ML20138F240

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Forwards Application to Eliminate Arbitrary Intermediate Pipe Breaks in High Energy Lines.Review & Concurrence Requested Prior to Upcoming Refueling outages.Marked-up Tech Specs Also Encl.Fee Paid
ML20138F240
Person / Time
Site: McGuire, Mcguire  Duke Energy icon.png
Issue date: 10/22/1985
From: Tucker H
DUKE POWER CO.
To: Adensam E, Harold Denton
Office of Nuclear Reactor Regulation
References
TAC-60026, TAC-60027, NUDOCS 8510250218
Download: ML20138F240 (55)


Text

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         *Re Peasement                                                                      TELEPHONE
       = = = = = = = =                            October 22, 1985                         U M*8 Mr. Harold R. Denton, Director l            Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 j           Attention:         Ms. E. Adensam Licensing Branch No. 4

Subject:

McGuire Nuclear Station. Units 1 and 2 Docket Nos. 50-369 and 50-370

Dear Mr. Denton:

Please find attached a technical document which discusse's the elimination of the arbitrary intermediate pipe breaks in high energy lines for McGuire Nuclear Station. The NRC has previously reviewed the climination of arbitrary intermediate i pipe breaks analysis at Catawba Nuclear Station, Unit 2 (Docket No. 50-414). The results of this evaluation were provided in a Safety Evaluation completed f in March 1984, transmitted to Duke by letter dated April 2, 1984 and included in Supplement No. 2 of NURG-0954, the Safety Evaluation Report for Catawba Nuclear Station. June 1984. For McGuire Nuclear Sation. Duke Power Company will apply an alternative pipe break criteria (excluding the RCS primary loop) as follows: '

1. Arbitrary interecdiate pipe breaks in all high energy piping systems be l

eliminated from the structural design basis when the following criteria are satisfied

a. For all piping systema, the stress criteria in McGuire FSAR Section 3.6.2 are not exceeded.
b. For Class 1 piping systems, the stress critoria and the usage factors in McGuire FSAR Section 3.6.2 are not exceeded.

2. The dynamic effects (pipe whip, jet impingement, and compartment pressur-iration loads) associated with arbitrary intermediate pipe breaks be excluded from the plant design basis.

3. Pipe whip restraints and jet shields associated with previously post lated arbitrary intermediate pipe breaks be climinated.

S 9. O l4)N Y/

Mr. Harold R. Denton, Director October 22, 1985 Page 2 ( ) The attached report provides the detailed technical justification which supports . ! this action. It is concluded that due to system design and operating procedures. ! , the probability of stress corrosion, thermal and vibrational fatigue or water l j hammer in the arbitrary intermediate pipe breaks locations is not significant. i Furthermore, the environmental analysis for McGuire is performed independent of the high energy pipe whip / jet impingement analysis. Therefore, the environ- ] mental analysis methods and results will not be affected by this change. l ) Duke Power has reviewed this proposed change in accordance with 10 CFR 50.59. Although no change to the Technical Specification is indicated by our review, we nevertheless request NRC prior review and concurrence on the acceptability j of this action. Prompt review and approval of this action is requested in ] order that the benefits identified might be achieved beginning with the forth

.      coming refueling outages of each McGuire unit. Following approval by NRC.

j the station modification process will be utilized to control the physical removal of the unneeded restraints and assure drawings are properly revised. Pursuant to 10 CFR 170.12. a license fee of $150.00 is enclosed. i i i j Very truly yours.  ! i kb. Y */Yl l Ital B. Tucker i l RLG/hrp i Attachment I [ 1 cc Dr. J. Nelson Grace. Regional Administrator i U. S. Nuclear Regulatory Commission , Region 11 l ! 101 Marietta Street. NW Suite 2900 i Atlanta, Georgia 30323 4 Mr. W. T. Orders NRC Resident inspector McGuire Nuclear Station

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1 i . 1 i I ! i 1 I l I

Duke Power Company McGuire Nuclear Station Application to Eliminate Arbitrary Intermmediate Pipe Breaks 1.0 Introduction At an ACRS Subconmittee meeting on March 28, 1983 and at the full ACRS Committee meeting on June 9, 1983, the staff proposed that arbitrary intermediate pipe breaks in high energy piping systems could be eliminated. Duke Power Company has determined that considerable benefit can be achieved by applying this NRC proposal at McGuire Nuclear Station while improving overall safety and piping system reliability. l In a letter from W.11. Owen to W. J. Dircks dated September 19, 1983, it was stated these changes will benefit McGuire Nuclear Station in a number of ways. Occupational radiation exposure will be reduced over the life of the station. Relief of congestion will improve access for operation and maintenance. Piping heat loss at whip restraint locations will be reduced. Overall plant safety will be improved, including a reduction in unanticipated restraint of piping thermal growth and seismic movement. Section 2.0 of this document provides the technical justification for elimination of the arbitrary intermediate pipe break analysis and a listing of the arbitrary intermediate break location proposed to be

 !      eliminated.

Section 3.0 provides a further discussion of the benefits to be derived at McGuire following NRC approval of this rejected action. Draft revisions to the McGuire FSAR are included as an Appendix to this report for informational purposes. Following NRC approval of this requested action, the FSAR will be updated appropriately. 2.0 Technical Justification The NRC has previously reviewed the elimination of arbitrary intermediate pipe break analysis at Catawba Nuclear Station, Unit 2 (Docket No. 50-414). The results of the evaluation were provided in a Safety Evaluation (SE) completed in March, 1984, transmitted to Duke by letter dated April 2, 1984 and included in Supplement No. 2 of NUREG-0954, the i Safety Evaluation Report for Catawba Nuclear Station, June,1984. The above SE adequately reflects several reasons previously provided by Duke for elimination of arbitrary intermediate pipe break analysis. The SE provides a thorough staff evil'iation of the Duke request and provides the l following conclusions: i 1. ASME_CodeClat.sjlPipingSystems "For Class 1 piping, a considerable amount of quality assurance in design, analysts, fabrication, installation, examination, testing, L

and documentation is provided which ensures that the safety concerns associated with the uncertainties discussed above are significantly reduced. Based on the staff evaluation of the design considerations given to Class 1 piping, the stress and fatigue limits provided in the MEB 3-1 break criteria, and the relatively small degree of uncertainty in the loadings, the staff finds that the need to postulate arbitrary intermed! ate pipe breaks in ASME Code Class 1 piping in which large unanticipated dynamic loads, stress corrosion cracking, and thermal fatigue such as in mixing situations are not present and in which all equipment has been environmentally qualified is not compensated for by an increased level of safety. In addition, systems may actually perform more reliably for the life of the plant if the request to postulate arbitrary intermediate break criteria for ASME Code Class 1 piping is eliminated."

2. ASME Code Class 2 and 3 Piping Systems "It can thus be concluded that when the piping designers have appropriately considered the fatigue effects for Class 2 and 3 piping systems in accordance with NC/ND-3645, the likelihood of a fatigue failure in Class 2 and 3 piping caused by unanticipated cyclic loadings can be significantly reduced.
    " Based on the staff evaluation of the design considerations given to Class 2 and 3 piping, the stress limits provided in the SRP break criterion, and the degree of uncertainty in unanticipated loadings, the staff finds that dispensing with arbitrary intermediate pipe breaks is justified for Class 2 and 3 piping in which stress corrosion cracking, large unanticipated dynamic loads, or thermal fatigue in fluid mixing situations are not expected to occur provided
1) the piping designers have appropriately considered the effects of local welded attachments per NC/ND-3645, and 2) all safety-related equipment in the vicinity of Class 2 and 3 piping systems have been environmentally qualified for the non-dynamic effects of a non-mechanistic pipe break with the greatest consequences on the equipment."

The following discussion addresses these piping system conditions for McGuire Nuclear Station. 2.1 Environmental Analysis The environmental analysis for McGuire Nuclear Station is performed independent of the high energy pipe whip / jet impingement analysis. Therefore, the environmental analysis methods and results will not be affected by the elimination of arbitrary intermediate breaks. 2.2 Protection from Stress Corrosion Cracking In order for stress corrosion cracking (SCC) to occur in piping, the following three conditions must exist simultaneously: high tensile stresses, a susceptible material, and a corrosive environment (NUREG-0691). Since some residual stresses and some t 2

degree of material susceptibility exist in any stainless or carbon steel piping, Duke Power minimizes the potential for stress corrosion by preventing the occurrence of a corrosive environment. Strict pipe cleaning standards prior to operation and careful control of water chemistry during plant operation are used to prevent the occurrence of this environment. All piping involved in the elimination of arbitrary breaks at McGuire is either austenitic stainless steel or carbon steel, as shown in Table 1. The Stainless steel is Type 304 and Type 316, and as such the carbon content is limited to a maximum of 0.08 weight percent. None of the higher carbon content types (304H, 316H) have been used. The corrodents known to increase the susceptibility of austenitic stainless steel to stress corrosion are (NUREG-0691): oxygen, fluorides, chlorides, hydroxides, hydrogen peroxide, and reduced forms of sulfur (e.g. , sulfides, sulfites, and thionates). In carbon steel, these same corrodents plus a few additional substances such as caustics and nitrates are thought to increase susceptibility. Prior to being put into service, piping at McGuire Nuclear Station is cleaned internally and externally, and water chemistry during flushes and pre-operational testing is controlled to maintain this cleanness according to written specifications. External cleaning for Duke Class A stainless steel piping includes patch tests to monitor and control chloride and fluoride levels. For preoperational flushes influent water chemistry is controlled with requirements on chlorides, fluorides, conductivity, and pH being included in the acceptance criteria for piping of the material type and class included in Table 1. During plant operation, primary and secondary side water chemistry is monitored in the carbon steel and stainless steel piping. Contaminant concentrations are kept below the thresholds known to be conducive to stress corrosion cracking. Table 1 shows the major water chemistry control standards for the lines in which arbitrary breaks were previously postulated. Oxygen content, the primary cause of stress corrosion cracking in BWR reactors, is more strictly controlled in the McGuire PWR environment than in a BWR environment. Oxygen concentration in the fluid in the McGuire stainless steel piping is expected to be less than 0.005 ppm during normal power operation, whereas the steady state oxygen content in a BWR system is approximately 0.2 ppm. Thus, this condition which facilitates cracking in BWR's is not present at McGuire. Note that a number of the lines involve operating temperatures less than 200'F. Any stress corrosion at these temperatures would be extremely slow. It is an industry-wide assumption that stress corrosion is not a problem at temperatures this low. Also note that steam generator water chemistry is the major m factor controlling steam generator' blowdown and main steam chemical composition. SCC has not proven to be a generic problem in PWR units. Over hundreds of PWR operating years of experience, there have been less'than two dozen SCC incidents reported to the NRC (compared with over 300 BWR incidents). In every case, an aggressive-

         - environment was created by a corrodent which is either not present or strictly controlled and maintained in the piping systems at McGuire. Thus, there is no evidence to suggest that SCC will occur at McGuire.

l l i-l' I i I i 4

4 WATER CHEMISTRY REQUIREMENTS DURING PLANT OPERATION Pcge 1.of 2 TABLE l' FOR LINES WITH PREVIOUSLY POSTULATED ARBITRARY BREAKS MAX. DUKE CHLORIDES & MAX. CATION NO. ARBITRARY PIPING PIPING OPERATING MAX. 02 HYDROGEN FLUORIDES CONDUCTIVITY

  • BREAKS ELIM.-

PIPING SYSTEM + MATERIAL CLASS TEMP (*F) (ppa) CONCENTRATION (ppm) (paho/cm) pH* MCGUIRE 1&2 Reactor Coolant' SS A 557 0.10** 25-50cc/kg(H 2O) 0.15 - - 15 Rssidual

Heat Removal SS A 557 0.10** 25-50cc/kg(H 2O) 0.15 - -

0 l Auxiliary Feedwater CS B&C 134-445 <0.10 - - - 8.8-9.3 0 i Msin 4 Feedwater CS B 445 <0.005 - - - 8.8-9.3 22 1 i Safety j- Injection (1) SS A 618 0.10** 15-50cc/kg(H 2O) 0.15 - - 20 Safety ! Injection (2) SS A Ambient ++ - - 0.15 - Low 7 1 Safety l Injection (3) SS B Ambient ++ - - 0.15 - Low 0

;       Steam Generator 0.02                     8.5-9.3

~ Blowdown (1) SS B 557 - - 0.8 23 j Ches & Vol. Control (1) SS A&B 130++ - - 0.15 - - 22 Chen & Vol.

