J.E.N, 2

Guillermo VelardeFrancisco AguilarCarolina AhnertJosé M. AragonésManuel GómezJesús GuerraÁngel PalmeroJuan Serrano

JUNTA DE ENERGÍA NUCLEAR

Madrid, May 31", 1972

Toda correspondencia en relación con este traba-jo debe dirigirse al Servicio de Documentación Bibliotecay Publicaciones, Junta de Energía Nuclear, Ciudad Univer_sitaría. Madrid-3, ESPAÑA.

Las solicitudes de ejemplares deben dirigirse aeste mismo Servicio.

Las publicaciones señaladas con la signatura /Ipertenecen a la categoría a. "Memorias Científicas Or̂Lginales"; las señaladas con la signatura ,/N pertenecena la categoría b, "Publicaciones Provisionales o NotasIniciales"; y los señalados con las signaturas /C, /CM,/B, /Conf pertenecen a la categoría c, "Estudios Re-capitulativos" de acuerdo con la recomendación GC(VII)/RES/150 del OIEA, y la UNESCO/NS/1 77.

Se autoriza la reproducción de los resúmenes analiticos que aparecen en esta publicación.

Este trabajo se ha recibido para su impresión enJunio de 1. 972.

Depósito legal n° M-20853- 1 972.

1.- INTRODUCTION

2.- REQUESTED INFORMATION2 • 1 • ~ Core DesJ.gr. Data

2.1.1.- Nuclear Design Data

2.1.2.- Thermal and Hydraulic Design Data

2.1.3.- List of Figures relative to Core Description

2.2.- Addixional Data

2.2.1.- Operaiional Characreristics and Dynamics Data

2.2.2.- Reactor Coolant System

2.2.3.- In-Core Insrrumentation

2.2.4.- Listof Figures relative to Additional Systems

2.3,- Design Results

2.3.1.- Nuclear Design Results

2.3.2.- Thermo-Hydraulic Results

2.3.3.- Transient and Accident Analysis

2 . 3 . ̂ . - List of Figures relative to Design Resuxxs

2.4.- General Reauests

-1-

1.- INTRODUCCIÓN

1.1.-

Uno de los principales objetivos, que en el campo de la

Tecnología Nuclear, pueden realizar las naciones en vias de de-

sarrollo, es la gestión y el diseño de los elementos combustible

de los reactores nucleares que paulatinamente se vayan importan-

do. Para llevar a cabo este objetivo, se requieren tres condicio

nes :

i) Disponer de un equipo de ingenieros, con suficiente

experiencia en el diseño de elementos combustibles.

ii) Disponer de un computador con una memoria de unas

140000 palabras de capacidad (CDC-6600, Univac 1108)9 y de un

grupo de códigos para la gestión y diseño de los elementos com

bustibles, con la posibilidad de actualizarles periódicamente.

iii) Disponer de determinados parámetros de proyecto.

1.2.-

Teniendo en cuenta el nivel de conocimientos que en el

campo de la energía nuclear, se adquieren actualmente en la Uni

versidad, la formación del personal postgraduado, podría reali-

zarse en dos ottres años.

1.3.-

El grupo mínimo de códigos necesarios para la gestión y

diseño de elementos combustibles puede estar compuesto por:

LeopardLáserAssault

, Fogi) nuclear { Nutrix

NuflowDTF-IVPDQ-7

\ T i m o c

/ Bolero. . .. . , . „ . . , . J Caramban ) t ermoniaraulic o < _

i Forcir

I Plankin

iii) termomecánico j Cygro

-2-

iv) economía < Fuel Cost Ii y IV

La descripción de estos códigos se hará en un informe pos_

terior.

La JEN dispone actualmente de todos ellos, excepto de los

códigos clave: PDQ-7 y Cygro, cuya exportación, fuera de los E.U.A

está prohibida.

