1. RPV Surveillance Programmes
Antonio Ballesteros
Tecnatom S.A.

2. RPV

3. Embrittlement / ageing

4. CHANGES IN MECHANICAL PROPERTIES DUE TO NEUTRON IRRADIATION
PROPERTIES VALUE
FLUENCE 1019 n/cm2

5. Neutron embrittlement
neutron fluence
impurities (late 60s: Cu & later P - L.E. Steele and co-workers at ONRL)
alloying elements
Ni (> ~1 wt%) identified ~10 y later
Today, indications on Mn
- Prediction models & formulas
- Surveillance programmes

6. PTS PTS transient must occur on RPV beltline Flaw exists Critical size
Near inner surface of RPV wall at limiting location
Beltline region suffers embrittlement
Neutron Irradiation
Results in reduced fracture resistence

7. Concentration of Cu, Ni and P in the steel. Spanish reactors

8. Surveillance Capsule

9. Spanish reactors in operation
Nuclear
power
plant
Type of
reactor
Supplier
Initiation
operation
Beltline
limiting
material for brittle
fracture
Surveillance
capsules
analysed at 7
/200 6
José Cabrera
PWR
Westinghouse
1968 W
eld 4
Almaraz I
1981
Base metal
Almaraz II
1983
Ascó I 4
Ascó II
PWR
Westinghouse
1985
Base metal
Vandellós II
1988
Weld 3
Trillo I
Siemens -
KWU 1
988
Santa Mª de
Garoña
BWR
General
Electric
1971 2
Cofrentes
1984

10. LOCATION OF SURVEILLANCE CAPSULES IN ALMARAZ, ASCÓ AND VANDELLÓS REACTORS

11. PWR Surveillance Capsule

12. Capsule Location in BWR-6

13. DOSIMETER AND SPECIMEN CAPSULES. BWR-6

14. CAPSULES LOCATION IN TRILLO I REACTOR

15. SURVEILLANCE CAPSULES IN TRILLO I REACTOR

16. CHARPY IMPACT TEST
E23 – Methods for Notched Bar Impact Testing of Metallic Materials
Data obtained: E
USE
30 ft-lb temperature
50 ft-lb temperature
35 Mil lat. expansion
Upper-Shelf Energy
Transition Region
LSE T

17. EFFECTS OF NEUTRON RADIATION ON CHARPY IMPACT PROPERTIES
DRTNDT
DUSE
Temperature
Charpy Impact Energy

18. FRACTURE TEST (LINEAR ELASTIC/PLAIN STRAIN)K
KIC
T-RTNDT
Data obtained: Fracture Toughness KIC

20. TENSION TEST Data obtained: Ultimate Strengh Yield Strengh
Fracture Strengh
Uniform Elongation
Toltal Elongation
Reduction of Area
Ultimate
Fracture
Load
Yield
Strain
Uniform
Total

21. EFFECTS OF NEUTRON RADIATION ON TENSION TEST PROPERTIES
IRRADIATED L
UNIRRADIATED E

22. Lack of non-hardening embrittlement in the Spanish PWR PV steels

23. Methodology RT SURVEILLANCE CAPSULE NDT TREND CURVES
LIMITS VERIFICATION
R.G. 1.99
10CFR50 20 40 60 80
100
120
140
160
180
200
Temperatura Indicada (ºC)
OPERACIÓN NO
PERMITIDA
OPERACIÓN
CALCULADAS CON LA RT
NDT
A 32 EFPY

24. Operating Window

25. Codes, Guides and Standards (1/2)
10CFR50, Appendix G, Fracture Toughness
10CFR50, Appendix H, Surveillance Programs
10CFR50.61, PTS
10CFR50.66, Vessel Annealing
ASME XI, Appendix G, P-T Curves Calculation
NUREG-0800, Section 5.3.2: PT limit curves
Branch Technical Position MTEB 5-2
ASTM-E 185, Surveillance Programs
Regulatory Guide 1.99 Rev 2, RTNDT and USE

26. Codes, Guides and Standards (2/2)
Regulatory Guide 1.154, Detailed analysis of PTS
Regulatory Guide 1.190, Neutron Calculations
ASME Code Case N-514
ASME Code Case N-588
ASME Code Case N-640
ASME Code Case N-641
Generic Letter 96-03: Relocation of PT limit curves
KTA
KTA 3203 revision 06/2001

