ML16314B082

Criteria for Protection Against Dynamic Effects Associated with a Postulated Rupture of Piping

ML16314B082
Person / Time
Site:	Diablo Canyon
Issue date:	11/09/2016
From:	Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
Download: ML16314B082 (14)

Text

3.6 CRITERIA FOR PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH A POSTULATED RUPTURE OF PIPING Special measures have been taken in the design and construction of the plant to protect the public against the consequences of major mechanical accidents, including a design basis loss-of-coolant accident (LOCA). The containment and all essential equipment within or outside of the containment, particularly components of the reactor coolant pressure boundary and other safety related components, have been protected against the effects of blowdown get, reactive forces, and pipe whip resulting from postulated rupture of piping inside and outside of containment. Analysis is currently underway to analyze the effects of the rupture of any high energy pipe outside the containment. The criteria for locating and minimizing those effects are described in this Section 3.6.

S stems in Which Desi n Basis Pi in Breaks are Postulated to Occur Inside the Containment The following systems inside the containment have been evaluated, with regard to the dynamic effects associated with a ruptured pipe:

1. Reactor Coolant System a) Primary Coolant Loop b) Pressurizer Surge Line c) Pressurizer Spray Line d) Pressurizer Relief and Safety Valve Lines e) Remote Temperature Detector Bypass Lines f) Drains

2. Chemical and Volume Control System a) Charging Line and Auxiliary Spray Line b) Reactor Coolant Pump Seal Water Injection c) Letdown Line d) Excess Letdown Line e) Reactor Coolant Pump No. 1 Seal Vent and Leakoff 3.6-1

3. Safety Injection System a) Accumulator Injection Lines b) Safety Injection Lines

4. Residual Heat Removal System a) Residual Heat Removal Supply b) Residual Heat Removal Return (Accumulator Injection)

Outside the Containment Evaluation is currently in progress at this time to evaluate the effects of a broken pipe carrying high energy fluid outside the containment. All systems 0

having a temperature greater than 200 F or a pressure greater than 275 psig during plant operational conditions are considered. Crack breaks are postu-lated at all locations in pip'ing having fluid temperature or pressure above these levels, and an evaluation is made to determine effects of blowdown jet forces. Design basis breaks, in addition to crack breaks, are postulated in the high energy systems, those systems in which both temperature and pressure exceed these levels. The criteria for determining the location of design basis breaks in defined on the following pages. At each of these locations an evaluation is made to determine the effects of-blowdown jet and reactive forces, and pipe whip resulting from a postulated design basis break.

Piping either encased in concrete or protected from safety and shutdown systems by barriers is not considered. Piping physically remote from safety and shutdown systems such that unrestrained 'motion (pipe whip) in any direction about a plastic hinge formed after a pipe rupture could not impact any safety or shutdown system is not considered.

The system outside the containment for which pipe breaks are assumed to occur are as follows:

3.6-2

Condensate System Feedwater System Turbine Steam Supply System Extraction Steam and Heater Drip System Chemical and Volume Control System Safety Injection System Residual Heat Removal System Turbine and Generator Associated Systems Auxiliary Steam System Other systems outside the containment which carry high pressure or high temperature fluid have been determined to be physically remote from any systems or structures required to function following a design basis accident and failure of those lines themselves would not cause an unsafe operating condition. These lines will not be further evaluated for piping rupture.

Pi in Break Criteria The criteria used to postulate pipe breaks differ depending on whether the pipe break directly involves a loss of reactor coolant accident and whether the pipe I.

postulated to break is inside or outside the containment.

Break Resultin in Loss of Coolant A loss of reactor coolant accident is assumed to occur for a pipe break down to the restraint of the second normally open automatic isolation valve on outgoing lines and down to and including the second check valve on incoming lines normally with flow. A pipe break beyond the restraint or second check valve if will not result in an uncontrolled loss of reactor coolant either of the two valves in the line close. To insure integrity and design adequacy of the primary reactor coolant loop piping and equipment supports system in the event of a highly improbable pipe rupture accident, a number of pipe rupture break locations- are postulated. The primary reactor coolant loop has been analyzed for the design pipe breaks shown in the following table and Figure 3.6.4.

