RS-11-100, Dresden, Units 2 and 3, Updated Final Safety Analysis Report (Ufsar), Revision 9, Chapter 5.0, Reactor Coolant System and Connected Systems

Dresden, Units 2 and 3, Updated Final Safety Analysis Report (Ufsar), Revision 9, Chapter 5.0, Reactor Coolant System and Connected Systems

ML11202A181
Person / Time
Site:	Dresden
Issue date:	06/29/2011
From:	Exelon Generation Co, Exelon Nuclear
To:	NRC/FSME
References
RS-11-100
Download: ML11202A181 (153)

Text

DRESDEN - UFSAR 5.1-1 5.0 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS

5.1

SUMMARY

DESCRIPTION

The equipment and evaluations presented in this chapter are applicable to either unit. When the data presented apply only to one unit, the applicable unit is identified.

The purpose of the boiling water reactor is to generate steam from the light water moderator and coolant to drive the main turbine and generator to produce electricity. The topics of the reactor coolant system (RCS) and connected systems discussed in this chapter include the reactor coolant system pressure boundary (RCPB) integrity, the reactor pressure vessel (RPV) and appurtenances, and the major RCPB allied systems. Pertinent design information for some of the major RCS components is presented in Table 5.1-1.

The reactor coolant system includes those systems and components which contain or transport reactor coolant, in the form of water or steam, to and from the reactor pressure vessel. These systems form a major portion of the reactor coolant pressure boundary. Chapter 5 provides information regarding the RCS and pressure-containing appendages out to and including the outermost isolation valve in the main steam and feedwater piping. The major components within the RCS are the reactor vessel; the reactor coolant recirculation system with its pumps, piping, and valves; the relief valves (RVs); the safety valves (SVs); the safety relief valve (SRV); the feedwater and steam piping; the feedwater and steam piping isolation valves; and portions of reactor auxiliary systems piping (see Figure 5.1-1). In addition, the isolation condenser, the reactor water cleanup system, the hydrogen water chemistry system, and the shutdown cooling system are included and addressed in Section 5.4.

The RCPB, as defined in 10 CFR 50.2, includes all those pressure-containing components (such as the RPV, piping, pumps, and valves) which are part of the RCS or are connected to the RCS up to and including any and all of the following:

A. The outermost containment isolation valve in system piping which penetrates the primary containment;

B. The second of the two valves normally closed during normal reactor operation in system piping (such as vent and drain lines on the reactor recirculation system) which does not penetrate the primary containment; and C. The RCS safety relief valve, relief valves, and safety valves.

The RCPB extends to and includes the outermost containment isolation valve in the main steam and feedwater piping.

The integrity of the RCPB is addressed in Section 5.2. Items of discussion include the overpressurization protection system, the RCPB materials, and the reactor water chemistry including the hydrogen water chemistry program. Inservice inspection and inservice testing programs are also addressed. Numerous leakage detection methods are also addressed.

DRESDEN - UFSAR Rev. 4 5.1-2 The reactor vessel and its appurtenances are addressed in Section 5.3. Items of discussion include the vessel materials of construction, compliance with fabrication codes and regulatory guides, and methods of fabrication. Sensit ized stainless steel and intergranular stress corrosion cracking (IGSCC) associated with the safe ends and vessel internal brackets are addressed in Section 5.3; the reactor vessel materi als surveillance program for the shift in the nil ductility transition (NDT) temperature is also covered in Section 5.3. The NDT temperature effect on the pressure-temperature operating curves is also addressed. Compliance with the intent of 10 CFR 50, Appendices G and H is also addressed.

The associated systems interfacing with or acting as a part of the RCS are addressed in Section 5.4. The systems and/or components discussed are the reactor recirculation system, the hydrogen water chemistry system, the isolation condenser, the main steam line isolation system and flow restrictors, the reactor shutdown cooling system, the reactor water cleanup system, and the main steam line and feedwater piping up to and including the outermost isolation valves. The main steam system, feedwater system, and condensate system are addressed in more detail in Chapter 10.

5.1.1 Schematic Flow Diagram

The typical reactor coolant system is shown in Figure 5.1-1. Nominal flowrates, temperatures, and pressures for both units are listed. Coolant volumes are listed in Table 5.1-3.

5.1.2 Piping and Instrumentation Diagrams The P&IDs applicable to the RCS and connected systems are identified in Table 5.1-2. This table is organized according to the drawing topic first and then the applicable unit.

5.1.3 Elevation Drawings

The elevation drawings and plan view section drawings for the RCS and other associated equipment are addressed in Drawings M-2 through M-10, M-2A, and M-10A through M-10E.

