ML18053A740
| ML18053A740 | |
| Person / Time | |
|---|---|
| Site: | Lee  |
| Issue date: | 12/19/2017 |
| From: | Donahue J Duke Energy Carolinas |
| To: | Office of New Reactors |
| Hughes B | |
| References | |
| DUKE, DUKE.SUBMISSION.15, LEE.NP, LEE.NP.1 | |
| Download: ML18053A740 (197) | |
Text
UFSAR Table of Contents 1 Introduction and General Description of the Plant 2 Site Characteristics 3 Design of Structures, Components, Equipment and Systems 4 Reactor 5 Reactor Coolant System and Connected Systems 6 Engineered Safety Features 7 Instrumentation and Controls 8 Electric Power 9 Auxiliary Systems 10 Steam and Power Conversion 11 Radioactive Waste Management 12 Radiation Protection 13 Conduct of Operation 14 Initial Test Program 15 Accident Analyses 16 Technical Specifications 17 Quality Assurance 18 Human Factors Engineering 19 Probabilistic Risk Assessment UFSAR Formatting Legend Description Original Westinghouse AP1000 DCD Revision 19 content Departures from AP1000 DCD Revision 19 content Standard FSAR content Site-specific FSAR content Linked cross-references (chapters, appendices, sections, subsections, tables, figures, and references)
14.1 Specific Information to be Included in Preliminary/Final Safety Analysis Reports ........................................................................................................ 14.1-1 14.2 Specific Information to be Included in Standard Safety Analysis Reports ... 14.2-1 14.2.1 Summary of Test Program and Objectives ................................. 14.2-1 14.2.1.1 Construction and Installation Test Program Objectives ................................................................ 14.2-2 14.2.1.2 Preoperational Test Program Objectives ................. 14.2-2 14.2.1.3 Startup Test Program Objectives ............................. 14.2-3 14.2.1.4 Testing of First of a Kind Design Features .............. 14.2-3 14.2.1.5 Credit for Previously Performed Testing of First of a Kind Design Features .............................................. 14.2-4 14.2.2 Organization, Staffing, and Responsibilities ............................... 14.2-4 14.2.2.1 PT&O Organization.................................................. 14.2-4 14.2.2.2 PT&O Organization Personnel Qualifications and Training .................................................................... 14.2-5 14.2.2.3 Joint Test Working Group ........................................ 14.2-6 14.2.2.4 Site Construction Group (Architect Engineer).......... 14.2-9 14.2.2.5 Site Preoperational Test Group ............................. 14.2-10 14.2.2.6 Site Startup Test Group ......................................... 14.2-10 14.2.3 Test Specifications and Test Procedures ................................. 14.2-11 14.2.3.1 Conduct of Test Program ....................................... 14.2-12 14.2.3.2 Review of Test Results .......................................... 14.2-15 14.2.3.3 Test Records ......................................................... 14.2-16 14.2.4 Compliance of Test Program with Regulatory Guides .............. 14.2-16 14.2.5 Utilization of Reactor Operating and Testing Experience in the Development of Test Program .................................................. 14.2-16 14.2.5.1 Use of OE During Test Procedure Preparation ..... 14.2-19 14.2.5.2 Sources and Types of Information Reviewed for ITP Development ................................................... 14.2-19 14.2.5.3 Conclusions from Review ...................................... 14.2-19 14.2.5.4 Summary of Test Program Features Influenced by the Review ............................................................. 14.2-20 14.2.5.5 Use of OE during Conduct of ITP .......................... 14.2-20 14.2.6 Use of Plant Operating and Emergency Procedures ................ 14.2-20 14.2.6.1 Operator Training and Participation during Certain Initial Tests (TMI Action Plan Item I.G.1, NUREG-0737) ....................................................... 14.2-21 14.2.7 Initial Fuel Loading and Initial Criticality ................................... 14.2-21 14.2.7.1 Initial Fuel Loading ................................................. 14.2-21 14.2.7.2 Initial Criticality ....................................................... 14.2-22 14.2.7.3 Power Ascension ................................................... 14.2-23 14.2.8 Test Program Schedule ............................................................ 14.2-24 14.2.9 Preoperational Test Descriptions ............................................. 14.2-25 14.2.9.1 Preoperational Tests of Systems with Safety-Related Functions .................................................. 14.2-26 14.2.9.2 Preoperational Testing of Defense-in-Depth Systems ................................................................. 14.2-52 14-i Revision 1
Radioactive Systems ............................................. 14.2-73 14.2.9.4 Preoperational Tests of Additional Nonsafety-Related Systems .................................................... 14.2-78 14.2.10 Startup Test Procedures ........................................................... 14.2-93 14.2.10.1 Initial Fuel Loading and Precritical Tests ............... 14.2-94 14.2.10.2 Initial Criticality Tests ........................................... 14.2-107 14.2.10.3 Low Power Tests ................................................. 14.2-110 14.2.10.4 Power Ascension Tests ....................................... 14.2-116 14.3 Certified Design Material ............................................................................. 14.3-1 14.3.1 CDM Section 1.0, Introduction .................................................... 14.3-2 14.3.2 CDM Section 2.0, System Based Design Descriptions and ITAAC ......................................................................................... 14.3-2 14.3.2.1 Design Descriptions ................................................. 14.3-3 14.3.2.2 Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) ....................................................... 14.3-6 14.3.2.3 Site-Specific ITAAC (SS-ITAAC) ............................. 14.3-8 14.3.3 CDM Section 3.0, Non-System Based Design Descriptions and ITAAC ................................................................................ 14.3-10 14.3.3.1 Waterproof Membrane ITAAC ............................... 14.3-10 14.3.3.2 Pipe Rupture Hazard Analysis ITAAC ................... 14.3-10 14.3.3.3 Piping Design ITAAC ............................................. 14.3-11 14.3.4 Certified Design Material Section 4.0, Interface Requirements ........................................................................... 14.3-11 14.3.5 CDM Section 5.0, Site Parameters ........................................... 14.3-11 14.3.6 Initial Test Program .................................................................. 14.3-12 14.3.7 Elements of AP1000 Design Material Incorporated into the Certified Design Material .......................................................... 14.3-12 14.3.8 Summary .................................................................................. 14.3-13 14.3.9 References ............................................................................... 14.3-13 14.4 Combined License Applicant Responsibilities ............................................. 14.4-1 14.4.1 Organization and Staffing ........................................................... 14.4-1 14.4.2 Test Specifications and Procedures ........................................... 14.4-1 14.4.3 Conduct of Test Program ........................................................... 14.4-1 14.4.4 Review and Evaluation of Test Results ...................................... 14.4-1 14.4.5 Interface Requirements .............................................................. 14.4-1 14.4.6 First-Plant-Only and Three-Plant-Only Tests .............................. 14.4-1 PENDIX 14A DESIGN ACCEPTANCE CRITERIA/ITAAC CLOSURE PROCESS...........14A-1 14-ii Revision 1
3-2 Design Basis Accident Analysis ................................................................ 14.3-17 3-3 Anticipated Transient Without Scram ........................................................ 14.3-34 3-4 Fire Protection ........................................................................................... 14.3-35 3-5 Flood Protection ........................................................................................ 14.3-37 3-6 Probabilistic Risk Assessment .................................................................. 14.3-38 3-7 Radiological Analysis ................................................................................ 14.3-48 3-8 Severe Accident Analysis .......................................................................... 14.3-51 3-201 Not Used ................................................................................................... 14.3-52 3-202 Not Used ................................................................................................... 14.3-53 3-203 Not Used ................................................................................................... 14.3-54 3-204 Not Used ................................................................................................... 14.3-55 14-iii Revision 1
Reports applicable to the AP1000.
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purpose of this section is to describe the test program that is performed during initial startup of AP1000 plant.
overall objective of the test program is to demonstrate that the plant has been constructed as igned, that the systems perform consistent with the plant design, and that activities culminating in ration at full licensed power including initial fuel load, initial criticality, and power ascension are ormed in a controlled and safe manner.
operational and/or startup testing is performed on those systems that are:
Relied upon for safe shutdown and cooldown of the reactor under normal plant conditions and for maintaining the reactor in a safe condition for an extended shutdown period; Relied upon for safe shutdown and cooldown of the reactor under transient and postulated accident conditions and for maintaining the reactor in a safe condition for an extended shutdown period following such conditions; Relied upon for establishing conformance with safety limits or limiting conditions for operation that will be included in the facility technical specifications; Classified as engineered safety features actuation systems (ESFAS) or are relied upon to support or ensure operation of engineered safety features actuation systems within design limits; Assumed to function or for which credit is taken in the accident analysis of the AP1000 as described in this Design Control Document.
Used to process, store, control, or limit the release of radioactive materials.
Other systems identified in Regulatory Guide 1.68, Revision 2, Appendix A that are in the AP1000 and are not captured by criteria a) through f).
inspections, tests, analyses and acceptance criteria of 10 CFR 52.47 (a)(1)(vi) relating to the 000 design are found in the AP1000 Certified Design Material (see Section 14.3).
initial plant test program consists of a series of tests categorized as construction and installation, operational, and startup tests. These tests are discussed in Section 14.4.
Construction and installation tests are performed to determine that plant structures, components, and systems have been constructed or installed correctly and are operational.
Preoperational tests are performed after construction and installation tests, but prior to initial fuel loading to demonstrate the capability of plant systems to meet performance requirements.
Startup tests begin with the initial fuel loading and are performed to demonstrate the capability of individual systems, as well as the integrated plant, to meet performance requirements.
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de 1.206, Part I, Section C.I.14.2.
ITP is applied to structures, systems, and components that perform the functions described in Regulatory Guide 1.68 evaluation in Section 1.9. The ITP is also applied to other structures, ems and components. The Startup Administrative Manual includes a list of the AP1000 ctures, systems and components to which the ITP is applied.
2.1.1 Construction and Installation Test Program Objectives adequacy of construction, installation, and preliminary operation of components and systems is fied by a construction and installation test program.
is program, various electrical and mechanical tests are performed including the following:
Cleaning and flushing Hydrostatic testing Checks of electrical wiring Valve testing Energization and operation of equipment Calibration of instrumentation a system basis, completion of this program demonstrates that the system is ready for operational testing.
tracts for tests constituting the construction and installation test program are not provided in port of Design Certification. Development of the construction and installation tests is based on the ineering information for the equipment and systems installed.
2.1.2 Preoperational Test Program Objectives owing construction and installation testing, preoperational tests are performed to demonstrate equipment and systems perform in accordance with design criteria so that initial fuel loading, al criticality, and subsequent power operation can be safely undertaken. Preoperational tests at ated pressure and temperature are referred to as hot functional tests.
general objectives of the preoperational test program are the following:
Demonstrate that essential plant components and systems, including alarms and indications, meet appropriate criteria based on the design Provide documentation of the performance and condition of equipment and systems Provide baseline test and operating data on equipment and systems for future use and reference Operate equipment for a sufficient period to demonstrate performance Demonstrate that plant systems operate on an integrated basis tracts for the preoperational tests for portions of systems/components that perform safety-related tions; perform defense-in-depth functions; contain, transport, or isolate radioactive material; and 14.2-2 Revision 1
nt operating, emergency, and surveillance procedures are incorporated into the initial test gram procedures. These procedures are verified through use, to the extent practicable, during the operational test program and revised if necessary, prior to fuel loading.
nt equipment used in the performance of preoperational tests is operated in accordance with ropriate operating procedures, thereby giving the plant operating staff an opportunity to gain erience in using these procedures and demonstrating their adequacy prior to plant initial criticality.
2.1.3 Startup Test Program Objectives startup test program begins with initial fuel loading after the preoperational testing has been cessfully completed.
tup tests can be grouped into four broad categories:
Tests related to initial fuel loading Tests performed after initial fuel loading but prior to initial criticality Tests related to initial criticality and those performed at low power (less than 5 percent)
Tests performed at power levels greater than 5 percent ing performance of the startup test program, the plant operating staff has the opportunity to obtain tical experience in the use of normal and abnormal operating procedures while the plant gresses through heatup, criticality, and power operations.
general objectives of the startup test program are:
Install the nuclear fuel in the reactor vessel in a controlled and safe manner.
Verify that the reactor core and components, equipment, and systems required for control and shutdown have been assembled according to design and meet specified performance requirements.
Achieve initial criticality and operation at power in a controlled and safe manner.
Verify that the operating characteristics of the reactor core and associated control and protection equipment are consistent with design requirements and accident analysis assumptions.
Obtain the required data and calibrate equipment used to control and protect the plant.
Verify that the plant is operating within the limits imposed by the Technical Specifications.
tracts of the startup tests are provided in this section.
2.1.4 Testing of First of a Kind Design Features t of a kind (FOAK) testing may occur in any of the phases, depending on the nature of the testing required sequencing of the tests. When testing FOAK design features, applicable operating erience from previous test performance on other AP1000 plants is reviewed, where available, and ITP modified as needed based on those lessons learned.
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andard design. In such cases, credit may be taken for the previously performed tests. A ussion is included in the startup test reports of the results of those tests that are credited.
2.2 Organization, Staffing, and Responsibilities AP1000 plant test and operations (PT&O) organization is described in Subsection 14.2.2.1. The anization for operating and maintaining the AP1000 plant is described in Section 13.1.
PT&O organization structure (organizational chart) is included in the Startup Administrative ual.
le 13.4-201 provides milestones for ITP implementation.
2.2.1 PT&O Organization ITP is the responsibility of the PT&O Organization. The ITP includes three phases of testing:
Construction and Installation Testing Preoperational Testing Startup Testing 2.2.1.1 Manager In Charge of PT&O manager in charge of PT&O reports directly to the plant manager. The manager in charge of O manages the ITP. The manager in charge of PT&O is responsible for:
Staffing the PT&O Organization.
Developing, reviewing, and approving the administrative and technical procedures associated with the preoperational and startup phases.
Managing the ITP and personnel.
Implementing the ITP schedule.
Managing contracts associated with the ITP.
2.2.1.2 Functional Manager In Charge of PT&O Support functional manager in charge of PT&O support reports directly to the manager in charge of O. The functional manager in charge of PT&O support plans and schedules procedure elopment to support startup. The functional manager in charge of PT&O support verifies that the documents conform to the approved project procedures.
functional manager in charge of PT&O support reviews and approves test procedures. These edures are used to demonstrate that a system and its components meet the design and ormance criteria.
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ineers are responsible for developing system test procedures.
2.2.1.4 Functional Manager In Charge of Startup functional manager in charge of startup reports directly to the manager in charge of PT&O. The tional manager in charge of startup manages the preoperational and startup testing. The tional manager in charge of startup is responsible for:
Participating in the Joint Test Working Group (JTWG) and ensuring that the JTWG reviews and approves administrative and test procedures. The JTWG structure and responsibilities are defined in Subsection 14.2.2.3.
Preparing a detailed preoperational and startup testing schedule.
Coordinating construction turnover to the PT&O organization.
Informing the functional manager in charge of PT&O when vendor support essential to preoperational and startup testing is required, and coordinating vendor participation.
Supervising and directing the startup engineers.
Involving operations personnel in testing activities. Utilizing operations personnel, to the extent practical, as test witnesses or test performers to provide the operations personnel with experience and knowledge.
Developing and implementing administrative controls to address system and equipment configuration control.
Maintaining the startup schedule.
Maintaining a daily startup log and issuing periodic progress reports that identify overall progress and potential challenges.
2.2.1.5 Startup Engineers startup engineers report directly to the functional manager in charge of startup. The startup ineers are responsible for:
Complying with administrative controls.
Identifying any special or temporary equipment or services needed to support testing.
Coordinating testing with involved groups.
Reviewing and evaluating test results.
2.2.2 PT&O Organization Personnel Qualifications and Training cedures are prepared to confirm that test personnel have adequate training, qualification and ification. Records are kept for extent of experience, involvement in procedure and test elopment, training programs, and level of qualification. The training organization qualifies Test 14.2-5 Revision 1
eptable qualifications of non-supervisory test engineers follow the guidance provided in ulatory Guide 1.28 as discussed in Appendix 1A, i.e., ASME NQA-1-1994, Appendix 2A-1, mandatory Guidance on the Qualification of Inspection and Test Personnel.
training program/procedures shall include:
The education, training, experience, and qualification requirements of supervisory personnel, test personnel, and other major participating organizations responsible for managing, developing, or conducting each test phase, or development of testing, operating, and emergency procedures.
The establishment of a training program for each organizational unit, with regard to the scheduled preoperational and initial startup testing. This training program provides meaningful technical information beyond that obtained in the normal startup test program and provide supplemental operator training. This program also satisfies the criteria described in TMI Action Plan Item I.G.1 of NUREG-0660 and NUREG-0737.
Startup Administrative Manual (Procedure) shall include:
The implementation of measures to verify that personnel formulating and conducting test activities are not the same personnel who designed or are responsible for satisfactory performance of the system(s) or design features(s) being tested. This provision does not preclude members of the design organization from participating in test activities. This description also includes considerations of staffing effects that could result from overlapping initial test programs at multi-unit sites.
2.2.3 Joint Test Working Group Joint Test Working Group (JTWG) consists of an organizational group of authorized esentative personnel from the Plants operations and support group functions, Westinghouse tric Company (WEC), Architect Engineer (AE) and other test support groups as identified below.
Licensee has the overall responsibility for conduct of the Startup Test Program. The stinghouse Startup Manager may be assigned overall responsibility and authority for technical ction of the Startup Test Program and may act as the JTWG Chairman.
JTWG Chairman reports to the Chairman of the Plant Owners Operations Review Committee RC) or qualified designee for matters of Startup test authority and acceptance.
JTWG provides the following administrative oversight activities associated with the Startup Test gram:
Review, evaluate and approve Startup Test Program administrative and test procedures.
Oversee the implementation of the Preoperational Test Program and the Startup Test Program, including planning, scheduling and performance of Preoperational and Startup testing.
Review and evaluate Construction, Preoperational and Startup test results and test turnover packages.
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Licensees Operations Group Licensees Maintenance Group Site Preoperational Test Group Site Startup Test Group Licensees Engineering Group Licensees Corrective Action Organization Westinghouse Site Engineering Group Licensees Health Physics/Chemistry Group Licensees Quality Assurance Group following are additional generic details of the key responsibilities, authorities and interfaces of Licensee Organizations delineated above:
Operations Group The Operations Group has the overall responsibility for Plant Operations, including administrative control and tagouts subsequent to system turnover. Their primary interfaces are with the Licensee Engineering and Technical Support organizations as well as the Westinghouse Engineering Organization, Preoperational and Startup Testing Teams and Construction Services Group.
Maintenance Group The Maintenance Group has the overall responsibility for the Maintenance of Plant systems and components subsequent to System Turnover. They are key participants and maintainers of system maintenance control and tagouts. Their primary interfaces are with the Licensee Operations Group and Technical Support organizations, as well as the Westinghouse Engineering Organization, Preoperational and Startup Testing Teams and Construction Services Group.
Corrective Action Organization The Corrective Action Organization may be an organization specific to itself or may be a part of the Performance Assessment organization, the Quality Organization or another organization.
This organization, together with every other site organization, is responsible for the administration and management of the corrective action program, as well as the identification of conditions adverse to quality. This organization interfaces with site organizations and identifies and documents conditions which need to be documented in the corrective action program.
Engineering Group This group has the primary responsibility for site engineering and design oversight of the plant components and systems, as well as interfacing with the vendor engineering organization. This 14.2-7 Revision 1
Assurance Program is delegated and implemented as agreed to by Westinghouse.
Westinghouse test personnel training is per certified design.
Health Physics/Chemistry Group This Technical Support organization has the responsibility and authority to maintain Health Physics and system chemistry conditions at the plant, particularly after system turnover. This organization primarily interfaces with the Licensee Operations Group, as well as the Westinghouse Engineering Organization, Preoperational and Startup Testing Teams and Construction Services Group.
Quality Assurance Group This group has the responsibility to verify that the applicable site Quality commitments are met within the scope of work performed at the site. This includes meeting the Criteria of 10 CFR 50 Appendix B. The primary interfaces for this group are the Licensee Operations Group and Technical Support organizations, including Quality Control and other quality organizations, as well as the Westinghouse Engineering Organization, Preoperational and Startup Testing Teams and Construction Services Group.
Site Preoperational Test Group This group has the primary responsibility for the development, maintenance and performance of the site preoperational procedures at the site. The primary interfaces for this group are the Licensee Operations Group and Technical Support organizations, as well as the Westinghouse Engineering Organization, Startup Testing Teams and the Construction Services Group.
Additional specific information regarding this organizations responsibilities and interfaces is described in Subsection 14.2.2.5, below. Once preoperational testing is complete, this group turns systems over to the Startup Group.
Site Startup Test Group This group has the primary responsibility for the development, maintenance and performance of the site startup procedures at the site. The primary interfaces for this group are the Licensees Operations Group and Technical Support organizations, as well as the Westinghouse Engineering Organization, Preoperational Testing Team and the Construction Services Group.
Additional specific information regarding this organizations responsibilities and interfaces is described in Subsection 14.2.2.6, below. The Startup Test Group turns over systems to the licensee when testing is complete.
Westinghouse Site Engineering Group This group has the primary responsibility for the vendor interface between the site and the vendors home offices, as well as the design authority for the primary vendors components and systems. The various Westinghouse site leads for specific disciplines are a part of this organization. This organization primarily interfaces with Licensee Operations Group, as well as the Westinghouse Engineering Organization, Preoperational and Startup Testing Teams and Construction Services Group. The responsibility for training the testing personnel in accordance with the applicable Quality Assurance Program is delegated and implemented as agreed to by Westinghouse and the Licensee. Westinghouse test personnel training is per certified design.
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gram:
Construction Group The Construction group has the primary responsibility for the construction and construction testing of the Balance of Plant (BOP) engineering systems and components. During Construction and Construction Testing, this group has authority over administrative control and tagouts of these systems. Their main interface is with the System Preoperational and Startup Testing Groups, as well as the Licensee Operations Group. The Construction Group is responsible for addressing open items in the system turnover punch lists to address turnover acceptability of the system.
Construction Services Group The Construction Services Group primarily supports the Construction Group with activities necessary to support construction of systems and testing of the BOP systems and components, including the construction of scaffolding, installation and removal of insulation, and similar activities. With agreement between the necessary parties, this group may also support the Westinghouse Site Engineering Group with similar activities on the primary side. The primary interfaces of this group are the Construction Group and the organizations of the JTWG.
Construction Services Procurement Group The Construction Services Procurement Group is responsible for the quality procurement of components and equipment necessary to support plant construction and testing. The primary interfaces of this group include the Construction Services Group and the Construction Services Quality Group.
Construction Services Quality Group The Construction Services Quality Group is responsible for the oversight of the Quality Program during Construction Activities, including those pertinent to 10 CFR 50 Appendix B and the disposition of Significant Construction Deficiencies, 10 CFR 50.55(e) reports as necessary. This group primarily interfaces with the Construction and Services groups as well as the Westinghouse Site Engineering group and the JTWG.
Construction Services Training Group This group is primarily responsible for the training and qualification of Site Construction Personnel in accordance with the applicable Quality Assurance Program. Their primary interface is with the qualified Construction personnel.
Site Construction Group performs the following functions and scope of work, as necessary to port the Site Startup Test Program:
Construction Installation and Testing, including management of construction testing documentation.
Construction and Installation activities required to support Preoperational and Startup Test Programs.
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Turnover of Construction and Installation tested equipment, systems, and testing documentation to the Site Preoperational Test Group.
2.2.5 Site Preoperational Test Group Site Preoperational Test Group consists of the following, as necessary to support the Site Startup Program:
Engineering Leads Preoperational Test Teams Site Preoperational Test Group performs the following functions and scope of work, as essary to support the Site Startup Test Program:
Coordinate tagging and maintenance prior to turnover to the Licensee to support system acceptance testing.
Accept systems for turnover from the installation organization.
Plan, scope and schedule plant systems for test to support the plant Preoperational Test Program.
Manage and oversee the testing of plant systems to support the Plant Hot-Functional Test Program.
Resolve open items and exceptions identified during implementation of the Preoperational Test Program.
Accept and turn over Preoperational Test Packages to the Site Licensee.
Support completion of Hot-Functional Test Program.
Coordinate other support tasks required during Startup Testing activities with responsible groups (e.g., Licensees Organization).
2.2.6 Site Startup Test Group Site Startup Test Group consists of the following, as necessary to support the Site Startup Test gram:
Engineering Leads Startup Test Teams Site Startup Test Group performs the following functions and scope of work, as necessary to port the Site Startup Test Program:
Coordinate tagging and maintenance as required to support system and equipment acceptance testing.
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Plant Startup.
Manage and oversee the testing of plant systems, structures and components to support the Plant Power Ascension Test Program.
Resolve open items and exceptions identified during implementation of the Startup Test Program.
Accept and turn over Startup Test Packages to the Site Licensee.
Coordinate other support tasks required during Startup Testing activities with responsible groups (e.g., Licensees Organization).
2.3 Test Specifications and Test Procedures operational and startup tests are performed using test specifications and test procedures.
the preoperational and startup tests, test specifications are written to specify the following:
Objectives for performing the test Test prerequisites Initial test conditions Data requirements Criteria for test results evaluation and reconciliation methods and analysis as required each test, the test procedure specifies the following:
Objectives for performing the test Prerequisites that must be completed before the test can be performed Initial conditions under which the test is started Special precautions required for the safety of personnel or equipment Instructions delineating how the test is to be performed Identification of the required data to be obtained and the methods for documentation Data reduction analysis methods as appropriate specifications and procedures are developed and reviewed by personnel with appropriate nical backgrounds and experience. This includes the participation of principle design anizations in the establishment of test performance requirements and acceptance criteria.
cifically, the principle design organizations will provide scoping documents (i.e., preoperational startup test specifications) containing testing objectives and acceptance criteria applicable to its pe of design responsibility as discussed in Subsection 14.4.5.
ilable information on operating or testing experiences of operating reactors is factored into the specifications and test procedures as appropriate.
ies of the test specifications and test procedures for the startup tests are provided to NRC ection personnel not less than 60 days prior to the scheduled fuel loading date.
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Tests of systems/components that perform safety-related functions Tests of systems/components that are nonsafety-related but perform defense in-depth functions.
specifications and test procedures for preoperational tests described in Subsections 14.2.9.3 14.2.9.4 of the plant systems/components which perform no safety-related or defense-in-depth tions are available to NRC inspection personnel prior to the scheduled performance of these s.
operational and startup tests are performed with the quality assurance requirements as specified ection 17.5.
Startup Administrative Manual shall include the following controls:
Controls to provide test procedures that include appropriate prerequisites, objectives, safety precautions, initial test conditions, methods to direct and control test performance, and acceptance criteria by which the test is evaluated.
Controls for the format of individual test procedures to provide consistency with the guidance contained in Regulatory Guide 1.68; or provide justifications for any exceptions.
Controls to provide for participation of the principal design organizations in establishing test objectives, test acceptance criteria, and related performance requirements during the development of detailed test procedures. Each test procedure should include acceptance criteria that account for the uncertainties used in transient and accident analyses. The participating system designers should include the nuclear steam supply system vendor, Architect Engineer, and other major contractors, subcontractors, and vendors, as applicable.
Controls to provide for personnel with appropriate technical backgrounds and experience to develop and review test procedures. Persons filling designated management positions should perform final procedure review and approval.
Controls to make the approved test procedures for satisfying FSAR testing commitments are made available to the NRC inspectors approximately 60 days prior to their intended use.
2.3.1 Conduct of Test Program inistrative procedures and requirements that govern the activities of the conduct of the initial test gram include the following:
Format and content of test procedures Process for both initial issue and subsequent revisions of test procedures Review process for test results Process for resolution of failures to meet performance criteria and of other operational problems or design deficiencies 14.2-12 Revision 1
Controls to monitor the as-tested status of each system and modifications including retest requirements deemed necessary for systems undergoing or already having completed testing Qualifications and responsibilities of the positions within the startup group startup administrative procedures supplement normal plant administrative procedures by ressing those administrative issues that are unique to the startup program.
Startup Administrative Manual (procedure) governs the initial testing and is issued no later than ays prior to the beginning of the pre-operational phase. Testing during all phases of the test gram is conducted using approved test procedures.
2.3.1.1 Procedure Verification e procedures may be approved for implementation weeks or months in advance of the eduled test date, a review of the approved test procedure is required before commencement of ing. The test engineer is responsible for verifying:
Drawing and document revision numbers listed in the reference section of the test procedure agree with the latest revisions.
The procedure text reflects any design and licensing (i.e., FSAR and Technical Specifications) changes made since the procedure was originally approved for implementation.
Any new (since preparation of the procedure) Operating Experience lessons learned are incorporated into individual test procedures.
cedures require signoff verification for prerequisites and instruction steps. This signoff includes tification of the person doing the signoff and the date and time of completion.
engineers maintain chronological logs of test status to facilitate turnover and aid in maintaining rational configuration control. These logs become part of the test documentation.
re is a documented turnover process to make known the test status and equipment configuration n personnel transfer responsibilities, such as during a shift change.
briefings are conducted for each test in accordance with administrative procedures. When a shift nge occurs before test completion, another briefing occurs before resumption or continuation of test.
a collected is marked or identified with test, date, and person collecting data. This data becomes of the test documentation.
plant corrective action program is used to document deficiencies, discrepancies, exceptions,
-conformances and failures (collectively known as test exceptions) identified in the ITP. The ective action documentation becomes part of the test documentation. WEC and/or other design anizations participate in the resolution of design-related problems that result in, or contribute to, a re to meet test acceptance criteria.
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inistrative procedures detail the test documentation review and approval. Review and approval st documentation includes the test engineer, testing supervisor, Startup Group manager, WEC representative or appropriate vendor, and JTWG. Final approval is by the plant manager.
nt readiness reviews are conducted to assure that the plant staff and equipment are ready to eed to the next test phase or plateau.
2.3.1.2 Work Control Startup Group is responsible for preparing work requests when assistance is required from the struction organization. Work requests are issued in accordance with site-specific procedures erning the work management process. The plant staff, upon identifying a need for Construction anization assistance, coordinates their requirements through the appropriate Startup Test ineer.
vities requiring Construction organization work efforts are performed under the plant tagging edures. Tagging requests are governed by a site-specific procedure for equipment clearance.
ging procedures shall be used for protection of personnel and equipment and for jurisdictional or odial conditions that have been turned over in accordance with the turnover procedure.
Startup Group is responsible for supervising minor repairs and modifications, changing ipment settings, and disconnecting and reconnecting electrical terminations as stipulated in a cific test procedure. Startup Test Engineers may perform independent verification of changes e in accordance with approved test procedures.
2.3.1.3 System Turnover ing the construction phase, systems, subsystems, and equipment are completed and turned over n orderly and well-coordinated manner. Guidelines are established to define the boundary and rface between related system/subsystem and are used to generate boundary scope documents; example, marked-up piping and instrument diagrams (P&IDs) and electrical schematic diagrams provided for scheduling and subsequent development of component and system turnover kages. The system turnover process includes requirements for the following:
Documenting inspections performed by the construction organization (e.g., highlighted drawings showing areas inspected).
Documenting results of construction testing.
Determining the construction-related inspections and tests that need to be completed before preoperational testing begins. Any open items are evaluated for acceptability of commencing preoperational testing.
Developing and implementing plans for correcting adverse conditions and open items, and means for tracking such conditions and items.
Verifying completeness of construction and documentation of incomplete items.
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test procedures. The test procedures contain restoration steps and retesting necessary to confirm sfactory restoration to the required configuration. Modifications may be performed by the struction organization or the plant staff processes prior to NRC issuance of the 10 CFR 52.103(g) ing. If the modification invalidates a previously completed ITAAC, then that ITAAC is erformed. Each modification is reviewed to determine the scope of post-modification testing that be performed. Testing is conducted and documented to maintain the validity of preoperational ing and ITAAC. Alterations made following NRC issuance of the 10 CFR 52.103(g) finding are in ordance with plant processes and meet license conditions. Modifications that require changes to AC require NRC approval of the ITAAC change.
2.3.1.5 Conduct of Maintenance During the Initial Test Program rective or preventive maintenance activities are reviewed to determine the scope of post-ntenance testing to be performed. Prior to NRC issuance of the 10 CFR 52.103(g) finding, post-ntenance testing is conducted and documented to maintain validity of associated preoperational ing and ITAAC remain valid. Maintenance performed following NRC issuance of the CFR 52.103(g) finding is in accordance with plant staff processes and meets license conditions.
2.3.2 Review of Test Results l review of the individual tests is discussed in Section 14.4.
2.3.2.1 Review and Approval Responsibilities n completion of a test, the startup engineer is responsible for:
Reviewing the test data.
Evaluating the test results.
Verifying that the acceptance criteria are met.
Verifying that the test results that do not meet acceptance criteria are entered into the corrective action program.
Verifying that the results of retesting do not invalidate ITAAC acceptance criteria.
results are reviewed and approved by the JTWG. Review and approval of test results are kept ent such that succeeding tests are not dependent on systems or components that have not been quately tested. Test exceptions which do not meet acceptance criteria are identified to the cted and responsible design organizations and entered into the corrective action program.
lementation of corrective actions and retests are performed as required.
r to initial fuel load, the results of the preoperational test phase are comprehensively reviewed by PT&O organization and the JTWG to verify the results indicate that the required plant structures, ems, and components are capable of supporting the initial fuel load and subsequent startup ing. The plant manager approves fuel loading.
h area of startup testing is reviewed and evaluated by the PT&O organization and the JTWG. The results at each power ascension testing power plateau are reviewed and evaluated by the PT&O anization and the JTWG and approved by the plant manager before proceeding to the next 14.2-15 Revision 1
reactor vendor is responsible for reviewing and approving the results of the tests of supplied ipment. Architect Engineer representatives review and approve the results of the tests of plied equipment. Other vendors' representatives review and approve the results of the tests of plied equipment. Final approval of individual test completion is by the plant manager after roval by the Joint Test Working Group (JTWG).
2.3.2.2 Technical Evaluation h completed test package is reviewed by technically qualified personnel to confirm satisfactory onstration of plant, system or component performance and compliance with design and license ria.
2.3.3 Test Records ention periods for test records are based on considerations of their usefulness in documenting al plant performance characteristics, and are retained in accordance with Regulatory Guide 1.28.
2.3.3.1 Startup Test Reports tup test reports are generated describing and summarizing the completion of tests performed ng the ITP. A startup report is submitted at the earliest of:
9 months following initial criticality, 90 days after completion of the ITP, or 90 days after start of commercial operations. If one report does not cover all three events, then supplemental reports are submitted every three months until all three events are completed.
These reports:
Address each ITP test described in the FSAR.
Provide a general description of measured values of operating conditions or characteristics obtained from the ITP as compared to design or specification values.
Describe any corrective actions that were required to achieve satisfactory operation.
Include any other information required by license conditions.
2.4 Compliance of Test Program with Regulatory Guides section 1.9.1 and Table 1.9-1 discuss compliance with the applicable NRC regulatory guides.
2.5 Utilization of Reactor Operating and Testing Experience in the Development of Test Program design, testing, startup, and operating experience from previous pressurized water reactor plants ilized in the development of the initial preoperational and startup test program for the AP1000
- t. Other sources of experience reported and described in documents such as NRC reports, uding Inspection and Enforcement bulletins and Institute of Nuclear Power Operations (INPO) 14.2-16 Revision 1
cial tests to further establish a unique phenomenological performance parameter of the AP1000 ign features beyond testing performed for Design Certification of the AP600 and that will not nge from plant to plant, are performed for the first plant only. Because of the standardization of AP1000 design, these special tests (designated as first plant only tests) are not required on follow ts. These first plant only tests are identified in the individual test descriptions. (See sections 14.2.9 and 14.2.10.) The following is a listing of the first plant only tests, and the esponding section in which they appear t Plant Only Test Section WST Heatup Test 14.2.9.1.3 Item (h) ssurizer Surge Line Stratification Evaluation 14.2.9.1.7 Item (d) actor Vessel Internals Vibration Testing 14.2.9.1.9 - Prototype Test tural Circulation Tests]* 14.2.10.3.6, [14.2.10.3.7]*
d Cluster Control Assembly Out of Bank Measurements 14.2.10.4.6 d Follow Demonstration 14.2.10.4.22 er special tests which further establish a unique phenomenological performance parameter of the 000 design features beyond testing performed for Design Certification for the AP600 and that will change from plant to plant, are performed for the first three plants. Because of the standardization e AP1000 design, once these special tests have affirmed consistent passive system function are not required on follow plants. These tests required on the first three plants are identified in individual test descriptions (See Subsection 14.2.9). The following is a listing of the tests required he first three plants, and the corresponding section in which they appear.
st Three Plant Tests Section re Makeup Tank Heated Recirculation Tests 14.2.9.1.3 Items (k) and (w)
S Blowdown Test 14.2.9.1.3 Item (s) subsequent plants, the COL holder shall either perform the subject test, or justification shall be ided that the results of the first-plant-only tests or first-three-plant tests are applicable to the sequent plant.]*
justifications for the first-plant-only tests and the first-three-plant tests are provided below:
ST Heatup Test (14.2.9.1.3 item (h))
ing preoperational testing of the passive core cooling system, a natural circulation test of the sive residual heat removal (PRHR) heat exchanger is conducted (item f). For the first plant only, mocouples are placed in the IRWST to observe the thermal profile developed during the heatup e IRWST water during PRHR heat exchanger operation. This test will be useful in confirming the lts of the AP600 Design Certification Program PRHR tests with regards to IRWST mixing, and is ful in quantifying the conservatism in the Chapter 15 transient analyses.
to the standardization of the AP1000, the heatup and thermal stratification characteristics of the ST will not vary from plant to plant. The PRHR heat exchanger design, and the size and figuration of the IRWST are standardized, such that the heatup characteristics will not ificantly change from plant to plant.
Staff approval is required prior to implementing a change in this information.
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e Makeup Tank Heated Recirculation Tests (14.2.9.1.3 Items (k) and (w))
ing preoperational testing of the passive core cooling system, a test is performed for each plant to fy the CMT inlet piping resistances. In addition, cold draining tests of the CMTs are conducted that fy the discharge piping resistance and proper drain rate of the CMTs for each plant. For the first e plants, two additional CMT tests are conducted during hot functional testing of the RCS. These s are a natural circulation heatup of the CMTs followed by a test to verify the ability of the CMTs to sition from a recirculation mode to a draindown mode while at elevated temperature and sure.
ration of the CMTs in their natural circulation mode is conducted on the first three plants only for following reasons:
Natural circulation of the CMTs will not vary from plant to plant, provided that the other verifications discussed above are performed as specified.
Natural circulation testing of the CMTs was extensively tested as part of the Design Certification Tests.
Performance of this test results in significant thermal transients on Class 1 components including the CMTs and the direct vessel injection nozzles.
S Blowdown Test (14.2.9.1.3 Item (s))
ing preoperational testing of the passive core cooling system, the resistance of the automatic ressurization system Stage 1, 2, 3 flow path(s) is verified. For the first three plants only, an matic depressurization blowdown test is performed to verify proper operation of the ADS valves, demonstrate the proper operation of the ADS spargers to limit the hydrodynamic loads in tainment to less than design limits. This test is performed on only the first three plants for the wing reasons:
The operation of the ADS, and the resultant hydrodynamic loads will not vary significantly from plant to plant.
Full scale automatic depressurization testing was performed in the AP600 Design Certification Program. Testing was conducted to conservatively bound ADS flow rates and resultant hydrodynamic loads that will be experienced by the plant during ADS operation.
Performance of this test results in significant thermal transients on Class 1 components including the primary components. It also results in hydrodynamic loads in containment including the IRWST.
ssurizer Surge Line Stratification Evaluation (14.2.9.1.7 Item (d))
part of the AP1000 conformance to NRC Bulletin 88-11, a monitoring program will be lemented by the COL Applicant for the first AP1000 to record temperature distributions and mal displacements of the surge line piping during hot functional testing and during the first fuel e, as discussed in Subsection 3.9.3.
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essment program. This program is discussed in Subsection 3.9.2.
ural Circulation Tests (14.2.10.3.6, 14.2.10.3.7) ural circulation tests using the steam generators and the passive residual heat removal heat hanger are performed at low core power during the startup test phase of the initial test program he first AP1000. This testing of the heat removal systems meets the intent of the requirement to orm natural circulation testing and the results of this testing is factored into the operator training iscussed in Subsection 1.9.4, Item I.G.1. This test is only required to be performed once because urpose is to obtain data to benchmark the operator training simulator.
d Cluster Control Assembly Out of Bank Measurements (14.2.10.4.6) cluster control assembly out of bank measurements are performed during power ascension
- s. The test is performed at the 30-percent to 50-percent power level so the plant does not exceed king factor limits. The test is required to be performed only for the first plant because its purpose validate calculation tools and instrument responses.
d Follow Demonstration (14.2.10.4.22) ad follow demonstration test is not required by Regulatory Guide 1.68. However, the AP1000 orms load follow with grey rods, as opposed to current Westinghouse PWRs which manipulate S boron concentration to perform load follow operations. Therefore, Westinghouse has included a follow test for the first AP1000, to demonstrate the ability of the AP1000 plant to load follow.
ization of Operating Experience inistrative procedures provide methodologies for evaluating and initiating action for operating erience information (OE). This subsection describes the general use of operating experience by C in the development of the test program.
