ML20136F372
| ML20136F372 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 11/20/1985 |
| From: | Tucker H DUKE POWER CO. |
| To: | Adensam E, Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8511220107 | |
| Download: ML20136F372 (62) | |
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=A m-DUKE Powrm GOMMNY P.O. ISOX 33180 CHARLOTTE, N.C. 28242 HALB. TUCKER TELEPHONE vnos rumsment (704) 073-4tK31 stumaa emeworson November 20, 1985 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.
S. Nuclear Regulatory Commission Washington, D.
C.
20555 Attention:
Ms.
E.
G. Adensam, Chief Licensing Branch No. 4 Catawba Nuclear Station 50-413 and 56 Units 1 and 2 Re:
Docket No.
414
Dear Mr. Denton:
The purpose of this submittal is to request elimination of arbitrary intermediate breaks for Catawba Nuclear Station, Unit 1 and additional arbitrary intermediate breaks for Catawba Unit 2 (Pressurizer surge line and feedwater lines).
The elimination of arbitrary intermediate breaks on Catawba Unit 2 was addressed by a letter (SER) from T. M. Novak to H. B.
Tucker dated April 2, 1984.
We understand that other applications have been permitted prior to the NRC completing all of the changes in regulatory requirements.
The criteria changes will benefit Catawba 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.
Based on the above reasoning as further explained and justified in Attachment E, Duke Power is submitting this request for elimination of arbitrary intermediate breaks.
At the ACRS Subcommittee 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 Catawba Nuclear Station Unit 1 and for additional breaks for Unit 2 while improving overall safety and piping system reliability.
A summarv of the potential benefits which can be realized specifically Trom the elimination of arbitrary intermediate breaks for Catawba Nuclear Station Units 1 and 2 is provided in Attachment A.
Technical justification for this action is provided in Attachment B.
8511220107 851120 I g DR ADOCK 0500 3
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o Mr. Harold R. Denton, Director November 20, 1985 Page Two For Catawba Nuclear Station, Duke Power Company requests NRC approval for the application of alternative pipe break criteria (excluding the RCS primary loop) as follows:
1.
Arbitrary intermediate pipe breaks in all high energy piping systems be eliminated from the structural design basis when the following criteria are satisfied:
a.
For all piping systems, the stress criteria in Catawba FSAR Section 3.5.2 are not exceeded.
b.
For Class 1 piping systems, the stress criteria and the usage factors in Catawba FSAR Section 3.6.2 are not exceeded.
2.
The dynamic effects (pipe whip, et impingement, and compartment pressurization loads associated with arbitrary intermediate pipe breaks be excl ded from the plant design basis.
3.
The requirement for pipe whip restraints and jet shields associated with previously postulated arbitrary intermediate pipe breaks be eliminated.
Environmental qualification criteria will not be affected by elimination of any arbitrary breaks.
A summary of the currently postulated arbitrary intermediate pipe breaks to be climinated for Catawba Nuclear Station, Unit 1 and additional arbitrary intermediate breaks for Catawba Unit 2 is provided in Attachment C-1, and C-2; respectively.
In order to implement the revised criteria, we hereby take exception to those portions of the Standard Review Plan Sections 3.6.1 (ASB 3-1) - and 3. 6. 2 (MEB 3-1) which deal with this type of pipe break.
An exact description of the exception is given in Table 3.6.1-3 of the Catawba FSAR.
Attachment D is the Catawba FSAR revision associated with the elimination of arbitrary intermediate breaks.
Once approved, these changes will be incorporated into the next applicable annual update to the Catawba FSAR.
We request a decision on this proposal by January 31, 1986.
In accordance with 10 CFR 170.21, attached is a check for $150.00.
Very truly yours, 24NV%.
Hal B. Tucker ROS:slb Attachments
Mr. Harold R.
Denton, Director November 20, 1985 Page Three cc:
Dr. J. Nelson Grace Regional Administrator U. S. Nuclear Regula, tory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 NRC Resident Inspector Catawba Nuclear Station Dr. J.
N. Jabbour Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.
C.
20555
I ATTACHMENT A Summary of Benefits From the Elimination of Arbitrary Intermediate Pipe Breaks on Catawba Nuclear Station Units 1 & 2 Category Benefit 1)
Relief of congestion, 41 person-rem reduction in improving access for operation radiation exposure over life of and maintenance unit including removal of existing protective devices.
($37,000).
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 $276,000.
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 Intermediate Break elimination.
devices that would be required for these modifications or additions is 6.
Total Savings:
$313,000 and 41 person-rem for a unit life of 40 years.
MN30130D/4
a ATTACHMENT B Technical Justification for Elimination of Arbitrary Intermediate Pipe Break Postulation The following reasons provide generic technical justification for eliminating the arbitrary intermediate pipe break postulation required by Standard Review Plan 3.6:
1.
The pipe rupture " threshold" for all nuclear class piping is 80% of the ASME Code stress allowables.
All arbitrary intermediate breaks involve stresses below this level.
Hence a large conservatism exists.
2.
Pipe rupture is recognized in Branch Technical Position MEB 3-1 as being a
" rare event which may only occur under unanticipated conditions."
3.
There is no technical or other justification for postulating arbitrary intermediate breaks, other than providing additional conservatism.
4.
The additional pipe rupture devices resulting from this additional " layer" of conservatism may actually reduce rather than improve plant safety.
This has been demonstrated in " Effects of Postulated Event Devices on Normal Operation of Piping Systems in Nuclear Power Plant," NUREG/CR-2136, Teledyne Services, 1981.
Included among other improvements from arbitrary break elimination is improvement in performing ISI and a reduction in unanticipated restraint of piping due to thermal growth and seismic movement.
5.
Due to system design and operating procedures at Catawba Nuclear Station, the probability of stress corrosion, thermal and vibrational fatigue, or water hammer in the arbitrary intermediate break lines is not significant.
Technical justification for this position is given in Attachments B-1, B-2, and B-3.
6.
The environmental analysis for Catawba 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.
It is concluded that the elimination of arbitrary intermediate break postula-tion is technically justifiable for the foregoing reasons.
k MN30130D/5
l ATTACHMENT B-1 Protection of Arbitrary Break Lines From Stress Corrosion 4
In order for stress corrosion cracking (SCC) to occur in piping, the following three conditions must exist simultaneously:
high tensile stresses, a suscepti-ble material, and a corrosive environment (NUREG-0691).
Since some residual stresses and some 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 4
standards prior to operation and careful control of water chemistry during plant operation are used to prevent the occurrence of this environment.
t
- All piping involved in the elimination of arbitrary breaks at Catawba 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 Catawba 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 specifi-cations.
External cleaning for Duke Class A stainless steel piping includes patch tests to monitor and control chloride and flucride levels.
For preopera-tional 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 r
cause of stress corrosion cracking in BWR reactors, is more strictly controlled in the Catawba PWR environment than in a BWR environment. Oxygen concentration in the fluid in the Catawba stainless steel piping is expected to be less than 0.005 ppa 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 Catawba.
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 factor controlling steam generator blowdown and main steam chemical composition.
MN301300/6 f
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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 inci-
' dents 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 Catawba Thus, there is no evidence to suggest that SCC will occur at Catawba.
l MN301300/7
Attrchment B-1 WATER CHEMISTRY REQUIREMENTS DURING PLANT OPERATION Page 1 cf 2 TABLE 1 FOR LINES WITH PREVIOUSLY POSTULATED ARBITRARY BREAKS MAX.
DUKE CHLORIDES & MAX. CATION NO. ARBITRARY PIPING SYSTEM +
PIPING PIPING OPERATING MAX. 0 HYDROGEN FLOURIDES CONDUCTIVITY
- BREAKS ELIM.-
2 UNIT 1 MATERIAL CLASS TEMP (*F) (ppm)
CONCENTRATION (ppm)
(pmho/cm) pH*
CATAWBA R::ctor Ccolent SS A
557 0.10**
25-50cc/kg(H0) 0.15 2
15 R:sidual H:at Removal SS A
557 0.10**
25-50cc/kg(H 0) 0.15 6
134-445 0.003 0.2 8.8-9.3 8
445 0.003 0.2 8.8-9.3 12 Safety Injection (1)
SS A
618 0.10"*
25-50cc/kg(H 0) 0.15 2
14 Safety Injection (2)
SS A
Ambient ++
0.15 Low 14 Safety Injection (3)
SS B
Ambient ++
0.15 Low 12 Stram Generator Blowdown (1)
SS B
557 0.02 0.8 8.5-9.3 5
Steaa Generator Blowdown (2)
CS F
557 0.02 0.8 8.5-9.3 7
Ch":s & Vol.
Control (1)
SS A&B 130++
24 Chzm & Vol.
Control (2)
SS B
290 0.15 4
Main Steam CS B&F 557 0.3 8.8-9.3 15 MN301300/8
' Attachment B-1' WATER CHEMISTRY REQUIREMENTS DURING PLANT OPERATION Page 2 ef.2 TABLE 1 FOR LINES WITH PREVIOUSLY POSTULATED ARBITRARY BREAKS MAX.
DUKE CHLORIDES & MAX. CATION NO. ARBITRARY PIPING SYSTEM + : PIPING PIPING OPERATING MAX. 0 HYDROGEN FLOURIDES CONDUCTIVITY
- BREAKS ELIM.-
2 UNIT 1 MATERIAL CLASS
. TEMP (*F) (ppm)
CONCENTRATION (ppm):
(paho/cm) pH*
CATAWBA Main Steam to Aux Equip CS B
557 0.3 8.8-9.3 4
^
Vent to Atmos CS B
557 0.3 8.8-9.3 8-UNIT 2 Reactor Coolant SS A
557 0.10**
25-50cc/kg(H 0) 0.15-2 4
Main l
Fesdwater CS B
445 0.003 8.8-9.3 12
- Stress corrosion not considered a problem at this operating temperature.
- Besed on EPRI-NP-2704-SR (Steam Generator Owner's Group Secondary Water Chemistry Guidelines.
Thsso standards are met at Catawba Nuclear Station).
I
- Data is for the portions of the system where arbitrary breaks were previously postulated.
- Tech Spec limit; 0xygen concentration expected to be less than 0.005 ppm during most power operation.
}
1 i
+
e MN30130D/9
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ATTACHMENT B-2 THERMAL AND VIBRATION FATIGUE IN PIPING ASSOCIATED WITH CATAWBA ARBITRARY INTERMEDIATE BREAKS For Catawba 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 i
piping components, fatigue failure protection is provided for by the allowable l
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 Catawba 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 and monitoring temperature near the nozzle to alarm 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,below11%poweror25(ffeedwatertemperature.
At this condition, temperature is monitored near the main nozzle to alarm should inflow of low temperature feedwater occur.
