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DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.1 Page 1.1-1 of 1.1-3
1.1 INTRODUCTION
1.1.1 GENERAL On October 19, 2017, Entergy Nuclear Operations, Inc. (Entergy) notified the U.S. Nuclear Regulatory Commission (NRC) that it would permanently cease power operations at Palisades Nuclear Plant (PNP) no later than May 31, 2022. On June 13, 2022, Entergy submitted certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel to the NRC in accordance with 10 CFR 50.82(a)(1)(i) and (ii).
Following the NRC docketing those certifications, the 10 CFR Part 50 license no longer permits operation of the reactor or placement of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2).
This Defueled Safety Analysis Report (DSAR) is derived from Revision 35 of the PNP Updated Final Safety Analysis Report (UFSAR). The DSAR is a licensing basis document that reflects the permanently defueled condition of PNP and supersedes revision 35 of the UFSAR. The DSAR is intended to serve the same function during SAFSTOR and decommissioning that the UFSAR served during operation of the facility.
For the purposes of 10 CFR 50.59 reviews or other activities that reference the UFSAR, the DSAR constitutes the safety analysis report reflective of the permanently shut down and defueled facility following the docketing of the certifications required in 10 CFR 50.82(a)(1) in accordance with 10 CFR 50.82(a)(2). The term DSAR is utilized in lieu of the term UFSAR. The DSAR is updated consistent with the requirements of 10 CFR 50.71(e).
By NRC order dated April 11, 2007 (Reference 1.1-1), Consumers Powers ownership/operation of PNP was transferred to Entergy Nuclear Palisades, LLC, as the owner of PNP, and Entergy Nuclear Operations, Inc. as the operator of PNP. Consequently, references to Consumers Power (or derivatives thereof) in this document remain only when used in historical context.
The remainder of the sections of Chapter 1 summarize the principal design features and parameters of the facility.
1.1.2 LICENSING HISTORY Consumers Power filed, on Docket 50-255, a Construction Permit and Operating License (CPOL) Application (which included the PSAR) to the AEC on June 2, 1966 for the Palisades Plant to be located near South Haven, Michigan. The application was for development of a 2,650 MWt (design core power) commercial nuclear-powered electrical generating facility to be operated at 2,200 MWt or an equivalent electrical output of 700 MWe. On March 14, 1967, the AEC issued Consumers Power a Class 104 Construction Permit CPPR-25, pursuant to Section 104(b) of the Atomic Energy Act, to construct a Combustion Engineering pressurized water reactor (PWR) with a
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.1 Page 1.1-2 of 1.1-3 full-power design rating of 2,650 MWt. Subsequent to the original CPOL/PSAR application, eight additional PSAR amendments were filed addressing NRC concerns and addition of the Technical Specifications. On November 1, 1968, Consumers Power filed with the AEC an Operating License (OL) Application (which included the FSAR) as Amendment 9 to the CPOL Docket 50-255, to operate the Palisades Plant at 2,200 MWt core power. Following submission of the initial FSAR as Amendment 9 to the CPOL, 23 subsequent amendments to the FSAR were submitted to the NRC.
They were identified as Amendments 10 through 32 to the CPOL Docket 50-255, the most extensive of which was the Full-Term Operating License (FTOL) Application, Amendment 28. In addition, nine minor revisions were subsequently submitted to the NRC.
On March 24, 1971, the NRC issued Interim Provisional Operating License IDPR-20 to be effective for 1-1/2 years to operate Palisades up to 1 MWt.
Subsequent amendments to IDPR-20 were issued on November 20, 1971 to operate up to 440 MWt (20% power); March 10, 1972 to operate up to 1,320 MWt (60% power); September 1, 1972 continued operation at 1,320 MWt (60% power); October 16, 1972 authorized operations for 2,200 MWt (100% power - limited to 60% power); and March 23, 1973 authorized operations for 2,200 MWt (100% power - limited to 85% power).
That Operating License has since been amended numerous times to keep the Facility current with NRC standards and to reflect Facility modifications.
On January 22, 1974, Consumers Power requested conversion of the Provisional Operating License DPR-20 to a Full-Term Operating License to operate at 2,638 MWt (845 MWe gross) for a period of 40 years from the date of the issuance of the Construction Permit. As part of the FOL Application, which was submitted as Amendment 28 to Docket 50-255, a complete amendment to the FSAR was provided. This FSAR amendment included major revisions based upon:
- 1.
Incorporation of information related to the once-through circulating water conversion to a closed-cycle system with mechanical draft cooling towers and water treatment chemistry changes
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Increase in operating power to 2,638 MWt core power
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Incorporation of information related to Radwaste System modifications implemented to obtain conformance to Appendix I of 10 CFR 50 NRC action on the request for an authorization to increase operating power and a Full-Term Operating License was delayed. The Provisional Operating License remained in effect indefinitely beyond its expiration date, however, under 10CFR2.109.
On August 12, 1977, Consumers Power requested that the Provisional Operating License limit of 2,200 MWt be increased to 2,530 MWt based upon reanalysis of safety evaluations and the improvements made with steam
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.1 Page 1.1-3 of 1.1-3 generator repairs. On November 1, 1977, the NRC granted Amendment 31 to DPR-20, authorizing operation of the facility at 2,530 MWt core power.
On February 21, 1991 the NRC issued the Full Term Operating License. This license was based on an Environmental Assessment dated October 22, 1990 and an SER issued as NUREG 1424 on November 21, 1990. The license expiration date is specified as midnight on March 14, 2007.
On November 30, 1999, the NRC issued Amendment 189 to the Palisades Operating Licensee to approve the conversion of Appendix A, Technical Specifications, from the original plant-specific format to a format more consistent with Standard Technical Specifications, Combustion Engineering Plants,NUREG-1432. These were referred to as Improved Technical Specification until implementation, which occurred on October 24, 2000.
On December 14, 2000, the NRC issued Amendment 192 to the Palisades Operating License to extend the license expiration date from March 14, 2007 to March 24, 2011. This action recaptured the Palisades construction period and provided for 40 years of licensed operation.
On June 23, 2004, the NRC issued Amendment 216 to the Palisades Operating License to authorize operation of the facility at steady state reactor core power levels up to 2565.4 Megawatts thermal.
On January 17, 2007, the Renewed Facility Operating License was issued by the NRC, extending the license expiration date to March 24, 2031.
On April 11, 2007, the NRC issued Amendment 224 to the Palisades Renewed Facility Operating License to reflect the transfer of ownership to Entergy Nuclear Palisades, LLC, and operating authority Entergy Nuclear Operations, Inc.
On May 13, 2022, the NRC issued Amendment 272 to the Palisades Renewed Facility Operating License to reflect the permanent cessation of operations at PNP and permanent removal of fuel from the PNP reactor vessel. On June 13, 2022, the licensee certified to the NRC that the Palisades Nuclear Plant had both permanently ceased operations (final shutdown May 20, 2022) and that all fuel had been removed from the reactor vessel and placed in the spent fuel pool On December 13, 2021, the NRC approved transfer of ownership of the Palisades facility from Entergy Nuclear Palisades, LLC, to Holtec Palisades, LLC (Holtec Palisades) and the transfer of operating authority of Palisades facility to Holtec Decommissioning International, LLC (HDI). On June 28, 2022, the NRC issued Amendment 273 to the Palisades Renewed Facility Operating License to reflect the transfer of ownership/operation to Holtec Palisades and HDI, respectively.
A chronological history of these license events is provided on Table 1-1.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.2 Page 1.2-1 of 1.2-4 1.2 GENERAL FACILITY DESCRIPTION 1.2.1 FACILITY SITE The site for the Palisades Facility consists of approximately 432 acres on the eastern shore of Lake Michigan, in Covert Township, approximately four and one-half miles south of South Haven, Michigan. The area adjacent to the site is sparsely populated and is primarily farmland. The population along the lake increases during the summer months. See Subsection 2.1.2 for details on demography and Figure 2-2 for site layouts.
The exclusion area for Palisades is defined as the property boundary shown on Figure 2-2. The minimum exclusion distance for the site is approximately 2,300 feet (667 meters) and the nearest population center area of more than 24,000 residents is constituted by the cities of Benton Harbor and St Joseph which are approximately 16 miles south of the site.
1.2.2 FACILITY ARRANGEMENT Figure 1-1, Facility Site Plan and Facility Area Plan, displays the primary power block structures arrangement. The turbine building for the Palisades Facility is oriented parallel and adjacent to the shoreline of Lake Michigan, with the reactor containment building located on the east, or landward, side of the turbine building. The office and auxiliary facilities are situated east of the north end of the turbine building so that the entire complex is L-shaped. The reactor containment structure is located inside the corner of this "L."
Equipment layouts are shown in Figures 1-2 through 1-18.
The containment building houses the NSSS, consisting of the reactor, steam generators, primary coolant pumps, pressurizer and some of the reactor auxiliaries. The containment building is served by a circular bridge crane.
The turbine building houses the turbine generator, condenser, feedwater heaters, condensate and feed pumps, turbine auxiliaries and certain of the switchgear assemblies. The north end of the turbine building provides additional shop, laboratory and office space.
The auxiliary building and auxiliary building addition (radioactive waste building) houses the waste treatment facilities, engineered safeguards components, heating and ventilating system components, the emergency diesel generators, switchgear, laboratories, offices and the control room. The spent fuel pool and the new fuel storage facilities are located in a separate section of the auxiliary building (Chapter 9) which is under controlled ventilation whenever spent fuel is being moved or stored in that section.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.2 Page 1.2-2 of 1.2-4 The condensate and makeup demineralizer building (feedwater purity building) was constructed during the feedwater purity modification. It houses the raw water filtration system, the reverse osmosis pretreatment system, the makeup demineralizer system, various components of the condensate demineralizer system, regeneration chemicals handling system, feedwater purity service and instrument air, chemical storage and a boiler room.
