NUREG-0609

Asymmetric Blowdown Loads on PWR

Primary Systems

Resolution of Generic Task Action Plan A-2

Manuscript Completed: November 1980

Date Published: January 1981

S. B. Hosford, Task Manager, R. Mattu,

R. 0. Meyer, E. D. Throm, C. G. Tinkler

Division of Safety Technology

Office of Nuclear Reactor Regulation

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555

ABSTRACT

The NRC staff, after being informed of newly identified asymmetric loadings

resulting from postulated ruptures of primary piping, initiated a generic

investigation, Task Action Plan A-2. This investigation was limited to

pressurized-water-reactor (PWR) plants because of their higher primary system

pressures and potential for larger asymmetric loads. The intent of the investigation was to develop a better understanding of this complicated phenomenon

and then to develop acceptable criteria and guidelines for the staff to use

for evaluating plant analyses.

The staff has completed its investigation and concludes that an acceptable

basis is provided in this report for performing and reviewing plant analyses

for asymmetric LOCA (loss-of-coolant accident) loads. Guidelines and criteria

were developed for the following areas: evaluation of the loading transients,

evaluation of the structural components, and evaluation of the fuel assembly.

The staff reviewed and approved computer programs and modeling techniques

submitted by each PWR nuclear-steam-supply-system vendor for development of

the subcooled blowdown- and cavity-pressure components of the loading transient.

Audit models were developed to evaluate the structural computer programs and

modeling techniques. The audit evaluation has been completed and the methods

have been approved for the structural-analysis method submitted by Westinghouse

for the Indian Point Unit 3 plant. Structural-analysis methods developed by

Babcock and Wilcox and Combustion Engineering will be evaluated when their

plant analyses are submitted to NRC. Criteria and guidelines are provided to

perform a detailed evaluation of the fuel assembly. Acceptance criteria are

also provided so deformed fuel-assembly spacer grids may be evaluated.

iii

CONTENTS

Page

ABSTRACT ......................................................... iii

1 INTRODUCTION ................................................ 1

2 BACKGROUND .................................................. 2

3 DEVELOPMENT OF LOADING FUNCTIONS ............................ 4

3.1 Internal Subcooled Blowdown Loads ...................... 7

3.2 Cavity-Pressure Loads .................................. 13

4 STRUCTURAL EVALUATION ....................................... 24

4.1 Primary-System Analytical Development .................. 24

4.2 Structural Acceptance Criteria ......................... 33

5 FUEL-ASSEMBLY EVALUATION .................................... 35

REFERENCES ....................................................... 36

APPENDIX A Task Action Plan A-2 .............................. A-i

APPENDIX B Status of of Operating Plant Evaluation ........... B-1

APPENDIX C Characteristics of Pipe Breaks .................... C-1

APPENDIX D Letters to Applicants/Licensees ................... D-1

APPENDIX E Fuel-System Design (Appendix A to SRP 4.2) ........ E-1

LIST OF FIGURES

1. Pressurization of Reactor Annulus ........................... 5

2. Example of Asymmetrical Internals Load ...................... 6

3. Westinghouse Three-Loop PWR Loads Evaluation, Cold-Leg

Nozzle Break ................................................ 8

4. Uncoupled Versus Coupled Fluid-Structure Response ........... 10

5. Sample Nodal Model for a Reactor Cavity ..................... 18

6. Steam-Generator Support ..................................... 28

7. Typical Piping-Loop Model ................................... 29

8. Reactor-Pressure Vessel and Internals Model ................. 30

9. Core Model .................................................. 31

V.
ABBREVIATIONS

ACRS Advisory Committee on Reactor Safeguards

B&W Babcock and Wilcox

BWR boiling-water reactor

CE Combustion Engineering

CFR Code of Federal Regulations

CP construction permit

DOR Division of Operating Reactors, NRC

DSS Division of Systems Safety

EB Engineering Branch, DOR

ECCS emergency-core-cooling system

EDO Office of the Executive Director for Operations

EPN equivalent pipe networks

HDR superheated steam reactor

HEM homogeneous equilibrium

ISI inservice inspection

LOCA loss-of-coolant accident

NRC U.S. Nuclear Regulatory Commission

NSSS nuclear-steam-supply system

O.D. outside diameter

OL operating license

PWR pressurized-water reactor

RPV reactor pressure vessel

SAI Science Applications, Inc.

SRP Standard Review Plan

SRSS square root of the sum of squares

SSE safe shutdown earthquake

TAP task action plan

W Westinghouse Electric Coporation

vi
ASYMMETRIC BLOWDOWN LOADS ON PWR PRIMARY SYSTEMS:

RESOLUTION OF GENERIC TASK ACTION PLAN A-2

1 INTRODUCTION

The NRC staff was informed on May 7, 1975 of newly identified asymmetric

loadings which could act on the reactor's primary system. These asymmetric

loadings result from a postulated double-ended rupture of the piping in the

primary coolant system. The magnitude of these loads is potentially large

enough to damage the supports of the reactor vessel, the reactor internals,

and other primary components of the system. Because this newly identified

phenomenon was complicated and not well understood at that time, the staff

initiated a generic study, Task Action Plan (TAP) A-2, to gain a better understanding of these loads and to develop criteria for an evaluation of the

response of the primary systems in pressurized-water reactors (PWRs) to

these loads.

The staff investigation, TAP A-2, was limited to primary systems of PWRs.

Although similar loads associated with a postulated rupture of piping in

primary systems in boiling-water reactors (BWRs) are expected to occur, the

overall safety significance is considered to be much less because of the lower

operating pressures in primary systems in BWRs. If necessary, a staff program

to resolve this issue in BWR plants will be considered, once the evaluation of

PWR plants is completed.

This investigation was concerned primarily with the loading phenomenon

resulting from a postulated rupture of the piping in a primary loop. Although

they act on the primary system, additional loadings, such as thermal and

seismic, are not the object of this investigation nor are they included in

this report. The results of the asymmetric loss-of-coolant accident (LOCA)

evaluation must be combined with other loading responses, as appropriate, by

applicable staff-design requirements.

Concurrent with the staff investigation, licensees of PWR plants were notified

by letter (see Appendix D) that an evaluation of their primary system for

asymmetric LOCA loads would be required.

The staff investigation (TAP A-2) has been completed. Criteria and guidance

for conducting an evaluation for asymmetric LOCA loads were developed in this

study and are included in this report. In addition, the staff has provided an

acceptable basis (provided in this report) for the evaluation and review of

the response of PWR primary systems to asymmetric LOCA loads.

The report is organized in the same way as is a typical evaluation or review.

Section 2 provides a detailed discussion about the nature of these loads on

the primary system, as well as a chronology of staff actions. Section 3

offers a detailed discussion of the analytical development of these loads. In

Section 4 a discussion of the structural analysis is found, and Section 5

details the fuel-assembly evaluation.

1
2 BACKGROUND

On May 7, 1975, NRC was informed by the Virginia Electric and Power Company

that in the original design of the support system for reactor vessels for

North Anna Units 1 and 2, neither Westinghouse nor Stone and Webster had

considered an asymmetric loading on the reactor vessel supports resulting

from a postulated rupture of reactor-coolant piping at a specific location

(for example, the nozzle of the reactor vessel). In the event of a postulated instantaneous, double-ended, offset rupture at the vessel's nozzle,

asymmetric loading could result from forces induced on the reactor internals

by transient differential pressures across the core barrel and by forces on

the vessel caused by transient differential pressures in the reactor cavity.

With the advent of more-sophisticated computer programs and the development

of more-detailed analytical models, it became apparent that such differential

pressures, although of short duration, could place a significant load on the

supports of the reactor vessel and on other components, possibly affecting

their integrity. This potential safety concern, first identified during the

review of the North Anna facilities, was determined to have generic implications for all PWRs.

Upon closer examination of this situation, it appeared that postulated breaks

in a reactor-coolant pipe at nozzles of reactor pressure vessels (RPVs) were

not the only area of concern, but that other pipe breaks in the reactor coolant

system could cause internal and external transient loads to act upon both the

reactor vessel and other components. For the postulated pipe break in the

cold leg, asymmetric pressure changes could take place in the annulus between

the core barrel and the vessel. Decompression could take place on the side of

the RPV annulus nearest the pipe break before the pressure on the opposite

side of the RPV changed. This momentary differential pressure across the core

barrel could induce lateral loads both on the core barrel and on the reactor

vessel. Vertical loads could also be applied to the core internals and to the

vessel because of the vertical flow resistance through the core and asymmetric axial decompression of the vessel. Simultaneously, for breaks in RPV

nozzles, the annulus between the reactor and biological shield wall could

become asymmetrically pressurized, resulting in a differential pressure across

the reactor vessel causing additional horizontal and vertical external loads

on the reactor vessel. In addition, the reactor vessel could be loaded

simultaneously by the effects of strain-energy release and blowdown thrust

at the pipe break. For breaks at reactor vessel outlets, the same type of

loadings could occur, but the internal loads would be predominantly vertical

because of the more-rapid decompression of the upper plenum. Similar asymmetric forces could also be generated by postulated pipe breaks located at

the steam generator and reactor-coolant pump. Therefore, the scope of this

problem was not limited to the vessel support system itself.

On October 15, 1975, the NRC staff sent out letters surveying all of the

operating PWRs (Ref. 1) and determined from the survey responses that these

asymmetric loads had not been considered in the design of any PWR primary

systems.

In June 1976, the NRC requested (Ref. 2) all operating PWR licensees to

evaluate the adequacy of reactor system components and supports at their

facilities, with respect to these newly identified loads.

.2
In response to our request, most licensees with Westinghouse plants proposed

to augment the inservice inspection (ISI) (Ref. 3) of the weld area between

reactor-vessel nozzle and safe-end in lieu of providing an evaluation of

postulated piping failures. In September 1976, licensees with Combustion

Engineering (CE) plants submitted a probability study (prepared by Science

Applications, Inc. (SAI)) (Ref. 4) in support of their conclusion that a

pipe break at a particular location (vessel nozzle) was so unlikely to occur

that no further analysis was necessary. A similar study was conducted by

SAI for Babcock and Wilcox (B&W) plants in September 1977 (Ref. 5). When

the Westinghouse and (Ref. 3) CE owners-group reports were received in

September 1976, NRC formed a special review task group to evaluate the

alternative proposals. In addition, EG&G Idaho, Inc., was contracted to

perform an independent review of the SAI probability study submitted for

the CE owners group.

This review effort resulted in a substantial number of questions which have

been provided to representatives of each group. Based on the nature of these

questions, the NRC staff did not accept those reports as resolving the

asymmetric LOCA-load generic issue. Based on NRC review, the staff concluded

that a sufficient data base did not exist within the nuclear industry to provide satisfactory answers. Several long-term experimental programs would be

required to provide much of this information. Although the probability study

submitted in September 1977 by SAI (Ref. 5) for certain B&W owners did respond

to some of these questions, NRC staff determined that the more fundamental

questions still remained unanswered.

A second and equally important reason for not accepting probability/ISI

approaches as a solution concerned our need and industry's need to obtain a

better understanding of the problem. We considered it essential that an

understanding of the important breaks and associated consequences be known

before applying any remedy--be it pipe restraints, probability, ISI, or some

combination of these measures.

Upon reaching these conclusions, NRC staff initiated an inhouse investigation

of this phenomenon (TAP A-2) (see Appendix A) and notified each PWR licensee

(January 25, 1978) (see Appendix D) of NRC's requirement to have an evaluation

of their facilities for these loads.

3
3 DEVELOPMENT OF LOADING FUNCTIONS

The first consideration in conducting an evaluation of the primary system for

LOCA loads is to define and develop the loading transients associated with the

postulated pipe rupture. The loading transients depend on the postulated

break conditions, the postulated break locations, and on the operating conditions and geometry of the system. Typical loads acting on the reactor

pressure vessel and internals are illustrated in Figures 1 and 2. A summary

of the loading components follows.

(1) Subcooled Blowdown Loads--Dynamic hydraulic forces occur throughout the

primary system due to the rapid subcooled depressurization of the system.

These forces occur in the piping and component internals throughout the

system.

(2) Cavity-Pressure Loads--A rupture of the primary-loop piping at a component nozzle can pressurize the cavity between the vessel (reactorpressure vessel, steam generator, or reactor-coolant pump) and the cavity

wall. Lateral and vertical forces are exerted on the component from this

cavity pressurization.

(3) Jet Impingement and Jet-Thrust Loads--Jet impingement and thrust loads

also act on the primary system at the break location.

(4) Strain-Energy-Release Loads--The pipe rupture is postulated to occur

during normal operating conditions when the strain energy has built up in

the system. Some of this strain energy is released when the rupture

occurs.

Typically, the strain-energy-release loads are determined by applying the

steady-state, normally operating, hydraulic loads and the forces that

result from the weight of the structure in a static analysis of the

intact system. For this static analysis, snubbers are assumed to be

inactive. The steady-state displacements from this static analysis are

then used as initial conditions for the dynamic analysis where the broken

leg is disconnected from the vessel and allowed to move, and the snubbers

are active.

All of the loading conditions discussed must be included in the evaluation of

the primary system. These loads provide the information needed to evaluate

the dynamic response of the supports of the reactor-pressure vessel, the

reactor-core structures, and the primary-coolant loop and its loop components.

A detailed discussion of the subcooled blowdown (hydraulic) and cavity loads

is provided in Sections 3.1 and 3.2.

4
;. M76.

Figure 1. Pressurization of Reactor Annulus.

5
Upper

plenum

bc

Zone of

reduced pressure

Failure close to

the inlet nozzle

of the primary

coolant circuit

Figure 2. Example of Asymmetrical Internals Load.

6
3.1 Internal Subcooled Blowdown Loads

This section describes the procedures currently used by NRC to evaluate the

methods utilized in a submittal for the calculation of the hydrodynamic loadings on a reactor-coolant system undergoing a postulated LOCA.

Development of these loads involves several steps:

(1) The first step in the development of these loads is the determination of

the transient pressure and velocity distribution throughout the system,

using an appropriate thermal-hydraulic, computer program.

(2) The second step involves the conversion of these transient pressure and

velocity data into equivalent transient forces throughout the primary

system. The transient forces developed in these two steps are then

included, along with the remaining LOCA loads, in a time-history structural analysis of the primary system.

The staff has reviewed and approved Westinghouse (Ref. 6), Babcock and Wilcox

(Ref. 7), and Combustion Engineering (Ref. 8) topical reports for the development of these hydrodynamic loads. The procedures and acceptance criteria used

by the staff for those reviews are discussed in Sections 3.1.1 through 3.1.4.

3.1.1 General Review Procedures

NRC staff review covers the analytical methods used by the licensee for the

evaluation of the PWR primary-coolant system subjected to a LOCA with emphasis

on the subcooled, transition, and early saturation periods of the decompression transient. It is during this timespan, on the order of 0.05 to 0.20

seconds, that the induced hydraulic loads are greatest. The analytical methods

used must account for the acoustic wave phenomenon associated with the subcooled blowdown, the transmission, and the reflection of pressure waves

throughout the system.

Modeling of the vessel's downcomer region is of major importance to this

review. The application of current, one-dimensional computer programs to the

calculation of the multidimensional pressure field of the downcomer annulus is

given significant attention. It is this pressure field which has the greatest

influence on the lateral loads on the reactor vessel, core-support barrel, and

fuel assemblies.

Figure 3 provides a summary of the effects of modeling on the predicted

hydraulic loads. A Westinghouse three-loop plant was analyzed using the NRC

staff program WHAM/MOD-007. Two models were used; one was similar to the

Westinghouse model, and the other was developed by the staff for coarse- and

fine-noding comparisons, respectively. Also shown in Figure 3 is the effect

on the calculated results of increasing the break-opening time.

In principle, the effects of a sudden decompression of a PWR primary system

can be determined by a straightforward calculation considering the LOCA

pressure wave and its resulting interaction with internal structures of the

7
m

.13

0

0

SOLID LINES ARE FIT TO WESTINGHOUSE MODEL

DASHED LINES ARE FIT TO STAFF MODEL

BREAK AT COLD LEG NOZZLE

to'

-1B

Up

- -" LEGEND

0= WESTINGHOUSE MODEL 1 MSEC BREAK

SA= WESTINGHOUSE MODEL 10 MSEC BREAK

0 = STfiPF MODEL I MSEC BREAK

V = STAFF MODEL 10 MSEC BREAK

= MULTIFLEX DATA 1 MSEC BREAK

( = MULTI FLEX DATA 10 MSEC BREAK

M=MULTIFLEX RIGID WALL DATA o A

0 100 200

BREAK

300 400

SIZE (SQ.IN.)

500 600

Figure 3. Westinghouse Three-Loop PWR Loads Evaluation,

Cold-Leg Nozzle Break.