2.5.1 Use of OE During Test Procedure Preparation inistrative procedures require review of recent internal and external operating experience when paring test procedures.
2.5.2 Sources and Types of Information Reviewed for ITP Development tiple sources of operating experience were reviewed to develop this description of the ITP inistration program. These included INPO Reports, INPO Guidelines, INPO Significant Event orts, INPO Significant Operating Experience Reports and NRC Regulatory Guide 1.68.
2.5.3 Conclusions from Review following conclusions are a result of the OE review conducted to develop this ITP administration gram description:
The test procedures should provide guidance as to the expected plant response and instructions concerning what conditions warrant aborting the test. Errors and problems with the procedures should be anticipated. A means for prompt but controlled approval of changes to test procedures is needed. Critical test procedures should provide specific criteria for test termination and specific steps to conduct termination is conducted in a safe and orderly manner. Providing procedural guidance for aborting the test could prevent delays in plant 14.2-19 Revision 1
Plant simulators may prove useful in preparing for special tests and verifying procedures.
Appropriate component/system operability should be verified prior to critical tests.
The need to perform physics tests that can produce severe power tilts should be evaluated, particularly if tests at other similar reactors have provided sufficient data to verify the adequacy of the nuclear physics analysis.
Compensatory measures should be implemented in accordance with guidance for infrequently performed tests or evolutions, where appropriate.
2.5.4 Summary of Test Program Features Influenced by the Review conclusions from the preceding section were incorporated in Section 14.2.
2.5.5 Use of OE during Conduct of ITP inistrative procedures require discussion of operating experience when performing pre-job briefs ediately prior to the conduct of a test.
2.6 Use of Plant Operating and Emergency Procedures appropriate and to the extent practicable, plant normal, abnormal, and emergency operating edures are used when performing preoperational startup tests.
use of these procedures is intended:
To demonstrate the adequacy of the specific procedure or to identify changes that may be required To increase the level of knowledge of plant personnel on the systems being tested st procedure using a normal, abnormal, or emergency operating procedure references the edure directly or extracts a series of steps from the procedure in the way that accomplishes the rator training goals while safely and efficiently performing the specified testing.
se procedures are used extensively in the Human-Machine Interface Testing, which is integrated part of the Control Room Design finalization. Additionally, the AP1000 plant operating and rgency procedures are developed to support the following design finalization activities:
Human Factors Engineering Operational Task Analysis Training Simulator Development Verification and Validation of the Procedures and the Training Material AP1000 emergency, abnormal and some normal operating procedures, along with some Alarm ponse Procedures and surveillance procedures, are exercised and verified in the processes neated above and in the Control Room design finalization process.
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cedures while preoperational and startup tests are performed, which adds to their validity and the t operators training.
2.6.1 Operator Training and Participation during Certain Initial Tests (TMI Action Plan Item I.G.1, NUREG-0737) objective of operator participation is to increase the capability of shift crews to operate facilities in fe and competent manner by assuring that training for plant changes and off-normal events is ducted.
rators are trained on the specifics of the ITP schedule, administrative requirements and tests.
cific Just In Time training is conducted for selected startup tests.
ITP may result in the discovery of an acceptable plant or system response that differs from the ected response. Test results are reviewed to identify these differences and the training for rators is changed to reflect them. Training is conducted as soon as is practicable in accordance training procedures.
2.7 Initial Fuel Loading and Initial Criticality al fuel loading and subsequent initial criticality and power ascension to full licensed power are ormed during the startup test program. Prior to the initiation of these operations, the systems and ditions necessary to bring the plant into compliance with the Technical Specifications must be rable and satisfied. These operations are performed in a controlled and safe manner by using test edures that specify:
Required prerequisite testing Operational status of required systems Step-by-step instructions Precautions which must be observed Actions to be taken in the event of unanticipated or abnormal response 2.7.1 Initial Fuel Loading minimum conditions for initial core loading include:
The composition, duties, and emergency procedure responsibilities of the fuel handling crew are established.
Radiation monitors, nuclear instrumentation, manual initiation controls, and other devices to actuate alarms and ventilation controls are tested and verified to be operable.
The status of systems required for fuel loading is established and verified.
The status of protection systems, interlocks, alarms, and radiation protection equipment is established and verified for fuel loading.
Inspections of fuel and control rods have been made.
Containment integrity has been established to the extent required by the Technical Specifications.
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Required fuel handling tools are available, operational, and calibrated to include indexing of the manipulator crane with a dummy fuel element. The fuel handling tools have been successfully tested.
Reactor coolant water quality requirements are established and the reactor coolant water quality is verified.
The reactor vessel is filled with water to a level approximately equal to the center of the vessel outlet nozzles. The reactor coolant water is circulating at a rate which provides uniform mixing.
The boron concentration in the reactor coolant is verified to be equal to or greater than required by the plant Technical Specifications for refueling and is being maintained under a surveillance program.
Sources of unborated water to the reactor coolant system have been isolated and are under a surveillance program.
At least two neutron detectors are calibrated, operable, and located in such a way that changes in core reactivity can be detected and recorded. One detector is connected to an audible count rate indicator and a containment alarm.
A response check of nuclear instruments to a neutron source is required within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to loading (or resumption of loading if delayed for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or more).
l assemblies together with inserted components (control rods, burnable poison assemblies, ary and secondary neutron sources) are placed in the reactor vessel, according to an blished and approved sequence.
ing and following the insertion of each fuel assembly, until the last fuel assembly has been ed, the response of the neutron detectors is observed and compared with previous fuel loading or calculations to verify that the observed changes in core reactivity are as expected. Specific ructions are provided if unexpected changes in reactivity are observed.
ause of the unique conditions that exist during initial fuel loading, temporary neutron detectors be used in the reactor vessel to provide additional reactivity monitoring. Credit for the use of porary detectors may be taken in meeting Technical Specifications requirements on the number perable source range channels.
2.7.2 Initial Criticality owing initial fuel loading, the reactor upper internals and the pressure vessel head are installed.
itional mechanical and electrical tests are performed in preparation for critical and power rations. The following conditions exist prior to initial criticality:
The reactor coolant system is filled and vented.
Tests are completed on the control rod drive system that demonstrate that the control rods have been latched, that the control and position indication systems are functioning properly, and that the rod drop time under hot full flow conditions is less than the Technical Specifications limit.
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The reactor coolant system is at hot no-load temperature and pressure. The reactor coolant boron concentration is such that the shutdown margin requirements of the Technical Specifications are satisfied for the safe shutdown condition.
al criticality is achieved in an orderly, controlled fashion by the combination of shutdown and trol bank withdrawal and reactor coolant system boron concentration reduction.
ing the approach to initial criticality, the response of the source range nuclear instruments is used n indication of the rate of reactivity addition and the proximity to a critical condition so that cality is achieved in a controlled, predictable fashion.
es for rod withdrawal and boron reduction are specified in such a way that the startup rate is less one decade per minute.
owing criticality and prior to operation at power levels greater than 5 percent of rated power, sics tests are performed to verify that the operating characteristics of the reactor core are sistent with design predictions. During these tests, values are obtained for the reactivity worth of trol and shutdown rod banks, isothermal temperature coefficient, and critical boron concentration elected rod bank configurations.
er tests at low power include verification of the response of the nuclear instrumentation system radiation surveys.
2.7.3 Power Ascension r the operating characteristics of the reactor have been verified by low-power testing, a power ension program brings the unit to its full rated power level in successive stages. At each cessive stage, hold points are provided to evaluate and approve test results prior to proceeding to next stage. The minimum test requirements for each successive stage of power ascension are cified in the applicable startup test procedures.
ing the power ascension program, tests are performed at various power levels as follows:
Statepoint data, including secondary system heat balance measurements, are obtained at various power levels up to full licensed power. This information is used to project plant performance during power escalation, provide calibration data for the various plant control and protection systems, and provide the bases for plant trip setpoints.
At prescribed power levels, the dynamic response characteristics of the primary and secondary systems are evaluated. System response characteristics are measured for design step load changes, rapid load reductions, and plant trips.
Adequacy of the radiation shielding is verified by gamma and neutron radiation surveys.
Periodic sampling is performed to verify the chemical and radiochemical analysis of the reactor coolant.
Using the incore instrumentation as appropriate, the power distribution of the reactor core is measured to verify consistency with design predictions and Technical Specifications limits on peaking factors.
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table for generation, review, and approval of procedures as well as the actual testing and lysis of results.
operational testing is performed as system and equipment availability allows. The rdependence of systems is also considered.
uencing of the startup tests depends on specified power and flow conditions and intersystem equisites. The startup test schedule establishes that, prior to core load, the test requirements are for those plant structures, systems, and components that are relied upon to prevent, limit, or gate the consequences of postulated accidents. Testing is sequenced so that the safety of the t is not dependent on untested systems, components, or features.
te-specific ITP schedule will be provided to the NRC after issuance of the COL. This schedule will ress each major phase of the test program (including tests that are required to be completed re fuel load), as well as the organizational impact of any overlap of first unit initial testing with al testing of the second unit.
sequential schedule for individual startup tests should establish that testing is completed in ordance with plant technical specification requirements for structures, systems, and components C) operability before changing plant modes. Additionally, the schedule establishes that the safety e plant is not dependent on the performance of untested SSCs. Guidance provided in Regulatory de 1.68 is used for development of the schedule.
Startup Administrative Manual shall include the following controls:
Test Procedure Development Schedule:
- Controls to establish a schedule for the development of detailed testing, plant operating, and emergency procedures. These procedures, to the extent practical, are trial-tested and corrected during the ITP prior to fuel loading in order to establish their adequacy.
- Controls to confirm that approved test procedures are in a form suitable for review by NRC inspectors at least 60 days prior to their intended use, or at least 60 days prior to fuel loading for fuel loading and startup test procedures.
- Controls to provide timely notification to the NRC of changes in approved test procedures previously available for NRC review.
Initial Test Program Schedule:
- Controls to establish a schedule to conduct the major phases of the ITP, relative to the expected fuel loading date. This is covered in the COL.
- Controls to allow at least 9 months for conducting preoperational testing.
- Controls to allow at least 3 months for conducting startup testing, including fuel loading, low-power tests, and power-ascension tests.
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- Controls to sequence the schedule for individual startup tests, insofar as is practicable, such that testing is completed prior to exceeding 25 percent power for the plant SSCs that are relied upon to prevent, limit, or mitigate the consequences of postulated accidents. The schedule should establish that, insofar as is practicable, testing is accomplished as early in the test program as is feasible and that the safety of the plant is not dependent on the performance of untested SSCs.
milestone schedule for developing plant operating procedures is presented in Table 13.4-201.
operating and emergency procedures are available prior to start of licensed operator training
, therefore, are available for use during the ITP. Required or desired procedure changes may be tified during their use. Administrative procedures describe the process for revising plant rating procedures.
2.9 Preoperational Test Descriptions ing preoperational testing, it may be necessary to return system control to Construction anization to repair or modify the system or to correct new problems. Administrative procedures ude direction for:
Means of releasing control of systems and or components to construction.
Methods used for documenting actual work performed and determining impact on testing.
Identification of required testing to restore the system to operability/functionality/availability status, and to identify tests to be re-performed based on the impact of the work performed.
Authorizing and tracking operability and unavailability determinations.
Verifying retests stay in compliance with ITAAC.
abstracts are provided for the preoperational testing of systems/components that perform safety-ted functions; that are nonsafety-related but perform functions designated to provide defense epth; systems/components that may contain radioactive material; and other applicable nonsafety-ted systems in accordance with Regulatory Guide 1.68, Revision 2, Appendix A. A limited ber of these testing abstracts establish performance parameters of AP1000 design features that not change from plant to plant. Because the AP1000 design is standardized, these tests need be performed on the first AP1000 plant. These testing abstracts are clearly identified.
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pose purpose of the reactor coolant system testing is to verify that the as-installed reactor coolant em properly performs the following safety-related functions:
Provide reactor coolant system pressure boundary integrity as described in Section 5.2 Provide core cooling and boration in conjunction with the passive core cooling system as described in Sections 5.1 and 6.3 Measure process parameters required for safety-related actuations and safe shutdown as described in Sections 7.2, 7.3 and 7.4 Measure selected process parameters required for post-accident monitoring as described in Section 7.5 Vent the reactor vessel head as discussed in Subsection 5.4.12 ing is also performed to verify that the system properly performs the following defense-in-depth tions described in Section 5.2:
Provide forced circulation cooling of the reactor core in conjunction with heat removal by the steam generator(s) as described in Section 5.1 Provide core cooling by natural circulation of coolant in conjunction with heat removal by the steam generator(s) as described in Section 5.1 In conjunction with the steam generator(s) and normal residual heat removal system, provide the capability to remove core decay heat and cool the reactor coolant to permit the reactor to be refueled and started up in a controlled manner Provide pressurizer pressure control during normal operation Provide pressurizer level control in conjunction with the chemical and volume control system Provide pressurizer spray requisites construction testing of the reactor coolant system has been successfully completed. The operational testing of the component cooling water system, service water system, chemical and me control system, main ac power electrical power system, and required interfacing systems is pleted to the extent sufficient to support the specified testing. The reactor coolant system is filled, ted, and pressurized above the minimum required pressure for reactor coolant pump operation, component cooling water flow to the reactor coolant pumps is initiated prior to starting the ps.
reparation for the hydrostatic test of the reactor coolant system, the reactor vessel lower and er internals and the closure head are installed. The closure head studs are properly tensioned for hydrostatic test pressure. The pressurizer safety valves and instrumentation within the test ndary are either removed, recalibrated or verified to be able to withstand the hydrostatic test 14.2-26 Revision 1
neral Test Method and Acceptance Criteria ctor coolant system performance is observed and recorded during a series of individual ponent and system tests. The following testing demonstrates that the reactor coolant system can orm the functions described above and in appropriate design specifications:
The integrity and leaktightness of the reactor coolant system and the high-pressure portions of associated systems is verified by performing a cold hydrostatic pressure test in conformance with Section III of the American Society of Mechanical Engineers (ASME)
Code. The reactor coolant system is pressurized in stages by operation of the temporary hydrostatic test pump, while monitoring system welds, piping, and components for leaks at each stage. The hydrostatic test verifies that there are no leaks at welds or piping within the test boundaries during the final inspection. Any identified pressure boundary leaks (i.e. piping walls, vessel walls, welds, valve bodies, etc.) are repaired and the hydrostatic test repeated.
Leakage through valve seats, valve packing, flanges, and threaded or mechanical fittings is acceptable during the hydrostatic test as long as the hydrostatic test pump can maintain the proper test pressure. Leakage through these items may, as necessary and practical, be isolated, repaired, and retested at a later date.
Proper operation of the safety-related reactor coolant system and reactor coolant pressure boundary valves is verified by the performance of baseline in-service tests as described in Subsection 3.9.6.
The operability of the pressurizer safety valves is demonstrated by a bench test at temperature and pressure with steam as the pressurizing fluid or with a suitable in-situ test.
This testing verifies that each pressurizer safety valve actuates at the required set pressure, with appropriate tolerance as specified in the Technical Specifications. The safety valve rated capacity, as recorded on the valve vendor code plates, is verified to be greater than or equal to that described in Section 5.4.
During hot functional testing, reactor coolant system leakage is verified to be within the limits specified in the Technical Specifications. Proper calibration and operation of instrumentation controls, actuations, and interlocks related to reactor coolant system leak detection are verified. The pressurizer water level is set to the no-load level, the chemical and volume control system makeup pumps and letdown line do not operate, and no primary system samples are taken. During this test, the identified and unidentified reactor coolant system leakage rates are determined by monitoring the reactor coolant system water inventory, reactor coolant drain tank level, containment sump level, and other leak detection instrumentation as described in Subsection 5.2.5 over a specified period of time.
The leakage across individual valves between high pressure and low pressure systems, as specified in the Technical Specifications, is verified to be less than design requirements.
The as-installed safety valve discharge chamber rupture disks are inspected to verify the manufacturers stamped set pressure is within the limits specified in the appropriate design specifications.
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z Hot leg and cold leg resistance temperature detectors z Flow instrumentation at selected locations in the reactor coolant loop z Reactor coolant system wide range pressure transmitters z Hot leg level instruments z Pressurizer pressure and level instruments z Reactor coolant pump bearing water temperature detectors z Reactor coolant pump speed sensor instruments z Reactor vessel head vent valve controls This testing includes demonstration of proper actuation of safety-related functions from the main control room.
Automatic trip of the reactor coolant pumps following appropriate safety-related actuation signals is demonstrated.
Proper operation of the reactor vessel head vent valves is verified with the reactor coolant system pressurized.
following testing demonstrates that the system properly performs the defense-in-depth functions cribed above and in appropriate design specifications:
The pressurizer spray valves are verified to operate properly over the range of reactor coolant system operating temperatures and with the reactor coolant pumps operating.
Proper calibration and operation of defense-in-depth related instrumentation, controls, actuation signals and interlocks are verified. This testing includes actuation of the pressurizer spray valves on receipt of appropriate signals, as well as actuation from the main control room.
Reactor coolant pump and motor performance and operating characteristics are initially verified with the reactor coolant system at cold conditions. This testing includes verification of the proper flow through the reactor coolant system when all four reactor coolant pumps are operated in various combinations and speeds as specified in the appropriate design specifications and operating procedures. In addition, the proper operation of the pump motor instrumentation, alarms, and interlocks is verified including:
z Motor current z Motor power z Pump vibration z Motor Stator temperature z Proper transfer from variable speed startup operation The reactor coolant system is heated from cold conditions to hot standby conditions by operating the reactor coolant pumps and the pressurizer heaters. The reactor coolant system is operated at full flow conditions for at least 240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> prior to core loading. The reactor coolant temperature is maintained at or above 515°F for at least one-half of this operating time. In addition to facilitating the reactor coolant system tests that are required to be performed hot and pressurized, these hot functional testing conditions allow the plant operators to control the plant using the plant operating procedures for the reactor coolant system, secondary side systems, and auxiliary systems.
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During hot functional testing, the reactor coolant pump and motor operating characteristics are measured and recorded at various temperature plateaus during reactor coolant system heatup to verify proper operation over their operating temperature range. This testing includes verification of the proper pump flow; proper motor current, power, and stator temperature; and pump vibration level.
The pressurizer spray continuous flow rate is established, and the proper spray line temperature is verified for each pressurizer spray line.
The proper operation of the pressurizer heaters, pressurizer spray, and pressure control functions and alarms is verified during the heatup, operation at hot functional test conditions, and cooldown of the reactor coolant system.
The proper operation of the pressurizer level control functions and alarms is verified during the heatup, operation at hot functional test conditions, and cooldown of the reactor coolant system.
The pressure drops across the major components of the reactor coolant system are measured and recorded using temporary instrumentation during flow testing, and verified to be in accordance with appropriate design specifications.
s associated with the automatic depressurization functions of reactor coolant system ponents are described in Subsection 14.2.9.1.3.
2.9.1.2 Steam Generator System Testing pose purpose of the steam generator system testing is to verify that the as-installed components perly perform the following safety-related functions as described in Sections 5.4, 10.3 and 10.4:
Provide steam generator isolation, including isolation of the main steam lines, feedwater lines, and blowdown lines Remove heat from the reactor coolant system and provide secondary side overpressure protection Measure process parameters required for safety-related actuations as described in Sections 7.2, 7.3, and 7.4 Measure process parameters required for post-accident monitoring as described in Section 7.5 testing also verifies that the as-installed components properly perform the following defense-in-th functions as described in Section 10.4:
Provide heat removal from the reactor coolant system Provide overpressure protection for the steam generators to minimize required actuations of the spring-loaded safety valves Measure process parameters and provide actuation signals for the diverse actuation system 14.2-29 Revision 1
erator system testing is performed during the plant hot functional tests. Prerequisite testing of uired interfacing systems are completed to the extent sufficient to support the specified testing the appropriate system configuration. Construction and installation testing of the special itoring system has been completed to the extent necessary to support preoperational testing.
uired electrical power supplies are energized and operational.
neral Test Method and Acceptance Criteria performance of the steam generator system is observed and recorded during a series of vidual component and integrated system testing that characterizes its modes of operation. The wing testing demonstrates that the steam generator system operates as specified in tions 10.3 and 10.4, and appropriate design specifications:
Proper operation of the steam generator system safety-related valves is verified by the performance of baseline in-service tests as described in Subsection 3.9.6. In addition, the ability of these valves to perform their safety related functions is verified during hot functional testing with the steam generators at normal operating pressure and temperature. The following valves are tested:
z Steam line condensate drain control and isolation valves z Main steam line isolation valves z Main and startup feedwater isolation valves z Steam generator blowdown isolation valves z Steam generator power-operated relief valves z Main steam isolation valve bypass isolation valves z Main and startup feedwater control valves This testing includes verification of the capability of the steam generator power operated relief valves to provide the required heat removal rate from steam generators/reactor coolant system.
Proper operation of safety-related and defense-in-depth instrumentation, controls, actuation signals, and interlocks is verified. This testing includes actuation of equipment from the main control room.
The proper operation of the steam generator safety valves is demonstrated in a bench test at temperature and pressure with steam as the pressurizing fluid or with suitable in-situ testing.
The safety valve rated capacity recorded on the valve vendor code plates is verified to be greater than or equal to the required relief capacity.
t transfer performance of the steam generator system is verified by startup testing of the reactor lant system described in other sections.
2.9.1.3 Passive Core Cooling System Testing pose purpose of the passive core cooling system testing is to verify that the as-installed components their associated piping and valves properly perform the following safety functions, described in tion 6.3:
Emergency core decay heat removal Reactor coolant system emergency makeup and boration 14.2-30 Revision 1
requisites construction testing of the passive core cooling system, or of a specific portion of the system to ested, is successfully completed. The preoperational testing of the reactor coolant system, mal residual heat removal system, chemical and volume control system, the refueling cavity, the ss 1E dc and uninterruptible power supply, the ac electrical power and distribution systems, and r interfacing systems required for operation of the above systems is completed as needed to port the specified testing and system configurations. A source of water, of a quality acceptable for g the passive core cooling system components and the reactor coolant system, is available.
neral Test Method and Acceptance Criteria performance of the passive core cooling system is observed and recorded during a series of vidual component testing and testing with the reactor coolant system. The following testing onstrates that the passive core cooling system operates as described in Section 6.3 and ropriate design specifications.
Proper operation of safety-related valves is verified by the performance of baseline in-service tests as described in Subsection 3.9.6. Also, the proper operation of non-safety-related valves is verified including manual valve locking devices. This testing does not include actuation of the squib valves, which is discussed in Item t, below.
Proper calibration and operation of safety-related instrumentation, controls, actuation signals, and safety related interlocks as specified in Section 7.6, is verified. This testing includes the following:
z Passive residual heat removal heat exchanger flow z Core makeup tank level z In-containment refueling water storage tank level z Containment floodup level z Core makeup tank inlet/outlet valve controls z Passive residual heat removal heat exchanger inlet/outlet valve controls z In-containment refueling water storage tank outlet valve controls z Containment recirculation valve controls z Automatic depressurization valve controls z In-containment refueling water storage tank gutter isolation valve controls This testing includes demonstration of proper actuation of safety-related functions from the main control room.
Proper calibration and operation of instrumentation, controls, and interlocks required to demonstrate readiness of a safety-related component is verified. This testing includes the following:
z Accumulator pressure and level and alarms z Passive residual heat removal heat exchanger temperatures z Passive residual heat removal heat exchanger high point vent level z Core makeup tank inlet line temperatures z Core makeup tank inlet line high point levels z Direct vessel injection line temperatures z In-containment refueling water storage tank level and temperatures 14.2-31 Revision 1
z CMT level z CMT flow and balance line temperatures z PRHR supply line temperatures z Accumulator wide range level z In-containment refueling water storage tank and sump-recirculation flow z ADS piping differential pressure passive core cooling system emergency core decay heat removal function is verified by the wing testing of the passive residual heat removal heat exchanger.
During hot functional testing of the reactor coolant system, the heat exchanger supply and return line piping water temperatures are recorded to verify that natural circulation flow initiates.
The heat transfer capability of the passive residual heat removal heat exchanger is verified by measuring natural circulation flow rate and the heat exchanger inlet and outlet temperatures while the reactor coolant system is cooled to 420°F. This testing is performed during hot functional testing with the reactor coolant system initial temperature 540°F and the reactor coolant pumps not running. The acceptance criteria for the PRHR HX heat transfer under natural circulation conditions are that the heat transfer rate is 1.78 E+08 Btu/hr based on a 520°F hot leg temperature and 1.11 E+08 Btu/hr based on 420°F hot leg temperature with 80°F IRWST temperature and the design number of tubes plugged. These plant conditions are selected to be close to the expected test conditions and are different than those listed in Table 6.3-2. The PRHR HX heat transfer rate has been adjusted to account for these different conditions. The heat transfer rate measured in the test should be adjusted to account for differences in the hot leg and IRWST temperatures and number of tubes plugged.
The proper operation of the passive residual heat removal heat exchanger and its heat transfer capability with forced flow is verified by initiating and operating the heat exchanger with all four reactor coolant pumps running. This testing is performed during hot functional testing with the reactor coolant system at an elevated initial temperature 350°F. The heat exchanger heat transfer is determined by measuring the heat exchanger flow rate and its inlet and outlet temperatures while the reactor coolant system is cooled to 250°F. The acceptance criteria for the PRHR HX heat transfer under forced circulation conditions are listed in Table 3.9-17. The heat transfer rate measured in the test should be adjusted to account for differences in the hot leg and IRWST temperatures and number of tubes plugged.
The heatup characteristics of the in-containment refueling water storage tank water are verified by measuring the vertical water temperature gradient that occurs in the in-containment refueling water storage tank water at the passive residual heat removal heat exchanger tube bundle and at several distances from the tube bundle, during testing in Item e), above. Note that this verification is required only for the first plant. The acceptance criterion for the IRWST heatup characteristics is that they support meeting the RCS safe shutdown temperature criteria (refer to Subsection 19E.4.10.2).
passive core cooling system emergency makeup and boration function is verified by the wing testing of the core makeup tanks.
The resistance of the core makeup tank cold leg balance lines is determined by filling the core makeup tanks with flow from the cold legs. This testing is performed by filling the cold, depressurized reactor coolant system using a constant, measured discharge flow from the 14.2-32 Revision 1
the resistance of the balance lines. The acceptance criterion for the resistance of these lines is 7.21 x 10-6 ft/gpm2.
During hot functional testing of the reactor coolant system, the core makeup tank cold leg balance line piping water temperature at various locations is recorded to verify that the water in this line is sufficiently heated to initiate recirculation flow through the CMTs.
[Proper operation of the core makeup tanks to perform their reactor water makeup and boration function is verified by initiating recirculation flow through the tanks during hot functional testing with the reactor coolant system at 530°F. This testing is initiated by simulating a safety signal which opens the tank discharge isolation valves, and stops reactor coolant pumps after the appropriate time delay. The proper tank recirculation flow after the pumps have coasted down is verified. Based on the cold leg temperature, CMT discharge temperature, and temporary CMT flow instrumentation, the net mass injection rate into the reactor is verified. Note that this verification is required only for the first three plants.]*
passive core cooling system safety injection function is verified by the following testing of the makeup tanks, accumulators, in-containment refueling water storage tank, containment sump, matic depressurization, and their associated piping and valves.
Proper flow resistance of each of the core makeup tank injection lines is verified by gravity draining each tank filled with cold water through the direct vessel injection flow path, while measuring the CMT level (driving head) and discharge flow rate. Air enters the top of the draining tank from the reactor coolant system cold leg via the cold leg balance line. If necessary, the flow limiting orifice in the core makeup tank discharge line is to be resized, and the core makeup tank retested to obtain the required line resistance. The acceptance criteria for the resistance of these lines are 2.25 x 10-5 ft/gpm2 and 1.81 x 10-5 ft/gpm2 with all valves open.
The proper flow resistance of each of the accumulator injection lines is verified by performing a blowdown from a partially pressurized accumulator through the direct vessel injection flow path, while measuring the change in accumulator level and pressure. If necessary, the flow orifice in the accumulator discharge line is to be resized and the accumulator retested to obtain the required discharge line resistance. The acceptance criteria for the resistance of these lines are 1.83 x 10-5 ft/gpm2 and 1.47 x 10-5 ft/gpm2.
The proper flow resistance of each of the in-containment refueling water storage tank injection lines is verified by gravity draining water from the tank through the direct vessel injection flow path, while measuring the water level (driving head) and discharge flow rate using temporary instrumentation. A test fixture with prototypical resistance may be used to simulate the squib valves in the flow paths tested. The acceptance criteria for the resistance of these lines are 9.20 x 10-6 ft/gpm2 and 5.53 x 10-6 ft/gpm2 for line A and 1.03 x 10-5 ft/gpm2 and 6.21 x 10-6 ft/gpm2 for line B with all valves open.
The flow resistance of each of the flow paths from the in-containment refueling water storage tank to each containment sump, and from each containment sump to the reactor is verified by a series of tests. These tests gravity drain water from the in-containment refueling water storage tank to the containment sump, and from the sump through the direct vessel injection flow path, while measuring the storage tank water level (driving head) and injection flow rate using temporary instrumentation. This testing is performed using temporary piping to prevent flooding of the containment. A test fixture with prototypical resistance may be used to Staff approval is required prior to implementing a change in this information.
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the resistance of the lines between the IRWST and each containment sump is 4.07 x 10-6 ft/gpm2.
The resistance of each automatic depressurization stage 1, 2, and 3 flow path and flow path combination is verified by pumping cold water from the in-containment refueling water storage tank into the cold, depressurized, water-filled reactor coolant system; and back to the in-containment refueling water storage tank using the normal residual heat removal pump(s).
The resistances are determined by measuring the residual heat removal pump flow rate and the pressure drop across the flow paths tested using temporary instrumentation. The acceptance criteria for the resistance of these lines is 2.91 x 10-6 ft/gpm2 for each ADS stage 1, 2, 3 group with all valves open.
The resistance of each automatic depressurization stage 4 flow path and their flow path combinations is verified by pumping cold water from the in-containment refueling water storage tank into the cold, depressurized, water-filled reactor coolant system using the normal residual heat removal pump(s). The resistances are determined by measuring the residual heat removal pump flow rate and the pressure drop across the flow paths tested using temporary instrumentation. A test fixture with prototypical resistance may be used to simulate the squib valves in the flow paths tested. The acceptance criteria for the resistance of these lines are 1.70 x 10-7 ft/gpm2 for ADS stage 4 on loop 1 and 1.57 x 10-7 ft/gpm2 for ADS stage 4 on loop 2 with all valves open.
The proper operation of the vacuum breakers in the automatic depressurization discharge lines is verified.
[During hot functional testing of the reactor coolant system, proper operation of automatic depressurization is verified by blowing down the reactor coolant system. This testing verifies proper operation of the stage 1, 2, and 3 components including the ability of the spargers to limit loads imposed on the in-containment refueling water storage tank by the blowdown.
Proper operation of the stage 1, 2 and 3 valves is demonstrated during blowdown conditions.
Note that this verification is required only for the first three plants.]*
The proper operation of at least one of each squib valve size and type including a containment recirculation, in-containment refueling water storage tank injection, and a stage 4 automatic depressurization squib valve is demonstrated. The squib valve performance and the flow resistance of the actuated squib valves is compared to the squib valve qualification testing results. This test does not have to be performed in the plant.
The proper operation of the containment sump instrumentation is demonstrated by simulating the containment flood-up water levels.
The proper operation of the CMT level instrumentation is demonstrated during the draindown testing of the CMTs, specified in Item l) above.
[In conjunction with the verification of the core makeup tanks to perform their reactor water makeup function and boration function described in item k) above, the proper operation of the core makeup tanks to transition from their recirculation mode of operation to their draindown mode of operation after heatup will be verified. This testing will also verify the proper operation of the core makeup tank level instrumentation to operate during draining of the heated tank fluid. The in-containment refueling water storage tank initial level is reduced to at least 3 feet below the spillway level as a prerequisite condition for this testing in order to Staff approval is required prior to implementing a change in this information.
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The recirculation operation in Item k) above, should be continued until the core makeup tank fluid has been heated to 350°F. The core makeup tank isolation valves are then closed, the reactor coolant pumps are started, and the reactor coolant system is reheated up to hot functional testing conditions. This testing is initiated by shutting off the reactor coolant pumps, opening the core makeup tank isolation valves, and by opening one of the automatic depressurization stage 1 flow paths to the in-containment refueling water storage tank. This will initiate a large loss of mass from the reactor coolant system, depressurization of the reactor coolant system to the bulk fluid saturation pressure, and additional recirculation through the core makeup tank. Core makeup tank draindown initiates in response to the continued depressurization and mass loss from the reactor coolant system. The automatic depressurization stage 1 flow path is closed after the core makeup tank level has decreased below the level at which stage 4 actuation occurs. Note that this verification is required only for the first three plants.]*
2.9.1.4 Passive Containment Cooling System Testing pose purpose of the passive containment cooling system testing is to verify that the as-installed ponents perform properly to accomplish their safety-related functions to transfer heat from inside containment to the environment, as described in Subsection 6.2.2. The passive containment ling water storage tank also provides a safety-related source of makeup water for the spent fuel l, and provides a seismically qualified source of water for the fire protection system. Testing of e functions are discussed in Subsections 14.2.9.2.7 Spent Fuel Pool Cooling System Testing, 14.2.9.2.8 Fire Protection System Testing.
requisites construction testing of the passive containment cooling system is successfully completed. The operational testing of the Class 1E dc electrical power and uninterruptible power supply systems, non-Class 1E electrical power supply system, the compressed and instrument air system, and r interfacing systems required for operation of the above systems is available as needed to port the specified testing and system configurations. Additionally, a sufficient quantity of eptable quality water for filling the passive containment cooling water storage tank and draining the containment is available, and a means of filling the tank is available.
neral Test Acceptance Criteria and Methods sive containment cooling system performance is observed and recorded during a series of vidual component testing that characterizes passive containment cooling system operation. The wing testing demonstrates that the passive containment cooling system operates as described in tion 6.2 and appropriate design specifications:
Proper operation of safety-related valves is verified by the performance of baseline in-service tests as described in Subsection 3.9.6.
Proper calibration and operation of safety-related, defense-in-depth, and system readiness instrumentation, controls, actuation signals and interlocks as discussed in Sections 7.3 and 7.5 are verified. This testing includes the following:
z Normal range containment pressure z High range containment pressure z Passive containment cooling water flow rate z Passive containment cooling water storage tank level Staff approval is required prior to implementing a change in this information.
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z Air inlet and shield plate freeze protection heater controls This testing includes demonstration of proper actuation of these functions from the main control room.
Flow testing is performed to demonstrate proper system flow rates by draining the passive containment cooling system water storage tank. This testing demonstrates the proper resistance of the four passive containment cooling water storage tank delivery flow paths.
This testing also demonstrates that water is supplied at the specified flow rates and times for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> consistent with the design basis analyses presented in Subsection 6.2.1.
The proper operation of the passive containment cooling water distribution bucket and weirs is verified and proper wetting of the containment is observed and recorded during draindown testing in Item c, above. Water delivery and coverage is verified at the initial minimum water level and as each of the first two standpipes is uncovered. Water coverage is measured at the spring line and the base of the upper annulus as described in Subsection 6.2.2.4.2.
The proper operation of the drains in the upper containment/shield building annulus to drain the containment cooling water from the annulus floor is verified.
The resistance of the passive containment cooling air flow path is verified by measuring the wind induced driving head developed from the air inlet plenum region of the shield building to the air exhaust at several locations along the flow path and at several circumferential locations, and measurement of the induced air flow velocity. Temporary instrumentation is used for this testing.
Sample coupons from the containment shell with and without an appropriate coating of paint are laboratory tested to determine their conductivity.
The proper operation of each of the PCS water storage tank recirculation/makeup pumps to makeup sufficient water to the PCS water storage tank from the PCS ancillary water storage tank is verified.
2.9.1.5 Chemical and Volume Control System Isolation Testing pose purpose of the chemical and volume control system isolation testing is to verify that the nstalled components properly perform the following safety-related isolation functions, described ection 9.3:
Termination of inadvertent dilution of the reactor coolant boron concentration Isolation of unborated water sources for reactor makeup Reactor coolant system pressure boundary isolation Isolation/termination of excessive makeup to the reactor requisites construction testing of the chemical and volume control system has been successfully pleted. The required preoperational testing of appropriate support and interfacing systems is pleted. Data collection is available as needed to support the specified testing and system figurations.
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ation modes of operation. The following testing demonstrates that the chemical and volume trol system properly performs the safety-related isolations as specified in Section 9.3 and ropriate design specifications:
Proper operation of the safety-related valves is verified by the performance of baseline in-service tests as described in Subsection 3.9.6, including:
z Purification loop isolation valves z Letdown isolation valves z Demineralized water isolation valves z Makeup isolation valves z Auxiliary spray isolation valve Proper calibration and operation of safety-related instrumentation, controls, actuation signals and interlocks is verified. This testing includes the following:
z Purification isolation valve controls z Letdown isolation valve controls z Demineralized water isolation controls z Makeup isolation valve controls This testing includes demonstration of proper actuation of safety-related functions from the main control room.
2.9.1.6 Main Control Room Emergency Habitability System Testing pose purpose of the main control room emergency habitability system testing is to verify that the nstalled components properly perform the safety-related functions described in Section 6.4, uding the following:
Provide sufficient breathable quality air to the main control room Maintain the main control room at positive pressure Provide passive cooling of designated equipment ddition, the following safety-related functions performed by the nuclear island nonradioactive tilation system described in Subsection 9.4.1 are tested:
Provide isolation of the main control room from the surrounding areas and outside environment during a design basis accident if the nuclear island nonradioactive ventilation system becomes inoperable.
Monitor the radioactivity in the main control room normal air supply and provide signals to isolate the incoming air and actuate the main control room emergency habitability system.
ddition, the following safety-related functions performed by the potable water system, described ubsection 9.2.5; the sanitary drainage system, described in Subsection 9.2.6; and the waste er system, described in Subsection 9.2.9, are tested:
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requisites construction testing of the main control room habitability system has been successfully pleted. The required preoperational testing of the compressed and instrument air system, ss 1E electrical power and uninterruptible power supply systems, normal control room ventilation em, and other interfacing systems required for operation of the above systems is available as ded to support the specified testing and system configurations. The main control room air supply s are filled with air acceptable for breathing. The main control room construction is complete and eak-tight barriers are in place.
neral Test Acceptance Criteria and Methods ormance of the main control room habitability system is observed and recorded during a series of vidual component and integrated system testing. The following testing demonstrates that the itability system operates as specified in Section 6.4 and as specified in the appropriate design cifications:
Proper operation of safety-related valves is verified by the performance of baseline in-service tests as described in Subsection 3.9.6.
Proper calibration and operation of safety-related and system readiness instrumentation, controls, actuation signals and interlocks is verified. This testing includes the following:
z Air storage tank pressure z Refill line connection pressure z Main control room differential pressure z Air supply line flow rate z Controls for the main control room pressure relief valves z Controls for the air supply isolation valves z Controls for the main control room air inlet isolation valves z Air intake radiation z Passive filtration line flow rate z Filter performance z Sanitary drainage system vent isolation valves The proper flow rate of emergency air to the main control room is verified, demonstrating proper sizing of each air flow limiting orifice, proper operation of each air supply pressure regulator, and the ability to maintain proper control room air quality. The MCR passive filtration system flow rate and filter performance will also be verified at this time to ensure a filtration flow rate of at least 600 cfm. This testing demonstrates the control room pollutant concentrations during the first 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of operation. To determine the control room air quality at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the CO2 concentrations can be predicted based on calculations. The other pollutants described in Table 1 and Appendix C, Table 1 of ASHRAE Standard 62-1989 can be predicted by extrapolating their concentrations for the entire 72-hour period.
The ability of the emergency air supply to maintain the main control room at the proper positive pressure is demonstrated, verifying proper operation of the main control room pressure relief dampers.
The ability of the emergency air supply to limit air inleakage to the main control room is verified by inleakage testing as specified in Subsection 6.4.5.4.
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0 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> with the actual heat loads from the battery powered equipment and personnel specified for this time period (for the MCR [room 12401], there is automatic deenergization of specific non-safety MCR heat loads). The control room temperature versus time versus heat load data are used to verify the analysis basis used to assure that the control room conditions remain within specified limits for the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time period. Periodic grab samples will be taken of the control room air environment to support analyses to confirm that specified limits would not be exceeded for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
The ability to maintain temperatures in the protection and safety monitoring system cabinet and emergency switchgear rooms within specified limits for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (Reference Subsection 6.4.3.2) is verified with a test simulating a loss of the nuclear island nonradioactive ventilation system. This testing demonstrates the room heatup from 0 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> with the actual heat loads from battery powered equipment. The room temperature versus time versus heat load data are used to verify the analysis basis used to assure that the room temperature will not exceed the specified limit for the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time period.
