ML17033B575

Redacted Updated Final Safety Analysis Report Chapter 14

ML17033B575
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Site:	North Anna
Issue date:	09/29/2016
From:	V Sreenivas Plant Licensing Branch II
To:	Heacock D Virginia Electric & Power Co (VEPCO)
Sreenivas V, NRR/DORL/LPL2-1, 415-2597
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North Anna Power Station Updated Final Safety Analysis Report Chapter 14 Intentionally Blank

Intentionally Blank Revision 52-09/29/2016 NAPS UFSAR 14-iChapter 14: Initial Tests and OperationTable of ContentsSectionTitle Page14INITIAL TESTS AND OPERATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.0-114.0.1Administration of the Preoperational Test Program. . . . . . . . . . . . . . . . . . . . . .

14.0-114.0.2Administration of the Start-Up Test Program. . . . . . . . . . . . . . . . . . . . . . . . . . .

14.0-214.1TEST PROGRAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.1-114.1.1Pre-Operational Test Program. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.1-114.1.2Initial Start-Up Test Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.1-2 14.1.2.1Initial Fuel Loading. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.1-214.1.2.2Initial Postloading Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.1-414.1.2.3Initial Criticality and Low-Power Physics Tests. . . . . . . . . . . . . . . . . . . . . . .14.1-5 14.1.2.4Power Level Escalation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.1-614.1.3Start-Up Physics Test Program Differences Between Unit 1 and Unit 2 . . . . . .14.1-714.1.4Special Low-Power Tests - Unit 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.1-814.1Reference Drawings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.1-914.2AUGMENTATION OF VEPCO'S STAFF FOR INITIAL TEST AND OPERATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-1 Appendix 14ANRC Questions and VEPCO's Responses Regarding the North Anna Power Station Unit 2 Modified Startup Physics Testing Program. . . . . . . . . . . . . . . . . . . . . . .14A-i Revision 52-09/29/2016 NAPS UFSAR 14-iiChapter 14: Initial Tests and OperationList of TablesTableTitle PageTable 14.1-1List of Preoperational Tests. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.1-10Table 14.1-2Lists of Start-Up Tests. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.1-34Table 14.1-3Unit 2 Start-Up Physics Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.1-45 Table 14.1-4Physics Tests that Have Been Deleted For Unit 2. . . . . . . . . . . . . . . . .14.1-46 Table 14.1-5Summary of Unit 1 Measured Values, Design Values, Design Tolerance, and Accident Analysis Crit eria for Physics Tests That Have Been Deleted for Unit 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.1-47Table 14A-1Unit 2 Isotherm al Temperature Coefficient, Boron Endpoint, Rod Worth Reactivity, and Boron Worth Tests and Review Criteria. . . . . . . . . . . .14A-7 Revision 52-09/29/2016 NAPS UFSAR 14-iiiChapter 14: Initial Tests and OperationList of Figures FigureTitle PageFigure 14.1-1Typical Pre-Operational Test Sequence. . . . . . . . . . . . . . . . . . . . . . . .14.1-50Figure 14.1-2Typical Start-Up Test Sequence . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.1-51Figure 14.1-3Cycle 1 BOL Physics Test Doppler Power Coefficient Used in Accident Analysis Unit 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.1-52Figure 14A-1Secondary Source Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14A-8Figure 14A-2Part Length Control Rod Locations (Unit 1). . . . . . . . . . . . . . . . . . . .14A-9 Revision 52-09/29/2016 NAPS UFSAR 14-iv Intentionally Blank

Intentionally Blank Revision 52-09/29/2016 NAPS UFSAR 14.0-1CHAPTER 14INITIAL TESTS AND OPERATIONThis chapter describes the sc ope of tests and operations pe rformed over the time period when construction was sufficiently complete to operate and test individual components and systems through the acceptance run at full power. Th is time period is divi ded into two categories:1.Pre-operational testing: tests performed before the initial core loading.2.Initial start-up testing: tests and operations from the initial core loading through the acceptance tests.

The preoperational and start-up programs, as outlined in Ta bles 14.1-1 and 14.1-2 , comply with the intent of Regulatory Guide 1.68, Preoperational and Initial Start-up Test Programs for Water-Cooled Power Reactors , dated November 1973, in most cases, and use the same wording as much as possible in order to more clearly address the NR C guide requirem ents. Detailed acceptance criteria were provided in each test procedure that was written to fulfill the testing requirements. The detailed criter ia of acceptability were based on various sources, such as equipment technical manuals, system descriptions, plant drawings, manufacturer specifications, and the North Anna Units 1 and 2 FSAR. The tests and their objectives are listed in Ta bles 14.1-1 and 14.1-2, which also provide a summary of each test. The acceptability of a test is contingent on the successful attainment of the objectives stated in Ta bles 14.1-1 and 14.1-2.Because of similarities and differences in the fuel and co re characteristics between the two units, certain tests performed for Unit 1 were not repeated for Unit 2, while specific tests were performed for Unit 2 only. A discussion of the start-up physics program diff er ences appears in Section 14.1.3.14.0.1 Administration of the Preoperational Test Program The management and direction of the preoperational test program was under the direct control of VEPCO, with the principal responsib ility lying with the Supervisor - Engineering Services. In most cases written preoperational test pr ocedures were prepar ed by the station engineering staff under the direction of the Supe rvisor - Engineering Services. In those areas where the station engineering staff was not know ledgeable, proced ures were provided by the architect-engineer or outside consultants, based on their expertise in the particular areas of concern. Test procedure format generally included the purpose of the test, in itial condition requirements, precautions and limi tations, instructions, and criteria for acceptability of data. Prior to issuance of test procedures fo r use in the field, they were reviewed by the Joint Test Group and approved by the Station Nuclear Safety and Operating Committee.

For those procedures provided by the architec t-engineer or outside consultants, the preoperational test procedure was used as a cover sheet to their proce dure in order to ensure review by the Joint Test Group and approval by the Station Nuclear Safety and Operating Revision 52-09/29/2016 NAPS UFSAR 14.0-2Committee. In some instances the preoperational te st procedures were used to review and approve test data from testing performed by equipment vendors off the site (e.g., vendor certifications).

In most instances the conduct and direction of th e preoperational test s were the direct responsibility of the VEPCO test engineers designated by the techni cal supervision at the station. In some instances architect-engineer personnel or outside consultants we re responsible for the conduct of tests under the dire ction of VEPCO by means of written administrative controls.

Changes to approved test procedures were documen ted and became part of the final test results.

Administrative controls for making changes to procedures prepar ed by the station engineering staff were provided in the Nuclear Power Station Quality Assu rance Manual. Administrative procedures for making changes to procedures provided by the architect-engineer or outside consultants were formulated by the architect-engineer a nd approved by VEPCO.

For preoperational testing, the Supervisor - Engineering Services and the Joint Te st Group reviewed and analyzed the test results. Assistance from the VEPCO system office, the nuclear steam supply system vendor, and the architect-engineer was solicited as deemed necessary. The test results and evaluations were reviewed by the Station Nuclear Safety and Operating Committee and approved if they were satisfactory. In inst ances where performance of components or systems deviated from predicted results, furthe r engineering evaluations were made to resolve the discrepancies before the test was considered satisfactory. Systems that had to be modified as a result of the preoperational tests were then retested to verify acceptable performance.

The completed test procedures , along with data and conclu sions, were documented and filed as part of the permanent plant records.

Minimum qualifications for the VEPC O test engineers were as follows:1.A bachelor's degree in engin eering or the physical sciences or the equivalent, and at least 1 year of applicable nuclear power plant experience, or:2.A high school diploma or the equivalent, and at least 3 years of applicable nuclear power plant experience. Credit for up to 2 years of nuclear experience may be given for related technical training on a one-for-one time basis.

Additional information relative to the preope rational test progra m is provided in the Nuclear Power Station Quality Assurance Manual, in VEPCO stat ion administrative procedures, and in the architect-engineer's administrative procedures.

14.0.2 Administration of the Start-Up Test Program The management and direction of the start-up test program ha s been under the direct control of VEPCO, with principal res ponsibility lying with the Supervisor - Engineering Services. Wr itten start-up test procedures were prepared by the station reactor engineers under the direction Revision 52-09/29/2016 NAPS UFSAR 14.0-3of the Supervisor - Engineering Services. Procedures from the Nuclear St eam Supply System Start-up Manual and assistance from Westinghouse personnel were ut ilized in many cases. Prior to issuance of test procedures for use in the fi eld, they were approved by the Station Nuclear Safety and Operating Committee.

The conduct and directi on of the start-up tests were th e responsibility of the reactor engineers designated by the Supervisor - Engineering Servi ces. Changes to approved test procedures were documente d and became part of the final results. Administ rative procedures for making these changes, includin g the review and approvals, we re formulated and utilized by VEPCO.For start-up testing the react or engineers and the Supervisor - Engineering Services reviewed and analyzed th e test results. The measurements a nd data analysis for start-up physics tests were performed by the VEPCO Fuel Resources Departme nt. Assistance from the VEPCO system office, the nuclear steam supply system vendor , and the archit ect-engineer wa s solicited as deemed necessary. Approval of test results was the responsibility of the Station Nuclear Safety and Operating Committee. The completed test pr ocedures, along with data and conclusions, were documented and filed as part of the permanent plant records.

The minimum qualifications for the reactor e ngineers, in terms of educational background and experience, are stated in Section 13.1. Minimum qualifications for the test engineers responsible for the preparat ion and performance of start-up tests were as follows:1.A bachelor's degree in engineering or the physical sciences or the equivalent and 2 years of applicable power plant experi ence, of which at least 1 year shall be appli cable nuclear power plant experience, or2.A high school diploma or the equivalent and 5 years of applicable power plant experience, of which at least 2 years shall be applicable nuclear power plant expe rience. Credit for up to 2 years of non-nuclear experi ence may be given for rela ted technical training on a one-for-one time basis.

Additional information relative to the start-up test progra m is provided in the Nu clear Power Station Quality Assurance Manual a nd in station administrative procedures.

Revision 52-09/29/2016 NAPS UFSAR 14.0-4 Intentionally Blank Revision 52-09/29/2016 NAPS UFSAR 14.1-1The following information is HISTORICAL and is not intended or expec ted to be updated for the life of the plant.14.1 TEST PROGRAM14.1.1 Pre-Operational Test ProgramThe pre-operational test program included tests, adjustments, calibrations, and system operations necessary to ensure that initial fuel loading, initial criticality , and subsequent power operation could be safely undertaken.

After installation of individual components and systems was comp leted, the installed components and systems were tested and evalua ted according to approve d testing procedures or check-off lists. Analyses of test results were made to verify that systems and components were performing satisfactorily or , if not, to provide a basis fo r recommended corrective action.

Whenever possible, these tests were perfo rmed under the same conditions to be experienced under subsequent station opera tions. During system tests for which unit parameters were not available, the systems were operationally tested as far as possible without these parameters. The remainder of the tests were performed under plan t conditions when the parameters were available. Ab normal unit conditions were simula ted during testing as required and when such conditions did no t endanger personnel or equipment, or contaminate systems

whose cleanliness had been established.

In general, pre-operational testing was comp leted before core loading. As individual systems were completed, pre-opera tional tests were performed to verify as nearly as possible the performance of the system under actual operating conditions. Where required, simulated signals or inputs were used to ve rify the full operating range of the system and to calibrate and align the systems and instruments at these conditions. Later, systems th at were used during normal operation were verified under actual operating conditions. Systems that are not used during normal plant operation, but s hould be in a state of readine ss to perform safety functions, were tested before plant start-up. Examples of these systems are the reactor trip system and engineered safety features system logic, operation checks, and se tpoint verifications.Testing performed during the pr e-operational test program is outlined in Table 14.1-1. A typical sequence of performance for operational test s is shown in Figure 14.1-1. The actual sequence of tests was formulated before the performance of the tests, considering equipment and system availability. In some cases, it was necessary to complete certain pre-operational tests after core loading. These included such tests as those performed on the complete rod control system, rod position indication, and complete incore mo vable detector system. These tests have been identified in Table 14.1-1.

