Diablo Canyon Unit 1 Operational Readiness

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August 10, 1983Enclosure

DIABLO CANYON UNIT 1

OPERATIONAL READINESS

8308i5026i 8308i0PDR ADOCK 05000275

PDR

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CONTENTS

1. Design Change Review

2. Test and Inspection

3. Containment Integrity

4. Security Return-to-Service

5. Emergency Preparedness

6. Plant Staff

7. Licensed Operators

ATTACHMENTS

A. Excerpts from Unit 1 Technical

B. Surveillance Tests Required By

C. Surveillance Tests Required By

D. Surveillance Tests Required By

E. Return-to-Servi ce Inspections

F. Plant Staff Organization

Specifications

Fuel Load

Entry Into Mode 4 (Above 200 F)

Entry Into Mode 3 (Above 350 F)

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DIABLO CANYON UNIT IOPERATIONAL READINESS

The following discussion provides information on seven important areas relatedto the operational readiness of the Diablo Canyon Power Plant (DCPP). The

actions in these areas, which have been carefully analyzed and reviewed, willassure the operational readiness of the DCPP for fuel load and low-powertesting.

l. Desi n Chan e Review

All design modifications made since the Facility Operating License was issuedhave been reviewed by the engineering and operating departments to determineif the modification affected a system that is important to safety or safetyrelated. The Plant Staff Review Committee (PSRC) reviewed all safety relatedor important to safety design changes to determine if an unreviewed safetyquestion existed in accordance with 10 CFR 50.59. Prior to the implementationof a change in design or procedures, review and approval by the plant staffand plant manager was obtained. During the review of proposed design changes,all changes were examined to determine if procedural changes, training, orpost-installation testing was required. The design change was then issued forconstruction with appropriate instructions regarding implementation inaccordance with applicable codes, standards, and procedures.

In addition to the'plant staff review, the construction startup organizationreviewed all design chang'es to determine the impact on system function and toidentify test requirements. Following implementation of the design changes,the operating staff reviewed the design change for (1) any test requirementsthat had not been previously identified, (2) to ensure that the properdocumentation had been accomplished, and (3) to ensure that any affectedprocedures had been revised. The overall result of this review processassures that operations and maintenance personnel are aware of the conditionof the plant and that all modified systems have been adequately tested.

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2. Test and Ins ection

A comprehensive test and inspection program covering prefuel load, fuel load,cold system operation, heatup, and hot system operation will be undertaken

during the startup of Diablo Canyon. The objectives of the test and

inspection program are to: (1) assure the functionality and operability ofplant systems and components, (2) verify the design bases and predictionsfrom the FSAR, and (3) ensure the physical integrity and operationalfunctionality of systems, structures, and components following construction.All procedures prepared in support of startup activities have been thoroughlyreviewed by the plant operating staff prior to review and approval by the PSRC.

Surveillance tests on plant systems and components required for normal plantoperation have been performed on schedule since receiving the FacilityOperating License in September 1981. Any systems or components that have been

modified will also be tested and maintained as part of this surveillancetesting program to assure operability prior to their actual need to supportplant operations.

A comprehensive explanation of the surveillance testing program that will be

current before heatup is contained in Attachment A, "Technical SpecificationExcerpts." Attachment B provides a listing of surveillance tests to be

verified current or to be performed prior to fuel load. Attachment C is a

listing of surveillance tests to be verified current or to be performed beforetrode 4 (heatup above 200 F). Attachment D is a listing of surveillancetests to be verified current or to be performed before Mode 3 (heatup above

350 F).

A complete inspection of piping system modifications in systems operating atgreater than ambient temperatures will be performed during plant heatup and

hot operations as final verification of correct behavior. These final pipingmovement verifications will be performed during post-core-load hot system

tests. All post-core-load subcritical activities, which includeinstrumentation systems tests, piping inspection, and system and equipment

operability verification, are to be performed during post-core-load hot system

tests. The interim post-LOCA sampling system will also be tested during thehot systems tests.

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Refueling boron concentration will be established and maintained from fuelload through the completion of hot system testing; therefore, there is no

realistic possibility of attaining criticality. For these tests, the plantwill be subcritical. Since the plant has never been critical, no source ofradioactive materials is present to be released and, therefore, there is no

danger to the health and safety of the public during this test program.

The operating characteristics and the operability of plant systems importantto maintaining a high level of safety have been verified by the performance ofthree previous hot functional tests. The first hot functional was performed

in 1976, the second in 1977, and the third in 1979. Together, these testsverified that plant performance was as predicted and confirmed that systems

functioned properly at temperature.

Those systems that have had modifications since these hot functional testswere conducted have been verified to be functionally correct through design

review, tests, and inspections. Tests which are equivalent to a hotfunctional test will be performed on all plant systems during the normal

course of startup, heatup and hot system operation.

The combination of previously performed hot functional tests and testsperformed during heatup and hot system operation, together with the design

reviews, tests, and inspections which have been or will be performed on thosesystems modified since the last hot functional test, support PGandE's

conclusion that these is no need to repeat hot functional tests prior to fuelload.

3. Containment Inte rit

To ensure that the containment pressure boundary was not jeopardized duringthe modifications performed, any design changes that could directly orindirectly affect the containment pressure boundary were reviewed under the

auspices of the PSRC prior to the implementation of the change. Ho design

change involving a penetration of the containment liner or a breach of thecontainment has been proposed.

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The procedures used to (1) perform welding on the containment liner,(2) inspect weld repair, and (3) document the results were approved by the

PSRC. To maintain the correct thickness of the containment liner, any

potential or actual thinning of the liner was repaired and inspected inaccordance with Section XI of the ASME Boiler and Pressure Vessel Code.

Prior to fuel load, those penetrations providing access from the containment

atmosphere to the outside atmosphere, which have provision for local testing,will be leak tested and demonstrated to be operational. These penetrationsinclude the airlock doors, equipment door, containment ventilation valves, and

the external hydrogen recombiner isolation valves. Prior to plant heatup

(Mode 4), all containment penetration isolation valve leak tests, airlock leak

tests, and sealing system leak tests will have been performed to establishcontainment integrity. The containment electrical penetrations that employ

the continuous leakage monitoring system have been pressurized and the leakage

rates are determined quarterly. The containment integrated leak-rate test and

penetration leak test required by 10 CFR 50 Appendix J were completed

satisfactorily in 1982. The next scheduled containment integrated leak ratetest will be conducted within the 40+10 month interval specified in the

Technical Specifications. The test currently is expected to be performed

during the first refueling outage.

In summary, the design review and inspection actions described above give a

high level of confidence that no containment liner leakage path is induced by

modifications made in containment. The leak tests that will be performed

before fuel load and heatup on isolation valves, equipment hatchs, air locks,and electrical penetrations will provide a high level of confidence that no

active component leakage path has been induced by modifications made incontainment. The actions described above make a full integrated leak-ratetest of the containment unnecessary. at this time.

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4. Security Return-to-Service

Amendment 4 to the DCPP Unit 1 Operating License DPR-76 granted temporary

exemption from certain portions of the requirements of 10 CFR 73.55. This

amendment also required that "Thirty days prior to fuel loading the licenseeshall implement the "Return to Service" provisions as described in Section F,.pages 4 through 6 of the request for exemption dated February 25, 1983."

The reimplementation of the security program will be initiated approximately 6

weeks prior to fuel load to assure a smooth transition to all Security Plan

commitments at least 30 days prior to fuel load. In addition, PGandE isinitiating plant turnover from construction to operations on an area basis as

construction work is completed in plant areas. This process has alreadybegun. The fuel handling building has been turned over, and an access controlprocedure is in effect. The extensive return-to-service provisions specifiedin the February 25, 1983 license amendment request are reiterated inAttachment E.

Thus, an orderly plan exists to reinstate security. Following securityreinstatement, plant systems will be checked in accordance with the test and

inspection program to provide assurance that no sabotage occurred during the

exemption.

5. Emer enc Pre aredness

A comprehensive emergency preparedness program has been ongoing, withcontinued training of personnel, upgrading of plans and procedures, and

assessment of the adequacy of emergency facilities and equipment.

PGandE's goals are to continue a high level of preparedness and to assist,interface, and coordinate with local and state governments.

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Successful full-scale emergency exercises were conducted in 1981 and 1982.

These exercises involved the response and mobilization of more than 100 State,County, and PGandE personnel. Each exercise was witnessed by 40 or more NRC

and Federal Emergency Management Agency evaluators. These exercisesdemonstrated the adequacy of plans, resources and procedures to protect publichealth and safety. PGandE has provided and will continue to provixle extensivemanpower and resources to San Luis Obispo County to assur e an integrated,response effort.

The San Luis Obispo County-Cities Nuclear Power Plant Emergency Response Plan

was approved by the Board of Supervisors in September 1982. Training ofemergency response personnel is held on a scheduled basis. Joint trainingsessions are held with County and PGandE decisionmakers, dose-assessment

personnel, field monitoring teams and public information personnel. The Stateof California, in conjunction with PGandE and the County, has developed a

special course for law enforcement, fire departments, and other emergency

personnel. County personnel have been trained to operate the Emergency

Operations Center (EOC), which is the focal point for assessment of emergency

situations and recommendations on implementation of actions to protect publichealth and safety.

PGandE installed an Early Warning Siren system consisting of 94 sirens. The

sirens were successfully tested on August 7, 1982 and will be continuallymaintained and tested annually by PGandE. The system is activated from the

San Luis Obispo County Sheriff 's office. PGandE has provided extensiveradiation monitoring and communications equipment to the County. An emergency

response plan booklet is distributed annually to residents of the plume

exposure planning zone. A newsletter containing supplemental emergency

information is distributed to the public on a periodic basis. A system ofposted signs and stickers in public areas is in place to inform the transientpopulation of expected emergency actions.

