ML19316A219

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Analysis of Effects Resulting from Postulated Piping Breaks Outside Containment.
ML19316A219
Person / Time
Site: Oconee  Duke Energy icon.png
Issue date: 04/25/1973
From:
DUKE POWER CO.
To:
References
05-27B, 5-27B, OS-73.2, NUDOCS 7912030337
Download: ML19316A219 (100)


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      -            ANALYSIS OF EFFECTS RESULTING FROM POSTULATED PIPING BREAKS OUTSIDE CONTAINMENT                                                                                           l FOR OCONEE NUCLEAR STATION, UNITS I, 2 s 3 a

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                                                                               .                                                h DEPARTMENTS CHARLOTTE, NORTH CAROLINA APRIL 25, 1973 i,o 9

FILE NO. OS-27B ,./e' f . o MDS REPORT NO. 05-73 2 [s,y cE,m, %b 2 APR 271973 > g l H2

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f 1 TABLE OF CONTENTS Section Page

1.0 INTRODUCTION

AND PURPOSE 1.1-1 1.1 HISTORY OF OCONEE DESIGN 1.1-1 1.2 DESIGN OF HIGH ENERGY SYSTEMS 1.2-1 1.2.1 GENERAL DESIGN PROCEDURES FOR HIGH ENERGY SYSTEMS 1.2-1 1.2.2 MAIN STEAM AND FEEDWATER SYSTEM DESIGN 1.2-2 1.2.3 QUALITY OF MAIN STEAM AND FEEDWATER SYSTEMS 1.2-3 2.0 DESIGN ANALYSIS OF HIGH ENERGY SYSTEMS 2.1-1 2.1 GENERAL REVIEW CRITERI A 2.1-1 2.1.1 POSTULATED PIPING BREAK LOCATIONS 2.1-1 2.1.2 POSTULATED PIPING BREAK SIZES 2.1-2 2.1 3 LINE SIZE CONSIDERATIONS FOR POSTULATED PIPING 2.1-2 BREAKS 2.2 ANALYTICAL CONSIDERATIONS FOR POSTULATED PIPING 2.2-1 BREAKS 2.2.1 STRESS CRITERIA 2.2-1

                                                                ^

2.2.2 DYNAM.*. ANALYSIS 2.2-2 2.2.3 STRUCTURAL ANALYSIS 2.2-3 2.3 CONSEQUENCES OF POSTULATED LINE BREAKS 2.3-1 2.3.1 ENVIRONMENTAL EFFECTS 2.3-1 2.3.2 POTENTIAL PHYSICAL DAMAGE TO STATION 2.3-2 30 OPERATIONAL ANALYSIS OF STATION 3.1-1 3.1 OPERATIONAL STATUS AND MITIGATION OF ACCIDENT 3 1-1 RESULTING FROM POSTULATED PIPING BREAK 3.1.1 MAIN STEAM SYSTEM , 3 1-1 3.1.2 MAIN FEEDWATER SYSTEM (TURBINE BUILDING) 3.1-3 3.1 3 MAIN FEEDWATER SYSTEM (PENETRATION ROOM) 3 1-4

,    i TABLE OF CONTENTS Section                                                         Page 3.1.4          AUXILIARY STEAM SYSTEM                           3.1-5 3 1.5          TURBINE EXTRACTION SYSTEM                        3.1-5 3.1.6          CONDENSATE SYSTEM, BOOSTER PUMP DISCHARGE        3.1-5 3.1.7          FEEDWATER HEATER DRAINS AND VENTS                3.1-5 3 1.8          HIGH PRESSURE INJECTION SYSTEM, SEAL SUPPLY AND  3 1-5 NORMAL MAKEUP 3.1.9          HIGH PRESSURE INJECTION SYSTEM, LETDOWN          3 1-6 3 1.10         EMERGENCY FEEDWATER SYSTEM, PUMP DISCHAkGE       3.1-6 3 1.11         EMERGENCY FEEDWATER SYSTEM, PUMP DISCHARGE TO    3.1-6 STEAM GENERATORS 3 1.12         HIGH PRESSURE INJECTION SYSTEM, PUMP C DISCHARGE 3.1-7 3.1.13         LOW PRESSURE INJECTION SYSTEM                    3.1-7 3 1.14         REACTOR BUILDING SPRAY SYSTEM                    3 1-7 3.2        SUHMARY OF POSTULATED BREAK OPERATIONAL ANALYSIS     3 2-1 4.0     PROTECTIVE METHODS AND STATION MODIFICATIONS FOR        4.1-1 POSTULATED PIPING BREAKS 4.1         INTERIM MEASURES                                    4.1-1 4.2        DESIGN CHANGES                                       4.1-2 4.3        EMERGENCY OPERATING PROCEDURES                       4.1-2 4.4        STATION MODIFICATIONS SCHEDULE                       4.1-2 50      

SUMMARY

5,o_] TABLES FIGURES 3 I I

1.0 INTRODUCTION

AND PURPOSE On December 15, 1972, the Atomic Energy Commission forwarded Duke Power Company a letter requesting additional information on the consequences of postulated piping breaks outside containment. . Basically, Duke has been asked to assure - that the reactor can be safely shut down and maintained in a safe shutdown con-dition subsequent to such a postulated accident. On January 18, 1973, Duke met with the AEC and explained the results of prelimi-nary studies to that time. On January 26, 1973, Duke formally responded to the January 23, 1973, request f rom Mr. Al Schwencer concerning interim measures to be implemented until necessary design changes and construction could be completed. The purpose of this report is to describe the analysis conducted, the conclusions reached and the proposed design changes and operational measures to be taken spect-fically in response to the request for additional information. 1.1 HISTORY OF OCONEE DESIGN

   -   t onee Nuclear Station utilizing three B & W pressurized light water reactors with accompanying 886,000 kwe GE turbine generators is located in Oconee County, South Carolina, near Clemson, South Carolina.

The station is a part of Duke Power Company's Keowee-Toxaway project made up of Keowee Hydro Station (2 - 140,000 kw uni ts) and the Jocassee Pumped Storage Hydro Station (4 - 152,500 kw units). Application to construct the Oconee Nuclear Station and the Preliminary Safety Analysis Report were submitted to the Atomic Faergy Commission on November 28, 1966, with the Construction Permit being eJ Jn November 6,1967 The Final Safety Analysis Report was submitted to the Atomic /nergy Commission on January 2,1969, and the Operating License for Unit I was g 9nted on February 6, 1973 During the detailed design and construction period of 1966 until the present, the station has undergone the influences of major new and evolving design cri-teria with success but not without delay. In every instance, Duke has responded positively and adequately to significant analysis requests such as the one repre-sented by this report.  ; l l 1 l l 1.1-1 l

1.2 DESIGN OF HIGH ENERGY SYSTEMS Recognizing reliability requirements and perforr.ance aspects of high energy power piping systems, Duke conservatively designs these systems as described in 1.2.1. The Main Steam and Feedwater System design is discussed in additional detail in 1.2.2. 1.2.1 GENERAL DESIGN PROCEDURES FOR HIGH ENERGY SYSTEMS High energy power piping system design for the Oconee Nuclear Station was accomplished by completing the following sequential design process:

a. Establish design criteria for the system including AEC General Design Criteria, as required considering physical separation, missile protection, seismic protection and control room integrity, as applicable.
b. Classify the system based on design criteria described above. Clas fication defines the applicable power piping code and design requireinents.

Table IC-1 of the Oconee Nuclear Station, Final Safety Analysis Report (FSAR), as follows, was utilized for the Oconee design. Table IC-1 System Piping Classification Designed For Piping Class Design Criteria Seismic Loading A Class I, USAS B31.7 Yes B Class 11, USAS B31.7 Yes C Class ill, USAS B31.7 Yes D Class 11, USAS B31.7 No E Class Ili, USAS B31.7 ifo F USAS B31.1.0 Yes G USAS B31.1.0 No H Good Industry Practice No

c. Lay out and route the high energy system in accordance with the design criteria defined by a. above, giving proper attention to support- j restraint requirements and flexibility considerations.
d. Prepare, review and approve material specifications for the system in accordance with applicable Codes and Cuke's quality Assurance Program.
e. Review and approve the system for functional operability.
f. Review and approve the system for dimensional Integrity.
g. Analyze, review and approve the system for conformance with stress criteria in areas of:

i) su ' orts and restraints, II) ;4ermal flexibility, and , iii) seismic operability. l.2-1 l

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  • l l
h. Procure materials and/or shop-fabricated subassemblies from suppliers who have been previously qualified under Duke's Quality Assurance Program.

All high energy systems including the Main Steam and Feedwater systems are designed to preclude pipe ruptures based on conservative engineering practices; and although this report reviews the consequences of postulated pipe ruptures, such ruptures are not considered credible. 1.2.2 MAIN STEAM AND FEEDWATER SYSTEM DESIGN in addition to meeting the General Design Procedure described in 1.2.1 above, supplementary design information for the Main Steam and Main Feedwater Systems is described In'this portion of the report. Figures 1.2-1 and 1.2-2 are sim-plified Isometrics of the safety-related portions of the Main Steam and Main Feedwater Systems. The primary basis for postulating a line rupture is stress level. The following describes representative stress conditions for the Main Steam and Main Feedwater Systems at Oconee.

a. Main Steam lines are 100 percent cold pulled so that as the line heats up, all thermal expansion stresses are essentially eliminated throughout the system. For example, at the Reactor Building penetration (terminal end), there will be only 1100 psi maximum thermal stress during normal operation; this is only about 4 percent of the ANSI B31.1.0 (1967) Code allowable stress. This fact coupled with the scfety factor built into Code allowable stress values indicates a tremenduas amount of conserva-tism. The Main Feedwater System is not cold pulles' since it operates at a lower temperature; however, there will be only 364.' psi maximum thermal stress during normal operation at the Reactor Building penetration (terminal end). This is only about 16 percent of the ANSI B31.1.0 (1967) Code allowable stress.
b. Overpressure capability of the piping based on actual wall thicknesses is as follows:

Normal Operating Actual Code Pressure Percent Pressure Capability Margin Main Steam: 910 psig 1093 psig 20 Feedwater: ic70 psig 1383 psig 29 it should be noted that these figures are highly conservative being based on ANSI B31.1.0 (1967) Code equations. The above is presented in terms of the original design for illustrative purposes. In accordance with the AEC Guidelines transmitted to Duke on December 15, 1973, the following stress analysis summary has been prepared which clearly demonstrates the acceptability of the design and the low probability of a postulated piping break. In all cawss. the maximum combination of stresses was less than the allowables established in tne AEC Guidelines and the following summary represents the maximum stresses for both systems: - 1.2-2

S S3 (Max) Sp + S s(Allow) Percent ST(Max) S7 (Allow) Percent System pps+i psi Margin psi psi Margin Main Steam: 11,295 29,375 71 1,100 21,000 95 Feedwater: 23,963 33,750 29 16,417 18,000 9 The above Sp+Ss (Allow) and ST (Allow) established by the AEC Guidelines were calculated f rom the equations described in detail in 2.2.1 of this report for Class F systems. 1.2.3 QUALITY OF MAIN STEAM AND FEEDWATER SYSTEMS Safety-related portions of the Main Steam and Main Feedwater systems are Duke Class F (see 1.2. lb. , above) . Materials for construction of Class F systems were procured, fabricated, tested, and documented similar to a nuclear system as follows:

1) P i p i ng Ma ter i a l s-- -----------------Traceab l e li) Welding Filler Metal----------------Traceable lii) Non -Des tructive Examination--------- 100 percent X-ray Iv) P i ping Ma teria l s QA-----------------i nspect ion a t fab ri ca tors plant and site receiving inspection v) Documentation-----------------------Required vi) Support Design QA-------------------As outlined in FSAR IC.3.4.5 and described below.

