11. STORAGE FOR FREE RELEASE AND RECYCLING OF MATERIAL
11.6. Experience on amount of recovered material
Statistically more than 95% can be free released from metallic scrap; it may be in the form of small pieces (delivered in drums or baskets) or larger LLW objects such as turbines and heat exchangers. As regards SGs, 85% can be recovered for recycling — except for the tubes (Fig. 67). If the SG, however, is pre-decontaminated
by system decontamination while the reactor is still in operation, the tubes may become clean enough for a more improved treatment, i.e. either volume reduction by melting or even ready for free release after additional forced secondary cleaning. Tubes made of Ni–Cr alloys provide valuable material which is worth retrieving if possible (ALARA). This may then also contribute to an economic refund.
Blasting cabin Tumble blaster
FIG. 66. Minimization of secondary waste by decontamination and crushing.
310 te 400 m³
80-85 %
Melting for volume reduction Melting for free release Blasting
chamber
Tube blasting robot Cutting cell
FIG. 67. SG waste treatment concept in Sweden.
Experience on radiation exposure to the personnel during decommissioning
An SG (typically 400 m3, 310 t from Ringhals in Sweden) without decontamination may draw 30–60 mman·Sv in collective dose during a waste recycling treatment as demonstrated during 2006–2007. Such results can be achieved if a careful ‘dose budget’ is established before the treatment in order to investigate any of the dose demanding phases and feasible counter measures proposed to minimize personal dose exposition to the workers.
Overall personal dose exposure when treating radioactive material from nuclear power plants (main nuclide 60Co) as well as from fuel fabrication factories (main nuclides from uranium and 241Am) was demon-strated to be less than 0.8 mSv, a mean value over a year’s handling, with a maximum of about 3 mSv for any person in the melting facility.
In Sweden, the SGs are treated in a specifically designed ‘cutting cell’ with necessary shielding and compu-terized and robotized tools inside. An operator stand equipped with lead glass windows allows visual control when necessary. Normally, remote supervision by cameras, internal lights and positioning control is applied successfully. Also, inside packaging of remotely segmented pieces into different boxes for further treatment — controlled by robots — allows optimized handling to provide a maximum of free releasable material for recycling.
REFERENCES
[1] INTERNATIONAL ATOMIC ENERGY AGENCY, Plant Life Management for Long Term Operation of Light Water Reactors, Technical Reports Series No. 448, IAEA, Vienna (2006).
[2] INTERNATIONAL ATOMIC ENERGY AGENCY, Nuclear Power Plant Life Management Processes: Guidelines and Practices for Heavy Water Reactors, IAEA-TECDOC-1503, IAEA, Vienna (2006).
[3] INTERNATIONAL ATOMIC ENERGY AGENCY, Assessment and Management of Ageing of Major Nuclear Power Plant Components Important to Safety: Steam Generators, IAEA-TECDOC-981, IAEA, Vienna (1997).
[4] INTERNATIONAL ATOMIC ENERGY AGENCY, Assessment and Management of Ageing of Major Nuclear Power Plant Components Important to Safety: PWR Pressure Vessels, IAEA-TECDOC-1120, IAEA, Vienna (1999).
[5] INTERNATIONAL ATOMIC ENERGY AGENCY, Assessment and Management of Ageing of Major Nuclear Power Plant Components Important to Safety: Primary Piping in PWRs, IAEA-TECDOC-1361, IAEA, Vienna (2003).
[6] TSUBOTA, M., et al., Effect of Cold Work on the SCC Susceptibility of Austenitic Stainless Steels” Proceedings of 7th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reac-tors,Vol.1, 519-527 (1995)
[7] INTERNATIONAL ATOMIC ENERGY AGENCY, Assessment and Management of Ageing of Major Nuclear Power Plant Components Important to Safety: BWR Pressure Vessel Internals, IAEA-TECDOC-1471, IAEA, Vienna (2005).
[8] INTERNATIONAL ATOMIC ENERGY AGENCY, Assessment and Management of Ageing of Major Nuclear Power Plant Components Important to Safety: CANDU Pressure Tubes, IAEA-TECDOC-1037, IAEA, Vienna (1998).