PGandE has been continually updating and improving the Emergency Response

Facilities. When Generic Letter 81-10, "Post-TMI Requirement for the

Emergency Operations Facility" and NUREG-0696, "Functional Criteria for

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Emergency Response Facilities" were issued, PGandE evaluated the installed and

proposed facilities and determined that major modifications and upgrading offacilities and equipment were required to meet the new guidelines. PGandE

designed and commenced construction of a new Emergency Operations Facility(EOF) to meet the guidelines of NUREG-0696. An upgraded Emergency Response

Facility Data System is currently being installed.

The EOF is a two story building that is scheduled for completion in December,

1983. The first floor will be used as the San Luis Obispo County Sheriff'scommunication center and the County EOC. The second floor will be used as

PGandE's EOF. An interim EOF consisting of trailers is available until thepermanent EOF is completed. This facility has been successfully tested duringtwo full scale annual exercises.

The Emergency Response Facility Data System and the Emergency Assessment and

Response System (EARS) are the two major equipment improvements in the area ofEmergency Preparedness. The Emergency Response Facility Data System is a

computer based data acquisition, recording, and display system. The SafetyParameter Display System (SPDS) is part of this system. The Emergency

Response Facility Display System will be available in the Technical SupportCenter (TSC) and EOF. The equipment is presently located in the interim EOF.

The EARS is a computerized dose calculation system, which has direct real-timeaccess to the plant radiation monitoring system, meteorological computer, and

on-line continuous pressurized ion chambers located throughout the plume

exposure planning zone. EARS output information is transmitted to terminalslocated in the control room, TSC, EOF, Corporate Incident Response Center inSan Francisco, and the State EOC in Sacramento.

The TSC has also been successfully tested during two full scale annual

exercises. The Emergency Response Facility Data System (including the SPDS),

EARS, communication equipment, and a closed circuit television system withcameras in the control room are the major equipment installed in the TSC.

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6. Plant Staff

PGandE has selected and trained the plant staff at Diablo Canyon to assure itis both sufficient and qualified to attain a high level of safety and

reliability in plant operations. The present Nuclear Plant Operations staffat Diablo Canyon is 479 employees. It is planned to increase this total tothe presently authorized level of staffing for operation of Units 1 and 2 of603 at the time Unit 2 becomes operational.

The plant staff is organized as shown in Attachment F. The organization chartshows the number of authorized positions and the number of personnel presentlyat DCPP. The authorized positions are the total required for operation ofboth units. All positions required to operate Unit 1 are filled. The PlantManager reports to the Manager of Nuclear Plant Operations who is located inthe PGandE San Francisco headquarters office. The supervisors of the major

plant support groups, including Personnel and General Services, Security, and

Technical Assistance, as well as the Plant Superintendent, report to the Plant

Manager. Four supervisors, responsible for the operations, plant engineering,maintenance, and chemisty and radiation protection functions, report to the

Plant Superintendent, and the guality Control Supervisor reports to theTechnical Assistant.

At present, all departments at the plant are adequately staffed for operationof Unit 1. Unit 1 has been functioning under the conditions of the OperatingLicense since September 1981. Technical Specification surveillance testingrequirements have been kept current. Periodic preventive maintenance and

corrective maintenance programs have been in effect to assure equipment

readiness. The operator rotating shift schedule has been maintained withestablished shift turnover and equipment check lists being followed. PGandE

intends to augment the plant staff during initial fuel load through startuptesting and ascension to full power. Individuals experienced in Pressurized

Mater Reactor (PWR) operation on comparable PWRs will be assigned to each

shift. Three of these individuals are presently on site and a fourth will be

assigned before fuel load. The Chemistry and Radiation Protection Department

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will also be augmented by the addition of approximately 25 to 30 experienced

Radiological and Chemistry Technicians to assist with startup health physics

surveys and radiochemical analysis. In addition, a contract is being

established with Battelle-Northwest Laboratories to perform neutron dose-rate

and spectral measurements within the Unit 1 containment during startup and

ascension to full power.

In summary, Diablo Canyon Unit 1 has been functioning as nearly as possibleto an operating condition since issuance of the Facility License, and the

plant staff personnel are well prepared for transition into fuel load, low

power testing, power ascension, and commerical operation modes.

7. Licensed 0 erators

An aggressive program has been established to provide well-qualified operatorsto Diablo Canyon and assure their capabilities are kept current.

Thirty-three personnel hold Senior Operator Licenses and 10 personnel holdReactor Operator Licenses at DCPP. Twenty-seven of these 43 operators are

assigned to operating shifts. The remaining 16 personnel are assigned to the

plant staff in engineering or supervisory positions. Eighteen licensedoperators are necessary for fuel load and the operation of DCPP Unit 1

assuming the minimum four shift rotation. However, 22 licensed operators arenecessary for our current five shift rotation. The 27 licensed shiftoperators and 16 licensed staff personnel ensure that an adequate number ofpersonnel are available to operate DCPP safely. An additional 20 personnel

have completed most of the license training program, including simulatortraining, and are tentatively scheduled to take the licensing examination thisfall.

An effective operator training and requalification program has been

established at DCPP. The operator training staff consists of 12 full time

training people, 5 of which hold Senior Reactor Operator Licenses for DCPP.

DCPP has adopted a five shift rotation, which requires that each shift has a

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full week of requalification training in each five week period. During the

requalification training, topics are covered pertaining to significant plantmodifications, revised operating procedures, newly installed equipment, and

operating experience events that occurred at other nuclear facilities. The

program also includes an annual simulator training session'onducted at the

Westinghouse Zion Training Center. PGandE has contracted for the design and

construction of a plant-specific training simulator that is scheduled to be

available during the summer of 1984. The Operator Licensing Branch of the NRC

inspected the requalification program at DCPP in November 1982 and June 1983;

the inspection results indicated that the training program met all NRC

requirements.

To provide the licensed operators with additional operating experience, the

majority of the licensed operators were sent to the Trojan Nuclear Power Plant

in the spring and summer of 1982. While at Trojan, the operators observed

control room operations and various plant evolutions during a major plantoutage.

These ongoing programs ensure that all licensed operators have and willcontinue to maintain a high degree of readiness to support fuel load and plantoperation.

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ATTACHMENT APage 1 of 17

EXCERPTS FROM UNIT 1 TECHNICAL SPECIFICATIONS

The following excerpts from the Diablo Canyon Technical Specificationsare intended to convey the extent and type of testing performed toassure that systems required for fuel load and transition to systemheatup and hot standby (Mode 3) are functioning properly prior to thetime they are required to perform their intended safety

function.'echnical

S ecification T.S. 3.1 - Reactivit Control

a ~ Shutdown Margin: The shutdown margin ensures that reactor can bemade subcritical from all operating conditions and reactivitytransients associated with postulated accidents are controllablewithin acceptable limits, and the reactor will be maintainedsufficiently subcritical to preclude inadvertent criticality inthe shutdown condition. This is accomplished at least once per24 hours by evaluating boron concentration, control rod position,reactor coolant temperatures, fuel burnup, and Xenon and'amariumconcentrations;

b.

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Boration Systems: These systems ensure that negative reactivitycontrol is available during each mode of facility operation.Operability of these systems is determined by:

( 1) Analysis of the borated water concentration;

(2) Confirming a sufficient volume of borated water;

(3) Confirming Operability of charging pumps;

(4) Maintaining separate and redundant flow paths of water;

(5) Confirming Operability of boric acid transfer pumps;

(6) Confirming Operability of associated heat tracing; and

(7) Confirming adequate emergency power supply from operablediesel generators.

Rod Position Indication System: This ensures that a minimumshutdown margi'n is maintained by determination of control rodposition. This requirement is met by the performance of aChannel Functional Test.

'See Glossary at end of Attachment A for explanation of terms.

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T. S. 3.3 - Instrumentation See Attachments 8, C and 0 for Com leteTest Listin

Reactor Trip System and Engineered Safety Feature ActuationSystem Instrumentation: Operability of instrumentation in thesesystems ensures:

(1) That the associated action and/or reactor trip will beinitiated when the parameter monitored by each channelreaches its setpoint;

(2) The specified coincidence logic is maintained;

(3) Sufficient redundancy is maintained to permit a channel tobe out of service for testing or maintenance; and

(4) Sufficient system functional capability is available fromdiverse parameters.

Compliance with this section is demonstrated by Actuation LogicTests, Analog Channel Operational Tests, Channel Calibrations,Channel Checks, Engineered Safety Feature Response Time Tests,Master Relay Tests, Reactor Trip Response Time Tests, Slave RelayTests, Source Checks, Trip Actuating Device Operational Tests,and Channel Functional tests.

b.

A brief description of the requirements of each of these tests isincluded in the attached Glossary of Terms.

Radiation Monitoring Instrumentation: The Operability of thisinstrumentation ensures that the radiation levels are continuallymonitored in the areas served by. individual detectors, and thatthe alarms or automatic action is initiated when the radiationlevel trip setpoint is exceeded. Operability of each instrumentis demonstrated by performance of Channel Checks, Calibrations,and Channel Functional Tests.

c ~ Seismic Monitoring Instrumentation: The Operability of theseismic instrumentation ensures that sufficient capability isavailable to promptly determine the magnitude of a seismic eventand evaluate the response of those features important to safety.This capability is required to permit comparison of'he measuredresponse to that used in the design basis for the facility todetermine if plant shutdown is required.