The intake structure houses the service water and fire protection pumps.
Prior to converting the Facility from once-through cooling to closed-cycle cooling, this building contained the circulating water pumps.
The cooling tower pump house contains two vertical pumps with sufficient head capacity to circulate the tube side condenser cooling water up to the cooling tower inlet near the tower top. The cooling tower basins are elevated some 20 feet above the Facility.
The circulating water cooling towers are cross-flow mechanical draft, located approximately 500 and 1,000 feet from the Facility. One tower contains 18 cells and the other tower contains 16 cells. Both towers are designed for a 32qF range.
1.2.3 CONTAINMENT The containment building uses a prestressed concrete design. The building is a vertical right cylindrical structure with a dome and a flat base. The building interior is lined with carbon steel plate for leak tightness. Inside the structure, the reactor and other NSSS components are shielded with concrete. An unlined steel ventilation stack is attached to the outside of the containment building and extends to an elevation equal to the top of the containment dome.
The original structure design conditions are an internal pressure of 55 psig, a coincident temperature of 283qF and a leak rate of 0.1% per day by weight at design temperature and pressure.
1.2.4 NUCLEAR STEAM SUPPLY SYSTEM (NSSS)
The NSSS consists of a pressurized water reactor with two closed loops. The principal components and supporting systems of the NSSS are the reactor vessel, internals, control rods, control rod drives, slightly enriched fuel, two "U" tube steam generators, four primary coolant pumps, primary system piping, pressurizer, quench tank, Chemical and Volume Control System, Safety Injection System, nuclear and process instrumentation, and the Reactor Protective System. See Table 1-2 for equipment design.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.2 Page 1.2-3 of 1.2-4
- 1.
Reactor Vessel and Internals The reactor vessel and its removable hemispherical closure head are fabricated from carbon steel and are lined with 308/309 stainless steel.
In the areas of internal attachments, the interior is clad with Ni-Cr-Fe alloy. A fixed hemispherical head is attached to the lower end of the shell. The reactor vessel is supported on three pads welded to the underside of the coolant nozzles.
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Steam Generators The two steam generators are vertical shell and "U" tube units (see Table 4-4).
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Primary Coolant Pumps The coolant in the primary loop is circulated by four primary coolant pumps of the single suction centrifugal type. The pump shafts are sealed by mechanical seals.
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Primary System Piping Each of the two loops which make up the Primary Coolant System consists of one 42-inch ID pipe and two 30-inch ID pipes.
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Nuclear Control and Instrumentation
- a.
Reactor Neutron Monitoring The nuclear instrumentation consists of excore and incore flux monitoring chambers.
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Process Instruments The facility gaseous and liquid effluents are monitored for radioactivity. The levels are displayed and recorded and high values are annunciated. Area monitoring stations are provided to monitor radioactivity at selected locations around the Facility.
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Safety Injection System Four safety injection tanks are provided, each connected to one of the four reactor inlet lines. Each tank has a volume of approximately 2,000 cubic feet.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.2 Page 1.2-4 of 1.2-4
- 7.
Shielding Shielding is provided so that radiation exposure of personnel will not exceed the recommended limits of 10 CFR, Part 20. The design of radiation shielding is dependent both on the extent of access required to a particular location and on the sources of radiation adjacent to that location.
The control room is shielded to permit continuous occupancy following any accidental release of radioactivity.
1.2.5 TURBINE GENERATOR The turbine is an 1,800 r/min tandem-compound unit with external moisture separation and live steam reheating.
The feedwater cycle is of the closed type with deaeration effected in the condenser. Feedwater heaters are arranged in two parallel trains, each with one high-pressure and five low-pressure heaters.
The 1,800 r/min, hydrogen inner-cooled generator is rated at 955,000 kVA at 75 psig hydrogen pressure, 0.85 power factor and 0.62 short circuit ratio.
Field excitation is provided by a brushless exciter directly coupled to the generator shaft.
The turbine generator has a guaranteed capability of 811,776 kWe gross at 1.8 inches Hg absolute back pressure and 0.25% makeup with inlet steam conditions of 735 psia and 509°F. The maximum calculated capacity of the turbine generator is 865 MWe gross.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.3 Page 1.3-1 of 1.3-1 1.3 IDENTIFICATION OF CONTRACTORS Consumers Power engaged Combustion Engineering, Inc (Combustion Engineering) to design and supply the nuclear fuel and the NSSS. The NSSS includes the primary system (eg, reactor vessel, steam generators, pressurizer, pumps), reactor auxiliary system components, nuclear and certain process instrumentation and the Reactor Protective System. Bechtel Corporation and its affiliate, Bechtel Company, were engaged to design and supply the balance of the Plant equipment, systems and structures. Bechtel Corporation performed the onsite construction of the original Plant, with technical advice and consultation provided by Combustion Engineering for installation of the NSSS. Subsequent to the initial Plant start-up and turnover to Consumers Power, several major modifications involving other contractors have been undertaken. Those contractors are identified in Section 1.5.
Under its contract with Consumers Power, Combustion Engineering furnished Bechtel with the design data for the NSSS. Bechtel and Consumers Power could request that Combustion Engineering make changes in the NSSS design, but Combustion Engineering did not need to accede to any such request if the proposed change, in Combustion Engineering's judgment, would be unsafe or technically unsound.
Because of the interdependence of the NSSS and certain balance-of-Plant equipment, systems and structures, Combustion Engineering furnished Bechtel with certain functional requirements for such balance-of-Plant items that affect the operability and maintainability of the NSSS or the nuclear safety of the Palisades Plant. As Bechtel's engineering work progressed, Combustion Engineering reviewed Bechtel drawings, specifications and data and Combustion Engineering was satisfied that Bechtel has understood and applied the functional requirements specified by Combustion Engineering and was satisfied that the balance-of-Plant items are compatible with the NSSS and with nuclear safety.
Palisades' original fuel vendor for cycle 1 was Combustion Engineering.
Starting with cycle 2, Exxon Nuclear Corporation designed and manufactured all fuel for the reactor. Over the years, Exxon Nuclear has undergone the following company name changes: from Exxon Nuclear Corporation (ENC), to Advanced Nuclear Fuels (ANF) Corporation, to Siemens Nuclear Power (SNP), to Siemens Power Corporation (SPC), to Framatome ANP, to AREVA NP Inc., to the present name Framatome Inc.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.4 Page 1.4-1 of 1.4-5 1.4 PRINCIPAL DESIGN CRITERIA 1.4.1 FACILITY DESIGN Principal structures and equipment which are necessary either to prevent accidents or to mitigate their consequences were designed, fabricated and erected in accordance with applicable codes and to withstand the effects of the most severe earthquakes, flooding conditions, windstorms, ice conditions, temperature and other deleterious natural phenomena which could be expected at the site during the lifetime of this unit.
Consult the specific DSAR chapters on systems or transient analyses for more detailed discussions. Section 5.1 details Palisades' conformance to General Design Criteria per 10 CFR 50, Appendix A. Section 5.2 specifies Design Codes, Structures/Systems/ Components Classification and establishes the basis for "CP Co Design Class" terminology.
1.4.2 DELETED 1.4.3 DELETED 1.4.4 CONTAINMENT SYSTEM The reactor containment is a steel-lined reinforced concrete cylinder with a hemispherical dome and a flat base.
Ground accelerations for the operational basis earthquake used for containment design purposes and all seismic Class I structures (Section 1.7) are 0.10g applied horizontally and 0.07g applied vertically. In addition, ground accelerations of 0.2g horizontal and 0.13g vertical are used for the design basis earthquake. In the permanently shut down and defueled condition, the containment must retain its structural integrity during natural phenomenon events to ensure that it does not impact the safe storage of spent fuel in the spent fuel pool.
1.4.5 DELETED 1.4.6 DELETED 1.4.7 ELECTRICAL SYSTEMS Facility power is provided by 21-kV/4.16-kV or 345-kV/4.16-kV transformers.
Standby power (diesels) is included to ensure further continuity of electrical power for critical loads.
The function of the auxiliary electrical system is to provide reliable power to those auxiliaries required during any normal facility conditions.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.4 Page 1.4-2 of 1.4-5 The system design provides sufficient independence, isolation capability, and redundancy between the different power sources to avoid complete loss of auxiliary power.
1.4.8 RADIOACTIVE WASTES AND RADIATION PROTECTION The radioactive waste treatment system was designed so that discharge of radioactivity to the environment is in accordance with the requirements of 10 CFR, Part 20, and Appendix I to 10 CFR 50.
The Facility was provided with a centralized control room having adequate shielding to permit occupancy during all credible accident situations. The radiation shielding in the Facility, in combination with Facility radiation control procedures, ensures that personnel do not receive radiation exposures in excess of the applicable limits of 10 CFR, Part 20, during normal evolutions and maintenance.
1.4.9 FUEL HANDLING AND STORAGE Fuel handling and storage facilities were provided for the safe handling, storage and shipment of fuel and will preclude accidental criticality.
1.4.10 FIRE PROTECTION A "Fire Protection Program" (FPP) consisting of Facility design considerations, fire detection and suppression equipment, and Facility procedures assures that the Facility can respond to fires that could result in a radiological hazard. The FPP complies with 10 CFR 50.48(f).
1.4.11 DELETED 1.4.12 SECURITY Access and egress to all "protected" areas of the Facility are monitored/controlled through the utilization of card readers. Access to the Facility is controlled at the security entrance via explosive detectors, metal detectors, guards and card readers. A physical security force is always present. Details of conformance are identified in the commission-approved physical security, safeguards contingency, and guard training and qualification plans.