2.9.1.7 Expansion, Vibration and Dynamic Effects Testing pose purpose of the expansion, vibration and dynamic effects testing is to verify that the safety-ted, high energy piping and components are properly installed and supported such that expected ement due to thermal expansion during normal heatup and cooldown, and as a result of sients; thermal stratification and thermal cycling; as well as vibrations caused by steady-state or amic effects do not result in excessive stress or fatigue to safety-related plant systems and ipment, as described in Section 3.9.
requisites construction testing and preoperational testing of the reactor coolant system at cold conditions been successfully completed. Required portions of the chemical and volume control system, sive core cooling system, normal residual heat removal system, main feedwater system, startup water system, steam generator system, and steam generator blowdown system are operational.
ng and components within the reactor coolant system and steam generator system pressure ndaries and their associated supports and restraints have been inspected and determined to be alled as designed. Permanently installed support devices have been verified to be in their ected cold, static positions and temporary restraining devices such as hanger locking pins have n removed. The instrumentation required for this testing is installed.
neral Test Method and Acceptance Criteria ing hot functional testing, verifications that ASME Code Class 1, 2, and 3 high-energy piping em components, piping, support, and restraint deflections are unobstructed and within design is functional requirements. The systems to be monitored during preoperational vibration and amic effects tests include:
ASME Code, Class 1, 2, and 3 piping High-energy piping systems inside seismic Category I structures High-energy portions of systems whose failure could reduce the functioning of seismic Category I plant features to an unacceptable safety level 14.2-39 Revision 1
high-temperature portions of the following systems are considered for inclusion in this test:
Reactor coolant system Chemical and volume control system Passive core cooling system Steam generator system (including the safety-related portions of main steam system, main and startup feedwater systems, and steam generator blowdown system)
Normal residual heat removal system Thermal expansion testing during the preoperational testing phase consists of displacement measurements on the above systems during heatup and cooldown of the reactor coolant system and associated systems (including heatup and cooldown of the passive core cooling system). The testing is performed in accordance with ASME OM Standard, Part 7 as discussed in Subsection 3.9.2.1.2 and consists of a combination of visual inspections and local and remote displacement measurements. This testing includes the inspection and measurement of deflection data associated with support thermal movements to verify support swing clearance at specified heatup and cooldown intervals; that there is no evidence of blocking of the thermal expansion of any piping or components, other than by installed supports, restraints, and hangers; that spring hanger movements remain within the hot and cold setpoints; that moveable supports do not become fully retracted or extended; and that piping and components return to their approximate baseline cold positions.
Vibration testing is performed on safety-related and high-energy system piping and components during both cold and hot conditions to demonstrate that steady-state vibrations are within acceptable limits. See Subsection 3.9.2.1.1 for the acceptable standard for alternating stress intensity due to steady-state vibration. This testing includes visual observation and local and remote monitoring in critical steady-state operating modes. Results are acceptable when visual observations show no signs of excessive vibration and when measured vibration amplitudes are within acceptable levels.
Testing for significant vibrations caused by dynamic effects is conducted during hot functional testing and may be performed as part of other specified preoperational tests. This testing is conducted to verify that stress analyses of safety-related and high-energy system piping under transient conditions are acceptable. See Subsection 3.9.2.1.1 for the acceptable standard for alternating stress intensity due to dynamic effects vibration. These tests are performed to verify that the dynamic effects caused by transients such as pump starts and stops, valve stroking, and significant process flow changes are within expected values. These tests include anticipated normal operating evolutions with system differential temperatures, such as startup, which could induce dynamic effects. Suitable instrumentation is used to monitor for the occurrence of water hammer noise and vibration. Visual inspections are performed to confirm the integrity of system piping and supports.
Deflection measurements during various plant transients are recorded and compared to acceptance limits and it is confirmed that no effects due to water hammer are detected.
As described in Subsection 3.9.3, temperature sensors are installed on the pressurizer surge line and pressurizer spray line for monitoring thermal stratification and thermal cycling during 14.2-40 Revision 1
main control room habitability system is classified as a high energy system based on the sure criteria not temperature. Tests that measure thermal movements are not required. Vibration ing of the high pressure portion of the main control room habitability system is performed during ing of the air delivery rate provided to the control room. See Subsection 14.2.9.1.6 for information he testing of the main control room habitability system.
2.9.1.8 Control Rod Drive System pose purpose of the control rod drive system testing is to verify the proper operation of the control rod e mechanisms, motor-generator sets and system components as described in Subsection 3.9.4 Section 4.6, and in appropriate design specifications.
requisites construction tests of the control rod system have been completed. Required interfacing systems, eeded, are completed to the extent sufficient to support the specified testing and the appropriate em configuration. Required electrical power supplies are energized and operational.
the control rod drive mechanism cooling test, the plant is at or near normal operating temperature pressure, and post-core hot functional testing is in progress. The integrated head and control rod e mechanism cooling system are in their normal operational alignment.
the control rod drive mechanism motor-generator sets tests, a three-phase load bank is available motor generator set testing under loaded conditions.
neral Test Methods and Acceptance Criteria ormance is observed and recorded during a series of individual component and integrated em tests. The following tests verify that the control rod drive system operates properly:
Tests are conducted to verify the current command sequence, timing, and rod speed signal voltages by initiating control rod drive mechanism withdrawal and insertion. Proper operation of the bank overlap unit to control rod bank sequence and movement is verified.
Tests are conducted to verify the adequacy of the integrated head and control rod drive mechanism cooling system for maintaining control rod drive mechanism temperature. This test is conducted by measuring control rod drive mechanism coil resistances and calculating the coil temperatures.
Tests are conducted to verify control rod drive mechanism motor-generator set and system component control circuits, including interlock and alarm functions.
Tests are conducted to verify generator phasing for parallel generator operation. Operation of the control rod drive mechanism motor generator sets and control system during starting, running, and parallel operations is verified.
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AP1000 reactor internals testing is part of a comprehensive vibration assessment program ormed in accordance with Regulatory Guide 1.20 as discussed in Subsection 3.9.2.4. This ing obtains data to verify the structural integrity of the AP1000 reactor internals with regard to
-induced vibrations, as part of an internals vibration assessment program. This program also udes visual examination of the reactor internals after testing is completed, and analysis of the test
. Testing is performed for the first plant only.
000 plants subsequent to the first plant are visually inspected before and after the hot functional to confirm that the internals are functioning correctly. The major features of the reactor internals ined in Subsection 3.9.2.4 are visually inspected for signs of abnormal wear and structural nges.
requisites construction testing of the reactor coolant system has been completed. The testing and bration of the required test instrumentation has been completed. The test instrumentation has n installed on the internals as specified in Table 3.9-4 and the internals pre-test visual inspection been completed. The internals, test instrumentation, and instrumentation lead wires are installed e reactor vessel. The reactor vessel head is installed in preparation for the cold hydrostatic test e reactor coolant system and instrument leads have been properly sealed. The proper operation calibration of the test instrumentation and recording equipment is verified during the hydrostatic ing of the reactor coolant system.
neral Test Method and Acceptance Criteria ctor vessel internals testing is performed for the first plant only by measuring and recording ins or accelerations of components in order to determine actual displacements that occur with the tor coolant pumps operating. This testing is performed at several reactor coolant system peratures during the system hot functional test. The analysis of data obtained from this testing, bined with a pre-test and post-test visual inspection of the internals, are intended to confirm that stresses and wear on the AP1000 internals, due to flow induced vibration during plant operation, acceptably low. The criteria for evaluating testing results are established in the AP1000 reactor rnals flow-induced vibration assessment program (see Section 7 of WCAP-15949), and ropriate design specifications.
the first plant only, the internals are instrumented to obtain data during the following reactor lant system operating conditions:
Background noise in the instrumentation and recording equipment is recorded with no reactor coolant pumps running Data is recorded during the initial startup of the reactor coolant pumps and with all four pumps operating and with the reactor coolant at cold temperature Data is recorded at several increasing coolant temperatures with the pumps operating Data is recorded at the hot functional testing temperature with all four pumps operating Data is recorded at the hot functional testing temperature with the appropriate combinations of reactor coolant pumps operating, including pump start and stop transients 14.2-42 Revision 1
cturally adequate and sound for operation. If such indications are detected, further evaluation is uired.
2.9.1.10 Containment Isolation and Leak Rate Testing pose purpose of the containment isolation and leak rate testing is to demonstrate that the as-installed tainment isolation valves, piping and electrical containment penetrations, and hatches, and the tainment vessel properly perform the following safety functions as described in Section 6.2:
Automatic isolation of the piping penetrating containment required to assure containment integrity The containment vessel, penetration, and isolation valve leakage is less than the design basis leakage at or near the containment design pressure consistent with 10 CFR 50, Appendix J pressure test requirements.
requisites construction testing of the containment, containment hatches/airlocks and containment etrations including the containment pressure test as specified in Subsection 3.8.2.7 has been pleted. The construction testing of the piping and isolation valves or electrical wiring through the etrations, has been completed. The instrumentation to be used in performing the Type A, B, C testing is calibrated and available, including their associated data processing equipment. The uired preoperational testing of the protection and safety monitoring system, plant control system, Class 1E electrical power uninterruptible power supply, and other interfacing systems required for ration of the containment isolation devices and data collection is available.
neral Test Acceptance Criteria and Methods tainment isolation functions, leak rate, and structural integrity performance are observed and rded during a series of individual component and integrated system testing. The following testing onstrates that the containment functions as described in Section 6.2 and the appropriate design cifications are achieved. The testing is in accordance with the Containment System Leakage ing Program and is discussed in Subsection 6.2.6, which meets the requirements of SI/ANS-56.8-1994, as appropriate.
Proper operation of safety-related containment isolation valves, listed in Table 6.2.3-1, is verified by the performance of baseline in-service tests as specified in Subsection 3.9.6.
Proper calibration and operation of safety-related containment isolation instrumentation, controls, actuation signals and interlocks is verified. This testing includes actuation of the containment isolation valves from the main control room, and upon receipt of a containment isolation signal.
The appropriate Type C leakage testing is performed for each piping path penetrating the containment boundary, verifying the leakage for each containment isolation valve (listed in Table 6.2.3-1) or set of isolation valves. This testing for individual isolation valves may be performed in conjunction with the associated system test.
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A baseline in-service test/inspection of the accessible interior and exterior surfaces of the containment structure and components is performed as specified in Subsection 3.8.2.
A Type A integrated leak rate test is performed to verify that the actual containment leak rate does not exceed the design basis leak rate specified in the Technical Specifications.
2.9.1.11 Containment Hydrogen Control System Testing pose purpose of the containment hydrogen control system testing is to verify that the system properly orms the following safety-related and non-safety defense-in-depth functions described in tion 6.2:
Prevent the concentration of hydrogen in containment from reaching the flammability limit.
Prevent the concentration of hydrogen in containment from reaching the detonation limit.
Monitor the containment hydrogen concentration as required by Regulatory Guide 1.97.
requisites construction testing of the containment hydrogen control system is completed. The Class 1E lectrical power and uninterruptible power supply systems, the non-Class 1E electrical supply em, and other interfacing systems required for operation of the above systems and calibrated collection instrumentation are available as needed to support the specified testing.
neral Test Acceptance Criteria and Methods ormance of the containment hydrogen control system is observed and recorded during a series dividual component testing. The following testing verifies that the system operates as described ubsection 6.2.4 and as specified in the appropriate design specifications:
Proper operation of both the Class 1E safety-related and non Class 1E containment hydrogen concentration instrumentation and alarms is verified.
The ability of the passive autocatalytic recombiners to properly respond to a known inlet hydrogen/air mixture is verified by removing and testing one plate or cartridge from each manufacturing lot of catalyst material, contained in each recombiner unit. This verification is performed in accordance with the guidance provided in Subsection 6.2.4.5.1 using a manufacturers standard test device and test procedure. Plate performance is verified to be consistent with the response obtained in manufacturers tests.
Manual actuation and operation of the hydrogen igniters confirm that the igniters are supplied by two power groups from two subsystems of the non-Class 1E dc and UPS system.
Operability of the igniters is confirmed by verification that the igniter surface temperature exceeds the temperature specified in Subsection 6.2.4.
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purpose of the protection and safety monitoring system preoperational testing is to verify that the nstalled components properly perform the following safety-related functions, described in tion 7.1:
Receive and analyze sensor inputs required for reactor trip and automatically initiate reactor trip signals when plant conditions reach the appropriate setpoints Provide actuation signals to the engineered safety features to limit the consequences of design basis accidents Provide instrumentation and display systems to monitor the safety-related functions of the plant during and following the occurrence of design basis accidents in accordance with Regulatory Guide 1.97 operational testing is also performed to verify proper operation of the following defense-in-depth tions, described in Section 7.1:
Provide data from the safety-related sensors to the plant control system Provide information to the data display and processing system Provide data to the monitor bus for use by other systems within the plant requisites struction and installation testing of the protection and safety monitoring system cabinets has n completed. Related system interfaces are available or simulated as necessary to support the cified test configurations. Component testing and instrument calibrations have been completed.
gramming has been completed and the initial software diagnostics tests have been completed.
uired electrical power supplies and control circuits are energized and operational. Plant systems omponents which are to be operated during testing are specifically identified in the preoperational procedures, are properly aligned, and have proper support systems operating prior to actuation e particular system or component. Equipment or components which can not be actuated without age or upsetting the plant are isolated using the test switches provided by the Protection and ety Monitoring System to block device actuation. Continuity of wiring up to the actuation ipment is verified.
neral Test Methods and Acceptance Criteria ormance of the protection and safety monitoring system is observed and recorded during a es of individual component and integrated tests designed to verify operation of the system ponents. The following testing verifies that the system operates as described in Section 7.1 and ropriate design specifications:
Processing of the analog and digital signals is verified by injecting reference signals and verifying the outputs at various locations in the system.
Capability to process sensor data and main control room manual inputs resulting in the initiation of appropriate reactor trip signals is demonstrated by simulating inputs for each of the trip functions. Response times are verified by demonstrating that the applicable trip, actuate, permissive or interlock signal reaches the actuated equipment within the maximum allowable period following a defined step change in the applicable simulated input, above or below the trip, actuate, permissive or interlock setpoint. Operation of the protection cabinet trip/normal/bypass switches and indicators for each of the reactor trip functions is 14.2-45 Revision 1
nonsafety function of the Plant Control System (PLS) will be verified.
Operation of the reactor trip breakers, including breaker interlock, alarm, and tripping functions and verification that reactor trip response times are less than the specified maximum allowable response times is performed by initiating a manual reactor trip from the main control room. The capability of the undervoltage coil and the shunt trip coil functions to independently trip the reactor trip breakers is verified during this test using the test capabilities provided by the reactor trip switchgear interface.
The capability to trip the reactor from the remote shutdown workstation is demonstrated by verifying actuation of the reactor trip breaker undervoltage and shunt trip attachments upon initiation of a reactor trip at the remote shutdown workstation location.
The capability of the protection and safety monitoring system to process sensor data and manual inputs, resulting in appropriate engineered safety features actuation at design setpoints, is demonstrated by verifying that injection of simulated inputs for each of the engineered safety features actuation functions results in the proper output as indicated by contact operation, component actuation, or electrical test. Response times associated with the engineered safety features actuation functions are evaluated during these tests to provide verification that the applicable trip, actuate, permissive or interlock signal reaches the actuated equipment within the maximum allowable period following a defined step change in the applicable simulated input above or below the trip, actuate, permissive or interlock setpoint. Operation of the manual actuation/bypass switches and indicators for each of the engineered safety features functions is verified by demonstrating appropriate system outputs.
Verification that the engineered safety features bypass logic satisfies the single failure criteria is demonstrated by operating the bypass switches while simulating channel failures. Correct input processing and calculational accuracy of the redundant actuation equipment and operator interface features is verified for each defined engineered safety features actuation function using simulated inputs. Proper operation of the engineered safety features reset functions will be verified.
Correct processing of inputs by redundant equipment and operation of the processing, permissive, interlock, display and operator interface features is verified by demonstrating that simulated command inputs result in correct output or actuation functions as indicated by contact operation, component actuation, or electrical test.
Accurate processing of component-level manual actuation commands from the main control room to the protection logic cabinets is verified by simulating main control room commands.
Processing of component status information is demonstrated by simulating protection logic cabinet outputs to the main control room.
Processing of component-level actuation commands from the remote shutdown workstation to the protection logic cabinets is verified by simulating remote shutdown workstation commands. Processing of component status information is verified by simulating protection logic cabinet outputs to the remote shutdown workstation.
Operation of the automatic testing features provided in the protection and safety monitoring system is verified by observing the automatic test functions while simulating component failures and utilizing man-machine interface capabilities to evaluate system performance.
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representing feedback from actuation devices and position indicators. Communication of information via the plant monitor bus/data display and processing system, such as channel input quality, neutron flux detector high voltage, partial trip/actuation, permissive, interlock, block, reset, bypass, automatic test, reactor trip switchgear and system level actuation status, from the protection and safety monitoring system to external systems is verified by evaluating system response to injected reference signals and operating applicable block and bypass controls.
Operation of the qualified data processing equipment is verified by monitoring outputs and qualified display indications generated in response to simulated inputs representing data from the integrated protection cabinets and sensor inputs to the qualified data processing I/O cabinets.
Operation of the isolated data links and data highways used for communication between the engineered safety features actuation cabinets, main control room multiplexer cabinets, remote workstation multiplexer cabinets and protection logic cabinets is verified.
Preoperational testing of plant sensors used to provide data related to plant equipment monitored by the protection and safety monitoring system is performed in conjunction with testing of the respective systems in which these sensors are located.
The capability of the protection and safety monitoring system to provide data from the safety-related sensors to the plant control system is verified by injecting reference signals into the integrated protection cabinets and monitoring the plant control system signal selector outputs.
2.9.1.13 Incore Instrumentation System Testing pose purpose of the incore instrumentation system preoperational testing is to verify that the nstalled components properly perform the following safety-related functions, described in tion 7.1:
Provide reactor coolant system pressure boundary integrity for the incore instrumentation thimble assemblies which penetrate the upper head of the reactor vessel Provide the protection and safety monitoring system with the core exit temperature signals required for post-accident monitoring ing is also performed to verify the following nonsafety-related defense-in-depth functions, cribed in Subsection 4.4.6:
Provide core exit temperature signals to the diverse actuation system dedicated display in the main control room requisites ated system interfaces are available or simulated as necessary to support the specified test figurations. Component testing and instrument calibrations have been completed. Required trical power supplies are energized and operational.
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ide the reactor vessel. The following testing verifies that the system operates as described in tion 7.1 and the appropriate design specifications:
Reactor coolant system pressure boundary integrity at the incore instrumentation reactor vessel head penetrations is verified during hydrostatic testing of the reactor coolant system.
Processing of the incore thermocouple signals is verified by thermocouple signals at the incore instrumentation thimble assembly connectors and verifying the thermocouple signal paths.
2.9.1.14 Class 1E DC Power and Uninterruptible Power Supply Testing pose purpose of the Class 1E dc power and uninterruptible power supply testing is to verify that the nstalled components properly perform the following safety-related functions described in tion 8.3:
Provide the electrical power required for the operation of the plant safety-related equipment, equipment controls, and instrumentation Provide the required safety-related electrical power for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following a design basis event, independent of both offsite and onsite ac electrical power supplies Provide separation and independence of Class 1E power divisions from other Class 1E divisions and non-Class 1E systems ing is also performed to verify proper operation of the following defense-in-depth functions cribed in Subsection 8.3.2:
The capability to recharge the batteries from the onsite or offsite ac electrical sources is verified so that safety-related functions can be supported for an indefinite time requisites construction testing of the Class 1E dc power and interruptible power supply components has n completed. The necessary permanently installed and test instrumentation is calibrated and rational. The 480V ac electrical power system is in operation and supplying power to the battery rgers and regulating transformers. A test load is available for the performance of battery capacity s.
neral Test Methods and Acceptance Criteria ormance of the Class 1E dc power and interruptible power supply is observed and recorded ng a series of individual component and integrated system tests that characterize the operation of system. The following testing verifies that this system operates as described in Section 8.3 and ropriate design specifications:
The capability of each of the seven Class 1E batteries to provide the required momentary and continuous load is verified by a battery service test performed in accordance with IEEE Standard 450. Following this discharge testing, the voltage of each cell is verified to be greater than or equal to the specified minimum cell voltage.
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greater than or equal to the specified minimum cell voltage.
The capability of each of the seven battery chargers to charge its associated battery at the required rate is verified. This testing includes verification that the individual voltage of each cell is within the specified limits for a charged battery.
The capability of each of the six inverters to provide the required output current, frequency, and voltage is verified.
The capability of each of the four regulating transformers to provide the proper ac current to the Class 1E ac distribution panels is verified.
The capability of each of the static transfer switches to automatically transfer the electrical loads supplied by each inverter to its associated regulating transformer is verified.
The separation and independence of each redundant division of the Class 1E dc power and interruptible power supply is verified by successively powering only one division at a time and verifying power to the proper loads and the absence of voltage at the bus and loads not under test.
The proper calibration and operation of instrumentation and alarms, electrical ground detection, and permissive and prohibitive interlocks is verified.
2.9.1.15 Fuel Handling and Reactor Component Servicing Equipment Test pose erify proper operation of the fuel-handling and reactor component servicing equipment as cribed in Section 9.1. This includes the refueling machine, fuel handling machine, fuel transfer em, and refueling tools used to lift, transport, or otherwise manipulate fuel, control rods and other re instruments.
requisites construction tests have been completed. Prerequisites of the required interfacing systems are pleted to the extent sufficient to support the specified testing. Required electrical power supplies energized and operational. Compressed air, as required for tool operation, is available. The tor vessel head has been removed, the reactor vessel and refueling cavity are drained, the eling cavity gate is open, and the area in which the refueling machine moves is free of structures omponents that could interfere with fuel handling operations.
spent fuel pool and fuel transfer canal are drained, and the area in which the fuel handling hine moves is free of any structures or components that interfere with design fuel handling rations.
fuel transfer system is operable and capable of transporting a dummy fuel assembly from the nt fuel pool to containment. A dummy fuel assembly, resembling an actual fuel assembly in ght, envelope, and mating hardware, is available for use. The fuel transfer system and new fuel ator are operable as required to permit testing of fuel handling machine functions.
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The refueling machine is operated to simulate actual refueling operations, using a dummy fuel assembly. This testing includes manual and automatic modes of operation, displays, interlocks, and limits. These tests verify:
z The ability to move a fuel assembly from the fuel transfer system to the reactor vessel and back z The consistency of measured trolley, bridge, and hoist speeds with each mode of operation z The operability of interlocks limiting motion, speed, and weight, including interlocks with other plant equipment z The operability of displays indicating position, mode, alarm status, and load z The adequacy of indexing (by placing the dummy fuel assembly in selected core locations)
A known weight or a calibrated spring scale is used to calibrate and set the load limits for the refueling machine load cells. A static load test or the manufacturers test results are used to verify the ability of the refueling machine hoists to support 125 percent of their rated loads.
following tests are performed to verify the operation of the fuel handling machine:
The fuel handling machine is operated to simulate actual refueling operations, using a dummy fuel assembly. These tests verify:
z The ability to transfer fuel assemblies between the new fuel elevator, fuel transfer system, fuel storage racks, and other areas of the pool where fuel is serviced or stored z The consistency of measured trolley, bridge, and hoist speeds with each mode of operation z The operability of interlocks limiting motion, speed, and weight, including interlocks with other plant equipment z The operability of displays indicating position, mode, alarm status, and load The fuel handling machine is operated to verify its capability to transfer fuel between the new fuel elevator, fuel transfer system, fuel storage racks, and other areas of the pool where fuel is serviced or stored.
A known weight or a calibrated spring scale is used to calibrate and set the load limits for the fuel handling machine load cells. A static load test or the manufacturers test results are used to verify the ability of the refueling machine hoists to support 125 percent of their rated loads.
following tests are performed to verify the proper operation of the fuel transfer system and eling tools:
Using appropriate plant operating procedures, the operability of the new fuel elevator is verified. Testing is performed to demonstrate the proper operation of controls, displays, and 14.2-50 Revision 1
Using appropriate plant operating procedures, the fuel transfer system is operated to simulate actual refueling operations, using a dummy fuel assembly. During these operations, the following items are verified:
z The ability to move fuel assemblies between the fuel building and containment, including proper operation of upenders in both locations z The operability and setpoints of limit switches and of interlocks between stations and with other plant equipment z The operability of displays indicating mode of operation and status Tests are performed to verify that the refueling tools operate properly. Included are tools for handling new fuel assemblies, fuel assembly inserts, irradiation specimens, control rod drive shafts, as well as tools for such operations as control rod drive shaft latching and reactor vessel stud tensioning. As applicable, power is applied to each tool to verify proper operation of controls, limit switches, actuators, and indicators. Stud tensioning equipment is checked when assembling the reactor for hot functional testing. The new fuel handling tool is tested with the dummy fuel assembly during the test of the new fuel elevator.
2.9.1.16 Long-Term Safety-Related System Support Testing pose purpose of this testing is verify the capability to perform the following functions for maintaining extended operation of the safety-related systems and components as described in Section 1.9:
Supply makeup water to the passive containment cooling system.
Supply makeup water to the spent fuel pool.
Provide electrical power for post-accident instrumentation, control room lighting and ventilation, division B and C I&C room ventilation, passive containment cooling system pumps, ancillary generator room lights, ancillary generator tank heaters.
Provide ventilation cooling to the main control room.
Provide ventilation cooling to the Class 1E cabinets for post-accident instrumentation.
requisites construction tests of the safety-related systems and/or components designed for long-term ons have been successfully completed. The preoperational testing of these systems and/or ponents, including instrument calibrations, has been completed as required for the specified ing, system configurations, and operations. Equipment required for data collection is available operable. Water used in this testing should be of a quality suitable for filling the specified ponents. Equipment used to provide the required long-term actions is available.
neral Test Method and Acceptance Criteria ability to perform the required long-term actions is observed and recorded during a series of vidual component and integrated system testing. The following testing verifies that the long-term 14.2-51 Revision 1
The ability to provide makeup water to the passive containment cooling water storage tank as described in Subsection 6.2.2 is verified.
The ability to provide electrical power to the post-accident monitoring instrumentation, control room lighting and ventilation, division B and C I&C room ventilation, passive containment cooling system pumps, ancillary generator room lights, ancillary generator tank heaters, using the ancillary diesel generators as described in Section 8.3 is verified.
The ability to provide main control room ventilation cooling using ancillary fans as described in Subsection 9.4.1 is verified.
The ability to provide ventilation cooling to post-accident monitoring instrumentation equipment rooms using ancillary fans as described in Subsection 9.4.1 is verified.
The ability to provide makeup water to the spent fuel pool via the safety-related makeup connection from the passive containment cooling system water storage tank, as described in Subsection 9.1.3, is verified.
2.9.2 Preoperational Testing of Defense-in-Depth Systems 2.9.2.1 Main Steam System Testing pose purpose of the main steam system testing is to verify that the as-installed system properly orms the following defense-in-depth function, as described in Section 10.3 and appropriate ign specifications:
Provide backup isolation of the steam lines to prevent blowdown of steam from the steam generators following an event where steam line isolation is required requisites construction tests of the as-installed main steam system have been completed. Prerequisites of required interfacing systems are completed to the extent sufficient to support the specified testing the appropriate system configuration.
neral Test Method and Acceptance Criteria n steam system performance is observed and recorded during a series of individual component integrated system testing. The following testing demonstrates that the system operates as cribed in Section 10.4 and appropriate design specifications:
per operation of the following system valves is verified.
Turbine steam stop valves Turbine bypass valves Auxiliary steam system supply header isolation valve Main steam moisture separator reheater 2nd stage steam isolation valve Extraction steam isolation and non-return valves testing includes actuation of these valves from the main control room. The ability of these valves olate steam flow is verified during hot functional testing.
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purpose of the main and startup feedwater system testing is to verify that the as-installed system perly performs the following nonsafety-related defense-in-depth function, as described in sections 10.4.7 and 10.4.9:
Provide startup feedwater to the steam generators to remove heat from the reactor coolant system following the loss of normal feedwater requisites construction tests have been completed. The component testing of the main and startup water system components and instruments, or specific portion to be tested has been completed.
uired interfacing systems are available.
neral Test Method and Acceptance Criteria main and startup feedwater system performance is observed and recorded during a series of vidual component and integrated system testing. The following defense-in-depth testing onstrates that the system operates as described in Subsections 10.4.7 and 10.4.9 and ropriate design specifications:
Proper operation of defense-in-depth instrumentation, controls, actuation signals and interlocks is verified. This testing includes actuation of startup feedwater pumps and remotely-operated valves from the main control room including isolation of the main feedwater system.
The capability of the startup feedwater pumps to operate properly when performing their defense-in-depth function and main feedwater pumps are verified with the steam generator at normal operating pressure.
The capability of the startup feedwater pumps to operate properly with miniflow to the condensate storage tank is verified.
The capability to restore normal steam generator water level from the low narrow range water level, without causing unacceptable feedwater or steam generator water hammer, is demonstrated (refer to Subsections 14.2.9.1.7 and 14.2.10.4.18).
2.9.2.3 Chemical and Volume Control System Testing pose purpose of the chemical and volume control system testing is to verify that the as-installed em properly performs the following defense-in-depth functions described in Subsection 9.3.6 and ropriate design specifications:
Provide makeup water to the reactor coolant system Provide boration of the reactor coolant system Provide auxiliary pressurizer spray requisites construction testing of the as-installed chemical and volume control system is completed. The wing interfacing and support systems are available as necessary to support testing: component ling water system; service water system; reactor coolant system; electrical power and distribution 14.2-53 Revision 1
neral Test Acceptance Criteria and Methods mical and volume control system performance is observed and recorded during a series of vidual component and integrated system testing. The following testing verifies the system perly performs the defense-in-depth functions described in Subsection 9.3.6 and appropriate ign specifications:
Operation of pumps and valves which perform defense-in-depth functions is verified, including:
z Makeup pumps z Boric acid mixing control valve z Makeup flow control valve Calibration and operation of defense-in-depth related instrumentation, controls, actuation signals and interlocks is verified, including:
z Automatic makeup pump actuation and shutoff z Automatic alignment of the boric acid tank z Pressurizer auxiliary spray initiation and termination z Letdown/purification isolation This testing includes actuation of defense-in-depth pumps and remotely-operated valves from the main control room. Pressurizer level control testing is described in Subsection 14.2.9.1.1.
The capability of the makeup pumps to operate when performing their normal makeup and pressurizer spray functions is verified with the reactor coolant system at normal operating pressure.
The capability of the makeup pumps to operate at miniflow and the operation of the miniflow heat exchanger is verified.
The proper purification loop flowrate through the demineralizers and filters is verified.
2.9.2.4 Normal Residual Heat Removal System Testing pose purpose of the normal residual heat removal system testing is to verify that the as-installed ponents and associated piping, valves, and instrumentation properly perform the following nse-in-depth functions, as discussed in Section 5.4:
Remove reactor core decay heat and cool the reactor coolant system during shutdown operations at low pressure and temperature Remove reactor core decay heat from the reactor coolant system during reduced reactor coolant inventory operations in Modes 5 and 6 Following actuation of the automatic depressurization system, provide makeup to the reactor coolant system at low pressure 14.2-54 Revision 1
Provide low temperature overpressure protection for the reactor coolant system Remove reactor core decay heat and cool the spent fuel pool during refueling operations when the core is off-loaded from the reactor vessel to the spent fuel pool.
requisites construction testing of the normal residual heat removal system is completed. The required operational testing of the in-containment refueling water storage tank, reactor coolant system, sive core cooling system, component cooling water system, service water system, ac electrical er and distribution systems, and other interfacing systems required for operation of the above ems and data collection is available as needed to support the specified testing and system figurations. The reactor coolant system and the in-containment refueling water storage tank have dequate water inventory to support testing.
neral Test Acceptance Criteria and Methods mal residual heat removal system performance is observed and recorded during a series of vidual component and system testing, that characterizes system operation. The following testing fies that the normal residual heat removal system performs its defense-in-depth functions as cribed in Subsection 5.4.7.6.1 and appropriate design specifications:
Operation of valves to open, to close, or to control flow as required to perform the above defense-in-depth functions is verified.
Operation of system controls, alarms, instrumentation, and interlocks associated with performing the above defense-in-depth functions is verified. In addition, the proper operation of the normal residual heat removal system/reactor coolant system isolation valve interlocks specified in Section 7.6 is verified.
The normal residual heat removal system pumps testing includes verification that the pump flow rate corresponds to the expected system alignment, proper pump miniflow operation, and verification that adequate net positive suction head is available for the configurations tested. The following system configurations are tested with each pump operating individually and with two pumps operating:
z Recirculation from and to the reactor coolant system with the reactor coolant system at mid-loop hot leg water level and atmospheric pressure z Makeup to the reactor from the in-containment refueling water storage tank with approximately 4 feet of water in the tank z Makeup to the reactor from the cask loading pit with water in the pit at a sufficient level to support pump operation z Recirculation from and to the spent fuel pool with the pool at normal minimum level.
During the verifications of normal residual heat removal system flow to the reactor coolant system, verify that the pumped flow provides sufficient back pressure to maintain a water level in the CMT.
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pump/heat exchanger operating individually.
The capability of the normal residual heat removal heat exchangers to provide the required heat removal rate from the spent fuel pool is verified. Since the spent fuel pool is not heated during pre-operational testing, this verification can be made based on the flowrate from Item c and heat removal capability from Item e, above.
Operation of the normal residual heat removal system relief valve which provides low temperature overpressure protection for the reactor coolant system is verified by the performance of baseline in-service testing, as specified in Subsection 3.9.6. The acceptance criteria are based on the valve performance criteria specified in Subsection 5.4.9.
Operation of the system to facilitate draining the reactor coolant system water level to near the centerline of the hot leg for reduced inventory operations is verified. This test is performed in conjunction with the chemical and volume control system, and is used to demonstrate the performance of the reactor coolant system hot leg level instruments as discussed in Subsection 14.2.9.1.1.
2.9.2.5 Component Cooling Water System Testing pose purpose of the component cooling water system testing is to verify that the as-installed system perly performs the following defense-in-depth functions as described in Subsection 9.2.2:
Provide cooling water to defense-in-depth components and transfer heat to the service water system. In addition, this system provides cooling water to other nonsafety-related components for heat removal.
requisites construction testing of the component cooling water system is completed. Preoperational testing e cooled components has been completed as necessary to support testing of the component ling water system. Required support systems are available, including applicable portions of the ice water system and electrical power and distribution systems. Data collection is available as ded to support the specified testing and system configurations.
neral Test Acceptance Criteria and Methods ponent cooling water system performance is observed and recorded during a series of individual ponent and integrated system testing that characterizes the various modes of system operation.
following testing demonstrates that the system operates as described in Subsection 9.2.2 and in ropriate design specifications:
Proper operation of the component cooling water pumps is verified.
Proper operation of defense-in-depth related instrumentation, controls, actuation signals and interlocks is verified, including:
z Automatic pump actuation if an operating pump stops z Pump flow rate z Pump discharge pressure z Surge tank water level and control 14.2-56 Revision 1
This testing includes actuation of the system pumps and remotely-operated valves from the main control room as appropriate.
The capability to provide the expected cooling water flow rates to and from the required components with both pumps operating, and with either individual pump and heat exchanger operating as specified in the appropriate design specifications is verified.
In conjunction with Item c above, the pump(s) runout flow rate is verified to be properly limited, and adequate net positive suction head is verified to be available during its operating modes.
The capability of the heat exchanger(s) to transfer heat properly to the service water system is verified under simulated plant conditions during plant hot functional testing. Testing conditions assume both pumps/heat exchangers in operation and with either one of the pumps/heat exchangers operating.
2.9.2.6 Service Water System Testing pose purpose of the service water system testing is to verify the capability of the as-installed system to orm the following defense-in-depth function as described in Subsection 9.2.1:
Transfer heat from the component cooling water heat exchangers to the environment requisites construction testing of the service water system is completed. Preoperational testing of the ponent cooling water heat exchangers so that they can receive service water has been pleted, as well as the electrical power and distribution systems, and other interfacing systems uired for operation of the service water system. Data collection is available as needed to support specified testing and system configurations. The component cooling water system and ponents it cools are functional and hot preoperational testing of the reactor coolant system is in gress in order to confirm the service water system heat removal and heat rejection capability.
neral Test Acceptance Criteria and Methods vice water system performance is observed and recorded during a series of individual component integrated system testing. The following testing demonstrates that the service water system perly performs its defense-in-depth functions, as described in Subsection 9.2.1 and appropriate ign specifications:
Proper operation of the service water pumps, valves, strainers, cooling tower fans, and freeze protection provisions are verified.
Proper operation of the defense-in-depth related instrumentation, controls, actuation signals and interlocks is verified, including:
z Automatic pump actuation if an operating pump stops z Pump flow rate z Pump discharge pressure z Cooling tower water level and control z Cooling tower basin water temperature and control 14.2-57 Revision 1
This testing includes actuation of defense-in-depth pumps and remotely-operated valves from the main control room as appropriate.
The capability of the pumps to provide the expected cooling flow rates to and from the component cooling water heat exchangers is verified. Testing conditions include both pumps operating and either individual pump operating.
In conjunction with Item c above, the pump(s) runout flow rate is verified to be properly limited, and adequate net positive suction head is verified to be available during appropriate operating modes.
The heat removal and heat rejection capability of the service water system during the conditions of the plant hot functional testing is verified. Testing conditions include both pumps/cooling towers cells in operation and with either one of the pumps/cooling tower cells operating.
2.9.2.7 Spent Fuel Pool Cooling System Testing pose purpose of the spent fuel pool cooling system testing is to verify that the system properly orms the following defense-in-depth function described in Subsection 9.1.3:
Remove heat from the spent fuel stored in the spent fuel pool Prevent back flow through refueling canal drain lines when other in-containment compartments have been flooded requisites construction testing of the spent fuel pool cooling system has been completed. The spent fuel l is filled with water of acceptable quality and chemistry. The ac electrical power and distribution ems and other interfacing systems required for operation of the pumps and for data collection are ilable as needed to support the specified testing and system configurations.
neral Test Acceptance Criteria and Methods nt fuel pool cooling system performance is observed and recorded during a series of individual ponent and integrated system testing. The following testing demonstrates that the system perly performs its defense-in-depth function as described in Subsection 9.1.3 and appropriate ign specifications:
Proper operation of the spent fuel pool cooling pumps, valves, and strainers is verified.
Proper operation of the instrumentation, controls, actuation signals, and interlocks is verified, including:
z Automatic pump actuation if an operating pump stops z Pump flow rate z Pump discharge pressure z Spent fuel pool water level and control 14.2-58 Revision 1
This testing includes operation of the system pumps from the main control room.
The capability of the pumps to provide the expected cooling flow rates to and from the pool is verified; with both pumps operating, with either individual pump operating, and with either heat exchanger operating.
In conjunction with Item c above, the pump(s) runout flow rate is verified to be properly limited, and adequate net positive suction head is verified to be available during the appropriate operating modes.
The proper operation of the spent fuel pool siphon breakers is verified.
The proper operation of the spent fuel pool post-72 hour gravity drain flowpaths from the cask washdown pit and the passive containment cooling water storage tank is verified.
The gates, drains, bellows, and gaskets in the refueling canal and fuel storage pool are checked for unacceptable leakage.
2.9.2.8 Fire Protection System Testing pose purpose of the fire protection system testing is to verify the system properly performs the wing defense-in-depth function as described in Subsection 9.5.1:
Provide equipment for manual fire fighting in areas containing safe shutdown equipment Provide automatic fire suppression in areas containing selected non-safety-related equipment.
Provide a nonsafety-related containment spray to reduce offsite dose following a severe accident requisites construction tests of the fire protection system have been completed. Required preoperational ing of the ac power and distribution systems and other interfacing systems required for operation e fire protection system. Data collection is available as needed to support the specified testing system configurations.
neral Test Method and Acceptance Criteria protection system performance is observed and recorded during a series of individual ponent and integrated system testing to verify the system performs its defense-in-depth function.
following testing demonstrates that the system performs its defense-in-depth functions specified ubsection 9.5.1 and as specified in appropriate design specifications:
The capability of the seismic standpipes to supply the required fire water quantity and adequate water pressure for effective hose streams as the required flow rate is verified.
The operability of the fire detection equipment is verified to be able to properly detect fires and alert personnel.
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The proper operation of the fire pumps, fire water storage tank, and fire water supply piping, valves, and instrumentation to provide the as-designed fire water supply is verified.
The proper installation and operation of automatic fire suppression equipment is verified.
The proper installation and operation of electrical isolation devices for non-safety related equipment in opposite divisional fire areas is verified.
Operation of the containment spray remotely operated valve and the continuity of a flow path through the containment spray piping is verified.
2.9.2.9 Central Chilled Water System Testing pose purpose of the central chilled water system testing is to verify that the as-installed low capacity ion of this system properly performs the following defense-in-depth function, as described in section 9.2.7:
Provide chilled water to cool air used to cool safety-related or defense-in-depth equipment rooms proper function of the high capacity portion of this system is also verified.
requisites construction testing of the low capacity subsystem of the central chilled water system has been pleted. The required preoperational testing of the component cooling and service water systems, lectrical power and distribution systems, and other interfacing systems required for operation of central chilled water system has been completed. Data collection is available as needed to port the specified testing and system configurations.
neral Test Acceptance Criteria and Methods tral chilled water system performance is observed and recorded during a series of individual ponent and integrated system testing. The following testing demonstrates that the central chilled er system performs its defense-in-depth functions described in Subsection 9.2.7 and appropriate ign specifications:
Proper operation of the low capacity portion of the central chilled water system equipment is verified, including chillers, pumps, and valves.