This system is demonstrated Operable by the performance ofChannel Checks, Channel Calibration, and Channel FunctionalTests.

d. Meteorological Ins trumentation: The Operability of themeteorological instrumentation ensures that sufficientmeteorological data is available for estimating potentialradiation doses to the public as a result of routine or

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accidental release of radioactive materials to the atmosphere.This capability is required to evaluate the need for initiatingprotective measures to protect the health and safety of thepublic.

Operability is demonstrated by the performance of Channel Checksand Channel Calibrations.

Remote Shutdown Instrumentation: The Operability of the remoteshutdown instrumentation ensures that sufficient capability isavailable to permit shutdown and maintenance of Hot Standby ofthe facility from locations outside of the control room. Thiscapability is required in the event control room habitability islost.

Operability is demonstrated by performance of Channel Checks andChannel Calibrations.

Accident Monitoring Instrumentation: The Operability of theaccident monitoring instrumentation ensures that sufficientinformation is available on selected plant parameters to monitorand assess these variables following an accident. The normalplant instrument channels specified are suitable for use aspost-accident instruments.

Operability of these instruments is demonstrated by theperformance of Channel Checks and Calibrations.

Chlorine Detection System: The Operability of the chlorinedetection system ensures that sufficient capability is availableto promptly detect and initiate protective action in the event ofan accidental chlorine release. This capability is required toprotect control room personnel.

Operability of this system is demonstrated by a Channel Check anda Channel Functional Test including the following: cleaning thesensing cells, filters and airlines; checking for proper flow andverifying that the detector responds to chlorine.

Fire Detection Instrumentation: Operability of the firedetection instrumentation ensures that adequate warningcapability is available for the prompt detection of fires. Thiscapability is required in order to detect and locate fires intheir early stages. Prompt detection of fires will reduce thepotential for damage to safety related equipment and is anintegral element in the overall facility fire protection program.

Operability is demonstrated by Channel Functional Tests.

Radioactive Effluent Monitoring Instrumentation: The radioactiveeffluent instrumentation is provided to monitor and control, asapplicable, the releases of radioactive materials in effluentsduring actual or potential releases. The alarm/trip setpointsfor these instruments shall be calculated in accordance with the

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procedures in the Offsite Dose Calculation Procedure (ODCP) toensure that the alarm/trip will occur prior to exceeding thelimits of 10 CFR 20. This instrumentation also includesprovisions foW monitoring (and controlling) the concentrations ofpotentially explosive gas mixtures in the waste gas holdupsystem.

The radioactive effluent monitoring system is maintained Operableby performance of Channel Checks, Source Checks, ChannelCalibrations, and Channel Functional Tests.

Turbine Overspeed Protection: This specification is provided toensure that the turbine overspeed protection instrumentation andthe turbine speed control valves are Operable and will protectthe turbine from excessive overspeed. Protection from turbineexcessive overspeed is required since excessive overspeed of theturbine could generate potentially damaging missiles which couldimpact and damage safety related components, equipment orstructures.

Protection is maintained Operable by Channel Checks, valvecycling and disassembly.

T. S. 3.4 - Reactor Coolant S stem RCS

a ~ Loop/Residual Heat Removal (RHR) Operation

(I) Hot Standby (Mode 3)

In Mode 3, a single reactor coolant loop provides sufficientheat removal capability for removing core decay heat;however, single failure considerations require that twoloops be Operable. H

(2) Hot Shutdown (Mode 4)

In Mode 4 with reactor coolant loops filled, a singlereactor coolant loop or RHR train provides sufficient heatremoval capability for removing decay heat; but singlefailure considerations require that at least two loops(ei ther RHR or RCS) be Operable.

(3) Cold Shutdown (Mode 5 ) Loops Not Filled

In Mode 5, with reactor coolant loops not filled, a singleRHR train provides sufficient heat removal capability forremoving decay heat; but single failure considerations andthe inavailability of the steam generator as a heat removingcomponent require that at least two RHR trains be Operable.

(4) Cold Shutdown (Mode 5 ) Loops filledIn Mode 5 with reactor coolant loops filled, a singlereactor coolant loop or RHR train provides sufficient heat

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removal capability for removing decay heat; but singlefailure considerations require that at least two loops(either RHR or RCS) be Operable.

The Operability of the RCS Loops/RHR trains are determined byverifying power is available, water in the steam generators atthe appropriate level for that particular mode, reactor coolantpump(s) or RHR train(s) in operation.

b. Safety Valves: The pressurizer code safety valves operate toprevent the RCS from being pressurized above its Safety Limit of2735 psig. Each safety valve is designed to relieve 420,000 lbsper hour of saturated steam at 110% of the valve's set point.

The Operability of the pressurizer safety valves are determinedby compliance with American Society of Mechanical Engineers(ASME) Section XI and applicable addenda.

c ~ Pressurizer: The limit on the maximum water volume in thepressurizer assures that the parameter is maintained within thenormal steady state envelope of operation. The maximum watervolume also ensures that a steam bubble is formed and, thus, theRCS is not a hydraulically solid system. The requirement that aminimum number of pressurizer heaters be Operable enhances thecapability of the plant to'ontrol reactor coolant systempressure and establish natural circulation.

The Operability of the pressurizer is determined by observinglevel and by measuring heater group power.

d. Relief Valves: The power operated relief valves and steam bubblefunction to relieve RCS pressure during all design transients upto and including the design step load decrease with steam dump.Operation of the power operated relief valves minimizes theundesirable opening of the spring-loaded pressurizer code safetyvalves.

Operability of the relief valves is demonstrated by operating thevalve through a full cycle, performing a Channel Calibration ofthe actuation instrumentation and by complying with allrequirements of ASME Section XI and the applicable addenda.

e. Leakage Oetection System and Operational Leakage: The RCSleakage detection systems required by this specification areprovided to monitor and detect leakage from the Reactor CoolantPressure Boundary.

The RCS leakage is demonstrated to be within its limits bymonitoring containment particulate or gaseous activity and sumplevels. Also by measuring the leakage and by the performance ofa water inventory balance. Additionally, RCS and ECCS pressureisolation valves are leak checked after each refueling outageduring startup or after maintenance and within 24 hours of theiractuation. Likewise, compliance with ASME Section XIrequirements apply to these 18 valves.

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Chemistry and Specific Activity: The limitations on RCS

chemistry ensure that corrosion of the RCS is minimized andreduces the potential for RCS leakage or failure due to stresscorrosion.

T. S.

The limitations on the specific activity of the primary coolantensure that the resulting two hour doses at the site boundarywill not exceed an appropriately small fraction following a steamgenerator tube rupture accident in conjunction withprimary-to-secondary steam generator leak.

3.5 - Emer enc Core Coolin S stem

a ~ Accumulators: The Operability of each RCS accumulator ensuresthat a sufficient volume of borated water will be immediatelyforced into the reactor core through each of the cold legs in theevent the RCS pressure falls below the pressure of theaccumulators.

b.

Operability of each accumulator is demonstrated by verifyingwater level, pressure on the tank, valves properly aligned anddemonstrating proper valve actuation. Additionally, the pressureand water level channels are demonstrated Operable by theperformance of Channel Calibrations and Channel Functional Tests.

Emergency Core Cooling System (ECCS) Subsystems: The Operabilityof the ECCS subsystems ensures that sufficient emergency corecooling capability will be available in the event of a LOCA.Either subsystem operating in conjunction with the accumulatorsis capable of supplying sufficient core cooling to limit the peakcladding temperatures within acceptable limits for all postulatedbreak sizes ranging from the double ended break of the largestRCS cold leg pipe downward. In addition, each ECCS subsystemprovides long term core cooling capability in the recirculationmode during the accident recovery period.

Operability is determined by testing the associated charging,safety injection and residual heat exchanger pumps, heatexchangers, 15 specific valves in the system plus the additionalvalves required to be Operable in accordance with ASME codeSection XI for this code class system. Additionally, the ECCSsystem is verified full of water, all valves are in their correctposition, and by visual inspections of the system. Additionally,the system pumps are tested pursuant to ASME code Section XIdeveloping minimum prescribed discharge pressure. Injection flowrates into the RCS is also measured and determined to be within1 imi ts.

C. Boron Injection System: The Operability of the boron injectionsystem as part of the ECCS ensures that sufficient negativereactivity is injected into the core to counteract any positiveincrease in reactivity caused by RCS system cooldown. RCScooldown can be caused by inadvertent depressurization, a LOCA ora steam line rupture. The redundant heat tracing channels

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d.

'ssociated with the boron injection system ensure that thesolubility of the boron solution will be maintained above thesolubility limit of 135'F at 21,000 ppm boron.

This system is demonstrated Operable by verifying the containedvolume of water, the boron concentration, the water temperature,the solution temperature and energizing the heat trace leads.

Refueling Water Storage Tank (RWST): The Operability of theRWST as part of the ECCS ensures that a sufficient supply ofborated water is available for injection by the ECCS in the eventof a LOCA.

The RWST is demonstrated Operable by verifying the containedvolume, boron concentration, and minimum water temperature.

T. S. 3.6 - Containment S stems

'a ~

b.

c ~

Primary Integrity: Primary containment integrity ensures thatthe release of radioactive materials from the containmentatmosphere will be restricted to those leakage paths andassociated leak rates assumed in the accident analyses.

Primary containment integrity is demonstrated by verifying propervalve lineup and verifying Operable each air lock. Also includedin the requirement by AStlE code Section XI is the valve leaktests. These leak tests are performed by Diablo Canyon PowerPlant under guidance of the V-600 test series. This series oftests includes approximately 100 valves above and beyond thosepreviously discussed.