Revision 52-09/29/2016 NAPS UFSAR 14.1-214.1.2 Initial Start-Up Test Program Fuel loading was begun when all prerequisite system tests and operations were satisfactorily completed. Upon completion of fuel loading, the reactor upper internals and pressure vessel head were installed, and additional mechanical and electrical tests were performed as discussed in pre-ope rational testing. The purpose of this phase of activities was to prepare the system for nuclear operation and to establish that all design requirements necessary for operation were achieved. The core-loading and postloading tests are described below.

14.1.2.1 Initial Fuel Loading The reactor containment structure was completed and tested before initial fuel loading.

Fuel-handling tools and equipment were checked out and dry runs con ducted in the use and operation of equipment.

The reactor vessel and associated components were in a state of readiness to receive fuel. Water level was maintained above the bottom of the nozzles.

The overall responsibility and direction for the initial core loading was exercised by the Station Manager assisted by the Superintendent - Station Op eration. The overall process of initial core loading was, in general, directed from the operating floor of the containment structure. Procedures for the control of personne l and the maintenance of containment security were in effect during initial fuel loading.The as-loaded core configuration was specif ied as part of the core design studies conducted in advance of core loading. In the event mechan ical damage to a fuel assembly occurred during core-loading ope rations, an evaluation would have been performed and a replacement assembly would have been procured if deemed necessary.The core was assembled in the reactor vessel, containing reactor-grade water with dissolved boric acid to maintain a calculated core effective multiplication factor of 0.95 or lower. The refueling cavity was kept dry duri ng the initial core loading. Core moderator chemistry conditions (particularly, boron concentration) were prescribed in the core-loading procedure document and were verified periodically by chemical analyses of moderator samples taken before and during core-loading operations.The following information is HISTORICAL and is not intended or expec ted to be updated for the life of the plant.

Revision 52-09/29/2016 NAPS UFSAR 14.1-3Core-loading instrumentatio n consisted of two permanently instal led source range (pulse-type) nuclear channels and three te mporary incore source range channels. The permanent channels, when responding, were mon itored in the main cont rol room by licensed station operators; the temporary channels were monitored by fuel-loading personnel. One permanent channel was equipped with an audible count rate indicator. The neutron flux level from both plant channels was di splayed on a strip chart recorder. The temporary channels were indicated on rate meters with one channel recorded on a strip chart recorder. Minimum count rates of two counts per sec, attributable to core neutrons, were required on at least two of the five available nuclear source channels at all times following instal lation of the initial nucleus of eight fuel assemblies.

Fuel assemblies together with inserted components (contro l rod assemblies, burnable poison inserts, source spider, or thimble plugging device s) were placed in th e reactor vessel one at a time according to a previously establis hed and approved sequence developed to provide reliable core monitoring with minimum possibility of core mechan ical damage. The core-loading procedure documents included deta iled tabular check sheets that prescribed and were used to verify the successi ve movements of each fuel asse mbly and its specified inserts from its initial position in the storage racks to its final position in the core. Multiple checks were made of component serial numbers and types at successive transfer points to guard against possible inadvertent ex changes or substitutio ns of components, and fuel assembly status boards were maintained th roughout the core-loading operation.

An initial nucleus of eight fuel assemblies, the first of which contained an activated neutron source, is the minimum source-fuel nuc leus that permits subsequent meaningful inverse count rate ratio monitori ng. This initial nucleus has been determined by calculation and previous experience to be markedly subcritical ( less than or equal to 0.90) under the required conditions of loading.

Each subsequent fuel addition was acco mpanied by detailed neutron count rate monitoring to determine that the just-loaded fu el assembly did not ex cessively increase the count rate and that the extrapolated inverse c ount rate ratio was not decreasing for unexplained reasons. The results of each lo ading step were evaluated before the next prescribed step was started.Criteria for safe loading require that loading operations stop immediately if:1.An unanticipated increase in the neutron count rates by a factor of two occurs on all responding nuclear channels during any single lo ading step after the initial nucleus of eight fuel assemblies is loaded (excluding anticipated changes due to detector and/or source movement).The following information is HISTORICAL and is not intended or expec ted to be updated for the life of the plant.

k eff Revision 52-09/29/2016 NAPS UFSAR 14.1-42.The neutron count rate on any individual nucle ar channel increases by a factor of five during any single loading step after the initial nucleus of eight fuel assemblies is loaded (excluding anticipated cha nges due to detector and/

or source movements).

An alarm in the containment and main cont rol room is coupled to the source range channels with a setpoint equal to or less than five times the baseline count rate. This alarm automatically alerts the loading operation personnel of high count rate and requires an immediate stop of all operations unt il the situation is evaluated.

Core-loading procedures specif ied the condition of fluid systems to prevent inadvertent dilution of the reactor coolant, specified the movement of fuel to preclude the possibility of mechanical damage, prescribed the conditions under which loading could proceed, identified responsibility and authority, and provided fo r continuous and complete fuel and core component accountability.

14.1.2.2 Initial Postloading Tests Upon completion of core loading, the reactor upper internals and the pressure vessel head were installed, and additional mechanical and electrical checks were performed before initial criticality. The final pressure test was conducted after filling and venting were completed to check the integrity of the vessel head installation.

Mechanical and electrical te sts were performed on the co ntrol rod drive mechanisms.

These tests included a complete operational checkout of the mech anisms and calibration of the individual rod position indication.Tests were performed on the reactor trip circui ts to test manual trip operation. The actual control rod assembly drop times were measured for each control rod assembly. The reactor control and protection system wa s checked with simula ted signals to produce a trip signal for the various conditions that require plant trip.At all times when the control rod drive mechanisms were being tested, the boron concentration in the coolant-moderator was main tained such that the reactor would remain adequately shut down with all cont rol rod assemblies fully withdrawn.

A complete functional electrical and mechanic al check was made of the incore nuclear flux mapping system, and reactor coolant system flow measurements were take n to relate reactor coolant pump input power and elbow tap pressure differen tial to actual reactor coolant loop flow.The following information is HISTORICAL and is not intended or expec ted to be updated for the life of the plant.

Revision 52-09/29/2016 NAPS UFSAR 14.1-5 14.1.2.3 Initial Criticality and Low-Power Physics Tests On completion of postloading te sts, nuclear operation of the reactor was begun. This final phase of start-up and testing included initial criticality, low-power tes ting, and power level escalation. The purpose of these tests was to establish the operational characteristics of the unit and core, to acquire data for the proper calibration of se tpoints, and to ensure that operation was within license requirements. A brief description of the testing is presen ted in this section. Table 14.1-2 summarizes the major tests that were performed from initial core loading to rated power. Figure 14.1-2 depicts a typical sequence for these tests; the actual sequence of tests was formulated by station engineeri ng and operating personnel, consid ering test requirements and equipment availability.Initial criticality was estab lished by sequentially withdraw ing the shutdown and control banks of control rod assemblies from the core, l eaving the last withdraw n control bank inserted far enough in the core to provide effective contro l when criticality would later be achieved, and then diluting the heavily borated reactor coolant until criticality was achieved.

Successive stages of control rod assembly ba nk withdrawal and of boron concentration dilution were monitored by obse rving changes in neut ron count rate as indicated by the normal plant source range nuclear instru mentation as functions of ba nk position during rod motion and, subsequently, of reactor coolant boron concentr ation and primary-water addition to the reactor coolant system during dilution.

Throughout this period, sample s of the primary coolant were obtained and analyzed for boron concentration.

Inverse count rate ratio monito ring was used as an indication of the proximity and rate of approach to criticality of th e core during control rod assemb ly bank withdrawal and during reactor coolant boron dilution. The rate of dilution was reduced as the reactor approached the boron concentration extrapolated for criticality to ensure that effective control was maintained at all times. Written procedures specified the plant conditions, precautions, and specific instructions for the approach to criticality.After initial criticality, a prescribed program of reactor physics measurements was undertaken to verify that the basi c static and kinetic characteristics of the core were as expected and that the values of the kinetic coefficients assumed in the safeguards analysis were indeed conservative.The following information is HISTORICAL and is not intended or expec ted to be updated for the life of the plant.

Revision 52-09/29/2016 NAPS UFSAR 14.1-6The measurements were made at low po wer and primarily at or near operating temperature and pressure. The measurements included verification of calculated values of control rod assembly bank reactivity worths, of isothermal temperature coefficient under various core conditions, of differential boron concentration reactivity worth, and of critical boron concentrations as functions of control rod assembly bank configuration.

In addition, measurements of the relative power distributions were made.

Concurrent tests were conducted on the instrumentation, including the source and intermediate range nuclear channels.Procedures were prepared to specify the sequence of test s and measurements conducted and the conditions under which each was to be performed to ensure both safety of operation and the validity and cons istency of the results ob tained. Had significant deviations from design predictions existed, or had unacceptable behavior be en revealed, or had apparent anomalies developed, then testing would have been suspe nded and the situation reviewed to determine whether a question of safety was invol ved before the resumption of testing.

14.1.2.4 Power Level Escalation When the operating characteristi cs of the reactor and unit were verified by low-power testing, a program of power level escalation in successive stages was used to bring the unit to its full rated ther mal power level. Both pr imary and secondary operati onal characteristics were examined at each stage of the power escalation program.

Measurements were made to determine the re lative power distribu tion in the core as functions of power level and control assembly bank position.

Secondary system heat balances ensured that the indica tions of power level were consistent and provided bases for calibration of the power range nuclear channels. The ability of the reactor coolant system to respond eff ectively to signals from primary and secondary instrumentation under a variety of conditions encountered in no rmal operations was verified.

At prescribed power levels the dynamic res ponse characteristics of the reactor coolant and steam systems were evaluated. The responses of the systems were measured for design step load changes of 10%, rapid 50% load re duction, and 50% and 100% power plant trips.

Adequacy of radiation shielding was de termined by gamma and neutron radiation surveys at selected points inside the containment and the outside area immediately adjacent to the containment at various power levels. Pe riodic sampling was performed to verify the chemical and radiochemical anal yses of the reactor coolant.The following information is HISTORICAL and is not intended or expec ted to be updated for the life of the plant.

Revision 52-09/29/2016 NAPS UFSAR 14.1-7All precritical tests were completed and the re sults evaluated before initial criticality.

Prerequisites for performing a test were spec ified in the individual test procedure. The sequence of testing was outlined in a start-up test sequence, such that required prerequisite testing was completed before s ubsequent testing. Any special te st instruments required were specified to be installed, calibrated, and checked in the test procedure that specified the test equipment.14.1.3 Start-Up Physics Test Program Differences Between Un it 1 and Unit 2 After the initial start-up physics program for North Anna Unit 1 was completed, several changes to the program were made before the initial start-up physics program for Unit 2. Table 14.1-3 lists the physics tests that were performed as part of the Unit 2 start-up program. These tests were chosen to:1.Verify that the core was correctly loaded and that there were no anomalies present that could cause problems later in the cycle.2.Verify that the calculational model that had been used would correctly predict core behavior during the cycle.3.Verify the reactivity worth of the control rod banks.

4.Provide data for nuclear instrumentation calibration.

5.Demonstrate the sensitivity of this in strumentation to abnormal core conditions.