Control (2) SS B&C 290 - -

0.15 - - 9 i 1 Main l Steam CS B&F 557 - - - 0.3 8.8-9.3 23 i I i i

WATER CHEMISTRY REQUIREMENTS DURING PLANT OPERATION Page 2 of 2 TABLE 1 FOR LINES WITH PREVIOUSLY POSTULATED ARBITRARY BREAKS MAX. DUKE CHLORIDES & MAX. CATION NO. ARBITRARY PIPING PIPING OPERATING MAX. 02 HYDROGEN FLUORIDES CONDUCTIVITY

  • BREAKS ELIM.-

PIPING SYSTEM + MATERIAL CLASS TEMP (*F) (ppm) CONCENTRATION (ppm) (pmho/cm) pH* MCGUIRE 1&2 Main Steam to Aux. Equip. CS B 557 - - - 0.3 8.8-9.3 I Main Steam Vent to Atmos CS B 557 - - - 0.3 8.8-9.3 16

     ++ Stress corrosion not considered a problem at this operating temperature.
  • Based on EPRI-NP-2704-SR (Steam Generator Owner's Group Secondary Water Chemistry Guidelines.

These standards are met at McGuire Nuclear Station).

     + Data is for the portions of the system where arbitrary breaks were previously postulated.
     ** Tech. Spec. limit; Oxygen concentration expected to be less than 0.005 ppm during most power operation.

2.3 Protection from Thermal and Vibration Fatigue For McGuire Nuclear Station non-class 1 ASME Code lines in general, the Code design allowables are intended to prevent fatigue failure. For Class 2 and 3 piping components, fatigue failure protection is provided for by the allowable stress range check for thermal expansion stress. This stress is included in the break stress ratio for all non-class 1 breaks. And even after elimination of the arbitrary intermediate breaks, the cut-off for postulating mandatory breaks (" threshold") is still 80% of the Code allowables. For Class 1 (Duke Class A) lines the conservatism allowed for fatigue failure is even more obvious. The ASME Code limit for the Cumulative Usage Factor (CUF) is 1.0 to assure that pipe failure will not occur. The pipe break postulation limit is 10% of this number, and most of the Class 1 arbitrary intermediate break locations involve CUF's far below this limit. NUREG-0691 reported a number of instances of pipe cracking due to thermal fatigue in PWR systems. All cases involved either feedwater or auxiliary feedwater piping and occurred in the vicinity of steam generator nozzles. The cracking was attributed to cyclic thermal stresses in horizontal runs at the nozzles. Cyclic thermal stress is minimized in the main and auxiliary feedwater piping at McGuire Nuclear Station by limiting mixing of low velocity, low temperature feedwater with high temperature water in the steam generator nozzles. Mixing is prevented in the auxiliary feedwater supply by providing vertical or upward sloping piping in the vicinity of the steam generator. Auxiliary feedwater instrumentation is provided near the inlet nozzle to monitor high temperature backflow. Mixing of low velocity, low temperature feedwater with high temperature water is prevented in the main nozzle by isolating flow to the main nozzle and using the auxiliary feedwater nozzle, at a power level below 17%. Before transferring feedwater flow to the main nozzle, cold water is purged from the main feedwater line. A controlled rate of high temperature flow from the Steam Generator main nozzle in the reverse flow direction, around the containment check valve, through the 1 inch bypass line, through the feedwater isolation bypass valve and to the feedwater bypass tempering lines purges cold water away from the steam generators to the condensers. To ensure that water in the line is above the required temperature before the feedwater isolation valves are opened, temperature is monitored in the main feedwater lines. At McGuire Units 1 and 2, twenty-two (22) of the approximately 158 eliminated arbitrary intermediate breaks occur in the main l feedwater system (none in auxiliary feedwater system), and none j of these are located in horizontal pipe runs at the nazzles. l t

Cyclic thermal stress is prevented in the arbitrary break lines in the remaining systems (listed below) by maintaining uniform temperatures with no mixing:

1. Steam Generator Blowdown
2. Residual Heat Removal
3. Main Steam Supply to Auxiliary Equipment
4. Main Steam
5. Main Steam Vent to Atmosphere
6. Reactor Coolant
7. Safety Injection
8. Chemical and Volume Control The potential for vibration fatigue in McGuire piping systems is minimized through pre-operational vibration tests. The purpose of the tests is to verify the following:

A. Piping layout and support / restraints are adequate to withstand normal transients without damage to piping system, and B. Flow induced vibration is sufficiently small to cause no fatigue or stress failures in the piping system. During the tests, points on the piping systems with large displacements are selected for measuremer.t of piping velocity. These measurements are evaluated with respect to the following acceptance criteria and any required modifications to achieve acceptability are made: A. Steady State Vibration Testing Acceptance ~ criteria are based on conservatively estimated stresses which are derived from measured velocities and conservatively assumed mode shapes. B. Transient Vibration Testing

a. No permanent deformation or damage in any system, structure, or component important to nuclear safety is observed.
b. All suppressors and restraints respond within their allowable ranges.

. c. The measured piping vibration for Reactor Coolant System during reactor coolant pump starts and trips do

not exceed the values-specified by Duke Power Company Design Engineering Department.

All piping systems that contain rupture devices to be deleted under our proposal are included in the Vibration Test Program. McGuire FSAR Table 3.9.2-1 lists the systems to be included in 1 the steady state test programs. This table is included on the following page. Transient response of piping due to valve closures, pump starts, and other changing configurations are observed as part of the preoperational test program. Furthermore, the systems which are subject to rapid changes in temperature during plant transients such as the Main Steam System and the Feedwater System also receive additional preoperational visual inspections as described in McGuire FSAR Chapter 14. Based on the information presented in this attachment, no problems from thermal or vibrational fatigue would be expected i at McGuire Nuclear Station. 1 2 ? i i 4 1 i [ l i i i I [ I

!                                                          FSAR Table 3.9.1-1 l                                   Piping Systems Included in Vibration Test Program 1

System t Reactor Coolant System

                  ' Safety Injection System Residual Heat Removal System Containment Spray System Chemical and Volume Control System 4                   Boron Recycle System 4
Boron Thermal Regeneration System Component Cooling System

[ Liquid Waste Disposal System Fuel Pool Cooling and Cleanup System 4 Diesel Generator Fuel Oil System l Diesel Generator Cooling Water System Diesel Generator Lube Oil System Nuclear Service Water System Refueling Water System j Main' Steam System Feedwater System Auxiliary Feedwater System Steam Dump System Containment Ventilation Cooling Water System Control Area Chilled Water System Steam Generator Blowdown Recycle System l Recirculation Cooling Water System

2.4 Protection from Water / Steam Hammer The potential for water hammer may exist in some of the systems involving lines where arbitrary intermediate breaks are being eliminated (e.g. , Auxiliary Feedwater and Steam Generator Blowdown). However, steps have been taken to minimize the probability of significant water hammer actually occurring in these lines. Water hammer in each of the systems involved in elimination of arbitrary breaks is discussed below:

1. Steam Generator Blowdown Fluid flow through the Steam Generator Blowdown Lines is normally two phase and of 0-10% quality. There is very little chance of water hammer in these lines inside Containment. A greater susceptibility to water hammer exists outside Containment, with the greatest potential occurring upon reinitiation of flow following containment isolation. Some problems have occurred in this area at McGuire. However, design and operating procedures have been improved to provide provisions to gradually repressurize the downstream piping before establishing full flow. Since these changes have been initiated, there has been no recurrence of water hammer induced damage.
2. Auxiliary Feedwater & Steam Generator Wet Layup Recirculation The Auxiliary Feedwater System is designed to minimize the probability of significant water hammer occurrence. The nozzle at the steam generator involves a 90' elbow connecting immediately to a verticle run of pipe to minimize steam voids. Also, tempering flow is maintained so that the line will be filled with water at all times.

However, it is recognized that some potential for water hammer in the auxiliary feedwater lines exists. Consequently, the temperature in these lines is monitored so that they may be filled slowly and flow initiated gradually when steam voids are t.uspected. The piping involved in the arbitrary interaediate breaks contains thermowells allowing such temperature increases to be detected and the proper operating procedures to be implemented; also the associated valves are periodically checked for leaks. McGuire has experienced no water problems to date. A short segment of Steam Generator Wet Layup Recirculation Piping and associated valves are required to isolate the Auxiliary Feedwater System. This portion of the system is part of the Auxiliary Feedwater System Pressure boundary. No other Steam Generator Wet Lay-up Recirculation Piping is involved in arbitrary intermediate breaks since there is no flow through the system during power operation.

3. Reactor Coolant Overall, there is a very low potential for water hammer in the Reactor Coolant System since it is designed to preclude steam void formation.

Some problems were experienced initially at McGuire, but shut-down procedures have been modified to eliminate water hammer.

4. Residual Heat Removal The portions of the Residual Heat Removal System having arbitrary intermediate breaks are within the Reactor Coolant System boundary prior to (or at) the first valve off the Reactor Coolant Loops. Therefore, the explanation in part 3 above is applicable here.
5. Safety Injection The Safety Injection System lines are all either water solid or gaseous lines. In general, valves in the system are slow acting and operating procedures are designed to prevent water hammer. There would obviously be no problem in the gaseous lines. Therefore, there is a low probability of water hammer problems in this system.
6. Chemical & Volume control & Boron Thermal Regeneration The majority of the arbitrary intermediate breaks in the Chemical and Volume Control System are in low temperature (130*F operating) lines. These normally water solid lines would have a very small probability of steam void formation, and no water hammer events would be expected.

For the breaks (outside Containment) in higher temperature lines (290*F), operating procedures for the system have been developed which minimize the probability of water hammer occurrence.

7. Main Steam Supply to Auxiliary Equipment These lines are provided with extra drain capacity to remove all condensate thereby preventing water hammer.

Periodic testing verifies that the drains are functioning properly.

8. Main Steam The main steam lines are provided with drains to continuously remove condensate. Significant water or steam hammer will not occur without liquid.

The drain lines contain condensate normally, but could also contain steam. No water hammer problems exist in these lines which are used to dump condensate to the condenser.

9. Main Steam Vent to Atmosphere There is no chance of water hammer in the Main Steam Vent Atmosphere System as long as the Main Steam line stays drained. Experience-has shown this to be the case.
10. Main Feedwater McGuire has Westinghouse preheat steam generators.

Westinghouse recommendations have been followed in the design and operation of the feedwater system to prevent water hammer in the steam generators. Procedures have been incorporated to control temperature and pressure to prevent void formation in the feedwater system piping during normal modes of operation. Following feedwater system trips, void formation is possible; hcwever, operating procedures and design considerations prevent water hammer during restart by providing a controlled flow to flush the voids from the system slowly. 2.5 Summary of Breaks to be Eliminated Table 2 provides the arbitrary intermediate breaks proposed to be eliminated by the requested change. l .I l

Table 2 Postulated Arbitrary Intermediate Breaks to be Eliminated on McGuire Nuclear Station Unit 1 Estimated No. Devices Eliminated Pipe Number Breaks Rupture Jet Piping System Diameter Location Eliminated Restraints Deflectors Steam Generator 2" IC 6 5 1 Blowdown 2 OC 2 0 0 Main 2 OC 5 3 0 Feedwater 6 OC 1 0 0 18 OC 8 10 0 Reactor 2 IC 2 0 0 Coolant 3 IC 2 0 0 14 IC 2 3 0 Safety 2 IC 2 4 0 Inj ection 6 IC 4 4 0 8 IC 4 0 0 Chemical & Volume 1 IC 1 0 0 Control 2 IC 6 0 0 3 IC 1 1 0 2 OC 1 0 0 3 OC 3 0 0 Main Steam 2 OC 7 0 0 Main Steam Vent 6 OC 8 0 0 to Atmosphere Total 65 30 1

Table 2 (Cont'd) Postulated Arbitrary Intermediate Breaks to be Eliminated on McGuire Nuclear Station Unit 2 Estimated No. Devices Eliminated Pipe Number Breaks Rupture Jet Piping System Diameter Location Eliminated Restraints Deflectors Steam Generator 2" IC 9 19 1 Blowdown 2 OC 6 2 3 Reactor 2 IC 7 0 2 Coolant 14 IC 2 0 3 Safety 2 IC 5 5 0 Inj ection 5_ IC 4 0 0 6 IC 4 0 0 8 IC 4 0 0

 . Chemical & Volume       2           IC            11            4            1 Control                  3           IC             2            0            1 2'          OC             4            0            0 3           OC             2           -0            0 Main Steam               2           OC            16            0            0 Main Steam Vent          6           0C             8            0            0 to Atmosphere Main Steam Supply        6           OC.            I            1            0 To Auxiliary Equip.

