ML17341A552

Forwards Seismic Qualification of Auxiliary Feedwater Sys, in Response to 810210 Generic Ltr 81-14.Results of Walkdown Revealed Existence of Minor non-seismic Category 1 Components

ML17341A552
Person / Time
Site:	Turkey Point
Issue date:	09/18/1981
From:	UHRIG R E FLORIDA POWER & LIGHT CO.
To:	EISENHUT D G Office of Nuclear Reactor Regulation
References
GL-81-14, L-81-405, NUDOCS 8109280310
Download: ML17341A552 (48)

Text

REGULAiTQR INFORMAiT'ION>

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~'>g P.O.BOX 529100 MIAMI, FL 33152 Pp>w i'd%%FLORIDA POWER&LIGHT COMPANY September 18, 1981 L-81-405 Office of Nuclear Reactor Regulation Attention:

Mr.Darrell G.Eisenhut, Director Division of Licensing U.S.Nuclear Regulatory Commission Washington, D.C.20555

Dear Mr.Eisenhut:

Re: Turkey Point Units 3 8 4 Docket Nos.50-250 8I 50-251 Generic Letter No.81-14 Sei smi c ual i f i cati on of Aux i 1 i a r Feedwater S stem F+v/>-Please find attached our report providing the information concerning auxiliary feedwater seismic design that was requested in Generic Letter No.81-14, dated February 10, 1981.The Turkey Point Units 3 5 4 Auxiliary Feedwater System (AFWS)is a seismically designed system as described in the FSAR.It is designed, constructed, and maintained in a manner consistent with other safety grade systems in the plant.Our architect-engineer has verified the seismic qualification of each of the AFWS components.

and supporting systems.A walkdown, as requested in your letter, was performed for those portions of the AFWS where sufficient information was not retrievab'le to verify its seismic qualification.

The results of the walkdown, indicated that the AFWS as currently exists contains several minor non-seismic Category I components.

These items, and the corrective actions taken to upgrade the system are summarized in Section IV of the attachment.

Very truly yours, Robert E.Uhrig Vice President Advanced Systems 5 Technology REU/J EM/ras Poyr S cc: Mr.J.P.O'Reil.ly, Region II Harold F.Rei s, Esquire af092803i0 Si09i8 PDR ADOCK 05000250 ,PDR PEOPLE...SERVING PEOPLE STATE OF FLORXDA))COUNTY OF DADE')ss Robert E.Uhrig, being first duly sworn, deposes and says: That.he is a Vice President of Florida Power 5 Light Company, the Licensee herein;That he has executed the foregoing document;that the state-ments made in this said document are true and correct to the best of his knowledge, information, and beli'ef, and that he is authorized to execute the document on behalf of said Licensee-Robert E.Uhrig Subscribed and sworn to before me this l9 Fl~P~~'":7O'LYt."-Y PUDLTC, n and for the county of Dade, State of Florida Notary Pubtic, State of Fiorida at Large My Commission Expires October 30;1983 Bonded thru Maynard Bonding Agency Ny commission expires:

SEISMIC QUALIFICATION OF THE AUXILIARY FEEDWATER'YSTEM CONTENTS I.INTRODUCTION II.SYSTEM DESCRIPTION III..SEISMIC MEZHODOLOGY IV.NON-SEISMIC CATEGORY I COMPONENTS'IGURES:

FIGURE 1-AFW SYSTEM ATTACHMENTS:

ATTACHMENT (A)SEISMIC CRITERIA

,I'NTRODUCTION The Turkey Point Units 3 and 4 Auxiliary Feedwater System is classified as a seismic system and was included within the scope of NRC I.E.Bul-letins 79-02, 79>>04, 79-07, 79-14, and 80-11, and I'.Information Notice 80-21.Although the system was not originally designed and classified as,a seismic system, it was upgraded and reclassified prior to the Turkey Point Plant receiving an operating license to meet the seismic criteria imposed'by the AEC at that time.This report contains the additional information requested by Generic Letter 81-14, Seismic Qualification of Auxiliary Feedwater Systems.It includes a brief system description, a.discussion of the methodologies, and a list of non-seismic components.

II.SYSTEM DESCRIPTION (Extracted verbatim from NRC L'etter to FPL dated 10/16/79 and updated.Portions updated noted by a bar and asterisk in righthand margin.)Confi uration-Overall Desi n The auxiliary feedwater system (AFWS)for the Turkey Point Plant (Units 3 and 4), as shown in Figure 1, consists of three steam turbine driven pumps, i.e., one pump normally aligned to each unit and the third pump is a shared standby for either unit.Each pump normally delivers 600 gpm (9 2775 ft.head)feedwater to the three steam generators (SG)in each unit.Also, the control room operator can manually direct flow from any pump to all three steam generators of either unit.Under a design basis accident, only one pump would be required in order to cool the plant down to a condition where the RHR system can be put into operation to continue the safe plant shutdown process.Primary water supply for the AFWS comes from the Seismic Category I condensate storage tanks (CST)of both units.Each CST has a capacity of 250,000 gallons with a minimum reserved storage capacity of 185,000 gallons of demineralized water.With this quantity of water, the licensee indicated that the unit can be kept at hot standby condition for 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> and then cooled to 350,R, at which point the RHR 0 system can be put in service, or the unit can be kept at hot standby condition for about 23 additional hours.All the.manually operated valves associated with CST's are locked open.A secondary water supply comes from the non-seismic Category 1 water treatment system.An additional feedwater supply can be provided from the main feedwater system of the adjacent Units 1&2 (non-nuclear power plant).Com onents-Desi n Classification 0 The AVOWS is designed according to seismic Category I requirements.

The APWS is classified as an engineered.safety related system and its associated instrumentation and controls are designed accordingly.

