CANAL user's manual

A. Faya,* L. Wolft and N. Todreast J.

Energy Laboratory Report No. MIT-EL 79-028 November 1979

*Ph.D. Candidate, Department of Nuclear Engineering tAssociate Professor, Department of Nuclear Engineering t_{Professor, Department of Nuclear Engineering}

REPORTS IN REACTOP THERMAL HYDRAULICS RELATED TO THE MIT ENERGY LABORATORY ELECTRIC POWER PROGRAM

A. Topical Reports (For availability check Energy Laboratory

Headquarters, Room E19-439, MIT, Cambridge,

Massachusetts, 02139)

A.1 General Applications A.2 PWR Applications A.3 BWR Applications A.4 LMFBR Applications

A.1 M. assoud, "A Condensed Review of Nuclear Reactor Thermal-Hydraulic Computer Codes for Two-Phase Flow Analysis", MIT Energy Laboratory Report MIT-EL-79-018, April 1979.

J.E. Kelly and M.S. Kazimi, "Development and Testing of the Three Dimensional, Two-Fluid Code THERMIT for LWR Core and Subchannel Applications", MIT Energy Laboratory Report, MIT-EL-79-046, December 1979.

A.2 P. Moreno, C. Chiu, R. Bowring, E. Khan, J. Liu, N. Todreas, "Methods for Steady-State Thermal/Hydraulic Analysis

of PWR Cores", Report MIT-EL-76-006, Rev. 1, July 1977 (Orig. 3/77).

J.E. Kelly, J. Loomis, L. Wolf, "LWR Core Thermal-Hydraulic Analysis--Assessment and Comparison of the Range of

Applicability of the Codes COBRA IIIC/M1IT and COBRA IV-l", Report MIT-EL-78-026, September 1978.

J. Liu, N. Todreas, "Transient Thermal Analysis of PWR's by a Single Pass Procedure Using a Simplified Nodal Layout", Report MIT-EL-77-008, Final, February 1979, (Draft, June 1977). J. Liu, N. Todreas, "The Comparison of Available Data on

PWR Assembly Thermal Behavior with Analytic Predictions",

Report MIT-EL-77-009, Final, February 1979, (Draft, June 1977).

A.3 L. Guillebaud, A. Levin, W. Boyd, A. Faya, L. Wolf, "OSUB

-A Subchannel Code for Steady-State and Transient Thermal-Hydraulic Analysis of Boiling Water Reactor Fuel Bundles", Vol. II, User's Manual, MIT-EL-78-024, July 1977.

L. Wolf, A. Faya, A. Levin, W. Boyd, L. Guillebaud, "WOSUB -A Subchannel Code for Steady-State and Transient

Thermal-Hydraulic Analysis of Boiling Water Reactor Fuel Pin Bundles", Vol. III, Assessment and Comparison, MIT-EL-78-025, October 1977, L. Wolf, A. Faya, A. Levin, L. Guillebaud, "WOSUB - A Subchannel Code for Steady-State Reactor Fuel Pin Bundles", Vol. I, Model

Description, MIT-EL-78-023, September 1978.

A. Faya, L. Wolf and N. Todreas, "Development of a Method for BWR Subchannel Analysis", MIT-EL 79-027, November 1979. A. Faya, L. Wolf and N. Todreas, "CANAL User's Manual",

MIT-EL 79-028, November 1979.

A.4 W.D. Hinkle, "Water Tests for Determining Post-Voiding Behavior in the LMFBR", MIT Energy Laboratory Report MIT-EL-76-005, June 1976.

W.D. Hinkle, ed., "LMFBR Safety & Sodium Boiling - A State of the Art Report", Draft DOE Report, June 1978.

M.R. Granziera, P. Griffith-, W.D. Hinkle, M.S. Kazimi, A. Levin, M. Manahan, A. Schor, N. Todreas, G. Wilson, "Development

of Computer Code for Multi-dimensional Analysis of Sodium Voiding in the LMFBR", Preliminary Draft Report, July 1979.

---B. Papers

B.1 General Applications B.2 PWR Applications B.3 BWR Applications B.4 LMFBR Applications

B.1 J.E. Kelly and M.S. Kazimi, "Development of the Two-Fluid Multi-Dimensional Code THEPRIT for LWR Analysis", accepted for presentation 19th National Heat Transfer Conference, Orlando, Florida, August 1980.

3. 2 P. Moreno, J. Liu, E. Khan and N. Todreas, "Steady-State

Thermal Analysis of PWR's by a Simplified Method,"

American Nuclear Society Transactions, Vol. 26, 1977, p. 465. P. Moreno,J. Liu, E. Khan, N. Todreas, "Steady-State

Thermal Analysis of PWR's by a Single Pass Procedure Using a Simplified Nodal Layout," Nuclear Engineering

and Design, Vol. 47, 1978, pp. 35-48.