At Catawba Unit 1 only eight of the approximately 148 eliminated arbitrary intermediate breaks occur in the auxiliary feedwater system and 12 breaks occur in feedwater system.
None of these breaks, including additional feed-water system breaks for Catawba Unit 2, are located in horizontal pipe runs at the nozzles.
1 Cyclic thermal stress is prevented in the arbitrary break lines in the remain-ing systems (listed below) by maintaining uniform temperatures with no mixing:
1.
Steam Generator Blowdown 2.
Main Steam Supply to Auxiliary Equipment 4.
Main Steam 5.
Main Steam Vent to Atmosphere 6.
Safety Injection 8.
Chemical and Volume Control MN30130D/10
o The potential for vibration fatigue in Catawba piping systems is minimized through pre-operational vibration tests.
Table 14.2.12-1 of the Catawba FSAR describes the Piping System Vibration Test which is part of the Preoperational Test Program at Catawbo.
The purpose of the test is to verify the following:
A.
Piping layout and support / restraints are adequate to withstand normal transients without damage to piping system, and 8.
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 measurement of piping velocity.
These measurements are evaluated with respect to the following acceptance criteria and any required modifica-tions 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, between stops or with indicators on scale.
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.
Catawba FSAR Tables 3.9.2-1 and 3.9.2-1 (a) lists the systems to be included in the steady state test programs and transient test programs, respectively.
These tables are included on the following two pages.
Based on the information presented in this attachment, no problems from thermal or vibrational fatigue would be expected at Catawba Nuclear Station.
MN30130D/11
A
. :.=
r Table 3.9.2-1 Piping Systems Included In Vibration Test Program
' System Reactor: Coolant System Safety In'jection System Residual Heat Removal System Containment Spray System Chemical and Volume Control System Boron. Recycle System
. Boron Thermal Regeneration System
-Component = Cooling System
-f Liquid Radwaste System Fuel Pool Cooling and Cleanup System Diesel Generator Fuel Oil; System Diesel Generator Cooling Water System i
-Diesel Generator Lub Oil System I
Nuclear l Service Water System
-Refueling Water' System l
Main Steam' System l:
L Feedwater System Auxiliary _Feedwater System lSteamDumpSystem
' Control Area Chilled Water System
. Steam Generator Blowdown Recycle System Recirculated Cooling Water System p-Rev. 1
m TABLE 3.9.2-la
' Piping Systems Included In Transient Vibration Test Program System Transient Type Type Measurement Simultaneous Test NC NC Pump Start Vibration measurement HFT at selected points NC Pump Trip Vibration measurement HFT at selected points NC PROV Cycling Post transient inspection HFT BB Initiation of S/G Post transient inspection S/G BD Test Blowdown Isolation of S/G Post transient inspection S/G BD Test Blowdown CA (Motor driven Pump Post transient inspection Aux. FDW F.T.
Start)
(Motor driven Pump Post transient inspection Aux. FDW F.T.
trip AFWPT Cold Start Post transient inspection Aux. FDW F.T.
AFWPT Trip Post transient inspection Aux. FDW F.T.
'CF Isolation Valve Post transient inspection HFT Closure NI NI Pump _ Start Post transient inspection ESF of pump discharge piping l
CCP' Pump Start Post transient inspection ESF of pump discharge piping NV Letdown isolation Post transient inspection HFT SM Main Steam Isolation Post transient inspection SM isolation HFT (individually)
Main Steam PORV Post transient inspection HFT
. Discharge l
CF, SM Loss of Electrical Post transient inspection Power Escalation Load 100% FP Testing CF, SM Turbine Trip Post transient inspection Power Escalation
$70% FP Testing Rev. 9
ATTACHMENT B-3 WATER / STEAM HAMMER IN PIPING ASSOCIATED WITH CATAWBA ARBITARARY INTERMEDIATE BREAKS 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 Gene qter Blowdown Fluid flow through the Steam Generator Blowdown Lines is normally two phase and of 0-10% quality.
There is very little chanca of water hammer in these lines inside Containment.
A greater susceptibility to water hammer exists outside Containment, with the greatest poteotial occurring upon reinitiation of flow following containment isolat'.an.
Some problems have occurred in this area at McGuire Nuclear Station.
However, design and operating procedures have been improved for McGuire and Catawba, and minimal water hammer problems are expected due to provisions to gradually repressurize the downstream piping before establishing full flow.
Since these changes have been initiated at McGuire,; there has been no recurrence of water hammer induced damage.
2.
Auxiliary Feedwater 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 vertical run of pipe to minimize steam voids.
Also, hmpering 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 suspected.
The piping involved in the arbitrary intermediate breaks contains thermowells allowing such tempera-ture increases to be detected and the proper operating procedures to be implemented; also the associated valves are periodically checked for leaks.
The Auxiliary Feedwater System at the McGuire Nuclear Station is similar to the one at Catawba, and McGuire has experienced no water hammer problems to date.
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 proce-dures have been modified to eliminate water hammer at McGuire and Catawba.
MN301300/14
i 4.
Residual Heat Removal The portions of the Residual Heat Removal System having arbitrary interme-diate breaks are within the Reactor Coolant System boundary prior to (or at) the first valve off the Reactor Coolant Loops.
Therefore, the explana-tion in part 3 above is applicable here.
5.
Safety Injection The Safety Injection System lines are all either water solid or gaseous lines.
Steam voids would not be expected, especia",1y since approximately 60% of the arbitrary intermediate breaks are in ambient temperature 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 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.
7c Main Steam Supply to Auxiliary Equipment These lines have redundant heat tracing to prevent water accumulation.
l Although the heat tracing is not nuclear safety related, it is required to be functional for system operation.
Therefore, the potential for water hammer problems is very low.
8.
Main Steam Because of the design and function of the two-inch Main Steam System lines containing arbitrary intermediate breaks, no water hammer problem exists.
The,two-inch warm-up lines are normally filled with steam and the steam drain lines are designed for water flow with no accumulation.
9.
Main Steam Vent to Atmosphere The breaks in the Main Steam Vent to Atmosphere System are in the vicinity of the PORV's which release steam to the atmosphere.
There is no chance of water hammer as long as the Main Steam line stays drained.
McGuire experience has shown this to be the case.
There is only very minimal probability of-significant steam hammer occurring
-in any of the lines involved in elimination of arbitrary breaks.
The function of the main steam line drains is to remove any liquid from the main steam lines.
Significant water or steam hammer will not occur withcut liquid.
MN30130D/15
- 10. Main Feedwater Catawba has Westinghouse preheat steam generators. Westinghouse recommen-dations 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 possi-ble; however, operating procedures and design considerations prevent water hammer during restart by providing a controlled flow to flush the voids from the system slowly.
MN30130D/16
ATTACHMENT C-1 POSTULATED ARBITRARY INTERMEDIATE BREAKS TO BE ELIMINATED ON CATAWBA UNIT 1 Estimated No.
Devices Eliminated Number Breaks Rupture Jet Piping System Pipe Diameter Location Eliminated Restraints Deflectors Steam Generator 2
IC 3
0 0
Blowdown 4
IC 2
3 0
2 OC 6
0 0
4 OC 1
0 0
Auxiliary 6
IC 8
18 1
Feedwater Reactor 2
IC 11 0
0 Coolant 14 IC 4
0 1
Residual Heat 12 IC 6
0 1
Removal Safety 2
IC 4
0 0
Injection 5
IC 4
0 0
6 IC 5
2 2
8 IC 4
0 0
10 IC 11 1
3 1
OC 3
0 0
2 OC 5
0 0
12 OC 4
0 0
Chemical & Volume 2
IC 14 5
1 Control 3
IC 2
11 1
2 OC 10 0
0 3
OC 2
0 0
Main Steam Supply 6
0C 4
0 0
to Auxiliary Equip.
OC 15 0
0 Main Steam Vent 6
0C 8
0 0
at Atmosphere Main Feedwater 18 OC 12 18 5
Total 148 58 15 I.C. = Inside Containment 0.C. - Outside Containment MN30130D/17
ATTACHMENT C-2 Additional Poetulated Arbitrary Intermediate Breaks TO BE ELIMINATED AT CATAWBA NUCLEAR STATION UNIT 2 Estimated No.
Devices Eliminated Number Breaks Rupture Jet Piping System Pipe Diameter Location Eliminated Restraints Deflectors t
Reactor Coolant 14 IC 4
0 0
Main Feedwater 18 OC 10 14 6
Total 14 14 6
MN301300/18
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2 ATTACHMENT D REVISION TO CATAWBA FSAR 4
FOR ARBITRARY INTERMEDIATE BREAK CRITERIA CHANGE i
i 4
4 5
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5 e
L 3
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- 3. 6 -
PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING a
General Design Criterion 4 of Appendix A_to 10CFR50 required that structures, systems, and components important to_ safety be protected from the dynamic effects of pipe failure.
This section describes the design bases and design measures to ensure that.the containment vessel and all essential eouipment in-i
. side or,outside the' containment, including components of the reactor coolant pressure boundary, have been adequately protected against the effects of blow-i down jet and reactive forces and pipe whip resulting from postulated rupture of piping.
Criteria presented herein regarding break size, shape, orientation, and location are in accordance-with the guidelines established by NRC Regulatory Guide 1.46, and include considerations which are further clarified in NRC Branch Technical Position MEB 3-1 and APCSB 3-1 where appropriate.
These criteria are intended
'to be' conservative and allow a high margin of safety.
For.those pipe failures where_ portions of these criteria lead to unacceptable consequences, further analyses will be performed.
However, any alternative criteria will be j
adequately justified and fully documented.
.Some criteria as stated in this section applies only to Unit 1, as certain exemptions to the above requirements have been granted for Unit 2 by the NRC.
In particular, Reference 5 allows the elimination of postulated arbitrary
_ intermediate breaks for Unit 2, and Reference 6 allows the deletion of the 4
requirement to consider the dynamic effects of large primary loop ruptures.
. ~.
3.6.1
. POSTULATED PIPING FAILURES IN FLUID SYSTEMS INSIDE AND OUTSIDE CONTAINMENT
- 3. 6.1.1 Design Bases 3.6.1.1.1 Reactor Coolant System The Reactor Coolant System, as used in Section 3.6 of the Safety Analysis Report, is limited to the main coolant loop piping and all branch connection nozzles out to the first butt weld.
The particular arrangement of the Reactor Coolant System, building structures, and mechanical restraints preclude the formation of 4-plastic hinges for breaks postulated to occur in the Reactor Coolant System.
Consequently, pipe whip and jet impingement effects of the postulated pipe MEB break will not result in unacceptable consequences to essential components.
Q 102 This restraint configuration, along with the particular arrangement of the
. Reactor Coolant System and building structures, mitigates the effects of the
_ jet from the given break such that no unacceptable consequences to essential components are experienced.