DRESDEN - UFSAR (Sheet 1 of 3)

Table 5.1-1

REACTOR COOLANT SYSTEM DATA Reactor Vessel Internal height 68 ft, 7 5/8 in. Internal diameter 20 ft, 11 in. Design pressure and temperature 1250 psig at 575ºF Maximum heatup/cooldown rate 100ºF in 1-hour period. Maximum uncontrolled cooldown rate (based on one-time transient) 240ºF/hr Base metal material SA-302 Grade B, modified with Code Case 1339 Wall thickness 6 1/8 in. minimum Design lifetime 40 years Base metal initial NDT temperature 40ºF maximum Cladding material Weld-deposited E-308L electrode Cladding thickness 1/8-in. minimum Design code ASME Section III, Class A Recirculation Loops Number 2 Material Unit 2 Unit 3 304 stainless steel 316NG stainless steel Design pressure and temperature Suction Discharge 1175 psig at 565ºF 1325 psig at 580ºF Design code ASME Section I and USAS B 31.1.0 Recirculation Pumps Number 2 Type Vertical, centrifugal, single-stage Power rating 6000 hp DRESDEN - UFSAR Rev. 2 Table 5.1-1 (Continued)

REACTOR COOLANT SYSTEM DATA (Sheet 2 of 3) Speed Variable Flowrate 45,000 gal/min Design pressure and temperature 1450 psig at 575ºF Developed head 570 ft Design code ASME Section III, Class C

Recirculation Valves Number Unit 2 Unit 3 8 4 Type Motor-operated gate Design code ASME Section I and USAS B 31.1.0

Jet Pumps Number 20 Material Stainless steel Overall height (top of nozzle to diffuser discharge) 18 ft, 7 in. Diffuser diameter 20 3/4 in. Design APED-5460 (General Electric)

Main Steam Lines Number 4 Diameter 20 in. Material Carbon steel Design code ASME Section I and USAS B 31.1.0

DRESDEN - UFSAR Rev. 5 January 2003 Table 5.1-1 (Continued) REACTOR COOLANT SYSTEM DATA (Sheet 3 of 3) Electromatic Relief Valves Number 4 Capacity (each) 540,000 lb/hr Pressure setting

< 1112 or <

1135 psig Design code USAS B 31.1.0 Target Rock Safety Relief Valve Number 1 Capacity 622,000 lb/hr Pressure setting, relief <

1135 psig Pressure setting, safety 1135 psig Design code ASME Section III, 1968 Safety Valves Number 8 Capacity (each) Varies with setpoint

- see Table 5.2-1 Pressure setting Varies 1240 to 1260 psig Design code ASME Section III and USAS B 31.1.0

ISOLATION CONDENSER Number 1 Number of tube bundles 2

Design pressure Shell Tubing 25 psig at 300 o F 1250 psig at 575 o F Design code

Shell Tubing ASME Section VIII ASME Section III, Class A

• See Table 5.2-1 for specific value setpoint information DRESDEN - UFSAR Rev. 4 (Sheet 1 of 1)

Table 5.1-2 APPLICABLE REACTOR COOLANT SYSTEM P&IDs Unit 2 Unit 3 Topic P&ID No. P&ID No. Main steam piping M-12 M-345 Standby liquid control piping M-33 M-364 Condensate piping M-15 M-348 Condensate booster piping M-16 M-349 Reactor feed piping M-14 M-347 Pressure suppression piping M-25 M-356 Reactor recirculation piping M-26 M-357 Core spray piping M-27 M-358 Isolation condenser piping M-28 M-359 DRESDEN - UFSAR Rev. 3 (Sheet 1 of 1)

Table 5.1-3 Coolant Volumes (ft

3)

Unit 2 Unit 3 Lower Plenum 2196 2196 Upper Plenum 1216 1216

Steam Dome 6672 6672 Steam Line Piping (up to MSIVs) 1759 1759

DRESDEN - UFSAR Rev. 7 June 2007 5.3-1 5.3 REACTOR VESSELS This section presents pertinent data on the reacto r vessels. Unless otherwise noted, the information applies to both vessels.

The reactor vessel is a vertical cylindrical pressure vessel as shown in Figures 5.3-1 and 5.3-2.

The control rod drive housings and the incore instrumentation housings are welded to the bottom head of the reactor vessel.

The reactor vessel is supported by a steel skirt. The top of the skirt is welded to the bottom head of the vessel. The base of the skirt is continuously supported by a ring girder fastened to a concrete foundation, which carries the load through the drywell to the reactor building foundation slab (see Figure 6.2-1).

The reactor vessel head is flanged to the vessel and sealed with two concentric O-rings. The steam outlet lines are from the vessel body, below the reactor vessel flange.

The structural integrity of the reactor system is maintained at the level required by the ASME Section XI. The Systematic Evaluation Program (SEP) Topic V-6 reviewed aspects such as fracture toughness, surveillance programs, and neutron irradiation against ASME Section III, 1977 Edition including addenda through Summer 1978; 10 CFR 50, Appendices G and H; and Regulatory Guide 1.99-implemented 10 CFR 50.55a(g) requirements to assure reactor vessel integrity.

5.3.1 Reactor Vessel Materials The reactor pressure vessel (RPV) materials and fabrication methods conform to ASME Section III, Class A, 1963 Edition including Summer 1964 Addenda and including code case interpretations pertaining to primary nuclear reactor vessels applicable on the date of the purchase contract (see Reference 15). Inservice inspection (ISI) techniques conform to ASME Section XI with approved exceptions as noted in Section 5.2.4.

5.3.1.1 Reactor Vessel Materials Specification

The reactor vessel material supplied was ASME SA-302 Grade B, modified in accordance with ASME Code Case 1339, Paragraph 1. The nozzles and other attachments are as specified in the purchase specification in Reference 15.

5.3.1.2 Special Processes Used for Manufacturing and Fabrication The longitudinal weld joints were accomplished using the electroslag welding process. Details of this fabrication method are contained in Reference 16. Other welding processes, as applicable, were used in the fabrication of the reactor vessels.