27. Listing of ASTM Standards applicable to Licensing Requirements of a LWR Pressure Vessel (Pre-operation phase)
E8 Tension Testing
E21 Elevated Temperature Tension Test
E23 Charpy Impact Testing
E185 Surveillance Program Requirements
E208* Drop-Weight/Nil-Ductility Testing
E399 Plane-Strain Fracture Testing
E636 Conducting Supplemental Surveillance Tests
E844 Sensor Set Design
E900 Predicting Neutron Radiation Damage
E1214 Melt Wire Temperature Monitors
E1921 Master Curve Testing
A370 Mechanical Testing of Steel Products
A533* Pressure Vessel Plates, Alloy Steel
A508* Alloy Steel Forgings for Pressure Vessels
*Standars not used after all fabrication, construction and design analysis is completed

28. Listing of ASTM Standards applicable to support continued Licensing Requirements of a LWR Pressure Vessel
E8 Tension Testing
E21 Elevated Temperature Tension Test
E23 Charpy Impact Testing
E185 Surveillance Program Requirements
E399 Plane-Strain Fracture Testing
E509 In-service Annealing Guide
E813 JIC (Measure of Fracture Toughness)
E900 Predicting Neutron Radiation Damage
E992 Equivalent Energy Fracture Toughness
E1152 Determining J-R Curves
E1214 Melt Wire Temperature Monitors
E1921 Master Curve Testing
Plus:
The standards related to the analysis and determination of the surveillance capsule of reactor vessel fluence, etc.
The reference standards which provide supporting data and information that make possible the interpretation of the irradiated materials data

29. DESIGN OF RV SURVEILLANCE PROGRAM
Regulation
10CFR50, Appendix H and by reference ASTM E185
ASTM E185 - Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels
Main requirements defined
Material Selection
Capsule Number
Dosimeter Selection
Thermal Monitors
Specimen Selection
Capsule Location
Capsule schedule
Primary Supporting ASTM Standards
E844
E854
E482
E900
E560
E583
E1005
E1214

30. DESIGN OF RV SURVEILLANCE PROGRAM
Material Selection
A. Actual materials used to fabricate vessels
B. Minimum: one base metal, one weld metal
1. Most limiting per Guide E900
2. Low charpy upper-shelf energy
C. Correlation monitor or reference steel

31. DESIGN OF RV SURVEILLANCE PROGRAM
B. Specimen Orientation
Location of Test Specimens Within Weld Test Material

32. DESIGN OF RV SURVEILLANCE PROGRAM
Neutron Dosimeter Selection
Selected according to Guide E482 and E844
Thermal Monitor Selection
Selected according to Guide E1214
Specimen Selection
A. Type and minimum number
Material
Charpy
Tension
Base metal 12 3
Weld Metal
Correlation
B. Specimen orientation

33. DESIGN OF RV SURVEILLANCE PROGRAM
Number of Capsules
Table 1
(A) Or at the time when the acccumulated neutron fluence of the capsule exceeds 5x 10E-22 n/m 2(5 x 10E-18 n/cm2 ), or at the time when the highest predicted ΔRTNDT of all encapsulated materials is aproximately 28ºC (50 ºF), whichever comes first.
(B) Or at the time when the acccumulated neutron fluence of the capsule corresponds to the approximate EOL fluence at the reactor vessel inner wall location, whichever comes first.
(C) Or at the time when the acccumulated neutron fluence of the capsule corresponds to the approximate EOL fluence at the reactor vessel ¼ T location, whichever comes first.
(D) Or at the time when the acccumulated neutron fluence of the capsule corresponds to a value midway between than of the first and third capsules.
(E) Not less than once or greater than twice the peak EOL vessel fluence. This may be modified on the basis of previous tests. This capsule maybe held without testing following withdrawall

34. DESIGN OF RV SURVEILLANCE PROGRAM
Capsule Location (Wall/Accelerated)
Vessel Wall Location-Lead Factor < 5
Accelerated Capsules-Optional. Lead Factor greater than wall capsules
Capsule Withdrawall schedule

35. Appendix G to 10CFR50
RTNDT is evaluated according to the procedures in the ASME Code, Paragraph NB-2331.
For the reactor vessel beltline materials, including welds, plates and forgings, the values of RTNDT and Charpy upper-shelf energy must account for the effects of neutron radiation, including the results of the surveillance program (Appendix H).
The effects of neutron radiation must consider the radiation conditions (i.e., the fluence) at the deepest point on the crack front of the flaw assumed in the analysis.

