ML20081K061
| ML20081K061 | |
| Person / Time | |
|---|---|
| Site: | 05200005 |
| Issue date: | 10/21/1994 |
| From: | Grover J, Sciacca F, Shaffer C SCIENCE & ENGINEERING ASSOCIATES, INC. |
| To: | NRC |
| Shared Package | |
| ML20081K038 | List: |
| References | |
| CON-FIN-J-2002, CON-NRC-03-93-032, CON-NRC-3-93-32 SEA-93-704-04-A, SEA-93-704-04-A:3, SEA-93-704-4-A, SEA-93-704-4-A:3, NUDOCS 9503280382 | |
| Download: ML20081K061 (104) | |
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REVIEW OF MATERIALS-RELATED ISSUES FOR
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REACTOR INTERNALS AND HORIZONTAL PRESSURE TUBE DESIGN ! FOR CANDU 3 REACTOR DESIGN i i i l ! i i October 21,1994 i
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Jeffrey L Grover* l Frank W. Sciacca l Clinton J. Shaffer ; Science and Engineering Associates,Inc. j i l
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t i l I SEA 93-704-04-A3 i I : I ; Technical Evaluation Report ! i REVIEW OF MATERIALS RELATED ISSUES FOR REACTOR INTERNALS AND HORIZONTAL PRESSURE TUBE DESIGN FOR CANDU 3 REACTOR DESIGN October 21,1994 F Jeffrey L Grover l Frank W. Sciacca ! Clintc.n J. Shaffer ! Science and Engineering Associates,Inc. [
- SEA Consultant '
I Contract: NRC-03-93-032 Job Code No:J-2002 Task Order No:4 j I
ABSTRACT Science and Engine (ring Associates, Inc. (SEA) has, under contract to the NRC, reviewed documents pertaining to the Canadian Deuterium Uranium 3 (CANDU 3) reactor design in preparation of the design certification review. These documents were submitted to the NRC by AECL Technologies, the U.S. sponsors of the design for its designers, the Atomic Energy of Canada, Ltd. (AECL). SEA was contracted to provide a technical review of the reactor internals and core support materials including the pressure tubes, calandria tubes, calandria, shield tank, feeder tubes, reactivity control mechanisms, and moderator I piping materials. The focus of the review was, however, on the pressure tubes and the fuel channel assemblics. This report discusses the findings of the technical review including several materials-related issues which likely will present challenges to the licensing review process. ( I I I I I I I I I ... ul
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l I EXECUTIVE
SUMMARY
The NRC is reviewing the Canadian Deuterium Uranium 3 (CANDU 3) reactor design in preparation of the design certification review. CANDU 3 is being designed by Atomic Energy of Canada, Ltd (AECL) and is sponsored in the U.S. by AECL Technologies. Science and Engineering Associates, Inc. was contracted to technically review documentation submitted to the NRC by AECL Technologies as the documentation relates to reactor internal and core support materials including the pressure tubes, calandria tubes, calandria, shield tank, feeder tubes, reactivity control mechanisms, and moderator piping materials. The focus of the review was the horizontal pressure tubes and the fuel channel assemblies. The documentation reviewed generally represents the conceptual design as it existed in 1989. Since the design has continued to develop beyond the state represented by these documents, this review does not necessarily represent the current design. However, it is likely that the materials-related issues identified in this report will not change substantially. Many of the design features of the CANDU reactors are not covered under U.S. Codes, Regulations, and Standards. These include the use of the multiple small diameter pressure tubes, roll expanded joints in the I reactor coolant pressure boundary, and a refueling machine that clamps onto the fuel channel and refuels the reactor while it is on-line. In addition, many of the components are made from materials that are not approved by the U.S. Codes. All of the U.S. Codes, Standards, and Regulations are written explicitly for light water reactors. Although they are not directly applicable to the CANDU reactor design, they provide guidance in the issues of concern. Each of the key reactor components willbe described briefly below, along with a summary of the key findings of this review. Further, Table A contains a list of the key findings that are likely to challenge the licensing review process.
*ihe basic component of the CANDU reactor is the fuel channel, which consists of the fvel bundles walled inside a small diameter pressure tube made from Zr-2.5% Nb. Heavy water acts as the reactor coolant and flows through this tube. The pressure tube is located concentrically inside another tube, which is called the calandria tube. The calandria tube is surrounded by relatively cool heavy water, which acts as the moderator. To maintain separation of the pressure tube and calandria tube, small toroidal springs (garter springs) are placed around the outside of the pressure tubes at four locations along the length. The annulus between the pressure tube and calandria tube is filled with dry CO 2gas that acts as a thermal insulator and is also used in leak detection.
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E as Under normal conditions, the life of the pressure tube is limited by irradiation induced creep and growth, which may cause the pressure tube to sag and contact the calandria tube. Although design changes have been made to mirumize the effects of creep and growth, pressure tube /calandria tube contact is still a possibility. When this happens, boiling may occur in the moderator, but in addition, a thermal gradient may be set up that leads to delayed hydride cracking of the pressure tube. Although the pressure tube is designed for leak before break, this particular failure mechanism results in a long, shallow flaw that ultimately ruptures without prior leakage. When the pressure tube fails, it is also likely that the calandria tube will fail because it is not designed for the dynam;c pressure resulting from a pressure tube rupture. Although there have been several cases of p c sure tubes that leaked without failing, there is at least one case where the pressure tube faded without prior leakage. This occurred on an older CANDU plant that had Zircaloy 2 pressure tubes rather than Zr-2.5% Nb pressure tubes. That failure was initiated by an annulus spacer that slipped out of position, allowing the pressure tube to contact the calandria tube. The past reliability of zirennium pressure tubes worldwide has been 4x104 leaks /Irr-EFPY (pressure tube effective full power yea) and 2x105 ruptures /PT-EFPY. The reliability of CANDU production reactors is similar, with a failure rate of 9x105 leaks /PT-EFPY and 3x10 4ruptures /PT-EFPY. Because the d CANDU 3 has 232 pressure tubes, a failure rate of 10 stillleads to a 2% probability of failure of a single tube per reactor year. The high probability of a pressure tube failure is somewhat mitigated by the fact g that the consequences of a single pressure tube failure are much smaller than typically assumed for a a failure of the reactor coolant pressure boundary in a light water reactor. The calandria tube is also subject to creep, but because it operates at a lower temperature than the pressure I tube, the thermal creep component will be smaller. A potential materials-related concern is that if the calandria tube sags too much,it could interfere with the reactivity control units as well. In addition, the material for the calandria tube has not been approved for a reactor pressure boundary by the American Society of Mechanical Engineers (ASME) Code. Further,it should be noted that the calandria tube is not designed to tolerate the failure of a pressure tube, which is a reasonably foreseeable event, which has already occurred. The rupture of a pressure tube at Bruce 2 in 1985 resulted in rupture of the calandria tube, spilling fuelinto the calandria tank. It should also be noted that a pressure tube failure in Pickering 2 in 1983 resulted in failure of the annulus bellows, spilling reactor coolant down the face of the reactor. The garter spring that acts as the annulus spacer between the pressure tube and the calandria tube has been shown to embrittle after over 20 years of radiation exposure in a test reactor. If the springs break, they would nolonger function properly in maintaining separation of the pressure tube and the calandria tube, leading to the possible failure scenario described above. VI
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The ends of the fuel channels are capped by end fittings made from martensitic stainless steels. 'Ihe material used for the end fittings does not meet the toughness requirements (i.e., lateral expansion) of the I ASME Code. The adequacy of the material has not been demonstrated by fracture mechanics principles, using the 1/4T (thickness) flaw required by Appendix G of 10 CFR 50. However, AECL has demonstrated by test that a factor of safety of at least three exists for the assumed flaw of Appendix G. The issue of the fracture toughness of the end fittings needs to be examined further. The end fitting also contains a removable closure plug, that allows the fueling machine to obtain access to the pressure tube for refueling. At various times during the refueling cycle, both the closure plug and the safety lock of the fueling machine become parts of the reactor coolant pressure boundary. However, these components are not approved by the ASME Code. In addition, the fueling machine is supported by a movable support which is not covered by the ASME Code. Movement of the fueling machine supports while the machine islatched to a fuel channel assembly could damage the end fittings. Failures of the end fittings and fueling machine are considered to be bounded by failures of the pressure tubes. The pressure tubes are connected to the end fittings by a roll expanded joint. Roll expanded joints are not allowed by the ASME Code for Class I applications. However, AECL has designed these components to ASME Section III and has performed prototype testing to demonstrate the adequacy of the joints. Early I joints experienced delayed hydride cracking that led to leaks. The cause of the cracking was an incorrect rolling procedure that resulted in high residual stresses. A revised rolling procedure and the change to Zr-2.5% Nb pressure tubes has effectively elimmated the cracking at the rolled joints. Even though many of the CANDU components are not covered by the ASME rules, the components have been designed to ASME Section III wherever possible. CANDU reactors have a large experience base of safe operation. The problems identified here are worst case scenarios, but are well within the realm of possibilities for the CANDU reactor design. A significant surveillance and inspection program will be required to ensure the safe operation of these reactors. I Still to be to decided is whether the pressure tube design should be held to the same high standards as the light water reactor design on which current regulations are based. The basis for not holding the pressure tubes to the same standards as a light water reactor vessel is that the consequences of failure are significantly lower for the pressure tube failure than the consequences of failure of the reactor pressure boundary of light water reactors. That decision will require review of safety analyses, which is well j beyond the scope of this review. vu
The necessary requirements for inspection and surveillanca of the CANDU 3 reactor design will also have to be developed, particularly with regard to hydrogen absorption (which can lead to delayed hydride cracking) and creep. The pressure tubes of a CANDU reactor have a finite life due to creep and/or hydrogen absorption. AECL recognizes that the life of the pressure tubes may be less than the desired life of the plant, and has designed the fuel channels to allow for their replacement when they reach the end oflife. A surveillance program will be necessary to ensure that the safety of the plant is not compromised as these components reach the end of life. This report should help in identifying the features of a surveillance program. It is clear that there are several major hurdles for AECL Technologies to overcome prior to licensing the CANDU 3 reactor design in the U.S. These include the use of: non-ASME materials, non-ASME components, and in-service degradation. The lack of ASME approval for using Zirconium alloys in Class 1 applications arises from the fact that they have never been evaluated on their technical merits. The adequacy of using martensitic stainless steel for the end fittings will have to be demonstrated since this material does not meet the lateral expansion requirements of the ASME Code. The rolled joints connecting the Zirconium alloy pressure tubes to the stainless steel end fittings are not allowed in Class I service per current code requirements. There are no LWR components equivalent to the non-ASME approved components associated with on-power refueling. Relevant in-service degradation mechanisms inclurie irradiation induced creep and growth, delayed hydride cracking (DHC), and stress corrosion cracking (SCC). Of these,only SCC is generally applicable to LWRs. The approval of the use of these materials in the intended CANDU 3 applications will require review by appropriate industry codes and standards groups. Alternatively, the NRC may need to sponsor research efforts to assure that these materials-related issues are satisfactorily resolved. I I I I v. I
I I Table A: Summary of Licensing R2 view Challenges Component Material Licensing Review Challenges I Pressure Tubes Zr-2.5wt% Nb
- Tube Sag Caused by Irradiation Induced Creep / Growth Irading to Contact with Moderator Tube and Delayed Hydride Cracking and Rupture (Failure Probability of 2%/ reactor-year)
- Roll Expanded Joints Connecting Pressure Tubes with End-Fittings Not Allowed by ASME Code for Class 1 Applications Which Includes Primary Heat Transport Systems I
- Fracture Resistance Reduced by Hydrogen Embrittlement
- Material Specification for Zr-2.Swt% Nb Not ASM
- Canadian Standards Do Not include Requirements for Impacs Testing or Provide a Given Level of Fracture Toughness
- Creep Degradation in Zr Alloys Not Considered by U.S. Codes, Standards, and Regulations End-Fittings Stainless Steel
- Fracture Toughness Not Adequately Demonstrated, e.g., the Martensitic Stainless Steels Do Not Meet ASME Code Toughness Requirements I
- Removable End-Fitting Closure Flug and Refueling Machine Safety Lock Not Approved by ASME Code Feeder Carbon Steel
- Unable to Verify That This HTS Component Meets ASME Tubes Requirements Garter InconelCoil
- Potential Failure Due to Radiation Induced Embrittlement After O Springs with Zr-2 Girdle 20 Years Imading to Contact Between Pressure and Calandria g Wire Tubes
- PotentialIncreased Corrosion if Moisture Leaks into Annular Cas System Calandria Zr-2
- Tube Sag Due to Creep May Interfere with Horizontal Reactivity Tubes Control Mechanisms
- Material Specification for Zr-2 Not ASME Approved
- Not Design to Tolerate Pressure Tube Rupture Reactivity Perforated Zr-2
- Potential Irradiation Induced Creep ard Growth Problem in Control or Zr-4 Horizontal Unit _.
