ML061880482

NRC Regulatory Guide 1.68, Initial Test Programs for Water-Cooled Nuclear Power Plants, Revision 2 (August 1978)

ML061880482
Person / Time
Site:	Vermont Yankee
Issue date:	08/31/1978
From:	NRC/OGC
To:
Julian E
References
50-271-OLA, ASLBP 04-832-02-OLA, RAS 11812 RG-1.068, Rev 2
Download: ML061880482 (23)

Text

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IlRevlsi . Ion 2 0 U.S. NUCLEAR REGULATORY COMMISSION August 1978

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OFF EG U LATSOAA 1%,.* OFFICE OF STANDARDS DEVELOPMENT DEVE GLUOIME REGULATORY GUIDE 1.68 (F

INITIAL TEST PROGRAMS FOR WATER-COOLED NUCLEAR POWER PLANTS A. INTRODUCTION tested to (I) provide for safe normal operation and high tolerance for system malfunctions and transients, (ý) en-Section 50.34. "Contents of Applications: Technical sure that, in the event of errors, malfunctions, and off-Information," of 10 CFR Part 50, "Domestic Licensing normal conditions, the reactor protection systems and of Production and Utilization Facilities," requires, in other design features will arrest the event or limit its part, that an applicant for a license to operate a produc- consequences to defined and acceptable levels, and (3)

• tion or utilization facility include the.principal design ensure that adequate safety margin exists for events of criteria for the proposed facility in thb safety analysis extremely low probability or for arbitrarily postulated report (SAR). The Introduction to Appendix A, "Gen- hypothetical events without substantial reduction in the eral Design Criteria for Nuclear Power Plants," to 10 CFR safety margin for the protection of public health and Par- 50 states that these principal design criteria are to safety.

establish the necessary design, fabrication, construe-tion, testing, and performance requirements for struc- While it is required that all structures, systems, and tures, systems, and componenls important to safety, components important to safety be tested, it is not re-i.e., structures, systems, and components that provide quired that all of them be tested to the same stringent reasonable assurance that the facility can be operated requirements. Specifically, Criterion I of Appendix A without undue risk to the health and safety of the to 10 CFR Part 50 requires, in pan, that structures, sys-public. tems, and components important to safety be tested to quality standards commensurate with the importance of Section XI, "Test Control," of Appendix B, the safety functions to be performed.

"Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50 re- A graded approach is also inherent in the testing re-quires that a test program be established to ensure that quiremcnts of Criterion X1 of Appendix B to 10 CFR structures, systems, and components will perform satis- Part 50.

factorily in service. Since all functions designated in the general design criteia (GDC) are important to safety, Section 50.34 of 10 CFR Part 50 also requires, in part, that the applicant include plans for preoperational all structures, systems, and components required to per- testing and initial operations in the final safety analysis form these functions need to be tested to ensure that they will perform properly. These functions, as noted report (FSAR). Chapter 14 of Regulatory Guide 1.70, throughout the specific GDC, are those necessary to en- "Standard Format and Content of Safety Analysis Re-sure that specified design conditions of the facility are ports for Nuclear Power Plants," provides guidance on not exceeded during any condition of normal operation, the information pertaining to initial test programs to be included in both the preliminary safety analysis report including anticipated operational occurrences, or as a re- (PSAR) and the FSAR for the NRC staff to perform its sult of postulated accident conditions. safety evaluations for construction permits and operat-The GDC and this guide recognize and provide for ing licenses.

successive levels of plant features for achieving safety of the facility. This is to provide for a systematic ap- This guide describes the general scope and depth of proach to the "defense-in-depth" concept. This concept initial test programs acceptable to the NRC staff for requires that the plant be designed, constructed, and light-water-cooled nuclear power plants. Appendix A to this guide provides a representative listing of the plant o Lines Indicate substanilve changes from previous Issue, structures, systems, and components and the design fea-USNRC REGULATORY GUIDES Cun,anirnnthiald "wit ~

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tures and performance capability tests that should be complished in a cost-effective manner. Documentation demonstratcd during the initial test program. No par. associated with testing such as procedures and records ticular significance should be attached to the order in should be commensurate with the importance to safety

• which the tests are listed, although, in general, those of the item being tested.

under "I. Preoperational Testing" should precede thos:

listed under "2. Initial Fuel Loading and Precritical To provide for the development and safe execution of Tests," and so on. Appendix B to this guide provides the initial test program, the applicant should formulate information on inspections, relating to initial test pro- advance plans for the eatire testing program prior to grams, that will be performed by the NRC Office of completion of the NRC staff's construction permit re-Inspection and Enforcement. Appcndix C to this guide view. Because of the complexity of these tests and the provides guidance on the preparation of procedures for large amount of manpower needed for developing and the conduct of initial test programs. executing the complete program, it is iml5ortant for the applicant to give early consideration to the following:

The Advisory Committee on Reactor Safeguards has been consulted concerning this guide and has concurred 1. Defining the responsibilities of the organization in the regulatory position. that will carry out the program. This should include the degree of participation of principal design organizations B. DISCUSSION in formulating test objectives and acceptance criteria.

The applicant for a construction permit or operating 2. Developing realistic schedules for the preparation license is responsible for ensuring that a suitable initial of detailed testing, plant operating, and emergency pro-(preoperational and startup) test program will be con- cedures. Schedules need to be established for conduct-ducted for the facility. The primary objectives of a suit- ing the major phases of the test program relative to the able program are (I) to provide additional assurance expected fuel loading date.

that the facility has been adequately designed and, to the extent practical, to validate the analytical models 3. Establishing methods or plans for providing the and to verify the correctness or conservatism of assump- necessary manpower at the times needed to maintain the tions used for predicting plant responses to anticipated schedules. If service contracts are to be used, it is transients and postulated accidents and (2) to provide necessary to have sufficient trained staff for good con-assurance that constiruction and installation of equip- tract management. Hiring and training schedules for the ment in the facility hay: been accomplished in accord- plant operating and technical staff need to be estab-ance with design. Other key objectives are to lished so that experienced and qualified personnel will familiarize the plant operating and technical staff with be available for the development of testing, operating, the operation of the facility and to verify by trial use, to and emergency procedures. It is importzint to consider the extent practical, that the facility operating proce- the effects on staffing that could result from overlapping dures and the emergency procedures are adequate. Ini- initial test programs at multi-unit sites.

tial test programs satisfying these objectives should provide the necessary assurance that the facility can be 4. Formulating administrative controls to govern the operated in accordance with design requirements and in development and conduct of the initial test program in-a manner that will not endanger the health and safety of cluding (a) controls that will provide for orderly turn-the public. over of plant systems and components from construc-tion forces or other preliminary checkout groups to the As mentioned in the Introduction, the test program is preoperational testing group for testing and (b) controls required to include suitable testing of all structures, sys- that will ensure that general prerequisites such as com-tems. and components important to safety. Both Ap- pletion of construction, construc:ion or preliminary pcndices A and B to 10 CFR Part 50 recognize that tests, and inspections will be satisfied prior to preopera-some structures, systems, and components are more tional and/or startup tests of individual systems or com-important to safety than others. For example, those ponents.

structures, systems, and components that are designated as Seismic Category I by Regulatory Guide 1.29, Establishment of early plans for using available in-

"Seismic Design Classification," arc considered more formation on operating experience, including reportable important to safety than some of the other structures, occurrences from other operating power reactors, is im-systems, and components that are identified as impor- portant in the development and conduct of the test pro-tant to safety in the functional design criteria of Appen- gram to help minimize recurrence of significant prob-I dix A to 10 CFR Part 50. It is not intended that the lems that could have been avoided by more complete I same test requirements be established for all structures, testing. If new, unique, or first-of-a-kind principal de-systems, and components important to safety. A graded sign.features will be used in the facility, the in-plant approach to testing should be implemented in order that functional testing requirements necessary to verify their adequate assurance, considering the importance to performance need to be identified at an early date to safety of the item, is provided that the item will perform permit these test requirements to be appropriately ac- 4 satisfactorily while, at the same time, the testing is ac- counted for in the final design.

1.68-2

The initial test program consists of preoperational and C. REGULATORY POSITION initial startup tests. Preoperational testing, as used in this guide, consists of those tests conducted f6llowing 1. Criteria for Selection of Plant Features To Be completion of construction and construction-related in- Tested spections and tests, but prior to fuel loading, to demon- Each applicant or licensee should prepare and conduct stratc, to the extent practical, the capability of struc- an initial test program to demonstrate that the plant can tures, systems, and components to meet performance be operated in accordance with design requirements im-requirements to satisfy design criteria. portant to safety, as defined by Appendix A to 10 CFR Initial startup testing, as used in this guide, consists Part 50. Suitable tests should be conducted to verify the of those lest activities scheduled to be performed during performance capabilities, as delineated in Appendix A and following fuel loading. These activities include fuel to 10 CFR Part 50, of structures, systems, and compo-loading, precritical tests, initial criticality, low-power nents that:

tests, and power-ascension tests that confirm the design a. Will be used for shutdown and cooldown of the bases and demonstrate, to the extent practical, that the reactor under normal plant conditions and for maintain-plant will operate in accordance with design and is cap- ing the reactor in a safe condition for an extended shut-able of responding as designed to anticipated transients down period.

and postulated accidents as specified in the SAR.

b. Will be used for shutdown and cooldown of the The initial test program should be designed to dem- reactor under transient (infrequent or moderately fre-onstrate the performance of structures, systems, com- quent events) conditions and postulated accident condi-ponents, and design features that will be used during tions and for maintaining the reactor in a safe condition normal operations of the facility and also demonstrate for an extended shutdown period following such condi-the performance of standby. systems and features that tions.

must function to maintain the plant in a safe condition in the event of malfunctions or accidents. It is very im- c. Will be used for establishing conformance with portant that the sequence of startup tests be ordered so safety limits or limiting conditions for operation that the safety of the plant is never totally dependent on that will be included in the facility technical the performance of untested structures, systems, and specifications.

components. d. Are classified as engineered safety features or will The NRC staff's safety evaluations of initial test pro- be relied on to support or ensure the operations of en-grams are based on information provided in the PSAR gineered safety features within design limits.

