SpISSN 0081-3397

porEmilio MínguezÁngel EstebanManuel GómezGuillermo LelraRafael MartínezJuan Serrano

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 in-forme con vistas a su recuperación. Para más detalles consultese el informe IAEA-INIS-12 (INIS: Manual de Indización)y IAEA-INIS-13 (INIS:Thesauro) publicado por el OrganismoInternacional de Energía Atómica*

Se autoriza la reproducción de los resúmenes analí-ticos que aparecen en esta publicación

Este trabajo se ha recibido para su impresión enMarzo de 1, 975,

Depósito legal nQ M-12026-1975

-1. INTRODUCCIÓN.

2. REQUESTED INFORMATION

2.1. Reactor Core Design Data

2.1.1. Nuclear Design Bata

2.1.2. Thermal and Hydraulic Design Data

2.1.3. List of Figures relativa to Core Descripticn

2.2. Additional Data

2.2.1. Reactor Coolant System

2.2.2. Dynamics Data

2.2.3. Instrumentation ans Safety Features

2.2.U, List of Figures relativo to Additiona] Data

2.3. Design Results

2.3.1. Nuclear Design Results

2.3.2. Thermal and Hydraulic Results

2.3.3. Transient and Accident Analysis

2.3.4. List of Figures re latí ve to Design "RP. svlts

2 . >+ . General Requests.

-1-

1. INTRODUCCIÓN.

1.1. Cuando este mismo trabajo se realizo para reactores

de tipo PWR, uno de los principales objetivos fue dar

una descripción detallada de los "parámetros de proyec-

to, que debe suministrar el fabricante del reactor a

las Empresas Eléctricas.

De la laisma forma, ahora se hace extensivo aquel pa-

ra reactores del tipo BWR. Muchos de los parámetros

de diseño son válidos para ambos tipos, por tener

muchas características comunes, üin embargo, las

propiedades físicas del refrigerante obligan a dise-

ños diferentes.

1.2. El grupo de códigos necesario para hacer frente a la

gestión y diseño de elementos combustibles, que de-

be poseer cualquier empresa que se dedique a estas

misiones, tiene q_ue estar compuesta por:

C LEOPARDC LÁSERC ASSAULTC FOG

''C CITATION( NUTRÍ X( NUFLOtíC DTF-IVC TIMOC

i) Nuclear

(BOLERO¿CARAMBA

ii) Termohidráulico .(FQRCIR(FLODISTCFIGRO

iii) Termome cánico CYGRO

iv) Económicos ,, FUEL COST II y IV

(-COSTAX-BWR

v) Cinética .,..,, (SPLOSH(RELAP

Todos estos códigos están disponibles actualmente en la

División de Tecnología de Reactores de la J.E.N. , excep-

to algunos que son restringidos como FIGRO y CYGRO.

1.3. Las necesidades de máquina para el buen uso de estos

programas deben ser las de un computador con una capa-

cidad de memoria de 280.000 palabras o más, tales como

CDC-6600, UNIVAC 1108.

Respecto a los parámetros de proyecto, en este informe

se relacionan aquéllos 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 Eléctri-

cas Españolas, para que estos parámetros sean exigidos

al Fabricante de la serie de reactores que actualmente

se van a contratar, los cuales, desgraciadamente, no fue-

ron exigidos en el contrato de los anteriores reactores.

Guillermo VELARDE, febrero 1.975

-3-

2. REQUESTED fNFORMAT^ON.

2'1' REACTOR CORE DESIGN DATA

2.1.1. NUCLEAR DESIGN DATA*

Core dará

Reference design thermal powers MWt.