Main Feedwater 18 OC 8 10 3 Total 93 41 14 l

3.0 Benefits from Elimination of Arbitrary Intermediate Breaks This information, along wit', that provided in Section 2.0, the basis for the conclusion provides that an overall gain in plant safety can be achieved by deleting arbitrary intermediate breaks and their associated protective devices. The elimination of arbitrary intermediate breaks would, at any point throughout the plant life, increase our confidence in the integrity of plant systems. The removal of rupture restraints and jet deflectors associated with these arbitrary intermediate pipe breaks would increase the visibility and detectability of piping related problems (e.g., inadvertent restraint) during routine maintenance activities, as well as provide more access for ISI activities. Occupational radiation exposure will be reduced over the life of the station. Piping heat loss at whip restraint locations will be reduced. Overall plant safety will be improved, including a reduction in unanticipated restraint of piping thermal growth and seismic movement. If welded piping attachments (other than shear lugs) that could contribute to high local stress concentrations are located in the vicinity of any arbitrary intermediate breaks, then these breaks will not be eliminated. Alternatively, unacceptable welded attachments will be relocated out of the vicinity of eliminated arbitrary intermediate breaks. Since there is a certain amount of ductility in any of the stainless steel or carbon eteel piping associated with the piping systems contain these arbitrary intermediate breaks, any piping failure is expected to be a leak-before-break phenomena. The piping materials are not types where 4 sudden catastrophic failures can be expected. Therefore, especially inside containment where leak detection capabilities consistent with Regulatory Guide 1.45 requirements are provided, leak detection would aid in protecting against catastrophic double-ended pipe rupture. Outside containment the additional mitigating factor of physical separation of redundant trains exists. There, structural design and plant layout provide important added layers of protection from piping failures of any type. , Elimination of arbitrary intermediate breaks at McGuire Nuclear Station i will result in an approximate 20 percent reduction in the total number of pipe rupture protection devices. The eliminated devices are scattered throughout the unit in the same proportion as the original device

~

requirements. The percent of devices inside and outside containment remains the same after the deletions. The distribution inside containment tends to be relatively even throughout the four quadrants because the piping math models involved are associated with the four Reactor Coolant System loops. Consequently, there will be no concentration of eliminated arbitrary intermediate break devices in any one area. Hence, we conclude that an adequate level of protection from pipe rupture will remain. The eliminated devices will be removed over a period of time as advantageous to plant operation and maintenance activities. In the highly unlikely event that a pipe break were to occur, there exists a degree of protection even without pipe rupture devices. The currently I

i designed support system (which is not, in any way, altered by this submittal) would provide a measure of restraint, if needed. These structures are conservatively designed to withstand forces such as piping dead weight, thermal loads, anchor movements, and earthquake loads. The density of the supports on the piping is such that one or more supports (e.g., struts, snubbers, rigid supports) will absorb a portion of the energy from any pipe break. The foregoing discussion is presented in s'ipport of the overall safety for McGuire Nuclear Station for the elimination. of arbitrary intermediate breaks. A summary ir provided in Table 3 attached. 5 Table 3 Summary of Benefits from the Elimination of Arbitrary Intermediate Pipe Breaks on McGuire Nuclear Station Units 1 & 2 Category Benefit

1) Relief of congestion, 48 person-rem reduction in improving access for operation radiation exposure over life of and maintenance plant including removal of exist-ing protective devices.

($36,000).

2) Reduction in piping heat loss Not quantitatively assessed, at whip restraint locations Insulation can be installed on piping at current locations of arbitrary break pipe whip restraints.
3) Improvement in overall plant Improvement in ISI quality.

safety (NUREG/CR-2136) Elimination of potential for restricted thermal movement.

4) Design, material and erection Estimated savings of $460,000, cost associated with future Additional protection for target modifications or additions pipe and/or source pipe relocation requiring protective devices for can be minimized through arbitrary intermediate breaks. implementation of Arbitrary Estimated number of protective Intereediate Break elimination.

devices that would be required for these modifications or additions is 10. Total Savings: $496,000 and 48 person-rem for a plant life of 40 years.

APPENDIX I Revision to McGuire FSAR For Arbitrary Intermediate Break Criteria Change 7 b 1 l J

3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING 3.6.1 SYSTEMS IN WHICH DESIGN BASIS PIPING BREAKS OCCUR

3. 6.1.1 Reactor Coolant System ,

In the design of a pressurized water reactor, special provisions are made for protecting the public against the consequences of major mechanical accidents, including a loss-of-coolant or steam line break accident. Consideration is given to: Reactor coolant branch line breaks (Ref. 3.6.1.2) Reactor coolant pipe breaks

'      Section 3.6 of this SAR defines the extent of the allowable mechanical damage considered in these accidents, the various systems and equipment which are necessary to recover from these accidents and the mechanical provisions which are provided to prevent unacceptable extension of the accident consequences.

The particular arrangement of the Reactor Coolant System, building structures and mechanical restraints preclude the formation of plastic hinges for breaks postulated to occur in the Reactor Coolant System. Consequently, pipe whip and jet impingement effects of the postulated pipe break will not damage necessary safety-related structures, mechanical or electrical systems and equipment required to mitigate the consequences of the postulated break. The Reactor Coolant System as used in this portion of the SAR is limited to the main coolant loop piping and all branch connection nozzles out to the first butt weld. The application of criteria applied for protection against the effects of postulated breaks in the Reactor Coolant System' Pressure Boundary in accor-dance with WCAP-8172 (1) results in a system response which can be accom- ,. modated directly by the supporting structures of the reactor vessel, the steam generator and the reactor coolant pumps including two additional pipe supports. The design basis for postulated breaks in the Reactor Coolant System are discussed in 3.6.2.1.

3. 6.1. 2 All Other Mechanical Pioing Systems This subdivision discusses all piping systems excluding the Reactor Coolant System as described in 3.6.1.1 and is in accordance with t!RC Branca Technical Position APSCSB 3-1 and Regulatory Guide 1.46 except as noted in Table 3.6.1-3.

4 Other mechanical piping systems, both inside and outside Containment, which are reviewed and considered in the design with respect to a postulated pipe ' break are those normally operating high energy and moderate energy lines which are safety related or which pass near safety-related structures, systems or components, and include the Reactor Coolant System branch piping terminating at the main coolant loop piping nozzle. t - 3.6-1 12/63

                                                                                                  , ,                               m - .
   --        -   - + - -       -       a    ng n ,- ---n-n   ------w, , ap<--nm . , - --we- - - -    ----,n e_,-,--,gwm. -:,-,-,   --m m  - . -em4

High-energy piping systems are those systems, or portions of systems, that during normal plant conditions are either in operation or maintained pres-surized under conditions where either or both of the following are met:

                                                                                      )

a) Maximum temperature exceeds 200 F, or b) Maximum pressure exceeds 275 psig.

                        ~

Exceptthat(I)n'on-l'iquidpipingsystems(air, gas, steam)withamaximum pressure less than or equal to 275 psig are not considered high energy regardless of the temperature, and (2) for liquid systems other than water, the atmospheric boiling temperature can be applied. Systems are classified as moderate energy if the total time that either of the above conditions are met is less than either of the following. a) One (1) percent of the normal operating lifespan of the plant, or b) Two (2) percent of the time period required to accomplish it's system design function. Moderate energy lines are defined as those which have: a) A maximum operating temperature less than or equal to 200 F, and b) A maximum operating pressure less than or equal to,275 psig. Systems which do not contain mechanical pressurization equipment are excluded from moderate energy lines; i.e., systems without pumps, pressurizing tanks, boilers, etc. , and wnich operate only from gravity flow or storage tank water ( head are not considered moderate energy. Open ended vents and drains and piping furnished as a part of equipment are also not considered moderate energy. ( Systems or. appropriate portions which fall in either or both of the above categories are analyzed as described in 3.6.4.2 and protected in accordance with 3.6.5.2. Table 3.6.1-1 lists High Energy Systems or portions thereof and Table 3.6.1-2 lists moderate energy systems or portions thereof in- accordance. with -the above definitions that are analyzed for the station. Subdivision 3.9.2.8 discusses Containment i.Megrity with respect to breaks involving mechanical penetrations. 3.6.2 DESIGN BASIS PIPIf1G BREAK CRITERIA 3.6.2.1 Postulated Systen Pioino Brea< Location Criteria for the Reactor Coolant The design basis for postulated pipe breaks of the reactor coolant loop piping should include not on!y the break criteria, b'st also the criteria to protect other piping and vital systems from the effects of. the postulated break. ' A loss of reactos coolant accident is assumed to occur for a pipe break down to the restraint.of the second normally open automatic isolation valve (Case 11 3.6-2 12/83 Ri. ~ u%

                                                                                                                                               ~        ..                - - . - _ -

k i in Figure 3.6.2-1) on outgoing line (*) and down to and including the second check valve (Case 111 Figure 3.6.2-1) on incoming lines normally with flow. A pipe break beyond the restraint or second check valve does not result in an i uncontrolled loss of reactor coolant if either of the two valves in the line close. Accordingly, both of the automatic isolation valves are suitably protected and restrained as close to the valves as possible so that a pipe break beyond the restraint does not jeopardize the integrity and operability of the valves. i l Further, periodic testing capability vf the valves to perform their intended function is essential. This criterion takes credit for only one of the two valves performing its intended function. For normally closed isolation or incoming check valves (Case I and IV in Figure 3.6.2-1) a loss of reactor coolant accident is assumed to occur for pipe breaks on the reactor side of the valve. 1 4 In any given piping system, there is a limited number of locations which are i more susceptible to failure by virtue of stress or fatigue than the remainder of the system. These postulated break locations are defined in 3.6.2.1.1 considering normal and upset operating conditions (defined by the applicable

  • Design Specification as required by ASME Code, Section III).

Engineered Safety Features are provided for core cooling and boration pressure reduction, and activity confinement in the event of a loss of reactor coolant or steam or feedwater line break accident to ensure that the public is protected in accordance with 10CFR100 guidelines. These safety systems have been designed i to provide protection for a Reactor Coolant System pipe rupture of a size up to { and including a double ended severance of the Reactor Coolant System main loop. Branch lines connected to the Reactor Coolant System are defined as "small" if they have an inside diameter equal to or less than 4 inches. This size is such that Emergency Core Cooling System analyses using realistic assumptions show that no clad damage is expected for a break area of up to 12.5 square inches corresponding to 4 inches inside diameter piping. l Inordertoassurethecontinuedintegrityofthevitalcomponentsandthe[ngi-neered safety systems, consideration is given to the consequential effects of the pipe break itself to the extent that; a. The minimum performance capabilities of the engineered safety systems are not reduced below that required to protect against the postulated break; b. The Containment leaktightness is not decreased below the design value, if the break leads to a loss of reactor coolant; (**) and (*) It is assumed that motion of the unsupported line containing the isolation valves could cause failure of the operators of both valves. 4 i i (**) The Containment i I s here defined as the Containment vessel and penetrations, i ( and the steam generator shell, the steam generator steam side instrumenta-(- tion connections, the steam feedwater, blowdown and steam generator drain pipes within the Containment structure. 3.6-3 12/83 I l e y , 4 ',f'* -

       =,--nm- -g,    ---,,-,,..,ww-<wy,       , - --     ,.~,-,,,,m          e,      -     e     v--raw %e    ---*-*a-?p r - e
  • v- --W-- ww- *w--M-ec~-++-e w -w-r * -- y v

l

c. Propagation of damage is limited in type and/or degree to the extent that:

1)- A pipe break which is not a loss of reactor coolant does not cause a loss of reactor coolant or steam or feedwater line break, and (!