These discrete break locations and types were determiend by an engineering 3.6-3

approach which employs, as its basis, stress and fatigue analyses, system considerations, operational characteristics, and loading conditions. The dynamic analyses of the primary reactor coolant loop piping and equipment supports system for each of the break locations in the table and Figure 3.6-4 assure that 'public health and safety will be adequately protected.

Location and e of Postulated Primar Coolant Loo Failure

1. Straight portion of hot leg piping guillotine

2. Straight portion of cold leg piping - guillotine

3. Steam generator inlet nozzle guillotine

4. Steam generator outlet nozzle guillotine

5. Reactor coolant pump inlet nozzle guillotine

6. 50 elbow - split"

7. Flow entrance to .the 90 0 elbow guillotine

8. RHR primary loop connection - guillotine

9. Safety injection/primary coolant loop connection - guillotine

10. Pressurizer surge/primary coolant loop connection guillotine ll. Loop closure weld in cross'over leg guillotine The break area for both guillotine and longitudinal breaks can be assumed to be less than the cross sectional area of the pipe when analytically or experimen-tally substantiated. In the absence of this data, the break area is the cross sectional area of the pipe. The break length for the longitudinal breaks is considered to be equal to two.pipe diameters. For the breaks listed in the table and Figure 3.6-4, the break area was conservatively assumed to be the cross, sectional area of the pipe.

Other Pi e Breaks Inside the Containment Piping other than the reactor',coolant loop inside the containment is postulated to rupture and the effect such a rupture would have on the safe shutdown of the reactor has been evaluated. The pipe breaks can be assumed to occur anywhere in the reactor auxiliaries'iping and the piping design criteria is to prevent propagation of a rupture pipe to the extent the loss of coolant would be greater than that postulated for a double ended severance of a main coolant 3.6-4

loop. The analysis made to determine if a break in a line inside the containement could eventually result in a loss of coolant exceeding that predicted for a double ended severance of a main loop is described later in this section. Pipe break sizes are assumed to be equal to the complete severance of the pipe being evaluated.

Pi e Breaks Outside the Containment f

The criteria for postulating pipe breaks outside the containment is based on results of the piping stress analysis. These analyses consider effects of pressure, deadweight, thermal expansion during normal operating, upset and test conditions and the Design Earthquake (DE).

Design basis breaks in straight or curved pipe four inches in diameter or greater are assumed to be longitudinal or circumferential, with the break area equal to the flow area of the pipe. Longitudinal breaks may have any orient-ation on the circumference of the pipe. Design basis breaks in pipe one to four inches in diameter are assumed to be circumferential only, with break area equal to flow area of t'e pipe. Design basis breaks at branch points are assumed to be circumferential in branch lines and longitudinal in run lines, with break area equal to flow area of the branch. The criteria for selection of design basis break locations in each piping run are as follows:

A. Postulate breaks at all terminal points (anchors or rigid equipment) .

B. Postulate breaks at all branch points (terminal).

C. Postulate intermediate breaks between terminal points whenever the~1 expansion stresses exceed 80% of SA where SA is as defined in ANSI B31.1 1967.

D. Postulate intermediate breaks between terminal points whenever primary stresses (pressure, weight, DE) plus thermal expansion stress exceeds 80% of (Sh + SA), where Sh and SA are defined in ANSI B31.1 - 1967.

The summation of stresses (pressure + weight + DE + thermal) is referred to as "combined stress." For those portions of piping where DE stress analyses are not available, breaks are postulated at locations where pipe break would yield most mevere consequences.

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E. As a minimum, two intermediate breaks are selected at locations of highest stress.

F. Additional intermediate breaks are selected at locations where pipe break would yield severe consequences and stress approaches that at highest stress location selected above.

Crack breaks are assumed to have a flow area equal to one half the pipe diameter times one half the pipe thickness and are postulated in the most adverse orientations and locations throughout'he piping.

Desi n Loadin Combinations Reactor Coolant Loo The primary coolant loop/supports system is dy'namically analyzed by the time history method for each of the above break locations to determine component and component support loadings (see Section 5.5.14). These design basis loadings (LOCA) are combined with the design basis earthquake and normal condition primary loadings and the resulting stresses are held within the faulted condition allowable stress limits (See Section 5.2).