Mechanisms
- Control Rod Guide Tubes Not Considered to be Class 1 Calandria Stainless Steel
- No Specific Material Concerns Identified I Shield Tank Carbon and Stainless Steels
- No Specific Material Concerns Identified Moderator Not Clearly *P; ecific Material Concerns Identified l Piping Specified l I
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I ; CONTEN15 ABSTRACT u: I EXECUTIVE
SUMMARY
v CONTENTS xi LIST OF TABLES xiv LIST OF FIGURES xiv ACRONYMS xv
1.0 INTRODUCTION
1 1.1 BackgroundInformation 1 1.2 Statement of Work 1 1.3 Review Process 1 1.4 Documents Reviewed 2 2.0
SUMMARY
OF MATERIALS 5 2.1 Overview of CANDU 3 Design 5 2.1.1 Fuel Bundles 6 2.1.2 Fuel Channels 6 I 2.13 Tube Sheets 6 2.1.4 End Fittings 9 2.1.5 FuelChannelInternalComponents 9 I 2.1.6 Fueling Machine 2.1.7 Heat Transport System Piping 14 14 2.1.8 Calandria Vessel 14 2.1.9 Moderator Piping 19 2.1.10 Shield Tank 19 2.1.11 Shield Water Piping 19 , 2.2 Component Materials 21 23 SpecialMaterialJunctions 21 2.3.1 Welding and Brazing 21 2.3.2 Rolled Joints 24 2.3.3 Interference Fits 26 2.3.4 Bolted Flanged Joints 26 2.3.5 Other MechanicalJoints 26 2.4 Material Environmerits 27 2.5 Generic CANDU Materials Challenges 28 I a I
E 2.5.1 Creep of Zirconium Alloys 30 2.5.2 Corrosion of Zirconium Alloys 34 1.93 Delayed Hydride Crackiag in Zirconium Alloys 36 2.5.4 Rolled Joints 37 2.5.5 Inspection and Leak Detection 40 2.5.6 Other Materialsissues 41 2.6 Operating Experience of CANDU Reactors 41 2.6.1 Experimental Reactors 42 2.6.2 Soviet Reactors 43 2.63 CANDU Reactors 43 2.7 Design Features Unique to CANDU 3 45 2.8 Future Development 46 3.0 ACCEPTANCECRITERIAREVIEW 49 3.1 US. Nuclear Regulatory Commission Acceptance Criteria 49 3.1.1 Reactor Coolant Pressure Boundary Materials 54 3.1.1.1 MaterialSpecifications 55 3.1.1.2 Reactor Coolant Pressure Boundary Materials 55 3.l.13 Fabrication and Processing of Ferritic Materials 56 3].1.4 Fabrication and Processing cf Austenitic Stainless Steels 57 3.1.1.5 Nondestructive Examination and Testing 57 3.1.1.6 Material Surveillance 58 3.1.1.7 Reactor Vessel Fasteners 58 3.1.2 Reactivity Control Mechanism Materials 58 3.13 Reactor Intemals and Core Support Materials 59 g
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3.2 Canadian Standards 59 33 U.S. NRC and Canadian Acceptance Criteria Differences 63 4.0 TECHNICAL REVIEW 65 4.1 Fuel Channel Assemblies 66 4.1.1 PressureTubes 67 4.1.2 Rolled Joints 70 4.13 End-Fittings 71 4.1.4 Annulus Spacer 76 4.1.5 Calandria Tubes 77 = 4.1.6 Bellows 78 X11
I 'I 4.2 Feeder Piping 79 4.3 Reactivity ControlMechanisms 79 4.4 Reactor Vessels, Support Structures, and Associated Piping 79 4.5 Fueling Machine 80 ; i 5.0 DISCUSSION OF ASME CODE APPLICATION TO CANDU 3 81 5.1 Non-ASME Materials 81 ; 5.2 Non-ASME Components 82 5.3 Material Degradation 82 ; 5.4 Recommendations 83
6.0 REFERENCES
85 I I I ! I I i I l i I l 1 I ! ,I
E LIST OF TABLES TABLE PAGE A Summary of Licensing Review Challenges ix 2-1 Summary of Materials for CANDU Reactor Internals 22 2-2 Water Chemistry Requirements for Primary Heat Transport System 29 2-3 Water Chemistry Requirements for Moderator System 29 2-4 Electrochemical Potential of Various Metals (Galvanic Series) 39 g 2-5 Comparison of CANDU 3 Design With Other CANDU Plants 45 5, 3-1 U.S. General Design Criteria 51 l 4-1 Required Room Temperature Tensile Properties 77 LIST OF FIGURES I FIGURE PAGE 2-1 CANDU 37 Element Fuel Bundle 7 2-2 Pressure Tube to Calandne Tube Annulus Spacer 8 2-3 Section Through End Shield and Calandria 10 2-4 Calandria Tube and ExtensionJoint 11 2-5 CANDU 3 Fuel Channel Hardware 12 2-6 Channelinlet Hardware 13 2-7 Channel Outlet Hardware 15 2-8 Channel Closure Plug 16 17 2-9 Sequence of Events During Refueling g 2-10 Overall View of CANDU Reactor 18 5 2-11 Front View of CANDU Reactor 20 2-12 Irradiation Growth of Cold Worked / Stress Relieved Zr-2.5% Nb 31 2-13 Effects of Creep and Growth on CANDU Reactor Components 32 2-14 Routes of Hydrogen Ingress into the Pressure Tube 35 4-1 Charpy V-Notch Impact Properties of AISI 403 Stainless Steel 73 4-2 Comparison of Fracture Toughness Data for 403 Stainless Steel with K Curve 74 xiv I
F I ACRONYMS LIST OF ABBREVIATIONS AE Acoustic Emission AECL Atomic Energy of Canada, Ltd. AISI American Iron and SteelInstitute ASME American Society of Mechanical Engineers ASMEIII Section III of ASME Boiler and Pressure Vessel Code ASTM American Society for Testing and Materials BLIP Blister and Spacer Location Inspection with PIPE CANDU Canadian Deuterium Uranium CFR Code of Federal Regulations CIGAR ChannelInspection and Gauging Apparatus for Reactors CSA Canadian Standards Association DO2 Deuterium (Heavy Water) DDE Design Basis Earthquake DHC Delayed Hydride Cracking I FSAR Final Safety Analysis Report GDC General Design Criteria HO2 Light Water HTS Heat Transport System ISI In-Service Inspection LOCA Loss of Coolant Accident LWR Light Water Reactor NDE Nondestructive Examination NRC U.S. Nuclear Regulatory Commission OBE Operating Base Earthquake PHTS Primary Heat Transport System PIP PeriodicInspection Program PIPE Packaged Inspection Probe PSAR Prelmunary Safety Analysis Report PT-EFPY Pressure Tube - Effective Full Power Year QA Quality Assurance SAR Safety Analysis Report SCC Stress Corrosion Cracking PBESSURE TU8E
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I ,g B surrounded by a second cylinder called the shield tank, which is constructed of carbon steel. The ends of the shield tank are a second set of stainless steel tube sheets through which the fuel channel must pass, as shown in Figure 2-3. The Zircaloy calandria tube stops after penetrating the inboard tube sheet, but the pressure boundary is continued through to the outer tube sheet by means of a stainless steel calandria tube extension that is attached to the calandria tube. The calandria tube is rolled into the calandria tube extension under a stainless steel insert as shown in Figure 2-4. The calandria tube extension is contained within a stainless steel lattice tube that connects the inner and outer tube sheet. At this point, there are three sets of concentric tubes (pressure tube, calandria tube extension, and lattice tube). 2.1.4 End Fittings At each end of the reactor, the fuel channelis terminated by a stainless steel end fitting. On the inlet end, the end fitting is attached to the lattice tube by bellows to accommodate axial grow 6 due to thermal expansion and creep. The inlet piping for the heat transport system is attached axially to the inlet end fitting by a flanged connection. he outlet end fitting is much more complex, as it accommodates the fueling machine that connects to the fuel channel to allow refueling while the reactor is online. Because the fueling machine requires axial access to the outlet end fitting, the feeder piping for the heat transport system exits the outlet end fitting at a right angle. Figure 2-5 shows the complete fuel channel assembly, including internals. 2.1.5 Fuel Channel Internal Components The intemal components provide several functions. They provide shielding for the ends of the I reactor, and allow for refueling. Figure 2-6 shows the shield plug and the fuel pusher that are contained on the inlet end of the fuel channel. The central section of the shield plug is tapered at both ends to allow flow smoothly around it, but the central section provides radiation shielding for the inis end of the reactor. In other CANDU designs, refueling is performed from both ends, where one machine pushes the new fuel in one end, and the new fuel pushes the used fuel out the other end, where it is received by the other fueling machine. In the CANDU 3 design, refueling is performed only from the outlet end. The fuel pusher shown in Figure 2-6 has grooves and channels that result in a flow restriction that provides the driving force to push the fuel bundles downstream. Refueling can only be accomplished when there is flow in the channel to push the fuel out. I 1 I 9 i i
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Figure 2-7 shows the internal components on the outlet end of the fuel channel. There is a shield plug similar to the inlet shield plug, except that it has slots in the outer sections. The slots allow flow to the outlet piping (which exits at 90 ) even when the shield plug is being removed to allow for refueling. The latched spacer plug is removed during refueling to make room for the fuel bundles. At the outboard end of the outlet end fitting is a removable closure plug, which has a flexible seal disc that bears against a replaceable seal ring insert in the end fitting. The closure plug has a three-jaw latching arrangement that is actuated by the fueling machine (Figure 2-8). Dunng the refueling operation, the fueling machine releases the closure plug latch mechanism and removes the closure plug, storing it in the magazine of the fueling machine. The latched spacer plug is similarly removed and stored in the fueling machine magazine during refueling. Figure 2-9 illustrates the sequence of events during a refueling cycle. 2.1.6 Fueling Machine The fueling machine consists of four main assemblies: the snout assembly, the magazine assembly, the separators, and the ram as;embly. The snout assembly is the mechanism that attaches the fueling machine to the end fitting. The magazine assembly stores the replacement fuel, closure plug, latched g spacer, and used fuel during the refueling cycle. The separator assembly is used to hold the internal E components and fuel in place while the ram assemb!y removes components from the fuel channel and places them in the magazine. The snout assembly and the magazine are part of the reactor pressure boundary during refueling. 2.1.7 Heat Transport System Piping The heat transport system piping consists of an array of small diameter carbon steel pipes, as shown in Figure 2-10. The feeder pipes are bolted to flanges on the inlet and outlet end fittings. The feeder piping is installed as four separate assemblies consisting of a header with 116 feeders attached (232 feeders on each end). The piping from t!.e headers to and from the steam generators is also carbon steel. 2.1.8 Calandria Vessel As described previously, the fuel channels are arranged horizontally within a cylinder called the calandria. The calandria is filled with heavy water that acts as the moderator. The calandria consists 14 I,
1 l l l (a) OUTLET SHIELD PLUG L i a 1 I a (b) LATCHED SPACER PLUG i .1 Figure 2-7 Channel Outlet liardware"' 15
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I I of a main shell welded to an annular plate at each end that connects to smaller diameter sub shells at each end, as shown in Figure 2-3. The calandria sub shells are welded to the inboard tube sheet. The annular plate provides flexibility at each end of the calandria to accommodate differential thermal expansion between the shell and the calandria tubes. A short vertical cylindrical vessel is set onto and welded to the calandria shell. This calandria extension is filled with helium moderator cover gas to control the deuterium gas concentration and to regulate the pressure in the moderator system. I 2.1.9 Moderator Piping I No information was provided on the moderator system piping. However, the statement is made that "Austenitic stainless steel is used for all moderator system components in contact with heavy water.'" 2.1.10 Shield Tank I The calandria is surrounded by a second cylinder called the shield tank, which is constructed of carbon steel. The shield tank is filled with light water (H O). 2 The outer tubesheet serves as the end of the shield tank. The space between the inner and outer tubesheet is filled with carbon steel balls and light water is circulated to act as shielding for the ends of the reactor. I The shield tank has a flat top that is joined to the top of the calandria extension, and is flexibly joined to a vertical shield tank assembly similar to the calandria tube extension. At the top of the shield tank is the reactivity mechanism deck, which supports the actuating mecharusms for the vertical reactivity control units. The vertical reactivity control unit thimbles extend down from the reactivity mechanism deck to the bottom plate of the shield tank extension, as shown in Figure 2-3. Horizontal reactivity control unit thimbles are welded to the calandria shell and attached to the shield tank shell with bellows. No information is provided on the material for the bellows. The reactivity control units can be seen in the views of the overall reactor assembly shown in Figures 2-10 and 2-11. 2.1.11 Shield Water Piping i No information was provided on the shield water piping. Because it carries light water, it is likely that the shield water system piping is made from carbon steel. I 'I 19 I
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I 2.2 Component Materials l Because the CANDU reactors use natural uranium as the fuel and heavy water as the moderator, the in-core components of the reactor must be made of a material with low neutron capture cross section. I The additional requirements of corrosion resistance and strength necessitated the use of Zirconium alloys. However, Zirconium alloys undergo changes in mechanical properties, chemistry and dimensions. It was therefore concluded that the reactor core components would have limited life, and the reactor was designed to allow for easy replacement of fuel channels. I Table 2-1 lists all of the material specifications for the components described above. As noted above, the components of the fuel channels are made from Zirconium alloys. 'Ihe pressure tubes were fabricated from Zircaloy 2 in the early CANDU reactors, but have been changed to Zr-2.5% Nb because of its improved corrosion properties. The calandria tubes remain Zircaloy 2, however. Most of the components that are in contact with heavy water are made from Zirconium alloys or stainless steel, however the feeder piping in the primary heat transport system is made from carbon steel. The material for the moderator piping is not specified in the available documentation. The specific issues relating to the suitability of the materials are described in the following sections on material junctions, material environments, and generic CANDU meterials problems. 2.3 Special Material Junctions I There are several types of joints between the various CANDU reactor internal components, including welding, brazing, rolled joints, interference fits, and bolted flanges. Each of the different joining I processes will be described below, with examples of the applications of each joining process, and discussion of the problems associated with each. I 2.3.1 Welding and Brazing Zirconium alloys are very weldable, however, they have a high affinity for contaminants, such as nitrogen, hydrogen, and oxygen, which results in brittle weldments. As a result, welds are usually made in a controlled environment, such as argon or helium gas, to prevent contamination.A However, zirconium can be resistance welded with case." Resistance welding is the technique used I to join the Zircaloy fuel sheaths to the end cap. No shielding is required for these resistance welds. The calandria tubes are fabricated from Zircaloy strips that have been rolled into a tube and seam I 21 I
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Table 2-1: Summary of Materials for CANDU Reactor Internals COMPONENT MATERIAL FUEL CHANNEL ASSEMBLIES Pressure Tubes Zirconium-2.5 wt% Niobium Gas Annulus Components End Fitting Bearings AISI Type A2 Tool Steel Pressure Tube-Calandria Tube Spacer Coil Inconel X 750 (AMS 5698D) Girdle Wire Cold-Drawn Zr-2 Channel Annulus Bellows Inconel 600 Channel Annulus Bellows End Piece Type 304 Stainless Steel Feeder Connections Flange Cr-Mo Steel (ASTM 541 or S42) Hub Carbon Steel (ASME SA 105) Seal Ring AISI Type 410 Stainless Steel Capscrews SA 193 Gr. B7 End Fittings , Body Type 403 Stainless Steel Liner Tubes Type 410 Stainless Steel , Bellows Attachment Ring Type 304 Stainless Steel Closure Seal Insert Ring Custom 455 Hardened Stainless Steel (Gold Plated) Calandria Tube Assemblies Calandria Tube Zircaloy 2 - Annealed Rolled Joint Insert StainlessSteel Calandria Tube Extension Type XM-19 (Nitronic 50) per SA 312 Fuel Channel Intemals Channel Closure Assembly 17-4 PH Stainless Steel Shield Plugs Type 410 Stainless Steel Fuel Pusher Zircaloy and Stainless Steel Latched Spacer Plug Assembly Type 410 Stainless Steel Latch Mechanism 17-4 PH 410 Stainless Steel 22 s-
I , , CALANDRIA AND SHIELD TANK ASSEMBLIES Calandria Shell Welded Stain! css Steel End Shield Assembly Shell and Tube Sheets Stainless Steel Shielding Carbon Steel Balls Lattice Tubes StainlessSteel Shield Tank Assembly Shield Tank Bolted and Welded Carbon Steel Upper Shield Tank Extension Carbon Steel Lower Shield Tank Extension Stainless Steel Thimbles StainlessSteel Flexible Connection StainlessSteel MODERATOR AND SHIELD COOLING SYSTEM PIPING Moderator -- Sideld - REACTIVITY CONTROL MECHANISMS Control Rod Guide Tubes Zircaloy, Perforated to 35% Void Control Rod Sheaths Stainless Steel HEAT TRANSPORT SYSTEM Feeders ASTM A106 Grade B Headers ASTM A106 Grade B , 'g Piping ASTM A106 Grade B 3 Steam Generrtor Tubes Incoloy 800 FUELING MACHINE , Snout Assembly 17-4 PH Stainless Steel Snout Plug StainlessSteel Magazine Inconel 600 FUEL BUNDLES Fuel Uranium Dioxide Cladding Zircoloy 4 I . n .
F E. welded. ne CSA N285.6.4 Standard requires that the " Welding of the tubes or strip shall be performed using an automatic or semiautomatic process. Atmospheric control during welding shall be such that neither weld nor parent metal is contaminated by impurities." Drazing is used on the fuel sheaths to provide a spacer between adjacent fuel elements within a fuel bundle. Although sound brazed joints can be made, they typically form brittle intermetallic compounds. As a result, it has been concluded that "It does not appear likely that any corrosion-resistant brazed joint can be developed which is strong and tough, since all materials which might be useable appear to have been tested with negative results"
- The concern here is primarily the ability to create a strong, tough joint. However, the brazing used in the fuel elements is not a structural joint; it simply provides a buildup of material that holds the fuel elements apart. The strength and toughness may not be as much of a conccrn as its corrosion resistance. The brazing material used in the CANDU fuel elements reportedly contains Beryllium. A Zr-5 Be filler metal tested in 360 C water showed considerably better corrosion resistance than othen zirconium alloy filler metals.M Welding is also used to join the bellows to the calandria tubesheet. He bellows are made from inconel 600 and the tubesheet is stainless steel. Le outboard end of the bellows is attached to the end fitt ng, which is made from martensitic stainless steel (type 403). Martensitic stainless steels are g difficult to weld without cracking. As a result, the outboard end of the bellows is welded to an 5 austenitic stainless steel ring (type 304) which is A unk fit on the end fitting (see the discussion on interference fits below). There are no significant problems associated with welding Inconel 600 to type 304 stainless steel.