[9] CHANUSSOT, M., THÉVENET, R., “Replacement of heavy components of the main primary system – recent inno-vations made by AREVA-NP”, Proceedings of the ENC 2005, Paris (2005)
[10] KASTL, H., “Steam generator replacement from ALARA aspects”, IAEA Regional Asia Training Course on Steam Generator in Nuclear Power Plants, November 10-21, 2003, Pusan, Republic of Korea (2003) [11] KASTL, H., “Mechanical Decontamination of Primary Pipe Ends during Steam Generator Replacement Project”
IAEA Regional Asia Training Course on Steam Generator in Nuclear Power Plants, November 10-21, 2003, Pusan, Republic of Korea (2003)
[12] TERRY, I., KASTL, H., “Steam Generator Replacement (1) and (2)” IAEA Regional Asia Training Course on Steam Generator in Nuclear Power Plants, November 10-21, 2003, Pusan, Republic of Korea (2003) [13] BREZNIK, B., KASTL, H., “Steam Generator Replacement in NPP Krško”, Krško, Slovenia (2003)
[14] CHARMPIGNY, et al., ”Ten Years Experience on Inconel Alloys Since the Bugey Leak in 1991”, ASME PVP Vol.
444, (2002)
[15] YAMASHITA, H., OHIDE, A., MIYANO, H., DUINK, S., SAGAWA, W., “Core shroud replacement technique &
procedure on TEPCO 1F-3” Proc.6th Int. Conf. on Nuclear Engineering, ICONE-6404 (1998)
[16] SATO, Y., INAMI, I., KURIBAYASHI, N., SAKAI, T., WILLE, H., FRIEDRICH, E., “Chemical decontamination of reactor pressure vessel and internals in TEPCO 1F-3” Proc.6th Int. Conf. on Nuclear Engineering, ICONE-6405 (1998)
[17] NUREG-0313, "Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," U.S. Nuclear Regulatory Commission, (Rev. 2) (1988)
[18] MIYANO, H., “Installation technology of reactor internals on Shroud Replacement Work” Proc.7th Int. Conf. on Nuclear Engineering, Special Lecture TOSHIBA (1999)
[19] MARUSKA, C.C., “Steam Generator Life Cycle Management Ontario Power Generation (OPG) Experience”, Proceedings of the 4th International Steam Generator and Heat Exchanger Conference, Toronto (2002)
[20] NICKERSON, J., MARUSKA, C.C., “Steam Generator Management at Ontario Hydro Nuclear Stations”, Proceed-ings of the 3rd International Steam Generator and Heat Exchanger Conference, Toronto (1998)
[21] BURNS, B., MILLMAN, J., “Bruce A Steam Generator (Boiler) Replacement”, Proceedings of the 5th CNS Interna-tional Steam Generator Conference, Toronto (2006)
[22] HART, R., “Steam Generator Replacement at Bruce A Unit 1 and Unit 2”, Proceedings of the 5th CNS International Steam Generator Conference, Toronto (2006)
[23] TAPPING, R., “Steam Generator Aging in CANDUs: 30 years of Operation and R&D”, Proceedings of the 5th CNS International Steam Generator Conference, Toronto (2006)
[24] INTERNATIONAL ATOMIC ENERGY AGENCY, Application of the Concepts of Exclusion, Exemption and Clearance Safety Guide, IAEA Safety Standards Series No. RS-G-1.7, IAEA, Vienna (2004).
ABBREVIATIONS
CRDM Control rod driving mechanism EDM Electric discharge machining EPRI Electric power research institute IGSCC Intergranular stress corrosion cracking LSFCR Large scale fuel channel replacement
NDE Non-destructive examination
OSG Original steam generator
PWSCC Primary water stress corrosion cracking RCL Reactor coolant line/loop
RCS Reactor coolant system
RFQ Request for quotation
RPV Reactor pressure vessel
RVHR Reactor vessel head replacement
RVI Reactor vessel internals
SCC Stress corrosion cracking
SFCR Single fuel channel replacement
SG Steam generator
SGR Steam generator replacement
TGSCC Transgranular stress corrosion cracking
UT Ultrasonic test
WBS Work breakdown structure
Appendix I
SAFETY, LICENSING AND REGULATORY ISSUES IN JAPAN
I.1. LICENSING PROCESS
The procedure for obtaining the permits and licences required for the replacement of equipment is described in this appendix.
I.2. LAWS AND AGREEMENTS
The major laws and agreements to consider when replacing equipment are listed in Table 3. To ensure that the design of the replacement equipment is appropriate, changes must be made to the nuclear reactor instal-lation permit and construction work plan licence application forms in accordance with Act on the Reguinstal-lation of Source Material, Nuclear Fuel Material and Reactors and Electricity Utilities Industry Act. Excessive worker exposure to radiation must also be minimized, as well as the safety of cranes assured, in accordance with Industrial Safety and Health Law, Laws Concerning the Prevention of Radiation Hazards due to Radioisotopes and others, and other applicable laws.