Other Pi in Inside and Outside the Containment For pipe restraints dead load, live load, earthquake forces, and loads associated with accidental pipe rupture are combined as follows:

1.0 Y = D + L+ TA + RA + YR + DDE Where D = dead load L = live load T = thermal load on rupture restraint generated by a postulated accident

= pipe reaction on rupture restraint from unbroken RA h pipe generated by postulated pipe break conditions Y = reaction on rupture restraint from -broken pipe generated by postulated pipe break, including an appropriate dynamic load factor DDE = load from double design earthquake 3.6-6

Thermal loads are neglected when it can be shown that they are secondary and self-limiting in nature and the material is ductile. Due to the high rate of strain that a pipe restraint would experience in the event of a pipe break, and partly due to the strain hardening effects, the static yield strength of the material used is increased by 15 percent.

In general, strains of up to 50 percent of ultimate strain are acceptable, provided there is no loss of function. Where buckling is critical in com-pression members, the load on the members 1s limited to 90 percent of the buckling load.

Jet forces on structures are calculated as F =C (12PA) p where F ~ get force acting on a structure C = factor to account for the dynamic nature of the load.

In determining the value of C , inelastic behavior is assumed.

Protective Measures Where necessary to satisfy the criteria presented in this Section (3.6), pipe restraints are provided. The location of p1pe restraints in the Unit 1 Containment Structure are shown in Figure 3.6-2. Restraints in the Unit 2 W

Containment Structure are the same. A typical pipe restraint is shown in Figure 3. 6-3.

Pi in Inside the Containment The piping layout was planned such that whipping of two free sect1ons cannot reach equipment or other pipes for which protection is required. In planning the piping layout and arrangement of restraints, barriers are utilized, where available, to prevent the whipping pipe from impacting on equipment or piping requiring protection. For example, the crane wall, operating floor, and n

3.6-7 II ~

~

1

refueling cavity walls serve as barriers between the reactor coolant loops s and the containment liner. Except for the Emergency Core Cooling System lines, which must circulate cooling water to the vessel, the engineered safety features are located outside of the crane wall. The Emergency Core Cooling System lines that penetrate the crane wall are routed outside of the crane wall so as to penetrate it in the vicinity of the loop to which they are attached. Also, the results of tests and/or analyses are utilized to demon-strate that the whipping pipe will not cause damage in excess of acceptable limits.

Engineered safety features are provided for core cooling and boration, pressure reduction, and activity confinement in the event of a loss of reactor coolant or steam or feedwater line break accident to assure that the public is protected in accordance with 10 CFR 100 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 severance of a Reactor Coolant System main loop.

In order to assure the continued integrity of the vital components and the engineered safety systems, consideration is given to the consequential effects of the pipe break itself to the extent that:

1. The minimum performance capabilities of the engineered safety systems are not reduced below that required to protect against the postulated break; k

2. The containment leaktightness is not decreased below the design value, if the break leads to a loss of reactor coolant;* and

• The containment is here defined as the containment structure liner and penetrations, and the steam generator shell, the steam generator steam side instrumentation connections, 'the steam feedwater, blowdown and steam generator drain pipes within the containment structure.

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

a. A pipe break which is not a loss of reactor coolant will not cause a loss of reactor coolant or steam or feedwater line break.

b. A Reactor Coolant System pipe break will not cause a steam-feedwater system pipe break and vice versa.

4. Branch lines connected to the Reactor Coolant System are defined as "large" if they have an inside diameter greater than 4 inches up to the largest connecting line, which is the pressurizer surge line. Rupture of these lines results in a rapid blowdown from the Reactor Coolant System, and protection is basically provided by the accumulators and the low head safety injection pumps (residual heat removal pumps).

In addition to the above (1 through 3), large piping is restrained so that:

a. Propagation of the break to the unaffected loops is prevented to assure the delivery capacity of the accumulators and low head pumps.

b. Propagation of the break in the affected loop is permitted to occur but will not exceed 20 percent of the area of the line which initially ruptured. This criterion has been voluntarily applied so as not to

,, substantially increase the severity of the loss of coolant.