Proper calibration and operation of defense-in-depth related instrumentation, controls, actuation signals and interlocks are verified, including:
z Temperature control of the chilled water z Chiller and chilled water pump actuation z Chilled water pump flow and discharge pressure z Chilled water flow control to air handling units This testing includes actuation of the defense-in-depth pumps and remotely operated valves from the main control room.
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In conjunction with Item c above, the pump(s) runout flow rate is verified to be properly limited, and adequate net positive suction head is verified to be available during the appropriate operating modes.
The heat removal capability of the air-cooled chillers is verified when the component areas cooled by the nuclear island nonradioactive ventilation system air handling units are operating.
ddition, the operability of the high capacity portion of the central chilled water system described in section 9.2.7 and appropriate design specifications, is verified.
2.9.2.10 Nuclear Island Nonradioactive Ventilation System Testing pose purpose of the nuclear island nonradioactive ventilation system testing is to verify that the nstalled system properly performs the following defense-in-depth functions, as described in section 9.4.1:
Protect the main control room and control support area from smoke infiltration Provide the capability to remove smoke from the main control room, control support area, and Class 1E electrical equipment rooms Provide heating, ventilation, and cooling for the main control room, control support area, and Class 1E electrical equipment rooms Provide air filtration to limit radioactivity in the main control room and control support area Maintain passive heat sinks at acceptably low initial temperatures Maintain the main control room and control support area at positive pressure safety-related functions associated with this system are tested as part of the main control room rgency habitability testing described in Subsection 14.2.9.1.6.
requisites construction testing of the nuclear island nonradioactive ventilation system has been completed.
required preoperational testing of central chilled water system, the hot water heating system, the lectrical power and distribution systems, and other interfacing systems required for operation of above systems has been completed. Data collection is available as needed to support the cified testing and system configurations.
neral Test Acceptance Criteria and Methods lear island nonradioactive ventilation system performance is observed and recorded during a es of individual component and integrated system testing to verify the system performs its nse-in-depth functions. The following testing demonstrates that the system performs its defense-epth functions as described in Subsection 9.4.1 and appropriate design specifications:
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verified. This testing includes the following:
z Smoke detectors and alarms z Air handling unit and fan flows, controls, and alarms z Differential air pressures and alarms z Air and air filtration unit charcoal temperatures, controls, and alarms z Air relative humidity measurements, controls, and alarms z Isolation/shutoff damper controls z Fire/smoke damper controls This testing includes operation from the main control room.
The proper air flows from and through each air handling unit, as well as to and from the main control room, control support area, and other equipment rooms is established for each mode of operation.
The main control room and control support area are verified to be maintained at the proper positive pressure.
The main control room, control support area, class 1E equipment rooms, and passive heat sink areas are verified to be maintained at their proper temperature during hot functional testing.
Air inleakage into the main control room and control support area is measured using a tracer gas.
2.9.2.11 Radiologically Controlled Area Ventilation System pose purpose of the radiologically controlled area ventilation system testing is to verify that the nstalled system properly performs the following defense-in-depth function, as described in section 9.4.3:
In conjunction with the low capacity portion of the central chilled water system, maintain the normal residual heat removal system and chemical and volume control system pump rooms at proper temperature during pump operation requisites construction testing of the radiologically controlled area ventilation system has been completed.
required preoperational testing of the central chilled water system, the ac electrical power and ribution systems, and other interfacing systems required for operation of the radiologically trolled area ventilation system has been completed. Data collection is available as needed to port the specified testing and system configurations.
neral Test Acceptance Criteria and Methods iologically controlled area ventilation system performance is observed and recorded during a es of individual component and integrated system testing to verify the system performs its nse-in-depth function as described in Subsection 9.4.3 and appropriate design specifications:
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and interlocks is verified. This testing includes operation of the normal residual heat removal and chemical and volume control pump room cooler/fans from the main control room.
The proper air flow and cooling capability of the normal residual heat removal and chemical and volume control pump room cooler/fans is verified.
The proper actuation of the normal residual heat removal and chemical and volume control pump room cooler fans in response to pump operation or high room temperature is verified.
2.9.2.12 Plant Control System Testing pose purpose of the plant control system testing is to verify that the as-installed components perform following nonsafety-related defense-in-depth functions, described in Section 7.1:
Provide control and coordination of the plant during startup, ascent to power, power operation and shutdown conditions by integrating the automatic and manual control of the reactor, reactor coolant and reactor support processes required for normal and off-normal conditions.
This includes rod control, pressurizer pressure and level control, steam generator water level control, steam dump (turbine bypass) control and rapid power reduction.
Provide control of other defense-in-depth systems and components.
requisites struction and installation testing of the plant control system has been completed. Related system rfaces are available or simulated as necessary to support the specified test configurations.
ponent testing and instrument calibrations have been completed. The reactor vessel integrated d package is in place, all control rod drive mechanism cables are connected and the integrated d and control rod drive mechanism cooling system is operational. Programming has been pleted and the initial software diagnostics tests have been completed. Required electrical power plies and control circuits are energized and operational. Required plant control system field wiring ectrically isolated to prevent operation of components controlled by the plant control system.
ipment or components that cannot be operated without damage or upsetting the plant are ated, either by using test switches provided by the Plant Control System or by racking out power uit breakers, to block device operation. Continuity of wiring up to the equipment is verified.
neral Test Methods and Acceptance Criteria ormance of the plant control system hardware and software is observed and recorded during a es of individual component and integrated tests designed to verify operation of defense-in-depth tions. The following testing demonstrates that the system operates as described in Section 7.1 applicable design specifications:
Processing of analog and digital signals is verified by injecting reference signals and monitoring the outputs of the plant control system.
Interfaces with other applicable plant equipment and systems such as reactor power control, feedwater control and turbine control are verified by demonstrating that injection of simulated inputs for each of the control functions provided in the main control room results in the proper output as indicated by contact operation, component actuation, or electrical test.
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actuation, or electrical test.
Proper operation of defense-in-depth processing, signal selector processing, monitoring, display and operator interface features provided by the plant control system is demonstrated by monitoring system outputs in response to simulated inputs, including simulated device or data highway failures, and utilization of provided self-test functions.
Proper functioning of the rod control system is verified by evaluating response to simulated demands from the plant control system and protection and safety monitoring system, including group selection and interlocking functions.
Proper calibration and operation of the rod position indication system is demonstrated by evaluating system response to simulated rod control logic inputs, utilizing applicable displays, annunciators and alarms.
Proper operation of logic and controls for the pressurizer level and pressure control functions, including interlocks and equipment protective devices, is demonstrated by injecting simulated input signals representing anticipated pressurizer level and pressure transients.
2.9.2.13 Data Display and Processing System Testing pose purpose of the data display and processing system testing is to verify that the as-installed ponents properly perform the following nonsafety-related defense-in-depth functions, described ection 7.1:
Display plant parameters for normal and emergency operations Provide plant alarm functions for normal and emergency plant operations Provide operational support for plant personnel, including computerized, interactive plant procedures Provide analysis, logging and historical storage and retrieval of plant data Provide a redundant communications network for transmission of plant parameters, plant status, displays, alarms and data files requisites struction and installation testing of the data display and processing system has been completed.
ated system interfaces are available or simulated as necessary to support the specified test figurations. Component testing and instrument calibrations have been completed. Programming been completed and the initial software diagnostics tests have been determined acceptable.
uired electrical power supplies are energized and operational. Required system interfaces are nected and available or simulated as necessary to support the specified test configurations.
neral Test Methods and Acceptance Criteria ormance of the data display and processing system hardware and software is observed and rded during a series of individual component and integrated tests designed to verify that the data 14.2-64 Revision 1
Initial operation of installed devices is verified by completing the diagnostic tests provided for the components and equipment.
Proper operation of the data display and processing system software and hardware is demonstrated by utilizing the data display and processing system to provide the processing, monitoring, display and operator interface features required during preoperational testing of associated plant instrumentation and control systems.
Verification that the time periods associated with accessing displays, displaying data after it has been made available on the plant monitor bus and display refresh or update rates are within the maximum allowable times is demonstrated. This verification is performed while utilizing the data display and processing system to provide the processing, monitoring, display and operator interface features required during preoperational testing of associated plant instrumentation and control systems.
2.9.2.14 Diverse Actuation System Testing pose purpose of the diverse actuation system preoperational testing is to verify that the as-installed ponents properly perform the following nonsafety-related defense-in-depth functions, described ection 7.7:
Provide diverse (from the safety-related protection and safety monitoring system) automatic actuation of the following:
- Reactor/turbine trip
- Passive residual heat removal heat exchanger
- Core makeup tanks/reactor coolant pump trip
- Passive containment cooling
- Isolation of selected containment penetrations Provide a diverse, alternate means for manual actuation of reactor trip and engineered safety features functions Provide a diverse system for monitoring selected plant parameters used to provide guidance for manual operation and confirmation of reactor trip and selected engineered safety features actuation requisites struction and installation testing of the diverse actuation system has been completed to the nt necessary to support preoperational testing. Related system interfaces are available or ulated as necessary to support the specified test configurations. Component testing and rument calibrations have been completed. Programming has been completed and initial system nostics tests have been determined acceptable. Required electrical power supplies and control uits are energized and operational. Required field wiring is electrically isolated to prevent ration of components controlled by the diverse actuation system. Exceptions are specifically tified in the preoperational test procedures if plant systems or components are to be operated ng testing and these systems or components are to be properly aligned and have proper support ems operating prior to actuation of the particular system or component. Equipment or ponents that cannot be actuated without damage or upsetting the plant are isolated using the test 14.2-65 Revision 1
neral Test Methods and Acceptance Criteria ormance of the diverse actuation system is observed and recorded during a series of individual ponent and integrated tests designed to verify operation of the system components. The wing testing demonstrates that the system operates as described in Section 7.7 and applicable ign specifications:
Processing of the analog and digital signals is verified by injecting reference signals and verifying the outputs at various locations in the system.
Correct outputs or actuation functions, for the automatic actuation logic mode, are verified by demonstrating that injection of simulated inputs for each of the specified actuation functions results in the proper output as indicated by contact operation, component actuation, or electrical test.
Correct outputs or actuation functions, for the manual actuation logic mode, are verified by demonstrating that each manual actuation function results in the proper output as indicated by contact operation, component actuation, or electrical test.
Proper operation of indications and alarms for the specified inputs, including those which provide reactor trip or engineered safety features actuation status, are verified by injecting simulated input signals.
2.9.2.15 Main AC Power System Testing pose purpose of the main ac power system testing is to verify that the as-installed components perly perform the following nonsafety-related function:
Provide ac electrical power to plant nonsafety-related loads as described in Subsection 8.3.1; and the following nonsafety-related function:
Provide onsite power for post-72 hour electrical requirements.
requisites construction tests for the individual components associated with the main ac power system have n completed. The required test instrumentation is properly calibrated and operational.
itionally, the plant offsite grid connection is complete and available.
neral Test Methods and Acceptance Criteria capability of the main ac power system to provide power to plant loads under various plant rating conditions is verified. The system components to be tested include the ancillary diesel erator, the medium and low voltage power system, load centers, motor control centers, and rumentation and controls. The following tests verify that the main ac power system provides its tions as specified in Subsection 8.3.1 and appropriate design specifications:
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supplies. Verify the bus voltages are within design limits. This test can be performed in conjunction with the testing of the standby diesel generator.
Energize the medium voltage buses from their associated unit auxiliary transformer. Verify the bus voltages are within design limits.
Energize each medium voltage bus from the reserve auxiliary transformer. Verify the bus voltages are within design limits.
Operate the automatic and maintenance bus transfer schemes. Verify successful transfer and return operation.
Verify correct operation of the manual controls, annunciation, and instrumentation for the 480 V load centers and their 6900 V feeder breakers.
Simulate fault conditions at the 480 V load centers and verify alarms and operation of trip devices and protective relays.
Energize the 480 V load centers. Verify the bus voltages are within design limits.
Verify the operability of motor control center supply breakers.
Simulate fault conditions at the motor control centers and verify alarms and operation of trip devices and protective relays.
Energize the motor control centers. Verify the bus voltages are within design limits.
Start ancillary diesel generators, energize voltage regulating transformers. Verify the input voltages to the regulating transformers are within design limits.
2.9.2.16 Non-Class 1E dc and Uninterruptible Power Supply System Testing pose purpose of the non-Class 1E dc and uninterruptible power supply system testing is to verify the ty to provide continuous, reliable power for the non-Class 1E control and instrumentation nse-in-depth loads.
requisites construction tests for the individual components associated with the non-Class 1E dc and terruptible power supply system have been completed. Permanently installed and test rumentation are properly calibrated and operational. The 480 V ac system is in operation to ply power to the battery chargers. Additionally, a test load is available for the performance of ery capacity tests.
neral Test Methods and Acceptance Criteria non-Class 1E dc and uninterruptible power supply system consists of electrical equipment uding batteries, battery chargers, inverters, static transfer switches, and associated rumentation and alarms that is used to supply power for the non-Class 1E control and rumentation loads. Performance is observed and recorded during a series of individual ponent and integrated system tests. These tests verify that the non-Class 1E dc and 14.2-67 Revision 1
The capability of each of the three non-Class 1E batteries serving defense-in-depth loads is verified to meet or exceed the required ampere-hour rating by a battery performance test in accordance with IEEE 450. Following this discharge, the voltage of each cell is verified to be greater than or equal to the specified minimum cell voltage.
The capability of each of the three chargers serving defense-in-depth loads to meet the rating specified by Table 8.3.2-6 is verified. This testing includes a verification that the charger output voltage is within design limits.
The capability of each inverter to meet the rating specified by Table 8.3.2-6 is verified. This testing includes a verification that the output frequency and voltage to be within the limits specified in Table 8.3.2-6.
The proper operation and calibration of instrumentation and alarms, electrical ground detection, and permissive and prohibitive interlocks is verified.
2.9.2.17 Standby Diesel Generator Testing pose purpose of the standby diesel generator testing is to verify the capability to provide electrical er to plant nonsafety-related loads that enhance an orderly plant shutdown if off-site ac power is available.
requisites construction tests have been completed. The necessary permanently installed instrumentation is perly calibrated and operational. Appropriate electrical power sources and diesel generator ding heating and ventilation system are available for use. The plant control system is available for ration as applicable to the diesel generators. Sufficient diesel fuel is available, on site or readily essible, to perform the tests.
neral Test Methods and Acceptance Criteria ormance is observed and recorded during a series of individual component and integrated tests.
se tests verify that the diesel generators operate properly as specified in Sections 8.3 and 9.5 ugh the following testing:
Verify the operability of generator protection features described in Subsection 8.3.1.1.2.2.
Simulate the loss of ac voltage and verify proper operation of undervoltage relay. Verify sequencer control logic support the description in Tables 8.3.1-1 and 8.3.1-2.
Verify the diesel generators fuel transfer pumps start and stop automatically in response to simulated day tank low level and high level signals.
Transfer fuel oil from the fuel oil storage tank to the diesel fuel oil day tanks by means of the transfer pumps. Verify flow parameters are within design limits.
Verify proper operation of diesel generators building heating and ventilation system fans and dampers, manual and automatic controls, alarms, and indicating instruments, as described in Subsection 9.4.10.
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Verify the diesel generator lockout features (turning gear engaged, emergency stop).
Verify that the diesel generator air starting system has sufficient capacity for cranking the engine for prescribed number of automatic or manual starts without recharging.
Start the diesel generators. Verify voltage and frequency control.
Verify the full load-carrying capability for a period of not less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, of which 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at are at a load equivalent to the 2-hour (Standby) rating of the diesel generators and 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> at a load equivalent to the continuous rating of the diesel generators. Verify the voltage and frequency requirements are maintained. Verify that the diesel generator cooling system functions within design limits.
Following the full-load capability test, simulate loss of ac voltage and verify proper automatic startup, sequencing, and operation of the diesel generators. Verify diesel generators bus de-energization and load shedding. Verify diesel generators attain frequency and voltage within design limits within the time described in Subsection 8.3.1.1.2.3. Verify sequencer control logic meets the description in Tables 8.3.1-1 and 8.3.1-2. Verify that the diesel generators continuous rating is not exceeded. Verify voltage and frequency requirements are maintained.
Verify that the rate of fuel consumption and the operation of the fuel transfer pumps and associated components, while providing power to the load equivalent to those specified in Table 8.3.1-1 or 8.3.1-2, are such that the design capacity of the fuel oil storage tanks meets the Subsection 9.5.4 requirement for 7-day storage inventory.
With each diesel generator bus supplied only by the diesel generator and supplying loads up to its continuous rating, trip a load equivalent to the largest single load in Table 8.3.1-1 or 8.3.1-2. Verify that the voltage and frequency values are maintained within design limits.
With each diesel generator supplying loads up to its continuous rating, trip the generator breaker that supplies power to the diesel generator bus. Verify that the diesel engine continues to run and does not trip on overspeed.
2.9.2.18 Radiation Monitoring System Testing pose purpose of the radiation monitoring system testing is to verify that the as-installed radiation itors perform their defense-in-depth function as described in Section 11.5.
requisites construction testing of the radiation monitoring system has been completed. The radiation itors have been calibrated and the monitor check sources are installed, as appropriate. The uired preoperational testing of the protection and safety monitoring system, plant control system, electrical power and distribution systems, and other interfacing systems required for operation data collection is available as needed to support the specified testing.
neral Test Acceptance Criteria and Methods iation monitoring system performance is observed and recorded during a series of individual ponent and integrated system testing to verify the system performs its defense-in-depth 14.2-69 Revision 1
The proper calibration and operation of each radiation detector assembly and associated equipment using a standard radiation source or portable calibration unit are verified.
Proper operation of the monitoring equipment and controls required for manually initiated operation of the monitor check sources is verified.
Proper operation of the local processors that process and transmit radiation monitoring data to the protection and safety monitoring system or plant control system, as appropriate, is verified.
Proper actuation of alarms and signals for actuation of equipment responses following receipt of a high radiation signal is verified.
preoperational testing discussed in Subsection 11.5.7 is performed following successful pletion of the testing described above.
2.9.2.19 Plant Lighting System Testing pose purpose of plant lighting system testing is to verify that the system can perform its defense-in-th function of providing emergency lighting in the main control room and remote shutdown kstation area to illuminate these areas for emergency operations upon loss of normal lighting, as cribed in Subsection 9.5.3. In addition, the operability of lighting for emergency ingress and ess is verified.
requisites construction testing of the plant lighting system is completed. The required preoperational testing e interfacing and support systems required for testing the emergency lighting function is ilable as needed to support the specified testing and system configurations including the ss 1E dc and uninterruptible power supply system, and the main ac power system.
neral Test Acceptance Criteria and Methods nt lighting system performance is observed during a series of individual component and integrated em testing to verify the system capability to perform its defense-in-depth functions. The following ing verifies that the system operates as described in Subsection 9.5.3 and in appropriate design cifications:
The proper operation of the plant lighting system emergency lighting is verified when powered from the Class 1E dc and uninterruptible power supply system.
Self-contained emergency lighting units are verified to be operable and installed into the proper ingress and egress paths, standby diesel generator rooms, switchgear rooms (annex and turbine buildings), fire pump rooms, access route between the main control room and remote shutdown workstation, and appropriate connecting corridors and stairwells.
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purpose of the primary sampling system testing is to verify that the as installed components perly perform the following nonsafety-related defense-in-depth functions described in section 9.3.3:
Provide the capability to obtain samples of the reactor coolant, passive core cooling system, containment sump water, and containment atmosphere Provide the capability to analyze and measure samples.
requisites struction testing of the primary sampling system has been completed. Component cooling water eing provided to the sample cooler when samples are taken from the reactor coolant system n it is at elevated temperature. The systems/components to be sampled are filled and at their mal pressure and temperature. The liquid radwaste system is available to receive discharged ple fluid. Electrical power is available for operation of the system components and a source of pressed gas is available for operation of the gas sample eductor.
neral Test Method and Acceptance Criteria performance of the primary sampling system is observed and recorded during a series of vidual component tests and testing in conjunction with the reactor coolant system and passive cooling system operation. The following testing demonstrates that the primary sampling system orms its defense-in-depth functions as described in Subsection 9.3.3 and appropriate design cifications.
Proper operation of the systems remotely-operated valves and eductor supply pump is verified.
Proper calibration and operation of instrumentation, controls, actuation signals, and interlocks are verified.
Verify the capability to obtain samples from the reactor coolant system, core makeup tanks, accumulators, containment sump, and containment atmosphere.
Verify the ability to return the sample stream fluid to the containment sump or liquid radwaste system, as appropriate.
Verify the capability to route sample streams to the laboratory.
Verify the operability of the test laboratory equipment used to analyze or measure radiation levels and radioactivity concentrations.
2.9.2.21 Annex/Auxiliary Building Nonradioactive HVAC System pose purpose of the annex/auxiliary non-radioactive HVAC system testing is to verify that the as alled system properly performs the defense-in-depth function, as described in Subsection 9.4.2, rovide conditioned air to maintain the diesel bus switchgear rooms and battery charger rooms taining DC switchgear) within their design temperature range during operation of the onsite dby power system.
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e above system is completed and these systems are available as needed to support the cified testing and system configurations.
neral Test Acceptance Criteria and Methods annex/auxiliary building non-radioactive HVAC system performance is observed and recorded ng a series of individual component and integrated system testing. The following testing verifies the system functions as described in Subsection 9.4.2 and appropriate design specifications:
Proper function of the fans, filters, and dampers is verified.
Proper operation of instrumentation, controls, actuation signals, and alarms and interlocks is verified. This testing includes the following:
z Air handling unit and fan flows, controls, and alarms z Air temperatures, alarms, and controls z Damper open, close and modulate control in response to monitored parameters This testing includes operation from the main control room.
The ventilated areas are verified to be maintained at a slightly positive pressure relative to the outside air pressure and other areas of the auxiliary building.
The switchgear and equipment room subsystem air handling unit supply and return fans are verified to be automatically connected to the onsite standby power supplies on a loss of power to the buses powered by the standby diesels.
2.9.2.22 Pressurizer Surge Line Testing (First Plant Only) pose purpose of the pressurizer surge line testing is: a) to obtain data to verify the proper operation of perature sensors installed on the pressurizer surge line and pressurizer spray line, and b) to in Reactor Coolant System piping displacement measurements for baseline data, as described ubsections 3.9.3, 14.2.5, and 14.2.9.1.7 item (d).
requisites construction tests for the individual components associated with the Reactor Coolant System e been completed. The testing and calibration of the required test instrumentation has been pleted. The temporary sensors and instrumentation lead wires required for monitoring thermal tification, cycling, and striping have been installed. The calibration of the transducers and the rability of the data acquisition equipment have been verified. Prior to testing of the piping system, etest walk-down shall be performed to verify that the anticipated piping movement is not tructed by objects not designed to restrain the motion of the system (including instrumentation branch lines). The system walk-down shall also verify that supports are set in accordance with design.
neral Test Methods and Acceptance Criteria performance of the Reactor Coolant System is observed and recorded during a series of vidual tests that characterize the various modes of system operation. This testing verifies that the 14.2-72 Revision 1
Verify the proper operation of temperature sensors installed on the pressurizer surge line and pressurizer spray line.
Record sensor data at specified intervals throughout hot functional testing of the RCS system, including during the drawing and collapsing of the bubble in the pressurizer.
Retain the following plant parameters time history for the same data recording period:
z Hot leg temperature z Reactor Coolant System pressure z Reactor coolant pump status z Pressurizer level z Pressurizer temperature (liquid and steam) z Pressurizer spray temperature z Pressurizer spray and auxiliary spray flow z Normal residual heat removal system flow rate z Passive core cooling system - passive residual heat removal flow rate.
Monitor pressurizer surge line and pressurizer spray line for valve leakage.
Remove the transducers and associated hardware after the completion of testing.
Proper operation of the temperature sensors in the pressurizer surge and spray lines is verified.
2.9.3 Preoperational Testing of Nonsafety-Related Radioactive Systems 2.9.3.1 Liquid Radwaste System Testing pose purpose of the liquid radwaste system testing is to verify that the as-installed components and ociated piping, valves, and instrumentation properly perform the following safety-related function cribed in Subsection 11.2.1.1:
Drain the passive core cooling system compartments to the containment sump to prevent flooding of these compartments and possible immersion of safety-related components Prevent back flow through the drain lines from the containment sump to the chemical and volume control system compartment and the passive core cooling system compartments, in order to prevent cross flooding of these compartments liquid radwaste system testing is performed to verify that the as-installed components and ociated piping, valves, and instrumentation properly perform the nonsafety-related functions 14.2-73 Revision 1
requisites construction testing of the liquid radwaste system is completed. The required preoperational ing of the interfacing and support systems required for testing has been completed. Data ection is available as needed to support the specified testing and system configurations.
neral Test Acceptance Criteria and Methods id radwaste system performance is observed and recorded during a series of individual ponent and system testing that characterizes system operation. This testing verifies that the em operates as specified in Section 11.2 and appropriate design specifications.
The drain lines from the passive core cooling system compartments and the refueling cavity are verified to provide a flow path to the reactor compartment.
Proper operation of the backflow prevention check valves is verified by the performance of baseline in-service tests, as specified in Subsection 3.9.6.
Proper operation of the system pumps and valves is verified, including:
z Effluent holdup tank pumps z Waste holdup tank pumps z Degasifier separator pumps z Chemical waste tank pump z Monitor tank pumps z Reactor coolant drain tank pumps Proper calibration and operation of the system instrumentation, controls, actuation signals, and interlocks are verified, including:
z Pump controls and alarms z Tank level control and alarms z Valve and pump responses to safeguards signals z Valve and pump responses to high radiation isolation signals In conjunction with the gaseous radwaste system testing in Subsection 14.2.9.3.2, the proper operation of the degasifier is verified.
The proper operation of the liquid radwaste filters and ion exchangers is verified.
2.9.3.2 Gaseous Radwaste System Testing pose purpose of the gaseous radwaste system testing is to verify that the as-installed components associated piping, valves, and instrumentation properly perform the following nonsafety-related tions described in Section 11.3.
Collect waste gases that contain radioactivity or hydrogen Provide holdup for radioactive waste gases as appropriate 14.2-74 Revision 1
ilable as needed to support the specified testing and system configurations. In addition, a source ydrogen and calibration gases is available.
neral Test Acceptance Criteria and Methods performance of the gaseous radwaste system is observed and recorded during a series of vidual component and system tests that characterizes the various modes of system operation.
testing verifies that the gaseous radwaste system operates as described in Section 11.3 and ropriate design specifications:
System and component control circuits, including response to normal control, interlock, and alarm signals are verified. The gaseous radwaste system instrumentation, controls, valves, and interlocks are verified to respond to various inputs and provide proper isolation and alarm signals. Appropriate automatic control functions are verified in response to abnormal conditions inputs.
Nitrogen, hydrogen, and calibration gases are routed through the system. Performance characteristics of the instrumentation and control systems are verified, and the delay bed operation is verified.
Moist test gas is routed through the system to verify proper moisture removal and detection.
The degasifer vacuum pump is verified to operate properly. Manual override of the automatic control functions of the drainpot and moisture separator drain and isolation valves is verified.
Sample pumps are operated and the sample flow meter indication is observed.
The proper operation of the degasifier moisture separator is demonstrated.
2.9.3.3 Solid Radwaste System Testing pose purpose of the solid radwaste system testing is to verify that the as-installed components and ociated piping, valves, and instrumentation operate properly to prepare waste generated during normal operation of the plant for processing, packaging, and shipment as described in section 11.4.1.2.
requisites construction testing of the solid radwaste system is completed. The interfacing and support ems required for testing and data collection are available as needed to support the specified ing and system configurations.
neral Test Method and Acceptance Criteria performance of the solid radwaste system is observed and recorded during a series of individual ponent and system tests that characterizes the various modes of system operation. This testing fies that the solid radwaste system operates as described in Section 11.4 and in appropriate ign specifications:
Tests are performed to verify that manual and automatic system controls, alarms, and instruments are functional; the system instrumentation, controls, valves, and interlocks 14.2-75 Revision 1
Tests are performed to verify proper system process rates as described in Section 11.4, and that no free liquids are present in packaged waste.
The capability to properly transfer and retain spent resins is verified.
The capability to properly handle filter cartridges in a manner that minimizes personnel radiation exposure is demonstrated.
2.9.3.4 Radioactive Waste Drain System Testing pose purpose of the radioactive waste drain system testing is to verify that the as-installed ponents and associated piping, valves, and instrumentation properly perform the following tions, described in Section 11.2 and Subsection 9.3.5:
Drain floor and equipment compartments Collect drainage and transfer drainage to the liquid radwaste system requisites construction testing of the radioactive waste drain system is completed. The interfacing and port systems required for testing and data collection are available as needed to support the cified testing and system configurations, including the liquid radwaste system and compressed air ply.
neral Test Method and Acceptance Criteria performance of the radioactive drain system is observed and recorded during a series of vidual component and system tests that characterizes the various modes of system operation.
testing verifies that the system operates as described in Section 11.2 and Subsection 9.3.5, and ppropriate design specifications:
Proper operation of system instrumentation, controls, alarms, and interlocks is verified.
Proper operation of the system pumps and valves is verified.
Proper system and component flow paths and flowrates, including pump capacities and sump tank volumes, is verified.
Flow water in each drain path to verify that the drains discharge to their designated destination and that proper drain path segregation is maintained.
2.9.3.5 Steam Generator Blowdown System Testing pose purpose of the steam generator blowdown system testing is to verify that the as-installed ponents and associated piping, valves, and instrumentation operate properly to provide an atable flow path for the controlled removal of water from the secondary side of the steam erators as described in Section 10.4.
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cified testing and system configurations. A portion of this testing is performed during the hot tional testing of the plant, when the steam generators are at or near normal operating pressure temperature.
neral Test Method and Acceptance Criteria performance of the steam generator blowdown system is observed and recorded during a series dividual component and system tests that characterize the various modes of system operation.
testing demonstrates that the system operates as described in Section 10.4 and in appropriate ign specifications:
Proper operation of system instrumentation, controls, alarms, and interlocks is verified.
Proper operation of the system pump and valves is verified.
The proper operation of the electrodeionization units is verified.
The heat transfer capability of each blowdown heat exchanger is verified.
The automatic isolation of steam generator blowdown on low steam generator level is verified.
2.9.3.6 Waste Water System Testing pose purpose of the waste water system testing is to verify that the as-installed components and ociated piping, valves, and instrumentation operate properly to collect and perform appropriate essing of normally non-radioactive drains, as described in Section 11.2 and Subsection 9.2.9.
requisites construction testing of the waste water system is completed. The interfacing and support ems required for testing and data collection are available as needed to support the specified ing and system configurations.
neral Test Acceptance Criteria and Methods ste water system performance is observed and recorded during a series of individual component system testing that characterizes system operation. This testing verifies that the system operates escribed in Section 11.2 and Subsection 9.2.9 and appropriate design specifications.
Proper operation of the system pumps and valves is verified.
Proper calibration and operation of the system instrumentation, controls, actuation signals, and interlocks is verified.
Proper system and component flow paths and flowrates, including pump capacities and sump tank volumes is verified.
Verify the ability of the waste water system radiation alarm to trip the drain tank pumps and the waste water retention basin pumps, as appropriate.
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pose purpose of the condensate system testing is to verify that the as-installed components properly orm the system functions, described in Subsection 10.4.7, of delivering the required flow of ted water from the condenser hotwell to the feedwater system.
requisites construction testing of the condensate system has been completed. The construction testing of condenser is completed and a source of water of appropriate quality is available for filling the denser hotwell. The steam generator feedwater system is available to receive flow from the densate system. Required electrical power supplies and control circuits are operational.
neral Test Method and Acceptance Criteria densate system performance is observed and recorded during a series of individual component integrated system testing. The following testing verifies that the condensate system can perform unctions as described in Subsection 10.4.7 and appropriate design specifications:
Proper operation of the condensate pumps and system valves is verified.
Proper calibration and operation of the system instrumentation, controls, actuation signals, and interlocks are verified.
Proper operation of the heater drains is verified.
During the plant hot functional testing, the integrated operation of the condensate system in conjunction with the feedwater system is verified with the condenser and circulating water system in operation.
2.9.4.2 Condenser Air Removal System Testing pose purpose of the condenser air removal system testing is to verify that the as-installed components perly perform the system functions to establish and maintain the required vacuum in the main denser, as described in Subsection 10.4.2.
requisites construction testing of the condenser air removal system has been completed. The construction ing of the condenser has been completed and a source of water of appropriate quality is available illing the condenser hotwell. The turbine gland sealing system and exhaust blower are in ration. A source of steam such as the auxiliary boiler is available. Required support systems, trical power supplies and control circuits are operational.
neral Test Method and Acceptance Criteria denser air removal system performance is observed and recorded during a series of individual ponent and integrated system testing. The following testing verifies that the condensate system perform its functions as described in Subsection 10.4.2 and appropriate design specifications:
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and interlocks are verified.
The capability of the vacuum pumps to establish the required vacuum in the main condenser is verified.
2.9.4.3 Main Turbine System and Auxiliaries Testing pose purpose of the main turbine system testing is to verify that the as-installed main turbine and its iliary components properly perform their functions, described in Sections 10.2 and 10.4. This ing includes testing of the turbine gland sealing system, lube oil system, turning gear, turbine trols and protective functions, and moisture separator reheater.
requisites construction testing of the main turbine and its auxiliaries has been completed. The construction ing of the condenser is completed and a source of water of appropriate quality is available for g the condenser hotwell. The main turbine is on turning gear and the condenser air removal em is operable. A source of steam such as the auxiliary boiler is available. Required support ems, electrical power supplies and control circuits are operational.
neral Test Method and Acceptance Criteria ause this testing is performed using a temporary steam source, the extent to which the turbine be tested in preoperational testing is limited. However, the proper function of the turbine iliaries is verified to assure the turbine will operate properly when a greater amount of steam is ided.
n turbine system performance is observed and recorded during a series of individual component integrated system testing. The following testing verifies that the turbine and its auxiliaries tion as described in Sections 10.2 and 10.4 and in appropriate design specifications:
Proper operation of the turbine lube oil pump and turning gear motor, gland seal exhaust blower, and moisture separator and gland seal valves is verified.
Proper operation of system valves including the turbine control and intercept valves is verified.
Proper calibration and operation of the system instrumentation, controls, actuation signals, and interlocks are verified.
Proper turbine operation during the turning gear testing is verified. The turning gear engagement and disengagement functions are verified to operate properly.
Proper performance of the turbine trip functions is verified.
2.9.4.4 Main Generator System and Auxiliaries Testing pose purpose of the main generator system testing is to verify that the as-installed main generator and uxiliary components properly perform their functions, described in Sections 8.2 and 10.2. This 14.2-79 Revision 1
requisites construction testing of the main generator and its auxiliaries has been completed. The struction testing of the condenser is completed. The turbine cooling water system is operable, required support systems, electrical power supplies, and control circuits are operational.
neral Test Method and Acceptance Criteria ormance is observed and recorded during a series of individual component and integrated tests.
se tests verify that the generator operated as specified in Sections 8.2 and 10.2 through the wing testing:
Verify the operability of the generator protection features.
Verify proper cooling of the generator stator and rotor.
Verify MW, MVAR, and frequency control.
2.9.4.5 Turbine Building Closed Cooling Water System Testing pose purpose of the turbine building closed cooling water system testing is to verify that the nstalled components properly perform their functions of supplying adequate cooling water to the ignated turbine building components, as described in Subsection 9.2.8.
requisites construction testing of the turbine building closed cooling water system has been completed.
cooled components are operational and operating to the extent possible, especially for verifying heat exchanger capability. Required support systems, electrical power supplies and control uits are operational.
neral Test Method and Acceptance Criteria bine building closed cooling water system performance is observed and recorded during a series dividual component and integrated system testing. The following testing verifies that the system tions as described in Subsection 9.2.8 and appropriate design specifications:
Proper operation of the system pumps and valves is verified.
Proper operation of the system instrumentation, controls, actuation signals, and interlocks is verified.
2.9.4.6 Circulating Water System Testing pose purpose of the circulating water system testing is to verify that the as-installed components perly perform the functions of cooling and circulating adequate cooling water to the main denser and turbine building closed cooling water system heat exchangers as described in section 10.4.5.
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ems, electrical power supplies and control circuits are operational.
neral Test Method and Acceptance Criteria e there will be little, if any, heat rejected to the circulating water system, verification of the heat oval capability of the ultimate heat sink is performed during the startup testing of the plant when reactor is producing power.
ulating water system performance is observed and recorded during a series of individual ponent and integrated system testing. The following testing verifies that the system functions as cribed in Subsection 10.4.5 and appropriate design specifications:
Proper operation of the system pumps and valves is verified.
Proper calibration and operation of the system instrumentation, controls, actuation signals, and interlocks are verified.
The proper operation of the system freeze protection equipment is verified, as applicable.
2.9.4.7 Turbine Island Chemical Feed System Testing pose purpose of the turbine island chemical feed system testing is to verify that the as-installed ponents properly perform the functions of adding appropriate chemicals to the condensate, ice water, and auxiliary boiler in a controlled manner, as described in Subsection 10.4.11.
requisites construction testing of the chemical feed system has been completed. Required support ems, electrical power supplies and control circuits are operational.
neral Test Method and Acceptance Criteria bine island chemical feed system performance is observed and recorded during a series of vidual component and integrated system testing. The following testing verifies that the system tions as described in Subsection 10.4.11 and appropriate design specifications:
Proper operation of the system pumps and valves is verified.
Proper calibration and operation of the system instrumentation, controls, actuation signals, and interlocks are verified.
2.9.4.8 Condensate Polishing System Testing pose purpose of the condensate polishing system testing is to verify that the as-installed components perly perform the functions of removing corrosion products, dissolved solids, and other impurities the condensate system, as described in Subsection 10.4.6.
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ems are operational. Required support systems, electrical power supplies and control circuits are rational.
neral Test Method and Acceptance Criteria densate polishing system performance is observed and recorded during a series of individual ponent and integrated system testing. The following testing verifies that the system functions as cribed in Subsection 10.4.6 and appropriate design specifications:
Proper operation of the system valves is verified.
Proper calibration and operation of the system instrumentation, controls, actuation signals, and interlocks are verified.
2.9.4.9 Demineralized Water Transfer and Storage System Testing pose purpose of the demineralized water transfer and storage system testing is to verify that the nstalled components properly perform the function of providing reservoirs of demineralized water deliver deoxygenated, demineralized water to various plant users, as described in section 9.2.4.
requisites construction testing of the demineralized water transfer and storage system has been pleted. The demineralized water treatment system is operational and the equipment which uses ineralized water is able to accept water. Required support systems, electrical power supplies and trol circuits are operational.
neral Test Method and Acceptance Criteria ineralized water transfer and storage system performance is observed and recorded during a es of individual component and integrated system testing. The following defense-in-depth testing fies that the system functions as described in Subsection 9.2.4 and appropriate design cifications:
Proper operation of the system pumps, valves, blower, and is verified.
Proper calibration and operation of the system instrumentation, controls, actuation signals, and interlocks are verified.
2.9.4.10 Compressed and Instrument Air System Testing pose purpose of the compressed and instrument air system testing is to verify that the as-installed ponents properly perform the functions of providing compressed air at the required pressures to ous plant users, as described in the Compressed and Instrument Air portion of Section 9.3.
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uired support systems, electrical power supplies and control circuits are operational.
neral Test Method and Acceptance Criteria pressed and instrument air system performance is observed and recorded during a series of vidual component and integrated system testing. The following testing verifies that the system its plant users, where applicable, function as described in Subsection 9.3.1.4 and appropriate ign specifications:
Proper operation of the system compressors, receivers, prefilters, air dryers, afterfilters, purifiers, and valves is verified.
Proper calibration and operation of the system instrumentation, controls, actuation signals, and interlocks are verified.
Integral testing is performed to verify that the instrument air subsystem can provide sufficient air pressure to accommodate the maximum number of air-operated valves expected to operate simultaneously.
Testing is performed to verify the fail-safe positioning of safety-related air-operated valves for sudden loss of instrument air or gradual loss of pressure as described in Subsection 9.3.1.4.
Proper calibration is verified for system relief valves that protect the system from overpressure conditions.
2.9.4.11 Containment Recirculation Cooling System Testing pose purpose of the containment recirculation cooling system testing is to verify that the as-installed ponents properly perform the functions of maintaining the proper containment air temperature ng normal plant operation and during refueling and maintenance operations, as described in section 9.4.6.
requisites construction testing of the containment recirculation cooling system has been completed. The tral chilled water system and hot water heating system are operational. Required support ems, electrical power supplies and control circuits are operational.
neral Test Method and Acceptance Criteria tainment recirculation cooling system performance is observed and recorded during a series of vidual component and integrated system testing. The following testing verifies that the system tions as described in Subsection 9.4.6 and appropriate design specifications:
Proper operation of the system fans and dampers is verified.
Proper calibration and operation of the system instrumentation, controls, actuation signals, and interlocks are verified.