Con objeto de reducir en lo posible la intervención perso-

nal en el funcionamiento de estos códigos, se tiende actualmente

al empleo de códigos integrados, del tipo Citation, los cuales

combinan códigos análogos a los anteriores. Ello exije computado_

res de una capacidad de memoria superior a 140000 palabras.

Respecto a los parámetros de proyecto, en este informe se

relacionan aquellos que creemos son necesarios para la gestión y

diseño de los elementos combustibles, los cuales han de ser sumi_

nistrados por el Fabricante del Reactor, a la firma del contrato.

Nuestro propósito es colaborar con las Empresas Eléctricas

españolas, para que estos parámetros sean exigidos al Fabricante

de la serie de reactores aue actualmente se van a contratar, los

cuales, desgraciadamente, no fueron exigidos en el contrato de

los anteriores reactores.

2.1.- REACTOR CORE DESIGN DATA

2.1.1.- NUCLEAR DESIGN DATA

Active Core

Equivalent diameter

Active fuel height

Length-to-diameter ratio

Total cross-section área

Reflectors and Core Structure

Dimensions and material compcsition for

Core baffle

Core barrel

Thermal shield

Top reflector

Bottoip reflector

Side reflector

H_0/U volume ratio (average in core)

Fuel Assemblies

Fuel rods,

Number of fuel rods per assembly

Rod array

Rod pitch

Guide thimbles

Number per assembly

Material composition

Dimensions (upper and lower part)

Instrumentation guide thimble

Number per assembly

Material composition

Spacer grids

Number of grids per assembly, normal

with mixing vanes

Data specifications are for cold conditions

Tolerances are included (vrhen possible)

Material composition

Weight per grid

Dimensions

End fittings

Material composixion

Total weight

Dimensions

upper end fitting

lower end fitting

Number of fuel assemblies in core

Fuel assembly overall dimensions

Fuel assembly pitch

Fuel loading per assembly (as U0?)

Zircaloy weight

Total weigth

Fuel Rods

Total number

Ciad material

Ciad thickness

Ciad outside diameter

Gap filler gas, composition

pres sure

allowable leak rate

Fuel loading per rod (for each región)

Fu_el__Pellets

Mater ial

Density, inner región

middle región

outer región

Oxygen/Uranium ratio

Impurities and equivalent boron content

Moisture content

Pellet fuel loading per cm. of height

U-235 initial enrichment, inner región

middle región

outer region

Initial load (g/cm) and composition of burr.ahle poisor a:

if any

— 5 —

Pellet diaraeter (for each región)

Pellet height

Rod Cluster Control Assemblies

Assembly v.'sight (dry)

Absorber mal erial composition

Absorber diameter

Absorber active length

Ciad material composition

Ciad thickness

Ciad outside diameter

Number of control rods per cluster•L n !_-,- •+.-, ^ -, , rfull lengthNumber of assemblies with control rod { , . , ° .,

partial length

Burnable Poison Rods,

Number (total)

Material composition

Outside diameter

Inner tube , 0 . D.

Ciad material

Ciad thickness

Inner tube material

Poison loading, gm per CID of rod

2.1.2.- THERMAL AND HYDRAULIC DESIGN PARAMETERS

General Data

Total core heat output

Heat generated in fuel

Máximum thermal overpower

Nominal system pressure

Coolant Flow

Total coolant flow rate

Bypass coolant flow rate

Average mass velocity

Primary coolant heat removal

Coolant flow for heat removal only

-6-

Nominal assembly coolant flow

Máximum rated assembly coolant flow

Average coolant velocity along fuel rods

Mínimum coolant velocity along fuel rods

Core inlet pressure (mininum)

