Sp ISSN 0081-3397

porEmilio MínguezCarolina AnhertJosé M. AragonésÁngel EstebanManuel GómezGuillermo LeiraRafael Martínez

JUNTA DE ENERGÍA NUCLEAR

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

Las solicitudes de ejemplares deben dirigirse aeste mismo Servicio.

Los descriptores se han seleccionado del Thesaurodel INIS para describir las materias que contiene este infor_me con vistas a su recuperación. Para mas detalles cónsultese el informe IAEA-INIS-12 (INIS: Manual de Indización) yIAEA-INIS-13 (INIS:Thesauro) publicado por el OrganismoInternacional de Energía Atómica,

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

Este trabajo se ha recibido para su impresión enMarzo de 1.975.

Depósito legal nQ M-12027-1975

1. INTRODUCTION

2. REQUESTED INFORMATION

2.1. Core Design 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. Additional Data

2.2.1. Reactor Dynamics,

2.2.2. Reactor Coolant System.

2.2.3. Instrumentation and Safety Features.

2.2.4. List.ofFigures 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.

2.3.4. List of Figures relative to Design Results.

2.4. General Requests

1. INTRODUCCIÓN.

1.1.-

Uno de los principales objetivos, que en el campo de

la Tecnología Nuclear, pueden realizar las naciones en vías de

desarrollos es la gestión y el diseño da los elementos combus-

tibles de los reactores nucleares q_ue paulatinamente se vayan

importando. Para llevar a cabo este objetivo, se requieren

tres condiciones;

i] Disponer de un equipo de ingenieros, con suficien-

te experiencia en el diseño de elementos combustibles,

ii) Disponer de un computador con una memoria de unas

280000 palabras de capacidad CCDC-6600, Univac 1108), 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

Universidad, la formación del personal postgraduado, podría

real izarse en dos o tres años•

1.3.-

El grupo mínimo de códigos necesarios para la gestión

y diseño de elementos combustibles puede estar compuesto por:

LeopardLáserAss aultFog

. , Nutrixi ) Nuclear . , . , , . « , , . . „ ,-,

NuflowDTF-IVPDQ-7TimocCitation

Boleroii) Termohidráulico Caramba

Forcir

Figro

iii) Termome canico Cygro

iv) Economía Fuel Cost II y IV

PlankinN ,, . - . . , Paretv; Dinámica y accidentes . „ . ,J Splosh

Relap-4

La descripción de alguno de estos códigos se rea-

lizo en el informe JEN. 250.

La JEN dispone actualmente de todos ellos, excep-

to de los códigos clave: PDQ-7, Figro y Cygro, cuya exporta-

ción, fuera de los E.U.A. está prohibida.

i.4.-

Respecto a los parámetros de proyecto, en este in-

forme se relacionan aquellos que creemos son necesarios para

la gestión y diseño de los elementos combustibles, los cuales

han de ser suministrados por el Fabricante del Reactor, a la

firma del contrato.

Nuestro proposito es colaborar con las Empresas

Eléctricas españolas, para que estos parámetros sean exigidos

al Fabricante de la serie de reactores que actualmente se van

a contratar, los cuales, desgraciadamente, no fueron exigidos

en el contrato de los anteriores reactores.

Guillermo VELARDE, febrero 1.975

-3-

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, composition for

Core baffle

Core barrel

Thermal shield

Top reflector

Bottom reflector

Side reflector

H O/U volume ratio Caverage 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

Dimensions

Data specifications are for cold conditionsTolerances are included (when possible).

-n-

Sapcer grids

Number of grids per assembly, normal and withmixing vanes

Material composition

Wéxgiit per grid

Dimensions and location

Drag coefficient

End fittings

Material composition

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 Cas UO }

Zircaloy weight

Total weight

Fuel Rods

Total number

Ciad material

Ciad thickness

Ciad outside diameter

Gap filler gas, composition

pre ssure

allowable leak rate

Fuel loading per rod Cfor each región)

Plennum volume

Fuel Pellets

Material

Density, for each región

Oxygen/Uranium ratio

-5-

Impurities and equivalent boron content

Moistu-re content

Pellet fuel loading per cm. of height

U-235 initial enrichment, for each región

Initial load (g/cm) and composition of burnable poisonadded, if any

Pellet diameter (for each región)

Pellet height

Cluster Control Assemblies

Assembly weight (dry)

Absorber material composition

Absorber density

Absorber diameter

Absorber active length.