Power Sources The turbine driven pumps are supplied with steam from the main steam line of either or both units upstream of the MSIV.The operator normally selects the steam supply from the Unit which has lost its normal feedwater supply.The turbines have an atmosphere exhaust.Steam can also be supplied from the Unit having normal feedwater supply and from an auxiliary steam system connection to Units 1&2.The turbine driven pump steam supply line has a normally closed AC motor operated valve in series with a normally closed DC solenoid air operated valve.The pump discharge control valves are DC solenoid operated air valves.Instrumentation and Control Controls The steam generator water level is manually controlled by the control room operator using either one of the DC solenoid operated air valves.A seismically installed nitrogen back-up system supplements the non-seismic instrument air supply to these valves.Local manual operation of these valves can also be performed.

The AFP pump feedwater discharge rate is always greater than the turbine steam consumption when the steam pressure is higher than 120 psig.Qhen the steam pressure is reduced to 120 psig, the RHR system is started and the AFW pumps are shut down.Information Available to 0 erator Low water level'n the condensate storage tank will alarm and annunicate in the main control room.In addition, AFH flow indication, SG water'level, and control valve position indication are provided'n the control room.Initiatin Si nals for A'utomatic 0 eration All three APW pumps will automatically start by any of the following signals from either Unit: (a)safety in)ection (b)low-low water level in any of the three steam generators (c)loss of voltage on both 4160V buses (d)loss of both main feedwater pumps.Any one of these signals will also automatically open, the.normally closed motor operated and air operated valves in series which isolate the main steam, line from the steam supply header of each AFW pump turbine.Air to operate the AFW control valves'to the steam generators is supplied when the steam supply valves commence.opening.The ASS can also be started manually in the control room or from: the local station.In accordance with NUREG 0578 and 0737, the following AFW system modifications are in process: (a)design andinstallation of'a safety-related, initiation ,and flow indication system (b)design and installation of a qualified lube.oil cooling system for the AkW pumps (c)replacement of two AC operators on the AFW steam admission valves with DC operators for each unit (d), addition of redundant steam supply lines to the ABiT'ump turbines (e)addition of redundant discharge.

piping from the AFW pumps (f)addition of redundant safety grade.condensate storage tank level indication..

III.SEISMIC METHODOLOGY Attachment A is a reproduction from the FSAR for Turkey Point Units 3 and 4,-describing the seismic criteria and methodology used to qualify the majority of the Class I (seismic)structures, equipment and components.

In addition, reproductions of applicable follow-up questions are included.Additional information on methods of qualification and scope not specifically described in the before mentioned excerpts from the FSAR is provided below., Initially the pumps and drives were not, procured te any specific criteria.They were later certified, by the supplier to be capable of functioning under the imposed seismic loadings.Class I structures were designed for an OBE, but later checked for the maximum earthquake (SSE).Steam supply piping to the AFW turbines, suction piping from the.condensate storage tanks to the,AFW pumps, discharge piping from the pumps to a point downstream of the main feedwater isolat'i'on valves and AFV pump recirculation piping were considered to be within the scope of NRC Bulletins'9-02, 79<<04, 79-14 and 80-11.

Branch lines to the first valve were also included.Valves were analyzed with the piping taking into account the.C.G.of valve operators when applicable.

Electrical equipment was purchased under specifications that included a description of the seismic.design criteria for the plant.Instrumentation, controls and panels supplied by the NSSS are covered under WCAP"7397<<L and its supplements.

Conduit supports were.installed in accordance with written ,procedures and based on conduit manufacturer's recommendations.

Typi'cal conduit supports have been evaluated and comply with the seismic requirements of the Turkey Point FSAR.Transfer switches (120 VAC)mounted on the 120 VAC distribution panels were supplied by A'irpax Electronics Corporation.

These switches will withstand shocks of'00 G without tripping, while carrying full rated current when tested per MlL-STD-202C (Method 213)and will withstand'ibrations of 10 G without tripping while carrying full rated current when tested per MlL>>STD-202C (Method 201A).IV.NON"SEISMIC CATEGORY I COMPONENTS The following is a list of components which have been identified as presently.

being non-seismic Category I.The items are arranged in the order presented in Table 1 of Generic Letter 81-14 and those that are being upgraded are identified'.

Condensate

'Transfer Pum-The condensate transfer'ump, located in a branch line from the AFW pump'suction, acts as a pipe anchor.Piping, to the pump is seismically supported to maintain the nozzle loads on the pump nozzles to within good engineering limits.The pump is not required to function.This, branch line will be disconnected from the AFW pump suction as,a result of.new demineralized water system modifications..

Condensate Recover Transfer Pum>>The condensate recovery transfer pump located in a branch line from the condensate transfer pump suction, acts as a pipe anchor.Piping to the pump is seismically supported to maintain nozzle loads on the pump nozzles to within good engineering limits.The pump is not required to.function.This branch line will be disconnected from the AFW pump suction as a result of new demineralized water system modifications.

(2)~Pi tn Condenser Make-U Line-The.condenser make-up line, a branch 1'ine from the AFW pump suction does not have the required valving arrangement.

Currentl'y, a, new demin-eralized water system is being installed which will include a new condenser.make-up line.The branch line from the APW'uction will be cut and capped when the new system is placed in service.(3)Valves and Actuators Air 0 crated Vent Valve-Three>>quarter (3/4)inch valves located downstream of the motor-operated'.steam admission valves to the APW turbines are provided to prevent the AFW turbines from turning due to valve leakage..Pailure of the valves.in the open position will not prevent the APW system from per-forming its required function.

(4)Power Su lies No non-seismic power supplies were identified.

(5)Primar Water and Su 1 Paths Condensate Recover Tank" The condensate recovery tank supports the condensate recovery pump which acts as a seismic pipe anchor.The condensate recovery line will be disconnected from the AFW pump suction as a result of new demineralized water system modifications.