Zirconium alloys and steels can not be welded satisfactorily due to the formation of brittle intermetallic compoimds. For this reason, the joints between the components made from zirconium alloys and the stainless steel components are made using mechanical (rolled) connections. 2.3.2 Rolled joints Because of the difficulty in joining zirconium alloys and steels, roll expanded joints are used rather than weldments for the connections between the pressure tube and end fitting, and between the calandria tube and the reactor end shield. Roll expanded joints have a very significant operational experience base in boilers, heat exchangers, and condensers. ney are accepted and approved practice per the standards of the Tubular Exchanger Manufacturer's Association (TEMA) for tubular heat exchangers and for use in power boilers per Section I of the ASME Boiler and Pressure Vessel Code 24
I ' l (Power Boilers). However, they are not approved for the primary pressure boundary of the reactor coolant in nuclear reactors (Discussed further in Section 5). The connection between the pressure tube and the end fitting and the connection between the calandria tube and the lattice tube are both rolled joints. To make the joint between the pressure tube and end fitting, the end fitting is heated to a maximum of 427'C, the pressure tube is inserted and both components are allowed to cool, resulting in an interference fit. The tube is then roll expanded into the end fitting. No Code or Standard defines criteria for the expanding operation. Compressive residual stresses induced during the expansion process hold the pressure tube against the end fitting hub, sealing the joint, and providing pull-out resistance. However, the pressure tube material is susceptible to creep and there is the potential that this preload will relax. Therefore, the end fitting hub is designed with grooves. During the expansion, the tube is deformed plastically, with metal flowing into the grooves in the end fitting. The grooves provide sealing and resistance to pull out loads in the connection. The loss of preload due to creep does not effect the pull-out resistance. During the rolling operation, the thickness of the pressure tube is reduced 13.5 % 11.5 % locally. I Residual stresses of the order of 55 MPa (8 ksi) exist in the pressure tube. AECL feels that the level of residual stress is maintained low enough that it should not contribute to delayed cracking. Several tube failurer, have occurred adjacent to the roll transition region. It was concluded that high hoop residual stressca resulting from rolling the tube too far inboard contributed to delayed hydride cracking. Cracks initiated only in those tubes where the residual stress exceeded 550 MPa (80 ksi). The rolling procedure has been modified to prevent this type of occurrence. I The pressure tube /end fitting joint is not covered by the ASME Code. However, the joint design is subjected to a finite element stress analysis, to meet the requirements of ASME fection III, Subsection NB-3200, and is qualified by prototype testing, using the same tooling design and procedures as the production joints. In addition, each production joint is required to be inspected and leak tested using helium. The pressure tube /end fitting joint is not the only rolled joint in the fuel channel. The calandria tube / lattice tube connection, fuel channel annulus bellows / lattice tube connection, and the end fitting / liner tube connection are all rolled joints. They are developed, designed, and tested using the same philosophy as described above for the pressure tube /end fitting joint. I n I
E e E 2.3.3 Interference Fits W As noted above, a shrink fit is part of the rolled joint. However, there is another location where a shrink fit is the only means of connecting two components. As noted above, the bellows are welded to the calandria tube sheet on one end. The other end is welded to a stainless steel ring which is shrunk on the end fitting. No details of the shrink fit were provided. 2.3.4 Bohed Flanged Joints All of the primary heat transport system feeder piping is attached to the end fittings by means of bolted Grayloc connections. Grayloc is a proprietary joint design involving a solid metal gasket. The joint is designed to CSA N285.2, while the individual components meet the requirements of ASME Section Ill, Class 1. The feeder piping flange is made from Cr-Mo steel, attached to a carbon steel hub. The gasket is a martensitic stainless steel (type 410) and the joint is made up with SA193, Grade B7 capscrews. 23.5 Other Mechanical Joints The fueling machine acts as part of the reactor coolant pressure boundary during refueling. The fuelir<g machine snout assembly attaches to the end fitting by means of tapered wedges that are driven in behind the flange on the end fitting, forcing the end fitting face against a metallic ring which acts as a seal. 'Ihe fueling machine is said to be fabricated from 17-4 PH stainless steel, although it is not clear that all components within the fueling machine are also made from 17-4 PH. Precipitation hardened stainless steels tend to be susceptible to stress corrosion cracking and crevice corrosion in certain environments. Fasteners, including nuts, bolts, retaining rings, and spring rings, are also highly susceptible to stress corrosion cracking. The mechanical components that act to make up the connection between the fueling machine and the end fitting may therefore be susceptible to crevice corrosion or stress corrosion cracking. These components, however, are accessible for inspection when the fueling machine is not in use. There are no components in light water reactors that perform a function similar to the mechanical closure on the snout assembly of the fueling machine. As a result, the ASME has not considered this type of closure, and it is therefore not allowed by the ASME rules. There are no criteria for equivalent components that could easily be adapted for use on the fueling machine /end fitting closues. An inspection and preventative maintenance program would clearly be required to ensure the continued integrity of the connection. 26
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I When the fueling machine is not attached, the outlet end fitting is sealed by a removable closure plug. The closure plug and latch assembly are also made from 17-4 PH stainless steel and may therefore be susceptible to crevice corrosion and stress corrosion cracking. Although the channel closures could be removed for detailed inspection when the channel is drained of fluid, there is no other means of routine inspection of the mechanisms of the closures, unless the fueling machine was programmed to remove a closure plug and reinstall a different closure plug after refueling. The closure plug provides a function similar to the snout assembly on the fueling machine described above, in that it is a removable mechanical connection that is part of the reactor coolant pressure boundary. There are no components in light water reactors that perform a similar function. This type of closure is not allowed by the ASME code, and no criteria exist at the present that could be easily adapted for use on either type of closure. However, the criteria that are developed for the fueling machine snout assembly should be applicable to the closure plug as well. 2.4 Meterial Environments There are basically six different types of environments to which the various parts of the reactor may f be exposed. These are: Primary Heat Transoort System Heavy water (D2O) is pumped through the primary heat transport system which is designed to operate at a temperature of 316*C and a pressure of 10.7 MPa (1560 psi). Annulus Gas System Dry CO, gas is pumped through the annulus between the pressure tube and the calandria tube at an operating temperature is 250 C and a pressure of 103 kPa (15 psig). 1 I hioderator Svstem The heavy water reactor moderator in the calandria tank and its associated piping operates at a design temperature of 120 C (maximum operating temperature of 71*C) and a design pressure of 1.2 MPa (175 psi). I Helium Cover Cas Svstem A helium cover gas is maintains over the free surface of the moderator in the calandria extension to prevent corrosion and to provide pressure control in the calandria under various operating conditions. I Shield Tank The shield tank is filled with light water, with a peak temperature of 69'C and a design pressure of 1.1 MPa (160 psi). I 27 I <
E. as < Ambient All of the reactor external components (e.g., end fittings, feeder piping, end shields) are exposed to the ambient atmosphere within the containment building. Dissolved oxygen in water is a major variable in aqueous corrosion. Hydrogen is added to the heat transport system fluid to suppress the production of radiolytic oxidizing species in the core. This is known as hydrogen water chemistry. Of course, in heavy water, deuterium is added instead of hydrogen, but hydrogen and deuterium are often lumped together under the label " hydrogen". Sufficient hydrogen is added to maintain the oxygen level around 5 ppb. The water chemistry requirements for the primary heat transport system are shown in Table 2-2, and the water chemistry requirements for the moderator system are shown in Table 2-3. In addition to being in contact with potentially corrosive fluids, many of the components are exposed to high radiation fields. Radiation affects different materials in different ways, but the effects may include embrittlement, hardening, swelling, and creep. The zirconium alloys, the 300 series stainless steels and ferritic stects are adversely affected by radiation. Most of the components in the high radiation field in the core of the reactor are zirconium, although there are stainless steel components on the end shields, end fittings, and the primary heat transport system piping is carbon steel. These components will be exposed to lesser amounts of radiation, but radiation effects should be addressed in the design. 2.5 Generic CANDU Materials Challenges From a materials point of view, the most unique features of the CANDU reactors are the use of the Zirconium alloys as structural materials and the use of rolled joints in the reactor coolant pressure boundary. From a review of the literature, there appear to be three basic primary concems relating to the degradation of Zirconium alloys: creep, corrosion, and delayed hydride cracking. Rolled joints present their own unique materials problems, such as crevice corrosion, stress corrosion cracking, and in the case of zirconium alloys, hydride cracking. Each of these materials issues will be discussed in detail below. The AECL document "The Technology of CANDU Fuel Channels ** provides an excellent discussion of creep in Section 8 and corrosion and hydrogen ingress in Section 9. Although much of the discussion below comes from that document, independent data from the literature were td to confirm the conclusions stated in the AECL document. l l 28 I
I Table 2-2: Water Chemistry Requirements for Primary Heat Transport System I PARAMETER PERMISSIBLE RANGE DESIRED VALUE pHg (at 25'C) 10.2 - 10.8 2 10.2 Lithium as Li (mg Li/kg D 2O) 0.35 - 1.4 2 0.35 as LiOH (mg Li/kg D2 0) 1.2 - 4.8 2 1.2 Dissolved Deuterium as mg (mg D2 /kg D,0) 0.5-4.0 0 5 - 1.6 as mL (mg D2 /kg D 2O at NTP) 3 - 25 3 - 10 Dissolved Oxygen , as mg (mg O2 /kg D 2O) < 0.01 0.005 as mL (mg O2 /kg D 2O at NTP) < 0.007 0.00 Chloride (mg/kg D2 0) < 0.2 0.05 Suspended Solids (mg C1/kg D2 0) < 1.0 0.1 l Conductivity (mS/m at 25*C) 0.9-3.6 2 0.9
- In heavy water, the true unit of measure of acidity / alkalinity is pD, but what is measured ic an apparent pH, because a pH. meter calibrated with light water is used to monitor heavy water.
The relationships between the pH. reading and actual values of pH and pD are given tsy: pH. = pH + 0.46 and pH. = pD - 0.41. Table 2-3: Water Chemistry Requirements for Moderator System PARAMETER PERMISSIBLE DESIRED l RANGE VALUE phi (at 25'C) 4.5-6.5 < 7.0 Chloride (mg C1/D2 0) < 0.2 < 0.1 Conductivity (mS/m at 25'C) < 0.15 < 0.2 Boron (mg B/kg D 2O) < 10 Gadolinium (mg Gd/kg D 2O) <1
- see the note for Table 2-2
" after shutdown system No. 2 has operated I
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i E1 2.5.1 Creep of Zirconium Alloys Creep is a time-dependent phenomenon where the material undergoes irreversible plastic strain. There are actually three distinct mechanisms that result in time-dependent strains in the Zircaloy pressure tubes and calandria tubes. Two of them are actual creep phenomena (irradiation creep and thermal creep), while the third is an irradiation induced shape change, which is usually called growth. Irradiation -rowth results from the fact that the hexagonal closed packed crystal structure of Zirconiu Jetorms anisotropically when subjected to neutron irradiation, even when the material is not stressed. In other words, the crystal structure changes shape when irradiated, but the dimensional changes are different in d.tferent directions relative to the crystal structure. The volume of the crystal remains constant, so when there is growth in one direction, there must be shrinkage in one or more of the other directions. This can be seen from the data in Figure 2-12. It is difficult to physically separate the three mechanisms from field measurements of changes in dimension. For example, in order to achieve irradiation creep, the material must be irradiated, and it will also undergo dimensional changes due to irradiation growth. However, it appears that irradiation creep is the dominant mechanism of these two, as the net creep is positive in both the axial and transverse directions. If irradiation growth were dominant, then creep in one direction would have the opposite sign of creep in the other direction (see Figure 2-12). It is possible to break out thermal creep separately from irradiation related dimensional changes, but I it must be recognized that the thermal creep rate of irradiated material will not be the same as the thermal creep rate of unitradiated material. However, thermal creep does not normally become a significant fraction of the total creep rate at the range of operating temperatures (300*C). Dimension changes due to creep and growth in the pressure tube will be manifested in three ways: either as sag in the pressure tube (axial creep or elongation), an increase in the tube diameter (circumferential creep), or deformation of the end shield (Figure 2-13 shows a schematic representation of the effects of creep and growth) The primary concern is that the pressure tube could sag to the point that it will contact the calandria tube. This results in direct heat transfer to the calandria tube, which may cause boiling in the moderator. This is not a life-limiting concern, but results in operational problems. A secondary concem is that if the pressure tubes becomes too bowed, the movement of fuel will be restricted. This is an operational life-limiting concern, but not a structural life-limiting concern. 30
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I : i 1 i I i i Figure 212 Irradiation Growth of Cold Worked / Stress Relieved Zr-2.5% Nb'" ! f I 31 I
I NOT To SCALE REACTOR VESSEL (CALANDRIA) FIXED END SHIELD FREE END SHIELD 1 l 3 i PRESSURE TUBE I i
. CALANDRIA TUBE , ; ,
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{ I l; l Figure 213 Effects of Creep and Growth on CANDU Reactor Components
- i 32 l
The calandria tube is also subject to creep, but because it operates at a temperature lower than the pressure tubes, the thermal creep component will be smaller. Because the calandria tubes do not have an expansion capability like the pressure tubes, the effect of creep or growth in the calandria tubes is an outward bowing of the end shields (see Figure 2-13). Creep of the calandria tube also has an influence on the creep of the pressure tube. The pressure tube is supported at four locations along its length by the garter spring annulus spacers. These in turn are supported by the calandria tube. I Thus, sag of the calandria tube affects the supports for the pressure tube, thereby affecting the sag of the pressure tube. This will not have any effect on the possibility of pressure tube /calandria tube contact, but it may have an influence on the bowing of the pressure tube, which restricts the free movement of fuel, as described previously. If the calandria tube sags too much, it could interfere with the reactivity control units as well. AECL has developed a database of measurements of in-service curvature and sag measurements as a means of validating the analytical models of creep and growth. The models predict an elongation of 150 to 200 mm in the pressure tube, with diameter changes of as much as 3.5% 1he calandria tube I may sag 70 to 80 mm, which may result in interference with the horizontal reactivity control mechanisms. The measurements on pressure tubes have shown that the strain rates are linearly dependent on flux and stress, and are also strongly dependent on the orientation of the pressure tube relative to the flow. The back end of the pressure tube (the part that exited the extrusion press last) has a transverse strain rate that is 25% higher than the front end under the same operational conditions. It has therefore been concluded that the back end of the extrusion should be installed at the inlet end of the fuel channel to minimize the amount of creep strain. This is opposite current practice, which was based on the assumption that fracture tougnness was lower at the back end, and that it would be beneficial to I install the back end of the extrusion at the outlet end of the fuel channel (presumably because the higher operating temperature at the outlet end would offset the lower toughness). One aspect of creep that is not addressed in the AECL documents is the onset of tertiary creep. Most materials experience three phases of creep, ranging from the initial high rate of strain to shake out the high stresses (primary creep), through a steady state creep regime (secondary creep) to a region of accelerating creep ratesleading to failure (tertiary creep). Tertiary creep also typically involves the formation and coalescence of microvoids. The onset of tertiary creep in zirconium alloys typically occurs around 1.2% strain,"" which is less than the anticipated creep over the life of the pressure tube. I This may be mitigated by the fact that the onset of terti:ry creep is typically delayed by irradiation. 33 I
The impact of tertiary creep, including the possible interaction of creep voids with other mechanisms, such as corrosion and hydride cracking, needs to be addressed by AECL 2.5.2 Corrosion of Zirconium Alloys I The pressure tubes contain high temperature, high pressure heavv water, and are surrounded by CO2 gas on the outside. Thus, both surfaces are subject to oxidation (corrosion). The hydq;ca that results from the corrosion reaction is absorbed by the pressure tube." Figure 2-14 shows the possible routes of hydrogen ingress into the pressure tubes at the rolled joint. Although corrosion leads to metalloss and thinning, the life of the pressure tubes is probably controlled by hydrogen absorption rather than the corrosion rate.(") This is because Zirconium hydrides will precipitate out if the solubility limit of hydrogen is exceeded. These hydrides decrease the fracture resistance significantly (see the following section). 'Ihe metal loss due to corrosion is taken into account by specifying a corrosion allowance. One of the reasons that Zirconium alloys were selected for use in the pressure tubes is that they tend to be fairly corrosion resistant. A tight adherent Zirconium oxide layer is formed that protects the underlying metal. However, after extended exposure, the oxide layer exhibits a phenomenon called
" breakaway" corrosiono2> where the corrosion rate increases significantly. The rate of hydrogen pickup above this transition point may be as high as five times the pre-transition rate in Zircaloy 2.""
However, the rate of hydrogen pickup in Zr-2.5% Nb material is relatively constant over time. Although the pre-transition rate in Zr-2.5% Nb is higher than the rate for Zircaloy 2, the post-transition rate is much lower. As the lifetime is dominated by post-transition behavior, Zr-2.5% Nb is the better alloy for this application because it will not absorb as much hydrogen as the Zircaloy 2 material.(") The rate of hydrogen absorption in Zr-2.5% Nb pressure tubes is estimated to be 1 ppm l per year up to 3000 effective full power days, possibly increasing to as much as 4 ppm per year thereafter. This is 20 times less than that for Zircaloy 2 over the same period of time.c23 i i l Another factor that affects the rate of hydrogen pickup in Zirconium is contact with other materials. Tests have shown that Inconel acts as a " window" for hydrogen and local accelerated oxidation exists at locations where Zircaloy, Inconel, and water coexist."') The coil wire for the annulus spacer is i inconel X-750, and is in direct contact with Zr-2.5% Nb and Zircaloy 2 tubes. The potential for increased corrosion exists if moisture ever gets into the annulus as the result of a leak. Presumably l In heavy water reactors, deuterium is the hydrogen isotope that is absorbed by the pressure tube. Throughout the remamder of this discussion. the terms " hydrogen" and " hydrides" will refer to absorption of the deuterium isotope as well as hydrogen. M
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the leaking tube would be replaced, but the annulus gas system has common " pigtails" jo' ming one fuel channel to the next. The possibility of water ingress from a leaking tube to a non-leaking %be should be examined. The effect of radiation on corrosion of Zirconium alloys has also been examined. For Zircaloy 2, the corrosion rate is increased during irradiation only if significant amounts of oxygen are present?) In contrast, irradiation induces an aging phenomenon in Zr-2.5% Nb that reduces the corrosion rate slightly?" However, Zr-2.5% Nb is sensitive to oxygen content, regardless of the level ofirradiation?) The CANDU 3 reactors use additions of deuterium to the coolant (hydrogen water chemistry) to i maintain the oxygen level around 5 ppb. At these levels, the rate of hydrogen absorption should be minimal. Another variable that has a significant effect on the rate of corrosion of the pressure tubes is temperature. A 50 C increase in temperature, results in a factor of two increase in the corrosion rate. There is a temperature gradient of approximately 50 C along the length of the pressure tube,0* which would imply a factor of two variation in the corrosion rate along the length of a tube. A factor of two increase in oxide thickness has in fact been observed along the length of Zr-2.5% Nb pressure tubes removed from CANDU reactors?) There also is the potential for corrosion on the outside surface of the pressure tube. The annulus between the pressure tube and the calandria tube is filled with dry CO2 gas (nitrogen was used in earlier reactors). There has been at least one case of hydrogen ingress from the outside, but that was in an older reactor that used nitrogen gas?) Zirconium oxide provides a good barrier to hydrogen, so it has been concluded that an oxidizing environment is necessary on the outside of the tubes to develop an oxide layer as a barrier to hydrogen. A revised CO 2gas specification is being implemented to ensure that oxidation of the outside surface is continuous, but at a low rate. 2.5.3 Delayed Hydride Cracking in Zirconium Alloys In the previous section it was shown that zirconium alloys have an affinity for hydrogen and it was noted that the fracture resistance may be reduced significantly. This phenomenon, called delayed hydride cracking (DHC), will be discussed in detail in this section. I 36
r 'I I The solubility of hydrogen in zirconium is a strong function of temperature. At high temperatures (300*C), the solubility of hydrogen is 60 times that at room temperature."5) Thus, the hydrogen that exists in solutior at the operating temperature may precipitate out as a zirconium hydride during a shutdown. The hydrides are brittle and may fracture upon subsequent straining. The fractured hydrides then accelerate fracture by microvoid fonnation and coalescence.") The fact that the material is embrittled during a shutdown but may not fail until the subsequent startup diminishes the confidence in the inservice inspection program. Zirconium hydrides precipitate out as plate!-te ne normal orientation of the platelets is in the axial-I circumferential plane (i.e., parallel to the pipe inner and outer surfaces). In this orientation, they have negligible effect on the strength of the tube. However, if the hoop stress exceeds a threshold value during hydride precipitation, the platelets will form ir: the radial-axial plane (in other words, in the plane normal to the highest stress due to internal pressure). As a result, the hydride platelets have a significant adverse effect on fracture. For Zr-2.5% Nb, this threshold value is about twice the normal operating stress, while for Zircaloy 2, the threshold value is very close to the operating stress.") AECL has expressed the threshold value in terms of the stress intensity factor, K, rather than the level i of applied stress. A threshold value of K si = 4.5 to 7.0 MPa-m2 (4.2 to 6.5 ksi-in /2) is repcrted.m I However, this approach requires the assumption that a crack exists. It is difficult to postulate a crack size, and therefore to calculate a stress intensity factor, without better knowledge of the crack initiation mechanism. The threshold value Km is insensitive to temperature, hydrogen concentration and neutron fluence. If the stress intensity factor exceeds K , crack growth proceeds at a rate which is independent of K, but varies significantly with temperature and neutron fluence. AECL relies on the concept of leak before break to ensure that unstable rupture does not occur. There have been several failures of CANDU pressure tubes, and most of these involved tubes that leaked but did not rupture. However, I at least one of the failures ruptured without a leak. His will be discussed in more detail in a later section. I 2.5.4 Rolled Joints The previous discussion on the rolled joints discussed the pull-out strength of the joints. The joints are designed to ASME Section III by analysis, and prototype testing is performed. The pull-out strength is adequately covered by the design approach, and will not be discussed further here. 37
E Instead, we will discuss other materials related issues that may be present in the rolled joints. Rolled joints are typically not used in the reactor pressure vessels of light water reactors, but they are used in the steam generators. Experience with cracking in steam generators indicates the types of problems that could occur in the pressure tube rolled joints. "Ihese include crevice corrosion, stress corrosion cracking, and galvanic corrosion. In Table 2-3, we can see that LiOH is used to control the pH of the primary heat transport system fluid. Zirconium alloys have been shown to be susceptible to corrosion for lithium concentrations above 1 gram / liter?" The concentration shown in Table 2-3 is far below this, indicating that general corrosion is not likely. However, the same series of tests also showed that a significant crevice effect was present. The corrosion rate in a crevice between two samples of Zircaloy was ten times that of the control specimens. A similar effect, but to a lesser degree, was noted for Zircaloy in contact with stainless steel. Crevice corrosion is not explicitly addressed in the AECL reports. Zirconium alloys resist stress corrosion cracking (SCC) in many environments. This may be due to its high repassivation rate, where a break in the protective oxide is quickly healed if sufficient oxygen is present?' The threshold stress for SCC for zirconium alloys is typically high enough that SCC resistance can be achieved by stress relief. It is not anticipated that stress corrosion cracking will be a problem in the zirconiam alloys. 3 Galvanic corrosion is the result of an electrochemical cell set up between two dissimilar materials in an aqueous environment. The probability of galvanic corrosion in a material depends on its electrochemical potential. Table 2-4 shows the electrochemical potential for a number of metals in seawater?' Note that zirconium is near the top of the list. Materials near the top of the list are considered noble and act as the cathode in a galvanic cell. Materials lower on the list act as the anode, and are therefore subject to weight loss due to galvanic corrosion. Because zirconium is relatively noble, it is not susceptible to galvanic corrosion. However, zirconium is subject to hydrogen embrittlement, as noted previously, when it is the cathode of an electrochemical cell. Note that stainless steels are close to zirconium on the galvanic series in Table 2-4. This means that the electrical potential is small between zimonium and stainless steel, provided that the stainless steel is passivated. Also note that the stainless steels drop way down the galvanic series when they are active (Table 2-4). This means that active stainless steels are subject to galvanic cor osion almost to the same level as carbon steel. Martensitic stainless steels (such as types 403 and 410) and precipitation hardened stainless steels (such as 17-4 PH) are much more difficult to passivate than austenitic stainless steels (such as 304) and are therefore subject to galvanic corrosion. These materials 38
,I t 5 Table 2-4: Electrochemical Potential of Vadous Metals (Galvanic Series) Cathodic (Noble) Platinum Gold i Graphite Titanium Silver Zirconium Type 316,317 Stainless Steels (Pas 9ve) Type 3Gi Stainless Steel (Passive) Type 410 Stamless Steel (Passive) Nickel (Passive) Silver Solder f I Copper-Nickel (70-30) Bronzes Copper Brasses Nickel (Active) ; Tin , lead Type 316,317 Stainless Steels (Active) Type 304 Stainless Steel (Active) I Cast Iron Steel or Iron Aluminum Alloy 2024 Cadmium Aluminum Alloy 1100 Zinc Magnesium and Magnesium Alloys Anodic (Active) 39 I
E also tend to be more susceptibie to crevice co-rosion and stress corrosion cracking than the austenitic stainless steels. There are a number of parts of the fuel channel, including the end fittings, closure plug, and end fitting liner tubes that are made from these types of stainless steels. The corrosion behavior of these materials needs to be better characterized. 2.5.5 Inspection and Leak Detection he restricted access to the inside of the pressure tubes presents some obvious problems for l inspection. A number of pieces of equipment have been developed to perform these inspections. The g primary piece of equipment used to inspect the pressure tubes in the periodic inservice inspections 3 is known as CIGAR (ChannelInspection and Gauging Apparatus for Reactors). The CIGAR system is ha., sed with the fueling machine, and performs the pressure tube inspection with the reactor shut down, but the fuel channel being examined is still full of water. The CIGAR system is capable of?