'I

5. Branch lines connected to the Reactor Coolant System are defined as "small" if they have an inside diameter equal to or less than 4 inches.

This size has been based on Emergency Core Cooling System analyses which show that no cladding damage is expected up to a break area of 12.5 square inches corresponding to 4 inches inside diameter. (See Section 15.3.)

In reviewing the mechanical aspects of these lines, it has been demonstrated by Westinghouse Nuclear Energy System tests that lines hitting equal or larger size lines of same schedule will not cause failure of the line being hit; e.g., a 1 inch line, should it fail, will 3.6-9

not cause subsequent failure of a 1 inch or larger size line. The reverse, however, is assumed to be probable; i.e., a 4 inch line, should it fail and whip as a result of the fluid discharged through the line, could break smaller size lines such as neighboring 3 inch or 2 inch lines.

In this case, the total break area shall be less than 12.5'quare inches.

In the unlikely event that one of the small pressurized lines should fail and result in a loss of coolant accident, the piping is restrained or arranged to meet the following requirement in addition to (1 through 3) above.

a. Break propagation must be limited to the affected leg; i.e.,

propagation to the other leg of the affected loop and to the other loops shall be prevented.

b. Propagation of the break in the affected leg is permitted but must be limited to a total break area of 12.5 square inches (4 inch inside diameter) . The exception to this case is when the initiating small break is the high head safety injection line. Further propagation is not permitted for this case.

c. Damage to the high head safety injection lines connected to the other leg of the affected loop or to the other loops must be prevented.

d. Propagation of the break to high head safety injection line connected to the affected leg must be prevented if the line break results in a loss of core cooling capability due to a spilling injection line.

As stated above, a small pipe break in one of the legs of a given loop must not cause a break in the opposite leg of the same loop or in other loops. A rupture of the resistance thermocouple device (RTD) bypass loop would consti-tute an exception to this criterion. If a break occurs at any location in the RTD bypass, however, the resulting blowdown, because of the relatively high flow resistance in the line, is less severe than from a rupture directly adja-cent to the reactor coolant pipe. Safety injection capacity for this case is provided by the high head connections.

3.6-10

Where necessary to satisfy these five requirements, restraints are provided with the proper arrangement and spacing to prevent a plastic hinge mechanism (unrestrained rotation) from forming as a res'ult of the forces associated with a pipe rupture (See Figure 3.6-2). The thrust forces associated with a pipe rupture are calculated as described previously.

The moment required to form a plastic hinge is calculated as

~KS I R0 where Mp = plastic moment K = Mp/My where My is the moment which produces yield stress on the extreme fiber Sy = yield stress of pipe material I = moment of inertia of pipe 'cross section = (R "

'7T 4 0 R

1 )

R 0

= outside radius of pipe R

1

= inside radius of pipe Whipping in bending of a broken stainless steel pipe secti'on does not cause this section to become a missile. This design basis has been demonstrated by performing bending tests on large and small diameter, heavy and thin walled stainless steel pipes.

The fluid discharged from the'uptured piping will produce reaction and thrust forces in -the piping systems. The effects of these loadings are considered in assuring the continued integrity of the vital components and the engineered safety features.

Pi in Outside the Containment Piping outside the containment is currently being analyzed to determine the dynamic effects of ruptured pipes. Rupture restraints will be added to existing piping where it is necessary to prevent the reactive forces of a broken pipe from damaging a safety related component. Plastic hinge formation it will be prevented when is demonstrated that such formation could cause damage which would endanger the safe shutdown of the reactor. Barriers will 3.6-11

be provided if necessary to protect or isolate safety related equipment. The extent of flooding that is a result of a broken pipe will be determined and design changes will be made if analysis shows safety related equipment could malfunction as a result of flooding or adverse environmental conditions.

3.6-12

S.G. RV RCP I

I 10 Qv I

II I RV = REACTOR VESSEL UNlTS l AND 2 S G = STEAM GENERATOR DIABLO CANYON SITE RCP=REACTOR COOLANT PUMP FIGURE 3.6-4 PRIMARY COOLANT LOOP BREAKS