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DSAR CHAPTER 4 - PRIMARY COOLANT SYSTEM Revision 36 SECTION 4.1 Page 4.1-1 of 4.1-1
4.1 INTRODUCTION
The Primary Coolant System is comprised of two heat transfer loops connected in parallel to the reactor vessel. Each loop contains one steam generator, two circulating pumps, connecting pipe and instrumentation. A pressurizer is connected to one of the reactor vessel outlet pipes by means of a surge line and is provided with pressurizer relief and safety valves. All components of the Primary Coolant System are located within the containment building.
DSAR CHAPTER 4 - PRIMARY COOLANT SYSTEM Revision 36 SECTION 4.2 Page 4.2-1 of 4.2-1 4.2 DESIGN BASIS 4.2.1 DELETED 4.2.2 DELETED 4.2.3 DELETED 4.2.4 CODES ADHERED TO AND COMPONENT CLASSIFICATION Following the implementation of Permanently Defueled Technical Specifications, the PCS design information herein is considered historical and does not constitute design requirements unless otherwise noted to still be applicable in the defueled condition.
The original design, fabrication, construction, inspection, testing and classification of all reactor coolant system components was in accordance with the ASME Boiler and Pressure Vessel Code,Section III, 1965 edition, including all addenda through Winter 1965 (ASME B&PV Code,Section III, 1965, W65a), and the Code for Pressure Piping, ASA B31.1, 1955. The replacement steam generators installed during 1990 meet ASME Code Section III 1977 edition.
The codes adhered to and component classifications are listed in Table 4-2.
4.2.5 DELETED 4.2.6 DELETED
DSAR CHAPTER 4 - PRIMARY COOLANT SYSTEM Revision 36 SECTION 4.3 Page 4.3-1 of 4.3-8 4.3 SYSTEM DESIGN AND OPERATION Following the implementation of Permanently Defueled Technical Specifications, the PCS design and operation information herein is considered historical and does not constitute design requirements unless otherwise noted to be applicable in the defueled condition.
4.3.1 GENERAL DESCRIPTION All components of the Primary Coolant System are located within the containment building. A diagram of the Primary Coolant System is shown in Figure 4-1. The system includes two identical heat transfer loops connected in parallel to the reactor vessel. Each loop contains one steam generator, two circulating pumps, flow and temperature instrumentation, and connecting piping. A pressurizer is connected to one of the reactor vessel outlet pipes by means of a surge line. The pressurizer is located with its base at a higher elevation than the reactor vessel piping. The equipment arrangement relative to its supports and the surrounding concrete is shown in Figures 1-4, 1-5, 1-8 and 1-10.
4.3.2 INTERFACES WITH OTHER SYSTEMS An inlet nozzle is provided on each of four reactor vessel inlet pipes. An outlet nozzle is provided on one reactor vessel outlet pipe.
Drains from the Primary Coolant System to the radioactive waste disposal system are provided. Only cold leg 2B has a drain line and it contains normally closed loop drain valves. Downstream of the pair of loop drain valves is a spectacle flange. There is also one hot leg drain line directly connected to the primary system drain tank with two isolation valves. A connection is also provided on the quench tank.
Sampling system lines are provided from the primary coolant piping, the pressurizer and the quench tank. A connection to the quench tank from the nitrogen supply system is provided. A connection to the quench tank spray header from the demineralized water supply is provided.
4.3.3 REACTOR VESSEL The reactor vessel and top head assembly are shown in Figure 4-2. The reactor vessel and top head was designed in accordance with ASME B&PV Code,Section III, Class A, 1965, W65a. The reactor vessel parameters are listed in Table 4-3. As the facility is permanently defueled, these design requirements are no longer applicable.
The inner surface of the reactor is clad with 308/309 stainless steel. In the areas of internal attachments, the interior is clad with Ni-Cr-Fe alloy. The vessel closure flange is a forged ring with a machined ledge on the inside surface to support the reactor internals and the core. The flange is drilled and
DSAR CHAPTER 4 - PRIMARY COOLANT SYSTEM Revision 36 SECTION 4.3 Page 4.3-2 of 4.3-8 tapped to receive 54 7-inch diameter closure studs and is machined to provide a mating surface for the reactor vessel seal. An externally tapered transition section connects the flange to the cylindrical shell.