In addition, the chosen tests were selected to encompass the physics test goals listed in the NRC Branch Technical Position DOR-1, Guidance for Reload Submittals, Draft - Spring, 1978. Table 14.1-4 lists those physics tests that were performed during the Unit 1 start-up, and that were not repeated as part of the Unit 2 st art-up program. The deletion of these tests was justified for the following reasons:1.The successful performance of the abbreviated program was sufficient to achieve the physics testing program goals.2.The calculational model was verified as a result of the Unit 1 start-up.3.The fuel and core characteristics of Unit 2 are virtually identical to those of Unit 1, and the results obtained for these tests during the Unit 1 start-up demonstrated that a large margin exists between the measured parameter values and the design values used in the accident analyses. Evidence of this is shown in Table 14.1-5.

When the modified test program for Unit 2 was proposed, se veral questions were raised by the NRC relating to the modifications. The questions and VEPCO's responses are the subject of Appendix 14A.The following information is HISTORICAL and is not intended or expec ted to be updated for the life of the plant.

Revision 52-09/29/2016 NAPS UFSAR 14.1-814.1.4 Special Low-Power Tests - Unit 2 This test program consisted of a series of natural circulati on tests that demonstrated the plant's cooldown capability in se veral simulated degraded modes of operation at power levels of up to 3% of rated thermal power.The objectives of the above tests and the methods used are described below.1.Natural Circulation TestObjective: To demonstrate the capability to remove decay heat by natural circulation.

Method: The reactor is at a pproximately 3% power and al l reactor coolant pumps are operating. All reactor coolant pum ps are tripped simultaneously , with the establishment of natural circulation indicated by the core exit thermocouples a nd the wide-range resistance temperature detectors.2.Natural Circulation With Simulated Loss of Offsite PowerObjective: To demonstrate that following a loss of offsite ac power, natural circulation can be established and maintained while being powered from the emergency diesel generators.

Method: The reactor is at a pproximately 3% power and al l reactor coolant pumps are operating. All reactor coolant pumps are tripped and a sta tion blackout is simulated. Alternating current power is returned by the diesel generators and natural circulation is verified.3.Natural Circulation With Loss of Pressurizer HeatersObjective: To demonstrate the ability to maintain natural circulation and saturation margin with the loss of pressurizer heaters.

Method: Establish natural circulation as in Test 1, and turn off the pressurizer heaters at the main control board. Monitor th e system pressures to determine the saturation margin, the depressurization rate, and the effects of charging/letdown flow and steam generator pressure on the saturation margin.4.Effect of Steam Generator Secondary-Side Isolation on Natural CirculationObjective: To determine the effects of steam generator secondary-side isolation on natural circulation.

Method: Establish natural circulation conditions as in Test 1 but at 1% power. Isolate the feedwater and steam line for one steam genera tor and establish equ ilibrium. Return the steam generator to service.5.Natural Circulation at Reduced PressureThe following information is HISTORICAL and is not intended or expec ted to be updated for the life of the plant.

Revision 52-09/29/2016 NAPS UFSAR 14.1-9 14.1 REFERENCE DRAWINGSThe list of Station Drawings below is provided for information only. The referenced drawings are not part of the UFSAR. This is not intended to be a complete listing of all Station Drawings referenced from this section of the UFSAR. The contents of St ation Drawings are controlled by station procedure.

Drawing Number Description 1.11715-LSK-1-3AGenerator Breaker Closing 2.11715-LSK-1-3BLogic Diagram: Power Circuit Breaker Opening 3.11715-LSK-1-3CLogic Diagram: "

86" Protective Lockout Relays 4.11715-LSK-1-3DLogic Diagram: Main Tr ansformer Protection Relays, 86-TL 5.11715-LSK-1-3ELogic Diagram: Generator Lockout Relays 6.11715-LSK-1-3FLogic Diagram: Main Transformer Diff erential Lockout Relays, 86-GL & 86-PWIA 7.11715-LSK-1-3GLogic Diagram: Main Transformer Coolers 8.11715-LSK-1-2ALogic Diagram: External Turbine Trips, Sheet 1 9.11715-LSK-1-3HLogic Diagram: Main Transformer Alarms 10.11715-LSK-1-2BLogic Diagram: Turbine Trips, Sheet 211.11715-LSK-1-2CLogic Diagram: Turbine Trips, Sheet 3 12.11715-LSK-1-2DLogic Diagram: Turbine Trips, Sheet 4 13.11715-LSK-1-2ELogic Diagram: Turbine Trips, Sheet 5 14.11715-LSK-1-2FLogic Diagram: Turbine Trips, Sheet 6 15.11715-LSK-1-2GLogic Diagram: Turbine Trips, Sheet 7Objective: To demonstrate the ability to maintain natural circ ulation at reduced pressure and saturation margin. The accuracy of the saturation meter will also be verified.Method: The test method is the same as for Test 3, with the exceptio n that the pressure decrease can be acceler ated with the use of auxiliary pr essurizer sprays. The saturation margin will be decreased to approximately 20°F.The following information is HISTORICAL and is not intended or expec ted to be updated for the life of the plant.

Revision 52-09/29/2016 NAPS UFSAR 14.1-10The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testing I. Plant Instrumentation1.Nuclear instrumentation (out of core)

Before core loading and initial criticality Before core loading, nuclear instru ments were aligned and source range detector response to a neutron source chec ked. Just before initial criticality all channels were checked to verify high-level trip functions , alarm setpoints, audible count rates where ap plicable, and opera tion of strip chart recorders and any auxiliary equipment.2.Process instrumentation (temperature, pressure, level, and flow instruments)

Ambient and/or at temperatureEquipment was aligned per manufacturer's instructions and applicable test procedures. Applicable alarm, trip, a nd control setpoints were checked for conformance with specified values.3.Seismic instrumentationBefore core loadingThis instrumentat ion was verified for correct installation and operability. A calibration record test was performed to verify operability of the magnetic tape playback units II. Reactor Coolant System1.Vibration and amplitudeBefore core loadingVibration sensors were placed on th e main coolant pumps and main cooling piping in order to check for excessive vibration while star ting and stopping the pumps2.Expansion and restraintBefore core loading, during heatup, and during cooldown.During the heatup to operating temper ature, selected points on cooldown components and piping of the reactor cool ant system were checked at various temperatures to verify unrestricted expansion. Points of interference detected during the heatup were corrected before increasing the temperature. Following cooldown to ambient temperature, the piping and components were checked to

confirm that they returned to their approximate base points.a.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-11 II. Reactor Coolant System (continued)3.Integrated hot functional tests Heatup, at temperature, and during cooldown.

(Hydrostatic testing has been satisfactorily completed and reactor coolant system instruments aligned and operational. Associated

auxiliary systems were to be operational.)The reactor coolant system was tested using pump heat to check heatup and cooldown procedures and to demonstrate satisfactorily performance of components and systems exposed to reactor coolant system temperature.

Proper operation of instrumentation, c ontrollers, and al arms was checked against operating conditions of auxili ary systems and setpoints verified.

Among the demonstrations performed were:a.To verify that water can be charged by the Chemical and Volume Control System at rated flow against normal reactor coolant pressures.b.To check letdown design flow rate for each operating mode.c.To check response of system to change in pressurizer level.

d.To check response procedures and components used in boric acid batching and transfer operations.e.To check operation of the excess le tdown and seal-water flow paths.f.To check steam generator level in strumentation response to level changes.g.To check thermal expansion of system components and piping.

h.To perform isothermal calibration of resistance temperature detectors and incore thermocouples.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-12 II. Reactor Coolant System (continued)i.To operationally check out the residual heat removal system.j.Visual and surface examination of the upper and lower core structures after the heatup and cooldown were completed (see item 6 below.)4.Component testsa.PressurizerAt operating temperatureDuring the hot functional testing, the pr essure-controlling capability of the pressurizer was demonstrated to be within the controlling band. After core loading, with the reactor coolant pumps operating and with full spray, the

pressure-reducing capability of the pressurizer was verified. With the spray secured and all heaters energized, the pressure-increasing ca pability of the pressurizer was verified. Expected rate s of pressure decrease/increase with tolerances were specified in the test procedures.b.Reactor coolant system pumps and motors At ambient conditions, during heatup, and at temperature As the pumps and motors were placed in operation they were checked for:a.Direction of rotation.b.Vibration.c.Power requirements.

d.Lubrication e.Cooling.

f.Recirculation flow.

g.Flow and pressure characteristics.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-13 II. Reactor Coolant System (continued)h.Megger and hi-pot tests (as applicable)i.Overload protection.

j.Correct power supply voltage During the reactor coolant system cold hydrostatic and hot f unctional tests, the pumps were operated to verify prope r installation. Follo wing core loading, measurements were made to determin e flow and input power relationships.c.Steam generatorsAt ambient conditions, during heatup, and at temperature. (The

secondary system had been satisfactorily hydrostatically tested.)

The proper operation of instrumentat ion and control systems for steam generators was checked dur ing heatup and at temperature. The heat transfer capability of the steam generators wa s demonstrated. The functioning of the blowdown system was also checked.d.Pressurizer relief and safety valvesPressure conditionsThe setpoints of the relief and safety valves were verified from vendor certification data, by bench te sts, or by in-plant tests. When verified by in-plant

tests, setpoints were checked by using a pr essure assist device that adds to the force due to pressure.Once the valve started to lift, this assi st device was vented, allowing the valve to reseat immediately. Following li fting and blowdown of any valve, the reseating of the valve was verified.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-14 II. Reactor Coolant System (continued)e.Main steam stop valves At operating temperature (steam flow no required)

At hot conditions and with pressure equalized across the valve, the operation of the main steam stop valves was verifie

d. The operating times were verified to be within expected values as specified by the test procedure.f.Reactor coolant system (RCS) loop isolation valves At ambient conditions and at operating temperature and pressure conditions At ambient and at hot conditions, the oper ation of the loop isolation valves was checked.g.Main steam stop valve pipingBefore core loadingMain steam stop valve piping was checked for excessive vibration while closing the main steam stop valves with steam availabl e from the heatup of the reactor coolant by the reactor coolant pumps.h.Pressurizer relief valve discharge pipingPressure conditionsThe discharge piping associated with pressurizer relief valves was checked for excessive vibration during the opera tion (opening and closing) of the pressurizer relief valves. Following li fting and blowdown of any valve, the re-seating of the valve was verified. (N ote that this test may be done in conjunction with item d of this section.)The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-15 II. Reactor Coolant System (continued)5.Pressure boundary integrity testsa.Hydrostatic testsBelow 200°F (after verification of cleanliness, and fill of system)

Cold hydrostatic testing of the reacto r coolant system pressure boundary was performed at test pressures as specif ied by ASME standards for the system.

Prior to pressurization, the system was heated above the minimum temperature for pressurization. The pressure was then increased in increm ents and at each increment inspections were made for le akage. Leaky valves or mechanical joints were not a basis for rejecting th e test. Relief valves were provided to prevent inadvertent over-pressurization of the system.b.Baseline data for inservice inspection During preoperational

testing Systems and components that require inspections in accordance with

Section XI of the ASME Code were examined for baseline data either

following the cold hydrostatic test or following hot functional testing, depending on the system and component an d its availability and accessibility.

Data from these inspections provided ba seline data for inservice inspectionsc.Nondestructive testing of stainless steel safe ends and critical componentsBefore hydrostatic testAll reactor coolant system weld joints were nondestructively tested using liquid penetrant and/or radiographic tests as required by Section III of the ASME Code.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-16 II. Reactor Coolant System (continued)6.Vibration monitoring on reactor internals During and after hot functional testing Comprehensive vibration measurements had been made during hot functional testing before core loading for Carolina Power & Light Company's H. B. Robinson Unit 2. The results of these te sts had been documented and submitted to the Directorate of Reactor Licen sing. These data were the basis for acceptance of following plants, such as North Anna Units 1 and 2, without

repeating these tests. During hot functi onal testing, the plant was operated with full flow for a minimum of 240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> in order to achieve approximately 20 million cycles on the internal componen ts. Following hot f unctional testing, the internals were removed and inspected for vibration effects before core loading.