Proper piping system erection and function of safety-related hangers, supports and restraints are assured by several means:

1) The Construction Department reviews erection against design drawings.
11) For hangers inside the Containment, an additional QA surveillance of the system and supports is made for proper function by the engineer responsible for the design.

III) QA surveillance is conducted by the Hanger-Contractor on al.1 safety-related systems to verify correct location, direction of movement, and proper hardware installation. Detailed and specific design information provided in Tables 1.2-1 and 1.2-2 is presented to describe the high quality of the Main Steam and Main Feedwater Systems; however, these tables are also -typical for all Duke Class F systems. 1.2-3

2.0 DESIGN ANALYSIS OF HIGH ENERGY SYSTEMS This section describes in detail the engineering review , analysis and results o' analysis conducted for the Oconee Nuclear Station high energy systems. 2.1 GENERAL REVIEW CRITERIA All high energy 'aechanical piping systems outside the Reactor Building were reviewed and co9sidered in the analysis with respect to a postulated pipe break. Those high ene' gy lines which are safety-related or which pass near safety-related struccures, systems or components throughout the station were of primary l concern. High energy lines are defined as those which have either or both:

a. A normal service operating temperature greater than 200*F
b. A normal service operating pressure greater than 275 psig  !

Specific high energy line postulated breaks and associated engineering data  ! are tabulated in Table 2.1-1 for those systems which are normally in operation. Specific high energy line postulated breaks and associated engineering data are tabulated in Table 2.1-2 for those systems which are not normally in operation. In general, these systems are: i) safc'.r-related systems which are tested on a scheduled basis, Ili systems required in the startup or shutdown of the unit or both, or 5 i) systems required to operate while the unit is in a shutdown condition. As these systems are not normally in operation, it should be recognized that the probability of a postulated pipe break is even more remote than for those systems normally in operation. Reactor Building integrity with respect to postulated pipe breaks is assured in that the. mechanical penetrations and the Reactor Building were designed for the imposed loading of postulated pipe breaks. 2.1.1 POSTULATED PIPING BREAK LOCATIONS Systems identified as containing high energy piping as described in Tables 2.1-1 and 2.1-2 were examined by a detailed design drawing review for a postulated pipe break along their entire routing regardless of Code class. Specifically, all term!aal ends and intermediate locations (minimum of 2) between terminal ends were reviewed for the consequences of a pipe break. Figures 2.1-1.a thru 2.1-24 were prepared based on the above criteria for this report and correspond to Tables 2.1-1 and 2.1-2 by postulated break case numbers. Since all'iocations of consequence were reviewed and as detailed stress analysis information is extensive, stress analysis information was only reviewed for spec ial identified problem areas which might require additional restraints be placed on the system. Safety-related portions of the McIn Steam and Main Feed-water Systems were reviewed in detail against existing stress analysis information. 2.1-1

Types of breaks postulated for this analysis were:

a. Double ended breaks and equivalent area longitudinal breaks occurring at:
1) Terminal ends ll) Butt weld joints of ells, tees, laterals, etc.

lii) Nozzle weld joints

b. Critical cracks of area Dt/4 occuring at:

i) Any location along straight section of piping

  • li) Any location along curved members of piping where: 0 = nominal OD of the system oloing materials t = minimum wall thickness of the system piping material (t,)

2.1.2 POSTULATED PIPING BREAK SIZES " Double ended and equivalent area longitudinal pipe break areas are based on the nominal inside diameter (ID) of the piping system. Critical crack pipe b'eak r areas are based on a length equal to one-half the nominal outside diameter (1/2 00) and a width equal to one-half the minimum wall thickness (1/2 tm) of the system piping materials. 2.l.3 LINE SIZE CONSIDERATIONS FOR POSTULATED PIPING BREAKS Piping larger than 1" nominal pipe size (NPS) is reviewed for the consequences of a double ended break. Piping 4" NPS and larger is reviewed for the consequences of double ended and equivalent area longitudinal breaks. All piping larger than 1" NPS is reviewed for the consequences of critical cracks. 2.1-2  ! l l l

2.2 ANALYTICAL CONSIDERATIONS FOR POSTULATED PIPING BREAKS Stress criteria as established in the AEC Guideline and described in 2.2.1 below was utilized in developing the results discussed in 1.2.2 for the Main Steam and Main Feedwater systems. Additional information on the types of analyses, stress criteria and loading combinations for mechanical and structural components, systems, and structures is described in this portion of the report. 2.2.1 STRESS CRITERIA Piping breaks listed in Tables 2.1-1 anit 2.1-2 which pose safety related pro-blems to structures, systems or componer ts in the immediate area are either restrained to mitigate the consequences of the break or reviewed in detail against existing stress analyses. If tre following stress allowables are not exceeded, then the break is not consider ed credible.

a. Duke System Piping Class A (Le 1.2.1.b. for Code correlations)

Sp +Ss g 2.05, for ferritic steel Sp + S, s 2.4Sm f r austenitic steel where: Sp = prim w/ stress intensity, Ss = secondary stress intensity; circumferential or longitudinal derived on an elastically cal-culated basis under the combined loadings associated with 1/2 safe shutdown earthquake and operational station conditicas. Operational station conditions include normal reactor operation, anticipated operational occurrences (upset conditions) and test-Ing conditions; S m is design stress intensity per ASME 111. U S 0.1 . where: U is the cumulative usage factor per ASME 111 based on all normal, upset and testing station conditions. 1 There are no Duke Class A piping systems outside the Reactor Building.

b. Duke System Piping Class B, C, D, E and F (See 1.2.1.b. for Code corre- j lations)
               + $3 3,0.8 (Sh+S)

A S where: Sp = primary stress intensity, Ss = secondary stress intensity; l circumferential or longitudinal derived on an elastically cal- l culated basis under the combined loadings associated with 1/2 safe shutdown earthquake and operational station conditions. Operational station conditions include normal reactor operation, anticipated operational occurrences (upset conditions) and testing conditions. Sh = stress per ASME lil, par. NC-3600 and ND-3600

  • for Class 2 and 3 components, respectively, Winter 1972 Addenda.

, SA = allowable stress range for expansion stress calculated by , ! the rules of NC-3600 of ASME Ill or ANSI B31.1.0 - 1967 l I l 2.2-1 1

ST $E 0.8 S A where: ST = thermal expansion stress 2.2.2 DYNAMIC ANALYSIS The CONTEMPT - PS code is utilized in developing pressure and temperature transients for postulated piping breaks within enclosed areas and rooms. Assumptions and considerations utilized in the analysis as applicable are:

a. Time dependent pressure rise is assumed to be equal throughout the volume being analyzed.
b. Orifice effects are considered through free vent areas.
c. Effective mass is assigned to blowout panels or louvers that might te in the structure.
d. Effective mass is assigned to walls, etc., that would fall and move out of the way for a continued blowdown.
e. Frictional effects of piping are not considered in mass flow rate cal-culations.
f. Two phase mass flow (liquid and vapor phases) is assumed to be homogeneous.
g. Condensation effects due to heat sinks are considered.

Environmental effects for Main Steam and Main Feedwater System piping breaks based en the above analytical method are described in 2.3.1. Dynamic analysis of piping and supports for seismic and other operational loads is described in IC.3.4 of the FSAR. Dynamic analysis of piping and supports for postulated piping breaks was not accomplished for Oconee high energy systems. Instead,the extensive investigations described in 2.1 for each postulated pipe break listed in Tables 2.1-1 and 2.1-2 were conducted assuming the loads. jet impingment forces, thrusts and areas of concern based on an unrestrained line. Thrusts developed from both the double ended and critical crack type breaks are tabulated in Tables 2.1-1 and 2.1-2. Environmental effects resulting from the above analysis are described in 2.3.1 for all postulated line breaks. Consequences of each break to station mechanical and electrical equipment and structures are listed on a case-by-case basis in Tables 2.3-1 and 2.3-2, all corresponding to Tables 2.1-1 and 2.1-2. The integrity of'the control room is unaffected as the result of postulated pipe breaks. A structural reinforced . concrete wall exists betwee'n the control room and the penetration room. There are no large openings in the wall, and all small openings have pressure seals which preclude any steam or water paths to the control

       . room.

2.2-2

2.2.3 STRUCTURAL ANALYSIS Evaluations of the adequacy of structures subjected to loads as the result of postulated pipe breaks are based on the following: For the Auxiliary Building (Reinforced Concrete Structure):

a. The method of evaluating stresses is " Ultimate Strength" as outlined in the 1971 ACI Building Code.
b. Maximum allowable stresses are the yleid stresses of the materials used.
c. Load combinations will be all dead and equipment loads plus pipe rupture loads. No load factors will be applied to these loads.
d. Design loads utilized in the analysis for structural adequacy are:
1) Concrete--------------------------------150#/cu. ft.

II) Block Walls-----------------------------90#/cu. ft. Ill) Structural Steel------------------------490#/cu. ft. Iv) Normal Piping Dead Load-----------------100#/sq. ft. v) Pressure and Temp. Transients-----------in accordance with Figures 2.3-1, 2.3-2, 2.3-3 and 2 3-4. vi) Dynamic Piping Load---------------------Based on thrusts listed in Tables 2.1-1 and 2.1-2; damage as listed in Tables 2 3-1 and 2.3-2 is assumed. For the Turbine Building (Structural Steel Structure):

a. The method of evaluating stresses is as outlined in the AISC Specifications (1963) and in accordance with accepted design of structural steel members,
b. Allowable design stresses are yield stress.
c. Load combinations are dead load, normal operating live load and pipe rupture load. No load factors will be applied to these loads,
d. Design loads utilized .in the analysis for structural adequacy are:

i) Concrete--------------------------------150#/cu. ft. II) S t ructu ra l S tee l------------------------490#/cu . f t . Ili) Normal Opera ti ng Live Load--------------125#/sq . f t . (includes Piping Dead Load) Iv) Pressure and Temp. Tr( istents-----------Negligible v) Dynamic Piping Load---------------------Based or thrus ts l isted in Tabl es 2.1-1 and 2.1-2; damage as listed in Tables 2 3-1 and 2.3-2 is assumed. The only area subjected to reversal of load due to postulated pipe breaks is the penetration room ceiling at Elevation 838'-6" of the Auxiliary Bulding. Investi-gation of this area resulted in stresses less than yield, thus assuring the inte-grity of the ceiling. Based on the above evaluations, failure of civil structures cau' sed by postulated pipe breaks will not cause failure of any other structure in a manner to adversely affect the mitigation of the postulated accident or the capability to safely shut down the reactor and maintain it in a safe shutdown condition. 2.2-3