The application of criteria for protection against the effects of postulated breaks in the Reactor Coolant System in accordance with Reference 1 results-in a system response which can be accommodated directly by the supporting i
MEB l _ structures of the reactor vessel, the steam gsnerator, and the reactor coolant-Q 102' l pumps including seven additional pipe supports. The design bases for postulated breaks in the Reactor Coolant System are discussed in Section 3.6.2.1.
i 3.6-1 Rev. 13
CNS 3.6.1.1.2 All Other Mechanical Piping Systems
)
This section discusses all piping systems excluding the Reactor Coolant System as described in Section 3.6.1.1.1 and is in accordance with NRC Branch Technical Position APCSB 3-1 and Regulatory Guide 1.46 except as noted in Table 3.6.1-3.
Other mecnanical piping systems, both inside and outside containment, which are reviewed and considered in the design with respect to postulated pipe break are those normally operating high-energy and moderate-energy lines which are safety-related or pass near safety-related structures, systems or components, and include the Reactor Coolant System branch piping terminating at the main cool-ant loop nozzle piping.
High-energy piping systems are those sy.;tems, or portions of systems, that during normal plant conditions are either in operation or maintained pressurized under conditions where either one or both of the following are met:
a)
Maximum operating temperature exceeds 200'F or b)
Maximum operating pressure exceeds 275 psig.
Except that (1) non-liquid piping systems (air, gas, steam) with a maximum pres-sure 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 in place of the 200*F criterion.
Moderate-energy through-wall cracks as defined in Section 3.6.2.1.2.3.C are assumed in these piping systems and the environmental effects of pressure, temperature, humidity, flooding, and wetting of equipment are all considered in the station analysis.
Systems are classified as moderate-energy if the total time that either of the above conditions a) or b) are met is less than either of the following:
a)
One percent of the normal operating lifespan of the plant, or b)
Two percent of the time period required to accomplish its system design function.
Moderate-Energy Piping Systems are those systems, or portions of systems, that during normal plant conditions are either in operaison or maintained pressurized (above atmospheric pressure) under conditions where both of the following are met:
a)
Maximum operating temperature is 200 F or less (or less than the atmospheric boiling temperature for non-water systems), and b)
Maximum operating pressure is 275 psig or less.
CNS Systems which do not contain mechanical pressurization equipment are excluded C'
from moderate-energy classification (e.g., systems without pumps, pressurizing tanks, boilers, or those which operate only from gravity flow or storage tank o
water head), however, limited failures are assumed to occur for the purpose of considering the effects of flooding, spray, and wetting of equipment in the station analysis.
The identification of piping failure locations will be performed in accordance with Section 3.6.2.
3.6.1.1.2.1 Interaction Criteria The following criteria define how interactions shall be evaluated.
The safety evaluation of each interaction is described in Sections 3.6.1.3 and 3.6.1.1.5.
a)
Environmental Interaction An active component (electrical, mechanical, and instrumentation and control) is assumed incapable of performing its function upon experiencing environmental conditions exceeding any of its environmental ratings.
b)
Jet Impingement Interactions Active components (electrical, mechanical, and instrumentation and control) subjected to a jet are assumed failed unless the active component is en-closed in a qualified enclosure, the component is known to be insensitive to such an environment, or unless shown by analysis that the active function C
will not be impaired.
c)
Pipe Whip Interaction l
A whipping pipe is not considered to inflict unacceptable damage to other pipes of equal or greater size and wall. thickness.
A whipping pipe is only considered capable of developing through-wall leakage cracks in other pipes of equal or greater size with smaller wall thickness.
An active component (electrical, mechanical, and instrumentation and control) is assumed incapable of performing its active function following impact by any whipping pipe unless an analysis or test is conducted to show otherwise.
3.6.1.1.3 Protective Measures 3.6.1.1.3.1 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 3.6-3 Rev. 13
CNS are considered in assuring the continued integrity of the vital components and j
the engineered safety features.
/
To 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 or feedwater line break, propagation of damage from the original event
-is limited,~ and the components as needed, are protected and available.
Large Reactor Coolant System piping (6 inch nominal or larger piping) and all connecting piping out to the LOCA boundary valve (Figure 3.6.2-1) is restrained to meet'the following criteria:
a)
Propa'gation 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 affected loop is permitted to occur but is limited by piping separation and restraints so as not to 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 lof the loss of coolant.
(See also paragraph K.3 of Section 3.6.2.1.2).
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.
Additional pipe restraint design criteria are discussed in Reference 1.
[
\\_
In addition to pipe' restraints, barriers and layout are used to provide pro-tection 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 follow-ing. The polar crane wall 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 in a separate compartment; thereby preventing an accident in any loop from affecting another loop or the Containment.
The portion of the main steam and feedwater lines within the Containment has been routed behind barriers to separate these lines from reactor coolant piping.
The barriers described above are designed to withstand loadings resulting from jet and pipe whip impact forces.
Other than Emergency Core Cooling System lines, all 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 and then penetrate the crane wall in the vicinity of the loop to which they are attached.
f~
CNS N
3.6.'.1.3.2 All Other Mechanical Piping Systems Preferred measures to protect against pipe whip, jet impingement and resulting reactive forces to assure plant safety 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, alternately, enclosure of the essential
. equipment.
c)
Provision of system-redundant design features separated, or otherwise protected, from the effects of the postulated pipe rupture; or additional protection features such as rupture restraints and jet deflectors.
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.
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 condenser circulating water expansion joint, or the most severe Condensate System failure.
There are no pipe or cable chase entrances below the elevation of the top of the curbs.
This flooding condition does not render any essential system or component inoperable.
3.6.1.1.3.3' Main Steam And Feedwater System Design Design of the Main Steam and Feedwater Systems meets the general design criteria; however, additional specific information as follows applies to these systems, a)
Main Steam Lines are 100 percent cold pulled in the horizontal direction on Unit 1.
Unit 2 main steam lines are not cold pulled.
b)
Overpressure capability of the piping based on actual wall thicknesses is as follows:
Actual Normal Opera-Code Pressure ting Pressure Capability Margin Main Steam:
985 psig 1250 psig 19%
1165 psig 1420 psig 22%
Feedwater:
7 s
(
i 3.6-5 Rev. 7
/
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 Section III, Class 2.
d)
The guard pipe.on the main steam piping inside containment is extended over the length of the vertical portion of the piping to prevent jet impingement on the ice condenser doors following a postulated rupture.
e)
As a result of a Duke-NRC meeting in May 1976, guard pipe was removed from the main steam and main feedwater piping in the doghouse which was originally designed with guard pipe.
Breaks in the Main Steam piping will be postulated based on consequence except for the break exclusion region inside the doghouse.
For the main steam piping which is part of the break exclusion region an augmented inservice inspection will be performed, as discussed in Section 6.6, for welds where no break is selected.
Breaks in the main feedwater will be selected as outlined in Section 3.6.2.
3.6.1.1.3.4 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 low pressure, low volume heating steam.
On the west side, the control room is bounded by the computer room and supporting areas.
Piping in this area consists of sanitary waste and A
f vent piping, drinking water and instrument air; none of which are high-energy
(
systems.
Immediately below the control room is the cable room containing no piping.
Based on the above physical parameters, the control room is structurally isolated from areas containing high-energy systems; therefore, there are no unacceptable consequences to the control room from the postulated break of high-energy piping systems.
3.6.1.1.4 Acceptability Criteria Tho capability to eventually achieve a cold shutdown condition is not jeopardized even if the pipe failure is followed by a single active failure.
The system rew irements and available redundancy are determined on a shutdown logic diagram, or c required equipmer.t list for mitigating the effects cf the postulated failure.
Repair of tallures is considered to assure acnievement of the cold snutdown condition where such repairs can be shown to be practical and timely, and pro-vided the unit can be held in a safe shutdown state during the time required for the repair.
3.6.1.2
-Description of Piping System Arrangement Separation is the primary consideration in piping system layout and arrangement.
Where physical separation is not feasible, protective devices are provided to protect essential components.
3.6-6 Rev. 7
%) * * -
CNS x
Table 3.6.1-1 provides a listing of high-energy systems.
Moderate-energy systems are listed in Table 3.6.1-2.
Control room habitability is discussed in Section 3.6.1.1.3.4.
3.6.1.3 Safety Evaluation Safety functions are identified for each initiating event by the failure mode and effects analysis discussed in Section 3.6.2.1.2.
For ead postulated failure, every credible unacceptable interaction shall be evaluated.
In establishing system requirements for each postulated break, it is assumed that a single active component failure occurs concurrently with the postulated rupture.
3.6.2 DETERMINATION OF BREAK LOCATIONS AND DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING 3.6.2.1 Criteria Used To Define Break And Crack Location And Configuration 3.6.2.1.1 Postulated Piping Break Location Criteria for the Reactor Cool-ant System The design basis for postulated pipe breaks of the reactor coolant loop piping includes not only the break criteria, but also the criteria to protect other
,[
piping and vital systems from the effects of the postulated break.
'g A loss of reactor coolant accident is assumed to occur for a pipe break in piping down to the restraint of the second normally open automatic isolation valve (Case II in Figure 3.6.2-1) on outgoing lines (*) and down to and in-cluding the second check valve (Case III in Figure 3.6.2-1) on incoming lines norreally with flow. A pipe break beyond the restraint or second check valve does not result in an uncontrolled loss of reactor coolant assuming either of the two check valves in the line close.
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.
Periodic testing i., performed of the capability of the valves to perform their intended function.
This criterion takes credit for only one of the two valves performing its intended function.
For normally closed isolation or incoming check valves (Cases 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.
In any given piping system, there is a limited number of locations which are rore susceptible to failure by virtue of stress or fatigue than the remainde-of the system.
These postulated break locations are defined in Section 3.6.2.1.2 considering normal and upset operating conditions (defined by the applicable Design Specification as required by ASME Code,Section III).
.g'
(*)It is assumed that motion of the unsupported line containing the isolation valves could cause failure of the operators of both valves.
3.6-7 a
i b
Engineered Safety Features are provided for core cooling, 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 to provide protection for a Reactor Coolant System pipe rupture of a size up to and including a double-ended severence of the Reactor Coolant System main loop.
Branch lines connected to the Reactor Coolant System are defined as "small" if This size is based they have an inside diameter equal to or less than 4 inches.
on Emergency Core Cooling System analyses using realistic assumptions that show that no clad damage is expected for a break area of up to 12.5 square inches j
corresponding to 4 inches inside diameter piping, In order to assure the continued integrity of the vital components and the l
engineered safety systems, consideration is given to the consequential effects of-the pipe break to the extent that:
The minimum performance capabilities of the engineered safety systems are a) not reduced below that required to protect against the postulated break.