III. Reactivity Control System1.Chemical and volume control system At ambient and/or at operating conditions.

System components were

operationally checked out before fuel loading.

Makeup and letdown operations were conducted with the Chemical and Volume Control System to check out the difference modes of dilution and boration and verify flows in the different modes.The adequacy of heat tracing to maintain the highest concentration in solution was verified. The ability to adequate ly sample and the sampling techniques were demonstrated.2.Emergency boron shutdown system During hot functional

testingThe pressure/flow characteristics of the emergency boration system were verified by pumping into the reactor coolant system.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-17III. Reactivity Contro l System (continued)3.Automatic reactor power control system Preoperational conditions (installation checks had

been made)

The system alignment was ve rified at preoperational conditions to demonstrate the response of the system to simulated inputs. These tests were performed to verify that the systems would operate sa tisfactorily at power. The alignment of the system was verified at power by programmed step changes and under actual test transient conditions to verify that controlled parameters were within tolerances specified by test procedures.4.Incore monitor systema.Incore thermocouplesDuring heatup and at temperature During heatup and at temperature, the inco re thermocouples were calibrated to the average of the reactor coolant system resistance temperat ure detectors. All readout and temperature-compensatin g equipment was checked during the calibration, and isothermal corrections for the operative th ermocouples were determined.b.Movable detector systemAt ambient conditions, before core loading, and

after core loading and critical testingBefore core loading, the installation checkout of the movable detector system was completed. The response of each ch annel was verified using simulated detector inputs. After core loading and insertion of th e thimbles in the core, a dummy cable was used to check indexing and to ensure free passage to all positions and set the limit switches based on data obtained during critical testing. During flux mapping at power, the detector responses to neutron flux were verified.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-18III. Reactivity Contro l System (continued)5.Control rod systema.Rod control systemAmbient conditions, before core loading, and hot conditions after core loadingDuring the installation check of this system, it was energized and operationally checked out with mechanisms connected to each power supply. The ability of the system to step the mechanism was ve rified, the alarm a nd inhibit functions checked out, and correct values of system parameters adjusted to specified values. After core loading, the operation of each rod over its full range of travel was demonstrated.b.Rod drop testsCold and hot plant conditions after core loading At cold and hot plant conditions afte r core loading, the drop times of the full-length rods were measured. The dr op time was measured from the release of the rod until the rod enters the top of the dashpot. This time was verified to be less than the maximum value specified in the Technical Specifications.c.Rod position indicationA t ambient conditions and at temperature after core

loading During rod control system tests, the ro d position indication system was aligned to provide rod movement indication. Rod setpoints were also adjusted during

these tests. After plant heatup, individua l rod positions were calibrated to within tolerances specified by the test procedure.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-19III. Reactivity Contro l System (continued)6.Auxiliary start-up instrumentation testBefore core loadingThree separate temporary source range instruments were installed in the core during core-loading operations. One of these channels served as a spare to the other two channels. During the core lo ading operations, these detectors were relocated at specific loading steps to provide the most meaningful neutron count rate within minimum acceptable le vels, as specified by the core-loading procedures. The response of each channel to a neutron source was verified before core loading.IV. Protection System1.Reactor protection systemBefore core loading (installation checks had been performed)

Before core loading, the reactor trip system was tested to demonstrate operability, proper logic, redundancy, coincidence, independence, and safe failure on power loss. The protection cha nnels were verified through to tripping of the reactor trip breakers. The trip time of each reactor protection signal was

also measured from the output of the se nsor to tripping of the reactor trip breaker. These times were verified to be less than the values identified in the safety analysis report.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-20IV. Protection System (continued)2.Engineered safety featuresBefore core loading (installation checks had been performed)

Before core loading, the engi neered safety features logi c systems were tested to demonstrate operability, proper logic, redundancy, coincidence, and independence. The protection channels were verified through to actuation of the output relays. The response time of each protection signal was also measured from the output of the sensor to actuation of the output relay. Their times were verified to be less than the values identified in the safety analysis report. Operation of the engineered sa fety features components (i.e., motors, valves, diesel generators) was checked in other tests.V. Power Conversion System1.System testsa.Vibration frequency and amplitude Hot functional testing and

or plant heatup after initial criticalityWhen the main turbine was rolled, vibration readings were monitored. (Turbine vibrations were also monitored throughout the power escalation program.)

Major equipment (e.g., feedwater pumps and condensate pumps) was operated as it became available and was observed for indications of excessive vibration.b.Expansion and restraintDuring heatup, at temperature, and cooldown before core loading During heatup to operating temperature, selected points on the components and piping of the systems were checked at vari ous temperatures to verify that they could expand unrestricted. After cooldown, these components were verified to have returned to their approximate cold position.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-21V. Power Conversion System (continued)2.Components and individual systemsa.Steam generator pressure relief and safety valvesPressure conditionsThe setpoints of safety valves were verified from vendor certification data, by bench tests, or by in-plant tests, set points were checked by using a pressure assist device that adds to the force due to pressure. Once the valve left the seated position, the assist device was vented, allowing the valve to re-seat immediately. Steam relief valve set points were made during instrument alignment and verified by plant transient tests.b.Emergency feedwater (auxiliary) systemBefore initial criticalityDuring hot functional testing before initial criticality, the emergency feedwater system was checked out to verify its ability to feed the steam generators.

Automatic starting was verified during te sting of the safeguards logic system tests. The auxiliary feedwa ter piping was checked for excessive vibration while starting and stopping the auxiliary feedwater pumps with normal operation of the associated motor-operated and hand-control discharge valves in the auxiliary feedwater system.c.Turbine control and bypass valves Hot functional testing and/or power operation after initial criticalityDuring hot functional testing, the turbine control system was demonstrated in turbine operation up to and including a period of operation at synchronous speed. The turbine bypass valves to the condenser and their associated control

systems were operationally checked out before and during hot functional testing. Other testing on the turbine bypa ss valves was comple ted after initial criticality.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-22V. Power Conversion System (continued)d.Feedwater and feedwater control system Before hot functional testing and at power The feedwater and condensate pumps we re operationally checked out before hot functional testing. During power escalation, the power was slowly increased and the ability of the feed water pumps and control system to maintain level in the steam generators was verified. Steam generator level indicators were aligned befo re filling the system, and during fill the system was used to monitor the level in the steam generator. Before start-up, the feedwater-regulating valve control system was calibrated using simulated signals. During start-up when at power the ability of the system to control level within specified tolerances under tr ansient conditions was also verifiede.Condenser circulating waterBefore initial core loading and at powerBefore core loading, the main circulating water pumps and circulating water

system valves were tested to verify operability. During unit start-up, acceptable condenser operation was verified in accordance with operating procedures.f.Makeup water and chemical treatment systems During steam generator fill, hot functional testing, and at power The makeup system to the steam generato rs was checked out during fill of the steam generators during hot functional testing and at power. The chemical treatment system was checked out when chemicals were added to the steam generators at heatup and at operating conditions.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-23 VI. Auxiliary Systems1.Reactor coolant system makeup (Chemical and Volume Control System (CVCS))See III, item 1.2.Seal and pump cooling water (CVCS)

Before heatup and at temperature Before reactor coolant pump operation and with the system pr essurized, flow to the pump seals and cooling water flow was adjusted to specified values using

installed instruments. During hot f unctional testing when at operating temperature and pressure, seal and c ooling flows and temperatures were checked.3.Vent and drain systemDu ring initial primary fill and pressurization and during hot functional testingVenting of the reactor coolant system was done during initial filling by venting the reactor vessel head and pressurizer. During hot functional testing and after core loading, the secondary system was vented while pressurizing the secondary system. Secondary drains were tested for unrestricted flow in

accordance with operating procedures.4.Component cooling systemAmbient and/or hot plant conditions Component cooling flow to the various components in the system was adjusted, the system operationally checked out, and setpoints adjusted. Data were taken to verify that adequate cooling was provided to each cooled component and, when load was available, that temperature limits were being maintained.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-24 VI. Auxiliary Systems (continued)5.Residual heat removal system Before and during hot functional testing This system was tested by verifying pressure and flow characteristics of the pumps and operation of th e isolation valves. Du ring cooldown after hot functional testing, the heat removal capabi lity and cooldown rate of the system was demonstrated. The resi dual heat removal system piping was checked for excessive vibration while starting a nd stopping the residual heat removal pumps with normal operation of the valve used to control flow.6.Purification system (CVCS)Operating temperature before core loading During hot functional testing with the demineralizers charged with resin, operation of the purification system was demonstrated by verification of flow, pressure drops, temperatures, and conditioning of ion-exchange resins (see III, item 1).7.Fire protection systemBefore initial core loadingThe water fire protection syst em motor and diesel-driven pumps and pressure maintenance equipment were tested to verify proper operation in conformance with fire insurance requirements. The carbon dioxide fire protection system was tested by individual component checks and by puff tests in various fire-protected areas by simulating syst em initiating condi tions. The Halon 1301 fire protection system was checked fo r operability and proper installation.8.Service water systemBefore initial core loadingThe system was operationally checked out to verify pressure and flow. Service water flow was verified to components in the system.9.Auxiliary building ventilationBefore initial core loadingThe system was operated to test for leaks and air flows to the areas supplied from the system and to verify motor cu rrents and speeds, veri fy setpoints, and check alarms (see also IX.)The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-25 VI. Auxiliary Systems (continued)10.Compressed gas system (used for safety-related functions)Before initial core loadingThe instrument air system , including air receivers and compressors, was tested to verify proper operation. A loss-of-i nstrument-air test was conducted by securing the makeup air to each dedicated air accumulator supplying each safety-related component that is re quired to operate following a loss of instrument air. The capacity of each de dicated air accumulator was verified by operating the safety-related component a specified number of times over a specified time interval. Air-operated co mponents were tested to ensure that they fail in the safe mode upon loss of operating pressure. Other compressed gas systems were verified for proper operation.11.Control-rod drive mechanism and rod position indication coil cooling

system Before and/or during hot functional testing The system was operationally checked out to verify air flow, temperatures, motor current and speed.12.Neutron shield tank cooling systemBefore initial core loadingThe system was operationa lly checked out to verify pump and heat exchanger operability.13.Leak detection system (sensitivity and accuracy to detect leaks)

Before and during preoperational testsTemperature detectors and their alarm functions in the drain lines from pressurizer safety valves and the r eactor vessel head seal were checked.