Ciad material composition

Ciad th.ick.ness

Ciad outside diameter

Number of control rods per cluster•L. ^ VT • • 4.1. i j rfull lengthNumber of assemblies with control rod i . , "

partí al length

Burnable Poison Rods

Number (total)

Material composition

Poison content (w/0 in rod)

Outside diameter

Inner tube, 0.D.

Ciad material

Ciad thickness

Inner tube material

Poison loading, gm per cm of rod

Initial reactivity worth (% Ap, hot and cold)

Excess Reactivity

Máximum fuel assembly K (cold, clean, unborated water)

Máximum core reactivity (cold, zero power, BOC)

-6-

2.1.2. THERMAL AND HYDRAULIC DESIGN PARAMETERS

General Data

Nominal power, MWt

Total core heat output

Heat generated in fuel

Máximum thermal overpower

Nominal system pressure

Fraction of heat generated outside fuel rod

Coolant Flow

Total coolant flow rate

Bypass coolant flow rate (feet/seg)

Bypass coolant flow Clb/h)

Average mass velocity

Primary coolant heat removal

Coolant flow for heat removal only

Nominal assembly coolant flow

Máximum rated assembly coolant flow

Average coolant velocity along'fuel rods

Minimum coolant velocity along fuel rods

Core inlet pressure Cminimum)

Pressure drop plenum 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 distribution

Coolant Temperature or Enthalpy

Nominal inlet temperature at rated power

Máximum inlet temperature at rated power

Average rise in vessel at rated power

-7-

Average rise in core at ratea power

Average temperature in vessel at rated power

Average film coefficient at rate power

Average film temperature difference at rated power

Heat Transfer

Average power density

Average specific power

Average lineal heat rate

Máximum lineal heat rate

Rated power

Design overpower

Active heat transfer área

Máximum /Kd8 (hottest rod)

Average heat flux at rated power

Hot channel máximum heat flux

Rated power

Design overpower

Crude and oxide conductance expected in the ciad

Hot Channel Factors

Engineering hot channel factors

a) Heat flux hot channel factor (F )q.

This factor should contain sunfactors 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 rod pitch

Fuel rod bowine

Flow re distribuí ion due to high resistance in hot channels

Flow mixing inside a fuel asse.mbly

Maldistribution on inlet flow

Overpower factors

Heat balance error

Instrument error

Instrument uncertainly for power and temperature

Transient overshoot

Instrument dead band

Total design

Design Mínimum Margin to Incipient Fuel-Clad Damage

Minimum allowable DNBR and suitable correlation

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 Paratneters

Turbulent mixing parameter without mixing vanes

Turbulent mixing parameter with mixing vanes

Friction factor for diversión cross flow

Diversión momentum factor

Turbulent momentum factor

- 9 -

2 i l ' 3 t LIST OF FIGURES RELATIVE TO CORE DESCRIPTION

1. Reactor vessel and in te rná i s

2. Core cross sect ion

3. Core ba r re l assembly

4. Fuel assembly out l ine

5. Grid assembly

6. Guide tube assembly

7. Rod cluster control assembly outline

8. Burnable poison rod design

9. Fuel loading arrangement

10. Rod cluster control assembly pattern

11. Burnable poison loading pattern

12. Burnable poison rod arrangement within an assembly

13. Distribution (assembly-wise and within assembly) ofpoison added to the fuel during manufacturing ( i f any)

14. Pellets description wit dimensions

-10-

2.2. ADITIONAL DATA

2.2.1. REACTOR DYNAMICS

Control Rods

Total number of steps (axial positions)

Heigtht of each step

Máximum withdrawal speed

Normal withdrawal and insertion speed

Weight of control rod and drive line

General Data

Effective prompt neutrón lifetiae, and

Effective delayed neutrón fractions for each group

rBOC, HZP, AROs critical boron concentrationat EOC, HFP, EROS no boron

Neutrón source strength

Normal heat distribution Crod/coolant)

ídem for residual heat

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 temperature

Coolant outlet temperature

Flow rate

Pressure loss

Heat transfer área

Primary side water volume

- 1 1 -

Secondary side:

Steam pressnre at fu l l power

Fe.eC.vf a "ter temperature

Steam r^ow rate ( to taJ)

Shell O.D., upper/lower

Shell thickness, upper/lower

Tube material

Number of U-tuoes

U--cube oui-side díame té !•

Tui)-¿ 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 volumfe

Cold leg volume

Reactor in le t piping, I.D.

Reactor in le t piping, nominal thickness

Reactor outlet piping s I.D.