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purpose of the containment air filtration system testing is to verify that the as-installed ponents properly perform the functions of supplying and exhausting air to maintain the proper tainment air pressure, and filter exhaust air to minimize radiation release, as described in section 9.4.7.
requisites construction testing of the containment air filtration system has been completed. The portions of radiologically controlled area ventilation system connected to the air filtration system are rational. The hot water heating and chilled water systems are required for verification of the air tion heating and cooling functions. Required support systems, electrical power supplies and trol circuits are operational.
neral Test Method and Acceptance Criteria tainment air filtration system performance is observed and recorded during a series of individual ponent and integrated system testing. The following testing verifies that the system functions as cribed in Subsection 9.4.7 and appropriate design specifications:
Proper operation of the system fans and dampers is verified.
Proper calibration and operation of the system instrumentation, controls, actuation signals, and interlocks are verified.
Proper operation of the containment air filtration filters is verified.
2.9.4.13 Plant Communications System Testing pose purpose of the plant communications system testing is to verify that the as-installed components perly perform the functions of verifying the proper operation and adequacy of the plant munication systems used during normal and abnormal operations, as described in Section 9.5.
requisites construction testing of the communication system has been completed. Required support ems, electrical power supplies and control circuits are operational.
neral Test Method and Acceptance Criteria nt communications system performance is observed and recorded during a series of individual ponent and integrated system testing. The inplant communications system includes the following systems:
Wireless telephone system Telephone/page system Private Automatic Branch Exchange (PABX) System Sound Powered Phone System Emergency Offsite Communication System Security Communication System 14.2-84 Revision 1
Transmitters and receivers are verified to operate without excessive interference.
Proper operation of controls, switches, and interfaces is verified.
Proper operation of the public address, including the plant emergency alarms, is verified.
The proper operation of equipment expected to function under abnormal conditions such as a loss of electrical power, shutdown from outside the control room, or execution of the plant emergency plan is verified. This functional testing will be performed under conditions that simulate the maximum plant noise levels being generated during the various operating conditions, including fire and accident conditions, to demonstrate system capabilities.
2.9.4.14 Mechanical Handling System Crane Testing pose purpose of the mechanical handling system crane testing is to verify that the as-installed ponents properly perform their functions. The test ensures operation and adequacy of the tainment polar crane, which is used to lift and relocate components providing access to the tor fuel, vessel internals, and reactor components during refueling and servicing operations.
ddition, the following load handling systems described in Subsection 9.1.5 are tested; the ipment hatch hoist, the maintenance hatch hoist, and the cask handling crane.
requisites construction testing of the heavy lift cranes has been completed. Required support systems, trical power supplies and control circuits are operational. The heavy load analysis, defining the paths, has been completed.
neral Test Method and Acceptance Criteria vy load crane performance is observed and recorded during a series of individual component integrated system testing. The following testing verifies that the crane systems function as cribed in Subsection 9.1.5 and in appropriate design specifications:
Proper operation and assembly of the various cables, grapples, and hoists including brakes, limit switches, load cells, and other equipment protective devices are verified.
Proper operation of control, instrumentation, interlocks, and alarms is verified.
Dynamic and static load testing of cranes and hoists, and associated lifting and rigging equipment are performed including a static load test at 125 percent of rated load and full operational test at 100 percent of rated load.
2.9.4.15 Seismic Monitoring System Testing seismic monitoring system testing described in this section also applies to site-specific seismic sors.
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cribed in Section 3.7.
requisites construction testing of the seismic monitoring system has been completed. Required support ems, electrical power supplies and control circuits are operational.
neral Test Method and Acceptance Criteria mic monitoring system instrumentation performance is observed and recorded during a series of vidual component and integrated system testing. The following testing verifies that the system tions as described in Section 3.7 and appropriate design specifications:
Proper calibration and response of seismic instrumentation are verified, including verification of alarm and initiation setpoints.
Proper operation of internal calibration and test features are verified.
Proper integrated system response, including actuations, alarms, and annunciations, is verified.
Verify the proper operation of the recording and analysis functions on a loss of AC power sourced.
2.9.4.16 Special Monitoring System Testing pose purpose of the special monitoring system testing is to verify that the as-installed components perly perform the following nonsafety-related functions, described in Subsection 4.4.6:
Detect the presence of metallic debris in the reactor coolant system Obtain baseline data for metal impact monitoring prior to power operations requisites struction and installation testing of the special monitoring system has been completed to the nt necessary to support preoperational testing. Related system interfaces are available or ulated as necessary to support the specified test configurations. Component testing and rument calibrations have been completed. Programming has been completed and initial system nostics tests have been determined acceptable. Required electrical power supplies are rgized and operational.
neral Test Methods and Acceptance Criteria ormance of the special monitoring system is observed and recorded during a series of individual ponent and integrated tests designed to verify system operation in response to specified input ditions. The following testing demonstrates that the system operates as described in section 4.4.6 and the applicable design specifications:
Proper calibration and response of digital metal impact monitoring instrumentation are verified.
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Baseline response data is obtained for the metal impact monitoring system to serve as a reference for monitoring degradation of sensor response.
2.9.4.17 Secondary Sampling System Testing pose purpose of the secondary sampling system testing is to verify that the as-installed components perly perform the following nonsafety-related functions, described in Subsection 9.3.4:
Provide the capability to continuously monitor selected secondary water and steam process streams in order to establish and maintain proper water chemistry during plant operation Provide the capability to manually analyze additional secondary water and steam process streams requisites struction testing of the secondary sampling system has been completed. Cooling water is being ided to the sample coolers when samples are taken from sample points with fluid temperatures eeding 125°F. The systems/components to be sampled are filled and operating at their normal sure and temperature. Electrical power is available for operation of the on-line chemistry lyzers.
neral Test Method and Acceptance Criteria performance of the secondary sampling system is observed and recorded during a series of vidual component tests and testing in conjunction with the plant in operation at normal pressure temperature. The following testing verifies that the secondary sampling system operates as cribed in Subsection 9.3.4 and appropriate design specifications.
Proper calibration and operation of on-line continuous analyzers, data collection and display, controls, and actuation signals to the turbine island chemical feed system are verified.
Proper calibration and operation of the portable analyzer are verified.
Proper operation of the sample coolers is verified.
Capability to obtain grab samples from the sample points is verified.
2.9.4.18 Turbine Building Ventilation System pose purpose of the turbine building ventilation system testing is to verify that the as installed system perly performs the normal air conditioning and ventilation functions, as described in section 9.4.9.
requisites construction testing of the turbine building ventilation system has been successfully completed.
required preoperational testing of the central chilled water and hot water heating systems, and r interfacing systems required for the operation of the above systems and data collection is 14.2-87 Revision 1
neral Test Acceptance Criteria and Methods turbine building ventilation system performance is observed and recorded during a series of vidual component and integrated system testing. The following testing verifies that the system tions as described in Subsection 9.4.9 and appropriate design specifications:
Proper function of the fans, filters, heaters, coolers, and dampers is verified.
Proper operation of instrumentation, controls, actuation signals, and alarms and interlocks is verified. This testing includes the following:
z Air handling unit and fan flows, controls, and alarms z Damper open, close and modulate control This testing includes operation from the main control room.
2.9.4.19 Health Physics and Hot Machine Shop HVAC System pose purpose of the health physics and hot machine shop HVAC system testing is to verify that the as alled system properly performs the normal air conditioning and ventilation functions, as described ubsection 9.4.11.
requisites construction testing of the health physics and hot machine shop HVAC system has been cessfully completed. The required preoperational testing of the central chilled water and hot water ting systems, and other interfacing systems required for the operation of the above systems is pleted and these systems are available as needed to support the specified testing and system figurations.
neral Test Acceptance Criteria and Methods health physics and hot machine shop HVAC system performance is observed and recorded ng a series of individual component and integrated system testing. The following testing verifies the system functions as described in Subsection 9.4.11 and appropriate design specifications:
Proper function of the fans, filters, heaters, coolers, and dampers is verified.
Proper operation of instrumentation, controls, actuation signals, and alarms and interlocks is verified. This testing includes the following:
z Radiation detectors and alarms z Air handling unit and fan flows, controls, and alarms z Air temperatures, alarms, and controls z Differential air pressure and alarms z Damper open, close and modulate control This testing includes operation from the main control room.
The health physics and hot machine shop HVAC system is verified to maintain the access control area and hot machine shop at a slightly negative pressure with respect to outdoors 14.2-88 Revision 1
2.9.4.20 Radwaste Building HVAC System pose purpose of the radwaste building HVAC system testing is to verify that the as installed system perly performs the normal air conditioning and ventilation functions, as described in section 9.4.8, as required for personnel and equipment in serviced areas; and provides the per filtration of air from potentially contaminated areas.
requisites construction testing of the radwaste building HVAC system has been successfully completed.
required preoperational testing of the central chilled water and hot water heating systems, the ac trical power and distribution systems, and other interfacing systems required for the operation of above systems is completed and these systems are available as needed to support the specified ing and system configurations.
neral Test Acceptance Criteria and Methods radwaste building HVAC system performance is observed and recorded during a series of vidual component and integrated system testing. The following testing verifies that the system tions as described in Subsection 9.4.8 and appropriate design specifications:
Proper function of the fans, filters, heaters, coolers, and dampers is verified.
Proper operation of instrumentation, controls, actuation signals, and alarms and interlocks is verified. This testing includes the following:
z Radiation detectors and alarms z Air handling unit and fan flows, controls, and alarms z Air temperatures, alarms, and controls z Differential air pressures and alarms z Damper open, close and modulate control in response to monitored parameters This testing includes operation from the main control room.
The radwaste building is verified to be maintained at a slightly negative pressure with respect to outdoors to prevent unmonitored releases of radioactive contaminants.
2.9.4.21 Main, Unit Auxiliary and Reserve Auxiliary Transformer Test pose purpose of the main, unit auxiliary and reserve auxiliary transformer testing is to demonstrate the rgization of the transformers and the proper operation of associated protective relaying, alarms, control devices.
requisites construction tests for the individual components associated with the main, unit auxiliary and rve auxiliary transformers have been completed. The required test instrumentation is properly brated and operational. Additionally, the plant offsite grid connection is complete and available.
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transformers:
Energize the unit auxiliary transformers. Verify phase rotation. Verify phase voltages are within design limits.
Energize the reserve auxiliary transformers. Verify phase rotation. Verify phase voltages are within design limits.
Simulate fault conditions and verify alarms and operation of protective relaying circuits.
2.9.4.22 Storm Drains pose m drain system testing verifies that the drains prevent plant flooding by diverting storm water y from the plant, as described in Section 2.4.
requisites struction of the storm drain system is completed, and the system is operational.
neral Test Methods and Acceptance Criteria storm drain system is visually inspected to verify the flow path is unobstructed. The system is erved under simulated or actual precipitation events to verify that the runoff from roof drains and plant site and adjacent areas does not result in unacceptable soil erosion adjacent to, or flooding Seismic Category I structures.
2.9.4.23 Off-site AC Power Systems pose site alternating current (ac) power system testing demonstrates the energization and proper ration of the as-installed switchyard components, as described in Section 8.2.
requisites struction testing of plant off-site ac power systems, supporting systems, and components is pleted. The components are operational and the switchyard equipment is ready to be energized.
required test instrumentation is properly calibrated and operational. The off-site grid connection omplete and available.
neral Test Methods and Acceptance Criteria plant off-site ac power system components undergo a series of individual component and grated system tests to verify that the off-site ac power system performs in accordance with the ociated component design specifications. The individual component and integrated tests include:
Availability of ac and direct current (dc) power to the switchyard equipment is verified.
Operation of high voltage (HV) circuit breakers is verified.
Operation of HV disconnect switches and ground switches is verified.
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Operation of switchyard equipment controls, metering, interlocks, and alarms that affect plant off-site ac power system performance is verified.
Design limits of switchyard voltages and stability are verified.
Under simulated fault conditions, proper function of alarms and protective relaying circuits is verified.
Operation of instrumentation and control alarms used to monitor switchyard equipment status.
Proper operation and load carrying capability of breakers, switchgear, transformers, and cables, and verification of these items by a non-testing means such as a QC nameplate check of as built equipment where testing would not be practical or feasible.
Verification of proper operation of the automatic transfer capability of the preferred power supply to the maintenance power supply through the reserve auxiliary transformer.
Switchyard interface agreement and protocols are verified.
test results confirm that the off-site ac power systems meet the technical and operational uirements described in Section 8.2.
2.9.4.24 Raw Water System pose water system testing verifies that the as-installed components supply raw water to the ulating water cooling tower basin, service water system cooling tower basin, fire protection water age tanks, and other systems, as described in Subsection 9.2.11.
requisites struction testing of the raw water system is completed. The components are operational and the age tanks and cooling tower basins are able to accept water. Required support systems, trical power supplies, and control circuits are operational.
neral Test Methods and Acceptance Criteria raw water system component and integrated system performance is observed to verify that the em functions, as described in Subsection 9.2.11 and in appropriate design specifications. The vidual component and integrated system tests include:
Operation of the system pumps, traveling screens, automatic strainers, and valves is verified.
Operation of the system instrumentation, controls, actuation signals, alarms, and interlocks is verified.
Operation of heat tracing on system piping is verified.
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itary drainage system testing verifies that the as-installed components properly collect and harge sanitary waste, as described in Subsection 9.2.6.
requisites struction testing of the sanitary drainage system is completed. Required support systems, trical power supplies, and control circuits are operational.
neral Test Methods and Acceptance Criteria sanitary drainage system component and integrated system performance is observed to verify the system functions, as described in Subsection 9.2.6.2.1 and in appropriate design cifications. The individual component and integrated system tests include:
Operation of lift stations and valves is verified.
Operation of the system instrumentation, controls, actuation signals, and interlocks is verified.
2.9.4.26 Fire Brigade Support Equipment pose brigade support equipment testing verifies that the equipment operates and is available when ded to perform the fire brigade functions, as described in Section 9.5.
requisites ipment is ready and available for testing.
neral Test Methods and Acceptance Criteria fire brigade support equipment undergoes a series of inspections to verify availability and rability. Equipment is available for selection and use, based on the hazard. Fire brigade support ipment tests include:
Location of portable extinguishers is verified; portable extinguishers are verified fully charged.
Operation of portable ventilation equipment is verified.
Operation of portable communication equipment is verified.
Operation of portable lighting is verified.
Operation of self-contained breathing apparatus and face masks is verified.
Operation of keys to open locked fire area doors is verified.
Turnout gear functionality and availability is verified.
Compatibility of threads for hydrants, hose couplings, and standpipe risers with the local fire department equipment is verified, or alternatively, an adequate supply of readily available hose thread adaptors is verified.
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able personnel monitors and radiation survey instruments testing verifies that the devices rate in accordance with their intended function in support of the radiation protection program, as cribed in Chapter 12.
requisites able personnel monitors, radiation survey instruments, and appropriate certified test sources are ite.
neral Test Method and Acceptance Criteria portable personnel monitors and radiation survey instruments are source checked, tested, ntained, and calibrated in accordance with the manufacturers recommendations. The portable itors and instruments tests include:
Proper function of the monitors and instruments to respond to radiation is verified, as required.
Proper operation of instrumentation controls, battery, and alarms, if applicable.
2.10 Startup Test Procedures startup testing program is based on increasing power in discrete steps. Major testing is ormed at discrete power levels as described in Subsection 14.2.7. The first tests during Power ension Testing that verify movements and expansion of equipment are in accordance with design, are conducted at a power level as low as practical (approximately 5 percent).
governing Power Ascension Test Plan requires the following operations to be performed at ropriate steps in the power-ascension test phase:
Conduct any tests that are scheduled at the test condition or power plateau.
Confirm core performance parameters (core power distribution) are within expectations.
Determine reactor power by heat balance, calibrate nuclear instruments accordingly, and confirm the existence of adequate instrumentation overlap between the startup range and power range detectors.
Reset high-flux trips just prior to ascending to the next level to a value no greater than 20 percent beyond the power of the next level unless Technical Specification limits are more restrictive.
Perform general surveys of plant systems and equipment to confirm that they are operating within expected values.
Check for unexpected radioactivity in process systems and effluents.
Perform reactor coolant leak checks.
Review the completed testing program at each plateau; perform preliminary evaluations, including extrapolation of core performance parameters for the next power level; and obtain the required management approvals before ascending to the next power level or test condition.
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eptability of performing the test at higher powers. This extrapolation is included in the analysis ion of the lower power procedure.
veillance test procedures may be used to document portions of tests, and ITP tests or portions of s may be used to satisfy Technical Specifications surveillance requirements in accordance with inistrative procedures. At Startup Test Program completion, a plant capacity warranty test is ormed to satisfy the contract warranty and to confirm safe and stable plant operation.
se tests comprising the startup test phase are discussed in this subsection. For each test a eral description is provided for test objective, test prerequisites, test description, and test ormance criteria, where applicable. In describing a test, the operating and safety-related racteristics of the plant to be tested and evaluated are identified.
ere applicable, the relevant performance criteria for the test are discussed. Some of the criteria te to the value of process variables assigned in the design or analysis of the plant, component ems, and associated equipment. Other criteria may be associated with expectations relating to performance of systems.
specifics of the startup tests relating to test methodology, plant prerequisites, initial conditions, ormance criteria, and analysis techniques are discussed in Section 14.4 in the form of plant, em and component performance and testing procedures.
2.10.1 Initial Fuel Loading and Precritical Tests s performed after preoperational testing is complete but prior to initial criticality are described in section. These tests include those performed prior to core load to verify the plant is ready for core ing, the loading of the core and the tests performed under hot conditions after the core has been ed but prior to initial criticality.
s to be performed prior to and during initial core loading are described in Subsections 14.2.10.1.1 ugh 14.2.10.1.5. These tests verify the systems necessary to monitor the fuel loading process operational and that the core loading is conducted properly.
r core load, tests are performed at hot conditions to bring the plant to a final state of readiness r to initial criticality.
2.10.1.1 Fuel Loading Prerequisites and Periodic Checks ectives Specify the prerequisites for initial fuel loading, including the status of required systems, plant conditions, and special equipment Provide a checklist for periodic verification that the conditions required for fuel loading are being maintained requisites Plant systems required for initial fuel loading have been satisfactorily tested and turned over to the plant operating staff, and are in the status specified 14.2-94 Revision 1
t Method Prior to the beginning of fuel loading, verify and document the required status of test prerequisites Throughout fuel loading, verify through periodic checks that conditions required for safe fuel loading are being maintained formance Criterion required status of prerequisites for initial fuel loading is verified and documented prior to fuel ing and maintained throughout the loading process.
2.10.1.2 Reactor Systems Sampling for Fuel Loading ective Verify that the dissolved boron concentration in the reactor coolant system and directly connected portions of associated auxiliary systems is uniform and equals or exceeds the value required by the plant Technical Specifications for fuel loading.
requisites Plant Technical Specifications for fuel loading are complete and verified Boric acid storage tanks, transfer pumps, and associated piping and equipment are filled and operable The reactor vessel is filled with borated water to a level approximately equal to the centerline of the outlet nozzles The water in the reactor vessel and reactor coolant system piping, including all directly connected auxiliary systems, is borated to a value that equals or exceeds the value specified in the plant Technical Specifications for fuel loading, and that water is circulating through the normal residual heat removal system at a rate that provides reasonable assurance of a uniform concentration.
t Method Obtain and analyze samples from at least one representative point in each auxiliary system and at four equidistant depths in the reactor vessel for boron concentration Periodically repeat sampling until the performance criteria are met formance Criteria The minimum boron concentration of all samples equals or exceeds the value specified in the plant Technical Specifications for fuel loading. If the minimum boron concentration criteria is not met, the chemical and volume control system is used to increase the boron concentration to above the specified limit.
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2.10.1.3 Fuel Loading Instrumentation and Neutron Source Requirements ectives Verify alignment, calibration, and neutron response of the temporary core loading instrumentation prior to the start of fuel loading Verify the neutron response of the nuclear instrumentation system source range channels prior to the start of fuel loading Verify the neutron response of the temporary and nuclear instrumentation system source range instrumentation prior to resumption of fuel loading following any delay of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or more requisites The following special equipment is available:
- The temporary core loading package consisting of three complete counting channels, including preshipment alignment and calibration data
- A portable neutron source with sufficient strength to verify detector response Preoperational testing of the nuclear instrumentation system source range channels is completed t Method Prior to the start of fuel loading, verify the response of temporary and nuclear instrumentation system source range channels to neutrons by using a portable neutron source Verify proper alignment and calibration of the temporary channels by comparing the neutron response data to the data obtained during preshipment testing Prior to resumption of fuel loading following a delay of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or more, verify proper operation of the temporary and nuclear instrumentation system source range channels by performing a neutron response check (using the portable neutron source or by moving a fuel assembly containing a primary neutron source) or by statistical analysis of the count rate data formance Criterion ipment used for neutron monitoring during fuel loading is operating correctly and is responsive to nges in neutron flux levels. Minimum count rates of 1/2 counts per second, attributable to core trons, are required on at least two of the available pulse-type nuclear channels at all times wing installation of the initial nucleus of fuel assemblies (approximately eight fuel assemblies, of which contains a neutron source), which permits meaningful inverse count-rate monitoring.
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fy the neutron monitoring data obtained during initial fuel loading is consistent with calculations wing the predicted response and, for plants subsequent to the first plant, with data obtained ng a previous similar fuel loading.
requisites Temporary and plant source range nuclear instrumentation has been operational for a minimum of 60 minutes to allow the equipment to attain stable operating conditions The plant is prepared for initial fuel loading Neutron monitoring data from a previous similar initial fuel loading or calculations showing the predicted response of monitoring channels are available for evaluating monitoring data t Method Prior to inserting the first fuel assembly into the reactor vessel, obtain background count rates for each temporary and plant source range channel During the insertion of each fuel assembly, continuously observe the response of at least one channel for unexpected changes in count rate Construct a plot of inverse count rate ratio, versus fuel loading step number, from monitoring data obtained after each fuel assembly is loaded and used to assess the safety with which fuel loading may continue formance Criterion itoring data are consistent with calculations showing the predicted response and, for plants sequent to the first plant, with data obtained during a previous similar fuel loading. Each sequent fuel addition will be accompanied by detailed neutron count rate monitoring to determine the just loaded fuel assembly does not excessively increase the count rate and that the apolated ICRR is behaving as expected and not decreasing for unexplained reasons.
2.10.1.5 Initial Fuel Loading ectives Establish the conditions under which the initial fuel loading is to be accomplished Accomplish initial fuel loading in a safe manner requisites The nuclear design of the initial reactor core specifying the final core configuration of fuel assemblies and inserts is completed.
Preoperational testing is completed on systems specified as required for initial fuel loading.
Preoperational testing is completed on required fuel handling tools. Tools are available, operational, and calibrated, including indexing of the manipulator crane with a dummy fuel element.
Containment integrity is established.
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The boron concentration in the reactor coolant equals or exceeds the concentration required by the plant Technical Specifications for refueling. Core moderator chemistry conditions (particularly boron concentration) are prescribed in the core loading procedure document and are verified periodically by chemical analysis of moderator samples taken prior to and periodically during core loading operations.
Sources of unborated water to the reactor coolant are isolated.
Temporary and plant source range channels are operable as required to monitor changes in core reactivity.
A surveillance program verifies that the conditions for fuel loading are maintained throughout the fuel loading program.
Auxiliary system status is in accordance with Technical Specification requirements.
The overall process of initial fuel loading will be supervised by a licensed senior reactor operator with no other concurrent duties.
t Method Place fuel assemblies, together with inserted components (control rods, burnable poison elements, primary and secondary neutron sources), in the reactor vessel one at a time according to an established and approved sequence During and following the insertion of each fuel assembly and until the last fuel assembly has been loaded, the response of the neutron detectors is observed and compared to previous fuel loading data, or calculations, to verify that the observed changes in response are as expected Check sheets are completed at prescribed intervals verifying that the conditions required for initial fuel loading are being maintained Fuel assemblies, together with inserted components (control rod assemblies, burnable poison inserts, source spider, or thimble plugging devices) are placed in the reactor vessel one at a time according to a previously established and approved sequence, which was developed to provide reliable core monitoring with minimum possibility of core mechanical damage. The core loading procedure documents include detailed tabular check sheets that prescribe and verify the successive movements of each fuel assembly and its specified inserts from its initial position in the storage racks to its final position and orientation in the core. Multiple checks are made of component serial numbers and types at successive transfer points to guard against possible inadvertent exchanges or substitutions of components, and fuel assembly status boards are maintained throughout the core loading operation. The results of each loading step will be reviewed and evaluated before the next prescribed step is started.
The criteria for safe loading require that loading operations stop immediately if:
- An unanticipated increase in the neutron count rate by a factor of two occurs in all responding nuclear channels during any single loading step after the initial nucleus of fuel assemblies is loaded.
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- A decrease in boron concentration greater than 20 ppm is determined from two successive samples of reactor coolant system water until the decrease is explained.
formance Criteria uel assemblies have been loaded into the vessel in the correct location and orientation consistent the prespecified configuration for the initial reactor core. All fuel loading steps are documented, uding the final core configuration.
2.10.1.6 Post-Fuel Loading Precritical Test Sequence ective cify the sequence of events constituting the precritical test program.
requisites Plant system conditions are established as required by the individual test instructions within the precritical test sequence, as described in Subsections 14.2.10.1.7 through 14.2.10.1.20 The systems, structures, and components required by Technical Specifications shall be operable as required for the specified plant operational mode prior to initiation of precritical testing. Preoperational and precritical tests shall be completed to confirm the operability of required plant safety systems to support precritical testing prior to the initiation of the precritical tests.
t Method instructions establish the sequence for required testing after core loading, until the plant has pleted precritical testing.
formance Criteria ormance criteria are contained in the various individual tests conducted during this time bsections 14.2.10.1.7 through 14.2.10.1.23).
2.10.1.7 Incore Instrumentation System Precritical Verification ectives Verify that the incore instrumentation thimbles have been installed correctly following initial fuel loading Verify proper operation of the incore thermocouples prior to plant heatup requisites Initial fuel loading has been completed, all incore instrumentation thimble assemblies have been installed, and all mechanical and electrical connections have been completed.
The plant is at ambient temperature and pressure prior to heatup for initial criticality.
Incore instrumentation system signal processing software has been installed and is operational.
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verify proper installation and connection of the incore sensor strings.
Obtain incore thermocouple data and compare with the measured reactor coolant system temperature to verify proper operation of the incore thermocouples and signal processing.
formance Criteria Prior to plant heatup, proper connections to the incore instrumentation thimbles are verified and outputs from the incore thermocouple system are consistent with existing plant conditions, and are consistent with design requirements specified in Subsection 4.4.6 and Section 7.5 and applicable design specification.
Data required for calibration of other plant instrumentation are obtained.
2.10.1.8 Resistance Temperature Detectors-Incore Thermocouple Cross Calibration ectives Verify calibration coefficients for the resistance temperature detectors installed in the reactor coolant system.
Determine calibration coefficients for resistance temperature detectors replaced in the reactor coolant system following hot functional testing as required.
Determine calibration coefficients for the incore thermocouples that are part of the incore instrumentation system.
requisites Initial fuel loading has been completed and the reactor coolant system is filled and vented prior to heatup for initial criticality.
Reactor coolant system resistance temperature detectors that were replaced as a result of preoperational testing are operational, and an initial alignment has been completed according to the manufacturers calibration data.
The incore instrumentation system, including signal processing software, has been installed and is operational, and the preoperational testing has been completed.
Instrumentation and data collection equipment is operational and available for logging plant data.
t Method With the reactor coolant system at ambient temperature, and at isothermal conditions at specified temperature plateaus during heatup to normal operating temperature, measure the resistance of each resistance temperature detector installed in the reactor coolant system and the output from each installed incore thermocouple, along with supplemental plant data.
Using the calibration coefficients determined during hot functional testing and the manufacturers resistance versus temperature calibration data for the replaced resistance temperature detectors, determine the best-estimate temperature of each temperature plateau from the average of the derived resistance temperature detectors temperatures.
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Verify or recompute calibration coefficients for each resistance temperature detector, as required, based on the final plateau average temperatures.
Compute calibration coefficients for each incore thermocouple based on the final plateau average temperatures and supplemental data obtained during heatup.
formance Criteria For each resistance temperature detector, the adequacy of the final calibration coefficients is verified when the temperature derived from the resistance temperature detector resistance agrees with the plateau average temperatures within predetermined limits as described in Sections 7.2 and 7.3.
For each incore thermocouple, the adequacy of the final calibration coefficients is verified when the temperature derived from the thermocouple output agrees with the plateau average temperatures within predetermined limits, as described in Subsection 4.4.6, Section 7.2 (Table 7.2-1) and Section 7.3 (Table 7.3-4).
2.10.1.9 Nuclear Instrumentation System Precritical Verification ective ablish and determine voltage settings, trip settings, operational settings, alarm settings, and rlap of channels on source range instrumentation prior to initial criticality.
requisite nuclear instrumentation system is aligned according to the design requirements.
t Method Calibrate, test, and verify functions using permanently installed controls and adjustment mechanisms.
Set operational modes of the source range channels for their proper functions, in accordance with the test instructions.
formance Criterion nuclear instrumentation system operates in accordance with the design basis functional uirements, as discussed in Subsection 4.4.6.
2.10.1.10 Setpoint Precritical Verification ectives Prior to initial criticality, verify that initial values of instrumentation setpoints assumed in the design, operation, and safety analysis of the nuclear steam supply system have been installed correctly, and identify which of these are expected to be readjusted based on the results of startup testing and initial operations.
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requisites Initial alignment and calibration of plant instrumentation has been completed, and initial set points are installed per applicable design documentation.
Preoperational and startup testing of affected plant instrumentation has been completed, and test results are documented.
t Method Review applicable design documentation and generate a list of the instrumentation setpoints assumed in the design, operation, and safety analysis of the plant. Identify setpoints expected to be modified based on the results of initial startup tests and operations.
Prior to initial criticality, the results of preoperational and startup tests, as applicable, are reviewed to verify that initial setpoints have been installed correctly. Document the results of this review for future use.
Prior to initial criticality, summarize and document the setpoint values for future plant operations.
formance Criterion r to initial criticality, installed setpoint values are verified to be consistent with Technical cifications.
2.10.1.11 Rod Control System ective onstrate and document that the rod control system performs the required control and indication tions just prior to initial criticality.
requisites The reactor coolant system is at no-load operating temperature and pressure The nuclear instrumentation system source range channels are aligned and operable t Method With the reactor at no-load temperature and pressure, just prior to initial criticality, verify the operation of the rod control system in various modes including tests of control rod block and inhibit functions.
Verify the operation of status lights, alarms, and indicators formance Criteria The performance of the rod control system as described in Subsection 7.7.1.2.
The rod control system withdraws and inserts each rod bank The rod position and indication system tracks each rod bank as it is being moved 14.2-102 Revision 1
2.10.1.12 Rod Position Indication System ective fy that the rod position indication system satisfactorily performs required indication and alarm tions for each individual rod and that each rod operates satisfactorily over its entire range of el.
requisites The reactor coolant system is at no-load operating temperature and pressure At least one reactor coolant pump is in service, with reactor coolant boron concentration not less than specified in the Technical Specifications for refueling shutdown t Method vidually withdraw rod banks from the core and reinsert them, according to the test procedure.
ord rod position sensor output voltages, and rod position readouts and group step counters in the n control room.
formance Criterion rod position indication system performs the required indication and alarm functions as discussed ubsection 7.7.1.3, and each rod operates over its entire range of travel.
2.10.1.13 Control Rod Drive Mechanisms ectives Demonstrate operation of each control rod drive mechanism under both cold and hot standby conditions Provide verification of slave cycler timing requisites The reactor coolant system is filled and vented at cold shutdown Rods are fully inserted Nuclear instrumentation channels are available A fast-speed oscillograph, or equivalent, to monitor test parameters is available t Method With the reactor core installed and the reactor in the cold shutdown condition, confirm that the slave cycler devices supply operating signals to the control rod drive mechanism stepping magnet coils.
Verify operation of all control rod drive mechanisms under both cold and hot standby conditions. Record the control rod drive mechanism magnet coil currents.
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cifications.
2.10.1.14 Rod Drop Time Measurement ectives Determine the rod drop time of each rod cluster control assembly under cold no-flow and hot full-flow conditions, with the reactor at normal operating temperature and pressure.
Verify the operability of the control rod deceleration device.
requisites Initial core loading is completed Source range channels are in operation Rods are fully inserted Reactor coolant pumps are operational t Method Withdraw each rod cluster control assembly Interrupt the electrical power to the associated control rod drive mechanism Measure and record the rod drop time, and verify control rod deceleration Perform a minimum of three additional drops for each control rod whose drop time falls outside the two-sigma limit, as determined from the drop times obtained for each test condition formance Criteria Measured rod drop times are consistent with the design basis functional requirements and the applicable plant Technical Specifications The control rod is slowed by the control rod deceleration device during rod drop testing 2.10.1.15 Rapid Power Reduction System ective fy proper operation of the rapid power reduction system prior to power operations.
requisites The following systems are operable to the extent necessary to support the test: rod control system, rod position indication system, reactor trip breakers, and reactor protection system.
The reactor is shut down, the reactor coolant system boron concentration is such that Technical Specifications requirements for shutdown margin will be met with required rod withdrawal, and all control banks are near their fully inserted positions.
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Input signals simulating a rapid loss of load exceeding 50 percent power are input to the rapid power reduction system. Verify the response of the system.
Demonstrate procedures for returning the plant to power following a partial trip.
formance Criteria Performance of the rapid power reduction system is in accordance with Subsection 7.7.1.10.
In response to the simulated loss of load, gripper power is interrupted to a preselected grouping of control rods, so that rods drop freely into the core.
Gripper power to only those control rods selected for drop is interrupted.
Procedures for returning the plant to power operations without a reactor trip are verified.
2.10.1.16 Process Instrumentation Alignment ective n T and Tavg process instrumentation under isothermal conditions prior to initial criticality.
requisites Reactor coolant pumps are operating The reactor coolant system average temperature is at the hot no-load average temperature t Method Align T and Tavg according to test instructions at isothermal conditions prior to criticality formance Criterion indicated values for reactor coolant system Thot, Tcold, Tavg, and T under isothermal conditions within the limits of the applicable design requirements as discussed in Section 7.2 (Table 7.2-1)
Section 7.3 (Table 7.3-4).
2.10.1.17 Reactor Coolant System Flow Measurement ectives Prior to initial criticality, verify that the reactor coolant system flow rate is sufficient to permit operation at power.
requisites The core is installed and the plant is at normal operating temperature and pressure.
Special instrumentation is installed and calibrated for obtaining reactor coolant flow data.
t Method Prior to initial criticality, measure the reactor coolant flow measurement parameters with all four coolant pumps in operation. Estimate the reactor coolant flow rate using these data.
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ration.
2.10.1.18 Reactor Coolant System Flow Coastdown ectives Measure the rate at which reactor coolant loop flow and pump speed changes, subsequent to tripping all reactor coolant pumps.
Measure the rate at which reactor coolant loop flow and pump speed changes, subsequent to tripping two of four reactor coolant pumps.
requisites Required component testing and instrument calibration are complete Required electrical power supplies and control circuits are operational The reactor core is installed, and the plant is at normal operating temperature and pressure with all reactor coolant pumps running t Method Record loop flow, pump speeds following the trip of all reactor coolant pumps Record loop flows, pump speeds following the trip of two of four reactor coolant pumps formance Criterion loop flows and pump speed data are obtained for verification of the loss of flow analyses in sections 15.3.1 and 15.3.2.
2.10.1.19 Pressurizer Spray Capability and Continuous Spray Flow Verification ectives Establish the optimum continuous spray flow rate Determine the effectiveness of the normal control spray requisites The reactor coolant system is at no-load operating temperature and pressure.
All reactor coolant pumps are operating.
t Method While maintaining constant pressurizer level, adjust spray bypass valves until a minimum flow is achieved that maintains the temperature difference between the spray line and the pressurizer within acceptable limits.
With the pressurizer heaters de-energized, fully open both spray valves, and record the time to lower the pressurizer pressure a specified amount.
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The pressurizer pressure response to the opening of the pressurizer spray valves is within design basis functional limits as specified in Subsection 7.7.1.6 and the appropriate pressure control system design specification documentation.
2.10.1.20 Feedwater Valve Stroke Test ective fy proper operation of the main and startup feedwater control valves prior to the start of power rations.
requisites Preoperational testing of the feedwater control systems has been completed Main and startup feedwater pumps are off Initial fuel loading has been completed prior to initial criticality.
t Method each main and startup feedwater flow control valve, the following tests are performed:
Using simulated signals for several valve demand positions covering the range from fully closed to fully open, verify the actual valve position to be consistent with the demand signal.
For selected valve position changes, measure the time required from the initiation of the demand signal until the valve reaches the final position. Typical demands changes are the following: fully closed to fully open, fully open to fully closed, 25 percent open to 75 percent open, and 75 percent open to 25 percent open.
formance Criteria main and startup feedwater valves operate as described in Subsection 7.7.1.8 and appropriate ign specifications including:
The differences between the measured actual and demand valve positions, over the range of travel, are less than prespecified tolerances.
The time between the initiation of the demand signal and the final valve position for each of the demand changes is within specified ranges as discussed in applicable design specifications.
For demand changes to intermediate valve positions, the amount of overshoot is less than specified limits as discussed in applicable design specifications.
2.10.2 Initial Criticality Tests al criticality testing is described in this section. Following completion of the core loading and riticality testing, the plant is brought to initial criticality, according to the test procedures in section 14.2.10.2.1.
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ne the sequence of tests and operations to bring the core to initial criticality.
requisite nt system conditions are established as required by the individual test instructions within this uence.
t Method ndividual test instruction will establish the plant conditions required for initial criticality.
formance Criteria evant performance criteria are provided in each of the test procedure abstracts.
2.10.2.2 Initial Criticality ective ieve initial criticality in a controlled manner.
requisites The nuclear instrumentation is verified to be operating properly (See Subsection 14.2.10.2.3)
The reactor coolant system temperature and pressure are stable at the normal hot no-load values Control rod banks are inserted, and shutdown rod banks are withdrawn The reactor coolant system boron concentration is sufficiently high so the reactor is shut down by at least 1000 pcm with all banks withdrawn t Method Accomplish initial criticality by the controlled withdrawal of the rods using the same rod withdrawal sequence used for normal plant startup, followed by the dilution of the reactor coolant system boron concentration.
At preselected points during rod withdrawal and/or boron dilution, gather data to plot the inverse count rate ratio to monitor the approach to critical evolution for reactivity monitoring.
As criticality is approached, slow or stop dilution rate to allow criticality to occur during mixing or by withdrawal of rods that have been slightly inserted for control.
formance Criterion reactor is critical.
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ablish and determine voltage settings, trip settings, operational settings, alarm settings, and rlap of channels on source and intermediate range instrumentation, from prior to initial criticality during initial criticality.
requisite nuclear instrumentation system is aligned according to the design requirements.
t Method Calibrate, test, and verify functions using permanently installed controls and adjustment mechanisms.
Set operational modes of the source and intermediate range channels for their proper functions, in accordance with the test instructions.
formance Criteria The nuclear instrumentation system operates in accordance with the design basis functional requirements, as discussed in Subsection 4.4.6.
The nuclear instrumentation system demonstrates an overlap of indication between the source and intermediate range instrumentation.
The nuclear instrumentation minimum neutron count rate and noise to signal ratio are within appropriate design specifications.
2.10.2.4 Post-Critical Reactivity Computer Checkout ective onstrate proper operation of the reactivity computer through a dynamic test using neutron flux als.
requisites The reactor is critical with the neutron flux level within the range for low-power physics testing The reactor coolant system temperature and pressure are stable at the normal no-load values The neutron flux level and reactor coolant system boron concentration are stable The reactivity computer is installed, checked out, and operational, and input flux signals are representative of the core average neutron flux level The controlling rod bank is positioned in such a way that the required reactivity insertion can be made by rod motion alone The systems, structures, and components required by Technical Specifications shall be operable as required for the specified plant operational mode prior to initiation of precritical, low power physics, and power ascension testing. Verification of proper operation of source-range and intermediate-range excore nuclear instrumentation and associated alarms and 14.2-109 Revision 1
t Method By control rod motion, add positive reactivity to the core in accordance with design requirements as discussed in Section 7.7.
During the resultant increase in flux level, make two independent measurements of core reactivity; one using the reactivity computer, and one using an analysis of the rate of change of flux level (for example, reactor period or doubling time).
formance Criterion h measurement deviation between the two independent sources of reactivity is within design rances. Adjustment and recalibration or repair of the reactivity computer may be required if the iation between the two independent sources of reactivity is not within design tolerances.
2.10.3 Low Power Tests owing successful completion of the initial criticality tests, low power tests are conducted, typically ower levels less than 5 percent, to measure physics characteristics of the reactor system and to fy the operability of the plant systems at low power levels.
2.10.3.1 Low-Power Test Sequence ective ne the sequence of tests and operations that constitutes the low-power testing program.
requisite nt system conditions are established as required by the individual test instructions within this uence.
t Method vidual test instruction will establish the plant conditions required for and during the low-power ing program following initial criticality.
formance Criteria evant performance criteria are provided in each of the test procedure abstracts.
