TXX-6086, Forwards FSAR Revs Reflecting Mods to Design Basis in Response to Modified GDC 4.Revs Will Be Documented in Future FSAR Amend.Cp Amend Not Required

Forwards FSAR Revs Reflecting Mods to Design Basis in Response to Modified GDC 4.Revs Will Be Documented in Future FSAR Amend.Cp Amend Not Required

ML20213E622
Person / Time
Site:	Comanche Peak
Issue date:	10/10/1986
From:	Counsil W TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:	Noonan V NRC - COMANCHE PEAK PROJECT (TECHNICAL REVIEW TEAM)
References
TXX-6086, NUDOCS 8611130233
Download: ML20213E622 (85)

Text

i Log # TXX-6086 File # 10010 TEXAS UTILITIES GENERATING COMPANY SKYWAY TOWER . 400 NORTH OLIVE STREET LB. 88 e DALLAS. TEXAS 75301 November 10, 1986 00^" $ff.UE Mr. Vince S. Noonan, Director Nuclear Reactor Regulation Comanche Peak Project Division of Licensing U. S. Nuclear Regulatory Commission Washington D. C. 20555

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)

DOCKET NOS. 50-445 AND 50-446 IMPLEMENTATION OF REVISED GDC-4

Dear Mr. Noonan:

The NRC issued a final rule on April 11, 1986, modifying GDC-4. Applicants for operating licenses seeking to modify design features to take advantage of the rule are required to document the revised design in an amendment to the pending FSAR. In addition, applicants should review the PSAR to determine whether the proposed change modifies design criteria set forth in the PSAR, which may require an amendment to the construction permit.

We have modified the design basis for CPSES to take advantage of the rule.

The dynamic effects associated with eight postulated primary loop breaks (per loop) have been removed from the design basis of CPSES Units 1 and 2. The dynamic effects associated with the rupture of the primary coolant large branch nozzles are still considered. These new design bases and the revised analyses performed using these new bases show that the primary coolant piping whip restraints and the reactor cavity non-crushable insulation are not needed and may be removed. The jet impingement barriers associated with the eight postulated primary coolant breaks (per loop) are also unnecessary and will not be installed in either Unit 1 or Unit 2 when these units are operating. We are evaluating the best time to remove the primary coolant piping whip restraints and reactor cavity non-crushable insulation and will document their removal in a future FSAR amendment. These changes will not effect containment design, environmental qualification or emergency core cooling system design.

Attachment 1 is an advance copy of FSAR revisions which will document these changes in a future FSAR amendment. These revisions describe:

(1) The new design bases that eliminate the dynamic effects of the eight reactor coolant loop pipe breaks (per loop),

(2) How the affected analyses have been performed, (3) The results of the new design bases and analyses (i.e., that the respective jet shields, piping whip restraints and reactor cavity non-crushable insulation are no longer needed),

8611130233 861010 PDR ADOCK 05000445 A PDR

TXX-6086 November 10, 1986 Page 2 of 2 (4) The elimination of the reactor coolant loop pipe break jet shields from the design, and (5) The references and a summary that describe the detailed analyses that justify the implementation of the modified GDC-4 for the design of CPSES Units 1 and 2.

This advance copy of FSAR revisions also includes some editorial corrections, clarifications, an update on how the break locations were chosen in the RHR piping and an update of the list of Code Cases used in the ASME Class 1 piping requalification. The update on the RHR pipe break locations is a conservative change. The new Code Cases listed are approved by Regulatory Guide 1.84 and, in the case of N-411 and N-397, by a docket specific letter. All changes are identified on their respective FSAR pages by a bar and "Rev." in the adjacent margin. A page by page description of the changes is provided in Attachment 2.

We have reviewed the design changes which take advantage of the modified GDC-4 against design criteria set forth in the PSAR and have determined that a construction permit amendment is not required.

Very truly yours, W. G. Counsil BSD/amb Attachments (2) c - NRC (0 + 40)

C. Trammell A. Vietti-Cook 1

o

TXX-6086 November 10, 1986 ATTACHMENT I TO TXX-6086 0F NOVEMBER 10, 1986 ADVANCE COPY OF FSAR REVISIONS 4

6 e

CPSES/FSAR TABLE 1.3-1 (Sheet 1 of 9)

DESIGN COMPARIS0N Systems and CPSES/FSAR Significant Significant Chapter Comoonents Section Similarities Differences 4.0 Reactor Systems Fuel 4.2 W. B. McGuire, none Watts Bar and Trojan Nuclear design 4.3 W. B. McGuire, none Watts Bar and Trojan Reactor vessel 4.3 W. B. McGuire, none 4.4 and Watts Bar Thermal-hydrau- 4.4 W. B. McGuire, none lic design Watts Bar and Trojan Reactivity con- 4.6 W. B. McGuire Part length control rods trol and Watts Bar utilized in the CPSES.

5.0 RCS and Connected 5.1 W. B. McGuire none Systems 5.2 and Watts Bar Reactor vessel

• 5.3 W. B. McGuire none and Watts Bar Reactor coolant 5.4.1 W. B. McGuire none pumps

• and Watts Bar Steam generators

• 5.4.2 W. B. McGuire CPSES utilizes model D4 and Watts Bar steam generators in Unit 1 and model D5 steam generators in Unit 2.

Piping

• 5.4.3 W. B. McGuire Reactor coolant loop pipe breaks are excluded from the CPSES Rev.

design basis for dynamic effects (Section 3.1.1.4)

CPSES/FSAR TABLE 1.3-2 (Sheet 1 of 29)

DESIGN CHANGES SINCE PSAR SUBMITTAL Systems CPSES/FSAR Comconents Section Chanaes I. Structures Category I Structures 3.8 Leak chase system behind liner of outdoor (other than containment) Category I tanks has been eliminated.

The requirement for roof blow out panels above the main steam lines in the Safeguards Building has been deleted.

Control Room 3.11B Limiting environmental conditions have been reduced to 800F and 60 percent relative humidity.

6.4 Positive pressure of 1/2 inch water gauge 3.11B changed to 0.1 inch water gauge during an accident.

Containment Systems 3.8 The containment external pressure design has been changed from a 3 psi differential to a 5 psi differential pressure.

The containment internal structure through liner anchors was eliminated.

The containment dome liner has been increased from 3/8 inch thickness to 1/2 i inch.

The containment liner paint has been changed.

1 3.9 The break flow in the reactor cavity I analysis is limited to 144 in2 by pipe whip rgstraints instead of the previous 150 ind. (See Section 3.9N.I.4.6 for application of leak-before-break to GDC-4.) Rev.

l a i

CPSES/FSAR TABLE 1.3-2 (Sheet 7 of 29)

DESIGN CHANGES ELNCE PSAR SUBMITTAL Systems CPSES/FSAR Components Section Chanaes 5.7, 7.6 An RCS cold overpressure control system is employed to provide for the mitigation of potential cold overpressurization transients, utilizing existing power operated relief valves with modifications to their actuation logic.

Steam Generators 5.4.3 CPSES utilizes model D5 steam generators in Unit 2. Thermal sleeves in reactor coolant loop branch nozzles have been deleted.

RHR systems 3.6 The RHR system is no longer identified as a high-energy fluid system located outside the containment.

Safety Injection 3.1 Change in SIS signal from coincident low System pressurizer pressure and water level to low pressurizer pressure.

Reactor Coolant 3.1 The leak-before-break technology has been and Whip Restraints 3.6 applied to exclude from the design basis 3.9N the dynamic effects of postulate ruptures 5.4 in the reactor coolant loop piping. Re v.

I

1 CPSES/FSAR TABLE 1.6-1 (Sheet 17 of 17)

Reference Review Report Section(1) Status

.)

" Reactor Vessel Head Drop Analyses," 9.1 U WCAP-9198, January 1978

" Westinghouse ECCS Evaluation Model, 15.6 AE February 1978 Version," WCAP-9220-P-A (Proprietary Version), WCAP-9221-P-A (Non-Proprietary Version), February 1978.

" Westinghouse Emergency Core Cooling System U Evaluation Model" - Modified October 1975 Version," WCAP-9168 (Proprietary) and WCAP-9169 (Non-Proprietary), September 1977.

" Technical basis for eliminating large primary 3.6B A primary loop pipe ruptures as the structural design basis for Comanche Peak Units 1 and 2" Rev.

WCAP-10528 (Proprietary) and WCAP-10528 (Non-Proprietary)

\

n _ _ . _ - - _

CPSES/FSAR pipe whipping, and discharging fluids, that may result from equipmznt failures and from events and conditions outside the nuclear power unit. However, the dynamic effects associated with postulated pipe ruptures or primary coolant loop piping in pressurized water reactors Rev. may be excluded from the design basis when analyses demonstrate the probability of rupturing such piping is extremely low under design basis conditions.[1]"

Discussion The station's structures, systems, and components important to safety are designed to accommodate the effects of, and to be compatible with, the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accident (LOCA). Environmental conditions are described in Section 3.11.

These structures, systems. and components are appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that can result from equipment failures and from events and conditions outside the nuclear power unit.

Details of the design, environmental testing, and construction of these systems, structures, and components are included in Chapters 3, 5, 6, 7, 8, 9 and 10. Evaluation of the performance of safety features is contained in Chapter 15.

The leak before break methodology demonstrates that the probability of rupturing primary coolant piping is extremely low under design basis conditions. It has been applied to CPSES Units 1 and 2 to exclude from the design basis the dynamic effects of postulated ruptures in the primary coolant loop piping, as discussed in Section 3.6B.2.5.1.

Rev. Implementation of this technology eliminates the need for primary coolant loop piping whip restraints and jet impingement barriers.

Containment design, emergency core cooling and environmental qualification requirements are not influenced by this modification.

3.1 - 6 a _ _ . -

CPSES/FSAR 3.1.1.5 Criterion 5 - Sharina of Structures. Systems, and Comnonents

" Structures, systems, and components important to safety shall not be shared between nuclear power units unless it can be shown that su:h sharing will not significantly impair their ability to perform their safety functions including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining unit [1]."

)

3.1- 6a 2 a

CPSES/FSAR abnormal leakage, of rapidly propagating failure, and of gross rupture

[1]."

Discussion The RCS pressure boundary is designed to accommodate the system pressures and temperatures attained under the expected modes of plant operation, including anticipated transients, and to maintain the stresses within applicable stress limits (Section 5.2.). Also, reactor coolant pressure boundary (RCPB) materials and selection and ,

fabrication techniques ensure a low probability of gross rupture of significant leakage.

In addition to the loads imposed on the system under normal operating conditions, consideration is also given to Abnormal loading conditions, such as seismic and pipe rupture, which are discussed in Sections 3.6 and 3.7. t L

The leak before break methodology demonstrates that the probability of '

rupturing primary coolant piping is extremely low under design basis conditions. It has been applied to CPSES Units 1 and 2 to exclude from the design basis the dynamic effects of postulated ruptures in the primary. coolant loop piping, as discussed in Section 3.6B.2.5.1. gey, Implementation of this technology eliminates the need for primary coolant loop piping whip restraints and jet impingement barriers.

Containment design, emergency core cooling and environmental qualification requirements are not influenced by this modification.

The system is protected from overpressure by means of pressure-relieving devices as required by applicable codes.

In conclusion, the RCS boundary has provisions fer inspection, testing, and surveillance of critical areas to assess the structural and leaktight integrity (Section 5.2). For the reactor vessel, a material surveillance program conforming to applicable codes is provided (Section 5.3).