- Flaw detection using 45' shear wave ultrasonic inspection (volumetric)
I
- Ultrasonic diameter measurements for determuung growth and ovality
- Ultrasonic wall thickness measurements Measurements of sag calculated by integration from inclinometer measurements g
- Anr alus spacer location using a send-differential-receive eddy current coil 3
- Suri tce profilometry using a special head CIGAR is the primary tool used for the periodic inservice inspections required by the Canadian Code.
I l There is another type of inspection, called the in-service inspection (ISI) that is used to provide specific l information about known or suspected problems, such as locating the annulus spacers, detection of l cracks in the rolled joint region, and detection of blisters. For example, the prcfilometry head on the CIGAR system is used for ISI inspections only. Other inspection systems have been developed for ISI inspections. Dry channel gauging equipment was originally developed for inspection of pressure tube deformation. The CIGAR system allows the same information to be obtained without draining the fuel channel. This reduces the radiation exposure for the inspection personnel. As a result, the dry channel gauging equipment is now used only to measure deformation of calandria tubes when the pressure tube has been removed The Packaged Inspection Probe (PIPE) is used to perform ultrasonic inspection of pressure tube rolled joints. The Blister and Spacer Location Inspection with PIPE (known as BLIP) uses eddy current techniques and ultrasonics to find blisters and inspect for cracks. 40
Several techniques have been used to attempt to measure the deformation of the end shields resulting from elongation of the calandria tubes. None of the techniques have produced conclusive results. The predicted deformation of the end shields is small, which may explain some of the problems associated with these measurements. I The annulus gas system is the primary leak detection system for the pressure tubes. The moisture content of the gas is constantly monitored. The air in the reactor vault is circulated through driers and the amount of moisture is monitored. An increase in the moisture content in the annulus gas I system or in the driers indicates a leak. Analysis of the chemical, radiochemical, and isotopic content of the collected water indicates which system is leaking (e.g., PHTS or moderator). Acoustic emission (AE) equipment is used to locate the specific leaking component. Low frequency airbome noise is indicative of closure plug leaks. Higher frequency noise, detected when the AE sensor is pressed against the end fitting, is used to identify leaks in the feeder connection or the pressure tube rolled joint. This system can not always isolate the leak to a specific channel, but can isolate it to a group of channels. Several other techniques have been attempted, but none can isolate the specific leaking channel. One pressure tube has ruptured after a leak was detected, but before the specific tube could be isolated (the tube actually failed during repressurization to utilize the AE I equipment). Improvements may need to be made in the area of leaking tube identification. 2.5.6 Other Materials Issues There are several other potential materials related issues that are not adequately addressed in the AECL documentation. These include: a wear in the pressure tubes due to fuel movement
= corrosion of the carbon steel feeder piping I =
=
dissimilar metal welds (Inconel 600 bellows to 304 stainless steel attachment ring) radiation damage to the carbon steel piping and stainless steel components outside the core 2.6 Operating Experience of CANDU Reactors l I Unlike a new reactor design, the CANDU reactors have a large experience base from which to draw (; conclusions regarding the suitability of materials. The CANDU 3 design is a new design, but relies i very heavily on the design of previous CANDU models. The differentes between CANDU 3 and other CANDU reactors will be discussed in the section following this one. 41
E a Over 70 reactors have been built around the world using zirconium pressure tubes. Nearly 30 reactors have been built using the basic CANDU design. These include approximately 20 in Canada, while the remainder are located in Pakistan, India, Argentina, and Korea. Although the first few CANDU plants were built with Zircaloy 2 pressure tubes, the remainder were built with Zr-2.5% Nb pressure tubes. In addition to the CANDU design, nearly 40 plants have been built with other designs that utilize pressure tubes made from zirconium alloys. These include the PRTR, CVTR and N-Reactors built in the U.S., as well as reactors in the United Kingdom, Switzerland, West Germany, Japan, India, and the Soviet Union. Most of these reactors were built with Zircaloy 2 pressure tubes, except for those in the Soviet Union. The Soviet Union has significant experience with Zr-2.5% Nb pressure g tubes (it was data provided by the Soviet Union that prompted the Canadians to consider the switch a to Zr-2.5% Nb tubes). The majority of pressure tube reactors are graphite moderated light water reactors. Only the CANDU and Hanford N-Reactors utilize horizontal pressure tubes; the others use vertical tubes. Pressure tube failures have occurred in most types of reactors. Some resulted from operation outside I design limits, some from material defects, and some from the degradation of the tube material. In addition, four reactors were re.noved from service because it was determined that the pressure tube material had degraded to a point that it was no longer suitable for service, and two reactors were g removed from service because of dimensional changes (creep). The four reactors that were removed 5 from service for material degradation used Zircaloy 2 pressure tubes, while the two reactors that experienced creep problems used Zr-2.5% Nb pressure tubes. 2.6.1 Experimental Reactors Serious pressure tube failures occurred in the 1960s in two small experimental reactors: the Plutonium Recycle Test Reactor (PRTR) at Hanford and the Lucens reactor in Switzerland. The failure of the PRTR reactor was the result of a test involving a fuel pin with an intentional defect. The pressure tube overheated due to oxidation inside the fuel pin and the system pressure blew a hole in the pressure tube. The failure of the Lucens reactor was the result of flow blockage cause by the re-entrant design of the pressure tube. The fuel overheated, melted and ignited, causing rupture of the pressure tube, the calandria tube, the calandria tank, and caused the other calandria tubes to collapse onto the pressure tubes. Neither reactor has been operated subsequent to the pressure tube failures. 42 I
I 2.6.2 Soviet Reactors The Soviet Union has sixteen production RBMK reactors utihzmg Zr-2.5% Nb pressure tubes. The Russians have reported 26 pressure tube leaks and one tube rupture. The cracks were due predominately to delayed hydride cracking facilitated by high roller straightening stresses. In one reactor, the stress relief step was omitted on the control rod tubes, and nearly all of the control rod tubes had to be replaced due to delayed hydride cracking. The life of the RBMK pressure tubes is limited by diametral creep. If the tubes expands too much, they can not be easily removed. 2.6.3 CANDU Reactors Three CANDU reactors have developed tube leaks, two of them more than once, and one CANDU reactor has experienced an in-service tube rupture. The cracks occurred predominately at the end of the roll joint transition, and were determined to be caused by incorrect rolling practice that resulted in high hoop residual stresses. Cracks formed in the axial-radial plane due to delayed hydride cracking between installation and start-up, and continued to propagate during shutdowns. Subsequent inspection revealed many other tubes with similar cracks that had not yet leaked. All of these incidents involved Zr-2.5% Nb pressure tubes. I In one instance, the leaking tube could not be positively identified, except that it came from a group of 11 tubes. To facilitate identification of the leaking channel by acoustic emission, the tubes were repressurized, and the tube ruptured, rupturing the calandria tube and spilling fuel elements into the moderator. The failure was due to a lap-type fabrication defect that is difficult to detect with ultrasonic testing. . Examination of fabrication records indicated that 6 tubes with similar defects had been installed in reactors and two others were found that had not yet been installed. The tube rupture occurred at the Pickering 1 plant in 1983. The pressure tubs in Pickering I were .I made from Zircaloy 2. During installation, one of the annulus spacers shifted, allowing the pressure tube to sag and contact the calandria tube. The resulting thermal gradient allowed hydrogen to migrate to the coldest point and form blisters. The blisters eventually tumed to cracks as a result of volume expansion of the hydrides, and the cracks propagated by delayed hydride cracking. The other surface of the pressure tube remained hot, which meant that the hydrogen would dissolve out. As a result, the crack did not propagate in the through-wall direction, but continued to grow axially until it reached a critical size and the tube ruptured. 43 ) I '
I. AECL has concluded from their review of pressure tube failure experience
- that there is an experience base of approximately 267,000 pressure tube effective full power years (PT-EFPY). Of these CANDU represents approximately 25% and the Soviet RBMK reactors represent approximately 62%. One hundred five (105) tube leaks have been identified, and five pressure tube ruptures. The past pressure tube reliability worldwide has therefore been 4x10-* leaks /PT-EFPY and 2x104 ruptures /PT-EFPY.
CANDU production reactors have experienced 68,000 PT-EFPY of operation, with six pressure tube leaks, including two that ruptured. The reliability of CANDU production reactors is therefore 9x104 4 leaks /PT-EFPY and 3x10 ruptures /PT-EFPY. AECL concludes that this is consistent with failure rates for piping and pressure vessels and implies that this is acceptable. De failure rates are far below the target reliability of 104 per year used in the U.S. for reactor pressure vessels. However, this has to be weighed against the different consequences between a pressure tube failure and an LWR pressure vessel failure. AECL also notes that all of the failures in CANDU reactors occurred in early CANDU units (there have been no presure tube failures since 1986) and all were attributed to specific problems that have now been solved.m Although AECL should be given credit for improving the quality of the pressure tubes, their statement that the problems have been solved is an overstatenient. When a reactor has 232 pressure tubes, there is a real possibility that some out of tolerance condition (combination of out of tolerance conditions) will lead to a failure. For example, a failure rate of 104 still leads to a 2% probability of failure of a single tube per reactor-year. AECL has not directly addressed the reliability of the refueling system. There have been a number I of problems associated with fueling machine operation,* but most of these are related to problems with the hydraulic system for the snout clamp mechanism (in several cases, difficulties occurred in removing the fueling machine from the fuel channel end fitting) or the ball screw drive mechanism. However, there was at least one case where the bridge that supports the fueling machine shifted while the fueling machine was clamped to the end fitting. This incident resulted in leakage of 20,000 liters g of coolant. 5 There is also some discussion of leaks from improperly seated closure plugs.m Although these incidents are not true failures, they result in the leakage of coolant into the reactor building. Frorn a safety point of view, AECL considers this type of event as being bounded by a complete fuel channel failure. However, these types of events need to be considered in the overall reliability and risk analysis of the plant. The only quantitative information provided by AECL in this regard is that " station incapability due to fuel handling is traditionally less than 1%".* 44
r-I 2.7 Design Features Unique to CANDU 3 We have already identified a number of unique design features of the CANDU reactors, including the horizontal pressure tubes and on-power refueling. The CANDU 3 reactor design utilizes all of the features of the CANDU reactors, but with some subtle differences. The CANDU 3 reactor is designed to be a smaller version of the CANDU series. Table 2-5 compares the CANDU 3 design with other CANDU reactors.m Table 2-5: Comparison of CANDU 3 Design with Other CANDU Plants CANDU Net Number Number of Reactor Maximum Reactor Plant Electrical of Fuel Elements Outlet Channel Outlet Power Channels in Fuel Header Flow Header Output Bundle Pressure Quality (MW) (MPa) (kg/s) (%) Pickering A 515 390 28 8.7 23.0 0 Pickering B 516 380 28 8.7 23.0 0 Bruce A and B 850 480 37 9.1 24.0 0.7 CANDU 6 665 380 37 9.9 24.0 4 Darlington A 881 480 37 9.9 25.2 2 CANDU3 450 232 37 9.9 27.0 3.6 I Although the CANDU 3 design has the fewest number of fuel channels, it has almost the same net power output as Pickering A and B. The net power output per fuel channel is higher in the CANDU 3 design than in any of the previous designs. The basis for the higher power density is not clear, due to conflicting and incomplete data.M lt is not clear from the available documentation whether this higher power density is the result of increased neutron flux, but if that is the case, additional radiation exposure would be expected to have an adverse effect, primarily on the amount of creep, which appears to be the life-limiting mechanism in the CANDU 3 design. In addition to the overall size of the plant and fewer number of fuel channels, the CANDU 3 has one I other basic design difference. In all other CANDU reactors, refueling is performed using a fueling machine at each end of the fuel channel. One fueling machine pushes new fuel in and the other receives the old fuel. The CANDU 3 design utilizes a single fueling machine on the outlet end of the
.I 45 I
l E; i reactor. lhis reduces the overall size of the reactor building, as space for the fueling operations is only required on one end of the reactor. The use of one fueling machine has two implications on the safety of the plant. The first is that the I number of times a fueling machine attaches to a fuel channelis reduced by half. The risks associated with this activity (e.g., the channel closure does not seat properly after refueling or the fueling machine does not lock on the end fitting properly) are therefore reduced by a factor of two. However, the use of two fueling machines allowed one machine to remove the fuel in case the other machine is damaged. With a single fueling machine, fewer options are available in case of a malfunction during refueling. , There are other differences between CANDU 3 and the other CANDU designs, but most of these are due to the one-ended fueling and the reduced size of the plant. For example, the primary heat transport system piping uses a single loop, with flow moving in one direction through the reactor (to provide the motive force for fuel removal) whereas in the older CANDU designs, two loops were used that had flow in altemate directions in adjacent tubes. The bridge that supports the fueling l machine is floor supported rather than crane mounted as in other CANDU designs. But the basic reactor design is same as in the CANDU 6 design. At this point, there appears to be no plans to g change the basic fuel channel materials, despite an ongoing research program involving attemate 5 materials. 2.8 Future Development AECL is continuing to develop the technical basis for the CANDU reactors. A review of the fabrication process of the pressure tubes was recently completed to identify trace materials that can ' reduce the fracture resistance of the tubes, and a series of manufacturing innovations and controls that minimize the adverse effects of the trace elements has been developed."7) Analytical and empirical programs have evaluated the basis for safety in the CANDU reactors."8) Research is ongoing in improving the properties of the CANDU pressure tubes. The primary areas of interest are creep / growth and sensitivity to delayed hydride cracking. Among the research being done are the following activities:
- A program is being run by the CANDU Owners' Group to produc e Zr-2.5% Nb tubes that are cold worked to 40% rather than the current range of 24-3(W. as a n eans of reducing the creep elongation.m 46 I
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- The Owner's Group is also funding research to develop techniques for measuring the hydrogen concentration in the pressure tubes without taking metal scrapings. Electrical ,
resistivity and ultrasonic shear wave techniques are being investigated.m
= 1mproved inspection capabilities, including measurement of the gap between the calandria tube and pressure tube, eddy current flaw detection, and underwater video inspections, are being developed for the CIGAR system.A
- Yttrium sinks are being considered as a means of preventing deuterium accumulation in the I pressure tubes (Yttrium has a greater affinity for deuterium than zirconium alloys).m ,
a The forging practices are being reviewed to determine methods of mmmuzmg tle variability in fracture toughness.A
- Consideration of altemate materials has led to the development of a new alloy called " Excel".