1.4.13 EMERGENCY PLANNING In the unlikely event of a Facility accident resulting in, or potentially capable of allowing, offsite releases of radioactivity in excess of federal regulations, a system of emergency warning sirens is in place. Established "emergency implementing procedures" in conjunction with the Facility's "Emergency Plan" have been developed to assure minimum risk to the general public in compliance with 10 CFR 50.54(q) and 10 CFR 50, Appendix E.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.4 Page 1.4-3 of 1.4-5 1.4.14 PLANT OPERATION The Facility Operating License requires operation to be in accordance with the Technical Specifications, which are contained in Appendix A to that license. Technical Specifications contain Limiting Conditions for Operation, Surveillance Requirements, Design Features, Programs and Manuals, and Administrative Controls, in accordance with the Code of Federal Regulations, Title 10, Part 50.36 (10CFR50).
Operation of the Facility in accordance with the Limiting Conditions for Operation assures that facility operation will remain within the assumptions and initial conditions of the safety analyses. The Administrative Controls provide NRC requirements for plant staff Responsibilities, Organization, Qualifications, Procedures, Programs and Manuals, Reporting Requirements to the NRC, and High Radiation Area Control.
1.4.15 STRUCTURES Facility structures were designed in accordance with the design criteria identified in Chapter 5. Structures were identified as CP Co Design Class 1, 2 or 3 according to Section 5.2. Specific design criteria for containment is discussed in Section 5.8, and other CP Co Design Class 1 structures are discussed in Section 5.9.
1.4.16 SINGLE FAILURE CRITERIA 1.4.16.1 Licensing Basis Palisades submitted application for an operating license in 1968. At that time, the General Design Criteria (GDCs) were in draft form. The original FSAR contained Appendix I, which presented a comparison of plant design features with the 1967 draft GDCs. From the wording of Criterion 39, "Emergency Power," and Criterion 41, "Engineered Safety Features Performance," of the original FSAR, it is clear that design considerations for single failure concerns were limited to "failure of a single active component."
Palisades was not designed with system redundancy (electrical or fluid systems) comparable to newer plant designs. As such, and in general, only the failure of a single active component (and not a passive failure) was considered.
In 1977, the NRC initiated the Systematic Evaluation Program (SEP) to review the designs of older operating plants. The review provided: 1) an assessment of the significance of differences between the then-current technical positions on safety issues and positions that existed when a particular plant was licensed; 2) a basis for deciding how these differences would be resolved; and 3) documented evaluations of plant safety. Palisades was one plant selected for the SEP reviews. Based on the SEP reviews,
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.4 Page 1.4-4 of 1.4-5 topics were closed based on the adequacy of the existing system designs or, in some cases, after the licensees made procedural or design changes.
Single failure criteria adequacy (electrical and fluid systems) was evaluated in several topics and no requirement to address passive failures on a plant-wide, system level basis was backfit by NRC or committed to by Palisades.
Specific issues were addressed on a case-by-case basis.
Also of concern is the assumed timing of a failure. The NRC Safety Evaluation Report for Design Basis Events (Reference 6) shows that the types of failures considered as most limiting for design basis events were assumed to occur at the time of the demand for the components being called upon to function. NRC Information Notice 93-17 Revision 1 (Reference 7) addressed the issue by acknowledging that some plants safety systems have been designed to respond properly to a failure upon demand but not for other possible sequences. The notice stated that no backfitting was intended or approved, and that the generic issue was dropped based on extremely low probability of occurrence.
In general, when performing design modifications or evaluating system performance, the licensing basis for Palisades for failures in electrical and fluid systems only considers single active failures, and the failure is only considered at the time the demand is placed on the component to function.
Exceptions to this treatment of single failures have been addressed on a case-by-case basis as design requirements have evolved over time and significant safety concerns have been addressed. Where it has been determined to be applicable, specific criteria have been incorporated into the licensing basis. For example, the single failure criterion of IEEE 279-1971 has been applied to protection systems between the sensors and the actuation devices (Chapters 7 and 8). Future design modifications should consider current guidance per Subsection 1.4.16.3.
1.4.16.2 Active and Passive Failures Active failures considered in Palisades design require a malfunction of a component that relies on a mechanical movement to complete its intended function upon demand. For example, the inadvertent opening of a normally closed breaker, absent a component fault, would be considered a passive failure of the breaker.
NRC Information Report, SECY-77-439, August 17, 1977, Single Failure Criteria, (Reference 8) gives the following definition and example of a passive failure:
A passive failure in a fluid system means a breach in the fluid pressure boundary or a mechanical failure which adversely affects a flow path. Examples include the failure of a simple check valve to move to its correct position when required,
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.4 Page 1.4-5 of 1.4-5 As shown by this statement, the NRC considered the failure of a check valve as a passive failure. Check valves were considered passive components when Palisades was originally designed and constructed.
1.4.16.3 Current Design Considerations Standards and requirements have changed since the original licensing of Palisades. The plant has been reevaluated in light of these newer standards and requirements, as well as industry experience, and has at times been required to make changes on a case-by-case basis. Therefore, as new issues arise, the plant staff should use current guidance and review criteria in order to make appropriate decisions regarding the adequacy of systems design and performance.
While not part of the Palisades licensing basis, ANSI/ANS-58.9-1981, American National Standard, Single Failure Criteria for Light Water Reactor Safety-Related Fluid Systems, provides current guidance. That standard provides the following definition of an active failure:
An active failure is a malfunction, excluding passive failures, of a component that relies on mechanical movement to complete its intended function upon demand.
Per that standard, failure of a check valve to move to its correct position is an example of an active failure, which is different from the Palisades original licensing basis. The standard also provides a method for taking exceptions by stating:
Where the proper active function of a component can be demonstrated despite any credible condition, then that component may be considered exempt from active failure. Examples of such component functions may include opening of code safety valves and certain swing check valves. Where such exemption is taken, the basis for the exemption shall be documented in the single failure analysis.
Though an issue may not be part of the Palisades licensing/design basis, if current standards would require a different treatment of a particular type of failure, it should be evaluated on a case-by-case basis to assess the safety significance and determine if it would be prudent to adopt the current standard. While failure at the time of demand is the licensing basis for consideration of single active failures, other credible single active failures should not automatically be eliminated.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.5 Page 1.5-1 of 1.5-3 1.5 MAJOR PLANT MODIFICATIONS (DESIGN/CONSTRUCTION)
Following initial completion of the Palisades Plant in 1971, several major facility modifications have been made to improve the safety and operability of the Plant. These modifications are briefly outlined below, with references to the appropriate DSAR Update section and identification of the designer/constructor.
STEAM GENERATOR REPLACEMENT - BECHTEL/BECHTEL Switch over to an all-volatile secondary water chemistry decreased the rate of tube degradation but over time examination revealed further intergrannular attack (IGA) and other growing problems related to denting at tube support plates. With excess outage times and plant operation nearing the point of power limitation due to plugged tubes, replacement of both steam generators was undertaken in late 1990. See Subsection 4.3.4.2.
FEEDWATER PURITY BUILDING ADDITION - BECHTEL/J A JONES A completely new secondary side feedwater (condensate) purity system was installed to provide full flow condensate demineralization system utilizing powdered ion exchange resins and on-line resin body feed capability. This new system is housed in the feedwater purity building addition. See Subsection 10.2.3.2.
COOLING TOWERS ADDITION Initially, the Facility was designed for a once-through Circulating Water System for providing cooling water to the condenser. For environmental reasons, the system was converted in 1974 to a closed-cycle system using two Ecodyne mechanical draft cooling towers and blowdown dilution. A cooling tower pump house was constructed to enclose the cooling tower pumps. See Subsections 10.2.4 and 10.2.4.1.
In 2012, the "A" cooling tower was replaced with an SPX Marley cooling tower. The replacement tower has 16 cells as apposed to the 18 cells of the original Ecodyne design. The new tower is a pultruded fiberglass design.
In 2017, the "B" tower was also replaced with an SPX Marley cooling tower similar in design to the "A" tower. The replacement tower retained all 18 cells, however.
RADWASTE SYSTEM MODIFICATIONS/AUXILIARY BUILDING ADDITION -
BECHTEL/BECHTEL During 1971-1973 the liquid waste management system was modified to reduce liquid discharges to "Near Zero" and meet the requirements of 10 CFR 50, Appendix I. The auxiliary building was expanded to enclose much of the new equipment required. The 1972-1973 service building
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.5 Page 1.5-2 of 1.5-3 addition, in conjunction with the aforementioned changes, was made to accommodate solid radwaste system changes designed by Protective Packaging Inc (PPI). This system was subsequently replaced by a molten bitumen immobilizing system for waste concentrates. In 1996, the molten bitumen immobilizing system was replaced by a concentrated waste drying system. See Chapter 11 for details.
SPENT FUEL POOL STORAGE MODIFICATIONS - NUS/J A JONES 1977 -
WESTINGHOUSE/WESTINGHOUSE 1987 In 1977, the spent fuel pool storage capacity was increased from a capacity of 272 assemblies to 798. In 1987, Amendment 105, dated July 24, 1987, authorized replacing existing racks with six high-density spent fuel racks that increased the storage capacity from 798 to 892 fuel assemblies. See Section 9.4 and Subsection 9.11.3.
HIGH-PRESSURE AIR ADDITION - BECHTEL/J A JONES In 1977, the compressed air system was augmented by the addition of a high-pressure air system (325 psig) for supply to safety-related air-operated valves and components. See Section 9.5.