2) A reactor Coolant System pipe break does not cause a steam-feedwater i

system pipe break and vice versa. In the unlikely event that one of the small pressurized lines should fail and - result in a loss of reactor coolant accident, the piping is restrained or arranged to meet the following additional criteria in addition to (a. through c.) above: ) a. Break propagation must be limited to the affected leg, i.e. , propagation to the other leg of the affected loop and to other loops is prevented; The RTD Bypass line are exceptions to this particular criteria. A break in one of these line is allowed to propagate to either of the other RTD lines in the affected loop. 1 This is permitted since:

1) A break in a RTD line effectively results in propagation to the other legs by flow through the unbroken RTD lines, ,
2) An analysis of dual aperture breaks has been performed and the results indicate that the limiting small break case bounds the possible multiply break cases for the RTD lines. (Reference 2)
b. Propagation of the break in the affected leg is parmitted but is limited s

to a total break area of 12.5 square inches (4 inch inside diameter). The exception to this case is when the initiating small break is the high head safety injection line. Further prcpagation is not permitted for d this case; i e c. Damage to the high head safety injection lines connected to the other leg of the af fected loop or to the other loops is prevented; and

d. Propagation of the break to high' head safety injection line connected to affected leg is prevented if the line break results in a loss of core ,

cooling capability due to a spilling injection line. 3.6.2.1.1 Postulated Piping Break Locations and Orientations In each leg of the Reactor Coolant System, a minimum of three postulated l rupture locations shall be selected in the following manner: Breaks shall be postulated at the terminal points and at all locations in a run or branch in which the cumulative usage f actor exceeds 0.2 for normal ano upset operating conditions or in unich the range of primary plus seconocry stress intensity for normal and upset operating conditions exceeos 80 percent ! of the ASME Section III Code allowable on an elastic basis (2.4 S ). In the event that a location between the terminal points cannot be chose 0 in this i manner, the point of highest fatigue usage shall be used to obtain a total of three break locations. 3.6-4 12/83

                                                                                                                                ..             ~           a                 -

_ , , , . , _ . , , - , - - - - - - - ~,-+w * ~ ~ ' ' ' ' ' - ~ - " * ~ ' " " ' " '" ~ "

At each possible break location, consideration must be given to the occurrence of either a circumferential or longitudinal break. As discussed in WCAP-8172 a circumferential rupture is more likely than a longitudinal rupture for reactor coolant lated. piping. Only in the case of one elbow is a longitudinal rupture postu-Circumferential breaks are perpendicular to the longitudinal axis of the pipe. Longitudinal breaks are parallel to the longitudinal axis of the pipe. Certain longitudinal break orientations may be excluded on the basis of the state of stress at the location considered. For the main reactor coolant piping system, eleven discrete break locations were determined by stress and fatigue analyses. The locations are given in Table 3.6.2-1 and shown in Figure 3.6.2-2. The postulated locations conform to the criteria stated above and are discussed in WCAP-8172. Break orientation at each discrete break location is presented in Table 3.6.2-1. The results of the analyses which lead to the break orientations are discussed in WCAP-8172. The primary plus secondary stress intensity ranges and the fatigue cumulative usage factors at the design break locations specified in WCAP-8172 are given in Table 3.6.2-2 for a reference fatigue analysis. The reference analysis has been prepared to be applicable for many plants. It uses seismic umbrella moments which are higher than those used in WCAP-8172 such that the primary stress is equal to the limits of equation 9 in NB 3650 (Section III of the ASME code) at many locations in the system. therefore, the results of the reference analysis may differ slightly from WCAP-8172, but the philosophy and conclusions of the WCAP are still maintained. There are no ether locations in the model used in the reference fatigue analysis, consistent with WCAP-8172, where the stress intensity ranges and/or usage factors exceed the criteria of 2.4 S,and 0.2, respectively. Actual moments for the McGuire units are also given in Table 3.6.2-2 so that the reference fatigue analysis can be shown to be applicable for McGuire. By showing actual moments to be no greater than those used in the reference analysis, it follows that the stress intensity ranges and usage factors for McGuire model. plant will be less than those for comparable locations in the reference By this means it is shown that there are no locations other than those identified in WCAP-8172 where the stress intensity ranges and/or factors for McGuire might exceed the criteria of 2.4 S and 0 applicability of WCAP-8172 to McGuire has 8een ver.2, respectively. Thus, the ified. 3.6.2.1.2 Postulated Piping Break Sizes For a circumferential break, the break area is the cross-sectional area of the pipe at the break location, unless pipe displacement is shown to be less by analysis, experiment or physical restraint. For a longitudinal break, a break area less than the cross sectional area of the pipe may be assumed when analytically or experimentally substantiated. In e ! 3.6-5 1984 Update l L

the absence of this data, the break area shall be assumed to be the cross-sectional area of the pipe and the break length shall be assumed to be two pipe diameters. 3.6.2.1.3 Line Size Considerations for Postulated Piping Breaks Branch lines connected to the Reactor Coolant System are defined as "large" for the purpose of this criteria as having an inside diameter greater than 4 inches ur to the largest connecting line, generally the pressurizer surge line. Pipe break of these lines results in a rapid blowdown of the Reactor Coolant System and protection is basically piovided bf the accumulators and the low head safety injection pumps (residual heat removai pumps). 3.6.2.2 General Desien Criteria for Postulated Pininn Breaks Other Than Reactor Coolant System

a. Station design considers and accommodates the effects of postulated pipe breaks with resoect to pice whip, jet imoinaement and resultino reactive forces for piping both inside anc outside Containment. The anaiyticai method utilized to assure that concurrent single active comoonent failure and pipe break effuts du not jeopcrdiza the safe shutdown of the reacter are outlinadin Figure 3.6.2-3.
b. Station general arrangement and layout design of high energy systems utilizes the possible combination of pnysical separation, pipe bends, pipe whip restraints and encased or jacketed piping for the most practical design of the station. These possible design comoinations cecrease postu-lated piping break consequences to minimum and acceptable levels. In all cases, the design is of a nature to miticate the consequences of the (

( break so that the reactor can ce shutdown safely and maintained in a safety shutdown condition.

c. The environmental effects of pressure, temperature and floccing are con-trolled to acceptable levels utilizing restraints, level alarms and/or other warning cevices, vent openings, etc. ',
d. Plant Operating Canaitions
1) Powar Level - At the time of the costulated pipe break, the plant is assumea .o be in the nor:al ,nde of plant coeration, in which the piping un':er investiratien eroeriertes the maximum ccnditions of pres-sure and tercerature. In cases wnere tnis moae is fuli puwer opera-tion, the cc.mr level assu ed is that assumed in the evaluation of the loss- c- w: nt c: nnt, narnim trea< ccc: cent. P 19euwater line break acticent, in Chapter 15 of the safety analysis report.
2) Of f site Pour - If the pipe creak resuits in a loss of-coolant accident, '

steam eine creak accicent, or f ?edsa ter ' ine nr,ak ;ccicar.t, a loss of offsite couer is assume <f to accur cuasecuent to the pipe ructure.

3) Seismic Loiciros equivalent to eithir '.n? site ?,utdown Earthquaxe (53E) or tn d ~Irating Basis P irthau'ke i rP.E ), ,s appropriate, will be usea in '.n2 analysis ci ? pint i l, quir ,ent, n otectiva devices, etc. -

3.5-6 12/83

e. Consideration is given to the potential for a random single failure of an active component subsequent to the postulated pipe rupture. Where the

(- postulated piping break is assumed to occur in one of two or more redun-dant trains of a dual purpose moderate-energy essential system, i.e., one required to operate during normal plant conditions as well as to shut down the reactor and mitigate the consequences of the piping rupture, single failures of components in the other train or trains of the system only are not assumed, provided the system is designed to seismic Category 1 standards, is powered from both offsite and onsite sources, and is con- . structed, operated, and inspected to quality assurance, testing, and in service inspection standards appropriate for nuclear safety systems.

f. In the event of a postulated break in the piping in one unit, safe reacto" shutdown of the affected unit cannot preclude the capability for safeshutdown}fthereactoroftheunaffectedunit(s).
g. Containment structural integrity is maintained by limiting the combination of break sizes and types to the design basis capability (i.e., temperature, pressure, and leakage rate) of the containment.
h. For any postulated pipe break the structural integrity of the containment structure shall be maintained. In addition, for those postulated breaks classified as a loss of reactor coolant the design leak tightness of the containment fission product barrier shall be mainta,ined.
1. The conditions within the control room or any other location wnere manual action is required to assure safe shutdown to the cold condition is such

( as to assure habitability and comply with the requirements of General Design Criterion 19.

j. A whipping pipe or jet is assumed not to cause failure of other pipes of equal or greater size and equal or greater thickness. Smaller and thinner pipes are assumed to encounter unacceptable damage upon impact. A whipping pipe or jet is considered capable of developing througn-wall leakage cracks in larger nominal pipe sizes with thinner wall thicknesses, except where experimental or analytical data for the expected range of impact energies demonstrate the capability to withstand the impact without failure,
k. Piping Breaks Within The LOCA Boundary
1) All LOCA breaks are allowed to damage any non-LOCA line except essential systems, and steam and feedwater lines.
2) Pipe breaks within the LOCA counoary are allowed to damage ECCS lines connecting to the ruptured line, providing the ECCS flow to other loops is maintained.
3) For breaks in 6" nominal or larger piping, propage. tion of the break in the affected loop is not permitted if the resultant break area is more than 120". of the originating break area. If the originating break is a Reactor Coolant System main loop break, propagation is permitted to occur but must not exceed the design basis for calculating L

3.6-7 12/83

containment and subcompartment pressure, loop hydraulic fc.rces, reactor internals, reaction loads, primary equipment support loads, or ECCS performance. Propagation to any other loop is not permitted in any Case.

4) Pipe breaks within the LOCA boundary that are equal to or less than
                                -4" nominal pipe size must meet the following criteria:

(a) Break propagation to the other leg of the affected loop and to other loops must be prevented. The RTD bypass lines are exceptions to this particular criteria. A break in one of these lines is allowed to propagate to either of the other RTD line in the affected loop. This is permitted since.

1) A break in a RTD line effectively results in propagation to the other legs by flow througn tne unoroxen RTD lines,
2) An analysis of dual aperture breaks has been perforced ana the results indicate that the limiting small break case bound the possible multiplg break cases for the RTD lines. (ieference 2)

(b) Propagation of the break in the affected !cg is oermitted but is limited to a total break area of 12.5 square incnes (4-inch inside diameter). The exception to this case is when tne initiating small break is the high head safety injection line. Further propagation f is not permitted for this case. ( (c) Damage to the high head safety injection lines connected to the other leg of the affected loop or to the other loops is prevented. (d) Propagation of the break to high head safety injection line connected to the affected leg is prevented if the line break results in a loss of core cooling capability oue to a spilling injection line.

1. Piping Breaks Outside the LOCA Boundary ('lon-LCCA)
1) A pipe break which is not a loss-of-reactor-coolcnt accicent cannot cause a loss-of-reactor-ccolant accident er steam or feedwater line break.
2) All non-LOCA breaks (exceot steam and feedsater line breaks) are allowed to damage the non-LOCA portion of a single train of an ESF ~

system, provided that unit shutdown can te achiveec.

3) All non-LOCA breaks (excluding steam ano (cacwate,.ljan breaks) are allowed to damage any non-LOCA, non essential ilnes (except steam and

{ feedwater lines). 1 ( 3.6-8 12/83

 . ~ - - _ . ~ _ - _ , ,            , - . - . - , - - < -   -w.,m,e,,y,- v -yw., -m   . . , .,% .. ww-.,wm,,,,.,.--,, . ,  ,,#-. %,.,   -.,%--_, . - - , . . ,
4) .

A pipe break in one train of a redundant essential system or a pipe break which damages one train of a redundant essential system cannot result in system. essential damage to the opposite train of that system or any other 5) A pipe break in a non seismic system (Ouke System Piping Class D,E,G,H) cannot result in damage to an essential system. m. Piping Breaks in Steam and Feedwater Lines 1) Steam and feedwater line breaks are allowed to damage steam and feedwater lines, respectively, of the same steam generator, provided that the aggregate break size does not exceed the applicable maximum break size considered in the safety analysis. 2) Steam requiredand feedwater essential systemline breaks can damage any non-LOCA lines excep lines. n. Failure of any structure caused by the postulated line break is not allowed to adversely affect the mitigation of the consequences of the break nor the capability to safely shut down and maintain the reactor in a safe shutdown condition. o. Loss of required redundancy in the protective system, engineered safety feature equipment, cable penetrations or their interconnecting cables due to postulated line breaks is not allowed to adversely affect the mitigation of the consequences of the break nor the capability to safely shutdown and maintain the reactor in a safe shutdown condition. p. Loss of ability to cope with subsequent line ruptures due to an initial postulated line rupture is not allowed in electrical components. q. Internal flow account fluidrestrictors. energy level associated with the pipe break may take into r. Environmental operability is assured for all electrical equipment in the immediate piping break area by the equi;> ment specification requirements based on conservstive design conditions. dasleee ' s. Duke's Ste m-Production Department prepares adequate emergency o procedures high that would energy systems as be followed af ter a postulated piping break for required. 3.6.2.2.1 Postulated Piping Break Locations and Orientations Systems identified as containing high energy or moderate energy piping are examined by a detailed design drawing review for a p Code class. listed in Table 3.6.1-1 and 3.6.1-2. Systems analyzed for consequences of a. Breaks in Duke Class A piping are postulated at the following locations: (See Table 3.2.2-3 for class correlations). 3.6-9 1984 Update l

                                                               . M *', , y 4, e a MAb   ' = "* '

e-

1) The terminal ends of the pressurized portions of the run.
2) At intermedia +c.

methods: locations selected by either one of the following (a) At each location of potential high stress and fatigue such as pipe fittings (elbows, tees, reducers, etc.), valves, flanges, and welded attachments, or (b) At all intermediate locations between terminal ends where the following stress and fatigue limits are exceeded, , (1) The maximum stress range should not exceed 2.4 Sm except as noted below.

                                          -(2) The maximum stress range between any two load sets (including) the zero load set) should be calculated by Eq. (10) in Paragraph NB-3653, ASME Code, Section III, for normal and upset plant                                    ,

conditions and an operating basis earthquake (OBE) event

                                                 -transient.