This material has similar strength properties to Zr-2.5% Nb, but has in-reactor creep and growth rates about 1/3 the amount for Zr-2.5% Nb. Excel has a composition of Zr - 3.5% Sn - 0.8% Nb - 0.8% Mo." I I I I I I ; I 47 1 I
I I 3.0 ACCEPTANCE CRITERIA REVIEW The U.S. Nuclear Regulatory Comnussion's regulations for licensing of commercial power plants are contained primarily in 10 CFR Parts 50 and 100, and were developed for light water reactors (LWRs), as that is the only type of commercial reactor built in the linited States. These address safety concerns of LWRs and frequently refer to LWRs, or the two types of LWRs, boiling water reactors (BWRs) and pressurized water reactors (PWRs). Because of differences in design, CANDU reactors have different safety-related characteristics than LWRs. Phenomena that are safety concems in LWRs may not occur or be safety concerns in CANDU reactors, and are therefore not reflected in the design Codes and I Standards in Canada. Similarly, some CANDU safety issues may not be appropriate for LWRs. This section of the report attempts to bridge the gaps between U.S. and Canadian Codes, Standards, and Regulations relating +o materials issues in reactor intemals. Previous reports" have addressed the differences in the regulatory approach in the two countries, but have focussed primarily on system design issues and accident analysis. Although this report will provide a brief overview of the overall regulatory processes in each of the two countries, the primary focus will be on those aspects relating directly to materials issues of reactor intemals. I 3.1 U.S. Nuclear Regulatory Commission Acceptance Criteria In the United States, the Nuclear Regulatory Commission has the responsibility of implementing and enforcing the requirements in Title 10 of the Code of Federal Regulations (CFR) to ensure the safety of nuclear reactors. The key requirements are delineated in Parts 50 and 100. Part 100 relates to site evaluation criteria, and will therefore be excluded from the scope of this review. Section 50.34 requires that all applicants for construction licenses for nuclear reactors must submit a preliminary safety analysis report (PSAR) to the NRC, and applicants for a license to operate a nuclear reactor must submit a final safety analysis report (FSAR). Although Section 50.55a requires that systems and I components of BWRs and PWRs meet the requirements of the ASME Code Sections III and XI, there is no such requirement for other types of reactors. The application for the construction license must include the principal design cri+eria for the proposed facility in the design application. Appendix A to 10 CFR 50 contains a list of General Design Criteria (GDC) which establish the minimum requirements for the design criteria of water-cooled nuclear I power plants similar in design and location to plants previously licensed by the NRC. Although the General Design Criteria are defined strictly for light water reactors of the type built in the United States, the criteria are considered to be generally applicable to other types of nuclear power plants, 49 I
E and are intended to provide guidance in establishing the general design criteria for those other type of plants. Table 3-1 contains a list of the General Design Criteria defined in 10 CFR 50 Appendix A. Only I General Design Criteria 1,4,10,14,15,30,31, and 32 are considered applicable to this review. Other criteria, while important to the design of a nuclear reactor, are considered outside the scope of this review either because they relate to systems design, such as safety features, or relate specifically to components other than the reactor internals. General Design Criterion 1 requires that structures, systems, and components important to safety be designed, fabricated, erected and tested to quality standards commensurate with the importance of the safety functions to be performed, and requires the establishment of a quality assurance program. Criterion 4 further requires that these structures, systems, and components be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents. Criterion 30 similarly specifies the use of the highest quality standards practical for components of the reactor coolant pressure boundary. Appendix B to 10 CFR 50 provides the nccessary criteria for a quality assurance program. Criterion 10 requires that the reactor core, coolant, control and protection systems be designed with appropriate margin to assure that fuel design limits are not exceeded during any anticipated operational condition. Criterion 15 imposes similar requirements c,n the reactor coolant system to assure that the design conditions of the reactor coolant pressure boundary are not exceeded. Criterion 14 requires that the reactor coolant pressure boundary be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture. The terms " extremely low" and " abnormal" are not defined. General Design Criteria 30,31, and 32 get more specific. Criterion 30 requires a leak detection system capable of locating the source of reacto coolant to the extent practical. Criterion 32 requires that the design permit pdoCC inspection, leak testing and a surveillance program. Criterion 31 requires that the material of the reactor coolant pressure boundary behaves in a non-brittle manner and that the probability of a rapidly propagating fracture is nunmuzed. Criterion 31 further requires consideration of service temperatures and uncertainties in the material properties, effects of irradiation, residual, steady state and transient stresses, and the size of flaws. I
, I a
I. I I Table 3-1: U.S. General Design Criteria I. OVERALL REQUIREMENTS Criterion 1 - Quality Standards and Records Criterion 2 - Design Bases for Protection Against Natural Phenomena Criterion 3 - Fire Protection Criterion 4 - Environmental and Dynamic Effects Design Bases Criterion 5 - Sharing of Structures, Systems, and Components I II. PROTECTION BY MULTIPLE FISSION PRODUCT BARRIERS Criterion 10 - Reactor Design I Criterion 11 - Reactor Inherent Protection Criterion 12 - Suppression of Reactor Power Osculations Criterion 13 - Instrumentation and Control Criterion 14 - Reactor Coolant Pressure Boundary Criterion 15 - Reactor Coolant System Design Criterion 16 - Containment Design Criterion 17 - Electric Power Systems Criterion 18 - Inspection and Testing of Electric Power Systems Criterion 19 - Control Room III. PROTECTION AND REACTIVnY CONTROL SYSTEMS Criterion 20 - Protection System Functions I Criterion 21 - Protection System Reliability and Testing Criterion 22 - Protection System Independence Criterion 23 - Protection System Failure Modes Criterion 24 - Separation of Protection and Control Systems Criterion 25 - Protection System Requirements for Reactivity Control Malfunctions Criterion 26 - Reactivity Control System Redundancy and Capability Criterion 27 - Combined Reactivity Control Systems Capability Criterion 28 - Reactivity Limits Criterion 29 - Protection Against Anticipated Operational Occurrences I I 51 I
E IV. FLUID SYSTEMS Criterion 30 - Quality of Reactor Coolant Boundary Criterion 31 - Fracture Prevention of Reactor Coolant Pressure Boundary Criterion 32 - Inspection of Reactor Coolant Pressure Boundary Criterion 33 - Reactor Coolant Makeup Criterion 34 - Residual Heat Removal Criterion 35 - Emergency Core Cooling Criterion 36 - Inspection of Emergency Core Cooling System Criterion 37 - Testing of Emergency Core Cooling System Criterion 38 - Containment Heat Removal Criterion 39 - Inspection of Containment Heat Removal System Criterion 40 - Testing of Containment Heat Removal System Criterion 41 - Containment Atmosphere Cleanup , a Criterion 42 - Inspection of Containment Atmosphere Cleanup System b Criterion 43 - Testing of Containment Atmosphere Cleanup System Criterion 44 - Cooling Water Criterion 45 - Inspection of Cooling Water System Criterion 46 - Testing of Cooling Water System V. REACTOR CONTAINMENT Criterion 50 - Containment Design Basis Criterion 51 - Fracture Prevention of Containment Pressure Boundary Criterion 52 - Capability for Containment Leakage Rate Testing Criterion 53 - Provisions for Containment Testing and Inspection Criterion 54 - Piping Systems Penetra v , Contamment Criterion 55 - Reactor Coolant Presse e Boundary Penetrating Containment Criterion 56 - Primary Containment Isolation Criterion 57 - Closed System Iso'ation Valves VI. FUEL AND RADIOACTP/ITY CONTROL Criterion 60 - Control of Releases of Radioactive Materials to the Environment Criterion 61 - Fuel Storage and Handling and Radioactivity Control Criterion 62 - Prevention of Criticality in Fuel Storage and Handling Criterion 63 - Monitoring Fuel and Waste Storage , Criterion M - Monitoring Radioactive Releases 52 i
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L [ In addition to the Genera ~. Design Criteria, the NRC also utilizes a dccument called the Standard Review Plan" to guide them in their review of the safety analysis reports provided with the licensing request. The purpose of the Standard Review Plan (SRP) is to assure the q.zality and uniformity of [ NRC license application reviews and to provide a well defined basis for performing those reviews. { Two aspects of the SRP should be noted relative to this review of the CANDU 3 reactor design. First, this review is not part of a licensing submittal, and we are therefore not reviewing a complete safety analysis report. Certain aspects of the SRP apply to licensing activities, such'as identification of quality groups and Code classes, and are not appropriate at this level of review. Other aspects are related to stress limits and Code acceptability. An insufficient level of detail has been provided at this time to attempt evaluation of stress limits, other than to examine where the effects of stress may exacerbate a materials problem. Those aspects of the SRP that are not covered by this review will, ( of course, become important during the actual license application review. Second, the SRP review of a plant feature may be performed on a generic basis if it is felt that the { design is sufficiently similar to previously approved plants. One of the basic aspects of the CANDU 3 approach is to provide a common design, which requires only muumal changes to adapt to site specific requirements. As a result, once the SRP review is performed of a key feature such as the pressure tubes, the review may not need to be repeated for subsequent plants, provided that the design does not change and the operational experience of existing pit.nts does not demonstrate en unforeseen problem. The specific requirements of a safety analysis report are delineated in the NRC Regulatory Guide 1.70," which provides a standstd format for SARs. The sections of the SRP are numbered to { match the sections of the standard fermct. "Ihe complete bst of SRP sections wu reviewed and the following sections were considered to be opplicable to this review: 3.2.2 System Quality Group Classification 3.9.3 ASME Code Class 1,2, and 3 Components, Component Supports, and Core Support Structures ( 3.9.4 Control Rod Drive Systems 3.9.5 Reactor Pressure Vessel Internals 4.5.1 Control Rod Drive Structural Materials { 4.5.2 Reactor Internal and Core Support Mr,terials 5.2.1.1 Compliance with the Codes and Standard Rule,10 CFR S 50.55a 5.2.1.2 Applicable Code Cases 53 {
E 5.23 Reactor Coolant Pressure Boundary Materias 5.2.4 Reactor Coolant Pressure Boundary Inscryice Inspection and Testing 5.2.5 Reactor Coolant Pressure Boundary leakage Detection 53.1 Reactor Vessel Materials 53.2 Pressure Temperature Limits 533 Reactor Vessel Integrity 17.1 Quality Assurance During the Design and Construction Phase It should be noted that several of the SRP sections listed above are related primarily to classification of component quality classes and loading combinations (e.g., 3.2.2, 3.93, 5.2.1.1, 5.2.2.2, and 17.1) rather than directly defining criteria for the various components. They are included here only to indicate that such criteria exist in the U.S. Regulations, and for subsequent comparison with Canadian Regulations. It should also be noted that the ASME Code "does not provide guidance in the selection of a specific classification to fit a component in a given system. Such guidance is derived from systems safety criteria for specific types of nuclear power systems, such as pressurized water reactm, boiling water reactors, or high temperature gas cooled reactors, and may be found in engineering standards or in g the requirements of regulatory and enforcement authorities having jurisdiction at the nuclear power 5 sitel'A For purposes of this review, we have divided the scope of the applicable SRP sections into three broad arers, based on the types of components involved: the reactor coolant pressure boundary, the reactivity control mechanisms, and the reactor internals (meloding core support materials). Each of these will be discussed below. 3.1.1 Reador Coolant Pressure Boundary Materials There are basically three components to the reactor cohnt pressure boundary: the reactor pressure I vessel (s), tne piping, and the primary side of the steam generator. The steam generator is outside the scur of this review. The NRC considers pressure boundary components such as the piping and steam generator separately from the reactor pressure vessel. The primary requirements for reactor coolant pressure boundary components are delineated in SRP Sections 5.23, 5.2.4, and 5.2.5. The criteria for the reactor pressure vessel are generally similar, but more rigorous, and are contained in SRP Sections 53.1,5 5.2, and 533. Five specific areas are common to the requirements of both the reactor coolant pressere boundary and the pressure vessel. These are: material specifications, 54 _I
compatibility of the materials with the reactor coolant, fabrication and processing of ferritic materials, fabrication and processing of austenitic stainless steel, and inspection / testing. Each of these are discussed below. 3.1.1.1 Material Specifications The material specifications for pressure retaining co nponents (including ferritic materials, nonferrous metals, austenitic stainless steels, and weld metals) are reviewed to determine the adequacy and suitability of the material for the relevant applications. The NRC review primarily consists of a determmation that the materials used in the design are identified in the ASME Code, Section III, Appendix I, or described in detail in the ASME Code, Section II, Parts A, B, and C. I The Zirconium alloys used in the CANDU 3 design are not currently included in Section III, Appendix I of the ASME Code. There are two approved grades of Zirconium pipe material in ASME Section II, Part B, but neither of the grades matches the UNS grades used for the pressure tubes or calandria tubes. Regulatory Cuide 1.85 describes the ASME Code Cases involving material specifications that l are acceptable to the NRC, but there are no approved Code Cases involving these alloys to date. Because the materials are not approved a priori, the SRP allows the NRC to evaluate the materials as I exceptions to the Code, provided the basis for the exce;ption is clearly identified. The fact that a s material is excluded from the ASME Code does not automatically exclude its approval by the NRC. Article IV-1000 of the ASME Code (Section III Appendices) provides a procedure, which is referenced in SRP Section 5.3.1, that describes the requirements for obtaining ASME Code approval of a material which has not been previously approved by the Code. It is the policy of the Code to approve only I material specifications that have been approved by the American Society for Testing and Materials (ASTM). If the material specification has not yet been approved by ASTM, the ASME Code will consider the new material only if the specification is before the ASTM for approval. In either case, the Code Committee will require info;mation on mechanical properties, weldability, and physical changes of the material, in sufficient detail to allow the definition of allowable stress values. Service experience is also useful tc the Code Committee. 3.1.1.2 Compatibility of Materials with the Reactor Coolant The water chemistry of the reactor coolant system, including additives, is reviewed to ensure that the I materials employed in the reactor coolant pressure boundary are compatible with the water I 55 I
O environment. The primary concerns delineated in the SRP are general corrosion of ferritic low alloy and carbon steels and stress corrosion cracking of sensitized austenitic stainless steels. For the ferritic materials, paragraph NB-3120 of the ASME Code, Section III addresses the required corrosion allowances. It also notes that the fatigue design curves of the ASME Code do not include tests in the presence of a corrosive environment which might accelerate fatigue (i.e., corrosion fatigue). Stress corrosion cracking (SCC) of austenitic stainless steels is addressed in the SRP through Regulatory Guide 1.44 and NUREG 0313. Control of SCC is indicated by proper material selection (avoiding sensitized microstructures), reducing stress levels, and reducing the oxygen level of the coolant. 3 Although not explicitly addressed in the SRP, it is also important to examine the compatibility of one material with another in the water environment. Galvanic corrosion can occur if the two materials have significantly different electrochemical potentials. Crevice corrosion is also possible at mechanical joints between components, where the local geometry restricts the flow of water, and a local corrosive environment is set up, typically resulting in some form of stress corrosion cracking. 3.1.1.3 Fabrication and Processing of Ferritic Materials I This part of SRP Section 5.2.3 discusses the fracture toughness, welding, and nondestructive examination criteria for ferritic materials in the reactor coolant pressure boundary. The fracture toughness requirements are delineated in Appendix G of 10 CFR 50, which is intended to ensure that the material will not failin a brittle manner. It should be noted that the criteria of Appendix G were developed for very thick walled pressure vessels (approximately 250 mm, or 10 inches thick) whereas the pressure tubes in the CANDU reactors are only 4.19 mm (0.164 in) thick. The Appendix G criteria may not be totally applicable. For the reactor pressure vessel, the criteria of 10 CFR 50 Appendix G are supplemented by Regulatory Guide 1.99, which addresses the effects of irradiation on the fracture toughness. The material composition, welding procedure, and projected radiation exposure are used to predict the shift in the nil ductility temperature and the resulting change in fractv.re toughness over time. This is used with SRP Section 5.3.2 to evaluate pressure-temperature *umits for the reactor pressure vessel material. The NRC also reviews the weld procedures to be used on low alloy steels to ensure that adequate precautions are taken to minimize weld defects. Compliance with ASME Code Sections III and IX is examined to ensure that high quality welds are made. The ASME Code provisions are supplemented with three Regulatory Guides that address specific aspects of welding low alloy steels. The criteria 56 a