One procedure in Japan is that the consent of the local communities involved, based on a safety agreement, must also be obtained. This is not stipulated by law, but to ensure the safety of these communities, it is necessary to forge an agreement with the local governing body, obtain the consent of the communities to the plans, ahead of time, in accordance with the agreement, and report to the communities on the progress of the construction work that is performed.
I.3. FLOW OF PERMITS AND LICENCES
Using the replacement of an SG as an example, Fig. 68 shows the procedural flow involving permits and licences based on the Act on the Regulation of Source Material, Nuclear Fuel Material and Reactors and Electricity Utilities Industry Act, from among the laws and agreements listed in the previous section. There are two perspectives from which to consider procedures covering the replacement of SGs: (1) the replacement of the SGs themselves and (2) the storage of the SGs after their replacement.
(1) Nuclear reactor installation change permit application forms: When SGs are to be replaced, this gives rise to changes to items in the main text (heat transfer pipe dimensions and steam generation amounts are added), and nuclear reactor installation change permit application forms become a requirement. Similar applications are also required for items stored in the SG storage facilities.
(2) Construction work plan licence application forms: When SGs are to be replaced, construction work plan licence application forms must be filed, covering changes in the materials used by the SGs themselves, changes in the heat transfer surfaces and internal structures, and changes in the strength and resistance to earthquakes calculations accompanying weight increases and other factors. In addition, applications for repairs in regard to the provisional openings of nuclear reactor containment vessels, provisional openings of external shields, and annulus seal cutting, which are all required for the work, need to be filed. Applica-tions for construction work plan licences concerning the capacities, shields and other aspects of the SG storage facilities are also filed.
I.4. IMPACT OF REPLACEMENT
After the nuclear reactor installation change permit applications have been filed, they go through the
Following the application process for the nuclear reactor installation change permits, an additional month or two is needed for the construction work plan licence application to be processed. An overview of the processes is shown in Fig. 69.
I.5. REGULATORY POSITIONS AND UPFRONT LICENSING
In Japan, the replacement of equipment is not subject to regulations. These replacements of SGs and other equipment have been implemented in accordance with the maintenance plans of the operators.
TABLE 3. LIST OF PERMITS AND LICENCES
Laws/agreements Permit/licence types Remarks
Act on the Regulation of Source Material, Nuclear Fuel Material
Applicable to changes in safety analysis results
Electricity Utilities Industry Act Construction work plan licence application form
Applicable to usage of sites for storage facilities
Factory Location Act Green spaces Pertaining to usage of sites for storage facilities
Building Standards Law Construction verification Applicable to storage facility structural parameters
Applicable to plans concerning safety of construction work
Disposal of concrete waste materials by burying on premises
Plans and performance reports Ship Safety Law Marine transportation plans Foreign Exchange Control Law Introduction of technologies
Replacement of SGs
Any changes in the main text?
Y
N
N
Nuclear reactor installation change permit application
Nuclear reactor installation change permits Y
Y
N Important equipment?
Any remodelling?
Construction work plan licence application
Construction work plan licences
Commencement of construction work Construction work plan licence notification
N
Y Any changes in the items set
forth on the installation permit application forms?
Construction work plan licences
FIG. 68. Flow of permits and licences.
7 9 11 1 3 5 7 9 11 1 3 5 7 9 11 1 3 5 7 9 11 1 3 5 7 9 11 1 3 5 7 9 11 1997
1992 1993 1994 1995 1996
EP
CP (Storage facility) CP (SGR) SG manufacturing Construction(Storage facility)
SGR Dcision of replacement
Agreement of safety
FIG. 69. Overview of permit and licence processes.