A boss located around each outlet nozzle on the inside diameter of the vessel wall provides a mating surface for the internal structure. This boss and the outlet sleeve on the core support barrel are machined to a common contour.
A fixed hemispherical head is attached to the lower end of the shell. There are no penetrations in the lower head.
The removable upper closure head is hemispherical. The head flange is drilled to match the vessel flange stud bolt locations.
The 2.76-inch ID control rod drive mechanism nozzles (Ni-Cr-Fe alloy through the head, stainless steel flange) terminate with bolted and seal-welded flanges at the upper end which are aligned on a single plane. There are eight 4-inch nominal instrumentation nozzles of similar construction to the control rod drive mechanism nozzles. In addition to these nozzles, there is a 3/4-inch vent connection.
All of the reactor head nozzles are made of a Ni-Cr-Fe alloy, known as Alloy 600.
The core is supported from the reactor vessel flange. The control rod drive mechanisms are supported by the nozzles in the reactor vessel head. The reactor vessel is supported on three pads welded to the underside of the coolant nozzles.
The reactor vessel internals are constructed with wetted parts of Stellite, Ni-Cr-Fe, stainless steel or Zircaloy. The control rod drive mechanisms' housings are stainless steel.
4.3.4 STEAM GENERATOR The nuclear steam supply system utilizes two steam generators shown in Figure 4-3. The parameters for the steam generators are given in Table 4-4.
The steam generator is a vertical U-tube heat exchanger and was designed in accordance with ASME B&PV Code,Section III, 1977 edition.
The steam generator has an inlet nozzle, 3/4-inch OD U-tubes and two outlet nozzles. Vertical partition plates in the lower head separate the inlet and outlet plenums. The plenums are stainless steel clad, while the primary side of the tube sheet is Ni-Cr-Fe clad. The vertical U-tubes are Inconel-600. The tubes are rolled in the full depth of the tube sheet and then the tube-to-tube sheet joint is welded on the primary side.
The steam generator has an inlet feedwater nozzle, a feedwater distribution ring with bottom apertures and a downcomer. The downcomer is an annular
DSAR CHAPTER 4 - PRIMARY COOLANT SYSTEM Revision 36 SECTION 4.3 Page 4.3-3 of 4.3-8 passage formed by the inner surface of the steam generator shell and the cylindrical shell which encloses the vertical U-tubes.
On the shell side of the steam generators and the main steam line piping up to the inlet of the turbine stop valve are twenty-four (24) safety valves. These valves are ASME B&PV Code spring-loaded, open bonnet, safety valves.
Twelve safety valves are mounted on each of the main steam lines upstream of the steam line isolation valves but outside the containment. Parameters for the secondary safety valves are given in Table 4-5.
The steam generator shell is constructed of carbon steel. Manways and handholes are provided for access to the steam generator internals.
4.3.4.1 Deleted 4.3.4.2 Steam Generator Replacement Due to the tube degradation problems noted above, replacement of both steam generators was undertaken in late 1990. The replacement steam generators are designed to physically match the essential parameters of the existing steam generators and to be compatible with the performance characteristics utilized in the FSAR and the license for operation at 2,565.4 MWt. Consistent with other PCS equipment, the replacement steam generators are designed for operation at 2650 MWt should an increase in the licensed power level be pursued in the future.
Some component design changes were made to improve the replacement units:
- 1. Tube wall thickness was reduced slightly (.042 vs .048) to improve heat transfer. Combined with 308 less tubes, the overall steam generator heat transfer effect is unchanged.
- 2. Tube support design was changed from solid plate to eggcrate dividers and other features to minimize corrosion crevices and denting.
- 3. Blowdown capability was improved through an internal center duct and increase in blowdown nozzle size. Sampling improvements were also made.
- 4. Hand holes and inspection ports were added for future internals inspection. Manway sizes were also enlarged.
- 5. Integral flow restrictor nozzles were added to the main steam outlets to restrict blowdown flow in the event of a ruptured steam line.
DSAR CHAPTER 4 - PRIMARY COOLANT SYSTEM Revision 36 SECTION 4.3 Page 4.3-4 of 4.3-8 4.3.5 PRIMARY COOLANT PUMPS The primary coolant is circulated by four pumps, Figure 4-4, which are of the vertical single suction, centrifugal type. The discharge nozzle is horizontal and the suction nozzle is in the bottom vertical position. The original pressure containing components were designed and fabricated in accordance with the ASME B&PV Code,Section III, Class A, 1965, W65a.