The Containment leak tightness is not decreased below the design value, if b) the break leads to a loss of reactor coolant (*).
Propagation of damage is limited in type and/or degree to the extent that:
c)
A pipe break which is not a loss of reactor coolant does not cause N
1) a loss of reactor coolant or steam or feedwater line break, and A Reactor Coolant System pipe break does not cause a steam-feedwater 2) system pipe break and vice versa.
In the unlikely event that one of the small (as defined above) pressurized lines should fail and result in a loss of reactor ccolant accident, the piping is re-strained or arranged to meet the following criteria in addition to a. through c.
above:
Break propagation must be Jimited to the affected leg (i.e., propagation a) to the other legs of the affect loop and to other loops is prevented) withtheexceptionofthe['RTD ass Lines which are allowed to rupture either of the other RTD li'ne in the affected ioop.
Propagation of the break in the affected leg is permitted but is limited b) by piping separation and restraints to a total break area of 12.5 square The exception to this case is when the inches (4 inch inside diameter).
initiating small break is the high head safety injection line; further propagation is not permitted for this case.
(*)The containment is defined here as the Containment vessel and pentrations, the steam generator shell, the steam generator steam side instrumentation connections, and the steam, feedwater, blowdown, and steam generator drain pipes within the Containment Structure.
3.6-8
O 7)
(
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 due to a spilling injection line.
3.6.2.1.1.1 Postulated Piping Break Locations and Orientations In each leg of the Reactor Coolant System, a minimum of three postulated rupture locations shall be selected in the following manner:
MEB For the Reactor Coolant System breaks are postulated as described in WCAP-Q13 8082 (Reference 1).
At each possible break location, consideration must be given to the occurrence of either a circumferential or longitudinal break.
As discussed in Reference 1, a circumferential rupture is more likely than a longitudinal rupture for reactor coolant piping.
Only in the case of one elbow is a longitudinal rupture post-ulated.
Circumferential breaks are perpendicular to the longitudinal axis of the pipe.
lingitudinal breaks are parallel to the longitudinal axis of the pipe.
Certain C),
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 Reference 1.
Break type at each discrete break location are presented in Table 3.6.2-1.
The results of the analyses which lead to the break orientations are discussed in Reference 1.
3.6.2.1.1.2 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 limited by analysis, experiment or physical restraint.
For a longitudinal break, the break area is the cross sectional area of the pipe at the break location unless analytically or experimentally shown other-Q 5 l wise.
A longitudinal break-area equal to the cross sectional area of the
.?
3.6-9 Rev. 7 e
CNS pipe is assumed for Break 7 in Table 3.6.2-1 (50 elbow on the Intrados).
Tne analytical and experimental bases of the break area assumption are supplied in Reference 1.
3.6.2.1.1.3 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 up 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 provided by the accumulators and the low head safety injection pumps (residual heat removal pumps).
3.6.2.1.2 General Design Criteria for Postulated Piping Breaks Other Than Reactor Coolant System a)
Station desigr considers and accommodates the effects of postulated pipe breaks with respect to pipe whip, jet impingement and resulting reactive forces for piping both inside and outside Containment.
The analytical methods utilized to assure that concurrent single active component failure and pipe break effects do not jeopardize the safe shutdown of the reactor are outlined in Section 3.6.2.3.
b)
Station general arrangement and layout design of high-energy systems
[
utilize the possible combination of physical separation, pipe bends,
(
pipe whip restraints and encased or jacketed piping for the most practical design of the station.
These possible design combinations decrease postu-lated piping break consequences to minimum and acceptable levels.
In all cases, the design is of a nature to mitigate the consequences of the break so that the reactor can be shutdown safely and eventually maintained in a cold shutdown condition.
c)
The environmental effects of pressure, temperature and flooding are con-trolled to acceptable levels utilizing restraints, level alarms and/or other warning devices, and vent openings.
d)
Plant Operating Conditions 1)
Power Level - At the time of the postulated pipe break, the plant is assumed to be in the normal mode of plant operation in which the piping under investigation experiences the maximum conditions of pressure and temperature.
In cases where this mode is full power operation, the power level assumed is that assumed in the evaluation of the loss-of-coolant accident, steamline break accident, or feed-water line break.
2)
Offsite Power - is assumed to be unavailable if a trip of the Turbine-Generator System or Reactor Protection System is a direct consequence of the postulated piping failure, (e.g., a loss-of-coolant accident, steam line break or feedwater line break).
3.6-10
7 CNS C.
3)
Seismic loadings equivalent to the Operating Basis Earthquake (0BE) are used in the analysis of piping systems for the purpose of post-ulating break locations.
Protective structures are designed to with-stand the effects of the postulated piping failure in combination with loadings associated with the Operating Basis Earthquake (0BE) and Safe Shutdown Earthquake (SSE) within the respective design load limits for the structures.
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 that system only are not assumed; provided the system is designed to seismic Category I 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 reactor shutdown of the affected unit cannot preclude the capability for safe shutdown of the reactor of the unaffected unit (s).
g)
Containment structural integrity is maintained by limiting the combination of break sizes and types to the design basis capability (i.e., ticperature, pressure, and leakage rate) of the containment.
h)
For those postulated breaks classified as a loss-of-reactor coolant, the design leak tightness of the containment fission product barrier shall be maintained.
i)
The conditions within the control room or any other location where manual action is required to assure safe shutdown to the cold condition are such as to assure habitability and comply with the require 2cnts 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 thi-kness.
Smaller and thinner pipes are assumed to encounter unacceptable damage upon impact. A whipping pipe or jet is considered capable of developing through-wall leakage cracks in equal or 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.
. If such exception is taken, the analytical technique or experimental data used will be documented in the FSAR.
The analytical technique for l
selected jet load evaluation is:
a) jet loads on lines smaller (or larger I
with thinner walls) than the source line are calculated in accordance with section 3.6.2.2.2.3, and, b) evaluated by including the jet loads in the target piping stress analysis.
3.6-11 Rev. 12
5 CNS
~
l k)
Piping Breaks Within The LOCA Boundary (See Figure 3.6.2-1) 1)
All LOCA breaks are allowed to damage any non-LOCA line except essen-tial systems, and steam and feedwater lines.
l 1
-=%.g 3.6-11a Rev. 10
CNS C
2)
Pipe breaks within the LOCA boundary are allowed to damage ECCS lines connecting to the ruptured line, providing that ECCS flow to other loops is maintained.
3)
For breaks in 6 inch nominal or larger piping, propagation of the break in the affected loop is not permitted if the resultant break area is more than 120 percent of the originating break area.
If the originating break is a Reactor Coolant System main loop break, pro-pagation is permitted to occur but must not exceed the design basis for calculating containment and subcompartment pressures, loop hy-draulic forces, reactor internals, reaction loads,-primary equipment support loads, or ECCS performance.
Propagation of the break is limited by the use of piping separation and restraints.
(See also Section 3.6.1.1.3.1).
Propagation to any other loop is not permitted in any case.
A rupture of the piping between the pressurizer and the code safety and power operated relief valves results in release of saturated steam and is less severe than for an equal size break through which reactor coolant fluid discharges because of the higher specific volume of the steam.
Because of this difference the following criteria apply to this specific case:
a)
Propagation of the break should not result in release of reactor coolant (liquid phase).
b)
Except for the upper level taps as discussed below, unlimited break propagation of the lines connected to the pressurizer above the normal water level is acceptable, c)
Propagation of the rupture must not result in breakage of more than one upper level tap.
4)
Pipe breaks within the LOCA boundary that are equal to or less than 4 inch nominal pipe size must meet the criteria as outlined in Section 3.6.2.1.1.
1)
Piping Breaks Outside the LOCA Boundary (Non-LOCA) 1)
A pipe break which is not a loss-of-coolant accident cannot cause a loss-of-coolant accident or steam or feedwater line break.
2)
All non-LOCA breaks (except steam and feedwater line breaks) are allowed to damage the non-LOCA portion of a single train of an ESF system, provided that unit shutdown can be achieved when considering a single active failure as described in item e) of this section.
3)
All non-LOCA breaks (excluding steam and feedwater line breaks) tre allowed to damage any non-LOCA, non-essential lines (except steam and feedwater lines).
3.6-12
- d y
CNS 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 damage to the opposite train of that system or any other essential system.
5)
A pipe break in a non-seismic system (Duke System Piping Class 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 ej feedwater lines, respectively, of the same steam generator, provided.{
that the aggregate break size does not exceed the applicable maximum ],
break size consi.dered in the safety analysis.
~
m 2)
Steam and feedwater line breaks can damage any non-LOCA lines except required essential system 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 safety shutdown and eventually maintain the reactor in a cold 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 eventually maintain the reactor in a cold shutdown condition.
p)
Minimum essential component and systems performance is provided as required for the type of break.
q)
The effects of pipe ruptures are not allowed to result in offsite doses in excess of 10CFR100 allowable limits.
r)
Operability in an environment is assured for all equipment. required to mitigate the break by the equipment specification requirements based on conservative design conditions.
s)
Emergency procedures are prepared that are to be followed after a postulated piping break for high-energy systems as required.
3.6.2.1.2.1 Postulated Piping Break Locations For High-Energy Piping Systems Systems identified as containing high-energy piping are examined by detailed design drawing review for postulated pipe breaks as defined below.
Systems analyzed for consequences of postulated piping breaks are listed in Table 3.6.1-1.
-L-l 3.6-13
+
CNS Terminal ends.are considered as piping originating 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 terminal ends.
The. branch connection to the main run is one of the terminal ends of a branch run, except intersections of runs of comparable size and fixity which have a significant effect on the main run need not be considered terminal ends when the stress analysis model-includes both the run and branch piping and the intersection is not rigidly constrained to the building structure.
a)
Breaks in Duke Class A piping are postulated at the following locations (see Table 3.2.2-3 for class correlations):
1)
The terminal ends of the pressurized portions of the run.
2)
At intermediate locations selected by either one of.the following methods:
i)
At each weld location of potential high stress or fatigue, such as pipe fittings (elbows, tees, reducers, etc.), valves, flanges, and welded attachments, or ii) At all intermediate locations between terminal ends where the following stress and fatigue limits are exceeded, a)
The maximum stress range shall not exceed 2.4 S,except as noted below, b)
The maximum stress range between any two load sets (includ-ing the zero load set) shall 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 S ) but is not greater than 3 S,,
the limit of U < 0.1 shall*be met.
i If the calculated maximum stress range of Eq. (10) exceeds 3 S*(, the stress ranges calculated by both Eq. (12) and Eq.
- 13) in Pa'ragraph NB-3653 shall not exceed 2.4 S r
m the limit of U < 0.1.
i l
c)
U shall not exceed 0.1.
where:
Primary plus secondary stress-intensity range, as S
=
n calculated from Equation (10) in Subarticle NB-3600
^
of the ASME Boiler and Pressure Vessel Code,Section III.