DRESDEN - UFSAR Rev. 7 June 2007 5.3-2 Such processes as forging, extruding, and casting were applied to the fabrication of items used in the assembly of the reactor vessels.

5.3.1.3 Special Methods for Nondestructive Examination The reactor vessel plate was 100% ultrasonically examined before fabrication of the vessel but after forming and heat treatment. The plate areas where attachments are welded were ultrasonically inspected prior to joining the attachments.

The closure studs, nuts, bushings, and washers were ultrasonically inspected. Longitudinal and shear wave techniques were used. The longitudinal wave examination was performed on 100% of the stud cylindrical surface and from both ends of each stud.

Liquid penetrant examinations and magnetic particle examinations were utilized on the forgings and on the cladding and flange sealing surfaces. The cladding was ultrasonically inspected per the General Electric specification.

All full penetration welds on the vessels received 100% radiographic examination. Either magnetic particle or liquid penetrant examination was performed on the final pass of structural welds.

5.3.1.4 Special Controls for Ferritic and Austenitic Stainless Steels Regulatory Guides, as such, did not exist at the time when these reactor vessels were fabricated. Information related to specific Regulatory Guides (as requested in Regulatory Guide 1.70, Revision 3) is provided below to correlate actual past practice with current requirements. Unless otherwise stated, there has been no commitment to the Regulatory Guides.

5.3.1.4.1 Regulatory Guide 1.31

Application of Regulatory Guide 1.31 was not a consideration in fabrication of the reactor vessels.

The stainless steel cladding is ER-308L weld deposit using arcosite S-4 flux with the submerged arc welding process.

5.3.1.4.2 Regulatory Guide 1.34 Electroslag welding of longitudinal seams was performed in accordance with ASME Section III, Class A and Code Case 1355 as discussed in Reference 16.

DRESDEN - UFSAR Rev. 6 June 2005 5.3-3 5.3.1.4.3 Regulatory Guide 1.43 Control of undercladding cracking was achieved by use of a base material, chosen for its fine grain properties, which was not susceptible to the undercladding cracking observed on other materials with a coarse grain structure.

5.3.1.4.4 Regulatory Guide 1.44

Control of sensitized stainless steel usage was not considered in the fabrication of these vessels. As a result of several incidents elsewhere, the Unit 3 vessel underwent major modification and/or replacement of nozzle safe ends prior to installation and entering service. These modifications are addressed further in Sections 5.2.3.4 and 5.2.3.5. The Unit 3 safe end to the vessel nozzle weld received a weld overlay on the internal surface for the two recirculation outlet nozzles and the two steam outlet nozzles for the isolation condenser and high pressure coolant injection system.

5.3.1.4.5 Regulatory Guide 1.50 Preheat temperatures used when welding low-allow steel components (shell plates, flanges, nozzles) met applicable requirements or had contract variations approved by General Electric, the vendor responsible for supplying the reactor vessels.

5.3.1.4.6 Regulatory Guide 1.71

Welders were qualified as required by ASME Section III, Class A, and ASME Section IX. Any special requirements per the General Electric specification were also used in qualifying the welders and welding procedures.

5.3.1.4.7 Regulatory Guide 1.99

Regulatory Guide 1.99 states the regulatory position and discusses the methods used by the NRC staff in evaluating all predictions of radiation embrittlement of the reactor vessel beltline materials for implementation of 10 CFR 50, Appendices G an d H. Section 5.3.2 addresses further compliance with the methodology in Regulatory Guide 1.99, Revision 2.

5.3.1.4.8 Regulatory Guide 1.190 Regulatory Guide (RG) 1.190 provides state of the art calculation and measurement procedures that are acceptable to the NRC for determining Reactor Pressure Vessel (RPV) neutron fluence. RPV fluence has been evaluated using a method in accordance with the recommendations of RG 1.190. Future evaluations of RPV fluence will be completed using a method in accordance with the recommendation of RG 1.190 (as noted in Reference 15).

5.3.1.5 Fracture Toughness These reactor vessels were designed before specific rules for brittle fracture control in nuclear components were developed and before the material fracture properties in pressurized systems, other than the reactor vessel, were generally measured or DRESDEN - UFSAR Rev. 7 June 2007 5.3-4 controlled. The applicable industry codes and standards for the station, at that time, did not contain these rules for brittle fracture.

A report describing the ductile yielding analysis of the reactor vessel, including a discussion of the assumptions, methods of analysis, and conclusions, has been prepared. The report also addresses thermal shock and brittle fracture.

[1] A comprehensive tabulation of fracture toughness results on the reactor vessel plate material and welds is contained in Reference 15. Details of the qualification results for the electroslag welds are given in Reference 16. Section 5.3.2 presents additional details on the provisions for fracture toughness evaluations during the operating life of the reactor vessels. Section 5.3.1.6 addresses the reactor vessel material surveillance program.

5.3.1.6 Material Surveillance

Vessel material surveillance samples are located within the reactor vessel to enable periodic monitoring of material properties with exposure. The program includes specimens of the base metal, weld metal, and heat-affected zone metal. These specimens receive higher neutron fluxes than the vessel wall 1/4 T location and, therefore, lead it in integrated neutron flux. About 400 samples were initially inserted in the vessel; samples are periodically removed for Charpy V-notch and tensile strength tests.