3.1 11 l

l l

n

CPSES/FSAR 3.1.2.6 Criterion 15 - Reactor Coolant System Desian 1

"The reactor coolant system and associated auxiliary, control, and protection systems shall be designated with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences [1]."

Discussion The design pressure and temperature for each component in the RCS and associated auxiliary, control, and protection systems are selected to be above the maximum coolant pressure and temperature under all normal and anticipated transient load conditions.

In addition, RCPB components achieve a large margin or safety by the use of proven American Society of Mechanical Engineers (ASME) materials and design codes, use of proven fabrication techniques, nondestructive shop testing, and of integrated hydrostatic testing of assembled components.

Chapter 5 discusses the RCS design.

3.1.2.7 Criterion 16 - Containment Desian

" Reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require [1]."

Discussion A steel-lined, reinforced concrete containment structure encloses the entire RCS and is designed to withstand the pressures and temperatures resulting from a spectrum of postulated LOCAs and secondary system breaks.

3.1- 12 e

r i CPSES/FSAR l Th2 Emergency Ccre Cooling System cosis the reactor core and limits the release of radioactive materials to the environment.

3.1 12a e

CPSES/FSAR Accordingly, both of the automatic isolation valves are suitably protected and restrained as close to the valves as possible so that a pipe break beyond the restraint will not jeopardize the integrity and operability of the valves. Further, periodic testing capability of the valves to perform their intended function is essential. This criterion takes credit for only one of the two valves performing its intended function. For.normally closed isolation or incoming check valves (Cases I and IV in Figure 3.6B-10) a loss of reactor ecolant accident is assumed to occur for pipe breaks on the reactor side of the valve.

Branch lines connected to the Reactor Coolant System are defined as "large" for the purpose of these criteria if they have an inside diameter greater than 4 inches. Rupture of these lines results in a rapid blowdown from the Reactor Coolant System and protection is 15 basically provided by the accumulators and the low head safety injection pumps (residual heat removal pumps).

Branch lines connected to the Reactor Coolant System are defined as "small" if they have an inside diameter equal to or less than 4 inches. This size is such that Emergency Core Cooling System analyses using realistic assumptions show that no clad damage is expected for a break area of up to 12.5 square inches.

In order to assure the design function of essential systems in the event of a l0CA, break propagation within the affected loop shall be limited as follows:

A. Reactor Coolant Loop Piping Propagation of damage resulting from rupture of the main reactor coolant loop is permitted to occur but must not exceed the design basis for the containment, environmental qualification of Ray, equipment and Emergency Core Cooling System performance.

3.68 6

CPSES/FSAR The application of Leak-Before-Break technology to CPSES Unit 1 and 2, discussed in Section 3.68.2.1.1, has shown that the dynamic effects of main reactor coolant loop rupture, such as loop hydraulic forces, reactor internals reaction loads and primary equipment support loads, can be excluded from the design basis, as allowed by General Design Criterion 4 (Section g,y, 3.1.1.4).

B. LargeBranchL'fnes In the event of a rupture of a large branch line resulting in a LOCA, propagation of the break in the affected loop must not exceed 20 percent of the. flow area of the line which initially ruptured.

C. Small Branch Lines In the event of a rupture of a small branch line resulting in a LOCA:

1

1. Break propagation must be limited to the affected leg of the affected loop.

15

2. Propagation of the break in the affected leg must be limited to a total break area of 12.5 square inches.

However, no propagation is permitted when the initiating small break is the high head safety injection line.

3. Propagation of the break to a high head safety injection line connected to the affected leg must be prevented.

3.68.1.2.3 Environmental Analysis to Determine HELB Temperature Transients for Purposes of Equipment Qualification 40 (OutsideContainment) 3.68-7 a

CPSES/FSAR A. Design Bases Environmental analyses are performed to model the thermal effects of HELBs on safety related areas and equipment outside containment that are needed for safe shutdown of the plant. This 40 information is used to determine that the plant can be safely shutdown following a high energy line break.

I l

l 1

3.6B-7a

-- a

CPSES/FSAR 3.E8.2 DETERMINATION OF BREAK LOCATIONS AND DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING (INSIDE AND OUTSIDE CONTAINMENT) k The design bases for locating postulated breaks and cracks in piping

, inside and outside of the Containment, the procedure used to define

, the jet thrust reaction at the break or crack location and the jet impingement loading on adjacent structures, equipment, systems, and components are described as follows.

3.6B.2.1 Criteria Used to Define Break and Crack location And Confiauration This section provides criteria for the location and configuration of postulated pipe breaks in high energy piping systems.

3.6B.2.1.1 Reactor Coolant System (RCS) Main Loop Piping The generic Leak-Before-Break technology described in NUREG-1061 Volume 3 [18] has been applied to the CPSES Units 1 and 2 RCS main loop piping. This application of Leak-Before-Break methodology, allowed under the modified GDC-4, is discussed in reference [19]. The analyses show that the probability of RCS main loop piping breaks is extremely low, thus the dynamic effects of these breaks are not considered in the design basis of CPSES. CPSES leak-before-break was developed from detailed analysis of the following factors [19]: Rev.

a. The loads, material properties, transients, and geometry for CPSES Units 1 and 2 RCS primary loop are enveloped by the Westinghouse generic leak-before-break analyses.

i b. Stress corrosion cracking is precluded by use of fracture resistant materials in the piping system and controls on reactor coolant chemistry, temperature, pressure, and flow during normal operation.

, i 3.6B- 15

._ . 5 .. _ _ _ _ _ _ -

CPSES/FSAR ,

c. Water hammer is precluded in the RCS primary loop piping because of system design, testing, and operational considerations.

d. The effects of low and high cycle fatigue on the integrity of the primary piping are negligible.

e. Ample margin exists between the leak rate of the reference flaw and the criteria of Reg. Guide 1.45. (See FSAR Section IA(B) and 5.2.5).

f. Ample margin exists between the reference flaw chosen for leak detectability and the " critical" flaw.

Rev. g. Ample margin exists in the material properties used to demonstrate end-of-life (relative to aging) stability of the reference flaw.

The reference flaw will be stable throughout reactor life because of the ample margins in e, f, and g, above and will leak at a detectable rate which will assure a safe plant shutdown.

Although the dynamic effects of the RCS main loop piping breaks are not considered in the design basis, GDC 4 (FSAR Section 3.1.1.4) requires that primary coolant piping breaks be included in the design of the reactor containment, the ECCS, and in environmental qua'n fication. In order to comply with this requirement, main loop breaks are postulated as discussed in Sections 6.2, 6.3 and 3.11 respec.tively.

3.6B.2.1.2 High Energy Piping Other Than The RCS Main Loop High energy fluid systems are those that during normal plant conditions are either in operation or maintained pressurized under conditions where either or both of the following are met:

Maximum operating temperature exceeds 2000F 3.6B- 16

CPSES/FSAR Maximum operating pressure exceeds 275 psig.

It is noted that systems which operate within the pressure and temperature conditions specified for high energy fulid systems for less than two percent (2%) of the time are considered moderate energy.

Breaks for these systems are postulated as described in Section 3.6B.2.1.4. A list of high energy lines is given in Table 3.68-1.

Design basis break locations and types are postulated in accordance with NRC Regulatory Guide 1.46 and Branch Technical Position MEB 3-1.

Whore required, postulated pipe breaks are selected as described below and analyzed to demonstrate the capability for a safe shutdown of the plant.

A. ASME Section III, Code Class 1 Piping (Excluding Primary Reactor Coolant Loop Piping)

For high energy piping in the RC, CVC, RHR, and SI systems, the l Rev.

1977 Edition up to and including Summer 1979 Addenda of ASME B&PV Code Section III, Code Class I will be used to postulate -

pipe breaks. For all other high energy piping the 1974 Edition  !

up to and including Winter 1975 Addenda is applicable. The pipe 31 breaks are postulated to occur at terminal ends and at all intermediate locations in the piping system where none of the following criteria is met:

a) The primary plus secondary stress intensity range (equation (10) of ASME III Subparagraph NB-3653) derived on an elastically calculated basis under loadings associated with the OBE and normal and upset plant conditions does not exceed 2.4 Sm E 15 b) The primary plus secondary stress intensity range derived on an elastically calculated basis under loadings associated with the OBE and normal and upset plant conditions exceeds 2.4 Sm but is less than 3.0 Sm, and the cumulative usage factor is less than 0.1 g 3.68-17

CPSES/FSAR c) The primary plus secondary stress intensity range derived on an elastically calculated basis under loadings associated with the OBE and normal and upset plant conditions exceeds 3.0 Sm but the stress ranges computed by Equations (12) and (13) of Subparagraph NB-3653 of ASME III are less than 2.4 Sm and the cumulative usage factor 15 is less than 0.1.

1 Where intermediate break locations are not required, based upon the preceding criteria, two postulated pipe break locations will be selected on the basis of the highest Equation (10) stress. However, only one intermediate break l need be postulated in sections of straight pipe where there are no fittings, valves, or welded attachments.

B. ASME Section III, Code Class 2 And 3 Piping ASME B&PV Code,Section III, Class 2 and Class 3 piping breaks are postulated to occur at terminal ends and intermediate locations in each piping run or branch run. Breaks at intermediate locations are selected by either of the following 15 criteria:

a. At each location of potential high stress such as pipe Q112.2 fittings (elbow, tee, cross, flange, and nonstandard fittings), welded attachments, and valves.

b. At each location where the stresses associated with normal and upset plant conditions and an OBE event, calculated by i 15 the sum of Equations (9) and (10) of paragraph NC-3652 of '

the ASME B&PV Code,Section III, exceed 0.8 (1.2 Sh+

Sa)-

A minimum of two intermediate breaks are postulated between terminal 2 ends. If the stresses calculated between terminal ends are below 0.8 ,

Q112.2 (1.2 Sh + Sa), then breaks at intermediate locations are selected I at points of highest stress to satisfy the minimum requirements. If l 3.68-18 i

e

CPSES/FSAR there is only one intermediate point where the stresses exceed the value of 0.8 (1.2 Sh + Sa), then one additional intermediate break is postulated to occur at the location of the next highest stress.

The two locations are 3.68-18a a

CPSES/FSAR behavior. PIPERUP utilizes a direct step-by-step integration method to determine the time history response of the ruptured piping system.

A typical restraint impact curve is shown in Figure 3.6B-8. An incremental procedure is used to account for the nonlinear deformation and elastic-plastic effect of the pipe and restraints. The incremental equation of motion is evaluated by the Newmark's method (14].

3.6B.2.3 Dynamic Analysis Methods to Verify Intearity and Ooerability 3.6B.2.3.1 Reactor Coolant System Main loop The leak-before-break technology has been applied to CPSES Units 1 and 2 to exclude from the design basis the dynamic effects of postulated ruptures in the RCS main loop piping. This applies, in particular, to jet impingement loads on components and supports.

Rev.

Jet loads from large branch nozzle breaks are addressed in Section 3.68.2.3.2.

3.68.2.3.2 High-Energy Piping Other than the RCS Main Loop Pipe breaks are postulated in high-energy piping in accordance with the criteria in Section 3.6B.2.1.2. The analyses for determining the dynamic effects of pipe break are as follows:

A. Jet Impingement For breaks in systems excluding the primary loops, jet impingement force on any essential or primary component or support is computed using a jet model based on the geometry of the piping system and the characteristics including enthalpy, pressure, density, and similar properties. The force of the jet at a distance from the break is computed using the effective jet area and effective jet pressure at the target.

