Prediction of irradiation embrittlement in WWER-440 Reactor Pressure Vessel materials

Prediction of irradiation embrittlement of Reactor Pressure Vessel (RPV) materials is performed usually in accordance with relevant codes and standards that are based on large amount of information from surveillance and research programs. The existing Russian Code (Standard for Strength Calculations...

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Datum:	2007
Hauptverfasser:	Brumovsky, M., Kytka, M., Debarberis, L., Gillemot, F., Kryukov, A., Valo, M.
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Sprache:	English
Veröffentlicht:	Ядерний науково-дослідницький інститут Чехії 2007
Schriftenreihe:	Вопросы атомной науки и техники
Schlagworte:	Материалы реакторов на тепловых нейтронах
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Zitieren:	Prediction of irradiation embrittlement in WWER-440 Reactor Pressure Vessel materials / M. Brumovsky, M. Kytka, L. Debarberis, F. Gillemot, A. Kryukov, M. Valo // Вопросы атомной науки и техники. — 2007. — № 6. — С. 72-77. — Бібліогр.: 4 назв. — англ.

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spelling	irk-123456789-1105972017-01-06T03:02:32Z Prediction of irradiation embrittlement in WWER-440 Reactor Pressure Vessel materials Brumovsky, M. Kytka, M. Debarberis, L. Gillemot, F. Kryukov, A. Valo, M. Материалы реакторов на тепловых нейтронах Prediction of irradiation embrittlement of Reactor Pressure Vessel (RPV) materials is performed usually in accordance with relevant codes and standards that are based on large amount of information from surveillance and research programs. The existing Russian Code (Standard for Strength Calculations of Components and Piping in Nuclear Power Plants (NPPs) – PNAE G 7-002-86) for the WWER RPV irradiation embrittlement assessment was approved more than 20 years ago and based mostly on the experimental data obtained in research reactors with accelerated irradiation. The validation of the above Code has been made without the surveillance specimen results that were produced in 1980-1990s. Thus, new analysis of all available data was required for more precise prediction of radiation embrittlement of RPV materials. Based on the fact that large amount of data from surveillance program as well as some research programs, IAEA International Database on RPV Materials (IDRPVM) has been used for the detailed analysis radiation embrittlement of WWER RPV materials. Thus, the following activities have been performed within the IAEA Co-ordinated project: Collection of complete WWER-440 surveillance and other similarly important data into the IDRPVM, Analysis of radiation embrittlement data of WWER-440 RPV materials using IDRPVM database, Evaluation of predictive formulae depending on material chemical composition, neutron fluence and neutron flux, Development of the guidelines for prediction of radiation embrittlement of operating reactor pressure vessels of WWER-440 including methodology for evaluation of surveillance data of a specific operating unit. Прогнозування радіаційного окрихчення матеріалів внутрішньо корпусних пристроїв (ВКП) зазвичай виконується у відповідності з кодами та стандартами, заснованими на численній інформації, що була накопичена на базі модельних та дослідницьких програм. Існуючий Російський Код (Стандарт для обчислення міцності компонентів і трубопроводів в атомних електростанціях (АЕС)-PNAE G 7 –002-86) для оцінки радіаційного окрихчення ВКП реакторів ВВЕР зарекомендував себе на протязі більш, ніж 20 років; він заснован на експриментальних даних, отриманих в дослідницьких реакторах з прискореним опроміненням. Оцінка вище згаданого Кода була виконана без результатів із зразків-свідків, які були отримані у 1980-1990 роках. Таким чином, необхідно провести новий аналіз усіх наявних даних для більш точного прогнозування радіаційного окрихчення матеріалів ВКП. На підставі того факту, що було використано велику кількість даних з макетних та дослідницьких програм, Міжнародна база даних МАГАТЕ по матеріалам ВКП біла використана для докладного аналізу радіаційного окрихчення ВКП матеріалів для реакторів ВВЕР. Таким чином, в межах Координаційного проекту МАГАТЕ було виконано наступне: Збирання повних даних із зразків-свідків ВВЕР-440 та інших подібних важливих даних в Міжнародну базу даних, аналіз даних по радіаційному окрихченню ВКП матеріалів ВВЕР-440 з використанням міжднародної бази даних, оцінка формули прогнозування в залежності від хімічного складу матеріалу, флюенса нейтронів та нейтронного потоку, розробка основних положень для прогнозування радіаційного окрихчення експлуатуємих внутрішньо корпусних пристроїв ВВЕР-440, включно з методологією для оцінки контрольних даних конкретної діючої установки. Предсказание радиационного охрупчивания материалов внутрикорпустных устройств (ВКУ) обычно выполняется в соотвествии с кодами и стандартами, основанными на обширной информации, накопленной в ходе модельных и исследовательских программ. Существующий Российский Код (Стандарт для вычислений прочности и компонентов и трубопроводов в атомных электростанциях (АЭС) – PNAE G 7-002-86) для оценки радиационного охрупчивания