2.3 CONSEQUENCES OF POSTULATED LINE BREAKS Resultant environmental ef fects and physical damage to the station as the result of postulated oipe breaks are described in 2.3.1 and 2.3.2, respec-tively,0f this portion of the report. A detailed discussion is presented in 2.3.1 on the environmental effects of the Main Steam and Main Feedwater system line break inside the Auxiliary Building penetration room with respect to analytical model, assumptions, etc. As discussed in 2.1.1, Figures 2.1-1.a thru 2.1-24 which are cross-referenced to each postulated pipe break having a safety-related consequence are included in this report for visualization and clarification of the break areas. 231 ENVIRONMENTAL EFFECTS Environmental eifects resulting from postulated pipe breaks are listed in Tables 2.3-3 and 2.3-4 on a case by case basis corresponding to each postulated pipe break listed in Tables 2.1-1 and 2.1-2. Environmental ef fects of concern are resulting pressures, temperatures, and flooding which could occur af ter a postulated pipe break at the locations listed in Tables 2.1-1 and 2.1-2. As the result of extensive investigations described in 2.1 above, the penetration room of the Auxiliary Building at Elevation 809'-3" is the only area of the station which could experience appreciable pressurization in the event of a postulated piping break. Only in the event of a Main Steam or Main Feedwater break could the room be pressurized. Blowdown mass flow rates asso-clated with the Main Steam line break were those determined in FSAR Supplement 3, Figures 14.3.5-1 and 14.3.5-2. The analysis assumes that both ends of the af fected pipe discharge to the room. New bicwdown data was established for the Main Feedwater line break. l A total vent area of 873 square feet to the outside of the north end of the l East Penetration Room was used with the inertial effects of moving the walls i from the v9nt area being considered. In developing the final transients pre- I sented in Figures 2. 3-1, 2. 3-2, 2. 3-3 and 2. 3-4, these walls were treated I as blowout panels with an effective mass of one pound per square foot of vent area and a 3 psf design pressure in the direction of required failure. Thus, the integrity of the pentration room of the Auxiliary Building is assured. Two phase mass flow through the north and east wall vent area of the East Penetration Room was assumed to be homogenous and no phase exchange was assumed to take place between vap3r and liquid phases. Penetration room pressurization was attributed cnly to th'e pressure drop across the vent opening since fric- l tional effects of the room were found to be negligible. Although condensation on the buiding walls was considered, it was found to have little effect en the peak building pressure. Based on all-of the above assumptions and input data, the peak penetration room pressure would be - 2.80 psig for the unrestrained Main Feedwater line break and - 2.30 for the unrestrained Main Steam line break as can be seen from Figures 2. 3-2 and 2. 3-1, respectively. 2.3-1

      )

I 2.3.2 POTENTIAL PHYSICAL DAMAGE TO STATION , Physicel damage or the consequences of postulated breaks to mechanical, elec-trical, and structural portions of the station are listed on a case-by-case basis in Tables 2.3-1 and 2.3-2. Damages listed in Tables 2.3-1 and 2.3-2 are for unrestrained lines as discussed l In 2.2.2. Required protection from postulated pipe breaks is discussed in 4.2 of this report. Protection from the consequences of postulated. pipe breaks, as required, will be furnished such that mitigation of the accident and capability to safely shut down the reactor and maintain it in a safe shutdown condition, when necessary, are assured. l J l 4 i l 2.3-2 i _.__.m _

3.0 OPERATIONAL ANALYSIS This section will describe the operational status of the station resulting from a postulated piping break and oucline the procedures the operating staff will use to mitigate the postulated accident. Discussion is limited to those postu-lated piping breaks listed in Tables 2.1-1 and 2.1-2 which could affect the safety of the reactor coolant system or could cause the possible loss of Engi-neered Safety Features equipment. 31 OPERATIONAL STATUS AND MITIGATION OF ACCIDENT RESULTING FROM POSTULATED PIPING BREAKS 3 1.1 MAIN STEAM SYSTEM Case 1 of Table 2.1-1, " Postulated Pipe Breaks - Engineering Data." lists engi-neering data associated with various main steam line breaks. Analysis of a single steam line break is given in FSAR Supplement 3 in answer to question 14.3.3 A synopsis of that analysis is presented here for completeness. For a single steam line break case numbers 1.a and 1.b, assuming no integrated control system action, the following major assumptions were made for the analysis:

a. Plant operating at rated power
b. No repositioning of feedwater valves
c. No immediate operator action
d. Maximum negative moderator coefficient corresponding to end-of-life conditions of equilibrium cycle
e. Minimum rod worth tripped corresponding to Technical Specification Ilmit for minimum shutdown margin
f. Maximum worth rod stuck out
g. Increased capacity of LP and HP injection systems af ter actuation with decreased reactor coolant pressure neglected. Credit taken for only one HP pump and one LP pump operating
h. Cooldown rate of the Reactor Coolant System assumed to be independent of core power I. Boron injection assumed to be perfectly mixed with all reactor coolant before entering core J. Perfect heat transfer assumed in the affected steam generator af ter the initial part of the transient The steam line break would cause an increase in the heat transfer from the reactor coolant to the feedwater. This initiates a cooldown of the Reactor Coolant System which, in turn, starts a pressure decrease within the system.

The reactor trips on low pressure in approximately six seconds. The reactor trip initiates turbine stop valve closure, isolating the unaffected steam gen-erator on the steam side. This isolated loop continues to supply steam to drive the associated main feedwater pump. For the cooldown part of the analysis, it is assumed that all feedwater from this pump goes to the affected steam gen-erator. The equilibrium flow rate is'about 135 percent of the-rated flow in one steam generator. With the above assumptions, the resulting reactor coolant temperature decrease causes initiation of high pressure injection in approximately 30 seconds (including a 10-second delay) and the core flood tanks injection in approximately one minute. These injections will keep the core subcritical during 3.1-1

the ensuing cooldown until just t'efore initiation of low pressure injection at approximately 100 seconds. The neutron power peaks at 8 percent rated power before going subcritical again at approximately 166 seconds af ter the break. S$-iticality is caused by injection of borated water f rom the Low Pressure injection System. Therefore, the reactor trips shortly af ter a steam line failure, and essentially remains shutdown for the ensuing cooldown of the Reactor Coolant System. Peak power generated in the core will not cause any thermal limits to be approached; therefore, no reactor damage occurs. (FSAR, page 14-17)

   -s For a single steam line break outside the Reactor Building, the following systems are required from the above analysis: High Pressure Injection System, Core Flood-ing System, Low Pressure Injection System, Reactor Protective System (low pressure trip), and Turbine Scop Valve Closure Signals and Actuation System. The operator receives adequate Indication to determine the plant situation by observing the automatic action as identified above and main steam line pressure.        In addition, penetration room temperature will identify if the steam line break has occurred within the penetration room.

The operator need take no immediate manual action other than confirming that the automatic action has occurred. However, to conserve feedwater, the Main Feedwater valve' would be closed to the affected steam generator and minimum level in the ui .rfected steam generator established to control plant cooldown. The High Pressure injection System would be operated manual!y to control pres-surizer level and boron concentration. For a postulated single steam line break in Main Steam Line B (which does not run through the penetration room), all instrumentation and equipment needed for the proper mitigation of that accident is available. However, for steam-line A should the accident occur within the penetration room, valve HP-26 cable and one channel of LP Injection piping could be lost by direct Impingement from the steam line break. Therefore, for this particular accident, the single failure criteria . l would not be met. For this case, further discussion of interim measures and design changes is presented in Section 4.0. The above analysis assumes no decay heat input into the Reactor Coolant System. This provides a very conservative analysis. If decay heat is factored into the > analysis, the temperature and therefore the pressure of the Reactor Coolant Sys-tem will decrease at a slower rate allowing more boron to enter the Reactor Coolant System via HP and CF Injection. Neutron power peaks at 1.3 percent and , CF Injection returns the reactor subcritical. LP Injection is not needed to pre- ' vent the reactor from returning to power. Consequently, no design changes are required for the LP Injection System. Case numbers 1.c and 1.d of Table 2.1-1, " Postulated Pipe Breaks - Engineering Data," identify main steam line breaks which could possibly damage branch lines terminating in the adjacent main steam line. Consequently, this could lead to a double steam generator blowdown. The same assumptions that are presented above, except that the Integrated Control System (ICS) functions' properly to close the startup and main feedwater valves after the reactor trip, is used to " analyze the plant situation. The Reactor Coolant System again cools down . which initiates a reactor trip on low pressure at approximately six seconds 3.1-2

following the break. A reactor trip will cause a turbine trip after which the ICS will act to shut off feedwater to the steam generators. The High Pressure injection System is actuated at 26 seconds af ter the break (including a 10-second delay). Neutron power rises to 5 percent within two seconds prior to actuation of the core flooding tanks at 30 seconds. Neutron power then decreases but will again rise af ter core flooding and return to 7 percent just before the Low Pres-sure injection is actuated at 65 seconds af ter the break. The Low Pressure injection then returns the core subcritical at 125 seconds af ter the break. The decay heat is about 4 percent of rated power within a minute following reactor trip. Therefore, the peak power generated in the core following a double steam line break is approximately 11 percent of rated power. This power is low enough that peaking does not become a problem; thus, the core will remain within its thermal bounds and the reactor damage criteria (FSAR, page 14-17) are met. This plant situation requires that the following systems function to mitigate the accident: Reactor Protective System (low pressure trip), High Pressure injection, Low Pressure injection, Core Flooding System, and integrated Control System. Because the actuation sensors, cables, power, and equipment for the required systems are in the Reactor Building, Auxiliary Building and areas of the Turbine Building well away from the postulated break locations, no equipment is damaged that would be required. The operator would not need to take immediate action other than verify that the automatic actions have occurred. The operator within a few minutes will take action to regulate feedwater to steam generators to affect an orderly plant cooldown. The High Pressure injection System will be used to control pressurizer level and boron concentration. Long-term shutdown cooling can be accomplished by use of the Low Pressure injection System. The above analysis assumed proper actuation of the ICS. If no ICS operation is assumed and 100 percent feedwater flow is assumed to reach steam generators throughout the accident, the following would occur: reactor trip at six seconds, High Pressure injection at 26 seconds, core flooding actuation at 27 seconds, Low Pressure injection at 60 seconds, and subcriticality at 134 seconds. Neutron power would reach 31 percent at peak; however, no violation of core integrity is anticipated due to the very short time duration of the peaking. 3.1.2 MAIN FEEDWATER SYSTEM (TURBINE BUILDING) Case number 4.b of Table 2.1-1 Identifies postulated b. eaks in the Main Feedwater System piping from the feedwater pump discharge through the high pressure feed-water heaters to the Turbine Building wall. A feedwater line break with a subse-quent pipe whip could strike the emergency feedwater piping connecting to the adjacent feedwater line and effect the possible loss of both Main Feedwater lines and the Emergency Feedwater System feeding the steam generators. Also, due to dire : impingement, the 4160 volt switchgear ITC, ITD, and ITE may also be lost. The analysis of this accident is similar to that presented in Section 14.1.2.8.3, "Results of a Complete Loss of All Station Power Analysis," of the Final Safety. Analysis Report. As stated in that section, immediate operation of cooling water l Is not of a critical nature. Each reactor can sustain a complete loss of electric , power without emergency cooling for about 23 minutes before the steam volume in i the pressurizer is filled with reactor coolant. Beyond this time, reactor coolant l will boil off and an additional 83 minutes will elapse before the boll-off will I st&.'t to uncover the core. 3.1-3 e