3.68-32

.- _ t

CPSES/FSAR Steady-state jet force is determined either by the RELAP-3 program or by a simplified method of steady-state function as described in Section 3.68.2.2.2. For modeling the jet impingement forces, the following assumptions are made for circumferential and longitudinal breaks:

i r

d 1

3.68- 33

_ _ - - -__.-._.-__t__.. . ._ _ __ ___.

CPSES/FSAR INTENTIONALLY LEFT BLANK 3.6B-34 9 . .. _ . . _

CPSES/FSAR INTENTIONALLY LEFT BLANK 3.6B-35 R

CPSES/FSAR

5. The design adequacy of essential systems and components to ensure that their design-intended functions will not be impaired to an )

unacceptable level on integrity or operability as the result of l high-energy pipe breaks.

6. Description of protective enclosures provided to protect safety-related equipment from the effects of a possible rupture in a high energy fluid piping system, including openings in these enclosures.

The implementation of criteria for inservice inspection is discussed in Section 6.6.8.

3.6B.2.5.1 Reactor Coolant System Main loop Piping Table 3.68.2 and Figure 3.68.9 identify the RCS main loop break locations. The eight main loop piping break locations (breaks 1 to 8 in Table 3.68.2 and Figure 3.6B.9) are included in the CPSES design basis for containment design, ECCS and environmental qualification Rev. requirements. These eight breaks are not part of the design basis for dynamic effects, as discussed in section 3.68.2.1.1. The three large branch nozzle breaks (breaks 9 to 11 in Table 3.6B.2 and Figure 3.68.9) are included in the design basis for dynamic effects.

The primary plus secondary stress intensity ranges and the fatigue cumulative usage factors at the design break locations specified in Reference [5] are given in Table 3.68.3 for a reference fatigue analysis. The reference analysis has been prepared to be applicable for many plants. It utilizes seismic umbrella moments which are higher than those used in Reference [5] such that the primary stress is equal to the limits of equation (9) in NB-3650 (Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code) at many locations in the system where in Reference [5] one location was at the limit. Therefore, the results of the reference analysis may differ slightly from Reference [5], but the philosophy and conclusions of Reference [5] are valid. There are no other locations in the model used in the reference fatigue analysis, 3.68-44

.--_--.--L- - _ _

CPSES/FSAR j consistent with Reference [5], where the stress intensity ranges and/or usage factors exceed the criteria of 2.4 Sm and 0.2, respectively.

t I

)

i t

l 1  !

1  !

l t

i 3.6B 44a 4

j 1

- _ - _ _ _ . . - _ _ _ _ - , _ , , _ - . . _ _ - _ . , , . - . . . _ _ , - . . . , - _ , . . _ , - _ . - _ _ _ _ , ..__,.-m_ . . _ , _ _ . _ . . _ - . _ _ . , _ . _ _ _ _ _ _ _ _ . - , _ , . . , . .,.,-------..-r, - ...

CPSES/FSAR 4

Actual plant moments for the CPSES are also given in Table 3.6B.3 at the design basis break location so that the reference fatigue analysis can be shown to be applicable for this plant. By showing actual plant moments to be no greater than those used in the reference analysis, it follows that the stress intensity ranges and usage factors for the l CPSES will be less than those for comparable locations in the reference mode. By this means it is shown that there are no locations other than those identified in Reference [5] where the stress intensity ranges and/or usage factors for the CPSES might exceed the criteria of 2.4 Sm and 0.2, respectively. Thus, the i applicability of Reference [5] to the CPSES has been verified.

i

Design loading combinations and applicable criteria for ASME Class 1 components and supports are provided in Section 3.9N. The forces associated with rupture of the branch nozzles (9 to 11, Table 3.68.2) to the reactor coolant loop piping systems are considered in combination with normal operating loads and earthquake loads for the Rev.

reactor coolant loop design in order to assure continued integrity of vital components and engineered safety features. Pipe rupture loads

include not only the jet thrust forces acting on the piping but also jet impingement loads on the primary equipment supports.

Barriers and layout are used to provide protection from pipe whip, Rev.

blowdown jet and reactive forces. Some of the barriers utiH zed for protection against pipe whip are as follows. The steam generator compartment walls serve 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 steam generator 4

compartment walls enclosed each reactor coolant loop into a separate l compartment, thereby preventing an accident, which may occur in any loop, from affecting another loop or the containment liner. The portion of the steam and feedwater lines within the containment have been routed behind barriers which separate these lines from all reactor coolant piping. The barriers described above will withstand loadings caused by jet forces and pipe whip impact forces.

3.6B-45

, , - , - ,,...m---. - -n, ,- , , .e,--,,..----.--,, -r--,- ,.-,,a ,,,,,,-,,,,,.--wn.....-, m.--m,-. - - - - - - - - - - - - - . , , , , . . - - - - - - - - , , - - - - , - - - -

CPSES/FSAR

15. Maody, F. J., " Fluid Reaction and Impingement Loads", Vol. 1, ASCE Specialty Conference on Structural Design of Nuclear Plant Facilities, pp. 219-262, December 1973.

31

16. Webb, S. W., " Evaluation of Subcooled Water Thrust Forces" Nuclear Technology, Vol. 31, October 1976.

17. Regulatory Guide, G.J.E. Willcutt, Jr., J.L. Lunsford and J.S.

Gilbert, COMPARE-MODl: "A Code for the Transient Analysis of Volumes with Heat Sinks, Flowing Vents, and Doors, LA-7199-MS, 40 March 1978.

18. NUREG - 1061 Volume 3, " Report of the USNRC Piping Review Committee. Evaluation of Potential Pipe Breaks", November,1984.

19. WCAP-10527 (Proprietary), WCAP-10528 (Non-Proprietary) " Technical Rev.

Bases for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Comanche Peak Units 1 and 2", April, 1984.

l 3.6B- 71 e

CPSES/FSAR TABLE 3.6B-1 SHEET 25 HIGH ENERGY LINE LIST System RESIDUAL. HEAT REMOVAL (RHR)

Conditions Condition Flow Line Number Building 9 Time of Break of Jet Remarks Diagram Press Temp Sub 2 Phase Figure No.

psig 0F cooled Flow 12-RH-1-001-250lR-1 RB 2235 120 X Note 30 ' 5.4-6 Rev.

12-RH-1-002-2501R-1 RB 2235 120 X Note 30 5.4-6 a

CPSES/FSAR TABLE 3.6B-1 SHEET 36 HIGH ENERGY LINE LIST

28. To SIS Cold Legs Loops 1 through 4, it is assumed that check valves ISI-8819 A, B, C, and D and ISI-8818 A, B, C, and D, all hold.

29. These lines are assumed to be high energy only up to normally closed valves ISI-8955 A, B, C and D.

30. For normally closed isolation valves a loss of reactor coolant accident is assumed to occur for pipe breaks on the reactor side of the first valve. Rev.

2

s CPSES/FSAR TABLE 3.6B-2 POSTULATED BREAK LOCATIONS FOR THE LOCA ANALYSIS OF THE PRIM ARY COOLANT LOOP' Rew Location of Break Opening Postulated Rupture & lyge, Areaa

1. Reactor vessel inlet nozzle Guillotine Ef fective cross-sectional flow area of the loop pipe

2. Reactor vessel outlet nozzle Guillotine Ef fective cross-sectional flow area of t he loop pipe

3. Steam generator inlet nozzle Guillotine Cross-sectional flow area of the loop pipe

4. Steam generator outlet nozzle Guillotine Cross-sectional flow area of the loop pipe

5. Reactor coolant pump inlet nozzle Guillotine Cross-sectional flow area of the loop pipe

6. Reactor coolant pump outlet nozzle Guillotine Cross-sectional flow area of the loop pipe

7. 50* elbow on the intrados Longitudinal Cross-sectional flow area of the loop pipe

8. Loop closure veld in crossover leg Guillotine Cross-sectional flow area of the loop pipe

9. Residual beat rescval (RHR) line/ Guillotine (viewed Cross-sectional flow area of the RHR line primary coolant lccp connection from the RHR line)

10. Accumulator (ACC) line/prisary Guillotine (viewed Cross-sectional flow area of the ACC line coolant loop connection from the ACC line)

11. Pressurizer surge (PS) line/ Guillotine (viewed Cross-sectional flow area of the PS line primary coolant loop connection from the PS line) w

1. Refer to Figure 3.6B-9 for location of postulated breaks in reactor coolant loop.

2. Less break opening area will be used if justified by analysis, experiments, or considerations of physical restraints such as concrete walls or structural steel.

3. Rupture locations I through 8 are postulated for containment design. ECCS and Environmental Qualification. These breaks are not postulated for dynamic ef fects (Section 3.68.2.1.1)

Rew l

AMENDMENT 11 JULY 31, 1980

l i l l

l

@ O l REACTOR U U  !

l PRESSURE Y I GENERATOR VESSEL l

4 1 8 PUMP l5 10 6

h DENOTES BREAK LOCATION PLAN VIEW l

STEAM GENERATOR n=:

M -

i 1 1

' REACTOR PtNP i PRESSURE 4 7 9 VESSEL HD L

R

')

0 ELEVATION NOTE: Postulated main reactor coolant loop break locations 1 through 8 are considered I

for containment design, ECCS and environmental qualification. These breaks are not postulated for dynamic effects (Section 3.6B.2.1.1) COMANCHE PEAK S.E.S.

FNAL SAFETY ANALYSIS REPORT JULY 31,1980 UNITS 1 and 2 Location of Postulated Breaks in Main Reactor Coolant Loop FK3URE 3.68-9

- - . t - - _____ --

t, jf

I .

w y

• D i elt... I9 E5 5 *B 55Y t ]

I

? .

= tj' g li r-la 5

- e s t' c ar e

i

~

$:il I'-

~ gE I

~

ai

• s I

! [j E

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/ =

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1. Break locations FOR CouT. SEE FIG 5.66-66 lyt- stC ot o- 25o e st-, I to 11 per Table Fost couT. Set PIG. S.Gs.go e 3.6B-2.

. 2. Based on CDC-4, a

i

\ leak-before-break (Section 5.4.14) 3 4

/ '

the loop whip REAcTo= cootaur ,l f y g g tf g R -t s ruu- / I pi..s.c .sz > restraints and non-TBK- RC Perac.og l

/ ,

y crushable insulation A may be eliminated.

I F ky4ft *

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/ Pr

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4 SI' 1.D. Psre # l

( c mo s, s ove = > H t 2[g[r L M\ ]

j I '

\ /

= 4 5 S 7 2 < g. me Acyo sg

2. = c-i. ois. 2 S ei n .i [/* Tex- ate ec sev

o= cour. sea w G.s.Go-G, gff x

\' 12 - RC -I- 007- 2 50 t R-1 Wo st c o u T. s e t , ,

fig. S.66-G4 itPIffes(R 32, 1983 COMANCHE PEAK S E S

@

• SEE NorE i N M W M N EPORT UNITS I and 2 REACTOR COOT. ANT IDOP 81

!! t nuu i.sa-o, stone sii

~ ___ . _ _ _ _ _ _ _ -_ _-

NOTES :

1. Break locations 1 to 11 per Table 3.6B-2.

MD 2. Based on GDC-4, leak-before-break (Section 5.4.14) the loop whip restraints sTEAu may be eliminated.

y GEu . we.t 7 T T -* G.wc- -ot9- Tsoi n-e Fost cour sea Fic. S.ss-49 I

iYr.ne-t-ose.2soise : 3

\ pose couT sEE RE Acto t g ciG. a.ca- So co ot. AuT Pu 64 P sec - 75oi n.