Air Locks: The limitations on closure and leak rate for thecontainment air locks are required to meet the restrictions oncontainment integrity and containment leak rate. Surveillancetesting of the air lock seals provide assurance that the overallair lock leakage will not become excessive due to seal damageduring the intervals between air lock leakage tests.

Air locks are demonstrated Operable by measuring seal leakagewith the air space between the locks and seals pressurized andverifying that only one door in each air lock can be opened atany one time.

Internal Pressure: The limitations on containment internalpressure ensure that the containment structure is prevented fromexceeding its design negative pressure with respect to theoutside atmosphere and the containment peak pressure does notexceed the design pressure of 47 psig during LOCA conditions.

Pressure is periodically measured using instrumentationcalibrated under the instrumentation section of the TechnicalSpecifications.

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<)ATTACHMENT APage 8 of 17

d.

e.

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~ Air Temperature: The limitations on containment average airtemperature ensure that the overall containment average airtemperature does not exceed the initial temperature conditionassumed in the accident analysis for a LOCA.

Containment air temperature is determined to be within limitsusing calibrated instrumentation.

Structural Integrity: This limitation ensures that thestructural integrity, of the containment will be maintainedcomparable to the original design standards for the life of thefacility. Structural integrity is required to ensure that thecontainment will withstand the maximum pressure in the event of aLOCA. The visual examination of the concrete, liner, and theleakage test demonstrate this capability.

Ventilation: Use of the containment purge lines is restricted toone supply line and one exhaust line of the purge system or thevacuum/pressure relief line at a time to ensure that the siteboundary dose guidelines of 10 CFR 100 would not be exceeded inthe event of a LOCA during purging operations.

The Operability of these valves is tested under the V-600 seriesprocedures.

Spray Systems: The Operability of the containment spray systemensures that containment depressurization and cooling capabilitywill be available in the event of a LOCA.

h.

The system is demonstrated Operable by verifying correct valvelineup, correct automatic valve actuation and automatic pumpstarting. Valve actuation is verified in accordance with ASMEcode Section XI. The testing of the valves in the V2 and V3series of tests check stroke times, position indication andproper movement. Additionally, when a pump routine test is run,additional valve testing is included to ensure Operability.

Spray Additive System: The Operability of the spray additivesystem ensures that sufficient NaOH is added to the containmentspray in the event of a LOCA. The limits on NaOH minimum volumeand concentration ensure that the iodine removal efficiency ofthe spray water is maintained because of the increase in pHvalue, and corrosion effects on components within containment areminimized. This system is demonstrated Operable by verifying thevalves are in their correct position, quantity and concentrationof solution in the tank, correct automatic valve actuation, and afull flow test of the system through a test valve. Likewise, apump routine test is also used to determine pump operabilitywhich tests additional valves in its dedicated system.

Containment Cooling System: The Operability of the containmentfan cooler units ensures that 1) the containment air temperaturewill be maintained within limits during normal operation; and 2)adequate heat removal capacity is available when operated in

~ 4I

0ATTACHMENT APage 9 of 17

conjunction with the containment spray systems during post LOCAconditions, and 3) adequate mixing of the containment atmospherefollowing a LOCA to prevent localized accumulations of hydrogenfrom exceeding the flammable limit.

Operability of the containment fan cooler units is verified bystarting the unit, running i t for a prescribed time, verifyingsufficient cooling flow, verifying proper damper shifting whenrequired, and starting on an automatic signal.

j . Containment Isolation Valves: The Operability of the containmentisolation Valves ensures that the containment atmosphere will beisolated from the outside environment in the event of a releaseof radioactive material to the containment atmosphere orpressurization of the containment.

Each valve is demonstrated Operable by verifying that on acontainment isolation signal, all valves move to their commandedpositions. Likewise, prior to this test, each valve is testedutilizing V2, V3 and V600 series test procedures to verify thatupon issuance of a command that the valve moves to its commandedposition within its allowed time limit, and that once in itscommanded position the valve maintains its leak integrity inaccordance with T. S. 3.6 which sets the leakage limits for allvalves. Additional valves are tested using the same V2, V3series of tests on all code classified systems that have tofunction properly to bring the plant to a safe shutdown.

k. Hydrogen Analyzers/Monitors and Recombiners: The Operability ofthe equipment and systems required for the detection and controlof hydrogen gas ensures that this equipment will be available tomaintain the hydrogen concentration within containment below itsflammable limit during post-LOCA conditions. Either recombinerunit is capable of controlling the expected hydrogen generationassociated with 1) zirconium-water reactions; 2 ) radiolyticdecomposition of water; and 3) corrosion of metals withincontainment.

Operability is confirmed by use of Channel Calibrations, systemFunctional Tests, visual exams, and verification of equipmentintegrity by the performance of electrical checks.

T. S. 3.7 - Plant S stems

a. Steam Generator Safety Valves: The Operability of the main steamline code safety valves ensures that the secondary systempressure will be limited to not more than 1055 of its designpressure of 1065 psig during the most severe anticipated systemoperational transient.

These valves are demonstrated Operable by testing in accordancewith ASME code Section XI.

0


Auxiliary Feedwater System: The Operability of the auxiliaryfeedwater system ensures that the RCS can be cooled down to lessthan 350'F from normal operating conditions in the event of atotal loss of off-site power.

Each auxiliary feedwater pump is demonstrated Operable byverifying that proper discharge pressure and flow is obtainedunder test conditions in accordance with ASME code Section XI,verifying proper valve lineup, and verifying each pump starts andautomatic valves open upon receipt of a test signal.

Additional valves are tested during the performance of theroutine pump test to further verify system Operability.

Condensate Storage Tank: The Operability of the condensatestorage tank with the minimum water volume ensures thatsufficient water is available for cooldown of the RCS to lessthan 350'F in the event of a,total loss of off-site power. Theminimum water volume is sufficient to maintain the RCS at HotStandby conditions for 8 hours with steam discharge toatmosphere.

The tank is verified Operable by verifying its volume of water tobe within limits.

Activity: The limitations on secondary system specific activityensure that the resultant off-site radiation dose will be limitedto a small fraction of 10 CFR 100 limits in the event of a steamline rupture.

Activity is determined to be within limits by sampling andanalysis of the secondary coolant system.

Main Steam Isolation Valves: The Operability of the main steamline isolation valves ensures that no more than one steamgenerator will blowdown in the event of a steam line rupture.This restriction is required to 1) minimize the positivereactivity effects of the RCS cooldown associated with theblowdown, and 2 ) limit the pressure rise within containment inthe event the steam line rupture occurs within containment.

Each isolation valve is tested to ensure Operability inaccordance with ASME code Section XI.

Steam Generator Pressure/Temperature Limitati ons: The 1 imitati onon steam generator pressure and temperature ensures that thepressure induced stresses in the steam generators do not exceedthe maximum allowable fracture toughness stress limits. Thelimitations of 70'F and 200 psig are based on average steamgenerator impact values taken at 10'F and are sufficient toprevent brittle fracture.

This limitation is verified by measuring temperature and pressureof the steam generator.

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ATTACHMENT4Page ll of 17 "

Vital Component Cooling Water (CCW): The Operability of thevital CCW system ensures that sufficient cooling capacity isavailable for continued operation of safety related equipmentduring normal and accident conditions. The redundant coolingcapacity of this system, assuming a single failure, is consistentwith the assumptions used in the accident analyses.

The vital loops of the CCW system are determined Operable byverifying each valve is properly positioned and that eachautomatic valve, when commanded, moves to its commanded position.Each pump is tested in accordance with ASME code Section XI, andalong with the pump test, additional valves are tested to provesystem Operability.

Auxiliary Saltwater System: The Operability of the auxiliarysaltwater system ensures,.that sufficient cooling capacity isavailable for continued operation of safety related equipmentduring normal and accident conditions.

Operability is determined by verifying each valve is in itscorrect posi tion, routine pump tests including valve tests inaccordance with ASME Section XI.

Control Room Ventilation System: The Operability of the controlroom ventilation system ensures that the ambient air temperaturedoes not exceed the allowable temperature for continuous dutyrating for the equipment and instrumentation and the control roomwill remain habitable for operations personnel during andfollowing all credible accident conditions.

Operability of the control room ventilation system isdemonstrated by initiating flow through the filters, verifyingelectrical power is available, running each fan, verifying theability of the filter medium to perform properly, and verifyingthe ability of the system to pressurize the control room.

Auxiliary Building Safeguards Air Filtration System: TheOperability of the auxiliary building safeguards air filtrationsystem ensures that radioactive materials leaking from the ECCS

equipment within the auxiliary building following a LOCA arefiltered prior to reaching the environment.

Operability of this filtration system is verified by initiatingflow through the filters, verifying electrical power isavailable, running each fan, and verifying the ability of thefiltration medium to perform properly.

Snubbers: All snubbers are required Operable to ensure that thestructural integrity of the RCS and all other safety relatedsystems is maintained during and following a seismic or otherevent initiating dynamic loads.

Snubbers are demonstrated Operable by visual exam, manualexercise, and an acceptable manufacturer's functional test.

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n.

Sealed Source Contamination: The limitations on removablecontamination for sources requiring leak testing, including alphaemitters, is based on limits for plutonium.

Verification of these limits are made by methods which havedetection sensitivity to determine if leakage has occurred.

Fire Suppression Systems: The Operability of the firesuppression systems ensures that adequate fire suppressioncapability is available to confine and extinguish fires occurringin any portion of the facility where safety related equipment islocated. The fire suppression system consists of the watersystem, spray and/or sprinklers, COz, Halon and fire water hosestations. The collective capability of the fire suppressionsystems is adequate to minimize potential damage to safetyrelated equipment and is a major element in the facility fireprotection program.