Reactor outlet piping, nominal thickness

Reactor Vesse.1 Desjgn Parameters

Design/operating pressure

Design temperature

Reactor coolant in l e t temperature

Reactor coolant outlet temperature

Pressure losses through vessel including nozzles

Reactor outlet plenum volume

Reactor inlet plenum volume

Core bypass volume

-12

Reactor Coolant Pump Design Parameters

Number of pumps

Design pressure/operating pressure

Design temperature

Developed head

Capacity

Characteristic curves

Power (nameplate)

Pressurizer Design Parameters

Volume of pressurizer and surge line

ídem up to nominal water level

Pressurizer spray rate (máximum)

Pressurizer heater capacity

Pressurizer relief tank volume

Coolant Chemistry

Recommended valúes (or typical valúes) for:

PH

Conductivity

Ppm H , 0 , Cl~3 total solids, etc.

2.2.3. INSTRUMENTATION AND SAFETY FEATURES

Reactor Trip System (RTS)

Number of safety rod banks

Total insertion time of safety banks

Insertion tie up to dashpot entry

Variables wich actúale the RTS, with its correspondigsetpoints and time delays (table)

RTS interl&cks (table)

Engineered Safety Systems

Safety Injection System (SIS), number of pumps

Boron Concentration of SIS water

Water volume of refueling and injection tanks

-13-

Water volume in the accurnulators

Trischarge pressure of accumulators

-V.ar'iables and setpoints for SIS actuation (table)

"-Variables and setpoints for feedwater line isolation

IVsr'á-ahl&s and setpoints for steain line isolation

llnterlocks for these safety systems (table)

In-Core Instr.umentation

Total number of thermocopuples 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 selected assemblies for nuclear instrumenta-t ion (if fixed)

3. -Location of neutrón sources in the core

4. Dimensioned drawing of reactor coolant pumps

5. ídem for steam generators

6. Ideffi 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) Fuel grapple

b) Fuel transport machine

10. RCCA insertion v.s. time in scram"

1.1-. SLS flow rate v.s. reactor coolant pressure

12. Auxiliary feedwater flow v.s. time after trip

-14-

2.3. DESIGN RESULTS

2.3.1. NUCLEAR DESIGN RESULTS"

Excess reactivity distribution (BOC)

CZP, clean

HZPS clean

HFP, clean

HFP, Xe and Sm equilibrium

Shutdown boron concentrations

Refueling ahutdown (k = 0.90) ; ARI

Clean, CZP

Shuxdown (k = 0.99); ARI

Clean, CZP

Clean, HZP

Shutdown (k = 0.99); ARO

Clean, CZP

Clean, HZP

Shutdown (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-lenght group inserted

All control groups inserted

ARI means All Control R_ods InsertedARO means A_ll Control Rods 0_utCZP means C_old Zero P_owerClean means without fission producís (Xe, Sm, ...}HZP means Hot Z_ero P_owerHFP means H_ot F_ull PowerBOC means B_eginning 0_f first C_ycleEOC means End Of first Cycle

-15-

All shutdown groups inserted

All but one rod inserted

ARI

BOCS HFP, ARO

Clean

With equilibrium Xenón

With equilibrium Xenón and Samarium

Reduction in critical boron concentration with fuelburnup (ppm/Mwd/t)

First cycle

Reload cycle

Moderator Temperature Coefficient Ccore and each región)

At BOCS HZP5 ARO, cleans critical boron concentration

At E0C3 HFP3 ARO5 equilibrium Xenón and Samarium, no boron

Moderator Pressure Coefficient (core and each región)

At HZP, ARO3 critical boron3 clean

at BOC

at EOC

Doppler Coefficient (core and each región)

At HFPj ARO s equilibrium Xenón and Samariunij 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,

and redistribution)

Operational maneuvering band and control rod bite

Total control

-16-

Contaol rods worth

Integral worth of each control rod group

at B0C3 HZP5 clean, critical boron

at BOC, HFP, eq.uilibrium Xenón and Samarium, critical boron

at EOCj HFP, eq_uilibrium Xenón and Samarium, no boron

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

te HZPa BOC condition with the critical boron concentration

corresponding to full power

Máximum worth of an ajected rod and resulting radial peaking

factor

Máximum 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

Worth variation with burnup (AK/K) at HFP, equilibrium Xenón

and Sni) critical boron

Heat generation rate inside the rods,

Isotopic Inventory and Fuel Management

For each previous cycle;