Total weight of UO in core

Total weight of U in core

Circunscribed core diameter

Equivalent core diameter

Effective core volume

Active fuel height

Number of fuel assemblies

Number of control rods

Number of burnable curtains (If any)

Overall core description

From Utilities

Operation sctieduling

Assumed load factor or effective full power days

Replacement energy cost

Technical specifications

Core and Refflectors Structure

H O/U volume ratio Caverage in core)

Dimensions and material composition for;

Core baffle

Core barrel

Thermal shield

" All data specifications are for cold conditions

Radial reflector

Compos ition

Thi ckness

Top reflector

Compos ition

Thickness

Bottom reflector

Comp os ition

Thickness

Fuel Assembly Data

Fuel rod array (specifing enrichment and dishing)

Weight of UO per assembly:

UNDISHEDDISHED

Overall dimensions

Fuel assembly pitch

Nominal active fuel lenght

Fuel element channel CBundle Channel)

Material composition of the box

Thicknes s

Overall lenght

Cross sections dimensions

Channel exit área ratio

End fittings

Material composition

Total weight

Dimens ions

Lower and upper tie plates

Material composition

Lenght

-5-

Spacer grids

Number per assembly

Material composition

Weight of grids per assembly

Dimensions

Drag Coefficient

Puel rods

Number of fuel rods per assembly

Outside diameter

Rod pitch

Ciad thickness

Ciad material

Cladding process

Fuel loading per fuel rods (as UC> )

UNDISHEDDISHED

SPACER

Gap-pellet to ciad, inch

Gap filler gas (composition, pressure)

Lenght of gas plenum, inch

Spacer rod description

Fuel rods with Gd 0 (If any)

Number of fuel rods with Gd 0 per assembly

Loading of Gd 0, per fuel rod (g

Location of part-lenght fuel rods with Gd 0

Height of burnable poison (Gd 0 ) for part-lenght fuel rods

Location of full-lenght rods with Gd 0,

Density of Gd 0

Water rods (If any)

Number per assembly

Material composition

Dimens ions

Location

-6-

Fuel pellets

Material

Density, % of theoretical

Diameter

Height

U-235, initial enrichment

Control rods

Control material composition

Control rod description:

Shape

Blade thickness

Blade span

Cladding thickness of control rod

Number of control material tubes per control rod

Tube outside diameter

TuBe inside diameter

Poison density

Control rods pitch

Control rod location

Absorber active lenght

Temporary control curtains Clf any)

Shape

Material composition

Widthj inch

Thickness, inch

Active control lenght

Curtain locations

-7-

2.1.2. THERMAL AND HIDRAULIC 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 assembly

Fraction of heat generated outside fuel rod but insideassembly

Coolant flow

Total coolant flow rate

Bypass coolant flow rate (feet/seg)

Bypass coolant flow Clb/h)

Dimensions of hypass flow paths

Total coolant flow for heat removal

Coolant flow for heat removal

Quality of recirculation flow

Feedwater flow rate

Nominal assembly coolant flow

Máximum rated assembly coolant flow

Average coolant velocity along fuel rods

Máximum 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 lenght at entrance

Unheated channel lenght at exit

Core inlet coolant flow distribution

Water level above riser exit

Riser height

Riser flow área

Riser equivalent diameter

Dowucomer flow área

Coolant Temperature or Enthalpy

Nominal inlet temperature ar rated power

Máximum inlet temperature at rated power

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 rated power

Average film temperature difference ar rated power

Feedwater enthalpy

Feedwater temperature

Entrance orifices for fuel assembly

Orifice types

Orifice distribution

Orifice diameter

Entrance loss coefficient

Pressure drop across orifices

Heat Transfer

Average power density

Average specific power

Average lineal heat rate

Rated power

Design overpower

Active heat transfer área

Máximum /Kd<8 (hottest rod)

Average heat flox at rated power

Hot channel máximum heat flux

Rated power

Crude and oxide conductance expected in the ciad

Q ~

Hot Channel Factors

Engineering hot channel factors

a) Heat flux hot channel factor (F )q

This factor should contain subfactors to accoi'Dt for

Variations in pellet diameter

Variations in pellet density

Variations in pellet enrichmert

Excentricity of the pellet

Variations in ciad diameter

b) Enthalpy rise hot channel factor CFA )