-3.6-14 Rev. 7 m -.
F 1
CNS allowabledesignstresskntensityvalue,as
.~
S
=
defined in-Subarticle NB-3600 of the ASME Boiler and Pressure Vessel Code,Section III.
thecumulativ[usagefactor,ascalculatedin U
=
accordance with Subarticle N8-3600 of the ASME Boiler and Pressure Vessel Code,Section III.
K 3)
If there are no intermediate locations where S exceeds 2.4 S or U
=
exceeds 0.1, M !
N 8 'N
((t [m Ye N W $ ed upe hi@ert t=.
Intermediate breaks are not postulated in sections of straight M
/"*N[ pipe where there are no pipe fittings, flanges, valves or welded at-tachments.
4)
As a result of piping reanalysis which follows the completion of the original problem interaction analysis, the highest stress lo-cations may be shifted however, the initially determined inter-mediatebreaklocationf;3will not be changed unless one of the fol-lowing conditions exists.
ygg jg,;,g J,,ty9 ja,,gj,/e. bcd,
- 1) Maximum stress ranges or cumulative usage factors exceed the threshold levels in 2)ii) a), b) or c) above.
ii) A change is required in pipe parameters, such as major dif-ferences in pipe size or wall thickness.
iii) Breaks at the new highest stress locations are determined to have significantly higher stress values from the original locations, and result in consequences to safety related sys-tems requiring additional safety protection.
b)
Breaks in Duke Class B and C piping are postulated at the following locations (See Table 3.2.2-3 for class correlations):
1)
The terminal ends of the pressurized portions of the run.
2)
At intermediate locations selected by either one of the following methods:
1) at each weld location of potential high stress or fatigue, such as pipe fittings (elbows, tees, reducers, etc.), valves, flanges and welded attachments, or ii) at all locations where the stress, S, exceeds 0.8 (1.2Sh
- S )'
A where:
[
S
=
stresses under the combination of loadings associated A
with the normal and upset plant condition loadings and an OBE event, as calculated from the sum of equations 3.6-15 Rev. 7
CNS (9) and (10) in Subarticle NC-3600 of the ASME Boiler and Pressure Vessel Code,Section III.
y S
=
basic material allowable stress af maximum (hot) tem-h perature from the allowable stress tables in Appendix I of the ASME Boiler and Pressure Vessel Code,Section III.
y allowable stress range for expansion stresses, as de-
'}
S
=
A fined in Subarticle NC-3600 of the ASME Boiler and
=
Pressure Vessel Code,Section III.
3)
If there are -" "NP " " a intermediate locations where S exceeds 1
0.8 (1.2 Sh + S )' " id8Y,fde$bDP 'cN$j }gjatjg37agMsm A
L Ibssed upon-Mghest-srtress.
Intermediate breaks are not postulated in sections of straight pipe where there are no pipe fittings, flanges, Mg valves, or welded attachments.
The pattern of postulated intermediate
/gaf*[{ I reak locations is determined separately for the normal plant condition b
load combination and for that upset plant condition which has the high-est stress.
~4)
As a result of piping reanalysis which follows the completion of the original problem interaction analysis, the highest stress locations
(
may be shifted; however, the initially determined intermediate break exists. jwill not be changed unless one of the following conditions locations
, yj g,,,,.,4,g,j,;j,,,, ;g,,,,,4s j,aq i) Maximum stress ranges exceed the threshold level of 0.8 (1.2S h
+ S )"
A ii) A change is required in pipe parameters, such as major dif-ferences in pipe size or wall thickness.
iii) Breaks at the new highest stress locations are determined to l-have significantly higher stress values from the original lo-i cations, and result in consequences to safety related systems requiring additional safety protection.
c)
To assure protection of safety-related structures, systems or components, breaks in Duke Class E, F, G and H piping are postulated at the following locations (See Table 3.2.2-3 for class correlations) 1).
The terminal ends of the pressurized portions of the run.
2)
At intermediate locations selected by one of the following methods:
i)
For Class E, F, G, and H Piping:
At each intermediate weld location of potential high stress or fatigue.
3.6-16 Rev. 7
CNS ii) For Class F Piping:
C At all locations where the stress, S, Exceeds 0.8 (1.2 Sh + S )'-
A where:
S
= stresses under the combination of loadings associated with the normal and upset plant condition loadings and
~
an OBE event, as calculated from the sum of equations (9) and (10) in subarticle NC-3600 of the ASME Boiler and Pressure Vessel Code,Section III.
~.g.
S
= basic material allowable stress at maximum (hot)
T h
temperature, per ANSI B31.1.0.
'=
S
= allowable stress range for expansion stresses, per A
ANSI B31.1.0.
3Property "ANSI code" (as page type) with input value "ANSI B31.1.0.</br></br>3" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process.)
For Class F Piping:
A.Jo If there are s.d_mee s intermediate locations where S exceeds 0*8 (1*2 Sh + S ) ' " "SVGe2laFC' JMW'%%y'd)Js/e8.~ - - - --~
" - - - ^' '-- --
'- chncon
-en A
ry t,esed gcr 'igt.c t : tress.
Intermediate breaks are not postulated Ng f
in sections of straight pipe where there are no pipe fittings, flanges, f#fg valves or welded attachments.
4)
As a result of piping reanalysis which follows the completion of C
the original problem interaction analysis, the highest stress locations may be shifted however, the initially determined inter-mediate break locationsfw;ill not be changed unless one of the fol-lowing conditions exists.,
g.,,.ded. -4//9 /,r/<wd/a/c. deak _
i) Maximum stress ranges exceed the threshold level of 0.8 (1.2S h
+ S )"
- j. _
A l
- 11) A change is required in pipe parameters, such as major dif-ferences in pipe size or wall thickness.
iii) Breaks at the new highest stress locations are determined to l
have significantly higher stress values from the original lo-cations, and result in consequences to safety related systems requiring additional safety protection.
3.6.2.1.2.2 Postulated Piping Break Locations For Moderate-Energy Piping Systems Syst' ems identified as containing moderate-energy piping are examined by de-tailed drawing review for postulated through-wall cracks as defined below.
Systems analyzed for consequences of postulated piping cracks are listed in Table 3.6.1-2.
3.6-16a Rev. 7 w --
CNS C
Cracks in Duke Class B, C and F piping are postulated at the following, a)-
i locations:
1)
The'terminalendsofthepressurizedportiodsloftherun.
_?
2).
At intermediate individual locations of potential high stress or
' fatigue (e.g. pipe fittings, valves, flanges and welded attachments) l that result in the maximum effects from fluid spraying, flooding or environmental conditions except in portions of piping where the max-imum stress range is less than 0.4 (1.2 Sh
- S ) as defined in items l
A b)2)ii) and c)2)ii) of Section 3.6.2.1.2.1.
i
~
b)
Cracks in Duke Class E, G and H piping are postulated at the following locations:
f 1)
The terminal ends of the pressurized portions of the run.
2)
At intermediate individual locations of potential high stress or fatigue (e.g. pipe fittings, valves, flanges and welded attachments) that result in the maximum effects from fluid spraying, flooding j
or environmental conditions.
l 3.6.2.1.2.3 Postulated Break Type, Size, and Orientation l
a)
Circumferential Pipe Breaks The following circumferential breaks are postulated in high energy fluid system piping at the locations specified in Section 3.6.2.1.2.1.
1)
'Circumferential breaks are postulated in fluid system piping and branch. runs exceeding a nominal pipe size of 1 inch, except where the maximum stress' range exceeds the limits of Section 3.6.2.1.2.1, items b) and c)2)ii) but the circumferential stress' range is at least 1.5 times the axial stress range.
2)
Where break locations are selected in fittings in accordance with Section 3.6.2.1.2.1 without the benefit of detailed stress calcula-tions, breaks are postulated at each weld, in piping greather than one inch NPS, to the fitting, valve, or welded attachment.
Alter-nately, a single break location at the section of maximum stress range may be selected as determined by detailed stress analyses or tests on a pipe fitting.
3)
Circumferential breaks are assumed to result in pipe severance and separation amounting to at least a one-diameter lateral displacement MEB Q20 of the ruptured piping sections unless physically limited by piping restraints.
No limited break a eas will be used for compartment pressurization calculations.
If limited break areas are'used for jet impingement reviews, the basis will be the installation of rigid rupture restraints; and any such limited break areas along with their l
locations will be documented in the FSAR.
(See Table 3.6.2-5.)
3.6-17 Rev. 12
CNS 4)
.The dynamic force of.the jet discharge at the break location is based
[~
~
on.the effective cross-sectional flow area of the pipe and on a A
calculated fluid pressure as modified by an an_alytically or experi-mentally determined thrust coefficient.
Limited. pipe displacement at the break location, line restrictions, flow limiters,-positive Z
pump-controlled flow, and the 3bsence of energy reservoirs may be taken into account, as applicab.e, in..the reduction of jet discharge.
5)
Postulated pipe whip for target review will be defined by engineering judgement based on piping geometry, jet thrust direction, break lo-E cation analysis type, and hanger location and type.
When further
- 5 confirmation is required, postulated piping breaks and targets are er field reviewed after the drawing based analysis has been complcted.
For the purposes of analysis, breaks are assumed to reach. full open-ing size in one millisecond after break initiation.
b)
Longitudinal Pipe Breaks
-The following longitudinal breaks are postulated in high-energy fluid system piping ? the locations specified in Section 3.6.2.1.2.1.
1)
Longitudinal breaks in. fluid system piping and branch runs are pos-
.tulated in nominal pipe sizes 4 inches and larger, except where the maximum stress range exceeds the limits of Section 3.6.2.1.2.1, items b) and c)2)ii) but the axial stress range is at least 1.5 times the circumferential stress range.
2)
'Longitudinalbreaksare'notpostulatedatt3 89 bTerminalendsprovidedthepipingattheterminalendscontains no longitudinal pipe welds.
b)
^t St:rn di:t: 1:::ti::: th:r: th: ceiteri r. for a ::;inimum-l
-"-^:r ef bred lec: tier: -" t be ::tieft:1 3)
Longitudinal breaks are assumed to result in an axial split without j
pipe severance.
Splits are oriented (but not concurrently) at two l
diametrically-opposed points on the piping circumference such that i
the jet reaction causes out-of plane bending of the piping configu-ration. Alternately, a single split may be assumed at the section of highest tensile stress as determined by detailed stress analysis I
(e.g., finite element analysis).
4)
The dynamic force of the fluid jet discharge is based on a circular or elliptical (20 x 1/2D) break area equal to the effective cross-
. sectional flow area of the pipe at the break location and on a cal-
'culated fluid pressure modified by an analytically or experimentally determined thrust coefficient.