FIRE PROTECTION SYSTEM MODIFICATIONS - NUCLEAR SERVICES CORP/QUADREX/J A JONES Following the 1975 Browns Ferry fire and subsequent NRC revised guidelines, Consumers Power undertook a series of studies and resultant Facility fire protection features modifications. This has included the addition of fire-fighting equipment, separation of cables, addition of fire stops, preparation of procedures, etc. See Section 9.6 and Chapters 7 and 8.
METEOROLOGICAL PROGRAM IMPROVEMENTS - EG&G/CP CO Several modifications were made to the onsite meteorological towers including the addition and relocation of new towers. A final meteorological program was attained in 1977 following a 1975 study by EG&G Environmental Consultants. See Subsection 2.5.2.3 for a description of the new tower and meteorological program.
AUXILIARY BUILDING TSC/EER/HVAC ADDITION - BECHTEL/BECHTEL During 1983 an addition was added to the north side of the auxiliary building to house a Technical Support Center (TSC), an Electrical Equipment Room (EER) and a Heating, Ventilating and Air Conditioning (HVAC) area.
The TSC was required to fulfill the guidelines of NUREG-0696, the HVAC area as a result of the control room habitability requirements of NUREG-0737, and the EER area as a result of loads placed on the electrical system by the addition of the TSC and HVAC areas. See Section 9.8 for discussion of the
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.5 Page 1.5-3 of 1.5-3 HVAC system, Chapter 8 for discussion of the electrical equipment and the Site Emergency Plan for the functional discussion of the TSC.
INTERIM OLD STEAM GENERATOR STORAGE FACILITY -
BECHTEL/BECHTEL In 1990, a reinforced concrete building was constructed for interim storage of two old steam generators. This facility is located in the controlled area of the site approximately 2,200 feet northeast of the containment building. The storage facility design provides sufficient radiation shielding such that the onsite and offsite dose rate will not exceed the limits defined in 10 CFR 20 and 40 CFR 190, respectively. The facility is designated as a secondary restricted area. The old steam generators will remain in this facility until an ultimate disposition method is selected.
INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) - PACIFIC SIERRA NUCLEAR In 1993 Palisades constructed a suitable facility and began dry storage of spent nuclear fuel in casks under the General License provisions of 10CFR72. This facility is located north of the support building and is enclosed within the plant security fence. The ISFSI was planned to hold 25 Ventilated Storage Casks (VSCs) designed by Pacific Sierra Nuclear (later Sierra Nuclear) Corporation although other NRC-certified cask designs could also be utilized if the associated fuel handling equipment were procured.
In 2003, Palisades constructed an additional ISFSI pad for storage of NUHOMS casks under the General License provisions of 10CFR72. This facility is located east of the plant and is enclosed by a security fence. Other NRC-certified cask designs could also be stored at the pad.
REPLACEMENT OF REGION 1 CARBORUNDUM RACKS - HOLTEC INTERNATIONAL In 2013, Palisades replaced six of seven Carborundum-equipped Region 1 racks in the Spent Fuel Pool with six new MetamicTM-equipped Region 1 racks under the provisions of 10CFR50. The new racks have the same storage capacity (the same number of fuel assemblies stored) as the replaced racks. The new racks were designed and supplied by Holtec International.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.6 Page 1.6-1 of 1.6-2 1.6 INSERVICE INSPECTION 1.6.1 HISTORICAL BACKGROUND The Palisades Nuclear Power Plant was built in the late 1960s and was placed in commercial service on December 31, 1971. During the first 40-month life of the Plant, in order to comply with Paragraphs 4.3 and 4.12 of the Technical Specifications (dated September 1, 1972) of the Provisional Operating License DPR-20 for the Palisades Nuclear Plant, which discusses ISI requirements of ASME Class 1 components and systems, the nondestructive examinations were performed to satisfy the requirements of the ASME Boiler and Pressure Vessel Code,Section XI, 1971 Edition, including the Winter 1972 Addenda (ASME B&PV Code,Section XI, 71W72a). In February 1976, the NRC amended Paragraph 55a (g) of 10 CFR 50 to require nuclear plants to upgrade their Technical Specifications in the areas of the ISI requirements and the functional testing of pumps and valves. By amending Paragraph 55a (g) and by invoking Regulatory Guide 1.26, the NRC required nuclear plants to upgrade their ISI program to include not only ASME Class 1 systems, but also ASME Class 2 and ASME Class 3 systems.
1.6.2 GENERAL The Inservice Inspection Plan for the initial 10-year inservice intervals was developed by Southwest Research Institute and Consumers Power Company, and reviewed and approved by Consumers Power Company for use at the Palisades Nuclear Power Plant. Subsequent updating to remain responsive to industry requirements is anticipated.
The start of the first 10-year interval coincides with the date of first commercial operation, December 31, 1971. The length of the first 3-1/3-year period was extended to October 30, 1976 by adding 18 months cumulative shutdown time between August 1973 and April 1975 in accordance with ASME B&PV Code,Section XI, IS-241, 71W72a. The second period ran to June 1, 1980 due to the 1979/1980 extended refueling outage. The third period extended to November 9, 1983 per ASME B&PV Code,Section XI, IWA-2400(c), 77S78a.
The second 10-year interval began November 10, 1983, and lasted until May, 1995.
The third 10-year interval began May 12, 1995, and lasted until December 2006.
The fourth 10-year interval began December 13, 2006, and lasted until December 12, 2015.
The fifth 10-year interval began December 13, 2015.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.6 Page 1.6-2 of 1.6-2 See Section 6.9 for details of the Inservice Inspection Program.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.8 Page 1.8-1 of 1.8-8 1.8 SPECIAL MAJOR PROGRAMS As a result of continued NRC concern with the "health and safety of the public" and its relationship to the safe operation of all nuclear facilities, the Palisades Plant and operations have been subjected to considerable NRC review.
The Inspection and Enforcement Branch of the NRC routinely provides IE Bulletins (for utilities to review and respond) regarding the identification of generic problems that could have a safety impact. In addition, the NRC on occasion establishes technical review programs as a result of legal mandates following court actions or NRC initiated programs resulting from unusual events in the industry.
Programs of special interest to Palisades resulting from these circumstances are discussed in the remainder of this section.
1.8.1 SYSTEMATIC EVALUATION PROGRAM 1.8.1.1 Description of Program Between 1977 and the early 1980s, Consumers Power Company (CPCo) participated in the NRCs Systematic Evaluation Program (SEP). The purpose of this program was to confirm the adequacy of certain as-licensed design features of eleven plants, including Palisades, whose construction permits were issued before the final General Design Criteria (GDC)
(10 CFR 50 Appendix A), the associated Standard Review Plans (NUREG 75/087 and 0800), and other guidance documents. The program was also intended to provide a basis for converting the Provisional Operating Licenses (POL) held by some of the plants to Full Term Operating Licenses (FTOL). The program was designed to be conducted by the NRC with limited, voluntary support of licensees. The overall approach was to document major differences between each plants as-licensed design and then-current design criteria in written topic evaluation reports; to evaluate the safety significance of differences that were judged to be potentially significant; and, finally, to make decisions about whether any of those differences should be resolved through backfits, using an integrated safety assessment process with the assistance of probabilistic risk assessment (PRA) tools.
Topic evaluation reports were prepared for a number of subject areas, and these reports, in turn, provided inputs to the integrated safety assessment process. A separate integrated safety assessment report was issued to document the final NRC conclusions for selected issues raised in the topic evaluation reports which were judged to warrant discussion or action. The individual topic evaluation reports did not contain the final NRC conclusions about each topic, although they often did include the authors opinions or recommendations about perceived safety significance and the desirability of backfits for certain issues.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.8 Page 1.8-2 of 1.8-8 The integrated safety assessment process was an internal NRC activity.
Accordingly, the public record does not always contain detailed discussions of NRC bases for selecting certain design differences for additional study, of the NRC decisions process for selecting specific differences to be assessed as candidates for backfitting, nor of the internal NRC process for judging adequacy of actions proposed by CPCo. The outputs of the integrated safety assessment process, which documented the final agency conclusions for the evaluated topics, were published in NUREG 0820 (Integrated Plant Safety Assessment Report) (Reference 3), NUREG 0820 Supplement 1, NUREG 1424 {Safety Evaluation Report (SER) supporting conversion of license from POL to FTOL} (Reference 4), and in several later letters on topics that were not closed until after the Integrated Plant Safety Assessment Report was published.
Table 1-3 provides a listing of the final 90 topics reviewed in the SEP for Palisades, and summarizes the results of each review. The SEP review began by comparing the as-built plant design with the then current review criteria in 137 different areas defined as "topics."
During the review, 47 of the topics were deleted from consideration by SEP, based on one of the three following reasons: (1) topic was part of the Unresolved Safety Issue Program (USI), (2) topic was part of Three Mile Island Action Plan Tasks, or (3) the topic was not applicable to the Plant. The remaining 90 topics were reviewed for Palisades and are those listed in Table 1-3. Fifty-nine of the 90 topics met current criteria or were acceptable on another defined basis. These topics are identified as Status Code S on Table 1-3.
Thirty-one topics received further review and evaluation during the Integrated Assessment Program. A major part of the integrated assessment was the probabilistic risk assessment (PRA). PRA was used to determine which system failures would create an unacceptable risk because either a redundant system was not available or available systems were inadequate for the job required. During the integrated assessment, several of these topics were found to be acceptable and required no further work. These items are identified as Status Code 4 in Table 1-3.
The remaining topics evaluated using PRA, were each found to require one or more of the following modifications:
- 1.
Plant changes
- 2.
Technical Specifications changes
- 3.