(6)Secondar Water and Su 1 Not applicable to Turkey Point Units 3 and 4.(7)Initiation and Control S stems Condensate Stora e Tank Level Transmitter

-The condensate tank level transmitter was procured to control grade.Loss of function of the transmitter and indicator will not prevent the AFW system from performing its function.Redundant safety grade indication is currently being added.Local Pressure Indicators

<<Local pressure indicators were procured to control grade (industrial grade standards).

Loss of the pressure indicator (gauge)function will not prevent the AFW system from performing its required function.Pressure Switches-Pressure switches located upstream of the AFW turbine trip and throttle valves are currently used to initiate"the air supply to the normally closed turbine pressure reducing valve.The use of these pressure switches with the new high pressure AFH turbines has yet to be determined, since the normally closed turbine pressure reducing valves will be replaced by normally open trip and throttle valves.AFW Flow Control and Indication

-AFR flow control and indication was originally procured to control grade standards.

This is currently being upgraded to safety grade (Class XE).N Backu S stem>>N2 Backup system is provided to supplement the instrument air used for AFP control.The components of the system were procured to industrial standards and installed seismically.

(8)Structures Su ortin or Housin AFW S stem Items A.Turbine Buildin-The turbine building is not a Class I structure.

However, the turbine building is a substantial steel and concrete structure with considerable inherent rigidity and resistance to the low OBE and SSE loads for the Turkey Point Plant.Portions of the AFW system piping.is supported along the east side of the turbine building.These pipe'upports have been analyzed for seismic loads and the portions of the, turbine building to which they are attached.have been analyzed for the seismic loads and are within allowable stresses.B.Other Su ortin Structures

-The following structures were walked down to evaluate the current structural condition of concrete, steel and anchor bolts and to identify any readily recognized deficiencies in seismic resistance.

Several minor maintenance action items were identified and have been corrected.

As granted by Generic Letter 81-14 and its enclosures, engineering judgement was used to determine the adequacy of the following structures to withstand the low OBE and SSE loads at'the Turkey Point Plant.a)Condensate Transfer Pump Foundation; b)Condensate Recovery Pump.Attachment to the Condensate Recovery Tank.c)Condensate Recovery Tank Foundation.

d)Condensate Storage Tank Level Transmitter Attachment to Condensate Storage Tank.

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{I JNI I:IS)lt)WN)Attxtltitly I'tletlw/tter Syitt>>>>I III I<VV I l>>lit (I J>>tti I el>>ll 4)I'l/JI>>I~I ATTACHMENT A Compr is ing: FSAR Appendix 5A-Seismic Cl'assification 6 Design Basis for'Structures, Systems and Equipment for Turkey Point-(27 pages)'SAR Questions (5 pages)

APPENDIX 5A SEISMIC'LASSIFICATION

&DESIGN'ASIS STRUCTURES SYSTEMS AND E UIPMENT FOR TURKEY POINT The design bases for structures at normal operating conditions are governed by the applicable building design codes.The design bases for specific sys-tems and equipment are stated in the appropriate FSAR section.The design bases for the containment structure are contained in Appendix 5B.The basic design criterion for the maximum hypothetical accident and earthquake condi-tions is that there be no loss of function if that function is related to public safety.I.Classes of Structures S stems and E ui ent Class I structures, systems and equipment are those whose fail-ure could cause uncontrolled release of radioactivity in excess of the established guidelines as prescribed in 10 CFR 100, those essential for immediate and long-term operation following a loss-of-coolant accident to either cool the core or reduce the contain-ment pressure~those required to function after a loss of power occurrence or steam line break to permit a controlled NSSS cool-dovn~or those required for a safe shutdown.Associated with Class I structures~

systems and equipment are their supports, enclosures~

piping, wiring, controls, power sources and switch-gear.They are designed to withstand the appropriate earthquake loads applied simultaneously with other applicable loads without 5A-1 loss of function.Rhen a system as a.whole is referred to as Class I the'portions not associated with the loss of function of the system may be designated as Class III as appropriate.

There are no components or structures designated as being Class II.The following are Class I structures, systems and equipment:

1~Reactor Coolant S stem Reactor vessel Reactor vessel internals RCC assemblies and drive mechanisms Steam generators Reactor Coolant pumps Pressurizer and relief tank All reactor coolant piping, plus any other lines carrying reactor coolant under pressure.2.Containment S stem Containment structure I Containment penetrations All lines penetrating the containment, up to and including the first isolation valves.30 Main Steam&Feedwater Hnes within the Containment 4.5~Main Steam Safet Isolation and Atmos heric Dum Valves New Fuel Stora e Facilities 6~Auxilia Feedwater S stem Auxiliary feedwater pumps and turbine drivers Condensate storage tank Steam, condensate and feedwater lines of auxiliary feed-water system 5A-2

'Emer enc Diesel Generators

-Da Tanks and Stora e Tanks 8.Containment Polar Crane and'Rail'u ort (Unloaded) 9.~'10;Refuelin Water Stora e Tanks Zmer enc Containment Coolin and Fil'terin Units 11.Intake Coolin Water S stems Intake structure and'rane support:s Intafce cooling water pumps and motors Intake cooling water piping, from pumps to component cool-ing water heat exchanger inlets.12.Com onent,Coolin S stem Component cooling heat exchangers Component cooling.pumps and'motors 4 Residual heat removal;pumps and motors (low-head safety infection pumps)Residual heat removal heat exchangers 13.Component cooling surge tanks S ent.Fuel Stora e Facilities Spent fuel pit and racks Spent fuel pit pump and motor Spent fuel pit heat exchanger, 14.Safet In ection S stem'ontainment spray, pumps and motors'ow-head safety in)ection pumps and motors (residual heat'removal pumps)High-head safety infection pumps and motors.Containment spray headers Boron in)ection,tank BOron tn)eLn tank aooonnlator Accumulator tanks Containment recirculation sumps 15~.Chemical and Volume Control S stem Charging pumps Volume control tank Boric'acid blender Boric acid tanks Boric acid transfer pumps Boric acid filters 16.Fuel Transfer Tube 17..Post Accident Containment Ventin S stem Piping within containment and to the second valve outside containment Desi n Bases a)Class I Structure Desi n 1)Normal Operation-For loads to be encountered during nor-mal operation, Class T.,structures are designed in accord-ance with design methods of:accepted standards,and codes insofar as they are applicable.

