system alarms to indicate particles of '

combustion present in the area . Combustibles are such that a slow growth fire would be expected, so that- the possibility of extinguishing a small fire with 'an -

extinguisher is good. The fire brigade responds to insure no smoldering material is left or to use hose lines . to -

extinguish a fire that becomes too intense,

b. Without Detection & Alarm Systems Functioning A fire in the area could possibly cause loss of equipment if ,

+

the smoke detection system did not function. The. redundant equipment in Fire Area 101 (Division I, Channel C) provides redundant safe shutdown functions. The Fire Areas 101 and 102 are separated by three hour fire rated barriers with no communicating openings.

15. Compliance with Design Basis The fire protection features for Fire Area 102 achieve the Design Basis Goals outlined CESSAR-DC Section 9.5.1 " Fire Protection Systems."

b L

O

--- r -- . - - - - - - - - - - -- -- , ,

(3) Method of Qualification

'N Laboratory test or engineering analysis.

d. Penetration Seals (1) Location TBD (2) Rating Wall, floor and ceiling penetrations are sealed to maintain a three hour fire rating.

(3) Method of Qualification Laboratory test or engineering analysis

7. Method of Communication to Control Room /Public Address TBD

8. Personnel. Egress / Fire Brigade Access

a. Primary Stairwell at column P and column 12.

b. Secondary Stairwell at column C and column 10.

9. Potential Effects of Fixed Automatic _ Suppression System Not Applicable

10. Potential Effects of Fire Brigade Activities

a. Water Spray TBD

b. Particles of Cembustion TBD

11. Radiological Consequences of Fire TBD

12. Smoke Control Methods A smoke purge system is provided to remove products of combustion in the event of a fire in 'the area . During smoke purge, the smoke purge fan is started and the area is supplied with make-up air to provide a ence through ventilation system, The-smoke purge mode of operation is manually activated by the control room operator.

4 13 .. Summary of Fire Protection Features

a. Active Systems

. - . .. _ . ...__..m._. _~ . .- _ _ _ _ _ _ _ , . _ . _ _ . _ . . _ _ _ _ . _ _ _ _ . _ _ . . _ _ _ _ . _ . . _ . - _

c. Supervision

! TBD

4. Detection i

a. Type The equipment used is a UL listed or FM approved smoke detection type.

b. Selection for Hazard i

Ionization detector

5. Alarms / Pull Station i

a. Location of Device TBD (In accordance with NFPA 101)

b. Annunciation Location TBD i

6. Fire Barrier / Insulating Material i a. Walls / Floors / Ceiling l

j (1) Location i

As shown on CESSAR-DC Figure 9 . 5 .1 - 5 ..

(2) Rating i

Three hour fire rating for walls, floor and ceiling.

~

(3) Method of Qualification I Laboratory test or engineering analysis '

b. Doors 1

(1) Location

As shown on CESSAR-DC Figure 9.5.1-5.

(2) Rating Three hour fire rating  ;

s (3) Method of Qualification i

e Laboratory test or engineering analysis

c. Dampers 1

[ (1) Location

- TBD

' (2) Rating TBD

+e re m eer -ww er- - -- -

• G. CURBS, DRAINS, EQUIPMENT PEDESTALS

\

Electrical equipment is mounted on 6-inch pedestals to prevent water infiltration in the event of uncontrolled water release into the area.

H.

SUMMARY

OF COMBUSTIBLE MATERIALS

1. Insitu Combustible Materials

a. Cable Insulation (Approximately 1000 feet)

b. Combustible Duct / Pipe Insulation (Limited)

c. Bearing Lubricant (Recirculation Cooling Unit, Minimal)

2. Transient combustible Materials

a. Cleaning Solvents

b. Cloths

c. Janitorial Supplies I

d. Documents

e. Scaffolding Boards II. FIRE PROTECTION FEATURES A. FIXED AUTOMATIC SUPPRESSION SYSTEMS None B. MANUAL FIRE SUPPRESSION SYSTEMS

1. Hose Stations

a. Location O TBD (In accordance with NFPA 14)

b. Length Nominal 75 feet or 100 feet

c. Nozzle Nonadjustable spray nozzle with fixed angle of spray for use on electrical fire to avoid the possibility of applying a straight stream on the electrical equipment.

2. Fire Extinguisher

a. Location TBD (In accordance with NFPA 10)

b. Type Carbon Dioxide

3. Fire Suppression System Valves >

a. Location TBD

b. Control Function

.O TBD

_.. - _ . _ _ .__- _ _. _ .. _ . _ . . _ . _ _ _ _ _ . _ . _ _ _ . . _ _..___._._.._.m_. __

1

.l

+ 1 l

l (b) Decirculation Cooling Units 480 VAC (7) Heat Sensitive Equipment Electronic components are qualified to withstand a minimum ambient temperature of 55 F contirwun or 122 0F for 60 minutes without failure.

(8) HVAC a

During normal plant operations, HVAC is supplf 9d by one '

of two redundant air-handling units serving this area.

Instrumentation and controls provide manual operation of the system from local and/or remote loce.tions .

Indication of f an operating status, damper ; asitions, and high room temperature alarms are provided in the Control Room.

(9) Acceptable Level of Risk Category 2 B. OPERATOR ACTIONS ,

1. Normal Procedures TBD

2. Non-Fire Emergency Procedures TBD C. MAINTENANCE ACTIVITIES

1. Equipment Disassembly & Laydown TBD

2. Use of Solvents  :

TBD D. OTHER ACTIVITIES  :

1. Health Physics TBD 1

2. Chemistry TBD

3. Testing Battery testing I

E. RADIOLOGICAL / TOXIC MATERIAL l Battery f aults, inverters, etc. are possible sources of toxic material

. in the area.

F. POTENTIAL IGNITION SOURCE Electrical equipment is a potential ignition source.

i

)

FIRE AREA 102 (D GENERAL V' I.

l A. DESCRIPTION

1. Division II, Channel D tion-Essential Equipment Room, includes areas on EL. 50+0, EL. 70+0 and EL. 81+0 between columns O-P and rows 13-14, on EL. 91+9 between columns 0-Q and rows 13-15, on EL. 115+6 between columns 0-Q and rows 12-14 ,see CESSAR-DC Figures 9.5.1-2 through 6).

2. Construction Features

a. Walls Walls are constructed of reinforced concrete.

b. Ceiling / Floor Ceiling and floor are constructed of reinforced concrete,

c. Interior Finish There are no combustible interior finish materials in this area.

3. Occupancy a, The area contains non-essential electrical equipment. ,

(1) Safety Division / Channel Equipment 4n the area is. associated with safety Division II, Channel D, but is not safety related.

(2) Major Equipment (a) Multiplex Cabinets (b) Inverter (c) Battery Charger (d) Batteries (e) Power Panel Boards and Distribution Center (f) Recirculation Cooling Units (g) MCC (3) Function for Safe Shutdown 4

!!cne (4) Importance to plant operation Equipment in this area provides power, control, and instrumentation to various non-essential systems.

(5) Location of Redundant Systems / Equipment Equipment which provides redundant functions is located in Fire Area 101 (Division I, Channel C)

(6) High Energy Equipment / Voltages l

(a) MCC 480 VAC

. . ~ . . . . -. . . . - - . - - - - . . - . ~ _ . - - - _ _ . - . . - . - -_ - -. .._ .- . - - - -

• l

l l

l

-l Smoke detection '

(2) Secondary ,

Fire extinguisher and hose lines

b. Passive Systems ,

Fire barriers l

c. Defense-in-Depth The defense-in-depth philosophy involves a ecmbinationiof fire barriers and manual suppression systems (i.e. hose lines and extinguishers). '

14. Consequences of Fire

a. With Detection & Alarm Systems Functioning  !

The smoke detection system alarms to indicate particles of combustion present in the area. Combustibles are such that a slow growth fire would be expected, so that the possibility of extinguishing a small fire with. an extinguisher is good. The fire brigade responds to insure no smoldering material is left or to use hose lines to extinguish a fire that becomes too intense, ,

}

b. Without Detection & Alarm Systems Functioning .

A fire in the area could possibly cause. loss of equipment if .

the smoke detection system did not function. The redundant 6 equipment in Fire Area 102 (Division II, Channel D) provides redundant safe shutdown functions. The Fire Areas 101 and 102 are separated by three hour fire rated barriers with no communicating openings. i

15. Compliance with Design Basis The fire protection features for Fire Area 101 achieve the Design'  !

Basis Goals outlined CESSAR-DC Section 9.5.1 " Fire Protection-  ;

Systems.*

I s

I i

i I

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

t

, b Laboratory test or engineering analysis I

d. Penetration Seals i

(1) Location  !

TBD ,

(2) Rating ,

Wall, f:.cor and ceiling penetrations are sealed to- t maintain a three hour fire rating.

(3) Method of Qualification >

Laboratory test or engineering analysis l 2

7. Method of Communication to control Room /Public Address i

TBD

8. Personnel Egress / Fire Brigade Access

a. Primary Stairwell at column P and column 22. .

b. Secondary l

~;

Stairwell at column C and column 24.  ;

9. Potential Effects of Fixed Automatic Suppression System l O Not Applicable f

10. Potential Effects of Fire Brigade Activities  :

a. Water Spray ,

TBD

b. Particles of Combustion ,

TBD

11. Radiological Consequences of Fire TBD

12. Smoke Control Methods A smoke purge system is provided to remove prodacts of combustion in the event of a fire in the area. During smoke' purge, the smoke purge fan is started and the area is supplied with make-up l air to provide a once through ventilation system. The smoke j purge mode of operation is manually activated by the control room ,

operator. .

-\

13. Summary of Fire Protection Features

a. Active Systems (1) Primary v - 4+<-m- I w- ',F wvw%<

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

TBD

4. Detection

a. Type The equipment used is a UL listed or FM approved smoke-l detection type,

b. Selection for Hazard +

Ionization detector

5. Alarms / Pull Station

a. Location of Device TBD (In accordance with NFPA 101)

b. Annunciation Location i

TBD

6. Fire Barrier / Insulating Material

a. Walls / Floors / Ceiling ,

i (1) Location ,

As shown on CESSAR-DC Figure 9.5.1-5.

i- (2) Rating v Three hour fire rating for walls, floor and ceiling.  !

(3) Method of Qualification Lab 6ratory test or engineering analysis- ,

3

b. Doors (1) Location As shown on CESSAR-DC Figure 9.5.1-5.

! (2) Rating Three hour fire rating i (3) Method of Qualification Laboratory test or engineering analysis 3 c. Dampers ji (1) Location TBD I

(2) Rating

[

!. TBD (3) Method of Qualification

- , . ,,,,,,,,,w, , .,,...-n.

M ^r t

Electrical' equipment is-mounted on 6-inch pedestals to prevent water infiltration in the event of uncontrolled water release into the area.

.->O - H.

SUMMARY

OF COMBUSTIBLE MATERIALS

1. Insitu Combustible Materials

a. Cable Insulation (Approximately 1000 feet)

-- b . Combustible Duct / Pipe Insulation (Limited)

c. Bearing Lubricant (Recirculation Cooling Unit, Minimal)

2. Transient Combustible Materials

a. Cleaning Solvents

b. Cloths c, Janitorial Supplies

d. Documents

e. Scaffolding Boards ,

II. FIRE PROTECTION FEATURES A. FIXED AUTOMATIC SUPPRESSION SYSTEMS None B. MANUAL FIRE SUPPRESSION SYSTEMS

1. Hose Stations

a. Location TBD (In accordance with NFPA 14)

b. Length Nominal 75 feet or 100 feet

c. Nozzle

' ,c.Cjustable spray nozzle with fixed angle of spray for use on electrical fire to avoid the possibility of applying a straight stream on the electrical equipment.

2. Fire Extinguisher

a. Location TBD (In accordance with NFPA 10)

b. Type carbon Dioxide

3. Fire Suppression System Valves

a. Location TBD

b. Control Function

- TBD

c. Supervision

. . _ _ ..m. . . . _ ___ .. _ .. .

4:

(7) Heat Sensitive Equipment Electronic components are qualified to withstand a minimum ambient temperature of 55 #F continuous or 122 0F for 60 minuten without failure.

(8) HVAC During normal plant operations, HVAC is' supplied by one .

of two redundant air-handling units serving this area.

Instrumentation and controls provide manual operation.

of the system from local and/or remote - locations.

Indication of fan operating status, damper positions, and high room temperature alarms are provided in the control Room. ,

(9) Acceptable Level of Risk Category 2 B. OPERATOR ACTIONS

1. Normal Procedures TBD ,

2. Non-Fire Emergency Procedures .

TBD C. MAINTENANCE ACTIVITIES

1. Equipment Disassembly & Laydown TBD

2. Use of Solvents TBD D. OTHER ACTIVITIES

1. Health Physics TBD

2. Chemistry TBD

3. Testing Battery testing E. RADIOLOGICAL / TOXIC MATERIAL Battery faults, inverters etc. are possible sources of toxic material in the area.

F. POTENTIAL IGNITION SOURCE Electrical equipment is a potential ignition source.

O' O. CURBS, DRAINS, EQUIPMENT PEDESTALS w e er, - - ,m , g m m .

FIRE AREA 101 I I. GENERAL A. DESCRIPTION

1. Division I, Channel C Non-Essential Equipment Room, includes areas on EL. 50+0, and EL 70+0 between columns 0-P and rows 20-21: on EL. 91+9 and EL.115+6 between columns 0-Q and rows20-21a (See CESSAR-DC Figures 9.5.1-2 through 6).

2. Construction Features

a. Walls Walls are constructed of reinforced concrete.

b. Ceiling / Floor Ceiling and floor are constructed of reinforced concrete.

c. Interior Finish There are no combustible interior finish materials in this area.

3. Occupancy

a. The area contains non-essential electrical equipment, (1) Safety Division / Channel

,-%g

-- Equipment in the area is associated with safety but is not safety related.

Division I, Channel C, (2) Major Equipment (a) Multiplex Cabinets (b) Inverter (c) Battery Charger (d) Batteries (e) Power Panel Boards and Distribution Center (f) Recirculation Cooling Units (g) MCC (3) Function for Safe Shutdown None (4) Importance to plant operation Equipment in this area provides power, control, and )

instrumentation to various non-essential systems. /

)

(5) Location of Redundant Systems / Equipment Equipment which provides redandant functions is located l in Fire Area 102 (Division II, Channel D)

(6) High Energy Equipment / Voltages l

() (a)

(b)

MCC 480 VAC Recirculation Cooling Units 480 VAC j l

_MM - 4wm_- en m a h l LO I r

THIS NUMBER IS RESERVED FOR FUTURE ASSIGNMENT

.O t

,:OI L

i

11. Radiological Consequences of Fire ,

TBD <;

12. Smoke Control Methods  !

None

13. Summary of Fire Protection Features .

a. Active Systems Fire Extinguishers and Hose Lines

b. Passive Systems s Fire Barrier

c. Defense-in-Depth The def ense-in-depth philosophy involves a combination of

' fire barriers and manual suppression ' systems (i.e. hose lines and extinguishers).

14. Consequences of Fire ,

There are no combustibles in.this area to support fire ignition or propagation.

15. Compliance with Design Basis The fire protection features for Fire Area 24 achieve the Design ,

Basis Goals outlined CESSAR-DC Section 9.5.1 " Fire Protection Systems."

l

?

i l

i l

1 I

i

w- ~ .e - -. , , _ _ .

, ~ - . - . . - ... - . - . . . . . . .

. . ~ ~ . .. .. - . . - . . - - . . ..

j '

, t

'(3). Method of Qualification

~N Laboratory test or engineering analysis f(d b. Doors (1). Location TBD (2) Rating Three hour rating (3) Method of Qualification Laboratory. test or engineering analysis ,

t

c. Dampers ,

None

d. Penetration Seals ,

(1) Location TBD (2) Rating Three hour rating j I

(3) Method of Qualification Laboratory test or engineering analysis

7. Method of Communication to Control Room /Public Address ,

TBD

8. Personnel Egress / Fire Brigade Access  ;

a. Primary TBD

.f

b. Secondary j TBD ,

9. Potential Effects of Automatic Suppression System Not Applicable

10. Potential Effects of Fire Brigade Activities

a. Water Spray. ,

TBD

b. Particles of Combustion TBD l

_ _ .. - ~ . - . . . . _ . . _ . . - . . . - . _ . . _ . _ . . _ . . - . .__ __m i

?

i

. 1. Hose Stations j i

a. Location -

TBD (In accordance with NFPA 14)

b. Length l 4

Nominal 75 feet or 100 feet, TBD .

c. Nozzle  !

Adjustable spray nozzle

2. Fire Extinguisher ,

i

a. Location '

i TBD (In accordance with NFPA 10)  ;

b. Type .j Carbon Dioxide ,

3. Fire Suppression System Valves

?

a. Location -i TBD  !

i

b. Control Function  !

.. TBD j

c. Supervision l TBD

4. Detection f None }

5. Alarms / Pull Station

a. Location of Device

.i TBD (In accordance with NFPA 101)

b. Annunciation Location P TBD

6. Fire Barrier / Insulating Haterial {

a. Walls / Floors / Ceiling (1) Location-As shown on CESSAR-DC Figure 9.5.1-2 through 9.5.1-9 (2) Rating f Three hour rating i l

l l

I lI y

,m,-- g_. ..

g . _ - , ~ w -

s.

None (9) Acceptable Level of Risk Category 2 B. OPERATOR ACTIONS

1. Normal Procedures TBD

2. Non-Fire Emergency Procedures TBD C. MAINTENANCE ACTIVITIES

1. Equipment Disassembly & Laydown TBD

2. Use of Solvents TBD D. OTHER ACTIVITIES

1. Health Physics TBD

2. Chemistry TBD

3. Testing TBD E. RADIOLOGICAL / TOXIC MATERIAL Radiological or toxic material are not present in the area.

F. POTENTIAL IGNITION SOURCE None G. CURBS, DRAINS, EQUIPMENT PEDESTALS None H.

SUMMARY

OF COMBUSTIBLE MATERIALS None II. FIRE PROTECTION FEATURES A. FIIED AUTOMATIC SUPPRESSION SYSTEMS None B. MANUAL FIRE SUPPRESSION SYSTEMS O4 .

1 FIRE AREA 49 I. _CENERAL

.y. -

k A. DESCR*APTION

1. Pipe Chase Located on EL. 50+0 through 170+0 at column N-0 and column 22-23.

(See CESSAR-DC Figure 9.5.1-2 through 9.5.1-9) 2 Construction Features a, Walls Walls are constructed of reinforced concrete,

b. Ceiling / Floor Ceiling and floor are constructed of reinforced concrete.

c Interior Finish There are no combustible interior finish materials in this area.

3. Occupancy

a. This area contains piping.

(1) Safety Division / Channel Equipment Division II. in the area is associated with safety (2) Major Equipment (a) Piping (3) Function for Safe Shutdown This area will contain Component Cooling Water and other safety related piping.

(4) Importance to plant operation This area contains normal operating system piping. ,

(5) Location of Redundant Systems / Equipment Equipment which provides redundant safe shutdown functions is located in Fire Area 24 (Division II).

(6) High Energy Equipment / Voltages None (7) Heat Sensitive Equipment None i t

(8) KVAC

()

I t

+ - -

t

b. Particles of Combustion TBD

11. Radiological Consequences of Fire TBD e

12. Smoke Control Methods None

13. Summary of Fire Protection Features

a. Active Systems Fire Extinguishers and Hose Lines

b. Passive Systems Fire Barrier

c. Defense-in-Depth The defense-in-depth philosophy involves a combination of fire barriers and manual suppression systems (i.e. hose lines and extinguishers).

14. Consequences of Fire There are no combustibles in this area to support fire ignition or propagation.

15. Compliance with Design Basis The fire protection features for Fire Area 24 achieve the Design Basis Goals outlined CESSAR-DC Section 9.5.1 " Fire Protection Systems."

O .

. . . . _ . _ = . - . - .-. - . . . . _.

1) Hating Three hour rating i

(3) Method of Qualification Laboratory test or engineering analysis

b. Doors (1) Location TBD (2) Rating .

Three hour rating (3) Method of Qualification Laboratory test or engineering analysis +

1

c. Dampers

• None c Penetration Seals (1) Location TBD

'+

(2) Rating Three hour rating (3) Method of Qualification t Laboratory test or engineering analysis .

7. Method of.Commun'ication to Control Room /Public Address .

TBD

8. Personnel Egress / Fire Brigade Access

a. Primary TBD

b. Secondary i

TBD

9. Potential Ef fects of Automatic Suppression System-

.r Not Applicable

10. Potential Effects of Fire Brigade Activities

a. Water Spray l

. () TBD  ;

e l

l l

. . . . . . . . ~ . . .- . . . - . . - - - . .. ~. - . .-. . ... .. ., ..

L l

[.

, A

3. MANUAL TIRE SUPPRESSICN SYSTEMS

1. Hose Stations l l

a. Location  ;

TBD (In accordance with NFPA 14)

b. Length Nominal 75 feet or 100 feet, TBD t

c. Nozzle Adjustable spray nozzle

2. Fire Extinguisher

a. Location TBD (In accordance with NFPA 10) .;

b. Type Carbon Dioxide

3. Fire Suppression System Valves >

a. Location f

TBD e.

g b. Control Eunction TBD i *

c. Supervision TBD

4. Detection None

5. Alarms / Pull Station i

a. Location of Device ,

TBD (In accordance with NFPA 101)  :

b. Annunciation Location ,

TBD  :

6. Fire Barrier / Insulating Material ,

'f

a. Walls / Floors / Ceiling (1) Location j As shown on CESSAR-DC Figure 9.5.1-2 through 9.5.1-9 l

.k

)

-4

,-.e-. , ,,-- , - . . , , + , . , v .- - - - , . -

Laboratory test or engineering analysis m

( 7. Method of Communication to Control Room /Public Address TBD

8. Personnel Egress / Fire Erigade Access

a. Primary TBD

b. Secondary TBD

c. Emergency Lighting Emergency lighting will be provided in this area per NFPA 101.

9. Potential Effects of Fixed Automatic Suppression System Not Applicable

10. Potential Effects of Fire Brigade Activities

a. Water Spray TBD

b. Particles of Combustion TBD

11. Radiological Consequences of Fire Not applicable

12. Smoke Control Methods A smoke purge system is provided to remove products of combustion in the event of a fire in the area. During smoke purge, the smoke purge fan is started and the area is supplied withThe make-up smoke air to provide a once through ventilation system.

purge mode of operation is manually activated by the control room operator.

13. Summary of Fire Protection Features

a. Active Systems (1) Primary 1 i

Smoke detection 1 (2) Secondary Fire extinguisher and fire hose

b. Passive Systems )

l (D Fire barriers Q

l

)

-t

c. Defense-in-Depth The defense-in-depth philosophy involves a combination of fire barriers and manual suppression systems (i.e. hoce ,

lines and extinguishers). l Consequences of Fire

14.  ;

l

a. With Detection & Alarm Systems Functioning The smoke detection ~ system alarms to indicate particles of .

combustion present in the area. Combustibles are such that a slow growth fire is expected so that the' possibility of extinguishing a small fire with an extinguisher is good. The fire brigade responds to insure no smoldering material is left or to use the hose line to extinguish a fire that becomes too intense,

b. Without Detection & Alarm Systems Functioning A fire in the area could possibly cause loss of equipment'if the smoke detection system did not function. The redundant '

equipment is located in Fire Area 120 (Division II Channel D) . The Fire Areas 116 and 120 are separated by multiple layers of three hour fire rated barriers.

15. Compliance with Design Basis The fire protection features for Fire Area 116 achieve

" Firethe Design protection Basis Goals outlined CESSAR-DC Section '9.5.1 ,

Systems.*

O  :

t e

I' O

.J .- ,

4 FIRE AREA 120 O I. GENERAL A. DESCRIPTION

1. Division II Channel D Multiplexer Located on EL. 115 + 6 between columns Q-o and rows 13a-15 (See CESSAR-DC Figure 9.5.1-6).

2. Construction Features

a. Walls Walls are constructed of reinforced concrete.

b. Ceiling / Floor Ceiling and floor are constructed of reinforced concrete.

c. Interior Finish There are no combustible interior finish materials in this area.

3. Occupancy-

a. The area contains electrical equipment.

(1). Safety Division / Channel Equipment in the area is associated with safety Division I Channel C.

(2) Major Equipment (a) Multiplex Cabinets (b) MCC  ;

(3) Function for Safe Shutdown TBD (4) Importance to plant operation f i

TBD  !

(5) Location of Redundant Systems / Equipment i Equipment which provides redundant safe shutdown-functions is located in Fire Area 116 (Division I channe1 C).

(6) High Energy Equipment / Voltages 1

(a) MCC 480 VAC l Heat Sensitive Equipment l (7)

Multiplexer 0 MCC

, . .. -~ . . . - -. - . = - - _ . _ - . . _ . . . . . . - . - .-. -

l (8) INAC l

TBD i (9) Acceptable Level of Risk l h

Category 2 i B. OPERATOR ACTIONS

1. Normal Procedures TED

2. Non-Fire Emergency Procedures TBD ,

C. MAINTENANCE ACTIVITIES

1. Equipment Disassembly & Laydown TBD 1

2. Use of Solvents TBD D. OTHER ACTIVITIES

1. Health Physics TED

2. Chemistry TBD

3. Testing TBD E. RADIOLOGICAL / TOXIC MATERIAL TBD  !

F. POTENTIAL IGNITION SOURCE High voltage electrical equipment is a potential ignition source.

G. CURBS, DFAINS,-EQUIPMENT PEDESTALS Electrical. equipment is mounted on 6-inch pedestals to prevent water infi,ltration in the event of uncontrolled water release into the ' area.  ;

i l

H.

SUMMARY

OF COMBUSTIBLE MATERIALS

)

1. Insitu combustible Materials

a. Cable Insulation

2. Transient Combustible Materials

a. Cleaning Solvents j

. . . . -. . ..- . - .. . - . . -~ .-_- .. - - .. - - - .-- . . . .

b. Cloths

c. Janitorial Supplies

d. Documents II. FIRE PROTECTION FEATURES A. FIIED AUTOMATIC SUPPRESSION SYSTEMS None B. MANUAL FIRE SUPPRESSION SYSTEMS 1

1. Hose Stations

a. Location TBD (In accordance with NFPA 14) l

b. Length Nominal 75 feet or 100 feet, TBD

c. Nozzle Nonadjustable spray nozzle with fixed angle of spray for use .

on electrical fire to avoid the possibility of applying a straight stream on the electrical equipment.

2. Fire Extinguisher

a. Location TBD (In accordance with NFPA 10)

b. Type TBD

3. Fire Suppression System Valves

a. Location  ;

TBD

b. Control Function TBD c, Supervision TBD

4. Detection

a. Type The equipment used is a UL listed or FM approved smoke detection type.

b. Selection for Hazard Ionization detector

5. Alarms / Pull Station

~

l

-,n-. , - , , . -- , +---+ c - - - - , ,n

.w.

,__.._, _ . . . . = . . _ = _ = _ _ _ _ . _ . , . _ . . ..

a. Location of Device

()_ TBD (In accordance with NFPA 101)

b. Annunciation Location 1

TBD 1

6. Fire Barrier / Insulating Material

a. Walls / Floors / Ceiling (1) Location As shown on CESSAR-DC Figure 9.5.1-6.

(2) Rating Three hour fire rating for walls, floor and ceiling.

(3) Method of Qualification Laboratory test or engineering analysis

b. Doors (1) Location ,

As shown on CESSAR-DC Figure 9.5.1-6 (2) Rating

() (3)

Three hour fire rating Method of Qualification Laboratory test or engineering analysis *

c. Dampers ,

(1) Location .

TBD (2) Rating TBD (3) Method of Qualification Laboratory test or engineering analysis

d. Penetration Seals (1) Location TBD (2) Rating Wall, floor and ceiling penetrations are sealed to-maintain a three hour rating.

(3) Method of Qualification l

1 q

. . - - ~ - . - .- .. -.-. .-. - . _ - . . . . .-.~.- . = , - . ~

l i

Laboratory test or engineering analysis

7. Method of Communication to Control Room /Public Address TBD 8 Personnel Egress / Fire Brigade Access

a. Primary ,

P TBD

b. Secondary TBD

c. Emergency Lighting ,

Emergency lighting will be provided in this area per NFPA 101.

9. Potential Effects of Fixed Automatic Suppression System Not Applicable

10. Potential. Effects of Fire Brigade Activities

a. Water Spray TBD

b. Particles of Combustion TBD

11. Radiological Consequences of Fire  ;

Not applicable ,

12. Smoke Control Methods  !

A smoke purge system is provided to remove products of combustion ,

in the event of a fire in the area. During smoke purge, the smoke purge fan is started and the area is supplied with make-up air to provide a once through ventilation system. The smoke purge mode of operation is manually activated by the control room j

operator.

13. Summary of Fire Protection Features

a. Active Systems (1) Primary ,

smoke detection (2) Secondary .

Fire extinguisher and fire hose

b. Passive Systems Fire barriers  :

i,

's t

c. Defense-in-Depth ,

The defense-in-depth philosophy involves a combination of fire barriers and manual suppression systems (i.e. hose lines and extinguishers).

14. Consequences of Fire

a. With Detection & Alarm Systems Functioning. .

The smoke detection system alarms to indicate particles of combustion present in the area. Combustibles are such that a slow growth fire is expected so that the possibility of i extinguishing a small fire with an extinguisher is good. The '

fire brigade responds to insure no smoldering material is left or to use the hose line to extinguish .a fire that becomes too intense.

b. ..Ithout Detection & Alarm Systems Functioning  :

A fire in the area could possibly cause loss of equipment if the smoke detection system did not function. The redundant equipment is located in Fire Area 116 (Division I' Channel C). The Fire Areas 116 and 120 are separated by multiple layers of three hour fire rated barriers. ,

15. Compliance with Design Basis The fire protection features for Fire Area 120 achieve the Design- .s Basis Goals outlined CESSAR-DC Section 9.5.1 " Fire Protection' Systems.*

a E

l l

I i

l l

l

= ---a-x a.+ +- ,.w .

a _

6 O

t THIS NUMBER IS RESERVED FOR FUTURE ASSIGNMENT O

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ATTACHMENT 11 4

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J CESSAR 8lnincano.

TABLE OF CONTENTS.(Cont'd)

CHAPTER 2 _

Section Subiect Egge_No _

2.3.5 LONG-TERM (ROUTINE) DIFFUSION. 2.3-3 ,

ESTIMATES (x/Q) 2.3.6 ONSITE (ACCIDENT)~ DIFFUSION 2.3-4 l ESTIMATES (X/Q) 2.4 HYDROLOGIC ENGINEERING 2.4-1 ,

2.4.1 EXTERNAL FLOODS 2.4-1 ,

2.4.2 INTERNAL FLOODS . 2 .~ 4 -1 2.5 GEOLOGY. SEISMOLOGY. AND 2.5-1 GEOTECHNICAL ENGINEERING 2.5.1 BASIC GEOLOGIC AND SEISMIC 2 '. 5-3 INFORMATION 2.5.2 VIBRATORY GROUND MOTION '2.5-3 2.5.2.1 Seismicity 2.5-3 2.5.2.2 Geoloaic and Tectonic 2.5-3 Characteristics of Site and Recion 2.5.2.3 Correlation of Earthauake 2.5-3 Activity with Geoloaic Structure or Tectonic Provinces 2.5.2.4 Maximum Earthcuake Potential 2.5-3 Seismic Wave Transmission ,2.5-4 2.5.2.5

, Characteristics of the Site 2.5.2.5.1 Control Motion '2.5-4 2.5.2.5.2 Generic Soil Sites 2.5-7 A<S*A*S*A $lfe /fttfsnde 0;kW8, A* S** .;

d-[a 3~ I 5;/e .Spee;/*c .Tel.nnic -9*cN- d 5" '

.I Amendment N 11 April 1, 1993 I

.i

q CESSAR !!nLua 9 I i

2.5.2.5 Seismic Wave Transmission Characteristics of the Site 1

2.5.2.5.1 Control Motion The Control Motion design response spectra are anchored to a 0.3g peak ground acceleration. They were developed with the objective of being in full compliance with the SRP requirements as well as the EPRI ALWR recommendations report. Again, to cover a maximum range of possible sites where the System 80+ standard design may be constructed, three separate control motion spectra were developed. These are:

A. Control Motion Spectrum 1 (CMS 1) : This spectrum is included for application at the free-field ground surf ace. It is identical to Regulatory Guide 1.60 (R.G. 1.60) spectrum and it is considered in order to cover sites with deep soil deposits. Furthermore, because CMS 1 is a standardized response spectrum shape, it is considered as the control motion for both rock and soil sites.