2.10.3.2 Determination of Physics Testing Range ectives Determine the reactor power level at which the effects from fuel heating are detectable Establish the range of neutron flux in which zero power reactivity measurements are to be performed requisites The reactor is critical, and the neutron flux level is below the expected level of nuclear heating 14.2-110 Revision 1
The neutron flux level and reactor coolant system boron concentration are stable The reactivity computer is installed, checked out, and operational, and input flux signals are representative of the core average neutron flux level The controlling rod bank is positioned in such a way that the required reactivity insertion can be made by rod motion alone t Method Withdraw the control rod bank and allow the neutron flux level to increase until nuclear heating effects are indicated by the reactivity computer Record the reactivity flux level and the corresponding intermediate range channel currents at which nuclear heating occurs Multiply the measured reactivity flux level by 0.3 to determine the maximum value for the zero power testing range formance Criterion zero power testing range is determined.
2.10.3.3 Boron Endpoint Determination ective ermine the critical reactor coolant system boron concentration appropriate to an endpoint rod figuration.
requisites The reactor is critical, and the neutron flux level is within the range for low-power physics testing The reactor coolant system temperature and pressure are stable at the normal no-load values The neutron flux level and reactor coolant system boron concentration are stable Instrumentation and equipment used to measure and compute reactivity is installed, checked out, and operational, with input flux signals representative of the core average neutron flux level The controlling rod bank is positioned in such a way that limited reactivity insertion will be required to achieve the endpoint condition t Method Move the rods to the desired endpoint configuration without boron concentration adjustment Directly measure the just-critical boron concentration by chemical analysis 14.2-111 Revision 1
Add the changes to the just-critical boron concentration to yield the endpoint for the given rod configuration formance Criterion measured value for the boron endpoint is consistent with the design value within design limits as cified in the Technical Specifications.
2.10.3.4 Isothermal Temperature Coefficient Measurement ectives Determine the isothermal temperature coefficient Calculate the moderator temperature coefficient requisites The reactor is critical, and the neutron flux level is within the range for low-power physics testing The reactor coolant system temperature and pressure are stable at the normal no-load values The neutron flux level and reactor coolant system boron concentration are stable Instrumentation and equipment used to measure and compute reactivity is installed, checked out, and operational, with input flux signals representative of the core average neutron flux level The controlling rod bank is positioned near fully withdrawn t Method Vary reactor coolant system temperature (heatup/cooldown) while maintaining rods and boron concentration constant Monitor reactivity results and determine the isothermal temperature coefficient Calculate the moderator temperature coefficient using the isothermal temperature coefficient and design values formance Criterion The measured value for the moderator temperature coefficient is more negative than the Technical Specification limit 2.10.3.5 Bank Worth Measurement ective date design calculations of the reactivity worth of the rod cluster control banks.
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The reactor coolant system temperature and pressure are stable at the normal no-load values The neutron flux level and reactor coolant system boron concentration are stable Instrumentation and equipment used to measure and compute reactivity is installed and operational, with input flux signals representative of the core average neutron flux level t Method One of the following methods will be used to measure the worth of all of the individual control rod banks:
- A bank is stepwise inserted into the core from fully withdrawn and the worth is measured using the reactivity computer
- Exchange bank with another bank measured as above, with the worth determined from the critical positions and the worth of the reference bank formance Criteria The measured value for the individual bank worth is consistent with the design value within specified limits as discussed in Subsection 4.3.2.5.
The sum of the measured bank worth is consistent with the design value within the assumed uncertainty used in the shutdown margin calculation 2.10.3.6 Natural Circulation (First Plant Only) ective onstrate that core decay heat can be removed by the steam generators under the conditions of ral circulation (no reactor coolant pumps operating).
requisites The reactor is critical, and the neutron flux level is within the range for low-power physics testing The neutron flux level and reactor coolant system boron concentration and temperature are stable, and the controlling rod bank is positioned in such a way that an increase in core power level to approximately 3 percent can be achieved by rod motion alone Reactor coolant pumps are operating The reactivity computer is installed, checked out, and operational, with input flux signals representative of the core average neutron flux level Instrumentation and data collection equipment is operational and available for logging plant data 14.2-113 Revision 1
t Method Because this test is performed at beginning of life when the core fission product density is low, decay heat is simulated by reactor power By control rod motion, increase reactor power to approximately 3 percent of full power based on predictions of vessel T at full power With reactor coolant pumps running, obtain data for correlating nuclear flux level and loop temperatures with power Trip all reactor coolant pumps. Maintain core power at approximately 3 percent by control rod motion while cold leg temperatures remain relatively constant.
Verify natural circulation by observing the response of the hot leg temperature in each loop.
The plant is stable under natural circulation at this power level when hot leg temperature is constant.
Obtain data characterizing the plant under natural circulation conditions Restart reactor coolant pumps only after the reactor is shut down and isothermal conditions are re-established formance Criterion measured average vessel T under natural circulation conditions is equal to or less than limiting ign predictions for the measured reactor power level as specified in the applicable design cifications.
2.10.3.7 Passive Residual Heat Removal Heat Exchanger (First Plant Only) ective monstrate the heat removal capability of the passive residual heat removal heat exchanger with reactor coolant system at prototypic temperatures and natural circulation conditions.]* Note that test is performed in conjunction with the reactor coolant system natural circulation test with heat oval via the steam generators described in Subsection 14.2.10.3.6.
requisites described in Subsection 14.2.10.3.6, the following prerequisites have been met in preparation for natural circulation test with heat removal via the steam generators:
The reactor is critical and the neutron flux level is within the range for low power physics testing.
The neutron flux level and reactor coolant system boron concentration and temperature are stable, and the controlling rod bank is positioned in such a way that an increase in core power level to approximately 5 percent can be achieved by rod motion only.
Reactor coolant pumps are running.
Staff approval is required prior to implementing a change in this information.
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Instrumentation and data collection equipment is operational and available for logging plant data.
Special instrumentation is available to measure the reactor vessel T with high precision at low power levels.
The passive residual heat removal heat exchanger inlet and outlet temperature instrumentation and heat exchanger flow instrumentation are calibrated and operational.
The passive residual heat exchanger inlet isolation valve is operational and in its open position, and the heat exchanger outlet isolation valves are operational and in their closed position.
The startup feedwater system and controls are operating properly to maintain the steam generator secondary side water levels.
The steam generator steam dump system is operating properly to maintain steam generator pressure so that the reactor coolant system cold leg fluid is at its expected temperature.
The chemical volume control system auxiliary spray and letdown flow path are operable for controlling the pressurizer pressure and level, respectively after the reactor coolant pumps are shutoff.
t Method e that the following test steps are to be performed at the conclusion of the natural circulation test heat removal via the steam generators.
Verify that the natural circulation test with core power being removed by dumping steam from the steam generators has been completed.
Initiate flow through the passive residual heat removal heat exchanger by slowly opening one of the two parallel heat exchanger outlet isolation valves until it is fully open.
The steam generator steam dump will automatically reduce heat removal by the steam generators in response to passive residual heat exchanger operation. Manual operation of the control rods may be required to maintain core power at approximately 3 percent.
Obtain heat exchanger flow and inlet/outlet temperature data to characterize the heat removal capability of the heat exchanger and heatup of the in-containment refueling water storage tank water with one of two parallel isolation valves open.
Close the open heat exchanger isolation valve to terminate the heat exchanger test. The steam generator steam dump should automatically maintain the reactor coolant system fluid average temperature constant. Note that operation of the passive residual heat exchanger should be terminated before the in-containment refueling water storage tank average water temperature exceeds 150°F.
Shutdown the reactor by inserting the control rods. Restart reactor coolant pumps only after the reactor is shutdown and isothermal conditions are re-established.]*
Staff approval is required prior to implementing a change in this information.
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in-containment refueling water temperatures.]*
2.10.4 Power Ascension Tests r low power testing is completed, testing is performed at specified elevated power levels to onstrate the facility operates in accordance with design during normal steady-state operations, to the extent practical, during and following anticipated transients. During power ascension, tests performed to obtain operational data and to demonstrate the operational capabilities of the plant.
2.10.4.1 Test Sequence ective ne the sequence of operations, beginning at approximately 5 percent rated thermal power, that stitutes the power ascension testing program.
requisite nt system conditions are established, as required, by the individual test instruction within this uence.
t Method sent the sequence of operations and tests, along with instructions, specific plant conditions, and procedures.
formance Criteria evant performance criteria are provided in each of the test procedures.
2.10.4.2 Incore Instrumentation System ectives Obtain data for incore thermocouple and flux maps at various power levels during ascension to full power determine flux distributions and verify proper core loading and fuel enrichments.
requisites Incore instrumentation system signal processing software is installed and operational For incore thermocouple and flux mapping, the plant is at various power levels greater than approximately 20 percent of rated thermal power t Method With the plant at approximate power levels of 25, 50, 75 and 100 percent of rated thermal power, obtain data from the incore instrumentation system and process to produce incore thermocouple and flux maps. (Actual power levels will be specified in the power ascension program test sequence.)
Use data from the incore maps to verify that core power distribution is consistent with design predictions and the limits imposed by the plant Technical Specifications, including detection of Staff approval is required prior to implementing a change in this information.
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formance Criteria Core power peaking factors derived from the incore data are consistent with design predictions and the limitations of the plant Technical Specifications Data required for calibration of other plant instrumentation are obtained 2.10.4.3 Nuclear Instrumentation System ective ablish and determine voltage settings, trip settings, operational settings, alarm settings, and rlap of channels on intermediate range and power range instrumentation from zero power to at or r full rated thermal power.
requisite nuclear instrumentation system is aligned according to the design requirements.
t Method Calibrate, test, and verify functions using permanently installed controls and adjustment mechanisms Set operational modes of the intermediate range and power range channels for their proper functions, in accordance with the test instructions formance Criteria The nuclear instrumentation system operates in accordance with the design basis functional requirements as discussed in Subsection 4.4.6.
The nuclear instrumentation system demonstrates an overlap of indication between the intermediate and power range instrumentation.
2.10.4.4 Setpoint Verification ective ing power ascension, document final values of instrumentation setpoints as modified by initial tup testing, operations, or reanalysis to serve as a basis for future plant operations.
requisites Initial alignment and calibration of plant instrumentation have been completed, and initial set points are installed per applicable design documentation Preoperational and startup testing of affected plant instrumentation has been completed, and test results are documented The results of the precritical verification of the instrument setpoints are completed and documented 14.2-117 Revision 1
During power ascension testing, readjust specific setpoints noted for readjustment on the data sheets if required. Record final setpoint values.
formance Criterion point changes based on initial startup testing and operations are documented for future reference.
2.10.4.5 Startup Adjustments of Reactor Control Systems ectives Determine the adequacy of the reactor coolant system programmed Tavg Obtain plant data during power ascension which would provide the basis for any required changes to the Tavg program requisites The reactor coolant system is at no-load operating temperature and pressure The reactor coolant system temperature is being controlled by the steam dump valves t Method Obtain system temperature and steam pressure data at steady-state conditions for zero rated thermal power and at hold points during power escalations At approximately 75 percent rated thermal power, modify the Tavg program as required to achieve design steam generator pressure at full power, based on extrapolation of the data to the full power condition.
Reevaluate the Tavg program as above at approximately 90 and 100 percent rated thermal power making modifications to the Tavg program as required.
formance Criterion reactor coolant system Tavg program is established such that steam generator pressure at the rated thermal power condition is within design functional requirements as discussed in tion 5.1.
2.10.4.6 Rod Cluster Control Assembly Out of Bank Measurements (First Plant Only) ectives Demonstrate the sensitivity of the incore and excore instrumentation system to rod cluster control assembly (RCCA) misalignments Demonstrate the design conservatism for predicted power distributions with a fully misaligned rod cluster control assembly Monitor the power distribution following the recovery of a misaligned rod cluster control assembly 14.2-118 Revision 1
The reactor power level, reactor coolant system boron concentration, and temperature are stable.
The control and shutdown banks are positioned as required for the specific measurement, near fully withdrawn for rod cluster control assembly insertion, and at their respective insertion limits for rod cluster control assembly withdrawal.
t Method For the rod cluster control assembly insertion, insert a group of selected rod cluster control assemblies, one at a time, first to the limit of misalignment specified in Subsection 15.0.5, then fully inserted, and finally restored to the bank position. Compensate for reactivity changes by dilution and boration as required.
For the rod cluster control assembly withdrawal, withdraw one or more selected rod cluster control assemblies, one at a time, to the fully withdrawn position. Compensate for reactivity changes by boration and dilution as required.
Record incore and excore instrumentation signals to determine their response and to determine the power distribution and power peaking factors prior to rod cluster control assembly misalignment, at partial misalignment, at full misalignment, and periodically after restoration to normal.
formance Criteria Measured power distributions and power peaking factors are within Technical Specification limits and are consistent with the predictions.
The sensitivity of the incore and excore instrumentation to rod cluster control assembly misalignment is demonstrated by examination of the power distribution and power peaking factors measured for each misalignment.
2.10.4.7 Axial Flux Difference Instrumentation Calibration ectives Calibrate the power range nuclear instrumentation signals used as axial flux difference (delta flux) input to the reactor protection system Calibrate instrumentation used to display and monitor axial flux difference requisites The reactor is at a power level greater than 50 percent of rated thermal power The incore instrumentation system is available for obtaining incore power distribution data A preliminary calibration of the axial flux difference indication instrumentation is completed t Method Using control rod movement, xenon redistribution, or a combination of both, vary the axial power distribution of the core over a specified range of interest. At selected values of indicated axial flux difference, obtain reactor thermal power data along with the outputs from 14.2-119 Revision 1
Calibrate signals from the nuclear instrumentation power range channels based on incore power distribution and thermal power data.
formance Criterion l flux difference signals, derived from the nuclear instrumentation power range detectors and t to the reactor protection system, display, and monitoring instrumentation, reflect actual incore er distribution within specified limits, as discussed in Subsection 7.7.1.1.
2.10.4.8 Primary and Secondary Chemistry ective fy proper water quality in the reactor coolant system and secondary coolant system.
requisite plant is at the steady-state condition at approximately 0, 25, 50, 75, and 100 percent rated mal power.
t Method lyze samples to determine the chemical and radiochemical concentrations.
formance Criterion chemical and radiochemical control systems maintain the water chemistry within the applicable elines as discussed in Subsections 5.2.3.2 and 10.3.5.
2.10.4.9 Process Measurement Accuracy Verification ectives Measure the temperature variation in the reactor coolant loops resulting from non-uniform flow effects such as streaming Measure the sensitivity of the excore detectors to variations in control bank position and reactor coolant loop cold leg temperature requisites For the reactor coolant loop temperature measurements:
- Special temperature measuring equipment, including recording and indicating instrumentation, is installed, as required, on the reactor coolant loops hot and cold leg piping
- The reactor is at a stable power level of approximately 0, 50, 75 and 100 percent of rated thermal power For the excore detector measurements:
- The reactor is at a stable power level of approximately 25, 50 and 100 percent of rated thermal power 14.2-120 Revision 1
- Measure reactor power level, using calorimetric data
- Simultaneously, measure the hot and cold leg temperatures, using normal plant instrumentation and any other required instrumentation For the excore detector tests, with the reactor at constant power level:
- Measure the response of the excore detectors as selected control banks are moved over prescribed ranges of travel
- Measure excore detector response as the reactor coolant cold leg temperature is varied over a prescribed range
- Simultaneously, for each of the preceding measurements, obtain calorimetric data to verify reactor power level formance Criteria Uncertainties in reactor coolant loop temperature measurements resulting from non-uniform flow effects such as streaming are consistent with allowances used in the plant safety analyses.
Uncertainties in excore detector response resulting from control rod motion and reactor coolant loop cold leg temperature changes are consistent with allowances used in the plant safety analyses.
2.10.4.10 Process Instrumentation Alignment at Power Conditions ective n T and Tavg process instrumentation at power conditions.
requisites Reactor coolant pumps are operating.
The reactor system is operating at the required power level.
t Method Align T and Tavg according to test instructions at approximately 75 percent rated thermal power. Extrapolate the 75 percent data to determine T and Tavg values for the 100 percent plateau.
At or near 100 percent rated thermal power, check the alignment of the T and Tavg channels for agreement with the results of the thermal power measurement.
formance Criterion indicated values for reactor coolant system Thot, Tcold, Tavg, and T at or near full thermal power within the limits of the applicable design requirements, as discussed in Section 5.1.
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ower, verify that the reactor coolant flow equals or exceeds the minimum value required by the t Technical Specifications.
requisites The reactor is at power levels greater than 75 percent and up to and including 100 percent of rated thermal power Special instrumentation required for measuring reactor thermal power and reactor coolant inlet and outlet temperatures is installed and calibrated t Method h the reactor at steady-state power greater than 75 percent and up to and including 100 percent ted thermal power, measure the reactor thermal power and coolant inlet and outlet temperatures.
ermine the reactor coolant flow rate using the data in conjunction with hydraulic analysis of rential pressures at different locations in the reactor coolant system.
formance Criterion reactor coolant system flow determined from the measurements at approximately 100 percent d thermal power equals or exceeds the minimum value required by the plant Technical cifications.
2.10.4.12 Steam Dump Control System ective fy automatic operation of the Tavg steam dump control system, demonstrate controller setpoint quacy, and obtain final settings from steam pressure control of the condenser dump valves.
requisites Steam dump control system is aligned and calibrated to initial settings Plant is at no-load temperature and pressure Condenser vacuum is established Reactor is critical t Method Increase reactor power to less than 10 percent rated thermal power by rod withdrawal and steam dump to condenser to demonstrate setpoint adequacy Increase pressure controller setpoint prior to switching to Tavg control, which rapidly modulates open condenser dump valves Simulate turbine operating conditions with reactor at power, then simulate a turbine trip resulting in the rapid opening of the steam dump valves formance Criteria The plant trip controller responds to maintain a stable Tavg. After steady-state power is achieved, no divergent oscillations in temperature occur 14.2-122 Revision 1
The steam header pressure controller responds to maintain a stable pressure at normal no-load pressure 2.10.4.13 Steam Generator Level Control System ective fy the stability of the automatic steam generator level control system by introducing simulated sients at various power levels during escalation to full power.
requisites The reactor is critical and stable at various power levels during the power escalation test program. (Typical power levels are 30, 75 and 90 percent of full rated thermal power)
The steam generator level control system is checked and calibrated Steam generator alarm setpoints are set for each generator t Method At each power level, with the steam generator control system in manual mode, simulate level transients by changing the level setpoint. Verify the steam generator level control response when the control system is returned to automatic control.
Verify the variable speed features of the main feedwater pumps by manipulating controllers and test input signals.
formance Criteria During recovery from a simulated steam generator level transient, steam generator level control response is consistent with the design for the following: overshoot or undershoot to the new level, time required to achieve the new level, and error between the actual level and control setpoint.
Feedwater pump discharge pressure oscillations are less than design test limits The main feedwater control valves open and stabilize in response to various steam flow conditions in accordance with design requirements discussed in Subsection 7.7.1.8.
2.10.4.14 Radiation and Effluent Monitoring System ectives For monitors that:
- Are used for establishing conformance within the safety limits or limiting conditions for operation that are included in the Technical Specifications, or
- Are classified as engineered safety features, or are relied on to support operation of the engineered safety features within design limits, or 14.2-123 Revision 1
- Are used to process, store, control, or limit the release of radioactive materials The objectives are:
- Verify the calibration of the process and effluent radiation monitor against an acceptable standard
- Establish baseline activity and background levels
- Demonstrate that process and effluent radiation monitoring systems respond correctly by performing independent analyses requisites The plant is stable at the desired power level The sampling systems for the process and effluent radiation monitoring systems are operable t Method Perform calibrations with the use of radioactive sources to verify proper operation of the monitors and detectors Collect and analyze samples with laboratory instruments, and compare the results from the process and effluent monitor to verify proper monitor operation Establish background levels at low power (less than 5 percent rated thermal power)
Establish background levels and baseline activity levels determined by sampling at 100 percent rated thermal power to monitor the buildup of activity formance Criteria Radiation monitors are calibrated against radioactive standards Baseline activities are established Laboratory analyses agree, given sensitivity and energy response, with the process and effluent radiation monitors 2.10.4.15 Ventilation Capability ective fy that heating, ventilation, and air conditioning systems for the containment and areas housing ineered safety features continue to maintain design temperatures.
requisite plant is operating at or near the desired power (0, 50, and 100 percent of rated power).
t Method Record temperature readings in specified areas while operating with normal ventilation lineups 14.2-124 Revision 1
Record surface concrete temperatures adjacent to the high temperature piping penetrations and at selected locations on the concrete shielding (at 100 percent rated thermal power only) formance Criterion heating, ventilation and air conditioning systems for the containment and areas housing ineered safeguards features perform as designed in accordance with Subsections 9.4.1 9.4.6.
2.10.4.16 Biological Shield Survey ectives Document the radiation levels in accessible locations of the plant outside of the biological shield while at power Obtain baseline radiation levels for comparison with future measurements of level buildup with operation requisites Radiation survey instruments are calibrated Background radiation levels are measured in designated locations prior to initial criticality The plant is stable at the applicable power level t Method sure gamma and neutron radiation dose rates at designated locations at approximately 25, 50, and 100 percent rated thermal power.
formance Criterion iation levels are acceptable for full-power operation and consistent with design expectations.
2.10.4.17 Thermal Power Measurement and Statepoint Data Collection ective ain thermal power measurement and statepoint data at selected power levels during the power ension testing program, typically at 25, 50, 75, and 100 percent of rated thermal power.
requisites The following equipment is installed and is operational: sensors for measuring steam generator feedwater temperature, differential pressure measuring devices for determining feedwater flow to each steam generator, and pressure gauges to measure steam pressure at steam generator outlets.
The pressurizer pressure and level control system, and the steam generator level control system are in automatic mode.
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Reactor power is stable at the required level.
t Method required data are obtained using installed plant equipment, special test equipment, and the plant processing equipment. These data are subsequently used to determine reactor thermal power assess the performance of the plant.
formance Criterion ctor thermal power is stable at each power level and at the rated level at full power conditions.
rability of the pressurizer pressure and level control systems not previously verified as part of tor coolant system preoperational testing (Subsection 14.2.9.1.1) is demonstrated.
2.10.4.18 Dynamic Response ectives onstrate during power range testing that the stress analysis for selected systems and ponents, under transient conditions is within design functional requirements. Portions of systems meet the selection criteria for Subsection 14.2.9.1.7 for dynamic effects testing, but were not ed because system conditions during hot functional testing are not conducive to prototypical ems conditions, are tested.
requisites Temporary instrumentation is installed, as required, to monitor the deflections of components under test and the occurrence of water hammer noise and vibration.
Points are monitored and baseline data are established.
t Method Record deflection measurements during various plant transients.
Monitor for the occurrence of water hammer noise and vibration.
formance Criteria The movements due to flow-induced loads do not exceed the stress analysis of the monitored points. See Subsection 3.9.2.1.1 for the acceptable standard for alternating stress intensity due to vibration.
Flow-induced movements and loads do not cause malfunctions of plant equipment or instrumentation.
No effects due to water hammer are detected.
2.10.4.19 Reactor Power Control System ective onstrate the capability of the reactor power control system to respond to input signals.
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Setpoints and controls for the pressurizer, steam generator steam dump, and feedwater pump are checked and are set to proper values.
t Method y Tavg from the Tref setpoint to verify the transient recovery capabilities of the automatic reactor er control system.
formance Criterion returns to the Tref setpoint, within pre-specified limits and without manual intervention.
2.10.4.20 Load Swing Test ective fy nuclear plant transient response, including automatic control system performance, when ercent step-load changes are introduced to the turbine-generator at 30, 75, and 100 percent d thermal power levels.
requisite plant is operating in a steady-state condition at the desired thermal power level.
t Method nge the turbine-generator output as rapidly as possible to achieve a step 10 percent load ease or decrease. Monitor and record plant parameters of reactor power, reactor coolant system perature, pressurizer pressure and level, and steam generator pressure and level during the load sients. Core power should not exceed 100-percent power as indicated by the excore nuclear rumentation.
formance Criterion primary and secondary control systems, with no manual intervention, maintain reactor power, tor coolant system temperatures, pressurizer pressure and level, and steam generator levels pressures within acceptable ranges during and following the transient. Control system response viewed and compared to the control system setpoint and performance analysis, and adjustments e control systems are made, if necessary, prior to proceeding to the next power plateau.
2.10.4.21 100 Percent Load Rejection ective onstrate the ability of the AP1000 plant to accept a 100 percent load rejection from full power.
requisites The plant is operating at a stable power level of approximately 100 percent rated thermal power. Reactor and turbine control systems are in the automatic mode of operation. Plant temperatures, pressures, levels, and flow rates are within their normal range for full-power operation.
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The incore instrumentation system, including signal processing software, is operational, and all preoperational and startup testing is completed.
Instrumentation and data collection equipment is operational and available for logging plant data.
Special test instrumentation is installed and operational as required to augment normal data logging ability.
t Method With the plant at nominal full-power steady-state conditions, to effect a rejection of 100 percent load, manually place the main step-up transformer high side breaker in the trip position.
Prior to the load rejection, and until the plant stabilizes at the lower power level, record key plant parameters using the plant computer and special test instrumentation. The key plant parameters include plant temperatures, pressures, levels and flow rates for the primary and secondary systems.
formance Criteria The plant is capable of accepting a 100 percent load rejection from full rated thermal power without reactor trip or operation of the steam generator relief valves or pressurizer safety valves.
The turbine speed does not exceed 108% of rated speed.
The turbine is capable of continued stable operation at the minimum house loads.
2.10.4.22 Load Follow Demonstration (First Plant Only) ective Demonstrate the ability of the AP1000 plant to follow a design basis daily load follow cycle.
Demonstrate the ability of the plant to respond to grid frequency changes while in the load follow cycle.
requisites The plant is operating at a stable power level of approximately 100 percent power and has been at that power for a sufficient length of time to have reached an equilibrium xenon condition.
Startup testing of the reactor and turbine control and protection systems are completed, and final setpoints are installed.
The incore instrumentation system, including signal processing software, is operational. All preoperational and startup testing is completed.
Instrumentation and data collection equipment is operational and available for logging plant data.
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Using normal plant procedures, reduce turbine load at a rate such that a reactor thermal power level of approximately 50 percent is achieved linearly in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
After remaining at 50 percent rated thermal power for more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> but less than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, increase turbine load at a rate such that a reactor power level of approximately 100 percent rated thermal power is achieved linearly in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
At selected times during the power decrease, while at reduced power, during the power increase, and after reaching approximately full rated thermal power, obtain data from both incore and excore instrumentation to monitor plant performance.
While within the load-follow maneuver, demonstrate the ability to respond to grid frequency changes by increasing and decreasing load by as much as 10 percent, at a rate of 2 percent per minute.
formance Criteria Core power distribution limits, as specified in the plant Technical Specifications, are not exceeded when the plant power is varied according to the design basis load-follow cycle, or while in the cycle, responding to load changes simulating grid frequency changes.
Load follow maneuvers, including response to grid frequency changes, can be accomplished without changes to the reactor coolant boron concentration.
2.10.4.23 Hot Full Power Boron Endpoint ective sure the reactor coolant system critical boron concentration at beginning of cycle life for the all out, hot full power, xenon equilibrium condition.
requisites The reactor is operating at approximately 100 percent of full licensed power and has been at that power for a sufficient time to reach xenon equilibrium.
The reactor power level and reactor coolant system boron concentration and temperature are stable, and control and shutdown rod banks are in the near fully withdrawn position.
Current core burnup data are available.
t Method During the power ascension test program, and, as soon as practicable after achieving xenon equilibrium at full licensed power, obtain and analyze samples of reactor coolant for dissolved boron content.
Using plant calorimetric and statepoint data obtained at the same time as coolant sampling, correct the measured boron concentration, as required, for control rod insertion, xenon nonequilibrium, and any difference between Tavg and Tref.
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As permitted by the plant Technical Specifications, use the corrected measured boron concentration to renormalize the predicted curve of boron concentration as a function of core burnup.
formance Criterion reactivity equivalent of the difference between measured and predicted boron concentrations le 4.3-2) is less than the design limit shown in Subsection 4.3.3.3.
2.10.4.24 Plant Trip from 100 Percent Power ectives Verify the ability of the plant automatic control systems to sustain a trip from 100 percent rated thermal power and bring the plant to stable conditions following the transient.
Assess the dynamic response of the plant for the event that subjects the turbine to its maximum credible overspeed condition.
Determine the overall response time of the hot leg resistance temperature detector.
Optimize the control systems setpoints, if necessary.
requisite plant is operating in a steady-state condition at full rated thermal power.
t Method Trip the plant by opening the main generator breaker.
Monitor and record selected plant parameters.
If necessary, adjust the control systems setpoints to obtain optimal response.
formance Criteria Following the opening of the main generator breaker while at 100 percent rated thermal power, primary and secondary control systems and operator actions can stabilize reactor coolant system temperature, pressurizer pressure and level, and steam generator levels to no-load operating temperature and pressure.
The steam dump control system operates to prevent opening of primary and secondary safety valves.
The hot leg resistance temperature detector (RTD) time responses are verified to be less than or equal to values used in the safety analysis.
The turbine speed does not exceed 108% of rated speed.
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onstrate that essential nuclear steam supply system and balance-of-plant components can and without obstruction and that the expansion is in accordance with design. Also, during ldown, the components return to their approximate baseline cold position. Testing is conducted to lve discrepancies from hot functional testing as in Subsection 14.2.9.1.1, and to test ifications made since hot functional testing was completed. Portions of systems that meet the ction criteria for Subsection 14.2.9.1.7 for thermal dynamic testing, but were not tested because em conditions during hot functional testing are not conducive to prototypical system conditions tested.
requisite porary instrumentation is installed, as required, to monitor the deflections for the components er test.
t Method the components tested, the following apply:
During plant heatup and cooldown, record deflection data.
Verify support movements by recording hot and cold positions.
formance Criteria rmal expansion testing is performed in accordance with ASME OM Standard, Part 7 as discussed ubsection 3.9.2.1.2. For the components tested, the following apply:
There is no evidence of blocking of the thermal expansion of piping or component, other than by installed supports, restraints, and hangers.
Spring hanger movements must remain within the hot and cold setpoints and supports must not become fully retracted or extended.
Piping and components return to their approximate baseline cold position.
2.10.4.26 Loss of Offsite Power ective onstrate plant response following a plant trip with no offsite power available.
requisites The plant is at minimum power level supplying normal house loads through the unit auxiliary transformers.
The unit is disconnected from the electrical grid.
t Method The turbine is tripped and the generator output breaker opens, removing ac power from the unit auxiliary transformers.
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Both standby diesel generators start and pick up the required loads in the proper sequence.
Class 1E dc and non-1E dc loads are uninterrupted and are provided by the battery subsystems.
The primary plant is placed in a stable condition.
2.10.4.27 Feedwater Heater Loss and Out of Service Test ective onstrate the plant response to the loss of one of the feedwater heaters during power operation to single failure or operator error. Demonstrate the plant response to a pair of feedwater heaters n out of service during power operation. Verify the ability of operators to manually reduce steam and place a pair of feedwater heaters out of service while maintaining reactor power operation.
requisites plant is operating in a steady-state condition at the rated thermal powers described.
t Method S OF FEEDWATER HEATER With the plant operating at 50% power, isolate the extraction steam supply to one of the main feedwater heaters.
With the plant operating at 90% power, isolate the extraction steam supply to one of the main feedwater heaters.
DWATER OUT OF SERVICE TEST The operators calculate the appropriate steam flow reduction which will maintain the plant at the desired thermal load after the heaters have been taken out of service.
Reduce steam flow by the appropriate amount and allow plant conditions to reach a new steady-state (approximately 10 minutes).
Take a pair of feedwater heaters out of service.
formance Criteria plant control systems properly respond to the loss of a main feedwater heater, without reactor or ine trip.
operator successfully removes a pair of feedwater heaters from service without causing a tor trip.
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onstrate the ability of the operators to conduct a remote shutdown of the plant during a ulated main control room evacuation.
requisites roved operation procedures for performing a remote shutdown is available. Communication ts between the control room and the remote shutdown room. Procedures for transferring control k to the main control room are available if an emergency or unsafe condition develops during the ing that cannot be managed by the shutdown crew.
plant is operating in a steady-state condition at 10-20 percent of power.
t Method Using the appropriate operating procedures, the operators transfer control of the plant from the main control room to the remote shutdown workstation.
From the remote shutdown workstation, the operators bring the plant to hot standby, and maintain hot standby conditions for at least 30 minutes.
From the remote shutdown workstation, the operators lower the reactor coolant system pressure and temperature to the appropriate conditions, and place the normal residual heat removal system into service. The normal residual heat removal system, in conjunction with the component cooling water system and service water system are used to cool the plant at least 50°F without exceeding prescribed cooldown limits.
formance Criteria operators successfully demonstrate the ability transfer control of the plant to the remote tdown workstation, shut down the reactor, maintain hot standby, and then demonstrate the ability ansition to cold shutdown conditions, while performing these operations from the remote tdown workstation.
2.10.4.29 Cooling Tower(s) ectives Verify proper cooling tower(s) function. Provide thermal acceptance testing of the cooling towers heat removal capabilities.
requisites The cooling tower(s) is structurally complete and in good operating condition.
Circulating water system testing is complete.
Required support systems, electrical power supplies, and control circuits are operational.
t Method rmal performance of the cooling tower(s) is tested and verified using established industry test dards.
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14.2-134 Revision 1 ign Material (CDM). This document provides the principal design bases and design racteristics that are certified by the 10 CFR Part 52 rulemaking process and included in the ign certification rule.
top-level design information in the Certified Design Material is extracted directly from the 000 design information. Limiting the certified design contents to top-level information reflects the d approach to design certification endorsed by the U.S. Nuclear Regulatory Commission (see erences 1 through 5).
objective of this section is to define the bases and methods that were used to develop the tified Design Material for the AP1000. This section contains no new technical information arding the AP1000 design.
AP1000 Certified Design Material consists of the following:
An introduction section which defines terms used in the Certified Design Material and lists general provisions that are applicable to all Certified Design Material entries. Also included is a list of acronyms and legends used in the Certified Design Material. (Because this material is self-explanatory, it is not discussed in this section.)
Design descriptions for selected systems that are within the scope of the AP1000 design certification, and the applicable portions of those selected systems that are only partially within the scope of the AP1000 design certification. The Certified Design Material design descriptions delineate the principal design bases and principal design characteristics that are referenced in the design certification rule. The design descriptions are accompanied by the inspections, tests, analyses, and acceptance criteria (ITAAC) required by 10 CFR 52.47(a)(1)(vi) to be part of the design certification application. The ITAAC define verification activities that are to be performed for a facility with the objective of confirming that the plant is built and will operate in accordance with the design certification. Completion of these certified design ITAAC, together with the Combined License applicants ITAAC for the site-specific portions of the plant, will be the basis for NRC authorization to load fuel per the provisions of 10 CFR Part 52.103.
Design descriptions and their associated ITAAC for design and construction activities that are applicable to more than one system. Design-related processes have been included in the Certified Design Material for:
- Aspects of the AP1000 design likely to undergo rapid, beneficial technological developments in the lifetime of the design certification. Certifying the design processes associated with these areas of the design, rather than specific design details, permits future license applicants referencing the AP1000 design certification to take advantage of the best technology available at the time of combined license application and facility construction.
- Aspects of the design dependant upon characteristics of as-procured, as-installed systems, structures, and components. These characteristics are not available at the time of certification and, therefore, cannot be used to develop and certify design details.
- Aspects of the seismic, structural and piping design for which detailed design has not been developed. These details are not available at the time of certification and, therefore, cannot be used to certify design details. Certifying the design processes associated with these design details provides the basis for future license applicants referencing the AP1000 design 14.3-1 Revision 1
Interface requirements as defined by 10 CFR Part 52.47(a)(1)(vii). Interface requirements are defined as those which must be met by the site-specific portions of the complete nuclear power plant that are not within the scope of the certified design. These requirements define characteristics of the site-specific features that must be provided for the certified design to comply with certification commitments. AP1000 has no interfaces meeting this definition. The Certified Design Material does not include ITAAC or a requirement for COL developed ITAAC for interface requirements.
Site parameters used as the basis for AP1000 design presented in the Tier 2 Material. These parameters represent a bounding envelope of site conditions for any license application referencing the AP1000 design certification. No ITAAC are necessary for the site parameters entries because compliance with site parameters will be verified as part of issuance of a license for a plant that references the AP1000 design certification.
following is a description of the criteria and methods used to select specific technical entries for Certified Design Material. The structure of the description is based on the Certified Design erial report structure.
criteria and methods discussed in the following sections are guidelines only. For some matters, contents of the Certified Design Material may not directly correspond to these guidelines because cial considerations related to the matters may warrant a different approach. For such matters, a e-by-case determination is made regarding how or whether the matters should be addressed in Certified Design Material. These determinations are based upon the principles inherent in 52.
3.1 CDM Section 1.0, Introduction section provides definitions, general provisions, a figure legend, and a list of acronyms used in AP1000 Certified Design Material.
ection Criteria - Section 1.1 is used to define terms used throughout the Certified Design Material.
ection of entries is based on a judgment that a particular word/phrase merits definition - with icular emphasis on terms associated with implementation of the ITAAC. Section 1.2 contains a ure of provisions that is selected on the basis that the provision is necessary to either define nical requirements applicable to multiple systems in the Certified Design Material or to provide ification and guidance for future users of the Certified Design Material.
ection Methodology - Entries in the Definition section are made on the basis of a self-evident d for a term to be defined. These terms are accumulated during the preparation and review of the tified Design Material. Entries in the General Provisions section also are developed as part of the tified Design Material selection and review process. Each entry has a unique background, but the rall intent is to state the broad guidelines and interpretations that are used to prepare Certified ign Material for the AP1000.
3.2 CDM Section 2.0, System Based Design Descriptions and ITAAC section of the Certified Design Material has the design description and ITAAC material for the cted AP1000 systems. The intent of this list of AP1000 systems is to define at the Certified ign Material level the full scope of the certified design.
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performance standards that pertain to the safety of the plant and include descriptive text and porting figures. The intent of the Certified Design Material design descriptions is to define the 000 design characteristics referenced in the design certification rule as a result of the certification isions of 10 CFR Part 52.
ection Criteria - The following criteria are considered in determining the information included in certified design descriptions:
The information in the certified design descriptions is selected from the technical information presented in the Tier 2 Material. This reflects the approach that the Certified Design Material contains top-level design information and is based on the NRC directive in Reference 2 that there be less detail in a certification than in an application for certification. In this context, the certification is the Certified Design Material and the application for certification includes the Tier 2 Material.
The certified design descriptions contain only the information from the Tier 2 Material that is most significant to safety. The Tier 2 Material contains a wide spectrum of information on various aspects of the AP1000 design. Not all of this information is included in the certified design descriptions. This selection criterion reflects the NRC directive in Reference 2 that the certified design should encompass roughly the same design features that Section 50.59 prohibits changing without prior NRC approval. In determining those structures, systems, or components for which certified design descriptions and ITAAC must be prepared, the following questions are considered for each structure, system, or component:
- Are there any features or functions classified as Class A, B, or C?
- Are there any defense-in-depth features or functions provided?
- For nonsafety-related systems, are there any features or functions credited for mitigation of design basis events?
- For nonsafety-related systems, are there any features or functions that have been identified in Section 16.3 as candidates for additional regulatory oversight?
If the answer to the first question is yes, then a certified design description and ITAAC are prepared using the safety function stated in the Tier 2 Material and the parameters from the safety analysis.
If the answer to either of the next two questions is yes, then a certified design description and ITAAC are prepared using the functions stated in the Tier 2 Material and the parameters from the system design calculations.
If the answer to the last question is yes and the feature or function is not a programmatic requirement related to operations, maintenance or other programs, then a certified design description and ITAAC are prepared using the functions stated in the Tier 2 Material and the parameters from system design calculations.
In addition, the following questions were considered for each structure, system, or component not already selected for ITAAC using the above selection criteria:
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- Are there any features or functions that represent an important assumption for probabilistic risk assessment?
- Are any features or functions important in preventing or mitigating severe accidents?
- Are there any features or functions that have a significant impact on the safety and operation of the plant?
- Are any features or functions the subject of a provision in the Technical Specifications?
If the answer to any of the above questions is yes, then a design description and ITAAC are prepared using the appropriate functions stated in the Tier 2 material and the parameters from the system design calculations.
A summary of the AP1000 structures, systems, or components considered for selection is given in Table 14.3-1.
In general, safety-related and defense-in-depth features and functions of structures, systems, and components are discussed in the certified design descriptions. Structures, systems, and components that are not classified as safety-related or defense-in-depth are discussed in the certified design descriptions to the extent that they have features or functions that mitigate a design basis event.
The certified design descriptions for structures, systems, and components are limited to a discussion of design features and functions. The design bases of structures, systems, and components, and explanations of their importance to safety, are provided in the Tier 2 Material and are not included in the certified design descriptions. The Certified Design Material design descriptions define the certified design. Justification that the design meets regulatory requirements is presented in the Tier 2 Material.
The certified design descriptions focus on the physical characteristics of the facility. The certified design descriptions do not contain programmatic requirements related to operating conditions or to operations, maintenance, or other programs. These matters are controlled by other means such as the technical specifications.