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I in Regulatory Guide 1.50 ensure that cracks resulting from excess hydrogen, such as cold cracks and reheat cracks, are avoided. The criteria of Regulatory Guide 1.43 are used to limit the occurrence of underclad cracking in low alloy steel that is clad with stainless steel, and Regulatory Guide 1.71 criteria place controls on welds made in regions of limited accessibility to improve the quality of welds in those locations. 3.1.1.4 Fabrication and Processing of Austenitic Stainless Steels A similar review is performed for austenitic stainless steels, although the primary focus shifts from I brittle fracture to stress corrosion cracking. One of the key factors is the avoidance of a sensitized microstructure. As noted above, Regulatory Guide 1.44 and NUREG-0313 provide criteria to avoid stress corrosion cracking (it should be noted that NUREG-0313 is strictly only applicable to BWR components, as BWRs typically have a higher oxygen conter.t in the coolant, which leads to increased susceptibility of SCC). Additional controls on contamination cleaning, and an upper limit on yield strength are also examined to minimize the risk of SCC. For example, Regulatory Guide 1.37 provides I criteria for the quality of water used in cleaning or flushing the piping, and Regulatory Guide 1.36 provides criteria to establish compatibility of the austenitic material with its thermal insulation (if the insulation gets wet, due to a leak or other means, contaminants can leach out, resulting in SCC). I The welding procedures are also reviewed to ensure compliance with ASME Code Sections III and IX, and Regulatory Guides 1.31 and 1.71. Regulatory Guide 1.31 specifies controls on the delta ferrite content of austenitic stainless steel welds to prevent the occurrence of microfissures that may have an adverse effect on the integrity of the components. As noted above for ferritic welds, Regulatory Guide 1.71 places controls on welds made in regions of limited accessibility to improve the quality of welds in those locations. 3.1.1.5 Nondestructive Examination and Testing Section 5.2.4 of the SRP specifies that the components that are part of the reactor coolant pressure boundary must be designed to accommodate periodic inservice inspection, as detailed in ASME Section XI. Section 5.3.1 delineates inspection requirements for the reactor pressure vessel. I The procedures for performing nondestructive exammation (NDE) arr described in Section V of the ASME Code, and include radiographic, magnetic particle, liquid penetrant, and ultrasonic testing. I Additional techniques may be employed, provided that it is demonstrated that the inspection techniques are equivalent to or superior to the techniques approved in Section V. I 57
as In addition to the NDE requirements, Section 5.2.4 of the SRP requires a program for leak testing and hydrostatic pressure testing. Section 5.2.5 requires a means of detecting leakage inservice, and determining the source of the leak. The requirements for the leak detection system are delineated in Regulatory Guide 1.45. The preceding five sections of this report summarized the requirements conunon to both reactor coolant pressure boundary components and the reactor pressure vessel. The SRP provides additional r requirements unique to the reactor pressure vessel in Section 5.3.1 of the SRP. These are discussed below. 3.1.1.6 Material Surveillance In order to ensure that the strength and fracture toughness of the reactor pressure vessel material do I not degrade to a point that affects the safety of the plant, and to ensure compliance with GDC 32, SRP 5 Section 5.3.1 requires a surveillance program. If the peak end of life neutron fluence exceeds 10' n/cm', a surveillance program complying with ASTM Standard E185 (as modified by 10 CFR 50 Appendix H)is required. 3.1.1.7 Reactor Vessel Fasteners Bolts, washers and nuts that are used to hold the reactor vessel head may be susceptible to stress corrosion cracking. As a result, SRP Section 5.3.1 requires a review of bolting materials to determine their adequacy, including fracture toughness, and resistance to stress corrosion cracking. Specific requirements are given in 10 CFR 50 Appendix G and Regulatory Guide 1.65, which includes a list of suitable materials. The SA-193 Grade B-7 bolts specified for the pressure tube end fitting / feeder piping closure are approved materials. 3.1.2 Reactivity Control Mechanism Materials General Design Criteria 26 specifies that one of the reactivity control mechanisms must be control rods, and SRP Section 4.5.1 delineates the requirements for the control rod drive structural materials. 'The review areas are similar to those given in SRP Section 5.2.3 and described previously. The criteria for the review of materials specifications are similar to that of SRP Section 5.2.3, except that cold-worked austenitic stainless steels are limited to a yield strength no greater than 90,000 psi to reduce g the probability of stress corrosion cracking. Limits on the tempering temperature of martensitic 5 58
lI I stainless steels and the aging temperature of precipitation hardening stainless steels are also provided to minimize SCC. I 3.1.3 Reactor Internals and Core Support Materials I A similar review is also required for reactor internals and core support materials, as specified in SRP Section 4.5.2. The governing ASME criteria for core support structures are delineated in Section III, NG-2000. He review includes material specifications, welding controls, NDE, and special requirements for austenitic stainless steels to minimize stress corrosion cracking. The specific criteria are sinQ to those described previously. 3.2 Canadian Standards Nuclear reactors are regulated within Canada by the Atomic Energy Control Board (AECB). The AECB has issued regulations, known as Atomic Energy Control Regulations, which are equivalent to the Title 10 CFR regulations in the U.S. He Canadian approach to regulating nuclear reactors is completely different than in the U.S. W f mdamental principle in Canada is that the applicant / licensee bears the responsibility for safe'y, with the AECB primarily setting safety objectives I and performance requirements, and auditing t'ae licensee's performance. As an example of the difference in regulatory philosophy, the AECB licenses reactors for a period of six months to two years. Renewal of the license depends on the ability of the licensee to meet the safety objectives and performance requirements. I There are few formal regulatory requirements, and these are broad safety goals and objectives, rather , 1 than the prescriptive design and operational rules specified in 10 CFR 50 and the ASME Code. For ; example, Part II of the Atomic Energy Control Regulations, which pertains to " Nuclear Facilities" is l only two pages in length. The AECB has three other types of documents that delineate policy more explicitly. Generic License Conditions are " standard t.ets of conditions that are included in particular AECB licenses of a common type." Regulatory Policy Statements are " firm expressions that particular ' requirements' not expressed ) as Regulations or License Conditions be complied with or that any requirements be met in a particular manner but where AECB retains the discretion to allow deviations or to consider alternative means of attaining the same objectives where a satisfactory case is made." The Regulatory Policy Statements are issued to define safety requirements, but they do not have the force of law. Regulatory Guides >I 59 l
E a provide " guidance or advice on any aspect of the AECB's regulatory process that is given in a manner less rigid than that intended by Policy Statements." In addition, the AECB issues Consultative Documents which are, in effect, draft versions of Regulatory Policy Statements. Dere are currently only five Regulatory Policy Statements relating to reactor safety: R-7 Requirements for Containnient Systems for CANDU Nuclear Plants R-8 Requirements for Shutdown Systems for CANDU Nuclear Plants R-9 Requirements for Emergency Core Cooling Systems for CANDU Nuclear Plants R-10 The Use of Two Shutdown Systems in Reactors R-77 Overpressure Requirements for Primary Heat Transport Systems in CANDU Power Reactors Fit:ed with Two Shutdown Systems here are also currently several Consultr.cive Documents. The most relevant is C-6, " Requirements for the Safety Analysis of CANDU Nuclear Power Plants
- Among other things, C-6 requires an analysis of a massive failure of all pressure vessels, unless tne probability of such an event can be shown to be extremely low, and:
- the vessel is designed, fabricated, installed, and operated to ASME Section III Class 1
- the vessel connections are relatively few (reactor headers are not considered to be vessels for the safety analysis) e an in-place inservice inspection (ISI) program exists
- leak before break can be assured
- leaks can be detected to allow the operators to take action following the detection of a leak The Canadian Standards Association (CSA) produces the N285 series of Standards that roughly parallel the ASME Code. The general requirements for pressure retaining components in CANDU reactors are contained in the N285.0 standard, which defines the code classification of components, using an approach similar to the Class 1,2, and 3 components classifications of the ASME Code.
However, unlike the ASME Code, N285.0 defines the criteria for classifying a component into one of the classes. There are also differences between the Canadian and U.S. criteria for assignment of class boundaries.* The N285.0 standard also provides basic criteria for design, fabrication, inspection, and quality assurance of nuclear components. I e
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ne N285.1 standard covers Canadian Class 1,2, and 3 components, and imposes the requirements of the corresponding sections of the ASME Code (NB for Class 1, NC for Class 2, and ND for Class 3), except that quality assurance and inservice inspection are covered by the criteria in N285.0 rather than l the ASME Code. In addition, materials approved under Canadian Standard N285.6 may be used in addition to those approved by the ASME Code. I Because the CANDU design incorporates several features that are not adequately addressed by the ASME Code, the Canadian Standards also provide a supplemental set of Code classes for components that fall outside the boundaries of the ASME Code. The N285.0 standard defines Class IC,2C, or Class 3C components as those components that would be classified Class 1,2, or 3 if the ASME rules , adequately covered the CANDU design. These were formerly called Class Special, but this designation was changed as it implies that the "special" criteria are less rigid than the normal criteria. He use of non-ASME Code material does not necessarily mean that a component is classified Class IC,2C, or 3C. Canadian Standard N285.2 covers the Class IC,2C, and 3C components. Material selection criteria are the same as for Class 1,2, and 3 components. Specific requirements are given for the design, fabrication, inspection, examination, test, and installation of the following components: I
- Fuel Channel Assemblies
- Pressure Tube to End Fitting Joints
- Pressure Tubes
- Pressure Tube Supports (Annulus Spacers)
- Channel Closure
- Calandria Assembly
- Calandria Tube to Tubesheet Rolled Joint
- Calandria Tube
- Lattice Tube to Calandria Tubesheet Joint
- Lattice Tube to Fueling Tubesheet Joint
- Reactivity Control Units
- Control Rod Drive Housings
- Liquid Injection Noz21es
- Threaded Connections Joints Between Tubular Connections a Fuel Handling Equipment Elastomeric Hose Assemblies I 61 I
E O1
- Fuel Channel Closure Lock
- Fueling Machine Safety Lock
- Fueling Machine Supports
{ Most of these components are included as Class IC,2C, or 3C components because they provide a function that does not exist in LWRs, and the ASME Code therefore does not cover them. Control rod drive housings are typically covered by Subsection NCA-1271 of the ASME Code, because they are part of the reactor coolant boundary in LWRs. However, in the CANDU reactors, control rod guide housings are not part of the reactor coolant pressure boundary, and are therefore covered by Canadian rules rather than the ASME rules. 1 There are two types of inspection programs in Canada: the Periodic Inspection Program (PIP) and the In-Service Inspection (ISI) program. Canadian Standard 285.4 covers the periodic inspection requirements of the CANDU reactors. Le periodic inspection program evaluates a sample of the components of interest on a recurring basis. The sample size is tailored to the individual components based on the applied stress intensity, the fatigue usage factor, and the size of the postulated failure. The fundamental concept is that there are a large number of small components, and the failure of one component (e.g., the pressure tube) is considered to be a possible (and probable) event. The periodic E inspection program is intended to ensure the detection of generic defects rather than specific defects. 5 The consequences of a failure are mitigated by ensuring leak-before-break. ne second type of inspection is called an in-service Inspection (ISI) program, and is intended to I provide specific information about known or suspected problems, such as locating the annulus spacers, detection of cracks in the rolled joint region, and detection of blisters. The methods and techniques used for periodic inspection are required to meet ASME Section V, unless approval is obtained from the appropriate authorities. Supplemental inspection criteria are specified for the fuel channels (volumetric and dimensional examination) and the feeder piping (wall thickness checks). Material specifications for all of the non-ASME materials (i.e., those made from zirconium alloys) are given in the N285.6 series of standards. nese standards define the fabrication, inspection and material property requirements of the fuel channels (N285.6.1), control rod guide tubes (N285.6.2), liquid injection shutdown nozzles (N285.6.3), calandria tubes and reactivity control guide tubes (N285.6.4), and the fuel channel spacer wires (N285.6.5). Le N285.6 standards also provide the inspection criteria for zirconium alloys (N285.6.6), design data for zirconium alloys (N285.6.7) a 62 I m
I t materials specification for the martensitic stainless steel for tiie end fittings (N285.6.8), and a materials specification for component supports (N285.6.9). The material allowables of the martensitic stainless steel for the end fittings are derived directly from the ASME Code. Quality assurance (QA) requirements are defined in the N286 series of standards. These standards were not available for review at this time, however, the CANDU 3U Project Quality Assurance Manual was reviewed. 'Ihe CANDU 3U QA program is based on the requirements of Appendix B to 10 CFR 50. However, the program outlined in the QA Manual is minimal, and it is not clear that it complies with the requirements of 10 CFR 50. For example, there is no discussion of nonconforming items. Improvements in the QA program will be required for licensing in the U.S. 3.3 U.S. NRC and Canadian Acceptance Criteria Differences U.S. and Canadian regulatory positions are like apples and oranges. The fund unental philosophies are at extreme ends of the spectrum. One is very prescriptive, while the other provides only general guidelines, relying on the nuclear plant licensees to ensure safety. In the latter ca^.e, the burden is on the licensee to prove a safe design, whereas in the former, an extensive set of rales exists that may lead to the mistaken belief that simple compliance with all of the rules will ensure an adequate design. I Despite the regulatory differences, the implementation of the design process is very similar in both countries. 'Ihat is because the ASME Code, for all its deficiencies, is a very good model of how to design a nuclear power plant. As a result, the Canadians have not only used the ASME Code where it applies, they have designed a set of standards similar to the ASME Code where it does not apply. There are two reasons the ASME Code does not apply to CANDU reactors. The first is the use of non-ASME materials (e.g. Zirconium alloys for the pressure tubes and calandria tubes). The second is for unique design features that do not exist in light water reactors. The list of Class IC,2C, and 3C components above shows the design features unique to the CANDU design. Tne fundamental difference between U.S. and Canadian appr >ach is that the Canadian regulations do i not contain the detailed regulatory guidr.nce contained in 10 CFR 50 and the Regulatory Guides. l AECL has reviewed the General Design Criteria of 10 CFR 50 Appendix A and have concluded that "the safety objectives and the spirit of the GDC can be met by the CANDU 3."W However, they can l not point to specific criteria (such as the NRC Standard Review Plan or Regulatory Guides) that can be audited to ensure compliance. AECL concludes that their successful operating experience using this approach is an indication that the safay objectives of the GDC have been met. I 63 ! 1 I i
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I 4.0 TECHNICAL REVIEW In this section, each of the key reactor internal components will be examined, comparing the design criteria employed by AECL with the broad requirements of the General Design Criteria and the specific requirements of the Standard Review Plan and the Regulatory Guides. In addition, the performance of the materials will be examined to identify possible failure modes and damage mechanisms. This review is based not only on the information provided by AECL, but also on information available from the open literature. Some of the criteria outlined in the GDC are applicable to all components of the reactor design, especially the various parts that make up the fuel channel. Those criteria are oriented more toward system design than materials, but will be briefly examined here. Criterion 1 requhes that the components be designed to quality standards commensurate with the importance of the safety function to be performed. Our review of the CANDU 3U QA Manual indicates that, although a quality assurance program exists, it is not up to the normal standards of 10 CFR 50 Appendix B. There is no evidence to indicate that the quality of the product is inadequate, just the manner in which the QA program is implemented. I Criterion 14 requires that the reactor coolant pressure boundary be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormalleakage, of rapidly propagating failure, and of gross rupture. Criterion 15 further requires that the reactor coolant system be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded for any anticipated operational condition. From a design point of view, these requirements are accomplished by using the ASME Section III design criteria wherever possible. Historically, the reliability of zirconium alloy pressure tubes worldwide has been 4x104 leaks /FT-I EFPY and 2x104 ruptures /PT-EFPY. It is not clear if this qualifies as " extremely low", as this term has not been defined. Given the large number of pressure tubes in a CANDU reactor, a failure rate of 104 stillleads to a 2% probability of failure of a single tube per reactor-year. The acceptable frequency of fuel channel failures should be compared with that for small diameter piping rather than for the reactor pressure vessel or the main reactor coolant piping, as the consequences of a pressure tube failure more closely resembles a small break LOCA (loss-of-coolant accident) than the large break LOCA, that is the design basis for LWRs. That evaluation is outside the scope of this review.
'I 5 I
E E Criterion 30 requires a leak detection system capable of locating the source of reactor coolant leakage to the extent practical. Two leak detection systems are employed: one in the fuel channel annu us to detect leaks inside the fuel channel, and a second system based on moisture collectors in the reactor building to identify leaks outside the reactor. Criterion 32 requires that the design permit periodic inspection, leak testing and a surveillance program. These are all included in the CANDU program. The remaining requirements of the GDC are more directly applicable to the actual material performance than to system design. The key requirements are outlined in Criterion 31, which requires that the material of the reactor coolant pressure boundary behave in a non-brittle manner and that the probability of a rapidly propagating fracture is muunuzed. Criterion 31 further requires consideration of service temperatures and uncertainties in the material properties, effects of irradiation, residual, steady state and transtent stresses, and the size of flaws. The CANDU 3 reactor design will be evaluated on a component by component basis against these criteria and the corresponding requirements of the SRP, Regulatory Guides and the ASME Code. Many of the reactor components do not meet the requirements of the ASME Code, Section III, either because the type of component is not approved for use in a given Code class, or because the material is not Code approved (or both). The determination of the significance of a non-ASME type of component is beyond the scope of this review. However, the basis for that deternunation should include not only the design approach, but also the results of testing and actual experience, and the consequences of failure. The determination of the significance of non-ASME material will be addressed below. 4.1 Fuel Channel Assemblies The fuel channel is the fundamental design feature of the CANDU reactor. The purpose of the fuel channel is to?