B. Control Motion Spectrum 2 (CMS 2): This is a rock outcrop spectrum and is developed to cover sites typical of Eastern North America which could be subjected to earthquakes with high frequency content.

C. Control Motion Spectrum 3 (CMS 3): This is a rock outcrop spectrum and is developed based on recommendations.of the NUREG/CR-0098 (Reference 4) primarily to cover lower frequency motions which may not be covered by CMS 2. It is also greatly enhanced in the high frequency range to cover earthquakes with high frequency content. The maximum spectral acceleration range is extended to 15 Hz, as opposed to 8 Hz which is used in NUREG/CR-0098 motions A md.

All of the above Control Motion Spectra are shown in '

Figure 2.5-5. All three motions (CMS 1, CMS 2, CMS 3) are used for application at rock sites. For soil sites, CMS 2 and CMS 3 are Y

intended for application at the rock outcrop, and CMS 1 is intended for application at the free-field ground surface. All three motions are applied to each of the 13 sites to '

conservatively cover all combinations.

The logic..for selection process of each of these control motion spectra is described in more detail below:

Selection Process for CMS 1 1 The spectrum shape corresponding to this control motion is  ;

as per the requirements of R.G. 1.60. This spectrum shape is chosen in order to be in full compliance with the SRP Section 2.5 requirements as well as the EPRI ALWR recommendations, and is intended to cover deep soil sites.

l Amendment U 2.5-4 December 31, 1993

l I

I CESSAR nainem0,.

1 The control motion is anchored to a peak ground acceleration of 0.3g for the two horizontal directions and the vertical direction.

Selection Process for CMS 2 j The spectrum shape corresponding to this control motion is ,

for application at the rock outcrop surface, is an 1 84 percentile curve, and is developed considering NUREG/CR-0098 recommendations as well as ground motions deemed appropriate for the Eastern North American continent. The  ;

intent of this spectral shape is to cover various soil i sites over-laying a competent material as well as having rock outcrop motion characteristics typical of Eastern North America. The construction of this spectrum shape is shown in Figure 2.5-6. As can be noted from this figure, the spectral ordinates were kept equal to those obtained using NUREG/CR-0098 for frequencies lower than 3.3 Hz, with maximum ground velocity of 24 in/sec/g, which again is typical of expected earthquakes for the Eastern United States. For higher frequencies, particularly above 10 Hz, the selected spectral ordinates are based upon ground motion estimates appropriate for Eastern North America and, as can 1 be seen, are signi-ficantly higher than those obtained using )

I the NUREG/CR-0098.

This control motion is anchored to a peak ground acceler-ation of 0.3g and peak ground velocity of 7.2 in/sec for the two horizontal directions. In the vertical direction, the control motion is anchored to a peak ground acceleration of 0.2g and peak ground velocity of 4.8 in/sec. The selection  !

of 0.2g at the rock outcrop for the vertical direction leads  !

to vertical spectra at the ground surface that equal or ]

exceed the horizontal spectra at the ground surface over a ,

significant range of frequencies for most of the soil cases. l Selection Process for CMS 3 1

The spectrum shape corresponding to this control motion is  !

developed for application to rock outcrop surface, is an I 84 percentile curve, and is in full compliance with the 1 recom-mendations of NUREG/CR-0098 with maximum ground velocity of 36 in/sec/g representing typical sites in Western North America. CMS 3 is greatly enriched in the high  ;

frequency end of the spectrum to cover earthquakes with high  !

frequency content. The maximum spectral acceleration range l extends from 2.2 Hz to 15 Hz. Again, this control motion is i anchored to a peak ground acceleration of 0.3g for the two l horizontal directions and 0.2q for the vertical direction. 1 2.f. 2.f. 3 Sh'c /?CCephLMe &s%sH2. 1 The CMS 1, CMS 2, and CMS 3 control motions were developed for application in the seismic design of the System 80+ Standard v Design. As discussed in Section 2.5 B, for a site to be F acceptable for construction, the COL applicant must meet the 4

/V,WJw hh J.4~ 2.43 Amendment U l h//*N f8db 42.f.2,4 2- 2.5-5 December 31, 1993 '

CESSAR h"Jho,.

1 Facceptance criteria for the SSE control motion, as outlined ,

in Figure 2.5-38. According to these acceptance criteria:

1. For a rock site, the COL applicant will develop site-specific response spectra at 5% damping for the horizontal and vertical directions, and compare them to _

the envelope of the CMS 1, CMS 2, and CMS 3 ~ control I motions (all with 5% damping).

N <

If the site-specific response spectra are enveloped by 4

h*

q the envelope of the CMS 1, CMS 2, and CMS 3 response

,, spectra, the site is acceptable for coastruction.

. 7 ,

]'y ( If the site-specific ' spectra exceed the envelope of the CMS 1, CMS 2, and CMS 3 response spectra at any frequency

,G range, a limited site-specific evaluation will be l performed. Then, in-structure response spectra at six l $ r) j

'3 \ critical locations obtained from the limited site-N( specific evaluation will be compared to the design ,

response spectra (envelope of all generic rock and soil q

4 cases). If the in-structure spectra from the site-

1. ) specific evaluation are within 10% of'the envelope of g< the in-structure design spectra for each of the six ,f

" locations, the System 80+ is certified for the site. If Y M the in-structure spectra- .from the site-specific i evaluation exceed the envelope of the in-structure design spectra for each of the=six locations by more q .

than 10% at any frequency range, a confirmatory site-q o 1 x specific evaluation must be performed.

\ The critical locations are:

a. Foundation Basemat Elevation +50 ft.

b. Interior Structure Elevation +91.75 ft.

c. Control Room Elevation +115.5 ft. (Areas 1 and 2)

d. Top of Steel Containment Vessel Elevation +251 f t.

e. Interior Structure Elevation +146 ft.

f. Shield Building Elevation +263.5 ft.

2. For a deep or shallow soil site, the COL applicant will develop site-specific response. spectra at 5% damping for the horizontal and vertical directions at the, free-field ground surface.- ,The site-specific free-fi'e,ld surf ace spectra will then be compared to the envelope of the CMS 1 spectra and the surface spectra from- CMS 2- and CMS 3 control. motions (all with 5% damping). These envelope ground surface spectra are -

shown in Figures 2.5-39 and 2.5-40 for the horizontal ,

and the vertical directions, respectively.-

If the site-specific surface spectra are enveloped by the envelope of the CMS 1 spectra and the surface spectra from CMS 2 and CMS 3, the site is acceptable for e construction.

Amendment U' December 31, 1993 2.5-6

CESSAR nuincano,,

~O

('

If the site-specific spectra exceed the envelope of the

[ CMS 1 spectra and surface spectra from CMS 2 and CMS 3 at h )' any frequency range, a limited site-specific evaluation

.( x9d' will be performed. Then, in-structure response spectra at six critical locations obtained from the limited 3

i

.T h; site-specific evaluation will be compared to the in-structure design response spectra (envelope of all j

\ generic rock and soil cases). If the spectra from the

\ N ] )) j site-specific evaluation are within 10% of the envelope of the design spectra for each of the six locations, gf Y

\ ], ]g ) the System 80+ is certified for the site. If the y , j spectra from the site-specific evaluation exceed the envelope of the design spectra for each of the six

f

/ locations by more than 10% at any frequency range, a V confirmatory evaluation must be site-specific performed.

The same critical locations as outlined in Item 1 above are used.

Synthetic Time Histories Synthetic time histories were generated for each of the components, Horizontal-1, Horizontal-2 and Vertical, of each of the control motions CMS 1, CMS 2 and CMS 3, respectively.

The spectral ordinates calculated for each synthetic time history and the corresponding smooth spectra are shown in Figures 2.5-7 through 2.5-9 for the CMS 2 motion, Figures 2.5-28 through 2.5-30 for the CMS 1 motion, and Figures 2.5-31 through 2.5-33 for the CMS 3 motion. The spectral ordinates of each synthetic time history conservatively envelop the target smooth spectra at a sufficient number of frequency points to satisfy the SRP Section 2.5 criteria for development of synthetic time histories.

The characteristics of each synthetic time history (accelerogram, velocity and displacement time histories and Power Spectral Density (PSD) ) are presented in Appendix 2B.

The average PSD of CMS 1 fully complies.to the SRP Section 3.7.1, Appendix A guidelines for Power Spectral Densities of motions that are based on a Regulatory Guide 1.60 spectral-shape. For all three motions CMS 1, CMS 2 and CMS 3, the synthetic time histories in the three directions are statistically independent with correlation coefficients less than 0.2.

2.5.2.5.2 Generic soil Sites Generic soil sites were selected by first choosing four generic -1 site categories. These categories were chosen to represent appropriate total thickness of soil overlying bedrock. The four categories are shown schematically in Figure 2.5-1. Site Category A consists of 52 feet of soil overlying bedrock; 52 feet l

Amendment U l 2.5-7 December 31, 1993 )

1 l

CESSAR n5 JLno,. I 1

l I

i is the embedment depth selected for the System 80+. The soils in site Category B extend to a depth of 100 feet and those in Categories C and D extend to depths of 200 and 300 feet, respectively.

One case was selected for Category A and one case for Category D; these were designated Case A-1 and Case D-1. Four cases were initially selected for site Category B; these were designated Cases B-1, B-2, B-3 and B-4. Three cases were initially selected for site Category C; these were designated Cases C-1, C-2 and C-3. Upon examination of the results of the response analyses for these cases, three additional cases were added. The additional cases were designated Cases B-1.5, B-3.5 and C-1.5.

These latter cases were selected to provide an estimate of the response at frequencies that were not considered to be adequately covered by the other cases.

The variations of maximum shear wave velocities with depth assigned for each case are summarized in Appendix 2A Figures 2A-2 through 2A-13. The shear wave velocity distribution with depth was selected to provide a reasonably wide range and also to provide significant contrast in velocities at certain depths for a selected number of cases. The range of maximum shear wave velocities used for all the cases considered in this study is presented in Figure 2.5-2. More details about each case are given in Appendix 2B.

The variation of shear modulus with shear strain was based on I,

d using the upper curve from the range published by Seed and Idriss q l (Reference 5) as shown in Figure 2.5-3. The variations of Q damping with shear strain was based on the lowered curve from the y range published by the same authors, as shown in Figure 2.5-4. gf y

.v5 A 4 3

.5.2.6 Safe Shutdown Earthcuake For the Safe Shutdown Earthquake (SSE) , the following Peak Ground Accelerations (PGA) were considered:

CMS 1 motion:

Horizontal PGA = 0.3g Vertical PGA = 0.39 CMS 2 motion:

Horizontal PGA = 0.3g Vertical PGA = 0.2g CMS 3 motion:

Horizontal PGA = 0.3g i Vertical PGA = 0.2g i Amendment Q 2.5-8 June 30, 1993

g Insert 2.5.2.5.4 2.5.2.5.4 Site Specific Seismic Spectra The COL applicant will be required to develop / site-specific seismic design response spectra for use in the design and qualification of site-specific struc tures, systems, and components not included in the design certification scope for System 80& standard plants. The following criteria shall be used in developing the minimum site-specific seismic design requirement

1. The horizontal and vertical free-field ground surface site-spet'fic  ;

response spectra shall be developed using approved NRC procedures.

2. The System 80+ certified design horizontal and vertical Regulatory Guide 1.60 design response spectrum shapes anchored to 0.30g peak ,j ground acceleration shall be scaled throughout their entire frequency y range such that the minimum spectral amplitudes of the certified design spectra are equal to the maximum spectral amplitudes of the ,

horizontal and vertical site-specific ground motion spectra,respectively, in the 5 to 10 hertz frequency range.

l

3. The resulting design response spectra shall be defined as the minimum )

seismic design requirement for design and qualification of site specific '

structures, systems, and camponents for the System 80& standard plant. l l

I l

l

)

CESSAR !!nLuion LIST OF FIGURES APPENDIX 2B Figure Subject 2B-1 Synthetic Acceleration, Velocity and Displacement '

Time Histories, CMS 1, Component H1 2D-2 Synthetic Acceleration, Velocity and Displacement Time Histories, CMS 1, Component H2 2B-3 Synthetic Acceleration, Velocity and Displacement Time Histories, CMS 1, Vertical Component ,

1 2B-4 Average Power Spectral Densities, CMS 1, Components H1 and H2 2B-5 Average Power Spectral Density, CMS 1, Vertical Component 2B-6 Synthetic Acceleration Time History, CMS 2, component H1 2B-7 Velocity -- Synthetic Time History, CMS 2, Component H1 2B-8 Displacement --

Synthetic Time History, CMS 2, Component H1 2B-9 hf-ffePowerSpectralDensity# " % .. m u m Il - :i A A Y A CMS 2, Component H1 Time History, CMS 2, 2B-10 Synthetic Acceleration Component H2 ]

2B-11 Velocity --

Synthetic- Time History, CMS 2, Component H2 Displacement --

Synthetic Time History, CMS 2, 2B-12 Component H2 2B-13 /Avryt. Power Spectral Densityy v Si ..thetim 1-c n m r" /. l CMS 2, Component H2

~ 2B-14 Synthetic Acceleration Time History, -CMS 2, Vertical Component j Amendment N 11 April 1, 1993 4 ,

<-s. , -n... . . . - , 4 .

CESSAR !!!nneum LIST OF FIGURES (Cont'd)

APPENDIX 2B Figure Subject 2B-15 Velocity -- Synthetic Time History, CMS 2, Vertical Component 2B-16 Displacement -- Synthetic Time History, CMS 2, Vertical Component 2B-17 gyenjt e Power Spectral Densityy W S ..'

CMS 2, Vertical Component i htiv Il  !!i--$. 7 2B-18 Synthetic Acceleration, Velocity and Displacement Time Histories, CMS 3, Component H1 2B-19 Synthetic Acceleration, Velocity and Displacement Time Histories, CMS 3, Component H2 2B-20 Synthetic Acceleration, Velocity and Displacement Time Histories, CMS 3, Vertical Component 2B-21 Average Power Spectral Density, CMS 3, Component H1 2B-22 Average Power Spectral Density, CMS 3, Component H2 2B-23 Average Power Spectral Density, CMS 3, Vertical component b

i Amendment N 111 April 1,_ 1993

+ -

~ . - - .., . - , . . . . . - , - , . . ,

CESSAR !al%uio,.

APPENDIX 2B CHARACTERISTICS OF SELECTED CONTROL MOTIONS ABSTRACT The synthetic time histories generated to represent. the horizontal components H1, H2 and the vertical component of control motions CMS 1, CMS 2 and CMS 3 are presented in this Appendix.

The acceleration, velocity and displacement time histories of control motion CMS 1 are shown in Figures 28-1 through 2B-3. The average Power Spectral Densities (PSD) of the CMS 1 synthetic time Y'

historiesj are shown in Figures 2B-4 and 2B-5.

The acceleration, velocity and displacement time histories of control motion CMS 2 are shown in Figures 2B-6 through 2B-8, 2B-10 through 2B-12, and 2B-14 through 2B-16. The Power Spectral Densities (PSD) of the CMS 2 synthetic time historiesgare shown in /

Figures 2B-9, 2B-13 and 2B-17.

The acceleration, velocity and displacement time histories of' control motion CMS 3 are shown in Figures 2B-18 hrough 2B-20.

The average Power Spectral Densities (PSD) of the CMS 3 synthetic time histories are shown in Figures 2B-21 through 2B-23. /

4 ,

The selection process for CMS 1, CMS 2 and CMS 3 are given in Section 2.5.2.5.1.

4 V A Y Y d$r ('e$f dC / M,

$f* dl* f3D3 I

Iw+1 e

Amendment N 2B-1 April 1, 1993

Insert 1

% average PSDs for the CM81, CM82 and CMS 3 control time histories are developed using the procedure described in SRP, Section 3.7.1, Appendix A. The target PSD for CMS 1 (horizontal) motion (E Reg. Guide 1.60 horizontal) is obtained directly from SRP, Section 3.7.1, Appendix A. The methodology for the development of the target PSDs for CMS 1 (vertical), CM82 and CMSS is described below, pud B GN Methodolorv for Develonmant of Tawet Power Snectral Denaitien

~

The development of target PSDs fo CMS 1 (vertical) and the rock outcrop motions CMS 2 and CMSS is performed sing principles f Random Vibration Theory (RVT). Details of this metho as well as the mathematical formulation are described in Reference s asq. The br. i approach is that the target PSD is developed by an iterative proess. MC step of the iteration, the PSD is refined to produce a spectrum th: closely matches the target response spectrum.

Adjustn2ents to the PSD are made at the frequency ranges that do not produce a close spectral match, and the final target PSD is obtained when the desired spectrum convergence is achieved. The minimum check is set at 80% of the target PSD, consistent with SRP guidelines.

N development of the target PSDs is performed using the 2% damped spectrum as the target spectrum each control motion.

e

/

W5NAAOES fdA OffEW.Dff 88 I. Boore, D. M., " Stochastic Simulation of High Frequency Ground Motions Based on Seismological Modele of the Radiated Spectra", Bulletin of the Seismological Society of America, Volume 73, Number 6, pp.1865-1894.

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CESSARinWncum TABLE OF CONTENIS_ (Cont'd)

CHAPTER 3 1 1

l Section Subiect Pace No.

3.1.52 CRITERION 61 - FUEL STORAGE AND 3.1-40 HANDLING AND RADIO-ACTIVITY CONTROL 3.1.53 CRITERION 62 - PREVENTION OF 3.1-40 CRITICALITY IN FUEL STORAGE AND HANDLING 3.1.S4 CRITERION 63 - MONITORING FUEL AND 3.1-41 WASTE STORAGE 3.1.55 CR; TERION 64 - MONITORING RADIO- 3.1-41 ACTIVITY RELEASES 3.2 CLASSIFICATION OF STRUCTURES. 3.2-1 M ONENTS, AND SYSTEMS 3.2.1 SZI!!MIC CLASSIFICATION 3.2-1 -

3.2.2 SYSTEM QUALITY GROUP CLASSIFICATIONS 3.2-3 (SAFETY CLASS) 3.3 WIND AND TORNADO LOADINGS 3.3-1 3.3.1 WIND LOADINGS 3.3-1 3.3.1.1 Desian Wind Velocity 3.3-1 3.3.1.2 Determination of ADD 1ied Forces 3.3-1 .

3.3.2 TORNADO LOADINGS 3.3-1 3.3.2.1 Applicable Desian Parameters 3.3-1 1

3.3.2.2 Determination of Forces on Structures 3.3-2 3

3 . 3 .2::2: I'

-Effect of Failure of Structures or 3.3-2 l Components Not Desianed for7 Tornado Loa

[ l g

3.4 WATER LEVEL (FLOOD) DESIGN 3.4-1 l 3.4.1 FLOOD ELEVATIONS 3.4-1 Qjf /jnj ,g,,

Amendment D 4 v September 30,-1988 )

)

hhk ICAT12N )

TABLE OF CONTENTS (Cont'd)

CHAPTER 3

,sy*sh.<s n0 Section f Subiect Pace No.

3.7.2.7 Copbination of Modal Responses 3.7-16a 3.7.2.7.1 Seismic Category I Structures, Systems 3.7-16a and Components other Than NSSS , l 3.7.2.7.2 Nuclear Steam Supply System 3.7-16b l 3e ... Erl__ j 3.7.2.8 InteractionofNon-;..ispx(AN.f*f__ ,

3.7-16b structures %ith C_f t c Al Md Je/5Ned (s/gery .T* Structures fej tinses s Ens / 67,9ansxf s J 3.7.2.9 Effects of Parameter Variations on 3.7-17 l Floor Response Spectra l l

3.7.2.10 Use of Constant Vertical Static Factors 3.7-17 j i

3.7.2.11 Methods Used to Account for Torsional 3.7-17 Effects 3.7.2.12 Comparison of Responses 3.7-18a 3.7.2.13 Methods for Seismic Analysis of Dams 3.7-18a l 3.7.2.14 Determination of Safetv-Related 3.7-18a ,

Structure Overturnine Moments l l

3.7.2.15 Analysis Procedure for Damoina 3.7-18a l l

3.7.3 SEISMIC SUBSYSTEM ANALYSIS 3.7-19 l 3.7.3.1 Seismic Analysis Methods 3.7-19 3.7.3.2 Determination of Number of Earthauake 3.7-20 Cycles 3.7.3.3 Procedure Used for Modelina 3.7-20 i

3.7.3.4 Basis for Selection of Frecuencies 3.7-21 l l

l Amendment I xii December 21, 1990

.CESSAR !!nhm.

TABLE 3.2-1 (Cont'd)

-(sheet 22 of 26)

CLASSIFICATION OF STRUCTtstE's. SYSTEMS. $2 COFONENTS ,

Safety selseic Cm oonent identification Ctess .. Category toestion Gustity Class Nuclear Anwa 3 I NA 1 Control Area 3 NA 1 EFW Tank / Main Steam 1 valve House Area I 1 Emergency Diesel 3 MA Generator Areas CVCS/ Maintenance Ares 3 1 NA 1 l Fuel Handling Area 3 1 MA 1 NA/Rs 2 l Unit vent % NNS  !!

Turbine Building NNS  !! TB 2 Redunste Building (28) NNS 11 RW 2 Station Service Water 3 I SP 1 Pump / Intake Structure Camponent Coollrg Water 3 I CX/ 1 Y

Noet ExchangerStructure[g #

M 8p 75,2*C.k Diesel Fuel Storage structure 3 I DF 1 Station Services Building / Auxillary NNS NS SE 3

+

Boiler Str1Jcture Administration Building NNS NS ADS 3 Warehouse MNS NS WM 3 Fire Pump Nause HMS NS FP 3 YA 2 l Dlke (Holdp, Boric Acld Storage kNS  !!  ;

and Reactor Make@ Waterianks) (23)

Dike (Condensate Storege Tank) (28) WNS 11 YA 2 I

Cranee Polar Crane MNS 11 RC 2 Cask Handling Holst NNS  !! MA 2 New Fuel Handling Noist NNS II NA 2 Ccaponent Stgports (23) 1/2/3/NNS 1/MS ALL 1/2/3 Amendment U December 31, 1993

CESSAR annnemo,.

Maximum wind speed: 330 mph Rotational speed: 260 mph i

Translational velocity: 70 mph l

Radius: 150 feet ,

Maximum pressure differential: 2.4 psid Rate of pressure drop: 1.7 psi /second-Missile Spectra: See Table 3.5-2 3.3.2.2 Determination of Forces on Structures The forces on Seismic Category I structures due to tornado wind loadings are obtained using methods outlined in Section 3.3.1.2, with a wind velocity of 330 mph (vector sum of all component velocities -

assumed constant with height) . Velocity profiles are determined as outlined in Section 3.3.1.1. Effective '

pressure distribution loads are transformed into equivalent static building forces as outlined in Section 3.3.1.2. In determining tornado wind loadings, both the importance factor and gust factors are taken as unity.

Tornado loadings include tornado wind pressure,. internal pressure due to tornado created atmospheric pressure drop, - and forces generated due to the impact of credible tornado missiles. These loadings are combined with other loads -f s- daygti bed in Section 3 B-g = eIsfe /JderJins.]

3 . 3 .3:gr. _

Effect of Failure ofi , Structures or Components Eg1

.g p g, T Desicned forj ornadouLoada

-4 Adiacent structures,awill not be permitted to affect or degrade

( the capability of Seismic Category I structuresjto p_erform their t

intended safety functions,22 -  ::;1t :f t:rrr'- cr"ir; . This is accomplished by one of the following methods:

A. Designing the adjacent structurej to Seismic Category I toind n), tornado loadings. gg gemjpg,ng Y B. Investigating the effect of adjacent structural failure on Seismic Category I structures to determine that no impairment of function results. '

C. Designing a structural barrier to protect Seismic Category I structuresjfrom adiacent structural failure. _

gMk L S

..nnMidKY g ),<.S

.,,benzdo /oads i

f $ ,N- )Q Amendment'I ,

3.3-2 December 21, 1990

CESSAR!anen-lengths of piping runs. The RB Subsphere also provides for close proximity of equipment to reduce piping runs from containment.

Flood barriers have been integrated into the design to provide further flood protection while minimizing the impact on maintenance accessibility. The primary means of flood control in the Nuclear Annex and RB Subsphere is provided by the divisional wall which serves as a barrier between redundant trains of safe shutdown systems and components. Each half of the Subsphere is further divided into two quadrants to separate redundant safe shutdown components to the extent practical. Flood barriers provide separation between Subsphere quadrants, while maintaining equipment removal capability. Emergency Feedwater pumps are located in separate compartments within the quadrants with each compartment protected by flood barriers.

Penetrations are sealed and no doors are provided up to EL. 70+0, the maximum internal flood in the divisional wall that separates the Nuclear Annex and the Reactor Building Subsphere. Where flood doors are provided, open and close sensors are also provided with status indication. Flood barriers also provide separation between electrical equipment and fluid mechanical systems at the lowest elevation within the Nuclear Annex. At higher elevations, safety-related- electrical components are elevated above the floor so that flooding events will not affect components. Additional barriers (e.g., curbs, sealed penetrations) are provided or safety-related' electrical components are elevated, as necessary, to mitigate the effects of postulated pipe rupture addressed in Section 3.6.

Flood protection is also integrated into the floor drainage system. The floor drainage systems are separated by division and Safety Class 3 valves are provided to prevent backflow of water to areas containing safety-related equipment. Each subsphere quadrant is provided with redundant Safety Class 3 sump pumps and associated instrumentation, which are powered from the diesel generators in the event of loss of offsite power.

%j The Nuclear Annex floor drainage system is divisionally separated, with no common drain lines between divisions. Floors T 'I are gently sloped to allow good drainage to the divisional sumps.

I f --Y J

Flood protection is incorporated into the Component Cooling Water Heat Exchanger Structure. This structure is divisionally separated by a wall such that a flood in one division can not flood the other division.

The Diesel Generator Building floor drain sump pumps and associated instrumentation are Safety Class 3 to prevent flooding of the diesel generators. These pumps are also powered from the diesel generator in the event of loss of offsite power.

Amendment P 3.4-3 June 15, 1993

Insert to section 3.4

@ No water lines are routed above or through the control room and the computer room, y HVAC water lines contained in rooms around the control room are located in rooms with I raised curbs to prevent leakage from entering the control room.

I l

I l

CESSAR innneim.

Given the separation criteria above, and the pipe break criteria in Section 3.6.2.1.2, the effects of high-energy pipe breaks are not analyzed where it is determined that all essential systems, components, and structures are sufficiently physically remote from a postulated break in that piping run.

3.6.2 DETERMINATION OF BREAK LOCATIONS AND DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING Described herein are the design bases for locating breaks and cracks in piping inside and outside containment, the procedure used to define the thrust at the break location, the jet impingement loading criteria, and the dynamic response models.

The COL applicant will provide final designs of high- and moderate-energy fluid systems. The final designs and results of high- and moderate-energy piping. analyses will be documented'in i a pipe break analysis report. fan inspection of the as-built \/

high-energypipingsystemswildbeperformed.

The inspection of ,

the as-built high energy pipe reak features shall be performed to verify: gg g,g The location of pipe break mitigation devices (restraints, jet shields)

- Clearances / gaps between restraints and piping

- The location of nearby safety-related targets to be protected from high-energy line breaks.

Any differences be' t ween the as-built information and the as-j designed information will be reconciled and documented in a pipe break analysis report.

3.6.2.1 Criteria Used to Define Break and crack Locations and Confinurat1ons 3.6.2.1.1 General Requirements postulated pipe ruptures are considered in all plant piping systems and the associated potential for damage to required systems and components is evaluated on the basis of the energy in '

the system. System piping is classified as high-energy or moderate-energy, and postulated ruptures are classified as circumferential breaks, longitudinal breaks, leakage cracks, or through-wall cracks. Each postulated rupture is considered separately as a single postulated initiating event.

l l

Amendment U l 3.6-12 December 31, 1993 l

l INSERT TO 3.6.2 i

. l The Pipe Break Analysis Report shall provide the results of the l pipe break analyses. These analyses shall be based on criteria '

used to postulate cracks and breaks in high- and moderate-energy j piping systems as defined in Section 3.6.2 and shall employ the analytical methods described in Section 3.6.2 and Appendix 3.6A. I I

For postulated pipe breaks, the Pipe Break Analysis Report shall ,

confirm that:

i (1) piping stresses in the containment penetration area are within their allowable stress limits, (2) pipe whip restraints and jet shield designs are capable of mitigating pipe break loads, and / 1 i

(3) loads on safety-related systems, structures and 3 components are within their design load limits.

The Pipe Break Analysis Report shall also confirm that structures, systems and components required for safe shutdown can withstand the environmental effects of postulated cracks and breaks.

w - _

-o CESSAR iPencamn 14 Irrespective of the fact that the criteria in Section 3.6.2 may not require specific breaks, if a structure outside containment separates a high-energy line from an essential component, that separating structure is designed to withstand the consequences of the pipe break in the high-energy line that produces the greatest effect on the structure. Structures inside containment which are used to separate high-energy lines from essential-components are designed to withstand the dynamic load effects of postulated pipe breaks not eliminated by leak-before-break. In additioA

• base structures inside containment are adequately igned t withstand the greatest effect from (1) pipe breaks no elA minated by leak-before-break, (2) the largest through-wall 1 kage crack Y

in the high-energy line (minimum 10 gpm) whethe consideration of dynamic effects is eliminated by LBB for that line, or (3) the largest leak from another leak source, such as a valve or pump seal.

3.6.2.1.2 Postulated Rupture Descriptions A. Circumferential Break A circumferential break is assumed to result in pipe severance with full separation of the two severed pipe ends unless the extent of separation is limited by consideration of physical means. The break plane area (A,) is assumed perpendicular to the longitudinal axis of the pipe, and.is assumed to be the cross-sectional flow area of the pipe at '

the break location. The break flow area (Ag ) from each of the broken pipe segments for a circumferential break, with

' full separation of the two broken pipe segments, is equal to the break plane area (A ). The break flow area, discharge coefficient and discharge correlation are substantiated analytically or experimentally.  ;

B. Longitudinal Break A longitudinal break is assumed to result in a split of the  :

pipe wall along the pipe longitudinal axis, but without- l severance. The break plane area (A,,) is assumed parallel to )

the longitudinal axis of the pipe and equal to the 1 cross-sectional flow area of the pipe at the break location. l The break flow area (A f ) is equal to the break plane area 1 (A The break is assumed to be circular in shape or '

ell)iptical (2D x D/2) with its long axis parallel to the axis. The discharge coefficient and any other values used for the area or shape associated with a longitudinal break l are substantiated analytically or experimentally. l l

C. Leakage Crack A leakage crack is assumed to be a crack through the pipe wall where the size of the crack and corresponding flow rate j are determined by analysis and a leak detection system, as described in Section 3.6.3.