H

This factor should contain subfactors to account for

All the effects in part a) aboye

Variations in fuel rod piten

Fuel rod bowing

Fuel assembly bowingFlow redistributioi due to high resistance in hotchannels

Flow mixing inside a fue3 assemblv

Maldis trib ut i or¿ o~. i n .le t assepbly

Overpower factors

Heat balance error

Instrument error

Instrument uncertainly for povrer and te Tnoerat 'jre

Transient overshoot

Instrument dead band

Total design

Mixing Parameters in Channel

Turbulent mixing parameter

Friction factor for divergió? across f]ow

Diversión momentum factor

Turbulent momentum factor

-10-

2.1.3. LIST 0F FIGURES RELATIVE TO CORE DESCRIPTIQN

1. Reactor vessel cutaway

2. Typical core arrangement

3. Core lattice, with U-235 enrichments assembly dis-tribution, and type fuel rods distribution.

4. Fuel rod.

5. Fuel assembly

6. Grids description

7. Temporary curtain (If any)

8. Control rod

9. Fuel bundles with active lenght for burnable poison(If any)

10. Riser description

-11-

2.2. ADDITIONAL DATA

2.2.1. REACTOR COOLANT SYSTEM

Reactor Coolant Piping Design Parameters

Design pressure

Operating pressure

Design temperature

Hot leg volume

Cold 1-eg volume

Reactor inlet piping, I.D.

Reactor inlet piping, nominal thickness

Reactor outlet piping, I.D.

Reactor outlet piping, nominal thickness

Reactor Vessel Design Parameters

Design pressure

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 feedwater inlet pressure

Reactor feedwater inlet flow

Reactor feedwater inlet temperature

Core assemblies volume

Total reactor vessel volume

Reactor Coolant Pump Design Parameters

Number of recirculation pumps

Number of jet pumps

Design pressure

Operating pressure

Design temperature

Developped head

Cap aci ty

Re c i r c u l a t i on p ump flovr rate

Charac te r i s t i c curves

Power (naineplate)

Recirculation pumps inlet oressur?

Recirculation pumps inlet '•eaperature

Recirculation pumps outlet oressare

Re circulation pumps ouíleL L c T -- *- 5 A < -

Jet pumps volume

2.2.2. DINAMICS DATA

General Data

Effective prompt neutrón lifstire

Effective delayed neutro1" írac'io- \j gro<-(equilibrium)

Neutrón source strength

Normal heat distribution (r 1 c ' •=> ^̂

ídem for residual heax

Control Rods

Total number of axial steps

Height of each step

Máximum withdrawal speed

Normal withdrawal and insertion soee^

Weight of control rod and drive line

2.2,3. INSTRUMENTATION AND SAFETY FE ATURES

Reactor Trip System CRTS)

Number of safety rod banks

Total insertion time of safety fiaulcs

Instruments, setpoints and time del3,s ofthe Reactor Protection System Ct?ble)

Interlocks in the RTS (table)

Typical secundancy in this ins tr v-^.ei u atior

-13-

Engineered Safety Systems

Setpoints and time delays of the variables which actúatethe high-pressure core aspersión system (table)

ídem for low-pressure aspersión

ídem for the coolant inyection.system

ídem for automatic depressurization of the coolant system

ídem for the Reactor Vessel Isolation System

Number of pumps and/or valves in each system

Water volume and number of tanks feeding ECCS

Number of Dieseis in the Standby Electrical Supply System

Capability of the Residual-Heat Renoval System (water volume,number of pumps¡ nominal flow)

In-Core Neutrón Instrumentation

Number of in-core neutrón detectors (fixed)