ВКУ реакторов ВВЭР хорошо зарекомендовал себя на протяжении более чем 20 лет; он основан на экспериментальных данных, полученных в исследовательских реакторах с ускоренным облучением. Оценка упомянутого выше Кода была выполнена без результатов с образцов-свидетелей, которые были получены в 1980-1990 годах. Таким образом, необходим новый анализ всех имеющихся данных для более точного прогнозирования радиационного охрупчивания материалов ВКУ. На основании того факта, что было использовано большое количество данных с макетных и исследовательских программ, Международная База данных МАГАТЭ по материалам ВКУ была использована для подробного анализа радиационного охрупчивания ВКУ материалов для реакторов ВВЭР. Таким образом, в рамках Координационного проекта МАГАТЭ были выполнено следующее: сбор полных данных с образцов-свидетелей ВВЕР-440 и других подобных важных данных в Международную базу данных; анализ данных по радиационному охрупчиванию ВКУ материалов ВВЭР-440 с использованием международной базы данных; оценка формулы прогнозирования в зависимости от химического состава материала, флюенса нейтронов и нейтронного потока, разработка основных положений для предсказания радиационного охрупчивания эксплуатируемых внутрикорпусных устройств ВВЭР-440, включая методологию для оценки контрольных данных конкретной действующей установки. 2007 Article Prediction of irradiation embrittlement in WWER-440 Reactor Pressure Vessel materials / M. Brumovsky, M. Kytka, L. Debarberis, F. Gillemot, A. Kryukov, M. Valo // Вопросы атомной науки и техники. — 2007. — № 6. — С. 72-77. — Бібліогр.: 4 назв. — англ. 1562-6016 http://dspace.nbuv.gov.ua/handle/123456789/110597 621.039.577 en Вопросы атомной науки и техники Ядерний науково-дослідницький інститут Чехії
institution	Digital Library of Periodicals of National Academy of Sciences of Ukraine
collection	DSpace DC
language	English
topic	Материалы реакторов на тепловых нейтронах Материалы реакторов на тепловых нейтронах
spellingShingle	Материалы реакторов на тепловых нейтронах Материалы реакторов на тепловых нейтронах Brumovsky, M. Kytka, M. Debarberis, L. Gillemot, F. Kryukov, A. Valo, M. Prediction of irradiation embrittlement in WWER-440 Reactor Pressure Vessel materials Вопросы атомной науки и техники
description	Prediction of irradiation embrittlement of Reactor Pressure Vessel (RPV) materials is performed usually in accordance with relevant codes and standards that are based on large amount of information from surveillance and research programs. The existing Russian Code (Standard for Strength Calculations of Components and Piping in Nuclear Power Plants (NPPs) – PNAE G 7-002-86) for the WWER RPV irradiation embrittlement assessment was approved more than 20 years ago and based mostly on the experimental data obtained in research reactors with accelerated irradiation. The validation of the above Code has been made without the surveillance specimen results that were produced in 1980-1990s. Thus, new analysis of all available data was required for more precise prediction of radiation embrittlement of RPV materials. Based on the fact that large amount of data from surveillance program as well as some research programs, IAEA International Database on RPV Materials (IDRPVM) has been used for the detailed analysis radiation embrittlement of WWER RPV materials. Thus, the following activities have been performed within the IAEA Co-ordinated project: Collection of complete WWER-440 surveillance and other similarly important data into the IDRPVM, Analysis of radiation embrittlement data of WWER-440 RPV materials using IDRPVM database, Evaluation of predictive formulae depending on material chemical composition, neutron fluence and neutron flux, Development of the guidelines for prediction of radiation embrittlement of operating reactor pressure vessels of WWER-440 including methodology for evaluation of surveillance data of a specific operating unit.
format	Article
author	Brumovsky, M. Kytka, M. Debarberis, L. Gillemot, F. Kryukov, A. Valo, M.
author_facet	Brumovsky, M. Kytka, M. Debarberis, L. Gillemot, F. Kryukov, A. Valo, M.
author_sort	Brumovsky, M.
title	Prediction of irradiation embrittlement in WWER-440 Reactor Pressure Vessel materials
title_short	Prediction of irradiation embrittlement in WWER-440 Reactor Pressure Vessel materials
title_full	Prediction of irradiation embrittlement in WWER-440 Reactor Pressure Vessel materials
title_fullStr	Prediction of irradiation embrittlement in WWER-440 Reactor Pressure Vessel materials
title_full_unstemmed	Prediction of irradiation embrittlement in WWER-440 Reactor Pressure Vessel materials
title_sort	prediction of irradiation embrittlement in wwer-440 reactor pressure vessel materials
publisher	Ядерний науково-дослідницький інститут Чехії
publishDate	2007
topic_facet	Материалы реакторов на тепловых нейтронах
url	http://dspace.nbuv.gov.ua/handle/123456789/110597
citation_txt	Prediction of irradiation embrittlement in WWER-440 Reactor Pressure Vessel materials / M. Brumovsky, M. Kytka, L. Debarberis, F. Gillemot, A. Kryukov, M. Valo // Вопросы атомной науки и техники. — 2007. — № 6. — С. 72-77. — Бібліогр.: 4 назв. — англ.