Pressurizer relief tank level and temper ature sensors were calibrated and their associated alarms checked.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-26 VI. Auxiliary Systems (continued)14.Primary sampling systemBefore and/or during hot functional testing Operations were performed to:a.Established purge times.b.Demonstrate that liquid and gas sa mples can be obtained from sample points.c.Demonstrate that valves, instruments and controls function properly.

d.Verify proper functioning of the sample cooler.

e.Demonstrate that sample vessels ca n be removed and replaced without problems.15.Primary pressure relief system Before hot functional testing and at pressure

conditions The pressurizer relief tank, associated valves, and instrumentation were checked out to verify performance of design functions. (See II, item 4.3, for

testing of pressurizer re lief and safety valves.)VII. Electrical Systems1.Normal distribution test (transformers, motor, relay

switches, power supplies

etc.; phasing and meggering where applicable)

Throughout plant start-up

and before applicable equipment operation The integrity and operation of these co mponents were verified before being energized by meggering, hi-pot testing, continuity checks, and operational verification of controlling devices as applicable. After being energized, phasing and voltage regulation test s were performed and channel and train separation and redundancy features were verified as applicable.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-27 VII. Electrical Systems (continued)2.Vital bus test (full-load test using all power sources)Before initial core loadingVerif ication that the vital bus load could be supplie d under normal and power-failure conditions was made. In particular, transfers that take place under loss of power and redundant features function per design were verified.3.Direct current systems (full-load and duration test)Before core loadingThe redundant features of the battery, battery charger, and inverters were checked out. The capacity of the batter y and voltage regulation was verified. The recharging of a discharged batter y within a specified period was also verified, The ability of each inverter to maintain design output under varying direct current input was also verified.4.Communications systems (telephone, public address, intercoms, and evacuation

signals)Before fuel loading and during power operationTo verify proper communications between al l onsite stations and interconnection to commercial telephone service. To balance and adjust amplifiers and speakers and verify that evacuation alarms could be heard at all stations throughout the plant. Also , to verify that all temporary communications at the fuel-loading st ations and control stations were functioning properly.5.Emergency power systems (manual start and synchronization, full

automatic loading tests, under loss of all alternating current voltage)Before initial core loadingThe automatic starting and loading of the diesel generators was demonstrated under loss of emergency bus alternating current power. The operation of the logic and sequencing of circuit breake rs were demonstrated along with the proper safety-related bus stripping and separation of non-vital loads. Load duration tests were demonstrated ove r several hours of operation along with voltage and frequency regulation te sts under transient and steady-state conditions.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-28 VIII. Containment Systems1.Reactor containmentBefore core loadingA containment stru ctural test was performed. Containment Type A, B, and C leakage tests were performed in acco rdance with Appendix J to 10 CFR 50.2.Ventilation systemBefore and/or during hot functional testing The system was operated to balance air fl ows and to verify ability to maintain temperatures below maximum allowable limits.3.Post-accident heat removal system (containment sprays)

Before initial criticalityTests were performed to verify pum p operating characteristics and response to control signals, sequencing of the pumps, valves, and controllers, and to ensure that spray nozzles were unobstructed. The time required to actuate the system after a containment high-pressure signal is received was verified.4.Containment isolationBefore core loadingThe operation of actuation system s and components used for containment isolation was verified.5.Hydrogen removal systemBefore initial criticalityOperability of flow paths and heaters associat ed with the recombiners were verified.IX. Gaseous Radioactivity Removal Systems Filtration system (testing

performed on particulate filter system in containment and auxiliary structures for post-accident and routine release of gaseous effluent)Before core loadingTesting was performed to verify flows, pressure drops, and effectiveness of these systems in performing their function.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-29X. Emergency Core Cooling System a1.System tests (expansion and restraints, vibration)

Before and/or during hot functional testing Movement of piping that connects to th e reactor coolant system was checked by the test described in II, item 2. Pump s, motors, and piping were observed for excessive vibration.2.High-pressure safety injectionBefore core loadingThis system wa s operationally tested to adjust pressure/flow values. Tests were also conducted to check pump operating ch aracteristics and to verify operation from normal and emergency power sources. More specifically that:a.Valves installed for redundant fl ow paths operated as designed.b.Pump operating characteristics were verified and the capability of the high-head safety injection pumps to take suction from the low-head pumps was demonstrated with the reactor coolant systems at ambient conditions.c.Valves and pumps operated on operator initiation and/or automatically on initiation of a safety injection signal.d.The fail position on loss of power for each remotely operated valve was as specified.e.Level and pressure instrume nts were properly calibrated.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-30X. Emergency Core Cooling System a (continued)3.Low-pressure safety injectionBefore core loadingThe low-head safety injection system was checked to verify design flow, flow paths, and pump operating characteristics. Tests were conducted to verify operation from the normal power source. More specifically, that:a.Valves installed for redundant fl ow paths operated as designed.b.Pump operating characteristics were verified with the reactor coolant system at ambient conditions.c.Valves and motors operated on oper ator initiation and/or automatically on initiation of a safety injection signal.d.The fail position on loss of power for each remotely operated valve was as specified.e.Level and pressure instrume nts were properly calibrated.4.AccumulatorBefore core loadingFlow through the accumulator discharge lines was initiat ed to demonstrate that the motor-operated valves stroked properl y and the check valves were free to open. Tests were also made to verify that accumulator pressure could be maintained.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-31XI. Fuel Storage and Handling System1.Spent-fuel pit cooling and refueling purification systemBefore core loadingTests were pe rformed to verify operability of the spent-fuel pit cooling pumps; operability of the refueling purificati on pumps; flow through the spent-fuel pit heat exchange loops; operation of the skimmer loops; flows through the refueling purification filters and ion exchanger; alarm setpoints; and correct functioning of valves, instruments, and controls. 2.Refueling equipment (hand tools and power equipment, including protective

interlocks)

Before storage of new fuel and initial core loadingTests were performed before core loadi ng to demonstrate the functioning of the fuel transfer system and the fuel ha ndling equipment using a dummy assembly, in accordance with design dr awings and instruction manuals. The sections that involve the spent-fuel facility were chec ked before the storage of new fuel in the spent-fuel storage pool.3.Operability and leak tests of sectionalizing devices in fuel storage pool and

refueling canalBefore initial core loadingDuring th e initial filling of the spent-fuel storage pool, operability and leaking testing of the sectionali zing devices was performed.4.Spent fuel storage building ventilation systemBefore plant start-upThis is part of the auxiliar y building ventilation system (refer to VI, item 1).5.Spent-fuel storage radiation monitoring equipment Before plant start-upRefer to XIII, item 1.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-32 XII. Reactors Components Handling System Reactor components handling system (polar crane)Before use for installation

of components within the

containmentTesting was conducted on the polar crane in accordance with standard crane testing procedures during the construction of the station.

XIII. Radiation Protection System1.Process, criticality, and area monitors Before core loading and plant operation Before core loading, the radiation alarms associated with core loading were checked out and alarm setpoints were verified. Process and area monitor sensors and channels were calibr ated and alarm setpoints made.2.Personnel monitor and survey instruments Before core loading and/or initial criticality Before core loading and required equipm ent use, instruments were calibrated.

After this initial calibration, the in struments are periodically checked for recalibration.3.Laboratory equipmentBef ore core loading and initial criticalityLaboratory equipment was checked to verify equipment performance and calibration. Chemical analyses performe d on standard samples. During start-up the equipment received additiona l verification by normal usage.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-33XIV. Radioactive Waste SystemRadioactive waste systemBefore initial criticalityTests were performed to establish the satisfactorily performance of pumps and instruments, leaktightness of piping and equipment, and the operation of packaging and waste reduction equipmen t; and to verify proper operation of alarms and controls. More specifically, to ensure that:a.Manual and automatic valves were operableb.Instrument controllers operated properly.c.Alarms were operable.

d.Pumps performed their system function satisfactorily.e.The waste gas compressors operated properly.f.The gas analyzers operated properly.g.The waste evaporator was operational.h.The hydrogen and nitrogen supply packages were sufficient for operation of the system.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-1 (continued)LIST OF PREOPERATIONAL TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.The preoperational test program for the emergency core cooling system (ECCS) meets the requirements set forth in Regulatory Guide 1.79, June 1974, with clarifications noted in Section 3A.40.

Revision 52-09/29/2016 NAPS UFSAR 14.1-34The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-2 LISTS OF START-UP TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of TestingI. Precritical Tests After Fuel Loading1.Mechanical and instrument tests on control rod drive and rod position indicators Before initial criticalityOperational testing of the rod contro l systems was conducted to check the controlling features, adjust s setpoints, and verify rod speeds and sequencing of power to the rod drives. After core loading and installation of the rod mechanisms, tests were conducted to verify operation of the rod drive mechanisms, over their full travel, the latching and releasi ng features were demonstrated, and calibration of the pos ition indicators was performed over the rod full-range travel per toleran ces specified in the test procedure.2.Reactor trip circuit and manual trip tests Before initial criticalityOperational testing was conducted to ve rify the reactor protection circuits in the various modes of tripping, including manual reactor trip up to the tripping of the reactor trip breakers. After core loading, the release and insertion of each full-length mechan ism was demonstrated.3.Rod drop measurement cold and hot at rated flow and no flow Before initial criticality At cold and hot plant conditions afte r core loading, the drop times of the full-length rods were measured. Th e drop time was measured from the beginning decay of the statio nary gripper coil voltage until the rod entered the top of the dashpot. This time was verified to be less than the maximum value specified in the Technical Specifications. Ten additional measurements were made for the fastest and slowest rods.4.Pressure test of reactor coolant systemBefore initial criticalityAfter core loading and installation of the re actor vessel head and torquing of the reactor vessel head studs, pressure testing was performed at 100 psi above operating pressure to verify that no leakage occurred past the head and vessel seal.a.Reference Drawings 1 through 15 show the logics for initiation of turbine and generator trips. The logics show the sensed variables and all avai lable setpoints. See Figure 7.2-1 for a listing of symbols used in these figures.

Revision 52-09/29/2016 NAPS UFSAR 14.1-35I. Precritical Tests Afte r Fuel Loading (continued)5.Chemical tests (to establish water quality)Before heatupWater for reactor coolant system fill and makeup was analyzed for chloride content, conductivity, total suspended solids, pH, clarity, and fluorides to requirements specified by the ch emistry manual for NSSS. During pre-operational testing, hydr azine was added to scavenge oxygen. After core loading and before exceeding 250°F, hydrogen was added to scavenge oxygen during critical opera tion. After initially establishing chemistry, analysis was performed to verify requirements.6.Nuclear instrumentation calibration and neutron responseBefore core loadingBefore core loading, the source range channels were aligned and operational based on data derived from using a neutron source.

After a power history had been established on the core, the dete ctor anode and discriminator voltages were reset based on obtained data.7.Mechanical and electrical tests of incore movable

detectorsBefore initial criticality and during physics testing The movable detector systems were checked out in accordance with the operating procedures and ICPs. After core loading and insertion of the detector thimbles, the system was again operationally checked out by ensuring the free passage of detectors into all inserted thimbles. Electrical tests were performed using simulated signals to check out the recorders.

During physics measurements the syst em was operationally checked and limit switches set based on flux mappi ng data. Incore thermocouples were checked out during hot functional testing (see Table 14.1-1, III, item 4.a).The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-2 (continued)LISTS OF START-UP TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.Reference Drawings 1 through 15 show the logics for initiation of turbine and generator trips. The logics show the sensed variables and all avai lable setpoints. See Figure 7.2-1 for a listing of symbols used in these figures.

Revision 52-09/29/2016 NAPS UFSAR 14.1-36I. Precritical Tests Afte r Fuel Loading (continued)8.Reactor coolant flow measurement Before initial criticalityAfter core loading, measurements were made of elbow tap differential pressures to make relative comparison. At hot shutdo wn conditions after core loading, measurements of loop elbow differential pressure drops were made.

Using these data with the reactor coolant pump performance curve, the calculated flow was verified to the design flow. Flow coastdown and transients after reactor coolant pump trips were also determined at shutdown conditions after core loading.9.Pressurizer effectiveness testAt hot shutdown after core loading At hot no-load temperature and pressure the effectiveness of the pressurizer heaters in maintaining and increasing system was demonstrated. The heaters were energized and the pressure was compared with an expected pressure rise given in the procedure. The ability of the spray system to reduce pressure was also demonstrated. The spra y valves were opened and the pressure decrease compared with the expected pressure decrease given in the procedure.10.Vibration monitoring on reactor internals--No vibration monitoring was done after co re loading (refer to test identified in Table 14.1-1, II. item 6)The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-2 (continued)LISTS OF START-UP TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.Reference Drawings 1 through 15 show the logics for initiation of turbine and generator trips. The logics show the sensed variables and all avai lable setpoints. See Figure 7.2-1 for a listing of symbols used in these figures.