If the calculated maximum stress range of Eq. (10) exceeds

                                            ~

the limit (2.4 Sm) but is not greater than 3 Sm, the limit of U < 0.1 should be met. . If the calculated maximum stress range of Eq. (10) exceeds 3 Sm, the stress ranges calculated by both Eq. (12) and Eq. i (13) in Paragraph NB-3653 should not exceed 2.4 Sm and the limit of U < 0.1. where: i S n = Primary plus secondary stress-intensity range, as calculated from Equation (1) in Subarticle NB-3600 of l the ASME Boiler and Pressure Vessel Code, Section III. S, =. Allowable design stress-intensity value, as defined in Subarticle

                                                       . Vessel      Code,N8-3600    of the Section         III. ASME Boiler and Pressure U
                                                       = The cumulative usage factor, as calculated in accordance                                        ,

with Vessel Subarticle NB-3600 Code, Section III. of the ASME Boiler and Pressure

3) If there are no interr.5ediate locations exceeds 0.1, e
  • where 5 d A o- 'M '-
                                                                                                  =t'-- :h:9'excee '-    s 2.4  S or U==-

p ;aemeMe 4 _.  : cha r- h?rd

pea 6$g ==s 4 1ntermediate breaks are not postulated in sections B **
  • g h, of straight pipe where there are no pipe fittings, flanges, valves or welded attachments. Al Ee , ' .t . ..li i. x Z ;.

g i; , ' ,- y u _:.-  : s s i h ,m t^ mm

                                                                                                                         .._ - ' M the 1e,,uu, o r u,e
                                                                                                                                      , u 6.,

b. Breaks in Duke Class B, C, and 0 piping are postulated at the following locations: (see Table 3.2.2-3 for class correlations) ,

                                                                                                                                                     ,h 3.6-10                                          1984 update              .
                            ,   ,,b              .'r n a m w=- M & M
1) The terminal ends of the pressurized portions of the run.
2) Atmethods:

intermediate locations selected by either one of the following (a) At each location potential high stress or fatigue, such as pipe fittings (elbows, tees, reducers, etc.), valves, flanges and welded attachments, or (b) At all locations where the stress, S, exceeds 0.8 (1.2Sh

  • 3A )'

where:

            .           S = Stresses under the combination of loadings associated with the normal and upset plant condition loadings, as calculated from the sum of equations (9) and (10) in Subarticle NC-3600 of the ASME Boiler and Pressure Vessel Code, Section III.

Sh= Basic material allowable stress at maximum (hot) temperature from the allowable stress tables in Appendix 1 of the ASME Boiler and Pressure Vessel Code, Section III. SA= Allowable stress range for expansion stresses, as defined in Subarticle Vessel Code,NC-3600 Section ofIII.the AMSE Boiler and Pressure i.e.4; w s

3) If there are eo nor m b :

3 h + 3A ), a . . *. me m Ve %ec.k. intermediateJwhere

                                                                   - pas A,(.,W. S exceeds 0.8 (1.2
                                            . ;;u;;. u ; wc separatcu ;ccau en; sna;,. ac anuse: c=ca c          Isa;;c stress.

vi- u :: u t : u: n= If the piping cr brench rtr nn eniy one unemge pu:Lulei.cd. Of 0 = 4termediste= break :-h-1 M y The pattern of postulated intermediate break locations shall be determined separately for the normal plant condition load p.,j, aph combination and for that upset plant condition which was the highest stress. Intermediate breaks are not postulated in sections of straight pipe where there are no pipe fittings, valves, flanges or welded attachments. i!:u,

L:: =d!:Lu 2: e d ::: u pun uleteu in piping - u N hss-than-&4 (1. 3Sm +S n ,m,s f --_ .,B , %. . - %,g C[ Lhe fun., --- -

c. Breaks in Duke Class E, F, G, and H piping are postulated at the following locations: (see FSAR Table 3.2.2-3 for class correlations)

1) For Class E, F, G and H Piping:

At each location of potential high stress or fatigue, such as pipe welded fittings (elbows, attachments; or tees, reducers, etc.), valves, flanges, and

2) For Class F Piping: P a) At all locations where the stress, S, exceeds 0.8 (1.2 S h
  • SA ),

3.6-11 1984 Update . 1 - C.z S hbx; m s/M M M A 3 .e , .,i

where: S = Stresses under normal and upset plant loadings S h= Basic material allowable stress at maximum (hot) temperature, per ANSI B31.1.0. SA= Allowable B31.1.0. stress range for expansion stresses, per ANSI 6') If there are not -

                                                         ==-
- s=i exceeds 0.8 (1.2 S h + 5A ), ^* i"b'ntermediate iocations wngry s___

i de;cr b ed uper "e9

                                                                            =-  M' Ry-=**4*'^'[".      -       -

r me

ynen uren. If dm p:pi: 3 ::: 1:

of cirection, a minimum of ena i M ermeciate creak is postuieted. ly 2 s-5= ge y,s po ry"B 4 Intermediate breaks are not postulated in sections of straight pipe ments. there where ^ite, are no pipe fittings, valves, flanges or welded attach-ie: necide bmok: em-uui-yudulatem!ccattens Here 5 is less t W O_^ (1_?5 n k). For Cases a. , b. , and $.2.*., above - longitudinal and circumferential breaks shall be postulated, but not concurrently, unless from a detailed stress ' analysis most (e.g.type. probable , finite element analysis) the state of stress can identify the If the primary plus secondary stress in the axial direc-tion is found to be at least 1.5 times that in the circumferential direction circumferential break need be postulated.for the most severe normal an Conversely, if the primary plus secondary stress in the circumferential direction is found to be at lease 1.5 times that in the axial direction for the most severe normal and upset transients, then only a longitudinal break need be postulated. At terminal ends lated. where piping has no longitudinal welds, no longitudinal breaks are postu-Also, at intermediate locations where the criterion for a minimum number postulated.of break locations must be satisfied, only circumferential breaks are Where break locations are postulated at fittings without the benefit of a detailed to-fittingsstress weld. calcul.ation, breaks should be assumed to occur at each pipe-

If a detailed stress analyses or tests are performed, the
      . maximum E er; t-e ;:    stressed
                            ;;c:          location in the fittings may be selected as the break location.

b: ems ere pestdeted, bree'-- 1ecati, = ere seled:d

  • d d.
         ] = [_( 2(( : ::y? _ \ _ 9_ ? ? ? u _

er :: : r._ ._ r . r

         = ; eleu;d ;: d: besH

_ == - g ~ ~ ~[= ;.

                                                                  -' rt..,'[IjJ N= D' I r     m em_

IUC3t ! um m ,u:= p:p: : u: ur ef t h= n -t--5evere

                                                                           ,,sy=nces.

as limited by structural design features.A circumferential break results in ' The break shall be assumed perpen-dicular to the longitudinal axis of the pipe, and the break area assumed to be the cross-sectional flow area of the pipe at the break location. The break discharge coefficient used shall be substantiated analytically or experi-mentally. assumed to In be the 1.0. absence of this data, the discharge coefficient shall be A longitudinal break results in an axial split without severance. The split " shall be assumed to be orientated at any point about the circumference of the l - i 3.6-12 1984 update t ' m.W ' " '

pipe, or alternatively at the point of highest stress as justified by detailed stress analyses. For the purpose of design, the longitudinal break shall be C assumed to be circular or ellipical (20 x 1/2D) in shape, with an area equal to the largest piping cross sectional flow area at the point of the break and have a discharge coefficient of 1.0. Any other values used for the area, diameter and discharge coefficient associated with a longitudinal break shall be verified by test data which defines the limiting break geometry. For the purpose of analysis, circumferential and longitudinal breaks are . assumed to reach full size within one (1) millisecond af ter break initiation ' ' ' ' unless otherwise analytically or experimentally substantiated. Through-wall cracks are postulated in moderate energy piping systems outside

                                   ~

containment having a nominal diameter greater than one (1) inch. Cracks are not postulated in piping that contains no pressurization equipment; i.e. , systems without pumps, pressurizing tanks, boilers, etc. , and which operate only from gravity, flow or storage tank head. Also, cracks are not postulated in portions of Duke Class B, C, D, or F piping where the stresses are less - than 0.4 (1.2 Sh

  • SA ). Through wall cracks in moderate-energy piping systems are not postulated inside containment.

Terminal ends are considered at piping origina6ing at structures or components (such as vessel and equipment nozzles and structural piping anchors) that act as rigid constraint to the piping thermal expansion. Typically, the anchors assumed for the piping code stress analysis would be teiminal ends. The branch connection to the main run is one of the terminal ends of a branch run, except where the branch run may be classified as part of a main run. Crack openings shall be assumed as a circular orifice of cross sectional flow area equal to that of a rectangular one-half diameter in length and one-half pipe wall thickness in width. The orifice shall be assumed to be orientated at any point about the circumference of the pipe. Pipe sizes and locations of postulated piping breaks in Duke Class A (ASME Class I) piping other than the reactor coolant loop are 4.:27e nt M '-to the

      ~m            -
                             -   ,a
      '                                                                        EL.3 a a. .a 4- the E4Ceds 2 "}eM,D " Vpesedel             . pe. Fsk;hrt   Wes "s-a.$.
                                                               ' P.s w t N% E-40 89 Table 3.6.1-3 identifies differences between Duke criteria and NRC require-                                    -
    . Regulatory ments contained Guidein   1.46 Branch
                                         -(fthyTechnical 1973). Position APCSB 3-1 (March 1975) and The analytical interface between Duke and Westinghouse for RCS pressure boun-dary is fully described in detail in puke's ASME Class I piping design specifi-cation. The interface occurs at the weld end of all RC System branch nozzles.

Analytical interfaces are defined to the extent that both Duke and. Westinghouse are able to perform independent analysis without compromising allowable stress #p limits at the branch line connection. Duke Design Specification is available - at the Construction site and in Duke's Design Engineering Department for inspection by the ASME inspectors and NRC as required. 3.6.2.2.2 Postulated Piping Break Sizes [ Double ended and equivalent are longitudinal pipe break areas are based on A the nominal inside diameter (ID) of the piping system, i . e. , d' 3.6-13 12/83 m:% mm.mm.wwww

i 4 A-{(ID)2 Through-wall crack pipe break areas are based on length equal to one-half the i nominal outside diameter (1/2 ID) and a width equal to one-half the minimum wall thickness (1/2 t) of the system piping materials, i.e., A= t i 3.6.2.2.3 Line Size Considerations for Postulated Piping Breaks

  • For high energy systems, piping larger than 1" nominal pipe size (NPS) is reviewed for the consequences of a double ended break.

For high energy systems, piping 4" NPS and larger is reviewed for the conse-quences of double ended and equivalent area longitudinal breaks. For moderate energy system, piping larger than 1" NPS is reviewed for the consequence of through-wall cracks. l 3.6.2.3 Analysis and Results > The results of analyses of failure in fluid systems occurring inside and I outside containment for McGuire are presented in the " Summary Evaluation of the Effects of Postulated Pipe Failures", Report No. NE-1019. i 3.6.3 DESIGN LOADING COM8INATIONS

 ,                                                                                                                                                                                                                         r 3.6.3.1                                Reactor Coolant System Design Loadino Combinations As described in Section 5.2., the forces associated with rupture of reactor i                      piping systems are considered in the design of supports and restraints in order to assure continued integrity of vital components and Engineering Safety Features.

l Reaction forces used in the design of supports and restraints are computed on

<                                                                                                                                                                                                                         l 4

the basis of an assumed break equal to the cross sectional flow area of the pipe. ' L 4 The design stress limits applicable to postulated reactor coolant piping breaks j and supports are discussed in WCAP-8172 and are listed in Table 5.2.1-3.  : l 3.6.3.2 All Other Mechanical Piping Systems Design Loading Combinations i

'                                                                                                                                                                                                                         l i

Since all locations of consequences are reviewed and as detailed stress analysis  ; information is extremely extensive, stress analysis information is only i reviewed for special identified problem areas which might require additional restraints. I [ These additional consequential piping breaks posing safety related problems to structures, systems or components in the immediate area are either restrained [ 3 to stress mitigate analysis. the consequences of the break or reviewed in detail against existing i

If the stress allowables discussed in 3.6.2.2.1 are not exceeded, then the break is not considered to occur. .

[ } 3.6-14 1984 Update i l 1

          +r,--,9 .t      ---9.- ,ve,--  z, iy--w-wq--w--t--rw-r-wy,-----,vww+--y--,y-       -,.y-4.-,,,,---t-.-,--,,y-F   N gr*'W -         'WP -' ='wevue m w 7 W' f W P - *    *M*=r*   9'vF'--*'WWWw"WWTW~**rW'

Loading and stress criteria for pipe whip restraints is fully described in

3. 9.