series	Вопросы атомной науки и техники
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fulltext	УДК 621.039.577 PREDICTION OF IRRADIATION EMBRITTLEMENT IN WWER-440 REACTOR PRESSURE VESSEL MATERIALS M. Brumovsky1, M. Kytka1, L. Debarberis2, F. Gillemot3, A. Kryukov4, M. Valo5 1 Nuclear Research Institute Rez plc, Czech Republic; 2 EC JRC Institute of Energy, Petten, Netherland; 3 Atomic Energy Research Institute, Budapest, Hungary; 4 RRC Kurchatow Institute, Moscos, Russian Federation; 5 VTT Manufacturing Technology, Espoo, Finland Prediction of irradiation embrittlement of Reactor Pressure Vessel (RPV) materials is performed usually in accordance with relevant codes and standards that are based on large amount of information from surveillance and research programs. The existing Russian Code (Standard for Strength Calculations of Components and Piping in Nuclear Power Plants (NPPs) – PNAE G 7-002-86) for the WWER RPV irradiation embrittlement assessment was approved more than 20 years ago and based mostly on the experimental data obtained in research reactors with accelerated irradiation. The validation of the above Code has been made without the surveillance specimen results that were produced in 1980-1990s. Thus, new analysis of all available data was required for more precise prediction of radiation embrittlement of RPV materials. Based on the fact that large amount of data from surveillance program as well as some research programs, IAEA International Database on RPV Materials (IDRPVM) has been used for the detailed analysis radiation embrittlement of WWER RPV materials. Thus, the following activities have been performed within the IAEA Co-ordinated project: Collection of com- plete WWER-440 surveillance and other similarly important data into the IDRPVM, Analysis of radiation embrittlement data of WWER-440 RPV materials using IDRPVM database, Evaluation of predictive formulae depending on material chemical composition, neutron fluence and neutron flux, Development of the guidelines for prediction of radiation embrit- tlement of operating reactor pressure vessels of WWER-440 including methodology for evaluation of surveillance data of a specific operating unit. INTRODUCTION Prediction of irradiation embrittlement of Reactor Pres- sure Vessel (RPV) materials is performed usually in accor- dance with relevant codes and standards that are based on large amount of information from surveillance and re- search programs. The existing Russian Code [1] for the WWER RPV irradiation embrittlement assessment was ap- proved more than 20 years ago and based mostly on the experimental data obtained in research reactors with accel- erated irradiation. Nevertheless, it is still in good use and generally consistent with new data. The validation of the above Code has been made without the surveillance speci- men results that were produced in 1980-1990s. Up to now a lot of surveillance data have been ob- tained from operating WWER type reactors. Also some years ago the IAEA International Database on RPV Ma- terials (IDRPVM) has been created. A big amount of RPV irradiation embrittlement data have been obtained in the framework of national research programmes as well. Using these data with the surveillance results could considerably expand the WWER RPV material database and assist in the RPV integrity assessment. Reactor pressure vessels for WWER-440 type reactors were manufacture from the steel of 15H2MFA(A) grade which is of Cr-Mo-V type. Such steels are not used in any other RPVs in other type or reactors, thus there does not exist any other prediction formulae that could be applied for WWER-440 materials – their chemical composition is quite different when compared with usual LWR RPV ma- terials – ASTM A 533-B or A 508 that are of Ni-Mo-Cr type without any amount of vanadium. Thus, new analysis of all available data were re- quired for more precise prediction of radiation embrit- tlement of RPV materials. This analysis was performed within the IAEA Cvo-ordinated Research Program (CRP) with these main tasks: – Collection of complete WWER-440 surveillance and other similarly important data into the IDRPVM; – Analysis of radiation embrittlement data of WWER-440 RPV materials using IDRPVM database; – Evaluation of predictive formulae depending on material chemical composition, neutron fluence and neutron flux; – Development of the guidelines for prediction of radi- ation embrittlement of operating reactor pressure vessels of WWER-440 including methodology for evaluation of surveillance data of a specific operating unit. These IAEA Guidelines were finished in 2004 and are now in the printing procedure. 1. CURRENT PROCEDURE [1] The similar procedure is used for the radiation em- brittlement assessment in Russia, Bulgaria and Ukraine and is based on Russian Code [1] for the embrittlement assessment based on the brittle fracture temperature Tk . Its value should be evaluated experimentally by Charpy impact testing entirely. However, if it is not applicable, empirical relations given next have to be used: Weld metals: Tk=T0 + ∆TF; (1) ∆TF=AF 270 (Fx10-22)1/3 for 1022<F<3 x1024 (2) ________________________________________________________________________________ ВОПРОСЫ АТОМНОЙ НАУКИ И ТЕХНИКИ. 2007. № 6. Серия: Физика радиационных повреждений и радиационное материаловедение (91), с. 72-77. 