Revision 52-09/29/2016 NAPS UFSAR 14.1-37II. Initial Criticality and Low-Power Tests1.Initial criticalityPlant at hot shutdownThe objective was to bring the reactor critical for the first time from the plant conditions specified. Before the star t of rod withdrawal, the nuclear instrumentation had been aligned and ch ecked, and conservative reactor trip setpoints made per the test procedures. All rods were withdrawn except the last controlling bank, which was left partially inserted for control once criticality was achieved by boron dilu tion. At preselected points in rod withdrawal and boron dilution, data were taken and inverse count rate ratio pots were made to enable extrapolat ion to the expected critical point.2.Radiation surveysAt steady-state conditions during power escalation Radiation surveys were made during th e power escalation to determine dose rate levels at preselected points in side containment due to neutron and gamma radiation. Instrument s used were calibrated to known sources, and the calibration rechecked following the survey.3.Calibration of nuclear instruments with thermal power and determination of

overlap After start-up and during escalationAfter initial criticality and during escalation into the intermediate and power ranges, data were taken to verify ov erlap between the source, intermediate, and power range channels and to verify the alarm and protective functions.

These data were collected until the overl aps were firmly established. During low power escalation, the power range de tector currents were monitored and compared with the intermediate range currents to verify response of the power range detectors. The power range nuclear channels were calibrated to

reactor thermal output based on measur ement of secondary plant feedwater flow, feedwater temperatur e, and steam pressure.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-2 (continued)LISTS OF START-UP TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.Reference Drawings 1 through 15 show the logics for initiation of turbine and generator trips. The logics show the sensed variables and all avai lable setpoints. See Figure 7.2-1 for a listing of symbols used in these figures.

Revision 52-09/29/2016 NAPS UFSAR 14.1-38II. Initial Criticality and Low-Power Tests (continued)4.Effluent radiation monitors (calibration against known concentration)Before plant start-upThese instruments were calibrated to a known radiation source or to analog signals which had been calibra ted to known radiation sources.5.Moderator temperature reactivity coefficientHot zero powerAt normal no-lo ad temperature and no nuclear heating, reactor coolant system cooldown and heatup were acco mplished using the steam dump and reactor coolant pumps opera tion as required. An approximate 5°F change in temperature was initiated, and during these changes the average temperature and reactivity were recorded on an X-Y plotter. From these data the moderator temperature coefficient was determined.6.Pressure reactivity coefficient measurements--Direct measurements of the pressure coefficient of reactivity were not made, since the effects of pressu re on reactivity are of s econd order when compared with other effects.7.Control rod reactivity worth determination of differential and integral worth and

verification of worth for

shutdown capabilityHot zero powerUnder zero-power conditions at near operating te mperature and pressure, the nuclear design predictions for rod cl uster control assembly (RCCA) group differential worths were validated. These validati ons were made from boron concentration sampling data, RCCA bank positions, and recorder traces of reactivity. From this data the integr al RCCA group worths were determined, including verification of rod insertion limits to ensure adequate shutdown margin. The minimum boron concentration for maintaining the reactor shutdown with the most r eactive rod cluster control assembly stuck in the full-out position was determined for Unit 1. The determination was made from analysis of boron concentration and RCCS worths.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-2 (continued)LISTS OF START-UP TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.Reference Drawings 1 through 15 show the logics for initiation of turbine and generator trips. The logics show the sensed variables and all avai lable setpoints. See Figure 7.2-1 for a listing of symbols used in these figures.

Revision 52-09/29/2016 NAPS UFSAR 14.1-39II. Initial Criticality and Low-Power Tests (continued)8.Boron reactivity worth measurementZero powerDifferential boron worth meas urements were made by monotonically increasing or decreasing reactor c oolant boron concentration. Compensation for the reactivity effect of the boron concentration ch ange was made by withdrawing or inserting respective c ontrol rods to maintain moderator average temperature and power level constant and observing the resultant accumulated change in core reactiv ity corresponding to these successive rod movements.9.Determination of boron concentration of initial criticality and reactivity

allocationZero powerThese determinations are described under II, item 1 above.10.Flux distribution measurement with normal rod

patterns Zero power Flux distribution measurements with normal rod patterns were taken during the zero-power physics tests. 11.Chemical tests to demonstrate ability to control water quality Before criticality and during power escalationBefore criticality, the procedures and equipment for performing chemical analyses of primary and secondary systems were demonstrated. During power escalation, sampling was performed and analysis done to verify that plant chemistry was within specifications.12.Pseudo-rod-ejection test, to verify safety analysis (hot)

Zero power Incore measurements were made for Unit 1 under pseudo-ejected-rod conditions simulating the zero-power accident to determine the hot-channel factors and verify that they were with in assumptions made in the accident analysis.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-2 (continued)LISTS OF START-UP TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.Reference Drawings 1 through 15 show the logics for initiation of turbine and generator trips. The logics show the sensed variables and all avai lable setpoints. See Figure 7.2-1 for a listing of symbols used in these figures.

Revision 52-09/29/2016 NAPS UFSAR 14.1-40III. Power Ascension Tests1.Natural circulation test to confirm sufficient cooling capacity--The ability of natural circ ulation to remove decay he at has been demonstrated at the Carolina Power & Light Company's H. B. Robinson Unit 2. Tests have shown natural circulation flow to be more than adequate to remove decay heat, and such a test was not repeated on North Anna Unit 1. However, special tests were conducted for Unit 2, as described in Section 14.1.4.2.Power reactivity coefficient evaluation and power defect measurements (30, 50, 75 and

100%)During power escalationDuring each power escalati on for Unit 1, recorder traces were made of reactor power and reactivity change

s. From these traces, the power coefficient of reactivity and power defects were determined.3.Plant response to load swings, including automatic control system checkout (30, 50, 75

and 100%)During power escalationPlant response to the following load changes was demonstrated:

a.+/-10% step load change from 30, 75 and 100% power.b.50% load reduction from 75 and 100% power.c.Plant trip from 100% power level.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-2 (continued)LISTS OF START-UP TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.Reference Drawings 1 through 15 show the logics for initiation of turbine and generator trips. The logics show the sensed variables and all avai lable setpoints. See Figure 7.2-1 for a listing of symbols used in these figures.

Revision 52-09/29/2016 NAPS UFSAR 14.1-41III. Power Ascension Tests (continued)

The data collected from the performanc e of these tests were analyzed for control system behavior and requirements for realignment. Acceptance criteria, such as the plant not tripping (where applicable), relief and safety valves not lifting, and steam dump operating correctly, were identified in the individual procedures. At approximately 30% power, the automatic control systems were checked by initiating a perturbation and observing controller response. During the transi ent tests these systems we re operationally checked under actual design load changing conditions.4.Chemical analysis (30, 50, 75, 100%)During power escalationDuring low-power physics tests and at 30, 50, 75, and 100% power, samples of reactor coolant were taken and analys is performed to verify that coolant chemistry requirements could be maintained.5.Effluents and effluent monitoring systems (30, 50, 75, 100%)During power escalationInstalled effluent monitors were operated c ontinuously at selected locations in the plant to monitor for radioactive constituents in the effluents.

Instruments detected any changes in ac tivity and alerted the operator when radiochemical analysis should be performed6.Evaluation of core performance (30, 50, 75, 100%)During power escalationAt steady-state power points, incore data were obtained and analysis performed to verify that the core performance margins were within design predictions, for expected normal and abnormal rod configurations.7.Loss of flowBefore criticalityReact or coolant system response to loss of flow for various combinations of pump trips was determined fr om hot shutdown conditions.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-2 (continued)LISTS OF START-UP TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.Reference Drawings 1 through 15 show the logics for initiation of turbine and generator trips. The logics show the sensed variables and all avai lable setpoints. See Figure 7.2-1 for a listing of symbols used in these figures.

Revision 52-09/29/2016 NAPS UFSAR 14.1-42III. Power Ascension Tests (continued)8.Turbine trip (100%)

a At powerThis test verified that the pressurizer safety valves did not lift and that the plant could be maintained in a hot s hutdown condition. The turbine trip from 100% power was conducted as an integral part of the generator trip from 100% power.9.Generator trip (100%)

a At power Generator trip was performed at 100% power to verify the plant's capability of withstanding an instantaneous re duction in load from 100 to 0%. A generator trip was initiated by manual ly opening the main generator breakers. This would be automatically initiate a turbine trip.10. Shutdown from outside the control room Greater than or equal to

10% generator power The ability to bring the plant to a nd maintain the plant in hot shutdown conditions after a trip from greater than or equal to 10% power was demonstrated using instrumentation an d controls outside the control room.11. Loss of offsite powerGreater than or equal to 10% generator powerTests were performed in which loss of voltage was simulated. Starting of the diesels and connecting of the emergency loads on the emergency bus was

demonstrated.12. Radiation surveys and shielding effectiveness (50 and 100%)At powerThe surveys to determine the effe ctiveness of the shielding have been discussed under Radiation Survey (Item II.2). These surveys were conducted up to and including 100% power.13. Part-length rod insertion/withdrawal (75%)

Approximately 75%

powerTechnical Specifications re quired part-length rods for Unit 1 to remain fully withdrawn; therefore no testing was performed. No part-length rods were installed in Unit 2.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-2 (continued)LISTS OF START-UP TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.Reference Drawings 1 through 15 show the logics for initiation of turbine and generator trips. The logics show the sensed variables and all avai lable setpoints. See Figure 7.2-1 for a listing of symbols used in these figures.

Revision 52-09/29/2016 NAPS UFSAR 14.1-43III. Power Ascension Tests (continued)14. Dropped rod-effectiveness of instruments to detect dropped rod and verification of

associated automatic action Greater than or equal to 50% reactor power Automatic turbine runback and rod wit hdrawal stop is not necessary as a result of a dropped rod, and the circui try required for such action does not exist in this plant. Rod drop tests based on common failure criteria were performed dynamically to de monstrate the negative rate trip function from greater than or equal to 50% reactor power.15. Vibration measurements on reactor internals (30, 50, 75, and 100%)--These measurements were not performed at power. Refer to II, item 6, above for measurements before operation.16.Pseudo-rod-ejection test to verify safety analysis30% powerIncore measurements were made for Unit 1 with individual rods withdrawn out of bank position to determine the resulting hot-channel factors and verify that they are within expected limits. These determinations were made from movable detector and thermocouple data. This measurement was not performed on Unit 2 because of the negligible worth of a control rod withdrawn from its full-power insertion limit, and the large magnitude of margin remaining to hot-channel factor limits in the ejecte d-rod configuration (see Section 14.1.3). This was verified by the test performed for Unit 1.17.Evaluation of flux asymmetry50% power Incore flux measurements were made with a single rod assembly moving partially below bank position and fully inserted to demonstrate that core limits are not exceeded.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-2 (continued)LISTS OF START-UP TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.Reference Drawings 1 through 15 show the logics for initiation of turbine and generator trips. The logics show the sensed variables and all avai lable setpoints. See Figure 7.2-1 for a listing of symbols used in these figures.

Revision 52-09/29/2016 NAPS UFSAR 14.1-44III. Power Ascension Tests (continued)18.Process computer (30, 50, 100%)During power escalationWhen ava ilable during power escalation, the process co mputer was checked out and comparisons made between process signals a nd those assessed by the process computer. (No safety-relat ed functions are performed by the computer.)19.Moisture carryover measurementAt powerRadioactive tracer of sodium-24 was injected into the steam generators.

Samples obtained from the steam generato r upper shell, main steam line taps, and the feedwater system were analyzed.The following information is HISTORICAL and is not intende d or expected to be updated for the life of the plant.Table 14.1-2 (continued)LISTS OF START-UP TESTSTest or Measurement Plant Condition/PrerequisiteTest Objective and Summary of Testinga.Reference Drawings 1 through 15 show the logics for initiation of turbine and generator trips. The logics show the sensed variables and all avai lable setpoints. See Figure 7.2-1 for a listing of symbols used in these figures.