and FSAR. This information is used to support deci- e. Are assumed to function or for which credit is sions to issue a construction permit or operating license. taken in the accident analysis of the facility, as de-The information provided in SARs is also used by the scribed in the FSAR, and NRC Office of Inspection and Enforcement as a basis for the inspection activities associated with initial test f. Will be used to process, store, control, or limit the programs. The satisfactory performance of approved release of radioactive materials.

test programs provides the confirmation that adequate Appendix A to this guide provides a representative margins of safety exist such that there is no undue risk listing of systems to be tested and performance to the health and safety of the public as a result of facil- capabilities important to safety, as defined by Appendix ity operation. A to 10 CFR Part 50, that should be demonstrated for The power-ascension test phase of the initial test pro- light-water-cooled nuclear power plants. However, ap-gram should be completed in an orderly and expeditious plicants should also conduct in-plant testing to verify manner. Failure to complete the power-ascension test the adequacy of construction, installation, and design phase within a reasonable period of time may indicate for other systems and design features not listed in Ap-inadequacies in the applicant's operating and mainte- pendix A if the systems or design features meet any of nance capabilities or may result from basic design prob- the above criteria.

lems. Also, design or construction-related problems The initial test program may be developed and im-disclosed during power-ascension testing can be more plemented using a graded approach. The graded ap-readily rectified if the reactor power production, and proach should ensure that the greatest attention is given consequently the radioactive buildup, has been kept to a to the most important structures, systems, and compo-minimum during this testing phase. Baseline data on the nents such as those considered engineered safety fea-performance of plant systems obtained and documehted tu res.

early in the plant life will permit early determination of degradation or undesirable trends. 2. Prerequisites for Testing Appendix A references existing regulatory guides that The construction or installation of structures, sys-are applicable to initial test programs. The referenced tems, and components should be essentially completed guides provide detailed guidance for particular tests. (to the degree that outstanding construction items could

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-not be expected to affect the validity of test results). Enforcement approximately 60 days prior to their in-The designated cons.truction.related inspections and tended use.

tests should also be completed prior to beginning preop-erational tests. Prior to commencement of fuel loading, results or completed preoperational tests should be evaluated by Tests designated in the FSAR as preoperational tests personnel or groups designated by the applicant. Ap-should be completed and the results of such tests propriate remedial actions, including retesting, should evaluated and approved by the applicant prior to is- be taken if acceptance criteria are not satisfied.

suance of the Operating License. The overall test pro-gram should also include surveillance tests necessary to 5. Schedule demonstrate the proper operation of interlocks, set- Sufficient time should be scheduled to perform or-points, and other protective features, systems, and derly and comprehensive testing. The applicant's equipment required by the technical specifications. schedules for conducting the preoperational phase and Administrative controls should be established to ensure the initial startup phase should provide for a minimum adequate retest of systems or design features returned to time of approximately 9 months and 3 months, respec-construction custody, maintained, or modified during or tively.

following preoperational testing.

6. Participation of Plant Operating and Technical

3. Scope, Testing Conditions, and Length of Testing Staff The initial test program should include, to the extent I practical, simulation of the effects of control system and equipment failures or malfunctions that could rea-The applicant's plant operating and plant technical staff should participate, to the extent practical, in the development and conduct of the initial test program and sonably be expected to occur during the plant lifetime. the evaluation of the test results.

The test program should also include testing to deter-mine that the system and component interactions are in 7. Trial-Testing of Plant Operating and Emergency accordance with design. To the extent practical, the Procedures plant conditions during the tests should simulate the ac- Plant operating and emergency procedures should, to tual operating and emergency conditions to which the the extent practical, be developed, trial-tested, and cor-structure, system, or component may be subjected. To rected during the initial test program prior to fuel load-the extent practical, the duration of the tests should be ing to establish their adequacy.

sufficient to permit equipment to reach its normal equilibrium conditions, e.g., temperatures and pres- 8. Mllestones and Power Hold Points sures, and thus decrease the probability of failures, in- Appropriate hold points should be established by the cluding "run-in" type failures, from occurring during applicant at selected milestones throughout the power-plant operation. ascension test phase to ensure that relevant test results

4. Procedures are evaluated and approved by personnel or groups des-ignated by the applicant prior to progressing with the The initial test program should be conducted using power-ascension test phase. As a minimum, hold points test procedures that include appropriate checklists and should be established for PWRs at" approximately 25%,

signature blocks to control test performance and the se- 50%, and 75% power level test conditions and for quence of testing. The test procedures should be de- BWRs at appropriate power-to-flow test conditions.

veloped and reviewed by personnel with appropriate technical backgrounds and experience. The procedures 9. Test Reports should receive final approval by persons filling desig- The preoperational testing procedures and results nated management positions within the applicant's or-ganization. Acceptance criteria that account for the un-should be retained as part of the plant historical record.

A summary of the startup testing should be included in I

certainties used in transient and accident analyses should be included as part of each test procedure. Pro-a startup report as discussed in Regulatory Guide 1.16, ¶ "Reporting of Operating Information-Appendix A cedures should ensure that temporary instrument cables Technical Specifications." This summary should in-and test leads used during the startup test phase are clude:

routed in a manner that will not compromise electrical separation criteria. Principal design organizations a. A description of the test method and objectives for should participate in establishing test performance re- each test.

quirements and test acceptance criteria. Available in-

b. A comparison of applicable test data with the ac-formation on operating experience, including reportable ceptance criteria, including the response of the systems occurrences at operating power reactors, should be used to major plant transients such as scram and turbine trip.

appropriately in the development and execution of the test procedures. Approved test procedures for satisfying c. Deficiencies relating to design and construction I FSAR testing commitments should be made available to NRC staff personnel from the Office of Inspection and found during conduct of the tests, system modifications and corrective actions required to correct these deficien-1,68-4 I

I4 cies, and the schedule for implementing these modifica- formation to applicants regarding the NRC staff's plans tions and corrective actions unless previously reported for using this regulatory guide.

to the Nuclear Regulatory Commission.

Except in those cases in which the applicant proposes

d. Justification for acceptance of systems or compo- an acceptable alternative method for complying with nents not in conformance with design predictions or specified portions of the Commission's regulations, the performance requirements. method described herein will be used in the evaluation

e. Conclusions regarding system or component ade- of construction permit and operating license applica-quacy.

tions docketed after August 15, 1978.

If an applicant wishes to use this regulatory guide in D. IMPLEMENTATION developing submittals for applications docketed on or before August 15, 1978, the pertinent portions of the The purpose of this section is to provide further in- application will be evaluated on the basis of this guide.

4-1.68-5

APPENDIX A INITIAL TEST PROGRAM

1. Preoperational Testing rates, systems, and components as appropriate for the facility. Preoperational .tests should not be limited to Following plant construction, testing should be ac-the listing provided in items a. through o. since addi-complished to demonstrate the proper performance of structures, systems, components, and design features tional or different tests may be dictated by the par-ticular plant design and/or the nomenclature applied in the assembled plant. To ensure valid test results, to plant systems and features.

the preoperational tests should not proceed until the construction of the system has been essentially com- a. Reactor Coolant System pleted and the designated construction tests and inspections have been satisfactorily completed. Con- The reactor coolant system includes all those struction and preliminary tests and inspections typi- pressure-containing components such as pressure ves-cally consist of items such as initial instrument cali- sels, piping, pumps, and valves within the reactor bration, flushing, cleaning, wiring continuity and coolant pressure boundary as defined in paragraph separation checks, hydrostatic pressure tests, and 50.2(v) of 10 CFR Part 50.

functional tests of components. (1) Integrated Systems Test. Expansion and re-Preoperational tests should demonstrate that struc- straint tests to confirm acceptability of clearances and

• tures, systems, and components will operate in ac- displacements of vessels, piping, piping hangers, and cordance with design in all operating modes and seismic and other holddown, support, or restraining throughout the full design operating range. Testing devices in the as-built system during normal hot func- I should include, as appropriate, manual operation, op- tional testing plant conditions. Hot and/or cold test-eration of systems and components within systems, ing of the system with simultaneous operation of automatic operation, operation in all alternate or auxiliary systems.

secondary modes of control, and operation and ver- (2) Component Tests. Appropriate tests and meas-ification tests to demonstrate expected operation fol- urements of the following reactor coolant system lowing loss of power sources and degraded modes for components:

which the systems are designed to remain opera- (a) Pressurizer.

tional. Tests should also include, as appropriate, verifications of the proper functioning of instrumenta- (b) Pumps, motors, and associatcd power tion and controls, permissive and prohibit interlocks, sources.

and equipment protective devices whose malfunction (c) Steam generators.

or premature actuation may shut down or defeat the (d) Pressure relief valves and associated dump operation of systems or equipment. System vibration, tanks and supports and restraints for discharge-expansion (in discrete temperature step increments), piping.

and restraint tests should also be conducted. This testing should include verification by observations (e) Main steam isolation valves.

and measurements, as appropriate, that piping and (f) Other valves.

component movements, vibrations, and expansions (g) Instrumentation used for monitoring system are acceptable for (1) ASME Code Class 1, 2, and 3 performance or performing permissive and prohibit systems, (2) other high-energy piping systems inside interlock functions.

Seismic Category I structures, (3) high-energy por- (h) Reactor vessel and internals, including reac-tions of systems whose failure could reduce the func- tor internals vent valves.

tioning of any Seismic Category I plant feature to an unacceptable level, and (4) Seismic Category I por- (i) Safety valves.

tions of moderate-energy piping systems located out- j) Jet pumps.

side containment. (3) Vibration Tests. Vibration monitoring of reac-The structures, systems, components, and tests tor internals I and of other components such as piping listed in items a. through o. of this section are repre- systems, heat exchangers, and rotating machinery.

sentative of the plant features that should undergo (4) Pressure Boundary Integrity Tests. Hydrosta-preoperational testing. The listing is provided to indi- tic tests; obtain baseline data for subsequent inservice cate the extent of testing necessary to demonstrate testing.

that the facility can be operated in accordance with design requirements. In general, items a. through o.

make no distinction between pressurized water reac-I tors and boiling water reactors. An applicant may 'Regulatory Guide 1.20, "Comprehensive Vibration Assess-ment Program for Reactor Internals During Preoperational and combine tests of items listed in this appendix and Initial Stanup Testing," should be used as guidance for vibra-should include preoperational tests of the listed struc- ela monitoring of rrmator iwnals and ow comonentsu.

