ML20082D422

Amend to Application for OL Re leak-before-break Concept & Elimination of Arbitrary Intermediate Breaks.Revised FSAR Pages Encl.Decision on Proposal Requested by 840120

ML20082D422
Person / Time
Site:	Catawba
Issue date:	11/18/1983
From:	Tucker H DUKE POWER CO.
To:	Adensam E, Harold Denton Office of Nuclear Reactor Regulation
References
NUDOCS 8311220421
Download: ML20082D422 (27)

Text

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DUKE POWER GOMPANY r.o. nox aanse CHARLOTTE. N.C. 28242 soa wt (704 3 831

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November 18, 1983 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 Re: Catawba Nuclear Station Docket Nos. 50-413 and 50-414

Dear Mr. Denton:

In response to your letter to W. H. Owen dated October 17, 1983, Duke Power is extremely interested in pursuing the timely application of the leak-before-break concept and elimination of arbitrary intermediate breaks. We understand from your letter that applications will be pennitted prior to the NRC completing all of the changes in regulatory requirements and, hence, this submittal.

As stated in the previous letter from W. H. Owen to W. J. Dircks dated .

September 19, 1983, 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. We estimate a total reduction of design, material, and erection costs at Catawba Unit 2 to be in excess of $7 million.

In order to achieve the above advantages at Catawba, Duke Power Company will make several specific applications to the NRC for review and approval. .This letter contains the first such application, and it is anticipated -the remaining submittals will occur through mid-1984. Future requests are expected to include application of the " leak-before-break" concept to the Reactor Coolant loops,

Pressurizer Surge Line,10-inch Accumulator Injection Lines, and the 12-inch Residual Heat Removal Lines. We request a decision on each proposal within two months after submittal in order to allow realization of design'and construc-tion savings on Catawba Unit 2.

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 intennediate 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 while improving overall ~ safety and piping system reliability. A summary of the potential benefits which can be realized specifically from the elimination of arbitrary intermediate breaks 1

I 8311220421 831118 i PDR ADOCK 05000413 A PDR I

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Mr. Harold R. Denton, Director November 18,-1983 Page . 2 --

. for Catawba Nuclear Station Unit 2 is provided in Attachment A. Technical justification for.this action is provided in Attachment B.

For: Catawba Nuclear Station, Duke Power-Company requests NRC approval for the a loop)pplication as follows: of alternative pipe break criteria (excluding the RCS primary

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 3.6.2 are not exceeded.

b. For Class 1. piping systems, the usage factors in Catawba

-FSAR Section 3.6.2 are not exceeded.

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

. 3. . 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 off the currently postulated arbitrary intemediate pipe breaks to be eliminated for Catawba Nuclear Station, Unit 2

.is.provided in Attachment C.

Attachment D is the Catawba FSAR revision associated ~with the elimination of arbitrary intermediate. breaks. The current request for criteria. change is for implementation on-Unit 2 only; Duke Power will submit additional informa-tion prior to implementation on Unit 1.

In order to achieve the _ maximum advantage from the arbitrary break criteria L

change, we request a decision on this proposal by January 20, 1984. If I-can be of further assistance, or if a meeting with the Staff-is deemed beneficial

.for a final resolution of this matter, please contact me.

~

'Very truly yours, .,

$$ ,x ~

Hal.B. Tucker ROS/php.

H Attach 1ents

Mr. Harold R. Denton, Director November 18, 1983 Page 3 cc: Mr. James P. O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30303 NRC Resident Inspector Catawba Nuclear Station Mr. Robert Guild, Esq.