Revision 52-09/29/2016 NAPS UFSAR 14.1-45 The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.Table 14.1-3UNIT 2 START-UP PHYSICS TESTSI. Hot Zero Power Tests1.Reactivity computer checkout.

2.Isothermal temperature coefficient at ARO an d D bank in (also D, C banks in if MTC for D bank in is greater than or equal to 0 pcm/°F).3.Boron endpoints at ARO; D bank in; D, C banks in; D, C, B banks in; D, C, B, A banks in; shutdown bank B in with all ot her rods out; and shut down bank A in with all other rods out.4.Reactivity worths of all control and shutdown rod banks.

5.Boron worth over the range of control banks A through D moving during rod insertion and withdrawal.6.Power distribution measurem ents for ARO and D bank in.II. Power Ascension Tests 1.30% power flux map.

2.50% power flux map.

3.Pseudo-dropped-rod test (RCCA D-10) and associated power di stribution measurements at 50% power.

4.Incore/ex-core detector calib ration flux maps at 75% power.

5.APDMS flux maps at or below 95% power.6.Flux maps at 90% and 100%

power (equilibrium conditions).

Revision 52-09/29/2016 NAPS UFSAR 14.1-46 The following information is HIST ORICAL and is not intended or expected to be updated for the life of the plant.

T able 14.1-4PHYSICS TESTS THAT HAV E BEEN DELETED FOR UNIT 2I. Hot Zero Power Tests1.Isothermal temperature coefficient at D, C banks in; D, C, B banks in; and D, C, B, A banks in.2.Boron endpoint for the N-1 rods-in configuration.3.Reactivity worth of N-1 rods.

4.Pseudo-rod-ejection and associated power distribution measurements.

II. Power Ascension 1.Pseudo-rod-ejection and associated power distribution measurement at 30% power.

2.Pseudo-dropped-rod test (RCCA H-6) and as sociated power distribution measurement.3.Power coefficients.

4.Integral power defect.

5.Doppler-only power coefficients.

Revision 52-09/29/2016 NAPS UFSAR 14.1-47The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.Table 14.1-5

SUMMARY

OF UNIT 1 MEASURED VALUES, DESIGN VALUES, DESIGN TOLERANCE,AND ACCIDENT ANALYSIS CRITERIA FOR PHYSICS TESTS THAT HAVE BEEN DELETED FOR UNIT 2Test DescriptionCore ConditionParameterUnit 1 Measured ValueDesign Value (beginning of life (BOL), Best Estimate)

Design Tolerance Accident Analysis Criterion1.Isothermal temperature coefficient Banks D, C in

Banks D, C, B in

Banks D, C, B, A inTTT -7.86 pcm/°F

-13.48 pcm/°F

-14.07 pcm/°F

-8.9 pcm/°F

-14.1 pcm/°F

-13.8 pcm/°F

+/-3 pcm/°F+/-3 pcm/°F+/-3 pcm/°F -2.107 pcm/°F -2.134 pcm/°F -2.135 pcm/°F2.Boron endpointN-1 rods insertedC B601 ppm580 ppm

+/-50 ppm x C B 24000 pcm where = 11.08 pcm/ppm3.Rod worthN-1 rodsI N-1 a80157893 pcm

+/-789 pcm(IN-1)/1.04 5780 pcm4.Pseudo- ejected control rod hot zero power (HZP), Bank C at

120 steps, Bank D at 0 steps, RCCA B-8 at 228 steps.

F Q I B-8 a 6.85 443 pcm 10.8 464 pcm NA+/-46 pcm 13.0 (I B-8) x 1.04785 pcm30% power, Bank

D at 194 steps, RCCA B-8 at 228 steps.F Q I B-8 2.1 3 pcm 2.1 7 pcm NA+/-1 pcm b 7.07 (I B-8) x 1.04200 pcm5.Pseudo- dropped control rod50% power, RCCA H-6 I H-6 a 1.62 138 pcm 1.70 146 pcm NA+/-22 pcm 1.69 c (I H-6) x 1.04250 pcm F NH Revision 52-09/29/2016 NAPS UFSAR 14.1-486.Power coefficient30% power-15.24 pcm/%

power-14.02 pcm/%

power+/-4.57 pcm/%

power NA50% power-12.74 pcm/%

power-13.75 pcm/%

power+/-3.82 pcm/%

power NA75% power-13.57 pcm/%

power-13.39 pcm/%

power+/-4.07 pcm/%

power NA90% power-10.70 pcm/%

power-13.31 pcm/%

power+/-3.21 pcm/%

power NA7.Power defect0 - 100% powerReactivity worth1270 pcm1299 pcm

+/-191 pcmNA8.Doppler-only power coefficient 30% power-13.62 pcm/%

power-11.35 pcm/%

power+/-4.09 pcm/%

power Inferred value

+/-30% uncertainty must overlap allowance range of Figure 15.1-350% power-10.77 pcm/%

power-10.75 pcm/%

power+/-3.23 pcm/%

power Inferred value

+/-30% uncertainty must overlap

allowance range of

Figure 15.1-3 The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.Table 14.1-5 (continued)

SUMMARY

OF UNIT 1 MEASURED VALUES, DESIGN VALUES, DESIGN TOLERANCE,AND ACCIDENT ANALYSIS CRITERIA FOR PHYSICS TESTS THAT HAVE BEEN DELETED FOR UNIT 2Test DescriptionCore ConditionParameterUnit 1 Measured ValueDesign Value (beginning of life (BOL), Best Estimate)

Design Tolerance Accident Analysis CriterionQ()powerQ()powerQ()powerQ()powerQ()inferredDopplerQ()inferredDoppler Revision 52-09/29/2016 NAPS UFSAR 14.1-498.Doppler-only power coefficient (continued)75% power-11.08 pcm/%

power-9.96 pcm/%

power+/-3.32 pcm/%

power Inferred value

+/-30% uncertainty must overlap allowance range of Figure 15.1-390% power-7.59 pcm/%

power-9.38 pcm/%

power+/-2.28 pcm/%

powerAll inferred values fell within range of

Figure 15.1-3, as shown on Figure 14.1-3a.I N-1 = integrated reactivity worth of all control rods except the most reactive rod (N-1).

I B-8 = integrated reactivity worth of rod cluster control assembly (RCCA) B-8.

I H-6 = integrated reactivity worth of RCCA H-6.b.Violation of design tolerance was evaluated to be insignificant due to low value of reactivity worth.c.Accident analysis referenced to hot full power.The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.Table 14.1-5 (continued)

SUMMARY

OF UNIT 1 MEASURED VALUES, DESIGN VALUES, DESIGN TOLERANCE,AND ACCIDENT ANALYSIS CRITERIA FOR PHYSICS TESTS THAT HAVE BEEN DELETED FOR UNIT 2Test DescriptionCore ConditionParameterUnit 1 Measured ValueDesign Value (beginning of life (BOL), Best Estimate)

Design Tolerance Accident Analysis CriterionQ()inferredDopplerQ()inferredDoppler Revision 52-09/29/2016 NAPS UFSAR 14.1-50The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.

Figure 14.1-1TYPICAL PRE-OPERATIONAL TEST SEQUENCE Revision 52-09/29/2016 NAPS UFSAR 14.1-51The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.

Figure 14.1-2TYPICAL START-UP TEST SEQUENCE Revision 52-09/29/2016 NAPS UFSAR 14.1-52The following information is HISTORICAL and is not intended or expected to be updated for the life of the plant.

Figure 14.1-3CYCLE 1 BOL PHYSICS TEST DOPPLER POWER COEFFICIENT USED IN ACCIDENT ANALYSIS UNIT 1

Revision 52-09/29/2016 NAPS UFSAR 14.2-1 14.2 AUGMENTATION OF VEPCO'S STAFF FOR INITIAL TEST AND OPERATIONThe start-up organization used during the period of initia l operation consisted of all personnel in the St ation Oper ations Department, with a dditional support provided by the Engineering Services Department and the Chemistry and Health Physics Department.

In operations, there were, for one-unit operation, fi ve senior licensed shift supervisors, five licensed control room operato rs, and a combined group of 14 assistant control room operators and auxiliary operators with a minimum of two licensed assistant control room operators. For two-unit operations, there were five senior licensed shift s upervisors, five senior licensed assistant shift supervisors, 10 licensed control room operator s, and a combined group of 18 assistant control room operators and auxiliary operator s with a minimum of six licensed assistant control room operators. Shifts were scheduled to ensure that a minimum of three licensed reactor operators and two licensed senior reactor operators are on duty at all times during two-unit operations.Technical support was provided during start-up using the services of graduate-level engineers. A trained power engineer was assigne d through the architect-engineer to assist in preliminary operati ons for both units. In addi tion, engineering representa tives were assigned at the station by the supplier of the nuclear st eam supply system to render start-up support.

VEPCO had overall responsibility during plant star t-up, including precriticality tests, approach to criticality, and post criticality operations. The station staff was assisted by the architect-engineer and the supplier of the nuclear steam supply system. The Stone & Webster start-up engineer was assigned to the station from the start of flushing operations through commercial operation. The start-up engineer reported directly to the Station Manager and received instructions from him.Experienced Westinghouse reacto r engineers were al so assigned to the station for fuel loading, initial criticality, and physics testing. These reac tor engineers were qualified and knowledgeable in reactor operations.

They reported directly to the VEPCO react or engineers and received instructions from them. The Westinghouse reactor engi neers acted in an advisory capacity only; VEPCO retained responsibility for , and control of, the unit. Reactor specialist s (e.g., control engineers) were available and utilized as required.

Revision 52-09/29/2016 NAPS UFSAR 14.2-2 Intentionally Blank Revision 52-09/29/2016 NAPS UFSAR 14A-i Appendix 14ANRC Questions and VEPCO's Responses Regarding the North Anna Power Station Unit 2 Modified S tartup Physics Testing Program Revision 52-09/29/2016 NAPS UFSAR 14A-ii Intentionally Blank Revision 52-09/29/2016 NAPS UFSAR 14A-1 APPENDIX 14ANRC QUESTIONS AND VEPCO'S RESPONSES REGARDING THE NORTH ANNA POWER STATION UNIT 2 M ODIFIED STARTUP PHYSICS TESTING PROGRAM Question 1With respect to Item B.3 on Table 14.1-3 and Item B.2 on Table 14.1-4 , what is the reason for performing the pseudo-dropped-rod test with RCCA D-10 instead of RCCA H-6?ResponseThis test was performed twice for Unit 1, once using RCCA D-10 and once using RCCA H-6. The evaluation of the results associated with these tests indicated that of the two rods, RCCA D-10 resulted in the more limiting radial power distribution and consequently had the minimum margin to departure from nucleate boiling. This was the basis for choosing RCCA D-10 for the Unit 2 test.Question 2Describe any known differ ences between the fuel and core of North Anna Unit 1 and the fuel and core of North Anna Unit 2.ResponseThree differences have been identified between the Unit 1 and Unit 2 fuel and core. They are:1.The location of the secondary sources within the core.2.The part-length control rods have been removed.3.The fuel rods have been prepressurized to a diff erent pressure.

The core locations of the secondary sources for Unit 1 and Unit 2 are shown in Figure 14A-1. It is not expected that this change will lead to a measurable difference in the physics characteristics between the Unit 1 and Unit 2 cores.The core location of the part-length control rods for Unit 1 is shown in Figure 14A-2. This change will not lead to a measurable diff erence in the physics characteristics between the Unit 1 and Unit 2 cores because the use of th e part-length control rods is not permitted during Unit 1 core operation. It is planned that these rods will be removed from Unit 1 following the end of Cycle 1 operation.

Unit 2 fuel has a prepressurization value that is approximately 50 psi lower than that used for Un it 1 fuel. This difference will have no perceptible effect on the physics characteristics of the Revision 52-09/29/2016 NAPS UFSAR 14A-2 core. In addition, an evaluation has shown that there will be no adverse impact on fuel or core performance.