Amendment U l 3.6-13 December 31, 1993 i

'CESSAR nnincum i

1

'For hand calculation the break mass flow-rate is obtained from a critical flow correlation which predicts an upper bound flow rate

~for the rupture geometry and fluid state under consideration. '

Examples are the Moody correlation (two-phase and saturated steam conditions), the Homogeneous Equilibrium Model (single phase steam), and the Henry-Fauske correlation (subcooled liquid).

Blowdown flow rate is obtained from the following equation per l ANSI /ANS-56.10:

W = C oAG e where: W = mass flow rate Cp = discharge coefficient A = break area Ge = critical mass flux The break fluid enthalpy is set equal to the stagnation enthalpy of the fluid in the ruptured pipe. A flow discharge coefficient i of 1.0 is used unless a lower value is justified as required by

(

l ANSI /ANS-56.10.

For complex systems and where less conservative release rates are needed, computer analysis is employed. Initial conditions (e.g. ,

fluid pressure, fluid temperature) are chosen within normal  ;

operating limits such that the set which will result in the ,

largest release rates are used. A system model of appropriate ,

complexity is generated and computer programs of the RELAP4 type are used. To calculate the pipe break response, the fluid system is divided into discrete volumes (control volumes or nodes) which are connected to other volumes by a junction. The' equations'of conservation of mass and energy are solved in the nodes, and the one-dimensional momentum equation is solved in the flow paths.

A time history of system cond.itions is output ' by the code. i CEFLASH-4A (Section 3. 9.1. 2.M , RELAP4 / MOD 5, and RELAPS/ MOD 3 '

Y i '

(Reference 15) are computer / codes applicable to the generation of  ;

mass and energy releases j Also, SGNIII (Section 6. 2.1. 4. 4 ) may j be used in the case offrain steam line breaks. ]

3.6.2.5.3 g compartment Pressurization Analysis and Environmental Pressure and Temperature Analysis Compartment pressurization analysis is performed to determine i pressure loadings on building structures. Environmental pressure l and temperature response analysis defines pressure and temperature conditions for qualification of mechanical and electrical equipment.

Computer codes are generally used in some phase of this analysis.

Typically the'model includes a network of volumes and junctions.

volumes represent rooms, corridors, pipe chases, and other portions of buildings outside Containment. When appropriate,-

volumes also are- used-to simulate the - HVAC system and outside atmosphere. Junctions represent flow paths between the volumes.

Multinode analysis may be required within a compartment. The computer codes addressed below provide acceptable results for Amendment R 3.6-30 July 30, 1993

CESSARENnnean  :

both compartment pressurization and environmental pressure and temperature analyses, with appropriate assumptions and models char.ged to obtain conservative results.

The DDIFF-1 computer code (Reference 19) is used to predict subcompartment conditions following incident -initiation during which the maximum pressure differentials on structures or components would occur. The transient calculations include determination of mass flow rates, mass and energy inventories, absolute and differential pressures, and temperatures in .the subcompartment system. The subcompartment system is a control volume-flow path spatial network created based upon the geometry of the plant regions being analyzed.

RELAP4/ MODS, RELAP5/ MOD 3, and COMPARE may be used for these analyses. Another computer code which may be applied here is the multicompartment containment system analysis code CONTEMPT 4/ MOD 4 (Reference 17). It is used to predict the long-term thermal-hydraulic behavior of a series of standard compartments. The code calculates the time variation of compartment thermodynamic properties, temperature distributions in heat conducting structures, mass and energy inventories in compartments, and mass and energy transfer due to intercompartment junction flow by solving the' mass and energy balance equations.

The GOTHIC computer code (Reference 18) is a state-of-the-art program for modeling multiphase flow. It solves the conservation equations for mass, momentum and energy for multicomponent, two-phase flow. The code contain a flexible noding scheme that allows lumped parameter, one , two, or three-dimensional analysis or any combination of these to be conducted. Conservation j equations are solved for three fields: (1) steam-gas mixture (2) continuous liquid, and (3) liquid droplet. It calculates the relative velocities between these fields, including the effects of two-phase slip on pressure drop and heat transfer between phases and between surfaces and the fluid. l l

3.G.3 LEAK-BEFORE-BREAK EVALUhTION PROCRDURE l This section describes Leak-Before-Break (LBB) analysis for all l applicable piping. LBB analysis is used to eliminatex from the structural design bases the dynamic effects of double-ended

'Y guillotine breaks and equivalent longitudinal breaks for an applicable piping system.

LBB is demonstrated for the following System 80+ piping systems: l

1. Main Coolant Loop (MCL) ' piping, ' hot and cold legs

2. Surge Line (SL)

3. Direct Vessel Injection (DVI) Line (main run inside j containment) l

4. Shutdown Cooling Line (SC) (main run inside containment) i

5. Main Steam Line (MSL) (main run inside containment)

Amendment R i 3.6-31 July 30, 1993 i

A CESSAR !!ninema i l

2.0 PIPE RUPTURE EVALUATION l

2.1 GENERAL APPROACH ,

15 The approach used for evaluating the effects of pipe rupture -mee-based on Reference 4.9. The specific method employed for pipe Y

whip evaluation is generally determined by the nature of the problem and the size and pressure of the line being restrained:

• Energy balance analysis is the simplest form of analysis.

Its use is confined to conceptual design ~and to the evaluation of restraints for small or relatively low pressure lines, especially the qualification of standard  ;

small line restraints, e Simplified dynamic analyses are used to evaluate restraints for small and moderate size lines and to evaluate situations, such as concrete barrier impact, which are evaluated primarily by empirical relationships and which do not lend themselves to more detailed analysis.

• Detailed dynamic analyses are performed for all large line restraints and for the evaluation of containment penetration areas in any size line.

2.2 PROCEDURE FOR ENERGY BALANCE ANALYSIS Energy balance analysis equates the work done by the blowdown thrust force to the energy absorbed in the restraint. This permits a designer to readily size the energy absorbing component I and this approach is often used for initial restraint sizing.

The work done is based on a quasi-steady-state fluid' force times the distance traveled, including the deflection' of the restraint.  ;

Energy absorbed by the pipe, as at a plastic hinge, is j conservatively ignored. The steady-state fluid forcing function 1 is derived in accordance with Section III.2.c(4) of Reference 4.2. If the approach is used for final design, typically for small lines, the approach follows the requirements of Reference 4.2 and includes an amplification factor of 1.1 on the fluid forcing function to account for a possible maximum reaction beyond the first quarter cycle of response.

2.3 PROCEDURE FOR DYNAMIC ANALYSIS WITH SIMPLIFIED MODELS simplified dynamic analysis models involve closed-form solutions l for the pipe whip event, as detailed in Reference 4.9. Two forms of analysis are used, both being enhancements of the energy balance approar'. in which the time domain is explicitly considered. As in energy balance analysis, an amplification i factor of 1.1 is applied to the fluid forcing function.

Amendment U l 3.6A-4 December 31, 1993 1

CESSARiPacm. ,

1 l

l simulate the dynamic characteristics of the models used in i that analysis.  ;

l The mathematical model provides a three-dimensional representation of the dynamic response of the coupled components to seismic excitations in both the horizontal and vertical directions. The mass is distributed at the selected mass points and corresponding translational degrees of freedom are retained  !

to include rotary inertial effects of the components. The total mass of the entire coupled system is dynamically active in each of the three coordinate directions.

Surae Line A lumped parameter, multimass mathematical model is employed in the analysis of the surge line. A representative model is shown l schematically in Figure 3.7-25. The surge line is modeled as a three-dimensional piping run with end points anchored at the attachments to the pressurizer and the reactor vessel outlet piping. All supports defined for the surge line assembly are included in the mathematical model. The total mass of the surge line is dynamically active in each of the three coordinate directions. The surge line is analyzed as uncoupled from the reactor coolant system, using the motions of the hot leg, pressurizer and supports as input.

3.7.2.1.2.3 Analysis N Modeling and analysis of the coupled components of reactor coolant system and the pressurizer are performed usin ANSYS. A description of ANSYS is given in Section 3.9.1.2.1.)#. Modeling 7h .

and analysis of the surge line is performed using the SUPERPIPE code, a description of which is given in Section 3.9.1.2.1.4.

Time history data for all six possible components of motion are applied simultaneously to the coupled building model to analyze the coupled components of the reactor coolant system.

The responses to seismic excitation for the coupled components of the reactor coolant system are computed using the transient analysis capability of ANSYS. In the analysis of the coupled components of the RCS, excitations are input at selected points in the reactor building. For the coupled components of the RCS, the relative support displacements are inherently accounted for during the coupled analysis. The building motions derived from the soil-structure interaction analysis consist of six time histories at each location per soil case, three linear and three 1 rotational. For each soil case all six time history motions are applied at each selected point of the coupled building model to analyze the coupled components of the RCS. The calculated motions for input to subsequent subsystem analyses therefore include the motions caused by the foundation torsion and rocking.

W Amendment O 3.7-10 May 1, 1993

l CESSARSinL -

Where R is the maximum response of a given element, Rg is the peak response of the element due to the K h mode, and N is the number of significant modes. l If some of the modes are closely spaced the response of the individual modes is combined using the Ten Percent Method from Regulatory Guide 1.92. This can be expressed as:

N R= ( Z + 2I /2 gjg lR Ry $l )

k=1 Where R, Rg and N are as previously defined. The second summation ,

is performed on all i and j modes whose frequencies are closely  !

spaced to one another. Alternative summation methods given in ,

Regulatory Guide 1.92, such as the Double Sum Method, are i acceptable substitutes for the method described above. '

3.7.2.7.2 Nuclear Steam Supply System The SRSS method is the procedure normally used to combine the l modal responses when the modal analysis response spectrum method i of analysis is employed. The procedure, in accordance with Regulatory Guide 1.92, is modified in two cases:

A. In the analysis of simple systems where three or less 1' dynamic degrees of freedom are involved, the modal responses are combined by the summation of the absolute values method; B. In the analysis of complex systems where closely spaced i modal frequencies are encountered, the responses of the closely spaced modes are combined by the summation of the absolute values method and, in turn, combined with the  !

responses of the remaining significant modes by the SRSS method. Modal frequencies are considered closely spaced when their difference is less than 10 percent of the lower frequency. .Sysfen.t anol Oct' smit bn O Interaction of Non-:.fot7 ".;_ Q 5Structuresdwith ONP*""N -

3.7.2.8 Ss/sp;/c d4Ity#ryf" f etT ":12ted Structures .fasfew.c j

ad Swo, newts

%d b 9.cr sfety-related and n:r- sfety-related etn cturer -

connceted, N

t' Y integrally included in th:  ::d 1 the n;n safety rel:ted

.; hen d;tcr;ining the rtreture forcer ir Or g$ cafety-r lated stu cturec, JelsMic.6dejer,y.2" [

To ensure that the failure of a non-nf% al_d structure under the effect of a seismic event does not impair the integrity of an adjacent ;;f e.ty r:lat;d structure, the following procedures are used: SaisMic CstyoryZ >SystO1 or cousfoneni A. Sufficient separation between non-:sf ty 12ted structures andcafetyr-1;tedstructuresismaintained,ork NISMIC GY *!]I Y

@ sten uk ens %($elsmic,Cnisforg%)

1. Amendment N 3.7-16b April 1, 1993

CESSARn!% -

Sj$7$ks471A4OffM1CU.S Seismic (LYdpkg B. The non-: f et; r; ;_d structures re analyzed and designed l to prevent their failure under SSE conditions in a manner such that the margin of safety of these structures ~ is equivalent to that of c2f^t"--r:12ted structures. l Se}sslC 0ld

• SeisailMejep Y C. The ::)^*"-r[__f.7tc9 structure g designed to withstand loads due to collapse of the adjacenc non-Muy- ul tQstructure g should sufficien separation /^' t' . ; ___.;tiacta.ca not be achieved. .sys'a src Q;gg g M 3.7.2.9 Effects of Parameter Variations on Floor ResDonse 9i SDectra To account for the expected variation in structural properties, dampings and other parameter variations, the peaks of floor response spectrum curves are broadened by 15% and smoothed in accordance with Regulatory Guide 1.122.

Soil property related spectrum peaks are further broadened, where required, to conservatively account for all potential variations I

of soil properties within the envelope of site conditions.

I 3.7.2.10 Use of constant Vertical Static Factors )

l A constant seismic vertical load factor is not used for the l seismic design of Sei'smic Category I structures, systems,  ;

components and equipment. l l

The safety-related structures, systems, and components are l analyzed in the vertical direction using the methods described in Section 3.7.2.1. Based on the vertical seismic analysis, a vertical static factor is determined to design columns and shear walls. The vertical floor flexibilities are accounted for in the response spectra at each individual floor elevation of the building structures. The floor beams are designed statically for the acceleration value obtained per Reference 1.

3.7.2.11 Methods Used To Account for Torsional Effects The mathematical models used in analysis of Seismic Category I systems, components, and piping systems include sufficient mass points and corresponding dynamic degrees-of-freedom to provide a three-dimensional representation of the dynamic characteristics of the system. The distribution of mass and the selected Amendment R 3.7-17 July 30, 1993

1 Insert 3.7.2.8-1

'The interfaces between Seismic Category I and non-Seismic Category I structures, systems and components are designed for the dynamic loads gf  !

and displacements produced by both ~the Seismic Category I and y ,

non-Seismic Category I structures, systems and components.

v Insert 3.7.2.8-2 l l

The COL applicant shall describe the process for the design of plant specific and non-Seismic Category I structures, systems and components to reduce the potential for non-Seismic Category I to Seismic Category I (II/I) interactions and propose procedures for an evaluation of the as-built plant for II/I interactions.

\[

CESSAR linsema where:

n = total number of components,

$ = composite modal damping for mode j, 3

St =

critical modal damping associated with component i,

$ = mode shape vector,

{Mg ) = subregion of mass matrix associated with component i, and .  ;

= the mass matrix of the system.

(M)

For direct integration method, viscous damping proportional to the mass and stiffness matrix is used; thus I (C) = a(K) + $(M) d where (C) is the damping matrix, (K) is the stiffness matrix and V$ (M) is the mass matrix. The values of a and S are selected such )

g },- that the damping in the range of frequency of interest is i approximately equal to the damping of the structure.

3.7.3 SEIBMIC SUBSYSTEM ANALYSIS 3.7.3.1 Seismio Analysis Methods l

The seismic analysis of the seismic Category I structures, subsystems, and components other than piping is performed by either the response spectrum or time history method as described in Section 3.7.2.1.1 or an equivalent static method described in '

Section 3.7.3.5. I When analyzed using the response spectrum method, four options I are available for the choice of resp.onse spectra. These are described in Appendix 3.9A, Section 1.4.3.2.1.2. Appendix 3.7D ,

shows sample spectra for use in the three options not related to l plant specific analysis.

For Seismic Category I piping, each piping system is idealized as a mathematical model consisting of lumped masses connected by clastic membbrs. The stiffness matrix for the piping subsystem is determined using the elastic properties of the pipe. This includes 'the effects of torsional, bending, shear, and axial deformati6ns as well as changes in stiffness due to curved members. Generally, a response spectrum analysis is performed using the envelope of all applicable spectra to account for inertia effects. The effects of rocking and torsion are implicitly included because the spectra at the support points Amendment R u 0, N 3.7-19

l l

l l

INSERT 3.7.2.15

' 1 k Where composite modal damping is used for piping, the input damping for piping elements is in accordance with Table 3.7-1. That is, for the Safe Shutdown Earthquake, the damping is 2.0 percent of ,

critical damping for piping of diameter s 12 inches and is 3.0 percent of critical damping for piping of diameter > 12 inches.  ;

~

1

\

et l

CESSAR EL5Lmu TABLE OF CONTENTS APPENDIX 3.7B Bection Bubiect Pace No.

1.0 INTRODUCTION

3.7B-1 1.1 SASSI METHODOLOGY 3.7B-1 1.2 COMPUTATION OF IMPEDANCES 3.7B-3 1.3 COMPUTATION OF SCATTERING 3.7B-5 1.4 STRUCTURAL ANALYSIS AND GENERATION 3.7B-7 OF TRANSFER FUNCTIONS 1.5 SSI ANALYSIS CASES 3.7B-8 1.6 GENERATION OF RESPONSE SPECTRA 3.7B-9 1.7 COMPUTATION OF ELEMENT FORCES 3.7B-11 1.8 OUTPUT LOCATIONS 3.7B-11 1.9 ANALYSIS RESULTS 3.7B-12 2.0 PARAMETRIC STUDIES 3.7B-14 1'

2.1 SOIL PROPERTY VARIATION STUDY 3.7B-14 2.2 EFFECTS OF CRACKED CONCRETE 3.7B-15 .

,1 2.3 EFFECTS OF DEBONDING 3.7B-16 2.4 CONNECTION OF SUBSURFACE SLABS TO 3.7B-17 EXTERIOR WALLS 4

3.o gas-pyggyrs,oggs Sgz pff,psgs af,7g-j7 1(

o 1

i Amendment 0 i May 1, 1993 i

(

hyure J,7c _g CE SysienN Response Spectra at Elev.73.75' of CCW, All Soll Cases, All Motions, Horizontal Model, H2 Motion (N-S),2% Damping 5.00 4.50 m

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CESSAR neiacam,. i l

i The containment vessel is analyzed to determine the stress levels and stability factors of safety resulting from the application of specified loads. The vessel is analyzed )

using thin shell finite element methodology. The ANSYS computer code (Reference 2) is used to generate the geometry of the shell, determine stress levels in the shell, and evaluate shell stability.

3.8.2.4.1 Description of Finite Element Models 3.8.2.4.1.1 3-D Finite Element Model A pictorial presentation of the containment vessel 3-D finite element model is given in Figure 3.8-3. An eight node isoparametric thin shell element is used. Fixed boundary conditions are applied in the model at the 90'+3" elevation.

The weight of the personnel airlocks and the equipment hatch penetrations is included in the model by increasing the density of the shell in the region of the penetration. Live load is included in the weight of the penetrations. The penetrations themselves are not modelled explicitly. The thickness of the shell in the region of the personnel airlocks and the equipment hatch is increased using the area replacement rules in the ASME Code. The containment spray mass is included in the upper region of the model by distributing additional mass at the appropriate locations in the dome. The mass of the piping and the electrical penetrations in the lower region of the sphere is accounted for by increasing the density of the shell elements in that region. The stiffness of the compressible material at the base of the containment vessel is modeled as a two-directional spring.

Although the transition region of the SCV is 2 inches thick, the 3-D finite element model has a uniform shell thickness of 1-3/4 inches. The additional 1/4 inch of material in the

, transition region is for corrosion allowance only and credit 4 for this addittnal thickness is not included in the 3-D Y analysis. The axisymmetric model described in Section 3.8.2.4.1.2 is used to evaluate the effects of the change in naterial thickness in the transition region.

3.8.2.4.1.2 Axis tric Finite Element Model wH The containmen essel is modelled with thin shell axisymmetric finite lements. The model is fixed at the base, elevation 90'+3 , :nd : two-directional spring elements gf in 2ttrahed at +h cercrete curf::: el:; ti:n, 91'19", to y represent the compressible material at the base of the containment vessel. The meridian modeled is the one corresponding to the equipment hatch since it has the Amendment U 3.8-8 December 31, 1993

CESSAR ENEncam,.

l The primary membrano stress evaluation for Service evel A Y l is the same as the Design Condition.  ;

I When evaluating secondary stress effects, the reduced load combination in:

D+L+Ta+P a Service Level C:

Pipe reactions Ra and Ro are eliminated as described in the Design combination.

The stresses resulting from the operating pressure loads, i P,

o are enveloped by the accident pressure loads and I therefore are not analyzed separately.

The T, and T o loads are not included in the combination because thermal loads are considered as secondary stresses as described in the Design combination. The ASME code does not require an analysis of secondary stresses for Service Level C.

The reduced Service Level C loads are the same as the i reduced Service Level D loads. The ASME Service Level D allowable stresses are lower than the Service Level C allowable stresses; therefore, the analysis is performed for the reduced Service Level D loading combination and compared with the lower allowable stresses of Service Level D.

Service Level D:

Pipe reactions, operating loads, and thermal loads are eliminated as described in the Service Level C combination.

The pipe rupture loads, Yg , are eliminated in the design by the use of rupture restraints and guard pipes in the System 80+ design. The jet impingement loads, Y3 , are eliminated by the use of guard pipes and, where necessary, jet impingement protection devices. The containment shell is protected from the missile loads, Y,, _ by the crane wall inside the containment vessel and the head area cable tray system.

The reduced load combination is:

D + L + P + E' ]

D. Construction Loads  !

All loads in ' the combination given in Table 3.8-2 are l applicable to the System 80+ design. l l

l l

1 l

Amendment U '

3.8-10 December 31, 1993

.'l 2  :

CESSAREMUnce The Service Level C and D stability analyses are completed for two types of imperfections:

1) A full sine wave with a half-wavelength = 8 feet and a peak to peak amplitude of 1.75 inches.

2) A half sine wave with a half-wavelength = 16 feet and a peak amplitude of 1.75 inches.

The resulting Service Level C stability safety factors for these two types of imperfections are 2.70 and 2.74, respectively.

F. Ultimate Load Considerations (D + L + Pu)

The Ultimate Capacity is determined using an elastic analysis with the axisymmetric model. All loads are applied simultaneously in a static manner. The dead and live load is applied as an increase in the density in the appropriate regions. The internal pressure load, which is applied to the inside face of each element, is increased until the maximum stress intensity reaches the Service Level C allowable membrane stress intensity for the given temperature. The ASME Service Level C allowable stress intensity value is the nominal yield stress value for the temperature given. Temperature values of 150*F, 290'F (Design Basis Accident Temperature) , 350*F, and 450*F are evaluated. The material properties associated with the temperatures are used. The internal pressure value which results in a maximum stress intensity equal to the Service Level C allowable membrane stress intensity is the ultimate pressure capacity, P.

u The results are summarized in Table 3.8-3D.

G. Combustible Gas Load Considerations (D + L + Pg + P,)

Thn rewnhustible Gas Loading is evaluated using an

[ 7 elastic $nalysis with the axisymmetric model. The dead and /

live load is applied as an increase in the density in the appropriate regions. The peak pressure from hydrogen combustion and the design basis pressure is added together and applied as an internal pressure to the inside face of each element. All loads are applied simultaneously in a static manner to determine the maximum membrane stress intensity. The results are summarized in Table 3.8-3A.

11 . Containment Overturning and Sliding (D + L + E')

The containment is analyzed for sliding and overturning of the interior structures against the steel containment and the interior structures and steel containment against the lower concrete dish structure outside of containment. The interior structures and the steel containment are modelled Amendment U 3.8-16 December 31, 1993

CESSAR8!Khmn  !

l Deviations from the design are acceptable provided the following acceptance criteria are met:

1. An evaluation is performed (depending on the extent of the deviations, the evaluation may range from the documenting of an engineering judgement to performance l of a revised analysis and design), and

2. The structural design meets the requirements specified in Section 3.8.2.

The COL applicant will prepare an as-built structural analysis report for the steel containment vessel.

3.8.2.5.1 Welding and Weld Acceptance Criteria Welding activities shall be in accordance with the requirements of Section III, Subsection NE of the ASME Code.

Y (d [P 3.8.2.6 Materials, cuality control, and special gh Construction Technicrues 3.8.2.6.1 Materials The containment vessel materials are in accordance with Article NE-2000 of Subsection NE, " Class MC Components," of the ASME Boiler and Pressure Vessel Code,Section III, " Nuclear Power Plant Components."

The containment plate material is ASME SA537 Class 2. This material is exempt from post-weld heat treatment requirements l when plate thickness is less than or equal to 1.75 inches in accordance with Table NE-4 622.7 (b)-1 of the ASME Boiler and Pressure Vessel Code,Section III. When plate thickness exceeds 1.75 inches, post weld heat treatment shall be performed.- The material will be impact tested in accordance with Article NE-2300 of Section III of the ASME Code.

Fabrication and erection of the containment vessel are in accordance with Article NE-4000 of Section III of the ASME Code.

This includes welding procedures, procedure and operator performance qualifications, post weld heat treatment and tolerances.

Nondestructive examination of welds and materials is in accordance with Article NE-5000 of Section III of the ASME Code.

3.8.2.6.2 Quality Control The general provisions of the overall Quality Assurance program are outlined in Chapter 17. These are supplemented by the special provisions of the ASME Code for quality control as applicable to Class MC Components. The containment vessel is Amendment U 3.8-20 December 31, 1993 i l

i I

INSERT A (to Section 3.8.2.5.1) '

V Radiographic examinations will be accepted by the COL applicant's nondestructive' examination (NDE) Level III examiner prior to final acceptance.

Confirmation that facility welding activities are in compliance with the certified design commitments shall include verifications of the following by individuals other than those who performed the activity-

/ 1. Facility welding specifications and procedures meet the l applicable ASME Code requrements,

2. Facility welding activities are performed in accordance with the applicable ASME Code requirements, V

3. Welding activities related records are prepared, evaluated and maintained in accordance with the ASME requirements,

4. Welding processes used to weld dissimilar base metal and welding filler metal combinations are compatible for - the -

intended applications, ,

S. The facility has established procedures for qualifications of welders and welding operators in accordance with the applicable ASME Code requirements,

6. Approved procedures are available and are used for pre-heating and post-heating of welds, and those procedures meet the applicable requirements of the ASME Code,

{

7. Completed welds are examined in accordance with the applicable  ;

examination method required by the ASME Codt.

I

CESSAREin% =,.

The concrete is sealed to preclude moisture. A visual )

inspection of coatings is performed.

Visual inspections of containment base metal and welds are performed in accordance with ASME Section XI, Subsection IWE and 10CFR50 Appendix J. These are formal inservice inspection requirements. The portions of containment embedded in the concrete are exempt from these inspection requirements while the welds around the embedded penetrations are required to be inspected.

Collection of moisture in the transition region is prevented i by use of sloped floors and drains.

The compressible material which is placed in the transition region between the steel and concrete is removable. Once i removed the material and SCV -.. M inspected.

is Y l l No equipment or ductwork is located such that it inhibits a visual inspection at the steel concrete interface for corrosion.

For further precautionary measures and conservatism, the SCV is 2 inch in thickness in the transition region. This thickness is beyond design requirements and allows for a corrosion allowance of approximately 4 mils per year over a 60 year life. With an inspection program and maintenance of  ;

the coatings, corrosion is minimized in this region.

3.8.2.7 Testina and In-service Surveillance Recuirements The containment vessel, personnel airlocks and equipment hatch are inspected and tested in accordance with the ASME Boiler and Pressure Vessel Code,Section III, Subsection NE. Penetrations i

are pressure tested as required for Subsection NC of the ASME Code.

Periodic leakage rate tests of the containment are conducted in accordance with 10CFR50, Appendix J to verify leak tightness and integrity. These tests and other in-service inspection requirements are described in Section 6.2. Periodic in-service inspections are conducted in accordance with the ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWE.

3.8.3 CONCRETE AND STRUCTURAL STEEL INTERNAL STRUCTURES 3.8.3.1 Description of the Internal Structur31 The internal structure is a group of reinforced concrete structures that enclose the reactor vessel and primary system.

The internal structure provides biological shielding for the containment interior. The internal structure concrete base rests inside the lower portion of the containment vessel sphere. A Amendment T 3.8-22 November 15, 1993

~

t

)

CESSARLih m., i l

l 1.6 instead of 1.7 in load combination 11.

4. The following noto is added to Scotion Q1.S.8  ;

"For constrained (rotation and/or displacement) members l supporting safety related structuras, systems, or '

components, the stresses under load combinations 9, 10, and 11 should be limited to those allowed in Table Q1.5.7.1 as i l

modified by provision 3 above. Ductility factors of Table Q1.5.8.1 (or provision S below) should not be used in these ca5SS." l S. For ductility factorn 'y' in Sections Q1.5.7.2 and Q1.5.8, are substituted provisions of Appendix A, II.2 of SRP Section 3.5.3 in lieu of Table Q1.5.8.1. I

6. In load combination 9 of Section Q2.1, the load factor applied to load P is 1.5/1.1 = 1.37, instead of 1.25.

7. Sections Q1.24 and Q1.25.10 is supplemented with the following requirements regarding painting of structural steel:

a) Shop painting shall be in accordance with Section M3 of Reference 17.

b) All exposed areas after installation shall be field painted (or coated) in accordance with the applicable portion of Section M3 of Reference 17.

c) The quality assurance requirements for painting (or , l coating) of structural steel shall be in accordance with Reference 18 an endorsed by Regulatory Guide 1.54, i

" Quality Assurance Requirements for Protective Coatings Applied to Water Cooled Nuclear Power Plants".

Welding activition associated with Saismic Category I structural steel components and their connections shall be accomplished in accordance with written procedures and shall meet the requirements of - / r : ;;;; :._ " . i; n

• ?- The visual l \/'

cri e/"

acceptanco ria shall be as defined in NCIGc01 (R f ren a 24). gg g y f,j ( gg,g ,g gje g ,y.),

2. 4.5.3 conor.se aar ste.1 structur..-

In addition to satisfying the load combinations for structural adequacy against the design loadings, the load combinations to and flotation are checked to ensure overall stability ofensure safety Seismfactors against overturning, structures. The following events are checked as a minimum:

A. The overturning about the toe of the foundation supported on soil.

Amendment U ,

3.8-35 December 31, 1993 l

L CESSAR !!fencamn B. The foundation sliding on soil.

C. Floating of the foundation base mat.

D. The containment vessel slipping in the lower concrete support dish.

. E. The containment vessel overturning about the edge of the Y lower concrete support dish.

I7 F. The interior structure concrete slipping inside the containment vessel.

The safety factors which must be satisfied during any of these events are shown in 9minilre-es4P-9. // gn,//g J,pg,gggglp,/,

No increase in allowable stresses under service load conditions due to normal or severe load combinations is permitted due to wind loadings as identified in NUREG-0800, NRc Atandard Review plan. seetion 3.n a ne TT.s.

s,r.ps..y shuermt Ana/ysi.s repoet L 39Meddr,'bh A structural analysis report vil be prepared for Seismic Category I structures. This report will document that the structures meet the L " '-^ " specified in Section 3.8 and design changes and identified construction deviations, which could potentially affect the structural capability of the structure h4ve been incorporated ir$o the structural analysis y Contostent ,M*1h 1%e, Mehdc snd praeth.rs.r *f Serfin .LP.

The following records will be reviewed, as applicable:

1. Construction records stating material properties for concrete, reinforcing steel, and structural steel

2. As-b u i

.qs,1tstrucho**redhmensonsandarrangements,/88 Sq2rt o bu.i/ ings y/j s.,

l y. Design documents for the structure

[

Deviations from the design are acceptable provided the following acceptance criter tre,gey M,,,Q ,g) p,,,,,ju,,,,, ,pg,,f,*, p ,7g p g

1. An evaluationfis performed (depending on the extent of the deviations, th's evaluation may range from the documenting of an engineering judgement to performance of a revised analysis and design), and 2Me /sud Ch'NAA

2. The structural design meets the * ."-----t specified in Section 3.8, and

3. The seismic floor response spectra of the as-built structure

~

i does not exceed the design basis floor response spectra by more than 10%.

.3, fils Q// /o2A t K,**remsk3 EntdAYi fAtt hNW glo& prassu r //*m/whe. * *i' & f*" b 'l*P k

2.wA Qc ( fdKIr2IdE #~ Amendment T j .8-36 November 15, 1993

. ~ , . . - - - . . _ . _ _ _ _ ... ___ . ._.

CESSAR g include.

e in w M. .sM ' 9' The structural analysis report will summarize the results of the reviews, evaluations, and corrective actions, as applicable, and l conclude that the as-built structure is in accordance with the (design. -

Welding activities associated with the Refueling Cavity and Spent Fuel Pool liners shall be accomplished in accordance with the requirements of the American Welding Society (AWS) Structural Welding Code, D1.1 (Reference 25). The welded seams of the liner plates shall be spot radiographed where accessible, liquid penetrant and vacuum box examined af ter fabrica$on to ensure the liners do not leak. The acceptance criterif shall meet the acceptance criteria stated in Article NE-5200,Section III, Division I of the ASME Code.