Refined engineering analysis required These items are identified as Status Code 1, 2 or 3 in Table 1-3. The disposition column of Table 1-3 shows where specific topics are addressed.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.8 Page 1.8-3 of 1.8-8 1.8.1.2 SEP Reviews Confirmed Safety of Palisades Design The SEP topic evaluations and integrated plant safety assessment documents confirmed that the level of safety provided by the Palisades design was adequate even though the design differed from later design requirements embodied in the General Design Criteria and other documents.
The NRC letter of October 29, 1982, which transmitted the final report of the SEP review (NUREG 0820) to CPCo, specifically states, The review has provided for a documented evaluation of plant safety when all supplements to the IPSAR and the Safety Evaluation report for converting the license from a provisional to a full-term license have been issued. The fact that Integrated Plant Safety Assessment Report did not identify backfits or other actions for many of the identified design differences, and the fact that the SEP results later provided a part of the basis to convert the Palisades POL to a FTOL, provide de facto evidence that the licensed plant design was found to be adequate.
While the GDC were used as reference standards for the reviews, the SEP did not backfit a requirement to comply with the GDC. The NRC generic position on applicability of the GDC to plants of Palisades age was later summarized as follows: The General Design Criteria are not applicable to plants with construction permits issued prior to May 21, 1971. At the time of the promulgation of Appendix A, the Commission stressed that the GDC were not new requirements and were promulgated to more clearly articulate the licensing requirements and practice in effect at that time. While compliance with the intent of the GDC is important, each plant licensed before the GDC were formally adopted was evaluated on a plant specific bases, determined to be safe, and licensed by the Commission. Furthermore, current regulatory processes are sufficient to ensure that plants continue to be safe and comply with the intent of the GDC. (Reference 5) The DSAR should be used to determine the extent of Palisades commitments, if any, to any particular revision or criterion of the GDC.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.8 Page 1.8-4 of 1.8-8 1.8.1.3 Continuing Applicability and Interpretation of SEP Information The SEP topic evaluations and supporting information have continuing relevance in that they help to clarify the facilitys original licensing and design bases, and they provide documentation of NRC review. The topic evaluations sometimes include summaries of the design and regulatory philosophy which existed when Palisades was designed and licensed. At times the topic evaluations discuss explicit bases for the reviewers judgments about adequacy and safety significance of certain facility design features. However, the SEP topic evaluation should not be used independently from the final Integrated Plant Safety Assessment Report or other closure documents. The information in individual topic evaluations concerning perceived design weaknesses, or additional actions recommended by a topic evaluations author, were not formal NRC conclusions, and the recommendations contained therein did not create licensee commitments. Some of the specific issues raised in topic evaluations have no documented closure. Closure sometimes has to be inferred from the fact that NRC did not include a discussion of a specific issue in NUREG 0820, its supplement, subsequent docketed letters on selected topics, or in NUREG 1424. If an issue was recommended for additional action within a topic evaluation, but that issue or action was not documented in NUREG 0820 or successor documents, it can be assumed that the issue was screened out during NRCs integrated safety assessment process as not having sufficient safety significance to warrant further action or discussion.
Most of the actions taken to resolve issues raised during the SEP were voluntarily proposed by CPCo and do not represent permanent obligations.
When accepted by NRC, the CPCo-proposed actions were implemented as NRC commitments, and the actions were summarized by NRC in NUREG 0820 and its Supplement 1. The specific actions taken by CPCo as a result of the SEP, and the specific facility design and operating features which were reviewed during the SEP, can be changed with appropriate justification in accordance with 10 CFR 50.59, facility administrative requirements, and/or NRC-endorsed industry guidelines for NRC commitment management.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.8 Page 1.8-5 of 1.8-8 1.8.2 TMI ACTION ITEMS (NUREG-0737)
As a result of the incident at Three Mile Island Nuclear Power Plant, the NRC developed a list of requirements for other nuclear-powered generating stations. The list consists of 37 items, which are broken down into a total of 94 subitems as identified in NUREG-0737.
The list of items, compliance status and general description of how the item was to be resolved, are shown in Table 1-4. All of the 94 subitems have been closed out and are identified in the table as Status Code 1. The NRC SER for issuance of the Full Term Operating License, NUREG 1424 dated November 1990, confirms that the NRC views all TMI Action Items as being resolved for Palisades.
1.8.3 PIPE SUPPORT BASEPLATE DESIGNS USING CONCRETE EXPANSION ANCHOR BOLTS (IE BULLETIN 79-02)
Nuclear Regulatory Commission IE Bulletin 79-02, addressed Seismic Category I pipe supports using concrete expansion anchor bolts (CEBs) for loadings obtained from analysis of Seismic Category I piping systems.
All baseplates or structural steel members using CEBs for large piping identified in the course of responding to IE 79-02 were evaluated. The evaluation was performed in accordance to the load combinations specified in Section 5.10. Acceptance criteria were as specified in the Bulletin for CEBs and Chapter 5 for baseplates or structural steel members. Those baseplates or structural steel members and CEBs which did not satisfy acceptance criteria were modified (or will be modified).
Approximately 3,000 accessible CEBs for large bore piping were inspected and load tested. More than 96% of this population satisfied the load testing.
Approximately 4% of the CEBs for large bore piping were inaccessible for full testing and inspection. These CEBs and their baseplates or structural steel members were evaluated. If these CEBs and baseplates or structural steel members did not satisfy acceptance criteria, they were either modified or the piping support system was revised to yield acceptable results.
Small bore piping supports were designed using a conservative chart method.
A sample of CEBs used for support of small piping was inspected and tested.
This sample consisted of more than 1,000 CEBs (more than 2/3 of the population). This testing and inspection program used in conjunction with the conservative chart method, yields an acceptable confidence level for small piping.
Thus, the inspection, testing and evaluation performed for baseplates or structural steel members and CEBs for Seismic Category I piping, satisfy the requirements of IE 79-02, and the modifications have been completed.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.8 Page 1.8-6 of 1.8-8 1.8.4 SEISMIC ANALYSIS FOR AS-BUILT SAFETY-RELATED PIPING SYSTEMS (IE BULLETIN 79-14)
The bulletin required an inspection of approximately 18,100 feet of large diameter safety-related piping, 1,550 pipe supports and piping components at the Palisades Facility. Small piping systems (2 inches or less in diameter) were also inspected, noted items evaluated and a sample of the small piping systems was evaluated.
The Palisades Facility systems were reviewed and it was determined that 23 systems had safety-related piping. Data on and sketches of the safety-related systems were completed, potential nonconformance items were listed and the as-built data were evaluated. Approximately 320 piping support changes have been completed: (1) 45 new supports were added, (2) 23 supports were removed, and (3) 252 supports were modified.
There were approximately 3,250 listed conditions that were either questionable or constituted a discrepancy. These items were evaluated and resolved. About 75% of the items related to lack of or nonconformance with existing drawings. The remaining 750 items related to hardware conditions, such as nuts/bolts loose or missing, spring cans bottomed out or without load and bent/broken or missing pipe support components. Two Licensee Event Reports (LERs) were issued as a result of the program (LER 79-033 and LER 80-001). Corrective action has been completed on both items.
In 1989, discrepancies were noted in the original 79-14 bulletin analyses. It was determined that design/evaluation criteria, documentation and work quality of the original effort were not adequate. LER 89-23 and LER 89-23 Rev 1 were issued. They outlined a work scope to correct the deficiencies and established interim operability criteria to facilitate implementation of the effort. A major program was established "Safety Related Piping Reverification Program" for re-evaluating large bore piping. Subsequently, a small bore piping program was also established. These programs were developed to consist of walkdowns to establish as built inputs for piping analysis and support evaluation. When analysis or evaluation shows it is necessary, facility modifications are made to satisfy acceptance criteria.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.8 Page 1.8-7 of 1.8-8 1.8.5 UNRESOLVED SAFETY ISSUES (NUREG-0410)
Of the unresolved safety issues (also called generic safety issues) identified by the NRC and discussed in NUREG-0410, -0510, -0649, and -0705, 19 were considered by the NRC to require investigation for their potential impact on the Palisades Nuclear Plant. The 19 unresolved safety issues considered are listed in Table 1-5 with a cross-reference to the appropriate DSAR chapter wherein they are discussed. All 19 of the unresolved safety issues have been assessed by Consumers Power Company and are considered to have no undue risk to the health and safety of the public while longer term generic review of these issues is being conducted.
USI A-46, Seismic Qualification of Equipment in Operating Plants, has been resolved by the Seismic Qualification Utility Group (SQUG). Equipment in the Safe-Shutdown paths required for facility shutdown was evaluated using the SQUG Generic Implementation Procedure (GIP) (Reference 10). The "Report of SQUG Assessment at Palisades Nuclear Plant for the Resolution of USI A-46" was submitted to the NRC on May 19, 1995 (Reference 11). The NRC Safety Evaluation Report for the resolution of USI A-46 was issued on September 25, 1998 (Reference 12).
There were several outliers identified (ie, equipment items that did not meet the SQUG GIP initial screening criteria). Essentially all of the outliers were resolved over the next five years by testing, analysis, modification or some combination of these approaches The letter entitled Final Closeout of Unresolved Safety Issue A-46 Outliers, from Douglas E. Cooper to the USNRC Document Control Desk was submitted to the NRC on June 26, 2003 (Reference 13). This letter provided the final resolution of USI A-46 for the Palisades Nuclear Plant including the resolution of all outliers, including the SIRW tank outlier which was resolved by implementing an alternate path.