Postulated pipe breaks are considered a faulted condition with respect to the pipe whip restraint design and allowable restraint stresses. 3.6.4 DYNAMIC ANALYSIS 3.6.4.1 Reactor Coolant System Dynamic Analysis I This section summarizes the dynamic analysis as it applies to the LOCA resulting ' from the postulated design basis pipe breaks in the main reactor coolant piping system. Further discussion of the dynamic analysis methods used to verify the design adequacy given in WCAP-8172. of the reactor coolant loop piping, equipment and supports is i The particular arrangement of the Reactor Coolant System for the McGuire Nuclear Station is accurately modeled by the standard layout used in WCAP-8172 and the  ; postulated break locations do not change from those presented in WCAP-8172. l In addition, an analysis will be performed to demonstrate that at each design basis break location the motion of the pipe ends is limited so as to preclude , unacceptable damage due to the effects of pipe whip or large motion of any major components. The loads employed in the analysis will be based on full pipe areas discharge except where limited by major structures. The effects of l jet discharges will be analyzed to demonstrate that any structure, system or component required to safety shutdown the reactor or mitigate the consequences , of an accident will not be impaired. The dynamic analysis of the Reactor Coolant System employs displacement method, lumped parameter, stiffness matrix formulation and assumes that all components behave in a linear elastic manner. The analysis is performed on integrated analytical modals including the steam generator and reactor coolant pump, the associated supports and restraints, and the attached piping. An elastic-dynamic three-dimensional model of the Reactor Coolant System constructed. in general, the foundation concrete / support structure interface.The boundary of the a The antici-pated deformation of the reinforced concrete foundation supports is considered where applicable to the Reactor Coolant System model. The mathematical model . is shown in Figure 3.6.4-1. The steps in the analytical method are: a. The initial deflected position of the Reactor Coolant System model is defined by applying the general pressure analysis; b. Natural frequencies and normal modes of the broken loop are determined; l

c.  !

The initial deflection, natural frequencies, normal modes, and time-history forcing functions are used to determine the time-history dynamic deflec-tion response of the lumped mass representation of the Ractor Coolant System; l d. The forces imposed upon the supports by the loop are obtained by multi-plying the support stiffness matrix and the time-history of displacement vector at the support point; and 3.6-15 12/83 1

e. The time-history dynamic deflection at mass point are treated as an imposed daflection condition on the ruptured loop Reactor Coolant System model and internal forces, deflections, and stresses at each end of the members of the reactor coolant piping system are computed.

The results are used to verify the adequacy of the restraints. The general dynamic solution process is shown in Table 3.6.4-1. In order to determine the thrust and reactive force loads to be applied to the Reactor Coolant System during the postulated LOCA, it is necessary to have a detailed of the hydraulic transient. Hydraulic forcing functions are calculated for the ruptured and intact reactor coolant loops as a result of a postulated loss of coolant accident (LOCA). These forces result from the transient flow and pressure histories in the Reactor Coolant System. The calculation is performed in two steps. The first step is to calculate the transient pressure, mass flow rates, and other hydraulic properties as a , function of time. The second step uses the results obtained from the hydraulic analysis, along with input of areas and direction coordinates and is to calculate the time history of forces at appropriate locations in the reactor coolant loops. The hydraulic model represents the behavior of the coolant fluid within the entire reactor coolant system. Key parameters calculated by the hydraulic model are pressure, mass flow rate, and density. These are supplied to the thrust calculation, together with appropriate station layout information to determine the concentrated time-dependent loads exerted by the fluid on the loops. In evaluating the hydraulic forcing functions during a postulated LOCA, the pressure and momentum flux terms are dominant. The inertia and gravitational terms are taken into account only in the evaluation of the local fluid conditions in the hydraulic model. The blowdown hydraulic analysis is required to provide the basic information concerning the dynamic behavior of the reactor core environment for the loop forces, reactor kinetics and core cooling analysis. This requires the ability to predict the flow, quality, and pressure of the fluid throughout the reactor system. The SATAN-V code was developed with a capability to provide this information. The SATAN-V computer code performs a comprehensive space-time dependent analysis of a loss of coolant accident and is designed to treat all phases of the blow-down. The stages are: (i) a subcooled stage where the rapidly changing pres-sure gradients in the subcooled fluid exert an influence upon the Reactor Coolant System internals and support structures; and (ii) a two phase depres-surized stage. The code employes a one-dimensional analysis in which the entire Reactor Coolant System is divided into control volumes. The fluid properties are considered uniform and thermodynamic equilibrium is assumed to each element. Pump characteristics, pump coastdown and cavitation, core and steam generator heat transfer including the W-3 DNB correlation in addition to the reactor kinetics are incorporated in the code. The blowdown hydraulic loads on primary loop components are computed from fluid ' transient function: information calculated using the following time dependent forcing ! 3.6-16 12/83 1 l

F = 144A [(P - 14.7) + ( m2 )] pgAgl44 which includes both the static and dynamic effects. The symbols and units are: F = Force, Lb f A = Aperture area, Ft2 P = System Pressure, PSIA th = Mass flow rate, Lb,/Sec p = Density, Lb,/Ft3 g = Gravitational Constant = 32.174 lb,x Ft lb 7x Sec2 A,= Mass Flow Area, Ft2 The main Reactor Coolant System is represented by a similar nodal system as employed in the blowdown analysis. The entire loop layout is described in a gicbal coordinate system. Each node is fully described by: (i) blowdown hydraulic information and (ii) the orientation of the streamlines of the force nodes in the system, which includes flow area, and projection coefficients along the three axes of the global coordinate system. Eachnode is modeled as a separate control volume, with one or two flow apertures associated with it. Two apertures are used to simulate a change in flow direction and area. Each forte is divided into its x, y, and z components using the projection coeffi-cients. The force components are then summed over the total number of aper-tures force. in any one node to give a total x force, total y force, and total z analysis.These thrust forces serve as input to the piping / restraint dynamic Further details are given in WCAP-8172. The dynamic analysis described above for the reactor coolant loop piping has been completed and the results verify that the break locations and type postu-lated in WCAP-8082/8172 are the only ones that are required to be postulated for McGuire. 3.6.4.2 All Other Mechanical Piping Systems Dynamic Analysis Effects of pipe break are conservatively evaluated to determine the need for pipe whip restraints. impingement force Energy of the whipping pipe, its effect on targets, jet and temperatures, and compartment pressurization and tem-perature effects establish the need and requirement for pipe whip restraints. Dynamic analysis of Category 1 piping and supports is or is not performed depending on the conservatively determined consequences of the break. The need for dynamic analysis depends on the need for fully identifying the response of the system. The purpose of the analysis when required is to prove that the consequences of the break does not prevent mitigation of the break nor present the safe and continued shutdown of the reactor. 3.6-17 1984 Update

Dynamic analysis methods have been developed. These methods consider the energy of the whipping pipe using conservative forcing functions, gaps between pipe and restraint, and energy absorbers designed to absorb the major portion of the whipping pipe energy. The design of energy absorbers is based on test results under dynamic loading conditions. The response of the system with respect to its effect on Category 1 systems and equipment has been determined by analysis using a computer program such as PWHIP or equivalent. PWHIP is described in 3.9.2.3 of this SAR. The dynamic analysis model used was one or more of three acceptable models specified by the NRC. Any one of these models was used depending upon the particular piping system being analyzed. A lumped parameter model has been formulated and programmed, and is available for use should this option be elected. This model consists of lumped masses interconnected by bending stiffness spriegc. Modulus of elasticity for the bending stiffness springs 4 is represented by a bilinear stress strain curve. A suddenly applied load of constant value is currently programmed into the model with the constant value determiend as outlined by the NRC (i.e., F = KpA). A time-history numerical integration is performed using the Runge-Kutta-Gill technique. Newton's Second Law of Motion is applied to each of the lumped masses using the shear forces to accelerate the masses. From the accelerations, velocities are determined, and in turn displacements, elastic axis slopes, bending moments, and new shear forces are also determined. Extension of the model to include interaction with pipe whip restraints was accomplished once characteristics of the restraints were finalized. Restrain loadings were then determined. Associated jet impingement forces on an object are treated as a suddenly applied load constant value and not a varying function of time. In piping systems other than the Reactor Primary Coolant System the blowdown forces may be calculated by the following equation: T = MVE+PAEE Any other method used for determining blowdown forces or thrust coefficients was based on justifiable analytical and/or experimental data such as the work of Henry and Fauske and Moody. The specialabove equation conditions as is applicable to all fluid flow but can be simplified for follows:

a. Subcooled water:

Temperature <212 F 2PRAE T = lesser of MV W2y E* 1+K "#E* gA

b. Subcooled water:

1 3.6-18 12/83

Temperature 5212 F 2P A T = lesser of MV E* 1+K # NVE* gA

c. Water - Steam mixture, low quality:
                    .T = <     R AE 1+K d.

Water - Steam mixture, high quality or superheated steam: T = MVE+PAEE-The flow is assumed to choked at the break area based on isentropic expansion from reservoir maximum operating condition for K = 0. Where K # 0, Fanno , Lines are used to determine flow conditions at exit or break location. Fluid properties are based Homogeneous Equilibrium Model. The sonic velocity, mass flow rate, and thrust is calculated using SONVEL, which is a Fortran IV program written to solve, by iteration, the following equations. These equations are based on sonic flow through a convergent isentropic nozzle. Sonic velocity is calculated as follows: VR = 0.(Assumed) VE = 223.7 (hR -h)E Also, VE = 12VE E( )E9 SR*bE where subscript R denotes reservoir conditions and subscript E denotes exit (or break) location. . A = Break area (ini) g = Gravitational constant (ft/sec2) h = Specific enthalpy (BTU /#) K = Flow resistance coefficient based on flow velocity at exit ' M=E 9 P = Static pressure (PSIG) T = Thrust (Lb) v = Specific volume (ft3/#) 3.6-19 12/83

                                                                                     ? _,       na g
                                                            "'**NVO*'*
                                                                ',     N   '

V = Flow velocity (ft/sec) W = Mass flow rate (#/sec) ( ) = Rate of change of pressure with specific volume at point "E" at E constant entropy The TMD Code is utilized in developing pressure transients for postulated piping breaks within containment and the main steam /feedwater penetration rooms (doghouses). Assumption and considerations utilized in the analysis as applicable are:

a. The total volume being analyzed is subdivided into smaller compartments as required, and the time-dependent pressure rise for each individual subcompartment is assumed to be equal throughout;
b. Frictional effects, turning losses, vent losses, etc., are considered for flow through each subcompartment;
c. Condensation effects due to heat sinks are considered negligible for conservatism;
d. Calculations for mass flows from pipe ruptures do not consider frictional effects of piping; and e.

Two phase mass flow (liquid and vapor phase) is assumed to be homogeneous. The assumptions listed below are applied to the pressurization calculations for Auxiliary Building pipe ruptures. a. A homogeneous mixture of air and steam or gas in each compartment, and thermodynamic equilibrium, are attained instantaneously.

b. Homogeneous or separated 2 phase flow models are used. A break discharge coefficient of 1.0 is used for all break sizes in blowdown analyses in the source compartment.

The orifice discharge coefficient between compartments is assumed to be 0.6 unless other values can be justified, and is used for the determination of pressure differentials in the source compartment. c. Potential energy and kinetic energy are negligible, and flow work is recovered and stored as integral energy.

d. Passive and active heat sinks are considered when justified.

e. Initial state of the contents of both the compartment and the pipe are known. Final state is saturated or super-heated vapor with liquid phase, if existing, at saturated or subcooled conditions. A division of the building volume into compartments allows computation by computer program of the pressure buildup in each compartment due to the effects of postulated pipe break. The pressure at any point in time is cal-culated by obtaining the simultaneous solution of the mass balance, energy balance, and equations of state for each volume considered. This volume is 3.6-20 12/83

either a total compartment volume or an arbitrary control volume assumed for computational purposes. The results of the pipe rupture analysis for Category 1 piping systems other than the Reactor Coolant Loop are presented in Appendix 3P and Report No. MDS/PDG-77-1 for inside containment and outside containment, respectively. 3.6.4.3 Structural Analysis of Postulated Piping Breaks Evaluation utilized to demonstrate the adequacy of or in the design of Category 1 structures subject to loadings of postulated piping breaks include:

a. Method of evaluating stresses;
b. Allowable design stresses and/or strains;
c. Load factors and combinations;
d. Design loads including pressure and temperature transients; and
e. Load reversal effects.