Appendix II
SG REPLACEMENT HISTORY (PWR) 1979–2005
Country Plant OSG designer Reactor type SG model Year
France DAMPIERRE 2 FRAMATOME 3 loop 900 47-22 2005
USA ARKANSAS ONE 1 B&W 2 loops Model B 177 R OTSG / OTSG 2005
USA CALLAWAY 1 WESTINGHOUSE Model 412 F / 73-19T 2005
USA PALO VERDE 1 CE 2 loops - System 80 CE-80 2005
Belgium DOEL 2 WESTINGHOUSE 2 loops W 44 / MHI 2004
France TRICASTIN 4 FRAMATOME 3 loop 900 51M / 55-19 2004
USA OCONEE 2 B&W 2 loops Model B 177 R OTSG / OTSG 2004
USA OCONEE 3 B&W 2 loops Model B 177 R OTSG / OTSG 2004
USA PRAIRIE ISLAND 1 WESTINGHOUSE 2 loops W 51 / 56-19 2004
France ST LAURENT B-2 FRAMATOME 3 loop 900 51M / 55-19 2003
USA CALVERT CLIFFS 2 CE 2 loops CE CE-67 / x 2003
USA OCONEE 1 B&W 2 loops Model B 177 R OTSG / OTSG 2003
USA PALO VERDE 2 CE 2 loops - System 80 CE-80 / x 2003
USA SEQUOYAH 1 WESTINGHOUSE 4 loops – Ice containment 51 / D75 2003
France FESSENHEIM 1 FRAMATOME 3 loop 900 51A / 47-22 2002
USA CALVERT CLIFFS 1 CE 2 loops CE CE-67 / x 2002
USA SOUTH TEXAS 2 WESTINGHOUSE Model 414 E / D94S 2002
Belgium TIHANGE 2 FRAMATOME 3 loops FRA 51M / MHI 2001
France TRICASTIN 3 FRAMATOME 3 loop 900 51M / 47-22 2001
Japan IKATA 2 MHI 2 loops - 2nd generation 51M / MHI 2001
USA FARLEY 2 WESTINGHOUSE 3 loops W 51 / x 2001
France GRAVELINES B-4 FRAMATOME 3 loop 900 51M / 47-22 2000
USA ARKANSAS ONE 2 CE 2 loops Model *** ***** 2000
Japan GENKAI 2 MHI 2 loops - 2nd generation 51M / MHI 2000
Slovenia KRSKO WESTINGHOUSE Model 212 D4,2 / FRA 2000
USA DC COOK 1 WESTINGHOUSE 4 loops – Ice containment 51 / x 2000
USA FARLEY 1 WESTINGHOUSE 3 loops W 51 / x 2000
USA INDIAN POINT 2 WESTINGHOUSE 4 loops W 44 / 44F 2000
USA SOUTH TEXAS 1 WESTINGHOUSE Model 414 E / D94S 2000
Japan IKATA 1 MHI 2 loops - 1st generation 51 / x 1999
Switzerland BEZNAU 2 WESTINGHOUSE 2 loops W 33W / 33-19 1999
Belgium TIHANGE 3 WESTINGHOUSE Model 314 E1 / 70-19T 1998
France TRICASTIN 1 FRAMATOME 3 loop 900 51M / 47-22 1998
Korea KORI 1 WESTINGHOUSE Model 212 51 / KHIC 1998
USA BRAIDWOOD 1 WESTINGHOUSE 4 loops W D4,1 / D4 1998
USA SHEARON HARRIS 1 WESTINGHOUSE Model 312 D75 1998
USA ST LUCIE 1 CE 2 loops CE CE-67 / 67 1998
France TRICASTIN 2 FRAMATOME 3 loop 900 51M / 47-22 1997
Japan OHI 2 WESTINGHOUSE Model 412 51A / 54FA 1997
Spain ALMARAZ 2 WESTINGHOUSE Model 312 D3,2 / 61W-D3 1997
USA BYRON 1 WESTINGHOUSE 4 loops W D4,1 / D4 1997 USA MCGUIRE 1 WESTINGHOUSE 4 loops – Ice containment D2,1 / x 1997 USA MCGUIRE 2 WESTINGHOUSE 4 loops – Ice containment D3,1 / D3 1997
USA SALEM 1 WESTINGHOUSE 4 loops W 51 / x 1997
Belgium DOEL 4 WESTINGHOUSE Model 314 E1 / 79-19T 1996
France GRAVELINES B-2 FRAMATOME 3 loop 900 51M / 47-22 1996
Japan MIHAMA 3 MHI 3 loops - 1st generation 51 / 54F 1996
Japan TAKAHAMA 1 WESTINGHOUSE Model 312 51 / 54F 1996
Spain ALMARAZ 1 WESTINGHOUSE Model 312 D3,1 / 61W-D3 1996
Spain ASCO 2 WESTINGHOUSE Model 312 D3,2 / 61W-D3 1996
USA CATAWBA 1 WESTINGHOUSE 4 loops – Ice containment D3,2 / CFR-80 1996
USA R.E. GINNA WESTINGHOUSE 2 loops W 44 / x 1996