72 AF = 15°C for weld metal Sv-10KhMFT(U) (Tirradiation = 270°C); (3) = 800 (P + 0.07 Cu) for weld metal Sv-10KhMFT (Tirradiation = 270 °C). (4) Here Tk0 is the initial value of brittle fracture temper- ature from Acceptance Tests and ∆TF is irradiation in- duced shift in Tk, AF 270 is the irradiation embrittlement factor for irradiation at temperature 270 °C, F is the neutron fluence in m-2, E>0.5 MeV and P, Cu, is the concentration in mass %. The irradiation embrittlement factor AF is irradiation temperature dependent and its value for irradiation at 250 °C is equal to: AF 250 =800(P+0.07Cu) + 8. (5) Base metals: Formulae (1.1-1.2) are used in general to evaluate the brittle fracture temperature of the WWER-440/230 reactor pressure vessels. In some cases these results are being verified by testing of the material samples, templates, cut directly from the vessel wall. Subsized Charpy specimens are used for this purpose. The special procedure has been developed to correlate results of subsized and standard Charpy testing, and ob- tained data supporting the procedure. Coefficients AF for base metals are given in [1] as: AF = 18 °C for steel 15Kh2MFA (Tirradiation = 270 °C) (6) = 12 °C for steel 15Kh2MFAA (Tirradiation = 270 °C) (7) Nor in [1] neither in other official document there is a valid procedure for evaluation of results from the surveillance specimen programs for WWER reactor pressure vessels. There is also no recommendation for the use of such data in reactor pressure vessel integrity and lifetime assessment. 2. IRRADIATION EMBRITTLEMENT MODELING Relatively large Charpy-V surveillance data set for WWER-440 pressure vessel materials was collected in the CRP but only limited number of fracture toughness data. Hence trend curve fitting in this report is based on ISO Charpy-V data only. The weld data consists of 34 low flux and 87 high flux data points, i.e. of 121 weld data points altogether, and the base metal data of 24 low flux and 76 high flux data points, i.e. of 100 base metal data points altogether. 2.1. PRINCIPLES OF THE TREND CURVE DERIVATION Basic physics research has not resulted in justification of a unique physically based functional forms. The number of candidate functions, which fulfill the requirements given above, is naturally large. In the statistical fitting several combinations of commonly issued functions are applied in order to have a selection for choice. The following three basic functions are used in fitting: The power law function nFaT ∗=∆ ; (8) The exponential function )1( Φ∗−−∗=∆ neaT ; (9) The tanh function )] 2 tanh(2/12/1[1 0 c cT Φ−Φ∗+∗=∆ , (10) where ∆T is the transition temperature shift, F is the neutron fluence and a, n, c1, c2 and F0 are coefficients. The acceptable values of n in the power law function (1) are 0 < n < 1. The power law function never saturates fully and its derivative at zero fluence is infinite for the realistic val- ues of n (0< n < 1). The derivate of the exponential function at zero fluence is "a∗n" and it saturates fully to the value of "a" at high fluences. Hence the exponential function suits well for the characterization of early de- velopment of embrittlement. The embrittlement description may include also threshold values for fluence and impurity contents. These threshold values can be connected to real physical phenomena. Threshold fluence can describe the nucleation fluence, which is needed before a physical response can be observed. Threshold on chemistry may be linked to the amount of the chemical element permanently bound to some structures or to solubility of the element in the matrix at the operation temperature of the material. The data are characterized by the phosphorus and cop- per contents, neutron fluence and neutron flux. The distri- bution of these parameters in the database may limit the in- formation, which can be derived from the data. The mea- sured data are also compared to the prediction function (1.4) given in the Russian Code PNAE [1]. The distribution of weld data in the copper-phospho- rus plane is shown in Fig. 1. Cu [wt-%] 0.00 0.05 0.10 0.15 0.20 0.25 0.30 P [w t- % ] 0.00 0.01 0.02 0.03 0.04 0.05 0.06 High flux Low fluxPVSCU1.JNB Weld All data Fig. 1. Distribution of weld data in the Cu-P plane. Low flux marks are located on the top of high flux marks, i.e. open circles in bold include both flux values Low and high flux data points are identified in the figure. The measured transition temperature shifts of weld data are shown as a function of neutron fluence in Fig. 2. ________________________________________________________________________________ ВОПРОСЫ АТОМНОЙ НАУКИ И ТЕХНИКИ. 2007. № 6. Серия: Физика радиационных повреждений и радиационное материаловедение (91), с. 72-77. 73 Φ [n/cm2, E>0.5MeV] 0 100 200 300 400 500 600 700 ∆ T 42 J, 4 7J [ °C ] 0 50 100 150 200 250 300 High flux Low flux DTVSFLUX1.JNB Weld All data Fig. 2. The measured transition temperature shifts as a function of neutron fluence The measured weld data compared to the prediction by formula (1.4) are shown in Fig. 3. The formula (1.4) gives a relatively good description of the data but at high shifts the formula predicts on average too high shift values which can be related to different approach for transition temperature shifts as it was described in PNAE (47 J and 71 J) and in the analysed data set (41 J and 47 J). ∆T-47Pred [°C] 0 50 100 150 200 250 300 350 ∆ T -4 7 M e a s [ °C ] 0 50 100 150 200 250 300 Weld All data ∆T = 800·(P+0.07·Cu)·Φ 1/3 PALLRUS.JNB 1: 1 High flux Low flux Fig. 3. The measured shift versus the shift calculated from the formula (1.4), weld data The distribution of base metal data in the copper- phosphorus plane is shown in Fig. 4, where low and high flux data points are indicated. Cu [wt-%] 0.00 0.05 0.10 0.15 0.20 0.25 P [w t- % ] 0.005 0.010 0.015 0.020 0.025 0.030 High flux Low fluxPVSCUB.JNB Base material All data Fig. 4. Distribution of base metal data in the Cu-P plane. Low flux marks are located on the top of high flux marks, i.e. open circles in bold include both flux values The copper and phosphorus contents are concentrated approximately around P= 0.013 mass.