Pressure drop plenurn to plenum

Pressure drop across the inlet nozzle

Pressure drop across the exit nozzle

Pressure drops across the grids

Coolant flow área per assembly

Channel equivalent diameter

Unheated channel length at entrance

Unheated channel length at exit

Core inlet coolant flow distribución

Coolant Temperature or Enthal'py

Nominal inlet temperature at rated power

Máximum inlet temperature at rated oower

Average rise in vessel at rated power

Average rise in core at rated power

Average temperature in core at rated power

Average temperature in vessel at rated power

Average film coefficient at rate power

Average film temperature difference at rated power

Heat Transfer

Average powtx' density

Average specific power

Average lineal heat rate

Máximum lineal heat rate

Rated power

Design overpower

Active heat transfer área

Máximum |Kd0 (hottest rod)

Average heat flux at rated power

Hot channel máximum heat flux

Rated power

Design overpower

-7-

Hot Channel Factors

Engineering hot channel factors

a) Heat flux hot channel factor (F )q

This factor should contain subfactors to account for

Variations in pellet diameter

Variations in pellet density

Variations in pellet enrichment

Eccentricity of the pellet

Variations in ciad diameter£

b) Enthalpy rise hot channel factor (F.„)in

This factor should contain subfactors to account for

All the effects in part a) above

Variations in fuel r-od pitch

Fuel rod bowing

Fuel assembly bowing

Flow redistribution due to high resistance in hot channels

Flow mixing inside a fuel assembly

Maldistribution on inlet flow

Overpower factors

Heat balance error

Instrument error

Instrument uncertainty for power and temperature

Transient overshoot

Instrument dead band

Total designDesign Mínimum Margin to Incipient Fuel-Clad Damage

Minimum allowable DNBR

Rated power

Design overpower

Máximum fuel centerline temperature

Rated power

Design overpower

Average fuel temperature

Rated power

Average ciad temperature

Rated power

Máximum ciad surface temperature

Rated power

Design overpower

Mixing Parameters

Turbulent mixing parameter without mixing vanes

Turbulent mixing parameter with mixing vanes

Friction factor for diversión cross flow

Diversión momentum factor

Turbulent momentum factor

2.1.3.- LIST OF FIGURES RELATIVE TO CORE DESCRIPTION

1. Reactor vessel and internáis

2. Core cross section

3. Core barrel assembly

4. Fuel assembly outline

5. Grid assembly

6. Guide tube assembly

7. Rod cluster control assembly outline

8. Burnable poison rod design

9. Fuel loading arrangement

10. ?od cluster control assembly pattern

11. Burnable poison loading pattern

12. Burnable poison rod arrangement within an assembly

13. Distribution (assembly-wise and within assembly) of

poison added to the fuel during manufacturing (if any)

-9-

2.2.- ADITIONAL DATA

2.2.1.- OPERATIONAL CHARACTERISTICS AND DYNAMICS DATA

Control Rods

Total number of steps (axial positions)

Heigtht of each step

Máximum withdrawal speed

Normal withdrawal and insertion speed

Terminal sc.^j speed

Weight of control rod and drive line

Dynarnic Data

Effective prompt neutrón lifetime, and

Effective delayed neutrón fractions for each group

^ rBOC, HZP, ARO, critical boron concentrationa L 1EOC, HFP, ARO, no boron

2.2.2.- REACTOR COOLANT SYSTEM

Design parameters for the steam generator

Number of steam generators

Design pressure, reactor coolant/steam

Design temperature, reactor coolant/steam

Primary side:

Heat transfer rate (per unit)

Coolant inlet tenperature

Coolant outlet temperature

Flow rate

Pressure loss

Heat transfer área

Primary side water volume

Secondary side:

Steam pressure at full power

Feedwater temperature

Steam flow rate (total)

Shell O.D., upper/lower

Shell thickness, upper/lower

Tube material

-10-

Number of U-tubes

U-xube outside diameter

Tube wall thickness

Average tube length

Secondary side water volume

Secondary side steam volume

Reactor Coolant Piping Design Parameters

Design/operating pressure

Design temperature

Hot leg volume

Cold leg volume

Reactor inlet piping, I.D.