Number of in-core detector assemblies

Number of detectors per assembly

Number of Flux Mapping neutrón detectors

Number and type of in-core neutrón sources

Range Cand Number) of Detectors

Source Range Monitor

Intermedíate Range Monitor

Local Power Range Monitor

Average Power Range Monitor

2.2.4. LIST OF FIGURES RELATIVE TO ADDITIONAL DATA

1. Description of bypass flow poths

2. Description of downcomer

3. Steam separator

4. Location of in-core neutrón instrumentation

5. Dimensioned drawing of reactor coolant pumps

6. Distribution of instrumentation for:

a) Loop temperatures

b) Reactor ex-core flux detectors

7. Schematic Diagram of Coolant Systems including isolatingvalves and Safeguards

8. Safety rod insertion vs . time in scram

9. Negative reactivity vs . time in scram (conservative)

10. RTS logic (schematic)

11. ECCS logic (schematic)

12. Vessel pressure vs. flow for high-pressure core aspersión

13. ídem for low-pressure aspersión

14. ídem for the coolant inyection systern

- 1 5 -

2 . 3 . DESIGN RESULTS

2 . 3 . 1 . NUCLEAR DESIGN RESULTS

Isotopic Inventory

Core average burnup at BOC ( i f t h i s i s ' t the f i r s t cycle)

Core average burnup at EOC

Reloading pat tern

Number of fuel assemblies discharged

Average burnup in discharged assemblies

Total U-235 in discharged assemblies

Average U-235 enrichment in discharged assemblies

Total U in discharged assemblies2 39 2 41

T o t a l Pu + Pu in d i s c h a r g e d a s s e m b l i e s

E xce s s re ac t i v i ty .dj^J^rib^utj. qii C at^JB 0 C )

CZP, c lean

HZP3 clean

HFP3 clean

HFPj Xe and S i , equilibrium

Moderator tempe r a t ure _cps¡f f i_c i en t

At 68°F3 A.K/K - °F water (CZP, clean)

Hotj no void3 AK/K - °F water (Xenón and Samarium equilibr-i nr "i

at BOC

at EOC

BOC means Beginning Of Cycle

MOC means M_iddle £f f i r s t £ y c l e

EOC means E_nd Of f i r s t £yc l e

CZP means Cold Zero Power

HZP means H_ot Z_ero Power

HFP means H.ot F_ull Pover

Clean means w i t h o u t f i s s i o n p r o d u c í s (Xe3 Sm, . , . )

CHFP- ip.sans C r i t i c a l Haat Flux R a t i o

-16-

Moderator Void Coefficient

Hot, no void, AK/K - % void

At rated output, AK/K - % void

at BOC

at EOC

Fuel Temperature Doppler Coefficient

At 68°F, AK/K - °F fuel

Hots no void, AK/K - °F fuel

At rated output, AK/K - T fuel

at BOC

at EOC

Poison Burnable «orth, Gd 0 Clf any)

At BOC worth, CAK/K)

Hot

Cold

At EOC worth, (AK/K)

Heat Generation rate inside the rods

Temporary control curtain worth (If any)

At BOC worth, (AK/K)

Hot

Cold

At EOC worth, (AK/K)

Heat Generation rate inside the temporary curtain

Control rods worth

Integral worth of each control rod group:

At BOC, CZP

At BOC, HZP, clean

At BOC, equilibrium Xenón and Samarium

At BOC, HFP, average and máximum void (%)

At EOC, HFP, average and máximum void (%)

-17-

Shutdown margin:

At BOC, HZP

At BOC, HFP, average and máximum void (%)

At EOC, HFP, average and máximum void (%)

Reactivity requirements for control rods

Máximum worth of an stucked rod and resulting radial peakingfactor

Máximum peaking factor and negative reactivity resulting froman inserted rod at full power

Control rod bite and máximum insertion rate

Heat generation rate inside the rods

2.3,2. THERMAL AND HYDRAULIC RESULTS

Sisign Minimum Margin to lncipient Fuel-Clad Damage

Calculated minimum CHFR and suitable correlation

Rated Power

Design iverpower

Steady reactor conditions to give a minimum CHFR

Power

Inlet temperature or enthalpy

Steady reactor to cause fuel centerline melting in hottest rod

Steady reactor power to cause ciad damage due to excesive fueltemperature

Effects of fuel densification on CHFR

Effects of geometry on CHFR

Coolant Temperature or Enthalpy

Average active coolant outlet temperature or enthalpy at ratedp ower

Hot channel outlet temperature or enthalpy

Rated power

Design overpower

Hot channel outlet void fraction

Rated power

Design overpower

-18-

2.3.3. TRANSIENT AND ACCIDENT ANALYSIS

Turbine trip

Closing of main steam isolation valves

Failure of pressure regulator (open or closed)