Cy-cle lífeti^me at rated power

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 35

Total U in discharged assemblies

Total U in discharged assemblies2 35

Average U enrichment in discharged assemblies2 3 9 2 4-1

Total Pu + Pu in discharged assemblies

- 1 7 -

2.3.2. THERMO-HYDRAULIC RESULTS

Design Mínimum Margin to Incipient Fuel-Clad Damage

Calculated mínimum DNBR

Rated power

Design overpower

Steady reactor conditions to give a mínimum 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 temperatura

Effects of fuel densifícatión on DNBR

Coolanr Temperature or Enthalpy

Reduction in assumed hot channel anthalpy rise due to

interchannel mixing

Average active coolant outlet temperature or enthalpy at

rated power

Hot channel outlet temperatura or enthalpy

Rated power

Design overpower

Hot channel outlet void fraction

Rated poWer

Design overpower

Length to start of local boiling

2.3.3. TRANSIENT AND ACCIDENT ANALYSIS

. , , , . , , , rfrom subcritical condxtionUncontrolled rod withdrawali at power

Partial loss of fprced reactor coolant flow

Turbine trip

Loss of normal feed water

Excessive load increase

-18-

Accidental despressurization of the reactor coolant system

Rupture of a inain steam pipe

Inadvertent loading .of a fuel assembly into an improper position

Fuel handling accident

Major loss of coolant accident

Steam generator tube rupture

Single Reactor Coolant Puflip Locked Rotor

RCCA ejection

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

3. Máximum and mínimum 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 peakingfactors v.s. insertion3 at B0C3 clean, HZP, critical boron.

6. ídem, at BQC, HFP, equilibrium Xenón and Samarium, criticalboron.

7. Idem3 at EOC, HFP, equilibrium Xenón and Samarium, no boron.

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

9. ídem, at ARO, HFP.

10. Differential boron worth v.s. boron concentration at HFP,ARO, BOCj and several burnups (average for core and eachregión.

11. Power distribution and peaking factor at BOC, HZPs clean,ARO, critical boron.

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

13. ídem, at HFP, ARO, equilibrium Xenón and Samarium, criticalboron (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 groupin (.for each control rod group).

16. Doppler coefficient 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 andEOC, critical boron.

-19-

20. Moderator temperature reactivity coefficient v.s, mode-rator temperatura at nominal pressure, (HFP)S ARO, BOCseveral boron 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, équilibriúinXenón and Sainarium and critical boron Cfor different bur-nups),

23. Axial peaking factor v.s. time, for a typical Xenóntransient Cunstable and stabilized with part length rodmotion).

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 data(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 uranium dioxide.

29. Cladding ínternal pressure v.s. time.

30. Temperature rise in the channels of a rod bundle v.s,channel power density.

31. Fuel cladding and U0 temperature limits v.s, time orfuel bundle exposure.

32. Thermal conductivity of cladding.

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

34. Fuel rod heat flux limits v.s. time or fuel bundle expo-s ure .

35. Core inlet temperature v.s. power level program.

36. Fraction of fission gases released to the gap v.s. burnup.

37. Diametral gap v.s. burnup.

- 2 0 -

2 • ^ ' GENERAL REQUEST.

a) Of f i c i a l documents: FSAR, Tech Specs . . .

b) Codes for r e a c t o r s u r v e i l l a n c e and process ing of in -co re

ins t rumen ta t ion .

c) Programming (software) of process computer ( i f any) .

d) Main design r e p o r t s :

Core analyses for 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 capability

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

parameters3 e tc . ) .

g) Last versión of cr i t ical heat flux correlation.

J.E.N. 298 J.E.N. 298

Junta de Energía Nuclear, División de Tecnología de Reactores, Madrid"Información c i u e debe a p o r t a r el s u m i n i s t r a d o r dela ca lde ra nuc lear (NSSS) p a r a efectuar la gest ióndel combust ib le de un r e a c t o r del tipo PWR".MINGUEZ, £ . ; ANHERT, C ; ARAGONÉS, J . l t ; ESTEBAN, A . ; GÓMEZ, H . ; L E I R A , G . ;MARTÍNEZ, R . ( 1 9 7 5 ) 2 0 p p .

Se relaciona un conjunto de parámetros nucleares, tormohídráulicos y mecáni-

cos, actualizados según los nuevos diseños de reactores PWR, que últimamente

han aparecido. Todos ellos son necesarios para efectuar la gestión y diseño del

combustible, y deben ser suministrados por el Fabricante del reactor a la Em-

presa Eléctrica propietaria del misino.