8
reactor. However, the physical process is not as straightforward. The resulting motion of the core-support barrel to the decompression wave will feed back

into the local fluid pressure through the compressibility of the water. The

resulting hydrodynamic loads are reduced when this fluid-structure coupling is

considered.

lhe SOLA-FLX computer program is being developed to supply the staff with a

tool for evaluating the coupled system. Figure 4, taken from Reference 9,

shows the effect of the fluid-structure-coupled calculation. This figure

illustrates the reduced loads associated with a coupled system and also shows

the effect of modeling the downcomer width. The results show that the rigidwall assumption yields higher differential pressure, hence higher peak loads,

with a higher frequency response than results in the coupled system; the

decompression wave travels faster in the uncoupled case. The thickness of the

downcomer annulus can be represented as a single cell and still yield a converged solution. The analytical development of a coupled fluid-structure

solution, if employed by the licensee, must also be reviewed.

The review procedure can be broken down into three major categories:

(1) analytical development, (2) application and system modeling, and

(3) computer-program verification. These items are addressed below in

Sections 3.1.2, 3.1.3, and 3.1.4, respectively.

3.1.2 Analytical Development

The staff typically reviews a number of areas as part of the licensee's analytical development of the thermal-hydraulic computer program used for the

evaluation of the subcooled blowdown loads. This review begins with the basic

conservation equations of mass, momentum, and energy, and with the equation of

state.

The assumption generally used for these analyses is homogeneity of the steam/

water mixture. The subcooled, transition, and early saturated regimes of the

transient, where the homogeneous-fluid assumption is acceptable, are evaluated.

The possibility of potential nonequilibrium effects is also considered in the

review.

The hydrodynamic loads are directly proportional to the differential pressures

applied to a structure. If, as a result of nonequilibrium, the pressure at

the break plane falls below the saturation pressure of the fluid, prior to the

development of two-phase-flow conditions, an increased loading could result.

The effects of nonequilibrium would increase the loading on the core support

barrel structure if this pressure "undershoot" is transmitted to the downcomerannulus region.

The equation of state, the relationship between the fluid pressure and density,

and the sonic or acoustic wave speed are also reviewed. The range of pressures

from operational pressure to the saturation pressure should be adequately

described.

The assumptions used to reduce the basic conservation equations to the form

used by the computer program are evaluated, with respect to the problem being

9
DI STANCE (CM)

Figure 4. Uncoupled Versus Coupled Fluid-Structure Response,

1/25 Scale PWR Geometry at 240 Microseconds.

10
solved. The review includes the solution schemes used, convergence criteria

employed, steady-state initialization, and the demonstration that conservation

of mass, momentum, and energy are achieved. This last item is of primary

importance when a one-dimensional computer is used to model a multidimensional

region.

For those computer programs which include the fluid-structure coupling, the

method of incorporating the moving boundary into the conservation equations is

reviewed.

The modeling of system components, such as the heat transfer in the reactor

core and steam generator, the pressurizer, the reactor-coolant pumps, and

nonrecoverable energy losses are also considered.

The essential boundary conditions reviewed are the break-opening time and area

characteristics. The history of the break-opening time and the break area in

the nuclear-steam-supply system (NSSS) primary piping system are plant specific

and depend on such considerations as the location of the break, the stiffness

and mass of the piping system involved, and the type and location of pipe

restraints/supports being used. The break geometry postulated for this evaluation is a complete circumferential break of the primary coolant pipe. Appendix C of this report discusses characteristics of pipe breaks in more detail.

The discharge-flow model is an important aspect of this review. The dischargeflow model for the postulated break is the system-forcing function. As such,

the treatment of the subcooled critical flow and potential nonequilibrium

effects must be properly accounted for in the development of the discharge-flow

model.

3.1.3 Application and System Modeling

After the basic solution technique has been selected and equations have been

developed, the next stage of the review is oriented to using the computer

program to evaluate the decompression transient. The model of the system used

to track transmission and reflection of the acoustic waves is reviewed.

The most significant aspect of the review is the modeling of the asymmetric

pressure field in the downcomer annulus. At a minimum, a two-dimensional

pressure field is required to evaluate adequately the effects of the decompression waves as they travel into and around the downcomer annulus. Since

the current generation of computer programs being used for this application is

one dimensional, the application of the program to modeling multidimensional

regions must be reviewed.

Guidelines for the development of these models, referred to as equivalent pipe

networks (EPN) for the solution of the wave equation in characteristic form,

may be found in References 10-12. In addition to the characteristic form of

solution, the volume-flowpath approach (typical of codes such as RELAP) is

also used by some licensees. This approach has been reviewed and can be

demonstrated to be nearly equivalent to the vector-momentum equation governing

nearly incompressible, low-speed flow for a multidimensional analysis.

11.
The acoustic-wave phenomenon, the transmission and reflection of the pressure

waves, is reviewed in terms of the computer-program method of solution. For

the characteristic form, the transmission and reflection coefficients are

dependent on the area, or the sum of the areas, through which the waves must

pass. In the volume-flowpath approach, these coefficients are replaced by the

effective inertia of a given flowpath connecting two volumes. These

coefficients are reviewed to assure that the wave transmission into the downcomer annulus is conservatively evaluated.

The sensitivity of the model to spatial representation, time-step size, and to

the various convergence criteria is also reviewed.

3.1.4 Code Verification

The last part of the review procedure is code verification, to verify the

ability of a computer program to predict the subcooled decompression phenomenon

or acoustic-wave tracking. By comparing the licensee's program and modeling

procedure to selected problems and experimental data, significant modeling

assumptions and features can be identified. In addition, the staff has augmented the WHAM computer program (Ref. 13) for use in calculating these same

problems and experiments. The program version used by the staff is WHAM/MOD-007

(Ref. 14).

One of the major aspects of the review is the modeling of the downcomer annulus

in at least two-dimensional form with the one-dimensional computer programs.

Four sample problems were developed to aid in this review. They are described,

along with the WHAM/MOD-007 analyses, in Reference 15. Analyses of these

problems have also been performed by the staff's consultants at Sandia with

the CSQ computer program (Ref. 16). CSQ is a two-dimensional hydrodynamic

program. Comparisons of the two sets of analyses demonstrate that the modeling

techniques used for the WHAM/MOD-007 analyses are in good agreement with the

two-dimensional analyses.

Comparisons with applicable test data covering a wide range of system geometries and scales are also required as part of the code-verification review. A

partial list of acceptable test data is given in References 17-24. An HDR

(superheated steam reactor) test facility (Ref. 24) analysis is required by

all licensees as part of code verification. HDR is a nearly full-scale PWR

geometry which has been instrumented to obtain data during a subcooled decompression transient. Westinghouse, Babcock & Wilcox, and Combustion Engineering

will be performing analyses of an HDR test to provide additional verification

of their methodologies.

Audit comparison of a PWR subcooled decompression analysis with the current

staff models and computer program is also performed as part of the verification

program.

3.1.5 Summary

The staff review of the methodology used by the licensee to evaluate the

response of the PWR to a subcooled decompression transient involves a number

of items important to the development of the hydraulic loads. Although each

12
analysis, including code formulation and code predictions of standard subcompartment problems. The licensee must also provide ancillary evidence, in the form

of nodalization-sensitivity studies, to justify that the nodalization used in

the design-basis analysis is adequate for calculating the subcompartment

pressure transients. This requirement applies irrespective of the licensee's

adherence to the subcompartment guidelines contained in this report.

If the licensee has used a subcompartment computer code not previously found

acceptable by the staff, the staff will perform confirmatory analyses and

compare the results with those obtained by the applicant. Of primary interest

are the differential pressures across subcompartment walls and the transient

asymmetric pressure loads and their resultant location on enclosed components

(such as the reactor pressure vessel). Evaluation of the adequacy of structures and component supports should be based on the more severe of the loading

conditions calculated by the staff and the licensee following staff review and

acceptance of the nodalization model and input data.

3.2.2 Analytical Methods

Conventional computer codes for subcompartment analyses utilize a number of

assumptions to facilitate the solution of the equations that describe the

fluid-flow phenomena occurring within a subcompartment. Codes developed for

current use by the staff include the COMPARE code (Ref. 25) and the RELAP-4

MOD 5 Code (Ref. 26). These codes predict (in a conservative manner) experimental data obtained from subcompartment tests. The principal assumptions

made in currently accepted subcompartment-analysis codes are addressed below.

Containment subcompartments (for example, the reactor cavity) are analyzed by

subdividing them into a number of control volumes or nodes, that is, by nodalization of the subcompartment. This network of control volumes interconnected

by junctions or flowpaths describes the geometry of the subcompartment.

All thermodynamic properties within a control volume are assumed to be uniform;

as a result, pressure distributions are represented by mean values. Furthermore, control-volume cQnstituents--a two-phase, two-component, air-steam-water

mixture--are assumed completely mixed at stagnation conditions and at the same

temperature. Thus, thermodynamic calculations are based on the assumption of

homogeneous equilibrium. Additionally, the water expelled from the break is

assumed to be completely entrained in the downstream flow mixture.

The subsonic rates of mass flow between control volumes are obtained by solving

the incompressible one-dimensional momentum equation. Compressibility effects

in the vent flow are accounted for by adjusting vent-flow densities. The

momentum equation includes terms that account for inertia effects and pressure

loss because of wall friction and form losses. The calculation of criticalmass flows between control volumes is based on either the frictionless

Moody correlation with a multiplier of 0.6 or the homogeneous equilibrium

model (HEM) correlation as outlined in Standard Review Plan Section 6.2.1.2.

It should be noted that the use of the HEM critical-flow correlation and the

Moody correlation with a 0.6 multiplier has, in the past, been found acceptable

14
part may be considered on its own, the entire package must be considered in

determining the acceptability of the method and of the computer program. The

goal is to provide assurance that the development of the pressure and velocity

fields is applicable to the PWR geometry and that the loads obtained are

conservative and consistent with safe plant-shutdown requirements.

3.2 Cavity-Pressure Loads

A subcompartment is a fully or partially enclosed volume within the containment

that houses or adjoins high-energy piping systems and restricts the flow of

fluid to the main containment volume in the event of a postulated pipe rupture.

Analyses of pressure transients in containment subcompartments involve the

evaluation of the thermodynamic consequences of a postulated pipe rupture in

these compartmentalized regions of the containment. A spatially dependent,

short-term-pressure transient will occur inside a subcompartment following a

pipe rupture, and pressure differentials across subcompartment structures and

equipment will reach their maximum values generally within the first second

after blowdown begins. The safety issue deals with the localized pressure

gradients within a subcompartment and whether the subcompartment walls and

component supports for the primary reactor-coolant system can withstand these

gradients so that the reactor can be brought to a safe condition in the event

of a pipe rupture in the reactor system.

Therefore, subcompartment pressure analyses must be performed to determine the

asymmetric pressure loadings on major components and on subcompartment walls.

A discussion of the analytical methods, assumptions, and procedures for performing these analyses is provided in this section. The discussion also includes

a summary of the staff's review procedures and acceptance guidelines for

evaluating the subcompartment analyses.

3.2.1 Review Procedures

The staff procedure for evaluating a subcompartment analysis will follow one

of two paths depending on the subcompartment-analysis code used by an applicant.

An expedited review is possible if a licensee has used a subcompartment code

that has been reviewed by the staff and found acceptable. In this instance,

the staff will review the subcompartment-modeling procedures in light of the

recommendations contained in this report. The staff will review the nodalization of the subcompartment and the input parameters used to determine the

adequacy of the analytical model for calculating the pressure gradients and

component loads. If a licensee has adhered to these guidelines, sensitivity

studies are not required. NRC staff will also review the adequacy of the

model used to determine the component loads if it differs from the subcompartment model used to calculate pressure gradients.

If a licensee has used a subcompartment code not previously reviewed and found

acceptable by the staff, additional information must be provided to justify

the acceptability of the subcompartment analysis. In this regard, the

licensee should provide a description of the subcompartment code used for

13
because these models in general calculate low-mass-flow rates conservatively,

over the spectrum of fluid conditions. The limiting loads on structures or

equipment are for most cases determined by the loads induced by the rupture of

a high-energy line immediately adjacent to the structure or equipment in

question. Therefore, critical-flow correlations which yield a low estimate of

mass flow will result in higher pressures in the vicinity of the break and

lower pressures in regions further removed from the break. This translates to

maximizing the loads on the structures or components.

In those situations, however, where the component is most vulnerable to a

loading induced by the rupture of a pipe not immediately adjacent to the

component or where the worst loading results from an overturning moment created

by loads away from the break, it is incumbent upon the designer to consider

the effects of a flow correlation which would yield corresponding high values

of choked flow.

Additionally, the critical-flow models employed in state-of-the-art codes are

unable to mechanistically model vena contracta effects. Therefore, the licensee

should demonstrate that the vena contracta effects are unimportant to the

analysis or that this phenomenom is adequately treated in the flow calculation.

In summary, the staff will require that subcompartment analysis codes that are

used for calculating transient pressure gradients incorporate the following

features:

(1) Thermodynamic capabilities of mixture of air, steam, and water

(2) Homogeneous equilibrium of control-volume constituents

(3) Complete water entrainment for the determination of flow mixtures

(4) Subsonic flow calculations based on the incompressible momentum equation,

including terms for:

(a) temporal change of momentum

(b) pressure loss from friction, area change, turning losses, and

(c) compressibility effects (variation of upstream density)

(5) Vent-flow checks based on selection of minimum flow calculated from

subsonic calculation and critical-flow calculation

(6) Critical-flow calculation based on the HEM correlation or the Moody

correlation with a 0.6 multiplier

Subcompartment Modeling--Subcompartment modeling involves the nodalization

of the physical region to be analyzed and the calculation of the requisite

input data describing the nodalization arrangement for use in computer-code

analyses. In this regard, the following topics need to be discussed:

15
(1) Subcompartment nodalization

(a) general nodalization procedures

(b) nodalization of blowdown volume(s)

(2) Subcompartment-code input data

(a) control-volume data

(b) junction-flowpath parameters

(c) time steps and blowdown data

Md) insulation considerations

(3) Force and moment calculations

3.2.2.1 Subcompartment Nodalization

Subcompartment nodalization is the subdividing of a subcompartment into a

network of control volumes or nodes to assure an accurate prediction of the

transient-pressure response within the subcompartment. Consequently, the

staff's review of subcompartment analyses will include an evaluation of the

basis for nodalization of a subcompartment. The following discussion will

address the general procedure for nodalization of a containment subcompartment, with special emphasis on the reactor cavity, which is the annular volume

between the reactor vessel and the primary shield wall. Since it is not feasible to prescribe an acceptable, specific nodal arrangement applicable to all

configurations of the reactor cavity, nodalization in the reactor cavity will

be discussed in detail to identify a modeling technique that will result in

acceptable nodalization. The recommendations on subcompartment nodalization

are based on a study (Ref. 27) of several different designs of reactor cavities;

this study investigated the sensitivity of the results to nodalization arrangements and other subcompartment-code input parameters.

General Nodalization Procedures--The nodalization of a subcompartment should

represent, as much as practicable, the physical geometry and flowpaths in a

manner that is consistent with the assumptions used in the subcompartmentanalysis code. Boundaries between control volumes, which represent junctions

or vent paths, should be located at physical discontinuities where geometric

influences are expected to create pressure differentials. In the subcompartment of a reactor cavity, there are many geometric variations that merit

attention when developing the nodalization model. Examples of these are

listed below:

(1) Nozzles and piping of reactor vessels (for example, hot leg, cold leg,

core flood line)

(2) Supports of reactor vessels

16
(3) Instrumentation tubes (for example, neutron detectors)

(4) Shielding devices (for example, shield blocks, plugs, shield ring, or

shield tank)

(5) Variations in diameter of reactor vessels

(6) Variations in profile of primary shield wall

(7) Entrances and exits of penetration of shield wall

In addition to nodalization in regions where there are flow-area changes, it

is also necessary to define nodal boundaries and, hence, junctions in regions

of constant cross-sectional area so as to account for frictional or inertial

effects which may create significant pressure losses. These frictional effects

include wall friction (fl/d) and turning losses.

A sample nodal model for a reactor cavity is shown in Figure 5. The reactor

cavity as shown is folded out to represent the entire 3600 circumferential

span. The nodalization in Figure 5 was developed for design of a reactor

cavity where only two geometric discontinuities were assumed to exist; namely,

the flow-area change past the hot and cold legs of the reactor-coolant system

and the variation in flow area encountered at the location of the change in

diameter of the reactor vessel. These two geometric influences result in

horizontal nodal boundaries at elevations A and B. The reason for including

the horizontal boundary at elevation C was to permit a more accurate calculation of the pressure gradient in the long, constant-area, annular region below

the nozzles of the reactor pressure vessel. The boundary at elevation C is

established at the midpoint between elevations B and D. The junctions at the

bottom would correspond to the area change in flowing to the lower reactor

cavity. The top boundaries would represent either a closed top or area change

for flow to volume(s) above the annulus of the reactor cavity.