36. Charpy Upper-Shelf Energy Requirements
Reactor vessel beltline materials must have Charpy upper-shelf energy, in the transverse direction for base material and along the weld for weld material according to the ASME Code, of no less than 75 ft-lb (102 J) initially and must maintain Charpy upper-shelf energy throughout the life of the vessel of no less than 50 ft-lb (68 J).
DRTNDT
DUSE
Temperature
Charpy Impact Energy

37. RCS Component Engineering Project Manager Technical Staff
OVERVIEW OF INTERACTIONS AND IMPACT OF PREPARATION OF A SURVEILLANCE CAPSULE REPORT ON THE OPERATION LICENSING OF A NUCLEAR POWER PLANT
RCS Component Engineering
Establish Design P-T Limit Curve for Reactor Vessel
Licensing
New Tech Specif
P-T Operation
Curves for Reactor Vessel
Project Manager Technical Staff
Prepare Report on Capsule Evaluation
Laboratory
Capsule Evaluation
Customer
CSN
Fluence Analysis
Calculate Dosimeter Activity and compare with Measured Activity
Calculate Fluence
Plant Engineering
Adjust P-T Limit Curve for Reactor Vessel
Fuels Engr
Power Distribution
Nuclear Services
Power History

38. INCREASE OF REFERENCE TEMPERATURE

39. FORMULES FOR RTNDT CALCULATION
NAME
CHEMICAL FACTOR
FLUENCE FACTOR
APPLICATION
COMMENTS
ORIGIN
PWR-ES
Base Material: 14.7 and 52,6 for 0,1
Weldings: and for 0.1
(f/1E+19) ( log (f/1E+19) )
Almaraz I and II, Ascó I and II, Vandellós, Trillo, José Cabrera
Only for spanish PWR
SPAIN
PWR-ES/BC
( %Cu)
Cu<0.1%
Generic ( Cu)
R.G rev 1
5/9( (%P-0.008)+
+1000(%Cu-0.08)
(f/1E+19) 0.5
EEUU
R.G rev 2
Weld and Base Material Tables [ f(%Cu,%Ni) ]
(f/1E+19) ( log (f/1E+19) )
MIANNAY
[ %P (%Cu-0.08)+12.1
(%Ni-0.7)+48.31(%Cu-0.08)(%Ni-0.7)]
(f/1E+19) 0.7
FRANCE
RSEM Code (FIM)
[ (%P-0.008)+238(%Cu-0.08)+
191(%Cu)(%Ni)2
(f/1E+19) 0.35
10CFR50
0.55[48+( %Cu+350(%Cu)(%Ni)
(f/1E+19) 0.27
HOLLAND
JEPE
( %P+215%Cu+77(%Ni%Cu)0.5)
(26-24%Si-61%Ni+301(%Ni%Cu)0.5
(f/1E+19) ( log (f/1E+19) )
(f/1E+19) ( log (f/1E+19) )
Base Material
Weldings
JAPAN
RSEM Code (FIS)
8+[ (%P-0.008)+238(
%Cu-0.08) +191(%Cu)(%Ni)2]
Fluences between
2.1018– n/cm2
Provides an upper limit to RTNDT
RCC-M
[22+556(%Cu-0.08)+2778(%P-0.008)]
1.1018– n/cm2
KTA
Cu %
Fluences > n/cm2
Provides an UL
to RTNDT (graph)
GERMANY
PNAE
BaseMaterial:[20+230(%Cu+10%P)]
Welds:[ %Cu+%P)]
(f/1E+19) 1/3
Base Material: WWER-1000 (Tirr=290ºC)
Welds: WWER-440 (Tirr=270ºC)
RUSSIA

40. Calculation of RTNDT according to R.G. 1.99 Rev 2
Non-Irradiated Material Reference Temperature

41. Cálculo de RTNDT por Regulatory Guide 1.99, revisión 2

42. FLUENCE (n/cm2) (E>1 Mev)
UPPER SHELF ENERGY DECREMENT: APPLICATION TO THE RESULTS OF THE SURVEILLANCE PROGRAM
USE DECREMENT (%)
FLUENCE (n/cm2) (E>1 Mev)

43. KTA 3202, issued in December 2001

44. Irradiation Temperature monitoring
melting alloys

45. Advanced model (Eason, 2003) for Transition Temperature Shift

46. Values of T41J in Spanish reactors

47. RTNDT. RG 1.99 (CF experimental) versus Eason model

48. RTNDT. RG 1.99 (CF from tables) versus Eason model

49. USE Behaviour. RG 1.99 versus Eason Model

50. ORIENTATION OF SURVEILLANCE SPECIMENS Charpy Reconstitution
FORGING
L: rolling direction
T:long transverse
S: thickness direction (long transverse) S
PLATE
L:circunferencial direction of forging , main forginig direction
T:axial direction of forging
S: thickness direction L S T
Charpy Reconstitution