The pump impeller is pinned, bolted and locked to the shaft. Each pump is equipped with replaceable casing wear rings. A hydrostatic bearing is located in the fluid between the impeller and thermal barrier to provide shaft support.
The thermal barrier is located above the shaft seal assembly, which consists of four face-type mechanical seals, three full pressure seals mounted in tandem and a fourth low-pressure backup vapor seal designed to withstand operating system pressure with pumps stopped. The seal detail is shown in Figure 4-6.
The pump is provided with a flywheel. The pump motor assembly includes motor bearing oil coolers, seal chamber, controls and instruments. Cooling water is provided from the Component Cooling Water System. The parameters for the primary coolant pumps are given in Table 4-6.
The primary coolant pump and motor are supported by four support lugs welded to the scroll.
The pump is constructed of high alloy casting and stainless steel parts to minimize corrosion. These materials are listed in Table 4-6. The Type SU mechanical seals consist of a rotating titanium carbide ring riding over a hard carbon face. The Type N-9000 seals have a tungsten carbide ring which rides over a hard carbon face.
The pump seal consists of three pressure breakdown stages in series and one vapor stage.
4.3.6 PRIMARY COOLANT PIPING The primary coolant piping consists of lengths of 42-inch ID hot leg pipe from the reactor vessel outlet to the steam generator inlet and lengths of 30-inch ID cold leg pipe between the steam generator outlet and the pump suction nozzle and between the pump discharge and the reactor vessel inlets. The other major piece of primary coolant piping is the 12-inch, Schedule 140 surge line pipe between the pressurizer and one hot leg. Parameters for the main primary coolant piping are given in the piping list table, Table 4-7.
DSAR CHAPTER 4 - PRIMARY COOLANT SYSTEM Revision 36 SECTION 4.3 Page 4.3-5 of 4.3-8 The hot and cold leg pipes have no individual supports. The hot and cold legs are supported through their respective connections at the steam generator and reactor vessel nozzles with the cold legs having additional support through their connection to the reactor coolant pumps.
The primary coolant piping is of rolled bond clad plate construction, having a base metal of ASTM A 516, Grade 70, with a cladding of 304L stainless steel with a nominal thickness of 1/4 inch.
The 12-inch surge line is Type 316 stainless steel. Thermal sleeves are installed on all 2-inch or greater inlet nozzles to reduce thermal shock effects from auxiliary systems.
The 12-inch nominal diameter safety injection nozzles on the 30-inch ID cold leg pipes are constructed of carbon steel with a stainless steel clad interior.
The 12-inch nominal diameter shutdown cooling nozzle on the 42-inch ID hot leg pipe is of the same construction. The remaining nozzles, all 3 inches or less in diameter, are of Ni-Cr-Fe alloy. The large diameter nozzles on the primary coolant piping all use Ni-Cr-Fe alloy safe ends for the field weld to the connecting piping. The Ni-Cr-Fe alloy used for the small nozzles and large nozzle safe ends in the piping is known as Alloy 600.
The piping is shop fabricated and shop welded into subassemblies to the greatest extent practicable to minimize the amount of field welding.
Fabrication of piping and subassemblies is done by shop personnel experienced in making large heavy wall welds. Welding procedures and operations meet the requirements of the ASME B&PV Code,Section IX, 1965, W65a. All welds are 100% radiographed and magnetic particle or liquid penetrant tested to the acceptance criteria of the ASME B&PV Code,Section III, Class A, 1965, W65a. All primary coolant piping penetrations are attached in accordance with the requirements of the ASA B31.1, Power Piping Code, 1955. Heat treatment of the piping is performed after all fittings have been assembled, nozzle bosses welded and cladding deposited. Field welds are postweld heat-treated to the requirements of the ASME B&PV Code,Section III, Class A, 1965, W65a. Cleanliness standards consistent with nuclear service are maintained during fabrication and erection. As the facility is permanently defueled, these design requirements are no longer applicable.
All small piping connected to the Primary Coolant System, such as instrument lines, is Type 316 stainless steel and welded using the same specification limits as the major piping connections.