Sensitivity studies (Ref. 27) show that the horizontal boundary at elevation C

(shown in Figure 5) can be eliminated for certain cavity designs, when the

boundary is established in the absence of geometric restrictions. These

studies also showed that the resultant forces and moments (calculated about a

horizontal axis through the nozzles--elevation A) changed by less than 5%.

However, NRC is unable to conclude that such consolidation of nodes is uniformly

acceptable for all cavity designs, since the impact of such consolidation

depends on specific cavity dimensions and flowpaths. Therefore, an applicant

should demonstrate a similar insensitivity by performing noding studies to

preclude any need for this level of nodalization for a specific plant application.

The above discussion dealt with the criteria for establishing horizontal nodal

boundaries in a model of a reactor cavity. In establishing vertical boundaries

to account for pressure differentials created by circumferential flow about

the reactor pressure vessel, attention should be directed again at modeling

geometric discontinuities. For the design of the reactor cavity represented

in Figure 5, the only geometric variations in the circumferential direction

are those created by the piping of the coolant system. Therefore, the control

17
U P I - - - I I -

A P 

I -

I -U-----4-- 4 ~-e- 4 ---- 4 I -4------4- I

I

B

0

oS

C

D - - - - --

Figure 5. Sample Nodal Model for a Reactor Cavity.

18
volumes in Figure 5 encompass equal 600 segments about the reactor cavity

region. This same level of noding detail was retained in areas located away

from the nozzle region to adequately account for inertia and frictional effects

(including wall friction and turning loss). Also, the greater degree of

nodalization reduces the impact of the typical computer-code assumption that

flowpaths are one-dimensional in nature. Cavity designs which include restrictions to circumferential flow such as support columns or instrument tubes of

the pressure vessel should be modeled so that nodal boundaries reflect the

presence of these restrictions.

Sensitivity studies (Ref. 27) have also been performed to evaluate the impact

of the level of detail in circumferential nodalization. These studies demonstrated that nodalization could be made much coarser; that is, fewer circumferential nodes, without substantial changes in the calculated peak loads and

moments on the reactor vessel. However, recognizing that sensitivity to

nodalization detail will depend upon plant-specific geometry, it is recommended

that circumferential noding for a design-basis model should include vertical

boundaries through all nozzles of reactor pressure vessels for the height of

the reactor cavity. Sensitivity studies on a specific plant may be performed

by applicants in order to demonstrate acceptability for a model with less

detail.

There are alternative approaches to development of an acceptable nodalization

arrangement for subcompartment analysis of the reactor cavity. One approach

would be to model the reactor cavity with detailed nodalization accounting for

all obstructions, and resolution of detail in large, constant-area regions in

the manner described above. In this case, if an approved subcompartmentanalysis is used, further sensitivity studies are not required.

Another approach would be to model the cavity in less detail, but account for

all pressure losses, and provide sensitivity studies demonstrating the acceptability of the design-basis model. Such sensitivity studies must compare the

results of interest, whether it be the loading on subcompartment walls or the

forces and moments on components. These studies may be performed on a model

of either the plant under study or another plant having the same reactor

cavity configuration and approximately the same dimensions.

Nodalization of Blowdown Volumes--A critical aspect of modeling the reactor

cavity region is nodalization of the volume(s) into which the break effluent

mass and energy are released. Development of criteria for the modeling of

this control volume, commonly referred to as the break node or blowdown volume,

is made more complex by the diversity of designs of reactor cavities.

For reactor cavity designs where the gap between the primary shield wall and

the pressure vessel is large (approximately 2 ft) or where the shield wall is

recessed in the vicinity of the nozzles, the break effluent should be assumed

to enter the adjoining control volumes which have a projected area on the

reactor pressure vessel. For an arrangement similar to that shown in Figure 5,

the blowdown from a postulated guillotine rupture, neglecting jetting effects,

would be assumed to enter the four control volumes surrounding the ruptured

pipe.

19
There are some designs in which the location of the terminal ends of the

piping of the reactor-coolant system is within nodes or control volumes that

are away from the annular region of the reactor cavity. Thus, in the event of

a postulated pipe rupture, the blowdown effluent is assumed to enter a control

volume inside the primary shield wall (for example, pipe-penetration volume)

or otherwise radially removed from the annular space immediately adjacent to

the reactor vessel.

For models of reactor cavities where the break node is not assumed to be

directly projected onto the reactor vessel, the applicant must clearly

demonstrate the acceptability of the nodalization in the vicinity of the pipe

break. This includes justification with sufficiently detailed drawings of the

region of the vent areas away from the break node.

In selecting the break node(s), consideration must be given to the potential

for, and effects of, fluid jetting from the site of the pipe rupture. This is

important for all reactor-cavity designs but especially in those configurations where the pipe rupture is postulated to occur within a primary shield-wall

penetration. For such designs the applicant must either justify that direct

jetting into the reactor-cavity annulus would not occur or include the effect

of jetting in calculating the loading condition on the reactor vessel. An

acceptable approach is to demonstrate, if practicable, that pipe restraints

would limit lateral separation in the event of pipe rupture to less than the

pipe thickness.

3.2.2.2 Subcompartment Code Input Data

The following is a discussion of the data that are required for input into the

computer codes used in subcompartment analysis. The data can be described as

either control-volume data or flowpath data. Additional input parameters

addressed are time-step data, blowdown data, and assumptions regarding the

treatment of thermal insulation.

Control-Volume Data--The pertinent control-volume data are the free volume and

the initial temperature, pressure, and relative humidity of control volumes.

Initial conditions should be selected, within the range of anticipated values,

to conservatively predict the result of interest; that is, loads on structures

or loads and moments on primary system components.

The studies (Ref. 27) we have performed showed that the sensitivity to initial

pressure, temperature, and relative humidity variations is insignificant (less

than 5%). However, this study was done for a specific reactor cavity design

and may not apply to other subcompartment (for example, steam-generator enclosure) analyses or analysis involving different reactor cavity designs.

Uncertainties in the calculated free volume of each node shall be applied in

such a way as to obtain the most severe loading condition of interest.

Flowpath Parameters--The flowpath data of interest in modeling a flowpath

between control volumes typically include the flow area, loss coefficients,

and the inertia term (L/A ft-1).

20
The flow area that is used as input to subcompartment codes should be based on

the minimum flow area between control volumes. This is consistent with the

basic premise of subcompartment nodalization which is to place boundaries at

flow restrictions. Also, critical flow models are based on the minimum flow

area.

Uncertainties in flow areas should be applied in a manner that yields conservative results of interest. For example, if one would calculate lateral loads

on the reactor vessel, a low estimate of the flow areas should be used for

flowpaths that relieve the pressure near the break, for example, circumferential flowpaths. However, if the vertical location of the resultant lateral

force, that is, the moment, is of concern, then maximization of appropriate

flow areas for vertical flow on the same side of the reactor vessel as the

break should be considered.

If the subcompartment model includes vent areas that become available as a

result of fluid flow or pressurization effects after initiation of the accident,

the following criteria apply:

(1) The modified flowpath parameters (for example, vent area, resistance) as

a function of time or pressure after the break should be based on analysis

of the subcompartment pressure response.

(2) Availability of vent areas resulting from pressure or flow effects should

be justified analytically or experimentally.

(3) The potential for, and effects of, missiles that may be generated during

the transient should be evaluated.

Loss coefficients are conventionally used to account for total pressure losses

due to changes in flow area, turning losses, and wall friction. Coefficients

of hydraulic loss are available in the literature; the most commonly used

comprehensive itemization is found in Reference 28. Applicants should provide

a-discussion of the methods used to determine loss coefficients and a representative sample calculation of the loss coefficients.

Consideration should be given to the direction of flow when selecting loss

coefficients for a given flowpath, since values may change depending on the

direction of flow. If the analysis is done with a subcompartment code that

does not account for effects of flow direction, the specified flow direction

that is used as input must be verified by the analysis.

Attention should also be given to ensuring the consistency between the flow

area and loss coefficients that are used as input data. For loss coefficients

based on an area other than that used as input to the code, the loss coefficients must be appropriately modified (by the ratio of the areas squared).

It is recognized that loss coefficients used in subcompartment analyses are

based on hydraulic, steady-state flow and, as such, do not reflect consideration of the complexities associated with a compressible two-phase, two-component,

transient-flow problem. However, based on the reactor-cavity studies (Ref. 27)

NRC staff has performed to date, the sensitivity of reactor-cavity analysis to

21
the uncertainties associated with the steady-state, hydraulic loss coefficients is small. Therefore, the current method of selecting loss coefficients

in subcompartment analysis of reactor cavities is judged to be acceptable.

Fluid inertia has a significant effect on the pressure gradients calculated in

subcompartment analyses (Ref. 27). Inertia is typically accounted for by use

of the geometric (L/A ft-1) as input, which results from, solution of the

momentum equation. Currently, acceptable practice for calculation of the

inertia term is outlined in Reference 27.

Time steps used to calculate the transient distribution of mass from the pipe

rupture should be sufficiently small (a decrease in the time-step size should

be done with the aid of sensitivity studies on a specific plant or by comparison with studies on similar plants).

3.2.2.3 Other Factors

Insulation Behavior Considerations--Thermal insulation on piping and vessels

must be considered when performing subcompartment analysis. The relative

importance of assumed insulation behavior is magnified for the reactor-cavity

analysis since thermal insulation may comprise a significant volume within the

reactor-cavity annulus and the penetrations through the shield wall.

Assumptions related to postaccident insulation behavior must either be shown

to be conservative or, if insulation behavior is treated mechanistically, such

behavior must be justified analytically or experimentally.

In either case, the applicant must provide justification for the assumed

behavior of insulation and a detailed discussion of how insulation affects

volumes and flow areas both in thereactor-cavity annulus and in those flowpaths away from the reactor cavity (for example, penetrations of the shield

wall).

Blowdown Mass and Energy--The driving potential for asymmetric pressure loads

is the blowdown from a postulated pipe rupture. The blowdown characteristics

are typically input to the subcompartment computer codes in the form of a

mass-flow addition rate as a function of time and a corresponding enthalpy- or

energy-addition rate as a function of time. On the basis of studies performed

by the staff (Ref. 27) and a review of applicants' analyses, we find that the

blowdown mass-and-energy release rates are important parameters in the calculation of asymmetric loads and moments. Analyses have shown that the sensitivity

of peak calculated loads and moments of reactor pressure vessels can be directly

proportional to the short-term mass-and-energy release rates. Therefore,

subcompartment analysis should be based on conservatively derived blowdown

mass-and-energy release rates.

In general, calculations of the mass-and-energy release rates for a LOCA

should be performed in a manner that conservatively establishes the containment

subcompartment response. The analytical approach used to compute the mass-andenergy release profile for the first several seconds of blowdown for use in

subcompartment analysis is acceptable if the volume noding of the piping

22
system is similar to those of an approved emergency-core-cooling-system (ECCS)

analysis or if appropriate noding-sensitivity studies have been made. The

computer programs that are currently acceptable include SATAN-V (Ref. 29),

CRAFT-2 (Ref. 30), CE FLASH-4A (Ref. 31), and RELAP4 (Ref. 32), when a flow

multiplier of 1.0 is used with the applicable choked-flow correlations for the

above computer codes. An alternate acceptable approach is to assume a constant

blowdown profile using the initial conditions with the choked-flow correlations

in the above references.

3.2.2.4 Force-Moment Calculations

The calculation of forces and moments, as applicable, on components of reactor

systems or on subcompartment structures is the culmination of the subcompartment analysis and, therefore, requires the same detailed consideration as that

for subcompartment modeling.

The following requirements apply to the translation of calculated pressure

gradients into forces and moments.

(1) The subcompartment nodal model used for the calculation of forces and

moments should be the same model as that found acceptable for the calculation of pressure gradients. If the model used to calculate forces and

moments is different from the model used to calculate pressure gradients,

then the calculated forces and moments shall be shown to be equivalent to

those forces and moments that would have been calculated had the other

model been used.

(2) Projected areas onto a curved surface shall be based on projected planar

areas. Then multiplication of the projected area by the calculated nodal

pressure gives the force acting normal to the surface through the area

centroid.

(3) Force calculations for components such as the reactor pressure vessel

shall include loads resulting from differential pressures acting across

the vessel piping and nozzles.

23
4 STRUCTURAL EVALUATION

The structural evaluation of the primary system for asymmetric LOCA loads

involves the development of detailed finite element models, the application of

the loading transients (discussed in Section 3) to these models, and finally

the assessment of the resulting structural response against appropriate acceptance criteria. The finite element models required for this evaluation are

typically large and complex and may incorporate both nonlinear (geometric) and

inelastic (material) modeling aspects. Based on these considerations, the

staff chose an audit approach to verify various modeling assumptions and

sensitivity.

The staff evaluated six PWR primary systems both to gain a better understanding

of the primary-system response and to develop the capability for performing an

audit of plant-specific analyses. The six plants chosen represented two from

each NSSS vendor: one using conservative elastic-analysis techniques and one

incorporating inelastic-and nonlinear-analysis techniques. A detailed discussion of each of these six plant-audit calculations, listed below, is provided

in References 33 through 38.

Westinghouse Lead Plants

North Anna 1--elastic analysis (Ref. 33)

Indian Point 3--inelastic analysis (Ref. 34)

Combustion Engineering Lead Plants

San Onofre--elastic analysis (Ref. 35)

St. Lucie 1--inelastic analysis (Ref. 36)

Babcock and Wilcox Lead Plants

Erie--elastic analysis (Ref. 37)

Arkansas 1--inelastic analysis (Ref. 38)

A discussion of modeling considerations and review procedures for the structural evaluation are-provided in Section 4.1 and structural acceptance

criteria, although not developed as part of this task action plan, are provided in Section 4.2.

4.1 Primary-System Analytical Development

4.1.1 Modeling Aspects

Structural analysis of reactor primary coolant systems is performed to

demonstrate that the plant can be safely shut down in the unlikely event of a

postulated LOCA. Structural analysis for purposes of this discussion is

defined as the specification of mechanical response (loads, deflections, and

stresses) of the primary coolant piping, major components and their supports,

and reactor internals and fuel. Detailed finite element models are required

to predict these mechanical responses accurately, although more-conservative

simplified techniques may be applicable in special cases.

24
Modeling techniques for this analysis are not unique. They vary depending on

the computer codes, the geometry, and the analytical assumptions. Primary

systems responses, in general, are nonlinear in nature due to geometrical gaps

and various design characteristics of supports, such as cables which act in

only one direction. The analytical models may also require inelastic considerations to reflect material nonlinearities.

A single model representing the entire primary system and incorporating sufficient detail is not practical within current computer and economic limits.

The structural analysis is, therefore, completed in phases in which the

responses from a general system model are used as input to detailed structural

models (subsystems) of critical parts of the primary system. The system model

typically includes a representation of the reactor pressure vessel, its core

barrel,, the internals and fuel assemblies, the primary coolant piping, the

steam generator and supports, and the primary-coolant pump and supports.

Typical subsystems which are evaluated using input from the systems model are:

attached ECCS piping, detailed models of RPV and component supports, and

detailed evaluation of the fuel integrity.

Basically, the preceding paragraphs have been oriented toward a finite element

model of a single reactor system. Finite element representation of the primary

coolant system using a number of completely decoupled models may also constitute an acceptable approach. As an example, a reactor-coolant system could be

represented by the following three separate finite element submodels:

(1) Reactor vessel internals and vessel supports

(2) Broken piping loop, components, and component supports

(3) Intact loop, component, and component supports

If this approach is pursued, the interaction of the various submodels must be

accounted for in the evaluation.

A general discussion of modeling aspects is provided in this section as well

as a detailed discussion of one acceptable modeling technique (used by NRC

staff for the Indian Point 3 audit) to illustrate modeling considerations.

4.1.1.1 Vessels

Rigid-body representation of the reactor vessel and steam-generator vessel is

acceptable for the LOCA analysis. Models of the vessel shell should include

sufficient noding detail to allow application of the time-history loads,

incorporation of supports and vessel internals, and determination of boundary

conditions for attached ECCS systems. Nozzle flexibilities should also be

included at appropriate locations in the models. The models should represent

the mass distribution and inertia characteristics of the vessels. Mechanisms

of the control-rod drive should also be included in the RPV model.