Later analyses have shown that one High Pressure injection pump which began operation at 35 minutes af ter the accident can prevent the core from uncovering without the aid of secondary cooling. Assumptions made for this analysis are no mixing occurs between High Pressure Injection water and reactor coolant; HPl water has no cooling effect until it has replaced the entire volume of coolant between the HPl nozzles and top of the core, and HPI water is injected at 90*F. With these assumptions and start of the HPl pump at 35 minutes, the core will nearly be uncovered at 99 minutes. At this time, the cooling capacity of the HPI is more than adequate to handle the decay heat removal; the makeup rate will exceed the bol l-off rate and coolant level will begin rising. In order to safely place the reactor in cold shutdown condition, plant cooldown is initiated by manually starting the auxiliary service water pump and feeding lake water into the steam generators. A system description of the operation of the Auxillary Service Water System and its design bases is presented in Section 9.11 of the FSAR. Simultaneously with the manual operation of the Auxiliary Service Water System, a High Pressure injection pump can be started manually by connecting a power source from the 4l60/600 volt auxiliary service water switchgear to any one of the three High Pressure injection pumps. Power to this switchgear is not a part of the 4160 power affected by the accident but comes directly from the 4160 main feeder buses (see Figure 8-3 of the FSAR). The pressurizer level and boron concentration can then be controlled by the operation of the High Pressure injection pump. These actions can be easily accomplished within a 30-minute time period. Since the failure of the Auxiliary Service Water System pump could leave the plant without adequate long-term cooling, design changes for a redundant Emergency Feed-water System are described in Section 4. 3.1.3 MAIN FEEDWATER SYSTEM (PENETRATION ROOH) Tabic 2.1-1, Case number 4.c, identifies feedwater line breaks within the pene-tration room. The most serious break that can occur in this area is between the Reactor Building wall and the check valve. Breaks between the Turbine Building wall and check valve are less serious because a steam generator blow-down does not occur. The analysis for the worst case break is similar to that used in Section 3.1.1 for a single steam line break. The assumptions used for the analysis are the same. The reactor again shuts down due to a Low Pressure Reactor Protective System trip. The reactor trip initiates turbine stop valve closure isolating the unaffected steam generator on the steam side. The resul t-Ing Reactor Coolant System temperature decrease causes initiation of High Pres-sure injection and core flooding. The core flooding injection will maintain - the reactor subcritical. No immediate action by the operator is required to mitigate this accident; however, to conserve feedwater the affected line would be isolated within several minutes. Feedwater can continue to be ' supplied to the unaf fected steam generator by using the hotwell and condensate booster pumps. Long-term cooling can be accomplished by the use of the Low Pressure injection System. Pressurizer level and Reactor Coolant System boron control can be accomplished by the use of the High Pressure Injection System. 3.1-4

Due to direct impingement fron the feedwater break, valve HP-26 and its asso-clated single channel of High Pressure injecticn piping could be lost. Therefore, the single failure criteria for this accider.t would not be met. Further dis- ! cussion of interim measures and design changes is presented in Section 4. 3 1.4 AUXILIARY STEAM SYSTEM Table 2.1-1, case number 3 Identifies engineering data associated with an euxiliary steam line break. The analysis, consequences, and operat'onal pro-cedures of the postulated break are the same as for that presented in Section 3.1.2, Main Feedwater System (Turbine Building) . 3 1.5 TURBlNE EXTRACTION SYSTEM Case number 6 of Table 2.1-1 Identifies engineering data associated with turbine extraction line breaks within the Turbine Building. The possible consequence of a postulated break of one of these lines is possible loss of one channel of emergency feedwater. No safety consequences are involved; however, the turbine would be manually tripped, the reactor shut down, and the Reactor Coolant System cooldown would be Initiated by normal procedures. 3.1.6 CONDENSATE SYSTEM, BOOSTER PUMP DISCHARGE Case number 7 of Table 2.1-1 identifies engineering data associated with conden-sate booster pump discharge piping postulated breaks. The analysis of this situation is similar to that presented in Section 3.'.2 concerning the main feedwater line break within the Turbine Building except that the 4160 volt switchgear ITC, ITD, and ITE would not be affected. As stated in Section 3.1.2, cooldown of the Reactor Coolant System would be by the use of the auxiliary service water pump , and the Reactor Coolant System volume and boron control would be accomplished by the use of the High Pressure injec-tion System. Long-term cooling could be accomplished either by the auxiliary ' service water pump or by the Lower Pressure injection System. Because the failure of the Auxiliary Service W' ter a System pump could prevent cooldown of the Reactor Coolant System, interim measures and redesign are discussed in Section 4.0. 3 1.7 FEEDWATER HEATER LRAINS AND VENTS Case number 8 of Table 2.1-1 identifies engineering data related to postulated breaks in the feedwater heater drains and vents system. The consequence of these breaks is possible loss of the Main Feodwater System. Two systems, Emergency Feedwater and Auxiliary Service Water, are available to begin Reactor Coolant System cooldown. The operator would Isolate the break in the Main Feedwater System- and begin cooldown using the Emergency Feedwater System. Low Pressure injection would be available for long-term cooling. 3 1.8 HIGH PRESSURE INJECTION SYSTEM, SEAL SUPPLY AND NORMAL MALElP

Table 2.1-l., Case' numbers 13 and 14, Identify engineering data concerning postu-lated break of the High Pressure injection System lines for the reactor coolant pump seal supply and for the normal makeup to the Reactor Coolant System. If 3.1-5 i

e y- , -- -

                                                             ,               ,--      ,,_-.-y, _ --
     .                                                                                    I i

1 I l a break occurred in the seal supply line, the operator would isolate the break by closing one of the letdown line isolation valves and the reactor would be shut down by normal procedures. The reactor coolant pumps are able to run during the cooldown period with component cooling water and without seal supply. Makeup to the Reactor Coolant System would be by HP Injection line B. If the break occurred in the normal makeup line, worst case analysis would show  ! that the letdown storage tank would be pumped dry very culckley. The operator , wou'd shutdown the operating High Pressure injection pump to prevent damage to the pump and isolate the letdown flow. Normal cooldown procedures coJId then be used to place the plant in a cold shutdown condition. One of the High Pressure injection pumps lined up to Emergency injection Line B would provide makeup for the cooldown shrinksge. l 3.1.9 HIGH PRESSURE INJECTION SYSTEM, LETDOWN Table 2.1-1, case number 15, identifies a possible break in the High Pressure injection line between the Reactor Building and the letdown line orifice. The analysis assumes a complete severence of the 2 1/2 inch letdown line. No operator action is assumed. Coolant is assumed to flow out of the pipe untII the isolation valves are fully closed. Credit is not taken for the reduction in flow during the last 10 seconds while the valve is closing. Normal makeup system is assumed to function which results in a slightly longer time to reach 1500 psig and a correspondingly higher loss of inventory. The Engineered Safe-guards System is actuated by low ReactorCoolant System pressure at 1500 psig and consequently closes valves HP-3, HP-4, and HP-5 at approximately 160 seconds after the accident. Off-site releases are within acceptable limits for this accident. Af ter isolatior of the break by Engineering Safeguards actuation, the operator would take actit to cool the Reactor Coolant System down by normal procedures. 3 1.10 EMERGENCY FEEDWATER SYSTEM, PUMP DISCHARGE As identified in Table 2.1-2, case number 17, a postulated break in the a nergency feedwater pump discharge piping from the pump to the upper surge tank would pro-vide no mechanical, electrical, or structural damage to other safety systems. Unless the break occurred immediately at the pump discharge, the break could be isolated and utilization of the energency feedwater pump to the steam generators would not be impaired. For a break between the pump discharge and the first isolation valve, another unit's emergency feedwater pump could be made available on a hot standby basis while repairs are being made. EMERGENCY FEEDWATER SYSTEM, PUMP DISCHARGE TO STEAM GENERATORS 3 1.11 The emergency feedwater pump discharge to steam generator postulated piping breaks identified in Table 2.1-2, case number 18.b., analysis is similar to that presented in Section 3 1.2. Case numbers 18.a,18.c, and 38.d are of no-safety co'nsequence since flooding is the only effect and the ability of safety systems to perform their Intended function is not impaired. i i 3.1-6 l i O

The analysis and assumptions for case number 18.b are the same as those pre-sented in Section 3.1.2, Main Feedwater System (Turbine Building), except that one main feedwater line remains unaffected. However, due to loss of the 4160 volt switchgear ITC, ITD, and ITE, all hotwell and condensate booster pumps are lost. The mitigation of this accident is identical to that presented in Section 3 1. 2 and design changes identified in Section 4 for loss of main Feedwater piping apply. 3 1.12 HIGH PRESSURE INJECTION SYSTEM, PUMP C DISCHARGE For a break in the high pressure injection pump C discharge piping, case number 22 of Table 2.1-2, flooding would be the only effect. The only time this line would be pressurized would be during a quarterly test of HP Injection Pump C. The ability to place the reactor in a cold shutdown condition would not be impaired. Because one channel of High Pressure injection would be lost due to a break in this piping, the reactor would be placed en a cold shutdown condition using normal procedures until repairs could be completed. 3.1.13 LOW PRESSURE INJECTION SYSTEM Table 2.1-2, case number 23.a. Identifies Low Pressure injection System piping breaks engineering data. This system would be in operation and pressurized only during quarterly test of the LP pumps or during a reactor shutdown. The con-sequence of a postulated break in a LP pump discharge or LP pump suction piping or in one of the two lines to the RB emergency sump is as follows: one channel of low pressure injection piping could be lost but the ability to place the Reactor Coolant System in a cold shutdown is not impaired. The other LP channel would be used for final cooldown. For the loss of one channel, the Reactor Coolant System would be placed in a cold shutdown condition until repairs could be made or an acceptable redundant system could be made available. For a postulated break in the decay heat removal piping frora the Reactor Coolant System to the LP Injection pump suction header, the reactor would be already shutdown as this system is pressurized only during shutdown. Operationally, the Reactor Coolant System would be held in a shutdown condition at a pressura and tenperature so that the Reactor Coolant pumps could be operated,and the shutdown margin would be greater than 1%Ak/k until repairs could be completed. 3 1.14 REACTOR BUILDING SDRAY SYSTEM Case numbers 24.a and 24.d of Table 2.1-2 identify that for the loss of Reactor Building spray pump discharge the plant could suffer the possible loss of one LP Injection pump motor, one decay heat removal pump motor and one RB spray pump motor. In this situaticn, the plant would be placed in a cold shutdown condition l by normal procedures until repairs are completed. Normal cooldown via the secon-dary side would be used until transfer to the decay heat removal mode. l An alternate consequence to the above identified by case number 24.d of Table 2.1-2 would be the possible loss of two high pressure injection. pump motors due to flooding for a postulated break within the HP pump bays. The Reactor Coolant System would be placed in a cold shutdown condition until repairs :ould be et n-pleted. Case numbers 24.b and 24.c are of no safety consequence. 3.1-7 1 i

  • t ' .

d J . A i

                       . 3.2           

SUMMARY

OF POSTULATED BREAK ACCIDENT OPERATIONAL ANALYSIS Table 3 2 summarizes the consequences of each postulated break and Identifies those situations where design changes are required. 1 4 e i e 4 i i I 1 i i i

  • 4 E

1 f i k i i 1 f e f

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} i i-t i i ! 3.2-1 i i.