% and Cu = 0.10 mass.%. The measured transition temperature shifts of base metal are shown in Fig. 5 as a function of neutron fluence. The shifts correlate well with neutron fluence. Cu [wt-%] 0.00 0.05 0.10 0.15 0.20 0.25 P [w t- % ] 0.005 0.010 0.015 0.020 0.025 0.030 High flux Low fluxPVSCUB.JNB Base material All data Fig. 5. The measured transition temperature shifts of base metal as a function of neutron fluence The measured shifts of base metal are compared to the prediction (1.4) used for weld metal in Fig. 6. For most of the points the weld prediction formula (1.4) rep- resents an upper boundary for base metal but there are points in the range of ∆TPred = 80 to 140 oC, which ex- ceed the prediction for weld. ∆ T-47Pred [°C] 0 50 100 150 200 ∆T -4 7 M e a s [ °C ] 0 50 100 150 200 Base material All data ∆ T = 800·(P+0.07·Cu)·Φ 1/3 PALLRUSB.JNB 1: 1 High flux Low flux Fig. 6. The measured base metal shifts versus the shifts calculated from the formula (1.4) for weld metal 2.2. TREND CURVE FITTING The following five types of trial functions have been used in fitting: Fit-1: nCuaPaT Φ∗∗+∗=∆ )21( ; (11) Fit-2: 33)21( nn aCuaPaT Φ∗+Φ∗∗+∗=∆ ; (12) Fit-3: 321 321 nnn aCuaPaT Φ∗+Φ∗∗+Φ∗∗=∆ ; (13) Fit-4: 321 3)1(2)1(1 nnn aeCuaePaT Φ∗+−∗∗+−∗∗=∆ Φ∗−Φ∗− ; (14) Fit-5: )] 2 tanh(2/12/1[1)/(11 c FFcFFebFaT startsatn −∗+∗+    −−∗+∗=∆ . (15) Fit-1 has the same form as the Russian Code formula for weld metal. Fit-2 is a function, where a matrix damage term has been added to Fit-1. Fit-3 has separate power law exponents for the phos- phorus and the copper terms in addition to the matrix damage term. ________________________________________________________________________________ ВОПРОСЫ АТОМНОЙ НАУКИ И ТЕХНИКИ. 2007. № 6. Серия: Физика радиационных повреждений и радиационное материаловедение (91), с. 72-77. 74 Fit-4 describes the phosphorus and the copper terms with exponential function for both terms having own ex- ponent values. In addition, it contains the matrix dam- age term. Fit-5 describes the phosphorus term with tanh func- tion. 2.3. SUMMARY OF THE FITS: The following conclusions can be made as results of fitting the candidate functions: − In weld metal matrix damage term is needed. The addi- tion of the matrix term reduces the standard deviation of the fit by approximately 5 oC as compared to Fit-1, which has the functional form given in the Russian Code. − The use of phosphorus threshold level reduces the scatter of all fits approximately by 0.5 oC. The value of the derived threshold is P0 = 0.015 mass.% for weld and base metals and it is clearly the property of the analyzed data set. − The data can be described equally well with differ- ent types of functions. Standard deviation of all best fits made to the whole data set is about 18 oC. − The early development of copper and phosphorus re- sponse as a function of neutron fluence is fast but it cannot be determined quantitatively from the data. Fits were made also to low flux and high flux sub data sets separately and the analyses is described in the reference documents. Standard deviation of all data fits is typically 18 oC. Fits made to low flux data sub sets are typically 12 oC. Because the number of low flux data points is relatively low, the derived functions are not so well defined as the all data fits. 3. GUIDELINES FOR PREDICTION OF RADIATION EMBRITTLEMENT OF OPERATING WWER-440 RPVS This Guideline should be used for assessment of irra- diation embrittlement of RPV ferritic materials as a result of degradation during operation. Both approaches, i.e. transition temperatures based on Charpy impact notch toughness as well as based on static fracture toughness tests are to be used in RPV integrity evaluation. Integrity and lifetime assessment of reactor compo- nents are based, in principle, on application of fracture mechanics approach, thus determination of fracture toughness changes is of the main concern. Thus, two ways can be applied, depending on the number and type of specimens either in surveillance specimens or in the material qualification programme and other experimen- tal/material validation programmes: − direct determination of fracture toughness of RPV materials is required at certain periods during RPV operation, i.e. with given level of degradation – in this case, fracture toughness testing is performed and its temperature dependence is determined directly using either single- or multi-temperature testing ap- proaches. Fracture toughness of the degraded materi- als shall be determined and no initial initial proper- ties are required; − indirect determination of the fracture toughness of RPV material using Charpy V-notch impact test specimens and correlation formulae between brittle fracture temperature and temperature dependence of fracture toughness including their shifts. In this case, critical temperature of brittleness from Acceptance Tests, Tk0, must