USA POINT BEACH 2 WESTINGHOUSE 2 loops W 44 / 44F 1996
Belgium TIHANGE 1 FRAMATOME 3 loops FRA 51A / MHI 1995
France DAMPIERRE 3 FRAMATOME 3 loop 900 51M / 47-22 1995
France ST LAURENT B-1 FRAMATOME 3 loop 900 51M / 47-22 1995
Japan MIHAMA 1 WESTINGHOUSE 2 loops W CE-33 / 35F 1995
Spain ASCO 1 WESTINGHOUSE Model 312 D3,1 / 61W-D3 1995
Sweden RINGHALS 3 WESTINGHOUSE Model 312 D3,2 / 72W-D3 1995
USA NORTH ANNA 2 WESTINGHOUSE 3 loops W 51D / 54F 1995
France GRAVELINES B-1 FRAMATOME 3 loop 900 51M / 47-22 1994
Japan GENKAI 1 MHI 2 loops - 1st generation 51 / 52F 1994
Japan MIHAMA 2 MHI 2 loops - 1st generation 44 / 46F 1994
Japan OHI 1 WESTINGHOUSE Model 412 51A / 52FA 1994
Japan TAKAHAMA 2 MHI 3 loops - 1st generation 51 / 52F 1994
USA V.C. SUMMER 1 WESTINGHOUSE 3 loops W D3,1 / D75 1994
Belgium DOEL 3 FRAMATOME 3 loops FRA 51M / 61W 1993
France BUGEY 5 FRAMATOME 3 loop 900 51A / 51B 1993
Switzerland BEZNAU 1 WESTINGHOUSE 2 loops W 33W / 33-19 1993
USA MILLSTONE 2 CE 2 loops CE CE-67 / 67 1993
USA NORTH ANNA 1 WESTINGHOUSE 3 loops W 51 / 54F 1993
USA PALISADES CE 2 loops CE CE-67 / 67 1991
France DAMPIERRE 1 FRAMATOME 3 loop 900 51M / 51B 1990
Sweden RINGHALS 2 WESTINGHOUSE 3 loops W C51 / 51 1989
USA DC COOK 2 WESTINGHOUSE 4 loops – Ice containment 51D / 54F 1989
USA INDIAN POINT 3 WESTINGHOUSE 4 loops W 44S / 44F 1989
USA POINT BEACH 1 WESTINGHOUSE 2 loops W 44S / 44F 1984
USA H.B. ROBINSON 2 WESTINGHOUSE 3 loops W 44S / 44F 1984
Germany OBRIGHEIM (KWO) KWU 2 loops KWU KWU 1983
USA TURKEY POINT 4 WESTINGHOUSE 3 loops W 51 / 44F 1983
USA TURKEY POINT 3 WESTINGHOUSE 3 loops W 44S / 44F 1982
USA SURRY 1 WESTINGHOUSE 3 loops W 51D / 51F 1981
USA SURRY 2 WESTINGHOUSE 3 loops W 51D / 51F 1979
Country Plant OSG designer Reactor type SG model Year
Appendix III
RVHR LIST FROM THE ORIGIN 1993–2005
Country Unit Original design/
new RVH supplier Plant family RVHR year
France BUGEY 5 FRAMATOME/
France BELLEVILLE 2 FRAMATOME/
new RVH supplier Plant family RVHR year
France CHINON B-2 FRAMATOME/
USA Three Mile Island 1 B&W B&W 2003
Country Unit Original design/
new RVH supplier Plant family RVHR year
China GUANGDONG 1 FRAMATOME/
new RVH supplier Plant family RVHR year
CONTRIBUTORS TO DRAFTING AND REVIEW
Rabbat, R. Atomic Energy of Canada Ltd, Canada Bezdikian, G. Électricité de France, France
Thevenet, R.S. AREVA, France
Hasegawa, Y. Kansai Electric Power Co., Japan Ishimoto, S. Mitsubishi Heavy Industries Ltd, Japan Suezono, N. Toshiba Corporation Power Systems Co., Japan Yamashita, N. Tokyo Electric Power Co., Japan
Lorenzen, J. Studsvik Nuclear AB, Sweden
Heilker, W. Westinghouse Electric Co., United States of America James, W. Entergy Operations Inc., United States of America Rushing, L. Entergy Operations Inc., United States of America Kang, Ki-Sig International Atomic Energy Agency
Consultancy Meeting Vienna, Austria: 8–11 August 2006
Technical Meeting Vienna, Austria: 3–6 July 2007
7.16 mm
IAEA Nuclear Energy Series No. NP-T-3.2Heavy Component Replacement in Nuclear Power Plants: Experience and Guidelines