The certified design descriptions in Section 2.0 of the Certified Design Material discuss the functional arrangement and performance characteristics that the structures, systems, and components should have after construction is completed. In general, the certified design descriptions do not address the processes that will be used for designing and constructing a plant that references the AP1000 design certification. This is acceptable because the safety-function of a structure, system, or component is dependent upon its final as-built condition and not the processes used to achieve that condition. Exceptions to this criterion are the selected design and qualification processes defined in the instrumentation and control portions and piping portions of Section 2 and the piping, seismic, structural and human factors portion of Section 3.
The programmatic aspects of the design and construction processes (training, qualification of welders, and the like) are part of the licensees programs and are subject to commitments made at the time of combined license issuance. Consequently, these issues are not addressed in the AP1000 Certified Design Material.
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The certified AP1000 design descriptions do not discuss component types (for example, valve and instrument types), component internals, or component manufacturers. This approach is based on the premise that the safety function of a particular design element can be performed by a variety of component types from different manufacturers.
The certified design descriptions do not contain proprietary information.
For the applicant or licensee of a plant that references the AP1000 design certification to take advantage of improvements in technology, the certified design descriptions in general do not prescribe design features that are the subject of rapidly evolving technology.
The Certified Design Material design description is intended to be self-contained and does not make direct reference to the Tier 2 Material, industrial standards, regulatory requirements, or other documents. (There are some exceptions involving the ASME Code and the Code of Federal Regulations.) If these sources contain technical information of sufficient safety significance to warrant Certified Design Material treatment, the information is extracted from the source and included directly in the appropriate system design description.
This approach is appropriate because it is unambiguous and it avoids potential questions regarding how much of a referenced document is encompassed in, and becomes part of, the Certified Design Material.
Selection of the technical terminology to be used in the Certified Design Material is guided by the principle that the terminology should be as consistent as possible with that used in the Tier 2 Material and the body of regulatory requirements and industrial standards applicable to the nuclear industry. This approach is intended to minimize problems in interpreting Certified Design Material commitments.
view of those sections of the AP1000 Tier 2 Material that document plant safety evaluations was ducted. Specifically, reviews were conducted of the following chapters of the AP1000 Tier 2 erial; the flooding analysis in Chapter 5, the analysis of overpressure protection in Chapter 5, tainment analysis in Chapter 6, the core cooling analysis in Chapters 6 and 15, the analysis of fire ection in Chapter 9, the safety analysis of transients in Chapter 15, the analysis of anticipated sients without scram (ATWS) in Chapters 7 and 15, the radiological analysis in Chapter 15, the lution of unresolved or generic safety issues and Three Mile Island issues in Chapter 1, and the A and severe accident information in Chapter 19. These reviews were important in identifying ty-related system design information warranting consideration in the design descriptions and the ompanying design commitments.
ection Methodology - The Certified Design Material uses a system report structure. The ified design description entry for any system is based on review of the multiple sources having nical information related to that system. Using the selection criteria listed, design description erial is developed for each system by reviewing the Tier 2 Material, safety analysis, test grams, and design documents relating to that system.
lication of the criteria listed results in a graded treatment of the systems. This leads to variation in scope of the design description entries. The following lists the types of AP1000 systems and is a mary of this graded treatment:
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is accidents tems with defense-in-depth functions that Major defense-in-depth features and tribute to plant performance during design performance characteristics is accidents nsafety-related systems potentially Brief discussion of design features that acting safety prevent or mitigate the potential safety concern nsafety-related systems with no No discussion tionship to safety safety-related systems, application of this criteria results in design description entries that include following information, as applicable:
System name and scope System purpose Summary of the systems safety-significant components (usually shown by a figure)
Equipment seismic and ASME classifications Piping ASME classification and Leak-Before-Break criteria Type of electrical power provided for the system Systems important instruments, controls, and alarms to the extent located in the main control room or remote shutdown workstation Equipment to be qualified for harsh environments Motor-operated valves within the system that have an active safety-related function Other features or functions that are significant to safety certified design descriptions for nonsafety-related systems include the information listed to the nt that the information is relevant to the system and is significant to safety. Since much of this rmation is not relevant to nonsafety-related systems, the certified design descriptions for safety-related systems are less extensive than the descriptions for safety-related systems.
3.2.2 Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) ble of ITAAC entries is provided for each system that has design description entries. The intent of e ITAAC is to define activities that will be undertaken to verify the as-built system conforms with design features and characteristics defined in the design description. ITAAC are provided in es with the following three-column format:
Design Commitment Inspections, Tests, Analyses Acceptance Criteria 14.3-6 Revision 1
ign Acceptance Criteria (DAC)/ITAAC closure is outlined in Appendix 14A.
ection Criteria - The following are considered when determining what information is included in Certified Design Material ITAAC entries:
The scope and content of the ITAAC correspond to the scope and content of the certified design descriptions. There are no ITAAC for aspects of the design not addressed in the design description. This is appropriate because the objective of the ITAAC design certification entries is to verify that the as-built facility has the design features and performance characteristics defined in the Certified Design Material descriptions.
Each AP1000 system with a design description has an ITAAC table. This reflects the assessment that a design feature meriting a Certified Design Material description also merits an ITAAC entry to verify that the feature has been included in the as-built facility.
One inspection, test, or analysis may verify one or more provisions in the certified design description. An ITAAC that specifies a system functional test or an inspection may verify a number of provisions in the design description. There is not necessarily a one-to-one correspondence between the ITAAC and the design descriptions.
As required by 10 CFR 52.103, the inspections, tests, and analyses must be completed (and the acceptance criteria satisfied) prior to fuel loading. Therefore, the ITAAC do not include any inspections, tests, or analyses that are dependent upon conditions that exist only after fuel load.
Because the design descriptions are limited to fixed design features expected to be in place for the lifetime of the facility, the ITAAC are limited to a verification of fixtures in the plant. There are no ITAAC for nuclear fuel, fuel channels, and control rods because they are changed by a licensee.
The ITAAC verify the as-built configuration and performance characteristics of structures, systems, and components as identified in the Certified Design Material design descriptions.
ection Methodology - Using the selection criteria, ITAAC table entries are developed for each cted system. This is achieved by evaluating the design features and performance characteristics ned in the Certified Design Material design description and preparing an ITAAC table entry for the ign description criteria that satisfied the selection criteria. There is a close correlation between the hand column of the ITAAC table and the corresponding design description entries.
ITAAC table is completed by selecting the method to be used for verification (either a test, an ection, or an analysis [ITA]) and the acceptance criteria for the as-built feature.
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pection To be used when verification can be accomplished by visual observations, physical examinations, review of records based on visual observations, or physical examinations that compare the as-built structure, system, or component condition to one or more design description commitments.
t To be used when verification can be accomplished by the actuation or operation, or establishment of specified conditions, to evaluate the performance or integrity of the as-built structures, systems, or components.
The type of tests identified in the ITAAC tables includes activities such as factory testing, special test facility programs, and laboratory testing.
alysis To be used when verification can be accomplished by calculation, mathematical computation, or engineering or technical evaluations of the as-built structures, systems, or components.
proposed verification activity is identified in the middle column of the ITAAC table. Where ropriate, the Tier 2 Material provides details regarding implementation of the verification activity.
Tier 2 Material is not referenced in the Certified Design Material and is not part of the Certified ign Material; Tier 2 Material is considered as providing one of potentially several acceptable hods for completing the ITA.
ection of acceptance criteria is dependent upon the design characteristic being verified by the AC table entry: in most cases, the appropriate acceptance criteria is self-evident and is based n the Certified Design Material design description. For many of the AP1000 ITAAC, the eptance criteria is a statement that the as-built facility has the design feature or performance racteristic identified in the design description. A guiding principle for acceptance criteria paration is the recognition that the criteria should be objective and unambiguous. The use of ctive and unambiguous terms for the acceptance criteria will minimize opportunities for multiple, jective (and potentially conflicting) interpretations as to whether an acceptance criteria has, or has been met. In some cases, the ITAAC acceptance criteria contain numerical parameters from the 2 Material that are not specifically identified in the Certified Design Material design description or design commitment column of the ITAAC table. This is acceptable because the design cription defines the important design feature/performance that merits Certified Design Material tment. The acceptance criterion defines a measurement standard for determining if the as-built ity is in compliance with the Certified Design Material design description commitment. Where ropriate, the Tier 2 Material identifies criteria applicable to the same design feature or function is the subject of more general acceptance criteria in the ITAAC table.
numerical acceptance criteria, ranges and/or tolerances are included. This is necessary and eptable because of the following:
Specification of a single-value acceptance criteria is impractical because trivial deviations will represent unnecessary noncompliances.
Tolerances recognize that legitimate site variations can occur in complex construction projects.
Minor variations in plant parameters within the tolerance bounds have no impact on plant safety.
3.2.3 Site-Specific ITAAC (SS-ITAAC) ble of inspections, tests, analyses, and acceptance criteria (ITAAC) entries is provided for each
-specific system described in this FSAR that meets the selection criteria, and that is not included 14.3-8 Revision 1
sign Commitment Inspection, Tests, Acceptance Analyses Criteria h design commitment in the left-hand column of the ITAAC tables has associated inspections, s, or analyses (ITA) requirements specified in the middle column. The acceptance criteria for the are defined in the right-hand column.
ITAAC do not address ancillary buildings and structures on the site, such as administrative dings, parking lots, warehouses, training facilities, etc.
ection Criteria In determining those structures, systems, or components for which ITAAC must be prepared, the following questions are considered for each structure, system, or component:
- Are any features or functions classified as Class A, B, or C?
- Are any defense-in-depth features or functions provided?
- For nonsafety-related systems, are any features or functions credited for mitigation of design basis events?
- For nonsafety-related systems, are there any features or functions that have been identified in Section 16.3 as candidates for additional regulatory oversight?
If the answer to any of the above questions is yes, then ITAAC are prepared.
The scope and content of the ITAAC correspond to the scope and content of the site-specific system design description.
One ITA may verify one or more provisions in the system design description. An ITAAC that specifies a system functional test or an inspection may verify a number of provisions in the system design description. There is not necessarily a one-to-one correspondence between the ITAAC and the system design descriptions.
As required by 10 CFR 52.103, the ITA is completed (and the acceptance criteria satisfied) prior to initial fuel loading.
The ITAAC verify the as-built configuration and performance characteristics of structures, systems, and components as identified in the system design descriptions.
ection Methodology - Using the selection criteria, ITAAC table entries are developed for each cted system. This is achieved by evaluating the design features and performance characteristics ned in the system design descriptions and preparing an ITAAC table entry for each design cription criterion that satisfies the selection criteria. A close correlation exists between the left-d column of the ITAAC and the corresponding design description entries.
ITAAC table is completed by selecting the method to be used for verification (either a test, an ection, or an analysis) and the acceptance criteria for the as-built feature.
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3.2.3.1 Emergency Planning ITAAC (EP-ITAAC) ergency Planning ITAAC (EP-ITAAC) have been developed to address implementation of ments of the Emergency Plan. Site-specific EP-ITAAC are based on the generic ITAAC provided ppendix C.II.1-B of Regulatory Guide 1.206. These ITAAC have been tailored to the specific tor design and emergency planning program requirements.
3.2.3.2 Physical Security ITAAC (PS-ITAAC) eric Physical Security ITAAC (PS-ITAAC) have been developed in a coordinated effort between NRC and the Nuclear Energy Institute (NEI). These generic ITAAC have been tailored to the 000 design and site-specific security requirements.
3.2.3.3 Other Site-Specific Systems additional site-specific system has been determined to meet the ITAAC selection criteria, and AC have been included for the Transmission Switchyard and Offsite Power System (ZBS) as cated in Table 14.3-1. Systems not meeting the selection criteria are subject to the normal tional testing to verify that newly designed and installed systems, structures, or components orm as designed.
mmary of the AP1000 structures, systems, or components considered for selection is given in le 14.3-1.
3.3 CDM Section 3.0, Non-System Based Design Descriptions and ITAAC ies in this section of the Certified Design Material have the same structure as the system material ussed in Subsection 14.3.2; that is, design description text and figures and a table of ITAAC ies. The objective of this Certified Design Material is to address selected design and construction vities which are applicable to more than one system. There are six entries in Section 3.0 of the tified Design Material: nuclear island buildings, initial test program, emergency response facilities, an factors engineering, Design Reliability Assurance Program, and radiation protection.
3.3.1 Waterproof Membrane ITAAC design of the waterproof membrane beneath the nuclear island basemat is described in section 3.4.1.1.1.1. Waterproof Membrane ITAAC have been developed to address verification the mudmat-waterproofing interface beneath the nuclear island basemat has a minimum fficient of friction to resist sliding of 0.55.
3.3.2 Pipe Rupture Hazard Analysis ITAAC pe rupture hazard analysis is part of the piping design. The analyses will document that ctures, systems, and components (SSCs) which are required to be functional during and wing a design basis event have adequate highenergy and moderate-energy pipe break gation features. The locations of postulated ruptures and essential targets will be established and uired pipe whip restraint and jet shield designs will be included. The as-designed pipe rupture ards analysis will be based on the as-designed piping analysis and will be in accordance with the ria outlined in Subsections 3.6.1.3.2 and 3.6.2.5. The evaluation will address environmental and ding effects of cracks in high and moderate energy piping. The report of the pipe rupture hazard lysis shall conclude that, for each postulated piping failure, the systems, structures, and 14.3-10 Revision 1
as-built reconciliation of the pipe rupture hazards evaluation whip restraint and jet shield design ccordance with the criteria outlined in Subsections 3.6.1.3.2 and 3.6.2.5 are covered in as-built AC identified in DCD Tier 1 to demonstrate that the as-built pipe rupture hazards mitigation ures reflect the design, as reconciled. The reconciliation report will be made available for NRC ection or audit when it has been completed.
as-designed pipe rupture hazard analysis completed for the first standard AP1000 plant will be ilable to subsequent standard AP1000 plants under the one issue, one review, one position roach for closure.
3.3.3 Piping Design ITAAC piping design ITAAC consists of the piping analysis for safety-related ASME Code piping. The ng design is completed on a package-by-package basis for applicable systems. In order to port closure of the piping design ITAAC, information consisting of the as-designed piping analysis iping lines chosen to demonstrate all aspects of the piping design will be made available for NRC ew, inspection, and/or audit. This information will consist of a design report referencing the esigned piping calculation packages, including ASME Section III piping analysis, support luations and piping component fatigue analysis for Class I piping. The piping packages to be lyzed are identified in the DCD.
ASME Code prescribes certain procedures and requirements that are to be followed for pleting the piping design. The piping design ITAAC includes a verification of the ASME Code ign report to ensure that the appropriate code design requirements for each systems safety class e been implemented.
conciliation of the applicable safety-related as-built piping systems is covered in as-built ITAAC tified in DCD Tier 1 to demonstrate that the as-built piping reflects the design, as reconciled. The nciliation report will be made available for NRC inspection or audit when it has been completed.
piping design completed for the first standard AP1000 plant will be available to subsequent dard AP1000 plants under the one issue, one review, one position approach for closure.
3.4 Certified Design Material Section 4.0, Interface Requirements 000 is a plant design incorporating the nuclear island, the annex building and associated ipment, the diesel/generator building and associated equipment, the turbine/generator building, turbine/generator equipment, and the radwaste facilities. As a result, no interfaces need to be tified between or among these portions of the plant. There are no safety-related interfaces ween the AP1000 certified design and other portions of a facility with a combined license under CFR Part 52.
al testing of interfacing non-safety systems in portions of the plant outside the scope of design ification is as discussed in Section 14.4. Section 1.8, Table 1.8-1, lists the interfacing systems structures. Those systems that meet the requirements of 10 CFR 52.47(a)(1)(viii) are tabulated ubsection 14.4.5.
3.5 CDM Section 5.0, Site Parameters section of the Certified Design Material defines the site parameters used as a basis for the ign defined in the AP1000 certification application. These entries respond to the 14.3-11 Revision 1
ide additional analysis to show acceptability of deviations from the interface envelope.
-specific external events that relate to the acceptability of the design (and not to the acceptability e site) are not considered site parameters and are addressed as interface requirements in the ropriate system entry in Section 4 of the Certified Design Material.
tion 5.0 of the Certified Design Material does not include any ITAAC and is limited to defining the 000 site parameters. This is an appropriate approach because compliance of the site with these ameters is demonstrated by a license applicant prior to issuance of the license.
ection Criteria - Chapter 2, Table 2.0-201, provides the envelope of site design parameters used he AP1000 design. The corresponding Certified Design Material Section 5.0 is based on using le 2.0-201. Section 5.0 is limited to a tabular entry; no supporting text material is required.
3.6 Initial Test Program AP1000 Initial Test Program defines testing activities that will be conducted following completion onstruction and construction-related inspections and tests. The Initial Test Program extends ugh the start of commercial operation of the facility. This program is discussed in Chapter 14.
mmary of the Initial Test Program is included in Certified Design Material Section 3.4. This mary includes an overview of the Initial Test Program structure. This information is included in Certified Design Material because of the importance of the Initial Test Program defining pre- and t-fuel load testing for the as-built facility. Key pre-fuel load Initial Test Program testing for vidual systems is defined in the system ITAAC in Certified Design Material Sections 2 and 3.
TAAC entries have been included in the Certified Design Material for the Initial Test Program.
is acceptable because of the following:
The Initial Test Program activities involve testing with the reactor at various power levels and thus cannot be completed prior to fuel load (Part 52 requires ITAAC to be completed prior to fuel load).
Testing activities specified as part of the ITAAC in Certified Design Material Sections 2 and 3 must be performed prior to fuel load. Because these ITAAC testing activities address the design features and characteristics of safety significance, additional ITAAC for the Initial Test Program are not necessary to ensure that the as-built plant conforms with the certified design.
3.7 Elements of AP1000 Design Material Incorporated into the Certified Design Material les 14.3-2 through 14.3-8 summarize the design material that has been incorporated into the M in the areas of 1) Design Basis Accident Analysis, 2) Anticipated Transients Without Scram WS), 3) Fire Protection, 4) Flood Protection, 5) Probabilistic Risk Assessment, 6) Radiological lysis, and 7) Severe Accident Analysis. PRA assumptions incorporated into these tables ompass elements of the system design and assumptions that were expressly included in Tier 1 to their importance. Both types of PRA assumptions were included for completeness, but are not nguished in the tables. CDM falling outside of the seven subject areas are intentionally not rporated in these tables. However, the referenced AP1000 DCD sections may contain more rmation than encompassed by these seven subject areas. Each table may also include design 14.3-12 Revision 1
3.8 Summary element of the design certification processes deriving from 10 CFR Part 52 is the selection and umentation of the technical information to be included in the design certification rule as the ified design. The certified design material is a subset of the design information presented in the 2 Material. It includes the following:
Key, important safety-significant aspects of the design described in the certification application Inspections, tests, analyses, and acceptance criteria (ITAAC) that will be used to verify that the as-built facility conforms with the certified design Interface requirements and site parameters information presented in the AP1000 Certified Design Material is prepared using the selection ria and methodology described in this section and is intended to satisfy the above Part 52 uirements for design certification. The ITAAC entries in Sections 2.0 and 3.0 confirm that key ign performance characteristics and design features are implemented in the as-built facility.
3.9 References SECY-90-377, Requirements for Design Certification under 10 CFR Part 52, February 15, 1991.
10 CFR, Part 52, Statements of Consideration, (54 Federal Register 15372 [1989]).
SECY-90-241, Level of Detail Required for Design Certification under Part 52, August 31, 1990.
SECY-90-377, Requirements for Design Certification Under 10 CFR Part 52, November 8, 1990.
SECY-91-178, Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) for Design Certifications and Combined Licenses, June 12, 1991.
14.3-13 Revision 1
Structure/ Structure/ Selected for System Acronym System Description ITAAC ADS Automatic Depressurization System X ASS Auxiliary Steam Supply System X BDS Steam Generator Blowdown System X CAS Compressed Air System X CCS Component Cooling Water System X CDS Condensate System X CES Condenser Tube Cleaning System X CFS Turbine Island Chemical Feed System X CMS Condenser Air Removal System X CNS Containment System X CPS Condensate Polishing System X CVS Chemical and Volume Control System X CWS Circulating Water System X DAS Diverse Actuation System X DDS Data Display Processing System X DOS Standby Diesel Fuel Oil System X DRS Storm Drain System XX DTS Demineralized Water Treatment System X DWS Demineralized Water Transfer and Storage System X ECS Main AC Power System X EDS Non Class 1E DC and UPS System X EFS Communication System X EGS Grounding and Lightning Protection System X EHS Special Process Heat Tracing System X ELS Plant Lighting System X EQS Cathodic Protection System X FHS Fuel Handling System X FPS Fire Protection System X FWS Main and Startup Feedwater System X GSS Gland Seal System X HCS Generator Hydrogen and CO2 Systems X HDS Heater Drain System X 14.3-14 Revision 1
System Acronym System Description ITAAC HSS Hydrogen Seal Oil System X IDS Class 1E DC and UPS System X IIS Incore Instrumentation System X LOS Main Turbine and Generator Lube Oil System X MES Meteorological and Environmental Monitoring System XX MHS Mechanical Handling System X MSS Main Steam System X MTS Main Turbine System X OCS Operations and Control Centers X PCS Passive Containment Cooling System X PGS Plant Gas System X PLS Plant Control System X PMS Protection and Safety Monitoring System X PSS Primary Sampling System X PWS Potable Water System X PXS Passive Core Cooling System X RCS Reactor Coolant System X RDS Gravity and Roof Drain Collection System X RMS Radiation Monitoring System X RNS Normal Residual Heat Removal System X RWS Raw Water System XX RXS Reactor System X SDS Sanitary Drainage System X SES Plant Security System X SFS Spent Fuel Cooling System X SGS Steam Generator System X SJS Seismic Monitoring System X SMS Special Monitoring System X SSS Secondary Sampling System X SWS Service Water System X TCS Turbine Building Closed Cooling Water System X TDS Turbine Island Vents, Drains and Relief Systems X TOS Main Turbine Control and Diagnostics System X 14.3-15 Revision 1
System Acronym System Description ITAAC TVS Closed Circuit TV System XX VAS Radiologically Controlled Area Ventilation System X VBS Nuclear Island Nonradioactive Ventilation System X VCS Containment Recirculation Cooling System X VES Main Control Room Emergency Habitability System X VFS Containment Air Filtration System X VHS Health Physics and Hot Machine Shop HVAC System X VLS Containment Hydrogen Control System X VPS Pump House Building Ventilation System NA VRS Radwaste Building HVAC System X VTS Turbine Island Building Ventilation System X VUS Containment Leak Rate Test System X VWS Central Chilled Water System X VXS Annex/Auxiliary Nonradioactive Ventilation System X VYS Hot Water Heating System X VZS Diesel Generator Building Ventilation System X WGS Gaseous Radwaste System X WLS Liquid Radwaste System X WRS Radioactive Waste Drain System X WSS Solid Radwaste System X WWS Waste Water System X YFS Yard Fire Water System XX ZAS Main Generator System X ZBS Transmission Switchyard and Offsite Power System XX ZOS Onsite Standby Power System X ZRS Offsite Retail Power System NA ZVS Excitation and Voltage Regulation System X end: X = Selected for ITAAC X = Selected for ITAAC - title only, no entry for Design Certification XX = Site-specific system selected for ITAAC - title only.
XX = Selected for ITAAC NA = System is not part of Lee Nuclear Station design 14.3-16 Revision 1
Reference Design Feature Value tion 3.10 The protection and safety monitoring system equipment is seismically qualified to meet safe shutdown earthquake levels.
tion 3.11.3 The design of the protection and safety monitoring system equipment has margin to accommodate a loss of the normal HVAC.
tion 5.1.2 Safety valves are installed above and connected to the pressurizer to provide overpressure protection for the reactor coolant system.
tion 5.1.2 The RCS has two hot legs and four cold legs.
tion 5.1.2 The RCS has two steam generators and four reactor coolant pumps.
tion 5.1.2 The RCS contains a pressurizer and a surge line connected to one hot leg.
tion 5.1.3.3 Rotating inertia needed for flow coast-down, is provided.
le 5.1-3 Minimum measured flow rate with 10% tube plugging (gpm/loop) 150,835 le 5.1-3 Initial rated reactor core thermal power (MWt) 3400 tion 5.2.2 Reactor coolant system and steam system overpressure protection during power operation are provided by the pressurizer safety valves and the steam generator safety valves, in conjunction with the action of the PMS.
tion 5.2.2.1 Safety valve capacity exists to prevent exceeding 110 percent of system design pressure for the following events:
- Loss of electrical load and/or turbine trip
- Uncontrolled rod withdrawal at power
- Loss of reactor coolant flow
- Loss of normal feedwater
- Loss of offsite power to the station auxiliaries tion 5.2.2.1 Overpressure protection for the steam system is provided by steam generator safety valves tion 5.3.2.3 Non-destructive examination (NDE) of the reactor vessel and its appurtenances is conducted in accordance with ASME Code Section III requirements.
14.3-17 Revision 1
Reference Design Feature Value tion 5.3.2.5 The initial Charpy V-notch minimum upper shelf fracture energy levels for the reactor vessel beltline base metal transverse direction and welds are 75 foot-pounds, as required by Appendix G of 10 CFR 50.
tion 5.4.1.2.1 Resistance temperature detectors (RTDs) monitor motor cooling circuit water temperature. These detectors provide indication of anomalous bearing or motor operation. They also provide a system for automatic shutdown in the event of a prolonged loss of component cooling water.
tion 5.4.1.3.4 It is important to reactor protection that the reactor coolant continues to flow for a time after reactor trip and loss of electrical power. To provide this flow, each reactor coolant pump has a high-inertia rotor.
tion 5.4.1.3.4 A safety-related pump trip occurs on high bearing water temperature.
tion 5.4.5.2.3 Power to the pressurizer heaters is blocked when the core makeup tanks are actuated.
tion 5.4.6 Automatic depressurization system stage 1, 2 and 3 valves are connected to the pressurizer and discharge via the spargers to the in-containment refueling water storage tank.
tion 5.4.6 Automatic depressurization system stage 4 valves are connected to each hot leg.
tion 5.4.9.3 In the analysis of overpressure events, the pressurizer safety valves are assumed to actuate at 2500 psia. The safety valve flowrate assumed is based on full flow at 2575 psia, assuming 3 percent accumulation.
tion 5.4.9.3 The pressurizer safety valves prevent reactor coolant system pressure from exceeding 110% of system design pressure.
tion 5.4.12 The reactor head vent valves can be operated from the main control room to provide an emergency letdown path.
le 5.4-1 Minimum reactor coolant motor/pump moment of inertia sufficient to provide flow coastdown as given in Figure 15.3.2.
le 5.4-11 Reactor Coolant System Design Pressure Settings:
- Safety valves begin to open (psig) 2485 14.3-18 Revision 1
Reference Design Feature Value le 5.4-17 Pressurizer Safety Valves - Design Parameters:
- Number 2
- Minimum required relieving capacity per valve (lbm/hr) 750,000
- Set pressure (psig) 2485 +/- 25 tion 6.1.1.4 The exposed surfaces of the excore detectors are made of stainless steel or titanium.
le 6.1-2 The exterior of the containment vessel (above plant elevation 135 3) and the interior of the containment vessel (above 7 above the operating deck) is coated with an inorganic zinc coating.
tion 6.1.2.1.5 The nonsafety-related coatings used inside containment on 100 walls, floors, ceilings, structural steel which is part of the building structure, and on the polar crane have a minimum dry film density (lb/ft3).
ure 6.2.2-1 The passive containment cooling system consists of a water storage tank, cooling water flow discharge path to the containment shell, a water distribution system for the containment shell, and a cooling air flow path.
ure 6.2.2-1 The minimum duration the PCS cooling water flow is provided 72 from the PCCWST (hours).
le 6.2.2-1 The water coverage of the containment shell exceeds the amount used in the safety analysis.
le 6.2.2-1 The minimum drain flow rate capacity of the upper annulus drain 525 (gpm).
le 6.2.2-1 The minimum makeup flow rate capability from an external 100 source to the PCS water storage tank (gpm).
le 6.2.2-1 The minimum makeup flow rate capability from the PCS water 118 storage tank to the spent fuel pit (gpm).
le 6.2.2-1 The minimum PCS water storage tank volume for makeup to the 756,700 spent fuel pit (non-coincident with PCS operation) (gallons).
le 6.2.2-1 The minimum long term makeup capability from the PCCAWST 4 to the PCCWST (days).
14.3-19 Revision 1
Reference Design Feature Value le 6.2.2-1 The minimum long term makeup flow capability from the 100 PCCAWST to the PCCWST (gpm).
le 6.2.2-1 The minimum long term makeup flow capability from the > 35 PCCAWST to the spent fuel pool (gpm).
le 6.2.2-2 The first (i.e., tallest) standpipes elevation above the tank floor 24.1 +/- 0.2 (feet).
le 6.2.2-2 The second tallest standpipes elevation above the tank floor 20.3 +/- 0.2 (feet).
le 6.2.2-2 The third tallest standpipes elevation above the tank floor 16.8 +/- 0.2 (feet).
le 6.2.2-1 The minimum passive containment cooling water flow rate at a 109.6 PCCWST water level 4.0 ft. (+/- 0.2 ft.) above the tank floor. (This supports analysis that ensures that delivered flow at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> will be greater than 100.7 gpm.) (gpm).
le 6.2.2-1 The minimum passive containment cooling water flow rate when 151.4 the PCCWST water level uncovers the third tallest standpipe (gpm).
le 6.2.2-1 The minimum passive containment cooling water flow rate when 184.0 the PCCWST water level uncovers the second tallest standpipe (gpm).
le 6.2.2-1 The minimum passive containment cooling water flow rate when 238.4 the PCCWST water level uncovers the first (i.e., tallest) standpipe (gpm).
le 6.2.2-1 The minimum passive containment cooling water flow rate with 471.1 water inventory at a level of 27.4 ft + 0.2, -0.0 ft above the tank floor (gpm).
tion 6.3 The passive core cooling system provides core decay heat removal during design basis events.
tion 6.3 The passive core cooling system provides RCS makeup, boration, and safety injection during design basis events.
tion 6.3 The passive core cooling system provides pH adjustment of water flooding the containment following design bases events.
14.3-20 Revision 1
Reference Design Feature Value tion 6.3.1.1 The passive core cooling system is designed to provide emergency core cooling during events involving increases and decreases in secondary side heat removal and decreases in reactor coolant system inventory.
tion 6.3.2.1.1 The heat exchanger consists of a bank of C-tubes, connected to a tubesheet and channel heat arrangement at the top (inlet) and bottom (outlet). The passive exchanger connects to the reactor coolant system through an inlet line from one reactor coolant system hot leg and an outlet line to the associated steam generator cold leg plenum (reactor coolant pump suction).
tion 6.3.2.1.1 For the passive residual heat removal heat exchanger, the normal water temperature in the inlet line will be hotter than the discharge line.
tion 6.3.2.1.2 The actuation of the core makeup tanks following a steam line break provides injection of borated water via water recirculation to mitigate the reactivity transient and provide the required shutdown margin.
tion 6.3.2.2.1 The CMT inlet diffuser has a minimum flow area (in2). 165 tion 6.3.2.2.3 The in-containment refueling water storage tank contains one passive residual heat removal heat exchanger.
tion 6.3.2.2.6 The connection of the sparger branch arms to the sparger hub 11.5 are submerged below the in-containment refueling water storage tank overflow level (ft).
tion 6.3.2.2.6 Automatic depressurization system stage 1, 2 and 3 valves are connected to the pressurizer and discharge via the spargers to the in-containment refueling water storage tank.
tion 6.3.2.2.7.1 The containment recirculation screens have plates that are located no more than 1 foot above the top of the screens and extend out at least 10 feet in front and at least 7 feet to the side of the screens to prevent coating debris from reaching the screens.
tion 6.3.2.2.7.1 The type of insulation used on ASME Class 1 lines inside containment and on the reactor vessel, reactor coolant pumps, pressurizer and steam generators is a metal reflective or suitable equivalent insulation.
14.3-21 Revision 1
Reference Design Feature Value tion 6.3.2.2.7.3 The surface materials used in the vicinity of the containment recirculation screens are stainless steel. In the vicinity of the containment recirculation screens includes surfaces located above the bottom of the recirculation screens up to and including the bottom surface of the plate discussed in Subsection 6.3.2.2.7.1, and the surfaces 10 feet in front and 7 feet to the sides of the screen face.
tion 6.3.2.2.7.3 The bottom of the containment recirculation screens are located 2 above the loop compartment floor (ft).
tion 6.3.3.2.1 For a loss of main feedwater event, the passive residual heat removal heat exchanger is actuated. If the core makeup tanks are not initially actuated, they actuate later when passive residual heat exchanger cooling sufficiently reduces pressurizer level.
tion 6.3.3.2.2 For a feedwater system pipe failure event, the passive residual heat removal heat exchanger and the core makeup tanks are actuated.
tion 6.3.3.3.1 For a steam generator tube rupture event, the nonsafety-related makeup pumps are automatically actuated when reactor coolant system inventory decreases and a reactor trip occurs, followed by actuation of the startup feedwater pumps. Makeup pumps automatically function to maintain the programmed pressurizer level. The core makeup tanks subsequently actuate on low pressurizer level, if they are not already actuated. Actuation of the core makeup tanks automatically actuates the passive residual heat removal system heat exchanger.
tion 6.3.6.1 The piping resistances connecting the following PXS components and the RCS are bounded by the resistances assumed in the Chapter 15 safety analysis:
- Core makeup tanks
- In-containment refueling water storage tank injection
- Containment recirculation
- Automatic depressurization system valves tion 6.3.6.1.3 The bottom of the core makeup tanks are located above the 7.5 reactor vessel direct vessel injection nozzle centerline (ft).
14.3-22 Revision 1
Reference Design Feature Value ction 6.3.6.1.3 The bottom of the in-containment refueling water storage tank is 3.4 located above the direct vessel injection nozzle centerline (ft).
ction 6.3.6.1.3 The pH baskets are located below plant elevation 107 2.
ure 6.3-1 The passive core cooling system has two direct vessel injection lines.
le 6.3-2 The passive core cooling system has two core makeup tanks, 2500 each with a minimum required volume (ft3).
le 6.3-2 The passive core cooling system has two accumulators, each 2,000 with a minimum required volume (ft3) le 6.3-2 The passive core cooling system has an in-containment refueling 73,900 water storage tank with a minimum required water volume (ft3) ction 6.3.2.2.3 The containment floodup volume for a LOCA in PXS room B has 73,500 a maximum volume (ft3) (excluding the IRWST) below a containment elevation of 108 feet.
le 6.3-2 Each sparger has a minimum discharge flow area (in2). 274 le 6.3-2 The passive core cooling system has two pH adjustment baskets 280 each with a minimum required volume (ft3).
ction 14.2.9.1.3f The passive residual heat removal heat exchanger minimum natural circulation heat transfer rate (Btu/hr)
With 520°F hot leg and 80°F IRWST 1.78 E+08 With 420°F hot leg and 80°F IRWST 1.11 E+08 ction 6.3.6.1.3 The centerline of the HXs upper channel head is located above 26.3 the HL centerline (ft).
ure 6.3-1 The CMT level sensors (PXS-11A/B/C/D, -12A/B/C/D, 1 +/- 1
-13A/B/C/D, and -14A/B/C/D) upper level tap centerlines are located below the centerline of the upper level tap connection to the CMTs (in).
ure 6.3-1 The CMT inlet lines (cold leg to high point) have no downward sloping sections.
ure 6.3-1 The maximum elevation of the CMT injection lines between the connection to the CMT and the reactor vessel is the connection to the CMTs.
ure 6.3-1 The PRHR inlet line (hot leg to high point) has no downward sloping sections.
14.3-23 Revision 1
Reference Design Feature Value ure 6.3-1 The maximum elevation of the IRWST injection lines (from the connection to the IRWST to the reactor vessel) and the containment recirculation lines (from the containment to the IRWST injection lines) is less than the bottom inside surface of the IRWST.
ure 6.3-1 The maximum elevation of the PRHR outlet line (from the PRHR to the SG) is less than the PRHR lower channel head top inside surface.
tion 7.1.2.10 Isolation devices are used to maintain the electrical independence of divisions and to see that no interaction occurs between nonsafety-related systems and the safety-related system. Isolation devices serve to prevent credible faults in circuit from propagating to another circuit.
tion 7.1.4.2 The ability of the protection and safety monitoring system to initiate and accomplish protective functions is maintained despite degraded conditions caused by internal events such as fire, flooding, explosions, missiles, electrical faults and pipe whip.
tion 7.1.2 The flexibility of the protection and safety monitoring system enables physical separation of redundant divisions.
tion 7.2.2.2.1 The protection and safety monitoring system initiates a reactor trip whenever a condition monitored by the system reaches a preset level.
tion 7.2.2.2.8 The reactor is tripped by actuating one of two manual reactor trip controls from the main control room.
tion 7.3.1.2.2 The in-containment refueling water storage tank is aligned for injection upon actuation of the fourth stage automatic depressurization system via the protection and safety monitoring system.
tion 7.3.1.2.3 The core makeup tanks are aligned for operation on a safeguards actuation signal or on a low-2 pressurizer level signal via the protection and safety monitoring system.
tion 7.3.1.2.4 The fourth stage valves of the automatic depressurization system receive a signal to open upon the coincidence of a low-2 core makeup tank water level in either core makeup tank and low reactor coolant system pressure following a preset time delay after the third stage depressurization valves receive a signal to open via the protection and safety monitoring system.
14.3-24 Revision 1
Reference Design Feature Value tion 7.3.1.2.4 The first stage valves of the automatic depressurization system open upon receipt of a signal generated from a core makeup tank injection alignment signal coincident with core makeup tank water level less than the Low-1 setpoint in either core makeup tank via the protection and safety monitoring system.
tion 7.3.1.2.4 The second and third stage valves open on time delays following generation of the first stage actuation signal via the protection and safety monitoring system.
tion 7.3.1.2.5 The reactor coolant pumps are tripped upon generation of a safeguards actuation signal or upon generation of a low-2 pressurizer water level signal.
tion 7.3.1.2.7 The passive residual heat removal heat exchanger control valves are opened on low steam generator water level or on a CMT actuation signal via the protection and safety monitoring system.
tion 7.3.1.2.9 The containment recirculation isolation valves are opened on a safeguards actuation signal in coincidence with low-3 in-containment refueling water storage tank water level via the protection and safety monitoring system.
tion 7.3.1.2.14 The demineralized water system isolation valves close on a signal from the protection and safety monitoring system derived from either a reactor trip signal, a source range flux doubling signal, low input voltage to the 1E dc uninterruptible power supply battery chargers or if the source range flux doubling logic is blocked during shutdown.
tion 7.3.1.2.15 The chemical and volume control system makeup line isolation valves automatically close on a signal from the protection and monitoring system derived from a source range flux doubling, high-2 pressurizer level, high-2 steam generator level signal, a safeguards signal coincident with high-1 pressurizer level, or high-2 containment radioactivity.
tion 7.3.2.2.1 The protection and monitoring system automatically generate an actuation signal for an engineered safety feature whenever a monitored condition reaches a preset level.
tion 7.3.2.2.9 Manual initiation at the system-level exists for the engineered safety features actuation.
14.3-25 Revision 1
Reference Design Feature Value tion 7.4.3.1 If temporary evacuation of the main control room is required because of some abnormal main control room condition, the operators can establish and maintain safe shutdown conditions for the plant from outside the main control room through the use of controls and monitoring located at the remote shutdown workstation.
tion 7.4.3.1.1 The remote shutdown workstation equipment is similar to the operator workstations in the main control room and is designed to the same standards. One remote shutdown workstation is provided.
tion 7.4.3.1.3 The remote shutdown workstation achieves and maintains safe shutdown conditions from full power conditions and maintains safe shutdown conditions thereafter.
tion 7.5.4 The protection and safety monitoring system provides signal conditioning, communications, and display functions for Category 1 variables and for Category 2 variables that are energized from the Class 1E uninterruptible power supply system.
tion 7.6.1.1 An interlock is provided for the normally closed motor-operated normal residual heat removal system inner and outer suction isolation valves. Each valve is interlocked so that it cannot be opened unless the reactor coolant system pressure is below a preset pressure.
tion 8.2.2 Following a turbine trip during power operation, the 15 reverse-power relay will be blocked for a minimum time period (sec).
tion 8.3.2.1.2 The non-Class 1E dc and UPS system (EDS) consists of the electric power supply and distribution equipment that provides dc and uninterruptible ac power to nonsafety-related loads.