1.8.6 DELETED 1.8.7 DELETED 1.8.8 DELETED
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.8 Page 1.8-8 of 1.8-8 1.8.9 DELETED 1.8.10 DELETED 1.8.11 HEAVY LOADS An NRC Generic Letter dated December 22, 1980, requested licensees to prepare responses to indicate their degree of compliance with certain guidelines for NUREG-0612, Control of Heavy Loads. Phase I of this effort involved identifying the load handling equipment within the scope of NUREG-0612 and describing general heavy load handling program activities, such as safe load path identification, load handling procedures, operator training, the use of special and general purpose lift devices, the maintenance, testing and repair of the cranes, and crane design (Reference 89). These program activities are implemented and controlled by facility procedures.
Phase II of this effort involved further actions in response to additional NUREG-0612 guidance (Reference 90). The NRC accepted the response to Phase I for Palisades in an SER dated November 9, 1983 (Reference 15).
Subsequently, the NRC issued Generic Letter 85-11 (Reference 16), which stated that, based on improvements in the handling of heavy loads obtained from Phase I, further action to reduce the risks associated with the handling of heavy loads was not required, and the Phase II was considered to be completed.
The NRC issued Bulletin 96-02, Movement of Heavy Loads Over Spent Fuel, Over Fuel in the Reactor Core, or Over Safety-Related Equipment" (Reference 91) which requested licensees to review their capabilities to handle heavy loads while the facility is operating and to remind licensees of their responsibilities for ensuring that heavy load activities are performed safety and within requirements. Palisades responded to the Bulletin (Reference 92) and the response was accepted by the NRC (Reference 93).
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.9 Page 1.9-1 of 1.9-14 1.9 RENEWED FACILITY OPERATING LICENSE On January 17, 2007, the Renewed Facility Operating License was issued by the NRC, extending the expiration date to March 24, 2031. License Condition 2.H of the renewed license required that the FSAR be supplemented in accordance with 10 CFR 54.21(d) to incorporate summary descriptions of the programs and activities credited for managing the effects of aging (Aging Management Programs), and of the evaluation of Time-Limited Aging Analyses (TLAAs), for the period of extended operation.
The Application for Renewed Operating License (LRA) had been submitted on March 22, 2005. This section includes that supplement.
During the NRC review of the LRA, several commitments for future actions were made by the licensee. Table 1-9 provides the listing of final commitments as submitted by the licensee and confirmed in Appendix A of the Safety Evaluation Report Related to the License Renewal of Palisades Nuclear Plant (NUREG 1871). Some changes to the commitments were made through the commitment change process. All commitments that were required to be implemented prior to entering the period of extended operation were completed prior to entering the period.
Section 1.9.1 contains summary descriptions of the programs used to manage the effects of aging during the period of extended operation, and Section 1.9.2 contains summaries of TLAAs applicable to the period of extended operation.
Under the renewed operating license, Aging Management Programs were applied to certain non-safety related Systems, Structures, and Components (SSCs). To ensure appropriate controls are provided for non-safety related aging management activities, a commitment was made to apply certain Quality Assurance Program provisions to Aging Management Programs. These provisions apply to the program elements of corrective action, confirmation process, and administrative controls. DSAR Section 15.1.2 describes the generic quality assurance requirements to be applied to the Aging Management Programs.
1.9.1
SUMMARY
DESCRIPTIONS OF AGING MANAGEMENT PROGRAMS This section provides summaries of programs and activities credited in the License Renewal Application for managing the effects of aging during the period of extended operation.
The activities implemented to manage aging at the Palisades Facility may be performed under discrete programs as defined herein, or they may be incorporated into other facility programs. The program summaries should be interpreted as summaries of activities to be performed to manage aging, and not as specific commitments to maintain unique programs with the specific titles and content listed.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.9 Page 1.9-2 of 1.9-14 It should also be noted that these summaries do not specifically invoke or reference the Generic Aging Lessons Learned (GALL) Report, NUREG-1801.
The activities credited for managing aging at Palisades were developed, to a large extent, to be responsive to the revision of the GALL that existed at the time that the License Renewal Application was developed. It is expected that changes will be made to these programs in the future as a result of advances in the state of knowledge in the industry, facility modifications, and operating experience. However, no commitment is made to update any aging management program in response to changes in the GALL. Future changes that may occur to aging management programs or activities will be managed under 10 CFR 50.59, "Changes, Tests and Experiments," and/or other regulatory and administrative requirements appropriate to the changes being made.
1.9.1.1 Nickel Alloy Program The Nickel Alloy Program manages aging due to Primary Water Stress Corrosion Cracking (PWSCC) of the Primary Coolant System (PCS) pressure boundary Alloy 600 components, including Inconel 82/182 weld joints, reactor vessel head penetrations, etc. The program includes:
- a.
PWSCC susceptibility assessment using industry models to identify susceptible components
- b.
Monitoring and control of primary coolant chemistry to mitigate PWSCC
- c.
In-Service Inspections (ISI) of pressurizer penetrations, reactor vessel head penetrations and Alloy 82/182 PCS pressure boundary welds in accordance with American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code (BPV Code)Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Subsection IWB, Table IWB-2500-1
- d.
Augmented inspections or preemptive repair/replacement of susceptible components or welds The License Renewal Application included a commitment to submit for NRC review and approval a revised nickel alloy (ie, Alloy 600) aging management program that updates the PWSCC corrosion rate assessments and inspection program consistent with the latest NRC requirements and industry commitments (Reference 57). The revised nickel alloy aging management program was submitted on March 13, 2008 (Reference 101). The NRC determined that the revised program was acceptable in a safety evaluation dated August 15, 2012 (Reference 102).
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.9 Page 1.9-3 of 1.9-14 1.9.1.2 ASME Section XI, Subsections IWB, IWC, IWD, IWF Inservice Inspection Program The applicable ASME BPV Code for the fifth ten-year interval of the inservice inspection program at the Palisades Facility is ASME Section XI, 2007 edition, including 2008 addenda.
ASME Section XI IWB, IWC, IWD, and IWF Inservice Inspection Program facilitates inspections to identify and correct degradation in Class 1, 2, and 3 piping, components, and their supports and integral attachments. The program includes periodic visual, surface, and/or volumetric examinations and leakage tests of all Class 1, 2, and 3 pressure-retaining components and their supports and integral attachments, including welds, pump casings, valve bodies, pressure-retaining bolting, piping/component supports, and reactor head closure studs. These are identified in ASME Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, or commitments requiring augmented inservice inspections, and are within the scope of license renewal. This program is in accordance with 10 CFR 50.55a, "Codes and Standards."
1.9.1.3 Bolting Integrity Program Palisades Bolting Integrity Program relies on the guidelines delineated in NUREG-1339, "Resolution of Generic Safety Issue 29: Bolting Degradation or Failure in Nuclear Power Plants," EPRI NP-5769, "Degradation and Failure of Bolting in Nuclear Power Plants" (with the exceptions noted in NUREG-1339), for safety related bolting, and EPRI TR-104213, "Bolted Joint Maintenance & Applications Guide" (for non-safety related bolting. The program also includes repair/replacement controls for ASME Section XI related bolting and generic guidance regarding material selection, thread lubrication, and assembly of bolted joints. The program considers the guidelines delineated in NUREG-1339 for a bolting integrity program, EPRI NP-5769 (with the exceptions noted in NUREG-1339) for safety related bolting, and EPRI TR-104213 for non-safety related bolting.
1.9.1.4 Boric Acid Corrosion Program The Palisades Boric Acid Corrosion Program monitors component degradation due to boric acid leakage through the performance of periodic inspections. It implements the recommendations of NRC Generic Letter 88-05, "Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR plants." The program requires periodic visual inspection of all systems within the scope of license renewal that contain borated water for evidence of leakage, accumulations of dried boric acid, or boric acid wastage. The program also provides for visual inspections and early discovery of borated water leaks such that structures and electrical and mechanical components that may be contacted by leaking borated water will not be adversely affected such that their intended functions are impaired.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.9 Page 1.9-4 of 1.9-14 1.9.1.5 Buried Services Corrosion Monitoring Program The Buried Services Corrosion Monitoring Program manages aging effects on the external surfaces of carbon steel, low-alloy steel, and stainless steel components that are buried in soil or sand. This program includes (a) visual inspections of external surfaces of buried components for evidence of coating damage and substrate degradation to manage the effects of aging, (b) visual inspection of the external surfaces of buried stainless steel components for evidence of crevice corrosion, pitting, and Microbiologically Influenced Corrosion (MIC). The periodicity of these inspections for carbon, low-alloy, and stainless steel will be based on opportunities for inspection such as scheduled maintenance work.
1.9.1.6 Closed Cycle Cooling Water Program The Closed Cycle Cooling Water Program manages aging effects in closed cycle cooling water systems that are not subject to significant sources of contamination, in which water chemistry is controlled and heat is not directly rejected to the ultimate heat sink. The program includes (a) maintenance of system corrosion inhibitor concentrations to minimize degradation, and (b) periodic or one-time testing and inspections to assess component aging. This program is based on the guidelines in EPRI TR-107396, "Closed Cooling Water Chemistry Guideline." The program scope includes activities to manage aging in the Component Cooling Water (CCS) System, Emergency Diesel Generator (EDG) Jacket Cooling Water (Emergency Power System),
and Shield Cooling System (SCS).
1.9.1.7 Containment Inservice Inspection Program The Containment Inservice Inspection (ISI) Program is designed to ensure that containment shell concrete, the post-tensioning system, and steel pressure retaining elements continue to provide an acceptable level of structural integrity. In addition, it is designed to ensure that the liner (with associated moisture barriers), other leakage limiting steel barriers, and pressure retaining bolted connections have not degraded.
1.9.1.8 Deleted 1.9.1.9 Diesel Fuel Monitoring and Storage Program The Diesel Fuel Monitoring and Storage Program assures the continued availability and quality of fuel oil to be used in diesel generators and diesel fire pumps. The program includes:
- a.