4.3.7 PRESSURIZER The pressurizer was designed in accordance with ASME B&PV Code,Section III, Class A, 1965, W66a. Table 4-8 gives parameters for the pressurizer. The pressurizer is shown in Figure 4-8. As the facility is permanently defueled, these design requirements are no longer applicable.
DSAR CHAPTER 4 - PRIMARY COOLANT SYSTEM Revision 36 SECTION 4.3 Page 4.3-6 of 4.3-8 There is a total of 120 pressurizer heaters. They are single unit sheath-type immersion heaters which protrude vertically into the pressurizer through sleeves welded in the lower head.
The pressurizer spray system consists of 3-inch lines running from the PCP discharges (P50B and P50C) to two 3-inch diaphragm-operated spray control valves, which then combine to a single 4-inch line connected to a single spray head inside the top of the pressurizer.
The pressurizer is supported by a circumferential skirt welded around the lower head.
The pressurizer is constructed of ASTM A 533, Grade B, Class 1 steel plate.
The interior surface of the cylindrical shell and upper head is clad with Type 304 stainless steel. The lower head is clad with an Ni-Cr-Fe alloy to facilitate welding of the Ni-Cr-Fe alloy heater sleeves to the shell; stainless steel and Ni-Cr-Fe alloy safe ends are provided on the pressurizer nozzles, as required to provide for field welds to the connecting piping.
The Ni-Cr-Fe alloy used for the nozzles and safe ends on the pressurizer is known as Alloy 600.
4.3.8 QUENCH TANK Parameters for the pressurizer quench tank are given in Table 4-11.
Fabrication is in accordance with ASME B&PV Code,Section III, Class C, 1965, W65a. As the facility is permanently defueled, these design requirements are no longer applicable.
The tank is constructed of carbon steel and the interior surface is epoxy lined.
The sparger, spray header, nozzles and rupture disc fittings are stainless steel.
4.3.9 VALVES 4.3.9.1 Deleted 4.3.9.2 Deleted 4.3.9.3 Power-Operated Relief Valves (PORV) and Block Valves Parameters for these valves are given in Table 4-14.
The valves are solenoid-operated power relief valves. The two full capacity valves are located in parallel pipes which are connected to the single pressurizer relief valve nozzle on the inlet side and to the relief line piping to the quench tank on the outlet side. A motor-actuated isolation valve is provided upstream of each of the relief valves.
DSAR CHAPTER 4 - PRIMARY COOLANT SYSTEM Revision 36 SECTION 4.3 Page 4.3-7 of 4.3-8 4.3.9.4 Spring-Actuated Primary Safety Valves Three primary safety valves are located on the pressurizer. They are totally enclosed, back pressure compensated, spring-loaded safety valves meeting ASME B&PV Code,Section III, Article 9, 1965, W65a, requirements. As the facility is permanently defueled, these design requirements are no longer applicable.
The primary safety valves discharge through the relief line piping into the quench tank.
4.3.10 ENVIRONMENTAL PROTECTION
- 1. Flooding The containment building is of watertight construction and is inherently safe against external flooding. Refer to Subsection 5.4.1 for additional information.
- 2. Seismic The NSSS is designed to withstand the load imposed by the maximum seismic disturbance. Refer to Section 5.7 for additional information.
4.3.11 DELETED 4.3.12 INSULATION Carbon steel piping and equipment are insulated with a conventional material compatible with the temperature and functions involved.
A removable metal reflective-type thermal insulation is provided on the flange stud area of the reactor vessel closure head to permit access to the head studs for removal and reinstallation of the head.
Supports for the insulation consisting of carbon steel rings formed to fit the OD of the respective shells and necessary attachment brackets are provided.
The heads of respective vessels have internal tapped studs appropriately spaced for attaching the insulation. All insulation support attachments are attached prior to final stress relief.
4.3.13 SYSTEM CHEMICAL TREATMENT Control and variation of the primary coolant chemistry is a function of the Chemical and Volume Control System. Sampling system lines are provided from the primary coolant piping to provide a means for taking periodic samples of the coolant for chemical analysis. The water chemistry is maintained as indicated in Table 4-16.