Loss of a feedwater heater

Misfunctionning of residual heat cooling system (decresingtemperature

Accidental start up of the high-pressure core aspersión systempump

Continuous withdrawal of control rods at power

Continuous withdrawal of rods during start up

Accidental opening of a realief/safety valve

Loss of feedwater flow

Loss of auxiliary electrical power

Trip of the recirculation pumps

Control rod falling (ejected-rod accident)

Ruptures of piping inside containment Closs of coolant accident)

Ruptures of piping outside containment (steam-line ruptureaccident)

Ruptures of feedwater system piping

2.3.4. LIST OF FIGURES RELATIVE TO DESIGN RESULTS

1. Local Power Distribution at BOC (clean; equilibriumXenón and Samarium)

2. Local Power Distribution at MOC

3. Local Power Distribution at EOC

4. Typical Gross Peaking v.s. Exposure

5. Axial Void Fraction Distribution (at BOC, MOC, EOC)

6. Radial Power Distribution and rod pattern

7. Average Axial Power Distribution (at BOC, MOC, EOC)

8. Radial Exposure Distribution (at BOC, MOC, EOC)

9. Axial Exposure Distribution (at BOC, MOC, EOC)

10. Holling Power Distribution Calculations

11. Differential rod worth for each control group, and axialpeaking factor v.s. insertion

12. Reactivity coefficients (Doppler, Moderator temperature,moderator void)

-19 -

13. Ef fec t ive fuel tempera ture v . s . r e l a t i v e power

14. Ef fec t ive fuel tempera ture at HFP v . s . rod burnup

15. Product ion and consumption of h igher i s o t o p e s v . s . burnup

16. Nuclear hot channel f a c t o r s for enthalpy r i s e and forhea t f lux v . s . rod i n s e r t i o n for the d i f f e r e n t con t ro lrod groups

17. Máximum and minimum c o n t r o l group i n s e r t i o n s v . s . powerle vel

18. Thermal conduc t i v i t y of uranium dioxide

19. Cladding i n t e r n a l p r e s s u r e v . s . time

20. Temperature r i s e in the channels of a rod bundle v . s .channel power dens i t y

2 1 . Fuel c ladding and UO temperature l i m i t s v . s . time orfuel bundle exposure

22. Thermal conduc t i v i t y of c ladding

23. Gap heat t r a n s f e r c o e f f i c i e n t v . s . burnup

24. Fuel rod heat f lux l i m i t s v . s . time or fue l feundl^exposure

25. Core i n l e t tempera ture v . s . power l e v e l program

26. Frac t ion of f i s s i o n gases r e l e a s e d to the gap v . s . burnup

27. Diametral gap v . s . burnup

-20-

GENERAL REQUEST

a) Official documents: PSAR3 Tech Specs

b) Main design reports:

Core analysis for cycle 1

Basic lines for fuel management for following cycles

c) P.rogramming of process computer

d) Other studies for:

Fuel manegement analysis in BWR

Historie data on the fuel performance

Behabiour of operating BWR's designed by the vendor

e) Codes for reactor surveillance and processing of in-coreinstrumentation

f) Cooperation for obtaining in-house fuel management capability(computer codess general method, up-dated valúes of empiricalp arametersI.

J . E . N . 297 E . N . 297

Jimia rlc Energía Nuclear, División de locnología de Reactores, Madrid"Información que debe apo r t a r el sumin i s t r ado r dela ca lde ra nuclear (NSSS) p a r a efectuar la gestióndel combust ible de un r e a c t o r del tipo BWR"MINGUI7, C ; ESTEBAN, A . ; GOME, H . ; LEIRA, Q . ; MARTIMEZ, R . ; SRRANO, J . ( 1 9 7 I J )

2 0 p p .S2 relaciona un conjunto do parámetros nucleares, fccrroohídráulicos v mecánicos,

actualizados según los diseños de reactores BWR.

Estos parámetros son necesarios para efectuar la gestión y diseño d.'l combus-t ib le , j) deben ser suministrados por el fabricante del Reactor a la EmpresaEléctrica propietaria del mismo.