Junta de Energía Nuclear, División de Tecnología de Reactores, Madrid"Información que debe aportar el suministrador dela caldera nuclear (NSSS) para efectuar la gestióndel combustible de un reactor del tipo PWR".

MINGUEZ, E . ; ANHERT, C ; ARAGONÉS, J . M . ; ESTEBAN, A . ; GÓMEZ, M . ; LEIRA, G . ;MARTÍNEZ, R . ( 1 9 7 5 ) 2 0 p p .

So relaciona un conjunto de parámetros nucleares, termohidráulicos y mecáni-

cos, actualizados según los nuevos diseños de reactores PWR, que últimamente

han aparecido. Todos ellos son necesarios para efectuar la gestión y diseño del

combustible, y deben ser suministrados por el Fabricante del reactor a la Em-

presa Eléctrica propietaria del mismo.

J.E.N. 298

Junta de Energía Nuclear, División de Tecnología de Reactores, Madrid

"Información que debe aportar el suministrador de

la caldera nuclear (NSSS) para efectuar la gestión

del combustible de un reactor del tipo PWR".MINGUEZ, E . ; ANHERT, C ; ARAGONÉS, J . M . ; ESTEBAN, A . ; GÓMEZ, M . ; LEIRA, G . ;MARTÍNEZ, R . ( 1 9 7 5 ) 2 0 p p .

Ss relaciona un conjunto do parámetros nucleares, termohidráulicos y mecáni-cos, actualizados sogún los nuevos diseños de reactores PWR, que últimamentehan aparecido. Todos ellos son necesarios para efectuar la gestión y diseño delcombustible, y deben sor suministrados por el Fabricante del reactor a laEmpresa Eléctrica propietaria del mismo.

J.E.N. 298

Junta de Energía Nuclear, División de Tecnología de Reactores, Madrid

"Información que debe aportar el suministrador de

la caldera nuclear (NSSS) para efectuar la gestióndel combustible de un reactor del tipo PWR".MINGUEZ, E.; ANHERT, C ; ARAGONÉS, J.M.; ESTEBAN, A.; GÓMEZ, M.; LEIRA, G.;MARTÍNEZ, R. (1975) 20 pp.

Se relaciona un conjunto de parámetros nucleares, termohidráulicos y mecáni-cos, actualizados según los nucevos diseños de reactores PWR, que últimamentehan aparecido. Todos ellos son necesarios para efectuar la gestión y diseño delcombustible, y deben ser suministrados por el Fabricante del reactor a la Empre-sa Eléctrica propietaria del mismo.

Clasificación INIS y Descriptores. P31.- PWR fypo Rcaclors ; Fue! Managnm jnt ; C lasi ficaci 5n, INIS y Doscri plores. E31.- PWR Type Roactors.: fuel Manageiirnl;

Fucl [lcmcnls; Soeci f ical ions; Reactor Cores; Fuel Assemblies; Reactor Cooling , fue! Elcments; Spccificalions; Reactor Coros; Fuol Assomblios; Reactor Cooling

Syslcms; Reactor Opei'at ion; Muí li-Parainotor Analysis; Data; R"ac lor Safo Lv; , Systems; Reactor Opera I ion; Huí Li-Paramo l n r Analysis; Dala; Reactor Safety;

Reactor Accidonts. R;aclor Accidonls.

Cl asi I i cae ion INIS y Descriptores. [ 3 1 . - PWR Fype Reactors ; Hi. 1 Hanageraonl; Clasificación INIS y Descriptores. [31 . - PWR fype Reactora; Funl Hanagpmenl;

f i i . l Llcments; Speci f ications; Reactor Cores; Fuol Assemblies; Reaclor Cooliny Fue! Flenimt:,; Spoci f ical ions; Reactor Coros; fucl Assemblies; Reactor Cooling

Systems; Reactor Opera I ion; Mu 11?-Paramo lo r Analysis; Dala; Reactor Safely; Systoms; Reactor Operation; fluí ti-Parametor Analysis; Data; Ifcaclor Safely;

Reactor Accidenls. Reactor Accidonts.

J . E . N . 298

Junta do Energía Nuclear, División do Tecnología de Rpaclores, Madrid" I n f o r m a t i o n to be r e q u e s t e d f rom the NSSS v e n d o rfor fuel management capability for PWR's".MINGUE/, L ; ANIO1, C ; ARAGONÉS, J . l i ; ES1CBAN, A.; GÓMEZ, M.; LFiRA, G.;MARTÍNEZ, R. (1975) 20 pp.