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

4.0 ~ PROTECTIVE METHODS AND STATION MODIFICATIONS FOR POSTULATED PlPING BREAKS 4.1 INTERIM HEASURES Section 3.0, Operational Analysis, identified several situations where the single failure criteria was not met for the immediate protection of the reactor. The purpose of this section is to describe those interim measures which will be taken prior to the completion of design changes which are described in Section 4.2. For the main feedwater line postulated break (Turbine Building) discussed in Section 3.1.2, the following interim measures will be adopted:

1. Drill openings in the piping insulation near high stress locations identified by Figures 2.1-4b.3 and 2.1-4b.4. These areas will be visually observed once per shift to detect any leakage.
2. The auxiliary service water pump will be tested for a minimum time of one hour each week to assure operability.

3 A cable has been pulled from the 4160 volt auxiliary service water switchgear to the high pressure injection pumps. This cable will be available for quick, manual connection to any of the high preesure injection pumps if needed. For the main feedwater line postulated break (penetration room) discussed in Section 3.l.3, the following measures will be taken:

1. The Reactor Building Isolation valve in the high pressure injection line B (HP27) will be left open to reduce the probability of a single failure occurrence should high pressure injection be needed.
2. Insulation on the feedwater lines in the penetration room will be drilled near the terminal ends identified by Figure 2.1-4.c and visually observed once per shif t to detect any leakage.

3 An additional temperature detector has been installed in the penetration room to give the operator additional information as to the occurrence and location of a break. For the Main Steam line postulated break in the penetration room, the following measures will be taken: , 1. The Reactor Building isolation valve in the high pressure injection line B (HP27) will be left open to reduce the probability of a single failure occurrence. This is the redundant valve to HP27 'which was identified in Section 3.1.1 as a possible valve which could fall due to direct impinge-ment on electrical cable from the steam line bresk. ,

2. An additional temperature detector has been installed in the penetration room to give .the operator additional information as to the occurrence and location of a break.

l 4.1-1 1

                                                                                      - - - m__

For the auxiliary steam line postulated break identified in Section 3.1.4, openings will be drilled in the piping insulation near the high stress locations of consequence as identified in Figure 2.1-3 and these areas will be visually observed once per shif t to detect any leakage. For the condensate booster pump discharge piping postulated breaks discussed in Section 3.1.6, openings will be drilled in the piping insulation near locations of consequence as identified in Figure 2.1-7. These areas will be visually observed once per shif t to detect any leakage. For the emergency feedwater pump discharge postulated breaks discussed in 3.11, openings will be drilled in the piping insulation near terminal ends and locations of consequence as identified in Figure 2.1-18.b, and these areas will be visually observed once per shif t to detect any leakage, in adaltion to the above interim measures, a procedure is being developed for a surveillance program to inspect the areas of postulated ruptures. The procedure will require a visual examination of the metal surfaces at selected locations where postulated breaks could occur when systems necessary to cope with these breaks do not now meet the single failure criteria. Metal surface examination will be performed on the following piping: main steam line within the penetration room, main feedwater line within the penetration room, main feedwater line in the turbine building, condensate booster pump discharge piping near the Emergency FDW pump, emergency feedwater pump discharge piping in the area of the 4160 volt switchgear, and auxiliary steam lines near the 4160 volt swltchgear. The procedure will require that the insulation be removed and an inspection performed every 60 days. Acceptance criteria will be that no leakage exists at the inspection points and that no defects are found as determined by visual examination. 4.2 DESIGN CHANGES Several changes to the station will be made as the result of this review although Duke considers the probability of such occurrences highly unlikely. All high energy system postulated break consequences listed in Tables 2.3-1 and 2.3-2 have been analyzed,and design changes as described in Table 4.2 are currently being implemented. . As referenced in Table 4.2, appropriate figures describing proposed station design changes are included in this report, it is important to note that these figures are preliminary and the final designs may be somewhat different. 4.3 EMERGENCY OPERATING PROCEDURES The operating, emergency, and alarm procedures are being examined to assure that all situations described in Section 3.0 are adequately addressed, if additional procedures are needed, they will be developed and completed by July 1,1973. Procedures will also be reviewed after design changes are made to assure that the modified system operation is correctly addressed. 4.4 STATION MOD IFICATIONS SCHEC'J' e As discussed in 4.2 above, detailed design engineering is currently being I accomplished for the design changes outlined in Table 4.2. The estimated 4.1-2

s time to complete engineering, procure necessary materials and complete the construction is also listed in Table 4.2. Implementation of design and construction will be completed in a timely fash-lon such that all requirements of Duke's . Quality Assurance Program are met. it can be noted from Table 4.2 that all required station modifications are expected to be complete within a 6-month time period by November I,1973 for Unit 1. Units 2 and 3 are essentially duplicates of Unit I and the analysis and necessary design changes are being implemented. For Unit 2, all required station modifications are ' expected to be complete by February 1, 1974 or earlier; nevertheless, all interim measures described in section 4.1 of this report will

                                                                                      ~

be implemented for any incompleted design changes at time of startup. For Unit 3, all required station modifications will be completed prior to - unit startup, t e 1 O 1 4.1-3 -

L 5.0

SUMMARY

Upon completion of this review, Duke has reached the following conclusions:

a. Potential damage to the stat'on could result if the piping breaks postulated by AEC criteria did occur.
b. Of all the high energy lines examined at Oconee by stress criteria, none exceeded the stress levels allowed by AEC criteria,
c. The proability of actual occurrence of any of these postulated breaks is highly unlikely,
d. Nevertheless, all design modifications necessary to fully comply with the AEC criteria are being implemented in a timely fashion as discussed in 4.4 of this report.
e. Where modifications cannot be completed prior to operation of a unit, then appropriate interim measures have been defined and will be implemented.

I

                                                                                         \

l 5.0-1 i

Table 1.2-1 MAIN STEAM SYSTEM - QUALITY ASSURANCE PARAMETERS Design Conditions: 1050 psig e 630*F Safety Valve Settings (psig) Maximun Operating Conditions: l104 psig @ 595'F 2 @ 1050, 2 @ 1067, 2 @ 1080, Normal Operating Conditions: 910 psig @ 570*F 2 @ 1090, 2 @ 1100, 2 @ 1104 Pipe Spec: 26" and larger; ASTM A-155 Gr KC70 Class I with plate to ASTM A-515. Gr 70 FB 24" and smaller; ASTM A-106 Gr B Fittings: 26" and larger; ASTM A-234, Gr WPB-W with plate to ASTM A-515, Gr 70 FB 24" and smaller; ASTM A-234, Gr WPB

                        !!yes & Laterals: Cast ASTM A-216, Gr WCC with quality factor of 1.0 per ANSI B31.1.0 (1967) Par.102.4.6 ANSI B31.I(1967) lten                                                Duke Requirements                             Requirments
1. Longitudinal Weld Seam Ground flush inside for flow, elimination of weld Not required stress risers and any erosion problems.
2. Longitudinal and Girth Butt 100% X-Ray; traceable to exact location on line Not required Welds 3 Cast Wye and Lateral Fittings 100% X-Ray; traceable to exact location on fitting Not required
4. Material Certification Hill Test Reports for all materials Not required 5 Quality Assurance All X-Ray and illl Test Reports read and approved Not required by Duke prior to acceptance. This same check is done by manufacturer and Grinnell Corp. (3 checks)
6. Allowable Stress-Thermal, 100% of Code allowable, Sa, in cold position only Dead Weigist 100% of Sa for which is approximately 10% or less of life (100% 100% of time allowed cold pulled system) 7 Seismic Stress (OBE) Longitudinal pressure stress plus dead weight stress Same as Duke require-plus seismic stress & l.2 S ments a
8. Materials Surveillance Program None planned for System itself Not required 9 Hanger Support and Seismic Duke will inspect and document the support systems Not required Control System prior to initial operation, end of 1st year and every four years subsequent thereto. Hydraulic devices will be inspected per manufacturer's instructions in addition to the Duke program.
10. NDT Temperature Considerations Carbon Steel System with a NDTT something less than Does not speak to 0*F. ASME Section I and ANSI B31.1.0 both specify subject specifically allowable stress between -20*F and 400*F.

l l Table 1.2-2 Page I of I l HAIN FEEDWATER SYSTEM - QUALITY ASSURANCE PARAMETERS Design Conditions: 1275 psig @ 475*F Maximum Operating Condit ions: 1259 psig @ 460*F Normal Operating Conditions: 1070 psig @ 460*F Pipe Spec: 26" and larger; ASTM A-155 Gr KC70 Class I with plate to ASTM A-515, Gr 70 FB 24" and smaller; ASTM A-106 Gr B Fittings: 26" and larger; ASTM A-234,'Gr WPB-W with plate to ASTM A-515. Gr 70 FB 24" and smaller; ASTM A-234, Gr WPB ANSI 831. l(1967) Item Duke Requirements Requirements

1. Longitudinal Weld Seam Ground flush inside for flow, elimination Not required of weld stress risers and any erosion prchlems
2. Longitudinal and Girth Butt Welds 100% X-Ray; traceable to exact location on line Not required 3 Material Certification Mill Test Reports for all materials Not required
4. Quality Assurance All X-Ray and Mill Test Reports read and approved by Not required Duke prior to acceptance. This same check is done by manufacturer and Grinnell Corp. (3 checks) 5 Allowable Stress-Thermal, 100% of Code allowable, Sa, in hot position only 100% of aS for Dead Weight 100% of time
6. Seismic Stress (08E) Longitudinal pressure stress plus dead weight stress, Same as Duke plus seismic stress & l.2 S a requirements 7 Mat'e rials Surveillance None plar.ned for System itself Not required
8. Hanger.-Support and Seismic Duke will inspect and document the support system prior Not required Control System to initial operation, end of 1st year & -very four years subsequent thereto. Hydraulic devices will be inspected per manufacturer's instructions in addition to the Duke Program 9 NDT Temperature Considerations Carbon Steel System with a NOTT something less than O'F. Do's not speak-ASME Section 1 and ANSI B31.1.0 both specify allowable to subject speci-stress between - 20*F and 400*F fica 'f

Table 2.1-1 Page I of 3 POSTULATED PIPE BREAKS - ENGINEERING DATA----SYSTEMS NORMALLY IN OPERATION Double Ended Longitudinal System and Postulated Break Oper. Press. Temp. Nominal Line Size impingement Thrust Slot Impingement Location (Case No.) (psig) (*F) & WI. Thk.(OD)/(in) (Ibs) Thrust (Ibs) Main Steam, OlA 1.a. Outside Reactor Bldg., West Line - 914 570 36 1/2 - 1.164 1._30,000 14,425 terminal end I.b. Penetration Room East Line - terminal end I.c. Turbine Bldg. East and West Line - terminal ends