1.68-6

b. Reactivity Control Systems onstrate redundancy, electrical independence, 3 coin-(1) Control Rod System Tests. Demonstrate nor-mal operation and scram capability of the control cidence, and safe failure on loss of power. If appro-priate for the facility design, demonstrate'operability of backup scram solenoid valves and devices, includ-I rods (BWR) and control rod drive system. Demon- ing detectors, logic,'and final control elements to strate proper operation of functions such as control protect the facility for anticipated transients without I rod withdrawal inhibit features, runback features, rod withdrawal sequence control devices, and rod worth minimizers. Demonstrate proper operation of rod po-scram (ATWS).

d. Residual or Decay Heat Removal Systems sition instrumentation and proper interaction of the Verify operability of systems and design features control rod drive system with other systems and de- provided or relied on to dissipate or channel thermal sign features such as automatic reactor power control energy from the reactor to the atmosphere or to the systems and refueling equipment. Demonstrate main condenser or other systems following off-proper operation, including correct failurie mode on normal conditions or anticipated transients, including loss of power, for the control rod drive system and reactor scram. Verify operability of systems and de-proper operation of system alarms. sign features provided for makeup of coolant, to dis-sipate residual heat, to cool the reactor down to a (2) Chemical Control System Tests. Verify proper cold shutdown condition, and to maintain long-term blending of boron solution and water, uniform mix- cooling. Tests should be conducted as appropriate to ing, adequacy of sampling and analytical techniques, verify redundancy and electrical independence. 3 The operation of heaters and heat tracing, and operation following list is illustrative of the systems and com-of instrumentation, controls, interlocks, and alarms. ponents that should be tested:

Demonstrate proper rate injection into the reactor (I) Turbine bypass valves.

coolant system and rate of dilution from the primary system. Verify redundancy, electrical independence, (2) Steam line atmospheric dump valves.

and operability of system components. Demonstrate (3) Relief valves.

correct failure mode on loss of power to system components. (4) Safety valves.

(5) Decay or residual heat remo,,al system.

(3) Standby Liquid Control System Tests. Demon- (6) Reactor core isolation cooling system.

strate proper operation of the system with de-mineralized water. Verify proper mixing of solution (7) Main steam isolation valves, branch steam and adequacy of sampling system. Demonstrate isolation valves, and nonreturn valves.

operability of instrumentation, controls, interlocks, (8) Auxiliary feedwater systems. Testing should and alarms. Verify operability of heaters, air spar- include demonstrations that the systems will meet de-gers, and heat tracing. Conduct test firings of sign performance requirements at approximately squib-actuated valves, and demonstrate design normal operating primary and secondary coolant sys-injection capability. Tests should be conducted as tem pressures and temperatures and over the range of appropriate to verify redundancy and electrical expected steam generator levels. Operability cf sys-independence. tem pumps, valves, controls, and instrumentation

c. Reactor Protection System and Engineered.

Safety-Feature Actuation Systems should be demonstrated, and, to the extent practical, testing should provide reasonable assurance that flow instabilities, e.g., "water hammer," will not occur I

Verify by test the response time of each of the pro- in system components, piping, or inside the steam tection channels, including sensors.2 Acceptance generators during normal system startup and criteria for the response time of the protection chan- operation.

nels should account for the response time of the as- (9) Condensate storage system.

sociated hardware between the measured variable and the input to the sensor (snubbers, sensing lines, flow- (10) Emergency cooling towers.

limiting devices, etc.). Verify proper operittion in all (11) Cooling water systems.

combinations of logic: calibration and operability of primary sensors: proper trip and alarm settings; e. Power Conversion System proper operation of permissive, prohibit, and bypass The power conversion system includes all compo-functions; and operability of bypass switches. Dem- nents pirvided to channel the reactor thermal energy 2

Regulatory Guide 1.118, "Perizdir Testing of Electric Power 'Regulhtory Guide 1.41, "Preoperational Testing of Redundant and Protction Systems." provides a test cuiterion also aceept. Onsite Electric Power Systems to Verify Proper Load Group able for preoperational testing of protection channels, includ. Assignments." should be used as guidance for appropriate ing sensors. tests.

1.68-7

during normal operation from the boundaries 6f the d.c. systems. Appropriate system and component reactor coolant system to the main condenser and tests should be conducted to verify, to the extent those systems and components provided for return of practical, that these systems will operate in accord-condensate and feedwater4 from the main condenser ance with design.

to complete the cycle.

(1) Normal A.C. P~ower Distribution System.

Appropriate system expansion, restraint, and Demonstrate proper operation of protective devices, operability tests should be conducted, to the extent initiating devices, relaying and logic, transfer and practical, for the following systems and components: trip devices, permissive and prohibit interlocks, (1) Steam generators. instrurrentation and alarms, and load-shedding fea-tures. Testing should also be conducted to demon-(2) Main steam system. strate proper operation and load-carrying capability (3) Main steam isolation valves. of breakers, motor controllers, switchgear, transfor.

mers, and cables. This testing should simulate, as (4) Steam generator pressure relief and safety closely as practical, actual service conditions, e.g.,

valves. fully loading motor control centers and operation of supplied loads at rated conditions, etc. Redundancy (5) Steam extraction system. and electrical independence' should be demonstrated (6) Turbine stop, control, bypass, and intercept where appropriate.

valves. Tests should demonstrate that the integrated sys-(7) Main condenser hotwell level control system. tem will perform as designed to a simulated partial and full loss of offsite power sources. Tests should (8) Condensate system.

(9) Feedwater system.

also demonstrate the design capability to transfer from onsite to offsite power sources.

I (2) Emergency A.C. Power Distribution System.

(10) Feedwater heater and drain systems. Demonstirate proper operation of protective devices, (11) Makeup water and chemical treatment relaying and logic, transfer and trip devices, permis-systems. sive and prohibit interlocks, instrumentation and alarms, and load-shedding or stripping features. Test-(12) Main condenser auxiliaries used for maintain- ing should also be conducted .to demonstrate proper ing condenser vacuum. operation and load-carrying capability of breakers,

f. Waste Heat Rejection Systems motor controllers, switchgear, transformers, and ca-bles. This testing should simulate, as closely as prac-The waste heat rejection systems include systems tical, actual seryice conditions, e.g., fully loading and components provided to remove the unused or motor control centers and operation of supplied loads wasted thermal energy from systems such as the at rated conditions. Tests should demonstrate that power conversion and residual heat removal system emergency or vital loads will start in the proper se-and to channel or direct this energy to the environ- quence and operate under simulated accident condi-ment. Tests should be conducted as appropriate to tions with both the normal (preferred) a.c. power verify redundancy and electrical independence.3 Ap-source(s) and the emergency (standby) power source. I propriate system operability tests should also be con-ducted to demonstrate, to the extent practical, that Emergency loads should also be tested to dem-the following waste heat rejection systems and com- onstrate that they can start and operate with the ponents, including associated instrumentation and maximum and minimum design voltage available. To controls, will perform as designed: the extent practical, the testing of emergency or vital loads should be conducted for a sufficient period of (1) Circulating water system. time to provide assurance that equilibrium conditions (2) Cooling towers and associated auxiliaries. are attained. System redundancy and electrical inde-pendence should be verified by appropriate tests.3 (3) Raw water and service water cooling systems. Loads supplied from the system, such as motor

g. Electrical Systems generator (m-g) sets with flywheels, that are designed The plant electrical systems include the normal to provide noninterruptible power to plant loads should be tested to demonstrate proper operation. If a.c. power distribution system, the emergency a.c.

power distribution system including vital buses, the applicable for the facility design, testing should in-emergency a.c. power supplies or sources, and the clude underfrequency and undervoltage relays as-sociated with such m-g rts. Full-load tests for vital buses should be condlicted using normal and emer-

'Regulatory Guide 1.68.1, "Preoperational and Initial Startup Testing of Feedwater and Condensate Systems for Boiling gency sources of power supplies to the bus. Testing Water Reactor Power Plants," should be used as guidance for should also demonstrate the adequacy of the plant's appropriate tests. emergency and essential lighting system. Tests 1.68-8

should be conducted to demonstrate the proper opera- worth minimizers for boiling water reactors. Since tion of indicating and alarm devices used to monitor engineered safety features vary for different plant de-the availability of the emergency power system in the signs, the listing below is only illustrative of those control room. commonly used to prevent, limit, or mitigate the con-sequences of postulated accidents. If additional.or (3) Emergency or Standby A.C. Power Supplies. different types of engineered safety features are pro-Appropriate tests should be conducted for emergency vided than those listed below, they should also be a.c. power supplies to demonstrate system reliabil- appropriately tested. Additionally, it should be noted ity,s redundancy, electrical dependence, and proper that other categories of systems listed in Section 1 of voltage and frequency regulation under transient and this appendix include plant features commonly desig-steady-state conditions. Auxiliary systems such as nated as engineered safety features that should be ap-those used for starting, cooling, heating, ventilating, propriately tested; for example, emergency a.c.

lubricating, and fueling should also be appropriately power distribution system [Section 1.g (2)], emer-tested to demonstrate that their performance is in ac- gency or standby a.c. power supplies [Section l.g cordance with design. Testing should be conducted (3)], the d.c. system [Section l.g (4)], and primary for a sufficient period of tini to ensure that equilib- and secondary containments (Section L.i).

