Text

Proposed Tech Specs, Sections 3.0,4.0 & 6.0 Re Implementation of TMI Lessons Learned Category a Action Items
	ML19339A561
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	10/31/1980
From:	; PORTLAND GENERAL ELECTRIC CO.
To:	;
Shared Package
ML19339A559	List: ML19339A558; ML19339A560; ML19339A561;
References
	RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 8011040290
	Download: ML19339A561 (16)
	v • d • e

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION Page 3/4.2 POWER DISTRIBUTION LIMITS' 3/4.2.1 Axi al Fl ux Di f fe rence.............................

3/4 2-1 3/4.2.2 Heat Fl ux Hot Channel Factor......................

3/4 2-5 3/4.2.3 RCS Flowrate and F...............................

3/4 2-8 R

3/4.2.4 Quad rant Powe r Til t Ra ti o.........................

3/4 2-10 3/4.2.5 D NB P a ram e te rs....................................

3/4 2-12 3.4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYTEM INSTRUMENTATION................

3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION.................................

3/4 3-14 3/4.3.3 MONITORING INSTRUMENTATION Radiation Moni toring Instrumen'.ation..............

3/4 3-33 Movable Incore Detectors..........................

3/4 3-37 Seismic Instrumentation...........................

3/4 3-38 Meteorol ogi cal Instrumentation....................

3/4 3-41 Remote Shutdown Instrumentation...................

3/4 3-4'4 i

Chl ori ne Detection Sys tems........................

3/4 3-47 Fi.e Detection Instrumentation....................

3/4 3-48 De c o u pl e Swi tc h e s.................................

3/4 3-50 Accident Moni toring Instrumentation...............

3/4 3-51 l

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS No rma l 0 p e ra t i o n..................................

3/4 4-1

&O11040 W TROJAN-UNIT 1 Iy

l INDEX LIMITING CONDITIONS FGR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION Page 3/4.4.2 SAFETY VALVES - SHUTD0WN........................

3/4 4-3 3/4.4.3 SAFETY AND RELIEF VALVES - 03ERATING Safety Valves...................................

3/4 4-4 Rel i e f V a l v e s...................................

3/4 4-4a 3/4.4.4 PRESSURIZER.....................................

3/4 4-5 3/4.4.5 STEAM GENERATORS................................

3/4 4-6 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.......................

3/4 4-12 Operational Leakage.............................

3/4 4-14 3/4.4.7 CHEMISTRY.......................................

3/4 4-16 3/4.4.8 SPECIFIC ACTIVITY...............................

3/4 4-19 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System..........................

3/4 4-23

. P re s su r i z e r....................................

3/4 4-28 3/4.4.10 STRUCTURAL INTEGRITY Qual i ty Grca p 1 Compo ne nt s......................

3/4 4-29 Quality Group 2 Components......................

3/4 4-38 Quality Groups 3A and 3B Components.............

3/4 4-47 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS....................................

3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - Tayg 2.350 F..................

3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tavg < 350 F..................

3/4 5-7 3/4.5.4 BORON INJECTION SYSTEM Boron Injection Tank............................

3/4 5-8 Heat Tracing....................................

3/4 5-9 3/4.5.5 REFUELING WATER STORAGE TAMK....................

3/4 5-10 TROJAN-UNIT 1 V

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INSTRUMENTATION ACCIDENT MONITORING IflSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.9 The accident monitoring instrumentation channels shown in Table 3.3-11 shall be OPERABLE.

APPLICABILITY: f0 DES 1, 2, and 3.

ACTION:

a.

With the number of channels of operable accident monitoring instrumentation channels less than the MINIMUM NUMBER OF CHANNELS shown in Table 3.3-11, either restore the inoperable channel (s,)

to coerable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or:

1.

Establish an alternate method of monitoring the appropriate parameters, and 2.

Submit a Special Report in accordance with Specification 6.9.2 a) by telephone within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months b) confirmed by telegraph, mailgram or facsimile transmission no later than the first working day following the event, and c) in writing within 14 days following the event outlining the action taken, the cause of the inoperab'.lity and the plans and schedule for restoring the system to OPERABLE status.

b.

The provisions of Specification 3.0.4 are not applicable.

SURVEILL Att'.

RE0VIREMENTS 4.3.3.9 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-7.

TROJAN UNIT 1 3/4 3-51

TABLE 3.3-11 ACCIDEflT MONITORIfiG INSTRUMENTATION T TAL NO.

MINIMUM NUMBER INSTRUMEllT OF CHANtiELS OF CHANNELS 1.

Reactor Coolant System Subcooling (2)

(1)

Margin Monitor

2.

PORY Position Indicator 1/ valve 1/ valve

- 3.

PORV 31ock Valve Position Indicator 1/vaive 1/ valve 4.

Safety ~.'alve Position Indicator 1/ valve 1/ valve (Acoustical Flow)

flot applicable if the associated block valve is in the closed position.

- Not apolicable if the block valve is verified in the closed position end power removed.