                                                      'I.d.            Turbine Bldg. East and West Line -         _   __         J __

i Any of Consequence HP Turb - Holsture Separator LP Turb - Steam, OlB

2. All terminal ends and any of -------------------------(No Breaks of Safety Consequence)------------------------

consequence Auxiliary Steam for Startup, 02A 3 All terminal ends and any of 310 480 8/0.322 1,100 3 260 consequence Main Feedwater, 03 4.a. FDW Pump Discharge-terminal end 1043 460 24/1.219 174,900 3,505 4.b. FDW Pump Discharge thru HP FDW 24/l.219 174,900 3,505 Htrs. to Turb. Bldg. Wall - any. 30/l.157 288,500 4,160 of Consequence 4.c. FDW Pump Discharge from Turb. Bldg _ _ . 24/1.219 174,900 3,505 Wall to R. Bldg. - terminal end Holsture Separator - Rehtr. Drains, 058 5 All terminal ends and any of ----------------------(No Breaks of S.ifety Consequence)------------------------ consequence

d Table 2.1-1 Page 2 of 3 POSTULATED PIPE BREAKS - ENGINEERING DATA----SYSTEMS NORMALLY IN OPERATION (CONT) Double Ended Longitudinal System and Postulated Break Oper. Press. Temp. Nominal Line Size impingement Thrust Slot Impingement Location (Case No.) (psig) (*F) & W1. Thk.(OD)/(in) (Ibs) Thrust (Ibs) Turbine Extraction Steam to FDW Htrs, 06A and B - All terminal ends and any cf consequence 6.o. HP, A & B Extraction 522 475 20/0.500 182,700 1,600 309 422 18/0.375 90,000 650 6.b. LP, C, D & E Extraction 179 380 30/0.375 150.800 650 43 300 48/0.375 99,700 256 0.2 214 48/0.375 1,200 3 Condensate Booster Pump Discharge, 078 7 All terminal ends and any of 403 372 24/0.688 68,200 700 consequence 30/0.689 109,000 875 FDW Heater Drains and Vents, 10 8.a. D Htr brain Pump Discharge 608 287 24/0.562 16,900 140 8.b. E Htr Drain Pump Discharge 599 207 12/0.375 15 1,350 Electro-Hyd. Control, Turb. Gen, 12A 9 All terminal ends and any of -------------------(No Breaks of Safety Consequence)------------------------ consequence Condensate Heating, 20A

10. All termin'ai ends and any of -------------------(No Breaks of Safety Consequence)------------------------

cor. sequence FDW Pump Turb. Seal Steam 34A

11. All terminal ends and any of -------------------(Mo Breaks of Safety Consequence)------------------------

consequence Nitrogen, 48

12. All terminal ends and any of ------------------- (No B rea ks of Sa f e ty Cons eq uence) ------------------------

consequence

Table 2.1-1 Page 3 of 3 POSTULATED PIPE BREAKS - ENGINEERING DATA----SYSTEMS NORMALLY IN OPERATION (CONT) Double Ended Longitudinal System and Postulated Break Oper. Press. Temp. itominal Line Size impingement Thrust Slot Impingement Location (Case No.) (psig) (*F) & WI. Thk.(OD)/(in) (Ibs) Thrust (Ibs) HP Injection, RC Pump seal supply. SI A 13.a. Pump Discharge (A or B)- 3020 1]O 4/0 531 70 1,050 terminal end

                                                                                                   ~

13.b. R Bldg-terminal end i 1/2-0.281 450 -- 13.c. Any of consequence u _f_ 4/0.531 70 1,050 HP Injectlon, normal makeup, 51A 14.a. Pump Discharge (A or B)-terminal 3020 130 4/0.53I 70 1,050 end 14.b. R. Bldg. terminal end 14.c. Any of consequence ' _ _. J_ _ _ _ HP injection, letdown line to orifice,51A . 15.a. R. Bldg.-terminal end 2200 579 2 1/2-0.375 5,056 -- 15.b. -Press. Red. btation-terminal ends l l [ 15.c. Any of consequence U _.L. 8 Notes: 1. (No Breaks of Safety Consequence) as used throughout this table means that line rupture effects were analyzed and found to not adversely affect operation of the Reactor Coolant System although the operator L.ay elect to shut the unit down for repairs.

2. Systems listed in this table are all inclusive of any and all branch lines to auxiliary equipment, vents, drips and drains, bypass lines, cross connections between systems, instrumentation, etc.

3 Flow controlled by a i I/4" orifice. _ - _ --____ _ m__- _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _

Table 2.1-2 Page I of 2 POSTULATED PIPE BREAKS - ENGINEERING DATA----SYSTEMS NOT NORMALLY IN OPERAT10tj Double Ended Longitudinal Systen and Postulated Break Oper. Press. Temp. Nominal Line Size impingement Thrust Slot Impingement Location (Case No.) (psig) (*F) & WI. Thk.(OD)/(in) (Ibs) Thrust (Ibs.) Auxiliary Steam to RC Bleed and Misc. Waste Evrporators , 02A(1)

16. Terminal ends and any of consequence -------------------------(No Breaks of Safety Consequence)---------------------

Emergency FDW Pump Discharge to Upper Surge Tank, 03A 17 Terminal ends and any of consequence 1,120 90 6/0.432 233 1,600 Emergency FDW Pump Discharge to Steam G;nerators, 03A . 18.o. Emerg. FDW Pumo Discharge terminal end 1.120 90 6/0.432 233 1,600 18.b. Main FDW Header - terminal end 18.c. R. Bldg. -terminal end 18.d. All above Emerg. FDW Pump Discharge lines,- ' U N any of consequence Strtion Heating, 20 19 All terminal ends and any of consequence ------------------------- (No B r ea k s o f S a f e t y Cons eq u e nc e ) -------------------- Steam Blanketing for HP FDW Heaters, 20(l)

20. All terminal ends and any of consequence -------------------------(No Breaks of Safety Consequence)--------------------

Steam Blanketing for Holsture Separator Rhtr. , 02A(1)

21. A!! terminal ends and any of ' consequence ------------------------- (No B r ea k s o f Sa fe t y Con s equ e nc e ) -------------- ------

HP Injection, Pump C Discharge, SIA 22.0. Pump Discharge-terminal end 3,020 130 4/0.531 70 1,050 22.b. R. Bldg.- terminal end 22.c. Any of consequence o __ i 1

Table 2.1-2 Page 2 of 2 - POSTULATED PIPE BREAKS - ENGINEERING DATA----SYSTEMS NOT NORMALLY IN OPERATION Double Ended Longitudinal System and Postulated Break Oper. Press. Temp. Nominal Line Size impingement Thrust Slot Imping ement Location (Case No) (psig) (*F) & W1. Thk. (OD)/(in) (Ibs) Thrust (Ibs.) LP Injection 53A and B 23.a. Decay Heat Removal -all terminal ends 388 300 12/0.180 5,700 30 and any of consequence 14/0.250 7,800 35 23.b. System Test- all terminal ends and ---------------- (No Brea ks o f Sa f e t y Consequ ence) --------------------------- any of consequence RB Spray, 54A and B 24.a. Pump Discharge-terminal end 311 90 8/0.148 112 200 24.b. 3" Test Line-terminal end in East Penetration Room 24.c. 3" Test Line-terminal end in ' Jest Penetration Room 24.d. All above RB Spray lines -any of _L _L ' _M_ _ji consequence Fual Cask Decontamination , 61(23) 25 Pump Discharge-terminal end and any ---------------- (No B rea ks o f Sa f e t y Cons equen ce) --------------------------- of consequence RC System Pressurizer Sample, 64(l)

26. Outside R. Bldg. -terminal end and ----------------(No Breaks of Safety Consequence)---------------------------

any of consequence , Steam Generator Secondary Side Sample, 64(2)

27. Outside R. Bldg. -terminal end and ----------------- (No B r ea ks o f Sa f e t y C on s eq ue nc e ) ---------------- -----------

any of consequence Notes: 1. (No Breaks of Safety Consequence) as used throughout this table means that line rupture effects were analyzed and found to not adversely af fect operation of the Reactor Coolant System although the operator may elect to shut the unit down for repairs.

2. Systems listed in this table are all inclusive of any and all branch lines to auxiliary equipment, vents, drips and drains, bypass lines, cross connections between systems, instrumentation, etc.

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s - Table 2.3-1 Page 2 of 3 POSTULATED PIPE BREAKS - CONSEQUENCES----SYSTEMS NORMALLY IN OPERATION System and Postulated Break tocation (Case No. ) Mechanical Damage Electrical Damage Structural Damage

          . Moisture Separator - Rehtr. Drains, 05B 5      All terminal ends and any of consequence None                           None                    None Turbine Extraction Steam to FDW Htrs, 06A cnd B - All terminal ends and any cf consequence 6.a. HP, A & B Extraction                        Possible loss of 1 Emerg. None                      Possible local yielding of FDW Line                                            steel beams, columns and swag l                                     of floor above in Turb. Bldg.

6.b. LP, C, D & E Extraction f " i Condtnsate Booster Pump Discharge 078 7 All terminal ends and any of Possible loss of Emerg. None Possible local yielding of consequence and Main FDW System steel beams, columns and swag of floor above in Turb. Bldg. FDW Heater Drains and Vents, 10 8.c. D Htr Drain Pump Discharge Possible loss of Main None Possible local yielding of FDW System steel beams, columns and swag

                                                                           }                                       of floor above in Turb. Bldg.

8.b. E Htr Drain Pump Discharge r ' j Electro-Hyd. Control, Turb. Gen., 12A 9 All terminal ends and any of None None None consequence Condensate ileating, 20A

10. All terminal ends and any of None None None consequence FDW Pump Turb. Seal Steam, 34A II. All terminal ends and any of None None None consequence

Table 2.3-1 Page 3 of 3 POSTULATED PIPE BREAKS - CONSEQUENCES----SYSTEMS NORMALLY IN OPERATION System and Postulated Break Location (Case No.) Mechanical Damage Electrical Damage Structural Damage Nitrogen, 48

12. .All terminal ends and any of None None None consequence HP Injection, RC Pump seal supply. SI A 13.a. Pump Discharge (A or 8)- None None None terminal end 13.b. R. Bldg. -terminal end 13.c. Any of consequence " __ L 1 HP Injection, normal makeup, SIA 14.a. Pump Discharge (A or 8)-terminal end None None None 14.b. R. Bldg. -terminal end 14.c. Any of consequence J_ _ _ . _ ll HP Injection, letdown line to orifice, SIA 15.a. R. Bldg. -terminal end None None Nonc 15.b. Press. Red. Station-terminal ends 15.c. Any of consequen:<. __A_ _ 1._ J_

Note: Consequences listed in this table are those required to either mitigate the consequences of the postulated rupture or those required to safely shut down and maintain the reactor in a safe shutdown condition only.

Table 2 3-2 Page I of 3 POSTULATED PIPE BREAK - CONSEQUENCES----SYSTEMS NOT NORMALLY IN OPERATION System and Postulated Break Location (Case No.) Mechanical Damage Electricas Damage Structural Damage Auxiliary Steam to RC Bleed and Mixc. Waste Evaporators - 02A(1)

16. Terminal ends and any of None None None consequence Emergency FDW Pump Discharge to Upper Surge Tank, 03A 17 Terminal ends and any of None None None consequence Emergency FDW Pump Discharge to Stcam Generators, 03A 18.a. Emerg. FDW Pump Discharge - None None None terminal end 18.b. Main FDW Header - terminal ends Possible loss of Emerg. and Possible loss of 4160 V Possible local yielding of Main FDW Systen ES Power in Turb. Bldg. steel beams, columns and swag of floor above in Turb.