51. Electron Beam Reconstitution

52. Standard Formula for Attenuation within the RPV wall
Due to these changes in neutron spectrum, the use of neutron fluence may give a non-conservative estimate of the neutron damage attenuation within the vessel wall. If dpa is calculated, it can be used to obtain the effective vessel wall fluence for use in embrittlement trend curves (ASTM E900):
()x = ()IS [dpax/dpaIS]
Alternatively, the following exponential attenuation formula may be used (according to Regulatory Guide 1.99 rev. 2)
()x = ()IS exp(-0.24x)

53. GE formula for BWRs n = distance (Cm) between points 1 and 2
m = distance (Cm) between point 1 and the core equivalent radius

54. Comparison for different type of reactors

55. PWR comparison. Exact versus conservative calculations

56. BWR comparison. Exact versus conservative calculations

57. Gamma-Ray Interaction Effects
Gamma rays cause damage by three mechanism:
Ionization
Heating
Energized electron

58. Ionization Effects
The principal ionization effect is water radiolysis, which in turns affects stress-assisted corrosion cracking.
Neutron can also contribute to radiolysis through atomic displacement reactions (via neutron scattering or capture).
Photons should dominate the breaking of atomic and molecular bonds by means of electromagnetic interactions.
The test specimens in the surveillance capsules will not experience the water chemistry affects that the vessel wall will.

59. Effects of Heating
Photons induce heating, leading to change in alloy morphology (grain structure) and chemical diffusion.
Gamma heating could be deduced by adding thermoluminescent dosimeters (TLDs) in the surveillance capsules.

60. Atomic Displacement by Energized Electrons
Photons interact by electromagnetic interaction and are generally incapable of directly displacing atoms.
However, photons can set free energetic electrons which can transfer the eV needed to displace atoms:
Compton scattering
Photoelectric effect
Pair production

61. Compton scattering

62. Photoelectric Effect
Below energies of about 0.1 MeV the dominant mode of gamma interaction in medium and high Z material is the photoelectric process

63. Pair production
By interaction in the vicinity of the coulomb force of the nucleus, the energy of the incident photon is spontaneously converted into the mass of an electron-positron pair.

64. Gamma-induced dpa cross-section of iron for the three interactions

65. Neutron and Gamma DPA Values

66. Tecnatom neutron calculations methodology
EXPERIMENTAL
(DOSIMETRY)
CALCULATIONS
THEORETICAL
CALCULATIONS
ENDF - VI
STANDARD
TEMPERTURE
NJOY
NEUTRON
SPECTRUM
ENDF – VI
OPERATION
TEMPERATURE -
ACTIVITY OF
THE DOSIMETERS
MCNP 4C
FLATO
POWER
DISTRIBUTION
IRDF-90 REV.2
IRRADIATION
HISTORY
REACTOR
GEOMETRY
CORRECTION OF THE THEORETICAL CALCULATIONS WITH THE EXPERIMENTAL VALUES IN THE CAPSULE
DEFINITIVE NEUTRON FLUX

67. Reactor Geometry

68. BWRs (Supplier GE)
The capsules are place adjacent to the vessel internal wall. The lead factor is frequently < 1.
The dosimeters used employ two materials, iron (for irradiation periods that do not exceed two years) and copper. The nuclear reactions involved are:
Fe54(n, p)Mn54
Cu63(n, alpha)Co60
The BWR/6 plant is supplied with three irons wires in a separate vessel dosimeter which is placed in a holder adjacent to the specimen holders.

69. Dosimeters used in the Spanish PWR SP

70. Location of the wire dosimeters

71. 238U fission monitor capsule

72. 237Np fission monitor capsule

73. Nuclear parameters MONITOR MATERIAL REACTION OF INTEREST RESPONSE
MONITOR MATERIAL
REACTION OF
INTEREST
RESPONSE
RANGE
(MeV) ([1])
PRODUCT
HALF-LIFE
Copper
Cu‑63(n,)Co‑60
4.7 – 11.1
5.271 years
Iron
Fe-54(n,p)Mn‑54
312.5 days
Nickel
Ni‑58(n,p)Co‑58
70.78 days
Uranium‑238 ([2])
U‑238(n,f)Cs‑137
30.03 years
Neptunium‑237(2)
Np‑237(n,f)Cs‑137
Cobalt‑aluminium
Co‑59(n,g)Co‑60
([3])
([1]) Ref. ASTM E 261‑90, Table 4.
([2]) dosimeter shielded with Cd.
([3]) Without threshold energy (to measure thermal flux).