rium conditions are attained. Testing should also demonstrate the proper logic, correct setpoints for The testing of engineered safety features should trip devices, and proper operation of initiating devices demonstrate that such features will perform satisfac-and permissive and prohibit interlocks and should torily in all expected operating configurations or also demonstrate redundancy and electrical independ- modes. Testing should include demonstrations of ence. 3 Emergency loads supplied should be con- proper operation of initiating devices, correct logic firmed to be in agreement with design sizing assump- and setpoints, proper operation of bypasses, proper tions used for the power supplies.' operation of prohibit and pFrmissive interlocks, and proper operation of equipment protective devices that (4) D.C. System. Demonstrate proper calibration could shut down or defeat the operation or function-and trip settings of protective devices, including re- ing of such features. Concurrent testing of systems or laying, and proper operation of permissive and pro- features provided to ensure or support the operation hibit interlocks. Demonstrate design capability of bat- of engineered safety features should also be con-tery chargers, transfer devices and inverters, and the ducted to demonstrate that they meet design require-emergency lighting systems. Testing should also be ments with the minimum number of operable compo-conducted to demonstrate proper operation of break- nents available for which these systems are designed ers, transfer devices, inverters, and cables. This test- to function. Examples of these types of systems are ing should simulate, as closely as practical, actual heating, ventilation, and air-conditioning systems service conditions. Demonstrate operation of in- used to maintain the environment within design limits strumentation and alarms and ground detection in- in the spaces housing engineered safety features, strumentation. Demonstrate redundancy and electri- cooling* water and seal injection systems, and *pro-cal independence 3 and that actual total system am- tected compressed gas supplies. Appropriate tests perage loads are in agreement with design loads. A should also be conducted to verify the functioning of discharge test of each battery bank should be con- protective devices such as leaktight covers, stme-ducted at full load and for design duration to demon- tures, or housings (low pressure pneumatic or vac-C strate that the battery bank voltage minimum limit and uum tests) provided to protect engineered safety fea-I individual cell limits are not exceeded. tures from flooding or keep-full systems used to pre-

h. Engineered Safety Features vent water hammer and possible damage to fluid Engineered safety features are those plant design systems.

features provided to prevent, limit, or mitigate the Tests should be conducted as appropriate to verify consequences of postulated accidents that are de- redundancy and electrical independence. 3 The follow-scribed in the safety analysis r6port. For the purpose ing list is illustrative of the systems and components of this guide, engineered safety features include fea- that should be tested:

tures that prevent accidents from occurring or that bound accident assumptions such as cold water injec- (i) Emergency core cooling systems (ECCS)7 tion interlocks for pressurized water reactors and rod (a) Perform expansion and restraint tests.

(b) Demonstrate operability using normal and

'Regulatory Guide 1.10, "'Periodic Testing of Diesel emergency power supplies. I Generator Units Used As Onsite Electric Power Systems at Nu-clear Power Plants." should be used as guidance for applicable (c) Demonstrate operability in ;iJ modes of op-tests.

'Regulatory Guide 1.9. "Selection, Design, and Qualification for Diesel-Generator Units Used As Onsite Electric Power Sys- 'Regulatory Guide 1.79. "Preoperational Testing of Emergency tems at Nuclear Power Plants," should be used as guidance for Core Cooling Systems for Pressurized Water Reactors." pro-applicable tests. vides specific guidance for pressurized water reactors.

1.68-9

eration, including design pump/system runout condi- operate in accordance with design requirements at the tions and injection at required flow rate and pressure. containment design peak accident pressure.

(d) Demonstrate operability of interlocks and (10) Ultimate heat sink.

isolation valves provided for overpressure protection for low pressure cooling systems connected to the i. Primary and Secondary Containments reactor coolant system.

(e) Demonstrate operability, including proper Appropriate tests should be conducted to demon-flow rates, for systems used for dilution of boron in strate that primary and secondary containments will the reactor vessel during post-loss-of-coolant- function as designed. Testing methods and ace-opt-accident long-term cooling. ance criteria for such tests should give due considera-tion to all systems and components that must operate (2) Autodepressurization system. Testing should for the containments to function as designed. In cer-include items such as accumulator capacity, relief tain designs, normally operating or intermittently valves, and operability using all alternate power and operating systems may be required to shut down and pneumatic supplies. isolate to achieve containment isolation. For exam-(3) Containment postaccident heat removal sys- ple, the secondary containment ventilation system in tems. Testing of the containment spray system should BWRs is required to shut down and isolate the include demonstrations that the spray nozzles, spray normal ventilation paths to permit the standby gas*

headers, and piping ate free of debris; chemical addi- treatment system to perform its design function.

tion systems operate properly; and proper transfer to Therefore, appropriate testing should be conducted to the recirculation phase'can be accomplished. demonstrate the operability of all components, fea-tures, and systems required to operate for the primary (4) Containment combustible gas control system or secondary containment to function properly.

(includes the backup purge system). For containment combustible gas control systems located outside con- Due consideration should also be given to plant tainment, testing should include demonstration that features such as heating, ventilation, and air-the containment hydrogen monitoring is functional conditioning systems required to maintain environ-without the operation of the hydrogen recombiner. mental conditions within design limits for compo-For hydrogen recombiners shared between plants or nents or equipment provided to effect containment sites, tests should include demonstrations that the isolation. Testing should be sufficient to demonstrate shared recombiner can be transported and connected redundancy, electrical independence 3 requirements to the combustible gas control system within the time for isolation valves, and proper operation of features stated in the FSAR. (including proper operation of devices upon loss or failure of motive power) provided for isolation valves (5) Cold water interlocks, including logic, cir- and other devices. To the extent practical, it should cuitry, and final control devices used to prevent cold be demonstrated that isolation devices perform as re-water injection into the reactor vessel. quired under simulated accident conditions. The list-(6) Air returnfans used in ice con( .nser contain- ing below is illustrative of systems, features, and per-ments and suppression pool makeup systems used in formance demonstrations that should be included in BWR Mark III containments. the test program:

(7) Ventilation, recirculation, and filter systems (1) Containment design overpressure structural tests' and vacuum tests (for subatmospheric contain-provided to minimize radioactive releases as a result of postulated accidents, including fuel handling ments).

accidents.' (2) Containment isolation valve functional and I (8) Tanks and other sources of water used for closure timing tests.

ECCS. Testing should include demonstrations of proper operation of associated alarms, indicators, (3) Containment isolation valve leak rate tests t I controls, heating and chilling systems, and valves. and inleakage tests (for subatmospheric contain-ments).

(9) Containment recirculationfans (if used as part (4) Containment penetration leakage tests?10 of postaccident containment heat removal systems).

Testing should include demonstrations that fans can (5) Containment airlock leak rate tests.t 0 (6) Integrated containment leakage tests."

$These tests should be consistent with the provisions of Regula-tory Guide 1.52, "Design, Testing, and Maintenance Criteria 'Per Section Ill 'of the ASME Boiler and Ptessure Vessel Code.

for Engineered.Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water.Cooled Nuclear Tbe rNquirments for such tests am given in Appendix J to 10 Power Plants." CFR Part 50 1.69-10

In general, the sequences of testing should proceed bounded by the design of control and instrumentation from the low-pressure test to the accident-pressure systems (e.g., pressurizer level or feedwater flow test, or sufficient time should be allowed between control). In such cases, operation of the instrumenta-tests to ensure that outgassing from concrete or com- tion and controls over the design operating range ponents within the containment will not affect test should be performed, and the effects of limiting mal-results. functions or failures should be simulated to demon-strate the adequacy of design and installation and the (7) Main steam line leakage sealing systems. validity of accident analysis assumptions. Tests (8) Primary and secondary containment isolation should be conducted, as appropriate, to verify redun-initiation logic tests. dancy and electrical independence.3 (9) Containment purge system tests. The listing provided below is illustrative of in-strumentation and control systems that should be in-(10) Containment and containment annulus cluded in the test program (some of these tests can be vacuum-breaker tests. conducted in conjunction with the appropriate system (11) Containment supplementary leak collection level tests):

and exhaust system tests. (1) Pressurizer pressure and level control systems.

(12) Containment air purification and cleanup sys- (2) Main, auxiliary, and emergency feedwater con-tem tests. trol systems.

(13) Containment inerting system tests. (3) Secondary system steam'pressure control (14) Standby gas treatment system tests. system.

(15) Containment penetration pressurization sys- (4) Recirculation flow control system.

tem tests. (5) Reactor coolant system leak detection systems.

(16) Containment ventilation system tests. (6) Loose parts monitoring system.

(17) Secondary containment system ventilation (7) Leak detection systems used to detect failures tests. in ECCS and containment recirculating spray sysfems (18) Containment annulus and cleanup system located outside containment.

tests, including demonstrating the ability to maintain (8) Automatic reactor power.control system, inte-design pressure control in all modes of operation.

grated control system, and T-average control system.

(19) Bypass leakage tests on pressure suppression containments. (9) Pressure control systems used to maintain de-sign differential pressures to prevent leakage across (20) Ice-condenster containments. Sufficient meas- boundaries provided to contain fission products; for urements should be made to ensure that gross bypass example, those used to pressurize spaces between leakage paths are not present. containment isolation valves.

(21) Cootainment penetration cooling system tests. (10) Seismic instrumentation.

J. Instrumentation and Control Systems (11) Traversing incore probe system.

The nomenclature applied to instrumentation and (12) Failed fuel detection system.

control systems varies widely with different plant de-signs; however, the primary functions are similar for (13) Incore and excore neutron instrumentation.

all reactors. The principal functions of instrumenta- (14) Instrumentation and controls that effect trans.

tion and control systems are to (I) control the normal fers of water supplies to auxiliary feedwater pumps, operation of the facility within design limits, (2) pro- ECCS pumps, and containment spray pumps.

vide information and alarms in the control room to monitor the operation and status of the facility and to (15) Automatic dispatcher control systems.

permit corrective actions to be taken for off-normal plant conditions, (3) establish that the facility is (16) Hotwell level control system.

operating within design and license limits, (4) permit (17) Feedwater heater temperature, level, and by-or support the correct operation of enginecred safety pass control systems.

features, and (5) monitor and record important parameters during and following postulated (18) Auxiliary startup instrument tests (neutron re-accidents. sponse checks).

In the design of nuclear power plants, postulated (19) Instrumentption and controls used for shut-accident assumptions are often explicitly or implicitly down from outside the control room.

1.68-11

(20) Instrumentation used to detect external and and/or proper calibration of radiation detectors and internal flooding conditions that could result from monitors.

such sources as fluid system piping failures.

The following list is illustrative of the systems, (21) Reactor mode switch and associated func- components, and features whose operability should tions. be demonstrated during the test program:

(22) Instrumentation that Fan be used to track the (1) Liquid radioactive waste handling systems.

course of postulated accidents such as containment wide-range pressure indicators, reactor vessel water (2) Gaseous radioactive waste handling systems.

level monitors, containment sump or pressure sup- (3) Solid waste handling systems. Solidification pression level monitors, high-range radiation detec- system tests should include verification that no free I tion devices, and humidity monitors. liquids are present in packaged wastes.

(23) Postaccident hydrogen monitors and analyz- (4) Isolatiun features for steam generator blow-ers used in the combustible gas control system. down.

(24) Annunciators for reactor control and en- (5) Isolation features for condenser offgas gineered safety features. systems.