- a. Support and locate fuel in the reactor core
- b. Permit free passage of fuel through the reactor core during refueling
- c. Permit the PHTS coolant flow to remove fuel heat with low pressure drop and low vibration
- d. Form part of the PHIS pressure boundary
- e. Provide thermal insulation from the moderator and end shield coolant in normal operation
- f. Minimize neutron absorption
- g. Provide shielding to attenuate radiation where they pass through the end shields
- h. Retain shield plugs and channel closures
l I i. h fide for low leakage fueling machine connections onto the fuel channel ends ,
- j. Provide connections to the PHTS feeders
- k. Accommodate thermal dimensional changes, as well as changes due to creep and growth
- 1. Pr vide for detection of leakage from the pressure tube or calandria tube I The fuel channel is designed for a life of 30 years at 80% capacity, giving an effective life of 24 effective full power years (EFPY).
i The fuel channel consists primarily of the pressure tube, calandria tube, and end fittings. However, it also includes the garter springs that maintain the separation between the pressure tube and the calandria tube, and various components (such as the shield plugs and channel closures) required to reduce radiation exposure at the reactor face and to perform on-line refueling. Each of these will be discussed below. I 4.1.1 Pressure Tubes In the CANDU 3 reactor there are 232 pressure tubes, with an inside diameter of 103.4 mm (4.07 in), a thickness of 4.19 mm (0.164 in), approximately 6.35 meters (250 in) long. 'Ihe pressure tubes are I fabricated from cold worked Zr-2.5 wt% Nb tubing (UNS designation R60901), according to CSA standard N285.6.1. This material specification is not approved by ASME, but the American Society for Testing and Materials (ASTM) is currently reviewing the content of specification B353, which may cover cold worked Zr-2.5% Nb in an upcoming edition. This is one requirement listed in Article IV-1000 of the ASME Code for Code approval of a previously unapproved material. . I The Zr-2.5% Nb pressure tubes are extruded and cold worked between 20 and 30%. After cold working, the tubes are conditioned by removal of 0.05 mm (0.002 in) from the inside surface. The tubes are then stress relieved at 400*C i 15'C (752*F 27 F) for 24 hours. A prior vacuum stress relief at a higher temperature may be specified. Each tube is hydrostatically tested to 27.5 MPa (4000 psi). This pressure is intended to apply a stress of approximately 60% of the room temperature yield strength. l The minimum specified tensile strength is 435 MPa (63.1 ksi) and the specified muumum yield strength is 325 MPa (47.1 ksi), both tested at 300 C (572'F). No impact testing or fracture toughness requirements are specified. However, hardness tests are required in accordance with ASTM Standard f I 67 I
E e EIS (no acceptance criteria are given in N285.6.1). Corrosion tests in accordance with ASTM Standard G2 are also required. The strength of the extruded pipe varies with location and orientation. He axial strength is typically less than the hoop strength by about 12%, and the front end (relative to the orientation during extrusion) is typically weaker than the back end. CSA Standard 285.6.1 requires the strength to be measured at the front end in the longitudinal direction. This provides additional margin on strength, especially since hoop stress due to pressure is the goveming loading. Strength variation with temperature was determined by a best fit line of strength data from various heats of material (a heat of material includes all of the products produced from a single batch of molten material at the mill). The slope of the line was determined from the test data and was adjusted up or down to go through the minimum specified value. The design allowable stress is determined as the lower of 2/3 of the minimum specified yield stress and 1/3 of the minimum specified ultimate stress, which is consistent with the ASME Code approach. For this material, the ultimate stress criterion is lower, so the allowable stress is based on 1/3 the ultimate strength. The pressure tube is a Class 1C component, and is designed according to ASME Section III, Subsection NB-3200, except that the following additional requirements are imposed by N285.2:
- The wall thickness includes an allowance for intemal and extemal corrosion, creep, and wear ,
due to fuel movement.
- The material must be capable of sustaining a suberitical through-wall crack (leak-before-break)
- Creep and growth are considered in the design analysis
- Limits are placed on applied and residual stresses to muumize the potential for delayed hydride cracking Now let us examine how the pressure tubes compare against the materials requirements, using the NRC's Standard Review Plan as a guide. The SPP specifies that materials that make up the reactor coolant pressure boundary must be compat'! ale with the reactor coolant. The primary concerns expressed in the SRP are corrosion and stres<, corrosion cracking. The Zr-2.5% Nb pressure tubes have been shown to have a lower corrosion rate (as measured by the rate of hydrogen absorption) than the i Zircaloy 2 pressure tubes used previously, and no evidence has surfaced to indicate that stress corrosion cracking is a problem for this material. Radiation has very mmunal effect on the rate of
, i s
I corrosion in Zr-2.5% Nb pressure tubes, although it has a pronounced effect on Zircaloy 2 tubes. The rate of corrosion is highly dependent on the oxygen content of the coolant, but the proposed water chemistry specification, in conjunction with hydrogen water chemistry, should keep corrosion and hydrogen absorption to a manageable level. I The other concem delineated in the GDC is brittle fracture. There are no impact testing requirements in any of the Canadian standards for pressure tubes nor any requirements that the pressure tube material have a given level of fracture toughness as there is in Appendix G of 10 CFR 50. There is, however, a requirement in N285.2 that che " Pressure tube material shall be capable of sustaining a I suberitical through-wall crack in a nonductile condition (i.e., leak before breaking)." Exactly how this is to be achieved is unclear. Although most of incidents involving pressure tubes did in fact leak before breaking, in at least one incident the tube ruptured without prior leakage. In that case, an annulus spacer had moved out of position, allowing the pressure tube to contact the calandria tube, which set up a through-wall thermal gradient that allowed hydride formation on the outside surface only. 'Ihe end result was a flaw that was long, but shallow, which results in failure before the flaw ever penetrates the wall (i.e., no leak before break). The curumstances behind that incident are still within the realm of possibilities, and therefore leak before break can not be assured I under those conditions. Fracture toughness is highly variable in zirconium alloys, depending primarily on the concentration of hydrogen, temperature, and the level and orientation of stress at which the hydrides precipitate out. Fracture toughness levels as low as 25 MPa-m '2 (22.8 ksi-in /2) have been measured on samples of Zr-i 2 2.5% Nb that were intentionally charged with hydrogen, and as low as 35 and 40 MPa-m '2 (31.9 and 36.4 ksi-init2) for samples removed from two Zircaloy 2 tubes removed from service.ns.2c The requirements of 10 CFR 50 Appendix G only apply to ferritic materials. The reason for this is presumably because the only other materials considered were austenitic stainless steels, that have inherently high toughness. It is not clear whether the Appendix G criteria should apply to the zirconium pressure tubes. It is clear the pressure tubes do not meet the intent of Appendix G. There are two possible resolutions to this problem: either develop means of ensuring that the combination of conditions necessary for brittle fracture do not exist, or conclude that the criteria of Appendix G are overly conservative because the consequences of failure of a pressure tube are small compared to a large break LOCA which is assumed in the criteria that were developed for light water reactors. The latter option involves safety analyses and consequence analyses that are beyond the I 69 i l I , i
E scope of this project. As a result, the discussion herein will be limited to those actions that can be taken to ensure that the combination of conditions necessary for brittle fracture do not exist. Because the fracture toughness problem in zirconium alloys is tied to the level of absorbed hydrogen, I some nondestructive means of accurately measuring the absorbed hydrogen should be developed (there is work underway in this regard). In addition, there needs to be additional effort placed on preventing and detecting pressure tube /ca!andria tube contact, as this is the condition that leads to failure before the crack penetrates the wall thickness of the pressure tube. There is one mode of material degradation present in zirconium alloys that is not considered by U.S. codes, standards, and regulations, because it is not a problem in the materials used in light water reactors. Creep appears to be the nominal life limiting damage mechamsm for Zr-2.5% Nb pressure tubes. Canadian standard N285.2 requires that "The design and stress analyses shall include the effects of creep deformation. This includes axial creep, diametral creep, and creep-sag of the pressure tubes." The primary concern related to creep is contact between the pressure tube and the calandria tube, either resulting from sag or from diametral creep (ballooning). In both cases, the metal to metal contact leads to thermal gradients that can accelerate degradation of the material by hydride formation. In the latter case, flow restrictions could result that may have an adverse effect on cooling of the fuel. It is clear that an extensive inspection and surveillance program will be required for the pressure tubes to monitor the effects of hydride formation and creep (including growth). The Canadian standards provide for inspection and surveillance, but CANDU plants licensed in the U.S. will not be subject to Canadian regulations. Licensing of CANDU plants in the US. will require development of US. standards or license conditions to address the inspection, testing and surveillance requirements of pressure tubes to ensure early detection of unacceptable damage by hydride formation and creep. 4.1.2 Rolled joints Roll expanded joints are not allowed in ASME Class 1 components (Section 111, Subsection NB-3671.2). However, they are allowed for Class 2 and 3 applications (NC-3671.2 and ND-3671.2) provided that
" experience or test (NC-3649) has demonstrated that the joint is suitable for the Design loadings and when adequate provisions are made to prevent separation of the joint."
I 70 I_
I I There are several types of rolled joints in the fuel channel assembly, ne joint between the pressure tui,e and the end fitting is a Class IC component, whereas the other joints (e.g., the calandria tube / lattice tube, fuel channel annulus bellows / lattice tube, and end fitting / liner tube) can be categorized as Class IC,2C, or 3C. Rolled joints in Class 2C and 3C components are required to be designed to the corresponding section of the ASME Code (NC-3300 or ND-3300). The following requirements are imposed on rolled joints in Class IC components, or Class 2C components designed ,' to the rules of NB-3200:
= provision is made to prevent separation of the joints under all specified service loadings, and I = they are accessible for inspection, maintenance and/or removal, and replacement, and
= they are either designed by analysis to the rules of NB-3200 or prototype joints are subjected to performance tests to determine the structuralintegrity of the joints under simulated service conditions, except for the case of the pressure tube /end fitting joint, where both design by analysis and prototype testing are required
= production joints are produced using the same tooling design and procedures that were used to qualify the prototype joints.
The prototype testing includes heliur.t leak tests, residual stiess measurements, hot pressurized pullout I tests, and long-term thermal fatigue tests. In addition, post-service examination of tubes removed from service is done wherever possible. The field experience indicated problems due to high residual ] stresses in early CANDU reactors, due to an improper rolling operation (the rolling head was inserted ; too far into the tube, leaving high residual stresses). Modifications to the rolling practice have effectively eliminated problems due to high residual stresses. l 1 l I The tube material is expanded into grooves in the end fitting during rolling, which creates a mechanical joint and not just an interference fit. As a result, creep is not likely to have a significant i effect on either the leak resistance or the pullout strength of the rolled joints. There appears to be no fundamental materials problems associated with using roll expanded joints in Class I service. However, inspection and surveillance should be required to monitor the condition 1 of the rolled joints. I 4.1.3 End-Fittings I The end fittings are fabricated from a quenched and tempered martensitic stainless steel (a modified i l AISI 403). This material specification was selected based on mechanical properties (strength, I 71 I
as = toughness, and hardness), corrosion resistance, a coefficient of thermal expansion that matches zirconium as closely as possible, and resistance to neutron irradiation. A number of materials were examined, but AISI 403 was selected as the best compromise of the various material requirements listed above. The regt.ited room temperature tensile properties are 725 MPa (105 ksi) ultimate tensile strength, S85 I MPa (85 ksi) yield strength, with a 12% minimum elongation and 30% reduction in area. A review of data from early reactors showed that transverse tensile property data bounded longitudinal properties. However, Charpy impact requirements are specified in both the longitudinal and transverse directions: 27 joules (20 ft-lb) at 21*C for longitudinal specimens and 20 joules (15 ft-lb) 3 at 65'C for transverse specimens. When the material choice was made in 1963, the material specification for the end fittings met the I toughness requirements of the ASME Code. However, the Code has subsequently been revised to strengthen the toughness requirements. Subsection NB-2332 of the ASME Code requires 25 mils lateral expansion for the range of thicknesses of the end fitting. A review of lateral expansion data for end fittings already installed indicated that the lateral expansion requirement of the ASME Code could not be met consistently (see Figure 4-1). AECL's position on this issue is that the fracture toughness requirements of ASME Section III are for low strength high toughness materials, and should not be applied to quenched and tempered materials such as the modified 403 stainless steel used for the end fittings. The AISI 403 material approved by the ASME Code is for the annealed condition, where the yield strength is half that of the quenched and tempered material. AECL has attempted to demonstrate by calculation and by test that the end fitting material meets the requirements of 10 CFR 50 Appendix G to ensure protection against non-ductile failure. They make the weak argument that "the end fitting 403 is no worse than some pressure vessel steels" based on the fact that the measured toughness of 403 stainless steel falls above the lower bound Ku curve of the ASME Code (Figure 4-2). Burst tests were performed using a machined flaw to simulate the 1/4 T (thickness) flaw of 10 CFR 50 Appendix G. The burst tests demonstrated a factor of safety of three over the operating pressure. This provides some confidence in the fracture tolerance of the design. AECL has concluded that meeting the new requirements would be " difficult without a high and unnecessary reject rate'* and therefore does not intend to meet the ASME requirements. 72 I
=
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. FRACTURE (3
I O A I I I 0 200 250 300 350 400 450 500 550 TEMPERATURE (K) CHARPY V-NOTCH IMPACT PROPERTIES OF AISI TYPE 403 LOW RESIDUAL STAINLESS STEEL O N ADIATED 50 - Di G IRRADIATED 6.0 x 1.0 23 nim? >1. McV
- I Q IRRADIATED 6.2 x 1023 nj,2 33 g,y lg 5 40 -
I 5 m z l
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" -95 K G
I O G o , l 10 - o Gg - 50% 5 HEAR j o e O Q FRACTURE ee . 1 I 1 1 l 200 250 300 350 400 450 500 550 TEMPERATURE (K) Figure'4-1 Charpy V-Notch Impact Properties of AISI 403 Stainless Steelm 73 I 1
200
- REFERENCE STRESS INTENSITY (e 160 -
Q FACTOR RELATIVE TO RTH0T
=
a - M 140 - 5 - 150 *5 v IRRADIATED. h 0 -
- UNIRRADIATED k N N u-u- 100 -
[ AISI 403 STEEL ~ 100 U CRNL Kci DATA g 5 80 -
- ;" a E 1- m m 60 -
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" o
-160 -120 -80 -40 0 43 80 120 160 200 240 *C
-200 67 -45 -22 0 22 45 67 90 112 135 *F
- 112 TEMPERATURE RELATIVE TO RTuoi (T-RTH0TI Figure 4-2 Comparison of Fracture Toughness Data for 403 Stainless Steel with K., Curve"8 74 M M M M M M M M M M M M M M M M M EM
I Irradiation increases the strength of the 403 stainless steel, by approximately 50 to 100 MPa. This of course increases the margin on strength, but it is usually accompanied by a decrease in ductility, or , toughness. Irradiation induces a shift in the Charpy impact transition temperature of a material. The procedure for evaluating this is given in Regulatory Guide 1.99. The amount of shift depends on the chemical composition of the material and the fluence. An end of life shift in the transition temperature was estimated to be 50 C to 100 C. This was taken into consideration in the burst tests that were used to demonstrate compliance with Appendix G of 10 CFR 50. Note that the required transverse Charpy specimens are tested at 65*C, which is in the lower part of the transition region of the Charpy energy curve (see Figure 4-1). A 50 C to 100 C shift in the Charpy energy curve would mean that the material is in the lower shelf of the Charpy energy curve, which is usually indicative of brittle behavior. The burst tests described above were tested at -40*C to account for a shift due to irradiation. Examination of the fracture surfaces revealed that "the failure ) mode was the same for both these conditions
- although it is not stated whether the fracture surface is ductile or brittle. Although it has been demonstrated that there is adequate toughness for the 1/4T flaw, the data provided by AECL do not rule out non-ductile failure.
I AECL has reviewed the performance of the material with regard to corrosion, erosion, stress corrosion cracking, and irradiation, as noted above. ney have considered the effect of the environment on corrosion (both inside and outside) and stress corrosion cracking, including compatibility with the feeder cabinet insulation,in case chlorides leach out onto the end fittings. For general corrosion and erosion, they have concluded that the 403 stainless steel will see less than 0.025 mm of metal loss over a 30 year life. For stress corrosion cracking, they have concluded that 403 stainless steel is not susceptible to SCC in a chloride environment unless the material is stressed to more than 75% of its yield strength. As a result, corrosion, erosion, and stress corrosion cracking are not considered to be problems for the end fittings. I Although the material does not meet ASME Code requirements on toughness and non-ductile fracture, it appears to meet the intent of all of the other requirements of the General Design Criteria, including consideration of corrosion, and stress corrosion cracking. Acceptance of this material should be contingent upon a more detailed review of the fracture toughness and non-ductile fracture issue. I The end fittings contain another feature that is unique to the CANDU design and is therefore not approved by the ASME Code. This is the channel closure plug, which is a mechanical closure on the
,eac,_ian, p,_e b_ea,y mat may be op_e wh, me ,eac,o,1s _11ne ,o, me pu, poses 75 I
e of refueling. Although it is actually a part of the end fitting,it will be discussed further in the section on the s'neling machine below. 4.1.4 Annulus Spacer The annulus spacer is an important component in the fuel channel. It has already been shown that one spacer out of alignment led to the rupture of a pressure tube, due to the formation of hydrides caused by the thermal gradient at the point of calandria tube to pressure tube contact. The annulus spacer is a toroidal spring (garter spring) that encircles the pressure tube. In early CANDU reactors, g the spring was made from Zr-2.5 Nb-0.5 Cu. ~1his alloy was selected over Zircaloy 2 because it had 5 "the best combination of strength, hydride orientation and resistance to corrosion in moist annulus gas."m This comment is somewhat surprising, as the annulus gas is supposed to be dry gas. In the early CANDU reactors, the annulus spacer fit loosely over the pressure tube. However, problems developed with the zirconium spacers. Tests showed that hydrides could form at the point of contact with the relatively cool calandria tube, so the material was changed to Inconel X-750, despite the disadvantages of its neutron absorption. In addition, the springs were redesigned to fit tightly around the pressure tube to eliminate movement. A Zircaloy 2 girdle wire is inserted down the center of the spring, and encircles the pressure tube 1.5 times. "Ihe girdle wire should hold the spring in place in case of breakage, an' ,ws the locations of the springs to be determined by eddy current inspection from inside the pressure tubes. Vibration tests have shown that the new - annulus spaccus should not move. The annulus spacer is a unique component that has no parallel in light water reactor design. It is not clear what CSA Class is assigned to the annulus spacers, so it is difficult to determine the exact regulatory requirements. No information was provided by AECL on the corrosion or stress corrosion cracking performance of the material, or the fracture toughness. Our review of the literature did not reveal any significant problems with X-750, except that Inconel acts as a window for absorption of hydrogen by zirconium alloys when in contact with Inconel in the presence of water." Thus, the comment about moist gas noted earlier leads to a concem over hydride formation at the point of , contact with the pressure tube and calandria tube, inconel spacers were used on the Nuclear Power Demonstration (NPD) reactor (the first reactor of the CANDU design), and four springs have been removed from service after 22 to 25 years and examined. Three of the four springs broke during handling, indicating a loss of ductility. The fracture potential m of the garter springs requires further investigation. 76 I.