4.1.1.2 Component Supports

Major component supports are, in some cases, very complex structures. Simplified representation of these structures in an analysis of the reactor-coolant

system is acceptable (where significant nonlinearities do not exist) and, in

25
some cases, necessary. Constraints in the size of the finite element model

may make it necessary to incorporate a simplified representation of the steam

generator and pump supports.

In modeling the various component supports, substructuring may be a valid

means of reducing the size of the overall model. To explain the term "substructuring" as used in this report, consider the following example for a main

coolant pump with a complex structure or framework of beams comprising its

support to the foundation. A separate analysis of the support structure can

be conducted with loads applied at the locations where the pump attaches to

its support structure. If force versus deflection curves and moment versus

rotation curves can be developed for this "substructured" support and the mass

characteristics (natural frequencies) can be reasonably represented, then the

complex support system can be replaced with a simpler support system whose

load versus deflection, moment versus rotation, and mass characteristics are

the same as those for the complex system. The net result is then a simpler,

smaller model for the system analysis.

4.1.1.3 Internals

The goal of the internal portion of the system model is twofold; the first is

to determine the structural response of the core barrel and core supports; the

second is to provide appropriate input for a detailed evaluation of the fuel

(discussed in Chapter 5). The internal structures and terminology may vary,

depending upon NSSS vendor designs; but, in general, the internals model

should include the following:

(1) Core barrel

(2) Thermal shield

(3) Core-support structure

(4) Core

(5) Upper plenum assembly

(6) Guide-tube assembly

Modeling of these structures in the system model should include sufficient

detail to represent the mass distribution and inertia effect. Attention

should be given to the review of interface areas (such as the core-barrel-toRPV connection) which are part of the load path, to ensure an accurate modeling expectation.

4.1.1.4 Primary Piping

Representation of the piping should be sufficiently detailed to calculate the

maximum stress, permit representation of pipe supports, and provide input for

structural analysis of ECCS and other piping subsystems that connect to the

primary coolant system.

4.1.1.5 Indian Point Unit 3 Audit Model

A summary of the modeling techniques used for the Indian Point Unit 3 (Ref. 34)

lead plant audit is provided to illustrate the modeling detail and development.

The primary system of Indian Point Unit 3 is represented by an assemblage of

26
linear elastic, nonlinear elastic, and elastic-plastic truss and beam elements.

Finite element models used in staff audit of Indian Point Unit 3 are represented in Figures 1-9. As shown in Figure 6, the steam-generator supports are

represented by equivalent beams analogous to the steam-generator supports.

These equivalent structures were derived from the flexibility matrices and

vibration modes of detailed support models. The primary piping loop models,

Figure 7, consist of an elastic-plastic beam representation of the primary

piping with an elastic beam representation of the pump, steam generator, and

main steam piping. Figures 8 and 9 show the RPV and internals and core models.

The RPV and internals are represented by elastic beams, and the core is represented by elastic beams interconnected by nonlinear elastic truss elements or

gap elements. With this model, the fuel, represented by elastic beams, is

allowed to impact both the core barrel and adjacent fuel. The RPV supports

shown in Figure 8 are represented by a horizontal elastic-plastic truss element and a vertical nonlinear elastic or compression-only truss element. The

remaining structural components, the hot stops and snubbers on the steamgenerator supports, the base restraints of the pump supports, the tie rods,

and the primary piping restraints are represented by either nonlinear elastic

or elastic-plastic truss elements.

4.1.2 Review Procedure

The structural evaluation of the primary system must include, as a minimum, an

analysis of the following:

(1) Reactor pressure vessel

(2) Fuel assemblies, including grid structures

(3) Control and rod drives

(4) ECCS piping that is attached to the primary coolant piping

(5) Primary coolant piping

(6) Reactor-vessel, steam-generator, and pump supports

(7) Reactor internals

(8) Biological shield wall and neutron shield tank (where applicable)

(9) Steam-generator compartment wall

The review and determination of the adequacy of a plant-specific analysis for

asymmetric loads by the staff should include an assessment in the following

areas: the structural computer codes, the modeling techniques and assumptions,

and the acceptability of the responses.

The structural computer codes and modeling techniques and assumptions are

reviewed and approved as a package. Two approaches for performing this review

have been developed by the staff:

(1) Review and approve a general analytical technique applicable to more than

one plant.

(2) Perform an independent audit of a specific plant evaluation.

27
Unrestrained

motion direction

Snubber

Hot stop

but free to slide vertically * Nodes

® Nodes to which steam

generator attaches

Nodes to which tie rods

attach

-1 j-- Gap elements

(nonlinear elastic trusses)

-Ag- Elastic plastic trusses

-F Flexible beams

. -- -Rigid beams

T'To RPV

Rigid connection

To pump

Unrestrained

motion direction ,a-l Snubber

Hot stop

Figure 6. Steam Generator Support.

28
Main steam pipe

4 btea

gene

Primary

pump

I

Crossover

pipe

Iv

0 Nodes

® Nodes to which

supports attach

m

rator

Primary pipe restraints

RPV inlet nozzle

RPV outlet

nozzle K>VIO

r Primary pipe

restraints

Rigid beams

-Flexible beams

1 Gap elements

(nonlinear elastic trusses)

Figure 7. Typical Piping-Loop Model.

29
Support ledge frigid)

Thermal shield

0- ' Rigid

Lower internals

Figure 8. Reactor Pressure Vessel and Internals Model.

30
BF1V

Upper core plate

-I A

-1B

Section A-A

Fuel

Core plate

tJ

I

I

IH

,I "- --lIIF-

" "-4--

"• II

Core plate

Fuel

IFuel

Fuel

A

i • I

4 I

A

I A

Lower core plate

Section B-B

Figure 9. Core Model.

31
4.1.2.1 General Analytical Technique

Review and approval of a general analytical technique is applicable when a

number of plants are evaluated using the same computer codes and modeling

considerations. This approach was chosen by the majority of operating plants

who engaged their NSSS vendors to perform these evaluations. The review is

performed in two steps: first, the general analytical technique is reviewed

and approved and, second, each plant analysis is reviewed for consistency with

the approved general approach.

Typically, the general analytical technique consists of the structural computer

codes, the analytical modeling, and the result from a representative analysis.

The representative analysis may be of a specific plant or a generic plant and

should identify selected user options in the computer code and provide the

modeling detail and assumptions. The staff typically determines the acceptability of the general analytical technique by performing an independent audit

evaluation, discussed in Section 4.1.2.2. The audit evaluation may not be

required if sufficient noding-sensitivity studies are performed on the structural models and the computer codes have been verified as acceptable by the

staff.

The staff has currently reviewed and approved the general analytical approach

submitted by Westinghouse for the Indian Point 3 nuclear plant (Ref. 39).

Details of the audit evaluation are presented in an EG&G report (Ref. 34).

The staff has developed audit models for both a Combustion Engineering plant

(Ref. 34) and a Babcock and Wilcox plant (Ref. 38). Their general analytical

approach will be evaluated when their respective plant reports are submitted

for staff review.

Basically, evaluation of each plant submittal involves reviewing the structural

models for detail and completeness and applying the structural computer codes

for consistency with the approved general approach. Review of the modeling

should include the relative noding detail, the boundary conditions, and the

noding assumptions. The computer code version and user-specific options

should also be verified to be consistent with the approved approach.

4.1.2.2 Independent Staff Audit

The second approach is to perform an independent audit of the plant using a

staff-approved computer code and staff-developed models. Plant analyses which

either do not use an approved general analytical approach or modify an approved

approach significantly, may require an independent staff audit. Exceptions

may exist where sufficient supporting documentation is provided to enable the

staff to make an evaluation. Review for acceptability is fairly simple in

this approach and typically requires a comparison of the response characteristics. The information required from the licensee to perform the audit is

summarized below.

Piping--Geometric layout dimensions for piping centerline are required so

connectivity from the RPV to the steam generator (hot-leg segment), steamgenerator-to-pump inlet (crossover segment), pump outlet to RPV inlet (cold-leg

segment), can provide for proper interactions of the piping with its attached

32
components. Section properties for each segment of piping must be available

from the drawings so proper stiffness effects can be included. Materials

lists for all segments should be provided along with normal operating conditions (pressure and temperature) for each segment. Pipe supports must be well

defined with regard to location and type (material, stiffness, load limit,

behavior above load limit--failure versus constant load, action--one directional only or either direction, geometric gaps).

Steam Generators--Generally, section properties of the shell must be available

so the effective stiffness of the steam generators can be modeled. Additionally, weights of steam-generator components should also be provided (upper and

lower head tube sheets, tubes, water contained). Total weight of the steam

generator (dry and including normal operation water) should be available to

provide reasonable mass distribution for the steam generator. Steam-generator

support details must also be available, including support-to-foundation interface to provide a useful form of input for the analyst.

Pumps--The weight (including water weight), center-of-gravity location, and

rotary inertia should be provided for the pumps in order to represent their

effects on the piping. Pump supports should be defined and located including

snubbers and their behavior (that is, their physical characteristics: load

rating, behavior on exceeding rated load and failure load, and preload, if

any). For complex pump supports to foundation, details must be provided to

allow mass and stiffness properties to be included for that part of the

system.

Reactor Pressure Vessel--Details of the RPV shell must be available since

stiffness properties for the RPV are normally based on properties of the shell

section. Weights should be provided (including water weights and their distribution) to represent the proper mass distribution of the RPV. Support details

and support-to-foundation details and information should be provided, as for

steam generators and pumps.

RPV Internals--Detailed and assembly drawings should be supplied so interactions of various, major internal components to the RPV can be modeled. These

drawings should include section properties and material properties for the

fuel modules and internals of the core barrel. Weights for these components

should also be supplied. In addition, weights of core plates and of remaining

internals should be supplied, so proper mass distribution in the RPV can be

included.

Water weights for the various regions inside the RPV and fuel frequencies

should be supplied.

The final review consideration is the review of the structural response against

applicable acceptance criteria (Section 4.2).

4.2 Structural Acceptance Criteria

The structural integrity of the primary system including the reactor pressure

vessel, RPV internals, primary coolant loop, and components must be evaluated

against appropriate acceptance criteria to determine if acceptable margins of

33
safety exist. Allowable limits and appropriate loading combinations are set

forth in standard review plans (SRPs), which are listed in the table that

follows. The staff recognizes that in some specific cases, where "as built"

designs are being reevaluated for asymmetric LOCA loads, these design limits

may be exceeded. Acceptance of alternative allowable limits will be based on

a case-by-case evaluation of the safety margins.

Load combination criteria in general were not addressed as part of this study.

Currently the staff requires that seismic (SSE) and LOCA response be combined,

along with responses due to other loading as specified by the SRP. An acceptable method for combining elastically generated seismic and LOCA responses is

provided in NUREG-0484 (Ref. 40). Acceptable methods for combining response

generated by an inelastic LOCA analysis and elastic seismic analyses will be

evaluated on a case-by-case basis.

Item SRP

Reactor pressure vessel

Reactor internals

Primary coolant loop piping

ECCS piping

RPV, SG, pump supports

Biological shield wall

Steam-generator compartment wall

Neutron-shield tank

3.9.3

3.9.5, 3.9.1

3.9.3

3.9.3

3.8.3

3.8.3

3.8.3

3.8.3

34
5 FUEL-ASSEMBLY EVALUATION

Although core components are included in the structural analysis of the primary

system, a more-detailed structural analysis of the reactor fuel assembly is

performed to show that the fuel is maintained in a cool condition and that

insertion of control rods is not prevented. These are basic requirements that

follow from the General Design Criteria (10 CFR Part 50) and are described

explicitly in Section 4.2 of the Standard Review Plan (SRP).

To perform the detailed analysis for fuel-assembly structural response, input

information is obtained from results of the structural analysis of the primary

system. The input information includes primary-system motions (for example,

core plate, core shroud, and fuel-alignment plate) and transient pressure

differences as described earlier. Since the basic requirements for fuel

integrity and the input motions are similar to those for the seismic analysis,

and since the fuel analysis is done separately from the primary-system analysis, unified guidelines for the fuel structural analysis have been developed.

These guidelines are described in a proposed Appendix A to SRP Section 4.2.

The SRP Appendix is included as Appendix E of this report and has been issued

for public comment (Federal Register, Vol. 45, No. 40, 2/27/80, p. 12939).

Appendix A of SRP Section 4.2 addresses seismic loads of fuel assembly and

structural loads of LOCAs as well as offering an acceptable method for combining these responses for PWRs and BWRs. It thus includes, but goes beyond,

requirements related to asymmetric blowdown loads of Task A-2. The SRP

Appendix details analytical methods for predicting structural loads on fuel

assembly components (for example, spacer grids, fuel rods, and guide tubes),

determination of component strengths, and acceptance criteria for comparing

predicted loads with component strengths. The main points of the acceptance

criteria are summarized below; see Appendix E for details.

(1) Fuel rod (poison rod, control rod, guide tube, etc.) fragmentation does

not occur. Allowable material strength values used in design should

bound a large percentage (about 95%) of the distribution of strength

values.

(2) LOCA temperature and oxidation limits of Title 10 of CFR 50.46 for the

fuel-rod cladding are not exceeded. If the fuel-assembly grid-crushing

strength Pcrit (the force required to permanently deform the grid) is not

exceeded, the standard LOCA analysis is performed; otherwise the effects

of a deformed grid are accounted for in the LOCA analysis. The allowable

value of Pcrit is a mean of the distribution of measured Pcrit values.

(3) Seismic and LOCA loads are combined by taking the square root of the sum

of the squares (SRSS).

Technical background for the SRP Appendix is given in three reports referenced

in that appendix.

35
REFERENCES

1. Letter from Operating Reactor Branch Chiefs to all Operating PWR Facilities Subject: Survey of Operating Plants to see if they had included

Asymmetric Loads, dated October 15, 1975.

2. Letter from Operating Reactor Branch Chiefs to all Operating PWR Licensees,

Subject: Request to evaluate adequacy of reactor system components at

their facilities with respect to Newly identified loads, dated June 1976.

3. Westinghouse Electric Corporation, "Augmented Inservice Inspection of

Piping Welds at the Reactor Vessel Nozzles," WCAP-8802, August 1976.

4. Science Applications Inc., "An Analysis of the Relative Probability of

Pipe Rupture at Various Locations in the Primary Cooling Loop of a Pressurized Water Reactor Including the Effects of a Periodic Inspection,"

SAI 001-PA, June 1976.

5. Science Applications, Inc., "An Analysis of the Probability of Pipe

Rupture at Various Locations in the Primary Cooling Loop of a Babcock and

Wilcox 177 Fuel Assembly Pressurized Water Reactor Including the Effects

of a Periodic Inspection," SAI-050-77-PA, September 1977.

6. Westinghouse Electric Corporation, "Multiflex--A Fortran-IV Computer

Program for Analyzing Thermal-Hydraulic-Structure System Dynamics,"

WCAP-8708(P), WCAP-8709(NP), September 1977.

7. Babcock and Wilcox, "Analytical Methods Description--Reactor Coolant

System Hydrodynamic Loadings During a Loss of Coolant Accident,"

BAW-1032P, May 1979.

8. Combustion Engineering, "Blowdown Analysis Method--Method for Analysis of

Blowdown Induced Forces in a Reactor Vessel," CENPD-252-P, August 1978.

9. J. K. Dienes, C. W. Hint, and L. R. Stein, Los Alamos Scientific Laboratory, "Computer Simulation of the Hydroelastic Response of a Pressurized

Water Reactor to a Sudden Depressurization," LA-NUREG-6772-MS, April 1977.

10. V. L. Streeter and E. B. Wylie, "Two- and Three-Dimensional Fluid

Transients," ASME Journal of Basic Engineering 90, 501-510, December

1968.

11. S. Fabic, "Two- and Three-Dimensional Fluid Transients," Transactions of

the ANS 14, 360-361, 1971.

10. R. K. Cole, Sandia Laboratories, "An Analysis of the Method of Streeter

and Wylie for Multi-dimensional Wave Propagation," SAND 77-0549,

July 1977.

36
13. S. Fabic, Kaiser Engineering, "Computer Program WHAM for Calculations of

Pressure, Velocity and Force Transients in Liquid Filled Piping Networks,"

Report No. 67-49-R, November 1967.

14. Memorandum from E. D. Throm, NRC, to Z. R. Rosztoczy, NRC, "Availability

of the WHAM/MOD-007 Subcooled Blowdown Hydraulic Analysis Code," May 13,

1976.

15. D. Berman et al., Sandia Laboratories, "LOCA Analyses Quarterly Repo4tApril-June 1977," SAND 77-1530, December 1977, pp. 14-6D.

16. S. L. Thompson, Sandia Laboratories, "CSQ--A Two-Dimensional Hydrodynamic

Program With Energy and Material Strength," SAND 74-0122, August 1975.