$&*3.8.4.6 Y* S* Y Material. Ouality Control, and special Construction Y Techniaues The category I structures are poured-in-place reinforced concrete structures. The major materials that will be used in the construction are concrete, reinforcing bars and structural steel.

A brief description of these materials is given below.

3.8.4.6.1 Material 3.8.4.6.1.1 Concrete The basic ingredients of concrete are cement, fine aggregates, coarse aggregates, and mixing water. Admixtures will be used if needed.

Cement will be Type I or Type II conforming to " Standard Specification for Portland Cement," ASTM C150. For special circumstances, other. approved cements will be used.

Aggregates will conform to " Standard Specification for Concrete Aggregate," ASTM C33.

Water used in mixing concrete will be clean and free from injurious amounts of oils, acids, alkalis, salts, organic materials or other substances that may be deleterious to concrete or steel. A comparison of the proposed mixing water properties will be made with distilled water by performing the following tests:

A. Soundness, in accordance with " Standard Test Method for Autoclave Expansion of portland Cement," ASTM C151. The results obtained for the proposed mixing water will not exceed those obtained for distilled water by more than ten percent.

Amendment U 3.8-37 December 31, 1993

CESSAR !!!Mnema -

3.8.4.6.1.2 Reinforcing Steel Reinforcing steel will consist of deformed reinforcing bars i conforming to " Standard Specification for Deformed and Plain Billet - Steel Bars for Concrete Reinforcement," ASTM A615, Grade 60 or " Specifications for Low-alloy Steel Deformed Bars for Concrete Reinforcing," ASTM A706, Grade 60. The fabrication of reinforcing bars, including fabrication tolerances, will be in accordance with CRSI " Manual of Standard Practice" MSP-1. The placing of reinforcing bars, including spacing of bars, concrete protection of reinforcement, splicing of bars and- field tolerances will be in accordance with ACI 349. Epoxy coated reinforcing steel is used for areas where a corrosive environment is encountered.

3.8.4.6.1.3 Structural Steel The structural steel will essentially consist of low carbon steel shapes, plates and bars conforming to " Standard Specification for Structural Steel," ASTM A36. Other structural steels listed in ANSI /AISC N690 may also be used.

Fabrication and erection of structural steel in Seismic Category I structures will be in accordance with the requirements of ANSI /AISC N690. The structural connections will be either welded or bolted. Welding activities associated with Seismic Category I structural steel components and their connections shall meet the requirements in Section 3. 8.4.5. 2. Tall bolted lv/

connections will be made with high strength bolts conforming to one of the following cpecifications:

A. " Specification for High-Strength Bolts for Structural Steel Joints," ASTM A325.

B. " Specification for Heat-Treated Steel Structural Bolts, 150 l KSI Tensile Strength," ASTM A490. l Other bolts listed in ANSI /AISC N690 may also be used.

Quality Control l 3.8.4.G.2 1

The quality of materials will be controlled by requiring the l suppliers to furnish appropriate mill test reports as required under relevant ASTM Specifications as described in Subsection 3.8.4.6.1. These mill test reports will be reviewed and approved-in accordance with the general provisions of the overall Quality l Assurance Program outlined in Chapter 17 and supplemented by the special provisions of the appropriate codes and specifications for design listed in Table 3.8-4.

l Amendment U 3.8-39 December 31, 1993

CESSAR E!nificarien Erection tolerances, in general, will be in accordance with the referenced design code. Where special tolerances that influence the erection of equipment, etc., are required, they will be indicated on the drawings by the Engineer.

3.8.4.6.3 Special Construction Techniques No unique or untried construction techniques are contemplated.

Both the cylindrical and the dome portions of the shield building will be constructed using standard construction techniques.

3.8.4.7 Testina and In-service Surveillance Recuirements There will be no testing or in-service surveillance beyond those quality control tests performed during construction, which will be in accordance with ACI 349, ACI 301, ANSI /AISC N690 or ANSI N45.2.5 (Reference 8) as applicable.

3.8.5 POUNDATIONS 3.8.5.1 Description of the Foundations The foundations of the Category I structures are reinforced l concrete mats. The foundation of the Nuclear Island is approximately 10 feet thick, has a flat bottom and rests on soil or rock. The top of the Nuclear Island basemat is located 40.75 feet 1 foot below the finished grade elevation. The minimum foundation mat thicknesses for - the Diesel Generator Fuel Oil structure and Component Cooling Water Heat Exchanger structure are approximately 2 Esot and 4 s&, respectively. /

Ret The COL applicant will submit the site-specific foundation mat construction procedures in accordance with SRP 3.8.5.

3.8.5.2 App _licable codes, Standards, and Specifications.

Reinforced concrete foundations and supports of Category I structures are designed as described in Appendix 3.8A using the codes and criteria shown in Table 3.8-4.

3.8.5.3 Loads and Loadina combinations The design loads and loading ' combinations are described in l Section 3.8.4.3 and Appendix 3.8A.

3.8.5.4 Desien and Analysis Procedures The reinforced concrete foundations of Category I structures are

)

analyzed and designed for the reactions due to static, seismic and .all other significant loads at the base of the superstructures supported by the foundation in accordance with the criteria in Appendix 3.8A. The foundation mat is modeled as a three dimensional finite element structure as an integral part Amendment U 3.8-40 December 31, 1993

4

'(

. isnLE 5.5-3 eet 1 of 3) /

STRESS INTEN Y LIMITS FOR S CONTAINMENTS ,e l

l l Primary & ,/

,/

[ Categorie Load Gen.

Prima Stresses Local Mem.

Bp[ ding &

t'

/ ocal Mem.

Secondary' Peak Stresses /

/

Buckling

, Pt P3+Pt(6) Pt + P p.4 Q Pt+ P+Q 3 4

- Mting Pneumati 0.75S, 1.15S, / 1.15S, N/A (2) Consider'for (5) See )

fatigue Condition f f evMuation Design Condition /

[ 1.05 ,

/

/ $.5S, 1.55 ,

/

N/A [/A

/

[See(9)

Level 1. 1.55 , 1.55 , Consider f See (9)

Servi -

3.0S,/, [ fatigue Li (1) / evalu. ton f l l l l e g _

6&fe

• Amendment T November 15, 1993

(

~

in6LE3.5-3/ Cent'd)

[ S ee 2 of 3)

[ /

, STRESS INTENSITY LIMITS FOR STEEL EONTAIHMENTS' i

' / / '

/ /

/ ,

/ Pria / Bending &- Primary &

' Gen. Me6. y StressesLocal gcal Mem/ Mem. Secondary P Peak resses Load Cpt,egories P t ,/

Buckling

,- f j,

P +P t (6) +P 3 + Q't Pt+ 3+Q+F p

/

/ / /

(/Level C Not integfal

[ [ See (9) [

and Continuous- / f.0S , 1.)$, 1.5S , 3 . 0.S., )(/A /

Service -

j -

or

• or

• N/A See (9 Limit Iptegral and / l.85 N/A 1.25, or

• j1.55, 1.8S" (8) j/ 1.05, ,

1.5S i

[ Continuous (4) (7) / / 1 1.2S,Ir

• 1.8S , or

• 1.8S , pr

• N/A Se (9)

Le el D ,-

[ Not integr3[ and Continuouy 1.054 1.55, 1.5S, /

N/A S rvice / '

Integ. Sj' l.55, 1. 5,S', s(

limit

-Elastt[An.(3) z

/ A N/A See (9) .

j Cont.

Ine)dstic An. (3) ' S, S, ,'S, P t-Flooding 1.25, or */ 1.8S, or

• 1.85 , o[*2.05, N/A (2) See (9)-

1.05, 1. 5 S,- 1.5S, /

j l _

/

us l

Amendment I December 21 390 i

l

(

F TABLE ~;.0 0,(Cor."d) ,

(Sh t3o 3)

STRESS INTENSI LIMIT OR STEEL C0Kl'AlletENTS NOTES:

(1) e allowab > stress ntensities S,; S mc sh be those defined i ection NE the ASME Code.

(2 N/A - t6 evaluat n required.

(3) S is 85% of he general prpdry membra allowable permit d in Appendix . In the application f the rul Appendi F, S,g, if app)1 cable, sha be as specified Section NE of he ASME Code. l Those mits identifi y the ast isk (*) indicate choice of the 1 ger of the two lim s.

(5) Th number of tes sequences s dll not exceed I unless a fatigue aluation is consi -red.

(6) alues shown e for a soli rectangular syc ton. See NE-3220 or other than a id rectangular section.

(7 These strpis intensity mits apply to partial penetrat n welds also.

(8) Value shown are a plicable whpr( P < 0.675 Whe P > 0.675 use the large of the two mits,

[2. 1.5(Pg /S,)]l. mc r [2.5 - 1.5( g/Sy )]Sy . y.

must be demo 6strated ' at any axisymetric tech igues pro sed are applicabl to a vessel aving large (9) asymmetric openings, and hat the overall margin of safety to event buckling is adequate.

L _

)

Delele Amendment N April 1, 1993

C E S S A R EMancu e,.

TABLE 3.8-3C STABILITY EVALUATION FOR THE STEEL CONTAINMENT VESSEL Reduced Load Calculated Required. Safety Load Categories Combination Safety Factor Equation Factor level A D+L+ P,+T, 3.0 3.0 l y .

Level C D+L+ + E '

/ 2.7 2.5 l

'l 4

I l

Amendment U December 31, 1993 I

a. ~ . , . , - .- . _ . - . . - . - . . _ , .. . . - - , . ,

CESSARUNncm.

TABLE 3.8-5

/

(Sheet 1 of Je) /

LOAD COMBINATIONS FOR CATEGORY I BTRUCTURES b

IHdEX

. oad D initio t

1. Normal ' oads

~

2. Sever Enviro ntal Loa J.

. Extr e Envir nmental L ads

4. Ab rmal Lo s

5. O er Defi tions I. Loa Combina ons and cceptance Criteri for Cat gory I Co crete St ctures

. Se Ice Load C nditions

2. F tored Loa conditions I. Loa Combinatio s and Acc tance itoria or Catego y I S el Structur s

1. Servic Load Cond ions

a. lastic D ign

b. Plastic sign

2. actored Lo Condit ns

a. Ela ic Desi

b. P1 tic Des n IV. Load Comb ations - d Accept ce Cri ria fo Cate ry i Foundatio s by)$ jyl), lo Ak daw. JhAfoO h S iS M.lC 0ftgot' $ 8t=t<,,efWd3 2('d YdbM.'CY $M. h cwlY

3. M , su&n- 40 Amendment I December 21, 1990

.CESSAR EHence,.

-y-bd f& h '

(. TA LE 3.8-5 (Cont'd)

(Sheet 2 of 10)

LOAD COMBINATIONS FOR CATEGORY I STRUCTURES

1. Load Def1 itions All the majo loads to be encountered and/or to postulated in a Category I str ture are grouped into four categor es described below.

All the loads lis d, however, are not necessaril applicable to all the structures and the elements in the plant. L ds and the applicable load combinations f which each structure i designed will depend on

- the conditions to whi that particular str ure could be subjected.

1. Normal loads

Normal loads are those oads to be ncountered during normal plant operation and shutdown. They ir lude the following

}'

D ---

Dead loads or ir related internal moments and forces, includi any permanent equipment loads and

hydrostatic lo s.

l L ---

Live loads r their related internal moments and

forces, i uding an movable ecuipment loads and other loa which vary th intens<ty and occurrence, such as oil pressure.

4 F ---

Later and Vertical orces associated with hydr tatic loading, either nternal or external. For fac red load combinations, only pressures due to no al fluid levels- shall b combined with cther e . reme or abnormal loads.

I H ---

ateral loads produced by stati or seismic earth pressures.

T, -

Thermal effects and loads during no 1 operating or shutdown conditions, based on the ost critical transient or steady state condition.

R, ---

Pipe reactions during normal operating o shutdown conditions, based on the most critical tr sient or steady state condition.

2. Severe Environmental Loads Severe environmental loads are those loads that could infrequ tly be encountered during the plant life. Included in this cate ry are:

W ---

Loads generated by the design wind specified for th plant.

Amendment S September 30, 1993

CESSAR Heincomi.

.De/ ele TABLE 3.8-5 (Cont'd)

(Sheet 3 of 10)

LOAD COMBINATIONS FOR CATEGORY I STRUCTURES

3. Extreme Enviro... ental loads Extreme environment loads are those which are credible but ara highly improbable. The inci e:

E' ---

Loads genera d by the Safe Shutdown Ear i quake. The loads consist of thr directional loads, E', (N direction), E'y (E-W direction),E', ertical direction). .

The earthquake loads are combine to obtain the max' um stress results by one of the following combinations:

(i) E ' = (E ',2 + E 'y2 + E ',2) 2 or- i (ii) E'= E', t 0.4 (E'y E ',)

or E'45 - E 'y 0.4 (E', E ',) l or '

E ' = t E ', i 0. 4 ( E ' E 'y)

W, ---

Loads generated the Design Basis Tor do specified for th'e plant. They inc ude loads due to the torna wind pressure (W,), l a loads due to th tornado-created differential ressures (W p), and loads due to e tornado-generated missiles ( ).

The combined effect o ,, W and W is determined in a cons vative manner l for each particular ructure,,or porf, ion thereof, as applicable, by using one '

or more of the fo owing combinations as appropriate:

(1) W, = W. 1 (ii) W, =

(iii) W, W, (iv) , = W, + 0. 5 Wp (v W, = W, + W, vi) W, = W, + 0. 5 W, + W, 1

Amendment U December 31, 1993

CESSAR !!nincim.

@h&- _ -

TABLE 3.8-5 (Cont'd)  ;

(Sheet 4 of 10)

LOAD COMBINATIONS FOR CATEGORY I STRUCTURES j

4. Abno .1 Loads Abnormal ds are those loads generated by a post ated high energy pipe break accide within a building and/or compartme thereof. Included in '

this category a the following: -

~~

P, --- Pressure quivalent static ioad wi in or across a~ compartment I and/or bu ding, generated by he postulated break, and-including an appropriate dynami cad factor to account for the dynamic natur of the load.

T, ---

Thermal -loads mder the 1 conditions generated by the i postulated break nd inclu .ng T,.

R, ---

Pipe reactions un er ermal conditions generated by the postulated break an i luding R,.

Y e

Equivalent static 1 d on the structure generated by the reaction of the br en igh-energy pipe during the postulated l break, and inclu ng a appropriate dynamic load factor to )

account for the namic n ture of the load. j Yj ---

Jet impingeme equivalent s- tic load on a structure generated by the post ated break, and neluding an appropriate dynamic load factor to account for the- namic nature of the load. 1 i

Y, ---

Missile pact equivalent static ad on a structure generated i by or < ring the postulated break, .uch as pipe whipping, and inclu ,ng an appropriate dynamic load actor to account for the  ;

dyn ic nature of the load.

l In detennini an appropriate equivalent static loa for Yr, Yj, and Ym, i elastic-pla ic behavior may be assumed with appropriat uctility ratios as i long as e essive deflections will not result in loss function of any 3 safety r ated system. i

5. Other finitions 1 S ---

For structural steel, S is the required section str ngth based on the elastic design methods and the allowable stres s defined in ANSI /AISC N690.

l Amendment N l April 1, 1993 l

. . - _ - - ~ . _ . _ _ _.

_, I

CESSAR1!ninca- n_

q u

TABLE 3.8-5 (Cent'd)

~

(Sheet 5 of 10)

LOAD COMBINATIONS FOR CATEGORY I STRUCTURES U --- For concrete structures, U is the ction strength.

required to resist design loads base on the ultimate trength design method described in 349-85.-

Y --- Fo structural steel,-Y is the se ion strength required to sist design loads based on lastic design methods descr ed in ANSI /AISC N690-1984 II. Load Combinations and Acceo ance Criteria for Cat aory I Concrete Structures The following set of load co inations and al wable design limits is used for all Category I concrete s uctures:

1. Service load Conditions Service Load Conditions, re esen Normal, Severe Environmental and Normal / Severe Environmental 1 .

/

The Ultimate Strength Design eth is used with the following load 1 combinations: ,

1) U = 1.4D + 1.7F + 1.7 + 1.7H + . 7R, l

2) U = 1.4D + 1.7F + 1 L + 1.7H + 1 + 1.7W l If thermal stresses du to T are present, t e coefficients for each load category may b multiplied by 0.75 to satisfy the following '

combination:

3) U = (0.75) .4D ++1.7F l U = 1.05D 1.05F 1.3L++1.7L 1.3H+ +1.7H ++

1.3T, 1.1T +R,1.7R,3W+

+ 1. 1.7W) or In addition, e following combination is considere 3

4) U= D+1.7W Where load reduces the effects of other loads, the co esponding coeff cient for that load should be taken as 0.9 if i can be 1 dem strated that the load is always present or occurs simulta ously w h other loads. Otherwise the coefficient for the load shou be aken as zero.

Amendment U.  ;

December 31, 1993 i

CESSAR !!ninema

.De/e ;&

-=

TABLE 3.8-5 (Cont'd)

(Sheet 6 of 10)

L D COMBINATIONS FOR CATEGORY I STRUCTURES I

2. Factored load nditions Factor Load Condi ons represent Extreme Environmental, Abnormal, -

Abnormal / Severe Envi nmental and Abnormal xtreme Environmental loads.

The Ultimate Strength esign methgd is sed with the following load combinations:

1) U = 0 + L + F + H + T, + R, + -

2) U = 0 + L + F + H + T, + R, W,

3) U = D + L + F + H + T, + , + 1. 5 P,

~

4) U=D+L+F+H+T, R, 1. 0 P, + 1. 0 ( Y, + Y) + Y,) + 1. 0 E '

In factored load combinati ns (3) an (4), the maximum values of P ,

T,, Y Y _and Y i luding an ap ropriate dynamic load factor, are R , d;,unle, ss a l,i -history analy 's is performed to justify use otherwise. Factored 1 dcombinations(2 and (4) are satisfied first i without the tornado ssile load in (2), a without Y , Y3 , and Y, in (4). When consider' g these loads, howeve , local se,ction strength capacities may be axceeded under the effec of these concentrated r loads, provided t re will be no loss of functi of any safety related system.

I Where any lo reduces the effects of-other loads, he corresponding ,

coefficient for that load should be taken as 0. if it can be demonstra d that the load is always present or occurs ultaneously with ot r loads. Otherwise the coefficient for the loa should be ,

taken zero.

Wh e the structural effects of differential settlement, c ep, or s inkage may be significant, they should be included with t dead oad, D, as applicable.

Amendment U December-31, 1993

LCESSAR !!ninc.m. m

._ZA./e 7'e TABLE 3.8-5 (Cont'd)

(Sheet 7 of 10) l LOAD COMBINATIONS FOR CATEGORY I STRUCTURES III. Load Combinatio s and Acceptance Criteria for Cateoory I Steel Structures The following set load combinations and allowable d ign limits is used for  !

all Category I stee tructures: .l

1. Service load Condi

• ns _

Either the elastic worki stress desig methods or the plastic design methods of ANSI /AISC N690 . be used.

a. If the elastic working st ss sign methods are used:

1) S=D+L+F4 H+R S=D+L+F+H+

2) ,+

[

If thermal stresses an R, are present, the following combinations are alsd satisfie e to T,d:

3) 1.3 S = D + + F + H + T, + R,

4) 1.3 S - D L + F + H + T, + R, + W Both cases o L having its full value or bei completely absent are checked.

b. If plas c design methods are used:

1) = 1.7 D + 1.7 L + 1.7 F + 1.7 H + 1.7 R, Y = 1. 7 D + 1. 7 L + 1. 7 F + 1. 7 H + 1. 7 R, + 1. W-Amendment R July 30, 1993 1

~ _

-C E S S A R Minnca m n bd. k fe '

TABLE 3.8-5 (Cont'd)

(Sheet 8 of 10)

LOAD COMBINATIONS FOR CATEGORY I STRUCTURES If t rmal stresses due to To and Ro are pres t, the following combin tions are also -,atisfied:

3) Y 1. 3 (D + L + F + H + T, + R,)

_ 4) Y = 1. ( D + L + F + H + T, + R, +

Both cases of aving its full va e or being completely absent are checked.

2. Factored Load Conditions The following load combinations re s isfied:

a. If elastic working stress des' methods are used:

1) 1.6 S = D + F + L I + T, + R, + E '

2) 1.6 S = 0 + F L+H+ + R, + W,

3) 1.6 S = D + +L+H+T, R, + P,

4) 1. 7 S* = + F + L + H + T, + R, P, + 1.0(Y; + Yr + Y,) + E '

• For combina on (4), in computing the r uired section strength,- .;

S, the pla ic section modulus of steel s apes may be used. ,

i

b. If plastic d ign methods are used: l

1) - D + F + L + H + T, + R, + E '

Y' = 0 + F + L + H + T, + R, + W, l 3) Y* = D + F + L + H + T, + R, + 1. 5 P,

4) Y* = 0 + F + L + H + T, + R, + P, + 1.0(Y)+Y,+Y,) +

Amendment U December 31, 1993 i

CESSAR !!aine-De/de TABLE 3.8-5 (Cont'd)

(Sheet 9 of 10)

LOAD COMBINATIONS FOR CATEGORY I STRUCTURES In th above factored load combinations, t rmal loads can be neglect when it can be shown that th are secondary and self-limi ing in nature and where the ma rial is ductile.

Y(for the fa hould be multiplied by 0.90forthe)toredloadcombinations) ternal Structyres and .0 for other Category 1 structures.

In factored load mbinations (3 and (4), the minimum values of P T,, T,, Y Y d Y , incl ding an appropriate dynamic load fEctor,areu;, sed,,un1ss,ati -history analysis is performed to justify otherwise, ctore load combinations (2) and (4) are first satisfied witho e tornado missile load in (2), and without Y , Y3 , and Y, (4). When considering these loads, however, focaT section s engths may be exceeded under the effect of these concentrated oa s, provided there will be no loss of g/

function of any safet -rel ed system. T Where any load r duces t effects of other loads, the corresponding coef icient for at load should be taken as 0.9, if it can be de nstrated that the load is always present or occurs simult eously with ot r loads. Otherwise, the coefficient f that load should b taken as zero.

Where the ructural effect of diff ential settlement may be signific it should be included with he. dead load, D.

IV. Load Combinatio and Acceptance Criteria for Catea v I Foundations In addition the load combinations and acceptance iteria referenced above, all tegory I foundations are also checked ag inst sliding and overturnin. due to earthquakes, winds, and tornadoes and ainst flotation due to fl ds in accordance with the following:

Minimum Factors of Safe ,

load mbination Overturnino Slidina Flotation )

i D+ +W 1.5 1.5 -

D H + E' 1.1 1.1 -

+H+W, 1.1 1.1 -

+F - -

.1 l Amendment S September 30, 1993 l

1

n. - -

CESSAR !!nincuiu .,

I S8/4fe- l l

TABLE 3.8-5 (Cont'd)

(Sheet 10 of 10)

LOA OMBINATIONS FOR CATEGORY STRUCTURES Definitions:

i 0 -

Dead ad I l F -

rostatic Forces of ign Basis Flood  !

H -

Lateral Earth Pressure - '

~'

E' -

SSE Seismic Load '

Wind Load W, -

Tornado Load j t

i l

4 l

l l

l

l l

l l

'l l

Amendment S September 30,.1993 I

.q

CESSAR !a'ariumu TABLE OF CONTENTS (Cont'd)

APPENDIX 3.8A Subject Pace No.

Ses11on STRUCTURAL DESIGN LOADS AND LOAD COMBINATIONS 3.8 A-8 5.0 5.1 DESIGN LOADS 3.8A-8 NORMAL LOADS 3.8A-8 5.1.1 D - Dead Load 3.8A-9 5.1.1.1 ,

3 pac e w f 5.1.1.2 L - Live Loads I C 3.8 A-9 5.1.1.3 H-Soil ad 3.8A-11 [

5.1.1.4 F - Hydrostatic Loads 3.8A-12 To - Thermal Loads 3.8 A-12 5.1.1.5 Ro - Pipe Reactions 3.8A-13 5.1.1.6 SEVERE ENVIRONMENTAL LOADS 3.8A-13 5.1.2 W - Wind Loads 3.8A-13 5.1.2.1 EXTREME ENVIRONMENTAL LOADS 3.8A-13 5.1.3 E' - Safe Shutdown Eanhquake (SSE) 3.8A-14 5.1.3.1 W - Tornado Loads 3.8A-14 5.1.3.2 i 5.1.4 ABNORMAL LOADS 3.8A-16 5.2 DESIGN LOAD COMBINATIONS 3.8A-16 5.2.1 GENERAL 3.8A-16 5.2.2 LOADING COMBINATIONS FOR SEISMIC CATEGORY I 3.8A-17 l CONCRETE STRUCTURES LOADING COMBINATIONS FOR SEISMIC CATEGORY I 3.8 A-18 5.2.3 l l

STEEL STRUCTURES 1 5.2.3.1 Service lead Conditions 3.8A-18 Arnendment T November 15, 1993 ii 1

CESSAR Knema tka 3

'" gg,J

.*w. j gs Ib d .0*SV ,

1.0 INTRODUCTIO.N l l

nis appendix provides the criteria for the analysis and desi of structures that comprise the System 80+ Standard Plant.

i The information presented in this appendix shall be in the analysis and design of Seismic l i Cdm, CL d4 structural components Category comprising theISystem and H/9:j C:= 2,Plant 80 + Standard -d structures Fin C :;;p& m-l _ Ti.; 3.2 :. Design requirements V for individual structures are based upon their seismic category and safety classifications listed in Table 3.2-1. The criteria for the Steel Contamment Vessel are provided in Section 3.8.2 and are l excluded from this appendix, g y All structures required to shut down and maintain the reactor in a safe and orderly condition or prevent the uncontrolled release of excessive amounts of radioactivity following a Safe Shutdown Earthquake have a classi'ication of Seismic Category I. Rese strucmres shall be designed to withstand, without loss of function, the most severe postulated plant accident or natural phenomena for the site.

Safety classifications are defined in Section 3.2.2. Structural components required as part of the primary contamment pressure boundary or for its support and under the scope of the ASME Boiler and Pressure Vessel Code are Safety Class 2. All other structural components required to perform safety related functions are Safety Class 3. Safety Class 1 applies to the NSSS primary system components. Safety Classes 1 & 2 are not applicable within this appendix.

nose non-Seismic Category I structures capable of impairing the functioning of any Seismic Category I structures or component in the event of failure are classified as Seismic Category H.

Seismic Category H structures are designed to prevent failure in the direction of a Seismic Category I structure or component under extreme emnronmental or accident conditions. He seismic design requirements for Category H structures under these conditions is equivalent to tiat of %=k Category I structures. g q g p);g e 6 ed[ .

Sersmic Category I and H, Non-Nuclear Island structurebclude the ine Building, Diesel Fu Storage Structure, Component Cooling Water (CCW) Heat Exchanger tructure, CCW Pipe Tunnel, Radwaste Facility, met Service Water Pumphouse & Intake Structure. Also included is the concrete dike surrounding the outside CVCS Boric Acid Storage Tank (C;_;d Wij I, Cdrj CI-4, l Holdup Tank (Cd;, C:~Am3), and Reactor Makeup Tank (Cds, C:n FMC).

The dike surrounding the Station Service Water Pond is site specific and is not addressed within th V appendix. -

a i Primary structural components consist of concrete floors, roof slabs, foundation basemats, walls, beams, and columns. Steel beams and columns will be included within this appendix if their prunary function is to provide support to walls, floors, or roof slabs. H _. . . a , ,' 9 1 Y

Ammimm U 3.8A-1 December 31,1993

m .

C E S S A R H uine m a C es men +

b w'\ .sde*c h W tsamplepuentsuppuzt will meet the code requirements of this appendix,ler ecific load and functional requirements udst$e addressed,under specific design criteria / specifications.

W' ( k SectiesJ.P.212nd Appeseln Information presented in this appendix is sutticiently comprenensive in nature to;

a. provide the criteria necessary to perform an analysis and translate that analysis into a final design, and

b. provide a correlation of analysis, design, and construction requirements with those in Sections 3.8.3, 3.8.4, and 3.8.5.

Miscellaneous components, while not primary structural components, must be considered in the design of pnmary components as to their loads and method of attachment. Design of these components is based upon the allowable loads and design requirements found in the ACI, ANSI, ASME and/or other specialized codes.

Design parameters or information indicated "(by COL)" are delegated to the Combined Operating License Applicant for completion as part of the site specific final design.

2.0 DEFINITIONS AND ABBREVIATIONS 2.1 DEFINITIONS Combined Operatmg License Combined Construction Permit and Operating License with conditions for a nuclear power facility issued in accordance 10CFR part 52 Subpart C.

Design Engineer For this criteria, the person given responsibility by the Plant ,

Designer to provide final approval for any structural design activity.

Exceedance Value A value for a design parameter based upon a selected probability that the identified value will not be exceeded.

Plant Designer A team of Architect Engineers and NSSS vendors who have l

the responsibility to develop and complete the System 80+

Standard Plant design.

Quality Class QA program classifications as identified by ABB-CE and included in CESSAR-DC Table 3.2-1. Safety related Category I & II structures will be Quality Class 1.

Safety Class Relative importance of fluid system components and related  ;

equipment as classified in ANSI ANS 51.1 (reference CESSAR-DC Section 3.2.2) Safety Classes 1, 2, 3, and NNS.

Amendment U 3.8A-2 December 31,1993 1

CESSAREin%uia

.D98 Diese/ Germrtfor AreL y Nt*nS Discharge Structure EFW Emergency Feedwater EPRI Electric Power Research Instante FHA Fuel Handling Area FPH Fire Pump House FSAR Final Safety Analysis Report GDC General Design Critena/ Criterion HIC High Integrity Container HVAC Heating Ventilation and Air Conditioning HVT Holdup Volume Tank IAC(s) Instrumentation & Control (s)

ICI In< ore Instrumentation IRWST In-Contamment Refueling Water Storage Tank IS Intake Structure ITAAC Inspections, Tests, Analyses and Acceptance Criteria LBB Leak-Before-Break MB CVCS & Maintenance Area MS Main Steam Valve House MX Miscellaneous Buildings NA Nuclear Annex NFPA National Fire Protection Association NI Nuclear Island NRC Nuclear Regulatory Commmion NSSS Nuclear Steam Supply System NUREG NRC Technical Report Designation ,

PAP Personnel Access Portal PMF Probable Maximum Flood PMP Probable Maximum Precipitation PRT Pressunzer Relief Tank PRZ Pics-M QA Quality Assurance l RA Reactor Shield Building Annulus l RB/RXB Reactor Building  ;

RC Reactor Building Steel Containment Vessel l RCP Reacwr Coolant Pump I RDT Reactor Drain Tank l RFAI Relay House  ;

RS Reactor Building Subsphere i RW Radwaste Facility SAR Safety Analysis Report SB Station Service Building SCS Shutdown Cooling System i SD Station Service Water Discharge Structure  !

SER Safety Evaluation Report (NUREG-1462)

SF Spent Fuel Storage Area SG Switch Gear Building SI Station Service Water Pump Structure Amendment T 3.8A.4 November 15, 1993

CESSAR !!niflCATION 3.1.2.1 Containment Shield Building

%e Containment Shield Building (see Section 10.2 of this appendix) is the concrete structure that surrounds the steel Contamment Vessel and Reactor Building Subsphere and provides protection from postulated external missiles and other environmental effects. De Containment Shield Building provides an additional barrier against the release of fission products.