The reactor vessel is a primary barrier against the release of fission products to the environment. In order to provide assurance that this barrier is maintained with a high degree of integrity, a materials surveillance program was developed and initiated at the beginning of nuclear operation of the reactors. This surveillance program was designed to be in conformance with the requirements of ASTM E185-62 with one exception. The base metal specimens of the vessels were made with their longitudinal axes parallel to the principal rolling direction of the vessel plate. Two material specimens were removed and tested under this original program.

In 2003, the NRC approved Dresden's participation in the BWR Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) as described in BWRVIP-78 and BWRVIP-86 (Reference 14). The NRC approved the ISP for the industry in Reference 14 and approved Dresden's participation in Reference 15. The ISP meets the requirements of 10 CFR 50 Appendix H and provides several advantages over the original program. The surveillance materials in many plant-specific programs do not represent the best match with the limiting vessel beltline materials since some were established prior to 10 CFR 50 Appendix H requirements. Also, the ISP allows for better comparison to unirradiated material data to determine actual shifts in toughness. Finally, for many plants, ISP data will be available sooner to factor into plant operations since there are more sources of data.

The withdrawal schedules for both units are shown in Table 5.3-1. The original withdrawal schedule was based on the three capsule surveillance programs as defined in Section 11.C.3.a of 10 CFR 50, Appendix H. The accelerated capsule (near core top guide) is not required by Appendix H but was tested to provide additional information on the vessel material. The results of the tests and examinations of these samples are used to generate the information addressed in Section 5.3.2. The current withdrawal schedule for both units is based on the NRC approved revision of BWRVIP-86 (Reference 14). Any changes to the withdrawal schedule must be approved by the NRC. All capsules placed in storage mst be retained for future insertion.

In the SEP Topic V-6, reactor vessel integrity was evaluated and the material surveillance program was found acceptable.

DRESDEN - UFSAR 5.3-5 5.3.1.7 Reactor Vessel Fasteners

The reactor vessel head closure studs are 6 inches in outside diameter and approximately 65 inches long with a 1.0-inch diameter bore hole. There are 92 closure studs for each reactor vessel. The stud material is specified as ASTM A 320 Grade L43 (ASME Code Case 1335, Paragraph 4, was also applied) which is a quench and tempered low-alloy steel that is similar to AISI Grade 4340.

[2]

General Electric performed an ASME Section III, Appendix E, code margin assessment for these units. For the vessel design pressure, the minimum calculated numbe r of studs is 79, whereas, the actual number is 92. Therefore, there is a significant margin.

[3] 5.3.1.8 Reactor Vessel Nozzle Safe Ends The safe ends originally ordered for the vessel were Type 304 and 316 stainless steel with 0.08% maximum carbon. These stainless steel safe ends were furnace-sensitized as a result of the furnace heat treatment of the reactor vessel. The circumferential welds in the Type 304 stainless steel pipe were not solution heat treated.

A list of stainless steel materials in the Unit 2 vessel (and attachments) which were subjected to heat treatment for stress relief is given in Table 5.2-3. A list of items which have not subsequently been replaced is given in Table 5.2-5. The safe ends which have been replaced are given in Table 5.2-4. Most of the safe ends on Unit 3 and internal attachments were modified as described in Section 5.2.3 and as stated in Tables 5.2-6 and 5.2-7.

Intergranular stress corrosion cracking (IGSCC) can occur when a special combination of materials, environment, and stress exist. If any of these conditions is not present, IGSCC will not occur. The susceptibility of austenitic stainless steels to IGSCC is high for furnace-sensitized, cold-worked material, and weld-sensitized material. Solution-annealed material, high-ferrite weld metal, or castings are low in susceptibility.

[4]

After reports were received of cracking in furnace-sensitized 300-series stainless steel safe ends on the nozzles of the Elk River and LaCrosse units, it was decided that, to the extent possible, considering time and material limitations, the furnace-sensitized safe ends on the Unit 3 vessel should be replaced (see Table 5.2-6 for initial modifications). Only four safe ends were not replaced: the two 28-inch diameter recirculation outlet nozzles and the two 14-inch diameter steam outlet nozzles for the isolation condenser and the high pressure coolant injection (HPCI) systems. The safe ends were examined carefully and no evidence of intergranular attack was found. Additionally, the internal surfaces of the safe ends were clad with Type 308 stainless steel laid down as axial stringer beads after the piping was welded on. Table 5.2-6 provides the details of the changes that were made. In making these changes on Unit 3, the area of sensitized metal exposed to the hostile environments was only reduced, not eliminated, since the base metal in the heat-affected zones adjacent to the welds was sensitized. Subsequently, some of DRESDEN - UFSAR Rev. 7 June 2007 5.3-6 these safe ends and welds have been replaced (see Tables 5.2-6 and 5.2-7). The inservice inspection program is in accordance with ASME Section XI and includes the additional requirements of Generic Letter 88-01 (see Section 5.2.4) and BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75).

Experience indicates that in the event of IGSCC, detectable leakage will occur before the crack grows to a critical size. Leaks can be detected by noting an increase in flow from the drywell equipment drain or floor drain sumps, high-temperature indication in some equipment drain lines, or unusual radioactivity buildup in filters and charcoal canisters in the drywell air sampling manifold system which are routinely monitored for increases in radioactivity. These particulate filters and charcoal canisters are taken daily through one of the three drywell sample lines with automatic isolation valves.