(25) Process computers. (6) Isolation features for ventilation systems.

k. Radiation Protection Systems (7) Isolation features for liquid radwaste effluent systems.

Appropriate tests should be conducted to demon-strate the proper operation of the following types of (8) Plant sampling systems.

systems and components used to monitor or measure m. Fuel Storage and Handling Systems radiation levels, to provide for personnel protection, Appropriate tests should be conducted for equip-or to control or limit the release of radioactivity:

ment and components used to handle or cool

. (1) Process, criticality, effluent, and area radiation irradiated fuel and tb handle nonirradiaied fuel to monitor tests. demonstrate that they wi!l operate in accordance with (2) Personnel monitors and radiation survey in- design. Tests should be conducted as appropriate to verify redundancy and electrical independence. 3 The strument tests. following list is illustrative of .the equipment and (3) Laboratory equipment used to analyze or component tests that should be included in the test measure radiation levels and radioactivity concentra- program:

tions.

(1) Spent fuel pit cooling system tests, including (4) High Efficiency Particulate Air (HEPA) filter the testing of antisiphon devices, high radiation and charcoal adsorber efficiency and in-place leak alarms, and low water level alarms..

tests.'

(2) Refueling equipment tests, including hand Tests should be conducted as appropriate 3 to verify tools, power equipment, bridge and overhead cranes, redundancy and electrical independence. and grapples. Testing should demonstrate the opera-bility of protective interlocks and devices.

1. Radioactive Waste Handling and Storage (3) Operability and leak tests of sectionalizing de-Systems vices and drains and leak tests of gaskets or bellows Appropriate tests should be conducted to demon- in the refueling canal and fuel storage pool.

strate the functional operability and design flow rates (4) Dynamic and static load testing' of cranes, of systems and components used to process, store, hoists, and associated lifting and rigging equipment, I and release or control the release, of radioactive liq- including the fuel cask handling crane. Static testing uid, gaseous, and solid wastes. Testing should dem- at 125% of rated load and full operational testing at onstrate, to the extent practical, that-the pumps, 100% of rated load.

tanks, controls, valves, and other equipment, includ-ing automatic isolation and protective features and (5) Fuel transfer devices.

instrumentation and alarms, will operate and function (6) Irradiated fuel pool or building ventilation sys-in accordance with design. Testing or calculations should include, as alipropriate, verification of tank tem tests.

volumes, capacities, holdup times, and proper opera-tion and calibration of associated instrumentation. "IRegulatory Guide 1.104, "Overhead Crane Handling Systems Spiked samples of the typical media, or sources for Nuclear Power Plants." should be used as guidance for tests should be used where necessary to verify operability on single-failure-proof overhead crane handling systems.

1.68-12

11. Auxiliary and Miscellaneous Systems (d) Diesel generator buildings.

Appropriate tests should be conducted to demon- (e) Auxiliary buildings, reactor building, tur-strate the operability of auxiliary and miscellaneous bine building, and. radioactive waste handling systems. Tests should be conducted, as appropriate, building.

to verify redundancy and electrical independence.3 (f) Control room habitability systems. Testing should include, as appropriate, demonstrations of the The following list is illustrative of the types of sys- proper operation of smoke and toxic chemical detec-tems or features whose performance should be dem- tion systems and ventilation shutdown devices, in-onstrated by testing: cluding leaktightness of ducts and flow rates, proper (1) Service and raw water cooling systems. direction of airflows, and proper control of space temperatures.

(2) Closed loop cooling water systems.

(15) Shield cooling systems.

(3) Component cooling water systems.

(16) Cooling and heating systems for the refueling (4) Reactor coolant makeup system. water storage tank.

(5) Reactor coolant and secondary sampling (17) Equipment and controls for establishing and systems. maintaining subatmospheric pressures in subatmos-(6) Chemistry control systems for the reactor cool- pheric containments.

ant and secondary coolant systems. (18) Heat tracing and freeze protection systems. I (7) Fire protection systems, including demonstra- o. Reactor Components Handling Systems tions of proper manual and automatic operation of fire detection, alarm, and suppression systems. Include the following:

(8) Seal water systems. (1) Dynamic and static load testsII of cranes, hoists, and associated lifting and rigging equipment II (9) Vent and drain systems for contaminated or (e.g., slings and strongbacks used during refueling or potentially contaminated systems and areas and drain the preparation for refueling). Static testing at 125%

and pumping systems serving essential areas, e.g., of rated load and full operational testing at 100% of spaces housing diesel generators, essential electrical rated load.

equipment, and essential pumps.

(2) Demonstration of the operability of protective (10) Purification and cleanup systems for the reac- devices and interlocks.

tor coolant system.'

(3) Demonstration of the operability of safety de-( I) Compressed gas systems" supplying pneuma- vices on equipment.

tic equipment, components, or instrumentation that are required to function to support the normal opera-tion of the facility or are essential for the operation of 2. Initial Fuel Loading and Precritical Tests standby safety. equipment or engineered safety Licensees should conduct the initial fuel loading features. cautiously to preclude inadvertent criticality. To load (12) Boron recovery system. on this basis requires that specific safety measures be established and followed such as (a) ensuring that all (13) Communication systems. Tests should in- applicable technical specification requirements and clude demonstrations of the proper operation of other prerequisites have been satisfied, (b) establish.

evacuation and other alarms, the public address sys- ing requirements for continuous monitoring of the tem within the plant, systems that may be used if the neutron flux throughout the core loading so that all plant is required to be shut down from outside the changes in the multiplication factor are observed, (c) control room, and communication systems required establishing requirements for periodic data-taking, by the facility emergency plan. and (d) independently verifying that the fuel and con-trol components have been properly installed.

(14) Heating, cooling, and ventilation systems serving the following: Predictions of core reactivity should be prepared in (a) Spaces lousing engineered safety features. advance to aid in evaluating the measured responses to specified loading increments. Comparative data of (b) Primary containment. neutron detector responses from previous loadings of (c) Battery rooms. essentially identical core designs may be used in lieu of these predictions. Licensees should establish 13Regulatory Guide 1.10. "Preopentionat Testing of Instrument criteria and requirements for actions to be taken if the Air Systems." pmzvides detauled juidance on testing of instru. measured results deviate from expected values. Shut-ment air systems. down margin verifications should be performed at 1.68-13

appropriate loading intervals (BWR), including full c. Final functional testing of the reactor protection core shutdown margin tests. It should be established system to demonstrate proper trip points, logic, and that the required shutdown margin exists, without operability of scram breakers and valves. Demon-achieving criticality. strate operability of manual scram functions.

To provide further assurance of safe loading, licen- d. Final test of the reactor coolant system to verify sees should establish requirements for the operability that system leak rates are within specified limits.

of plant systems and components, including reactivity e. Measurements of the water quality 13 and boron control systems and other systems and components concentration (PWR) of the reactor coolant system.

necessary to ensure the safety of plant personnel and the public in the event of errors or malfunctions. The f. Reactor coolant system flow tests to establish initial core loading should be directly supervised by a that vibration levels are acceptable, that differential Senior Licensed Operator having no other concurrent pressures across the fully loaded core and major duties and the loading operation should be conducted components in the reactor coolant system are in ac-in strict accordance with detailed approved proce- cordance with design values, and that piping reac-dures. Typical prerequisites, precautions, and details tions to transient conditions (e.g., pump starting and that should be included in the initial fuel loading and stopping) and flows are as predicted for all allowable precritical check procedures are described in Appen- combinations of pump operation. Loss of flow tests dix C to this guide. should be conducted to measure flow coastdown.

(Differential pressure measurements across the fully After the core is fully loaded, sufficient tests and loaded core and major components need not be re-checks should be performed to ensure that the facility peated for plants using calculation models and de-is in a final state of readiness to achieve initial criti- signs identical to prototype plants.)

cality and to perform low-power tests. The list below g. Final calibration of source-range neutron flux is illustrative of the types of tests and verifications measuring instrumentation. Verification of proper opera-that should be conducted during or following initial tion of associated alarms and protective. functions of fuel loading: source- and intermediate-range monitors.

a. Shutdown margin verification for partially h. Mechanical and electrical tests of incore (BWR) and fully loaded core. monitors, including traversing incore monitors, if installed.

b. Testing of the control rod withdrawal and insert speeds and sequencers, control rod position indica- 3. Initial Criticality tion, protective interlocks, control functions, alarms, Licensees should conduct the initial approach to and scram timing (and friction tests for BWRs) of criticality in a deliberate and orderly manner using I control rods after the core is fully loaded. Scram time the same rod withdrawal sequences and patterns that tests should be sufficient to provide reasonable assur- will be used during subsequent startups. Neutron flux ance that the control rods will scram within the re- levels should be continuously monitored and periodi-quired time under plant conditions that bound those cally evaluated. A neutron count rate at least th count under which the control rods might be required to per second should register on the startup channels be-function to achieve plant shutdown. To the extent fore the startup begins, and the signal-to-noise ratio practical, testing should demonstrate control rod should be known to be greater th-n two. All systenis scram times at both hot zero power and cold temper- required for startup or protection of the plant, includ- I ature conditions, with flow and no-flow conditions in ing the reactor protection system and emergency the reactor coolant system as required to bound con- shutdown system, should be operable and in a state of ditions under which scram might be required. For readiness. The control rod or poison removal se-each test condition, those control rods whose scram quence should be accomplished using detailed proce-times fall outside the two-sigma limit of the scram dures approved by personnel or groups designated by time data for all control rods should be retested a suf- the licensee. For reactors that will achieve initial crit-ficient number of times (;t 3 times) to reasonably en- icality by boron dilution, control rods should be sure proper performance during subsequent plant op- withdrawn before dilution begins. The control rod in-erations. For facilities using more than one type of sertion limits defined in the technical specifications control element or control rod drive design, scram should be observed and complied with.

times should be compared with identical designs (e.g., two control rods attached to a single drive Criticality predictions for boron concentration mechanism.) (PWR) and control rod positions should be provided, and criteria and actions to be taken should be estab-Additionally, the proper ope'ration of decelerat-ing devices used to prevent mechanical damage to the "Design features of BWRs to maintain water quality are dis-control rods should be demonstrated during this cussed in Regulatory Guide 1.56, "'Maintenance of Water Pur.

testing. ity in Boiling Water Reactors."