Details of the structural analysis involving the above combinations are discussed in 3.8.1. 3.6.5 PROTECTIVE MEASURES 3.6.5.1 Reactor Coolant System The fluid discharged from broken Reactor Coolant System piping will produce reaction and thrust forces in branch line piping. The effects of these loadings are considered in assuring the continued integrity of the vital components and the engineered safety features. The accomplish this in the design, a combination of component restraints, barriers, and layout are utilized to ensure that for a loss of coolant or steam-feedwater line break, propagation of damage from the original event is limited, and the components as needed, are protected and available. 3.6.5.1.1 Postulated Pipe Break Restraint Design Criteria for Reactor Coolant System Large Reactor Coolant System piping and all connecting piping out to the LOCA boundary valve (Figure 3.6.2-1) is restrained to meet the following criteria. .

a. Propagation of the break to the unaffected loops is prevented to assure the delivery capacity of the accumulators and low head pumps;
b. Propagation of the break in the af fected loop is permitted to occur but must not exceed 20 percent of the area of the line which initially failed. This criterion is voluntarily applied so as not to substantially increase the severity of the loss of coolant; and 3.6-21 12/83
c. Where restraints on the lines are necessary in order to prevent impact on and subsequent damage to the neighboring equipment or piping, restraint type and spacing is chosen such that a plastic hinge on the pipe at the two support points closest to the break is not formed.

3.6.5.1.2 Protective Provisions for Vital Equipment In addition to pipe restraints, barriers and layout are used to provide protection from pipe whip, blowdown jet and reactive forces for postulated Reactor Coolant System piping breaks. Some of the barriers utilized for protection against pipe whip are the following. The polar crane serves as a barrier between the reactor coolant loops and the Containment liner. In addition, the refueling cavity walls, various structural beams, the operating floor, and the crane wall enclose each reactor coolant loop into a separate compartment; thereby preventing an accident, which may occur in any loop, from affecting another loop or the Containment. The portion of the steam and feedwater lines within the Containment have been routed behind barriers which separate these lines from all reactor coolant piping. The barriers described above will withstand loadings caused by jet forces, and pipe whip impact forces. Other than Emergency Core Cooling System lines, which must circulate coolirg water to the vessel, Engineered Safety Features are located outside the crane wall. The Emergency Core Cooling System lines which penetrate the crane wall are routed around and outside the crane wall to penetrate the crane wall in the vicinity of the loop to which they are attached. In reviewing the mechanical aspects of these lines, it has been demonstrated by Westinghouse Nuclear Energy System tests that lines hitting equal or larger size lines of same schedule do not cause failure of the line being hit, e.g., a one-inch line, should it fail, does not cause subsequent failure of a one-inch or larger size line. The reverse, however, is assumed to be probable, discharged through the line, could break smaller size lines such as neighboring three-inch or two-inch lines. In this case, the total break area shall be less than 12.5 square inches. Alternately, the layout is planned such that whipping of the two free sections cannot reach equipment or other pipes for which protection is required; plastic hinge formation can be allowed to form. As another alternative, barriers can be erectedprotection. requiring to prevent the whipping pipe from impacting on equipment or piping Finally, tests and/or analyses may be performed to demonstrate that the whipping pipe doe not cause damage in excess of the acceptable limits. Whipping in bending of a broken stainless teel pipe section such as used in the Reactor Coolant System does not cause this section to become a missile. This design basis has been demonstrated by performing bending tests on large and small diameter, heavy and thin walled stainless steel pipes. 3.6.5.1.3 Criteria for Separation of Redundant Features There are no redundant features associated with reactor coolant piping system. Redundant features of other mechanical piping systems are discussed in 3.6.5.2. 3.6-22 1984 update

3.6.5.1.4 Separation of Piping The Reactor Coolant System is separated from other piping systems and components by barriers, as discussed in 3.6.5.1.2. 3.6.5.1.5 Pipe Restraints and Locations For the Reactor Coolant System, a pipe restraint is located at each of the 90" elbows of the cross-over leg. 3.6.5.2 All Other Mechanical Piping Systems Measures to protect against pipe whip, jet impingement and resulting reactive forces to meet established criteria outlined in 3.6.2.2 are as follows:

a. Separation and remote location of fluid system piping from essential structures and equipment.
b. Structural enclosure of the fluid system piping with access provided for inservice inspection; or, alternatively, enclosure of the essential equipment.
c. Provision of system-redundant design features separated, or otherwise protected, from the ef fects of the postulated pipa. rupture; or additional protection features such as restraints and barriers.
d. Design of essential structures and equipment to withstand the effects of the postulated pipe rupture.

e. Addition of guard piping for the main purpose of diverting or restricting blowdown flow.

f. In areas where none of the above can be met, or where unacceptable, more severe problems may be creased, augmented inservice inspection may be used on a case by case basis to reduce the probability of failure to acceptable levels, and not postulate the failure. The augmented inservice i

inspection is in accordance with the guidelines presented in NRC MEB Branch Position No. 4 " Augmented Inservice Inspection and Secondary Pro-tective Measures." Table 3.6.5-1 identifies all cases where exceptions to the criteria of Section 3.6 have been taken. See Table 3.6.1-1 and 3.6.1-2 for protection methods on a system basis. Curbs are provided around passageways to the Auxiliary Building from the Turbine Building. These curbs are of adequate height to contain flood water caused by the break of the main consenser circulating water expansion joint, or the most severe Condensate System failure for a minimum of fifteen minutes. There are no pipe or cable chase entrances below the elevation of the top of the curus. This flooding condition does not render any essential system or component inoperable. 3.6-23 12/83

3.6.5.3 Main Steam and Feedwater System Desinn Design of the Main Steam and Feedwater System meets the general design criteria established in 3.6.2.2; however, additional specific information as follows applies to these systems. a. Main Steam Lines are 100 percent cold pulled so that as lines heat up, all thermal expansion stresses are essentially eliminated throughout the system; b. Overpressure capability of the piping based on actual wall thicknesses is as follows: Actual Normal Operating Code Pressur'e Pressure Caoability Margin Main Steam: 985 psig 1250 psig 19% Feedwater: 1165 psig 1420 psig 22% c. Safety-related portions of the Main Steam and Feedwater Systems are Duke Class B. Class B system materials, fabrication, nondestructive examinations and documentation are in accordance with ASME III, Class 2; d. Proper piping system erection and function of safety-related supports and restraints are assured by several means: 1. The Construction Department Reviews erection against design drawings, and CA 2. AQ surveillance is conducted by the Hanger-Contractor to verify correct location, direction of movement and proper hardware installation.

3. Compliance with requirements of IE Bulletin 79-14.

e. Figures 3.6.5-1, 3.6.5-2, 3.6.5-4, 3.6.5-5, 3.6.5-6, and 3.6.5-7 show de-sign routing of the Main Steam and Feedwater Systems outside Containment to the Turbine Building. Piping for these two systems is isolated from other safety-related systems, equipment and the Control Room by a missile barrier as can be noted from the above listed figures. 3.6.5.4 Control Room Protection from Postulated Piping Breaks The Control Room is located on the top floor of the Auxiliary building and is bounded on the north and south sides by Electrical Penetration Rooms which contain no piping. The east side of the Control Room is bounded by the equip-ment area housing the Control Room ventilation equipment. Piping in this area consists of low pressure, low volume chilled water and hw pressure, Icw volume heating steam. room and supporting Oneareas. the west side, the Control Room is bounded by the computer Piping in this area consists of sanitary waste and vent piping, drinking water and instrument air, none of which are high energy systems. i r 3.6-24 1984 Update i

Immediately below the Control Room is the cable room containing no piping. The Control Figures Room 3.6.5-1 andceiling is bounded by a missile barrier roof as denoted on 3.6.5-2. Penetrations into the Control Room area consists of ducts, electrical cables and instrument air only. Openings around such penetrations are sealed. Doors entering the Coitrol Room area will have pressure seals. A slight positive pressure is maintained in the Control Room by a pressurizing fan such that any leakage is out-leakage. Momentary loss of pressure is experienced during ingress and egress but air flow is outward, o.e. , air flow is from the Control Room to adjacent areas. Based on the above physical parameters, the Control Room is structurally iso- ' lated from areas containing high energy systems; therefore, there are no related consequences to the Control Room from the postulated ureak of high energy piping systems. 3.6.5.5 Postulated Pipe Break Restraint Desinn Criteria for All Other Mechanical Pipinn Systems Postulated pipe break restraints are considered to consist of four basic components. } These are: " Process Pipe," " Energy Absorbing Device," " Structure ^ Extension" and the " Anchorage" as further explained below. Related to the pipe break restraint is the " Structure" to which it is attached, which is also further discussed below. a. The process pipe is the pipe which is to be restrained and includes all integral pipe wall,attachments which are welded, cast, or forged directly to the b. The energy absorbing device is a structural, mechanical, hydraulic cushion or other energy absorbing device or material which is designed to minimize the forces imposed on the structure. In some cases, the process pipe itself may be the energy absorbing device if it can be quantitatively demonstrated that the local deformations of the pipe account for that portion of the induced energy required to maintain forces on the structure and structure extension below their design limits. In some ::ases, no energy absorbing device is employed when the structure and structure extension is designed the pipe break to withstand the entire resisting force impcsed by phenomenon. c.

'                                                                                     The structure extension is the structural assemblage which connects the anchorage to the energy absorbing device or process pipe. In general, it may be considered as an extension of the anchorage. It is designated as a separate component because it can be an extensive structure and may be designed than used using for thedifferent rules, appif cable to the type of material used, anchorage.
In rare cases, the process pipe or energy absorbing device may be directly connected to the anchorage in which case 4 there is not structure extension.

l d. I The anchorage is that component which connects the structure extension to the structure. Generally for a concrete structure, it is an embedded plate. bolting, For a steel structure, it generally consists of welding or i 4 3.6-25 12/83 i_

e. The structure is that feature of the building which is a necessary part of the building but also is designed to accommodate the loads transmitted through the anchorage caused by the postulated pipe break. It may be either a steel or concrete component and is characterized by being I relatively stiff and massive when compared to the pipe break restraint. ! f. ! Allowable stresses used in the design of the pipe break restraint compo- i t nents are consistent with the component function. In general, the i allowable stresses associated with the total reaction force, including impact, on the structure extension, anchorage and structure is taken as the minimum yield stress for structural steel and concrete embedments.  ; For those situations where structure load limiting features cannot be provided to maintain the allowable stresses to within yield, plastic deformation in structural components is tolerated so long as the struc-  ; ture is capable of continuing its functional requirement af ter the i deformation occurs. The upper design limit for pipe break restraint , i material ultimate strain, j 3.6.5.5.1 Typical Pipe Whip Restraints i i 5 A description of the typical pipe whip restraints and a summary of number and i location in W of =dall pipe ruptures requiring restraints in each system is presented 9md: l f* &res{ Ay wf s4. n tort..9.1k, %..g sc

  • f A fug g Af/.// 4 3.

6.6 REFERENCES

1. { Pipe Breaks for the LOCA Analysis of the Westinghouse Primary Coolant,

loop, 19/3.

July, WCAP-8082, June,1973 (Westinghouse NES Proprietary), and WCAP-8173,

  • 1
      .           t                                                                                                           1B ak ehtueSm                                                                                          5 Analysis," Letter from W. O. Parker, Jr.

f l I - l i l ? i 3.6-26 12/83 4

                 . , .a g 3 ,, t                                                                                                                   * ' ^ ,   .MD       '
                                                                                                                                                                            ~MN

Table 3.6.1-1 High Energy Mechanical Piping Systems Analyzed for Consecuences of Postulated Piping Breaks System or Portion Thereof Operating System Pipe Break During Normal Reactor Operation Identification Protection Method High Energy Safety Related Systems Steam Generator Blowdown Recycle System BB (a) (b) Auxiliary Feedwater System (Motor Driven Pump Portion) CA (a) (b) Reactor Coolant System NC (a) (b) Residual Heat Removal System ND (a) (b) Safety Injection System NI (a) (b) Nuclear Sampling System NM (a) (b) Baron Thermal Regeneration System NR (a) (b) Chemical and Volume Control System NV (Letdown Portion and Sealwater Injection (a) (b) Other High Energy Systems Feedwater System CF (a) (b) (c) Main Steam Supply to Auxiliary Equipment SA (a) (b) Main Steam System SM (a) (b) (c) Main Steam Vent to Atmosphere System SV (a) Pipe Whip Protection Methcds legend: 4 (a) Physical Separation , (b) Piping Restrair.ts (c) Enclosures, pipe break). structural, guard pipes, etc., (designed specifically for Note: High Energy Systems may contain moderate energy portions; however, for brevity, high energy systems are only listed in this table. ! 12/83 t