Attorney-at-Law P. O. Box 12097 Charleston, South Carolina 29412 Palmetto Alliance 21351 Devine Street Columbia, South Carolina 29205

~

Mr. Jesse L'. Riley Carolina Environmental Study Group-854 Henley Place Charlotte, North Carolina 28207 2 )

ATTACHMENT A 4

Summary of Benefits from the Elimination of Arbitrary Intermediate Pipe Breaks on Catawba Nuclear Station Unit 2 Category Benefit

1. ' Design, material and erection $4.4 million*

costs associated with 96 rupture devices

- 2. _ Relief of congestion, improvir.; 95 man-rem reduction in access for operation and maintenance. radiation exposure over life of Unit 2 ($240,000*)

3. 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.

4. Improvement in overall Improvement in ISI quality.

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

• Current (1983) dollars 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. For Class 1 piping, the allowable cumulative usage factor for pipe rupture

. postulation is 0.1; the 1974 ASME Code allowable is 1.0. This represents a large conservatism.

2. The pipe rupture " threshold" for all nuclear class piping is 80% of the 1974 ASME Code stress allowables. All arbitrary intermediate breaks involve stresses below this level. Hence a large conservatism exists.

3. Pipe rupture is recognized in Branch Technical Position MEB 3-1 as being a " rare event which may only occur under unanticipated conditions."

4. There is no technical or other justification for postulating arbitrary intermediate breaks, other than providing additional conservatism.

5. The additional pipe rupture devices resulting frar 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 Plants," 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.

It is concluded that the elimination of arbitrary intermediate break postulation is technically justifiable for the foregoing reasons.

=

ATTACHMENT'C Postulated Arbitrary Intermediate Breaks to be Eliminated on Catawba Unit 2 Estimated No.

Devices Eliminated Pipe Diameter Loca- Number Breaks Rupture. det Piping System - Di ameter - tion **- Eliminated Restraints Deflectors Steam Generator '2" IC 4 4 3 Blowdown' ~4" IC 2 3 1 2" OC 7* 6 4 4" OC 1* 1 2 Auxiliary 6" IC 8 13 '

8 Feedwater Reactor 2" IC 6 0 0 Coolant 3" IC 6 5 3 Residual Heat 6" IC 2 0 3 Removal 12" IC 6 0 6

Safety Injection 2" IC 2 1 2 6" IC 2 0 1 8" IC 4 0 0 10" . IC 8 0 4 li s" OC 3* 0 0 2" OC 5* 0 0 12" OC 3* 0 0-Chemical & Volume- 2"- IC 8 11 1 Control 3" IC 2 4 . 3 2" OC 9* 0 0 3" OC 2* 0 0 Main Steam Supply 6" OC- 4* 0 0 to Auxiliary Equip.

Main Steam 2" OC 14* 2 3 Main Steam Vent to Atmosphere 6" OC 8* 0 2 TOTALS 116 50 46-l

• Estimated-Evaluation not complete

• • I.C.=Inside Containment 0.C.=0utside Containment 4

ATTACHMENT D Revision to Catawba FSAR For Arbitrary Intermediate Break Criteria Change

CNS S*

= allowable design stress-intensity value, as defined in Subarticle NB-3600 of the ASME Boiler and Pressure Vessel Code,Section III.

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

3) If there are no intermediate locations where S exceeds 2.4 S, or U exceeds 0.1, no intermediate breaks are postulated. Intermediate breaks are not postulated in sections of straight pipe where there are no pipe fittings, flanges, valves or welded attachments.

-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:

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 11); at.all locations where the stress, S, exceeds 0.8 (1.2Sh

• bA )'

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.

=

S h basic material allowable stress at maximum (hot) tem-perature from the allowable stress tables in Appendix I of the ASME Boiler and Pressure Vessel Code,Section III.

S A

= allowable stress range for expansion stresses, as de-fined in Subarticle NC-3600 of the ASME Boiler and Pressure Vessel Code,Section III.

3.6-15 Rev. 9

CNS

3) If there are no intermediate locations where S exceeds 0.8 (1.2 S h + SA ), n intermediate breaks are postulated. Intermediate breaks are not postulated in sections of straight pipe where there are no pipe fittings, flanges, valves, or welded attachments. The pattern of postulated intermediate break locations is determined separately for the normal plant condition load combination and for that upset plant condition which has the highest stress.