He Shield Building has a 105' inside radius, 4 feet thick, cylindrical reinforced concrete shell extending from the foundation basemat at El. 50'-0* to El.146'-W.' The cylindrical wall extends upward from El.146' with a 3 ft thickness to the spring line at El.157'-0". The Shield Building is topped by a 3 feet thick reinforced concrete hemispherical roof. The outside apex of the dome is at elevation 265'-0".

3.1.2.2 Reactor Building Subsphere The Reactor Building Subsphere (see Section 10.3 of this appendix) is located inside the Shield Building and external to the Contamment Vessel. The Subsphere consists of reinforced concrete walls and slabs and the Containment Support Pedestal. He purpose of the subsphere structures is to support the contamment vessel and the Internal Structures and isolate safety related equipment.

3.1.2.3 Containment Internal Structures The Containment Internal Structures (see Section 10.4 of this appendix) are located inside the spherical steel containment vessel. He purpose of these internal structures is to provide stmetural support, radiation and missile shielding, and space for the IRWST. These structures are constructed of reinforced concrete and structural steel. Rese structures are described in Section 3.8.3.1.

3.1.3 NUCLEAR ANNEX ne Nuclear Annex (see Section 10.5 of this appendix) is a multi-level reinforced concrete structure surrounding the Reactor Building. He Nuclear Annex is integral with the Containment Shield

- Building and provides lateral bracing while providing partial tornado wind and missile protection.

He Nuclear Annex provides protected areas (Control Complex, Diesel Generator Area, Fuel Handling Area, CVCS Area, and Main Steam Valve House) for safety related equipment. Structural components provide biological shielding required as a result of handling nuclear fuel or processing radioactive wastes.

NosM244HRISL AND 1 3.2 SEISMIC CATEGORY I AND II STRUCTURES g Refer to Section 11.0 for detailed descriptions of the following:

• Diesel Fuel Storage Structure - Category I,

• Component Cooling Water Heat Exchanger Structure - Category I,

• Radwaste Facility - Category II,

• Service Water Pumphouse and Intake Structure - Category I,

• Turbine Building - Category II, Ammiment T 3.8A-6 November 15,1993

CESSAR Knem  !

h Men:f' dd sipJulfeA AZ A; femsnen} olessl had need no-} de.

I y .sicsib 2 fhcKed pmided ik .rize sud /salhx ne e-pecfad l c h rsMsin. Crr st'sd.- l l

5.1.1.1 D - Dead Imad Dead load refers to loads which are constant in magnitude and point of(pplication. Dead loads are the mass of the structure plus any permanenthpettmainsi equipment loads +"D" may also refer to the internal forces and moments due to dead loads. The effects of differential settlement shall be considered with dead loads. //gdes.rdt dc /s/de fra m . 4m5 /fs d /A,// 4.e/s V shs// b< cn.udd uiM dead b.zds.

Uniform dead loads represent the structural mass, miscellaneous equipment, and distribution system (electrical cable trays and mechanical piping or HVAC) loads. Specific loads for designated equipment are represented by concentrated loads at the point of application.

5.1.1.2 L - Ihe Loads Live load, also referred to as operating load, refers to any normal load that may vary.with intensity and/or location of occurrence. Variable loads include movable equipment or equipment that is likely to be moved. "L" may also refer to the internal forces and moments due to live loads.

Live loads are applied to the structure as either concentrated or uniform loads. For equipment supports, live loads should also consider contributory loads due to the effects of vibration and any support movement. L. -- , J = &- - :f i " :'-M '- -2 ; ; " " ' - A y

lueukunsMmeMEE Design drawings prepared by the COL applicant should show allowable loads for the designated laydown areas.

5.1.1.2.1 Precipitation ne minimum design live load due to precipitation (rain, snow, or ice) for Seismic Category I buildings shall be taken as 50 psf. His live load, equivalent to approximately 9%" of water, will be sufficient for the design peak rainfall of 19.4 in/hr or 6.2 in/5 min given in Table 2.0-1. He design load for rain shall also include the additional load that may result from ponding due to the deflection of the supporting roof or the blockage of the primary roof drains.

5.1.1.2.2 Compartmental Pressure Loads Compartments shall be evaluated for the potential for inwrnal pressurization. Pressure loads associated with tornadoes, LOCAs, or other explosive type loads shall be classified as extreme environmental or abnormal loads. See Sections 5.1.3.2.1 and 5.1.4.1.

5.1.1.2.3 Truck Imads Loads due to vehicular traffic in designated truck bays is in accordance with standard AASHO truck I loading or identified special loads. Special loads may consist of construction or maintenance loads or routine shipments of fuel casks or other high level radioactive waste. j Amendment T 3.8A-9 November 15, 1993 l

CESSAR !!aineam,.

h.6 fut/ we y}er.r 2n A heentnem+s AffwY d rey; red S 5.1.1.2.4 Rail Loads Design of th railltruck bays ~ controlled by anticipated shipping weights.

5.1.1.2.5 Cranesc Elevators, and Other Hoists SM b Ris criteria is maintenance as w 1 licable to ermanently installed cranes required for station operation and emporary construction cranes. The structural design shall consider the

/

placement of construction hoists on floors, walls, and columns. Design loads shall include the full rated capacity of the hoists plus impact loads as well as test load requirements.

Test loads shall be evaluated as 125% of the crane rated capacity. The test loads shall be increased by an additional 25% to account for impact. Test loads shall be checked in Service Load combinations with a factor of 1.1 applied instead of the 1.7 factor normally applied to live loads.

He factor is reduced because the test loads are known and the tests are performed under controlled conditions.

For construction cranes located adjacent to the structure, the structural design shall include soil surcharge loads produced by the fullload of the crane. Cranes permanently mounted to structures j shall be identified on general arrangement drawings.

Pendant operated traveling cranes and trolley hoists shall be designed for 110% of the rated load ,

l capacity, to account for impact as required by ANSI N690 Section Q1.3.2. Design loads for motor operated trolleys and cab operated traveling cranes shall be increased by 25% of the rated load capacity to account for impact in Service and Factored load combinations.- -l Mmimum lateral design loads on crane runways shall be 20% of the sum of the rated hoist capacity plus the weight of the crane trolley to account for the effects of the moving trolley. Load shall be applied at the top of the rail in either direction and distributed according to the relative stiffnesses of the end supports.

Mimmum longitudinal load on each crane rail shall be 10% of the maximum crane wheel loads.

1 Elevators live loads shall be increased by 100% for design of supports.

5.1.1.2.6 IAad Allowances for Cable Trays Loads to be applied in areas where multiple cable tray runs are identified include: .

1

• 7 kips at mid-span on steel beams and columns.

• 7 kips at a spacing of 8 ft on center for slabs.

Acceptability of these design loads will be determmed through review of the final electrical layout drawings prepared by the COL applicant.

Amendment T 3.PA-10 November 15, 1993

1 C E S S A R !a Wic a ,.

9/ Acceptslinfp of these de.ryn />21s aW/ ha. at:Ra.d 7%<>ql.

ren:es > f 4 O u l y /a d a n R p rz +t a ~ . _

5.1.1.2.7 Miscellaneous Equipment and large Bore Piping l The following load allowances shall be considered where multiple large bore piping runs are located l or where large temporary loads are identified.

• In addition to major equipment located on general arrangement drawings, a point load of 20 kips ""

should be applied at the midpoint of each concrete =g "="  :

floor slab and concrete beams (Case A). 30g case c

• A point load of 40 kips shall be applied at the i

• C* 02 l midpoint of steel collector beams providing pnmary 3er cese c gg framing (Case B). "

• Ccse B e

• A point load of 30 kips shall be applied to the E cm m midpoint of other steel collector beams or beams --

provided for support frammg (Case C). h .. ,,

3 g,,, g ,,

• A point load of 30 kips at midspan on prunary steel filler beams frammg into steel collector beams (Case D1) and 20 kips on other steel filler beams or stringers (Case D2). (Note: These loads are for added design margin on the beams and slabs and are not to be carried beyond the beam support connection to the supporting beam or column.)

• A contingency load of 80 kips should be applied to the top of each steel column. y 5.1.1.2.8 Miscellaneous Equipmmt, Small Bore Piping, Cable Tray, and HVAC l Ductwork The following load allowances should be included for areas with multiple runs of small bore piping, cable tray, or HVAC ducts.

• A load of 15 kips on steel collector beams

• A load of 5 kips on other steel beams

• A load of 50 kips on steel columns y 5.1.1.3 H - Soll Load 90 b so' pr re s allbe ased n th il d ity s 1in ude3he eff of to w rd ce th s 'on .1.1.4 f this pen .N ilI ads r&ll co ider gro w

\ Iev up 88' ",2' " belo pl finish yar el ion . 90 -9"). Thelate son g,gt? p s e sh be ased on tb folio mg s pro erties

• ' S D ity

[\'$

/

5,g, 25 ds ppr cub' foot ef), o 80 p . dry /

mo' soil ,

l 145 cf s Amendment U 1 3.8A-11 December 31,1993 l

l l

l

CESSAR EininCAMpd

.gdLGd6kw M

~

Rer> ' Angle o

ternal *on,4; 3

• Co scient of f etion, con ete on soil; use p = 0.5 assuming con e is poured ireedy on petent staletural bacistil without any intervening' material, su as waterpt fing.

/

Re at-r t soil pressu/re shall beg cafculated usin/

a coefficien of earth pr sure at rest, ,

of 0. f

• a coeffici, of passive ear'th pressure K p, shall be etermmed based upon angle of mternal 'on, c. K, ' tan (45' ,+' 4/2). He . ects of buildings, y es, ,

{

material ekpiles, etc. g as sur argeloads o e soiladjacent to exteri building is shall o be consider .

For fa ed load combi 'ons the i eral soilload I be bas 41pon satu soil ass lated with flood' and a grou ater leve ' 0" below th plant finis yard gr 5.1.1.4 F -Ilydrostatic Imads .g g j g f,,,

Hydrostatic loads are due to ground water, exterior flood waters, or fluid ^ ^[in internal compartments, including internal flooding.

Maximum flood level is specified to be l'-0* below fimshed plant grade. Site specific flood l elevations greater than this will be addressed by the COL applicant.

5.1.1.5 To - Thermal Loads Hermal effects consist of thermally induced forces and moments resulting from plant operation or enytronmental conditions. Thermal loads and their effects are based on the critical transient or steady ,

state condition. Gem 2/ eyansin- /*sa(s due A 1; tid / s \

leAir ressHiy paw &ms/g.ed;cJ.s .1As/ se cmaSere/ethf d. $$ veeYM \

ne following ambient temperature values during normal conditiom shall be used as a basis for l design. Site specific provisions may be taken to mmimm the effects of the structural temperature gradients produced by these conditions.

External ambient conditions, reference Table 2.0-1.

Outside air temperatures - 100'F max.

-10*F min.

Ground Temperature- 50*F Internal ambient conditions, reference Appendix 3.11A and Sections 10 and 11 of this appendix.

Thermal analysis may be performed to determme concrete surface temperatures.

Amendment U 3.8A-12 December 31,1993

771< fe h % a s f F M e n + .sh s /I b e .f u tler. ukaasl abet I inferytaly 27uhr/2/s 2re useol.

URNpm 3.B A 5.1.1.3 R-soll Load Lateral soil pressuro shall be based upon the soil density and  !

shall include the effects of ground water in accordance with section 5.1.1.4 of this appendix.

a ground water level u Normal soil loads shall consider 88'-9", 2'-0" below plant finished yard grade elevation (p to El.

Bl. 90'-9').  ;

. Soil Donalty (y) : Saturated Soil = 145 pcf (pounds / cubic foot)

Moist Soil = 125 pcf Dry Soil = 80 pcf Angle of Internal Friction: $ = 30' '

tan. (Au d, 0.57)

• Coefficient of friction: p assuming the concrete is poured direct y on competent structural backfill withoutanyinterveningmaterial,suchaswaterproofingj.tr,.ftex/

.if concretc a,s noff*ure direcfly onn 7Yte . sci /.

At-last lateral soil pressure coefficient Ko = 0.5 (Used in Service Load Combinations)

• Active lateral soil proccure coefficient : Ki =. tan (45' - $/2) a Passive lateral soil pressure coefficient: K, = tan (45* + $/2) 2

.' Active lateral earthquake soil pressure coefficient: K.

r ,- 81"' (% + 0 'O#)

a cos (O') sin (p) ein(p-O'-8) [1+$ sin sin (6 +8 ) sin (4-0/-a )

(@-6-B') ain(s +p) ): ,

Passive lateral earthquake soil pressure coefficient: K,,

K,,= sin 8G +0'-4) l i

sin ($ +6 ) sin (4+ s-O')

cos (0') sin * (p) sin (5 + p +0'-9 0)sin (1 (p $ +6 +0') sin (a+p)3 l

l where: k3 y,

(1-A,) l

. neraft k, = horizontal earthauake acceleration comnonent acceleration due to gravity, g JWLFA.f6  !

A =Jvertical earthnuake accoloration \

/'

acceleratlon due to gravity, comuonent. 9 J

G

<m_, ~.7 -, . -,

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

8.),)e$ (C#NNb '

a = the slope or angle of the backfill surface as measured from the horizontal p= (1) The. angle formod by the exterior face of the wall and the horizontal. (2) The angle shall be measured as 180" minus the angle formed by the exterior wall surface and the horizontal direction extending out under the backfill. (3) This value will be 90* for vertical walls.

8= the angle of wall friction is a quantitative value, expressed in degrees, used to define the level of friction between soil backfill and the retaining structure The total lateral earth pressure is calculated as; e

At-rest lateral soil pressure: P. = %KoyH 8 e Active lateral soil pressure: P, = %KayH8 V

. Passive lateral soil pressure: P, = %K,yH 2

. Active lateral earthquake soil pressure: P,, = %KanyH 8 (lik,)

. Passive lateral earthquake soil pressure: P, = %K,sTH8 (lik,)

where; y = soil density (pcf)

H = height of soil-wall interface (ft)

Rt.Fts. tact. : bh5, s.M. , Pswurns, of MudbnTrohl EnkuMc=Jg _

s ss** sa. r PWs-KLwr Neumone, ca. , Soyrant, M%. ~

The offects of buildings, vehicles, cranes, material stockpiles, etc. acting as surcharge loads on the soil adjacent to exterior building walls shall also be considered.

For factored load combinations the lateral soil' load shall be based upon saturated soil asacciated with flooding and a ground water level l'-0" below the finished plant yard.

CESSAR n!&"icari:n i

de -Ad/ potenf'2/ //h. /not sdd/ de used ;Ge /sez/

Rn1/ysed af .s&u?'a.n/ msxJers.

5.1.3.1 E' - Safe Shutdown Earthquake (SSE) l SSE loads are loads generated by an earthquake with a peak horizontal ground acceleration of 0.30g.

Refer to Section 2.5.2.5.1 of CESSAR-DC.

Total loads for E' shall consider simultaneous seismic accelerations acting in three orthogonal directions (two horizontal and one vertical). Each of the three directional components of the earthquakes will produce responses in all three directions. Colinear responses due to each of the 3 individual earthquakes may be combined using the " Square Root of the Sum of the Squares" (SRSS) method. The resultant nodal loads are applied simultaneously to the structure. The seismic forces l and moments may also be combined simultaneously using directional combination participation factors of 100%/40%/40% applied to the individualloads produced as a result of each earthquake to produce the design SSE loads. The critical load combination would use 100% of the loads due to one earthquake and 40% due to the other 2 earthquakes, i.e., E of F, due to 100%E', i40%E'y 40 %E',.

SSE loads are obtained by multiplying the dead load and 25% of th design live load by the structural ;j acceleration obtained from the seismic analysis of the structure. Amplification of these accelerations V due to flexibility of structural members should be considered. Construction loads are not required to be included when determining seismic loads. Other temporary loads must be evaluated for applicability on a case b case basis.

f Seirgic ssil /s2 s .S'At// de a. sed m. e m bin d 'an u d, ) J .,

c_ :a.:e = ==; e tc:a a h-_ :n. cL . .#._ ,

nt rm M """ fr cpp!!:di'ig n _ L, _ L~'.n

. /

4 SSE damping values used in design (Reference NRC Reg Guide 1.61 and Table 3.7-1) shall be as follows:

Structure Tvoe  % of Critical Damoine Welded Steel 4 Bolted Steel 7 Reinforced Concrete 7 Prestressed Concrete 5 Equipment (steel assembly) 3 Fluid sloshing loads in the IRWST, Spent Fuel Pool, and all other fluid reservoirs due to the SSE shall be considered in accordance with ASCE 4-86.

5.1.3.2 W,- Tornado Loads Loads generated by the design tornado are as identified in Section 3.3.2. Tornado loads include loads due to the tornado wind pressure (W,), the tornado created differential pressure (W,), and tornado-generated missiles (W,). Twenty-five percent of the design live load shall be considered with ,

tornado load combinations. t

,7'g f f ); g y y g g g fe has/ aalg.ses 7)feo/yg* .si / y,%dml iner,1hers . .

Amedmm U 3.8A-14 December 31,1993

CESSAR n!E"ioarian l

Design for missile impacts shall be in accordance with Section 3.5.3 and ACI 349, Appendix C.  ;

Minimum concrete wall and roof thicknesses shall be in accordance with Standard Review Plan 3.5.3 I Table 1. Non-Category I structures shall not be assumed to shield seismic Category I structures from tornado wind, differential pressure, or missile loads.

)

l 5.1.4 ABNORMAL LOADS  !

Abnormal loads are those loads generated by a postulated high-energy pipe break accident. Als event is classified as a " Design Basis Accident". Included in this category are:' Pressure loads (P.),

Thermal loads (T,), Pipe reactions (R,), Load on the structure generated by the reaction on the pipe (Y,), Jet impingement loads (Yp, and Missile impact loads (Y ). Rese loads are defined by:

• P, - Pressure equivalent static load within or across a compartment and/or building, generated by the postulated break, and including an appropriate dynamic load factor to account for the dynamic nature of the load.

T - Hermal loads generated by the postulated break and including T,.

R - Pipe reactions generated by the postulated break and including R .

• Y,- Equivalet stric load on the structure generated by reaction of the broken high-energy pipe during the postAated break, and including an appropriate dynamic load factor to account for the dynamic nature of the load.

• Yj- Jet impingement equivalent static load on a stmeture generated by the postulated break, and including an appropriate dynamic load factor to account for the dynamic nature of the load.

• Y, - Missile impact equivalent static load on the structure generated by or during the postulated break, such as pipe whipping, and including an appropriate dynamic load factor to  ;

account for the dynamic nature of the load. .j 5.2 DESIGN LOAD COMBINATIONS 5.2.1 GENERAL ne following loading combinations shall be used for analysis and design of Y

Category I structures and their components.

Live loads shall be applied (fully or partially), removed, or shifted in location and pattern as necessary to obtain the worst case loading conditions for maximizing internal moments and forces for all load combinations. Impact forces due to moving loads shall be applied where appropriate.

Where any load is determined to have a mitigating effect on the overall loading for a steel or concrete structural member, a load coefficient of 0.9 should be applied to that load component. He reducing coefficient should be used on13 for that load which can be demonstrated to be always present or occurring simultaneously with other loads. For loads which cannot be shown to be always present, Am* nam ~* T 3.8A-16 November 15,1993 i

CESSARMnema )

.I Ductility ratios determined from ACI 349 Appendix C should be used. Deflections shall be evaluated '

for potential loss of function for safety related systems.

Load combination 5.2.2.2 b) shall first be satisfied without the tornado missile load. Load combination,5.2.2.2 d) shall first be satisfied without the Y loads. When including these loads however, local section strength capacities may be exceeded under the effect of these concentrated ,

loads, provided there will be no loss of function of any safety related system.  :

.I Structural effects of differential settlement, creep, or shrinkage shall be included with the dead load.  :

l 5.23 LOADING COMBINATIONS FOR SEISMIC CATEGORY I STEEL l STRUCTURES I

'Ibe following set of load combinations define design requirements used for all Seismic Category I steel structures, j 5.2.3.1 Service IAad Conditions l

5.2.3.1.1 If elastic allowable strength design methods are used:

a) S=D+F+L+H b) S=D+F+L+H+W If thermal stresses due to T and R, are present, the following combinations are also satisfied:

d) ASS = D + f + L + // + b + 75 (Vension mentbers)

,ay 1.3 S = D + F + L + H + R, + T. (cempuss/ex maxbrs) d) /,5J6 13 S S ' D=+Df + + 6F++L{ +b H + +y

                                                  + 7,  R, + T + W (6ension armhers) Cow)*ressio For steel members, S is the required section strength based on the elastic design methods and the allowable stresses defined in Part 1 of ANSI /AISC N690 5.2.3.1.2             If plastic design methods are used:

a) Y = 1.7 (D + F + L + H) b) Y = 1.7 (D + F + L + H + W) c) Y = 1.3 (D + F + L + H + T + R ) d) Y = 1.3 (D + F + L + H + T + R, + W) For steel members Y is the section strength required to resist design loads based on the plastic design methods described in Part 2 of ANSI /AISC N690.' Amr.ndment T 3.8A-18 November 15,1993

CESSARin!L mu

                   /,6t S e.9+ffJ f h'+ 4 + Tx + Pa, (coyressson Member.s)
                    } 4 S' .Dtf+/ +H+ Rs + % + We                        (c*yre.ssisn mem.bers) 5.23.2                   Factored lead Conditions 5.2.3.2.1

2) /.t s .D +If Felastic allowable strength design methods are used:d'")

                                         + L + H + R + *& + F <tensin. M'
         .a'r     1.4 S = D + F + L + H + R, + To + E'                    (congressiox preWS)

( b)/.(e.k4 S = D + F + L + H + R, + T, + W. (fen.y/sn memhers)

                ' (refer                                                                                              y c M.,k4      5 = toD+  item F +5.2.2.2 L + H + b)   R,for
                                                           + T,components
                                                                + P,             of W,) Went/sv httMbdr$)

d)/, .k6 S = D + F + L + H + R, + T, + (Y,+Y)+ Y ) + E' + P, 8 ens /n umitet)

                  /.tnS s W+ f + J. 4M + R1 + 75.+ (Yr+                      Ys)+    1'+ Ps
                                                                                  >53for  this load combinatio .)6mpr<ssis astic section modulus for steel shapes may be us

-[ 5.23.2.2 If plastic design methods are used: a) Y* = 1.0 (D + F + L + H + Ro + To + E') b) Y* = 1.0 (D + F + L + H + R, + T + W) i (refer to item 5.2.2.2 b) for components of W,) c) Y* = 1.0 (D + F + L + H + R, + T, + 1.5 P.) l d) Y* = 1.0 (D + F + L + H + R, + T, + Y, + Y j + Y, + E' + P,) use 0.9Y for Internal Structures and 1.0 for all other Category I structures. - (Reference SRP 3.8.3.IL5) 5.2.4 LOADING COMBINATIONS FOR SLIDING, OVERTURNING, AND FLOTATION Minimum Factors of Safety l l Load Combination Overturmng Sliding Flotation D+H+W 1.5 1.5 - D+H+W. 1.1 1.1 - D + H + E' 1.1 1.1 - D+F - - 1.1

                     . . . . . - . . . . . . . . . . . . - .      ,.,.                                              /

5.2.5 CONSTRUCTION LOAD COMBINATIONS Service load combinations shall be used to evaluate construction methods and sequence and determine structural integrity of the partially erected structures. 5.2.6 APPLICABILITY OF LOADS Lateral loads due to soil bearing pressure shall apply to all exterior walls up to El. 90'-9". Ammament U 3.8A-19 December 31,1993

CESSARMu% - 1 l l Structure Interaction (SSI) analyses described in Section 3.7B. For each elevation these ZPA values I are enveloped from the values of each stick in the SSI model at the corresponding elevation. His  ! enveloping of ZPA values at each elevation is repeated for all control motions and all soil cases, and a final envelope of ZPA values for each elevation is determined. The ZPA values are further  ; amplified where necessary to account for floor slab flexibility. Figures 3.8A-3 through 3.8A-5 show this envelope profile for the NS, EW and vertical directions respectively. This envelope of ZPA ) values is applied as a uniform factor to the floor mass and contributing portions of the wall masses  ; at each elevation within the structure as the applied seismic loading. For the soft soil model this envelope loading is obtained from an envelope of soil cases B-2, C-2, C-3 and C-1.5. Rese soil cases represent the soft soil site category. Applying the envelope of the l soft soil ZPA values is compatible with the soil stiffness modeled for the soft soils. He enveloping ZPAs are used in the local analyses to determine the forces and moments from the  : inertia loads. The masses in the local models are accelerated by the appropriate ZPA value for the l elevation being analyzed and the forces are applied as static point loads, static body forces, or static j uniformly distributed loads. For each load the response from all three directional earthquakes are combined simultaneously. De  ! independent directional responses are combined using the square root of the sum of the squares j (SRSS) method or the 100-40-40 Percent Rule described in ASCE 4-86. He 100-40-40 Rule is based i on the observation that the maximum increase in the resultant for two orthogonal forces occurs when these forces are equal. De maximum value is 1.4 times one component. All possible combinations l of the three orthogonal responses are considered. He 100-40-40 combination is expressed .I mathematically as: R= (il.ORx i 0.4Ry i 0.4Rz) or, R= (i0.4Rx i 1.0Ry i 0.4Rz) or, R= (i0.4Rx i 0.4Ry i 1.0Rz) He 100-4040 Percent Rule may also be applied for combining responses in the same direction due to different components of motion. f Additional seismic loads due to accidental torsion is accounted for as required by SRP Section 3.7.2.II.11. This accounts for variations in material densities, member sizes, architectural variations, equipment loads, etc., from design assumptions. Due to these potential variations, an additional eccentricity of the mass at each floor equivalent to 5% of the maximum building dimension is included. The accidental torsion load is an additional shear force at each floor elevation determmed l based on a percentage of total accumulated shear at each elevation. j Re dynamic increment for horizontal soil loads on the exterior walls of the Nuclear Island, CCW beat exchanger structure and diesel fuel storage structure is ' ed from the 2D SSI analyses as described in Section 3.7. For other structures, the elastic method in ASCE 4-86 is used. I I Ammima* U 3.8A-21 December 31,1993

CESSAR EEnCATI::N 97o6>wek f -%%: ey}1.8,y.s:/. war >t andes sk/ med 1%e refkw

• Transverse reinforcing at the edges of wall panels shall be anchored in accordance with Paragraphs 21.5.3.5 and 21.5.3.6
• IAngitudinal reinforcing for beams shall be anchored according to Paragraph 21.6.1.3 with l hoop reinforcement per Paragraph 21.6.2.1
• Development lengths for reinforcing will be according to Paragraph 21.6.4.

Epoxy coated reinforcing shall be used for exterior walls and slabs when the existing groundwater is determined to be sufficiently corrosive so as to adversely affect the long term durability of the concrete structure. 'Ibe required splice length given in ACI 349 Section 12.2.2 shall be increased using factors provided in ACI 318 Section 12.2.4.3. When feasible, uniform reinforcement patterns should be used for sections with similar requirements, thickness and loading. 6.2.1.1.2 Concrete Expansion Anchors an rs s Ibe the edge, eeve, under design specifi in S on 3.8. 5. ni d ign s ety f rs all be-

          *       . for edge         si    e typ ancho

• 3.0 und t ty e ancho of 1.5 f coner e fail ewi p n an r ments - Iha a mini m factor f saf f

resp to or imum ile . ectio of ion chors all cons' er ener abso on cap ity e. d ' ity f th e ancho .  ; A specification for the design, installation, and use of expansion anchors should be developed by the COL Applicant and include;

• expansion anchor allowableloads,
• expansion anchor minimum spacing,
• spacing requirements for expansion anchors,
• procedures for addressing baseplate flexibility's in calculating design loads on expansion anchors,
• procedures for addressing shear tension interaction, and a required load reductions for cyclic loadings.

When high capacity concrete anchors are specified, they should be of the direct bearing or " undercut" type. Imad transfer for these anchors is achieved by bearing of the expanded embedded tip against the undercut concrete hole produced by a special flaring tool. Undercutting of the concrete is required for the anchor to provide the concrete shear capacity to match the high strength bolts. For smaller safety related or non-safety related applications expansion anchors referred to as

        " Sleeves" or " Wedges" may be used, subject to the safety factors given alem. In Sect /m 3, f. Y,5.1

• A==dm= U 3.8A-24 December 31,1993

CESSAR sinincoc,.  ! 6.2.1.2 Steel The design of Category I steel structures and/or components shall use Allowable Strength Design methods in accordance with ANS SC N694 S ;;5:m' x f_:xn (...' ._..~ aum 1 3.;. :.2 , as utende hg Sacho>s. 5,g, L S'.2. j

   /     *

• econd

                                                                                                                   ~

Stress applies nly to tem rature lo ings, Q1.0. . l

• dditio notes r Sectio 1.3. l
• Eff due to fferential ement s I be inct ed with d loads.

Off ' g1 in any I comb' on shall h e a load fa or of 0.0 ess they ar always p ent or et simul usly wi ther loads which e the factor hould be 0 . l

• tress 1 't coeffic' are m ified as s wn in S on 5.2.3. c) & d) d Section 1 5.2.3. 1, Table .5.7.1.

Ch eload fa r for P,in 'on 5.2. 2.2 to 1.3 ,, See Q2. (equivalent 1.5/1.1xP

                 ' ting re irements           given in S ion Q1.24 and Q1.2 shall be sup emented by e 11 0
                      '                                                                                                         f aintings o coatings for structural s I shall                 the requ'     ents of          atory Guide 1.        and ANSI        CE N10 4, "Quali           Assurance      r Protecti      Coatin Applied       Nuclear F ilities".          a suppl       tal require     t to AN      N690,s
                              ' s are to be op painted rior to d very, in a dance wi                       ection        of         l the       C MANU          OF STE        CONS         CTION, " lowable S ess Desi                and          ,

i Commen Steel s expos after install .on are to e field p or ated in ac dance with is same re ence.