17. R. T. Allemann et al., Battelle Pacific Northwest Laboratories, "AEC

Research and Development Report: Coolant Blowdown Studies of a Reactor

Simulator Vessel Containing a Simulated Reactor Core," BNWL-1524,

June 1971.

18. D. D. Christensen, EG&G Idaho, Inc., "United States Standard Problem 5,

Final Report," PG-R-77-44, December 1977.

19. H. M. Delaney, EG&G Idaho, Inc., "United States Standard Problem 6 and

International Standard Problem 4, Final Report," CVAP-TR-4-78, March 1978.

20. G. F. Brockett et al., EG&G Idaho, Inc., Experimental Investigations of

Reactor System Blowdown," IN-1348, September 1970.

21. H. C. Robinson, EG&G Idaho, Inc., "Experiment Data Report for LOFT

Nonnuclear Test L1-2," TREE-NUREG-1026, January 1977.

22. R. J. Fritz and E. Kiss, General Electric Company, "The Vibration Response

of a Cantilevered Cylinder Surrounded by an Annular Fluid," KAPL-M-6539

(RJF-1O), February 24, 1966.

23. D. A. Prelewicz, Westinghouse Electric Corporation, "Hydraulic Pressure

Pulses With Structural Flexibility Test and Analysis," WAPD-TM-1227,

April 1976.

24. R. Krieg, E. G. Schlechtendall, and K. H. School, "Design of HDR Experimental Program on Blowdown Loading and Dynamic Response of PWR Vessel

Internals," Nuclear Engineering and Design 43, 1977.

25. Los Alamos Scientific Laboratory "COMPARE MOD-l A Code for the Transient

Analysis of Volumes With Heat Sinks, Flowing Vents and Doors," Report

LA-7199-MS, March 1978.

26. Aerojet Nuclear Company, "Containment Supplement to: RELAP 4 A Computer

program for Transient Thermal-Hydraulic Analysis," Interim Report

1-214-75-03, June 1975.

37
27. Los Alamos Scientific Laboratory "Subcompartment Analysis Procedures,"

Report LA-8169-MS, December 1979.

28. I. E. Idel'chik, Handbook of Hydraulic Resistance. Coefficients of Local

Resistance and Friction, U.S. Atomic Energy Commission Report AEC-TR-6630,

1966.

29. Westinghouse Electric Corporation, "Westinghouse Mass and Energy Release

Data for Containment Design," WCAP-8312A, August 1975.

30. U.S. Nuclear Regulatory Commission, "Safety Evaluation Report Related to

the Preliminary Design of the BSAR-205 Standard Design by Babcock and

Wilcox," USNRC Report NUREG-0433, May 1978.*

31. U.S. Nuclear Regulatory Commission, "Safety Evaluation Report for the

Preliminary Design Standard, Approval of the Combustion Engineering

Standard Safety Analysis Report--CESSAR System 80," USNRC Report NUREG

75/112, December 1975.*

32. C. G. Bruch, ed., Idaho National Engineering Laboratory "RELAP4/MOD5: A

Computer Program for Transient Thermal-Hydraulic Analysis of Nuclear

Reactors and Related Systems, User's Manual." Supplement 1,

ANSR-NUREG-1335, September 1976.

33. R. E. Mathews, EG&G Idaho, Inc., "Pressurized Water Reactor Mechanical

Response Analysis North Anna Unit 1," RE-E-76-083, December 1976.

34. R. W. Macek, EG&G Idaho, Inc., "Nonlinear LOCA Dynamic Analysis of the

Indian Point Unit 3 Primary Coolant System," RE-A-77-106, May 1978.

35. R. F. Lippert, EG&G Idaho, Inc., "LOCA Mechanical Response Analysis of

San Onofre Unit 2 Pressurized Water Reactor," RE-A-78-252, October 1978.

36. R. W. Macek, EG&G Idaho, Inc., Nonlinear LOCA Dynamic Analysis of the St.

Lucie Unit 1 Primary Coolant. System," RE-A-79-024, February 1979.

37. G. L. Thinnes, EG&G Idaho, Inc., Pressurized Water Reactor Mechanical

Response Analysis Erie Unit 1," EA-5119, March 1980.

38. T. R. Thompson, EG&G Idaho, Inc., "Nonlinear LOCA Dynamic Analysis of the

Arkansas Nuclear One Unit 1 Primary Coolant System," RE-A-78-264, Rev. 1,

July 1979.

39. Westinghouse Electric Corporation, "Analysis of Reactor Coolant System

for Postulated Loss of Coolant Accident-Indian Point Unit 3 Nuclear Power

Plant," WCAP-9117, June 1977.

40. R. Cudlin, S. Hosford, R. Mattu, and K. Wichman, "Methodology for Combining Dynamic Responses," USNRC Report NUREG-0484, September 1978.*

*Available for purchase from the National Technical Information Service,

Springfield, VA 22161.

38
APPENDICES

39

APPENDIX A

TASK ACTION PLAN A-2

ASYMMETRIC BLOWDOWN LOADS ON REACTOR PRIMARY COOLANT SYSTEM

Lead NRR Organization: Division of Operating Reactors

(DOR)

Lead Supervisor: Darrell G. Eisenhut,

Acting Director, DOR

Task Manager: Steve B. Hosford, EB/DOR

Applicability: Light-Water Reactors (PWRs)

PWR Project Completion Date: October 1979

A-1

1. DESCRIPTION OF PROBLEM

On May 7, 1975, the NRC was informed by Virginia Electric & Power Company

that an asymmetric loading on the reactor vessel supports resulting from

a postulated reactor coolant pipe rupture at a specific location (e.g.,

the vessel nozzle) had not been considered by Westinghouse or Stone and

Webster in the original design of the reactor vessel support system for

North Anna, Units 1 and 2. In the event of a postulated LOCA at the

vessel nozzle, asymmetric LOCA loading could result from forces induced

on the reactor internals by transient differential pressures across the

core barrel and by forces on the vessel due to transient differential

pressures in the reactor cavity. With the advent of more sophisticated

computer codes and the accompanying more detailed analytical models, it

became apparent to Westinghouse that such differential pressures, although

of short duration, could place a significant load on the reactor vessel

supports, thereby affecting their integrity. Although first identified

at the North Anna facility, this concern has generic implications for all

PWRs.

Upon postulation of a break in a reactor coolant pipe, at the above

mentioned locations, several rapidly occurring events could cause internal

and external transient loads to act upon the reactor vessel. For the

reactor vessel pipe break at the inlet nozzle, asymmetric pressure changes

take place in the annulus between the core barrel and the vessel.

Decompression could occur on the side of the vessel annulus nearest the

pipe break before pressure on the opposite side changes. The momentary

momentary difference in pressure across the core barrel could induce

lateral loads in opposite directions on the core barrel and the reactor

vessel. Vertical loads could also be applied to the core internals and

to the vessel due to the vertical flow resistance through the core and

asymmetric axial decompression of the vessel. Simultaneously, as fluid

escapes through the break, the annulus between the reactor vessel and

biological shield wall could become asymmetrically pressurized resulting

in a difference in pressure across the vessel causing additional horizontal

and vertical external loads on the vessel. In addition, the vessel could

be loaded by the effects of initial tension release and blowdown thrust

at the pipe break. The loads occur simultaneously. For a reactor vessel

outlet break the same type of loadings could occur, but the internal

loads would be predominantly vertical due to more rapid decompression of

the upper plenum.

For each of the above-mentioned postulated breaks, the time history of

the reactor vessel support reactions due to the complete set of simultaneous horizontal and vertical loads should be calculated.

In the event that such loadings would result in a significant degree of

failure within the reactor pressure vessel support system and consequent

vessel movement, there is a potential that this could (1) result in

damage to the ECCS lines, (2) affect the capability of the control rods

to function properly, and (3) result in damage to other reactor coolant

A-3
system components (pump and steam generator supports). In addition, the

differential pressures occurring during subcooled blowdown could result

in stresses on fuel assemblies caused by lateral core barrel and core

plate motion. This could degrade heat transfer capability if fuel spacer

grids are deformed by impacting either each other or the core baffle.

(This loading can occur independently of vessel support failure.)

The above-described phenomena may also exist to some degree in BWR plants

but is expected to have a lesser safety significance due to the lower

operating pressures. Currently, a plan of resolution for the BWR plants

is being developed by DOR:EB and will be considered as a separate issue

outside the scope of this TAP.

2. PLAN FOR PROBLEM RESOLUTION

A. Background

Foilowing disclosure of this problem during the OL review of North

Anna Units 1 and 2, a survey of all operating PWR reactors was

conducted in October 1975. That survey showed that neither of the

above described transient differential pressures had been considered

in the design of the reactor vessel supports for any operating PWR

facility.

In June 1976, the NRC requested all operating PWR licensees to

proceed to assess the adequacy of the reactor vessel supports at

their facilities with respect to these newly-identified loads. Most

licensees having a common NSSS vendor took identical or similar

positions with respect to this request and did not respond as

requested.

Most licensees with Westinghouse plants proposed an augmented inservice

inspection program (ISI) of the reactor vessel safe-end-to-end pipe

welds in lieu of providing the detailed analysis we requested.

Licensees with Combustion Engineering plants submitted a probability

study (prepared by Science Applications, Inc.) in support of a

conclusion that a break at this location has such a low probability

of occurrence that no further analysis is necessary. B&W licensees

have engaged Science Applications, Inc. for a similar study.

When the W and CE owners group reports were received in September

1976, DOR formed a review Task Group consisting of members from DOR,

DSS and EDO to evaluate these alternate proposals. In addition,

EG&G Idaho, Inc. was contracted to perform an independent review of

the submitted probability study. A short review schedule was

established since it appeared that most licensees would hold off on

further analysis pending our consideration of their submittals.

Our review of the proposed alternates has been completed. The Task

Group and EG&G independently reached the same conclusion: that the

alternate proposals set forth in these reports should not be accepted

A-4
in lieu of the requested analyses. The basis is that a sufficient

data base does not exist within the nuclear industry to provide

satisfactory answers to many information needs we identified. This

information would be needed to support this "no-break" approach.

Further investigation of pipe break probabilities is planned by the

staff, see item (3) below.

B. Plan

(1) Letters will be sent to all licensees and applicants stating

that an analysis must be undertaken to assess the design adequacy

of the reactor vessel supports and other structures to withstand

the loads when asymmetric LOCA forces are taken into account.*

We will point out that it may be possible to group plants such

that only a limited number of plants need be analyzed, and that

it may be possible to provide a simple "fix" (e.g., pipe

restraints) which will permit bounding the problem. Therefore,

the letters will request licensees and applicants to submit

their schedule for completion of the task.

(2) The staff will meet with the licensees constituting both the W

and the CE owners group to explain why the probability study

reports could not be accepted. We will also provide them all

the questions that have been generated to date as a result of

our review of the W and CE topical reports. (We will not issue

formal requests for additional information on these topicals to

these groups of licensees.)

(3) We will review and approve vendor models and codes prior to

plant-specific application to the extent possible. The staff

or its consultants will audit the calculations on a plantspecific basis where a pre-approved code is not used (the

case-specific reviews are not part of this TAP).

(4) The staff will develop explicit guidelines and acceptance

criteria for the asymmetric LOCA load analysis, including load

combinations and acceptable alternatives where, depending on

the construction or operating status of a given plant, application of the guidelines per se could require modifications

that are judged to be a practical impossibility. Such alternative guidelines would be designed to provide an adequate and

acceptable LOCA load generic issue consistent with safe plant

shutdown requirements.

(5) The staff will conduct a pipe break probability study that will

encompass (a) advances that are being made in nondestructive

Including an assessment of asymmetric loads produced by large pipe breaks

outside the reactor vessel cavity.

A-5
examination techniques, (b) an improved flaw distribution data

base of actual NSSS materials, and (c) development of a realistic

break opening model to describe pipe break characteristics.

The pipe break probability study will be used to confirm the

adequacy of staff decisions related to the conLinued operation

of plants for the interim period until an analysis of these

loads is conducted.

(6) The staff will perform a series of sensitivity studies to

independently evaluate the effect of noding upon the magnitude

and distribution of pressures within typical reactor cavity

designs. Results of sensitivity studies will be utilized to

prepare guidelines for the evaluation of the volumes within the

confines of the reactor cavity.

3. BASIS FOR CONTINUED PLANT OPERATION AND LICENSING PENDING COMPLETION OF

TASK

The safety issue addressed by this Task Action Plan (TAP) is primarily

applicable to pressurized light water reactors (PWRs). As discussed in

Section 1, the potential asymmetric loading conditions for boiling water

reactors (BWRs) appear to be of a lesser safety significance. The

potential loads are not expected to be as large, since the reactor vessel

pressure and sub-cooling are substantially lower than for PWR plants

(i.e., 1000 psia and approximately 40 Btu/lb, respectively for BWRs

compared to 2200 psia and 100 Btu/lb for PWR plants).

With respect to PWRs, each NSSS vendor (i.e., Westinghouse, Babcock and

Wilcox, and Combustion Engineering) has submitted topical reports which

describe analysis methods for assessing the internal asymmetric blowdown

loads. In accordance with the schedule for completion of this TAP, staff

review and approval of these topical reports will be completed by December

1978 (the Westinghouse analysis methods have already been approved by the

staff).

Topical reports have not been submitted by all the vendors and architect

engineers that deal with asymmetric cavity and reactor vessel external

loads in PWRs. These reviews are conducted by the staff on a case by

case basis. Technical assistance programs are underway to resolve those

aspects related to the reactor vessel cavity asymmetric loads and structural

response methods and criteria. It is expected that these programs will

be completed in FY 79 and will provide acceptable methods and models for

independent staff analysis. Current staff evaluations include case-by-case

reviews of applicants' mass and energy blowdown calculations, reactor

cavity loads including pressurization and vessel internal loads, and the

resulting structural response. The results and conclusions drawn from

these ongoing staff activities are subject to confirmation in the course

of development of the generic methods and models that will be the output

of this generic task.

A-6
For PWRs currently under licensing review for a Construction Permit, the

staff is requiring applicants to commit to address this safety issue as

part of the subsequent applications for Operating Licenses. The CP applicants are also required to perform plant-specific analyses of mass and

energy release and reactor cavity pressurization for purposes of establishing asymmetric design loads at the CP stage while the generic methods

are being reviewed and developed as part of this task. Since staff

approval of the analysis methods is anticipated well in advance of the

time for consideration of this safety issue at the Operating License review

stage for plants now under CP review and since necessary modifications,*

if any, to the plant design can be accomplished while the plant is being

constructed, the staff has concluded that, pending completion of this

task, Construction Permits can be granted with reasonable assurance that

(1) there will be a satisfactory resolution of this concern prior to

operation, and (2) operation will not present undue risk to the health

and safety of the public.

CP applicants are required to perform plant specific analyses using the

best available methodology for their designs. Also, the staff reviews

and performs audit calculations to ensure the reasonableness of the

currently available models; e.g., mass and energy blowdown calculations,

cavity pressure loads, and methodology for load combinations,

For PWRs under review for an Operating License, the staff will require

applicants to perform a plant specific analysis of their facility using

the best available methodology and criteria for their design and to

implement any design modifications which are necessary prior to the

issuance of an Operating License. Applicants are also required to commit

to perform an evaluation of their facility following staff approval of

the generic analysis method should the methodology or criteria developed

from this task warrant such reevaluation. Based upon this commitment and

the rationale presented below for continued operation of licensed facilities, we have concluded that, pending completion of this task, Operating

Licenses can be granted for facilities within this category with reasonable

assurance that operation will not present an undue risk to the health and

safety of the public.

*Such modifications, i.e., installation of physical restraints to limit the

postulated break area within the reactor cavity and/or to provide additional

piping support, have been implemented late in the construction stage of several

facilities (North Anna Units Nos. 1 and 2, Farley Unit No. 1). Modifications

of this type can also be accomplished after construction is completed, as

evidenced by the modifications proposed by the licensee for Indian Point Unit

No. 3, an operating facility. It should be noted, however, that physical

space (access) limitations at certain older operating reactors may preclude

the accomplishment of such modifications.

A-7
For PWRs currently licensed for operation, we have concluded that there

is reasonable assurance that continued operation, pending completion of

this task, does not constitute an undue risk to the health and safety of

the public for the following reasons.

As discussed below, the likelihood of occurrence of an initiating event

of sufficient magnitude to seriously challenge the structural adequacy of

the vessel support members or other structures is low.