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 or 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,

11) For Class F Piping:

At all locations where the stress, 5, Exceeds 0.8 (1.2 Sh

• SA )'

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.

S h

= basic material allowable stress at maximum (hot) temperature, per ANSI B31.1.0.

S A

= allowable stress range for expansion stresses, per ANSI B31.1.0.

3) For Class F Piping:

If there are no intermediate locations where S exceeds 0.8 (1.2 S h

• SA), n intermediate breaks are postulated. Intermediate breaks are not postulated in sections of straight pipe where there are no pipe fittings, flanges, valves or welded attachments.

3.6.2.1.2.2 Postulated Piping Break Locations For Moderate-Energy Piping Systems 3.6-16 Rev. 9

CNS Systems 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.

a) Cracks in Duke Class B, C and F piping are postulated at the following locations:

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 or environmental conditions except in portions of piping where the max-imum stress range is less than 0.4 (1.2 S h + SA ) as defined in items b)2)ii) and c)2)ii) of Section 3.6.2.1.2.1.

b) Cracks in Duke Class E, G and H piping are postulated at the following locations:

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 attachements)

.that result in the maximum effects from fluid spraying, flooding or environmental conditions. ,

3.6.2.1.2.3 Postulated Break Type, Size, and Orientation I

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.

3.6-16a Rev. 9 l Carryover i

CNS

3) .Circumferential breaks are assumed to result in pipe severance and separation amounting to at least a one-diameter laterai displacement of the ruptured piping sections unless physically limited by piping restraints. No limited break areas 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 locations will be documented in the FSAR.

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 calculated fluid pressure as modified by an analytically or experi-mentally determined thrust coefficient. Limited pipe displacement at the break location, 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.

5) Postulated pipe whip for target review will be defined by engineering judgement based on piping geometry, jet thrust direction, break lo-cation analysis type, and hanger location and type. When further confirmation is required, postulated piping breaks and targets are field reviewed after the drawing based analysis has been completed.

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 at 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) Longitudinal breaks are not postulated at terminal ends provided the piping at the terminal ends contains no longitudinal pipe welds.

3) Longitudinal breaks are assumed to result in an axial split without pipe severance. Splits are oriented (but not concurrently) at two diametrically-opposed points on the piping circumference such that 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 (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-3.6-17 Rev. 9

CNS 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.

5) Piping movement is assumed to occur in the-diretticii Of the jet reaction unless limited by structural mesher u ' piping restraints, or piping stiffness as demonstrated by ineiasQ c ).imit analysis.

For the purpose of analysis, breaks are assumed 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.

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

• bA ). Throughwall cracks 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 e) Environmental Effects (Temperature, humidity, water spray)

2) T..eughwall leakage cracks aj Environmental Effects (Temperature, Humidity) b) Flooding 3.6-18 Rev. 9 Carryover

CNS lb) Moderate-Energy Piping

1) Through-wall leakage cracks a) Flooding b) Environmental Effects (Temperature, humidity, water spray)-

c) Water Spray 3.6.2.2 Analytical Methods to Define Forcing Functions and Response Models 3.6.2.2.1 Reactor Coolant-System Dynamic Analysis This section summarizes the dynamic analysis as it applies to the LOCA result-ing from the postulated design basis pipe breaks in the main reactor coolant piping system. Further discussion of the dynamic analysis methods used to verify the' design adequacy of the reactor coolant loop piping, equipment and supports is given in Reference 1.

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 Referen:e 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 unac-ceptable 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 methor, lumped parameter, stiffness matrix formulation and assumes that all components behave 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 mathe-matical 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 broker. 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; 3.6-19 Rev. 9

. Carryover

. _ _ , . _ - - - . _- ~ , .

CNS 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.