Reactor inlet piping, nominal thickness

Reactor outlet piping, I.D.

Reactor outlet piping, nominal thickness

Reactor V-'ssel Design Parameters

Design/operating pressure

Design temperature

Reactor coolant inlet temperature

Reactor coolant outlet temperature

Pressure losses through vessel including nozzles

Reactor outlet plenum volume

Reactor inlet plenum volume

Core bypass volume

Reactor Coolant Pump Design Parameters

Number of pumps

Design pressure/operating pressure

Design temperature

Developed head

Capacity

Characteristic curves

Power (nameplate)

Coolant Chemistry

Recommended valúes (or typical valúes)for:

-11-

PH

Conductivity

Ppm H , 0 , Cl~, total solids, etc

2.2.3.- IN-CORE INSTRUMENTATION

Total number of thermocouples in the core

Total number of flux thimbles (if fixed)

Total number of neutrón sources

2.2.4.- LIST OF ADITIONAL FIGURES

1. Location of thermocouples in the core

2. Location of selecred essemblies for nuclear instrumenta-

t ion (if fixed)

3. Locaiion of neutrón sources in the coreu. Dimensioned drawing of reactor coolant pumps

5. ídem for steam generators

6. ídem for pressurizer

7. Distribution of instrumentation for:

a) Loop temperatures

b) Pressurizer pressure control

c) Reactor ex-core flux detectors

8. Dimensioned drawing of control rod drives

9. Dimensioned drawings for fuel handling equipment :

a) Tuel grapple

b) Fuel transport machine

-12-

2.3.- DESIGN RESULTS

2.3.1.- NUCLEAR DESIGN RESULTS *

Excess reactivity distribution (BOC)

CZP, clean

HZP, clean

HFP5 clean

HFP, Xe and Sm equilibrium

Shutdown 1 orón concentrations

Refueling shutdown (k = 0.90); ARI

Clean, CZP

Shutdown (k = 0.99); ARI

Clean, CZP

Clean, HZP

Shutdown (k= 0.99); ARO

Clean, CZP

Clean, HZP

Shuxdown (k= 0.99); All but one control rod inserted

Clean, CZP

Clean, HZP

Critical Boron Concentrations

BOC, clean, CZP, ARO

BOC, clean, HZP:

ARO

Parth-length group inserted

Each full-length group inserted

All control groups inserted

All shutdown groups inserted

All but one rod inserted

ARI

ARI means A_ll Control R_ods _I_nsertedARO means _A11 Con t ro l ílods OutCZP means ¿_id Z_ero PowerClean means wi thou t f i s s i o n p r o d u c í s (Xe, SinHZP means H_ot Z_ero P_owerHFP means H_ot F u l l FowerBOC means B_eginiling £f f i r s t CycleEOC means E_nd 0_f f i r s t Cycle ~

-13-

BOC, HFP, ARO

Clean

with equilibrium Xenón

with equilibrium Xenón and Samarium

Moderator Temperature Coefficient (core and each región)

At BOC , HZP, ARO, clean, critical boron concentration

At EOC, HFP, ARO, equilibrium Xenón and Samarium, no boron

Moderator Pressure Coefficient (core and each región)

At HZP, ARO, critical boron, clean

at BOC

at EOC

Doppler Coefficient (core and each región)

At HFP, ARO, equilibrium Xenón and Samarium, critical boron

at BOC

at EOC

Reactivity requirements for control rods

(% AK/K, at BOC an EOC)

Control

Power defect (combined Doppler T , and void effects)