14.3-26 Revision 1
Reference Design Feature Value tion 9.1.1.2.1.C During normal fuel handling operations, a single failure-proof hoist, designed to meet the requirements of NUREG-0554, is the only hoist capable of moving new fuel above the operating floor. Per the design criteria contained in NUREG-0554, drops from a single failure-proof hoist are deemed unlikely and do not require further analysis. The consequences of such a drop are minimal since no safety-related equipment would be impacted and there are no radiological releases with new unirradiated fuel. Because the likelihood of a new fuel assembly being dropped into the new fuel pit and onto the new fuel racks is minimal, it is unnecessary to evaluate drop scenarios for the new fuel storage rack.
tion 9.1.3.5 The spent fuel pool is designed such that a water level is maintained above the spent fuel assemblies for at least 7 days following a loss of the spent fuel cooling system using only on-site makeup water sources (See Table 9.1-4).
tion 9.1.3.5 The spent fuel pool cooling system includes safety-related connections to establish safety-related makeup to the spent fuel pool following a design basis event including a seismic event.
tion 9.1.4.1.1 In the event of a safe shutdown earthquake (SSE), handling equipment cannot fail in such a manner as to prevent required function of seismic Category 1 equipment.
tion 9.3.6.3.7 The chemical and volume control system contains two redundant safety-related valves to isolate the demineralized water system from the makeup pump suction.
tion 9.3.6.3.7 The chemical and volume control system contains two safety-related valves to isolate the makeup flow to the reactor coolant system.
tion 9.3.6.4.5 The chemical and volume control system contains two safety-related valves to isolate the makeup flow to the reactor coolant system.
tion 9.3.6.4.5.1 The chemical and volume control system contains two redundant safety-related valves to isolate the demineralized water system from the makeup pump suction.
14.3-27 Revision 1
Reference Design Feature Value tion 9.3.6.7 The demineralized water system isolation valves close on a signal from the protection and safety monitoring system derived from either a reactor trip signal, a source range flux doubling signal, low input voltage to the 1E dc and uninterruptible power supply battery chargers, a safety injection signal, or if the source range flux doubling logic is blocked during shutdown conditions.
tion 9.3.6.7 The chemical and volume control system makeup line isolation valves automatically close on a signal from the protection and safety monitoring system derived from a source range flux doubling, high-2 pressurizer level, high steam generator level signal, or a safeguards signal coincident with high-1 pressurizer level.
tion 10.1.2 Safety valves are provided on both main steam lines.
tion 10.2.2.4.3 The flow of the main steam entering the high-pressure turbine is controlled by four stop valves and four governing control valves.
The stop valves are closed by actuation of the emergency trip system devices.
tion 10.3.1.1 The main steam supply system is provided with a main steam isolation valve and associated MSIV bypass valve on each main steam line from its respective steam generator.
tion 10.3.1.1 A main steam isolation valve (MSIV) on each main steam line prevents the uncontrolled blowdown of more than one steam generator and isolates nonsafety-related portions of the system.
tion 10.3.1.2 Power-operated atmospheric relief valves are provided to allow controlled cooldown of the steam generator and the reactor coolant system when the condenser is not available.
tion 10.3.2.1 The main steam supply system includes:
- One main steam isolation valve and one main steam isolation valve bypass valve per main steam line.
- Power-operated atmospheric relief valves and upstream isolation valves.
tion 10.3.2.3.2 In the event that a design basis accident occurs, which results in a large steam line break, the main steam isolation valves with associated main steam isolation bypass valves automatically close.
14.3-28 Revision 1
Reference Design Feature Value ure 10.3.2-1 The steam generator system consists of two main steam, two main feedwater, and two startup feedwater lines.
le 10.3.2-2 Design data for main steam supply safety system valves:
- Number per main steam line 6
- Minimum relieving capacity per valve at 110% of design 1,370,000 pressure (lb/hr) le 10.3.2-2 The flow capacity of the steam generator safety valves (lbm/hr) 8,240,000 at 110% of design pressure.
le 10.3.2-2 The maximum set pressure of the steam generator safety valves 1,242 (psig).
tion 10.4.8.3 The safety-related portions of the steam generator blowdown system are located in the containment and auxiliary buildings and are designed to remain functional after a safe shutdown earthquake.
tion 10.4.7.1.1 Double valve main feedwater isolation is provided via the main feedwater control valve and main feedwater isolation valve. Both valves close automatically on main feedwater isolation signals, an appropriate engineered safety features isolation signal, within the time established with the Technical Specifications, Section 16.1. The startup feedwater control valve also serves as a containment isolation valve.
tion 10.4.7.1.1 The condensate and feedwater system provides redundant isolation valves for the main feedwater lines routed into containment.
tion 10.4.7.1.1 For a main feedwater or main steam line break (MSLB) inside the containment, the condensate and feedwater system is designed to limit high energy fluid to the broken loop.
tion 10.4.7.1.2 The booster/main feedwater pumps are tripped simultaneously with the feedwater isolation signal to close the main feedwater isolation valves.
tion 10.4.7.2.1 The main feedwater pumps and booster pumps are tripped with the feedwater isolation signal that closes the main feedwater isolation valves. The same isolation signal closes the isolation valve in the cross connect line between the main feedwater pump discharge header and the startup feedwater pump discharge header.
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Reference Design Feature Value tion 10.4.7.2.2 One MFIV is installed in each of the two main feedwater lines outside the containment and downstream of the feedwater control valve. The MFIVs are installed to prevent uncontrolled blowdown from the steam generators in the event of a feedwater pipe rupture. The main feedwater check valve provides backup isolation. In the event of a secondary side pipe rupture inside the containment, the MFIVs limit the quantity of high energy fluid that enters the containment through the broken loop and limit cooldown. The MFCV provides backup isolation to limit cooldown and high energy fluid addition.
tion 10.4.7.2.2 In the event of a secondary side pipe rupture inside the containment, the main feedwater control valves provide a redundant isolation to the MFIVs to limit the quantity of high energy fluid that enters the containment through the broken loop.
tion 10.4.7.3 For a main feedwater line break inside the containment or a main steam line break, the MFIVs and the main feedwater control valves automatically close upon receipt of a feedwater isolation signal.
tion 10.4.7.3 For a steam generator tube rupture event, positive and redundant isolation is provided for the main feedwater (MFIV and MFCV) with isolation signals generated by the protection and safety monitoring system (PMS).
tion 10.4.8.2.2.7 Blowdown system isolation is actuated on low steam generator water levels. The isolation of steam generator blowdown provides for a continued availability of the steam generator as a heat sink for decay heat removal in conjunction with operation of the passive residual heat removal system and the startup feedwater system.
tion 10.4.8.3 The safety-related portions of the steam generator blowdown system located in the containment and auxiliary buildings are designed to remain functional after a safe shutdown earthquake.
tion 10.4.9.1.1 Double valve startup feedwater isolation is provided by the startup feedwater control valve and the startup feedwater isolation valve. Both valves close on a startup feedwater isolation signal, an appropriate engineered safeguards features signal, within the time established within the Technical Specifications, Section 16.1.
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Reference Design Feature Value tion 10.4.9.1.1 For a steam generator tube rupture event, positive and redundant isolation is provided for the startup feedwater system (startup feedwater isolation valve and startup feedwater control valve), with isolation signals generated by the protection and safety monitoring system.
tion 10.4.9.2.2 In the event of a steam generator tube rupture, the startup feedwater isolation valve and startup feedwater control valve limit overfill of the steam generator by terminating startup feed flow.
tion 10.4.9.2.2 In the event of a secondary pipe rupture inside containment, the startup feedwater isolation valve and startup feedwater control valve provide isolation to limit the quantity of high energy fluid that enters the containment.
tion 10.4.9.2.2 The startup feedwater isolation valve is provided to prevent the uncontrolled blowdown from more than one steam generator in the event of startup feedwater line rupture. The startup feedwater isolation valve provides backup isolation.
le 15.0-1 Initial core thermal power (MWt). 3400 le 15.0-3 Nominal values of pertinent plant parameters used in accident 296,000 analysis with 10% steam generator tube plugging
- Reactor coolant flow (gpm) tion 15.1.2.1 Continuous addition of excessive feedwater is prevented by the steam generator high-2 water level signal trip, which closes the feedwater isolation valves and feedwater control valves and trips the turbine, main feedwater pumps and reactor.
tion 15.1.4.1 For an inadvertent opening of a steam generator relief or safety valve, core makeup tank actuation occurs from one of four sources:
- Two out of four low pressurizer pressure signals
- Two out of four low-2 pressurizer level signals
- Two out of four low Tcold signals in any one loop
- Two out of four low steam line pressure signals in any one loop tion 15.1.4.1 After an inadvertent opening of a steam generator relief or safety valve, redundant isolation of the main feedwater lines closes the feedwater control valves and feedwater isolation valves, and trips the main feedwater pumps.
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Reference Design Feature Value tion 15.1.5.1 Following a steam line rupture, core makeup tank actuation occurs from one of five sources:
- Two out of four low pressurizer pressure signals
- Two out of four high-2 containment pressure signals
- Two out of four low steam line pressure signals in any loop
- Two out of four low Tcold signals in any one loop
- Two out of four low-2 pressurizer level signals tion 15.1.5.1 After a steam line rupture, redundant isolation of the main feedwater lines closes the feedwater control valves and feedwater isolation valves, and trips the main feedwater pumps.
tion 15.1.5.2.1 Core makeup tanks and the accumulators are the portions of the passive core cooling system used in mitigating a steam line rupture.
tion 15.1.6.1 The heat sink for the PRHR heat exchanger is provided by the IRWST, in which the PRHR heat exchanger is submerged.
tion 15.2.6.2.1 Following a loss of ac power, the PRHR heat exchanger is actuated by the low steam generator water level (wide range).
tion 15.2.8.2.1 Receipt of a low steam line pressure signal in at least one steam line initiates a steam line isolation signal that closes all main steam line and feed line isolation valves. This signal also gives a safeguards signal that initiates flow of cold borated water from the core makeup tanks to the reactor coolant system.
tion 15.3.3.2.2 The pressurizer safety valves are fully open at 2575 psia. Their capacity for steam relief is described in Section 5.4.
tion 15.4.6.2.2 A safety signal from the protection and safety monitoring system automatically isolates the potentially unborated water from the demineralized water transfer and storage system and thereby terminates the dilution.
tion 15.5.1.1 Following inadvertent operation of the core makeup tanks during power operation, the high-3 pressurizer level signal actuates the PRHR heat exchanger and blocks the pressurizer heaters.
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Reference Design Feature Value tion 15.5.2.1 The pressurizer heaters are blocked, and the main feedwater lines, steam lines, and chemical and volume control system are isolated.
le 15.6.5-10 ADS Valve Flow Areas (in2)
- ADS Stage 1 Control Valve 4.6
- ADS Stage 2 Control Valve 21
- ADS Stage 3 Control Valve 21
- ADS Stage 4A Valve 67
- ADS Stage 4B Valve 67 le 15.6.5-10 ADS Valve Opening Times (sec)
- ADS Stage 1 Control Valve 40
- ADS Stage 1 Isolation Valve 30
- ADS Stage 2 Control Valve 100
- ADS Stage 2 Isolation Valve 60
- ADS Stage 3 Control Valve 100
- ADS Stage 3 Isolation Valve 60 tion 18.8.3.2 The main control area includes the reactor operator workstations, the supervisors workstation, the dedicated safety panel and the wall panel information system.
tion 18.8.3.2 The human system interface resources available at each workstation are the plant information system displays, the control displays (soft controls), the alarm system support displays, procedure system, and the screen and component selector.
e:
valve closure times reflect the design basis of the AP1000. The applicable Chapter 15 accidents were evaluated for these ign basis valve closure times. The results of this evaluation have concluded that there is a small impact on the Chapter 15 lysis and the conclusions remain valid.
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Reference Design Feature Value tion 7.7.1.11 The diverse actuation system is a nonsafety-related system that provides a diverse backup to the protection and safety monitoring system.
tion 7.7.1.11 The diverse actuation system trips the reactor control rods and the turbine on low wide range steam generator water level, or on low pressurizer water level, or on high hot leg temperature.
tion 7.7.1.11 The diverse actuation system initiates passive residual heat removal on low wide range steam generator water level or high hot leg temperature; actuates core makeup tanks and trips the reactor coolant pumps on low pressurizer water level; and isolates selected containment penetrations and starts passive containment cooling on high containment temperature.
tion 7.7.1.11 The manual actuation function of the diverse actuation system is implemented by wiring the controls located in the main control room directly to the final loads in a way that bypasses the normal path through the control room multiplexers, the protection and safety monitoring system cabinets, and the diverse actuation system logic.
tion 7.7.1.11 The diverse actuation system uses microprocessor or special purpose logic processor boards different from those used in the protection and safety monitoring system.
tion 7.7.1.11 The diverse actuation system hardware implementation is different from that of the protection and safety monitoring system.
tion 7.7.1.11 The operating system and programming language of the diverse actuation system is different from that of the protection and safety monitoring system.
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Reference Design Feature Value tion 9A.3.1 Separation is maintained between Class 1E divisions and between Class 1E divisions and non-Class 1E cables in accordance with the fire areas.
tion 3.4.1.1.2 The AP1000 arrangement provides physical separation of redundant safety-related components and systems from each other and from nonsafety-related components.
tion 3.8.4.1.1 The conical roof supports the passive containment cooling system tank, which is constructed with a stainless steel liner on reinforced concrete walls.
tion 7.1.2 The ability of the protection and safety monitoring system to initiate and accomplish protective functions is maintained despite degraded conditions caused by internal events such as fire and flooding.
tion 7.4.3.1 If temporary evacuation of the main control room is required because of some abnormal main control room condition, the operators can establish and maintain safe shutdown conditions for the plant from outside the main control room through the use of controls and monitoring located at the remote shutdown workstation.
tion 7.4.3.1.1 The remote shutdown workstation equipment is similar to the operator workstations in the main control room and is designed to the same standards. One remote shutdown workstation is provided.
tion 7.4.3.1.3 The remote shutdown workstation achieves and maintains safe shutdown conditions from full power conditions and maintains safe shutdown conditions thereafter.
tion 8.3.2.2 The four divisions of Class 1E battery chargers and Class 1E voltage regulating transformers are independent, located in separate rooms, cannot be interconnected, and their circuits are routed in dedicated, physically separated raceways.
tion 8.3.2.3 Each safety-related circuit and raceway is given a unique identification number to distinguish between circuits and raceways of different voltage level or separation groups.
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Reference Design Feature Value tion 8.3.2.4.2 Cables of one separation group are run in separate raceway and physically separated from cables of other separation groups.
Group N raceways are separated from safety-related groups A, B, C, and D. Non-class 1E circuits are electrically isolated by isolation devices, shielding and wiring techniques, physical separation, or an appropriate combination thereof.
tion 9.5.1.2.1.1 Separation is maintained between redundant safe shutdown components, including equipment, electrical cables, and instrumentation controls, in accordance with the fire areas.
tion 9.5.1.2.1.5 The standpipe system is supplied with water from the safety-related passive containment cooling system storage tank and normally operates independently of the rest of the fire protection system. The supply line draws water from a portion of the storage tank, using water allocated for fire protection.
tion 9.5.1.2.1.5 The standpipe system serving areas containing equipment required for safe shutdown following a safe shutdown earthquake is designed and supported so that it can withstand the effects of a safe shutdown earthquake and remain functional.
tion 9.5.1.2.1.5 The volume of the water in the PCS tank is sufficient to supply 18,000 two hose streams, each with a flow of 75 gallons per minute, for two hours (gal).
le 9.5.1-2 Each fire pump is rated:
- Flow rate (gpm) 2000
- Total head (ft) 300 tion 18.8.3.2 The human system interface resources available at each workstation are the plant information system displays, the control displays (soft controls), the alarm system support displays, procedure system, and the screen and component selector.
tion 18.8.3.4 The mission of the remote shutdown workstation is to provide the resources to bring the plant to a safe shutdown condition after an evacuation of the main control room.
tion 18.12.3 The controls, displays, and alarms listed in Table 18.12.2-1 are retrievable from the remote shutdown workstation.
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Reference Design Feature Value tion Appendix 1-A The lowest level of the auxiliary building, elevation 66-6, RG 1.143 contains the components of the radwaste system within a Section C.1.1.3 common flood zone with watertight floors and walls. This volume Clarification of this enclosed flood zone is sufficient to contain the contents of the radwaste system.
le 2.0-201 Plant elevation for maximum flood level (ft). 100 tion 3.4.1.1.1 The seismic Category I structures below grade are protected against flooding by waterstops and a waterproofing system.
tion 3.4.1.1.2 The boundaries between mechanical equipment rooms and the electrical and instrumentation and control equipment rooms of the auxiliary building are designed to prevent flooding of rooms that contain safe shutdown equipment up to the maximum flood level for each room.
tion 3.4.1.2.2 The boundaries between mechanical equipment rooms inside containment and the electrical and instrumentation and control equipment rooms of the auxiliary building are designed to prevent flooding of rooms that contain safe shutdown equipment up to the maximum flood level for each room.
tion 3.4.1.2.2 Boundaries exist to prevent flooding between the following rooms which contain safety-related equipment: PXS valve/
accumulator room A, PXS valve/accumulator room B, and chemical and volume control room.
tion 3.4.1.2.2 The AP1000 arrangement provides physical separation of redundant safety-related components and systems from each other and from nonsafety-related components.
tion 3.4.1.2.2 The safety-related components available for safety shutdown are located in the auxiliary building and inside containment. No credit is taken for operation of sump pumps to mitigate the consequences of flooding.
tion 3.4.1.2.2.1 The PXS-A compartment, PXS-B compartment and the chemical and volume control system compartment are physically separated and isolated from each other by structural walls such that flooding in any one of these compartments cannot cause flooding in any of the other compartments at elevations up to the top of these compartments.
tion 3.6 In the event of a high- or moderate-energy pipe failure within the plant, adequate protection is provided so that essential structures, systems, or components are not impacted by the adverse effects of postulated pipe failure.
tion 7.1.2 The ability of the protection and safety monitoring system to initiate and accomplish protective functions is maintained despite degraded conditions caused by internal events such as fire and flooding.
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Reference Design Feature Value le 3.2-2 The Nuclear Island structures include the containment and the shield and auxiliary buildings. These structures are seismic Category I.
le 3.2-3 The components identified under Reactor Systems in Table 3.2-3, as ASME Code Section III are designed and constructed in accordance with ASME Code Section III Requirements.
le 3.2-3 The Nuclear Island structures include the containment and the Shield and Auxiliary Buildings. These structures are seismic Category I.
tion 3.4.1.1.2 The boundaries between mechanical equipment rooms and the electrical and instrumentation and control equipment rooms of the auxiliary building are designed to prevent flooding of rooms that contain safe shutdown equipment up to the maximum flood level for each room.
tion 3.4.1.1.2 The AP1000 arrangement provides physical separation of redundant safety-related components and systems from each other and from nonsafety-related components.
tion 9A.3.1 Separation is maintained between Class 1E divisions and between Class 1E divisions and non-Class 1E cables in accordance with the fire areas.
tion 3.4.1.2.2 Boundaries exist to prevent flooding between the following rooms which contain safety-related equipment: PXS valve/
accumulator room A, PXS valve/accumulator room B, and chemical and volume control room.
tion 3.4.1.2.2 The boundaries between mechanical equipment rooms inside containment and the electrical and instrumentation and control equipment rooms of the auxiliary building are designed to prevent flooding of rooms that contain safe shutdown equipment up to the maximum flood level for each room.
tion 3.4.1.2.2 The safety-related components available for safety shutdown are located in the auxiliary building and inside containment. No credit is taken for operation of sump pumps to mitigate the consequences of flooding.
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Reference Design Feature Value tion 3.4.1.2.2.1 The PXS-A compartment, PXS-B compartment and the chemical and volume control system compartment are physically separated and isolated from each other by structural walls such that flooding in any one of these compartments or in the reactor coolant system compartment cannot cause flooding in any of the other compartments.
tion 3.11.3 The design of the protection and safety monitoring system equipment has margin to accommodate a loss of the normal HVAC.
tion 3D.6 RXS equipment in Appendix 3D is seismically qualified.
tion 5.1.3.7 ADS has four stages. Each stage is arranged into two separate groups of valves and lines.
- Stages 1, 2, and 3 discharge from the top of the pressurizer to the IRWST.
- Each stage 4 discharges from a hot leg to the RCS loop compartment.
tion 5.3.1.1 The reactor vessel provides a high integrity pressure boundary to contain the reactor coolant, heat generating reactor core, and fuel fission products. The reactor vessel is the primary boundary for the reactor coolant and the secondary barrier against the release of radioactive fission products.
tion 5.4.6 ADS has four stages. Each stage is arranged into two separate groups of valves and lines.
- Stages 1, 2, and 3 discharge from the top of the pressurizer to the IRWST.
- Each stage 4 discharges from a hot leg to the RCS loop compartment.
tion 5.4.6.2 Each ADS stage 1, 2, and 3 line contains two normally closed motor-operated valves (MOVs).
tion 5.4.6.2 Each ADS stage 4 line contains a normally open MOV valve and a normally closed squib valve.
tion 5.4.7 The RNS removes heat from the core and reactor coolant system at reduced RCS pressure and temperature conditions after shutdown.
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Reference Design Feature Value tion 5.4.7 The normal residual heat removal system (RNS) provides a safety-related means of performing the following functions:
- Containment isolation for the RNS lines that penetrate the containment
- Long-term, post-accident makeup water to the RCS tion 5.4.7.1.1 The RNS containment isolation and pressure boundary valves are safety-related. The motor-operated valves are powered by Class 1E dc power.
tion 5.4.7.1.2.1 The component cooling water system (CCS) provides cooling to the RNS heat exchanger.
tion 6.2.4 The containment hydrogen control system provides nonsafety-related hydrogen igniters for control of the containment hydrogen concentration for beyond design basis accidents.
tion 6.2.4.2.3 At least 64 hydrogen igniters are provided.
tion 6.3.1.1.3 The automatic depressurization system provides a safety-related means of depressurizing the RCS.
tion 6.3 The in-containment refueling water storage tank subsystem provides a safety-related means of performing the following functions:
- Low-pressure safety injection
- Core decay heat sink during design basis events
- Flooding of the lower containment, the reactor cavity and the loop compartment by draining the IRWST into the containment.
- Borated water tion 6.3.1 The core makeup tanks provide safety-related means of safety injection of borated water to the RCS.
tion 6.3.1 Passive residual heat removal (PRHR) provides a safety-related means of removing core decay heat during design basis events.
tion 6.3.2 The ADS valves are powered from Class 1E dc power.
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Reference Design Feature Value tion 6.3.2 There are two CMTs, each with an injection line to the reactor vessel/DVI nozzle.
- Each CMT has a pressure balance line from an RCS cold leg.
- Each injection line is isolated with a parallel set of air-operated valves (AOVs).
- These AOVs open on loss of air.
- The injection line for each CMT also has two check valves in series.
tion 6.3.2 The IRWST subsystem has the following flow paths:
- Two (redundant) injection lines from the IRWST to the reactor vessel/DVI nozzle. Each line is isolated with a parallel set of valves; each set with a check valve in series with a squib valve.
- Two (redundant) recirculation lines from the containment to the IRWST injection line. Each recirculation line has two paths: one path contains a squib valve and an MOV, the other path contains a squib valve and a check valve.
- The two MOV/squib valve lines also provide the capability to flood the reactor cavity.
tion 6.3.2 There are screens for each IRWST injection line and recirculation line.
tion 6.3.2 PRHR is actuated by opening redundant, parallel air-operated valves. These air-operated valves open on loss of air.
tion 6.3.2.2 The passive core cooling system (PXS) is composed of the following:
- Accumulator subsystem
- Core makeup tank (CMT) subsystem
- In-containment refueling water storage tank (IRWST) subsystem
- Passive residual heat removal (PRHR) subsystem.
- The automatic depressurization system (ADS), which is a subsystem of the reactor coolant system (RCS), also supports passive core cooling functions.
tion 6.3.2.2.2 There are two accumulators, each with an injection line to the reactor vessel/direct vessel injection (DVI) nozzle. Each injection line has two check valves in series.
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Reference Design Feature Value tion 6.3.2.2.2 The accumulators provide a safety-related means of safety injection of borated water to the RCS.
tion 6.3.2.2.8.7 The accumulator discharge check valves are of a different type than the CMT discharge check valves.
tion 6.3.3 IRWST squib valves and MOVs are powered by Class 1E dc power.
tion 6.3.3 The CMT AOVs are automatically and manually actuated from PMS and DAS.
tion 6.3.3 The PRHR air-operated valves are automatically actuated and manually actuated from the control room by either PMS or DAS.
tion 6.3.3 The squib valves and MOVs for injection and recirculation are automatically and manually actuated via PMS, and manually actuated via DAS.
tion 6.3.3 The squib valves and MOVs for and reactor cavity flooding are manually actuated via PMS and DAS from the control room.
tion 6.3.7 The positions of the containment recirculation isolation MOVs are indicated in the control room.
tion 6.3.7 The position of the inlet PRHR valve is indicated in the control room.
tion 6.3.7.6.1 The ADS first-, second-, and third-stage valve positions are indicated in the control room.
tion 7.1.1 The diverse actuation system provides a nonsafety-related means of performing the following functions:
- Initiates automatic and manual reactor trip
- Automatic and manual actuation of selected engineered safety features
- Main control room display of selected plant parameters.
tion 7.1.1 The protection and safety monitoring system provides a safety-related means of performing the following functions:
- Automatic and manual reactor trip
- Automatic and manual actuation of engineered safety features (ESF).
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Reference Design Feature Value tion 7.1.1 PMS provides for the minimum inventory of fixed position controls and displays in the control room.
tion 7.1.2 Each PMS division is powered from its respective Class 1E dc division.
tion 7.1.2 PMS has four divisions of reactor trip and ESF actuation.
tion 7.1.2.5 PMS has two divisions of safety-related post-accident parameter display.
tion 7.1.2.9 PMS automatically blocks an attempt to bypass more than one channel of a function that uses 2-out-of-4 logic.
tion 7.1.2.14 The PMS hardware and software are developed using a planned design process which provides for specific design documentation and reviews during the design requirement, system definition, development, test and installation phases.
tion 7.1.4.2 The ability of the protection and safety monitoring system to initiate and accomplish protective functions is maintained despite degraded conditions caused by internal events such as fire and flooding.
tion 7.1.2 The flexibility of the protection and safety monitoring system enables physical separation of redundant divisions.
tion 7.2.2.2.1 The protection and safety monitoring system initiates a reactor trip whenever a condition monitored by the system reaches a preset level.
tion 7.4.3 The PMS allows for the transfer of control capability from the 18.12.2 main control room to the remote shutdown workstation. The minimum inventory of displays and controls in the remote shutdown workstation is provided.
tion 7.3.1 The ADS valves are powered from Class 1E dc power.
8.3.2.1.1 tion 7.7.1.11 The ADS valves are automatically and manually actuated via the 7.3.1.2.4 protection and safety monitoring system (PMS), and manually actuated via the diverse actuation system (DAS).
tion 7.3.1.2.3 The CMT AOVs are automatically and manually actuated from 7.7.1.11 PMS and DAS.
tion 7.3.1.2.2 The squib valves and MOVs for injection and recirculation are 7.3.1.2.9 automatically and manually actuated via PMS, and manually 7.7.1.11 actuated via DAS.
14.3-43 Revision 1
Reference Design Feature Value ure 7.2-1 (Sheets 16 The squib valves and MOVs for reactor cavity flooding are and 20) manually actuated via PMS and DAS from the control room.
tion 7.3.1.2.7 The PRHR air-operated valves are automatically actuated and 7.7.1.11 manually actuated from the control room by either PMS or DAS.
tion 7.3.1.2.20 The RNS containment isolation MOVs are actuated via PMS.
tion 7.5.4 PMS has two divisions of safety-related post-accident parameter display.
tion 7.6.1.1 An interlock is provided for the normally closed motor-operated normal residual heat removal system inner and outer suction isolation valves. Each valve is interlocked so that it cannot be opened unless the reactor coolant system pressure is below a preset pressure.
tion 7.7.1.11 The diverse actuation system is a nonsafety-related system that provides a diverse backup to the protection and safety monitoring system.
tion 7.7.1.11 The diverse actuation system trips the reactor control rods and the turbine on low wide range steam generator water level, or on low pressurizer water level, or on high hot leg temperature.
tion 7.7.1.11 DAS manual initiation functions are implemented in a manner that bypasses the signal processing equipment of the DAS.
tion 7.7.1.11 The DAS automatic actuation signals are generated in a functionally diverse manner from the PMS signals. Diversity between DAS and PMS is achieved by the use of different architectures, different hardware implementations, and different software, if any.
Software diversity between the DAS and PMS will be achieved through the use of different algorithms, logic, program architecture, executable operating system, and executable software/logic.
14.3-44 Revision 1
Reference Design Feature Value tion 8.3.1.1.1 On loss of power to a 6900V diesel-backed bus, the associated diesel generator automatically starts and produces ac power. The source circuit breakers and bus load circuit breakers are opened, and the generator is connected to the bus. Each generator has an automatic load sequencer to enable controlled loading on the associated buses.
tion 8.3.1.1.2.1 Two onsite standby diesel generator units provide power to the selected nonsafety-related ac loads.
tion 8.3.1.1.4 The main ac power system distributes non-Class 1E power from onsite sources to selected nonsafety-related loads.
tion 8.3.2.1 The Class 1E dc and uninterruptible power supply (UPS) system (IDS) provides dc and uninterruptible ac power for the safety-related equipment.
tion 8.3.2.1.1.1 There are four independent, Class 1E 250 Vdc divisions.
Divisions A and D are each composed of one battery bank, one switchboard, and one battery charger. Divisions B and C are each composed of two battery banks, two switchboards, and two battery chargers. The first battery bank in the four divisions is designated as the 24-hour battery bank. The second battery bank in Divisions B and C is designated as the 72-hour battery bank.
tion 8.3.2.1.1.1 Battery chargers are connected to dc switchboard buses. The input ac power for the Class 1E dc battery chargers is supplied from onsite diesel-generator-backed low-voltage ac power supplies.
tion 8.3.2.1.1.1 The 24-hour battery banks provide power to the loads for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without recharging. The 72-hour battery banks supply a dc switchboard bus load for a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without recharging.
tion 8.3.2.1.2 The non-Class 1E dc and UPS system (EDS) consists of the electric power supply and distribution equipment that provides dc and uninterruptible ac power to nonsafety-related loads.
tion 8.3.2.1.2 EDS load groups 1, 2, 3, and 4 provide 125 Vdc power to the associated inverter units that supply the ac power to the non-Class 1E uninterruptible power supply ac system.
14.3-45 Revision 1
Reference Design Feature Value tion 8.3.2.1.2 Battery chargers are connected to dc switchboard buses. The input ac power for the non-Class 1E dc battery chargers is supplied from onsite diesel-generator-backed low-voltage ac power supplies.
tion 8.3.2.1.2 The onsite standby diesel-generator-backed low-voltage ac power supply provides the normal ac power to the battery chargers.
tion 8.3.2.4.2 Separation is provided between Class 1E divisions, and between Class 1E divisions and non-Class 1E cables.
tion 9.2.1 The service water system is a nonsafety-related system that transfers heat from the component cooling water heat exchangers to the atmosphere.
tion 9.2.1.2.1 The SWS is arranged into two trains. Each train includes one pump and one cooling tower cell.
tion 9.2.2 The component cooling water system is a nonsafety-related system that removes heat from various components and transfers the heat to the service water system (SWS).
tion 9.2.2.2 The CCS is arranged into two trains. Each train includes one pump and one heat exchanger.
tion 9.3.6 The CVS provides a nonsafety-related means to perform the following functions:
- Makeup water to the RCS during normal plant operation
- Boration following a failure of reactor trip
- Coolant to the pressurizer auxiliary spray line.
tion 9.3.6.1.1 The chemical and volume control system (CVS) provides a safety-related means to terminate inadvertent RCS boron dilution.
tion 9.4.1 The main control room has its own ventilation system and is pressurized. The ventilation system for the remote shutdown room is independent of the ventilation system for the main control room.
tion 9.5.1.2.1.1 The PMS allows for the transfer of control capability from the main control room to the remote shutdown workstation. The minimum inventory of displays and controls at the remote shutdown workstation is provided.
14.3-46 Revision 1
Reference Design Feature Value tion 9.5.1.2.1.1 Class 1E divisional cables are routed in their respective divisional raceways.
tion 9.5.1.2.1.1 Separation is maintained between Class 1E divisions and between Class 1E divisions and non-Class 1E cables in accordance with the fire areas.
tion 17.4.1 D-RAP provides reasonable assurance that the design of risk-significant SSCs is consistent with their PRA assumptions.
tion 18.8.3.2 The main control area includes the reactor operator workstations, the supervisors workstation, the dedicated safety panel and the wall panel information system.
tion 18.12.2 The minimum inventory of instrumentation includes those displays, controls, and alarms that are used to monitor the status of the critical safety functions and to manually actuate the safety-related systems that achieve the critical safety functions. The minimum inventory resulting from the implementation of the selection criteria is provided in Table 18.12.2-1.
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Reference Design Feature Value le 2.0-201 Plant elevation for maximum flood level (ft) 100 tion 2.3.4 Atmospheric dispersion factors - /Q (sec/m3)
- Site Boundary /Q 5.1 x 10-4 0 - 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time interval
- Low Population Zone Boundary /Q 2.2 x 10-4 0 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.6 x 10-4 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.0 x 10-4 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 8.0 x 10-5 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> le 6.2.3-1 Containment penetration isolation features are configured as in Table 6.2.3-1 le 6.2.3-1 Maximum closure time for remotely operated containment purge 10 valves (seconds) le 6.2.3-1 Maximum closure time for all other remotely operated 60 containment isolation valves (seconds) tion 6.4.2.3 The minimum storage capacity of all storage tanks in the VES 327,574 (scf) eted tion 6.4.4 The maximum temperature in the instrumentation and control 120 rooms and dc equipment rooms following a loss of the nuclear island nonradioactive ventilation system remains over a 72-hour period (°F).
tion 6.4.4 The main control emergency habitability system nominally 65 +/- 5 provides 65 scfm of ventilation air to the main control room from the compressed air storage tanks.
tion 6.4.4 Sixty-five +/- five scfm of ventilation flow is sufficient to pressurize 1/8th the control room to 1/8th inch water gauge differential pressure (WIC).
tion 6.4.5.1 The maximum temperature in the main control room pressure 95 boundary following a loss of the nuclear island nonradioactive ventilation system over a 72-hour period (°F) (dry bulb temperature) ure 6.4-2 The main control room emergency habitability system consists of two sets of emergency air storage tanks and an air delivery system to the main control room.
tion 6.5.3 The passive heat removal process and the limited leakage from the containment result in offsite doses less than the regulatory guideline limits.
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Reference Design Feature Value tion 8.3.1.1.6 Electrical penetrations through the containment can withstand the maximum short-circuit currents available either continuously without exceeding their thermal limit, or at least longer than the field cables of the circuits so that the fault or overload currents are interrupted by the protective devices prior to a potential failure of a penetration.
ction 9.4.1.1.1 The VBS isolates the HVAC ductwork that penetrates the main control room boundary on High-2 particulate or iodine concentrations in the main control room supply air or on extended loss of ac power to support operation of the main control room emergency habitability system.
tion 12.3.2.2.1 During reactor operation, the shield building protects personnel occupying adjacent plant structures and yard areas from radiation originating in the reactor vessel and primary loop components. The concrete shield building wall and the reactor vessel and steam generator compartment shield walls reduce radiation levels outside the shield building to less than 0.25 mrem/hr from sources inside containment. The shield building completely surrounds the reactor components.
tion 12.3.2.2.2 The reactor vessel is shielded by the concrete primary shield and by the concrete secondary shield which also surrounds other primary loop components. The secondary shield is a structural module filled with concrete surrounding the reactor coolant system equipment, including piping, pumps and steam generators. Extensive shielding is provided for areas surrounding the refueling cavity and the fuel transfer canal to limit the radiation levels.
tion 12.3.2.2.3 Shielding is provided for the liquid radwaste, gaseous radwaste and spent resin handling systems consistent with the maximum postulated activity. Corridors are generally shielded to allow Zone II access, and operator areas for valve modules are generally Zone II or III for access. Shielding is provided to attenuate radiation from normal residual heat removal equipment during shutdown cooling operations to levels consistent with radiation zoning requirements of adjacent areas.
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Reference Design Feature Value tion 12.3.2.2.4 The concrete shield walls surrounding the spent fuel cask loading and decontamination areas, and the shield walls surrounding the fuel transfer and storage are sufficiently thick to limit radiation levels outside the shield walls in accessible areas to Zone II. The building walls are sufficient to shield external plant areas which are not controlled to Zone I.
tion 12.3.2.2.5 Shielding is provided as necessary for the waste storage areas in the radwaste building to meet the radiation zone and access requirements.
tion 12.3.2.2.7 Shielding combined with other engineered safety features is provided to permit access and occupancy of the control room following a postulated loss-of-coolant accident, so that radiation doses are limited to five rem whole body from contributing modes of exposure for the duration of the accident, in accordance with General Design Criteria 19.
tion 12.3.2.2.9 The spent fuel transfer tube is shielded to within adjacent area radiation limits, is completely enclosed in concrete, and there is no unshielded portion of the spent fuel transfer tube during the refueling operation.
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Reference Design Feature Value tion 1.2 The discharge from the IRWST vents located in the roof of the IRWST next to the containment vessel are oriented away from the containment vessel.
tion 5.3.1.2 There are no penetrations in the reactor vessel below the core.
tion 5.3.5 The reflective reactor vessel insulation provides an engineered flow path to allow the ingression of water and venting of steam for externally cooling the vessel.
- A flow path exists from the loop compartment to the reactor 6 vessel cavity (ft2).
- A flow path area to vent steam exists between the vessel 12 insulation and the reactor vessel (ft2).
tion 6.2.4.2.3 The hydrogen ignition subsystem consists of 64 hydrogen igniters strategically distributed throughout the containment.
le 6.2.4-3 The minimum surface temperature of the hydrogen igniters (°F). 1,600 tion 6.3 The ADS provides a safety-related means of depressurizing the RCS.
tion 6.3 The PXS provides a safety-related means of flooding the reactor cavity by draining the IRWST into the containment.
tion 7.3.1.2.9 Signals to align the IRWST containment recirculation isolation valves are generated by manual initiation.
tion 7.7.1.11 Initiation of containment recirculation is a diverse manual function.
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4.1 Organization and Staffing staff, staff responsibilities, authorities, and personnel qualifications for performing the AP1000 al test program are addressed in Section 14.2. This test organization is responsible for the ning, executing, and documenting of the plant initial testing and related activities that occur ween the completion of plant/system/component construction and commencement of plant mercial operation. Transfer and retention of experience and knowledge gained during initial ing for the subsequent commercial operation of the plant is an objective of the test program.
4.2 Test Specifications and Procedures operational and startup test specifications and procedures are provided to the NRC in accordance the requirements of Subsection 14.2.3. The controls for development of test specifications and edures are also described in Subsection 14.2.3.
oss reference list is provided between ITAACs and test procedures and/or sections of test edures.
4.3 Conduct of Test Program te-specific startup administration manual (procedure), which contains the administration edures and requirements that govern the activities associated with the plant initial test program, escribed in Section 14.2 is provided.
4.4 Review and Evaluation of Test Results iew and evaluation of individual test results, as well as final review of overall test results and cted milestones or hold points are addressed in Subsection 14.2.3.2. Test exceptions or results do not meet acceptance criteria are identified to the affected and responsible design anizations, and corrective actions and retests, as required, are performed.
4.5 Interface Requirements Test Specifications and acceptance criteria required of structures and systems which are outside scope of the design certification are addressed in Subsections 14.2.9.4.15, 14.2.9.4.22 through
.9.4.27, 14.2.10.4.29, and in the Physical Security Plan.
4.6 First-Plant-Only and Three-Plant-Only Tests e COL holder for the first plant and the first three plants will perform the tests listed in section 14.2.5. For subsequent plants, either tests listed in subsection 14.2.5 shall be performed, he COL applicant shall provide a justification that the results of the first-plant-only tests or first-e-plant tests are applicable to the subsequent plant.]*
Staff approval is required prior to implementing a change in this information.
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ided prior to preoperational testing.
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butes upon which the NRC relies, in a limited number of technical areas, in making a final safety rmination to support a design certification. DAC is to be objective (measurable, testable, or ject to analysis using pre-approved methods), and must be verified as a part of the ITAAC ormed to demonstrate that the as-built facility conforms to the certified design (SECY 92-053).
re are three process options for DAC/ITAAC resolution:
Resolve through amendment to design certification Resolve as part of COL review Resolve after COL is issued e first two options, the applicant will submit the design information and the NRC will document its ew in a safety evaluation. In the third option, the COL holder notifies the NRC of availability of ign information and the staff will document its review in an inspection report.
uld the third option be implemented for the first standard AP1000 plant, subsequent COL licants may reference the first standard plant closure documentation and close the DAC/ITAAC er the concept of one issue, one review, one position, identified in NRC guidance.
itionally, Westinghouse may submit licensing topical reports for NRC review of the material porting the DAC/ITAAC closure and request that the NRC issue a safety evaluation in conjunction a closure letter or inspection report concluding that the acceptance criteria of the DAC/ITAAC e been met. Subsequent COL applicants may reference these reports and NRC closure uments in an effort to close the DAC/ITAAC.
technical areas where DAC/ITAAC applies in the design certification rule, COL applicants will ide an ITAAC and associated closure schedule indicating the approach to be applied. For sequent COL applicants following the first standard AP1000 plant, the indication could be to rence the existing DAC/ITAAC closure documentation for the first standard plant.
C guidance for DAC/ITAAC is provided in Regulatory Guide 1.206, Section C.III.5. Further rmation on the staffs position of DAC/ITAAC being used as part of the 10 CFR Part 52 review ess is provided in SECY-92-053.
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