TROJAN-VitIT I 3/4 3-52

TABLE 4.3-7 ACCIDEliT MONITORIf1G IfiSTRUMEllTATION SURVEILLANCE REQUIRENENTS Yi:

CilAftitEL SE Cl!ANNEL CilAflNEL FUtiCTI0fiAL 5

IflSTRUMENT CifECK CALIBRATION TEST

~

1.

Reactor Coolant System Subcooling M

R flA Margin Monitor 2.

PORV Position Indicator flA NA R

d.

PORV Glock Valve Position Indicator

!!A flA R

4.

Safety Valve Position Indicator NA flA R

(Acoustical Flew)

W

~-

REACTOR COOLANT SYSTEM 3/4.4.3 SAFETY AND RELIEF VALVES - OPERATING

~

SAFETY VAL \\ES LIMITING CONDITION FOR OPERATION 3.4.3.1 All pressurizer code safety valves shall be OPERABLE with a l

lift setting of 2485 PSIG + 1%.

APPLICABILITY: MODES 1, 2 and 3.

1 ACTION:

With a pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.3.1 Each pressurizer code safety valve shall be demonstrated '0PERABLE l with a lift setting of 2485 PSIG + 1%, in accordance with Section XI of the ASME Boiler and Pressure Vessel Code,1974 Edition.

l-1

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(

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l 1

TROJAN-UNIT 1 3/4 4-4

REACTOR COOLANT SYSTEM 3/4.4.3 SAFETY AND RELIEF VALVES - OPERATING RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.3.2 Two power operated relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a.

With one or more PORV(s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV(s) to OPERABLE status or close the associated block valve (s) ard remove power from block valve (s), or close the PORV(s) and remova control power frca the PORV(s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With one or more block valve (s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the block valve (s) to OPERABLE status or close the block valve (s) and remove power from the block valve (s), or close the PORV(s) and remove control power from the PORV(s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and ir COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c.

The provisions of 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.3.2.1 Each 'PORV shall be demonstrated OPERABLE:

a.

At least once per 31 days, the control circuit to each valve shall be demonstrated to have circuit continuity.

b.

At least once per 13 months by performance of a test to verify that each valve opens at the proper setpoint.

4.4.3.2.2 Each bicck valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full t ra vel.

TROJAN-UNIT 1 3/4 4-4a

REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with at least 150 kw of pressurizer heaters and a water level above that i.ecessary for heater operation but less than or equal to 1795 cu ft (92 percent indicated).

APPLICABILITY: MODES 1, 2, and 3 ACTION:

a.

With the pressurizer inoperable due to an.noperable emergency power supply to the pressurizer heaters either restore the inoperable emergency power supply within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least F0T STAtlDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWil withi i the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the reactor trip breakers L;en within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.4.1 The pressurizer water volume shall be determined to be within its limit ct least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.4.2 The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by transferring power from the normal to the emergency power supply and energizing the heaters.

TROJAN-UNIT 1 3/4 4-5

b I

c M

ELECTRICAL POWER SY',TEMS

^

1 SURVEILLANCE REQUIREMENTS (Cortinued) 3.

Verifying that samples of diesel fuel from the day tanks l

.and the tuel storage tanks are within the acceptable limits specified in. Table 1 of ASTM 0975-68 when checked for viscosity, water and sediment, 4.

Verifying the fuel transfer pump can be c:arted and transfers fuel from the storage system to the day tank, 5.

Verifying the diesels start from ambient condition, I

6.

Verifying the generator is synchronized, loaded to > 1682 kw, and operates for > 60 minutes with both diesel engines operating, and 7.

Verifying the diesel generator set is aligned to provide standby power to the associated emergency busses.

b.

At least once per 18 months during shutdown by:

1.

Subjecting the diesels to an inspection in accordance with -

)

. procedures prepared in conjunction with its manufacturer's -

recommendations for this class of standby service, i

2.

Verifying the generator capability to reject a load of j

> 828 kw without tripping, 3.

Simulating a loss of offsite power with and without the j

presence of a safety injection signal, and:

a)

Verifying de-energization of the emergency busses and load shedding from the emergency busses.

b)

Verifying the diesels start from ambient condition on the auto-start signal, energize the emergency busses with permanently connected loads, energize the auto-connected emergency loads through the applicable load sequencer and operate for > 5 minutes while the ~

4 generator is loaded with the emergency loads.

t c)

Verifying that all diesel generator. trips, except l

engine overspeed, generator g ise overcurrent, j

generator neutral overcurrent or generator i

loss of field, are automatically bypassed upon loss 'of voltage on the emergency bus and/or safety i

injection actuation signal.

1-i t

i TROJAN-UNIT 1-3/4 8-3

~

. INSTRUMENTATION BASES 3/4.3.3.6 CHLORINE DETECTION SYSTEMS

~

The OPERABILITY of the chlorine detection system ensures that sufficient capability is available to promptly detect and initiate protective action in the event of an accidental chlorine release. This capability is required to protect control room personnel and is consistent with the recommendations of Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room p

Operators Against an Accidental Chlorine Release," February 1975.