Monitoring and trending of fuel oil chemistry to maintain fuel oil quality and mitigate corrosion
- b.
Periodic draining, cleaning, and internal inspection of fuel oil storage tanks
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.9 Page 1.9-5 of 1.9-14
- c.
Periodic ultrasonic measurement of thickness of the bottom of fuel oil storage tanks Fuel oil quality is maintained by monitoring and controlling fuel oil contamination in accordance with the guidelines of the American Society for Testing Materials (ASTM) Standards D 2276, D 2709, and D 4057, and by verifying the quality of new oil before its introduction into the storage tanks.
1.9.1.10 Fire Protection Program The Fire Protection Program includes:
- a.
Fire barrier inspections
- b.
Electric and diesel-driven fire pump tests
- c.
Periodic maintenance, testing, and inspection of water-based fire protection systems Periodic visual inspections of fire barrier penetration seals, fire dampers, fire barrier walls, and ceilings and floors and periodic visual inspections and functional tests of fire-rated doors are performed to ensure that functionality and operability is maintained. Periodic testing of the fire pumps ensures that an adequate flow of firewater is supplied and that there is no degradation of diesel fuel supply lines. Periodic maintenance, testing, and inspection activities of water-based fire protection systems provide reasonable assurance that fire water systems are capable of performing their intended function. Inspection and testing include periodic hydrant inspections, fire main flushing, sprinkler inspections, pipe wall thickness testing, and flow tests.
1.9.1.11 Flow Accelerated Corrosion Program The Flow Accelerated Corrosion Program manages aging effects due to Flow-Accelerated Corrosion (FAC) on the internal surfaces of carbon or low alloy steel piping, elbows, reducers, expanders, and valve bodies which contain high energy fluids (both single phase and two phase). The program implements the Electric Power Research Institute (EPRI) guidelines in NSAC-202L-R3, "Recommendations for an Effective Flow-Accelerated Corrosion Program," for an effective FAC program and includes:
- a.
An analysis using a predictive code such as CHECWORKSTM to determine critical locations
- b.
Baseline inspections to determine the extent of thinning at these locations
- c.
Follow-up inspections to confirm the predictions
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.9 Page 1.9-6 of 1.9-14
- d.
Repairing or replacing components, as necessary 1.9.1.12 Non-EQ Electrical Commodities Condition Monitoring Program The Non-EQ Electrical Commodities Condition Monitoring Program manages aging in selected non-EQ commodity groups within the scope of 10 CFR 54, "Requirements for Renewal of Operating Licenses for Nuclear Power Plants."
Features of the program include periodic inspection and/or testing of the following commodity groups:
- a.
Accessible insulated cables and connections in scope of license renewal installed in adverse localized environments.
- b.
Sensitive instrumentation cables and connections in scope of license renewal.
- c.
Inaccessible medium voltage cables in scope of license renewal (not designed for submergence), subject to long periods of high moisture conditions and voltage stress.
- d.
Underground manholes for the accumulation of water over medium voltage cables in scope of license renewal.
- e.
Non-segregated bus and connections in scope of license renewal for insulation degradation, bus enclosure for degradation, and bus supports for structural integrity.
- f.
Representative sample of bolted electrical connections in scope of license renewal.
1.9.1.13 One-Time Inspection Program The One-Time Inspection Program addressed potentially long incubation periods for certain aging effects, including various corrosion mechanisms, cracking, and selective leaching, and provided a means of verifying that an aging effect is either not occurring or progressing so slowly as to have negligible effect on the intended function of the structure or component.
Hence, the One-Time Inspection Program provided measures for verifying an aging management program is not needed, verifying the effectiveness of an existing program, or determining that degradation is occurring which required evaluation and corrective action.
The program included:
- a.
Determination of appropriate inspection sample size
- b.
Identification of inspection locations
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.9 Page 1.9-7 of 1.9-14
- c.
Selection of examination technique, with acceptance criteria
- d.
Evaluation of results to determine the need for additional inspections or other corrective actions The inspection sample included locations where the most severe aging effect(s) would be expected to occur. Inspection methods included visual (or remote visual), surface or volumetric examinations, or other established Non-Destructive Examination (NDE) techniques.
This program was used for a variety of purposes, including the following:
To verify the effectiveness of water chemistry control for managing the effects of aging in stagnant or low-flow portions of piping or components, exposed to a treated water environment To manage the aging effects of loss of material due to aging mechanisms such as general, crevice, pitting, and galvanic corrosion; selective leaching; and MIC To verify that cracking due to stress corrosion cracking or cyclic loading, in small bore (< 4" NPS) ASME class 1 piping, is not occurring To verify, for components in the Compressed Air System, that there are no aging effects requiring management in the dry air environment
[This aspect of the program was superseded by the creation of a Compressed Air Monitoring Program as described in licensee letter dated October 31, 2005. This program is further described in Section 1.9.1.23.]
To verify, for carbon steel storage tanks supported on earthen or concrete foundations, that excessive corrosion is not occurring on the bottom surfaces of the tanks The One Time Inspection Program was completed on March 14, 2011. This completion date conforms with the program implementation schedule date of prior to the period of extended operation as described in NUREG-1871, "Safety Evaluation Report Related to the License Renewal of Palisades Nuclear Plant," (Reference 69) and guidance in NUREG-1801, "Generic Aging Lessons Learned (GALL) Report," (Reference 68) that the population of components be inspected before the end of the current operating term.
Palisades entered the period of extended operation on March 24, 2011.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.9 Page 1.9-8 of 1.9-14 1.9.1.14 Open Cycle Cooling Water Program The Open Cycle Cooling Water Program manages aging effects such as loss of material due to general, pitting, and crevice corrosion, erosion, MIC, and loss of heat transfer due to biological/corrosion product fouling (e.g.,
sedimentation, silting) caused by exposure of internal surfaces of metallic components to raw, untreated (e.g., service) water. The program scope includes activities to manage aging in the Service Water System (SWS) and Circulating Water system (CWS).
The aging effects are managed through:
- a.
Monitoring and control of biofouling
- b.
Flow balancing and flushing
- c.
Heat exchanger testing
- d.
Routine inspection and maintenance program activities
- e.
System walkdowns
- f.
Review of maintenance, operating, and training practices and procedures to ensure that aging effects do not impair component intended function Inspection methods include visual (VT), ultrasonic (UT), radiographic (RT),
and eddy current (ECT). This program is responsive to NRC GL 89-13, "Service Water System Problems Affecting Safety-Related Equipment."
1.9.1.15 Overhead Load Handling Systems Inspection Program The Overhead Load Handling Systems Inspection Program provides for inspections of the structural components and rails of cranes and fuel handling machines associated with heavy load handling that are subject to the requirements of NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," and are within the scope of license renewal requiring aging management. For Palisades these are the Containment Building Polar Crane, the Spent Fuel Pool Overhead Crane, the Containment Building jib and boom cranes, and the reactor and spent fuel pool fuel handling machines. These cranes comply with the Maintenance Rule requirements provided in 10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." The Overhead Load Handling Systems Inspections Program is primarily focused on structural components that make up the bridge and trolley of the overhead cranes that are within the scope of NUREG-0612.
1.9.1.16 Deleted
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.9 Page 1.9-9 of 1.9-14 1.9.1.17 Deleted 1.9.1.18 Deleted 1.9.1.19 Structural Monitoring Program The Structural Monitoring Program is designed to ensure that age related (as well as other) deterioration of facility structures (including masonry walls) and components within its scope is appropriately managed to ensure that each such structure or component retains the ability to perform its intended function. The program is implemented through visual examination of these structures, components, and other specified items. In addition, the program provides for inspections of opportunity of normally inaccessible below grade concrete when excavation work uncovers a significant depth (several feet or more) to provide access for inspection. Damage or degradation found during visual examination may be further evaluated by measurements and testing techniques as appropriate. As part of the Structural Monitoring Program, groundwater sampling for pH, chlorides, and sulfates will be performed, with a periodicity not to exceed every 5 years, to ensure the below grade environment remains non-aggressive.
This program also implements provisions of the Maintenance Rule, 10 CFR 50.65, that relate to masonry walls and water-control structures. It conforms to the guidance contained in RG 1.160, "Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," and NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," as well as Nuclear Energy Institute publication NEI 96-03, "Guideline for Monitoring the Condition of Structures at Nuclear Power Plants." This NEI document, which supplements NUMARC 93-01, contains additional guidance specific to the monitoring of structures. In addition, the program specifies that inspections for unreinforced block walls that are not contained by bracing will be performed on a more frequent basis than the normal frequency of once each 10-year interval specified for reinforced or braced block walls.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.9 Page 1.9-10 of 1.9-14 1.9.1.20 System Monitoring Program The System Monitoring Program manages aging effects for normally accessible, external surfaces of piping, tanks, and other components and equipment within the scope of License Renewal. These aging effects are managed through visual inspection and monitoring of external surfaces for leakage and evidence of material degradation. The program relies upon periodic system walkdowns to monitor degradation of the protective paint or coating, and/or the exterior steel surface area (if no paint or coatings exist, or if the existing protective paint and coatings are degraded to a point whereby the exterior steel surface is exposed). Palisades does not take credit for any above ground coating or paint for mitigating corrosion even though the tanks may be painted or coated. However, inspections of the above ground coating or paint will provide an indication of the condition of the material underneath the coating or paint.