Line restrictions, flow limiters, positive pump-controlled flow, and the absence of energy reservoirs may be taken into account, as applicable, in the reduction of jet discharge.
3.6-18 Rev. 7
CNS 5)
C Piping movement is assumed to occur in the direction of the jet reaction unless limited by structural members, piping restraints, "
or piping stiffness as demonstrated by inelastic limit analysis.
For the purpose of analysis, breaks are assume _d to reach full size one millisecond after break initiation.
c)
Through-Wall Leakage Cracks The following through-wall leakage cracks should be postulated in moderate-energy. fluid system piping at the locations specified in Section 3.6.2.1.2.2 9;-
1)
Cracks are postulated in moderate energy fluid system piping runs
[
exceeding a nominal pipe size of one inch.
2)
Fluid flow from a crack is based on a circular opening of area equal to that of a rectangle one-half pipe diameter in length and one-half pipe wall thickness in width.
3)
The flow from the crack is assumed to result in an environment that wets all unprotected components within the compartment, with conse-quent flooding in the compartment and communicating compartments.
4)
Cracks are not postulated in portions of Duke Class B, C, or F piping where the stresses are less than 0.4 (1.2 Sh
- b ).
Throughwall cracke A
are not postulated inside containment because environmental conse-quences are enveloped by high energy circumferential breaks.
3.6.2.1.3 Failure Consequences The interactions that are evaluated to determine the failure consequences are dependent on the energy level of the contained fluid.
They are as follows:
a)
High-Energy Piping 1)
Circumferential Breaks and Longitudinal Splits a)
Pipe Whip (displacement) b)
Jet Impingement c)
Compartment Pressurization d)
Flooding l
e)
Environmental Effects (Temperature, humidity) 2)
Throughwall leakage cracks a)
Environmental Effects (Temperature, Humidity) b)
Flooding b)
Moderate-Energy Piping 1)
Through-wal'1 leakage cracks a)
Flooding b)
Environmental Effects (Temperature, humidity, water spray) 3.6-19 Rev. 22
CNS 3.6.2.2 Analytical Methods to Define Forcing Functions and Response Models 3.6.2.2.1 Reactor Ccolant System Dynamic Analysis
- This section summarizes the dynamic analysis as it a'phlies' to the LOCA result-
?,
-ing from the postulated design basis pipe breaks in the main reactor coolant l
piping system.
Further discussion of the dynamic analysis methods used to verify the design adequacy of the reactor coolant loop piping, equipment and supports-is given in Reference
_l.
w-The particular arrangement of the Reactor Coolant System for the Catawba Nuclear Station is accurately modeled by the standard layout used in Reference 1 and
' (;
the postulated break locations do not change from those presented in Reference 1.
In addition, an analysis is 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 are based on full pipe area discharge except where limited by major structures.
'The dynamic analysis of the Reactor Coolant System employs displacement method, lumped parameter, stiffness matrix formulation and assumes that all components behavo in a linear elastic manner.
The analysis is performed on integrated analytical models 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 is constructed.
The boundary of the analytical model is, in general, the foundation concrete / support structure interface.
The anticipated deformation of the reinforced concrete foundation supports is considered where applicable to the Reactor Coolant System model.
The mathematical model is shown in Figure 3.6.2-4.
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; c)..The initial deflection, natural frequencies, normal modes, and time-history forcing functions are used to determine the time-history dynamic deflection response of the lumped mass representation of the Reactor Coolant System; 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 e)
The time-history dynamic deflections at mass points are treated as an imposed deflection 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.
3.6-20 Rev. 12 Carryover
CNS The results are used to verify the adequacy of the restraints.
The general dynamic solution process is shown in Figure 3.6.2-5.
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 description 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 T,
transient pressure, mass flow rates, and other hydraulic properties as a 2'
funct, ion of time.
The second step uses the results obtained from the hyd.aulic 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-IV (Reference 3) code was developed with a capability to provide this information.
The SATAN-IV computer code performs a comprehensive space-time dependent anal-ysis of a loss of coolant accident and is designed to treat all phases of the blowdown.
The stages are:
(i) a subcooled stage where the rapidly changing pressure gradients in the subcooled fluid exert an influence upon the Reactor Coolant System internals and support structures; (ii) a two phase depressur-ization stage; and (iii) the saturated stage.
The code employs 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 in each element.
Pump characteristics, pump coastdown and cavitation, and core and steam generator heat transfer including the W-3 DNB correlation in addition to the reactor kinetics are incorporated in the code.
The THRUST computer program was developed to compute the transient (b!owrown) loads resulting from a LOCA.
l 3.6-21 Rev. 12 Carryover
.a CNS
-The blowdown hydraulic loads on primary loop components are computed from the.
(F fluid transient information calculated using the following time dependent forcing function:
- 2-F = 144A'{(P -314.7) + (
.)}
2 pg A 144 c
which includes both the static and dynamic effects.
The symbols and units are:
OJ F a Force, Lb
'=
f A = Aperture area, Ft2 P = System Pressure, psia
-m = Mass flow rate, Lb,/Sec p = Density,' Lb,/Ft3 Lb" x Ft gf=GravitationalConstant=32.174 Lb x Sec 2 7
- 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 represented in a global 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 areas, and projection coefficients along the three axes of the global coordinate system.
Each node 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 force is divided into its x, y, and z components using the projection coefficients.
The force components are then summed over the total number of apertures in any one node to give a total x force, total y force, and total z force.
These thrust forces serve as' input to the piping / restraint dynamic analysis.
Further details are given in Reference 1.
3.6.2.2.2 All Other Mechanical Piping Systems Dynamic Analysis Effects of pipe break are conservatively evaluated to determine the need for pipe whip restraints.
Energy of the whipping pipe, its effect on targets, jet impingement forces and temperatures, compartment pressurization, and temperature effects establish the need for pipe whip restraints.
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 do not prevent mitigation of the break nor prevent the' safe and continued shutdown of the reactor.
3.6-22 Rev. 12 Carryover
4 CNS 3.6.2.2.2.1 Assumptions 1-a)
The thrust load acting on the pipe due to a blowdown jet is equal and opposite to the jet load.
b)
The discharge coefficient is equal to 1.0.
c)
The break opens to its defined size in 1 millisecond.
d) "For the purpose of estimating jet forces, the blowdown shall be to an infinite volume at standard ambient conditions.
e)
The initial fluid condition within the pipe prior to rupture is that for the worst case normal plant operating condition.
f)
The jet profile expansion half-angle is 10 degrees.
3.6.2.2.2.2 Blowdown Thrust Loads The thrust force at any time, T (t) is given by V2 T(t) = (pE E P })
A
,gp A
jE E
where:
fluid density at break at time t pE
=
fluid velocity at break at time t V
=
E pi e break exit area A
=
P jE control volume pressure at break at time t P
=
E ambient pressure P
=
A gra M adon constant
=
gg A simplified analysis may be conducted by assuming that the fluid is blowing down in a steady-state condition with frictionless flow from a reservoir at fixed absolute pressure P (P is the initial line pressure.) When the fluidissubcooled,nonfl3 shin 0 liquid,theflowwillnotbecriticalatthe break area so that P
=P and V = J2g (P - P)P.
If P <<P, the thrust g
A E
A forcemaybeconservktivekyapprbximate[1by i
/
l l
T = 2PdjE 3.6-23
I CNS
\\.
When the fluid is saturated, flashing or super-heated vapor,. the flui i can be
~
assumed to be a perfect gas.
The velocity for critical flow at the break area is given by VE = (Kg P /P )
cE E and K/K-1 P
=P
( 2
)
E K+1 where K = C /C is a ratio of specific heats p y C = specific heat at constant pressure p
C = specific heat at constant volume y
A value of K=1.26 is justified for steam as being conservative.
If P >>P E
A the thrust force may be conservatively approximated by:
T = 1.26 P, AjE 3.6.2.2.2.3 Jet Impingement Loads The loads on an object exposed to the jet from a pipe break can be determined from the blowdown thrust and the profile of the impinged object.
A g Y) = T.
_.Sp.DLF cos &
^j where l
Y) = Normal load applied to a target by the jet A
= Cross-sectional area of jet intercepted by target structure 4
A) = Total cross-sectional area of jet at the target structure S
= Shape factor p
DLF = Dynamic load factor T
= Total blowdown thrust at break as calculated in Section 3.6.2.2.2.2.
= Angle between jet axis and the target.
3.6-24
o C
CNS The ratio A /A) represents the portion of the total infss flow from the jet g
which is intercepted by target structure'.
A dynamic load factor of 2.0 shall be used in the absence of an analysis justifying a lower'value.
3.6.2.3 Dynamic A'1alysis Methods to Verify Integrity and'0perability 3.6.2.3.1 General Criteria for Pipe Whip Evaluation V
1).. The dynamic nature of the piping thrust load shall be considered.
In the r
absence of analytical justification, a dynamic load factor of 2.0 is applied in determining piping system response.
2)
(Elastic perfectly p astic) pipe and crushable material properties may ME8 be considered as applicable.
Consideration for crushable materials is Q27 described in Table 3.6.2-3.
3)
Pipe whip is considered to result in unrestrained notion of the pipe along a path governed by the hinge mechanism and the direction of the thrust force.
A maximum of 180 rotation may take place about any hinge.
4)
The effect of rapid strain rate of material properties is considered.
(
A 10 percent increase in yield strength is used to account for strain rate effects.
3.6.2.3.2 Analysis Methods The pressure time history, jet impingement load on targets, and the thrust resulting from the blowdown of postulated ruptures in piping systems is determined by thermal and hydraulic analyses or conservative simplified analyses.
In general, the loading that may result from a break in piping is determined using either a dynamic blowdown or a conservative static blowdown analysis.
The method for analyzing the interaction effects of a whipping pipe with a restraint will be one of the following:
1)
Equivalent static method 2)
Lumped parameter method 3)
Energy balance method In cases where time history or energy balance method is not used, a conserva-tive static analyses model will be assumed.
The lumped parameter method is carried out by utilizing a lumped mass model.
Lumped mass points are interconnected by springs to take into account inertia and stiffness properties of the system.
A dynamic forcing function or equiva-
/
lent static loads may be applied at each postulated break location with un-(
acceptable pipe whip interactions.
Clearances and inelastic effects are considered in the analyses.
3.6-25 Rev. 7
CNS Ihe energy balance method is based on the principle of conservation of energy."
The kinetic energy of the pipe generated during the firs _t quarter cycle of movement is assumed to be converted into equivalent strain energy, which is distributed to the pipe or the support.
The strain in the restraint is limited to 50 percent of the ultimate uniform strain.
3.6.2.3.3 Pipe Whip Restraint Design When required, restraints are designed to protect essential components from the e.-
e dynamic effects of pipe whip and jet impingement.
