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 ' \ ;
Rnaí lor A c u d í i t h . R at lo i Accid ni .
Clasi í icaLion
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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 ' ;
R>ai I oí Aci idcnl -