TBA- Mc PC Pc-o f 29'I.D. PIPE (e4oT LEG) 8 e is stE DETAIL"A Y e e,ESo s ric.346-es q,, ,___'

/io-ec-e-o31-25cin i c D'I.o 2 s=o.e cour. see o'D (sq,p'"E FIG. 5.G B-SS , 3 to Y- .I .

se c-7 5 oist-i i S s'- 1.D. F1 r"E C CE"oS

• OV ER )

" b I

<J L 2- RC-i-055 2 5 0iR- e

• k REAC"fDR x .= o .e co m. ... P ,. . .. . . . .. vax.=c-cev i_Ef.E LJ D AME lefe:E NT 42 SEPIIMRER 12. 1983 COMANCHE PE AK r, E S F94AL SAFETY ANALYSE FEPG 7 UNeTS I and 2 REAC10R CODIANT IDOP O2 MTRA IIT TIOld FI[aAE 3.65-90 elDOP 02:

NOTES a

1. Break locations 1 to 11 per Table 3.6B-2.

2. Based on CDC-4, leak-before-break (Section 5.4.14) the loop whip restraints may be eliminated.

N C DTEAM so-RC-s-055 2soie-t srom GEU.ut.5 CO*f ME 8'an. S.(, e - 5 6

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< J\'@ RC-250iR -t 34*1.D. PtPE a

( stEacTOst (f RCESovt R)

Tow- sec P c 8EV AMINDMINT 42 5tPIIMBER 12. 19R)

LEGEuo COMANCHE FE AK S E S FNAL SAFETV ANALYSIS FUCRT

@

• SEE NOTE 1 UNITS l and 2 REACTOR COOLANT IAOP G)

STRESS NODE & 84 EAR POINT AND RFCTM A f gMUM_

f usu 1. e.n-91 (I rwir 61)

__-.__ _m..___.___________..____.__ _____ ___ _ _ _ - -

4 Rc.e-o75-7soir. sh e c.s- oi ,z$oie..

roe couT SEE F Q S CE d,8 FOR cowT t,EE g eTggs 5 ec-i-ow,- z g o i = -i F'G 8 63-58 1. Break locations FOR couT.SEE Ficv 5.ss r5 8

1 to 11 per N so - ac- i- ov a - 25 oi s-l %

c M k 3.M 4.

'~

io Fom couT. ss E FlQ. A GS-)$ kg p I

leak-before-f j m

\

y a c- tsoi m -l _ \/ f/f break (Section

1. - n,5 g, o. ,,, , a

- 5.4.14) the loop (COLD LEG) G Whip restraints 3,g .. may be eliminate 1.

W EIG 3G'B"09 #C p,

C h

h /

14 FC-1-15b- T';OIR l FOR couT- SEE FIG. 5 68 GS a

k REACTOR s Tax- Rc e C = v

1. o'%+s 'a WD % '

{;

( 48 ,

STE AM f 6tu.ut4 kg g TBY- R,c Pet,G

@ -04 ,$ l 2 RC-e-o77 2SOIR-l 8 g For cowT EE E 12-RC-t- O(.9 25os R / ge Foa cour see F.sf-I S.G6-G4 ( ) j-p, q , g g,g . g ,

M (

s3 r o.

4 '&

1 i LE OE D P i , 1983 l CD - 5** "0T6 i COMANCHE PE AK S E S FNAL SAFETY ANALYSS KPORT UNITS I and 2 RfACTOR C(X) TANT IJX)P 84 A Nh T1 fE184 1.68-92 ( I AW)F 04)_

CPSES/FSAR 3.7N.3.6 Three Components for Earthauake Motion Methods used to account for three components of earthquake motion for subsystems in Westinghouse's scope of responsibility are given in Section 3.7N.2.6.

3.7N.3.7 Combination of Modal Resoonses Methods used to combine modal responses for subsystems in Westinghouse's scope of responsibility are given in Section 3.7N.2.7.

3.7N.3.8 Analytical Procedures for Pipina The Class 1 piping systems are analyzed to the rules of the American Society of Mechanical Engineers (ASME) Code,Section III, N8-3650.

When response spectrum msnthods are used to evaluate piping systems supported at different elevations, the following procedures are used.

The effect of differehtial seismic movement of piping supports is included in the piping analysis according to the rules of the ASME Code,Section III, NB-3653. According to ASME definitions, these displacements cause secondary stresses in the piping system. The response quantity of interest induced by differential seismic motion of the supports is computed statically and addresses the relative g,y, phase of the seismic motion at each support.

In the response spectrum dynamic analysis for evaluation of piping systems supported at different elevations, the envelope floor response l Rev.

spectrum corresponding to the support locations is used. Westinghouse does not have in their scope of analysis any piping systems interconnected between buildings.

3.7N I7m

CPSES/FSAR TABLE 3.7N-1 DAMPING VALUES USED FOR SEISMIC SYSTEMS ANALYSIS FOR WESTINGHOUSE SUPPLIED E0VIPMENT

.~

Damping (Percent of Critical)

Upset Faulted Conditions Condition 1115 (0BE) 11SE. DBA)

Primary coolant loop system 2 4 components and large pipinga,b Rev.

Small pipingb 1 2 Welded steel structures 2 4 Bolted and/or riveted 4 7 steel structures a Applicable to 12 inch or larger diameter piping.

b For piping analyzed by the response spectrum method, ASME Code Case N-411 damping values may also be used in lieu of the damping provided in this Rev.

table.

a

CPSES/FSAR plugging. The secondary side is then repressurized (to a higher pressure) and the underside of the tubesheet is again checked for leaks. This process is repeated until all the leaks are re-paired. The maximum (final) secondary side test pressure reached is 840 psig.

The total number of tube leakage test cycles is defined as 800 during the 40 year life of the ple.nt. Following is a breakdown of the anticipated number of occurrences at each secondary side test pressure:

Test Pressure (psig) Occurrences 200 400 400 200 600 120 840 80 Both the primary and secondary sides of the steam generators will be at ambient temperatures during these tests.

3.9N.l.2 Computer Programs Used in Analyses '

The following computer programs have been used in dynamic and static analyses to determine mechanical loads, stresses, and deformations of Seismic Category I components and equipment. These are described and verified in Reference [1] and [13].

Rev. 1. WESTDYN - static and dynamic analysis of piping systems.

2. FIXFM - time history response of three dimensional structures.

3.9N-24 m

1 CPSES/FSAR 3.9N.l.4.2 Analysis of the Reactor Coolant Loop and Supports The reactor coolant loop piping is evaluated in accordance with the 10 criteria of ASME III, NB-3650 and Appendix F. The loads included in the Q112.25 evaluation result from the SSE, deadweight, pressure, and LOCA loadings (loop hydraulic forces, asymmetric subcompartment pressurization forces, and reactor vessel motion). The results of the stress analysis of the reactor coolant loop piping are given in Table 3.6B-3.

The loads used in the analysis of the reactor coolant loop / supports System are described in detail below.

Pressure Pressure loading is identified as either membrane design pressure or general operating pressure, depending upon its application. The mem-brane design pressure is used in connection with the longitudinal pres-sure stress and minimum wall thickness calculations in accordance with the ASME Code.

The term operating pressure is used in connection with determination of the system deflections and support forces. The steady state operating hydraulic forces based on the system initial pressure are applied as general operating pressure loads to the reactor coolant loop model at change in direction or flow area.

Weight A dead weight analysis is performed to meet ASME Code requirements by applying a 1.0 g load downward on the complete piping system. The piping is assigned a distributed mass or weight as a function of its properties. This method provides a distributed loading to the piping system as a function of the weight of the pipe and contained fluid during normal operating conditions.

3.9tl-26

_ - 9 , -- - . , - . - -

CPSES/FSAR Seismic The forcing functions for the reactor coolant loop seismic piping analy-ses are derived from dynamic response analyses of the Containment Build-ing subjected to seismic ground motion. Input is in the form of floor response spectrum curves at various elevations within the Containment Building.

For the OBE and SSE seismic analyses, 2 and 4 percent critical damping, respectively, are used in the reactor coolant loop / supports system analysis.

In the response spectrum method of analysis, the total response loading obtained from the seismic analysis consists of two parts: the inertia response loading of the piping system and the differential anchor move-ments loading. Two sets of seismic moments are required to perform an ASME Code analysis. The first set includes only the moments resulting from inertia effects and these moments are used in the resultant moment (Mi) value for equations 9 and 13 of K8-3650. The second set includes the moments resulting from seismic anchor motions and are used in equa-tions 10 and 11 of N8-3650. Differential anchor movement is discussed in Section 3.7N.

Loss of Coolant Accident Blowdown loads are developed in the broken and unbroken reactor coolant loops as a result of transient flow and pressure fluctuations following a postulated pipe break to the large branch nozzles of the reactor coolant loop (breaks 9 to 11, Table 3.68-2). Structural consideration of dynamic effects of postulated pipe breaks requires postulation of a Rev.

finite number of break locations. Postulated pipe break locations are given in Section 3.68.

Broken loop time history dynamic analysis is performed for these postulated break cases. Hydraulic models are used to generate time-3.9N-27 T

CPSES/FSAR dependent hydraulic forcing functions used in the analysis of the reactor coolant loop for each break case. For a further description of R:v. l the hydraulic forcing functions, refer to Section 3.68.2.2.1.

Transients The ASME Code requires satisfaction of certain requirements relative to j operating transient conditions. Operating transients are tabulated in Section 3.9N.l.l.

The vertical thermal growth of the reactor pressure vessel nozzle centerlines is considered in the thermal analysis to account for equipment nozzle displacement as an external movement.

i The hot moduli of elasticity E, the coefficient of thermal expansion at the metal temperaturea , the external movements transmitted to the piping due to vessel growth, and the temperature rise above the ambient temperature a T, define the required input data to perform the flexibility analysis for thermal expansion.

To provide the necessary high degree of integrity for the RCS, the tran-sient conditions selected for fatigue evaluation are based on conserva-tive estimates of the magnitude and anticipated frequency of occurrence of the temperature and pressure transients resulting from various plant operation conditions.

3.9N.l.4.3 Reactor Coolant Loop Models and Methods l The analytical methods used in obtaining the solution consists of the transfer matrix method and stiffness matrix formulation for the static structural analysis, the response spectra method for seismic dynamic analysis, and time history integration method for the loss of coolant accident dynamic analysis.

3.9N-28 s

CPSES/FSAR The integrated reactor coolant loop / supports system model is the basic system model used to compute loadings on components, component supports, and piping. The system model includes the stiffness and mass characteristics of the reactor coolant loop piping and components, the stiffness of supports, the stiffnesses of auxiliary line piping which affect the system and the stiffness of piping restraints. The deflection solution of the entire system is obtained for the various loading cases from which the internal member forces and piping stresses are calculated.

Static The reactor coolant loop / supports system model, constructed for the WESTD(N computer program, is represented by an ordered set of data which lRev.

numerically describes the physical system. Figure 3.9N-1 shows an isometric line schematic of this mathematical model. The steam generator and reactor coolant pump vertical and lateral support members are described in Section 5.4.14.