Table 3.6.1-2 (Sheet 1 of 2) Moderate Enercy Mechanical Pipina Systems ' Analyzed for Consecuences of Postulated Pioino Breaks System or Portion Thereof Operating System Pipe Break Durino Normal Reactor Ooeration Identification Protection Method Moderate Energy Safety Related Systems Auxiliary Feedwater System CA (Turbine Driven Portion) (a) Diesel Fuel Oil System FD (a) Refueling Water system FW (a) Component Cocling System KC (a) Diesel Generator Cooling Water System KO (a) Spent Fuel Cooling System KF (a) Diesel Generator Lube Oil System LO (a) Boron Recycle System r NB (a) ( Residual Heat Removal System ND (a) Containment Spray System NS (a) Nuclear Service Water System RN ' (a) Containment Ventilation Cooling Water RV System (a) Main Steam Supply to Aux. Equipment SA (a) FWP Turbine Exhaust TE (a) Diesel Generator Starting Air System VG (a) Gaseous Waste Recycle System WG (a) Liquid Waste Recycle System WL (a) Solid Waste Olsposal System WS (a) Filtered Water System YF (a) 12/83

Table 3.6.1-2 (Sheet 2 of 2) Moderate Eneroy Mechanical Pipino Svstens Analyzed for Consecuences of Postulatea Pipino Breaks System System Pipe Break Identification Protecticn Method Other Moderate Energy Systems Auxiliary Steam System AS (a) Recirculated Cooling Water System KR (a) Ice Condenser Refrigeration System NF (a) Fire Protection System RF (a) Equipment Decontamination System WE (a) Liquid Waste Monitor & Disposal System WM' (a) Chemical Addition System YA (a) Plant Heating System YH (a) Mike up Demineralizer System YM (a) Pipe Whip Protection Methods Legends: (a) Physical Separation ' (b) Piping Restraints (c) Enclosures, pipe break) structural, guard pipes, etc. , (Designed specifically for L Revision 21 l

7 ___ v, JC;9: ?:~f y M Jn:S

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  • C , ;I E' ) -
                                                                                                                                          . 'f       . . ,
                                                                                                                                                                             .           'l Table 3.6.1-3 I                                                                                                                                              *         ~

i l' P: - l Comparison of Duke Pipe Rupture Criteria And " 2

     ~

IIRC Requirements of Branch Technical Position t APCSB 3-1 (March 1973) and NRC Regulatory Guide 1.46 (May 1973) NRC Criteria - 4 Duke Criteria - > APCSB 3-1, Section.B.2:c SAR Section 3.6.2.2  ? i Section B.2.c. requires that piping between containment o isolation valves be provided with pipe whip restraints Duke criteria is roughly equivalent to NRC criteria capable of resisting bending and torsional moments pro- as clarified below: duced by a postulated failure either' upstream or down-stream of the valves. Also, the restraints should be The containment structural integrity is provided for-designed to withstand the loadings from postulated all postulated pipe ruptures. In addition, for any failures so that neither isolation valve operability postulated rupture classified as a loss of coolant nor the leaktight integrity of the containment will be accident, the design leaktightness of the containment impaired. - fission product barrier will be maintained. 1 Terminal ends should be considered to originate at a Penetration design is discussed in SAR Section 3.9.2.8. point adjacent to the required pipe whip restraints. This section also discussed penetration guard, pipe design criteria. j Terminal ends are defined as piping originating at

                                 ,                                                                                       structure or component that act as rigid constraint to the piping thermal expansion.                               .
!         APCSB 3-1, Section B.2.d

! SAR Section 5.2.8 (1) The protective measures, structures, and guard .

pipes should not prevent the access required to Duke criteria is different than the NRC criteria due l conduct inservice inspection examination. to the code effective date as described below:

(2) For portions of piping between containment isola- ASME Class 2 piping welds will.be inspected in accor-tion valves, the extent of inservice examinations dance with Tables ISC-251 of Section XI (1971), i completed during each, inspection interval should through Winter 1971 Addenda, of the ASME Code, as '

, provide 100 percent volumetric examination of accessibility permits. Inservice inspection program j circumferential and longitudinal pipe welds. requirements are given in SAR Section 5.2.8.

i . 3 a n. m . , 12/83 [' .. i ' t

p ,- h, f ( s .,:- 7, Table 3.6.1-5 Jont'd). \:/ 0, ,

                                                                                                                                                     . ".. "'i .':' 7'[ '                              O-
                                                                                                                                                            <        -                           ( 2 c. a)         M, (3) Inspection ports should be provided in guard pipes to '!                                                      *
                                                                                                                                                                    .s
  • 9 g

permit the required examination of circumferential welds. m ' ' U' Inspection ports should not be located in that portion ' ' i of guard pipe passing .through the annulus. ' i . (4) The areas subject to examination should be defined ' in accordance with Examination Categories C-F and , i C-G for Class 2 piping welds in Tables IWC-2520. - I' T. I. i-APCSB 3-1, Appendix A

                                                                                                               'SAR Section 3.6.1.2                           o wI l
      -          High Energy fluid systems are defined as those systems that, during normal plant condition <, ar,e either in "                                        Duke criteria is the same as NRC criteria with operation or maintained pressurized under conditions                                           expansion of definition as clarified below:                                                  $-

i where either or both of the following are met: a. Non-liquid systems with a maximum normal

a. t pressure lessithan 275 psig are not con-maximum operating temperature exceeds 200 F, or sidered high energy regardless of the temp-1 b.

maximum operating pressure exceeds 275 psig. t erature. Suchlowpressuresystem(i.e., Auxiliary Steam, 50 psig, 298 F) do not contain sufficient sensible energy to . develop _ sudden, catastrophic failures. ! ~ Propagation of a crack to a full failure is extremely unlikely. 1 b. Exception to the 200 F threshold for high

 )                                                                                                                       energy systems is taken for non water systems such as ethylene glycol. Such systems e         that operate at less than their boiling temperature are considered moderate energy.

APCSB 3-1, Appendix A 1

                                                          .                                                  SAR Section 3.6.2.2.1

) In piping runs which are maintained pressurized I i-1 during normal plant conditions for only a portion Duke criteria is different from NRC cri,teria as of the run (i.e., up to ti]e first normally shut described and justified below: *

valve) a terminal end of such runs is the piping .

! connection to this closed valve. Terminal ends are considered at piping originating at structure or components that act as ri

                                          .                                                               straint to the piping thermal expansion. gid                                             con-Typically,
, - the anchors assumed for the code stress analysis i would be terminal ends. Stresses in the system
l. - , either side of the closed valve will be about the
                                   .               4,                                                     same; therefore, terminal end classification b'ased
!'                          ,                       :..                                                   on constraint and high stresses are not applicable.
                                                                                                     ..                                  1 12/83 1    .
i .

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                                                                                                                                                                                                      .t 4;

_ . . - Table 3.6.h3 (Cont'd)  ;  ; APCSB 3-1 Appendix Il and C . . (3 of 6) t SAR Section 3.6.2.42.1' '

  • In Appendix B, pipe break locations are spechfied
              '-l
   ,.           .             for ASi1E Section 111 Code Class,1, 2, aruf 3 piping                        Duke criteria specifies that if the threshold 3

such that a minimum of two intermediate breaks are stress icvels are not exceeded, then no inter-

        .s                l  selected per run although threshold limits are not                          mediate breaks are postulated.

' ) 3 piping). (forInASl1E exceeded Section III Code Class 1. 2, and Appendix C, a minimum of either two . - ' a.. or one intermediate breaks within the boundary of each compartment is specified. , HEB 3-1, Section B.I.b(6) .. i , j . e SAR Section 3.9.2 8 Section B.I.b(6) requires that guard pipe assemblies bet'.veen containment isolation valves ' meet the following Dukedescribedcriteria is different justifiedfrom NRC criteria as ) requirements: and below: a. The design pressure and temperature should not be Guard pipes provided between containment isolation Icss than the maximum operating temperature and pressure of the enclosed pipe under normal plant valves is desi0 ned in accordance with SAR 5ection 3.9.2.8. cunditions. . Guardpipe thicknesses were developed using the criteria of ASME Code Case 1606 and the

b. appropriate loading combination and stress limits The desiC n stress limits of Paragraph NE-3131 of Table 3.9.2-2. Guard pip 9s are subjected to a shoulci not be exceeded under the loading associated (c) pressure test as required by the material specifica-with design pressure and temperature in combination tion before welding? to the penetration assembly.

with the safe shutdown earthquakes.

c. It is impractical.to test gulird pipes in the finished Guard pipe assemblies should be subjected to a penetration assembly due to the configuration and single pressure test at a pressure equal to design pressure. potential damage to internal process' pipe and associated insulation. Independent design analysis t have been conducted to provide assurance that Duke penetration designs are acceptable. .'In addition, -

the extent of NOT conducted on guard pipes to flued head butt weld is such to assure integrity of design.

                             !!ED 3-1, Section B.I.c(1)                                                                                                       ,
                                                                                                        'SAR Section Intermediate breaks in Class I piping are postulated
3. 6. 2.'.3. 2.1....

at the two highest stress locations based on Duke criteria states that if there are no Equation (10) if two intermediate locations

  • intermediate locations where S exceeds 2.4 S ,

cannot be determined by application of Equations or il exceeds 0.1, no intermediate breaks are (10), (12), and (13) or U>0.1. postulated. i

      ~                                                                                            i'                -

s

euuse .s.u. r a p uot. u) (4 of b) i - HEB 3-1, 55ctirn 8.1.c(2) '* SAR Secticn 3.6.2.A 2.~1 -

                                                                                                                                                          ,     ' , y+,

Intermedicte breaks in Class 2 and 3 piping are ~ Duke criteria specifies that if the ihresho'id '"

                                                                                     +

stress levels are'not exceeded, then'no inter-hig)heststress. S but at not less than postulated two locations based where Where the piping consists of a ce the stresses exceed mediale breaks 0.8 (1.25,! are postulated, straight run without fittings, welded attachments, " ,

                                                                                                                                                   i and valves, and all stresses are less than 0.8                      s
                                                                                                                      ~

(1.25 i 5 ),3a minimum of one ,locat,i.op should be l chose,!s based on highest stress. f - 3 , .. MEB 3-1. Sections B.I.c(/) C M O ) SAR Section 3.6.2.2.1 i Breaks in non nuclear piping should be postulated.- at the following location: * ' Duke criteria is ' roughly equivalent' to 'NRC criteria as described and just'ified below: s ,

a. Termglends,
b. _ Breaks i'n Duke Class F' piping (non-nucl' ear,"shic)

At each intermediate pipe fitting, welded attachment, are postulated at terminal ends and at intermediate and valve. , locations based on the use of ASME Section III analysis techniques, the same as Duke Class B and C piping. Duke Class F piping is constructed in accordance with ANSI.831.1 and is dynamically analyzed and restrained for seismic loadings similar to ASME Section III piping. Materials are specified, procured, received, stored, and issued under Duke's QA program similar to ASME Section III materials except that certificate of compliance in lieu of mill test reports are acceptable on minor components, and construction documentation for erected materials is not

                                           -                                                 uniquely maintained.       Construction documentation for erected materials is generically maintained.
                                                                                        ,,  HIR are required for the bulk.of piping materials.

MEB 3-1, Section B.2.e SAR Section 3.6.1.2 Thru-wall cracks may be postulated instead of breaks in those fluid systems that qualify as high energy Duke criteria is roughly equivalent to NRC criteria fluid syste:as for short operational periods. This as clarified below: cperational period id defined as about 2 percent of the time that the system operates as a moderate energy The operational period that classifies such systems fluid system. as moderate energy in either:

a. One percent of the normal operating lifespan of the plant, or
b. Two percent of the time period required to M accomplish its system desig'n function.

y .. . . . ..

                                                                                 ~ ,~ 4:

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                                                                                                                                                           -s Table 3.6.1-3 (Cont'd)                                           -

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                                                                                                                                                                    -(6 of 6) f                      Regulatory Guide 1.46 .-                      - - -                       -
                                                                                                         ~

SAR Section 3.6.1.2 i i  !!aasures for restraint against pipe whipping need not be provided for piping where: Duke criteria is roughly equivalent to fiRC Branch i Technical Position APCSB 3-1 and differs frcm (1) the design temperature is 200*F or less, and Regulatory Guide 1.46 with expansion of definition as described below: (2) the design pressure is 275 psig or less. i liigh energy piping is reviewed for pipe whipping and l

                                                                                    ,                -                 is defined as those systems that during normal plant conditions are either in operation or maintained l

pressurized under conditions where either or both of the folicwing are cet: d h a. j " naximum terperature exceeds 200 F, or 4 b. i maximum pressure exceeds 275 psig, except that (1) r.on-liquid piping system with a maximum pressure less than or equal to 275 psig are not considered high energy regardless of the tempera-ture, and (2) for liquid systems other than water, the atmospheric boiling temperature can be applied. Systems are classified as moderate energy if the total tire that less thaneither either:of the above conditions are met is a. cne (1) percent of the operating lifespan of the plant, or b. two (2) percent of the tice period required to accomplish its system design function. 12/83

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