The results are .used .to verify the adequacy of the restraints. The general dynamic s'olution 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 cal-culation is performed in two steps. The first step is to calculate the transient pressure, mass flow rates, and other hydraulic properties as a function of time.

The second step uses the results obtained from the hydraulic analysis, along with input of areas and direction coordinates and is to calculate the time history of forces at appropriate locations in the reactor coolant loops.

The hydraulic model represents the behavior of the coolant fluid within the entire reactor coolant system. Key parameters calculated by the hydraulic model are pressure, mass flow rate, and density. These are supplied to the thrust calculation, together with appropriate < station layout information to determine the concentrated time-dependent loads exerted by the fluid on the loops. In evaluating the hydraulic forcing functions during a postulated LOCA, the pressure and momentum flux terms are dominant. The inertia and gravitational terms are taken into accou'nt 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 require:; 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 twc phase depressur-ization stage; and (iii) the saturated stage.

3.6-20 Rev. 9 Carryover

3 ,

W' lJ4

)

M! -s CNS 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 DN8 correlation in addition to the reactor kinetics are incorporated in the code.

The THRUST computer program was developed to compute the transient (blewdown) loads resulting from a LOCA.

The blowdown hydraulic loads on primary loop components are computed from the fluid transient information calculated using the following time dependent forcing function:

"2 F r 144A {(P - 14.7) + ( )}

pg A2 744 c

!)

kwhich includes both the static and dynamic effects. The symbols and units are:

F = Force, Lb 4

f u

,- A = Aperture area, Ft2 l

'$ P = System Pressure, psia s

m = Mass flow rate, Lb,/Sec p = Density, Lb,/Ft3 gc= Gravitational Constant = 32.174 Lb,x Ft Lbf x Sec 2 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 numoer of apertures in any one node to give a total x force, total y force, and total z fcrce. These thrust forces serve as input to the piping / restraint dynamic analysis. Further details are given in Reference 1.

3.6-21 Rev. 9 Carryover

i_

CNS l-3.6.2.2.2 All Other Mechanical Piping Systems Dynamic Analysis Effects of pipe break are conservatively evaluated to determine the need for

~p ipe 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.

3.6.2.2.2.1 Assumptions 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 ,gp p_ p }) A g A jE where:

pE = fluid density at break at time t VE ,= fluid vel city at break at time t A

jE = pipe break exit area P

E

= control volume pressure at break at time t P

A

= ambient pressure  ;

3.6-22 Rev. 9 Carryover

CNS gC

= gravitation constant 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 g (P is the initial line pressure.) When the o

fluid is subcooled, nonflashing liquid, the flow will not be critical at the break area so that PE* A and V E= 2gc (P g PA )/PE. If PA<<P, the thrust force may be conservatively approximated by T = 2PgAjE When the fluid is saturated, flashing or super-heated vapor, the fluid can be assumed to be a perfect gas. The velocity for critical flow at the break area is given by VE = (KgcE P !PE) and PE* o ( 2 )

K+1 where K = Cp/Cy is a ratio of specific heats Cp = specific heat at constant pressure Cy = specific heat at constant volume

-A value of K=1.26 is justified for steam as being conservative. If PE nPA '

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; Y) = T . 2.Sp.DLF cos &

^j 3.6-23 Rev. 9 Carryover

_ _ - - _ _ _ _ _ _ _ _ _ - - - - - - - - - - - - -- - J

CNS where Y) = Normal load applied to a i.arget by the jet Ag = Cross-sectional area of jet intercepted by target structure A

j = Total cross-sectional area of jet at the target structure S

p = Shape factor Dtp = 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.

The ratio Ag/A3 represents the portion of the total mass flow from the jet 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 Analysis Methods to Verify Integrity and Operability 3.6.2.3.1 General Criteria for Pipe Whip Evaluation

1) The dynamic nature of the piping thrust load shall be considered. In the absence of analytical justification, a dynamic load factor of 2.0 is applied in determining piping. system response.