Operability of this system is verified by routine pump tests, Y2,V3 and V600, all of which are in accordance with ASME codeSection XI. Operability is also achieved by verifyingsufficient water volume, proper valve lineup, performing routinesystem flushes, performance of flow tests, verifying properdamper response, and functional and flow tests.

Fire Barrier Penetrations: The functional integrity of the firebarrier penetrations ensures that fires will be confined oradequately retarded from spreading to adjacent portions of thefacility. This design feature minimizes the possibility of asingle fire rapidly involving several areas of the facility priorto detection and extinguishment.

Verification of fire barrier penetrations are completed utilizingvisual exams.

0 ~

T. S.

Area Temperature Monitoring: The area temperature limitationsensure that safety-related equipment will not be subjected totemperatures in excess of their environmental qualificationtemperatures. Exposure to excessive temperatures may degradeequipment and can cause loss of its Operability.

The temperatures are determined to be within limits periodically.

3.8 - Electrical Power S stems

a ~ A.C. 5 D.C. Sources: The Operability of the A.C. and D.C. powersources and associated distribution systems during operationensures that sufficient power will be available to supply thesafety related equipment required for 1) the safe shutdown of thefacility; and 2) the mitigation and control of accidentconditions within the facility.

I ~I

C ~ )


The Operability of the minimum specified A.C. and D.C. powersources and associated distribution systems during shutdown andrefueling ensures that 1) the facility can be maintained in theshutdown or refueling condition for extended time periods; and 2)sufficient instrumentation and control capability is availablefor monitoring and maintaining the facility status.

The Surveillance Requirements for demonstrating the Operabilityof the diesel generators are in accordance with therecommendations of Regulatory Guides 1.9, "Selection of DieselGenerator Set Capacity for Standby Power Supplies," March 10,1971; 1. 108, "Periodic Testing of Diesel Generator Units Used asOnsite Electric Power Systems at Nuclear Power Plants," Revision1, August 1977, where applicable; and 1.137, "Fuel Oil Systemsfor Standby Diesel Generators," Revision 1, October 1979.

The Surveillance Requirements for demonstrating the Operabilityof the batteries are based on the recommendations of RegulatoryGuide 1. 129, "Maintenance Testing and Replacement of Large LeadStorage Batteries for Nucl'ear Power Plants," February 1978, andIEEE Std 450-1980, " IEEE Recommended Practice for Maintenance,Testing, and Replacement of Large Lead Storage Batteries forGenerating Stations and Substations."

Verification of operability of these systems include thefollowing tests, but are not limited by this partial list:(1) Instrumentation and relays (T. S. 3.3)

(2) Diesel Engine Performance Testing

(3) Generator Performance Testing

(4) Battery Performance Testing

(5) Verification of electrical breaker alignment

(6) Verification of voltage on buss

(7) All routine maintenance of the above equipment to ensureOperability

(8) ASME code Section XI valve testing (V2 and V3 series)

(9) ASME code Section XI pump testing

(10) Verified volumes of fuel oil and lubricating oil on hand foremergency power sources

(11) Diesel Generator protective features

(12) Ability of Diesel Generators to pick up loads

4 ~

ATTACHMENT APage 14'f 17

( 13) Verification that loads on generators do not exceed theirsteady state rating

(14) Operability of battery chargers.

b. Electrical Equipment Protective Devices: The Operability of themotor operated valves thermal overload protection and bypassdevices ensures that these devices will not prevent safetyrelated valves from performing their function.

Operability is determined by Channel Functional Tests and ChannelCalibrations.

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ATTACHMENT APage 1'5 of 17

GLOSSARY OF TERMS

ENGINEERED SAFETY FEATURE RESPONSE TIME

The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time intervalfrom when the monitored parameter exceeds its ESF actuation setpoint at thechannel sensor until the ESF equipment is capable of performing its safetyfunction (i.e., the valves travel to their required positions, pumpdischarge pressures reach th'eir required values, etc). Times shall includediesel generator starting and sequence loading delays where applicable.

GASEOUS RADLJASTE SYSTEM

A GASEOUS RADMASTE SYSTEM shall be any system designed and installed toreduce radioactive gaseous effluents by collecting primary coolant systemoffgases from the primary system and providing for delay or holdup for thepurpose of reducing the total radioactivity prior to release to theenvironment.

IDENTIFIED LEAKAGE

IDENTIFIED LEAKAGE shall be:

a. Leakage, except CONTROLLED LEAKAGE, into closed systems, such as pumpseal or valve packing leaks that are captured and conducted to a sumpor collecting tank, or

b. Leakage into the containment atmosphere from sources that are bothspecifically located and known either not to interfere with theoperation of leakage detection systems or not to be PRESSURE BOUNDARYLEAKAGE, or

c. Reactor Coo lant System leakage through a steam generator to thesecondary system.

MASTER RELAY TEST

A MASTER RELAY TEST shall be the energization of each master relay andverification of Operability of each relay. The MASTER RELAY TEST shallinclude a continuity check of each associated slave relay.

OPERABLE - OPERABILITY

A system, subsystem, train, component or device shall be OPERABLE orhave OPERABILITY when it is capable of performing its specifiedfunction(s) and when all necessary attendant instrumentation,controls, electric power, cooling or seal water, lubrication or otherauxiliary equipment that are required for the system, subsystem,train, component or device to perform its function(s) are also capableof performing their related support function(s).

~ I


GLOSSARY OF TERMS (Cont'd)

OPERATIONAL MODE — MODE

An OPERATIONAL MODE (i.e., NODE) shall correspond to any one inclusivecombination of core reactivity condition, power level and average reactorcoolant temperature.

PHYSICS TESTS

PHYSICS TESTS shall be those tests performed to measure the fundamentalnuclear characteristics of the reactor core and related instrumentation and1) described in Chapter 14.0 of the FSAR; 2) authorized under theprovisions of 10 CFR 50.59; or 3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE

PRESSURE BOUNDARY LEAKAGE shall be leakage, except steam generator tubeleakage, through a non-isolable fault in a Reactor Coolant System componentbody, pipe wall or vessel wall.

REACTOR TRIP SYSTEM RESPONSE TIME

The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from whenthe monitored parameter exceeds its trip setpoint at the channel sensoruntil loss of stationary gripper coil voltage.

SHUTDOWN MARGIN

SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by whichthe reactor is subcritical or would be subcritical from its presentcondition assuming all full length rod cluster assemblies (shutdown andcontrol) are fully inserted except for the single rod cluster assembly ofhighest reactivity worth which is assumed to be fully withdrawn.

SLAVE RELAY TEST

A SLAVE RELAY TEST shall be the energization of each slave relay andverification of Operability of each relay. The SLAVE RELAY TEST shallinclude a continuity check, as a minimum, of associated testable actuationdevices.

SOURCE CHECK

A SOURCE CHECK shall be the qualitative assessment of channel response whenthe channel sensor is exposed to a radioactive source.

4

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GLOSSARY OF TERMS (Cont'd)

STAGGERED TEST BASIS

A STAGGERED TEST BASIS shall consist of:

a. A test schedule for n systems, subsystems, trains or other designatedcomponents obtained by dividing the specified test interval into nequal subintervals;

b. The testing of one system, subsystem, train or other designatedcomponent at the beginning of each subinterval.

TRIP ACTUATING DEVICE OPERATIONAL TEST

A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating theTrip Actuating Device and verifying Operability of alarm, interlock and/ortrip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall includeadjustment, as necessary, of the Trip Actuating Device such that itactuates at the required setpoint within the required accuracy.

UNIDENTIFIED LEAKAGE

UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGEor CONTROLLED LEAKAGE.

CHANNEL FUNCTIONAL TEST

A CHANNEL FUNCTION TEST shall be:

a. Analog channels - the injection of a simulated signal into the channelas close to the sensor as practicable to verify Operability includingalarm and/or trip functions.

b. Bistable channels - the injection of a simulated signal into thesensor to verify Operability including alarm and/or trip functions.

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~ S

TACHMENT 8age I of 9

SURVEILLANCE TESTS REQUIRED BY FUEL LOAD

STP I-lc

STP I-1D

STP M-8F

STP M-8G

STP V-2Q

STP V-2S

STP I-4A

'TP I-48

STP I-483

STP I-16A

STP I-168

STP I-34

STP I-34C

STP I-34G

STP I-34H

STP M-19A

STP M-198

STP M-39

STP M-63A

STP M»638

Routine Weekly Checks

Routine Monthly Checks

Leak Rate Test of Personnel Air Lock Seals

Leak Rate Test of Emergency Air Lock Seals

Exercise Containment Ventilation Valves Inside Containment

Exercise Chemical and Volume Control System Valves InsideContainment

Operational and Functional Test of Nuclear Source RangeInstrumentation

Calibration of Source Range Protection and Safe Guards ChannelN-31 and N-32

Determination of Source Range Detector Characteristic Curvesfor Channel N-31 and N-32