1.9.1.21 Water Chemistry Program The Water Chemistry Program manages aging effects such as loss-of-material due to general, pitting, and crevice corrosion; cracking due to SCC; and steam generator tube degradation caused by denting, Intergranular Attack (IGA), and Outer Diameter Stress Corrosion Cracking (ODSCC), by controlling the environment to which internal surfaces of systems and components are exposed. The aging effects are minimized by controlling the chemical species that cause the underlying mechanisms that result in these aging effects. The program provides assurance that an elevated level of contaminants and, where applicable, oxygen does not exist in the systems and components covered by the program, thus minimizing the occurrences of aging effects, and maintaining each components ability to perform the intended functions. The program is based on the guidelines in EPRI Topical Report, "PWR Primary Water Chemistry Guidelines," and EPRI Topical Report, "PWR Secondary Water Chemistry Guidelines."
1.9.1.22 Inspections of Opportunity for Internal Surfaces of Selected Components and Corrosion Under Insulation Internal surfaces of selected systems and components which are exposed during periodic system and component surveillances, or during the performance of maintenance activities, are subject to visual inspections of opportunity. These inspections are applicable to components in-scope for license renewal that have an internal environment of water, are constructed of materials that are potentially susceptible to internal aging degradation in a wetted environment, but are not subject to an Aging Management Program (e.g., Water Chemistry) that would manage the internal environment such that aging degradation of the internal surfaces would not be expected. Visual inspections are performed to assure that existing environmental conditions are not causing material degradation that could result in a loss of a component intended function. Inspection activities are performed by qualified personnel looking for corrosion (General, Pitting, Crevice, MIC) and fouling.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.9 Page 1.9-11 of 1.9-14 Degraded conditions are documented in the Corrective Action Program and evaluated for acceptability, repair, or replacement.
External surfaces of selected insulated piping and components, which are exposed when insulation is removed for maintenance or surveillance, are subject to visual inspections of opportunity. The piping and components of interest are those within the scope of the System Monitoring Program, constructed of carbon or low alloy steel, with low normal operating temperatures in an indoor or outdoor environment such that the piping could be wetted under its insulation (e.g., from condensation or rain water) for extended periods without being detected. Degraded conditions are documented in the Corrective Action Program and evaluated for acceptability, repair, or replacement.
1.9.1.23 Compressed Air Monitoring Program The Compressed Air Monitoring Program manages aging affects on the internal surfaces of carbon steel, low-alloy steel, copper alloys, and stainless steel components within the scope of License Renewal that are exposed to a compressed air environment. These include components such as piping, traps, heat exchangers, filter housings, dryer housings, accumulators, and valve bodies made of materials such as carbon steel, low alloy steel, copper alloys, and stainless steel. The program manages the aging effects of General, Crevice, and Pitting Corrosion and Stress Corrosion Cracking. The program includes maintenance of the compressors, dryers, and filters associated with the facility Instrument Air System, High Pressure Air System, Feedwater Purity Air System, and associated back-up systems.
1.9.1.24 Oil Sampling and Analysis For selected components, in-scope for License Renewal, that have an internal environment of oil, and are constructed of materials that are potentially susceptible to internal aging degradation in that environment, the oil shall be subject to periodic sampling and analysis. The purpose of these activities is to ensure that oil system contaminants (primarily water and particulates) are maintained within acceptable limits, thereby preserving an environment that is not conducive to loss of material or reduction of heat transfer. Associated activities include:
- a.
Determination of appropriate analysis to be performed
- b.
Frequency of analysis
- c.
Acceptance criteria
- d.
Trending of results
- e.
Corrective actions, if required These activities ensure that the lubricating oil environment of these components is maintained such that water and contaminants are minimized.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.9 Page 1.9-12 of 1.9-14 1.9.1.25 Electrical Equipment Qualification Program The Electrical Equipment Qualification Program is an existing program that implements the requirements of 10 CFR 50.49, Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants, at Palisades. 10 CFR 50.49 defines the scope of components to be included, requires the preparation and maintenance of a list of in-scope components, and requires the preparation and maintenance of a qualification file that includes component performance specifications, electrical characteristics, the environmental conditions to which the components could be subjected, and the basis for qualification. 10 CFR 50.49(e)(5) contains provisions for aging that require, in part, consideration of all significant types of aging degradation that can affect component functional capability. 10 CFR 50.49(e)(5) also requires replacement or refurbishment of qualified components prior to the end of its designated life, unless additional life is established through ongoing qualification. EQ programs manage component thermal, radiation, and cyclical aging through the use of aging evaluations based on 10 CFR 50.49(f) qualification methods.
1.9.1.26 Fatigue Monitoring Program The Fatigue Monitoring Program is a new program that ensures that limits on fatigue usage are not exceeded during the renewal term. The program monitors and tracks selected cyclic loading transients (cycle counting) and their effects on susceptible components. Palisades has selected this option under 10 CFR 54.21, "Contents of Application--Technical Information," to manage cracking due to metal fatigue of the reactor coolant pressure boundary during the extended period of operation.
The Fatigue Monitoring Program provides cycle counting activities for confirming analytically derived cumulative usage values for applicable locations. Specific locations that may be subject to cyclic loading that could cause fatigue cracking are monitored using a computer-based monitoring program provided by EPRI, called FatiguePro. If warranted, other monitoring methods in addition to cycle counting may also be employed under this program to monitor specific locations.
1.9.2
SUMMARY
DESCRIPTIONS OF TIME-LIMITED AGING ANALYSES As part of a License Renewal Application, 10 CFR 54.21(c) requires that an evaluation of Time-Limited Aging Analyses (TLAAs) for the period of extended operation be provided. The following TLAAs were identified and evaluated during the license renewal process to meet this requirement.
1.9.2.1 Deleted
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.9 Page 1.9-13 of 1.9-14 1.9.2.2 Metal Fatigue The Fatigue Monitoring Program will ensure that the effects of aging will be adequately managed for the period of extended operation under 10 CFR 54.21(c)(1)(iii) by assuring that a reanalysis or other appropriate corrective action is taken if a design basis cycle count limit is reached at any time during the extended licensed operating period.
1.9.2.3 Deleted 1.9.2.4 Deleted 1.9.2.5 Other Plant-Specific Time-Limited Aging Analyses Fuel Handling Crane Load Cycles A crane evaluation to the Crane Manufacturers Association of America Standard CMAA-70 assumes a number of rated lifts in the design lifetime in order to establish the design Service Level, and hence the allowable stresses.
At Palisades, two cranes have been reanalyzed to CMAA-70 design criteria.
The NUREG-0612 heavy loads evaluation of the reactor building polar crane was performed to CMAA 70. A redesign of the Spent Fuel Pool Crane for dry fuel storage included a NUREG-0612 evaluation to CMAA-70 design criteria.
The limiting components of the containment polar crane (135 tons) and the redesigned spent fuel pool crane (110 tons) are now rated for CMAA-70 Service Level A - Standby or Infrequent Service. Service Level A cranes are designed to stress limits which assume either 20,000 to 100,000, or 200,000, rated lifts in a design lifetime.
Analysis Containment Polar Crane: The polar crane was originally designed to Electric Overhead Crane Institute Specification 61. The subsequent NUREG-0612 heavy loads evaluation of the polar crane was performed to CMAA 70 (1975).
The minimally-rated components are CMAA 70 Service Level A. Since the minimally-rated components are Service Level A, the effective crane design life for fatigue or allowed number of rated lifts depends on this classification, which assumes 20,000 to 100,000 rated lifts in a design lifetime.
Separate evaluations have been performed of polar crane planned engineered lifts (over the rated capacity). The evaluations were done to ANSI/ASME Standard B30.2 (1996). Lifts have been evaluated and approved up to 140 T, less than 4 percent over the 135 T rating.
DSAR CHAPTER 1 - INTRODUCTION & GENERAL DESCRIPTION OF FACILITY Revision 36 SECTION 1.9 Page 1.9-14 of 1.9-14 Spent Fuel Pool Crane: The redesign to 110 T for dry cask storage included a NUREG-0612 evaluation to CMAA-70 Service Level A design criteria, and other considerations; and replacement of the trolley with a 110 T single-failure-proof trolley meeting NUREG-0554 guidelines. The structural redesign and evaluation considered the load combinations of ASME NOG 1, Section NOG 4140, Load Combinations.
Disposition: 10 CFR 54.21(c)(1)(i)
Polar Crane (L-1): Polar crane rated or near-rated lifts are limited to the reactor head plus CRDMs, insulation, and shielding, in accordance with site procedures. Only a few rated lifts are performed each refueling outage, and none during operation. Therefore this machine cannot realistically approach the 20,000 to 100,000 rated lifts, assumed for components evaluated to CMAA 70 (1975) Service Level A, during a 60-year licensed operating period.
Spent Fuel Pool Crane: Approximately 11 dry cask storage campaigns are expected between rerating and the end of the 60-year extended license. This will require loading about 64 casks. Each will require about two lifts of 100 T or more per cask, and some additional lifts of between 50 and 100 T. The total for 64 casks and 11 campaigns is about 140 lifts of 100 T or more, and about 162 lifts between 50 and 100 T. Other lifts, and lifts prior to rerating the crane, were determined to be inconsequential. Therefore, this machine can not realistically approach the 20,000 to 100,000 rated lifts assumed for its design evaluation during the 60-year extended licensed operating period.
1.9.3 NEWLY IDENTIFIED STRUCTURES, SYSTEMS, AND COMPONENTS After a renewed license is issued, 10 CFR 54.37(b) requires that DSAR updates include any Systems, Structures, or Components (SCCs) newly identified that would have been subject to an aging management review or evaluation of time-related aging analyses in accordance with 10 CFR 54.21.
The FSAR update must describe how the effects of aging will be managed such that the intended functions in 10 CFR 54.4(b) will be effectively maintained during the period of extended operation.
In accordance with 10 CFR 54.37(b), Table 1-10 provides a listing of newly identified SSCs and describes how the effects of aging will be managed such that intended functions will be maintained during the period of extended operation.