Question 3Describe any differences between the physic s test methods that were used for Unit 1 and the physics test methods that will be used for Unit 2.Response For Unit 1, the reactivity wo rth of shutdown bank B was determined with control banks A through D fully inserted into the core, using th e dilution-boration technique. The worth of shutdown bank A was determined with control banks A through D and shutdown bank B fully inserted into the core. Shutdown bank A underwent an exchange wi th the most reactive rod (RCCA B-8) followed by a dilution of the reactor coolant system in order to fully insert shutdown bank A.For Unit 2, the reactivity worths of shutdown banks A and B will be determined individually with all other c ontrol rod banks out of the co re. The worth of shutdown bank B will be determined using the conventional di lution/boration techni que. A boron endpoint determination will be made for this control rod configuration. The worth of shutdown bank A will be determined by using rod exchange with one of the rod banks, and if necessary, dilution/boration of the reactor coolant system in order to reach the desired state point, i.e., shutdown bank A fully inserted w ith all other rods out. A boron e ndpoint determination will be made for this control rod configuration.

Question 4 For each of the start-up physics te sts that were performed for Unit 1 but are not going to be performed for Unit 2, give the technical basis for not performing those tests.

Response

The tests that are not going to be performed for Unit 2 are listed in Table 14.1-4. Deletion of these physics tests from the start-up program is justified for the following reasons:1.The successful performance of the abbreviated program is suff icient toa.Verify that the core was correctly loaded and that there are no anomalies present that could cause problems later in the cycle.b.Verify that the calculational model that has been used will correctly predict core behavior during the cycle.c.Verify the reactivity worth of the control rod banks.

Revision 52-09/29/2016 NAPS UFSAR 14A-3d.Provide data for nuclear instrumentation calibration.e.Demonstrate the sensitivity of this in strumentation to abnormal core conditions.2.The calculation model was verified as a result of the Unit 1 start-up.

3.The fuel and core characteristics of Unit 2 are virtually identical to those of Unit 1, and the results obtained for these tests during the Unit 1 start-up demonstrated that a large mar gin exists between the measured parameter values and the design values used in the safety analyses.Each of the tests that are not going to be performed for Unit 2 is listed below, together with the specific technical basis for not performing these te sts as part of the Unit 2 start-up physics testing program.A1.Isothermal Temperature coefficient at D, C banks in, D, C, B banks i n, and D, C, B, A banks in The Core Operating Limits Report will require that a nonpositive va lue for the moderator temperature coefficient be maintained during norm al operation. Base d on the results of design calculations and the Unit 1 moderator temperature coefficient test results, it is expected that performance of the moderator temperature coefficien t tests with all rods out and with control bank D in will be adequate to demonstrate a nonpositive moderato r temperature coef ficient value and also provide enough data to establish control rod withdrawal limits, should they be necessary. The successful completion of these test s will verify the design model used to predict the isothermal temperature coefficient values. Additionally, revi ew criteria have been developed for the Unit 2 isothermal temperature coef ficien t tests that are based on the Unit 1 test results (Table 14A-1). The acceptability of the Unit 2 tests with respect to these review criteria will further demonstrate the similarity between the Unit 1 and Unit 2 cores. All of the Unit 1 measured temperature coefficient values were acceptable. The Unit 1 test results and review criteria for the isothermal temperature coefficients at D, C banks in, D, C, B banks in, and D, C, B, A banks in are listed in Table 14.1-5 , Item 1.A2.Boron Endpoint for the N-1 rods-in configurationBoron endpoint measurements will be made fo llowing each of the rod worth tests. The successful completion of these tests will verify th e design model used to predict the boron endpoint values. Additionally, review criteria have been developed for the Unit 2 boron endpoint measurements that are based on the Unit 1 test results (Table 14A-1). The acceptability of the Unit 2 tests with respect to these review criteria will further demonstrate the similarity between the Unit 1 and Unit 2 cores. All Unit 1 measured endpoint values were acceptable. The Unit 1 test results and review criterion for the boron endpoint for the N-1 rods-i n configuration are listed in Table 14.1-5 , Item 2.

Revision 52-09/29/2016 NAPS UFSAR 14A-4A3.Reactivity worth of N-1 rods As described in the response to Question 3 , the reactivity worth of the control and shutdown rod banks will be measured as part of the physics testing program. The successful completion of these measurements will verify the design models used to predic t the re activity wo rth of the rod banks. Additionally, review criteria have been developed for the Unit 2 rod bank reactivity worth tests that are based on the Unit 1 test results (Table 14A-1). The acceptability of the Unit 2 tests with respect to these review criteria will fu rther demonstrate the similarity between the Unit 1 and Unit 2 cores. All Unit 1 measured rod worth values (includi ng the reactivity worth of N-1 rods) were acceptable, and demonstrated that a large margin existed with re spect to the shutdown margin limit. The Unit 1 test results and review criterion for the reactivit y worth of N-1 rods are listed in Table 14.1-5 , Item 3.A4.Pseudo-rod-ejection and associated power distribution measurements (HZP)

The successful completion of the rod bank reactivity worth measurements for the four control banks and the two shutdown banks for Unit 2 will verify the design model used to calculate rod worths. Sinc e the same design model is used to predict all rod worths, including the worth of an ejected rod, additiona l verification of the design model is not required. Additionally, review criteria have been developed for the Unit 2 rod bank reactivity worth tests that are based on the Unit 1 test results (Table 14A-1). The acceptability of the Unit 2 tests with respect to these review criteria will further demonstr ate the similarity between the Unit 1 and Unit 2 cores. The Unit 1 test results and review criteria for the pseudo-rod-ejection and associated power distribution measurements (HZP) are listed in Table 14.1-5 , Item 4. These results indicated that the measured rod worth value and the heat flux hot-channel factor value were acceptable, and demonstrated a large margin with respect to the values used in the safety analysis.

B1.Pseudo-rod-ejection and asso ciated power distribution measurement at 30% powerThe successful completion of the rod bank reactivity worth measurements for the four control banks and the two shutdown banks for Unit 2 will verify the design model used to calculate rod worths. Sinc e the same design model is used to predict all rod worths, including the worth of an ejected rod, additiona l verification of the design model is not required. Additionally, review criteria have been developed for the Unit 2 rod bank reactivity worth tests that are based on the Unit 1 test results (Table 14A-1). The acceptability of the Unit 2 tests with respect to these review criteria will further demonstr ate the similarity between the Unit 1 and Unit 2 cores. The Unit 1 test results and review criteria for the pseudo-rod-ejection and associated power distribution measurement at 30% power are listed in Table 14.1-5 , Item 4. These results indicated that the measured rod worth value and the heat flux hot-c hannel factor valu e were acceptable, and demonstrated a large margin with respect to the values used in the safety analysis.

Revision 52-09/29/2016 NAPS UFSAR 14A-5B2.Pseudo-dropped-rod test (RCCA H-6) and associated power distribution measurement As described in the response to Ques tion 1, this test was performed twice for Unit 1, once using RCCA D-10 and once using RCCA H-6. For Unit 2, this test will be performed using the limiting rod, RCCA D-10. The successful completion of this test will verify the design models and demonstrate margin to the values used in the safety analysis. Additionally, the use of review criteria that are based on Unit 1 test results for other Unit 2 tests (Table 14A-1) will further demonstrate the similar ity between the Unit 1 and Unit 2 cores. The Unit 1 results for both dropped rod tests were acceptable and demonstrated marg in with resp ect to the values used in the safety analysis. The Unit 1 test results and review criteria for the pseudo-dropped-rod test and associated power distribut ion measurement using RCCA H-6 are listed in Table 14.1-5 , Item 5.B3.Power coefficient testsB4.Integral power defectB5.Doppler-only power coefficientsThe successful completion of the isothermal temperature coefficient tests, the boron endpoint measurements, a nd the rod bank reacti vity tests, together with the acceptability of these tests with respect to their respective review criteria, will service to further demonstrate the similarity between the Unit 1 and Unit 2 cores. As indicated in Table 14.1-5 , Items 6, 7, and 8, the Unit 1 measured values for the total power coeffi cient, the integral power defect, and the Doppler-only power coef ficient veri fied the design models used to predict the values of these parameters and were acceptable wi th respect to the values used in the safety analyses. An additional description of the Unit 1 test results and their evaluation has been provided in a letter from Mr. C. M. Stallings, VEPCO, to Mr. H. R. Denton, US NRC, Ser. No.

169, dated March 20, 1979.In summary, this information provides a suffic ient technical basis fo r the deletion of these tests from the Unit 2 physics testing program.

Question 5 It is suggested that review criteria be established, where appropriate, to compare the Unit 2 test results with the Unit 1 test results for the isothermal temperature coefficient measurements, the boron endpoint measurements, the rod bank reactivity worth measurements, and the boron worth measurement. For each of these tests, list the specific review criteria that will be used. Also, indicate the action that will be take n if the review criteria are not met.

ResponseThe tests and the specific test review criteria that wi ll be used are listed on Table 14A-1. As described in the response to Question 3 , the reactivity worth and bo ron endpoint measurements associated with shutdown banks A and B that will be performed for Unit 2 are not direct Revision 52-09/29/2016 NAPS UFSAR 14A-6 duplicates of the tests that were performed during the Unit 1 testing program. Therefore, review criteria based on Unit 1 test results would be inappropriate. The results of these tests will be reviewed, instead, with respect to design values (best-estimate predictions) and the standard design tolerances.

As stated in the main body of the FSAR and as required by the VEPCO Nuclear Power Station Quality Assurance Manual, test results will be reviewed and evaluated by the Station Nuclear Safety and Operating Commit tee. Should the results of any of these tests fail to meet the review criteria, the Committee may decide to pe rform additional testing. This additional testing may be a repeat of the original test or may be the performance of a test that had been deleted from the Unit 2 physics testing program.

In addition, the NRC Region II Resident Inspector will be notified verbally in a timely manner, and a report will be sent to Nuclear Reactor Regulation (NRR).

Revision 52-09/29/2016 NAPS UFSAR 14A-7Table 14A-1 UNIT 2 ISOTHERMAL TEMPERATURE COEFFICIENT, BORON ENDPOINT, ROD WORTH REACTIVITY, AND BORON WORTH TESTS AND REVIEW CRITERIA Review CriteriaTest Description Unit 1 Measured ValueToleranceIsothermal temperature coefficient ARO 0.98 pcm/°F

+/-2 pcm/°F D bank in-4.29 pcm/°F

+/-2 pcm/°F Boron endpoint ARO 1322 ppm+/-24 ppm D bank in1193 ppm+/-24 ppm D, C banks in 1075 ppm+/-24 ppm D, C, B banks in 884 ppm+/-24 ppm D, C, B, A banks in 781 ppm+/-24 ppm Shutdown bank A in 1220 ppm (design)

+/-21 ppm Shutdown bank B in 1224 ppm (design)

+/-20 ppm Control rod worth D bank 1463 pcm+/-100 pcm C bank 1303 pcm+/-98 pcm B bank 2036 pcm+/-153 pcm A bank 1309 pcm+/-98 pcmTotal D through A6111 pcm+/-306 pcm Shutdown bank A1114 pcm (design)

+/-111 pcm Shutdown bank B 1043 pcm (design)

+/-104 pcm Boron worth ARO through A bank 1 1.08 +/-0.55 pcm ppm----------

-pcm ppm----------

-

Revision 52-09/29/2016 NAPS UFSAR 14A-8 Figure 14A-1SECONDARY SOURCE LOCATIONS Revision 52-09/29/2016 NAPS UFSAR 14A-9 Figure 14A-2PART LENGTH CONTROL ROD LOCATIONS (UNIT

1)

Revision 52-09/29/2016 NAPS UFSAR 14A-10 Intentionally Blank