The reactor vessels are designed and fabricated in accordance with ASME Section III. Paragraph N151 of ASME Section III defines the boundary between the vessel and the piping as the first circumferential joint, exclusive of the connecting weld in welded connections. This means that the nozzles are part of the vessel and they must be designed, manufactured, and attached in the vessel wall in accordance with ASME Section III; however, the safe ends and the welds attaching them to the nozzle are not a part of the vessel but a part of the piping system.

The Illinois Board of Boiler Rules recognized that, at the time the Dresden units were purchased, there was no code for nuclear piping applicable in their jurisdiction. Therefore, the board required that the piping attached to the vessel and extending out to and including the first shutoff valve should be designed, fabricated, and installed in accordance with ASME Section I, as well as USAS B31.1. The board also required that the hydrostatic test pressure should be 125% of design pressure for the system comprising the boiler rather than the 150% specified in ASME Section I.

The furnace-sensitized safe ends were carefully removed from the Unit 3 vessel nozzles and the new safe ends were welded to the nozzles using welders and procedures qualified in accordance with ASME Section IX. The Illinois Board of Boiler Rules had agreed that since no nuclear piping code was available to their jurisdiction, stamping would not be required. Instead, the board decided that modified P4A forms or letter certificates of compliance signed by a qualified inspector would be acceptable evidence that the piping contained approved materials, that it had been installed by an ASME Code qualified contractor using qualified welders working in accordance with a qualified procedure, and that the work had been inspected by a qualified inspector.

Since inservice inspections on Unit 3 revealed indications of IGSCC, piping replacement was performed. IGSCC-susceptible piping and the associated safe ends were replaced during the 1985-86 recirculation pipe replacement (RPR) outage (Reference 17 and Table 5.2-6). The material employed was Type 316 Nuclear Grade stainless steel whose chemical properties provide more resistance to weld sensitization than the old material. Solution heat treatment was performed on all shop welds (completed at Mannesmann Manufacturing located in Germany), and mechanical stress improvement (MSIP) was performed on a number of core spray and isolation condenser system welds (not replaced during the outage). Table 5.2-6 provides the details of those safe ends replaced.

DRESDEN - UFSAR Rev. 8 June 2009 5.3-7 Some safe ends have been replaced on Unit 2. These were not replaced prior to unit operation, as on Unit 3; rather, the Unit 2 safe ends were replaced during subsequent outages. A list of material in the Unit 2 vessel which was originally furnace-sensitized stainless steel (see Table 5.2-3) but which has not been replaced is given in Table 5.2-5. The Unit 2 safe ends which have been replaced are given in Table 5.2-4.

The core spray piping out to the second isolation valve, the core spray nozzle safe ends, and the core spray thermal sleeves were replaced (see further the discussion in Section 5.2.3).

Feedwater nozzle cladding cracks were observed during vessel nozzle inspections. The observed cracks were caused by thermal cycling which occurred in the annulus between the nozzle and sparger thermal sleeve. The stainless steel cladding flaws were ground out and interference fit thermal sleeves were installed. The final modification to the feedwater nozzle sparger thermal sleeves resulted in the removal of the stainless steel cladding from the feedwater nozzle and installation of the dual-seal, triple-sleeve sparger configuration supplied by General Electric and recommended by NUREG-0619.

Routine inspections of the feedwater nozzles and spargers are performed in accordance with Table 2 of the NRC SER (TAC M94090) of the BWR Owners Group proposed alternate inspections to NUREG-0619 (General Electric report GE-NE-523-A71-0594).

5.3.2 Pressure-Temperature Limits The design temperature for various system components varies according to the specific operating condition. The design temperature for the reactor vessel is based on the saturation temperature corresponding to the design pressure. Therefore, no specific system temperatures are designated as safety or operating limits.

Neutron radiation exposure above 10 17 n/cm 2 (greater than 1 MeV) begins to affect the mechanical properties of ferritic steel. The most important consideration is that of the change in the temperature below which ferritic steel breaks in a brittle rather than a ductile mode (referred to as the NDT temperature). The NDT temperature increases with increasing neutron exposure. ASME Section III, N-446 specifies that, for neutron irradiated areas of the vessel, there should be no nozzle or other structural discontinuities. The design conditions for determination of the NDT temperature is specified in ASME Section III, N-330. Extensive tests have established the magnitude of changes in the NDT temperature as a function of the integrated neutron dosage.

The present pressure temperature operating curves shown in Figure 5.3-3 through 5.3-7 are calculated using the methodology and data from Regulatory Guide 1.99, Revision 2.

DRESDEN - UFSAR Rev. 3 5.3-8 5.3.2.1 Limit Curves

The reactor is a primary barrier against the release of fission products to the environment. In order to provide assurance that this barrier is maintained at a high degree of integrity, pressure and temperature limits have been established for the operating conditions to which the reactor vessel may be subjected. Figure 5.3-3 through 5.3-7 presents the pressure-versus-temperature curves for the operating conditions: Pressure testing (5.3-3 through 5.3-5), nonnuclear heatup and cooldown (5.3-6), and core critical operation (5.3-7). These curves have been established to be in conformance with 10 CFR 50, Appendix G, and Regulatory Guide 1.99, Revision 2. The curves take into account the change in the reference nil-ductility transition temperature (RT NDT) as a result of neutron embrittlement. The adjusted reference temperature (ART) of the limiting vessel material is used to account for irradiation effects.