DSAR CHAPTER 4 - PRIMARY COOLANT SYSTEM Revision 36 SECTION 4.3 Page 4.3-8 of 4.3-8 The solids content is maintained below the design level by minimizing corrosion through careful selection of materials, chemistry control and continuous purification of the letdown stream of primary coolant through filters and demineralizers. Hydrogen is maintained in the reactor coolant to chemically combine with the oxygen produced by the radiolysis of water. The primary coolant pH is controlled by the addition of lithium hydroxide (LiOH).
Hydrazine may be added during initial start-up for oxygen scavenging.
Zinc (depleted in Zn-64) is added to primary coolant through the Zinc Addition System for the removal of radioactive cobalt ions from PCS piping (inner walls). Removal of the radioactive cobalt ions reduces dose to personnel from PCS piping.
All wetted surfaces in the Primary Coolant System are compatible with the above water chemistry. Specific component material is in the sections describing individual components.
DSAR CHAPTER 4 - PRIMARY COOLANT SYSTEM Revision 36 SECTION 4.5 Page 4.5-1 of 4.5-7 4.5 INSPECTIONS Following the implementation of Permanently Defueled Technical Specifications, the test and inspection information herein is considered historical and does not constitute design requirements unless otherwise noted to still be applicable in the defueled condition.
4.5.1 GENERAL Shop inspection and tests of all major components were performed at the vendor's plant prior to shipment. An inspection at the site was performed to assure that no damage has occurred in transit. Testing of the Primary Coolant Systems was performed at the site upon completion of the Plant construction. These tests included hydrostatic tests of primary and secondary loops. A complete visual inspection of all welds and joints was performed prior to the installation of the insulation. All field welds were radiographically and dye penetrant inspected in accordance with the requirements of the ASME B&PV Code,Section III, Class A, 1965, W65a and special erection specifications prepared by Combustion Engineering.
A hot flow test was made of the primary loop up to zero power operating pressure and temperature without the core installed. The system was checked for vibration and cleanliness. Auxiliary systems were checked for performance.
4.5.2 DELETED 4.5.3 DELETED 4.5.4 NONDESTRUCTIVE TESTS Prior to and during fabrication of the reactor vessel, nondestructive tests based upon the ASME B&PV Code,Section III, 1965, W65a, were performed on all welds, forgings and plates as follows:
All full penetration pressure containing welds were 100% radiographed to the standards of the ASME B&PV Code,Section III, Subparagraph N-624.8, 1965, W65a. Other pressure containing welds such as used for the attachment of mechanism housings, vents and instrument housings to the reactor vessel head were inspected by liquid penetrant tests of the root pass, each 1/2 inch of weld deposit and the final surface.
DSAR CHAPTER 4 - PRIMARY COOLANT SYSTEM Revision 36 SECTION 4.5 Page 4.5-2 of 4.5-7 All forgings were inspected by ultrasonic testing, using longitudinal beam techniques. In addition, ring forgings were tested using shear wave techniques. Rejection under longitudinal beam inspection, with calibration so that the first back reflection is at least 75% of screen height, was based on interpretation of indications causing complete loss of back reflection.
Rejection under shear wave inspection was based on indications, exceeding in amplitude the indication from a calibration notch whose depth is 3% of the forging thickness, not exceeding 3/8 inch with a length of 1 inch.
All forgings were also subjected to magnetic particle examination. Rejection was based on relevant indications of:
- 1. Any cracks and linear indications
- 2. Rounded indications with dimensions greater than 3/16 inch Plates were ultrasonically tested using longitudinal and shear wave ultrasonic testing techniques. Rejection under longitudinal beam testing performed in accordance with ASME B&PV Code,Section II, SA-435, 1965, with calibration so that the first back reflection is at least 50% of screen height, was based on defects causing complete loss of back reflection. Any defect which showed a total loss of back reflection which could not be contained within a circle whose diameter is the greater of 3 inches or one-half the plate thickness was unacceptable. Two or more defects smaller than described above which caused a complete loss of back reflection were unacceptable unless separated by a minimum distance equal to the greatest diameter of the larger defect unless the defects were contained within the area described above. For shear wave testing, the maximum permissible flaw was one which did not exceed that from a calibrated notch having a depth of 3% of the plate thickness and 1 inch long.
Nondestructive testing of the vessel was performed during several stages of fabrication with strict quality control in critical areas such as constant calibration of test instruments, metallurgical inspection of all weld rod and wire, and strict adherence to the nondestructive testing requirements of the ASME B&PV Code,Section III, Class A, 1965, W65a.