The loadings on the restraint
- 3 are determined by one of the methods outlined in Sections 3.6.2.2 and 3.6.2.3.
The d'dsign of these restraints follows the guidelines of AISC (Ref. 4); however, since pipe rupture is associated with the faulted plant condition, higher stress allowables are permitted as identified in Table 3.6.2-3.
Where a restraint is also designed to functi m as a piping support, the discussion in Section MEB 3.9.3.1.5 is applicable.
Rupture loads with a dynamic load factor of 2.0 shall 027 be added to the faulted loads and the support designed for faulted condition per Table 3.9.3-11.
3.6.2.4 Mechanical Penetrations Mechanical penetrations are treated as fabricated piping assemblies meeting the requirements of ASME Section III, Subsections NC and NE and which are assigned the same classification as the piping system that includes the assembly (i.e., Class A through H as defined in Table 3.2.2-3 except that
(
Class C through H lines are upgraded to Class B between Containment isolation valves).
The process line making up the pressure boundary is consistent with the system piping materials, fabrication, inspection, and analysis requirements of ASME Section III, Subsection NC.
Critical high temperature lines and selected engineered safety system and auxiliary lines (regardless of temperature) require the " Hot Penetration" assembly as shown on Figure 3.6.2-8 which features the exterior guard pipe for the purpose of returning any fluid leakage to the Containment and for protection of other penetrations in the building annular space.
Other lines are treated as cold penetrations since a leak into the annular space would not cause a personnel hazard or damage other penetrations in the immediate area.
Penetration assemblies and their anchorages are analyzed in accorcance with Table 3.2.2-3 and applicable response spectra curves (0.5 percent damping) as developed f' rom the method described in Section 3.7.2 and enveloped for con-servatism.
Loading combinations and stress criteria for penetrations are shown in Table 3.6.2-2.
The design of guard pipes considers the simultaneous effects of pressure and jet loadings resulting from a rupture within the guard pipe and the SSE loadings.
3.6.2.4.1 General Design Information for All Mechanical Penetrations The following definitions are utilized to distinguish the categories of mech-anical penetrations.
3.6-26 Rev. 7
CNS a)
Primary System - Reactor Coolant System and any 1-ine connecting to same which penetrates the Containment.
b)
Secondary System - All other piping penetrations and systems within the Reactor Building; this includes the Nuclear Auxiliary Systems.
Design requirements as follow are applicable to piping between the Contain-ment boundary (steel Containment shell or concrete wall, whichever is appli '
~ 3-cabl,e for anchorage) and the crane wall only.
a)
All penetrations are designed to maintain Containment integrity for any loss-of-coolant accident combination of Containment pressures and tempera-tures.
b)
All primary penetrations and all secondary penetrations that would be
. damaged by a primary break are designed to maintain Containment integ-rity.
.c)
All secondary lines whose break could damage a primary line and also break Containment are designed to maintain Containment integrity.
d)
Quality assurance measures for penetration design calculations, criteria,
()-
documentation and procedures are in accordance with the design control requirements of Chapter 'J.
e)
Flued head design is based on the same criteria as the guard pipe design.
Design criteria for bellows expansion joints consider operational differential movements between primary and secondary containment as appropriate.
f)
Mechanical penetration design features for precluding bypass leakage are as follows:
1)
All mechanical penetrations are designed, fabricated, non-destruc-tively examined and erected to the requirements of ASME Section III, Subsections NC and NE.
2)
All mechanical penetrations and their anchorages are analyzed in accordance with the requirements of ASME Section III, Class 2, Subsection NC for pipe whip, and associated loadings to assure containment integrity for any loss of coolant accident.
'3)
All bellows expansion joints are of two ply construction with a wire mesh between plys for testability of bellows and bellows weld to piping.
3.6.2.4.2 Hot Penetrations Typical hot penetration assemblies as shown on Figure 3.6.2-8 consist of three major components; a) process line and flued head, b) guard pipe, and c) expan-sion joint Containment seal.
3.6-27
CNS Design requirements for hot penetrations are as follow:
I.
a)
The guard pipe and bellows assembly constitute an extension of the Containment and as such meet Containment de. sign conditions.
b)
A guard pipe is required for lines that can overpressurize the annulus and/or release unacceptable amounts of radioactivity to the atomosphere.
y.-
c)
Guard pipe contains and returns any process line leakage back to the Cohtainment.
d)
Bellows design accommodates both axial and lateral displacements between the Containment and Reactor Building for thermal, seismic, and Contain-ment test conditions.
e)
The guard pipe and process line are anchored and guided so as to act as a single unit under thermal, seismic, and pipe rupture loads, f)
Stress levels for process lines meet requirements of Section 3.9.3.
g)
Stress levels for guard pipes and other penetration structural components meet the requirements of Section 3.9.3.
h)
Exterior bellows cover and impingement plate protects the bellows l
assembly from foreign objects during construction and station operation.
i)
The process pipe is designed to meet the requirement of Table 3.9.3-8 for stress levels and applicable loading combinations.
The process pipe is of seamless construction made from SA376 GR304 or GR316 stainless steel, except for Main Steam and Main Feedwater penetrations which are A106 GRB.
(See Figure 3.6.2-7)
Design codes applicable to hot penetrations are as described below, a)
Penetration boundaries are in accordance with ASME III, Subsection NE, Paragraph NE-1100.
Process lines including flued head, guard pipe, and bellows assemblies including dished heads, are designed, fabricated, and inspected to ASME III, Subsection NE, with the allowable stresses as de-fined above.
The guard pipe wall thickness design complies with the re-quirements of NE-3324.3a of the ASME Code when using the design pressure and temperature of the enclosed process pipe.
b)
The Reactor Building anchor section is considered a structural component.
Attachment welds to the guard pipe meet and are inspected to ASME Section III, Subsection NE.
Field welds between the guard pipe attachment and Re-actor Building anchor section are structural welds.
Field welds between the bellows and Containment meet and are inspected to ASME Section III, Subsection NE.
/
3.6-28
3.6.2.4.3 Residual Heat Removal Recirculation Line Penetration Residual heat removal recirculation line penetrations.-are of the cold-penetration type.
(See Figure 3.6.2-6)
Design requirements for these penetrations are as follow:
a) The recirculation line is an extension of Containment up through the first valve.
<y-b) These valves are Safety Class 2 and are conservatively designed (600 ps'ig
~ 9 design pressure) to withstand the containment design pressure of 15 psig.
c) Valves are located in an accessible area for maintenance during the post-accident period.
d) Expansion joints are utilized in the penetration design.
The stress analysis for the two recirculation lines between containment and their sump isolation valves shows that the stress in these sections is below values that would require postulation of breaks occurring if this pipe were normally in use (i.e., stress does not exceed 0.4 (1.2SH + S ))-
A For this reason it is acceptable not to provide guard pipe for these sections of piping (see SRP 6.2.4 Acceptance Criteria 6.e.).
Any postulated leakage from the valve bonnet via the valve stem is contained via a leakoff
(.
that is directed to the Recycle Holdup Tank (the RHT is equipped with a diaphragm that would contain any gases released from solution). The valves themselves are designed to withstand 600 psig at 400*F which is well in excess of values to which they would be exposed after an accident (approximately 36 psig, 190 F).
3.6.2.4.4 Access for Periodic Examination A description of the method of providing access to permit periodic examinations of process pipe welds within the protective assembly as required by the plant inservice inspection program is discussed in Section 6.6.
3.6.2.5 Summary of Dynamic Analyses Results MEB A summary of postulated circumferential and longitudinal break locations is Q103 shown on Figures 3.6.2-9 to 3.6.2-207.
MEB A summary of the dynamic analyses, resulting from postulated pipe breaks in Q108 high-energy piping systems, is comprised of the following information:
a) System pipe routing - Figures 3.6.2-9 thru 3.6.2-207 b) Location of postulated breaks - Figures 3.6.2-9 thru 3.6.2-207 c) Location of postulated pipe rupture restraints - Figures 3.6.2-9 thru 3.6.2-207 (Jet barriers are located near target to intercept jet) d) Summary of protection requirements - Table 3.6.2-4.
e) Summary of combined stresses at break locations - Table 3.6.2-4.
3.6-29 Rev. 11
s CNS REFERENCES FOR SECTION 3.6 1 -.
.~
1.
" Pipe Breaks for the Loca Analysis of the Westinghous~e Primary Coolant Loop", WCAP-8082-P-A, January,1975 (Proprietary) and WCAP-8172-A (Non-Proprietary), January,1975.
~
2.
" Documentation of Selected Westinghouse Structrual Analysis Computer
?I" Codes", WCAP-8252, Revision 1, May, 1977.
'n'
=
- 3. **Bordelon, F.M., "A Comprehensive Space-Time Dependent Analysis of Loss of Coolant (SATAN IV Digital Code)", WCAP-7263, August,1971 (Proprietary) and WCAP-7750, August, 1971 (Non-Proprietary).
4.
American Institute for Steel Construction, " Specifications for the Design, Fabrication, and Erection of Structural Steel for Buildings",
Februrary 12, 1969.
(
3.6-30
Table 3.6.1-1 High-Energy Mechanical Piping Systems Analyzed for Consequences of Postulated Piping-Breaks
~
System ~
Pipe Break High-Energy Piping System
- Identification Protection Method Steam Generator Blowdown Recycle System BB (a) (b) i*f Auxiliary Feedwater System CA (a) (b)
(Motor Driven Pump Portion) i Main Feedwater System CF (a) (b) 4 Reactor Coolant System NC (a) (b)
Safety Injection System NI (a) (b) (c)
Boron Thermal Regeneration System NR (a) (b)
Chemical and Volume Control System NV (a) (b)
(Letdown Portion and Sealwater Injection)
Main Steam Supply to Auxiliary Equipment SA (a)
(b)
System Main Steam System SM (a)
(b) (c)
Main Steam Vent to Atmosphere SV (a) (b)
System Boron Recycle System NB (a)
Liquid Radwaste System WL (a)
Solid Radwaste System WS (a)
Pipe Break Protection Methods Legend:
(a) Physical Separation (b) Piping Restraints (c; Enclosures, structural, guard pipes, etc., (designed specifically for pipe break)
Note:
High-Energy Systems may contain moderate-energy portions; however, for brevity, systems only are listed in this table.