C. Chiu, P. Moreno, R. Bowring, N. Todreas, "Enthalpy Transfer Between PWR Fuel Assemblies in Analysis by the Lumped Subchannel Model," Nuclear Engineering and Design, Vol. 53, 1979, 165-186.

B. 3 L. Wolf and A. Faya, "A BWR Subchannel Code with Drift Flux and Vapor Diffusion Transport," American Nuclear Society Transactions, Vol. 28, 1978, p. 553.

B.4 W.D. Hinkle, (MIT), P.M. Tschamper (GE), M.H. Fontana, (ORNL), R.E. Henry, (ANL), and A. Padilla, (HEDL), for

U.S. Department of Energy, "LMFBR Safety & Sodium Boiling," paper presented at the ENS/ANS International Topical

Meeting on Nuclear Reactor Safety, October 16-19, 1978, Brussels, Belgium.

M.I. Autruffe, G.J. Wilson, B. Stewart and M.S. Kazimi, "A Proposed Momentum Exchange Coefficient for Two-Phase

Modeling of Sodium Boiling", Proc. Int. Meeting Fast Reactor Safety Technology, Vol. 4, 2512-2521, Seatle, Washington, August, 1979.

M.R. Granziera and M.S. Kazimi, "NATOF-2D: A Two Dimensional Two-Fluid Model for Sodium Flow Transient Analysis", Trans. ANS, 33, 515, November 1979.

ABSTRACT

This report gives a detailed description of the input data and contains a listing of the computer program CANAL.

Brief History

In the fall of 1976 Louis Guillebaud performed the first consistent check on the models and method of solution employed by the computer program WOSUB [1,2,3] which is an extension of the MATTEO code[4]. Alan Levin provided the additional subroutines for calculating the heat transfer coefficients and critical heat flux thus enabling WOSUB to present data beyond the scope of the MATTEO code.

In the spring of 1977 William Boyd concentrated his work on a parametric sensitivity study of the empirical parameters of the WOSUB code and their effects upon the overall results [3]. He was succeeded by Artur Faya who added to the code a fuel pin model based on the collocation method [1].

In the fall of 1977 Artur Faya started the development of the CANAL code which is the subject of this report. New

physical models were necessary because WOSUB results for some important experiments were not satisfactory [3]. Besides the physical models of WOSUB tend to overestimate the transport of vapor for bulk boiling conditions. This in turn leads to

numerical instabilities in some cases.

The similarities between WOSUB and CANAL reside only on the numerical scheme, heat transfer coefficient package and fuel pin model. CANAL and WOSUB differ in the following main points:

· mixing model

· vapor generation rate

· liquid and vapor are treated as compressible in CANAL and incompressible in WOSUB

· correlations for fluid physical properties . correlations for friction pressure losses

L. Wolf supervised the foregoing efforts. N. Todreas assisted in the supervision of the final stages of

Artur Faya's work.

References

1. L. Wolf et al. "WOSUB - A Subchannel Code for Steady-State and Transient Thermal Hydraulic Analysis of BWR Fuel Pin

Bundles. Volume I - Model Description", MIT-EL-78-023 (1978) 2. L. Wolf et al. "WOSUB - A Subchannel Code for Steady-State

and Transient Thermal Hydraulic Analysis of BWR Fuel Pin Bundles. Volume II - User's Manual", MIT-EL 78-024 (1978)

3. L. Wolf et al. "WOSUB - A Subchannel Code for Steady-State and Transient Thermal Hydraulic Analysis of BWR Pin Bundles. Volume III - Assessment and Comparison", MIT-EL 78-025 (1978)

4. G. Forti and J.M. Gonzalez-Santalo, "A Model for Subchannel Analysis of BWR Rod Bundles in Steady-State and Transient", Int. Conf. Reactor Heat Transfer, Karlsruhe, Germany (1973)

Contents

1. Introduction

2. Input Description

3. CANAL Flow Diagram

4. Description of Subroutines 5. Sample Problem 6. Listing of CANAL 7. References 1 1 8 10 12 18 90

1. Introduction

CANAL is a subchannel computer program for the steady-state and transient thermal hydraulic analysis of BWR fuel rod bundles. The physical models and numerical scheme used in the code are described in /1/.

The purpose of this manual is to introduce the user into the mechanism of running the code by providing information about the input data and options.