The detection of flaws in irregular geometries was facilitated because most nondestructive testing of the materials was completed while the material is in its simplest form. Nondestructive inspection during fabrication was scheduled so that full penetration welds are capable of being radiographed to the extent required by ASME B&PV Code,Section III, Class A, 1965, W65a.
Each of the vessel studs received two ultrasonic tests and one magnetic particle inspection during the manufacturing process.
DSAR CHAPTER 4 - PRIMARY COOLANT SYSTEM Revision 36 SECTION 4.5 Page 4.5-3 of 4.5-7 The first ultrasonic test was a radial longitudinal beam inspection and the standard for rejection was 100% loss of back reflection or an indication which reduced the adjusted back reflection by greater than 20%. The second ultrasonic test was a radial inspection using the angle beam technique with the rejection standard the same as for forgings. The use of these techniques insured that only such materials were accepted that have flaws no greater than 1/2 inch and no observable cracks or sharply defined linear defects.
Magnetic particle inspection was performed on the finished studs. Axially aligned defects whose depths are greater than thread depth and nonaxial defects were unacceptable.
The vessel closure contains 54 studs, 7 inches in diameter with 8 threads per inch. The stud material is ASTM A 540-65, Grade B24, with a minimum yield strength of 130,000 psi. The tensile stress in each stud when elongated for operational conditions is approximately 40 ksi. Calculations show that 32 uniformly distributed studs can fail before the closure will separate at design pressure. However, 16 uniformly distributed broken studs or 4 adjacent broken studs will cause O-ring leakage. Failure of at least 16 adjacent studs is necessary before the closure will fail by "zippering" open.
The vessel studs are stressed as they are elongated by the stud tensioners during the initial installation of the vessel head and at each refueling. The amount of elongation versus hydraulic pressure on the tensioner will be compared with previous readings to detect any significant changes in the elongation properties of the studs. Studs which yield questionable data during the head installation, or receive damage to the threads, will be replaced before returning the vessel to pressure operations.
Table 4-18 summarizes the inspection program by component.
4.5.5 ADDITIONAL TESTS During design and fabrication of the reactor vessel, a number of operations over and above the requirements of the ASME B&PV Code,Section III, Class A, 1965, W65a, were performed by the vendor. Table 4-19 summarizes the additional tests by component.
During the design of the reactor vessel, detailed calculations were performed to assure that the final product would have adequate design margins. The design adequacy was established by stress concentration factors which have been obtained through the use of photoelastic models for areas which are not amenable to calculation. A detailed fatigue analysis of the vessel for all design conditions has been performed. In addition, Combustion Engineering has performed test programs for the determination, solution and verification of analytical solutions to thermal stress problems. Also, fracture mechanics and brittle fracture evaluations have been performed.
DSAR CHAPTER 4 - PRIMARY COOLANT SYSTEM Revision 36 SECTION 4.5 Page 4.5-4 of 4.5-7 All material used in the reactor vessel was carefully selected and precautions were taken by the vessel fabricator to ensure that all material specifications were adhered to. To assure compliance, the quality control staff of Combustion Engineering reviewed the mill test reports and the fabricator's testing procedures.
All welding methods, materials, techniques and inspections comply with ASME B&PV Code, Sections III and IX, 1965, W65a. Before fabrication was begun, detailed qualified welding procedures including methods of joint preparation, together with certified procedure qualification test reports, were prepared. Also, prior to fabrication, certified performance qualification tests were obtained for each welder and welding operator. Quality control was exercised for all welding rod and wire by subjection to a complete and thorough testing program in order to insure maximum quality of welded joints.
During the manufacture of the reactor vessel, quality control by the vendor, in addition to and in areas not covered by the ASME B&PV Code,Section III, 1965, W65a, included: Preparation of detailed purchase specifications which included cooling rates for test samples; requiring vacuum degassing for all ferritic plates and forgings; specification of fabrication instructions for plates and forgings to provide control of material prior to receipt and during fabrication; use of written instructions and manufacturing procedures which enabled continual review based on past and current manufacturing experiences; performance of chemical analysis of welding electrodes, welding wire and materials for automatic welding, thereby providing continuous control over welding materials; the determination of NDT temperature through use of drop weight testing methods and test programs on fabrication of plates up to 15 inches thick to provide information about material properties as thickness increases. Shear wave and longitudinal wave ultrasonic testing was performed on 100% of all plate material.