Bldg. 18.c. R. Bldg.- terminal end None None None 18.d. All above Emerg. FDW Pump i i __[_ Discharge lines -any of consequence Station Heating, 20 19 All terminal ends and any of None None None consequence Steam Blanketing for HP FDW Htrs. 20(l)

20. All terminal ends and any of None None None consequence Steam Blanketing for Moisture Sepa-rater Rhtr., 02A(1)
21. All terminal ends and any of None None None consequence

Table 2.3-2 Page 2 of 3' POSTULATED PIPE BREAK - CONSEQUENCES----SYSTEMS NOT NORMALLY IN OPERATION

' System and Postulated Break Location (Case No.)                                  Mechanical Damage              Electrical Damage                   Structural Damage HP injection, Pump C Discharge, SIA 22.a. Pump Discharge - terminal end                      None                           None                                 None
 .22.b. : R. Bldg.      terminal end                           e                                                                   i 22.c. Any of consequence                                  i                             e                                     i LP Injection, 53A and B 23.a. Decay Heat Removal - all                           None                           Possible loss of I LP Inj. and       None terminal ends and any of                                                        I DHR pump motor or.2 HP Inj.
          -consequence                                                                     pump motors in Aux. Bldg.
'23.b. Sys tem . Test - all ' terminal                     u                           None                                 g_

ends and any of consequence .RB Spray, 54A and B 24.a. Pump Discharge - terminal.end None Possible loss of I LP inj., None , 1 DHR and 1 RB Spray pump motor in Aux. Bldg. 24.b. 3" test line - terminal end In None East Penetration Room . l , 24.c. 3" Test line - terminal end in s

    .. West Penetration Room 24.d. All above RB Spray lines - any                     U Possible loss of I LP Inj.,          _.l L of consequence.                                                                 I DHR and 1 RB Spray pump motor or 2 HP inj. pump motors in Aux. Bldg.

Fual Cask Decontaminatlon, 61(23)

25. Pump Discharge - termir.al end and None None None any of. consequence

_ _ _ - - _ - - T__ -m__- -

e e Table 2. 3-2 Page 3 of 3 POSTULATED PIPE BREAK - CONSEQUENCES----SYSTEMS NOT NORMALLY IN OPERATION System and Postulated Break location (Case No.) . Mechanical Damage Electrical Damage Structural Damage RC System Pressurizer Sample , 64(l)

26. Outside Reactor Bldg.-terminal None None None end and any of consequence Steam ' Generator. Secondary Side Sample , .64(2) 27 Outside Reactor Bldg-terminal None None None end and any of consequence Note: Consequences listed in this table are those'requir-d to either mitigate. the consequences of the postulated rupture or those required to safely shut down and maintain the reactor in a safe shutdown condition.

L_______.___ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _

Table 2.3-3 Page I of 3 POSTULATED PIPE BREAKS - ENVIRONMENTAL EFFECTS----SYSTEMS NORMALLY IN OPERATION System and Postulated Break Pressure Temperature Flooding Location (Case No.) (psig) (*F) (gallons and/or inches) Main Steam, OlA None None None 1.a. Outside Reactor Bldg., West Line - terminal end 2 1.b. Penetration Room East Line - terminal end 2.30 380 None Turbine Bldg. East and West Line - Negligible Negligible Negligible 1.c. terminal ends [ l g 1.d. Turbine Bldg. East and West Line - any of I I consequence HP Turb. - Moisture Separator LP Turb. - Steam, OlB

2. All terminal ends and any of consequence -------------- (No B r ea ks o f Sa f e ty Cons eq u ence ) -----------------

Auxiliary Steam for Startup, 02A All terminal ends and any of consequence Negligible Negligible Negilgible 3 Main Feedwater, 03 FDW Pump Discharge - terminal end Negligible Negligible 365,585 gal. max., 3.6" max. 4.a. in Turb. Bldg. Basement l [ r 4.b. FDW Pump Discharge thru HP FDW Htrs. to f Turb. Bldg. Wall - any of consequence 4.c. FDW Pump Discharge from Turb. Bldg. Wall 2.80 212 Negligible, North wall of to R. Bldg. - terminal end East Penetration Room falls due to pressure, relieving water to outside'the Aux. Bldg Moisture Separator - Rehtr. Drains, 058 5 All terminal ends and any.of consequence --------------(No Breaks of Safety Consequence)----------------- Turbine Extraction Steam to FDW Htrs, 06A and B All terminal ends and any of consequence Negligible Negligible Negligible 6.a. HP, A & B Extraction i e i 6.b. LP, C, D & E Extraction

Table 2.3-3 Page 2 of 3 POSTULATED PIPE BREAKS - ENVIRONMENTAL EFFECTS----SYSTEMS NORMALLY IN OPERATION System and Postulated Break Pressure Temperature Flooding Location (Case No.) (psig) (*F) (gallons and/or inches) Condensate Booster Pump Discharge, 078

7. .All terminal ends and any of consequence Negligible Negligible 365,585 gal. max., 3.6" max. In Turb. Bldg. Basement FDW Heater Drains and Vents, 10 8.a. D Htr Drain Pump Discharge Negligible Neglipible Negligible 8.b. E Htr Drain Pump Discharge I e i Electro-Hyd. Control, Turb. Gen., 12A 9 All terminal ends and any of consequence --------------- (No B r ea ks o f Sa f e t y Cons equ enc e )----------------

Condensate Heating, 20A

10. All terminal ends and any of consequence Negligible Negligible Negligible FDW Pump Turb. Seal Steam, 34A
11. All terminal ends and any of consequence Negligible Negligible Negligible Nitrogen, 48
12. All terminal ends and any of consequence --------------- (No B rea ks o f Sa f e ty Cons equ enc e )----------------

HP Injection, RC Pump seal supply, SI A 13.a. Pump Discharge (A or B) - terminal end None None 2400 gal. max., 2.6" mak. in HP Inj. Pump Bay 13.b. R. Bldg. - terminal end ~700 ga l . max. , 0. 5" max. In East Penetration Room 13.c. Any of consequence u u 2400 gal. max., 2.6" max. In HP inj. Pump Bay or 1.l" max. In East Penetration Room HP Injection, normal makeup, SI A '14.a. Pump Discharge (A or 8) - terminal end None None 2400 gal, max., 2.6" rex in HP inJ. Pump Bay 14.b. R. Bldg. - tenminal end 2400 gal. max., 1.1" max. In East Penetration Room 14.c. .Any of consequence U u 2400 gal. max., 2.6" max. In HP Im]. Pump Bay or 1.l" max. In East Penetration Room

L Table 2.3-3 Page 3 of 3 POSTULATED PIPE BREAKS - ENVIRONMENTAL EFFECTS----SYSTEMS NORMALLY IN OPERATION System and Postulated Break Pressure Temperature Flooding (Psig) (*F) (gallons and/or inches) Location (Case No.) HP Injection,. letdown line to orifice, SIA 2,633 gal, max., l.?" max. Negligible Negligible 15.a. R. Bldg. - terminal end In East Penetration Room Press. Red. -Station - terminal ends 2,633 gal max., 2.8" max. 15.b. In HP Inj. Pump Bay Any of consequence 4 1 2,633 gal max., 1.2" max. 15.c. In East Penetration Room or 2.8" max. In HP Inj. Pump Bay Notes: 1. (No Breaks of Safety Consequence) as used throughout this table means that line rupture effects

                               ,were analyzed and found to be negligible and insignificant with respect to the surrounding environment.
2. Maximum temperature of steam entering penetration room from isolated end of Main Steam line for period less than 5 seconds.

Table 2 3-4 Page 1 of 3 POSTULATED PIPE BREAKS - ENVIRONMENTAL EFFECTS----SYSTEMS NOT NORMALLY IN OPERATION System and Postulated Break Pressure Temperature Flooding Location (Case No.) (psig) (*F) (Gallons and/or inches) Aux. Stm. to RC Bleed / Misc. Waste Evaporators, 02A(I)

16. Terminal ends tnd any of consequence ------------- (No Brea ks of Sa fe ty Consequence)-------------

Emergency FDW Pump Discharge to Upper Surge Tank, 03A 17 Term *,al' ends and any.of consequence None None Negligible Emergency FDW Pump Discharge to Stean Generators, 03A 18.a. Emerg. FDW Pump Discharge terminal end None None 3380 gal . max. , 0.03 " max. in Turb. Bldg. basement 18.b. Main FDW Header - terminal end 3380 ga l . max. , 0.03 " max. In Turb. Bldg. Basement 18.c. R. Bldg.-terminal end 31,500 gal . max. , 14.6" max. in East Penetration Room or 46.4 " max. In West Pene-tration Room.2 18.d. All above Emerg. FDW Pump Discharge lines - U " 31,500 gal . max. , 0.3" any of consequence max. In Turbine Bldg. Basement or 31,500 gal, max., 14.6" max. In East Penetration Room or 46.4" max. In Penetration Room.2 Station Heating, 20

19. .All terminal ends and any of consequence Negligible Negligible Negligible

_ Steam Blanketing for HP FDW Heaters , 20(l)

20. All terminal ends and any of consequence ---------- -- (No Breaks of Sa fe ty Cons equ ence)--------------

Steam Blanketing for Moisture Separator Rhtr., 02A(I)

21. All terminal ends and any of consequence ------------- (No B reaks of Sa fe ty Cons equence )--------------

t Table 2.3-4 Page 2 of 3 POSTULATED PIPE BREAKS - ENVIRONMENTAL EFFECTS----SYSTEMS NOT NORMALLY IN OPERATION System and Postulated Break Pressure Temperature Flooding Location (Case No.) (psig) (*F) (Gallons and/or i nc hes ) HP Injection, Pump C Discharge, SI A 22.a. Pump Discharge terminal end None None 3,380 gal. max., 3.6" max. In Hr inj. Pump Bay 22.b. R. Bldg. - terminal end None 22.c. Any of consequence 4 3,380 ga l . max. , 3.6 " max. In HP Inj. Pump Bay or 10,125 gal. max., 14.9" max. In West Penetration Room LP Injection, 53A and B 23.a. Decay Heat Removal - all terminal ends and Negligible Negligible 24,632 cal. max., 47.9" max. In any of consequence LP Inj. Pump Bay; 26.2" max in HP Inj . Pump Bay, 36.3" ma x. in West Penetration Roon 4 None None 16,000 gal. max., 31" max.

 ]3.b. System Test - all terminal ends and any of                                                   in LP inJ. Pump Bay; 7 5 ma consequence inEastPenetrationRoom; max. In West Penetration Room or 17" max in HP Inj. Pump By RB Spray, 54A and B 24.a. Pump Discharge-terminal end                                 None               None          16,000 gal. max., 31" max.