Functional Test of Solid State Protection System Logic Train A

Functional Test of Solid State Protection System Logic Train 8

Service and Functional Test Fire Detection System

Functional Test Fire Detection Supervisory Panel A and Panel 8

Check Fire Detection Panel A and 8 Voltage

Fire Detection Panel Voltage Check Panel A

Halon Storage Tank Weight and Pressure Verification

Halon Functional Test

Routine Test C02 Fire System Protection

Fire Water System Flush

Containment Fire Water Flush

0108L/0112P-1

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TACHMENT B

age 2 of 9

SURVEILLANCE TESTS RE(}UIRED BY FUEL LOAD

STP I-1881

STP I-1882

STP I-18V2

STP I-18C1

~ STP I-18C2

STP I-18D1

STP I-18D2

STP I-18F1

STP I-18F2

STP I-18Il

STP I-18I2

STP I-18K1

STP I-18M1

STP I-18M2

STP I-18P1

STP I-18P2

STP I-76A

STP I-76B

Plant Vent Radiogas Monitor RM-28 A and B and RM-14 A and B

Functional Test

Plant Vent Radioactive Gas Monitor-Calibration RM-14A andRM-14B

Plant Vent Air Particulate Monitor-Calibration RM-28A andRM-28B

Waste System Discharge Liquid Monitor Functional Test RM-18

Waste System Discharge Liquid Monitor-Calibration RM-18

Gas Decay Tank Discharge Gas Monitor Functional Test RM-22

Gas Decay Tank Discharge Gas Monitor-Calibration RM-22

Functional Test Steam Generator Blowdown Tank Vent andDischarge Monitor RM-23 and RM-27

Steam Generator Blowdown Tank Liquid and Vent GasMonitors-Calibration RM-23 and RM-27

Nuclear Fuel Storage Area Monitors Functional Test RM-5 andRM-9 UN 1 and UN 2

Spent Fuel Pool and New Fuel Storage Area Monitors-CalibrationRM-5 and RM-9

Functional Test of Plant Vent Iodine Monitor Flow Switch FIS 24

Control Room Air Intake Monitor Functional Test RM-25 andRM-26 Unit 1

Calibration of Control Room Ventilation Systems IntakeRadiation Monitor RM-25 and RM-26 Unit 1

Oily Water Separator Effluent Monitor Functional Test Rfl-3

Oily Water Separator Discharge Liquid Monitor-Calibration RM-3

Liquid Radwaste Effluent Line Flow Recorder Functional TestFR-20

Calibration of Liquid Radwaste Discharge Flow Indication Loop19-7

0108L/0112P-2

\g I P

TACHMENT B

age 3 of 9


STP I-77A

STP I-77B

STP I-78A

STP I-78B

STP I-79A

STP I-79B

STP I-85A

STP I-858

STP I-86

STP M-85A

STP M-85B

STP M-85C

STP I-35A

STP I-35B

STP I-37A

STP I-378

STP I-37C

STP I-44C

STP I-48

STP M-10B

Steam Generators Blowdown Effluent Line Flow RecorderFunctional Test FR-53

Calibration of Steam Generator Blowdown Tank Discharge FlowLoop 4-61

Oily Water Separator System Flow Totalizer FR-251 FunctionalTest

Calibration of Oi.ly Water Separator Flow

Functional Test of Waste Gas System Oxygen Analyzers Channel75 and 76

r

Calibration of Waste Gas System Oxygen Analyzers Channel 75

and 76

Plant Vent Flow Recorder FR-12 Functional Test

Calibration of Plant Vent System Flow Monitor Channel 12, Loop23-24

Calibration of Iodine Sampler Flow

Liquid Radwaste System Operability Verification

Gaseous Radwaste System Operability Verification

Solid Radwaste System Operability Verification

Functional Test of Control Room Chlorine Monitor

Inspection and Maintenance of Control Room Chlorine Monitor

Functional Test of Seismic Monitor System

Calibrate Seismic Monitoring System

Recording Triaxial Accelerometer Channel Check

Meteorological Instrumentation Calibration

Calibration of RWST Storage Tank Level Channels 920, 921, 922

Diesel Fuel Oil Analysis for Viscosity, Water and Sediment

0108L/0112P-3

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TACHMENT 8Page 4 of 9


STP M-10A

STP M-6

STP M-13A

STP M-13B

STP M-42

STP M-53

STP M-9A

STP M-9D

STP M-9E

STP M-9G

STP M-9H

STP M-9J

STP M-9L

STP M-Sl

STP M-9M

STP M-llB

STP M-11C

STP M-12A

STP M-128

STP M-12C

Diesel Fuel Oil Storage Tank Inventory

Routine Surveillance Test of Control Room Vent System

Functional Test of Offsite Power to 4KV Vital Busses

Functional Test of Engineered Safeguards Automatic TimersSetting

Load Test Manipulator Crane

Control Room Vent System HEPA and Charcoal Banks PenetrationTests

Diesel Engine Generator Routine Surveillance Test

Diesel Load Rejection Test

Diesel Trip Circuitry Bypass Verification

Diesel Generator 24-Hour Load Test

Diesel Generator Interdependence

Diesel Loss and Reload

Shutdown Relay Test of Diesel Generator

Diesel Generator Inspection

Verification of Auto Connected Loads ~ 2750KW

Measurement of Station Battery Voltage and Specific Gravity

Battery Terminal Resistance Measurement

Battery Capacity Performance Test

Battery Charger Performance Test

Station Battery Service Test

0108L/0112P-4

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. iTACHMENT 8Page 5 of 9


STP M-11A

STP M-15

STP M-16A

STP M-16M

STP M-16N

STP M-65

STP M-66

STP M-67

STP M-68

STP M-69

STP M-70

STP M-71

STP M-79

STP M-80

STP P-28

STP P-178

STP P-38

STP P-78

STP P-88

STP P-128

STP P-138

Routine Surveil 1 ance

Routine Surveillance

Test Diesel Fuel Oil Transfer Pumps

Test Fire Pumps

Measurement of Station Battery Pilot Cell Voltage and SpecificGravity

Int. Test of Engineered Safeguards and Diesel Generators

Operation of Slave Relays Resulting in Charging and/or Letdown. Isol ati on

Operation of Slave Relays K622A and 8, K625A and 8, Cont.Vent. Isolation and Phase 8 Isolation