The disruptivegailure of a reactor vessel itself has been estimated to

lie between 10 and 10 per reactor year -- so low that it is not

considered as a design basis event. The rupture probability of pipes is

estimated to be higher. The data base used by WASH-1400 indicates a

median value of 10 - for LOCA initiating ruptures per plant-year for al•

pipe sizes 6" and greater (with a lower and upper bound of 10 and 10

respectively). We believe that considering the large size of the pipt in

question4 (up to 50" O.D. and 4-1/8" thick), a median value nearer 10

than 10 is more appropriate using the same data base. In addition, the

quality control of the piping used in nuclear power plants is somewhat

better than that of the piping used in the WASH-1400 data base.

Because (1) the break of primary concern must be large (and therefore of

low probability); (2) only certain break locations lead to high loads;

and (3) these welds are currently subject to preservice and inservice

inspection by volumetric and surface techniques in accordance with ASME

Code Section XI, we conclude that the probability of a pipe break resulting

in substantial transient loads on the vessel support system or other

structures is acceptably small and that reactor operation and licensing

of facilities for operation can continue during the interim period of

approximately 2 years while this matter is being resolved.

It is anticipated that the plant-unique analyses, which Will be performed

following staff approval of the generic analysis methods, will indicate

that design modifications may be necessary to restore the originally

intended safety margins at certain operating facilities. Such modifications, i.e., installation of physical restraints to limit the postulated

break area within the reactor cavity, have already been shown to be

feasible. In the event that design modifications are judged to be

impossible for specific operating facilities, alternative solutions,

including such things as augmented inservice inspection, will be required

to ensure adequate safety margins.

4. NRR TECHNICAL ORGANIZATIONS INVOLVED

A. Analysis Branch, Division of Systems Safety. Has lead responsibility

for review of vendor hydrodynamic analysis methods and codes.

Manpower Estimates: 0.5 man-months in FY 1979

B. Core Performance Branch, Division of Systems Safety. Has lead

responsibility for reviewing vendor analysis methods for calculating

loads on fuel assemblies resulting from decompression for plants

under CP and OL review.

A-8
Manpower Estimates: 1.75 man-months in FY 1979

C. Containment Systems Branch, Division of Systems Safety. Responsible

for reviewing vendor models and methods for calculating asymmetric

cavity loads for all plants, and associated vendor models.

Manpower Estimates: 4.6 man-months in FY 1979

D. Mechanical Engineering Branch, Division of Systems Safety. Responsible

for review of structural aspects of vendor analysis methods and

codes for plants not licensed for operation. Responsible for

developing structural acceptance criteria (with Engineering Branch,

DOR).

Manpower Estimates: 1.5 man-months in FY 1979

E. Engineering Branch, Division of Operating Reactors. Responsible for

review of structural aspects of analysis methods and codes applicable to operating reactors (including loads on fuel assemblies).

Responsible for development of structural acceptance criteria (with

Mechanical Engineering Branch, DSS).

Manpower Estimates: 1.0 man-months in FY 1979

F. Engineering Branch, Division of Operating Reactors. Responsible for

the coordination and management of this Technical Activity.

Manpower Estimates: 1.5 man-months in FY 1979

5. TECHNICAL ASSISTANCE

A. Managed by DOR (Engineering Branch):

Contractor: EG&G Idaho, Inc.

Funds Available: $70K in FY 1979.

This is an NRC program to independently model representative Westinghouse 4-loop (Indian Point 3), B&W (Arkansas Nuclear One Unit 1),

and CE (St. Lucie 1) plants for the purpose of assessing the loads

on all major structures and components resulting from asymmetric

LOCA loads. The purpose of this program is to develop an independent

NRC capability for performing inelastic dynamic analyses. Sensitivity

studies will be performed to evaluate the effects of various break

opening times, effects of component stiffness, and three-dimensional

coupling effects.

B. Managed by DSS (Mechanical Engineering Branch):

Contractor: EG&G Idaho, Inc.

Funds Available: 80K in FY 1979

A-9
This is an NRC/DSS program to provide the staff with the analytical

tools necessary to independently verify the selection of design

basis pipe rupture locations; and to verify that the criteria for

assurance of integrity under LOCA & SSE loads for reactor coolant

piping, the reactor vessel, steam generators, main coolant pumps and

the supports for these components have been implemented correctly.

Verification analyses for a B&W plant (Erie), and a BWR plant

(Zimmer 1) will be run to verify results reported by the applicants.

Support models will be designed to be revised as necessary to

represent various support configurations utilized by Architect/

Engineers of the plants under CP/OL review.

C. Managed by DSS (Core Performance Branch)

Contractor: EG&G, Inc.

Funds Available: 60K in FY 1979

This is an NRC/DSS program to perform independent calculations of

fuel assembly response from LOCA as well as seismic loads. Audit

calculations are performed to assist NRC in reviewing vendor

analytical methods. Evaluation on Westinghouse and General

Electric Co. analytical methods for fuel assembly response have been

completed.

In addition, the contractor will assist DSS/CPB in developing

acceptance criteria for the fuel assembly evaluation.

6. INTERACTIONS WITH OUTSIDE ORGANIZATIONS

A. W Owners Group of licensees

The W owners group of licensees is an ad hoc organization of most

(but7not all) owners of operating W pliants, formed for the purpose

of sponsoring and proposing the augmented inservice inspection

program (WCAP-8802) in lieu of furnishing the detailed analysis

requested by NRC. This group of licensees has engaged Westinghouse

Electric Corporation as its principal consultant.

With the advent of the NRC decision to request all licensees for a

detailed analysis and to set aside - at least for the present--the

ISI proposal, the continued role of this licensee group is undetermined.

B. CE Owners Group of Licensees

The CE owners group of licensees is also an ad hoc organization of

most owners of operating CE plants. This group sponsored the probability study prepared by Science Applications, Inc., which concluded

that the probability of severe pipe breaks that could trigger the

loads under consideration is below the threshold of concern. The

future role of this licensee group is also undetermined.

A- 10
C. B&W Owners Group of Licensees

This group is composed of owners of B&W plants having nuclear steam

supply systems of the same design (177 fuel assemblies, skirt supported

vessels.) This group has engaged SAI and B&W as its consultant for

the preparation of a probability study similar to the one done by

SAI for the CE owners group. This report has not yet been submitted.

D. ACRS

This task is closely related to one of the generic items identified

by the ACRS and, accordingly, will be coordinated with the Committee

as the task progresses.

7. ASSISTANCE REQUIREMENTS FROM OTHER NRC OFFICES

None.

8. POTENTIAL PROBLEMS

A. Three owners groups representing most operating PWRs have been

formed and either will propose or have proposed solutions different

from the requested analysis (augmented ISI, probability studies).

Therefore, strong industry resistance to our request for some form

of analysis is possible.

B. Rigorous application of the generic structural acceptance criteria

may require modifications that are judged to be impossible for some

older plants. For these cases, alternative solutions may be required.

A-I1

APPENDIX B

STATUS OF OPERATING PLANT EVALUATION

November 30, 1980

Projected

Licensee Safety

Evaluation Evaluation

Plant Submitted Schedule

Arkansas 1 Yes 5/81

Arkansas 2 Yes 10/81

Beaver Valley 1 Yes 5/81

Calvert Cliffs 1 Yes 7/81

Calvert Cliffs 2 Yes 7/81

Cook 1 Yes 3/81

Cook 2 Yes 3/81

Crystal River 3 Yes 5/81

Davis-Besse 1 Yes 5/81

Farley 1 Yes 9/81

Fort Calhoun 1 Yes 7/81

Ginna Yes 3/81

Haddam Neck Yes 3/81

Indian Point 1* No --

Indian Point 2 Yes 12/80

Indian Point 3 Yes 12/80

Kewaunee** Yes 5/81

Maine Yankee** Yes 10/81

Millstone 2 Yes 7/81

North Anna Yes 5/81

Oconee 1 Yes 5/81

Oconee 2 Yes 5/81

Oconee 3 Yes 5/81

Palisades Yes 7/81

Point Beach 1 Yes 3/81

Point Beach 2 Yes 3/81

Prairie Island 1 Yes 5/81

Prairie Island 2 Yes 5/81

Rancho Seco Yes 5/81

Robinson 2 Yes 3/81

Salem 1 Yes 9/81

Salem 2 Yes 9/81

San Onofre 1 Yes 3/81

St. Lucie 1 Yes 5/81

Surry 1 Yes 3/81

Surry 2 Yes 3/81

TMI-1 Yes 5/81

TMI-2 Yes 5/81

Trojan Yes 9/81

Turkey Point 3 Yes 3/81

Turkey Point 4 Yes 3/81

Yankee Rowe 1 Yes 3/81

Zion 1 Yes 3/81

Zion 2 Yes 3/81

* Shut down.

** Modification complete.

B-1

APPENDIX C

CHARACTERISTICS OF PIPE BREAKS

The staff performed a study at EG&G to investigate pipe-break area and time

characteristics. Pipe-break flow area and the time associated with development

of flow area are major considerations in the development of asymmetric LOCA

loads. Historically, the staff has required conservative assumptions in these

areas to be used in the development of loading transients. A task in the

generic action plan A-2 required the staff to investigate realistic characteristics of pipe breaks.

The study performed by EG&G included a review of pipe-break models submitted

by each PWR NSSS vendor and the development of pipe-break characteristics at

four specific plants. A summary of the EG&G report* and its conclusions documenting that study are presented below.

This report documents an analysis performed by EG&G Idaho on break-opening

times and associated flow areas due to pipe separation during a postulated

loss-of-coolant accident (LOCA). This effort supports the resolution of the

U.S. Nuclear Regulatory Commission (NRC) Category A Task Action Plan A-2 on

asymmetric LOCA loads and provides NRC Division of Operating Reactors (DOR)

the capability of independently verifying vendor analyses.

A postulated LOCA for a nuclear-power plant is initiated by a severance,

or break, in the piping of the primary coolant system, inducing severe

transient loads on piping, major components, and supports in the system.

The severity of the depressurization transient is directly dependent upon

the combination of break size and length of time required for pipe separation.

In general, LOCA thermal hydraulic analyses have taken the conservative

approach by assuming break openings to occur instantaneously (that is,

10 msec or less). If it can be shown that this approach is unrealistic,

(that is, break openings are not instantaneous), the loads associated with

the LOCA transients may be greatly reduced.

In this report, break-opening times and areas are examined on a plantspecific basis. The analysis covers four nuclear-power plants designed

by three'different vendors: San Onofre Unit 2 and St. Lucie Unit 1

designed by Combustion Engineering (CE), Arkansas Nuclear One Unit 1

designed by Babcock and Wilcox (B&W), and Indian Point Unit 3 designed by

Westinghouse. These four plants were chosen for two reasons: (1) mechanical

dynamic response analyses of all four plants have previously been performed

for postulated LOCA applied loads by EG&G Idaho in support of NRC licensing

audits, and (2) the four plants represent a good cross-section of nuclearplant design. Information from the previous analyses was utilized in

*R. F. Lippert, EG&G Idaho, Inc., "Study of Break Opening Area/Time Determination

for Postulated LOCA Guillotine Pipe Ruptures of Nuclear Primary Coolant Systems,"

RE-A-79-093, July 1979.

C-i
expediting development of models of the necessary segments of the primary

coolant loops and in providing appropriate hydraulic transients. The

structural models include material and geometric nonlinearities when

applicable.

Subsequent sections of the report include: (1) an investigation of

various techniques for geometrically calculating break-opening areas, (2)

a description of the several analytical structural models employed in the

analysis, (3) a description of the applied LOCA loadings associated with

the various postulated pipe breaks analyzed, and (4) a presentation of

the resulting pipe-break areas as a function of time and associated

observations. In addition, the appendices include vendor break-area

procedures, plant geometries, and plots of break-area time-history results.

The EG&G analysis of four nuclear plants to study break opening areas and

times for postulated LOCA pipe ruptures presented in this report provides the

following observations. Times for break-opening areas to develop are significantly greater than the 1- and 1O-msec times presently assumed, and, in the

cases of limited breaks, the break areas developed were significantly less

than those to which the plants were designed. Both results indicate significant

conservatism in the design and analysis approaches generally employed. For

the results of this study to be applicable to nuclear-power plants other than

those analyzed herein, careful consideration must be given to the particular

plant's size, power, dimensions, and geometry as well as to the type of

structural analysis and LOCA load development.

C-2
APPENDIX D

LETTERS TO APPLICANTS/LICENSEES

D-1

-0 UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D. C. 20555

All PWR Licensees

Gentlemen:

In October of 1975, the NRC staff notified each licensee of an operating

PWR facility of a potential safety problem concerning the design of the

reactor pressure vessel support system. Those letters requested each

licensee to review the design basis for the reactor vessel support system

for each of its PWR facilities to determine whether certain transient

loads, which were described in the enclosure to the letter, had been

appropriately taken into account in the design. Furthermore, these

letters indicated that, on the basis of the results of licensees' reviews,

a reassessment of the reactor vessel support design for each operating

PWR facility may be required.

Licensee responses to that request indicated that these postulated

asymmetric loads have not been considered in the design basis for

the reactor vessel support system, reactor internals including the

fuel, steam generator supports, pump supports, emergency core cooling

system (ECCS) lines, reactor coolant system piping, or control rod

drives.

Subsequently in June 1976, the NRC staff informed each PWR licensee

that a reassessment of the reactor vessel support system design for

each of its facilities was required. While the emphasis of these

letters was primarily focused on the need to reassess the vessel

support design for transient differential pressures in the annular

region between the reactor vessel and the cavity shield wall and

across the core barrel, we indicated that our generic review may

extend to other areas in the nuclear steam supply system (NSSS) and

that further evaluation may be required.

For your information, Enclosure 1 is a summary of the background

and current status of our review efforts related to this generic

concern.

D-3
All PWR Licensees -2- January 20, 1978

We have now determined that an assessment of the potential for

damage to other NSSS component supports (e.g., steam generator and -

pump supports), the fuel assemblies, control rod drives, and ECCS

piping attached to the reactor coolant system due to loadings

associated with postulated coolant system piping breaks is required.

Our request for additional information transmitted to you in June 1976

has been revised both to clarify our original request and to identify

the extension of our concerns to other areas in the NSSS, as

identified above. A copy of this revised request for additional

information is provided as Enclosure 2.

The revised request for additional information identifies a requirement

that your assessment of potential damage to the reactor vessel and other

NSSS component supports, reactor vessel, fuel and internals, attached ECCS

lines and the control rod drives should include consideration of breaks

both inside and outside of the reactor pressure vessel cavity. This

assessment should be made for postulated breaks in the reactor coolant

piping system, (secondary systems are not to be included), including the

following locations:

a) Reactor vessel hot and cold lea nozzle safe ends

b) Pump discharge nozzle

c) Crossover leg

d) Hot leg at the stean generator (B&W and CE plants only)

A number of licensees, have presented to the NRC staff alternate proposals,

other than to conduct a detailed analyses, to resolve this concern. Based

upon our review of these proposals, we have concluded that these alternative

proposals do not establish an acceptable basis for long term operation

without a detailed assess-nent of the risk resulting from these postulated

transient loading conditions. We have, however, concluded that the low

probability for occurrence of an event which could result in these loads

establishes an adequate basis to justify continued operation for a short

term period.

The IIRC staff will consider an analysis that is applicable to more than

one specific plant if it can be adequately demonstrated that such an

analysis is either representative or bounding for each plant concerned.

Additional guidance regarding loading combinations (safe shutdown earthquake

loads, loss of coolant accident loads), will be provided by about March 1,

1978, following the conclusion of staff investigations in this area.

D-4
All PWR Licensees -3- January 20, 1978

Please respond within 30 days of receipt of this letter, indicating your

intent to proceed with an evaluation of the overall asymmetric loss of

coolant accident (LOCA) loads as described herein. In addition, please

submit to us, within 90 days, your detailed schedule for providing the

required evaluation. Your schedule should be consistent with our desire

to resolve this problem within two years and should clearly state your

intent to demonstrate the safety of lono term continued operation.

We are transmitting information copies of this letter to the Westinghouse,

Combustion Engineering and Babcock & Wilcox Companies. If you have any

questions or want any clarification on this matter, please call your NRC

Project Manager.

Sincerely,

Victor Ste Director

Division of Operating Reactors

Office of Nuclear Reactor Regulation

Enclosures:

1. Background and Current Status

2. Revised Request for Additional

Information

cc w/enclosure:

See attached listing

D-5
January 20, 1978

ENCLOSURE 1

BACKGROUND AND CURRENT STATUS OF THE NRC STAFF REVIEW

OF ASYMMETRIC LOCA LOADS FOR PWP FACILITIES

On May 7, 1975, the NRC was informed by Virginia Electric & Power Company

that an asymmetric loading on the reactor vessel Supports resulting from

a postulated reactor coolant pipe rupture at a specific location (e.g.,

the vessel nozzle) had not been considered by Westinghouse or Stone &

Webster in the original design of the reactor vessel support system for

North Anna, Units 1 and 2. It had been identified that in-the event of

a postulated instantaneous, double-ended offset LOCA at the vessel nozzle,

asymmetric loading could result from forces induced on the reactor internals by transient differential pressure across the core barrel and by

forces on the vessel due to transient differential pressures in the reactor

cavity. With the advent of more sophisticated computer codes and the

accompanying mnore detailed analytical models, it became apparent that such

differential pressures, although of short duration, could place a significant load on the reactor vessel supports and on other components, thereby possibly affecting their integrity. Although this potential safety

concern was first identified during the review of the North Anna facilities,

it has generic implications for all PWRs.