The spatial geometric description of the reactor coolant loop model is based upon the reactor coolant loop piping layout and equipment drawings. The node point coordinates and incremental lengths of the members are determined from these drawings. Geometrical properties of the piping and elbows along with the modulus of elasticity E, the coefficient of thermal expansion a , the average temperature change from ambient temperature AT, and the weight per unit length are specified for each element. The primary equipment supports are represented by stiffness matrics which define restraint characteristics of the supports. Due to the symmetry of the static loadings, the reactor pressure vessel centerline is represented by a fixed boundary in the system mathematical model. The horizontal thermal growth of the reactor pressure vessel (RPV) is included in the loop model by modeling the RPV as an equivalent pipe from the RPV centerline to the RPV nozzle. The 20 vertical thermal gruwth at the RPV nozzle location i

3.9N-29 R

CPSES/FSAR priate response spectra value to give the modal amplitude for each mode.

The total modal amplitude is obtained by taking the square ront of the sum of the squres of the contributions for each direction.

The modal amplitudes are then converted to displacements in the global coordinate system and applied to the corresponding mass point. From these data the forces, moments, deflections, rotations, support reactions and piping stresses are calculated for all significant modes.

The total seismic response is computed by combining the contributions of the significant modes by the square root of the sum of the squares Rsv. l method and accounting for closely spaced modes as described in Section 3.7N.

Loss of Coolant Accident The mathematical model used in the static analyses is modified for the loss of coolant accident analyses to represent the severance of the Rev. reactor coolant loop piping nozzles at the postulated break location.

Modifications include addition of the mass characteristic of the piping and equipment. To obtain the proper dynamic solution, two masses, each containing six dynamic degrees of freedom and located on each side of the break, are included in the mathematical model. The natural fre-quencies and eigenvectors are determined from this broken loop model.

The time-history hydraulic forces at the node points are combined to obtain the forces and moments acting at the corresponding structural lumped mass node points.

The dynamic structural solution for the full power loss of coolant accident and steam line break is obtained by using a modified-predictor-corrector-integration technique and normal mode theory.

When elements of the system can be represented as single acting members (tension or compression members), they are considered as nonlinear 3.9N-32 3 -.

T CPSES/FSAR i

I elements, which are represented mathematically by the combination of a j gap, a spring, and a viscous damper. The force in this nonlinear i element is treated as an externally applied force in the overall normal j mode solution. Multipled nonlinear elements can be applied at the same  !

node, if necessary.  !

The time-history solution is performed in subprogram FIXFM3. The input 10 to this subprogram consists of the natural frequencies, normal modes, Q112.25 applied forces and nonlinear elements. The natural frequencies and normal modes for the modified reactor coolant loop dynamic model are determined with the WESTDYN program. To properly simulate the release Rev.

of the strain energy in the pipe, the internal forces in the system at the postulated break location due to the initial steady state hydraulic forces, thermal forces, and weight forces are determined. The release of the strain energy is accounted for by applying the negative of these internal forces as a step function loading. The initial conditions are equal to zero because the solution is only for the transient problem (the dynamic response of the system from the static equilibrium position). The time-history displacement solution of all dynamic degrees of freedom is obtained using subprogram FIXFM and employing 4 percent critical damping.

, The loss of coolant accident displacements of the reactor vessel are

{

applied in time-history form as input to the dynamic analysis of the L

reactor coolant loop. The loss of coolant accident analysis of the reactor vessel includes all the forces acting on the vessel including internals reactions, cavity pressure loads, and loop mechanical loads.

The reactor vessel analysis is described in Section 3.9N.1.4.6.

} The main loop piping breaks and, therefore, the vessel cavity pressurization effects are not part of the design basis (because they are excluded as dynamic effects, Section 3.1.1.4). However, they are Rev.

included as a conservative estimate of vessel motion due to LOCA. If required, vessel motion for the postulated branch nozzle breaks can be 3.9N-33 J

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

CPSES/FSAR used in place of the conservative main loop piping breaks discussed in Rev. this section.

The resultant asymmetric external pressure loads on the RCP and steam generator resulting from a postulated pipe rupture and pressure buildup in the loop compartments are applied to the same integrated RCL/ supports system model used to compute loadings on the components, component supports, and RCL piping as discussed above. The response of the entire system is obtained for the various external pressure loading cases from which the internal member forces and piping stresses are calculated.

For each pipe break case considered, the equipment support loads and piping stresses resulting from the external pressure loading are added to the support loads and piping stresses calculated using the loop LOCA hydraulic forces and RPV motion.

10 The break locations considered for subcompartment pressurization are Q112.25 those postulated for the RCL LOCA analysis as discussed in Section 3.6N and WCAP-8172 (Reference [1] of Section 3.6N). The asymmetric subcompartment pressure loads are provided to Westinghouse by Gibbs &

Hill, Inc. The analysis to determine these loads is discussed in Section 6.2.

The time-history displacement response of the loop is used in computing support loads and in performing stress evaluation of the reactor cool-ant loop piping.

The time-history displacements of the FIXFM3 program are used as input to program WESOYN to determine the internal forces, deflections, and stresses at each end of the piping elements. For this calculation the displacements are treated as imposed deflections on the reactor coolant loop masses.

Transient Operating transients in a nuclear power plant cause thermal and/or pressure fluctuations in the reactor coolant fluid. The thermal 3.9N-34 t -, - .__ . - - - -

CPSES/FSAR 4

transients cause time varying temperature distributions across the pipe wall. These temperature distributions resulting in pipe wall stresses j may be further subdivided in accordance with the Code into three parts, a uniform, a linear, and a nonlinear portion. The uniform portion

results in general expansion loads. The linear portion causes a bending moment across the wall and the nonlinear portion causes a skin stress.

f I

1 I

1 I

a i

3.9N-34a ce

CPSES/FSAR

4. Discontinuity temperature (T A - T B) represents the difference in average temperature at the cross sections on each side of a discontinuity.

Each transient is described by at least two load sets representing the maximum and minimum stress state during each transient. The construction of the load sets is accomplished by combining the following to yield the maximum (minimum) stress state during each transient.

1. AT.1

2. AT-2

3. a AT A - oB TB-

4. Moment loads due to TA -

5. Pressure loads.

This procedure produces at least twice as many load sets as transients for each point.

As a result of the normal mode spectral technique employed in the seismic analysis, the load components cannot be given signed values.

Seismic loads are considered in the fatique analysis as having either a Rev.

positive or negative sign, thus ensuring the most conservative combination of seismic loads are used in the stress evaluation.

For all possible load set combinations, the primary plus secondary and peak stress intensities, fatigue reduction factors (Ke) and cumulative usage factors, U, are calculated. The WESTDYN program is used to l Rev.

perform this analysis in accordance with the ASME Code,Section III, Subsection NB-3650. Alternatively, detailed finite element stress analyses may be used to determine primary plus secondary and peak stress 3.9 N-37 l

[ a j

CPSES/FSAR intensities, for the load set combinations. Since it is impossible to predict the order of occurrence of the transients over a 40 year life it is assumed that the transients can occur in any sequence. This is a very conservative assumption.

The combination of load sets yielding the highest alternating stress intensity range is used to calculate the incremental usage factor. The next most severe combination is then determined and the incremental usage factor calculated. This procedure is repeated until all combinations having allowable cycles <106 are formed. The total cumulative usage factor at a point is the summation of the incremental usage factors.

3.9N.1.4.4 Primary Component Supports Models and Methods The static and dynamic structural analyses employ the matrix method and normal mode theory for the solution of lumped parameter, multimass structural models. The equipment support structure models are dual purpose since they are required: 1) to quantitatively represent the clastic restraints which the supports impose upon the loop, and 2) to evaluate the individual support member stresses due to the forces imposed upon the supports by the loop.

A description of the supports is found in Section 5.4.14. Detailed models are developed using beam elements and plate elements, where 12.25 applicable. The reactor vessel supports are modeled using the WECAN computer program. Structure geometry, topology and member properties are used in the modeling. Steam generator and reactor coolant pump supports are modeled as linear or non-linear springs.

For each operating condition, the loads (obtained from the RCL analysis) ,

acting on the support structures are appropriately combined. Although Rsv. the main loop piping breaks are not part of the design basis (excluded as dynamic effects, Section 3.1.1.4), they are conservatively included 3.9N-38 a

CPSES/FSAR in the load and stress evaluation tables for the primary component Rev.

support. Reactor coolant loop normal and upset condition thermal expansion loads are treated as primary loadings for the primary component supports. The adequacy of each member of the steam generator supports, reactor coolant pump supports, and piping restraints is verified by solving the ASME III Subsection NF stress and interaction equations. The adequacy of the RPV support structure is verified using the WECAN computer program and comparing the resultant stresses to the criteria given in ASME III Subsection NF.

12.25 Tables 3.9N-14 through 3.9N-17 present maximum stresses in each member of the steam generator, reactor coolant pump, and pressurizer support structures expressed as a percentage of maximum permissible values for all operating condition loadings. The loads on the reactor vessel l Rev.

supports and the resulting stresses are shown in Table 3.9N-19. The above loads and stresses include the effects of loads resulting from asymmetric subcompartment pressurization.

3.9N.1.4.5 Analysis of Primary Components 13 Equipment which serves as part of the pressure boundary in the reactor coolant loop includes the steam generators, the reactor coolant pumps, the pressurizer, and the reactor vessel. This equipment is evaluated for the loading combinations outlined in Table 3.9N-2. The equipment is analyzed for: 1) the normal loads of deadweight, pressure and thermal,

2) mechanical transients of 08E, SSE, and pipe ruptures, including the effects of asymmetric subcompartment pressurization and 3) pressure and temperature transients outlined in Section 3.9N.I.l.

The results of the reactor coolant loop analysis are used to determine the loads acting on the nozzles and the support / component interface locations. These loads are supplied for all loading conditions on an

" umbrella" load basis. That is, on the basis of previous plant analyses, a set of loads are determined which should be larger than those seen in any single plant analaysis. The umbrella loads represent a Con-3.9N-39

. .-. . _ . -- - .-__ - . .--- - t . --

CPSES/FSAR the procedure discussed above in given in Reference [3]. This report discusses the evaluation procedure in detail as applied to a severe faulted condition (a postulated loss of coolant accident), and concludes that the integrity of the reactor coolant pressure boundary would be maintained in the event of such an accident.

The pressure boundary portions of Class 1 valves in the RCS are designed and analyzed according to the requirements of NB-3500 of the ASME Code,Section III. These valves are identified in Section 3.9N.3.2.

Valves in sample lines connected to the RCS are not considered to be ANS Safety Class 1 nor ASME Class 1. This is because the nozzles where the line connect to the primary system piping are orificed to a 3/8 inch hole. This hole restricts the flow such that loss through a severance of one of these lines can be made up by normal charging flow.

3.9N.l.4.6 Dynamic Analysis of Reactor Pressure Vessel for Postulated Loss of Coolant Accident

1. Introduction 10 Q112.25 This section presents the method of computing the reactor pressure vessel response to a postulated loss of coolant accident (LOCA). The dynamic analysis of the reactor vessel was performed prior to the leak-before-break update to GDC-4 and therefore included the effects of main l loop piping breaks and reactor vessel cavity pressurization. These  !

effects, no longer required for the dynamic analysis, are conservative loadings for the postulated LOCA analysis of the reactor vessel and internals discussed in this section and identified in Table 3.68-2. The structural analysis considers simultaneous application of the time-10 history loads on the reactor vessel resulting from the reactor coolant Q112.25 loop mechanical loads, internal hydraulic pressure transients, and  !

reactor cavity pressurization (for postulated breaks in the reactor coolant pipe at the vessel nozzles). The vessel is restrained by 3.9N-44

._--_--_.-s - - - _ _ - - - . -_- - -

CPSES/FSAR reactor vessel support pads and shoes beneath four of the reactor vessel nozzles and the reactor coolant loops with the primary supports of the steam generators and the reactor coolant pumps.