be well known, as surveillance Charpy specimens could serve only for determina- tion of a shift of the transition temperature 3.1. FRACTURE MECHANICS TEMPERATURES 3.1.1. MASTER CURVE APPROACH Reference temperature T0 is determined from static fracture toughness tests using a single- or multiple-tem- perature „Master Curve“ approach in accordance with the standard ASTM E 1921-02 and its application is given in [2, 3]. Then, a chosen lower tolerance bound (usually 5%) should be applied for determination of fracture toughness temperature dependence to be used in integrity/lifetime calculations. In principle, transition temperature T0 is usually de- termined for the required fluence for the RPV integrity assessment, i.e. for end-of-life fluence or for extended life fluence In addition, in special cases during RPV de- sign or for prediction of RPV behaviour during future operation, these temperatures could be also evaluated using similar procedure as for critical temperature of brittleness, together with shifts of brittle fracture tem- perature instead of shifts ΔTJC or separately only for trend curves of shifts ∆T0. Similarly, this temperature could be determined also for a given time of ageing, i.e. for characterisation of thermal ageing of materials. The reference temperature, T0, as determined in accordance with the standard ASTM E 1921-02 shall be increased by a margin, equal to a standard deviation σ (defined below) only for the tested condition, i.e. either initial or for a given degrada- tion state. Reference temperature T0 is defined from ex- perimentally determined values of static fracture tough- ness, KJC, adjusted to the thickness of 25 mm. Margin σ1 is added to cover the uncertainty in T0 as- sociated with using of only a few specimens to establish T0 while margin δ TM characterizes the scatter of the properties of forgings and welds. The total standard de- viation σ of the estimate of T0 is given by: σ=(σ1 2+ δ TM 2)1/2, (16) where margin σ1 is defined as σ1 = β / N0.5, oC, (17) where N = total number of specimens used to establish the value of T0, β = + 18 °C. If the value of δ TM is not available from Qualification tests of given material, the use of the following fixed val- ues is suggested: δ TM1 = 10°C for the base material, δ TM2 = 16°C for weld metals. ________________________________________________________________________________ ВОПРОСЫ АТОМНОЙ НАУКИ И ТЕХНИКИ. 2007. № 6. Серия: Физика радиационных повреждений и радиационное материаловедение (91), с. 72-77. 75 Thus, reference temperature that will be used in RPV integrity evaluation, RT0, is defined as: RT0 = T0 + σ. (18) In the case when surveillance specimens for determination of temperature T0 were irradiated at a different neutron fluence as required (i.e. end-of-life or extended life), an interpolation between results from two adjacent fluences is permitted. In this case, interpolation can be performed using power law formula. 3.1.2. BASE CURVE APPROACH “Base Curve” approach is approved as a national procedure [4] in Russian Federation. It contains the fol- lowingparts: − A new procedure of RPV brittle fracture resistance calculation; − A prediction procedure of fracture toughness tem- perature dependence on the base of testing small specimens (Prometey Probabilistic Model). 3.2. BRITTLE FRACTURE TEMPERATURE To determine both the initial and actual temperature de- pendences of fracture toughness KJC, respectively, the brit- tle fracture temperature T k has to be used. The brittle frac- ture temperature T k is usually determined for the fluence corresponding to the design or extended end of life. The brittle fracture temperature T k as a result of irradiation embrittlement is given by the following relationship: T k = ,Fko T + T ∆ (3.4) where T ko critical temperature of brittleness, [°C], FT∆ shift of the brittle fracture temperature due to irradiation, °C. The values of the brittle fracture temperature Tk0 was to be obtained from the Acceptance Tests. If such value is unknown, then the so-called “guaranteed” value from the corresponding Technical Requirements or from the Code for a given material is used. When using “guaranteed” val- ues of Tk0 in RPV integrity evaluation, then no tempera- ture margin should be added as this value is considered to be enough conservative. If the experimentally determined values of the initial brittle fracture temperature Tk0 from component Accep- tance Tests are known (based on component Passport) these should be increased by the temperature margin δ TM ; the margin has to take into account the scatter of the values of mechanical properties in forgings and welds; δ TM. Value of δ TM is the standard deviation of Tk0 deter- mined for the given forgings or weld metal in the frame of Qualification Tests or in the frame of a set of identical ma- terials established during production of the component by the identical technology. If this value is not available the application of the following values is suggested δ TM1 = 10°C for the base material, δ TM2 = 16°C for weld metals. This margin should be applied to the temperature Tk determined in accordance with equation (3.4). 