A sel of [\\P nucí par, thormal-hydraulic, and iiiechanical parainr>lers is l i s lod ;Ihis cinta aro I he inain paramolers for I he no»/ design of lhu last PWR's. Al 1 Ihisare noccosary lo per forra Lhu fuol elosimnls manaqemenI and di'sign, and i I mustbe supplied by Ihe Reactor Manufacturar lo the U t i l i l y .

J . E . N . 298

Junta de Energía Nuclear, División do Tecnología do Reactores, Madrid"Informat ion to be r e q u e s t e d from the NSSS vendorfor fuel managemen t capabi l i ty for P W R ' s " .MINGUE/, [ . ; ANIIERT, C ; ARAGONCS, J . M . ; ESTEBAN, A . ; GÓMEZ, M . ; LFIRA, 6 . ;MARMNEZ, R. ( I 9 7 r i ) 20 p p .

A set of tho nuclear, thmnal-hydra'ilic, and mechanical parameiers is us ted ;

Ihis data are tho rnain paramoters for the new design of tho last PWR's. Al l i h i s

are nnccesary lo perfora the fuol el ornen Is managomeni and design, and i t" must

be supplied by the Reactor Manufacturar to the U t i l i t y .

J.E.N. 298

Jimia de Energía Nuclear, División de Tecnología do Reactores, Madrid."Informat ion lo be r eques t ed [rom the NSSS vendorfor fuel m a n a g e m e n t capabiliLy fox- PWR's 1 1 .

M I N G U E / , r . ; A N I l f R l , C ; A R A G O N É S , J . M . ; E S I E B A N , A . ; G Ó M E Z , M . ; L f I R A , G . ; tM A R I I N C / , R . ( 1 9 7 b ) 7 0 p p .

A sel of the nuclear, Ihunnal-hyclnulit, and mechanical parameters is l ís led; '

Ihis dala are Ihu main paramolors Tor Ihe new dosign of the last PWR's. Al l this

are ni ccesary to pcplonn Ihe fuel eljinenls managomeni and dosign, and i I musí i

be supplied by Ihe Reactor Manufaclurer to thQ U l i l i t y . !

J.E.N. 298

Junta de Energía Nuclear, División de Tecnología de Reactores, Madrid."Informat ion to be r e q u e s t e d from the NSSS vendorfor fuel m a n a g e m e n t capabi l i ty for P W R ' s " .MINGUEZ, L ; A N I O T , C ; ARAGONÉS, J . M . ; ESTEBAN, A . ; GÓMEZ, M . ; L C I R A , G . ;M A R Í I N I 7 , R . ( 1 9 7 b ) 2 0 p p .

A set of tho nuclear, thermal-hydraulic, and mechanical parame le rs is Usted;

this data are the main parameiers for the now design of the lasl PiíR's. A l l Ihis

are noccesary Lo perlorm Ihe fuol elements managemenl and design, and i t " must

be supplied by Ihe Reactor Manufactiiror to tho U t i l i t y .

Cld->i l i cauon INIS y D s( r íp lu r J C . F I- Pl'ÍR I vpt Riaclors; [ m i Manjgpintnl, C h s i f i r ti 111 INIS y Dt 11 iploi ^s. F Ti. PWR I y pe Reactor ; H K I Managenu n I ,

f u t í [ lempnl c ; So tc ih rahon 0 , ' Re actor Con. ; H I P I As°i,inbl KO,* R acloi Coolinq f i l I I 111 r t , S J P C I I V J ion , Ri i 1 oí Co^ ; In 1 A st ubi ios, Reactor Cooling

S si un0 ; RJ v I r f 0|pral >nn; M i l h-Pi ranph i Arn l^ i ; D i h ; \b v I or ST U b ; Sv^lom ; R i i l o i On r -d io i r M 1^1-Pirampti r An i lys i ; D i ta ; Riaclor Sa fp ' \ ;

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t l a s i l i catión INIS y Doscr iptoie-,. L í l . - PWR lypr R i a c l o r ; Hiél Hanag me n ' ;

f i d r i i inenl ; Soi n í i r a h o n ; Rtat lor C o r e ; fucl Ássnnblios; Roaclor Coohng

Sy L L ni ; R a I nr O > i l i o n ; M J! t ? -Pat amo I f r Analvsis; D i ta ; R a d o r S i l i I ' ;

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