[5] 5.3.2.1.1 Beltline, Nonbeltline, Closure Flange Regions, and Bottom Head regions Four vessel regions are considered for the development of the pressure and temperature curves: the core beltline region, the nonbeltline region (other than the closure flange region and the bottom head region), the closure flange region, and the bottom head region. The beltline region is defined as that region of the reactor vessel that directly surrounds the effective height of the reactor core (between the bottom and top of active fuel) and is subject to a RT NDT adjustment to account for irradiation embrittlement. The nonbeltline, closure flange regions, and bottom head region receive insufficient fluence to necessitate a RT NDT adjustment. These regions contain components which include: the reactor vessel nozzles, closure flanges, top and bottom head plates, control rod drive penetrations, and shell plates that do not directly surround the reactor core. The closure flange region, and bottom head region, although not a nonbeltline regions, are treated separately from the nonbeltline region for the development of the pressure-temperature curves to address 10 CFR 50, Appendix G requirements.

5.3.2.1.1.1 Boltup Temperature The limiting initial RT NDT of the main closure flanges, the shell and head materials connecting to these flanges, connecting welds and the vertical electroslag welds which terminate immediately below the vessel flange is 23°F. Therefore, the minimum allowable boltup temperature is established as 83°F (RTNDT + 60°F) which includes a 60°F conservatism required by the original ASME Code.

5.3.2.1.1.2 Figures 5.3-3 through 5.3 Pressure Testing DRESDEN - UFSAR Rev. 3 5.3-9 As indicated in figures 5.3-3 through 5.3-5 for pressure testing, the minimum metal temperature of the rector vessel shell is 83°F for reactor pressures less than 312 psig. This 83°F minimum boltup temperature is based on a RT NDT of 23°F for the electroslag weld immediately below the vessel flange and a 60°F conservatism required by the original ASME Code. The bottom head region limit is established as 68 °F based on moderator temperature assumptions for shutdown margin analysis.

At reactor pressures greater than 312 psig, the minimum vessel metal temperature is established as 113°F. The 113°F minimum temperature is based on a closure flange region RT NDT of 40°F and 90°F conservatism required by 10 CFR 50, Appendix G.

5.3.2.1.1.3 Figure 5.3 Nonnuclear Heatup and Cooldown Figure 5.3-6 applies during heatups with nonnuclear heat (e.g., recirculation pump heat) and during cooldowns when the reactor is not critical (e.g., following a scram). The curve provides the minimum reactor vessel metal temperatures based on the most limiting vessel stress.

5.3.2.1.1.4 Figure 5.3 Core Critical Operation

The core critical operation curve shown in Figure 5.3-7, is generated in accordance with 10 CFR 50, Appendix G which requires core critical pressure and temperature limits to be 40°F above any Pressure Testing or Non Nuclear Heatup/Cooldown limits. Since Figure 5.3-6 is more limiting, Figure 5.3-7 is equal to Figure 5.3-6 plus 40°F.

5.3.2.1.2 Procedure for Updating Limit Curves When credible surveillance data from the reactor vessel are not available, calculation of the neutron radiation embrittlement of the beltline materials should be based on the requirements outlined in Regulatory Guide 1.99.

[6] When two or DRESDEN - UFSAR Rev 7 June 2007 5.3-10 more credible surveillance data sets are available for the reactor vessel, they may be used to determine the ART and the Charpy upper shelf energy for the beltline materials. The application of the methods described in Regulatory Guide 1.99

[6] and used in the General Electric report

[5] must also consider the chemistry factor as applied in the Regulatory Guide. In such applications the controlling factor for the pressure-versus-temperature curves may not be the beltline materials, as is evidenced for Dresden where the electroslag weld metal below the closure flange was found to be the controlling point.

5.3.2.2 Operating Procedures Reactor operating procedures are utilized to implement the operating criteria and limitations specified in the Technical Specifications. The heatup and cooldown rates and pressure changes are coordinated so as to remain within the specified limitations. The water chemistry control and hydrogen addition are implemented in accordance with operating procedures specific for each unit.

5.3.3 Reactor Vessel Integrity

The structural integrity of the primary system boundary shall be maintained at the level required by ASME Section XI.

The Babcock and Wilcox Company (B&W) designed and fabricated the reacto r vessels purchased by GE, who supplied the vessels to CECo for the Dresden Station. The Hartford Company had the responsibility for third party inspection at B&W and signed both the data reports and the ASME Code N-1A forms (Reference 15).

This section and the following subsections summarize the reactor vessel's purpose and the factors that contribute to its integrity.

5.3.3.1 Design 5.3.3.1.1 General Parameters The performance objectives of the reactor vessel are to contain the reactor core, the reactor internals, and the reactor core coolant-moderator and to serve as a high-integrity barrier against leakage of radioactive materials to the drywell. To achieve these objectives, the reactor vessel was designed using the following bases:

A. Design pressure 1250 psig B. Base metal ASME SA-302 Grade B, Modified (Code Case 1339)

C. Cladding material Weld deposited E-308 electrode

D. Design code ASME Section III-A DRESDEN - UFSAR 5.3-11 The nominal operating pressure of 1005 psig has been chosen on the basis of economic analyses for boiling water reactors. The reactor vessel design pressure of 1250 psig was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation (with additional allowances to accommodate transients above the operating pressure without causing actuation of the safety valves).