3.3. DETERMINATION OF THE EFFECT OF IRRADIATION EMBRITTLEMENT 3.3.1. THE SHIFT OF THE BRITTLE FRACTURE TEMPERATURE DUE TO IRRADIATION First method: Results of tests of the surveillance program for specimens of the material of the vessel are available: respectively, also results for other vessels containing identical materials – for example, identical heat of the welding wire and flux: Shift of the brittle fracture temperature is determined from the formula: ∆TF = TkF - Tki, (19) where TkF is a value of transition temperature for a fluence F, Tki is a value of transition temperature for initial conditions (unirradiated). In both cases, these temperatures are determined from similar sets of specimens (minimum 12) using similar test equipment and procedure. The difference in fluence between specimens of one set should be smaller than ± 15 % of the mean value, and the difference in irradiation temperatures of individual specimens should be within 10 °C. Finally, the mean value of irradiation temperature should be not higher than + 10 °C above the inner wall temperature of the reactor pressure vessel. Obtained experimental values of KV (impact notch energy) are evaluated using the following equation KV = A + B tanh [(T-T0)/C], (20) where A, B, C and T0 are constants derived by statistical evaluation. It is strongly recommended to set lower shelf energy at values between 0 and 5 J to avoid incorrect fitting when a small number of specimens are tested in the lower shelf energy temperature region. Upper shelf energy should be fixed to the mean value of ductile fractured specimens. Shift of the transition temperature is determined for the criterion or consistent with national procedures KV = 41 J. (21) This procedure results in valid values of ΔTF only when the upper shelf energy, derived from the formula (3.6) - i.e., sum of (A+B), - is greater than 68 J. The results of determinations of the shift in the brittle fracture temperature obtained at least for three different neutron fluences are to be evaluated by the least squares method using the relationship: ΔTF = AF exp. (F.10-22)n, (22) where F is the fluence of fast neutrons with the energy higher than 0.5 MeV, AF exp and n are empirical constants obtained by statistical evaluation of surveillance data. Determination of shifts ΔTF is to be based on unirradiated and irradiated test data obtained from the same type of testing equipment and using the identical procedures for statistically processed curves. Mean experimentally determined fluence dependence of ΔTk in accordance with the equation (3.8) from surveillance tests is compared with the prediction for a given chemical composition of the tested material. If the real shifts are larger by more than 30 °C (approx. 1.5 SD given in (3.9+3.10)) for end-of-life fluence than predicted value, analysis of the difference should be performed and evaluated. In addition, the mean line from (3.8) should be vertically shifted upward by the value of δ TM calculated according to 3.2. If any experimental point exceeds this ________________________________________________________________________________ ВОПРОСЫ АТОМНОЙ НАУКИ И ТЕХНИКИ. 2007. № 6. Серия: Физика радиационных повреждений и радиационное материаловедение (91), с. 72-77. 76 adjusted trend curve, the curve should be shifted further until it bounds all data. This upper boundary of the shifts is to be used in assessment of RPV resistance against fast fracture. It is not allowed to extrapolate shifts of the transient temperatures for the fluences higher than double of the maximum fluence used for the experiment. Second method: If there are insufficient number or no surveillance test results: In such a case, the following prediction formula can be used for prediction of the shift in brittle fracture temperature: Metal Formula SD Num- ber Weld metal* [ ] 29.03.51884 Φ∗∗+∗=∆ CuPT = 29.0)064.011.1(800 Φ∗∗+∗∗ CuP SD=22.6 oC (23) Base metal* ΔT = 8.37* F0.43 SD=21.7 °C (24) *Formulae are valid for neutron fluences in the range 1022 < F < 4 x 1024 m-2 SD = standard deviation Both formulae represent the mean trend line; this mean value with the margin should be used for RPV integrity assessment. 3.3.2. DETERMINATION OF THE REFERENCE TEMPERATURE T0 FOR REQUIRED TIME OF OPERATION If this cannot be determined directly by fracture tough- ness testing, then the following mixed way (i.e. combina- tion of static fracture toughness and Charpy V-notch im- pact test results) may be conservatively used for determina- tion of temperature T0 during operation, i.e. T0 operation = T0 initial + 1.1 ΔTF, (25) where ΔTF is determined by the same process as is shown in 3.3.1, i.e. using Charpy impact specimen test- ing and/or prediction using formula (3.8). In this case, the same margin than for the scatter of the material and the margin equal to standard deviation, Δσ, in accordance with the standard E 1921-02 should be applied for determination of T0 initial. CONCLUSION New Guidelines for prediction irradiation embrittlement in reactor pressure vessel materials of WWER-440 type reactors were prepared within the IAEA Co-ordinated Research project. These Guidelines are based on analysis of experimental data from irradiation of materials of these RPVs collected in the IAEA International Database on RPV Materials (IDRPVM). These Guidelines contain formulae for prediction irradiation embrittlement for base and weld metals of this type of reactors, either based on brittle transition temperature, Tk, or reference temperature of Master Curve approach, T0. Recommendation for the use of real experimental data from testing surveillance specimens from these RPVs was also elaborated and recommended. REFERENCES 1. PNAE-G0002-86 – Standard for strength calculations of components and piping in NPPs, PN AE G-7 002-86. M.: «Energoatomizdat», 1989. 