Cladding for the reactor vessel was a continuous integral surface of corrosion-resistant material, 1/4-inch nominal thickness. The detailed procedure used; ie, type of weld rod, welding position, speed of welding, nondestructive testing requirements, etc, was in compliance with the ASME B&PV Code, Sections III and IX, 1965, W65a.
Combustion Engineering has checked the cladding on completed reactor vessels and such tests have not shown the need for 100% ultrasonic testing for weld deposited cladding after fabrication. The clad surface is ultrasonically inspected transverse to the direction of welding for lack of bond at intervals of 12 inches or 1.4 times the base metal thickness, whichever is less.
Upon completion of all postweld heat treatments, the reactor vessel was hydrostatically tested, after which all weld surfaces, including those of welds used to repair material, were magnetic particle inspected in accordance with ASME B&PV Code,Section III, Paragraph N-618, 1965, W65a.
DSAR CHAPTER 4 - PRIMARY COOLANT SYSTEM Revision 36 SECTION 4.5 Page 4.5-5 of 4.5-7 Quality control by the licensee was also carried out during the manufacture of the vessel by a resident inspector. This work included independent review of all radiographs, magnetic particle tests, ultrasonic tests and dye penetrant tests conducted during the manufacture of the vessel. A review of material certifications, and vendor manufacturing and testing procedures was also conducted. This review included all manufacturers' records such as heat treat logs, personnel qualification files and deviation files.
4.5.6 INSERVICE INSPECTION Provision was made in the design to permit inservice inspection as may be required. The location of the more highly stressed portions of the reactor vessel was identified. These areas are equipped with removable insulation and portions may be inspected at various intervals, utilizing appropriate nondestructive testing techniques. In addition, the inside of the reactor vessel and the internals may be subjected to routine visual inspection during refueling outages. An inspection of accessible areas of the reactor vessel and internals, with a television camera or other suitable means, may be accomplished at any time when the reactor core is completely unloaded. The design permits all vessel internals except the flow skirt to be removed so that a complete internal vessel visual inspection would be possible. During refueling outages, the reactor vessel head and the closure sealing surfaces may be visually inspected. The internal parts of the vessel which are visible, including the cladding and components, may also be visually checked, as well as the accessible external surface of the vessel, nozzles and the vessel studs.
A combination of ultrasonic, dye penetrant, magnetic particle and visual inspections will be used to conduct the inspections. The planned inspection program takes into account the mechanisms which may lead to failure in the Primary Coolant System. Emphasis has been placed on the expected high stress areas as determined by a design evaluation and experience.
The major premises of this inspection program are:
- 1. Selected areas of expected maximum stress will be inspected at intervals in accordance with ASME B&PV Code Section XI, except as adjusted by NRC-approved code cases and relief requests. These inspections will serve to indicate potential problems before significant flaws develop there or at other areas.
- 2. If flaw initiation or growth is detected in one of the selected areas, all comparable areas in the primary system will be inspected.
- 3. Regardless of the results of inspection of the selected areas, all major discontinuity areas in the primary system will be inspected within a ten-year period, except as adjusted by NRC-approved code cases and relief requests.
DSAR CHAPTER 4 - PRIMARY COOLANT SYSTEM Revision 36 SECTION 4.5 Page 4.5-6 of 4.5-7 Refer to Section 6.9 for a detailed description of the inservice inspection program. A summary of the inspection program is as follows:
- 1. Areas of expected maximum stress selected for periodic inspections are:
- a. The flange-to-shell weld of the vessel
- b. The flange-to-torus weld of the vessel head
- c. The primary coolant outlet nozzle-to-shell welds and nozzle-shell radii on the vessel ID
- d. The dissimilar welds between the primary coolant piping and pumps
- e. Longitudinal and circumferential welds in the primary coolant piping
- f. Branch piping connections to the primary coolant piping
- g. The tube sheet-to-head weld of the steam generators
- h. The nozzle-to-head welds in the lower head of the steam generators
- i. The internal support stand welds in the lower head of the steam generators
- j. The support stand to steam generator head
- 2. Areas to be inspected within a ten-year period, except as adjusted by NRC-approved code cases and relief requests, are:
- a. The nozzle-to-shell welds of primary coolant nozzles not inspected on a more frequent basis
- b. The longitudinal and circumferential welds in the core region of the reactor pressure vessel
- c. The outlet nozzle-to-shell weld of the pressurizer
- d. The longitudinal weld in the pressurizer
DSAR CHAPTER 4 - PRIMARY COOLANT SYSTEM Revision 36 SECTION 4.5 Page 4.5-7 of 4.5-7 Ultrasonic inspection of components provides indications from discontinuities, impedance mismatches (such as a junction between Inconel weld metal and carbon steel) and from changes in component geometry. Baseline data to assist in interpretation of future inspection results will be acquired from a preservice inspection and pertinent shop data.