I I 4.1.5 Calandria Tubes f ) The calandria tubes are fabricated from Zircaloy 2 strip that is formed into a tube and seam welded. I As noted previously, the welds are made using a controlled atmosphere to limit contamination by impurities. The required room temperature tensile properties are provided in Table 4-1. The elongation must be at least 20%. No Charpy impact tests are required. H Table 4-1: Required Room Temperature Tensile Properties 7 Property Longitudinal Direction Transverse Direction Yield Stress 317 MPa (46 ksi) 317 MPa (46 ksi) Tensile Strength 427 MPa (62 ksi) 414 MPa (60 ksi) L y Many of the materials issues for the calandria tubes have already been discussed in the previous 1 discussions on the pressure tubes. Zircaloy 2 has a higher rate of corrosion than Zr-2.5% Nb for a given set of conditions, but the operating temperature of the calandria tubes is over 200*C colder than the pressure tubes. It was previously noted that a 50*C change in temperature corresponds with a factor of two change in the rate of corrosion. Therefore, the rate of corrosion of the calandria tubes l should be significantly less than the pressure tubes. This is supported by oxide thickness measurements of calandria tubes removed from service, where the oxide has been too thin to measure. l The creep rate of Zircaloy 2 is typically less than for Zr-2.5% Nb. This, in combination with the lower temperatures, means that creep is not likely to bt a problem .r the calandria tubes. Radiation growth i E measurements of the end shields have been inconclusive . suse of the small magnitude of growth. P The calandria tube is not part of the reactor coolant pressure boundary, so the Charpy impact requirements of 10 CFR 50 Appendix G are not appropriate. AECL does not have any data supporting the threshold stress intensity factor Km nor any measurements of the fracture toughness of hydride Zircaloy 2, in part because they have been unable to induce delayed hydride cracking in annealed Zircaloy 2. Although there is a lack of data to support this conclusion, it does not appear that degraded fracture toughness is a significant concern. This is because the calandria tubes are not as likely to be attacked by delayed hydride cracking as pressure tubes because the amount of hydrogen absorbed during the corrosion process is much smaller. i 1 77 j 1 __ ... d
E However, there is one concem regarding failure of the calandria tubes. If a p4 essure tube fails, the dynamic pressure inside the calandria tube may exceed the strength of the tube. Analytical models have shown that this is possible, and in fact this has occurred when a pressure tube ruptured at Bruce 2 in 1985. The calandria tube also burst, spilling fuelinto the moderator tank. Burst tests have shown that the Zircaloy 2 calandria tubes will most likely fail in the weld metal of the longitudinal seam. Because of the possibility that the calandria tube will rupture if a pressure tube ruptures, AECL does not take any credit for the integrity of the calandria tube in the safety analysis. They have performed tests to see if adjacent calandria tubes will be damaged by fuel ejected from the pressure tube. E Although there may be some local deformation and denting of the calandria tube, it is unlikely that 3 any damage would occur to the adjacent pressure tube. The calandria tubes are designed such that if they are dented or collapsed due to a dynamic event in the moderator, the deformation will remain in the elastic regime. When the dynamic event passes, the calandria tube will rebound elastically back into its round shape, and no permanent contact will exist between the pressure tube and calandria tube as a result of the failure of an adjacent fuel channel. The fact that the calandria tube is not designed to withstand a likely accident condition is not in agreement with the basic ASME Code philosophy or the requirements of the General Design Criteria E for components of the reactor coolant pressure boundary. However,it should be remembered that 5 the calandria tube is not part of the primary pressure boundary. 4.1.6 Bellows I ASME does not currently allow bellows or expansion joints in piping for Class I components (NB-3649.1), although it does indicate that rules are under development (that statement has existed in the ASME Code for at least nine years). It is not clear what CSA Code Class is assigned to the gas annulus bellows. However, because the bellows are not part of the primary reactor coolant pressure boundary, it is likely that the bellows are Class 2 or 3 components. In that case, the bellows could be designed to the appropriate sections of NC-3600 or ND-3600. Very little information is provided on the bellows in the documents provided by AECL, so the scope I of the review is limited. However, it was noted that the bellows suffer from the same deficiency as the calandria tubes. If the pressure tube fails, the bellows may not be able to withstand the primary system pressure. A pressure tube rupture in Pickering 2 in 1983 resulted in rupture of the bellows, spilling reactor coolant down the face of the reactor. No other significant materials issues were identified in this review. 78 s.
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- 4.2 Feeder Piping Very little information is provided regarding the feeder piping except that it is made from carbon steel (A-106 Grade B). Typical materials concems for this type of material would include general corrosion and fracture toughness. A corrosion allowance is specified in the design of the feeder piping to account for general corrosion, and preventative measures are taken during maintenance to prevent internal corrosion when the piping is drained (e.g., the piping is filled with an inert gas such as nitrogen to prevent corrosion). No information is provided on the fracture toughness of the material.
It is assumed that the feeder piping is considered a Class 1 component, as it is part of the primary reactor coolant pressure boundary. The many feeder tubes ultimately come together to form a single large heat transport system pipe, which is the only place that a large break LOCA can occur. As a result, it should have been designed, fabricated, and tested to the requirements of ASME Section III, including the fracture toughness requirements. 4.3 Reactor Control Mechanisms Very little information was provided by AECL on the control rod guide tubes and the liquid poison I injection systems for the moderator. Canadian standards N285.6.2 and N285.6.3 provide specifications for the materials for the control rods and liquid injection shutdown systems, respectively. The two specifications are nearly identical, allowing either Zircaloy 2 or Zircaloy 4. Unlike their use in light water reactors, the control rod guide tubes are not part of the reactor coolant pressure boundary. As a result, the control rod guide tubes are not considered to be Class I components. The materials issues related to zirconium alloys have been discussed at length earlier in this report. l These components would be susceptible to corrosion and creep, except that they operate in a less severe environment than the pressure tubes. The reactor control mechani:ms are located in the relatively cool moderator, so corrosion and hydrogen ingress will not h significant. Irradiation !
)
induced creep and growth may be a problem, especially for the horizontal reactivity control mechanisms. No other specific materials concerns have been identified in this review. 4.4 Reactor Vessels, Support Structures, and Associated Piping Very little detailed information was provided by AECL on the calandria tank, moderator system I piping, shield tank, and reactor supports. None of these are parts cf the reactor coolant pressure boundary, and therefore many of the requirements of the General Design Criteria do not apply. It I 79 I
E a is anticipated that relevant materials issuas include corrosion, stress corrosion cracking, and radiation embrittlement, although it is not likely that any of these are severe problems. 4.5 Fueling Machine I The fueling machine snout assembly attaches to the end fitting by means of tapered wedges that are I driven in behind the flange on the end fitting, forcing the end fitting face against a metallic ring which acts as a seal. When the fueling machine is not attached, the outlet end fitting is sealed by a removable closure plug (which is really part of the fuel channel). During refueling, the fueling machine acts as part of the reactor coolant pressure boundary and therefore should be elesigned to the requirements of the General Design Criteria. The specific CSA Code Class to which the fueling machine is designed is unclear, and various parts of the fueling machine may in fact be designed to different Code Classes. The fueling machine safety lock and be fuel channel closure safety lock are covered in N285.2, which requires that they be designed to ASME Section Ill, Appendix III. The fueling machine is said to be fabricated from 17-4 PH stainless steel, although it is not clear that all components within the fueling machine are also made from 17-1 PH. Precipitation hardened stainless steels tend to be susceptible to stress corrosion cracking and crevice corrosion in certain environmen'.s. Fasteners, including nuts, bolts, retaining rings, and spring rings, are also highly susceptible to stress corrosion cracking. The mechanical components that act to make up the connection between the fueling machine and the end fitting may therefore be susceptible to crevice corrosion or stress corrosion cracking. These components, however, are accessible for inspection when the fueling machine is not in use. 'Ihe ASME Code does not allow any closures such as this for Class I service, and there is no basis for establishing the acceptance criteria. An irepection and preventative maintenance program would clearly be required to ensure the continued integrity cf tj e connection. I The closure plug and latch assembly are also made from 17-4 PH stainless steel ar.d -may therefore be susceptible to crevice corrosion and stress corrosk n cracking. Althcugh the channel closures could be removed for detailed inspection when the channel is drained of fluid, there is no other means of routine inspection ; the mechanisms of the closures, unless the fueling machine was programmed to remove a closure plug and reinstall a different closure plug after refueling. This type of closure is not approved by the ASME Code. The basis for acceptance of the closure plug design will be similar to the basis for acceptance of the fueling machine snout assembly discussed above. 60
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,I 5.0 DISCUSSION OF ASME CODE APPLICATION TO CANDU 3 The CANDU 3 desigrt as presently defined, proposes to utilize materials and service conditions that will require additional review prior to their acceptance for licensmg of the design in the U.S. The key materials-related areas needing additicaal review indude: e non-ASME materials e non-ASME components e in-service degradation 5.1 Non-ASME Materials The first item is not a major obstacle. The NRC has in the past approved non-ASME materials for use in nudear plants. There is nolegal requirement that a heavy water reactor meet the ASME Code. However, AECL has attempted to meet the Code wherever possible. They performed an " extensive test program on the same basis as that for ASME Class 1 materials".m This is intended to meet the criteria of Artide IV of the ASME Code for approval of a previously unapproved material. They apparently have submitted an I application to ASTM forconsideration of the Zirconium alloys as part of ASTM specification B353, which is one of the other requirements of Artide IV. It should also be noted that the Zirconium alloys were never rejected by the ASME; their lack of ASME approval comes from the fact that they have never been evaluated on their technical merits. Efforts have been made by AECL to apply for ASME Code Cases to allow some of the unique aspects of the CANDU design. Progress was limited, so the Canadians had to develop their own standards. I The case of the martensitic stainless steel used for the end fittings is not as clear cut. The material used in the CANDU reactors does not meet the lateral expansion requirements of ASME. The material met the I AShE specification when it was first used, but ASME modified the requirements in 1971. AECL submitted a request to ASME for changes in the toughness requirements based on metallurgical differences between U.S. and Canadian materials, but were denied. In this case, there are two options available for AECL Technologies. This first is to change the material specification for the CANDU 3 reactors so that the toughness requirement can be met. However, AECL believes that the ASME requirements can not be met "without a high and unnecessary reject rate"." The other option is to demonstrate the adequacy of the material. AECL has attempted to do this by experimentally validating the material against the .I 81 I
E as requirements of Appendix G of 10 CFR 50. Although the experimental evidence appears to support their case, the analytical arguments are not conclusive. I 5.2 Nen-ASME Components The second issue noted above was the use of non-ASME components, including the fuel channel rolled I joints, the closure assemblies, the fueling machine, and the fueling machine supports. Of these components, only the rolled joint has a comparable use in light water reactors. Rolled joints are used in ste.m generators in light water reactors; however, the industry experience with nied joints in the steam generators has generally been poor. The ASME Code therefore allows the use of rolled joints in Gass 2 and Class 3 locations, but not in Class I locations. As a result, the ro'ic<' Mus tai in PWR steam generators are seal welded, and the seal weld acts as the Class 1 pressure bxa hry. % is not a viable option in the CANDU pressure tubes, as the Zirconium alloys are not weldabic to the :ainless steel end fittings. The pressure tube rolled joints in the CANDU reactors are designed by analysis to the rules for Gass I components (except, of course, for the clause that does not allow rolled joints in Cass 1 service) and are tested as required by Gass 2 rules for rolled joints. There are no LWR components equivalent to the other components unique to the CANDU design. Even if all of these components are designed with the factors of safe:y implicit in Gass 1 design, an accident during refueling operations could result in leakage of reactor coolant. Examples of refueling type of accidents include failure of the fueling machine to seal properly against the end fitting, failure of the closura plug to seal properly, and movement of the fueling machine support bridge while the fueling machine is attached. These are all accident scenarios that have occurred in CANDU plants. 5.3 Material Degradation Several of the components, including the pressure tutes, may reach the end of their useful life prior to the end of the plant design life. Relevant damage mechanisms include irradiation induced creep and growth, delayed hydride cracking (DHC), and stress corresion cracking (SCC). Of these, only stress corrosion cracking is generally applicable to light water reactors. Stress corrosion cracking typically occurs so rapidly that it is basically a binary system cither it occurs or it does not. As a result, the U.S. codes and standards attempt to prevent SCC by maintaining the stress below a threshold value and by controlling the envimnmental conditions to prevent its occurrence. Even then, SCC may occur due to wnditions outside the control of the plant. As a result, inspection is used to detected damage caused by 1CC. A similar 82 I
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I I- program can be developed for preventing SCC in the stainless steel components in the end fittings and fueling machines of the CANDU 3 design. I In many ways, delayed hyddde cracking is smular to SCC. Maintaining the stress below a given threshold and controlling the environment can greatly reduce the risk of DHC. Therefore, a program similar to SCC prevention and mitigation can be developed for DHC prevention and mitigation. However, there is one ; precursor to DHC that may be useful in DHC prevention: the presence of hydrogen. The absorbed hydrogen precipitates out when the reactor is cooled during a shutdown, thereby embritt!!ng the material. The cracking itself does not occur until the pressure tube is subsequently loaded when the reactor is restarted. If a nondestructive technique can be developed for detection of absorbed hydrogen during a shutdown, there may be a way to prevent the subsequent hydride cracking. Another approach to preventing and mitigating DHC would be to develop a comprehensive predictive model for DHC, and to use surveillance data to ensure that the model is accurate. An analogy can be drawn from the methods used for monitoring radiation embrittlement in light water reactors in the U.S. Regulatory Guide 1.99 contains an explicit model for predicting radiation embrittlement as a function of neutron flux and chemical composition. He> wever, a surveillance program is used to ensure that the I predictive model is accurate. A surveillance program typically consists of small pieces of the actual material used in the pressure vessel (in this case, the pressure tube) that are exposed to the operating environment with a known level of radiation, and can be removed for destructive testing. Irradiation induced creep and growth is a phenomenon unlike any other in light water reactors. The best
- method of dealing with this damage mechanism will be to develop a predictive model and use inspection and/or surveillance to ensure that the predictions are accurate. Because the primary concern is dimensional rather than structural, the surveillance program may not require separate specimens, became they do not have to be destructively tested. However,if a surveillance program were implemental for DHC, the same specimens could also be used for creep and growth measurements.
I 5.4 Recommendations A great deal of work will be required to demonstrate the adequacy of non-ASME materials and components for use in Class I service. The process for accomplishing this includes collection of enalysis and test data, comparison of factors of safety, identification of probh r.t areas, and development of plans for mitigating the problems, followed by third party review and reg datory acceptance. There are three i parties that could be involved in this process: AECL Technologies, (SME, and the NRC. Ideally, ASME S3 I l I __J
E would develop all of the necessary rules, based on information provided by AECL Technologies. Once ASME has agreed upon the rules, the NRC would review them to ensure that the new rules adequately address safety concems. Then AECL Technologies could submit a license application for a reactor design that meets the ASME Code. An altemative would be for AECL Technologies to develop the necessary requirements in the form of Code Cases for each exception to the Code. The level of detail of the submittals must be censistent with the nature of U.S. codes and regulations. As there is no legal requirement that heavy water reactors meet the ASME Code. AECL Technologies could develop a set of requirements / criteria (similar in scope to the ASME code) for the heavy water reactor that meet the safety objectives of the ASME Code. These criteria would be used for those features where the ASME Code's light water reactor criteria may not apply. Tis criteria should be supported by test data and analysis. The criteria should be more detailed than the existing Canadian standards, along the lines of existing U.S. codes and standards. Problem areas such as creep, DHC, and SCC should be identified and plans developed to mitigate the problems, including surveillance programs where appropriate. Once AECLTechnologies has developed a complete package of design criteria, then the NRC can evaluate the criteria and the supporting data in light of the safety objectives of the General Design g Criteria. 5 I I I I I I I 84
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6.0 REFERENCES
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I
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Pacific Northwest laboratories' Test Program Results", Electric Power Research Institute, NP-5132, RP 1250-4, April 1987.
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Atomic Energy of Canada Limited, Report AECL-10949, Chalk River Nuclear Laboratory, Chalk River, Ontario, March 1994,
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I
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Regulatory Document, R-9, Atomic Energy Control Board, Ottawa, Ontario, February 21,1991. I 31. " Overpressure Protection Requirements for Pnmary Heat Transport Systems in CANDU Power Reactors Fitted with Two Shutdown Systems", Regulatory Document, R-77, Atomic Energy Control Board, Ottawa, Ontario, October 20,1987. h i 87 II
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CAN/CSA-N285.3-88, Canadian Standards Association, February 1988.
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i .l 43. Douglass, DL,"Cor msion Mechanism of Columbium, Zirconium and Their Alloys: Diffusion of Oxygen in Columbium Pentoxide and Zirconium Dioxide", Corrosion of Reactor Materials, Salzburg, Austria, June 4-8,1962, Vol. 2, pp. 223-255.
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I 45. Evans, U.R.,' Hydrogen Troubles in Zirconium Alloys", The Corrosion and Oxidation ofMetals: First Supplementary Volume, St. Martins Press, New York,1968, pp.164-167.
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I 48. Simpson, LA., and Wilkins, B.J.S.," Prediction of Fast Fracture in Zr-2.5% Nb Pressure Tubes Using Elastic-Plastic Fracture Mechanics", Third International Conference on Mechanical Behaviour of Materials, Cambridge, E. gland, August 20-24,1979, Vol. 3, pp. 563-572.
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I 50. Ells, C.E., and Sawatzky, A.,"The Effect of Neutron Irradiation and Hydrogen Concentration on Tensile Properties of the CANDU CoolantTube Spacer Alloy", Atomic Energy of Canada Limited, I Report AECL-2272, Chalk River Nuclear Laboratory, Chalk River, Ontario, June 1965.
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- 52. Coleman, C.E., and Simpson, L.A., " Evaluation of a Leaking Crack in an Irradiated CANDU Pressure Tube", Atomic Energy of Canada Limited, Report AECL-9733, Chalk River Nudear Laboratory, Chalk River, Ontario, June 1988.
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- 54. Dunn, J.T., Kakaria, BX, Graham, J., and Jackman, A.H., "CANDU-PHW Fuel Channel Replacement Experience",1AEA Specialists' Meeting, Riso, Denmark, September 1982.
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