2. ASTM E 1921–Test Method for Determination of Reference Temperature, T0, for Ferritic Steels in the Transition Range, ASTM, West Conshohocken, PA, (-97 version published in 1998 and -02 version pub- lished in 2002) 3. IAEA TECDOC xxxx- IAEA Guidelines for Applica- tion of the Master Curve Approach to Reactor Pres- sure Vessel Integrity. IAEA TECDOC (in print). 4. Procedure for determination of residual lifetime of WWER reactor pressure vessels during operation. MRK-SKhR-2000. Sankt Peterburg-Moscow, 2000. ПРЕДСКАЗАНИЕ РАДИАЦИОННОГО ОХРУПЧИВАНИЯ ВНУТРИКОРПУСНЫХ МАТЕРИАЛОВ РЕАКТОРА ВВЕР-440 М. Брумовский, М. Куткаб Л. Дебарберис, Ф. Желлемот, А. Крайков, М. Вало Предсказание радиационного охрупчивания материалов внутрикорпустных устройств (ВКУ) обычно выполняется в соотвествии с кодами и стандар- тами, основанными на обширной информации, накопленной в ходе модельных и исследовательских программ. Существующий Российский Код (Стандарт для вычислений прочности и компонентов и трубопроводов в атомных электростанциях (АЭС) – PNAE G 7-002-86) для оценки радиационного охрупчива- ния ВКУ реакторов ВВЭР хорошо зарекомендовал себя на протяжении более чем 20 лет; он основан на экспериментальных данных, полученных в иссле- довательских реакторах с ускоренным облучением. Оценка упомянутого выше Кода была выполнена без результатов с образцов-свидетелей, которые были получены в 1980-1990 годах. Таким образом, необходим новый анализ всех имеющихся данных для более точного прогнозирования радиационного охрупчивания материалов ВКУ. На основании того факта, что было использовано большое количество данных с макетных и исследовательских программ, Международная База данных МАГАТЭ по материалам ВКУ была использована для подробного анализа радиационного охрупчивания ВКУ материалов для реакторов ВВЭР. Таким образом, в рамках Координационного проекта МАГАТЭ были выполнено следующее: сбор полных данных с образцов-свиде- телей ВВЕР-440 и других подобных важных данных в Международную базу данных; анализ данных по радиационному охрупчиванию ВКУ материалов ВВЭР-440 с использованием международной базы данных; оценка формулы прогнозирования в зависимости от химического состава материала, флюенса нейтронов и нейтронного потока, разработка основных положений для предсказания радиационного охрупчивания эксплуатируемых внутрикорпусных устройств ВВЭР-440, включая методологию для оценки контрольных данных конкретной действующей установки. ПРОГНОЗУВАННЯ РАДІАЦІЙНОГО ОКРИХЧЕННЯ ВНУТРІШНЬОКОРПУСНИХ МАТЕРІАЛІВ РЕАКТОРА ВВЕР-440 М. Брумовський, М. Кутка, Л. Дебарберис, Ф. Желлемот, А. Крайков, М. Вало Прогнозування радіаційного окрихчення матеріалів внутрішньо корпусних пристроїв (ВКП) зазвичай виконується у відповідності з кодами та стандартами, заснованими на численній інформації, що була накопичена на базі модельних та дослідницьких програм. Існуючий Російський Код (Стандарт для обчислення міцності компонентів і трубопроводів в атомних електростанціях (АЕС)-PNAE G 7 –002-86) для оцінки радіаційного окрихчення ВКП реакторів ВВЕР зарекомендував себе на протязі більш, ніж 20 років; він заснован на експриментальних даних, отриманих в дослідницьких реакторах з прискореним опроміненням. Оцінка вище згаданого Кода була виконана без результатів із зразків-свідків, які були отримані у 1980-1990 роках. Таким чином, необхідно провести новий аналіз усіх наявних даних для більш точного прогнозування радіаційного окрихчення матеріалів ВКП. На підставі того факту, що було використано велику кількість даних з макетних та дослідницьких програм, Міжнародна база даних МАГАТЕ по матеріалам ВКП біла використана для докладного аналізу радіаційного окрихчення ВКП матеріалів для реакторів ВВЕР. Таким чином, в межах Координаційного проекту МАГАТЕ було виконано наступне: Збирання повних даних із зразків-свідків ВВЕР-440 та інших подібних важливих даних в Міжнародну базу даних, аналіз даних по радіаційному окрихченню ВКП матеріалів ВВЕР-440 з використанням міжднародної бази даних, оцінка формули прогнозування в залежності від хімічного складу матеріалу, флюенса нейтронів та нейтронного потоку, розробка основних положень для прогнозування радіаційного окрихчення експлуатуємих внутрішньо корпусних пристроїв ВВЕР-440, включно з методологією для оцінки контрольних даних конкретної діючої установки. ________________________________________________________________________________ ВОПРОСЫ АТОМНОЙ НАУКИ И ТЕХНИКИ. 2007. № 6. Серия: Физика радиационных повреждений и радиационное материаловедение (91), с. 72-77. 77 УДК 621.039.577 PREDICTION OF IRRADIATION EMBRITTLEMENT IN WWER-440 REACTOR PRESSURE VESSEL MATERIALS INTRODUCTION 2. IRRADIATION EMBRITTLEMENT MODELING 2.1. PRINCIPLES OF THE TREND CURVE DERIVATION 2.2. TREND CURVE FITTING 2.3. SUMMARY OF THE FITS: 3. GUIDELINES FOR PREDICTION OF RADIATION EMBRITTLEMENT OF OPERATING WWER-440 RPVS 3.1. FRACTURE MECHANICS TEMPERATURES 3.1.1. MASTER CURVE APPROACH 3.1.2. BASE CURVE APPROACH 3.2. BRITTLE FRACTURE TEMPERATURE 3.3. DETERMINATION OF THE EFFECT OF IRRADIATION EMBRITTLEMENT 3.3.1. THE SHIFT OF THE BRITTLE FRACTURE TEMPERATURE DUE TO IRRADIATION CONCLUSION REFERENCES ПРЕДСКАЗАНИЕ РАДИАЦИОННОГО ОХРУПЧИВАНИЯ ВНУТРИКОРПУСНЫХ МАТЕРИАЛОВ РЕАКТОРА ВВЕР-440 М. Брумовський, М. Кутка, Л. Дебарберис, Ф. Желлемот, А. Крайков, М. Вало

Prediction of irradiation embrittlement in WWER-440 Reactor Pressure Vessel materials

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