Upon closer examination of this situation, it was determined that postulated breaks in a reactor coolant pipe at vessel nozzles were not the only

area of concern but rather that other pipe breaks in the reactor coolant

system could cause internal and external transient loads to act upon the

reactor vessel and other com1ponents. For the postulated pipe break in

the cold leg, asymmetric pressure changes could take place in the annulus

between the core barrel and the vessel. Decomnpression could occur on the

side of the vessel annulus nearest the pipe break before the pressure on

the opposite side of the vessel changes. This momentary differential

pressure across the core barrel could induce lateral loads both on the

core barrel and on the reactor vessel. Vertical loads could also be

applied to the core internals and to the vessel due to the vertical flow

resistance through the core and asymmetric axial decompression of the

vessel. Simultaneously, for vessel nozzle breaks, the annulus between

the reactor and biological shield wall could become asymmetrically

pressurized resulting in a differential pressure across the vessel

causing additional horizontal and vertical external loads on the vessel.

In addition, the vessel could be loaded by the effects of initial tension release and blowdown thrust at the pipe break. These loads could

occur simultaneously. For a reactor vessel outlet break, the same type

of loadings could occur, but the internal loads would be predominantly

vertical due to more rapid decompression of the upper plenum.

D-6
-2-

Although the NRC staff's original emphasis and concern were focused

primarily on the integrity of the reactor vessel support system with

respect to postulated breaks inside the reactor cavity (i.e., at a

nozzle), it has since become apparent that significant asymmetric forces

can also be generated by postulated pipe breaks outside the cavity

and that the scope of the problem is not limited to the vessel support

system itself. For such outside-cavity postulated breaks, the

aforementioned concerns, such as the integrity of fuel assemblies and

other structures, need to be examined.

In June 1976, the NRC requested all operating PWR licensees to evaluate

the adequacy of the reactor system components and their supports at

their facilities with respect to these newly-identified loads.

In response to our request, most licensees with Westinghouse plants

proposed an augmented inservice inspection program (ISI) of the

reactor vessel safe-end-to-end pipe welds in lieu of providing an

evaluation of postulated piping failures. Licensees with Combustion

Engineering plants submitted a probability study (prepared by Science

Applications, Inc.) in support of their conclusion that a break at a

particular location (vessel nozzle) has such a low probability of

occurrence that no further analysis is necessary. A similar study has

been recently submitted by Science Applications, Inc. (SAI) for B&W

plants.

When the Westinghouse and CE owners group reports were received in

September 1976, the NRC formed a special review task group to evaluate

these alternative proposals. In addition, EG&G Idaho, Inc., was

contracted to perform an independent review of the SAI probability

study submitted for the CE owners group.

This review effort resulted in a substantial number of questions

which previously have been provided to representatives of each group.

Based on the nature of these questions and other factors to be

discussed later in this report, we cannot accept these reports in

their present form as a resolution for the asymmnetric LOCA load generic

issue. Based on our review, we have concluded that a sufficient data

base does not presently exist within the nuclear industry to provide

satisfactory answers to these information needs. Several long-term

experimental programs would be required to provide much of this

information. Although the probability study recently submitted by

SAI for certain B&W owners does respond to some of the informal

questions raised during our review of the SAI report prepared by

CE plants, the more fundamental questions remain. Therefore, this

conclusion also applies to the SAI topical report for B&W plants

(SAI-050-77-PA).

b-7
-3-

A second - and equally important - reason for not acceptinq probability/ISI

approaches as a solution at this point concerns our need and industry's need

to oain a better understandina of the problem. We consider it essential

that an understandinq of the important breaks and associated conseauences

be known before applyinq any remedy - be it pipe restraints, Drobability,

ISI, or some combination of these measures. Only in this way will we have

a basis on which to judge the importance of the remedy with respect to

what it is desianed to prevent.

Although we have many questions on each of these tooical reports, this

does not mean that we view the probabilistic/ISI approach as completely

without merit. In fact, the results of a probabilistic evaluation serves

as the basis for continued operation and licensing of nuclear plants

during this interim period while additional evaluations can be performed

by vendors and licensees.

We believe that the justification for continued plant operation has as its

basic foundation the fact that the event in question, i.e., a hypothetical

double-ended instantaneous rupture of the main coolant pipe at a particular

location, has a very low probability of occurrence.

The disruptive failure probagility of 7a reactor vessel itself has been

estimated to lie between 10 and 10' per reactor year - so low that it

is not considered as a design basis event. The rupture probability of4

pipes is estimated to be hiqher. WASH-1400 used a median value of 10"4

for LOCA initiating ruptures per plant-year Ior all pipes sizes 6" and

qreater (with a lower and upper bound of 1O" and 10- , respectively).

We believe that considerino the large size of the pipes in question (up

to 50" O.D. and 4-1/8" thick), the lower bound is more appronriate since

these pipes are more like vessels in size. In addition, the quality control of this pipinq is the best available and somewhat better than that

of the piping used in the WASH-1400 study.

These factors, coupled with the facts that (1) the break of primary concern must be very laroe, (2) it must occur at a specific location, (3)

the break must occur essentially instantaneously, and (4) these welds

are currently subject to inservice inspection by volumetric and surface

techniques in accordance with ASME Code Section XI, lead us to conclude

that the probability of a pipe break resulting in substantial transient

loads on the vessel support system or other structures is acceptably

small such that continued reactor operation and continued licensinq of

facilities for operation can continue while this matter is beinq

resolved.

D-8
-4-

In support of the above, the staff has developed a short-term interim criterion to determine if an acceptable level of safety exists for operating

PWRs under conditions of a postulated pipe break. This interim criterion

is based on a simplified probabilistic model that incorporates elastic fracture mechanics techniaues to estimate the probability of a pipe break.

Critical flaw size and subcritical flaw qrowth rates were determined assumina

the presence of a surface flaw located in a circumferential weld of a

thick-walled pipe, Determination of the critical flaw size was based on

an estimated fracture toughness value of KIC at a minimum temnerature of

200 F and a uniform tensile stress equal to the consideration of various

operatino conditions producinq elastically calculated stresses ranaina in

value from I to 3 times the material minimum yield strenoth.

Then, using the calculated critical flaw size, the subcritical orowth

rate, and an estimated probability distribution of an undetected flaw in

thick-walled pipe wilds, the upoer bound probability of pipe break was

estimated to be 10- . This value is also supported by a recent publication by Dr. S. H. Brsh* which states that actual failure statistics

confirm rates of 10- to l0- per reactor-year in large pipes, with

higher rates as the pipe size decreases. Considerina these analyses,

we conclude that our conservative estimate on a pipe break in the primary

coolant system is in the ranae of 10" to 10'. This estimated pipe

break probability is considered acceptably low to justify short-term

operation of nuclear power plants.

In view of all previous discussions concerning this issue, the NRC staff

has concluded that an evaluation mujst be undertaken to assess the design

adeouacy of the reactor vessel supports and other affected structures and

systems to withstand asymmetric LOCA loads, includinn an assessment of

the effects of asymmetric loads produced by various pipe breaks both

inside and outside the reactor cavity. On nerforming these evaluations

the staff will permit the grouping of plants, where adeouate justification for such grouping exists, in order to limit the number of plants

to be analyzed. Alternatively, the staff will permit the analyzina of

a "prototypical" plant, which is sufficiently representative of a

group of plants, to provide the necessary information. Both of these

concepts have been discussed with the Westinghouse and CE Owners Groups,

and we believe that such approaches could save a significant amount of

time and effort in obtaining results on which to base any needed

corrective measures. The NRC staff is prepared to meet with PWR licensees to discuss such approaches, and has already done so. For example,

we met with the Westinahouse owners croup on October 19, 1977 for the

purpose of discussing a generic solution for breaks outside the reactor

cavity. It is expected that a similar meeting will be held in the near

*"Critical Factors in Blowdown Loads in the PWR Guillotine Nozzle

Break (Volume 2 - the Asymmetric Load Problem)", dated June 6, 1977

D-9
-5-

future to address breaks located inside the cavity. This "phased"

approach is acceptable to us, provided that it sheds light on and

serves to expedite consideration of the more limiting insidecavity breaks.

For your information, the NRC has a technical assistance contract with

EG&G Idaho, Inc., to independently model representative Westinghouse,

B&W, and CE plants for the purpose of assessing the loads on all major

structures and components resulting from asymmetric LOCA loads. We

believe that the results of this program which will include sensitivity

studies, will provide significant confirmatory information related to

this generic safety concern.

Although, as stated earlier, we believe that continued operation and

licensing of facilities for the short ter' is justified, we also believe

that efforts to resolve this issue should proceed without delay, with

the objective of both completing the necessary assessments and installing

any necessary plant modifications within two years. In making this statement, we wish to make it clear that plant modifications, if indicated by

licensee assessments, is the preferred approach. At the same time, we

recognize that there may be cases wherein appropriate modifications may

be judged to be unwarranted based on the consideration of overall risk.

In such cases, and only in such cases, we will be prepared to give further

consideration to alternate approaches, such as probability ISI, We feel,

however, that ISI techniques as they exist today could be considerably

improved, and, to the extent that such improvements could have a direct

bearing on this problem as well as an impact of nuclear safety in general,

we would welcome their development.

D-1O
January 20, 1978

ENCLOSURE 2

REVISED REQUEST FOR ADDITIONAL INFORMATION

Recent analyses have shown that certain reactor system components and

their supports may be subjected to previously underestimated asymmetric

loads under the conditions that result from the postulation of ruptures

of the reactor coolant piping at various locations. It is therefore

necessary to reassess the capability of these reactor system components

to assure that the calculated dynamic asymmetric loads resulting from

these postulated pipe ruptures will be within the bounds necessary to

provide high assurance that the reactor can be brought safely to a cold

shutdown condition. For the purpose of this request for additional information the reactor system components that require reassessment shall

include:

a. Reactor Pressure Vessel

b. Fuel Assemblies, Including Grid Structures

c. Control Rod Drives

d. ECCS Piping that is Attached'to the Primary Coolant Piping

e. Primary Coolant Piping

f. Reactor Vessel, Steam Generator and Pump Supports

g. Reactor Internal.s

h. Biological Shield Wall and Neutron Shield Tank (where applicable)

i. Steam Generator Compartment Wall

The following information should be included in your reassessment of the

effects of postulated asymmetric LOCA loads on the above-mentioned reactor

system components and the reactor cavity structure.

1. Provide arrangement drawings of the reactor vessel, the steam generator

and pump support systems to show the geometry of all principal elements

and materials of construction.

2. If a plant-specific analysis will not be submitted for your plant,

provide supporting information to demonstrate that the generic plant

analysis under consideration adequately bounds the postulated accidents

at your facility. Include a comparison of the geometric, structural,

mechanical and thermal hydraulic similarities between your facility

and the case analyzed. Discuss the effects of any differences.

3. Consider postulated breaks at the reactor vessel hot and cold leg

nozzle safe ends, pump discharge nozzle and crossover leg that result in the most severe loading conditions for the above-mentioned

D-11
-2-

systems.* Provide an assessment of the effects of asymmetric pressure differentials on these systems/components in combination with

.all external loadings including asymmetric cavity pressurization for

both the reactor vessel and steam generator which might result from

the required postulate. This assessment should consider:

a. limited displacement break areas where applicable

b. consideration of fluid-structure interaction

c. use of actual time-dependent forcing function

d. reactor support stiffness

e. break opening times.

4. If the results of the assessment required by 3 above indicate loads

leading to inelastic action in these systems or displacement exceeding

previous design limits provide an evaluation of the following:

a. Inelastic behavior (including strain hardening) of the material

used in the system design and the effect on the load transmitted

to the backup structures to which these systems are attached.

5. For all analysis performed, include the method of analysis, the structural and hydraulic computer codes employed, drawings of the models

employed and comparisons of the calculated to allowable stresses

and strains or deflections with a basis for the allowable values.

6. Provide an estimate of the total amount of permanent deforma-tion

sustained by the fuel spacer grids. Include a description of the

impact testing that was performed in support of your estimate.

Address the effects of operating temperatures, secondary impacts,

and irradiated material properties (strength and ductility) on the

amount of predicted deformation. Demonstrate that the fuel will

remain coolable for all predicted geometries.

7. Demonstrate that active components will perform their safety function

when subjected to the postulated loads resulting from a pipe break

in the reactor coolant system.

8. Demonstrate functionability of any essential piping where service

level B limits are exceeded.

In order to review the methods employed to compute the asymmetrical

pressure differences across the core support barrel during subcooled

portion of the blowdown analysis, the following information is requested:

*B&W and CE plant licensees shoild also consider breaks in tne hot leg

at the steam generator inlet.

D-12
-3-

1. A complete description of the hydraulic code(s) used including the

development of the equations being solved, the assumptions and

simplifications used to solve the equations, the limitations resulting from these assumptions and simplifications and the numerical

methods used to solve the final set of equations. Provide comparisons

with experimental data, covering a wide range of scales, to demonstrate

the applicability of the code and of the modeling procedures of

the subcooled blowdown portion of the transient. In addition, discuss

application of the code to the multi-dimensional aspects of the

reactor geometry.

If an approved vendor code is used to obtain the asymmetric pressure

difference across the core support barrel, state the name and version

of the code used and the date of the NRC acceptance of the code.

2. If the assessment of the asymmetric pressure difference across the

core support barrel is made without the use of a hydraulic blowdown

code, present the methodology used to evaluate the asymmetric loads

and provide justification that this assessment provides a conservative

estimate of the effects of the postulated LOCA.

A compartment multi-node, space-time pressure response analysis is

necessary to determine the external forces and moments on components.

Analyses should be performed to determine the pressure transient resulting

from postulated hot leg and cold leg reactor coolant system pipe ruptures

within the reactor cavity and any pipe penetrations. If applicable,

similar analyses should be performed for steam generator compartments

that may be subject to pressurization where significant component support

loads may result. This information can be provided to encompass a group

of similarly designed plants (generic approach) or a purely plant specific

(custom plant) evaluation can be developed. In either case, the proposed

method of evaluation and principal assumptions to be used in the analysis

should be provided for review in advance of the final load assessment.

For generic evaluations, perform a survey of the plants to be included

and identify the principle parameters which may vary from plant to plant.

For instance, this should include blowdown rate and geometrical variations in principle dimensions, volumes, vent areas, and vent locations.

A typical or lead plant should be selected to perform sensitivity and

envelope calculations. These analyses should include:

(1) nodal model development for the configuration representing the

most restrictive geometry; i.e., requiring the greatest nodalization;

(2) the most restrictive configuration regarding vent areas and

obstructions to flow should be analyzed; and,

(3) sensitivity to code data input should be evaluated; e.g., loss

coefficients, inertia terms, vent areas, nodal volumes, and any

other input data where there may be variations from plant to

plant or uncertainty for the given plant.

D ;-13
-4-

These studies should be directed at evaluating the maximum lateral and

vertical force and moment time functions, recognizing that models may

be different for lateral as opposed to vertical load definitions.

The following is the type of Unfarmation needed for both generic and

custom plant evaluations. Although this request was primarily developed

for reactor cavity analyses it may be applied to other component subcompartments by general application.

(1) Provide and justify the pipe break type, area, and location for

each analysis. Specify whether the pipe break was postulated for

the evaluation of the compartment structural design, component

supports design, or both.

(2) For each compartment, provide a table of blowdown mass flow rate

and energy release rate as a function of time for the break which

results in the maximum structural load, and for the break which

was used for the component supports evaluation.

(3) Provide a schematic drawing showing the compartment nodalization

for the determination of maximum structural loads, and for the

component supports evaluation. Provide sufficiently detailed

plan and section drawings for several views, including principal

dimensions, showing the arrangement of the compartment structure,

major components, piping, and other major obstructions and vent

areas to permit verification of the subcompartment nodalization

and vent locations.

(4) Provide a tabulation of the nodal net-free volumes and interconnecting

flow path areas. For each flow path, provide an L/A (ft-') ratio,

where L is the average distance the fluid flows in that flow path

and A is the effective cross sectional area. Provide and justify

values of vent loss coefficients and/or friction factors used to

calculate flow between nodal volumes. When a loss coefficient consists of more than one component, identify each component, its

value and the flow area at which the loss coefficient applies.