Operational maneuvering band and control rod bite

Total control

Control rods worth

Integral worth of each control rod group

at BOC, HZP, clean, critical boron

ataBOC, HFP, equilibrium Xenón and Samarium- critical boron

at EOC, HFP, equilibrium Xenón and Samarium5 no boron

Shutdown margin with the highest worth rod out of the core in

the HZP, BOC condition with the critical boron concentration

corresponding to full power

Máximum worth of an ajected rod and resulting radial peaking

factor

-14-

Maximum peaking factors and negative reactivity resulting from

a dropped rod at full power

Heat generation rate inside the rods

Burnable Poison Rod Worth

BOC worth (AK/K)

hot

cold

Heat generation rate inside the rods

Isotopic Invenxory

Summary at EOC

Cycle lifetime at rated pow r

Core average burnup at BOC

Core average burnup at EOC

Reloading pattern

Number of fuel assemblies discharged

Average burnup in discharged assemblies

Energy generated in discharged assemblies2 3 5

Total U in discharged assemblies

Total U in discharged assemblies2 3 5

Average U enrichment in discharged assemblies2 3 9 241

Total Pu + Pu in discharged assemblies

2.3.2.- THERMO-HYDRAULIC RESULTS

Design Mínimum Margin to Incipient Fuel-Clad Damage

Calculated minimum DNBI.

Rated power

Design overpower

Steady reactor conditions to give a minimum DNBR = 1.0

Power

Inlet temperature or enthalpy

Steady reactor power to cause fuel centerline melting in hottest

rod

Steady reactor power to cause ciad damage due to excessive fuel

temperature

-15-

Coolant Temperature or Enthalpy

Reduction in assumed hot channel enthalpy rise due to interchannel

mixing

Average active coolant outlet temperature or enthalpy at rated po-

wer

Hot channel outlet temperature or enthalpy

Rated power

Design overpower

Hot channel outlet void fraction

Rated power

Design overpower

2-3.3.- TRANSIENT AND ACCIDFNT ANALYSIS

rT n , , , . ,, ,rfrom subcritical conditionUncontrolled rod withdraw >il{ .at powerPartial loss of forced reactor coolant flow

Turbine trip

Loss of normal feed water

Excessive load increase

Accidental despressurization of the reactor coolant system

Rupture of a main steam pipe

Inadvertent loading of a fuel assembly into an improper position

Fuel handling accident

2.3.4.- LIST OF FIGURES RELATIVE TO DESIGN RESULTS

1. Required shutdown margin v.s. boron concentration,

2. Nuclear hot channel factors for enthalpy rise and for heat

flux v.s. rod insertion for the different control rod groups.

3. Máximum and minimum control group insertions v.s. power level

(for all loop operation).

4-. ídem, (for all minus one loop operation).

5. Differential worth of each control rod group and axial peaking

factors v.s. insertion, at BOC, clean, HZP, critical boron.

6. ídem, at BOC, HFP, equilibrium Xenón and Samarium, critical

boron.

7. ídem, at EOC, HFP, equilibrium Xenón and Samarium, no boron.

8. Critical boron v.s. burnup at ARO, HZP.

9. ídem, at ARO, HFP.

-16-

10. Differential boron worth v.s. boron concentration at HFP,

ARO, BOC, and several burnups (average for core and each

región) .

11. Power distribution and peaking factor at BOC, HZP, clean,

ARO, critical boron.

12. ídem, at BOC, HFP, clean, ARO, critical boron.

13. ídem, at HFP, ARO, equilibrium Xenón and Samarium, critical

boron (BOC, different burnups and EOC).

14. ídem, at BOC, HFP, part length rods in , critical boron.

15. ídem, at BOC, HFP, critical boron, one control rod group

in (for each control rod group).

16. Doppler coefficienr v.s. effective fuel temperature (BOC}.

17. Effective fuel temperature v.s. rod relative power (BOC).

18. Effective fuel temperature at HFP v.s. rod burnup.

19. Power coefficient v.s. power level at HFP, ARO, BOC and

EOC, critical boron.

20. Moderator temperature reactivity coefficient v.s. moderator

temperature at nominal pressure, (HFP), ARO, BOC several bo-

ron concentrations (core and each región).