2. Input Description

The input data for CANAL is divided into sections by type. The first card of each section contains the section number

(four-digit integer) which defines the type of input to follow. This permits the user to treat many problems in the same run and only the sections which are changing from the preceding problem need to be supplied. A blank card must be the last card of the first and each succeeding case in a deck of input

data. Every case must begin with a title card with any

alphanumeric information in columns 5 to 72. A negative

integer in columns 1 to 4 of the title card indicates that the preceding was the last case.

Case Control Cards

1 ICASE, TITLE(I) (I=1,17) (I4,17A4)

ICASE Problem case number

TITLE Identification of the problem

Section 003 - NSUBS, NRODS and options

1 NSUBS, NRODS (214)

NSUBS Number of subchannels in the lattice.

NRODS Number of heating rods in the lattice.

2 IUSYS, ITRAN, ICHF, ICISE, ISHAP,

ICOUPL, IPRI, IPRL, NPTR, NSPAC (1014) IUSYS System of units indicator.

IUSYS=O, metric =1, British ITRAN Type of transient.

ITRAN=O, steady-state

=1, mass flow transient. =2, power transient

=3, pressure transient ICHF CHFR Correlation indicator

ICHF=O, No CHF calculation =1, Hench-Levy

=2, Barnett =3, CISE

ICISE Order number of a center sub-channel for CISE CHF calculation.

Card Format ISHAP Axial power shape indicator

ISHAP=O, flat

=1, tabulated (Section 009) ICOUPL If ICOUPL>0, thermal coupling

between fuel and coolant. IPRI If IPRI>O, print input data IPRL If IPRL>O, long print out of

results.

NPTR Printout of results every NPTR times steps.

NSPAC Number of axial locations

where grid spacers are present.

Section 005 - Subchannel Layout

1 MSUB(I) , KISUB(I,K) (K=1,4) (614)

MSUB Total number of type I sub-channels in the bundle.

KISUB Adjacent subchannel identification number for up to 4 subchannels

(in ascending order).

2 LIROD(I,K) (K=1,4)

LIROD Adjacent rod identification number for up to 4 rods

Format Section 007 - Subchannel Geometrical

and Hydraulic Parameters

(Read NSUBS sets of cards 1,2,and 3 sequentially)

A(I), DEQ(I) (2E10.0)

A Subchannel flow area (in or m )

DEQ Subchannel equivalent hydraulic diameter (in or m)

PFRAC (I,K) (K=l, 4)

PFRAC Fraction of the perimeter of rod LIROD(I,K) bounding subchannel I. GAP(I,K) (K=, 4)

GAP Gap width between subchannel I and the adjacent subchannel

specified by KISUB(I,K) (in or m)

ZTOT, ZTOT DROD ACISE (4E10.0) (4E10.0)

DROD, ACISE (3E1(

Bundle height (in or m)

Diameter of heating rods (in or m)

Flow area of subchannel ICISE (in2 or m ) ).0)

Section 009 - Fuel Rod

MROD(I) (I=1,NRODS)

MROD Total number of type I heating rods in the bundle.

RPEAK(I) (I=1,NRODS)

RPEAK Radial peaking factor for rod I (1.) (1814) (7E10.0) Card 1 2 3 4 1 2

(card 3 is needed if ISHAP>0) SHAPEF(J) (J=1,JMAX)

SHAPEF Axial shaping factor (1.). (.card 4 and 5 are needed if ICOUPL>0) RFS, CONDF, CPF, RHOF, HGAP

RFS Fuel radius

CONDF Fuel thermal conductivity

(BTU/hr-ft F or W/m C).

CPF Fuel specific heat

(BTU/lb F or J/Kg C).

RHOF Fuel density (lb/ft3 or Kg/m3)

HGAP Effective Gap Conductance

(BTU/hr-ft2 F or W/m 2 - C)

RC, CONDC, CPC, RHOC RC Clad thickness.

CONDC Clad thermal conductivity. CPC Clad specific heat.

RHOC Clad density.

Section 011 - Calculational Parameters JMAX, ITMAX

JMAX Number of axial nodes.

ITMAX Maximum number of iterations

to find the mass crossflows (20). DT, TMAX, EPSP

DT Time increment.

TMAX Total time of calculation. EPSP Convergence criteria to

equalize pressure drops. Card 3 4 5 2 Format (7ElO O) (5E10.0) (4E10.0) (214) (3E10..0).

Format Section 013 - Fluidd Inlet Conditions

P, POWER, FINLET, HINLET

P System pressure (psia or N/m2). POWER Power added to the bundle (KW). FINLET Inlet mass flow rate

(Mlb/hr or Kg/sec).

HINLET Inlet subcooling (BTU/lb or J/Kg). Section 015 - Transient Specifications NTAB

NTAB

(4E10.0)

(I4) Number of tabulated points.