• ctility factors ) in Tab Q1.5.8.1 all not ap y to cons ned (ro 'on or isplacement) embers er load binations 5 .3.2.2d). Ductili facto fro Appendix A, .2 of SRP ection 3.5. shall be subs ted for T e Q1.5. 1.

k Uniform depths of steel beams and connections should be maintained. 1 Bolted connections should be used for field erection of structural steel beams and columns. Load ) indicator bolts are recommended. The design of bolted connections shall be in accordance with ANSI  ! N690 Section Q1.16 and the " Specification for Stmetural Joints Using ASTM A325 or A490 Bolts". l Bolted connections shall be designed to be " slip critical" unless ' tified otherwise. 7fg qs ds -Or allt AtAn*4. in Seah'*w %2.3 e elded connection { " b; d='g;;d in r:xd WIN of

                                                                                          .;s ..O! :1.tME!

lr r t"!EM d ( D!l.Set.yndW - j

                                                                                                                                    \

Maximum utilization of shop fabricated connections should be considered to avoid welding in ) hazardous environments. I i 4 Transverse welds across the flanges of rolled Sections of Seismic Category I or II steel members are prohibited without approval of the design engineer. All transverse welds on Category I or H  ; members shall be shown on approved drawings. Structural members with restrained end conditions and thermal loads shall be evaluated for potential buching.  ; Amandmaat U 3.8A-25 December 31,1993

CESSAREnWicm 6.2.1.3 Missile Protection l Exterior walls and roof slabs of Seismic Category I structures are required to function as missile barriers for tornado generated missiles. Design of missile barriers shall assure that the structure will not collapse under the missile load nor will there be penetration through the barrier. Safety related structures, systems and components shall be protected from secondary missiles as a result of backface scabbing. In-ferior tJx//s 2nd {/ows sjg// je enks/co( acud s/esped y 75 S nc6 4 as M.rsik bur;cr:sg/,yujg,(, 6.2.1.4 Mre Protection / Fire protection is provided in the form of fire rated walls and barriers as identified in Figure 3.8-5. In addition to passive fire protection offered by fire rated structural barriers, the structural design shall offer protection to the active fire suppression system to assure that they will not be made inoperable due to the failure of any structural member. 6.2.1.5 Flooding Flooding is addressed in Section 3.4.4. Flood barriers are identified in Figure 3.8-5. l Protection of the Seismic Category I structures agamst flooding shall be insured by;

• allowing no access openings in the exterior walls lower than 1 foot above plant grade a having no unsealed exterior wall or floor penetrations below plant flood level (El. 89'-9",

1 foot below finished yard grade)

• having water stops in all below grade exterior construction joints a providing floor drainage 6.2.1.6 Construction Support l

Cost saving may be achieved by reducing the duration of the construction schedule. Durations may be reduced by standardizing details and using modular designs that will allow offsite fabrication and l assembly. Modular designs must consider transportability to the point of installation. 1 Connections / fit-ups with previously erected components must be considered. 6.2.1.7 Security 10CFR Chapter I Part 73 provides the regulatory requirements for physical protection of the plant l against sabotage as a result of unauthorized access. Plant designs shall prevent use of unauthorized access routes. In accordance with Part 73 Section 45(f)(1)(1), barriers shall be provided to channel access through protected area entry control points or delay any unauthorized penetration anempt sufficiently to allow detection by security personnel. 6.2.2 SPECIAL DESIGN CRITERIA 6.2.2.1 Radiation / Contamination Control The design of strucmral elements shall provide surface features to prevent the spread of commi*ian and facilitate plant cleanup. Sumps for drain lines that may collect potentially contaminated liquids , I will be lined with stainless steel over the potentially wetted surface. Concrete surfaces should be protected by a smooth surface epoxy coating where the potential exists for contambtion. Amendment U l 3.8A-26 December 31,1993  ; I I l l

i l CESSAR E!!snem,. 6.2.2.6.2 Electrical Cable Tray S. nob /k/IO M sayh dufves g i/ m h d m 4 Y l Design ofj for support of cable traysphall meet the requirements of Appendix 3.9A. 6.2.2.6.3 Support / Restraints for Piping and Its Components l l Design ofg du/hj.sduAre8a -- 'piping for andsuppogaf its components shall meet the requirements of / l Appendix 3.9A. l 6.2.2.6.4 Fabricated Embedments l He walls and floors of Seismic Category I Structures shall be provided with embedments for the 1 mounting or attachment of structures and components. Additional typical embedments should be l l provided for welding structural attachments which will reduce the number of attachments utilizing expansion anchors. Tolerances for fabrication and installation of embedments shall be provided on design drawings or in specifications issued by the COL Applicant. l De anchorage for structural embedments shall be designed based upon ACI 349, Appendix B with  ; the following exception. He assumed concrete failure cone projects out at an angue of 35' instead i of 45*. He angle shall be measured from the plane normal to the axis of the embedtr e. T.3  ; exception applies to structural embedments and headed anchors, such as " NELSON Sttds", an,. i expansion anchors. He exception is to prevent an overlapping of the concrete shear cones ;im , anchors are spaced at a "2d" spacing (reference Section 3.8.4.5) and to avoid a less than required l minimum edge distance, i l i A reduction in load capacity for embedments shall be applied for placement of anchors in the tension zone of concrete members. l

7.0 CONSTRUCTION

FORMING. FABRICATTON. ANT) ERECTION  ! 7.1 CONCRETE i l Concrete work for Seismic Category I structures shall conform to all requirements of ACI 349 and ACI 301 except as modified by this appendix. 7.1.1 CONCRETE MIX DESIGN l Concrete mix design for Seismic Category I structures, see Section 9.2 of this appendix, shall be determmed based upon field testing of trial mixtures with the materials to be used. Testing shall evaluate;

• ultimata concrete strength as well as early strength in support of an aggressive construction schedule, l
• concrete workability and consistency, e required concrete Mmieres,
• heat of hydration and required temperature control for large or thick concrete pours, and
• special exposure requirements when identified on design drawings.

Am-iment U 3.8A-28 December 31,1993 l u

CESSAREnanc-l 7.1.2 CONCRETE PLACEMENT Requirements and/or limitations on concrete placement will be determmed in conjunction with the construction schedule. A site specific construction specification should be prepared by the COL i Applicant to address requirements and procedures for concrete placement. The concrete specification should address;

• desired volume of concrete pours and rate of deposition.
• special formmg requirements, a maximum height of pours, a temperature linutations; weather conditions and concrete mix, including approved methods for temperature control, and
• curing requirements and procedures.

7.1.3 REINFORCING Fabrication and placing of reinforcing bars for concrete in Seismic Category I structures shall l conform to the requirements and tolerances specified in ACI 349 Section 7.5 and in ACI 301 Sections 5.5, 5.6, and 5.7. Consideration shall be given for modular assemblies of reinforcing. Such assemblies shall be designed to be moved without changing their alignment. Lap splices shall be prohibited for locations with tension stresses normal to the plane for the splice and for bar sizes greater than #11, except as provide by ACI 349 Section 12.14.2.1. Welding conhnforeY SksJI of i fAe shall be prohibited ex$s ks rovided for in approve ta es- ulvens n).s.m.Q 'gkey ,) h)et);q Y 7.1.4 CONSTRUCTION UENCING  ; Construction sequence will be deternuned by the COL Applicant. Additional design requirements due to the construction sequence will be determined by the COL Applicant during the final design. 7.2 STRUCTURAL STEEL 7.2.1 STRUCTURAL STEEL; FABRICATION AND ERECITON Fabrication and erection of safety related steel members shall be in accordance with AISC N690, Sections Q1.23 and Ql.25. Additional requirements are applica.ble as provided for in this appendix. 7.2.2 HIGH STRENGTH BOLTED CONNECTIONS , Bolts shall be installed and tightened in accordance with Sxtion 8(d) of " Specification for Structural j ' Joints Using ASTM A325 or .A490 Bolts.' 'Ibe use of ' load indicator" bolts or washers should be used where possible. " Snug tight" installation of bolts in " slip critical" connections shall not be permitted. _ _ _ cada-te;~fors4 st. "is, yea i.e;&,.,;q .chall b <ltmn. n

   /W nfuciq     ' de mi.          JetAi[Lv,3.8A-29                                   Amendment D December 31.1993 l

l

CESSAR E!Encam,. 7.2.3 WELDED CONNECTIONS Welding activities associated with Seismic Category I structural steel and their connections shall be accomplished in accordance with written procedures and shall meet the requirements of4NeinMSC- l\/ 8.0 STRUCTURAL ACCEPTANCE CRITEM JPJSE/LT D Structural Acceptance Criteria are specified in Section 3.8.4.5. Separation Criteria for Seismic Category I and non-Seismic Category structures and components shall be verified. 9.0 MATERIAIS 9.1 GENEIML Material shall conform to requirements for Section 3.8.4.6.1 and this appendix. Materials used should be selected based upon a proven record of service in other nuclear facilities. Materials shall be specified based upon approved codes and standards. Additional matenal restrictions or requirements may be added by the design engineer to meet anticipated design or field conditions. . With suitable qualification and no applicable material restrictions, substitute materials may be used. Materials used shall be qualified to withstand environmental conditions for normal and accident conditions. Site specific design specifications prepared by the COL Applicant should identify required qualifying environmental conditions. 9.2 SPECIFICATIONS

    'Ibe materials identified below and in Section 3.8.4.6.1 shall be considered acceptable for the analysis and design of System 80+ Standard Plant structures.

Additional materials may be added to this criteria when qualified by appropriate codes and staMards, 9.2.1 CONCRETE Concrete - compressive strength == 4000 psi (5000 psi for the Nuclear Island superstructure) Normal weight concrete with a density of 135 to 160 pcf. Cement - material shall conform to ASTM C 150 per ACI 349 par. 3.2. Cement shall conform to Type I or Type II designations except where additional qualifications are conducted for special applications. Am*Mment U 3.8A-30 December 31,1993

1 i 1 l INSERT D (to Section 7.2.3 of Appendix 3.8A)

           . . . the AWS D1.1 Structural Welding Code.

criteria shall be as defined in NCIG-01, The visual acceptance

                                                           " Visual Acceptance. I

[ Criteria for Structural Welding of Nuclear Power Plants," Revision 2, EPRI NP-5380. h

                                                                                    ?

CESSAR !!niac m Aggregates - material shall confonn to ASTM C 33 per ACI 349 par. 3.3. ASTM specification C 637 may apply where deemed necessary for radiation shielding. Limestone based aggregates should be considered for use in the floor of the reactor cavity for core concrete interaction concerns. Admixtures - Admixtures conforming to applicable ASTM standards are acceptable when qualified by testing to verify required mix design. Water shall conform to requirements of ACI 349 Section 3.4 and Section 3.8.4.6.1.1. Use of non-potable water shall be restricted in accordance with ACI 349 Section 3.4.3. Reinforcing Steel- ASTM A615 Grade 60, Fy = 60,000 psi or - ASTM A706 Fy = 60,000 psi The use of welded splices and mechanical connections is addressed under Paragraph 12.14.3 of ACI 349. Mechanical reinforcing coupler devices may be used. Epoxy coating of reinforcing shall be in accordance with ASTM A775 (ACI 318 paragraph 3.5.3.7). l 9.2.2 STEEL t 9.2.2.1- Structural Steel Structural Shapes - ASTM-A36, Fy = 36,000 psi additional material per ANSI /AISC N690 Section Q1.4.1 (excluding round & tubular shapes) Structural Tubing - ASTM-A500 Grade B, Fy = 42,000 psi , Steel Plates - ASTM A240 Type 304L Stainless Steel (ASTM A36) y' 9.2.2.2 Structural Bolts Structural Bolts shall comply with ASTM material specifications identified in Section Q1.4.3 of the ANSI /AISC Standard N690 or other materials identified in the " Specification for Structural Bolting Using ASTM A325 ora 490 Bolts". Bolts shall have nuts and washers as identified below: (and R4/AC

• Bolts- A193 A320, A325, A490,ses.,//J5p # AN/ g/
• Nuts, for A325 A194 Grade 2 or 2H nuts or A563 Grade C, C3, D, DH, or DH3, F
• Washers - F436 hardened steel washers.

High strength threaded rods such as A193 Grade B7 or A320 Grade L43 may be used in lieu of A325 bolts with qualifying documentation identifying the installation.

         * &~S hr* /?/93, }/326, //.K.Cf 3h A Nf0 ~ )Y/9Y 6'tZb 2                                    *Y
                                                                 ~        sus n koitnon.

Amendment U 3.8A-31 December 31,1993 1

i C E S S A R Eininca m ,. l i i l l I Re inner face of the lower Pnmary Shield Wall will be provided with projecting reinforced concrete corbels to be used as the support bases for the Reactor Vessel steel support Columns. Corbels shall have symmetrical reinforcing in the top and bottom to resist the upward loads resulting from a -, potential ex-vessel steam explosion (Section 3.8.3.3.H). Refer to Table 3.8A-1 for additional design loads that are applicable to the Primary Shield Wall. l l 10.4.2 CRANE WALL (SECONDARY SHIELD WALL) 10.4.2.1 Description i ne Crane Wall is a reinforced concrete right cylinder with an inside diameter of 130 feet and height I of 118'-3" from its base. The top elevation is at El. 210'-0". He Crane Wall is a mmimum of four feet thick. 10.4.2.2 Design Requirements ne Crane Wall provides supports for the polar crane and protects the steel covenmment vessel from mternal missiles. In addition to providing biological shielding for the coolant loop and equipmmt, the Crane Wall also provides structural support for pipe supports / restraints and platforms at various levels. He design shall address the vertical alignment of the Crane Wall with the corresponding structure below the Contamment Vessel and pmvides special construction tolerances, as necessary, to ensure potential misalignment is appropriately considered. De design also considers potential differential basemat settlement and the effect on the Crane Wall alignment. 10.4.2.3 Design Loads (Reference Section 3.8.3.3) Refer to Table 3.8A-1 for additional loads that are applicable to the Crane Wall. 10.4.3 REFUELING CAVfTY J 10.4.3.1 Deswiption l l He Refueling Cavity is the reinforced concrete enclosure that provides a pool filled with borated water above the reactor vessel to facilitate the fuel handling operation without exceeding the acceptable level of radiation inside the Contamment Vessel. He Refueling Cavity has the following sub<ompartments.

• Storage Area for Upper Guide Structure
• Storage area for Core Support Barrel
• Refueling Canal Re Reactor Vessel flange is ,-- , sealed to the bottom of the Refueling Cavity to prevent Y 1 leakage of refueling water into the reactor cavity. He Fuel Transfer Tube connects the Refueling l Cavity to the Spent Fuel Pool. De shield walls that form the Refueling Cavity are a mmimum of six feet thick. l Amendment U 3.8A-36 December 31,1993 I

CESSAR Ennnenma l Railroad service is provided at the east end of the building with the track running through the inside of the building in the north-south direction. 11.5.3 ELEVATIONS rcknsnce 6hetko?ts eM]s s/ Turbine Building elevations are pr/MM E Ed: :.2. 9 Y 11.5.4 CODES AND STAhTARDS ne codes and standards applicable to Seismic Category II buildings shall be met. , 11.5.5 LOADS In addition to the mtmmum design loads requirements of Section 5.1 of this appendix, the following additional specific load requirements shall be met. Should conflicting values occur between this section and Section 5.1 of this appendix, the values specified in this section apply. 11.5.5.1 Dead Lead (D) ne mim*~i weights for major equipment are listed in Table 3.8A-7. 11.5.5.2 Ihe lead (L) ne live loads are specified in Table 3.8A-8. 11.5.5.3 Temperature loads (T,) The normal operanng temperature within the building ranges from 40*F to 100'F. He ambient temperature range outside of the building shall be -10*F to 100'F (Section 5.1.1.5 of this appendix). 11.5.5.4 Seismic Loads (E') ne seismic accelerations shall be as specified in the Table 3.8A-9. 11.5.5.5 Pipe loads Where the piping loads are not known at the time of design, beams and girders are designed for a concentrated load applied at midspan at indicated below.

1. In areas where the main steam and steam generater feedwater lines are located, use the weight ,

of the lines full of water.

2. In areas where large bore piping is heavily concentrated:

Girders (column to column) 55 kips Primary beams (column to column) 45 kips Secondary beams 30 kips Amendment T 3.8A-50 November 15, 1993

   ~-                               --                                                 ---        _

CESSAR 5=MA..,3. I TABLE 3JA-3 COMPONENT COOLING WATER HEAT EXCHANGER STRUCTURE SSE ACCELERATIONS IN Gs h  % Direction Short Direction Vertical

          ""'                  o.rar +a+-       ::: m,m o.a :.=                         l h Rm                      0.574      d.J72 -hM9-       g,4spea34             l  y    j B===                      0.5u       o, e   :. =       o gg+2a.              l      l i

i 1 Amendment U i December 31,1993 - l

CESSAR U1u?icavi:n

      .Du.eNll           rdinlMIy (Yf'nfhtenb5 h r C o n er e /c .5 d e k'o n 5 ga pnvidad kkthw 6.s s f +2.s me,,,tw 1.0     RBJFcavis AND SCOPE This Apper; dix presents analysis results and typical main reinforcing design for thirteen selected areas of the System 80+ Seismic Category I Nuclear Island structure, the Diesel Fuel Storage Stmeture, Component Cooling Water Heat Exchanger Structure, and Component Cooling Water Tunnel using the criteria in Appendix 3.8A. Based on the general arrangement of major structural elements and components, the thirteen Nuclear Island areas are selected to provide representative design details for structural elements having both typical and tmique design requirements. Design details for the steel contamment are included in Section 3.8.2.

In addition to the evaluation of the thirteen areas, shear requirements have been calculated and capacities demonstrated for all major shear walls of the Nuclear Island. The resulting design forces and moments presented in this Appendix are from use of a conservative envelope of design loads. Reinforcing details presented are typical details to develop the capacity q required to envelope these forces and moments.FThe design review demonstrates that it is feasible r to design and construct the structures as configured in the general arrangements presented in Chapter 1. The structural analysis repon prepared by the COL Applicant, Section 3.8.4.5.3, will document that the final design details for the Nuclear Island structure meet the analysis and design criteria of Section 3.8. Design and analysis details of the Diesel Fuel Storage Structure, Component Cooling Water Heat Exchanger Structure and Component Cooling Water Tunnel are provided in Section 7.0 of this appendtx.

2.0 DESCRIPTION

OF THE NUCLEAR ISLAND CRITICAL AREAS The location and description of the thirteen areas are identified in Table 3.8B-1. The areas are shown in Figure 3.8B-1, Sheets 1-5. 3.0 ANALYSIS METHODS The Nuclear Island is analyzed to account for both global and local effects of design basis loads described in Appendix 3.8A. The complete Nuclear Island is founded on a common basemat and is analyzed as a monolithic structure. A three dimensional finite element model of the Nuclear Island is developed and equivalent static global loading conditions are applied to the stmeture. These results are combined using the loading combinations identified in Section 5.2 of Appendix 3.8A. The global results from the three dimensional finite element model are combined with local analysis results to determine forces and moments for the design of the walls, columns and Cabs. The analysis methods are described in further detail in Appendix 3.8A, Section 6.1. 4.0 LOADS AND LOAD COMBINATIONS The loads evaluated for the Nuclear Island are addressed in Appendix 3.8A, Section 5.1. 3.8B-1 A endment U - 12/31/93

CESSARE! Enc =w l l i Area IB  ; l Shear (in-plane) 235 kips /ft l (out-of-plane) 24 kips /ft Moment 130 ft-kips /ft Axial (tension) 50 kips /ft (compression) 205 kips /ft The in-plane loads on area IC are predominantly shear loads from the SSE. The in-plane forces are obtained from output computed by the application of these loads to the static three dimensional fmite element model. The out-of-plane loads on the wall are predommantly from the accident temperature differential from a postulated Annulus Ventilation System failure. The out-of plane resultant forces and moments are determined by hand calculation.

  'Ibe design forces and moments for Area 1C are:

Shear (in-plane) 200 kips /ft (out-of-plane) 86 kips /ft Moment (2 way bending) 402 ft-kips /ft 118 ft-kips /ft Axial (tension) 140 kips /ft (compression) 250 kips /ft 5.1.5 TYPICAL REINFORCING DETAILS Area IA Wall Thickness 4 feet

  #18 at 12" vertical steel each face
  #18 at 12" horizontal steel each face Shear ties not required                                               ,                g.g.             Y Area IB        Wall 'Ibickness 4 feet                    gjg            -[              oy/ds2
  #14 at 12" vertical steel each face                      2 3 dd$crik e d 281 84#b
  #14 at 12" horizon:al steel each face                                                          .      -
                                                                              /4 Af[dA 'N*
                                                              'O #b Shear ties not required                                                                                Y Area 1C        Wall Thickness 4 feet
   #18 at 12" verdcal steel each face
   #14 at 12" horizontal steel each face Shear ties - #5 horizontal ties at 12" x 12" Y

3.8B-5 Ammammt U - 12/31/93

CESSAR E!Enemu  : 5.

1.6 CONCLUSION

The Area 1 concrete section strengths determined from the criteria in Appendix 3.8A are sufficient to resist the design basis loads. 5.2 AREA 2 - EAST END WALL ADJACENT TO TURBINE BUILDING 5.2.1 DESCRHTON OF AREA Area 2 is a segment of the exterior wall at the East end of the Nuclear Island adjacent to the Turbine Building. The wall extends from the top of the basemat at elevation 50'+0" to the top of the roof at elevation 146'+0". The walls in this area are four feet thick. Out-of-plane lateral support is provided to the walls by the floor slabs on the interior of the structure. 'Ihe wall is arranged and designed to function as a major structural shear wall in addition to providing protection for the safety related equipment. 5.2.2 GENERAL LOADS The loads applicable to Area 2 are summanzed in Appendix 3.8A, Table 3.8A-1. The out-of-plane passive soil pressure loads are the predommant loads. The Nuclear Island evaluation credits the passive soil pressure loads to resist sliding. 5.2.3 GOVERNING LOAD COMBINATIONS Area 2 Shear (in-plane) 4.1.2(a) (out-of-plane) 4.1.2(a) Bending 4.1.2(a) Axial (tension) 4.1.2(a) (compression) 4.1.2(a) 5.2.4 ANALYSIS METHODS AND RESULTS The Area 2 wall is analyzed as a stmetural shear wall. The in-plane forces are obtained from output computed by the application of these loads to the static three4imensional finite element model. The out-of-plane loads on the wall are predommantly soil pressure loads with the effect of the SSE. The out-of plane resultant forces and moments are determined by local two dimensional frame models. The design forces and moments for Area 2 are: 2 19 Shear (in-plane) 499 kips /ft (out-of-plane) 273 kips /ft I I 3.8B-6 W-t U - 12/31/93

CESSAR Heincomi Moment 910 ft-kips /ft Axial (tension) 50 kips /ft (compression) 277 kips /ft 5.2.5 TYPICAL REINFORCING DETAIIS Area 2 Wall Thickness 4 feet

 $$atts"pertit2/.sks/,2 /syars sack -fue Oe/ou destNEW 90 ,+' b
 #14 at 12" vertical steel,2 layers each face (Ahove s/sp,afp

• m fo#p.g ")

 #11 at 12" horizontal steel,2 layers each face Shear ties - #6 hori ntal ties at 12" x 2         Jhs#            .
 ))%fiNons/             $cti///f reinhor"cin.y)     S             tNine. SO A,P,P6    0 VN5bd)*rWided    f L' 5.

2.6 CONCLUSION

A**Y

• The Area 2 concrete section strengths determined from the criteria in Appendix 3.8A are sufficient to resist the design basis loads.

5.3 AREAS 3A AND 3B - EMERGENCY DIESEL ROOM INTERIOR AND EXTERIOR WALLS 5.

3.1 DESCRIPTION

OF AREA Diesel generator areas exist on the north and south side of the Nuclear Annex. The interior wall is Area 3A and exterior wall is Area 3B. Area 3A, the interior wall. extends from the top of the basemat at elevation 50'+0" to the top of the roof slab at elevation 91'+9". This four feet wall continues upward ending at the top of the roof slab at elevation 191'+0". all at Area 3A functions as an east-west structural shear wall. I f/' l Area 3B, the exterior wall. extends from the top to the basemat at elevation 50'+0" to the top of the I roof slab at elevation 9". This five feet exterior wall spans between the basemat and the roof [ l slab. This wall also functions as an East-West shear wall. l 5.3.2 GENERAL LOADS 'l l The loads applicable to Area 3 are summarized in Appendix 3.8A, Table 3.8A-1. The predominant I loads on the exterior wall are from the out-of-plane soil pressure loads. Passive soil pressure was considered in the design of the exterior walls. The Nuclear Island evaluation credits the passive soil pressure loads to resist sliding. Construction crane loads are also considered on the exterior wall. Lateral bracing of the exterior wall is considered during construction due to the vertical span of the wall. l 3.8B-7 Ammiment U - 12/31/93

CESSAR E!nincum The design forces and moments for Area 3B are: Shear (in-plane) 253 kips /ft (out-of-plane) 218 kips /ft Moment 1756 ft-kips /ft Axial (tension) 40 kips /ft (compression) 230 kips /ft 5.3.5 TYPICAL REINFORCING DETAIIS Area 3A Wall nickness 4 feet

 #14 at 12" venical steel each face
 #14 at 12" horizontal steel each face Shear ties not required y

Area 3B (fdd$onal duenh,G min *"iM Wall Thickness 5 feet SA g g g g a gge d fs d **p-L ggf//g 4,4 s/ -///.5 RgM.

 #18 at 12" vertical steel,3 layers each face
 #18 at 12" horizontal steel each face
                                                                                                  /

Shear ties - #5 horizontal ties at 4" x 12"/ Y 5.

3.6 CONCLUSION

The Area 3 concrete section strengths determined from the criteria in Appendix 3.8A are sufficient to resist the design basis loads. The exterior wall requires lateral shoring during construction to withstand the potential overburden pressure loads from construction crara:s. 5.4 AREAS 4. '.i AND 7 - CONTAINMENT PEDESTAL. DISH AND SUPPORT 5.4.1 DESCRIFIION OF AREA This area comprises the primary structural components supponing the Steel Containment Vessel (SCV) and its internal stmetures. The SCV is supponed by the pedestal and outer dish. The outer dish is supponed by the lower crane wall, the pedestal, the floor slab at elevation 91'+9", and the radial walls. This section addresses the design of these stmetural components, specifically described as follows: Area 7 Pedestal - Solid mass of concrete below the SCV, above the basemat, centered under the SCV, nommally 66 feet in diameter. 3.8B-9 A-det U - 12/31/93

CESSAR !!nincim. The design forces and moments for Area SC ne: Shear (in-plane) 400 kips /ft (out-of-plane) 45 kips /ft Moment 528 ft-kips /ft Axial (tension) 107 kips /ft (compression) 334 kips /ft 5.5.5 TYPICAL REINFORCING DETAILS Area SA Wall Thickness 4 feet (MSVH area)

 #18 at 12" vertical steel,2 layers each face
                                                                                /                     m
 #18 at 12" horizontal steel,2 layers each face
                                                            //dY/spf4//

o tut /<*///f At/MOr8Ag > Shear ties - #5 horizontal ties at 4" x 12" M ,y f . g g f pff,j jg Area SB Wall Thickness 4 feet

                                                          &cf'jf, 4,4 g/ 7%y agord
 #18 at 12" vertical steel cach face                                                                      '
 #18 at 12" horizontal steel each face Shear ties - #4 horizontal ties at 4" x 12"                                                            Y Area SC       Wall Thickness 6 feet (Inside Shield Building from top of basemat to bottom of slab at elevation 70'+0")
 #18 at 12" vertical steel,2 layers each face
 #18 at 12" horizontal steel,2 !ayers each face Shear ties - #4 horizontal ties at 4" x 12" 4                                                                                                   p 5.

5.6 CONCLUSION

The Area 5 concrete section strengths determined from the criteria in Appendix 3.8A are sufficient to resist the design basis loads. It is feasible to design and construct the structural components considered. The assumptions envelope the given parameters so that the design presented is adequate for any specific site conditions, within those parameters. The main steam line piping is assumed to be 32 inches in diameter, The main steam line anchor is assumed to have an 80 inch diameter bearing plate. A mimmum separation of 6' 6" from the centerline of the main steam line to any other discontinuity such as a wall, slab, opening or other possible failure plane should be maintained. Any separations less than 6'6" shall be analyzed and designed on a case by case basis. The 6'6" distance is the radius of the shear failure cone with the bearing plate assumed. 3.8B-23 Ammament U - 12/31/93

CESSAR Eininema 5.6.5 TYPICAL REINFORCING DETAILS Area 8 Wall Thickness 3 feet

    #18 at 12" vertical steel,2 layers each face
    #18 at 12" horizontal steel each face Shear ties - #4 horizontal ties at 4" x 12"      ,y AddM'asi duefiniy reinforeim)/fs//beprovidad as deserbd                                      n. f Serfim 4.0 s/ 7%'s godWN, .

5.

6.6 CONCLUSION

The Area 8 concrete section strengths determined from the criteria in Appendix 3.8A are sufficient to resist the design basis loads. It is feasible to design and construct the stmetural components considered. The assumptions envelope the given parameters so that the design presented is adequate for any specific site conditions, within those parameters. 5.7 AREA 9 - SPENT FUEL POOL WALL 5.7.1 DESCRIFFION OF AREA Area 9 is the wall between the spent fuel pool and the refueling canal, at Elevation 104'+0" to 146' +0", Column line 17-18 @ Column line T. The wall provides a barrier to isolate the spent fuel pool from the fuel transfer canal to allow maintenance on the fuel transfer system. A weir gate in the wall is removed to transfer fuel between the spent fuel pool and the refueling canal. 5.7.2 GENERAL LOADS The loads applicable to Area 9 are summarized in Appendix 3.8A, Table 3.8A-1. These loads include hydrodynamic and thermal loads from the spent fuel pool. 5.7.3 GOVERNING LOAD COMBINATIONS Area 9 Shear (in-plane) 4.1.2(d) (out-of-plane) 4.1.2(d) Bendmg 4.1.2(d) Axial (tension) 4.1.2(d) (compression) 4.1.2(d) 5.7.4 ANALYSIS METIIODS AND RESULTS The in-plane loads on Area 9 are predominantly shear loads from the SSE. The in-plane forces and moments are obtained from the global static three-dimensional finite element model results. 3.8B-25 A=M-t U - 12/31/93

CESSAR !!Encm,. Out-of-plane forces and moments are obtained by applying the out-of-plane loads to a local static three-dimensional finite element model of the wall. These forces are then considered in conjunction with the loads from the global finite element model results to determine design forces and moments , for the wall. Horizontal reinforcing is designed for the maximum out-of-plane bending about a vertical axis, due to local loads. Vertical reinforcing is designed for the maximum out-of-plane bending about a horizontal axis, due to local loads, combined with the maximum tension produced by global loads. Out-of-plane shear is determined by local analysis. The predominant forces are out-of-plane shear and bending forces from the hydrostatic and inertial forces associated with the water in the spent fuel pool, including sloshing effects. Also significant are the thermal effects from the heat generated by spent fuel. The design forces and moments for Area 9 are: Shear (in-plane) 282 kips /ft (out-of-plane) 179 kips /ft Moment (2 way bending) 2704 ft-kips /ft 4807 ft-kips /ft Axial (tension) 80 kips /ft 5.7.5 TYPICAL REINFORCING DETAILS , Area around weir gate notch in wall controls. Area 9 Wall Thickness 6 feet

  #18 at 8" vertical steel,2 layers each face
  #18 at 6" horizontal steel,3 layers each face Shear ties - #5 hon'rpntal ties at 18.5" x 6"       rja//

Add /Nnts/ MMef////_y fghtfpfgEL } fgpfeyhtg g$ fjCflb6 f 5,

7.6 CONCLUSION

S Seefan af,6 of dir AffON* The Area 9 concrete section strengths determined from the criteria in Appendix 3.8A are sufficient 3 to resist the design basis loads. It is feasible to design and construct the structural components considered. The assumptions envelope the given parameters so that the design presented is adequate for any specific site conditions, within those parameters. i l l 3.8B-26 Ammament U - 12/31/93

I CESSAR !!nincamn 5.8.5 TYPICAL REINFORCING DETAILS Area 10 Wall Thickness 5 feet

 #18 at 12" vertical steel,2 layers each face
 #18 at 12" horizontal steel, 3 layers each face Shear ties - 2 #5 horizontal ties at 4" x 12" Jllf*//

corwided As de,$ct'ib4d *. Y AddiNa>ts/ 5.