2)Hypothetical Accident,.

Wind and Earthquake.

Conditions-The Class X structures are proportioned to maintain elastic behavior.when sub)ected to various combinations of dead loads'ccident loads, thermal loads and wind or seismic loads.The upper limit of elastic behavior is considered to be the yield strength.of'he'effective load-carrying structural materials.

The yield strength for steel (includ-ing reinforcing steel)is considered to be the minimum as given in the appropriate ASSN Specification.

Concrete 4 structures are designed for ductile behavior whenever possible;that is, with steel stress controlling the design.The values for concrete, as given in the ultimate strength design portion of the ACI 318-63 Code, are used in determ-ining",Y", the required yield strength of the structure.

Limited yielding is allowable provided the deflection is checked to ensure'that the affected Class I systems and equipment (except reactor vessel internals under MHA load-ings)are not stressed beyond the values given below.The structure design loads are increased by load factors based on the probability and conservatism of the predicted normal design loads.The Class l structures outside the containment structure satisfy the most severe of the following:

Y~1/0 (1.25D+1.25E)"'i'~1/8 (1.25D+1.0R)1/g(1.25D+1.25H+1.25E)1/5 (1.0D+1.0E)where Y~required yield strength of the structures.

D dead load of structure and equipment plus any other permanent loads contributing stress, such as soil or hydrostatic loads.In addition, a portion of"live load" is added when such load is expected to be present when the unit is operating.

An allrmance is also made for future 5A-5 permanent loads.',~force or pressure on structure due to rupture of any one pipe.H force on.structure, due to restrained thermal expansion of pipes under operating conditions.

E~design earthquake load.E'~maximum earthquake load.,W~wind load (to replace E in the above load equations, whenever it produces higher stresses than E does).5~0.90 for reinforced concrete in flexure.5~0.85 for tension, shear, bond, and anchorage in reinforced concrete.g~0.75 for spirally reinforced concrete compression members.'g~0.70 for tied compression'embers.

5~0.90,for fabricated structural steel.b)Class I S stems and E ui ent Desi All Class X systems and equipment are designed to the standards, of the applicable Code.The loading combinations which are employed in the design of Class I systems and equipment are given in Table 5A-1.Table 5A-1 also indicate the stress limits which are used'n the~~design of the listed equipment for the various loading combinations.

To perform their function, i.e., alla@core shutdown and cooling, the reactor vessel internals must satisfy deformation lhnits which 5A-6

~~are more restrictive than the stress limits, shown on Table 5A l.For this reason the reactor vessel internals are treated separately.

Pi in and Vessels The reasoning for selection of.the load combinations and stress limits given in Table SA-1 is as follows: For the desi'gn earth-quake, the nuclear steam supply system is designed to be capable of continued safe operation, i.e.~for the combination of normal loads and design earthquake loading.Critical equipment needed for this purpose is required to operate within normal design i&nits.In the case of the maximum hypothetical earthquake, it is only necessary to ensure that critical components do not lose their capability to perform their safety function, i.e., shut the unit down and maintain it in a safe condition.

This capability is ensured by maintaining, the stress limits as shown in Table 5A-1.No rupture of a Class I pi'pe is caused by the occurrence of the maximum hypothetical earthquake.

Careful design and thorough quality control during manufacture and construction and'nspection during unit life, ensures that the independent occurrence of a reactor coolant pipe rupture is extremely remote.If it is assumed that a reactor coolant pipe ruptures, the stresses in the unbroken leg will be as noted in line 4 of Table 5A-1.5A-7 I(

TABLE SA-1 LOADING COMBINATIONS AND STRESS LIMITS 0 LOADING COMBINATIONS VESSELS-REACTOR COOLANT SYSTEM PIPIN REACTOR COOLANT SYSTEM OTHER CLASS I PIPING ormal Loads Pm-Sm L+B-15Sm C PL+PB S+p++g-S ormal+Design arthquake Loads Pm~S+P-1.5 S Pm 1.2 S L+PB-1.2 S Vp+'~g+<ad~i.2 ormal+Maximum otential Earth-'uake Loads P<1.2 S PL+PB+1.2 (1.5 S)P<1.2 S PL+PB~1.2 S (1)qp+Vg+Wsm<Sy ormal+Pipe upture Loads Pm 1.2 Sm r PL+PB~1.2 (1~5 Sm)Pm 1.2 S PL+PB~1.2 S Not applicable-See Pipe Restraint Criteria.Where: Pm~primary general membrane stress;or stress intensity PL=primary local membrane stress;or stress intensity PB primary bending stress;or stress intensity Sm~stress intensity value from ASME B&PV Code,Section III S~allowable stress from USAS'B31.1 Code for Pressure Piping longitudinal pressure stress P (j g~gravity-caused stress Ci sd seismic stress due to design earthquake sm seismic stress due to maximum potential earthquake Sy~Minimum yield strength at operating temperature Note (1)-This equation satisfies no loss of function criteria.Rev.1-3/16/70 5A-8 Reactor Vessel Internals Desi n Criteria for Normal 0 eration The internals and core are designed for normal operating conditions and sub)ected to 1'oad o'f mechanical, hydraulic,.and thermal origin.The response, of the structure under the design earthquake is included'n this category.The stress criteria established in the ASME Boiler and Pressure Vessel Code,'Section III, Article 4, have been adopted as a guide for the.design of.the internals and core with the exception of those fabrication techniques and materials which are not covered by the Code.Earthquake stresses are combined in the.most conservator;ve way and are considered primary stresses.The members are designed under the basic principles of: (1)maintaining distortions within acceptable limits, (2)keeping the stress level's within acceptable limits, and (3)prevention of fatigue failures..