Pipe displacement restraints installed in the primary shield wall limit the break opening area of the vessel nozzle pipe breaks to less than 144 square inches. This break area was determined to be an upper bound by using worst case vessel and pipe relative motions based on similar plant Q112.25 analyses. Detailed studies have shown that pipe breaks at the hot or cold leg reactor vessel nozzles, even with a limited break area, would give the highest reactor vessel support loads and the highest vessel displacements, primarily due to the influence of reactor cavity pressurization. The dynamic analysis of the reactor vessel was performed prior to the leak-before-break update to GDC-4 and therefore included the effects of main loop piping breaks and reactor vessel cavity pressurization. These effects, no longer required for the dynamic analysis, are conservative loadings for the postulated LOCA Rev.

analysis of the reactor vessel and internals discussed in this section and identified in Table 3.68-2. By considering these breaks, the most severe reactor vessel support loads are determined. For completeness, an additional break outside the shield wall, for which there is no cavity pressurization, was also analyzed, specifically, the pump outlet nozzle pipe break.

2. Interface Information Asymmetric reactor cavity pressurization loads were provided to Westinghouse by Gibbs and Hill, Inc.

10 All other input information was developed within Westinghouse. This Q112.25 information includes: reactor internals properties, loop mechanical loads and loop stiffness, internal hydraulic pressure transients, and ,

reactor support stiffnesses. These inputs allowed formulation of the  !

mathematical models and performance of the analyses, as will be described. l l

3.9N-45 l l

l a

i 1

CPSES/FSAR

3. Loading Conditions The dynamic analysis of the reactor vessel was performed prior to the leak-before-break update to GDC-4 and therefore included the effects of main loop piping breaks and reactor vessel cavity pressurization. These Rcv. effects, no longer required for the dynamic analysis, are conservative loadings for the postulated LOCA analysis of the reactor vessel and internals discussed in this section and identified in Table 3.68-2.

Following a postulated pipe rupture at the reactor vessel nozzle, the reactor vessel is excited by time-history forces. As previously mentioned, thse forces are the combined effect of three phenomena: (1) reactor coolant loop mechanical loads, (2) reactor cavity pressurization forces and (3) reactor internal hydraulic forces.

The reactor coolant loop mechanical forces are derived from the elastic analysis of the loop piping for the postulated break. This analysis is described in Section 3.9N.1.4.3. The loop mechanical forces which are 10 Q112.25 released at the broken nozzle are applied to the vessel in the RPV blowdown analysis.

Reactor cavity pressurization forces arise for the pipe breaks at the vessel nozzles from the steam and water which is released into the reactor cavity through the annulus around the broken pipe. The reactor cavity is pressurized asymmetrically with higher pressure on the side of the broken pipe resulting in horizontal forces applied to the reactor vessel. Smaller vertical forces arising from pressure on the bottom of the vessel and the vessel flanges are also applied to the reactor vessel. The cavity pressure analysis is described in Section 6.2.

The internals reaction forces develop from asymmetric pressure distributions inside the reactor vessel. For a vessel inlet nozzle break and pump outlet nozzle break, the depressurization wave path is through the broken loop inlet nozzle and into the region between the 3.9N-46 n

CPSES/FSAR core barrel and reactor vessel. This region is called the downcomer annulus. The initial waves propagate up, down and around the downcomer annulus and up through the fuel. In the case of an RPV outlet nozzle break the wave passes through the RPV outlet nozzle and directly into the upper internals region, depressurizes the core, and enters the downcomer annulus from the bottom of the vessel. Thus, for an outlet 10 Q112.25 nozzle break, the downcomer annulus is depressurized with much smaller differences in pressure horizontally across the core barrel than for the inlet break. For both the inlet and outlet nozzle breaks, the depressuriz(cion waves continue their propagation by reflection and translation through the reactor vessel fluid but the initial depressurization wave has the greatest effect on the loads.

The reactor internals hydraulic pressure transients were calculated including the assumption that the structural motion is coupled with the i

l 3.9N-46a

\

l G 1

CPSES/FSAR matrix, the flexibility matrix is determined. The flexibility matrix is multiplied by the negative of the load vector to determine the network point deflections due to the thermal and boundary force effects. Using the general transfer relationship, the deflections and internal forces are then aetermined at all node points in the system. The support loads are also computed by mutiplying the stiffness matrix by the displacement vector at the support point.

Seismic The models used in the static analyses are modified for use in the dynamic analyses by including the mass characteristics of the piping and equipment.

/

The lumping of the distributed mass of the piping systems is accomplished by locating the total mass at points in the system which will appropriately represent the response of the distributed system.

Effects of the primary equipment motion, that is, reactor vessel, steam Q112.25 generator, reactor coolant pump, and pressurizer, on the Class 1 piping system are obtained by modeling the mass and the stiffness characteristics of the primary equipment and loop piping in the overall system model.

The supports are represented by stiffness matrices in the system model for the dynamic analysis. Shock suppressors which resist rapid motions are also included in the analysis. The solution for the seismic disturbance employs the response spectra method. This method employs the lumped mass technique, linear elastic properties, and the principle

, of modal superposition.

The total response obtained from the seismic analysis consists of two parts: the inertia response of the piping system and the response from differential anchor motions. The stresses resulting from the anchor motions are considered to be secondary and, therefore, are included in the fatigue evaluation.

3.9N-52a

-- __ ._ _ .. -.3._ _ . - .. - .- ._ -

CPSES/FSAR Code is used as a basis for evaluating acceptability of calculated stresses. Both static and alternating stress intensities are considered.

It should be noted that the allowable stresses in Section III of the

ASME Code are based on unirradiated material properties. In view of the fact that irradiation increases the strength of the Type 304 stainless steel used for the internals, although decreasing its elongation, it is considered that use of the allowable stresses in Section III is appropriate and conservative for irradiated internal structures.

The allowable stress limits during the design basis accident used for the co*e support structures are based on the 1974 Edition of the ASME Code for Core Support Structures, Subsection NG, and the Criteria for Faulted Conditions.

In summary, the design and construction of the CPSES core support structures conforms to the requirements of the subsection NG of Section III of the ASME code, except that a) the internals are not " stamped" and b) a specific stress report is note required.

3.9N.6 INSERVICE TESTING 0F PUMPS AND VALVES Refer to Section 3.98.6.

REFERENCES

l. " Documentation of Selected Westinghouse Structural Analysis Computer Codes, WCAP-8252, Revision 1, July 1977.

2. " Sample Analysis of a Class 1 Nuclear Piping System," prepared by ASME Working Group on Piping, ASME Publication, 1972.

3.9N-109

._ . - . _ . - . - _ . -_ .J-,._-_. ._ -- . _ - .-

CPSES/FSAR TABLE 3.9N-1 (Sheet 3 of 4)

Emeraency Conditions

• Occurrences

1. Small loss of coolant accident 5

2. Small steam break 5

3. Complete loss of flow 5 Faulted Conditions *

1. Main reactor coolant pipe break 1 (dynamic effects, resulting from gey, large loss of coolant accident at the RHR, surge and accumulator nozzles, Section 3.6)

2. Large steam break 1

3. Feedwater line break 1

4. Reactor coolant pump locked rotor 1

5. Control rod ejection 1

6. Steam generator tube rupture (included under upset conditions, reactor trip from full power with safety injection)

7. Safe Shutdown Earthquake 1 In accordance with the ASME Nuclear Power Plant Components Code, emergency and faulted conditions are not included in fatigue evaluation.

t

CPSES/FSAR TABLE 3.9N-4 DESIGN LOADING COMBINATIONS FOR ASME CODE CLASS 2 AND 3 COMPONENTS AND COMPONENT SUPPORTS (EXCLUDING PIPING AND PIPE SUPPORTS)*** Rev.

Condition Classification loadino Combination Design and Normal Design pressure design temperature,*

deadweight, nozzle loads **

Upset Upset condition pressure, upset condition metal temperature,* deadweight, OBE, nozzle loads **

Emergency Emergency condition pressure, emergency condition metal temperature,*

deadweight, nozzle loads **

Faulted Faulted condition pressure, faulted condition metal temperature,* deadweight, SSE, nozzle loads **

• Temperature is used to determine allowable stress only.

• • Nozzle loads are those loads associated with the particular plant operating conditions for the component under consideration.

• • • For loading combinations and stress limits for ASME Class 2 and 3 piping, refer to Table 3.98-IB and Table 3.98-1C for Class 2 and 3 supports. Rev.

n

CPSES/FSAR TABLE 3.9N-12 Maximum Reactor Vessel Displacements at Reactor Vessel Centerline

• Maximum Horizontal Maximum Vertical Maximum Displacement Displacement Rotation (inches) (inches) (radians) 144 Square Inch +0.116 +0.095 +0.00057 RPV Inlet 0.0 -0.087 -0.00066 144 Square Inch 0.0 +0.0014 0.0 RPV Dutlet -0.0081 -0.046 -0.0029 Double Ended +0.074 +0.0029 +0.0031 Pump Outlet -0.00064 -0.06 -0.00047 a

• Although the main loop piping breaks are not part of the design basis (excluded as dynamic effects, gey, Section 3.1.1.4), they are conservatively included in the load and stress evaluation tables for the primary component supports.

CPSES/FSAR TABLE 3.9N-13 Maximum Reactor Vessel Support Loads For Postulated Pipe Ruoture Conditions

• LOCA Maximum Vertical Load Per LOCA Maximum Horizontal Load Per Support Including Deadweight Support (Kips) (Kips) 3171 2410 Although the main loop breaks are not part of the design basis (excluded as dynamic effects, Section 3.1.1.4), they are conservatively included in the load and stress evaluation tables for the primary Rev.

component supports.

CPSES/FSAR TARLE 3.9N-14 STEAM GENERATOR LOWER SUPPORT MEMBER STRESSES 0 Member Member Stresses (Percent of Allowable / Loading Condition)

Units (kips)

Normal Upset Faulted Percent Percent Percent Loada Stressed Loada Stressed Loada Break Stressed LS-1 115 4.84 667 28.06 3406 RV0N 92.27 (Bumper)

LS-2 (Beam) 22 --

645 62.77 1070 SGIESe 55.79 d 90.55 LS-3 0 0.0 623 26.19 1134 SGON 42.08 (Bumper)

Column 1 +0 0 +261 21.92 +579 SGONc 43.48

-33 2.20 -304 20.44 -679 SGONc 55.78 Column 2 +0 0 +11 0.88 +471 SGONc 44.36

-324 21.81 -639 43.03 -790 SGONc 69.21 Column 3 +0 0 +32 2.68 +583 X0Lc 34.17

-261 17.58 -527 35.48 -665 X0Lc 45.43 Column 4 +42 3.46 +343 28.83 +817 X0Lc 44,79

-0 0 -274 18.47 -747 SGINc 51.36 o Although the main loop piping breaks are not part of the design basis (excluded as dynamic effects, Section 3.1.1.4), they are conservatively included in the load and stress Re evaluation tables for the primary component supports.

NOTES:

a) (+) = tensions and (-) - compression for column loadings.

c) Includes the effects of main loop jet impingement.

d) The steam generator lateral beam was shown to be91 percent stressed with the attached safety injection line restraints and pipe hangers active.