Operation of Slave Relays K632A, K6328, K634A, K6348 TurbineAuxiliary Feed Pump

Functional Test Deluge System

Deluge System Nozzle Proof Test

Weekly Fire Hose Reel Station Inspection

Fire Hose Station Functional Test

Monthly C02 Hose Reel, Deluge Valve and Fire ExtinguisherInspector

Inspection of Fire Barrier Penetrations

Fire Water System Flow Test

Fire Hose Inspection, Gasket Replacement and Reracking

Fire Hose Hydrostatic Test

Routine Surveillance Test Centrifugal Charging Pumps

Routine Surveillance Test Reciprocating Charging Pump

Routine Surveillance Test Residual Heat Removal Pumps

Routine Surveillance Test Auxiliary Saltwater Pumps

Routine Surveillance Test Component Cooling Water Pumps

0108L/0112P-5

TACHMENT B

Page 6 of 9

SURVEILLANCE TESTS. REQUIRED BY FUEL LOAD

STP P-14B

STP P-15B

STP R-16

STP R-20

STP V-2A

STP V-2D

STP V-2E

STP V-2F

STP V-2I

STP V-2K

STP V-2V

STP V-3E1

STP Y-3E5

STP V-3E6

STP V-3F1

STP V-3F3

STP V-3F4

STP V-3F5

STP V-3H7

STP V-3H8

STP V-3H11

STP V-3H12

Exercise Valve

Exercise Valve

FCY430 and 431 CCW HX Outlet Valves

LCV69 and 70 Make Up Water to CC'W

Exercise Valve RCY16 CCW Surge Tank

Routine Surveillance Test Boric Acid Transfer Pumps

Routine Surveillance Test Make Up Water Transfer Pumps

Functional Test Boric Acid Heat Trace

Boric Acid Inventory

Exercise Auxiliary Saltwater Yalves

Exercise Auxiliary RHR Pump Recirc Yalves

Exercise Auxiliary Containment Yent Yalves

Exercise Auxiliary Component Cooling Water Valves

Exercise Charging and Letdown Valves

Exercise CVCS Boration Valves

Char ging Pump Suction from RWST

Exercise Valves FCV 110A and 8472, Boric Acid to Blender

Exercise Valves 8104 and 8445, Emergency Borate

Exercise Valve FCV128, CCP Flow Cont. Valve

Exercise Valve FCV495 ASW Pump 2 Cross Tie

Exercise Valve FCY601 ASW Cross Tie UN 1-2

Exercise Valve FCV602 CCW HX ¹1 Saltwater Inlet

Exercise Valve FCV603 CCW HX ¹2 Saltwater Inlet

Exercise Valve FCV364 and 365 RHR HX CCW Return Valves

0108L/0112P-6

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TACHMENT B

Page 7 of 9

SURVEILLANCE TESTS REQUIRED BY FUFL LOAD

STP V-3J1

STP V-3K5

STP V-3K8

STP V-3K9

STP V-3Ml

STP V-3M2

STP Y-3M3

STP V-3M4

STP V-3M7

STP V-3Q1

STP V-3Q2

STP V-3T1

STP V-11

STP I-38

STP I-39

STP I-45

STP V-657

STP V-661

STP I-74

STP V-3H4

STP V-3H5

Exercise Pressurizer Power Operated Relief Isolation Valves8000A, B, C

Exercise Valves, 8146 and 8147 Normal and Alternate Charging

Exercise Valves 8145 and 8148 Pressurizer Auxiliary Spray

Exercise Valves 8105 and 8106 CCP Recirc

Exercise Valves FCV641A and B RHR Pump Reci rc

Exercise Valves HCV 637 and 638 RHR HX Outlet

Exercise Valves HCV 670 RHR HX Bypass

Exercise Valves 8700A and B RHR Pump Suction

Exercise Valves 8716A and B RHR Train Crosstie

Exercise Valves Fire Water Loop Isolation 1 and 2

Exercise Valves Fire Water Inside Loop Isolation

Exercise Valves Phase B Containment

Routine Surveillance Phase B Containment Isolation Valves

Calibration

Calibration

of Firewater Storage Tank Level Channel 38

of Raw Water Reservoir Level Channel 36 and 37

Calibration of Boric Acid Tank Levels Channel 102 and 106

Penetration 57 and 82 Containment Isolation Valve Leak Testing

Penetration 61 and 62 Containment Isolation Valve Leak Testing

Calibration of Area Monitors

Exercise Valve FCV 356 CCW to RCP's and Vessel Support Coolers

Exercise Valve FCV 357 RCP Thermal Barrier

01 08L/01 12P-7

III

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TACHMENT B

Page 8 of 9


STP V-3H6

STP Y-3I1

STP Y-3M5

STP V-9

STP I-73

STP I-4C

STP I-18Z2

STP I-18S2

STP I-18AA1

STP I-18AA2

STP I-18AA2

STP I-27

STP I-29

STP I-34E

STP I-34F

STP I-44B

STP I-84A

STP I-84B

STP M-17

STP M-43

STP M-64

Exercise Valve FCV 363 RCP Oil Cooler/Support Cooler

Exercise Valve 8998A and B Cont. Spray Eductor Check Valves

Exercise Valve 8701 and 8702 RC Loop 4

Containment Vent Isolation Systems Operability Demonstration

Calibration of Gas Analyser 02 Monitor

Calibration of Audio Count Rate/Sealer Timer Channel

Calibration of TSC Area Rad Monitors

Calibration of CRPS Rad Monitors

Functional Test Plant Vent Iodine Monitors

Calibration of Plant Vent Iodine Monitors

Calibration of Post Accident Sample Room

Calibration of RHR Heat Exchanger Flow Channels 970A and B,971A and B

Functional Test Emergency Signals and Communications System

Functional Test Plant Elevator Smoke Detectors

Functional Test Security Building Smoke Detection System

Meteorological Instrumentation Meekly Functional Test

Functional Test CRPS CL Monitor

Calibration CRPS CL Monitor

Functional Test Emergency DC Lighting Systems andCommunications System

Fuel Handling Building Crane Verification and Interlocks

Deluge System Functional Test

0108L/0112P-8

I I ~~

TACHMENT B

age 9 of 9


STP V-2W

STP V-3E2

STP V-3E3

STP V-3G1

STP V-3K1

STP V-301

STP V-302

STP V-6

STP I-18R1

STP I-18R2

STP I-18X1

STP I-18X2

STP I-1821

STP I-18Z2

Boric Acid Tank Recirc Valves

Exercise Valves FCVllOB Blender to Charging Header

Exercise Valves FCV111A Primary H20 to Blender

Exercise Valves FCV410 Gas Decay Tank Plant Vent

Exercise Valves HCV133 RHR CVCS Letdown Crosstie

Exercise Diesel Turbo Air Receiver Check Valves

Exercise Diesel Starting Air Receiver Check Valves

Positive Verification of Penetration Test Connection Valves

Functional Test Main Steam Rad Monitors

Calibration of Main Steam Rad Monitors

Functional Test Gas Decay Tank Rad Monitors

Calibration of Gas Decay Tank Rad Monitors

Functional Test TSC Area Rad Monitors

Calibration of TSC Area Rad Monitors

0108L/0112P-9

I I

TTACHMENT C

age 1 of 6

SURVEILLANCE TESTS REQUIRED BY ENTRY INTO MODE 4 (above 200'F)

STP M-8E

STP M-54

STP M-8F

STP M-8G

STP V-2M

STP V-2N

STP V-2Q

STP V-20

STP V-272

STP I-16C

STP I-65

STP M-51

STP I-60

STP I-62

STP I-61

STP I-18A1

STP I-18A2

STP I-18T2

ST> M-1

STP M-3A

STP M-3B

STP M-3C

Airlock Door Interlock Verification

Measurement of RCP Seal Injection Flow

Personnel Air Lock Seal Leak Rate Test

Emergency Air Lock Seal Leak Rate Test

Exercise Valves Located Near the Pressurizer Relief Tank

Exercise S/G Blowdown and Excess Letdown Valve

Exercise Containment Vent Valve 1C

Exercise Containment Inside Isolation Valves

Exercise RX Head Vent Valves

Functional Test of Manual Initiation and Block of ReactorProtection

Calibration of Cont. Fan Coolers Moisture Collection" Measurement

Routine Test of Cont. Fan Coolers

Cal. Cont. Structure Sump Level Channel 60 and 61

Cal. Cont. RX Cavity Sump Level Channel 62

Cal. Cont. Structure Sump Flow Channels 40 and 41

Cont. Air Particulate and Radiogas Monitors Func. Text RM-11and 12

Cont. Air Particulate and Radiogas Monitors Calibration RM-12

Cont. Air Particulate and Radiogas Monitors Calibration RM-ll

Air Test of Containment Spray Nozzles

Auxiliary Building SFGDS Air Filtration System PenetrationTests

Auxiliary Building Vent Dampers Leak Test

Auxiliary Building Charcoal .Filter Preheater Test

0108L/0112P-10

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ATTACHMENT C

Page 2 of 6


STP M-4

STP M-7

STP M-45

STP M-56

STP M-78A

STP M-78B

STP I-33

STP M-16B

STP M-16C

STP M-16H

STP M-16I

STP M-16J

STP M-16K

STP M-16L

STP M-16P

STP M-16R

STP M-75

STP M-84A

STP M-84B

STP P-4B

STP P-23

STP V-2B

Auxiliary Building SFGDS Air Filter System RoutineSurveillance Tests

Containment Integrated Leak Rate Test

Containment Inspection

Spray Additive Tank Line-Verification of No Obstruction

Snubber Visual Inspection Program

Hydraulic Snubber Testing Program

Reactor Trip System and Engineered SFGDS Actuation System TimeResponse Tests

Operation of Slave Relays K604A, K604B (S.I)

Operation of Slave Relays K608A (SI)

Operation of Slave Relays K644A and B, K645A and B Cont. Spray

Operation of Slave Relays K606A and B, Phase A Cont. Isolation

Operation of Slave Relays K612 Train B Phase A Cont. Isolation

Operation of Slave Relays K613A and B Phase A Cont. Isolation

Operation of Slave Relays K614A and B Phase A Cont. Isolation

Continuity Testing of Equipment Slave Relay Actuating Circuitry

Slave Relay Response Time

Calibration of Vital 4KV Undervoltage Relays

Functional Test Mov Thermal Overload Bypass

Calibration of Mov Thermal Overload Bypass

Routine Surveillance Test Cont. Spray Pump

Acceleration Timing of Safety Related Pumps

Auxiliary Feedwater and Cont. Spray Valves

0108L/0112P-11

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4

ATTACHMENT C

Page 3 of 6


STPV-2G'TP

V-2H

STP V-2J

STP V-2X

STP V-2Y

STP V-3K2

STP V-3K7

STP V-3L1

STP V-3L2

STP V-3L3

STP V-3L5

STP V-3L6

STP V-3L10

STP V-3P4

STP V-3R1

STP V-3Sl

STP V-3S2

STP V-3S3

STP V-3S4

STP V-3S5

STP V-13

STP V-15

Exercise Spray Additive Tank Outlet Valves

Exercise Miscellaneous Auxiliary Building (85') Valves

Exercise Outside Cont. Isolation Valves

Exercise Manual Cont. Isolation Valves

Exercise Letdown HX Inlet and Letdown Orifice Isolation Valves

Exercise Valves 8100 and 8112 RCP Seal

Exercise Valves 8149A-C Letdown Orifice Isolation

Exercise Valves 8802A and B SI Pump Discharge Isolation to RCSHot Legs

Exercise Valves 8821A and B SI Pump Discharge Isolation to RCSCold Legs

Exercise Valves 8807A and B S. I Charging Pump Suction Crosstie

Exercise Valves 8809A and B RHR to RCS

Exercise Valves 8835 SI. Cold Leg Isolation

Exercise Valves 8923A and B S. I Pump Suction Valves

Exercise Valves FCV436 and 437 RWSR Supply to AuxiliaryFeedwater Pumps

Exercise Valves PVC 19, 20, 21, 22 100% Atmospheric Dump

Exercise Valves Phase A Containment ISO





Accumulator Isolation Valve Functional Test

Emer Core Cooling System Flow Balance Test

0108L/0112P-1 2

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ATTACHMENT C

Page 4 of 6


STP V-600

STP I-49

STP M-8A

STP M-8B

STP M-BC

STP M-80

STP I-52

STP V-619

STP V-621

STP V-623

STP V-630

STP V-635

STP V-641

STP V-645

STP V-646

STP V-649

STP V-650

STP V-651

STP V-651B

STP V-652

STP V-654

STP V-656

STP V-659

Valve and Penetration Leak Summary

Calibration of RCDT Level Channel 93

Personnel Air Lock Leak Rate Test

Leak Rate Test Electrical Penetrations

Leak Rate Test Equip Hatch, Fuel Transfer Tube Seals

Combined Type B and C Leakage Evaluation

Calibration of Containment Temperature

Penetrations 19 and 20 Cont. Isolation Valve Leak Testing

Penetrations 21 Cont. Isolation Valve Leak Testing




Penetrations 41, 42, 43 and 44 Cont. Isolation Valve LeakTesting






Penetrations 51B Cont. Isolation Valve Leak Testing

Penetrations 52A, B, D Cont. Isolation Valve Leak Testing



Penetrations 50A, B, C, 76A, B Cont. Isolation Valve LeakTesting

0108L/0112P -1 3

4

v h

~ yATTACHMENT C

Page 5 of 6

SURVEILLANCE TESTS RE(}UIRED BY ENTRY INTO MODE 4 (above 200'F)