If indications of defect initiation or growth are noted, the program will be reviewed and sufficient inspections performed to determine that defects are not being initiated or propagated in other areas of the pressure vessel or components.
The bases for the above inspection points and the frequency of inspection are the result of a review of design drawings, the test results available from the PVRC vessel test program conducted at Southwest Research Institute, the present knowledge available on the mechanics of failure of such systems, ASME Section XI Code requirements, and NRC-approved code cases and relief requests. They are also based on the fact that the component fabricator for this Facility has a history of successful vessel fabrication in accordance with the practices of the ASME B&PV Code and more restrictive self-imposed specifications. Code manufacturing procedures and inspection techniques precluded the initial presence of large flaws in the vessel. Therefore, it is believed that the most likely location of a failure would be at a point of expected maximum stress concentration and not at some random location.
Thus, high stress locations are selected for monitoring of initiation of flaws.
Furthermore, with baseline ultrasonic readings obtained on the pressure vessel and other inspection points in the reactor primary system, added assurance is attained that no significant flaws exist in the pressure boundary components of the Primary Coolant System.
Palisades was granted relief in a NRC Safety Evaluation dated February 11, 2009 (Reference 59) to extend the third inservice inspection interval for reactor vessel weld examinations until December 12, 2015.
4.5.7 DELETED 4.5.8 NONDESTRUCTIVE TESTS OF OTHER PRIMARY SYSTEM COMPONENTS Prior to and during fabrication of the original components of the Primary Coolant System, nondestructive testing based upon the requirements of the ASME B&PV Code,Section III, Class A, 1965, W65a, was used to determine the acceptance criteria for various size flaws. The requirements for the Class A vessels are the same as the reactor vessel. Vessels designated as Class C were fabricated to the standards of the ASME B&PV Code,Section III, Article 21, 1965, W65a. Requirements for replacement parts and components are as specified in Section 4.2.1.
DSAR CHAPTER 4 - PRIMARY COOLANT SYSTEM Revision 36 SECTION 4.8 Page 4.8-1 of 4.8-1 4.8 PRIMARY COOLANT GAS VENT SYSTEM Following the implementation of Permanently Defueled Technical Specifications, the Primary Coolant Gas Vent System information herein is considered historical and does not constitute design requirements unless otherwise noted to still be applicable in the defueled condition.
The Primary Coolant Gas Vent System (PCGVS), see Figure 4-1, consists of a flow-limiting orifice on both the reactor vessel vent and pressurizer vent lines, solenoid valves, a pressure transmitter for pressure indication, and connecting piping.
The entire PCGVS was designed for Seismic Category I. The primary coolant pressure boundary within the PCGVS, up to and including the second solenoid valve, is Safety Class 2 (Safety Class 1 upstream of the flow-limiting orifices). The piping was designed, fabricated, installed, and tested to ASME B&PV Code,Section III, Subsection NC, 1974, S76a. Supports were designed, fabricated, installed and tested in accordance with Subsection NF, ASME B&PV Code,Section III, 1974, S76a. The entire PCGVS was analyzed using the ADLPIPE Computer Code, Revision 3C. The PCGVS, up to and including the second normally closed solenoid valve, will be maintained as Quality Group A (Class 1) per Reg Guide 1.26 and ASME B&PV Code,Section XI. As the facility is permanently defueled, these design requirements are no longer applicable.
The PCGVS piping is AISI Type 304 or Type 316 stainless steel. The PCGVS solenoid valve bodies are AISI Type 316 stainless steel.