Seismic A~nal sis of Reactor Internals The maximum stresses are obtained by combining.

the contributions from the horizontal and vertical earthquakes in the.most conservative manner.The following paragraphs describe the horizontal and vertical contributions.

The reactor building.with the reactor vessel support, the reactor vessel,.and the reactor internals are included in this analysis.The mathematical model of the building, attached to ground, is similar to that used to evaluate the building structure.

5A-'9.((Rev.1-3/16/70 The reactor internals are mathematically modeled by beams, concentrated masses, and linear springs All masses, water, and metal are included in the mathematical model.All beam elements have the component weight or mass distributed uniformly, e.g., the fuel assembly mass and barrel mass.Additionally, wherever components are attached somewhat uniformly their mass is included as an additional uniform mass, e.g.,'baffles and formers acting on the core barrel.The water near and about the beam elements is included as a distributed mass.Horizontal components are considered as a concentrated mass acting on the barrel.These concentrated masses~also include components attached to the horizontal members since this is the media through which the reaction is transmitted.

The water near and about these separated components is considered as.being additive at these concentrated mass points.The concentrated masses attached to the barrel represent the following:

a)the upper core support structure, including the upper vessel head and one-half the upper internals; b)the upper core plate, including one-half the thermal shield and the other half of the.upper internals; c)the lower core plate, including one-half of the lower core support columns;d)the lower one-half of the thermal shield;and e)the lower core support, including the lower instrumentation and the remaining half of the lower core support columns.The modulus of elasticity is chosen at its hot value for the three ma)or materials found in the vessel, internals, and.,fuel assemblies.

In considering shear deformation, the appropriate cross-sectional areas are selected along with a value for Poisson's ratio.The fuel assembly moment of inertia is Rev.1-3/16/70 5A-9a Ae~~derived from experimental results by static and dynamic tests performed on fuel assembly models.These tests provide stiffness values for use in this analysis.The fuel assemblies aie assumed to act together and are represented by a single beam.The following assumptions are made in regard to connection restraints.

The vessel'is pinned to the vessel support which i's the surrounding concrete structure and part of the containment building.The barrel is clamped to the vessel at the barrel flange and spring connected to the vessel at the lower core barrel radial support.This spring corresponds to the radial support stiffness for two opposite supports acting together.The beam representing the fuel assemblies is pinned to the barrel at the locations of the upper and lower core plates.Modal analysis, plus the response spectrum method is used in (1)this analysis.The modal analysis is studied by the use of a transfer matrix method., The maximum deflection, acceleration, etc., is determined at each particular point by summing the absolute values obtained for all modes.With the shear forces and bending moments determined, the earthquake stresses are then calculated.

Figure 5A-3 shows the mathematical model studies.The reactor internals are modeled as a single degree of freedom system for vertical eathquake analysis.The maximum acceleration at the vessel support is increased by the amplification due to the building soil interaction.

(1)Shock and Vibration Handbook, edited by Harris and Crede, Volume 3, Chapter 50: "Vibration of Structures Induced by Seismic Waves" by George W.Housner.5A-9b Rev.1-3/16/70 Desi n Criteria for Abnormal 0 eration The abnormal design condition assumes blowdown:effects'ue to a reactor coolant pipe double-ended break.For this condition the criteria for acceptability are that the reactor be capable of safe shutdown and that the engineered safety.features are able to operate as designed.Consequently, the limitations established, on the internals for these types of'oads are concerned principally with the maximum allowable deflections.

The deflection criteria for critical maxima under abnormal operation are presented in Table 5A-2.5A-Qc Rev.1-3/16/70 TABLE 5A-2 INTERNALS DEFLECTIONS UNDER ABNORMAL OPERATION (Inches)/ion (to assure sufficient inlet flow area/and'o prevent the barrel from touching any guide tube to avoid disturbing the RCC guide structure).

Calculated Deflection Prelimina 0.072 Allowable Limit No Loss-of-Function Limit'6 U er Packa e, axial deflection (to maintain the control rod guide structure geometry).

0.005 CC Guide Tube, cross section distortion (to avoid interfer-ence between the RCC elements and the ides.0 0.0035 0.072 CC Guide Tube, deflection as a beam (to be consistent with conditions under which ability to trip has been tested).0.2 1.0 1.5 el Assembl Thimbles, cross ection distortion (to avoid nterference between the control ods and the guides)0 0.035 0.072 c)Mind and Earth uake Loads for Class I Structures S stems and~cCu~iment The wind loads are determined from the fastest mile of wind for a 100-year occurrence as shown in Figura 1(b)of Ref-erence 4.This is 122 mph at the Turkey Point site.The Class I structures're designed, however, to withstand a 5A-10 Rev.1-3/16/70 wind velocity at 145 mph.In addition, Class I structures are designed to resist the effects of a tornado.C Inadings due to a tornado to be used in the design of tornado-resistant structures are as follows, the loads to be applied simultaneously:

a.Differential pressure between inside and outside of b.Co enclosed areas-1.5 psi (bursting).

External forces resulting from a tornado wind velocity of 225 mph.Missiles as.defined in Appendix 5E.The forces due to the wind are calculated in accordance d with methods described in ASCE Paper No.3269 entitled,"Rind'orces on Structures"~Applicable pressure and shape coefficients are used.There is no variation with height or gust factor.The forces resulting from a tornado are combined with dead loads only.Dead loads include piping and all other perman-ently attached or located items.There will be sufficient time after sighting a tornado to remove significant live loads such as loads on cranes.Allowable stresses are limited to yield strength for struct-'"ural steel and reinforced concrete.Local crushing of con-crete is permitted at the missile impact zone.In all 5A-11 Rev.20-12/21/71'1 cases~structures are reviewed to assure no loss of func-tion for a tornado wind of 337 MPH combined with a pressure differential of 2.25 psi.2)Earth uake Forces E and E'EC Publication TZD 7024,"Nuclear Reactors and Earth-quakes"~as amplified in this Appendix is used as the basic design guide for earthquake analysis.Earthquake loads on structures, systems and equipment are determined by realistic evaluation of dynamic properties and the accelerations from the attached acceleration spec-trum curves.These spectrum curves are corrected for the design ground accelerations.