2). (Elastic perfectly plastic) pipe and crushable material properties may be considered as applicable. Consideration for crushable materials is described in Table 3.6.2-3.

3) Pipe whip is considered to result in unrestrained motion 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:

3.6-24 Rev. 9 Carryover

CNS

1) Equivalent static methoa

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- I 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.

The energy balance method is based on the principle of conservation of energy.

The kinetic energy of the pipe generated during the first 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 dynamic effects of pipe whip and jet impingement. The loadings on the restraint are determined by one of the methods outlined in Sections 3.6.2.2 and 3.6.2.3.

The design 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 function as a piping support, the discussion in Section 3.9.3.1.5 is applicable. Rupture loads with a dynamic load factor of 2.0 shall be added to the faulted leads 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 Containmert 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 "l:ot 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 3.6-25 Rev. 9 Carryover

CNS 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 accordance with Table 3.2.2-3 and applim ble response spectra curves (0.5 percent damping) ,

as developed from 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.

a) Primary System - Reactor Coolant System and any line 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-cable 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 17.

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.

3.6-26 Rev. 9 Carryover

CNS 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.

Design requirements for hot penetrations are as follow:

a) The guard pipe and bellows assembly constitute an extension of the Containment and as such meet Containment design conditions.

b) A guard pipe is required for lines that can overpressurize the annulus and/or release unacceptable amounts of radioactivity to the atomosphere.

c) Guard pipe contains and returns any process line leakage back to the l Containment.

d) Bellows design accommodates both axial and lateral displacements between l the Containment and Reactor Building for thern l, 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 assembly from foreign objects during construction and station operation.

3.6-21 Rev. 9 Carryover

CNS 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.

3.6.2.4.3 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 follows:

a) The recirculation line is an extension of Containment up through the first valve.

b) These valves are Safety Class 2 and are conservatively designed (600 psig 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.

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 A summary of postulated circumferential and longitudinal break locations are shown on Figures (later).

I 3.6-28 Rev. 9

4 CNS page deleted l

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i 3.6-29 Rev. 9

-<--4 ----m- w -- ---e y m y -

n Table 3.G.1-3 (Prge 3) 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 closed valve will be about the same; therefore, terminal end classification based on constraint and high stresses are not applicable. Duke  !

reviews these closed valve locations to assure l high stresses are not developed as a result of rigid constraint from nearby anchors of com-ponent connections in the non pressurized i portion of the piping.

APCSB 3-1, Appendix B and C SAR Section 3.6.2.1.2.1 In Appendix B, pipe break locations are specified Duke criteria specifies that if the threshold for ASME Section III Code Class, 1, 2, and 3 piping stress levels are not exceeded, then no inter-such that a minimum of two intermediate breaks are mediate breaks are postulated.

selected per run although threshold limits are not exceeded (for ASME Section III Code Class 1, 2, and l 3 piping). In Appendix C, a minimuin of either two I

or one intermediate breaks within the boundary of each compartment is specified.

MEB 3-1, Section B.1.b(6) SAR Section 3.6.2.4 Section B.1.b(6) requires that guard pipe assemblies Duke criteria is different from NRC criteria as between containment isolation valves meet the follow- described and justified below:

ing requirements:

Guard pipes provided between containment isola-

a. The design pressure and temperature should not tion valves are designed in accordance with SAR be less than the maximum operating temperature Section 3.6.2.4. Guard pipes are subjected to a pressure test as required by the material and pressure of the enclosed pipe under normal plant conditions. specification before welding to the penetration assembly.