1.68-14

lished if actual plant conditions deviate from pre- g. Determination of proper response of process dicted values. The reactivity addition sequence and effluent radiation monitors. To the extent practi-should be prescribed, and the procedure should re- cal, responses of installed process and effluent radia.

quire a cautious approach in achieving criticality to tion monitors should be verified by laboratory prevent passing through criticality on a period shorter analyses of samples-from the process and/or effluent than approximately 30 seconds (<I decade per systems.

minute).

h. Chemical and radiochemistry tests and meas-urements to demonstrate design capability of chemi-

4. Low-Power Testing cal control systems and installed analysis and alarm Following initial criticality, licensees should con- systems to maintain water quality within limits in the duct appropriate low-power tests (normally at less reactor coolant and. secondary coolant systems.

than 5% power) to (a) confirm the design and, to the i. Demonstration of the operability of control rod extent practical, validate the analytical models and withdrawal and insertion sequencers and control rod verify the correctness or conservatism of assumptions withdrawal inhibit or block functions over the reactoT used in the safety analyses for the facility and (b) power-level range during which such features must be confirm the operability of plant systems and design operable.

features that could not be completely tested during the preoperational test phase because of the lack of an j. Demonstration of the capability of primary con-adequate heat source for the reactor coolant system tainment ventilation system to maintain the contain-and main steam system. ment environment and imponant components in the containment within design limits with the reactor The listing below is illustrative of the tests that coolant system at rated temperature and with the should be conducted if they have not been previously minimum availability of ventilation system compo-completed during preoperational hot functional test- nents for which the system is designed to operate.

ing. Tests that are specific to one type of light-water reactor are noted by the symbols PWR for pressurized k. Demonstration of the operability of steam-water reactors and BWR for boiling water reactors. driven engineered safety features and steam-driven plant auxiliaries and power conversion equipment.

a. Determination of boron and moderator temper-ature reactivity coefficients over the temperature and 1. Demonstration of the operability, including boron concentration ranges in which the reactor may stroke times, of main steam line and branch steam initially be taken critical. (PWR) line valves and bypass valves used for protective iso-lation functions at rated temperature and pressure

b. Measurements of control rod and control rod conditions.

bank reactivity worths to (1) ensure that they are in accordance with design predictions and (2) confirm m. Demonstration of the operability of main steam by analysis that the rod insertion limits will be line isolation valve leakage control system.

adequate to ensure a shutdown margin consistent with (BWR--during hot standby conditions.)

accident analysis assumptions throughout core life, n. Demonstration of the operability of control with the greatest -North control rod stuck out of the room computer system.

core. (PWR) o. Control rod scram time testing at rated temper-

c. Pseudo-rod-ejection test to verify calculational ature in the reactor coolant system, if not previously models and accident analysis assumptions. (PWR) conducted.

d. Determination that adequate overlap of source- p. Demonstration of the operability of pressurizer and intermediate-range neutron instrumentation exists and main steam system relief valves at rated and verification that proper operations of associated temperature.

protective functions and alarms provide for plant pro- q. Demonstration of the operability of residual or tection in the low-power range (if not previously decay heat removal systems, including atmospheric Iperformed).

e. Determination of flux distribution for compari-steam dump valves (PWR) and turbine bypass valves.

r. Demonstration of the operability of reactor son with distribution assumptions or predictions to coolant system purification and cleanup systems.

provide a check for potential errors in the loading or s. Vibration measurements of reactor vessel inter-enrichment of fuel elements or lumped poison ele- nals " and reactor coolant system components, if not ments and to check for mispositioned or uncoupled previously conducted.

corrol rods. Measurements may be performed at a higher power level depending on the sensitivity of in- "Regulatory Guide 1.20. "Comprehensive Vibration Assessment core flux instrumentation. Program for Rteactor Internals During Preoperational and Initial Startup Testing." should be used as guidance for these meas-

f. Neutron and gamma radiation surveys. uirements.

1.68-15

t. Performance of natural circulation tests of the The following list is illustrative of the types of per-reactor coolant system to confirm that the design heat formance demonstrations, measurements, and tests removal capability exists or to verify that flow (with- that should be included in the power-ascension test out pumps) or temperature data are comparable to phase. Parenthetical numbers following the items prototype designs for which equivalent tests have listed below indicate the approximate power levels been successfully completed. (PWR) for conducting the tests. If no number follows the

u. Demonstration of the operability of major or listed item, the test should be performed at the lowest practical power level. Tests that are specific to one principal plant control systems, as appropriate.

type of light-water reactor are noted by the symbols PWR for pressurized water reactors and BWR for

5. Power-Ascension Tests boiling water reactors.

Licensees should complete low-power tests, as de- a. Determine that power reactivity coefficients scribed in the FSAR, and evaluate and approve the (PWR) or power vs. flow characteristics (BWR) are low-power test results prior to beginning power- in accordance with design values. (25%, 50%, 75%,

ascension tests. Power-ascension tests should demon- 100%)

strate that the facility operates in accordance with de-sign both during normal steady-state conditions and; to the extent practical, during and following antici-pated transients. To validate the analytical models

b. Determine that steady-state core performance is in accordance with design. Sufficient measurements and evaluations should be conducted to establish that

• I used for predicting plant responses to anticipated flux distributions, local surface heat flux, linear heat transients and postulated accidents, these tests should rate, departure from nucleate boiling ratio (DNBR),

establish that measured responses are in accordance radial and axial power peaking factors, maximum with predicted responses. The predicted responses average planar linear heat generation rate should be developed using real or expected values of (MAPLHGR), minimum critical power ratio items such as beginning-of-life core reactivity coeffi- (MCPR), quadrant power tilt, and other important cients, flow rates, pressures, temperatures, pump parameters gre in accordance with design values coastdown characteristics, and response times of throughout the permissible range of power-to-flow equipment and the actual status of the plant and not conditions. (25%, 50%, 75%, 100%)

those values or plant conditions assumed for conser- c. Demonstrate that core limits will not be ex-vative evaluations of postulated accidents. ceeded during or following exchange of control rod Tests and .acceptance criteria that demonstrate the patterns that will be permitted during operation (the ability of major or principal plant control systems to demonstration test should be conducted at the highest automatically control process variables within design power level at which control rod pattern exchanges limits should be prescribed. This should provide as- will be allowed during plant operation). (BWR) surance that the integrated dynamic response of the d. Demonstrate the capabilities of plant features facility is in accordance with design.for plant events such as part-length control rods and of procedures for such as reactor scram, turbine trip, reactor coolant controlling core xenon transients. Acceptance criteria pump trip, and loss of feedwater heaters or pumps. for the test should account for expected changes in Testing should be sufficiently comprehensive to es- core performance throughout core life. (75%-85%)

tablish that the facility can operate in all operating (PWR) Results of xenon oscillation tests performed at modes for which the facility has been designed to op- plants of essentially identical design can be used to erate; however, tests should not be conducted or substitute for or supplement this testing.

operating modes or plant configurations established if they have not been analyzed or if they fall outside the e. Pseudo-rod-ejection test to validate the rod ejec-range of assumptions used in analyzing postulated tion accident analysis. (Greater than 10% power with accidents in the FSAR for the facility. control rod banks at the full power rod insertion limit) (PWR) This test need not be repeated for Appropriate consideration should be given to test- facilities using calculational models and designs iden-ing at the extremes of possible operating modes for tical to prototype facilities.

facility systems. Testing under simulated conditions of maximum and minimum equipment availability f. Demonstrate that core thermal and nuclear within systems should be accomplished if the facility parameters are in accordance with predictions with a is intended to be operated in these modes, e.g., test- single high worth rod fully inserted and during and ing with different reactor coolant pump .configura- following return of the rod to its bank position. (50%)

tions, single loop reactor coolant system operation, (PWR) operation with the minimum allowable number of g. Demonstrate that control rod sequencers, con-pumps, heat exchangers, or control valves in the trol rod worth minimizers, and rod withdrawal block feedwater, condensate, circulating, and other cooling functions operate in accordance with design, if not water systems. previously demonstrated. (25%)

1.68-16

h. Check rod scram times from data recorded dur- r. Verify by review and evaluation of printouts ing scrams that occur during the startup test phase to and/or cathode ray tube (CRT) displays that the con-determine that the scram times remain within allowa- trol room or process computer is receiving correct in-ble limits. puts from process variables, and validate that per-

i. Demonstrate capability and/or sensitivity, as formance calculations performed by the computer are ap- correct. (25%, 50%,'75%, 100%)

propriate for the facility design of incore and excore neutron flux instrumentation, to detect a control rod s. Calibrate, as necessary, and verify the perform.

misalignment equal to or less than the technical spec- ance of major or principal plant control systems, in-cluding T-average controller, automatic reactor con-ification limits. (50%, 100%) (PWR)

j. Verify that plant performance is as expected for rod runback and partial scram.

trol system; boron addition systems (PWR); inte-grated control system; pressurizer control system; I

reactor coolant flow control system; main, auxiliary,

k. Demonstrate that ECCS high-pressure coolant and emergency feedwater control systems; hotwell injection systems can start under simulated accident level control systems; steam pressure control sys-conditions and inject into the reactor coolant system tems; and reactor coolant makeup and letdown con-as designed. (At a power level in the 25%-50% range trol systems. (25%, 50%, 75%, 100%)

for BWRs with steam-driven pumps and for BWRs t. If not previously accomplished, verify, as ap-with electric-driven pumps, if not previously con- propriate, the operability, response times, relieving ducted.) (BWR) (For PWRs, the testing should be in capacities, setpoints, and reset pressures for pres-accordance with.Regulatory Guide 1.79.7) surizer relief valves; main steam line relief valves;

1. Demonstrate design capability of all systems atmospheric steam dump valves; turbine bypass val-and components provided to remove residual or decay ves; and turbine stop, intercept, and control valves.

heat from the reactor coolant system, including tur- (25%) (During transient tests, verify operability, set-bine bypass system, atmospheric steam dump valves, points, and reset pressures of relief valves.)

residual heat removal (RHR) system in steam con- u. Verify operability and response times of main densing mode, reactor core isolation cooling (RCIC) steam line isolation and branch steam line isolation system, and auxiliary feedwater system. Testing of valves. For PWRs, justification for conducting this the auxiliary feedwater system should include provi- test at low power and/or a description of design qual-sions that will provide reasonable assurance that ex- ification tests for valves of the same size and design cessive flow instabilities (e.g., water hammer) will may be submitted. (25%)

not occur during subsequent normal system startup and operation. (Prior to exceeding 25% power) v. Verify that the main steam system and feedwa-ter systems operate in accordance with design per-

m. Demonstrate that the reactor coolant system formance requirements. (25%, 50%, 75%, 100%)

operates in accordance with design. Sufficient meas-urements and evaluations should be conducted with w. Demonstrate adequate beginning-of-life per-the plant at steady-state conditions to establish that formance margins for shielding and penetration cool-flow rates, reverse flows through idle loops or jet ing systems to provide assurance that they will be pumps, core flow, differential pressures across the capable of maintaining temperatures of cooled com-core and major components in the reactor coolant sys- ponents within design limits with the minimum de-tem, vibration levels of reactor coolant system com- sign capability of cooling system components avail-ponents, and other importa'nt parameters are in able. (100%)

agreement with design values, if not previously demonstrated.