(Card 2 is to be repeated NTAB times) TTAB(N), VTAB(N)

TTAB

VTAB

Time; first value must be always 0.

Value of the quantity specified by ITRAN (flow, pressure or

power) at time TTAB.

Section 017 - Empirical Parameters

YK, TETAM, CKNOT

YK Parameter K in the mixing model. TETAM Parameter 6M in the mixing

model.

CKNOT Drift-flux concentration parameter, C. Card 1 1 2 (2E10.0) 1 (3E10.0)

AFR, AFR BFR CFR

BFR, CFR

Parameters a, b and c, respec-tively, in the friction factor correlation (f = aRe-b+C)

Section 019 - Spacer Coefficients SPACL (N) (N=1,NSPAC)

SPACL Distance from spacer axial location N to the bottom of the core.

(card 2 is to be repeated NSPAC times) SPACER(I,N)

SPACER Grid spacer coeffient for sub-channel I at axial location N.

2 (3E10.0)

1

2

Read Input Data

Preliminary Calculations

1

Evaluate Fluid Physical Properties

Set Inlet Conditions and Guess Crossflows for First Axial Node

L

Criteria | no Iterate to Obtain

P

i-

avl ? New Crossflows

es

Compute Liquid and Vapor Densities, set Inlet Condi-MAX .. tions for Next Axial Node

and Guess Crossflows

Compute Heat Transfer Fuel Coupled? Y -Coefficients, Compute

Fuel Temperature Distribution

no IF,

Solve Conservation Equations

'-- r.

---I

I--`---MAIN Calls subroutines to read input data and performs preliminary calculations

READ1 Reads the input data

RITE1 Prints out input data

RITE2 Short printout of the results

RITE3 Prints out local quantities

RITE4 Prints out CHF information

TSTEP Controls time step

XMASS Computes mass exchange between subchannels

XENGY Computes energy exchange between subchannels

XMOMT Computes momentum exchange between subchannels

CONSV Solves conservation equations in each of the

subchannels

VAPSC Evaluates the vapor source term

PARAM Computes parameters that depend on physical

properties

CIRCUL Finds the circulation paths

MULTI Multiplies a square matrix by a vector

CONNCT Sets the subchannels connection matrix

CHEN Computes the heat transfer coefficient

CHF Computes the critical heat flux

TURB Computes the turbulent radial velocity

FREG Computes the two-phase mixing multiplier

FUEL Computes fuel temperature distribution

NODCO Evaluates nodal coefficients employed in the

fuel pin model

FVS12 Evaluates Hermite cubic polynomials and

CONLIQ Computes the thermal conductivity of the liquid

CPLIQ Computes the specific heat of the liquid VISLIQ Computes the viscosity of the liquid

SURTEN Computes the surface tension

TTSAT Computes the saturation temperature of water as a function of pressure

ROL Computes the density of the liquid ROV Computes the density of the vapor

HLIQS Computes the enthalpy of saturated water as function of pressure

HEVAP Computes the latent heat of evaporation th

POLY Evaluates a polynomial of n order

FFACT Computes the single-phase friction factor FIL02 Computes the two-phase friction multiplier FIL02L Computes the two-phase local friction

5. Sample Problem

Fig. 1 shows the GE 9-rod bundle /2/. The sample problem is the GE run 2D1 where the power distribution is uniform both axially and radially. One can see that for this particular case there is a 45° symmetry and the analysis can be performed using a configuration of the type shown in Fig. 2.

Other data of importance are

Pressure 1000 psia

Total bundle power 1064. KW

Inlet mass flow rate 0.01098 Mlb/hr

Inlet enthalpy -259.2 BTU/lb

Total height of the bundle 6 ft.

Subchannel Flow Area Hydraulic Diameter

(in2) (in)

Corner 0.0391 0.28028

Side 0.1824 0.4467

Center 0.1447 0.6465

-'---I--12a

- .. .;.L . A.

wsd-t? B

Fig. 2 Snimple subchannel layout to describe

input data

II_

_

_ _

·_I _II__

Input for the Sample Problem GE 201 0 0 0 0 3 .28028 .4467 .168

.e

L65

.57 4 4 l 1. 0. 1064000. iO. .25 L 1 0 0 3 0 2 1.

2

2 2 0003 3 0005 8 I 1 8 1 0007 0. 0391 .125 .1 35 .1 824 .5 .135 .I447 .5 .168 72 0009 1.' .001 011 30 0. 0013 1000. 0017 1 4 .32 0000 -1 .01098 259.2 1.3 -3

Output for the Sample Problem

4

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