Rev 12
Table 3.6.1-2 (Page 1)
Moderate-Energy Mechanical Piping Systems Analyzed for Consequences of Postulated Piping Breaks System Pipe Break System Identification Protection Method Moderate-Energy Safety Related Systems Auxiliary Feedwater System CA (a)
(b)
(Turbine Driven Portion)
Diesel Fuel Oil System FD (a)
(b)
Refueling Water System FW (a)
(b)
Component Cooling System KC (a)
(b)
Diesel Generator Cooling Water System KD (a)
(b)
Spent Fuel Cooling System KF (a)
(b)
Diesel Generator Lube Oil System LD (a)
(b)
Boron Recycle System NB (a)
(b)
Residual Heat Removal System ND (a)
(b)
Containment Spray System NS (a)
(b)
Nuclear Service Water System RN (a)
(b)
Main Steam Supply to Aux. Equipment SA (a)
(b)
FWP Turbine Exhaust TE (a)
(b)
Waste Gas System WG (a)
(b)
Liquid Radwaste System WL (a)
(b)
Solid Radwaste System WS (a)
(b)
Filtered Water System YF (a)
(b)
Auxiliary Bldg Cooling Water System YN (a)
(b)
Control Area Chilled Water System YC (a)
(b)
Rev 12
-,--m.,
Table 3.6.1-2 (Page 2)
(
Moderate-Energv Mechanical Piping Systems Analyzed for Consequences of Postulated Piping Breaks System Pipe Break System Identification Protection Method Other Moderate-Energy Systems Auxiliary Steam System AS (a)
(b)
{
hiecirculatedCoolingWaterSystem KR (a)
(b)
Ice Condenser Refrigeration System NF (a)
(b)
Fire Protection System RF (a)
(b)
Equipment Decontamination System WE (a)
(b)
Chemical Addition System YA (a)
(b)
Plant Heating System YH (a)
(b)
Make-up Demineralizer System YM (a)
(b)
Pipe Break Protection Methods Legends:
(a) Physical Separation (b) Piping Spray Shields (c) Enclosures, structural, guard pipes, etc., (designed specifically for pipe break) l 4
Rev 12
]
O Table 3.6.1-3 (Page 1)
Comparison of Duke Pipe Rupture Criteria And NRC Requirements of Branch Technical Positions APCSB 3-1 (November 1975), MEB 3-1 (November 1975), and NRC Regulatory Guide 1.46 (May 1973)
NRC Criteria Duke Criteria APCSB 3-1, Section B.2.c SAR Section 3.6.2 Section B.2.c. requires that piping between containment Duke criteria is generally equivalent to NRC isolation valves be provided with pipe whip restraints criteria as clarified below:
capable of resisting bending and torsional moments pro-duced by a postulated failure either upstream or down-The containment structural integrity is provided stream of the valves.
Also, the restraints should be for all postulated pipe ruptures.
In addition, designed to withstand the loadings from postulated for any postulated rupture classified as a loss failures so that neither isolation valve operability of coolant accident, the design leaktightness of nor the leaktight integrity of the containment will the containment fission product barrier will be be impaired.
maintained.
Terminal ends should bo considered to originate at a Penetration design is discussed in SAR Section point adjacent to the required pipe whip restraints.
3.6.2.4.
This section also discussed penetra-tion guard pipe design criteria.
Terminal ends are defined as piping originating at structure or component that act as rigid con-straint to the piping thermal expansion.
APCSB 3-1, Section B.2.d SAR Section 6.6 il (1) The protective measures, structures, and guard Duke criteria is different than the NRC criteria pipes should not prevent the access required to due to the code effective date as described below:
conduct inservice inspection examination.
ASME Class 2 piping welds will be inspected in (2) For portions of piping between containment isola-accordance with requirements given in 5AR Section tion valves, the extent of inservice examinations 6.6.
completed during each inspection interval should provide 100 percent volumetric examination of circumferential and longitudinal pipe welds.
i n.1
~
4
b Table 3.6.1-3 (Page 2)
(3) Inspection ports should be provided in guard pipes to permit the required examination of circumferen-tial welds.
Inspection ports should not be loca-ted in that portion of guard pipe passing through the annulus.
(4) The areas subject to examination should be defined in accordance with Examination Categories C-F and C-G for Class 2 piping welds in Tables IWC-2520.
APCSB 3-1, Appendix A SAR Section 3.6.1.1.2 1
High Energy fluid systems are defined as those systems Duke criteria is the same as NRC criteria with that, during normal plant conditions, are either in expansion of definition as clarified below:
operation or maintained pressurized under' conditions where either or both of the following are met:
a.
Non-liquid systems with a maximum normal pressure less than 275 psig are not consi-a.
maximum operating temperature exceeds 200 F, or dered high energy regardless of the temper-ature.
Such low pressure system (i.e.,
b.
maximum operating pressure exceeds 275 psig.
Auxiliary Steam, 50 psig, 298 F) do not contain sufficient sensible energy to deve-lop sudden, catastrophic failures.
Propa-l gation of a crack to a full failur is l
extremely unlikely.
l b.
Exception to the 200 F threshold for high l
energy systems is taken for non-water sys-l tems such as ethylene glycol.
Such systems that operate at less than their boiling temperature are considered moderate energy.
APCSB 3-1, Appendix A SAR Section 3.6.2.1.2.1 In piping runs which are maintained pressurized dur-Duke criteria is different from NRC criteria as ing normal plant conditions for only a portion of described and justified below:
4 the run (i.e., up to the first normally shut valve) a terminal end of such runs is the piping connection Terminal ends are considered,at' pip'ing origina-d !,
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Table 3.6.1-3 (Page 3)
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to this closed valve, ting at structure or components' that act as rigid constraint to the piping thermal expan-sion.
Typically, the anchors assumed for the code stress analysis would be terminal ends.
Stresses in the system either side of the I
closed valve will be about the same;-therefore, terminal end classification based on constraint and hi0h stresses are not applicable.
Duke neviews these closed valve locations to assure high stresses are not developed as a result of 1
rigid constraint from near by anchors of com-ponent connections in the non pressurized portion of the piping.
APCSB 3-1, Appendix B and C SAR Section 3.6.2.1.2.1 1
In Appendix B, pipe break locations are specified for A5i1E Section 111 Code Clas~s,1, 2, and 3 piping Duke criteria specifies that if the threshold such that a minimum of two intermediate breaks are stress levels are not exceeded, then no inter-
)
selected per run although threshold limits are not mediate breaks are postulated.
i exceeded (for A5f1E Section III Code Class 1, 2, and' 3 piping).
In Appendix C, a minimum of either two or one intermediate breaks within the boundary of each compartment is specified.
t
}
11EB 3-1, Section B.1.h(6)
SAR Section 3.6.2.4 Section B.I.b(6) requires that guard pipe assemblies between containment isolation valves meet the follow-Duke criteria is different from NRC criteria as
~
j i ing requirements:
described and justified below:
i The design pressure and temperature should not Guard pipes provided between containment isola-a.
tion valves are designed in accordance with SAR j'
be less than the maximum operating temperature Section 3.6.2.4.
Guard pipes are subjected to
,y and pressure of the enclosed pipe under normal l*,
plant conditions, a pressure test as required by the material
!i specification before welding to the penetration assembly.
ql-b.
The design stress limits of Paragraph NE-3131(c) should not be exceeded under the loading asso-It is impractical to test guard pipes in the ciated with design pressure and temperature in finished penetration assembly due to the con-combination with the safe shutdown earthquakes.
figuration and potential damage to internal process pipe and associated insulation.
I ntle-i1 1 k.
Rev. 9
a.
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,1' Table 3.6.1-3.(Page 4) l
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Guard pipe assemblies should be subjected to a
- yy c.
singic pressure test at a pressure equal to de-pendent design analysis have been conducted to sign pressure.
provide assurance that Duke penetration designs are acceptable.
In_ addition, the extent of flDI conducted on guard pipes to flued head' butt weld is such to assure integrity of design.
IIEB 3-1, Section B.I.c(1)
SAR Section 3.6.2.1.2.'1.
Intermediate breaks in Class I piping are postulated -
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 U exceeds 0.1, no intermediate breaks are" l
(10), (12), and (13) or U)0.1.
postulated.
i
!!EB 3-1, Section 8.1.c(2) i SAR Section 3.6.2.1.2.1 Intermediate breaks in Class 2 and 3 piping are l
postulated where the stresses exceed 0.B (1.2S Duke criteria specifies that if the threshold S ) but at not less than two locations based onh+
stress levels are not exceeded, then no inter-g highest stress. Where the piping consists of a mediate breaks are postulated.
i straight run without fittings, welded attachments, i
and valves, and all stresses are less than 0.8 1
(1.2Sh ' S ), a minimum of one location should be A
chosen based on highest stress.
liEB 3-1, Sections B.I.c(3)
SAR Section 3.6.2.1.2.1
\\
Breaks in non-nuclear piping should be postulated at the following location:
Duke criteria is generally equivalent to 11RC cri-O teria as described and justified below:
a.
Terminal ends, Breaks in Duke Class F piping (non-nuclear,
..[
b.
.At each intermediate pipe fitting, welded seismic) are postulated at terminal ends and at g
attachment, and valve.
intermediate locations based on the use of ASt!E y
Section III analysis techniques, the same as (I
Duke Class B and C piping.
Duke Class f piping i
is constructed in accordance with ANSI B31.1 and f
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at terminal ends where the piping has no longi-tudinal welds.
9 Duke criteria specifies that if the threshold stress levels are not exceeded, then no inter-mediate breaks are postulated.
Pggulatory G_uide 1.46 SAR Section 3.6.2.1.2 A whipping pipe s!.culd be considered capable of Duke criteria is the same as flRC Branch Techni-rupturing an impacted pipe of smaller nominal pipe size arn! lighter wall thickness.
cal Position APCSB 3-1 and roughly equivalent to Regulatory Guide 1.46 with expansion of def-inition as described below:
The energy associated with a whipping pipe is considered capable of (a) rupturing impacted pipes of smaller nominal pipe sizes, and (b) developing through-wall cracks in larger nominal pipe sizes with thinner wall thicknesses.
O e
Rev. 9
o l
t ATTACHMENT E
[
POSITIVE BENEFITS FROM ELIMINATION OF ARBITRARY INTERMEDIATE BREAKS AT CATAWBA NUCLEAR STATION This information along with Attachments A-D provides sufficient reason and justification to demonstrate 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 arbi-trary intermediate pipe breaks would increase the visibility and detectability of piping related problems (e.g., inadvertent restraint) during routine mainte-nance activities, as well as provide more access for ISI activities. Occupa-tional 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 elimi-nated arbitrary intermediate breaks.
Since there is a certain amount of ductility in any of the stainless steel or carbon steel piping associated with the piping systems which contain these arbitrary intermediate breaks, any piping failure is expected to be a leak-before-break phenomena.
The piping materials are not types where 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 Catawba Nuclear Station Unit 1 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 concludt 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.
MN30130D/20
r a
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 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 support of the overall safety for Catawba Nuclear Station for the elimination of arbitrary intermediate breaks.
l I
i MN301300/21 l
-