Damping factors are listed in the table belier.Earthquake forces are applied simultaneously in the vertical and any horizontal direction.

The vertical component of acceleration at any level is taken as two-thirds of the horizontal ground acceleration.

5A-12 I~(

~~DAMPIN CTORS FOR VARIOUS TYPES OF CO RU ION 7.Critical'Dam in Design Earthquake (E)Maximum Earthquake (E')(0.05g, Ground Surface (0.15g Ground Surface Acceleration Acceleration Welded, Steel Plate Assemblies Welded Steel Framed Structures

.2 Bolted S'teel Framed Structures'oncrete Equipment Supports on Another Structure Prestressed Concrete Containment

'Structure 2 Soil 10 Prestressed Containment Including Interior Concrete and Soil Composite 3.5 7.5 Reinforced Concrete.Frames and Buildings 3 Composite with Soil 5'-Steel Piping 0.5 d)Class III S stems and Enui ent Desi n 7.5 0.5 Class XII systems and equipment including pipe are not designed'to with-stand, any earthquake loads.The wind loads are as per South Florida Building Code which has a basic design pressure of 37 psf.Shape Factors are applied in accordance with the Reference 4.No tornado loads are considered.

e)Miscellaneous Loads: 'The units are designed for a temperature rang of+30F to+95F.No ice or snow loads are considered in the design of the various struc-tures and equipment.

The unit is'designed for a hurricane tide to an elevation of+20', with wave run up to an elevation of+22.5.'n.the east side of the unit.5A-1'3 Rev..1-.3/16/.70

~~4 The protection is afforded by a continuous barrier sisting of O building walls, florio walls, a flood embankment as shown in Fig.1.2-3.(g Door openings are protected by stop logs.The intake cooling water pumps located at the Intake Structure are protected'y thei'r elevation.

Flooding from rain water is prevented by an elaborate system of storm drains, catch basins, and sump pumps.All outdoor equipment is de-signed for such service.III.Hethod of Seismic Anal sis The method of seismic analysis for the containment structure is described in section 5.1.3.2(b).

Response spectrum curves are also generated for the control building.Response curves for floors at grade and for basement are as shown in Figures 5P-1 and~5A-For class I piping, floor response spectra for the connecting points are developed by the technique described in section 5.1.3.2.The pipe.loop itself is also idealized as a mathematical model consisting of lumped masses connected by elastic members, and.the frequencies and mode shapes for all significant modes of vibration are determined.

The distance from the pipe axis to the center of gravity of the valve and, operator is considered, with the mass of the valve and operator, for al'1 motor, air,'or gear operated valves.When necessary for the integrity of the piping', valve, or operation, the valve structure is ex-ternally supported.

The flexibility matrix for the pipe is developed to include the effects of torsional, bending, shear and axial deformations as well as change in flexibility due to r curved members and internal pressures.

Flexibilitv factors are calculated in accordance with USAS 831.1.The spectral ac-celeration is determined from the response spectra.Rev.1-3/16/70 2-6/26/70 11-2/25/71 5A-14 The following equations are successively used to determine the response for each mode, maximum displacement for each mode, and the total dis-placement for each mass point: Yn max R San D M w 2 n n in which: Yn max~response of the n mode th R=participation factor for the n mode~Z Mi th n i in Sa~spectral acceleration for the n mode th n D~earthquake direction matrix M~generalized mass matrix for the n mode'Z Mi th'2 n in (2)Vin~in Yn max in which: Vi=maximum displacement of mass i for pode n in (3)Vi~ZV in in which: Vi~maximum displacement of mass i due to all modes calculated The inertial forces for each direction of earthquake for each mode are then determined from: 5A-15 Rev.2-6/26/70 in which: Qn~inertia force matrix for mode n.V~displacement matrix corresponding to gn EacF mode's contribution to the total displacements, internal forces, moments and reactions in the pipe can be determined from standard structural analysis methods using the inertia forces for each mode as an external loading condition.

The total combined results are obtained by taking the square root of the sum of the squares of each parameter under consideration, in a manner similar to that done for displacements.

A representative number of critical piping runs have been analyzed by this method..Balance of the pipe runs have been evaluated.

by (i)closeness of similarity to the runs'fully analyzed, (ii)simplicity of layout lending to a visual examination for location of seismic restraints to remove the fundamental frequency away from the resonance range, and,-.(iii)Static analysis based on a uniform static load equal to the peak of the pertinent response spectrum curve.Electrical cable trays and D-C battery racks are being checked for'g'oading obtained from the spectrum curves of the supporting floors.Motor Control Centers and Load Centers have been shaker-table tested to demonstrate no-loss-of-function capability under the maximum hypothetical earthquake.

For additional information on instrumentation, see page B-37 in response to Request No.7.3.Rev.2-6/26/70 5A-16 Mechanical and electrical equipment has been purchased under specifi-cations that include a description of, the seismic design criteria for the plant.I Hydrodynamic analysis of the Refueling Water, Storage Tank has been performed using the methods of chapter'6 of:the U.S.Atomic Energy Commission

-TXD 7024..5A-17 Rev.2-6/26/70 9-11/24/70 3.Seismic.Loads The reactor coolant loop (RCL)which consists of the reactor vessel (RV), steam generator (SG), reactor coolant pump (RCP), the pipe con-necting these components, and the large component supports has been analyzed for seismic ldads., The components and piping are modeled as a system of lumped masses connected by springs whose'alues are com-puted from'elastic properties that are input.A simplified support model was arrived at by representing the structural support system as equivalent springs rather than as member beams and columns.The analysis was performed by using a proprietary computer code called WESTDYN.The code uses as input, system geometry, inertia values, member sectional properties, elastic characteristics, support and re-straint characteristics, and the appropriate seismic floor response spectrum fo 0.5%cr tical damping.The floor response spectrum k curves were generated at the appropriate support locations of the 4 equipment by a time history technique described in Section 5.1.3(b).Both horizontal and vertical components of the seismic response spectrum are applied simultaneously.