The strength required to withstand external and internal loadings, while maintaining a high degree of corrosion resistance, dictated the use of a high-strength carbon alloy steel, SA-302 Grade B, Modified (Code Case 1339) with an internal cladding of Type 308 stainless steel applied by weld overlay. The reactor vessel was designed for a 40-year life. It will not be exposed to more than 10 19 n/cm 2 of neutrons with energies exceeding 1 MeV.

The ASME Section III, Class A design criteria provides assurance that a vessel designed, built, and operated within its design limits has an extremely low probability of failure due to any known failure mechanism.

The reactor vessel was designed and built in accordance with ASME Section III, Class A. General Electric specified additional requirements. Records of material properties were developed and retained for later evaluation of the reactor vessel during all operating conditions.

5.3.3.1.2 Specific Criteria The design stress (general membrane) permitted by ASME Section III (1965) for design purposes is the lesser of one-third of the minimum ultimate strength (S ult) or two-thirds of the minimum yield strength (S y). For the pressure vessel material, SA 302 Grade B, Modified (Code Case 1339), at a temperature of 550°F, the design stress is found to be 1/3 S ult or 1/3 x 80,000 psi = 26,700 psi. Therefore, the maximum allowable pressure stress that can be tolerated without failure is estimated to be 80,000 psi (i.e., the ultimate strength of the material).

A vessel stress report was prepared by B&W. An independent review of this report was conducted by GE as each section was received, and a certified report was issued upon completion of the analysis.

Bell-mouthing of the reactor vessel is applicable only to vessels with breech closure, or a closure made by screwing the reactor vessel head into the reactor vessel. Since the reactor vessel head is flanged to the vessel bell-mouthing is not applicable.

5.3.3.1.3 Temperature and Pressure Cycles

The reactor vessel design cycles are presented in Section 3.9.

5.3.3.1.4 Static and Dynamic Loadings

DRESDEN - UFSAR Rev. 7 June 2007 5.3-12 Stabilizer brackets, located below the vessel flange, are connected with flexible couplings to tension bars mounted on the top of the reactor shield wall. The reactor shield wall is laterally supported by stabilizers which are attached to gibs passing through the drywell wall and embedded in the concrete structure outside the drywell. The lateral supports limit horizontal vibration and resist seismic and jet reaction forces, yet the tension bars will permit radial and axial expansion.

Vertical loads from the reactor vessel are transmitted to the foundation through the vessel skirt, support girder, and support pedestal. Lateral loads are transmitted to the building through vessel stabilizers. The vessel stabilizers are attached near the top third of the vessel and are connected to the top of the concrete and steel reactor shield wall. The reactor shield wall in turn is anchored at the base to the top of the vessel pedestal and restrained at the top by a horizontal tubular truss system. The lateral loads are transmitted thro ugh the truss system to the drywell shear lug mechanism. This shear lug mechanism permits vertical movement of the steel drywell, but restricts rotational movement. However, lateral loads are transmitted through the shear lug mechanism to the heavy concrete envelope around the drywell which is part of the reactor building. A portion of the lateral loads are transmitted from the reactor vessel to the vessel pedestal and then to the foundation. Additional details of the loadings and supports are addressed in Section 3.9.

5.3.3.2 Materials of Construction The materials of construction for the reactor vessels are addressed in Sections 5.2.3.1 and 5.3.1.1 and in the purchase order and vessel design report contained in Reference 15.

5.3.3.3 Fabrication Methods The methods used in fabricating the reactor vessels are addressed in Sections 5.2.3.3 and 5.2.3.4, Section 5.3.1, and Appendices 5A and 5B. The major fabrication processes involved are electroslag welding, submerged arc welding, hot rolling of thick vessel plate, forging of nozzles and vessel closure flanges, and stress-relieving heat treatments. A summary of the reactor vessel fabrication history is presented for both units in References 7 through 9.

5.3.3.4 Inspection and Testing Requirements The reactor vessel was stamped with an ASME Code N-symbol verifying that a hydrostatic test was satisfactorily made and all other required inspection and test ing was satisfactorily completed. Such application of the ASME Code N-symbol together with final certification confirms that all applicable ASME Code requirements have been complied with.

The reactor coolant system was given a system hydrostatic test in accordance with code requirements prior to initial reactor startup. Before pressurization, the system was heated to 60°F above the NDT temperature. Piping and support DRESDEN - UFSAR 5.3-13 hangers were checked while thermal expansion was in progress. Recirculation pump operation was also checked.

A system leakage test at operating pressure is made on the primary system following each removal and replacement of the reactor vessel head. The system is checked for leaks and abnormal conditions which are corrected before reactor startup. The minimum vessel temperature during

system hydrostatic testing and system leakage testing is governed by the Technical Specifications in accordance with the curves shown in Figure 5.3-3.

System hydrostatic tests are performed after repair or replacement to the system. The hydrostatic test pressure and testing conditions are detailed in ASME Section XI and Figure 5.3-3.

Periodic examinations and tests to ensure system integrity

[10, 11] are carried out as part of an ongoing ISI program as required by ASME Section XI. See Section 5.2.4 for a more detailed description.