1 e) Includes the effect of subcompartment pressurization.

u._ ___ ___ __ ____ ___ ~ _ . _ .

CPSES/FSAR TABLE 3.9N-15 STEAM GENERATOR UPPER SUPPORT MEMBER STRESSESO Member Member Stresses (Percent of Allowable / Loading Condition)

Units (kips)

Normal Upset Faulted Percent Percent Percent Load Stressed Load Stressed Load Break Stressed US-1 0 0.00 1229 54.62 2676 MSTL 53.51 (Snubbers)b US-2 7 0.21 1236 36.07 1468 RVIN 37.80 (Bumper)

US-3 7 0.21 1027 29.72 1552 MSTL 39.62 (Bumper)

US-4 7 (a) 1027 (a) 1500 MSTL (a)

(Beam) o Although the main loop piping breaks are not part of the design basis (excluded as dynamic effects, Section 3.1.1.4), they are conservatively included in the load and stress g, evaluation tables for the primary component supports.

NOTES:

a) Scope of evaluation by others.

b) The snubbers were qualified by a 450 kip upset capacity and a 1000 kips faulted capacity, per snubber, in accordance with the snubber stress report.[1]

[1] Paul-Monroe Hydraulics Inc., Report A-690623, Revision 0, entitled "Multiplant II 1000 kip Snubber Stress Report".

1 l

s

CPSES/FSAR TABLE 3.9N-16 REACTOR COOLANT PUMP SUPPORT MEMBER STRESSES

• Member Member Stresses (Percent of Allowable / Loading Condition)

Units (kips)

Normal Upset Faulted Percent Percent Percent Loada Stressed Loada Stressed Loada Break Stressed Tie Rod-A 358 3.11 202 18.00 946 SGON 57.73 Tie Rod-B 2118 18.76 459 40.91 1610 RVIN 98.25 Tie Rod-C 1708 15.10 486 43.32 1094 RCPONd 66.79 Column-1 +0 0 +150 12.57 +850 RCPINb 50.68

-224 14.56 -494 32.11 -798 SGON 45.79 Column-2 +162 13.58 +677 56.93 +965 RCPINb 82.25

-0 0 -412 26.80 -772 RCPINb 79.76 Column-3 +0 0 +340 28.55 +808 RCPINb 69.59

-223 14.47 -757 49.17 -1477 RCPINd 84.73 Although the main loop piping breaks are not part of the design basis (excluded as dynamic effects, Section 3.1.1.4), they are conservatively included in the load and stress Res evaluation tables for the primary component supports.

NOTES:

a) (+) - tensions and (-) - compression for column loadings.

b) Includes the effects of main loop jet impingement.

d) Includes the effect of subcompartment pressurization, e) Normal codition tie rod loads result from over-temperature transients.

a

CPSES/FSAR TABLE 3.9N-19 REACTOR VESSEL SUPPORT LOADS AND STRESSES

• REACTOR VESSEL SUPPORT LOADS Load (kips)

Loading Condition Vertical Tangential Dead Weight 581 --

Thermal 267 11

0. T. Thermal 7 52 Pressure 2 22 OBE 545 404 SSE 970 578 LOCA-1[a],[e] 2430 1790 LOCA-2[a],[f] 1630 1900 Normal [b] 855 84 Upset [c] 1399 488 Faulted-1 [d] 3015 1965 Faulted-2 [d] 2344 2070 REACTOR VESSEL SUPPORT STRESSES Loading Stress Ratio Condition Actual Stress (ksi) Allowable Stress (ksi) Pct of Allowable)

Normal PM - 10.97 Sm, - 23.3 47.1 PM+PB - 10.97 1.SSM - 34.95 31.4 Upset PM - 21.18 Sm - 23.3 90.91 Pg + PB - 21.23 1.5 Sm - 34.95 60.74 Faulted [d] PM - 48.24 0.70 Su = 49.0 98.44 PM+PB - 50.83 1.05 Su - 73.5 69.15

• Although the main loop piping breaks are not part of the design basis (excluded as dynamic effects, Section 3.1.1.4), they are conservatively included in the load and stress evaluation tables for the primary Rev.

component supports,

a. Includes dead weight

b. Dead weight + Thermal + 0.T. Thermal + Pressure

c. Normal + OBE

d. 2 Normal + (SSE2 + LOCA )l/2

e. LOCA-1: Maximum Vertical Load with corresponding Tangential Load

f. LOCA-2: Maximum Tangential Load with corresponding Vertical Load a

STE AM GENER ATOR II REACTOR STEAM GENERATOR PRESSURE $ UPPER SUPPM T VESSEL R E AC IOR C OOLA N T PUMP (D

C R E A C T OR C00'*"I YG PUMP J k I SUPPORT CROSSOVER E S U PPOR T RE RA I i LUMPED MASS CROSS 0W ER LiG RESTR AINI (1)

RCL GLOBAL COORDINATE ,3YSTEM CROSS 0WER LEG RESIRAINT (1)

NOTE (1): Based on CDC 4, leak-before-break application (Section 5.4.14) the loop restraints are not required and may be eliminated.

COMANCHE PE AK S.E.S.

FNAL SAFETY ANALYSIS REPORT UNITS 1 and 2 Reactor Coolant Loop Supports System, Dynamic Structural Model FGJRE 3.9N-1

CPSES/FSM

2. Tho upper lateral support consists of struts cantilevered off the compartment walls that bear against the " seismic lugs" provided on the pressurizer. The configuration of the lateral struts depends on the location of the concrete walls and piping within the compartment, as well as the orientation of the pressurizer.

5.4.14.2.5 Pipe Restraints The application of leak-before-break technology to the CPSES main loop piping has resulted in the exclusion of the main loop piping breaks from the design basis for dynamic effects (Sections 3.1.1-4 and 3.6B).

The reactor coolant loop piping is qualified for LOCA dynamic effects from the loop branch nozzle breaks (breaks 9,10 and 11, Table 3.6B-2) Rev.

without the main loop piping restraints (Figures 3.68-89 to 92).

These piping restraints, illustrated in Figures 5.4-16 to 5.4-19, are therefore not required for CPSES Units 1 and 2. These p ye restraints may be eliminated.

5.4- as i

I - _ _ _ _ _ _ _ . ,__

CPSES/FSAR 5.4.14.3 Evaluation Detailed evaluation ensures the design adequacy and structural integrity of the reactor coolant loop and the primary equipment supports system. This detailed evaluation is made by comparing the analytical results with established criteria for acceptability.

Structural analyses are performed to demonstrate design adequacy for safety and reliability of the plant in case of a large or small seismic disturbance and/or LOCA conditions. Loads which the system is expected to encounter often during its lifetime (thermal, weight, pressure) are applied and stresses are compared to allowable values as described in Section 3.9N.1.4.7.

The Safe Shutdown Earthquake and design basis LOCA resulting in a rapid depressurization of the system are required design conditions for 5.4-86

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(See Section 5.4.14.2.5) FNAL SAFETY ANALYSIS REPORT UNITS 1 and 2 Typical Crossover Leg Restraint FGURE 5.4-16

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NOTE: Hot and Cold Leg Restraint may be eliminated (See Section 5.4.14.2.5) FNAL SAFETY ANALYSIS REPORT UNITS 1 and 2 i

' TYPICAL LATERAL RESTRAINTS FIGURE 5.4-19 I

i i

CPSES/FSAR Pressurizer:

1528 - Same As Above Piping:

1432 Wrought Type 304 and 316 with Nitrogen Added, Sections I, III, VIII, Division I and 2 N-122 - January 21,1985 " Stress Indices for Integral Structural Attachments, Class 1".

N-391 - November 28, 1983 " Procedure for Evaluation of the Rev.

Design of Hollow Circular Cross Section Welded Attachments on Class 1 Piping".

N-397 - February 20,1984 " Alternative Rules to the Spectral Broadening Procedures of N-1226.3 for Class 1, 2 and 3 Piping".

N-411 - September 17,1984 " Alternative Damping Values for Seismic Analysis of Class 1, 2 and 3 Piping Systems".

Flux Thimble Tubing:

1612 - Use of Type 308 Stainless Steel Rod and Bar for Section III, Class 1, 2, 3 and CS construction Valves:

1649 - Modified SA 453-GR 660 for Class 1, 2, 3 and CS Construction 1553 Upset Heading and Roll Threading of SA-453 for High Temperature Bolting,Section III, Classes 1, 2, 3 and MC Q005-5 r

CPSES/FSAR Reactor Vessel Bottom-Mounted Instrumentation:

N-378 - Seal Table Material 59 Q005-5a

~

CPSES/FSAR allowable values.

6. Demonstrate that active components will perform their safety function when subjected to the combined loads resulting from the loss-of-coolant accident and the safe shutdown earthquake.

7. Demonstrate the functional capability of any essential piping when subjected to the combined loads resulting from the loss-of-coolant accident and the safe shutdown earthquake.

R112.25 1. Refer to existing drawings in sections 5.4 and 3.8 for Reactor Vessel Supports Systems arrangement drawings.

2. A specific analysis is not required. Refer to Rev.

Section 3.9N.I.4.6.

3. Refer to revised sections 3.9N.l.4.3, 3.9N.l.4.4 and 3.9N.1.4.6.

4. An evaluation is not required. Refer to Section Rev.

3.9N.1.4.6.

5. Refer to sections 3.9N.I.4.3, 3.9N.l.4.4, 3.9N.1.4.5, 3.9N.1.4.6, 3.9N.l.4.8, and 3.9N.2.5.

6. Refer to section 3.9N.3.2.

7. Refer to section 3.98.3.1.2 lRev.

i l

112.38 s

CPSES/FSAR TABLE 17A-1 SHEET 3 LIST OF OUALITY ASSURED STRUCTURES, SYSTEMS AND COMPONENTS Applicable Safety Code or Code Seismic Quality Re ference System and Components Class (7) Standard (12) Class Category Assurance Section Remark s j Refisctive insulation assemblies NNS Mfrs Stds - II Note 27,B 6.lB, 6.2.2 Note 13d (except for the portions installed on the RCS cold leg and hot leg

, pipes which are located inside the I

biological shield tunnels)

• Reactor vessel noriles non. crushable 1 Mfrs Stds -

I Note 17,A 3. 9N. I .4. 6 Note 13c, 64 Rev.

insulation Reflective insulation assemblies on NNS Mfrs Stds - NONE MONE 6.18 Note 139 RCS cold leg and hot leg pipes and located in biological shield tunnels Supports for Class I piping 1 ASME III 1 1 Note 27,A 3.9N Note 13c

{ Supports for Class 2 piping 2 ASME III 2 1 Note 27,A 3.9B Note 13c Supports for Class 5 piping NNS ANSI B31.1 - 11 Note 44,8 3.7B Note 13a, 13e

2. Chemical and Volume Control System (CVCS)

Regenerative heat exchanger 2 ASME III 2 I Note 3.A 9.3.4 Note 13a letdown Ikat Exchanger Tube side 2 ASME III 2 I Note 3,A 9.3.4 Note 13a Shell side 3 ASME Ill 3 I Note 3,A 9.3.4 Note Ic, 2, 13a Mixed-bed demineralizer 3 ASME III 3 NONE Note 4,A 9.3.4 Note 11, 13g Cation. bed demineralizer 3 ASME III 3 NOME Note 4 A 9.3.4 Note 11,13g Reactor coolant filter 2 ASME Ill 2 I Note 3,A 9.3.4 Note 13a Volme control tank 2 ASME Ill 2 I Note 3,A 9.3.4 Note 13a Centrifugal charging pump 2 ASME III 2 1 Note 3,A 9.3.4 Note 13b, la and Id, 2 Positive displacement pump 2 ASME 111 2 I Note 3,A 9.3.4 Note 13b, la and Ic, 2 FEBRUARY 10, 1984

. CPSES/FSAR TABLE 17A-1 SHEET 4g LIST OF QUALITY ASSURED STRUCTURES. SYSTEMS AND COMPONENTS

54. Fire dampers are functionally part of fire barriers but are shown on ventilation flow diagrams because they are physically located in ventilation ducts and penetrations. The safety class designation on ventilation 41 diagrams does not apply to the fire dampers. Fire dampers which are required to remain open after an SSE are designated seismic Category II by the specification and are qualified to remain open.