• x. Demonstrate adequate beginning-of-life per-formance margins for auxiliary systems required to l

n. Obtain baseline data for reactor coolant system loose parts monitoring system, if not previously done.

support the operation of engineered safety features or to maintain the environmeni in spaces that house en-gineered safety features to provide assurance that the I

engineered safety features will be capable of perform-

o. Calibrate instrumentation and demonstrate the ing their design functions over the range of design proper response of reactor coolant leak detection capability of operable components in these auxiliary systems, if not previously demonstrated. systems. (50%, 100%)

p. Conduct vibration monitoring of reactor inter-

y. Calibrate, as required, and verify the proper nals during steady-state anc transient operation to estab- operation of important instrumentation systems, in-lish that design limits are not exceeded (see Regula- cluding reactor coolant system flow; core flow, level, tory Guide 1.20'), if this testir. has not been pre- and temperature; incore and excore neutron flux; and viously completed. instruments and systems used to calculate thermal

q. Verify the proper operation of failed fuel detec- power level (heat balance) of the reactor. (25%,

tion systems. (25%, 100%) 50%, 75%, 100%)

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z. Demonstrate that process and effluent radiation initiating the pump trip or control valve closure monitoring systems are responding correctly by per-forming independent laboratory or other analyses.

should result in the fastest credible coastdown in flow for the system. (100%) I a.a. Demonstrate that chemical and radiochemical j.j. Demonstrate that the dynamic response of the control systems function in accordance with design, plant is in accordance with design for a condition of and sample to establish that reactor coolant system loss of turbine-generator coincident with loss of all and secondary coolant system limits are not ex- sources of offsite power (i.e., station blackout). (In ceeded. (25%, 50%, 75%, 100%) the 10-to-20% power range) b.b. Conduct neutron and gamma radiation sur- k.k. Demonstrate that the dynamic response of the veys to establish the. adequacy of shielding and to plant is in accordance with design for the loss of or identify high radiation zones as defined in 10 CFR bypassing of the feedwater heater(s) from a credible Part 20, "Standards for Protection Against Radia- single failure or operator error that would result in tion." (50%, 100%) the most severe case of feedwater temperature reduc-tion. (50%, 90%)

c.c. Demonstrate that gaseous and liquid radioac-tive waste processing, storage, and release systems 1.I. Demonstrate that the dynamic response of the plant operate in accordance with design. is in accordance with design requirements for turbine d.d. Demonstrate the capability to shut down and trip. This test may be combined with item n.n. below maintain the reactor in a hot standby condition from if a turbine trip is initiated directly by all remote-manual openings or automatic trips of the generator outside the control room, using the minimum shift crew, as well as the potential capability for placing main breaker, i.e., a direct electrical signal, not a the reactor in a cold shutdown condition." (Greater secondary effect such as a turbine overspeed. (100%)

than or equal to 10% generator load) m.m. Demonstrate that the dynamic response of the plant is in accordance with design for the case of e.e. Demonstrate that primary containment inerting automatic closure of all main steam line isolation Iand purge systems operate in accordance with design, if not previously demonstrated.

f.f. Demonstrate or verify that important ventila-valves. For PWRs, justification for conducting the test at a lower power level, while still demonstrating proper plant response to this transient, may be sub-tion and air-conditioning systems, including those for mitted for NRC staff review. (100%)

the primary containment and steam line tunnel, con-tinue to maintain their service areas within the design n.n. Demonstrate that the dynamic response of the limits. (50%, 100%) plant is in accordance with design for the case of full load rejection. The method used for opening of the g.g. If appropriate for the facility design, conduct generator main breakers (by simulating an automatic tests to determine operability of equipment provided or manual trip) should be selected 'such that the for anticipated transient without scram (ATWS). if turbine-generator will be subjected to the maximum not previously done. (25%) credible overspeed condition. The test should be ini-tiated with the plant's electrical distribution system h.h. Demonstrate that the dynamic response of the aligned for normal full power operation. (100%)

plant to the design load swings for the facility, in-cluding step and ramp changes, is in accordance with o.o. Verify by observations and measurements, as design. (25%, 50%, 75%, 100%) appropriate, that piping and component movements, vibrations, and expansions are acceptable for Li. Demonstrate that the dynamic response of the (1) ASME Code Class 1, 2, and 3 systems, (2) other plant is in accordance with design for limiting reactor I coolant pump trips and/or closure of reactor coolant system flow control valves (BWR). The method for high-energy piping systems inside Seismic Category I structures, (3) high-energy portions of systems whose failure could reduce the functioning of any Seismic Category I plant feature to an unacceptable level, and "SRegulatory Guide 1.68.2, "Initial Startup Test Program to (4) Seismic Category I portions of moderate-energy Demonstrate Remote Shutdown Capability for Watcr-Coolcd piping systems located outside containment. Tests Nuclear Power Plants," should be used as guidance for demon- performed earlier in the test program need not be stration of this capability. repeated.

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APPENDIX B INSPECTION BY THE OFFICE OF INSPECTION AND ENFORCEMENT The NRC Office of Inspection and Enforcement to determine whether -the applicant has established a conducts a series of inspections of the initial test pro- set of administratie procedures that will ensure that gram beginning before preoperational testing and the programs are carried out in accordance with the continuing throughout startup. These inspections are methods described in the FSAR.

intended to determine, on a selective basis, whether the applicant's test programs, as described in the 2. An examination of selected test procedures to FSAR, are adequately implemented and whether the re- ascertain whether the tests are designed to satisfy the sults of the tests demonstrate that the plant, proce. test objectives, whether test procedures contain ap-dures, and personnel are ready for safe operation. propriate acceptance criteria, and vhether the proce-The inspection effort focuses on the manner in which dures require the documentation of suff.cient infor-the applicant has fulfilled his commitments for ensur- mation to permit adequate evaluation of the results of the test. Also, an examination, on a selective basis, ing that adequate programs have been developed and that changes to approved test procedures have been carried out, as exemplified by the methods he has reviewed and authorized.

used for establishing procedures and the results that the methods have produced. 3. An examination of the fuel loading and startup For the NRC to implement this inspection pro- procedures to ascertain whether prerequisites, pre-gram, the applicant should have copies of the test scribed operations, and limitations are appropriately procedures available for examination by the NRC re- included to control the operation and whether the ap-plicant has implemented administrative controls iden-gional personnel approximately 60 days prior to the scheduled performance of the preoperational tests, tified in item I above. I and, not less than 60 days prior to the scheduled fuel 4. Confirmation that the applicant has evaluated loading date, copies of procedures for fuel loading, the results of the testing and has concluded that the initial startup testing, and supporting activities. results are satisfactory and meet the acceptance Drafts of these procedures should be made available criteria or has initiated corrective action.

as early as practical. Examination by NRC personnel does not constitute approval of the procedures. The 5. Confirmation that the applicant has reviewed possession of such procedures by NRC personnel the results of the fuel loading and initial operations.

should not impede the revision, review, and refine- 6. An independent examination of the results of m'ent of the procedures by the applicant. selected tests important to safety. This examination is intended primarily as an independent, selective audit The inspections by NRC personnel generally in- to determine whether information is being appro-clude the following: priately documented and evaluated by the applicant and whether the applicant's technical conclusions are

1. An examination of methods being used for pre- valid.

paring, reviewing, and approving procedures; for controlling the performance of tests; for recording, 7. Witnessing parts of preoperational, fuel load-evaluating, reviewing, approving, and retaining test ing, and startup tests to determine whether they are data and results; and for identifying and correcting being conducted in the manner described in the appli-deficiencies noted in systems and procedures. For the cant's administrative and test procedures and whether most part, this examination will be carried out prior they are being performed in a technically competent to the start of the formal test program and is intended manner.

1.68-19

APPENDIX C PREPARATION OF PROCEDURES This appendix provides guidance regarding prep- Flow and pressure characteristics aration and content of procedures for preoperational Lubrication tests, fuel loading and precritical tests, startup-to. Acceleration and coastdown critical low-power tests, and power-ascension tests.

(c) Motors and Generators Direction of rotation

1. Preoperational Test Procedures Vibration

a. Prerequisites Thermal overload protection, margins between setpoints, and full load run-Each test of the operation of a system normally re- ning amps quires.that certain other activities be performed first, Lubrication e.g., completion of construction, construction and/or preliminaz" tests, inspections, and certain other Megger or hi-pot tests preoperational tests or operations. The preoperational Supply voltage testing procedures should include, as appropriate, Phase-to-phase checks specific prerequisites. The following are typical pre- Neutral current requisites: Acceleration under load Temperature rise (I) Confirmation that construction activities as-sociated with the system have been completed and Phase currents documented. Field inspections should have been Load acceptance capability versus both made to ensure that the equipment is ready for opera- time and load (generators) tion, including inspection for proper fabrication and (d) Piping and Vessels cleanness: checkout of wiring continuity and electri- Hydrostatic test cal protective devices; adjustment of settings on Leaktightness torque-limiting devices and calibration of instru- Cleaning, flushing, and iayup 2 ments; verification that all instrument loops are oper- Clearance of obstructions able and respond within required response times: and Support adjustments adjustment and settings of temperature controllers Proper gasketing and limit switches.