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

"Información que debe aportar el suministrador dela caldera nuclear (NSSS) para eíectuar la gestióndel combustible de un reactor del tipo BWR"MINGUEZ, L ; ESTEBAN, A . ; 60HEZ, M . ; LEIRA, G . ; MARTINE7, R . ; SERRANO, J . ( 1 9 7 5 )20 p p .

Se relaciona un conjunto de parámetros nucleares, brmohidráulicos y mecánicos,

actuali/aiioo según los diseños de reactores BWR,

Estos paránrlros son necesarios para ^Tectuar la gestión y diseño del combus-t in l r y riebon ser suministrados por el labricanle del Reactor a la Empresa

1'clr ica propietaria del mismo.

J . E . N , 2 9 7

Junta de Energía Nuclear, División di1 Tecnología de Reactoras, Madrid."Información que debe aportar el suminislradoi* dela caldera nuclear (NSSS) para efectuar la gestióndel combustible de un reactor del tipo BWR"MINGULZ, l \ ; ESTEBAN, A.; GOMI/, H.; LEIRA, G.; fWtflNL/, R.; SFRRANO, I. (1975)¿0 pn.

S1 relaciona un conj mío do parámetros nucleares, tormohidráulicos v mecínicos,aciualirados según los diseños de reactores EMR.

rslos parámetros son necesarios para efectuar la gestión y diseño di'l combiib-l i b l e , ) deben si r suininisirados por el fabricante df'l Reactor a la ImnrvsaEléctrica propietaria del mismo.

J . E . N . 297

ínula d' Energía Nuclear, División de Tecnología de Rractores, Madrid"In tormacion que debe apo r t a r el s u m i n i s t r a d o r dela ca lde ra nuc lea r (NSSS) para efectuar Ja gest iónclol combust ib le ele un r e a c t o r del tipo BWR"MINHIJEV, I . ; ESIFBAN, A . ; GÓMEZ, M. ; LEIRA, G . ; M A R 1 I É / , R.; SERRANO, J . (1975)/O |,|..

Sr- n larinna un conj mío de parámetros nuJpqr 's, tu raoh i drául i eos y mecánicos,aHuali adus según los diseños do reaolorrs BWR.

tslos uarámelros son neresarios para " fect inr la gestión y diseño d>J coinbus—l ib io , y cioben :^r suministrados por el fahricaite del Reactor a la Lmnr'sa'1 ' c l r i ca ' iropielaria del mismo.

Clasificación INIS y Descriptora. EJ2.- Bffi Type Roactorc; ; RIHI Managenrn1; Clasificación INIS y Descriptores. [32.- RWR lype Reactors : Tuel Managjntonl;

fnul riemenis; Souci fica+íons; Ruaclor Cores; Diel Assembl les; Reactor Cooling > f'uel Elrmonts; Speci f ications; Reactor Cores; fue! Assembl ios; Reactor Cooling

Systems; Reactor OperaI ion; Mu H i-Parame ler Analysis; Dala; R°ac I or Safety; SVSLPIBS; Reactor Opera t ion; Mu H i-Parame te r Analysis; Dala," Reactor Safety;

Reactor Accidents., Reactor Accidents.

Clarif icación INIS y Descriptores. Ló?.- BWR Type Reactors ; fue! Management; Clasificación INIS y Descriptores. D i . - BWR lype Reaclors; fue! Management;

H.i'l ricinenls; SOL r i f ications; Reactor Cores; fue I Assembl i os; Reactor Cooling Hiél flemenls; Speci f ical ions; Reactor Cores; luel Assemblies; Reac lo r Cooling

Svstonis; Reactor Oporalion; Muí I i-Parameler Analysis; Data; Reactor Safoty; f Syslencí; Reaclor Operation; Muí li-Parameter Analysis; Dala; Reactor Safely;

Reactor An i denU. i Reai lor Accidents.