, s.

55. The tank boundary extends to the first weld connecting the penetration nozzles to system piping outside the tank. The tank and associated piping inside the tank is not N-stamped.

56. The application of IEEE-323 is limited to those channels identified in FSAR Section 7.5 as Category 2 radiation monitors.

57 The Class 5 piping and valves in these 2" and under seismic Category !! lines are seismically analysed and are supported by seismic Category Il supports. Installation requirements are the same as other Class 5. seismic Category II pipes.

58. Excluding buried yard piping.

59. QA for water extinguishers limited to UL listing.

60. Excluding Turbine Buldling.

61. The control panel and associated electrical components of the turbine driver are associated Class IE located in mild environment.

62. This component has been qualified as seismic category I by analysis. 55

63. App 1tcable where wall is supported by structural steel specified seismically. 59 Application of modified CDC-4 (leak-before-break) removes the need for this insulation. The insulation may

64. Rev.

be removed.

1 AMENDMENT 59 JUNE 1, 1986

TXX-6086 November 10, 1986 Page 1 of 7 Attachment 2 ATTACHMENT 2 TO TXX-6086 0F NOVEMBER 10, 1986 DESCRIPTION OF FSAR REVISIONS j

i l

l 1

I

TXX-6086 November 10, 1986 Page 2 of 7 This attachment includes a page by page description of the FSAR pages that will be revised to implement the modified GDC-4 for CPSES Units 1 and 2.

Page 3 of this attachment provides the technical bases and a summary description of the FSAR revisions that relate directly to the modified GDC-4.

The remaining pages in this attachment provide a page-by-page description of the FSAR revisions. For each page in which the revision relates to the discussion on page 3, the description for that page will include the identifier "GDC-4".

Pages included in Attachment 1, but not described in this attachment, contain no changes and are provided for clarity only.

i

TXX-6086 November 10, 1986 Page 3 of 7 Attachment 2 Technical Basis and Summary

Description:

The general leak-before-break technology described in NUREG 1061 Volume 3 has been applied to the CPSES Units 1 and 2 RCS main loop piping. This application of leak-before-break, discussed in WCAP-10527, shows that the RCS main loop piping breaks should not be considered in the design basis of CPSES.

This conclusion results from detailed analyses of the following factors:

a. The loads, material properties, transients, and geometry of Units 1 and 2 RCS primary loop are enveloped by generic Westinghouse leak-before-break analyses.

b. Stress corrosion cracking is precluded by use of fracture resistant materials in the piping systems and controls on reactor coolant chemistry, temperature, pressure, and flow during normal operation.

c. Water hammer is precluded in the RCS primary loop piping because of system 'esign, testing, and operational considerations.

d. The effects of low and high cycle fatigue on the integrity of the primary piping are negligible.

e. Ample margin exists between the leak rate of the reference flaw and the criteria of Reg. Guide 1.45.

f. Ample margin exists between the reference flaw chosen for leak detectability and the " critical" flaw.

g. Ample margin exists in the material properties used to demonstrate end-of-life (relative to aging) stability of the reference flaw.

The reference flaw will be stable throughout reactor life because of the ample margins, in e, f, and g, above, and will leak at a detectable rate which will assure a safe plant shutdown.

Although RCS main loop piping breaks are not considered in the structural design basis, GDC 4 requires that they be included in the design of the reactor containment, the ECCS and the environmental qualification.

The exclusion of the dynamic effects associated with the main loop piping breaks from the design of CPSES, permits the elimination of some or all of the main loop pipe whip restraints (crossover leg restraints, hot leg restraint, vessel nozzles restraints) indicated in FSAR Figures 5.4-16 to 5.4-19. The elimination of the main loop whip restraints will reduce congestion and therefore help reduce worker radiation exposures during routine plant inspection and maintenance activities.

The main loop piping and component supports analyses and qualification, as well as that of the main loop branch lines, will be based on the remaining branch nozzle breaks (RHR suction, pressurizer surge and accumulator injection) without the main loop whip restraints.

TXX-6086 November 11, 1986 Page 4 of 7 Attachment 2 Table GDC-4 States that the dynamic effects of RC loop pipe breaks are 1.3-1 excluded from the CPSES design basis. Provides reference to Section 3.1.1.4.

Table GDC-4 Lists FSAR Sections where leak-before-break /GDC-4 1.3.2 methodology has been used to exclude the dynamic effects of RC loop pipe breaks. (Sheet 7) Lists the WCAP references for the CPSES leak-before-break analyses.

(Sheet 17)

Table GDC-4 Lists the WCAP references for the CPSES leak-before-break 1.6-1 analyses.

3.1-6 GDC-4 Revised quoted version of GDC-4. States the new design basis.

3.1-11 GDC-4 States the new design basis as it relates to GDC-4.

3.6B-6 GDC-4 States the new design basis under Reactor Coolant Loop 3.6B-7 GDC-4 Piping discussion.

3.68-15 GDC-4 States new design basis and gives a summary of the CPSES 3.68-16 GDC-4 leak-before-break analysis.

3.6B-17 Clarification: Added a specific reference to denote that"RC system high energy piping breaks were postulated per the ASME Code 1977 Edition through Summer of 1979. The previous version implied that these lines were 1974 Edition through Winter 1975.

3.68-32 GDC-4 Restates new design basis as it affects jet loads.

3.6B-44 GDC-4 Defines how the new design basis is implemented with respect to designated breaks.

3.6B-45 GDC-4 Discusses how dynamic loads are now addressed for ASME Class 1 components and supports. Deletes a previous discussion on pipe whip restraints that is no longer required.

3.6B-71 GDC-4 Adds leak-before-break WCAP analysis and NUREG 1061 to FSAR references.

Table Update: Adds note 30 to indicate that although RHR piping between 3.6B-1 two normally closed isolation valves is considered high energy, the postulated break is assumed to be on the RCS side of the valve nearest to the RCS. This is conservative since there is no continuing source of energy if a break is postulated between the isolation valves.

Table GDC-4 Defines how the RC loop break locations are incorporated 3.6B-2 into the various analyses.

1 Figure GDC-4 Same as Table 3.68-2.

3.68-9

l l

TXX-6087 November 10, 1986 Page 5 of 7 Attachment 2 l

Figure Update: Revises figure to delete the previous notation that 3.68-64 indicated that the RHR pipe between the isolation valves )

was considered moderate energy piping. (See Table 3.6B-1)

Figures GDC-4 Notes that whip restraints and non-crushable insulation 3.6B-89 may be removed.

Thru 3.68-92 3.7N-27 Clarification: The maximum resultant building seismic motion is considered at each support point. The previous words, "on a made by mode basis", are clarified to state that the calculation considers the relative phase (in-phase or out-4 of-phase) of the seismic motion of each support.

3.7N-27 Clarification: The words "most severe" are replaced by " envelope" because the envelope of all applicable response spectra is used. The words "most severe" imply the use of a single

" worst" spectrum, whereas for different frequencies different spectra may be more severe.

Table Update: Note "b" added to allow use of Code Case N-411 for piping 3.7N-1 analyzed by the response spectrum method as approved by NRC letter of March 18, 1986.

3.9N-24 Correction: Computer code name WESTDYN-7 has been changed to WESTDYN. Also the word " redundant" has been removed from the code description because the code also applies to non-redundant systems.

3.9N-26 GDC-4 3.9N-27 GDC-4 Discusses how pipe break blowdown loads are considered.

3.9N-28 GDC-4 3.9N-29 Correction: Same as 3.9N-24 3.9N-32 Clarification: A reference is added to show that " closely spaced modes" were considered and provides FSAR section where it is discussed.

GDC-4 Discusses how stress analysis is performed.

3.9N-33 Correction: Same as 3.9N-24 3.9N-33 GDC-4 Defines design basis for vessel motion and describes how 3.9N-34 it was analyzed for CPSES.

3.9N-37 Clarification: The previous statement regarding eight permutations was confusing and has been changed to simply state that seismic loads can be either positive or negative. The most conservative seismic combination is used.

3.9N-37 Correction: Same as 3.9N-24.

TXX-6086 November 10, 1986 Page 6 of 7 3.9N-38 GDC-4 Defines design basis for primary component supports and 3.9N-39 GDC-4 describes how they were analyzed for CPSES. Deletes a discussion on pipe restraints which is no longer required.

3.9N-44 GDC-4 Defines the design basis for the reactor vessel and 3.9N-45 GDC-4 describes how that analysis was performed for CPSES.

3.9N-46 GDC-4 3.9N-52a Editorial only.

3.9N-109 Editorial only.

Table GDC-4 Updates the description of the Faulted Condition based on 3.9N-1 the new design basis.

Table Clarification: The Table was clarified to show that it does not 3.9N-4 address piping and to show where information can be found for Class 2 and 3 piping and pipe supports.

Table GDC-4 Defines design basis for primary component supports and 3.9N-12 describes how they were analyzed for CPSES.

Table GDC-4 See Table 3.9N-12.

3.9N-13 Table GDC-4 See Table 3.9N-12.

3.9N-14 Table GDC-4 See Table 3.9N-12.

3.9N-15 Table GDC-4 See Table 3.9N-12.

3.9N-16 Table GDC-4 See Table 3.9N-12.

3.9N-19 Figure GDC-4 Identified the piping analysis result whereby the shown 3.9N-1 loop restraint (s) may be eliminated.

5.4-85 GDC-4 See Figure 3.9N-1.

Figures GDC-4 See Figure 3.9N-1.

5.4-16 Thru 5.4-19

TXX-6086 November 10, 1986 Page 7 of 7 Q005-5 Update: This section has been updated to include additional Code Cases used in the requalification of Class 1 piping.

o Code Case N-122 was first approved by the NRC In R.G. 1.84 dated August 1976. The January 21, 1985 version of the Code Case remains approved in Revision 24 of R.G. 1.84.

This Code Case provides guidance (not contained in the Code) for the analysis of structural attachments.

o Code Case N-391 was first approved by the NRC in Revision 23 to R.G. 1.84. This Code Case provides guidance for evaluation of welded attachments.

o Code Case N-397 was specifically approved for use at CPSES by NRC letter from V.S. Noonan to W.C. Council dated March 18, 1986. This Code Case provides alternate rules for spectral broadening for Class 1, 2 and 3 piping.

o Code Case N-411 was specifically approved for use at CPSES by the above letter. This Code Case provides alternate damping values for the seismic analysis of Class 1, 2 and 3 piping systems.

Q112.25 GDC-4 Updates the response to this question to incorporate the new design basis.

Table GDC-4 Indicates that the reactor cavity non-crushable insulation 17A is no longer required and may be eliminated.

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