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. Inde-Rev. 9

Tcble 3.6.1-3 (Prge 4)

c. Guard pipe assemblies should be subjected to a pendent design analysis have been conducted to single pressure test at a pressure equal to de- provide assurance that Duke penetration designs sign pressure. are acceptable. In addition, the extent of NDT conducted on guard pipes to flued head butt weld is such to assure integrity of design.

l l

MEB 3-1, Section B.1.c(1) SAR Section 3.6.2.1.2.1. l l

Intermediate breaks in Class 1 piping are postulated Duke criteria states that if there are no l at the two highest stress locations based on intermediate locations where S exceeds 2.4 S  ;

Equation (10) if two intermediate locations or U exceeds 0.1, no intermediate breaks are" l cannot be determined by application of Equations postulated. l (10), (12), and (13) or U>0.1.

M,EB 3-1, Section 8.1.c(2) SAR Section 3.6.2.1.2.1 Intermediate breaks in Class 2 and 3 piping are Duke criteria specifies that if the threshold postulated where the stresses exceed 0.8 (1.2Sg+ stress levels are not exceeded, then no inter-Sg ) but at not less than two locations based oh mediate breaks are postulated.

highest stress. Where the piping consists of a straight run without fittings, welded attachments, and valves, and all stresses are less than 0.8 (1.2S3 + S ), a minimum of one location should be A

choseH based on highest stress.

MEB 3-1, Sections B.I.c(3) SAR Section 3.6.2.1.2.1 Breaks in non-nuclear piping should be postulated Duke criteria is generally equivalent to NRC cri-at the following location: teria as described and justified below:

a. Terminal ends, Breaks in Duke Class F piping (non-nuclear, seismic) are postulated at terminal ends and at

b. At each intermediate pipe fitting, welded intermediate locations based on the use of ASME attachment, and valve. Section III analysis techniques, the same as ~

Duke Class B and C piping. Duke Class F piping is constructed in accordance with ANSI B31.1 and Rev. 9

Table 3.6.1-3 (Pzge 5) is dynamically analyzed and restrained for seis-mic loadings similar to ASME Section III piping.

Materials are specified, procured, received, stored, and issued under Duke's QA program simi-lar to ASME Section III materials except that certificate of compliance in lieu of mill test reports are acceptable on minor components, and construction documentation for erected materials is not uniquely maintained. Construction docu-mentation for erectcd materials is generically maintained. MTR are required for the bulk of piping materials.

MEB 3-1, Section B.2.e SAR Section 3.6.1.1.2 Through-wall cracks may be postulated instead of Duke criteria is generally equivalent to NRC cri-breaks in those fluid systems that qualify as teria as clarified below:

high energy fluid systems for short operational periods. This operational period is defined as The operational period that classifies such sys-about 2 percent of the time that the system oper- tems as moderate energy is either:

ates as a moderate energy fluid system.

a. One percent of the normal operating life-span of the plant, or

b. Two percent of the time period required to accomplish the system design function.

Regulatory Guide 1.46 SAR Section 3.6.2.1.2.1 Longitudinal breaks are postulated in piping runs Duke criteria is the same as NRC Branch Techni-4 inches nominal pipe size and larger. Circum- cal Position APCSB 3-1 and roughly equivalent ferential breaks are postulated in piping runs to Regulatory Guide 1.46 with expansion of def-exceeding 1 inch nominal pipe size. inition as described below:

As a minimum, there should be two intermediate Longitudinal breaks are postulated in piping break locations for each piping run or branch runs 4 inches nominal pipe size and larger ex-run. cept that longitudinal breaks are not postulated Rev. 9

Table-3.6.1-3 (P ge Sa) at terminal ends where the piping has no longi-tudinal welds.

Duke criteria specifics that if the_ threshold

~

stress levels are not exceeded, then no inter-mediate breaks are postulated. I

. Regulatory Guide 1.46 SAR Section 3.6.2.1.2 A whipping pipe should be considered capable of Duke criteria is the same as NRC Branch Techni-rupturing an impacted pipe of smaller nominal cal Position APCSS 3-1 and roughly equivalent pipe size and Ifghter wall thickness. 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 l pipes of smaller nominal pipe sizes, and (b) developing'through-wall cracks in larger nominal pipe. sizes with thinner wall thicknesses.

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