(5) Describe the nodalization sensitivity study performed to determine

the minimum number of volume nodes required to conservatively predict

the maximum pressure load acting on the compartment structure. The

nodalization sensitivity study should include consideration of

spatial pressure variation; e.g., pressure variation circumferentially,

axially and radially within the compartment. The nodal model

development studies should show that a spatially convergent differential pressure distribution has been obtained for the selected evaluation model.

D-14
-5-

Describe the justify the nodalization sensitivity study performed

for the major component supports evaluated, if different from the

structural analysis model, where transient forces and moments acting

on the components are of concern. Where component loads are of

primary interest, show the effect of noding variations on the

transient forces and moments. Use this information to justify

the nodal model selected for use in the component supports

evaluation.

If the pressurization of subvolumes located in regions away from the

break location is of concern for plant safety, show that the selection of parameters which affect the calculations have been conservatively evaluated. This is particularly true for pressurization of

the volume beneath the reactor vessel. In this case, a model which

predicts the highest pressurization below the vessel should be

selected for the evaluation.

NOTE: It has been our experience that for the reactor cavity, three

regions should be considered (i.e., nodalized) when developing

a total model. These are:

(1) the volume around or in the vicinity of the break location out to a radius approximated by the adjacent

nozzles, and including portions of the penetration volume

for some plants;

(2) the volume or region covering the upper reactor cavity,

primarily the RPV nozzles other than the break nozzle;

and

(3) the region encompassing the lower reactor cavity and

other portions of the reactor-cavity not included in

Items (1) and (2).

(6) Discuss the manner in which movable obstructions to vent flow (such

as insulation, ducting, plugs, and seals) were treated. Provide

analytical and experimental justification that vent areas will not

be partially or completely plugged by displaced objects. Discuss

how insulation for piping and components was considered in determining

volumes and vent areas.

(7) Graphically show the pressure (psia) and differential pressure (psi)

response as functions of time for a representative number of nodes

to indicate the spatial pressure response. Discuss the basis for

establishing the differential pressure on structures and components.

D-15
-6-

(8) For the compartment structural design pressure evaluation, provide

the peak calculated differential pressure and time of peak pressure

for each node. Discuss whether the design differential pressure

is uniformly applied to the compartment structure or whether it is

spatially varied. If the design differential pressure varies

depending on the proximity of the pipe break location, discuss how

the vent areas and flow coefficients were determined to assure

that regions removed from the break location are conservatively

designed, particularly for the reactor cavity as discussed above.

(9) Provide the peak and transient loading on the major components used

to establish the adequacy of the support design. This should include

the load forcing functions (e.g., fx(t), fv(t), fz(t)) and transient

moments (e.g., Mx(t), My(t), Mz(t)) as resolved about a specific,

identified coordinate system. The centerline of the break nozzle

is recommended as the X coordinate and the center line of the

vessel as the Z axis. Provide the projected area used to calculate

these loads and identify the location of the area projections on

plan and section drawings in the selected coordinate system. This

information should be presented in such a manner that confirmatory

evaluations of the loads and moments can be made.

.D-16
APPENDIX E

FUEL-SYSTEM DESIGN

(Appendix A to SRP 4.2)

E-1

NUREG-0S00

(Formerly NUREG-75/0W7)

R .- U.& NUCLEAR REGULATORY COMMISSION

1STANDARD REVIEW PLAN

OFFICE OF NUCLEAR REACTOR REGULATION

Appendix E

U.S. Nuclear Regulatory Commission

Office of Nuclear Reactor Regulation PROPOSED ADDITION OF

APPENDIX A

EVALUATION OF FUEL ASSEMBLY STRUCTURAL RESPONSE

TO EXTERNALLY APPLIED FORCES

TO

STANDARD REVIEW PLAN PSRP-4.2, REVISION 2, DRAFT 1

A. BACKGROUND

Earthquakes and postulated pipe breaks in the reactor coolant system

would result in external forces on the fuel assembly. SRP Section 4.2

states that fuel system coolability should be maintained and that damage

should not be so severe as to prevent control rod insertion when required

during these low probability accidents. This Appendix describes the

review that should performed of the fuel assembly structural response to

seismic and LOCA loads. Background material for this Appendix is given

in Refs. 1-3.

This proposed revision of the Standard Review Plan and the support value/impact

statement have not received a complete staff review and approval and do not

represent an official NRC staff position. Public comments are being solicited

on both the revision and the value/impact statement (including any implementation

schedules) prior to a review by the Regulatory Requirements Review Committee

and their recommendation as to whether this revision should be approved.

Comments should be sent to the Secretary of the Commission, U.S. Nuclear

Regulatory Commission, Washington, D.C. 20555, Attention: Docketing and

Service Branch. All comments received by will be considered by the Regulatory Requirements Review Committee. A summary of the

meeting of the Committee at which this revision is considered, the Committee

recommendations and all of the associated documents and comments considered by

the Committee will be made publicly available prior to a decision by the

Director, Office of Nuclear Reactor Regulation, on whether to implement this

revision.

USNRC STANDARD REVIEW PLAN

Standard review plans are prepared for the guidance of the Office of Nuclear Reactor Regulation staff responsible for the review of

applications to construct and operate nuclear power plants. These documents ure made available to the public as part of the

Commission's policy to inform the nuclear industry and the general public of regulatory procedures and policies. Standard review

plans are not substitutes for regulatory guides or the Commission's regulations and compliance with them is not required. The

standard review plan sections are keyed to the Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants.

Not all sections of the Standard Format have a corresponding review plan.

Published standard review plans will be revised periodically, as appropriate, to accommodate comments and to reflect new information and experience. E-3

Comments and suggestions for Improvement will be considered and should be sent to the U.S. Nuclear Regulatory Commission,

Office of Nuclear Reactor Regulation. Washington, D.C. 2M5.
B. ANALYSIS OF LOADS

1, Input

Input for the fuel assembly structural analysis comes from results

of the primary coolant system and reactor internals structural

analysis, which is reviewed by the Mechanical Engineering Branch.

Input for the fuel assembly response to a LOCA should include (a)

motions of the core plate, core shroud, fuel alignment plate, or

other relevant structures; these motions should correspond to the

break that produced the peak fuel assembly loadings in the primary

coolant system and reactor internals analysis, and (b) transient

pressure differences that apply loads directly to the fuel assembly.

If the earthquake loads are large enough to produce a non-linear

fuel assembly response, input for the seismic analysis should use

structure motions corresponding to the reactor primary coolant

system analysis for the SSE; if a linear response is produced, a

spectral analysis may be used (see Regulatory Guide 1.60).

2. Methods

Analytical methods used in performing structural response analyses

must be reviewed. Justification should be supplied to show that the

numerical solution techniques are appropriate.

Linear and non-linear structural representations (i.e., the modeling)

must also be reviewed. Experimental verification of the analytical

representation of the fuel assembly components should be provided

when pratical.

A sample problem of a simplified nature must be worked by the applicant and compared by the reviewer with either hand calculations or

results generated by the reviewer with an independent code (2).

Although the sample problem should use a structural representation

that is as close as possible to the design in question (and, therefore, would vary from one vendor to another), simplifying assumptions

may be made (e.g., one might use a 3-assembly core region with

continuous sinusoidal input).

The sample problem should be designed to exercise various features

of the code and reveal their behavior. The sample problem comparison

is not, however, designed to show that one code is more conservative

than another, but rather to alert the reviewer to major discrepancies

so that an explanation can be sought.

3. Uncertainty Allowances

The fuel assembly structural models and analytical methods are

probably conservative and input parameters are also conservative.

However, to ensure that the fuel assembly analysis does not

introduce any non-conservatisms, two precautions should be taken:

(a) If it is not explicitly evaluated, impact loads from the PWR

E-4
LOCA analysis should be increased (by about 30%) to account for a

pressure pulse, which is associated with steam flashing that affects

only the PWR fuel assembly analysis. (b) Conservative margin should

be added if any part of the analysis (PWR or BWR) exhibits pronounced

sensitivity to input variations.

Variations in resultant loads should be determined for +10% variations

in input amplitude and frequency; variations in amplitude and frequency

should be made separately, not simultaneously. A factor should be

developed for resultant load magnitude variations of more than 15%.

For example, if +10% variations in input magnitude or frequency

produce a maximum resultant increase of 35%, the sensitivity factor

would be 1.2. Since resonances and pronounced sensitivities may be

plant-dependent, the sensitivity analysis should be performed on a

plant-by-plant basis until the reviewer is confident that further

sensitivity analyses are unnecessary or it is otherwise demonstrated

that the analyses performed are bounding.

4. Audit

Independent audit calculations for a typical full-sized core must be

performed by the reviewer to verify that the overall structural

representation is adequate. An independent audit code (2) should be

used for this audit during the generic review of the analytical

methods.

5. Combination of Loads

General Design Criterion 2 requires an appropriate combination of

loads from natural phenomena and accident conditions. Loads on fuel

assembly components should be calculated for each input (i.e.,

seismic and LOCA) as described above in Paragraph 1, and the resulting

loads should be added by the square-root-of-sum-of-squares (SRSS)

method. These combined loads should be compared with the component

strengths described in Section C according to the acceptance criteria

in Section D.

C. DETERMINATION OF STRENGTH

1. Grids

All modes of loading (e.g., in-grid and through-grid loadings)

should be considered, and the most damaging mode should be represented in the vendor's laboratory grid strength tests. Test

procedures and results should be reviewed to assure that the appropriate failure mode is being predicted. The review should also

confirm that (a) the testing impact velocities correspond to expected

fuel assembly velocities, and (b) the crushing load Pcrit has been

suitably selected from the load-vs- deflection curves. Because of

the potential for different test rigs to introduce measurement

variations, an evaluation of the grid strength test equipment will

be included as part of the review of the test procedure.

E-5
The consequences of grid deformation are small. Gross deformation

of grids in many PWR assemblies would be needed to interfere with

control rod insertion during an SSE (i.e., buckling of a few isolated

grids could not displace guide tubes significantly from their proper

location), and grid deformation (without channel deflection) would

not affect control blade insertion in a BWR. In a LOCA, gross

deformation of the hot channel in either a PWR or a BWR would result

in only small increases in peak cladding temperature. Therefore,

average values are appropriate, and the allowable crushing load

Pcrit should be the 95% confidence level on the true mean as taken

from the distribution of measurements on unirradiated production

grids at (or corrected to) operating temperature. While Pcrit will

increase with irradiation, ductility will be reduced. The extra

margin in Pcrit for irradiated grids is thus assumed to offset the

unknown deformation behavior of irradiated grids beyond Pcrit"

2. Components Other than Grids

Strengths of fuel assembly components other than spacer grids may be

deduced from fundamental material properties or experimentation.

Supporting evidence for strength values should be supplied. Since

structural failure of these components (e.g., fracturing of guide

tubes or fragmentation of fuel rods) could be more serious than grid

deformation, allowable values should bound a large percentage (about

95%) of the distribution of component strengths. Therefore, ASME

Boiler and Pressure Vessel Code values and procedures may be used

where appropriate for determining yield and ultimate strengths.

Specification of allowable values may follow the ASME Code requirements and should include consideration of buckling and fatigue

effects.

D. ACCEPTANCE CRITERIA

1. Loss-of-Coolant Accident

Two principal criteria apply for the LOCA: (a) fuel rod fragmentation

must not occur as a direct result of the blowdown loads, and (b) the

10 CFR 50.46 temperature and oxidation limits must not be exceeded.

The first criterion is satisfied if the combined loads on the fuel

rods and components other than grids remain below the allowable

values defined above. The second criterion is satisfied by an ECCS

analysis. If combined loads on the grids remain below Pcrit' as

defined above, then no significant distortion of the fuel assembly

would occur and the usual ECCS analysis is sufficient. If combined

grid loads exceed Pcrit' then grid deformation must be assumed and

the ECCS analysis must include the effects of distorted fuel

assemblies. An assumption of maximum credible deformation (i.e.,

fully collapsed grids) may be made unless other assumptions are

justified.

E-6
Control rod insertability is a third criterion that must be satisfied.

Loads from the worst-case LOCA that requires control rod insertion

must be combined with the SSE loads, and control rod insertability

must be demonstrated for that combined load. For a PWR, if combined

loads on the grids remain below Pcrit as defined above, then significant deformation of the fuel assembly would not occur and control

rod insertion would not be interfered with by lateral displacement

of the guide tubes. If combined loads on the grids exceed Pcrit'

then additional analysis is needed to show that deformation is not

severe enough to prevent control rod insertion.

For a BWR, several conditions must be met to demonstrate control

blade insertability: (a) combined loads on the channel box must

remain below the allowable value defined above for components other

than grids; otherwise, additional analysis is needed to show that

deformation is not severe enough to prevent control blade insertion,

and (b) vertical liftoff forces must not unseat the lower tieplate

from the fuel support piece such that the resulting loss of lateral

fuel bundle positioning could interfere with control blade insertion.

2. Safe Shutdown Earthquake

Two criteria apply for the SSE: (a) fuel rod fragmentation must not

occur as a result of the seismic loads, and (b) control rod insertability must be assured. The first criterion is satisfied by the

criteria in Paragraph 1. The second criterion must be satisfied for

SSE loads alone if no analysis for combined loads is required by

Paragraph 1.

E. REFERENCES

1. R. L. Grubb, "Review of LWR Fuel System Mechanical Response with

Recommendations for Component Acceptance Criteria," Idaho National

Engineering Laboratory, NUREG/CR-IOI8, September 1979.

2. R. L. Grubb, "Pressurized Water Reactor Lateral Core Response Routine,

FAMREC (Fuel Assembly Mechanical Response Code)," Idaho National

Engineering Laboratory, NUREG/CR-1019, September 1979.

3. R. L. Grubb, "Technical Evaluation of PWR Fuel Spacer Grid Response

Load Sensitivity Studies," Idaho National Engineering Laboratory,

NUREG/CR-1020, September 1979.

E-7

NRC FORM 335 usNN1. REPORT NUMBER (Assignedby DDC)

BIBLIOGRAPHIC DATA SHEET NUREII-0609

4. TITLE AND SUBTITLE (Add Volume No., if appropriate) 2. (Leave blank)

Asymmetric Blowdown Loads on PWR Primary Systems

Subtitle: Resolution of Generic Task Action Plan A-2 3. RECIPIENT'S ACCESSION NO.

7. AUTHOR(S) S. B. Hosford, 1asK Manager, K. MdtLLu, R. r. 5. DATE REPORT COMPLETED

E. D. Throm, C. G. Tinkler MONTH YEAR

November 1980

9. PERFORMING ORGANIZATION NAME AND MAILING ADDRESS (Include Zip Code) DATE REPORT ISSUED

Division of Safety Technology ONur

Office of Nuclear Reactor Regulation ny

U.S. Nuclear Regulatory Commission 6. (Leave blank)

Washington, DC 20555 8. (Leave blank)

12. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (Include Zip Code)

10. PROJECT/TASK/WORK UNIT NO.

Same as 9, above.

11. CONTRACT NO.

13. TYPE OF REPORT PERIOD COVERED (Inclusive dates)

NUREGn

15. SUPPLEMENTARY NOTES 14. (Leave blank)

Nolne

16. ABSTRACT 9200 words or less)

NRC staff, after being informed of newly identffied asymmetric loadings resulting from

postulated ruptures of primary piping, initiated a generic investigation, Task Action

Plan A-2, limited to pressurized-water-reactor (PWR) plants because of their hirher

primary system pressures. The intent of the investigation was to develop acceptable

criteria and auidelitnes for evaluating plant analyses.

The staff concludes that an acceptable basis is provided in this report for performing

and reviewing plant analyses. Criteria were developed for evaluatina loadinc transien'

structural components, and the fuel assembly.

The staff approved computer proarams and modeling techniques submitted by each PWR veni

for development of the subcooled blowdown and cavity-pressure loadina transients Audit

models were develooed to evaluate the structural computer programs and modeling

techniques. Methods have been approved for the structural-analysis method submitted

by Westinghouse for the Indian Point Unit 3 plant. Criteria and guidelines are providi

to perform a detailed evaluation of the fuel assembly. Acceptance criteria are also

nrovided so deformed fuel-assembly spacer grids may be evaluated.

17. KEY WORDS AND DOCUMENT ANALYSIS 17a. DESCRIPTORS

17b. IDENTIFIERS/OPEN-ENDED TERMS

18. AVAILABILITY STATEMENT 19. SECURITY C.AS (This report 21. NO. OF PAGES

UnclassifiTed

Unlimited 20. SECURITY CLASS (This page) 22. PRICE lUnclassified $

or

N RC FORM 335 (7-77)