Two directions, namely X and Z axes, were chosen for application of the horizontal component of the seismic response spectrum.The results of the two cases were combined to determine the most severe loading condition.

With this input data, the overall stiffness matrix fK)of the three dimensional piping system is generated (including translational and rotational stiffnesses).

Zero rows and columns representing restraints are deleted', and the stiffness matrix is inverted to give the flexi-bility matrix)F3 of the system.Rev.8-11/6/70 9-11/24/70 5A" 18~'\I ,\

A product matrix is formed by the multiplication of.the flexibility and mass matrices.This product matrix forms the dynamic matrix, fDg from which the modal matrix is computed.fD)-CF3 BQ The eigenvalues and eigenvectors representing the frequency and associated mode shape for each mode are generated using a modified Jacob i method.(W.fM]-[K3){X)0 Prom this information, the modal participation factor is combined with'the mode shapes and the appropriate seismic response spectrum values to give t'e structural response for each mode.Then the forces, moments, deflections, rotations, constraint reactions, and stresses are calculated I'or each significant mode.The maximum response of the system is obtained by combining the modal contributions using the root mean square method.The restraints, supports, and other constraints assumed for input into the seismic computer model are given below (see Figure 5A<<4 for axes 7 orientation..)

.Reactor Vessel Steam Generator The RV is rigid.The SG at the upper support point is permitted to translate along and rotate about the X, Y, and Z axes, but translations along X and Z are resisted by the springs representing the upper support.5A-19 Rev.8-11/6/70 9-11'/24/70 The SG at the lower support point is permitted to translate along and rotate about the X, Y, and Z axes, but all movements are resisted by springs representing the lower supports stiffness.

Reactor Coolant Pump The RCP is permitted to translate along ind rotate about the X, Y and Z axes, but all movements are resisted by springs representing the supports stiffness.

A susemzJJ og smxgmum pgpe stresses fs given 1n Table 5A-3.'ev.8-ll/6/70 9-11/24/70 5A-20

(%0 L.h 0 Op s L a py~V s.t 0'I I a~V SQ o 0 gJ Et.O 0'OV'b, aha 4 4 0 0 R cb VELOCITY>lH/5EC.0 0 DAMPED RESPONSE SPECTRA 5'/, ACCELERATION (I Fi.pure 5h-1

~a r.e~Q u O 0 I l l I I'~~!..'3<r)I}~tr<<}I I!),LP e/C<<I'4.:., i I X~4 y", I/I I I}I j/~e~~o!0 o o I tO O O/I I u>>~~<<'~YI~~I~Q r'9~<<, I~-j I~I oy~.)o;y/C" 4 I.~~~~I 4/I.~'rK.i"<4~4}L bo I~I I<<1/p~rj!r.'y~<<gC....>>9 r I I"u!e Q r~I O CL CP.o~U.X o}U o 6 O D o!0 I~I l 0 O 0 0 I I I~I I I j eg~I O 4 I~~I/I o o q ci 0 v g4 VE LC77" I T+I, iaJ./oopc, e~I I h~~o 0 0~e V, 0 4 QI 4 0~4 44.j~4 O O DAMPED RESPONSE SPECTRA I5'/, ACCELERATION Figure 5A-2 LEGEND: k~radial support spring constant rs k~rotational ground spring constant e k~translational ground spring constant~concentrated masses D distributed masses VESSEL BUILDING BARREL NOTE: THIS FIGURE IS ONLY ILLUSTRATIVE FUEL ASSEMBLIES rs k MATHEMATICAL MODEL FOR REACTOR VESSEL INTERNAL ANALYSIS (HORIZONTAL EXCITATION)

FIG.,'5A-3 REV.1-3/16/70

~I e<STEet CENERATOR 4 21 REACTOR VESSEL REACTOR COOLANT PUHP~~W CO lpt I t hJ Nv~0 Og g 0 0 Ogch 0 H R O 0 REACTOR COOLANT LOOP TABLE 5A-3 MAXIMUM STRESSES EXPECTED"IN REACTOR COOLANT SYSTEM PIPING DUE TO THE OPERATING.05 EARTH UAKE Location Reactor Cool'ant'Pump Inlet Reactor Coolant Pump Outlet 10 Inch Accumulator Line Steam Generator Outlet Reactor Vessel'nlet Reactor Vessel.Outlet Pressurizer Surge Line Connection Steam Generator Inlet Maximum Stress si 4085 3616 3201 2274.1289 182 78 71 Maximum Allowable Seismic Stress~13,125 psi (This value is the.result, after deadweight and pressure stresses have been subtracted from 1.2 times the material allowable stress.)Rev.9-ll/24/70 5.0 Structures REQUEST: 5,1 In reference to.Table 5A-1;,which contains loading'ombinations and stress limits..for Class.I systems and equipment'esign; provide the following information not presently included.therein: a+Stress limits for normal reactor'.operating, conditions.plus.faulted reactor operating conditions, i;e., normal+pipe-rupture'+

design basis earthquake loads.b.C~Quantitative stress limits under all loading combinations for equipment supports.Relate these limits to the requirements of paragraph N-473 of ASIIE Code Section III..,Identify.the source of and the bases for the piping'stress'imits:

p~1.2S, Pl+PB 1.2S, and Pl+PB~1.2 (1.5S)m RESPONSE: a.Refer to FSAR page 5A-7 and Table 5A-1 (Rev.).b.There is no relationship c.Refer to ASA 831.2 1955.REQUEST: 5.2 Identi'fy the issue(s)of the ASHE Boiler'and Pressure Vessel Code Section.III and addenda thereto spec'fied for the design and fab'rication of all applicable Class I components for the Turkey Point Units;Indicate any differences which exist between Units 3 and 4...RESPONSE:

1964 and addenda thereto through October 1965.REQUEST: 5.3'ailure of the bearings on'primary pump shaft could conceivably lead'o the generation'of missiles due to flywheel breaI:up.Provide the.results of an analysfs of the effects of applicable load combinations, including seismic.loads on the'bearings, and'indicate the margins against failure.RESPONSE: Refer to answer 4.2.3-.Con Ed Docket 50-247.B-28 Rev.1-3/16/70 2-6/26/70

~REQUEST: 5.8'Appendix 5A indicates that.*various..structures;-,systems"and-equipment,'because"of-their.

special'mportance-to-.pub).ic-safety;.

are"designed and built to more-exacting standards; thanmould-otherwise-be-necessary for" reliab3.e plant-operation'lone:-.

'lease-describe; in-detail, the'management reviews"apd approvals required'in-determining which-portions of.the plant-must be'f this.higher classification.

C RESPONSE: The design standards and criteria selected for safety related portions of the plant were subjected to the same management review chain as shown in Figure 1.9-1 for Quality Assurance.

Also, consultants (refer to FSAR page 1.7-'1)were engaged.Safety Analyses Reports of other projects were used during the design as references for those evaluating the Turkey Point design.REQUEST: 5.9 During the construction of the facility corrective concrete work was undertaken around the'tendon an-chorages.Provide a summary description of the out-come of this work including a discussion of any post-repair tests that were performed.

tAat special atten-tion will be given to monitoring for distress or failure in tendon bearing plate areas during tendon tensioning and containment structural proof testing?RESPONSE:.

Refer to reports to DRL of August 14, 1968, November 1, 1968, and February 9, 1970.REOUEST: 5.10 Provide information on the as-built foundations regarding any unexpected foundation conditions'encountered and any.changes brought about by these cond'itions, such.'s changes in elevation, types of foundations, and grouti'ng.

RESPONSE: None.B-30 Rev.1~~

v'EQUEST: 7.3 What are your seismic design bases for'the.reactor protection system, the emergency'lectric power syst: em,'nd the instrumentation and control'for both'he engineered safety features and'he decay'heat'emoval system?!Will the systems be designed to be capable:of:

actuating'eactor trip or engineered safety feature action during the maximum peak acceleration?

If a seismic'disturbance occurred after a ma)or accident, would emergency core'ooling be'nterrupted?

What tests and analyses will'be performed to assure that the seismic design bases are met?What seismic'specifica-tions are employed'n the instrumentation and control purchase order(s)?RESPONSE: The Westinghouse design bases for the protection grade equipment with respect to earthquakes

'is that, for design basis earthquake

'(DBE)or operational basis earthquake (OBE), the equipment will be designed to ensure that such equipment will not lose its capability to perform its design objective; namely, shut the unit down and/or maintain the unit in a safe shutdown condition.

It is conceivable that protection grade equipment may have permanent deformation due to stresses from the maximum potential earthquake; as such, the deformation will not impede its design objective.

If a seismic disturbance occurs subsequent to an accident, the instrumentation and electrical equipment associated with emergency core cooling will not be interrupted during this disturbance.

The manufacturer states that the 4160V.switchgear, including breaker contacts, instruments and relays will withstand, and still remain operable, a seismic acceleration force of 3G in any direction.

The manufacturer of the 480V.power supply equipment made tests as follows'.480V.Motor Control Center including starters, circuit breakers and relays-successful operation at accelerations of 0.5G to 1.25G at fundamental frequencies of 4 to 10 cps with shocks applied front to back and side to side.2.480V Switchgear including circuit breakers, relays and instruments-successful operation at accelerations of 0.'5G to 3.0G at fundamental frequencies of 4 to ll cps with shocks applied front to back and side to side.B-37 Rev.1-3/16/70 2-6/26/70 4-8/12/70 v 3.Power transformers withstand impacts of 4G in the vertical direction and 6G in the horizontal direction, as evidenced by impact meter.records.Mathematical models are not used for seismic design evaluation of instru-mentation.

Evaluation of such, equipment for its ability to withstand the seismic condition in accordance with the design ob')ective is done by actuil vibration type testing of typical protective grade equipment.

Documentation of the test program results is.available in a Westinghouse proprietary documentSupplement 2 to WCAP 7397-L, Seismic Testing of Electrical and Control Equipment (WCID Process Control Equipment), E.L.Vogeding, January 1971.No seismic specifications are employed in the instrumentation and control purchase orders.Type testing was reported as described'bove (WCAP 7397-3.)to provide verification of the seismic design objectives.

B-37a~~Rev.1-3/16/70 2-6/26/70.4-8/12/70 11-2/25/71.

9.0 Auxil'iar S stems REQUEST: 9.1 Provide a PE:I diagram of the'ntake cooling water system.RESPONSE: Refer to FSAR Figure 9.6-2 (new).REQUEST'9.2 Describe'he.

applicable codes.and'tandards to which'he piping and components of'he'ntake cooling water and'auxiliary-feedwater'ystems are designed.RESPONSE: Refer to FSAR Tables 9.6-.2'new), and 9.11-2 (new).REQUEST: 9;3 Describe the design features which will:prevent; loss'f.the'uel pool water as a result.of tornado generated winds or'missiles;,'main'urbine missiles or a'ropped fuel cask.Mhat'eans:are provided to maintain adequate cooling of stored fuel in the event fuel pool'water should'e lost'ESPONSE:

Refer to'SAR Sections 5.2 and 9.3.B-42.Rev.1