21. Production and consumption of higher isotopes v.s. burnup

(for each región).

22. Assembly wise burnup distribution at HFP, ARO, equilibrium

Xenón and Samarium and critical boron (for different burnups)

23. Axial peaking factor v.s. time, for a typical Xenón transient

(unstable and stabilized with part length rod motion).

24. W-3 correlation probability distribution curve.

25. Comparison of W-3 prediction and uniform flux data.

26. Comparison of W-3 correlation with rod bundle DNB da~a (simple

grid without mixing vane).

27. Comparison of W-3 correlation with rod bundle DNB data (simple

grid with mixing vane).

28. Thermal conductivity of uraniuir dioxide.

29. Cladding internal pressure v.s. time.

30. Temperature rise in the channels of a rod bur.cle v.s. channel

power density.

31. Fuel cladding and U09 temperature limits v.s. time or fuel

bunc le exposure .

32. Thermal conductivity of cladding.

33. Gap heat transfer coefficient v.s. burnup.

-17-

3^. Fuel rod heat flux limits v.s. time or fuel bundle exposure

35. Core inlet tetnperature v.s, power level program.

-18-

2.4.- GENERAL REQUESTS

a) Official documents: FSAR, Tech Specs, • .

b) Codes for reactor surveillance and processing of in-core i n s -

trumentat ion

c) Programming (software) of process computer (if any)

d) Main design reports:

Core analyses cov cycle 1

Basic lines of fuel management for following cycles

e) Other studies

Historie data on the fuel performance

Behaviour of operating PWR's designed by the vendor

f) Cooperation for obtaining in-house fuel management capabiliry

(computer codes, general method, up-dated valúes of empirical

parameters, etc.)

g) Last versión of critical heat flux correlation

J.E.N. 245 J.E.N. 245

Junta de Energía Nuclear, División de Física loórica, Madrid

"Información que debe aportar el suministradorde la caldera nuclear (NSSS) para poder efectuar lagestión del combustible".VEIARCC, 6 . ; AGUILAR, F.; AIINERT, C ; ARAGONÉS, J.H.; GÓMEZ, H.; GUERRA, J .

PALMERO, A.; SERRANO, J . (1972) 18 pp.

Se relaciona un conjunto de parámetros nucleares, terinohidráulicos y mecá-

nicos, necesarios para la gestión y diseño de los elementos combustibles en los

PWR, los cuales deben ser suministrados por el Fabricante del Reactor a la

Empresa Eléctrica.

Junta d>' Energía Nuclear, División de Física leórica, Madrid

"Información que debe aportar el suministradorde la caldera nuclear (NSSS) para poder efectuar lagestión del combustible".VELARCE, G.; AGUILAR, F.; AUNERf. C ; ARAGONÉS, J.M.; GÓMEZ, M.; GUERRA, J .

PALMERO, A.; SERRANO, J . (1972) 18 pp.

Se relaciona un conjunto de parámetros nucleares, termohidráulicos y mecá-

nicos, necesarios para la gestión y diseño de los elementos combustibles en los

PWR, los cuales deber, ser suministrados por el Fabricante del Reactor a la

Empresa Eléctrica.

J.E.N. 245

Junta de Energía Nuclear, División de Física Teórica, Madrid"Información que debe apor tar el suminis t rador

de la ca ldera nuclear (NSSS) p a r a poder efectuar lagestión del combust ible" .VELARDf, 6 . ; AGUILAR, F . ; AHNCRF, C ; ARAGONÉS, J . M . ; GOMLZ, M . ; GUERRA, J .PALMFRO, A . ; SERRANO, J . (1972 ) 18 p p .

Se relaciona un conjunto de parámetros nucleares, termohidráulicos y mecá-nicos» necesarios para la gestión y diseño de los elementos combuslibles en losPWR, los cuales deben sor suministrados por el Fabricante del Reactor a laImpresa Eléctrica.