J.E.N. 297 J.E.N. 297

Jurrta de Energía Nuclear. División.de Tecnología do Reactores, Madrid."Information to be r e q u e s t e d í r o m the NSSS vendorfox- fuel managemen t capabi l i ty for B W R ' s " .MINGUE7, [ . ; ESírBAN, A . ; G0ME7, H . ; LEIRA, G . ; MARÍINFZ, R; SERRANO, J . (1975)20 pp.

A sol of Ihp nuclear, therraaI-hydranlie, and mechanical parameIcrs neccesary

according to Ihe design oí BWR's, is l i s ted .

This parame* lors are noccesary Lo porform thp fue! elemenls management and

design, and i t musfc be suppiiod by the Reactor Hanufactiiror to the U t i l i t y .

Jiinta 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 fue] managemen t capabi l i ty for BWR's1 1 .MINGIJEZ, E.; ESTEBAN, A.; GÓMEZ, M.; LEIRA, G.; MARTÍNEZ, R.; SERRANO, J . (1975)

20 pp.

A sel of Iho nuclear, thermal-hydraulic, and mochanical parameters necessary

according to the design of BWR's, is l i s ted .

This paramelers are neccesary to perform the fuel elements managomont and

dosign, and i t must be supplied by the Reactor Manufaclurer to the U t i l i t y .

J.E.N. 297

Junta de Energía Nuclear. División de Tecnología de Reactores, Madrid."Information to be r e q u e s t e d from the NSSS vendorfor fue! managemen t capabi]iLy for B W R ' s ' .MINGUEZ, E . ; ESIEBAN, A . ; GÓMEZ, M. ; LEIRA, C ; MARIINEZ, R . ; SERRANO, J . ( 1 9 7 5 )2 0 pp.,

A set of Lhc nuclear, thermal-hydraulic, and raechanical paramelers necessaryaccording lo the design of BWR's, is listed»

This paramelurs are neccesary lo perform the fuel elements management anddtisign, and i b musí be suppliod by the Reactor Manufacturar to tho U t i l i l y .

J.E.N. 297

Junta de Fnergía Nuclear, División de Tecnología de Reactores, Madrid"Informat ion to be r eques t ed from the NSSS vendoi*for fuel managemen t capabi l i ty for B W R ' s " .MIMGUEZ, E . ; ESTEBAN, A . ; GÓMEZ, M . ; LEIRA, G . ; MARTÍNEZ, R . ; SERRANO, J . ( | 9 7 5 )2 0 p p .

A set of the nuclear, thermal-hydraulic, and mechanical paramsters necessaryacrording to the design of BWR's, is Usted.

This pararaeters aro neccesary to perform lhc fuel el ornenIs management anddesign, and 'ú' must be supplied by Ihe Reactor Manufacturer to Iho U t i l i t y .

Clasificación INIS y Descriptores. £Ó?.,- BWR Type Rcaclors; Fucl Managompnt;

funi Clornen te; Soccifications; Reactor Coros; Riel Assomblies; Reactor Cooling

Svslpffls; Reactor Operation; Mulli-ParaiPtur Analysis; Dala; Reactor Sarel-y;

Reactor Accidenta.

Clasificación INIS y Descriptores. £31.- BWR Typc Reactors; Riel Management;

Fucii ü croen Is; Spoci f ications; Reacior Cores; fue! Assemblies; Reactor Cooling

Systems; Reactor Operalion; Muíti-Parameter Analysis; Daia; Reactor Safety;

Reactor Accidente.

Clasificación INIS y Descriptores. Cal . - BWR fype Reactors; Riel Management;Hti-l f ] i ' im jnl ' ; Specificalions; Reactor' Cores; Riel Assemblics; Reaclor CoolingSysloiTi';; Reactor Opuraliun; Huí Li-Pararao'i'r Analysis; Dala; Reactor Safe ty;Reactor Accidents.

Clasificación INIS y Descriptores. C32.- BWR lype Reaclors; Fue! Management;

RIPI floments; Sppcifications; Reactor Cores; Fuel Assemblios; Reactor Cooling

SystoniG; Reactor Operalion; Mullí-Paramo lor Analysis; Data; Reactor Safety;

Reactor Accidento.