5.3.3.5 Shipment and Installation The reactor vessels, closure heads, closure head studs, and the nuts were packaged and shipped in accordance with the purchase specification. The reactor vessels were shipped on skids which were an aid in uprighting the vessels in preparation for setting them in place. Hoisting slings were provided for lifting both the reactor vessels and the closure heads. General Electric Quality Assurance personnel assured that all shipments and installation met the appropriate regulations and requirements.

5.3.3.6 Operating Conditions The reactor vessel is designed for the anticipated transients which are expected to occur or could occur during the designed 40-year life.

5.3.3.6.1 Original Design Basis Transients and Cycles

The original design basis transients and estimated cycles are presented in Table 3.9-1 and are addressed in Section 3.9.1.1.

5.3.3.6.2 Revised Design Basis Transients and Cycles

The revised design basis transients and cycles

[12] are presented in Table 3.9-1 and are addressed further in Section 3.9.1.1 along with the fatigue analysis of the reactor vessels.

DRESDEN - UFSAR Rev. 6 June 2005 5.3-14 5.3.3.7 Inservice Surveillance The ISI program delineates and implements the requirements of 10 CFR 50.55a and the ASME Code Section XI.

In the ISI plan, certain examination requirements as stated cannot be performed. Therefore, relief requests are filed in accordance with 10 CFR 50.55a(g)(6)(i).

The ISI and augmented inspection programs are addressed further in Section 5.2.4.

Section 5.3.1.6 addresses the material surveillance program and the withdrawal schedule for internal samples exposed at the beltline region of the reactor vessels.

The reactor material surveillance programs is addressed in Section 5.3.1.6.

DRESDEN - UFSAR Rev. 7 June 2007 5.3-15 5.3.4 References

1. L.C. Hsu, A Comprehensive Analysis of the Structural Integrity of GE-BWR Vessels Subject to the Design Basis Accident, November 1968.

2. Letter from M.H. Richter (CECo) to T.E. Murley (NRC), dated July 3, 1991, Reactor Vessel Head Closure Studs.

3. H.G. Mehta (GE), Fracture Mechanics Based Structural Margin Evaluation for Commonwealth Edison BWR Reactor Pressure Vessel Head Studs, GE-NE-523-93-0991, DRF 137-0010, September 1991.

4. H.H. Klepfer, et al., Investigation of Cause of Cracking in Austenitic Stainless Steel Piping, Volume 1, NEDO-21000, General Electric, July 1975, p. 8-1.

5. T.A. Caine (GE), Pressure-Temperature Curves per Regulatory Guide 1.99, Revision 2 for the Dresden and Quad Cities Nuclear Power Stations, SASR 89-54, DRF 137-0010, Revision 1, August 1989.

6. NRC Regulatory Guide 1.99, Revision 2, May 1988, Radiation Embrittlement of Reactor Vessel Materials.

7. Letter from R. Stols (CECo) to T.E. Murley (NRC), dated July 2, 1990, Reactor Vessel Fabrication History Summary (Transmitting Document 508-9006, Dresden II Upper Vessel Contract Variation Review by General Electric Company, June 29, 1990).

8. Letter from M.H. Richter (CECo) to T.E. Murley (NRC), dated September 4, 1990, Summary of Fabrication History for the Unit 3 Upper Reactor Vessel.

9. Letter from R. Stols (CECo) to T.E. Murley (NRC), dated January 3, 1991, Reactor Vessel Fabrication History Summary.

10. S.P. Selby and W.E. Brooks, "CRDM Nozzle Inspection," Nuclear Plant Journal, November/December 1992, pp. 56ff.

11. S. Ranganath and T.L. Chapman, "Inservice Inspection Experience in Boiling Water Reactors," Nuclear Plant Journal, November/December 1992, pp. 77ff.

12. T.A. Caine (GE), Tabulation of Thermal Cycles for Dresden Nuclear Power Station Units 2 and 3, SASR 89-111, Revision 2, November 1990.

13. BWRVIP-86-A: "BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP)," Final Report, October 2002.

14. Letter from M. Banerjee (U.S. NRC) to J. L. Skolds, dated September 29, 2003.

15. "Dresden 2 Reactor Pressure Vessel Design Exhibits."

16. "Dresden Station Units 2 and 3 Reactor Vessel Electroslag Weld Report."

17. Dresden Station Unit 3 Recirculation Pipe Replacement (RPR) Project Completion Report.

DRESDEN - UFSAR Rev. 6 June 2005 (Sheet 1 of 1)

Table 5.3-1

NEUTRON FLUX MONITOR AND BASE METAL SAMPLE WITHDRAWAL SCHEDULE

Withdrawal Schedule(1) Part Number Location Comments Unit 2 1977 6 Near core top guide - 180

° Accelerated sample 1980 8 Wall - 215

° 2003 7 Wall - 95

° 2003 9 Wall - 245

° 10 Wall - 275

° Standby Unit 3 1978 16 Near core top guide - 180

° Accelerated Sample 1981 18 Wall - 215

° 30 EFPY 19 Wall - 245

° 15 Wall - 65

° Standby 20 Wall - 275

° Standby

Note: 1. Withdrawals completed are listed by year withdrawn. Future withdrawals are listed by the effective full power years (EFPY) anticipated at withdrawal.