8.6 CONCLUSION

dxci/'///y IMinOw'Q be)Seefjen, d,e e{ dis yendW ' The Area 10 concrete section strengths determined from the criteria in Appendix 3.8A are sufficient to resist the design basis loads. It is feasible to design and construct the struennal components considered. The assumptions envelope the given parameters so that the design presented is adequate for any specific site conditions, within those parameters. To accommodate punching shear requirements, the anchor embedment design must incorporate excess punching shear, or alternately rupture loads may be reduced by more detailed analysis. The main steam line piping is assumed to be 32 inches in diameter. The main steam line anchor is assumed to have an 80 inch diameter bearmg plate. A mmimum separation of 7'6" from the centerline of the main steam line to any other discontinuity such as a wall, slab, opening or other possible failure plane should be maintained. The main feedwater line located in this area is assumed to be 24 inches in diameter and contains a 72 inch diameter beari' n plate. The minimum separation of the main feedwater centerline to the edge of any other disce tuity is 7'0". Any separation less than 7'6" for the main steam lines and 7'0" for the main feedwater line is analyzed and designed on a case by. case basis. These distances are the radii of the shear failure cones of the main steam line and main feedwater line with the bearing rings assumed. l 5.9 AREA 11 - NORTH-WEST END WALL 5.9.1 DESCRIPITON OF AREA Area 11 is the northern part of the west side end wall adjacent to the Radwaste Building. The wall extends from the top of the basemat at elevation 50' +0", to the top of the roof in the Fuel Handling area at elevation 191'+0*. The walls in this area are four feet thick. Out-of-plane lateral support is provided to the walls by the floor slabs on the interior of the structure. The wall is arranged and designed to function as a major stmetural shear wall in addition to providing protection for the safety related equipment. 5.9.2 GENERAL LOADS The loads applicable to Area 11 are summarized in Appendix 3.8A, Table 3.8A-1. The out-of-plane passive soil pressure loads are the predominant loads in the lower elevations of the wall. The Nuclear Island evaluation credits the passive soil pressure loads to resist sliding. Local loads , resulting from the wall mounted supports for the spent fuel pool bridge crane are also included. l 1 I 3.8B-28 W=nt U - 12/31/93 l l i

z CESSAR 6"encamn 5.9.3 GOVERNING LOAD COMBINATIONS Area 11 Shear (in-plane) 4.1.2(a) (out-of-plane) 4.1.2(a) Bending 4.1.2(a) Axial (tension) 4.1.2(a) (compression) 4.1.2(a) 5.9.4 ANALYSIS METHODS AND RESULTS The Area 11 wall is analyzed as a structural shear wall. The in-plane forces are obtained from output computed by the application of these loads to the static three-dimensional finite element model. The out-of-plane loads on the wall are predommantly soil pressure loads with the effect of the SSE in the lower elevations. The effects of the spent fuel pool bridge crane loads in combination with thermal loads are predominant in the upper portion of the wall. 'Ihe out-of plane resultant forces and moments are determined by local two dimensional frame models. The wall is analyzed and designed to resist the spent fuel pool bridge crane bending and axial loads. The bending effects dissipate below elevation 91'+9". The design forces and moments, excluding the spent fuel pool bridge crane loads for Area 11 are: Shear (in-plane) 274 kips /ft (out-of-plane) 400 kips /ft

                                 /$3 Moment                            938 ft-kips /ft Axial (tension)                   240 kips /ft (compression)             512 kips /ft 5.9.5          TYPICAL REINFORCING DETAILS l

Area 11 Main Steel > Wall Thickness 4 feet A e

 #14 at 12" vertical steel               each face    {nbwe dlts/2bbK W+3N
 #18 at 12" vertical steel        layer,feach laye               face (.de/eM e/43/af'on .$d eJ")
 #11 at 12" horizontal steeloesh4ees          9 f./ ers exe4 fue Shear ties - #5 horizontal ties at 12" x 12"        g/

Ridifpu/ k<Mity reinbre9A be ymidesL a hearthe.l in Saetion. 4.0 d -fAi.s pedy. Y j 3.8B-29 Ameut-r U - 12/31/93

CESSARinne.m l 5.11.4 ANALYSIS METIIODS AND RESULTS The Area 13 basemat responds in bending and shear loads from the SSE and dead load. The forces and moments are obtained from application of these loads to the static three-dimensional f' mite element model. The moment evaluated is the maximum moment experienced by the basemat. The l shear evaluated is from a representative area under a primary shear wall. Most of the basemat will not require any shear reinforcing. The design forces and moments for Area 13 are: Moment: 3545 ft-kips /ft Shear: 68.3 kips /ft ne basemat is symmetrically reinforced to resist the potential moments as a result of differential settlement of the foundation. The capacity of the basemat to withstand differential settlement is determined by calculating the deflection at the edge of the mat that would occur if the maximum moment were developed in the center. The maximum deflection in the basemat relative to the center of the Nuclear Island at the four exterior walls is: Wall DJdla North 20 in South 20.in East 25 in West 49 in 5.11.5 TYPICAL REINFORCING DETAILS Basemat Thickness 10 feet l

        #18 at 12" horizontal steel,2 layers each face each direction Shear ties - (When required)

Verh2Al

        #10 hesimental ties at 12" x 12" l j Most of the basemat will not r        ire any shear reinforcing.
                                                                                                  ',M-Mdifirmr/ d'edi/Hy rdin ni                           po ovidad 2 s ducrd.

• A y 5.

11.6 CONCLUSION

S $py Sec/Nn 4.0 of fAI.* Afjendi$. The Area 13 concrete basemat strength determined from the criteria in Appendix 3.8A is sufficient I to resist the design basis loads. It is feasible to design and construct the nuclear island foundation basemat. The design envelopes the given parameters so that the design presented is adequate for any , specific site conditions, within those parameters. Stress concentrations exist in the areas around I sumps that require additional detailed analyses and design.  ! 3.8B-32 Ammament U - 12/31/93

CESSAR !!n%ma Center Wall: The primary flexural reinforcing for this two-foot thick wall consists of a rectangular grid of #11 at 6 inches each way/each face, [i.e., 3.12 in2 /ft). No transverse shear reinforcing is required. Encf: ) The primary flexural reinforcing for these two-foot thick walls consists of a rectangular grid of #11  ; at 6 inches each way/each face [i.e., 3.12 in2 /ft]. , No transverse shear reinforcing is required. 7.

1.5 CONCLUSION

l The concrete and reinforcing steel section strengths of the Diesel Fuel Storage Structure are sufficient , to resist the design basis load and load combination criteria specified in Sections 3.8A.ll.1 and

3. 5 .

7~ c r n for.c; b,fg;), pg.g/ m in f y ,3gg f y 7.2 COMPONENT COOLING WATER HEAI_ EXCHANGER STRUCTURE , 7.

2.1 DESCRIPTION

OF STRUCTURE The Component Cooling Water Heat Exchanger Structure is a single bay, partially embedded, two story reinforced concrete building. The top floor houses two heat exchangers supported on saddles which spread the loadings to the supporting floor and column system. l The specified concrete compression strength is 4,000 psi and the specified minimtun yield strength of the reinforcing steel is 60,000 psi. 7.2.2 ANALYSIS METHODS The Component Cooling Water Heat Exchanger Structure is analyzed for the design loads described in Appendix 3.8A to determine the global and localized member forces for which the structure must be designed. The stmeture is analyaed using manual computations which consider the stmeture to be comprised of linear elastic one-way wall and slab panels. Thermal and equivalent static loads corresponding to the various individualloading conditions identified in Sections 3.8A.5.1 and 3.8A.11.2.5 are applied to the one-way panel models and resulting member forces and moments computed. The resulting i member forces are combined in accordance with the load combinations, specified in Section 5.2.2 of Appendix 3.8A, to determine the design loads for the critical sections. l 3.8B 35 Amendment U - 12/31/93

CESSAR!annema 1 No transverse shear reinforcing is required. Floor Slab at Elevation 90'-9"- I l The primary reinforcing for the three-foot floor slab consists of a rectangular grid of #10 at 10 inches , each way/each face, (i.e.,1.52 id/ft). I No transverse shear reinforcing is required.  : Roof Slab at Elevation 110'-9': . The prunary reinforcing for these two-foot thick roof consists of a rectangular grid of #11 at 10 inches each way/each face, (i.e.,1.87 id/ft). No transverse shear reinforcing is required. J 7.

2.5 CONCLUSION

The concrete and reinforcing steel section strengths of the Component Cooling Water Heat Exchanger Strucmre are sufficiem to resist the design basis load and load combination criteria specified in Sections 7F re/n fac/ M de/A//r Kre SAeA /* y Ryre.s 3.8A.11.23.9B- 7 and dr~3.8A.S.O. J.BB-9. bpkJ/ - 7.3 COMPONENT COOLING WATER TUNNEL 7.

3.1 DESCRIPTION

OF STRUCTURE The Component Cooling Water Tunnel is a single compartment, fully embedded, one-story reinforced concrete structure. The tunnel houses and protects the Component Cooling Water piping which is routed from the corresponding Nuclear Island pipe chase to the basement of the Component Cooiing Water Heat Exchanger Strucmre. The tunnel is artmehed at one end to the Nuclear Island Pipe Chase and the Cornponem Cooling Water Heat Exchanger Strucmre at the other end via flexible connections. The flexible connections allow differennal movement between the three structures without transfernng in=Aings The specified concrete compression strength is 4,000 psi and the specified mininnrm yield strength of the reinforcing steel is 60,000 psi. 7.3.2 ANALYSIS METHODS The Component Cooling Water Tunnel is analyzed for the design loads described in Appendix 3.8A to determme the global and localized member fortes for which the structure taust be designed.

      . The structure is analyzed using manual wpiations which consider the structure to be comprised of linear elastic one-way wall and slab panels. The lateral loads on the tunnel were evaluated using a linear elastic frame model with a unit width. Thermal and equivalent static loads corresponding              ,

to the various individualloading conditions idennfied in Sections 3.8A.5.1 and 3.8A.1 5 are applied / < to the equivalent frame model and resulting mmber forces and moments co uted. The resulting

                                                                                       ,7 3.8B-37                        A**'""* U - 12/31/93

                                                                     =

CESSAR !!nLwo North and South Walls: The primary reinforcing for these two-foot thick walls consists of a rectangular grid of #11 at 10 inches each way/each face, [i.e.,1.87 in2 /ft). No transverse shear reinforcing is required. Enaf: The primary reinforcing for these two-foot thick roof slabs consist of a rectangular grid of #11 at 10 inches each way/each face, [i.e.,1.87 in3 /ft], No transverse shear reinforcing is required. 7.

3.5 CONCLUSION

The concrete and reinforcing steel section strengths of the Component Cooling Water Tunnel are sufficient to resist the design basis load and load combination criteria specified in Sections 3.8A.11.7

      ""d' *^ 5*-

7y~ tax / rein % ; g detsj/s are .sku. ;,, ,' Ejwes J.8 8- /0 2nd 4.98-/), 3.8B-39 Amendment U - 12/31/93

                                                  =                                         ,

l CESSAR EEncue,. TABLE 3.8B-1 l AREAS IDENTIFIED FOR DETAILED DESIGN ) l Area Description Section Elevation Col. Line/ Azimuth 1 Shear & Shield Building Wall 1A 50 to'M D-F @ 17 IB 50 to M E17 1C 50 to FIN 16-18, E-F 2 East Wall @ Turb Building 2 Nto+3 O B14 3 Diesel Gen. Room Ext, & Int. 3A Mto 93 N23 3B 8to 93 N25 l 4 Subsphere Radial Wall 4 8to5[## 225*, R33-R65 5 Shear Wall and Slab @ SA kto M8 K12-EE ' Emerg. FDW Pump Room o SB to 130+6 K11 and CCW Pump Room SC M tom # K10-K13 6 SCV Anchorage Region 6 70 to## # #74.I 7 SCV Support Pedestal 7 50 to 62 d&R33 8 S/G Wing Wall @ IRWST 8 70 to 91+9 Agu% L15 9 Spent Fuel Refueling Canal 9 -93 to.149 T17-18 Wall W #6 10 - Main Steam Valve House 10 106 to 130 H23-25 Wall i 11 Nuclear Annex Wall @ 11 50 to 94+G- U19-20 Radwaste Building /$/  ; I 12 Interior Structure Steel 12 91+9 to M9 N/A Columns //A i 13 Basemat 13 40 to 50 N/A i Amendment U - 12/31/93

CESSAR isEZeu,. 13

                                                                            ).

influence oefficients are calculated for each dynamic ) degree-of- reedom of each mass point and for each degree-of-freedom o each support point. The ANSYS computer code (Section . 3.9.1.2.1. ) is also used as an alternate to MDC-STRUDL for Y l defining the dynamic characteristics of the reactor coolant I system and seismically analyzing it. j 1 The program can perform either time-history analysis or spectrum ) analysis using the modal super position technique. Support i reactions, member loads and joint acceleration are computed by back substituting from the modal coordinates to physical coordinates through the applicable transformation matrices and then combining modal contributions from each individual mode included in the response analysis. , 1 MDC STRUDL is a program which is commercially available and has had sufficient use to justify its applicability and validity. Extensive verification of the C-E version has been performed to supplement the public documentation. The version of the program in use at C-E was developed by the McDonnell Automation Company / Engineering Computer International and is run on the IBM computer system. MDC STRUDL is described in more detail in Reference 1. 3.9.1.2.1.2 C-E MARC The C-E MARC program is a general purpose nonlinear finite element program with structural and heat transfer capabilities. It is described in detail in Reference 2. C-E MARC is used for stress analysis of regions of vessels, piping or supports which may deform plastically under prescribed loadings. It is also used for elastic analyses of complex geometries where the graphics capability enables a well defined solution. The thermal capabilities of C-E MARC are used for complex geometries where simplification of input and graphical output are preferred. C-E MARC is the C-E modified version of the MARC program, which is commercially available and has had sufficient use to justify l its applicability and validity. Extensive verification of the C-E versi6n has bee.n performed to supplement the public documentation. 3.9.1.2.1.3 PICEP The PICEP program calculates the flow through a crack in a pipe. PICEP uses the simplified engineering approach for elastic-plastic fracture analysis for finding the crack opening displacement and area. Fluid calculation options include single and two-phase flow as well as allowance for friction. PICEP was developed by EPRI. Amendment S 3.9-5 September 30, 1993

CESSAR H!nnem. program was verified by comparisons of program results and hand-calculated solutions of classical problems. 3.9.1.2.1.10 CE105, Nossle Fatigue Program This program computes the redundant reactions forces, moments, and fatigue usage factors for nozzles in cylindrical shells. This program is used to perform the fatigue analysis of reactor

] vessel nozzles and steam generator feedwater nozzle. The program was verified by       comparisons of program        results   and  hand-calculated solutions of classical problems.

3.9.1.2.1.11 CEC 26, Edge Coefficients Program This code calculates the coefficients for edge deformations of conical cylinders and tapered cylinders when subjected to axisymmetric unit shears and moments applied at the edges. This program is used to perform the fatigue analysis _of reactor l vessel wall transition. The program was verified by comparisons of program results and hand-calculated solutions of classical problems. . 3.9.1.2.1.12 CE124, Generalised 4 x 4 Program This program computes the redundant reactions, forces, moments, stresses, and fatigue usage factors for the reactor vessel vall , at the transition from a thick to thinner section and at the ' l bottom head juncture. This program is used to perform fatigue analysis of . reactor l vessel . bottom head juncture. The program was verified by comparisons of program results and hand-calculated solutions of classical problems. 0 0.1 1.1.-- ' e;; 11 i he SEC pro am a oma a th flaw eval tion eth of SME B& , S tion I, pend A. This prog a pe orms the -ra ,, g owth analy es a as sses the rgin gain cr ica cr ck Y , ize ccord ng t the rit la i Appe ix A Th pro a a been erif ed b dir tc pari n of rogr m re its an h d cal lati ns. The rogr m is sed r 1 k-b ore- re t pe

r. lycer i

               /3                                                                I 3.9.1.2.1.),#         ANSYS y   J ANSYS is a large-scale, general-purpose, finite element program for linear and nonlinear structural and thermal analysis. This program     is commercially available.          Additional descriptive information on this code is provided in Section 3.9.1.2.2.2.

This program is used for numerous applications for all components Amendment S 3.9-8 September 30,1993

C E S S A R E!Minca m I in the areas of structural, fatigue, thermal and eigenvalue analysis. The program was verified by comparisons of program l results and hand-calculated solutions of classical problems. 3.9.1.2.1. CE301, The Structural Analysis for Partial Penetration Nossles, Heater Tube Plug Welds, and the Water Level Boundary of the Pressuriser Shell Program' This program computes various analytical parameters, primary plus secondary stresses and stress intensities, peak stresses and stress intensities, and the cyclic fatigue analysis with usage factors at cuts of interest. This program is utilized to satisfy the requirements of Section III, of the ASME B&PV Code. This program is used in the fatigue analysis of partial penetration nozzles in the pressurizer and piping. The program j was verified by comparisons of program results and hand-calculated solutions of classical problems. CE223, Primary Structure Interaction Program Y 3.9.1.2.1.)4 This code calculates redundant loads, stresses, and fatigue usage factors in the primary head, tubesheet, secondary shell, and stay cylinder for pressure and thermal loadings. This program is used in the fatigue analysis of the steam generator primary structure. The program was verified by l comparisons of program results and hand-calculated solutions of classical problems. 3.9.1.2.1. CE362, Tube-To-Tubasheet Wald Program This code performs a three body interaction analysis of the tube-to-tubesheet veld juncture. The code calculates primary, secondary, and peak stresses and computes range of stress and- l fatigue usage factors. l This program is used in the fatigue analysis of steam generator i tube-to-tubesheet weld. The program was verified by comparisons l ) of program results and hand-calculated solutions of classical ] problems. 17 3.9.1.2.1./ CE286, Support Skirt Loading Program / This code calculates the stresses in the conical support skirt of the steam generator for external loads. This program is used in the structural analysis of steam generator support skirt. The program was verified by comparisons l of program results and hand-calculated solutions of classical problems. i Amandment S 3.9-9 September 30,1993 l

CESSARn h m.

                 /s                                                           y CE210, Principal Stress Program 3.9.1.2.1.)4 This code sums stresses for three load. conditions and computes principal stress intensity, stress intensity range, and fatigue usage factor.

This program is used in the fatigue analysis,of steam generator l components. The program was verified by comparisons of program results and hand-calculated solutions of classical problems.

                  /$        CE211, Nossle Load Resolution Program             Y 3.9.1.2.1./

This is a special purpose code, us'ad to calculate stresses in nozzles produced by piping loads in combination with internal pressure. This program is used in the fatigue analysis of steam generator l nozzles. The program was verified by comparisons of program results and hand-calculated solutions of classical problems. 20 y 3.9.1.2.1.Af KINI2100 Program This is a general purpose finite difference heat transfer program. This program is used for steady-state and transient thermal analysis. This program is used in numerous thermal relaxation analyses for l all components. The program was verified by comparisons of program results and hand-calculated solutions of classical problems'. El Y 3.9.1.2.1 M CEFLASH-4A l This is a code used to calculate transient conditions resulting , from a flow line rupture in a water / steam flow system. The i program is used to calculate steam generator internal loadings following a postulated main steam line break. This program is used in a steam line break accident structural l analysis. The program was verified by comparisons 'of program-results and hand-calculated solutions of classical problems. 3 3.9.1.2.1. CRIBE This is a one-dimensional, two-phase thermal hydraulic code, utilizing a momentum integral model of the secondary flow. This code was used to establish the recirculation ratio and fluid mass - The code- is inventories as a function of power level. commercially available and has had suf ficient use to justify its applicability and validity. This program is used for determining by l steam generator performance. The program was verified comparisons.of program results and hand-calculated solutions of c.lassical problems. Amendment S 3.9-10 September 30,1993

CESSARinMeuiu i AS 3.9.1.2.~1.)4 FAST 2 [ FAST 2 is a computer code originated by She11 tech Associates for the analysis of vessel-nozzle intersections. It uses closed form asymptotic results for the. solutions of the thin shell equations. FAST 2 calculates stress and deflections of a cylindrical vessel or spherical head with a cylindrical pipe intersecting the vessel wall. Vessel geometries are idealized as a horizontal cylinder on two saddle supports, a horizontal cantilevered cylinder fixed on the left end, or a horizontal cantilevered spherical head. Spherical heads are simply supported at their base such that all points at l the base remain in a vertical plane. Radial expansion and local rotation is a function o,f the head stiffness and the stiffness of an attached cylindrical vessel which may be included in the model at the user's option. The loading conditions available in FAST 2 are nozzle loads, vessel end loads, internal pressure, and thermal loads. Nozzle loads are applied at the nozzle / vessel intersection. Shear loads are not considered. Vessel end loads are external landings applied at the right end of the vessel. Internal pressure can be applied to any combination of vessel and pipe. Thermal loads are uniform thermal expansion parameters for each portion of the defined model. The code has the capability of modeling stiffening rings at either or both ends of a cylindrical vessel, and at the vessel / head junction for a spherical head. End caps or vessel heads on a cylindrical vessel may be modeled by stiffening rings representing the equivalent stiffness of the head or cap. i FAST 2 has been used by Shelltech Associates in the development of WRC Bulletin No. 297.  ; 3.9.1.2.1. PC-PREPS y The evaluation and design of pipe support frames and baseplate is l performed using PC-PREPS. PC-PREPS is a personal computer based, integrated pipe support analysis software package. It is interactive, menu-driven, with built-in structural analysis and graphics capability. The package is totally self-contained, except for a word processor used - for the final calculation document production. All operations, including the finite element analysis, are performed on the personal computer. PC-PREPS allows a pipe support analyst to prepara data, view j associated graphics, and execute frame and baseplate analyses.  ; It can automatically perform load combinations and convert loads  ! computed with pipe stress sof tware to the pipe support frame, and i from the frame to any of the defined baseplate. The post- i processing capabilities of PREPS include AISC and NF Code checks, Amendment S' 3.9-lla September 30,1993

CESSAREm%ma maximum displacement checks, weld stress check, and local stress check. PC-PREPS has been qualified by comparison to other software performing similar calculations and to manual calculations. 24' 3 . b .1. 2 .1.g LIDOP y The LIDOP program computes the local crush characteristics of a pipe section for use in the analysis of pipe motion and subsequent impact on structural targets or pipe rupture restraint structures. , The program will generate crush rigidities and deformation energies for pressurized or unpressurized piping in the following geometries: A. Ring crush against flat rigid surface. B. Indent or straight pipe against rigid cylinder. C. 1.5D pipe elbow (extrados) against a flat rigid surface. D. Pipe bend (extrados) against a flat rigid surface. E. Indent of straight pipe against a rectangular block. , Both dynamic effects and material properties are considered in generation of the crush characteristics. Unpressurized force-displacement and energy-displacement characteristics of pipe and elbows are generated from empirical equations which are based on experimental data. Pressurization effects, based on fluid displacement during deformation, are superimposed on the unpressurized characteristics. The overall dimensions of the contact area, where applicable, are generated by empirically corrected geometric relationships. Dynamic effects of elbows are empirically determined from an experimental comparison of static and dynamic impact of spheres.. Dynamic effects of all other geometries and elbows in certain cases are based on the results of finite element computer simulations of rings impacting flat, rigid surfaces. The effects of material properties are determined from empirical relationships based'on computer predictions. 24 a TIMHIS6 Y 3.9.1.2.1.y The TIMHIS6 program performs modal superposition time history analysis for lumped mass / stick models and response spectra calculations. Amendment R 3.9-llb July 30, 1993

CESSARin h o 27 RELAP5 y 3.9.1.2.1.g RELAPS is used to perform transient analysis of thermal-hydraulic systems with water as the fluid. RELAPS uses a five equation ' two-phase flow continuity equations, two phasic momentum equations and an overall energy equation augmented by the requirement that one of the phases is assumed saturated. In this model, only two interphase constitutive relations are required, those for interphase drag and interphase mass exchange. Models are included for abrupt area changes, choking, mass transfer interphase drag, wall friction and branching. , The program requires numerical input data that completely describes the initial fluid conditions and geometry of the system being analyzed. The output consists of variables necessary to describe the transient state of the system being analyzed. 28 3.9.1.2.1.Jf REPIPE [ REPIPE computes the loading time histories on a piping network based upon the results from computer program RELAP5 hydrodynamic analysis of the contained fluid. The RELAP5 time-varying pressure, momentum flux and energy states throughout a fluid system containing water, steam, and/or a two phase mixture are used as in input to the REPIPE program to produce time histories , for input to the piping stress analysis program. REPIPE distributes the RELAPS control volume forces to the structural network nodes by a process based upon fluid momentum balance principle and newtons third law of motion. The output from REPIPE consists of dynamic loads on the pipe, organized into force- vs time tables. 29 sf , 3.9.1.2.1.pd CCN-318 r , CCN-318 is a computer program used to evaluate the design of rectangular cross section attachments on ASME Class 2 and 3  ! Piping following the requirements of ASME Code Case N-318. The l program checks for Code case limitations, calculates the required coefficients and then checks local stress in the pipe wall. In addition, it also evaluates the adequacy of fillet and partial penetration welds. .The results of the analysis are compared to ASME Code Allowables. JO 3.9.1.2.1.Jf CCN-392 [ CCN-392 is a computer program used to evaluate the design of circular cross section attachments on ASME Class 2 and 3 piping following the requirements of ASME Code Case N-392. The program I checks for code case limitations, calculates the required l coefficients and then checks the local stress in the pipe wall. In addition, it also evaluates the adequacy of finet and partial f Amendment R . 3.9-lle July 30, 1993

CESSAR nninc-penetration welds. The results of the analysis are compared to ASME Code Allowables. 3 . 9 .1. 2 .1. TRANS2A TRANS2A is a computer program which determines radial temperature distributions and gradients in a pipe wall experiencing fluid temperature excursions. TRANS2A determines these temperature distributions by solution of the unsteady one-dimensional axisymmetric heat transfer equation. For aid in Class 1 piping analysis values of the thermal gradients AT and AT: and the 3 average temperatures (K and/or K) are calculated (and printed) in accordance with ASME BPVC Section III Article NB-3650. To be l of more aid to the analyst in choosing values of the average and temperature gradient data to be input to the combined stress analysis, TRANS2A evaluates the actual histories of the thermal stress terms according to the equations of Section III, Article NB-3650 with as many as ten sets of stress indices and summarizes them in a table by extreme and time of occurrence. Amendment S 3.9-11d September 30,1993

CESSAREEac-The analyses performed for branch line breaks use the MDC STRUDL l (Section 3.9.1.2.1.1) or ANSYS (Section 3. 9.1. 2.1. 13 M) code. V The resultant component and support reactions are specified, in combination with tne appropriate normal operating and seismic reactions, for design verification by the methods discussed below and in Section 3.9.3. l The system or subsystem analysis used to establish, or confirm, 1 loads which are specified for the design of components and supports is performed on an elastic basis. When an elastic system analysis is employed to establish the loads which act on components and supports, elastic stress analysis methods are also used in the design calculations to evaluate the effects of the loads on the components and supports. In particular, inelastic methods such as plastic instability and limit analysis methods, as defined in Section III of the ASME Code, are not used in conjunction with an elastic system analysis. The RCS and its supports, which are analyzed using elastic methods, are shown in diagram form in Figure 3.9-1. Inelastic methods of analysis are used in cases where it is deemed desirable and appropriate to permit significant local inelastic response. In these cases, if any, the system or subsystem analysis performed to establish the loads which act on components and component supports are modified to include the inelastic strain compatibility in the local regions of the components and component supports at which significant local inelastic response is permitted. Inelastic methods defined in Section III of the ASME Code as plastic instability or limit analysis methods are not used. 3.9.1.4.1.1 Reactor Internals and CEDMs See Sections 3.7.3.14 and 3.9.2.5. 3.9.1.4.1.2 Non-Code Items The components not covered by the ASME Code but which are related to plant safety include: A. Reactor Internal Structures (Class IS). B. Puol. C. Control element drive mechanisms (CEDMs). D. Control element assemblies (CEAs). Amendment S l 3.9-20 September 30,1993

CESSAR8Ence - 3.9.2.5 Dynamic System Analysis of the Reactor and CEDMs Under Faulted Conditions Dynamic analyses are performed to determine blowdown loads and structural responses of the reactor core support, internals structures and fuel to postulated pipe break and SSE loadings and to verify the adequacy of their design. Because of Leak-Before-Break arguments, all main RCS loop pipe breaks and all major primary branch line pipe breaks have been eliminated from consideration of dynamic effects. Internal blowdown loads due to breaks in small primary side pipes (6 inch diameter and less are considered in the design of the reactor internals. The loads due to these small pipe breaks are combined with the SSE loads by the SRSS method, and are found to represent less than a 10% increase in the SSE loads. Stress intensities for f aulted conditions are governed by reactor vessel response motions from SSE and major secondary side branch line pipe breaks. Dynamic analyses are performed to determine the structural response of the Class CS and internal structures to assure that the criteria of Table 3.9-14 is achieved for the appropriate combination of pipe break and SSE loads. 3.9.3 ASME CODE CLASS 1, 2 AND 3 COMPONENTS, COMPONENT SUPPORTS AND CLASS CS CORE SUPPORT STRUCTURES ASME B&PV Code Section III Class 1, 2 and 3 Piping and Components are designed and constructed in accordance with Section III of the ASME Boiler and Pressure Vessel Code and Code Case (s). In accordance with ASME Code, a specification is provided for , piping supports which defines the jurisdictional boundary for the l NF portion of the piping support. i l For equipment component supports, such as those for pumps and i I vessels, the supports are generally furnished by the manufacturer along with the equipment. The supports are designed and l classified and meet ASME Code Section III, Subsection NF. I 4 1 Welding activities shall be performed in accordance with the l requirements of Section III of the ASME Code. Component supports I shall be fabricated in accordance with the requirements of Rubendien NF of Section III of the ASME Code.___. i...,--

 " Visual weld acceptance criteria shall be per the Nucle M onstrudinn Issue Group NCIG) standard NCIG-01 (Reference 51) .          gj Welding activitiWror nou(g Grade B tube steel shall be performed        V in accordance with the requirements of AWS D1.1,       " Structural Welding Code," (Reference 52).                                              j Amendaent U 3.9-32               December 31, 1993

t CESSAR Einineau 3.9.3.1.4 Piping and Piping Supports a Piping systems classified as ASME Code Section III Class 1, 2 or 3 are designed to maintain dimensional stability and functional integrity under design loadings expected to be experienced during , a 60-year design life. The COL applicant will reconcile the as-built piping with the as-designed piping configurations.

        .9.3.1.4.1          ASME Code Class 1                                                            i A. Piping Ig(bi        For ASME Code Class 1 piping, the combinations of design                                       ,

3g,3)*q loadings are categorized with respect to service levels, as identified as Level A, Level B, Level C, or Level D, shown in Table 3.9-10. The design stress limits for each of the loading combinations are found in ASME B&PV Code, Section III, NB-3600. B. Piping Supports For pipe supports, the design loading combinations are presented in Tables 3.8-5 and 3.9-12. Pipe support members are designed to meet the requirements defined by ASME Code, Section III, Subsection NF. See Appendix 3.9A, Section 1.7.4, for a further discussion. 3.9.3.1.4.2 ASME Code Class 2 and 3 A. Piping l For ASME Code Class 2 and 3 piping the combinations of design and service loadings are categorized with respect to system service levels identified as Design, Level A, B, C l and D as shown in Tables 3.9-11. The design stress limits l for each of the loading combinations are found in ASME B&PV Code, Section III, NC/ND-3600. B. Piping Supports For pipe supports, the design and service loading

             ~ombinations are presented in Tables 3.9-12. Pipe support members are designed to meet the requirements defined by ASME Code, Section III, Subsection NF.       See Appendix 3.9A, Section 1.7.4, for a further discussion.

I Amendment U  ! Dec d er 31, 1993 3.9-38

INSERT TO 3.9.3.1.4 M'

     !      The COL applicant will perform an as-built inspection of the pipe routing, location and orientation, the location, size, clearances 4      and orientation of piping supports, and the location and weight of pipe mounted equipment. The inspection will be perfomed by      /

reviewing the as-built drawings containing verification stamps, and by performing a visual inspectoin of the installed piping system. I I The piping configuration and component location, size, and

   /        orientation shall be within the tolerances specified in the            I certified as-built piping stress report. The tolerances to be used      I
  /                                                                                    f for reconciliation of the as-built piping system with the as-             y
 /

designed piping system are provided in Reference misse. A I I reconciliation analysis using the as-built and as-designed !' information shall be performed. The certified as- uilt stress i report shall document the results of the as-built r conciliation j analysis. ss. s w ===

         'CESSAR innncam fl Coyotenf su.ppod bu.lld'>tg .sfradu.res    see de.rc.yn e d ,g gee f-1 % dri,4r u in RyenA y .3.M .