STP V-663

STP V-668

STP V-670

STP V-671

STP V-678

STP V-679

STP V-682

STP V-683

STP V-3H3

STP V-3H10

STP V-3L13

STP V-3L14

STP V-3M5

STP I-89

STP I-90

STP I-29

STP I-50

STP M-77

STP M-83A

STP M-83B

STP M-83C

STP M-83D



Penetrations 70 Cont. Isolation Valve Leak Testingr


Penetrations 78 and 52E, F Cont. Isolation Valve Leak Testing




Exercise Valve FCV750 CCW Header C Isolation

Exercise Valve FCV749 RCP Oil CLR/Support Cooler CCW Return IC

Exercise Valve 8976 RWST to S. I Pump

Exercise Valve 8980 RWST to RHR Pump Suction

Exercise Valve 8701 and 8702 RC Loop 4

Calibration of Containment Wide Range Level

Calibration of Containment Wide Range Pressure Channel 938/939

Functional Test Emergency Signals and Communications System

Calibration of RC Drain Tank Flow Channel 43

Safety and Relief Valve Testing

Containment Penetration Conductor Protection Relay Calibration

Containment Penetration Conductor Protection Device IntegratedSystem Test

Containment Penetration Conductor Molded Case Circuit BreakerFunctional Test

Containment Penetration Conductor Cktbreaker Inspection andPreventatire Maintenance

0108L/0112P-14

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0ATTACHMENT C

Page 6 of 6


STP V-2R

STP V-3E4

STP V-3K6

STP V-3LB

STP V-3T3

Containment Fan Cooler Unit Dampers

Exercise Valves FCV111B Blend to Letdown Line

Exercise Valves 8166 and 8167 RCS Letdown

Exercise Valves 8916 Nitro Accum Fill Check

Exercise Valves Sample and Recombiner Valves

0108L/0112P-15

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ATTACHMENT D

Page I of 4


STP I-3A

STP I-38

STP I-5A

STP I-58

STP I-585

STP I-6A

STP I-68

Functional Test of Intermediate Range NIS Channel N-35 and N-36

Calibration of Intermediate Range NIS Channel N-35 and N-36

Functional Test Delta T Protection Channels

Calibration of Delta T, Overtemperature Delta T, OverpowerDelta T and T Average Channels

Response Test of Primary Loop RTD Channels

Functional Test of Pressurizer Pressure Channels

Calibration of Pressurizer Pressure Protection/SafeguardsChannels

STP I-683

STP I-684

STP I-78

Calibration of Pressurizer Pressure Transmitters

Response Time Test of Pressurizer Pressure Channels

Calibration of Pressurizer Level Protection and SafeguardsChannels

STP 'I-783

STP I-11A

STP I-1182

STP I-1183

STP I-1184

STP I-12A

STP I-128-2

STP I-128-3

STP I-128-4

Calibration of Pressurizer Level Transmitters LT-459 and LT-460

Functional Test of Steam Generator Water Level Channels

Calibration of Steam Generator Level Channels Protection andSFGDS and Post Accident Monitoring

Calibration of Steam Generator Leve] Transmitters

Response Steam Generator Water Level Transmitters

Functional Test of Steam Generator Flow and Pressure Protection

Calibration of Steam Generator Pressure Channels

Calibration of Steam Generator Pressure Protection Channels

Calibration of Steam Generator Steam Flow Channels

0108L/01 12P-1 6

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ATTACHMENT D

Page 2 of 4


STP I-12B6

STP I-1287

STP I-12B8

STP I-12810

STP I-12811

STP I-15A

STP I-15B

STP I-15B3

STP I-1584

STP I-16E

STP I-30A

STP I-30B

STP I-32A

STP I-32B

STP I-55

STP I-56

STP I-80

STP I-81

STP I-83

STP I-53

STP I-54

Calibration of Steam Generator Steam Flow and Pressure andSafeguards (and Alarms) Functions for Protection Sets 1-4

Calibration of Steam Flow Transmitters

Response of Steam Flow Transmitters

Calibration of Steam Pressure Transmitters

Response of Steam Pressure Transmitters

Functional Test Containment Pressure Channels

Calibration of Containment Pressure Channels

Calibration of Containment Transmitters

Response of Containment Transmitters

Functional Test SSPS P-4 Interlock

Functional Test Accumulator Level Channels

Calibration of Accumulator Level Channels

Functional Test of Accumulator Pressure Channels

Calibration of Accumulator Pressure Channels

Calibration of Steam Generator Wide Range Channels

Calibration of Auxiliary Feedwater Flow Channels

Calibration of Reactor Coolant System Subcooled Margin Monitor

Calibration of Reactor Coolant System Wide Range T Hot and TCold

. Calibration of Pressurizer Safety Valve Acoustic Detector

Calibration of First Stage Pressure Channels No. 505

Calibration of First Stage Pressure Channels No. 506

0108L/0112P-17

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ATTACHMENT D

Page 3 of 4


STP M-89

STP I-19

STP I-47

STP I-57

STP P-1B

STP P-5B

STP P-6B

STP V-2C

STP V-2U

STP V-3P5

STP V-3P6

STP I-18M1

STP I-18M2

STP V-3M5

STP I-34E

STP I-34F

STP I-44B

STP I-84A

Emergency Core Cooling System Venting

Charging Pump Flow Calibration

Calibration of Condensate Water Storage Tank Level Channel 40

Calibration of Emergency Borate Flow Channel No. 113

Routine Surveillance Test Safety Injection Pumps

Routine Surveillance Test Motor Driven Auxiliary FeedwaterPumps

Routine Surveillance Test Steam Driven Auxiliary FeedwaterPumps

Exercise Auxiliary Boron Injection Valves

Exercise Steam Generator Related Valves

Exercise Valves LCV 106, 107, 108 and 109 Auxiliary FeedwaterLevel Control

Exercise Valves LCV 110, ill, 113 and 115 Auxiliary FeedwaterLevel Control

Control Room Air Intake Monitors Functional Test RM-25, 26

Control Room Air Intake Monitors Calibration RM-25, 26

Exercise Valves 8701 and 8702 RC Loop 4

Functional Test Plant Elevator Smoke Detectors

Functional Test Security Building Smoke Detectors

Meteorological Instrumentation Weekly Functional Test

Functional Test CRPS CL Monitor

0108L/01 12P-1 8

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0ATTACHMENT D

Page 4 of 4


STP I-84B

STP M-43

STP M-64

STP V-2L

STP V-2P

Cal ibra tion of CRPS CL Moni tor

Fuel Handling Building Crane Verification of Interlocks

Deluge System Functional Test

Pressurizer Spray Valves

Accumulator Check Valve Leak Test Valves

0108L/0112P-19

~ P

ATTACHMENT EPage 1 of 2

RETURN-TO-SERVICE INSPECZIONS

Return-to-service inspections will be conducted on vital systems when requiredby the Technical Specifications. In particular, surveillance testing andpost~intenance testing of plant systems will be performed in accordance withthe requirements of the Technical Specifications, fuel load and low power testprogram and administrative procedures associated with maintenance andmodifications.

'Ihe measures employed will include the following:

1. Post~intenance functional testing will be performed on all active vital-components which have been modified or have undergone maintenance duringthe period affected by this change.

2. All areas in which vital equipment is located will be subjected tocaaprehensive visual inspection for evidence of sabotage or evidence ofplacement of contraband materials.

3. Work will be performed inassurance/control inspectionknowledgeable personnel willprogress. PGardE inspectionincreased above normal levels.

accordance with PGandE's qualityprogram. In addition, technically

patrol vital areas to observe work inand contractor supervision have been

In addition to the testing performed in (1) above, the TechnicalSpeciQ.cations (Appendix A to the Operating License) require veriQ.cationof Operability of all vital plant systems. Verification of Operabilitywill be performed prior to entry into the Operational Mode, as deQned bythe Technical Specifications, in which the system or component is requiredto be Operable. 'Ihe testing performed to verify Operability includesvital instrument channel functi.onal tests, pump and valve testingperformed in accordance with the ASME Boiler and Pressure Code,Section XI, fire protection system tests, containment isolation valve leaktests, containment airlock leakage tests, containment hatch seal l~etests, verification of proper vital system chemistry, electrical equipmentfunctional and integrated system testing, diesel engine generatorfunctional and integrated systems tests, and operating pressure leak testsof appropriate vital systems. 'Ihis is an extensive and comprehensivetesting program involving over 370 tests.

5. As required by plant operating and surveillance procedures, properalignment of all fluid system valves and instrument valves on thesesystems willbe verified prior to declaring the system Operable.

0~ f J

ATTACHMENT EPage 2 of 2

6. 'Lhe security perimeter intrusion system will normally be maintainedoperable with routine surveillance testing performed. However, if thesyst'm or portions of it are reanved from service, testing will beperformed on those portions removed from service prior to reimplementationof securi.ty. When areas are returned to vital status, testing andcheckout of the access control cardreaders and associated supervisioncircuitry will be performed.

7. Major components and piping systems will be monitored for proper 'thermalexpansion and freedom from interference during heatup.

0027B

t

PLANT STAFF ORGANIZATION

Plant Manager

(10/10)

Personnel andGeneral Services Plant Superintendent Security Technical Assistant

(66/52) (16/15) (33/29)

Quality Control

(24/19)

OperationsPlant Engineering and

I & C MaintenanceElectrical and Mechanical

MaintenanceChemistry and Radiation

Protection

(160/128) (85/65) (149/99) (60/39)

n

Authorized/Actual Staff Number of authorized positions indicated is total necessaryfor operation of both units. All positions required for operationof Unit 1 are filled.

I

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i{

Documents

Diablo Canyon Unit 1 Operational Readiness