Bolt torque (2) Confirmation that test equipment is operable Insulation and properly calibrated. Filling and venting (3) Tests of individual components or subsystems (e) Electrical and Instrumentation and Control to demonstrate that they meet their functional re- Verification that sensing lines are clear quirements. Typical items to consider for common for process sensors and that instru-types of equipment are: ment root valves are open (a) Valves' Voltage Leakage Frequency Opening and closing times Current Valve stroke Circuit breaker operation Position indication Power source identification Torque- and travel-limiting settings Bus transfers Operability against pressure Trip settings (b) Pumps I Operation of interlocks, prohibits, and Direction of rotation permissives Vibration Operation of logic systems Motor load versus time Calibration Seal or gland leakage Control transformer settines Seal cooling Temperature effects Range checks Response times ISection Xl of the ASME Boiler and Pressure Vessel Code pro- 'Regulatory Guide 1.37. "Quality Assurance Requirements for vides requirements for inservicc testing of pumps and valves in Cleaning or Fluid Systems and Associated Components of nuclear powet plants. The applicant should examine these re- Water-Cooled Nucleru Power Plants." should be used as guid-quirements for applicability to its preoperational test programs. ance.

1.68-20

b. Test Objectives interference with the proper testing of the as-built system.

Objectives of the test should be stated. Many sys-tems tests will be intended to demonstrate that each I. Documentation of Test Results of several initiation events will produce one or more expected responses. Thcse initiating events and the Records should identify each observer and/or data corresponding responses should be identified. recorder participating in the test, the type of observa-tion, the identifying numbers of test or measuring

c. Special Precautions equipment, the results, the acceptability, and the ac-tion taken to correct any deficiencies. Administrative Special precautions needed for safety of personnel procedures should specify the retention period of test or equipment or needed to ensure a reliable test result summaries and should require permanent reten-should be highlighted and clearly described in the test procedure. tion of documented summaries and evaluations.

d. System Initial Conditions 2. Fuel Loading Where appropriate, -instructions should be given This section provides guidance on typical informa-pertaining to the system configuration, the compo- tion to be included in the detailed fuel loading nents that should or should not be operating, and procedure.

other pertinent conditions that might affect the opera- a. Prerequisites for Fuel Loading tion of this system. (1) The composition, duties, and emergency pro-

e. Environmental Conditions cedure responsibilities of the fuel handling crew should be specified.

Most tests will be run at ambient conditions; how-ever, procedures should inclUde provisions to test the (2) Radiation monitors, nuclear instrumentation.

equipment under environmental conditions as close as manual initiation, and other devices to actuate build-practical to those the equipment will experience in ing evacuation alarm and ventilation control should both normal and accident situations. have been tested and verified to be operable.

(3) The status of all systems required for fuel load-

f. Acceptance Criteria ing should be specified.

The criteria against which the success or failure of (4) Inspections of fuel, control rods, and poison the test will be judged should be clearly identified curtains should have been made.

and should account for measurement errors and un-certainties. In some cases, these will be qualitative (5) Nuclear instruments should be calibrated, op-criteria. In other cases, quantitative values with ap- erable, and properly located (source-fuel-detector propriate tolerances should be designated as accept- geometry). One operating channel should have audi-ance criteria. ble indication or annunciation in the control room.

g. Data Collection (6) A response check of nuclear instruments to a neutron source should be required within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> The test procedures should prescribe the data to be prior to loading (or resumption of loading, if delayed collected and the form in which the data are to be for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or more).

recorded. All entries should be permanent entries.

The administrative controls should include an accept- (7) The status of containment should be specified able method for correcting an entry. and established.

(8) The status of the reactor vessel should be spec-

h. Detailed Procedures ified. Components should be either in place or out of Detailed step-by-step procedures should be pro- the vessel, as specified, to make it ready to receive vided for each test. To the extent practical, the test fuel.

procedures should use approved normal plant operat- (9) The vessel water level should be established ing procedures. and the minimum level prescribed for fuel loading Each procedure should require necessary nonstan- and unloading.

dard arrangements to be restored to their normal (10) Coolant circulation for borated reactors status after the test is completed. Control measures should be specified and established. Precautions such such as jumper logs and checkoff lists should be speci- as valve and pump lockouts should be taken to pre-fied. Nonstandard bypasses, valve configurations, vent deboration.

and instrument settings should be identified and high-lighted for return to normal. Nonstandard arrange-. (11) The emergency boron addition system (or ments should be carefully examined to ensure that other negative reactivity insertion system) should be temporary arrangements do not invalidate the test by operable.

1.68-21

(12) Fuel handling equipment should be checked nonstandard rod patterns or with operational inter-and dry runs performed. locks bypassed.

(13) The status of protection systems, interlocks, (9) Determination of the boron concentration in mode switch, alarms, and radiation protection borated reactors and frequency of determination. The equipment 'should be prescribed and verified. For frequency of determination should be commensurate reactors that have operable control rods during fuel with the worst possible dilution capability, as deter-loading, the high-flux trip points should be set for a mined by consideration of piping systems that attach relatively low-power level (normally not greater than to the reactor coolant system.

1% of full power).

(10) Actions, especially those pertaining to flux (14) Water quality should be established and limits monitoring, for periods when fuel loading is identified. interrupted.

(15) Fuel loading boron concentration should be (11) The maintaining of continuous voice com-established and verified. munication between control room and loading station.

b. Procedure Details (12) Minimum crew required to load fuel. The The procedure should include instructions or in- presence of at least two persons at any location where formation for the following areas: fuel handling is taking place should be required. A Senior Reactor Operator with no other concurrent (1) The loading sequence and pattern for fuel, con- duties should be in charge.

trol rods, poison curtains, and other components. It should also provide guidance on fuel addition incre- (13) Crew werk time. If personnel are scheduled ments and should, in general, require constituting the for consecutive daily duty, they should not normally core so that the reactivity worth of added individual be expected to work more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> out of each fuel elements becomes less as thc core is assembled. 24.

(14) Approvals required for changing the (2) The maintaining of a display for indicating the status of the core and fuel pool. Maintaining appro- procedure.

priate records of core loading. c. Limitations and Actions (3) Proper seating and orientation of fuel and (1) Criteria for stopping fuel loading should be es-components. A visual chcck of each assembly in each tablished. Some circumstances that might warrant core position should be specified. this are unexpected subcritical multiplication be-(4) Functional testing of each control rod im- havior, loss of communications between control room mediately following fuel loading. (BWR) and fuel loading station, inoperable source-range de-tector, and inoperability of the emergency boration (5) Nuclear instrumentation and neutron source system.

requirements for monitoring subcritical multiplica-tion, including source or detector relocation and nor- (2) Criteria for emergency boron injection should I malization of count rate after relocation. (Normally a be established.

minimum of three source-range monitors on a BWR (3) Criteria for containment evacuation should be and two on a PWR should be operable whenever established.

operations are performed that could affect core reactivity.) (4) Action to be followed in the event of fuel damage should be outlined.

(6) Flux monitoring, including counting times and frequencies and rules for plotting inverse multiplica- (5) Actions to be followed or approvals to be ob-tion and interpreting plots. The counting period for tained before routine loading may resume after one of count rates should be specified. An inverse multipli- the above limitations has been reached or invoked cation plot should be maintained. should be listed.

(7) The expected subcritical multiplication behavior. 3. Initial Criticality Procedures (8) Determination of adherence to the minimum This section provides some specific guidance for shutdown margin and rod worth tests in unborated the detailed procedure for operations associated with reactors and the frequency of determination. The bringing the reactor critical for the first tirn';. The minimum shutdown margin should be proved period- guidance provided in Section 1, "Preoperational Test ically during loading and at the completion of load- Procedures," of this appendix is also considered ap-ing. Shutdown margin verifications should not plicable. This procedure should include steps to en-involve a planned approach to criticality using sure that the startup will proceed in a deliberate and 1.68-22

c,,Jeriy manner, that the changes in reactivity will be tions necessary for conducting tests. The indjvidual continuously monitoTed, and that inverse multiplica- procedures should highlight these special cQnditions tion plots will be maintained and interpreted. A criti- and specifically provl1 for restoration to normal fol-cal rod position (boron concentration) should be pre- 16#/ing the test. The overall or governing power-dicted so that any anomalies may be noted and ascension test plan should typically require the fol-e'vaated All systems needed for startup should be lowing operations to be performed at appropriate aligned and in proper operation. The emergency liq- steps in the power-ascension test phase:

uid poison system should be. operable and in readi-ness. 'iechnical specification requirements must be a. Conduct any tests that are scheduled at the test met. condition or power plateau.

Nuclear instruments should be calibrated. A neut b. Examine the radial flux for symmetry, and ver-ron count rate (of at least 1/2 count per second) should register oni startup channels before the startup begins, ify that the axial flux is within expected values.

I and the signal-to-noise ratio should be known to be c. Determine reactor power by heat balance, I greater than two. A conservative startup rate limit (no shorter thanappioximately a 30-second period) should be established. High-flux scram trips should calibrate nuclear instruments accordingly, and deter-mine that adequate instrumentation overlap between the intermediate- and power-range detectors exists. I be set at their lowest value (approximately 5%-20%).

d. Just prior to ascending to the next level,-reset high-flux trips to a value no greater than 20% beyond

4. Low-Power and Power-Ascension Procedures the power of the next level unless technical specifica-1"his section provides guidance for the planning tion limits are more restrictive.

and preparation of procedures for conducting the ini-tial i.seension to rated power. The guidance provided e. Perform general surveys of plant systems and in Section 1, "Preoperational Test Procedures," of equipment to determine that they are operating within this appendix is also considered applicable. The pro- expected values.

grazi should be planned to increase power in discrete f. Check for unexpected radioactivity in process steps. Major testing should be performed at approxi- systems and effluents.

mately 25%. 50%, 75%, and 100% power levels.

lI tests intended to verify that movements and ex- g. Perform reactor coolant leak checks.

pansion of equipment are in accordance with design

h. Review the completed testing program at each are not conducted during hot functional tests and plateau, perform preliminary evaluations, including must be delayed until generation of nuclear heat, the extrapolation of minimum DNBR and maximum first power level for conducting such tests should be linear heat rate values to the high-flux trip setpoint as low as practical (approximately 5%). for the next power level, and obtain the required Individual test procedures should includc instruc- management approvals before ascending to the next tions and precautions for establishing special condi- power level or test condition.

1.68-23