3.9.3.4 o gonDonent E pporta Jurisdictional boundaries between ASME Section III Class 1, 2 and 3 component supports and the building structure are established in accordance with ASME Section III, Subsection NF. ASME B&PV Code Section III C.' ass 1, 2 and 3 component supports are designed and constructed in accordance with Section III of the ASME B&PV Code and Code Case (s). I A i ld Supports for ASME Section III Code Class 1, 2 and 3 components h are specified for design in accordance with the loads and loading combinstions discussed in Section 3.9.3.1 and presented in Table 3.9-2. component supports which are loaded during normal operation, seismic and following a pipe break (branch line breaks not eliminated by leak-before-break) are specified for design for loading combinations (A) through (D) of Section 3.9.3.1. Design stress limits applied in evaluating loading combinations (A),  ; (B), and (C) of Section 3.9.3.1 are consistent with the ASME Code, Section III.. The. design stress limits applied- in j evaluating loading combination (D) of Section 3.9.3.1 are in  ; accordance with the ASME B&PV Code, Section III. Loads in i compression members are limited to 2/3 of the critical buckling  ! load. j l l Concrete expansion anchors meet the requirements of ACI-349, i

            " Code     Requirements  for   Nuclear    Safety     Related    Concrete Structures" and IE Bulletin 79~02, Rev. 02, " Pipe Support Base Plate Design Using Concrete Expansion Anchor Bolts", November 8, 1979, with the provisions ' identified in Section 3.8.4.5 and further discussed in Appendix 3.9A.

See Appendix 3.9A, Section 1.7.4, for a discussion of concrete expansion anchors. Where required, snubber supports are used as shock arrestors for safety-related systems and components. Snubbers are used as structural supports during a dynamic event such as an earthquake or a pipe break, but during normal operation act as passive devices which accommodate normal expansions and contractions of the systems without resistance. For System 80+, snubbers are minimized, to the extent practical, through the use of design optimization procedures. , l Assurance of snubber operability is provided by incorporating l analytical, design, installation, in-service, and verification I criteria. The elements of snubber operability assurance for System 80+ include: A. Consideration of load cycles and' travel that each snubber will experience during normal plant operating conditions. Amendment R 3.9-52 July 30,- 1993

                                              - - ,,.      .- -      ~.           - . .     - - - - -

l

             - INSERT B (t.o'Section 3.9.3.4)                    _

Category I component supports are designed to meet the [ Seismic requirements'of Subsection NF, Section III of the ASME Code. Welding fabrication and installation, nondestructin examination (NDE) and acceptance standards shall be .in ac ,,rdance with Subsection NF, Section III of the ASME Code. In ailition. visual weld acceptance criteria shall be per the Nucler.r Construction Issue Group (NCIG) standard NCIG-01 (Reference 51). f - Radiographic examinations will be accepted by the COL applicant's nondestructive examination (NDE). Level III examiner . prior to final acceptance, f Confirmation that f acility welding activities are in compliance with the certified design commitments shall include verifications of the following by individuals other than those who performed the activity-

1. Facility welding specifications and ' procedures meet the applicable ASME Code requrements,

2. Facility welding activities are performed in accordance with the applicable ASME Code requirements,

3. Welding activities related records are prepared, evaluated and maintained in accordance with the ASME requirements,

4. Welding processes used to weld dissimilar base metal and welding filler metal combinations are compatible for the .

intended applications,

5. The f acility has established procedures for qualifications of welders and welding operators in accordance with the applicable ASME Code requirements, 1

6. Approved procedures are available and are used for pre-heating - .I and post-heating of welds, and those procedures meet the applicable requirements of the ASME Code, I

7. Completed welds are examined in accordance with the applicable examination method required by the ASME Code.

1 I l

                                     ~

l l

          ,y   p                -

• LCESSARUnamn

43. " Comprehensive Vibration . Assessment Program for Reactor Internals During Preoperational and Initial Startup Testing," USNRC Regulatory Guide 1.20 Rev. 2, May, 1976.

44. "A Comprehensive Vibration Assessment Program for Palo Verde Nuclear Generating Station Unit 1 (System 80 Prototype) ,' "

Combustion Engineering, Inc., CEN-263, Rev. 1 January, 1985 (Proprietary).

45. " Structural Analysis of Fuel Assemblies for Seismic and Loss-of-Coolant Accident Loading," Combustion Engineering, Inc., CENPD-178, Revision 1, August 1981.

46. " Random Vibrations, Elementary Theory, Structural Dynamics and Design, Signal Analysis and Testing", University of Arizona Seminar, October 29 to November 2, 1990.

47. " Flow Induced Vibration", R. D. Blevins, Second Edition, 1990.

48. "ATWS: A Reappraisal, Part 3: Frequency of Anticipated Transients", EPRI-NP-2230, January 1982.

49. " Development of Transient Initiating Event Frequencies for Use in Probabilistic Risk Assessment", NUREG/CR-3862, May 1985.

50. NRC Letter of September 11, 1992, " Safety Evaluation on the Use of Single Earthquake Design for Systems, Structures and Components in the'ABWR", Docket 52001.

51. NCIG-01 " Visual Weld Acceptance Criteria for Structural Welding of Nuclear Power Plants," Revision 2, EPRI NP-5380.

52. AWS D1.1, The American Welding Society, Structural Welding.

Code, 1990.

33. EPRI NP- 5639, "Gu"N AG b ?'YT 7'

               )?e a,rici % A ? ,t     Nla y / 9 N '

Amendment U Deceh 31, 1993  ; 3.9-85

CESSAR uninc- l

                                       ',h       l l

N TABLE 3.9-13 A STRESS LIMITS FOR CEDM PRESSURE HOUSINGS

                 / MSb Mrvk l We l                                   Stress Categories and          lV 0 :r: tin: 0: fitir-                    Limits of Stress Intensities (a)(b)
           'p .       Lev 41 A arn level A: No             1           igures)B-3221- and 3     -1,l        '

i erati Load plus rmal incl ding not . sj Oper ng & set.P1 Trans ts V p1 Safe utdow arthqu (c) i a( F rceo  ; f '. N t. Level D: Normal Operating Article F-1000, Appendix F, lV Loadings plus Faulted Plant Rules for Evaluation of Service Transients plus Safe Shutdown Conditions Loading with Level D Earthquake Forces plus Loads due Service Limits. to Design Basis Pipe Breaks and/or pipe breaks not eliminated by LBB. S, 't, Testina: Testing Plant Transients Paragraph NB-3226 For the above listed operating conditions, the following limits regarding. function apply:

1. Level A and level 8: The CEDMs are designed to function normally during  :

and after exposure to these conditions.

2. Level 0: For SSE plus Design Basis Pipe Breaks and/or pipe breaks not -l eliminated by LBB, the deflections of the CEDM pressure housing are ,

limited to the elastic design limits of Article F-1330, Appendix F 1 (defined above) so that the CEAs can be inserted after exposure to these I conditions.  ! NOTE: a. References listed are taken from Section III of the ASME Boiler and Pressure Vessel Code,

b. Dynamic loads including SSE, pipe breaks not eliminated by LBB l and Design Basis Pipe Breaks are combined by the SRSS method in )

accordance with the guidelines of NUREG-04B4.

c. Alternatively, a lower level of SSE motion may be used in accordance with Section 3.7.3.2.

I w.

                                                                                                                    ]

1 Amendment T I November. 15, 1993 l l

m - l INSERT A1 (Rev 2) to Table 3.9-13

       ~

1. Desian: Design Pressure, NB-3221 and Figure NB-3221-1, I Weight, Other Sustained including notes. l Mechanical Loads.

2. Level A: Normal Operating NB-3222 and Figure NB-3222-1, / I Loading plus Normal Operating including notes. '

Transients.

3. Level B: Normal Operating NB-3223 and Figures Loading plus Normal Operating NB-3221-1 and NB-3222-1,

    & Upset Transients plus Low         including notes.

Cycle Fatigue Loading due to Safe Shutdown Earthquake (SSE) Forces.

       ~.

      -                                                ~

CESSAR !!ninema TC one >cf su.gsr/ bu.l/ ding .S$ru ciures 3rd bs*fned 7$ nuat Q a derh k Apered y 2.ri9. attached to valve operators are also evaluated. The valve operator support does not support the pipe. 1.6.8 EXPANSION JOINT REQUIREMENTS Expansion joints are evaluated to ensure compliance with vendor allowables based on the stress report provided by the vendor. 1.6.9 WELDING AND WELD ACCEPTANCE CRITERIA f Welding fabrication and installation, nondentructive examination (NDE) and acceptance standards for ASME Code Class 1, 2, and 3 piping shall be in accordance with Articles 4000 and 5000 of Subsections NB, NC, and ND in Section III of the ASME Code.

        /                                                                                   Y ggtt        1.7           PIPE SUPPORT DESIGN REOUIREMINT_B_

1.7.1 GENERAL I Pipe supports are designed to meet the intended functional requirements of the stress analysis as well as the specified stress limits for the support components. Support components include typical structural steel members as well as manufactured catalog items for typical support components. r Supports are idealized in the piping analysis as providing restraint in the analyzed direction while providing unrestricted movement in the unrestrained direction. Since the design of supports cannot completely duplicate the idealized condition, supports are designed to minimize their effects on the piping analysis. Additionally, it is confirmed that the support design does not invalidate any assumptions used in the analysis of the piping system. In addition to loads defined by the stress analysis, any , additional forces the support are subjected to are considered in I the support qualification. I l 1.7.2 DESIGN CONSIDERATIONS

1. 7 . 2 .1' Deadweicht Loads Gravity loads of the pipe are typically restrained by two types of supports. The piping analysis defines whether the support is designed as a rigid or flexible support. Flexible supports are specified when the pipe must be restrained for its gravity weight, however must remain free to move during thermal expansion. Vendor supplied spring components with specified spring constants are typically provided in this application.

I Amendment U 3.9A-22 December 31, 1993

                                                   -~

INSERT C (to Section 1.6.9 of Appendix 3.9A) m [ Radiographic examinations will be accepted by the COL applicant's nondestructive examination (NDE) Level III examiner prior to final-acceptance.

         !   Confirmation that f acility welding activities are in compliance I   with the certified design commitments shall include verifications f    of the following by individuals other than those who performed the
       '     activity:

1. Facility welding specificatipns and procedures meet the applicable ASME Code requrements,

~ I \ l 2. Facility welding activities are performed in accordance with _he applicable ASME Code requirements, I

3. Welding activities related records are prepared, evaluated and y maintained in accordance with the ASME requirements,

4. Welding processes used to weld dissimilar base metal and welding filler metal combinations are compatible for the intended applications,

5. The facility has established procedures for qualifications of welders and welding operators in accordance with the applicable ASME Code requirements,

6. Approved procedures are available and are used for pre-heating and post-heating of welds, and those procedures meet the l applicable requirements of the ASME Code,

7. Completed welds are examined in accordance with the applicable examination method required by the ASME Code.

Welding activities involving non-ASME pressure retaining piping shall be accomplished in accordance with written procedures and shall meet the requirements of the ANSI B31.1 Code. The weld acceptance criteria shall be as defined for the applicable nondestructive examination method described in ANSI B31.1.

CESSARannncug - - x_ f guy a,wer f Jasne csyy19-Lz pYe mports l AS clefneef th Secl'tm l t

               .v]v b'                                                     _ _

induced into the pipe. Materials used as welded attachments are l compatible with the piping material. ' 1.7.2.13 Minimum Desian Loads In order to provide some uniformity in load carrying ability, all supports are designed to minimum loads. All supports are designed for the largest of the following three loads: e 100% of the Level A condition load from the piping stress analysis

• The weight of a standard ANSI B31.1 span of water filled, schedule 80 pipe ,

e Minimum value of 150 pounds f, 1.7.3 LOAD COMBINATIONS Load combinations are in accordance with Section 3.9.3.1 and are detailed in Table 3.9-12. For common supports, the SRSS method for combination of dynamic loads is used. 1.7.4 ACCEPTANCE CRITERIA Pipe supports are either linear or plate and shell type devices. A linear type component support is defined as acting under essentially a single component of direct stress. Such devices may also be subjected to shear stresses. Plate and shell type of supports are fabricated from plate and shell elements and are normally subjected to a biaxial stress. Categ rY I pipe - Seb WcArequirements support members are designed to meet the W defin d in ASME Code, Section III, Subsection NF. ' l For A500 Grade B t le steel, NF requirements are supplemented by the weld requiremerts of AWS D1,1. " Structural Welding Code" .g g (Reference 4.26). V f ldi oncle ruc ve f icat n d i amin ion NDE) and a ept ce st dard shal talla lon,3 ' A' e ac da e w S sect n NF Sec on I of eA ' IsW ' Cod . ' ad tion vis '1 we acc tanc critgia s 11 beyber pf$e , leay' Con ruc ' n sue c oup NCIG)/ stan . rd NCIG-g ge rency 4.25 . Category II pipe support members are designed to meet the requirements of the AISC Steel Construction Manual. Standard support manufactured catalog items are designed to meet the requirements of MSS-SP-58, " Pipe Hangers and Supports-Materials, Design and Manufacture." The application of catalog components is consistent with the manufacturer's requirements and are designed to meet the manufacturer's load rated capacities for the items. The piping design is consistent with the manufacturers' requirements for pipe deflection limits at pipe supports, such as requirements for travel in snubbers and hangers, or with industry practice, such as requirements for the Amendment U 3.9A-26 December 31, 1993

CESSAR inlinemu

   -     snubbers, mechanical or hydraulic;
   -     constant or variable spring support hangers;
   -     rigid supports consisting of anchors, guides, restraints, rolling or sliding supports, and rod type hangers;
   -     sway braces and vibration dampeners;
   -     structural attachments such as ears, shoes, lugs, rings, clamps slings, straps and clevises;
   -     any other NRC approved devices.

Concrete expansion anchors are designed to meet the requirements of ACI-349, " Code Requirements for Nuclear Safety Related Concrete Structures", . cit!. uo w- .ug oma u m..~ . m . -. . . a

                         ^ 2S     Mended bySech;m J.?. g.C/,

5i-c-- x r . factor of sa ety ac eptable to the NRC i applied to anchor llowab es. B. Prov olons e take for an or stre th red ctions v en the anc or is ocated n the ncrete ension one.

     . T     fail       e cone angle         ed is co sisten with re ent tes               (

ta fo the e'cific a licatio and ac ptable the NR . D. Emb, ment Aength calculati ns fo duct e anc ors de nstra a m' imum fac r of afety f 1.5 when termi pe ng the ullout st ngth of the co rete b ed on e mi imum te ile stren h of th anchor teel. E The/ energy bsorption apabili (defo ation apabil any afer yiel is cons cred for the an or mate ial an chor a eptable to he NRC aff for ductile pplica ions is chos . This assures that the design strength of concrete for a given expansion anchor or group of anchors is greater than the strength of the anchor steel, accounts for the effect of shear-tension ' interaction, and considers minimum edge distance and bolt spacing on expansion anchor capacity. Base plate flexibility is accounted for in the calculation of expansion anchor bolt loads. 1.7.5 JURISDICTIONAL BOUNDARIES The jurisdictional boundaries are defined in ASME Section III, Subsection NF. l 1.8 POSTULATED PIPE BREAKS l 1.8.1 CLASSIFICATION l 1.8.1.1 Hich Enerov High energy piping systems are those systems or portions of systems that are maintained pressurized at either temperatures in Amendment P 3.9A-27 June 15, 1993

CESSAR !!? dine-l 1 i does not meet the LBB criteria of the PED, the COL applicant will revise the design until the LBB criteria of the PED are met.

   % /Ce6neDiatien with4s-built pi                 g system pardmeter     will,also]

lbe mdde bp the' COL applican y .d6monstrating,- that the j dim'ensional- and material- propertfes of' each as-built pipisxf ji Af ' system are' consistent with the-parameters used'in the development l Y p0)* y,

        ,01LSd'fKthVPEDs    and (gttiat cnde adowab_    __ amithe' Eheas-built LBBrPEDpiping   responses.

cr,iteria/ / meet the 1.10 TUBING 1.10.1 GENERAL Design, analysis and loading considerations that are used for piping and supports are used for tubing. Due to the amount of tubing, bounding analyses are performed. This analysis method is also used for small-bore piping. These criteria apply to safety-related tubing. Non-safety related manifold valves, solenoid valves, and instruments located over or near safety-related equipment or components are supported using the same criteria, except where justified by analysis. This prevents damage, degradation, or interference with the performance of equipment required for safety functions. 1.10.2 SUPPORT AND MOUNTING REQUIREMENTS Two support mechanisms are used, free tube spans and tube track supports. Criteria for each tube support mechanism are determined as described above. The following are additional support and mounting considerations: A. Tubing that is routed in two or more Seismic Category I structures (i.e., Reactor Building, Containment, Main Steam Valve House, Nuclear Annex, Diesel Generator Building) are verified to have sufficient flexibility to allow for differential building displacements. B. Span lengths are chosen and supports and tube details are designed to accommodate heat tracing and/or insulation requirements. C. All reservoirs, valves, and other in-line components are independently supported. D. Movements of the root valve (SAM and TAM) between the pipe and the tubing are considered. Amendment U 3.9A-39a December 31, 1993

INSERT TO 1.9.6.6 OF 3.9A

    Reconciliation of the as-built piping systems with the final design will be documented by the COL applicant in a LBB Evaluation Report.

The LBB Evaluation Report shall contain results of the LBB evaluations for as-built piping. The LBB evaluations shall employ

 ',              methods described in Section 1.9 of this appendix. Reconciliation-of each as-built piping system qualified for LBB will be made by the COL applicant by demonstrating that:

( (1) the as-built piping system meets the screening criteria of Section 3,6.3, (2) the dimensional and material properties of the as-built piping system are consistent with the parameters used in

         ,                      the development of the final LBB PED (s) for that piping system, (3)     the as-built piping responses meet        the     ASME  Code allowables and the final LBB PED criteria.
               \
                 \
                               .. #               ~

l l l

CESSAR 8n'ancariau #M# 3.10.2 BEIBMIC AND DYNAMIC QUALIFICATION OF ELECTRICAL EQUIPMENT Instrumentation and electrical equipment used for post-accident monitoring, the Reactor Protective System (RPS), the Engineered Safety Features Actuation System (ESFAS), the actuation devices for ESF system actuated components, and the emergency power system are designed to seismic Category I requirements to ensure the ability to initiate required protective actions during, and following, a Safe Shutdown Earthquake (SSE) and for all static and dynamic loads from normal, transient and accident conditions; and, to supply power, following an SSE and for all static and dynamic loads from normal, transient and accident conditions, to components required to mitigate the consequences of events which require safety system operation. Instrumentation and electrical equipment designated Seismic Category II are shown to maintain their structural integrity and not adversely impact safety related equipment during an SSE and for all static and dynamic loads from normal, transient and accident conditions. g Methods and procedures for qualifying electrical equipment and instrumentation are escribed below, and meet the requirements of gf Regulatory Guide 71 03 Revision 2, and IEEE Standard 344-1987. Y 3.10.2.1 Methods and Procedures for Oualifyinc Soismic Catecorv I Electrical Ecuipment and Instrumentatin Seismic Category I instrumentation and electrical equipment required to perform a safety action during a seismic event and for all static and dynamic loads from normal, transient and accident conditions; after a seismic event and for all static and dynamic loads from normal, transient and accident conditions; or both are qualified with appropriate documentation in accordance with the requirements of the equipment specifications. These requirements are consistent with those of IEEE Standard 344-1987,

" Seismic Qualification of Class 1 Electrical Equipment for Nuclear Power Generating Stations", and Regulatory Guide 1.100, Rev. 2. The methods and procedures used for qualifying Seismic Category I electrical equipment and instrumentation include the        i following l

A. Testing and analyses are used to confirm the operability of l the instrumentation and electrical equipment during and 1 after an SSE and for all static and dynamic loads from normal, transient, and accident conditions. Prior to SSE qualification, it is demonstrated that the equipment can withstand the application of five (5) cycles of 1/2 SSE excitations without loss of structural integrity. Analyses alone, without testing, is used as a basis for qualification only if the necessary functional operability of the Amendment R 3.10-3 July 30, 1993

ATTACHMENT 12 e 9 J i i l 1

CESSAR E!!acma TABLE 1.84 (Sheet 1 of 2) SYSTEM 80+ INDUSTRIAL QQDES AND STANDARDS Code Edaion TMs ANSI /American Concrete insthuta (ACl) - 318 1989 Building' Code Requirements for Reinforced Concrete,1991 Printing. l 349 1985 Code Requirements for Nuclear Safety-Related Concrete Structures l ANSI /American insthirts of Steel Construction [AISC) N690 1984 Specification for the Dulun, Fabrication, and Erection of Steel Safety-Related Structures for Nuclear Facilhies , 1989 /)fbytua/ s;$ Stee/ OHsku.d'bN> $lbMAN Shs.s y

                   %)>s. , Mi>a% .fAL ANS!/American Nuclear Society IANS) 51.1     1983       Nucisar Safety Criteria for the Design of Stationary PWR Plants 56.2     1989       Containment isolation Provisions for Ruld Systems after a LOCA 58.1     1982       Plant Design Against Missiles 58.2     1988       Design Basis for Protection of LWRs against Effects of Pipe Rupture 58.8      1984      Time Response Design Critoria for Safety-Related Operator Action 58.9      1987       Single Failure Crtteria for LWR Safety Related Fluid Systems ANSl/American Petroleum Institute [ API) 650       1988      Weided Steel Tanks for Oil Storage               .

ANSI /American Society of Civil Engineers 7 1990 Minimum Deslon Loads for Building and Other Structures (ANSI A58.1) ANSI /American Society of Mechanical Engineers (ASMEl BPVC 1989 Section 11; Materials Specifications BPVC 1989 Secdon ill; Rules for Construedon of Nuclear Power Plant Components; Division I, Division 11 l BPVC 1989 Section V, Non Destructive Examination BPVC 1989 Section Vill; Rules for Construction of Pressure Vusets BPVC 1989 Section IX; Qualification Standard for Welding and Brazing BPVC 1989 Section XI: Rules for Inservice inspection of Nuclear Power Plant Components Editions and Addenda As Applicable AG 1 1991 Code on Nuctur Air and Gas Treaanent B31.1 1992 Power Piping OM S/G 1990 Standards and Guldes for Operation and Maintenance of Nuclear Power Plants; through 1992 Addenda. NOA-1 1989 Quality Assurance Program Requirements for Nuctur Facilities, and NOA 1b-1991 Addenda NOA 2 1989 Quality Assurance Requirements for Nuclear Power Plants, and NOA-2a-1990 Addenda

                                                                                     ~

i amandment.U December 31, 1993

C E S S A R E!?& ne m a i Internal Pressure Failure Pressure Level (Dsia) Probability l Design 68 0.00 l l 1.5 x Design 94 0.00 l l ASME Level "C" (Local 3D) 145-135 0.03 l ASME Level "C" (Global) 157-147 0.05 l l Nominal Yield (mean 172-160 0.50 l properties) l Maximum Yield (max. 187-174 1.00 l properties) This method was used to translate data obtained from containment stress analyses to fragility (probabilistic failure) curves at temperatures typical of both early and late containment failure. It was assumed that early failure stress curves allow greater strength because of the lower shell temperatures expected prior to containment failure. In these instances, containment failure is due to a rapid pressurization process to which the shell cannot thermally respond. The design basis accident (DBA) peak temperature (290*F) was selected as the conservative temperature for evaluation of the early containment failure. I Late containment failure includes a gradtal overpressurization process that takes from hours to days; therefore, failure is expected to occur with a " hot" wall. The late containment failure fragility curve for " wet" sequences was conservatively established assuming the 350*F peak containment environmental temperature. The dry cavity overpressurization scenario was a conservative upper bound of the median shell temperature (See Section 19.11.5). - l The fragility curve generited using the pressure-failure probability points of the above table are shown in Figure 19.11.3.1-3 for a containment environmental temperature of 290*F. This curve is conservatively biased in the low pressure . tail of the curve and consequently results in a modestly conservative ,j bias within the PRA. This is confirmed by comparison of the y  ; piecewise linear fragil y curve developed in this section with i alternate methodologies employing a lognormal containment l fragility curve construction. (See, for example, Reference 111) . l (su Ryuk M.nH) Amendment U 1 19.11-8 December 31, 1993 l l

I APPENDII 19.11H V COMMENTS ON THE CONSTRUCTION AND APPLICATION OF THE SYSTEM 80+ CONTAINMENT FRAGILITY CURVE ,

Introduction l The construction of the containment fragility cutve used in the FRA is described in Section 19.11.3.1.2.4. The construction is based on general guidance used in the support of NUREG-1150. In estimating failure, it was assumed that once the material yield point is reached using an axisymmetric shell model described in Section 19.11.3.1.2.3.1 and 19.11.3.1.2.3.2, the containment will fail. The above procedure has been biased to provide a conservative estimate of the containment failure probability. This conservative bias arises from the following assumptions / procedures:

1. All properties are evaluated at high mean shell temperatures. In fact, it is expected that during most containment challenges to which the fragility curve is applied the average containment shell temperature will be between 150 F and 250 F. This temperature range is based on the fact that the FRA containment challenges with sprays operational will maintain a cool containment atmosphere. For those transients where sprays are unavailable, the shell temperature prior to burn will be less than 250 F to ensure the containment atmosphere ,f is not inerted. While burn temperatures can be high, their short y duration (less than 30 seconds) and the large mass of the steel shell results in only minor increases in the mean shell temperature.

This assumption conservatively biases the median containment strength calculation from 2 to 10%.

2. In the fragility curve construction, the median material yield stress was taken to be 1.10 times the minimum expected yield stress.

Material data discussed in Reference 210 of CESSAR-DC indicates that the median shell stress is actually 1.167 times the minimum yield stress. The difference between these values was taken to approximately account for effects of material variations and modeling uncertainties.

3. The fragility curve used in the PRA assumed a linear fit between the points defined in the Table in Section 19.11.3.1.2.4 f e m m +a = =amaq6 This procedure overestimated the failure probability of the shell in the tail region of the fragility curve below the 31 failure point

                                                                                        ~

(in the pressure region between 94 and 145 pain). The fragility curve challenges for System 80+ were mostly confined to containment pressure below 145 psia. The highest containment challenge noted for the very low probability high pressure DCH event resulted in a pressure of 151 psia, See Figure 19.11.4.1.1.-4A. Comparison of C-E Fragility Curve with the Methodology of Reference 1 An alternate method of defining a fragility curve may be established by defining a logarithmic standard deviation for material properties and for modeling uncertainty. Given a failure pressure calculated from mean material properties a mean failure pressure probability curve can be developed. The methodology is generally analogous to the seismic strength analysis employed in Section 19.7.5. For the ultimate pressure fragilit'y curve, the true mean containment failure

presSWe U}dck (WSU]Ss in the w h12l y pressure (@ 290 F) based on the Reference 210 data would be 180.7 psia (166 psig). The beta factor based on the variation in the material yield point is

      .09. Material uncertainty in this range is typically consistent for fragility analyses. In order to account for other undefined property variations which are associated with the imperfect experimental modeling of a real structure (variations in plate thickness, 1,cundary conditions, velds, residual stresses, etc.) the material uncertainty is combined with a second factor of equal value

(.09). This factor is equivalent to the A parameter of Reference 1. This selection conservatively bounds the value of .05 used in that reference for this parameter. In addition, Reference 1 also suggests the use of a modeling uncertainty of .05 for a spherical shall geometry (see Reference 1, page 57). This selection is typically associated with the use of simplified Reference 1; modeling equation 5.8(, Calculations offyield stresser used in the System 80+ were based on use of the ANSYS computer code. Therefore, the snshD5dvariability e factor is not considered applicable, but was retained for conservatism. (In fact, Reference 1 indicates that ANSYS calcualtions tend to underpredict structural capability by approximately 101. This bias, as well as, bias associated with the high temperature material property selection provides additional conservatism which is not reflected in the above statistical treatmant.) Following the procedure indentified in Reference 1, a combined coefficient of variation,$, for the spherical shell model was found to be: 8* = (.09)* + (.09)* + (.05)* and 8 = 0.137 For illustration purposes a combined standard deviation of .135 was selected for evaluating the fragility curve. Assuming the fragility curve to be a lognormal distribution, the coefficient of variation, 8, is 8 = in (P,,,/P,) / K, A fragility curve explicitly accounting for material and modeling uncertainties can be then be evaluated as follows: P, = P,,,exp ( K, 8 ) where P,  : pressure with x probability of containment failure median / i P,, :seest failure pressure K,  : coefficient associated with x probability of containment failure 8  :. combined standard deviation

The results of this curve construction and the data used for the System 80+ PRA i fragility estimates are presented in Table 19.11H-1. Table 19.11H-1: Comparison of Estimated Pragility Curve Methods q Probability of Failure Linear Approximation Combined Beta Method  ! used in System 80+ PRA Pressure (psia) l l Pressure (psia) l 0.00 94 --- I 0.001 95.2 124 . 1 0.005 100 130.9 0.01 106 135.8 0.02 125 138.5 0.03 145 143.4 0.05 157 147.5 - 0.10 158 154.1 0.25 163.6 166.09 0.50 172 180.7 f Using the current PRA values, the failure probability is significantly l exaggerated in the low pressure region below 140 psia. Both methods yield ; similar results around 145 psia. In the pressure range from 145 to about 160 psia failure probabilities computed using the beta method are somewhat higher than that used for the FRA. A review of these differences illustrates that for the region the fragility below about 145 paia, the net consequence of the use of the System 80+ PRA curve is to conservatively bias the overall shall failure probability. As will be discussed . below. Containment fragility curves are used in evaluating three containment threats: hydrogen burn, DCH, and rapid steam generation. Impact on Hydrogen Burn Failure Potential Hydrogen Burn failure probabilities are shown in Tables 19.11.4.1.3-3 and 19.11.4.2.4-1. For early hydrogen burns the largest expected pressure threat was estincted to be below 106 psia. This was classified as having a containment failure probability of .006. Using the beta method, the probability is virtually zero. A review of the late hydrogen burn sequences produce similar conclusions. The late hydrogen burn pressures range are defined for three cases as, 103, 125.2, and 140 psia. This results in containment failure probabilities of .006, .0184,

and .0276. Using the beta method, the failure probabilities would be lower for l the first two cases ( less than .001) and about the same for case 3. Impact on DCH Containment Failure Potential l I The DCH containment threat is evaluated in Section 19.11.4.1. Figures l 19.11.4.1.1-4 (a through c) illustrate the use of the fragility curves and bounding pressures used in the quantification process. For all DCH events that result from an intermediate pressure RV failure, the largest containment threat is below 120 psia, and therefore use of the existing PRA model results in fragility estimates that ar.e consistently biased high. For the high pressure RV DCH, containment pressure threats are distributed between 99 and 151 psia. Of those threats fewer than 2% are above 145 psia. The net effect on using the existing PRA approach would produce higher DCH conditional containment failure probabilities than that using a beta approach. Impact oa Rapid Steam Generation Rapid steam generation issues are discussed in Section 19.11.4.1.2. Table 19.11.4.1.2-4 indicates that the highest containment threat is 98 psia. This produces a small conditional containment probability using the existing fragility curve. The beta developed curve would indicate-this failure probability to be zero.

                                                 ~~_         .

Reference

1. NUREC/CR-2442," Reliability Analysis of Steel Containment Strength",

Creimann,L.C., et. al., Ames Laboratory, June, 1982 SB 'fg n e

a ,A Eu. a A J A4.~... mM4,..J *J49.4J-@WAdR 94 A Bm &E4 i, A't-- g.h- 4...hilr # I'emem m,. 4L 7 .44..A 4w F 3 ATTACHMENT 13 l i I

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I l l i i 1 1 1 l

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