J.E.N. 245

Junta de Energía Nuclear, División de Física leórica, Madrid"Información que debe apor tar el sumin is t rador

de la ca ldera nuclear (NSSS) p a r a poder efectuar lagestión del combust ible" .VELARDE, G . ; AGUILAR, F . ; AIINERT, C ; ARAGONÉS, J . H . ; GÓMEZ, M . ; GUERRA, J .PALMERO, A . ; SERRANO, J . (1972) 1 8 p p .

Se relaciona un conjunto de parámetros nucleares, Lermohidráulicos y mecá-nicos, necesarios para la gestión y diseño de los elementos combuslibles en losPWR, los cuales deben ser suministrados por el Fabricante del reactor a laEmpresa Eléctrica.

J.E.N. 245

Junta de Energía Nuclear, División de Física Teórica, Madrid"Information to be reques ted í rom the NSSS vendor

for fuel management capabi l i ty" .VELARDE, Q., AGUIIAR, I . ; AIINIRI, C , ARAGONiS, .M., Wñ¿, M., GIJLRRA, i.

PALffRO, A., SERRANO, . (197/J. 1« pn.

A Si I oí lh nuclear, Ihormal-hydraulic, and mnchamcal parara I rs mct bar>

10 u rlorm Ihi f i 1 1 in n L manag>m ni and design for PWR's is l is lnd. Ih'.,

dala m'isl bi j i np l i d by lh' R'aclor Manulacl irer to lh i l u l i l , .

J . E . N . 245

'unía d r~n rgia Nut lear, División d-1 lísha hórua, Hadrid-" I n f o r m a t i o n to be r e q u e s t e d í r o m the NSSS vendor

for fuel m a n a g e m e n t c a p a b i l i t y " .

VELARDE, G. ; AGHIAR, L ; AIINIR1. C ; ARAbONi S, ' . M . ; GOMt/, M., GUHIRA, .PALMERO, A . ; Si RRANO. . ( W . ' j . 1 " .|,.

A set o í lh n ' i t l a r , Hunnal ! n d r a n h i , and ni i h a n i í a l naramclTS n ic>sarylo pLrlorm Ihi I i I 1 ra>nh manaq muí1 anrl d sign lo r PWR's i s l i s l e d . fl icsrda la musí b supnh d by lh R at l o r Marvilact i r r lo Ih

J.E.N. 245

Junla do Energía Nuclear, División de Física Teórica, Hadrid"Information to be reques ted from the NSSS vendor

for fuel management capabi l i ty" .VElARDf, G., AGUIIAR, F.; AHNERT, C , ARAGONÉS, J.M.; GOMCZ, M.; GUERRA, J .

PALMIRO, A., SI RRANO, J . (1972). 18 pp.

A sol of the nuclear, iherrnal-hydraulic, and mechanical parametors neccosary

lo pi rform Un-1 íuel - l i mcnts managemenl and design for PWR's is l i s ted . Ihese

dala nust b Sipplirf l by Ihp Rcaclor Manufacluror to Ihe U l i l i l y .

J.E.N. 245

liinla d> n> rgía Nuc h ar, División de Física Teórica, Madrid." In fo rma t ion to be r e q u e s t e d f rom the NSSS vendor

ii

. M . ; GOME/, H . ; GUERRA,

for luel management capabilityV I I A R D i , G . ; AGUIIAR, F . ; AIINER1, C ; ARAGONLS,PAIMERO, A . ; Si RRANO, '. U 9 7 2 j 1 8 p p .

A s.jl of lh ni.char, Ihcnnal hydraulit, and michanical pararaclors neccosarylo pnrlortii Mu f i I i l imi' i i ts iiianagmi ni and di sign for PWR's is l i s led . fhcsodala musí b snppliul by Un R> arlor Hanufatluri r lo li l i U l i l i l y .