In LP inj. Pump Bay 24.b. 3'.' Test Line-terminal end in East 16,000 gal . max. 7.5" max in Penetration Room East Penetration Roon 24.c. 3" Test Line-terminal end in West Pene- 16,000 gal. max., 24" max. tration Room. in West Penetration Roon' 24.d. All above RB Spray lines any of consequence 16,000 gal. max., 31" max, in LP Inj. Pump Bay; 7 5" max. In East Penetration Room; 24" max. In West Penetration Room or 17" max. In HP in). Pump Bay.

Table 2.3- 4 Page 3 of 3 POSTULATED PIPE BREAKS - ENGINEERING DATA----SYSTEMS NOT NORMALLY IN OPERATION System and Postulated Break Pressure Temperature Flooding Location (Case No.) (psig) (*F) (Gallons and/or inches) Fuel Cask Decontamination, 6)(23) 25 Pump Discharge-terminal end and any of ------------- (No Br eaks of Sa f e ty Consequ ence)------------ consequence .RC System Pressurizer Sample, 64(l)

26. Outside R. Bldg.-terminal end and any -------------(No Breaks of Safety Consequence)3------------

of consequence Steam Generator Secondary Side Sample, 64(2) 27 Outside R. Bldg.-terminal end and any -------------(No Breaks of Safety Consequence)3 ------------ of consequence Notes: 1. (No Breaks of Safety Consequence) as used throughout this table means that line rupture effects were analyzed and found to be negligible and insignificant with respect to the surrounding environment.

2. 46.4" max. water level in West Penetration Room will fall West concrete block wall relieving water to outside the Aux. Bldg. without adverse consequences.

3 These lines do not fall under the jurisdiction of the AEC guidelines; however, they were visually inspected in the station since they are field-routed lines and with results as described in Note 1,above

4. 36.3" max. water level in West Penetration Room will fall West concrete block wall relieving water to outside the Aux. Bldg. without adverse consequences.

Table 3.2 Page I of 4 Summary of Postulated Break Accident Analysis System and Postulated Break Location (Case No.) Station Situation Remedial Action Main Steam, OlA 1.a. Outside Reactor Bldg., West Line - Single Steam Generator Blowdown None terminal end- All systems, equipment necessary to mitigate accident available. 1.b. Penetration Room East Line - terminal Single Steam Generator Blowdown Protect HP-26 end Possible loss of I redundant channel of HP. l.c. Turbine Bidg. East and West Line - Double Steam Generator Blowdown None terminal ends All systems, equipment necessary to mitigate accident available. 1.d. Turbine Bldg. East and West Line - I Any of consequence Auxiliary Steam for Startup, 02A 3 All terminal ends and any of consequence Possible loss of Emerg. and Main Provide redundant sys FDW Systems. Possible loss of LP for long-term cooling and IIP. HP available manually. Aux. Service Water used for long-term cooling. Main Feedwater, 03 4.a. FDW Pump Discharge - terminal end Possible loss of one FDW line. All None systems, equipment necessary to mitigate

                     .                                      accident are available.

4.b. FDW Pump Discharge thru HP FDW Htrs. Possible loss of Emerg. and Main FDW Provide redundant sys to Turb. Bldg. Wall - any of consequence Systems. Possible loss of LP and HP. for long-term cooling HP available manually. Aux. Service Water used for long-term cooling. 4.c. FDW Pump Discharge.from Turb. Bldg. Wall Single Steam Generator Blowdown. Possible Prevent loss of HP to R. Bldg. - terminal end loss of I redundant channel of HP. by pipe restraints or protect HP

Table 3.2 Page 2 of 4 Summary of Postulated Break Accident Analysis System and Postulated Break Location (Case No.) Station Situation Remedial Action Turbine Extraction Steam to FDW Htrs, 06A and B - All terminal ends and any of consequence 6.a. HP, A & B Extraction Possible loss of I Emerg. FW line. None All systems, equipment necessary to mitigate accident are available. 6.b. LP, C, D & E Extraction I J i-Condensate Booster Pump Discharge, 078 7 All terminal ends and any of consequenco Possible loss of Emerg. and Main Feedwater. Provide redundant Auxiliary Service Water used for long-term sys for leag-term cooling. cooling FDW Heater Drains and Vents, 10 8.a. D Htr Drain Pump Discharge Possible loss of Main FDW. Emerg. FDW or None Auxiliary Service Water available for immedi-ate cooling; LP for long-term. E Htr Drain Pump Discharge i 8.b. HP injection, RC Pump seal supply, SIA 13.a. Puer.p Discharge (A or 8) - terminal end Loss of RC Pump seal supply. RC pumps None protected by canponent cooling during cooldown. System available for normal cooldown. 13.b. R. Bldg. - terminal end Any of consequence

                                                                                                     "                          _J i-13.c.

HP Injection, normal makeup, SI A 14.a. Pump Discharge (A or B) - term'nal end Loss of normal makeup. Use of HP inj. None line B for makeup during cooldown. System

                                                                                        ~

available for normal cooldown. 14.c, R. Bldg. - terminal end 1 14.c. Any of consequence

Table 3.2 Page 3 of 4 Summary of Postulated Break Accident Analysis System and Postulated Break Location (Case No.) Station Situation Remedial Action HP Injection, letdown line to orifice, SIA 15.a. R. Bldg. - terminal end Loss of letdown. RC System blowdown to None 1500 psig. HP-ES actuates to mitigate accident. Systems available for normal cooldown. 15.b. Press. Red. Station - terminal end 15.c. Any of consequence o _; ;_ Emergency FDW Pump Discharge to Upper Surge Tank, 03A 17 Terminal ends and cny of consequence Possible loss of Emerg. FDW. Normal None FDW system available. Emergency FDW Pump Discharge ti Steam Generators, 03A , 18.a. Emerg. FDW Pump Discharge - terminal end Possible loss of Emerg. FDW. Normal None FDW System available. 18.b. Main FDW Header  :?rminal end Possible loss of Emerg. and Main FDW Provide redundant sys Systems. Possible loss of LP and HP. for long-term cooling HP Avai lable manually. Aux. Service Water e, sed for long-term cooling. 18.c. R. Bldg. - terminal end Possible loss of Emerg. FDW. Normal None FDW system available. 18.d. All above Emerg. FDW Pump Discharge Lines l Aay of consequence I o HP injection, Pump C Discharge, SIA 22.a. Pump Discharge - terminal end Possible loss of I channel of HP. No rma l None shutdown. All systems, equipment for normal shutdown available. 22.b. R. Bldg. Isolation Valve inlet - krminal end 22.c. Any of consequence " _J __

Lble 3.2 Page 4 of 4 Summary of Postulated Break Accident Analysis System and Postulated Break Ren;edial Action Location (Case No. ) Station Situation LP injection, 53A and B 23.a. Decay Heat Removal - all terminal ends and Possible loss of I channel of LP In]. piping. None any of consequence Systems available for safe shutdown. RB Spray, 54A and B 24.a. Pump Discharge - terminal end Possible loss of I LP Injection, 1 DHR None and l RB Spray pump motors. Systems available for safe shutdown. 24.b. 3" Test Line - terminal end in East Loss of one channel of RB Spray. No safety Penetration Roc. problems. 24.c. 3" Test line - terminal end in West U Penetration Room 24.d. All above RB Spray lines, any of conse- Possible loss of I LP injection, 1 DHR _J L_ quence and 1 RB Spray pump motor. ' Systems avail-able for safe shutdown. Also possible loss of 2 HP Injection pump motors. System available for safe shutdown.

Table 4.2 Page 1 of 2 POSTULATED PIPE BREAK - STAT 10ll MODIFICATIONS AND SCHEDULE System and Postulated Break Estimated Time Lc.ca t ion (Case no.) Required Station Modification to implement Remarks Main Steam, OlA , i.b. Penetration. Room East Line - Terminal end Install light weight blewout panels in north and east wall 2  : of East Penetration Room 6 months See Fig. 4.2-Reinforce Battery Room walls, l.h remove existing restroom, shield & Fig. 4.2-1.b; LP inJ. line. 6 months

                                                                                           *Rarouta' electrical cable in East Penetration Room.                          3 months Auxiliary Stea.n for Startup, 02A 3      All terminal cnds and any of consequence                            Install Emerg. FDil Bypass lines around postulated pipe break area for both Stm Generators                   6 months     Sea Fi g. 4. 2                      Hain Feedwater,03 4.b.       FDU Pump Discharga thru HP FDW Htrs. to Turb.                   Install Emarg. FDW Evpass lines BlJg. i!all - any of consequence                               around postulated ruptu.             area for both steam generators                     6 monti.s
                                  ,.                                                                                                                      Sea Fig. 4.2-2 4.c.       FDW Pump Discharge from Turb. Bldg. wall to                     lastall !!ain FCU restraints                  6 months     See Fig. 4.2-4:
                             . R . Bldg. -termina l end                                       between R. Bldg. anchor and isolation check valvas Condentate Gooster Pump Discharge, 070 7      All tecminal ends' and cny of consequence                           Install Emerg. FDW Lypass line frca Unit 2 around postulated rupture area tying into bypass lines described in 3 and 4.b.

above. 6 months See Fig. 4.2-2 R av .' I 7-16-73

  • Cable will be rcroutedLin preference to previously proposed impingement deflector.

Table 4.2 Page 1 of 2 POSTULATED PIPE BREAK - STATION MODIFICATIONS AND SCHEDULE System and Postulated Break Estimated Time Location (Case no.) Required Station Modification to implement Remarks Main Steam, OlA 1.b. Penetration Room East Line - Terminal end install light weight blowout

           ,                                                  panels in north and east wall

___ of East Penetration Room 6 months See Fig. 4.2-Reinforce Battery Roon walls, l.h remove existing restroom, shield F. Fig. 4.2-1.b2 LP Inj. line. 6 months Install impingement deflector for electrical cable in East PenetrrjonRoom. 3 months Auxiliary Steam for Startup, 02A j

                                                                      /'

3 All terminal ends and any of consequence Install Emerg. FDW Bypass lines aro'nd u postulated pipe break area for both Stm Generators 6 months See Fig. 4.2-2 Main Feedwater,03 , 4.b. FDW Pump Discharne thru HP FDW Htrs. to Turb. Install Emerg. FDW Bypass lines Bldg.-Wall - any of consequence ' around postulated rupture area for both steam generators 6 months See Fig. 4.2-2 4.c. FDW Pump Discharge from Turb. Bldg. wall,.to Install Main F3W restraints 6 months See Fig. 4.2-4c R. Bldg. -terminal end ' between R. Bldg. anchor and

                                              ,e              isolation check valves 4

Condensate Booster Pump Discharge, 078,/ 7 All terminal ends and any of conpe'quence Install Emerg. FDW Bypass line j from Unit 2 around postulated rupture area tying into bypass lines described in 3 and 4.b. above. 6 months See Fig. 4.2-2

                                                                                             /                      ./

j ll [

Table 4.2 Page 2 of 2 POSTULATED PIPE BREAK - STATION MODIFICATIONS AND SCHEDULE -System and Postulated Break Estimated Time Location (Case No.) Required Station Hodification To implement Remarks Emergency FDW Pump Discharge to Steam Generators, 03A 18.b. Main FDW Header - terminal and Install Emerg. FDW Bypass lines 6 months See Fig. 4.2-2 around postulated rupture area for both steam generators Note: Postulated Rupture Locations listed in this table are only those which require change in the design and subsequent station modification.

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