]

3/4.3.3.7 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in their early s tages. Prompt detection of fires will reduce the potential for damage to safety related equipment and is an integral element in the overall facility fire protection program.

In the event that a portion of the fire detection instrumentation is inoperable, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumen-tation is restored to CPERABILITY.

3/4.3.3.8 DECOUPLE SWITCHES l

OPERABILITY of the decouple switches in the cable spreading room (CSR) ensures that the control cables passing through the CSR to certain equipment required for safe shutdown of the Plant will be isolated and local operation of the equipment can be achieved.

In the event that a portion of the decouple switches becomes inoperable, a fire watch will be established in the CSR until the inoperable equipment is restored to OPERABILITY.

j 3/4.3.3.9 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident.

This capabil-ity is consistent with the reccmmendation of HUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations".

TROJ AN-UNIT 1 B 3/4 3-3 n.

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3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with all reactor coolant loops in operation, and maintain DNBR above 1.73 during all normal operations and anticipated transients. With one reactor coolant loop not in operation, THERMAL POWER is restricted to < 38 percent of RATED THERMAL POWER until the Overtenperature AT trip is reset. Either action ensures that the DNBR will be maintained above 1.73.

A loss of flow in two loops will cause a reactor trip if operating above P-7 (10 percent of RATED THERMAL POWER) while a loss of flow in one loop will cause a reactor trip if operating abcve P-8 (35 percent of RATED THERMAL POWER).

A single reactor coolant loop provides sufficient heat removal capability for removing care decay heat while in HOT STANDBY:

however, single failure considerations require placing a RHR loop into operation in the shutdown cooling mode if component repairs and/or corrective actions cannot be made within the allowable out-of-service time.

3/4.4.2 SAFETY VALVES 3/4.4.3 SAFETY AND RELIEF VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.

Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam at 110%

of the valve s setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.

In the event that no safety valves are.0PERABLE, an operating RHP loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig.

The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip setpoint is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power-operated relief valves or steam dump valves.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code,1974 Edition.

TROJAN-UNIT 1 B 3/4 4-1 i

REACTOR COOLANT SYSTEM BASES The power operated relief valves (PCRVs) operate to re. f eve RCS pres-sure below the setting of the pressurizer code safety valves.

These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoparable. '

3/4.4.4 PRESSURIZER The requirement that 150 kw of pressurizer heaters and their associ-ated controls be capable of being supplied electrical power from an.q 4 emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation at HOT STANDBY.

3/4.4.5 STEAM GENERATORS One OPERABLE steam generator provides sufficient heat removal capa-bility to remove decay heat after a reactor shutdown.

The requirement for two OPERABLE steam generators, combined with other requirements of the Limiting Conditions for Operation ensures adequate decay heat removal capabilities for RCS temperatures greater than 350*F if one steam gen-erator becomes inoperable due to single failure considerations. Below 350 F, decay heat is removed by the RHR system.

The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained.

The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice inspection of steam generator tubing is s sential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degra-dation due to design, manufacturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those parameter limits found to result in negligible corrosion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these parameter limits, localized corrosion may likely result in stress corrosion cracking.

The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant I TROJAN-UNIT 1 B 3/4 4-2

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a b TABLE ~6.2-1 MINIMUM SHIFT CREW COMPOSITION # LICENSE APPLICABLE MODES CATEGORY 1,-2, 3 & 4 5&6 SOL 1 1* ' OL 2 1 'Non-Licensed 3 1 Shift Technical .1 None Required Advisor

Does not include the licensed Senior Reactor Operator or Senior t

Reactor Operator Limited to Fuel Handling supervising-l CORE ALTERATIONS after the initial fuel loading. i

Shif t crew composition may be less than the minimum requirements for a period 'of time not to exceed 2 hours in order to accommodate unexpected absence of on duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1.

f-1 i J r TROJAH-UNIT 1 6-4 l

4. ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF QUALIFICATIONS 4 6.3.1 Each member of the facility staff shall meet or exceed the t minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the Radiation Protection Supervisor who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975 and (2) the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific r. engineering discipline with specific j training in Plant design, in' - response and analysis of the Plant for transients and accidents; i 6.4 TRAINING 4 6.4.1 A retraining and replacement. training program for the facility staff shall be maintained under the direction of the General Manager and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55. ~ o.4.2 A training program for the Fire Brigade shall be maintained under the direction of the General Manager and shall meet or exceed the requirements of S3ction 27 of the NFPA Code-1975, except for Fire Brigade training sessions which shall be held at least quarterly. 6.5 REVIEW AND AUDIT 6.5.1 PLANT REVIEW BOARD (PRB) FUNCTION 6.5.1.1 The Plant Review Board shall function to advise the General Manager on all matters related to nuclear safety. I . TROJAN-UNIT 1 6-5 I n+ , se - e ~ r - w -m ,w-e v

c ADMINISTRATI D CONTROLS additional narrative material to provide complete explanation of the circumicances surrounding the event. n. Reactor protection system or _ engineered safety-feature instru-ment settings which are found to be less conservative than those established by the technical specifications but which i do not prevent the fulfillment of the functional requirements of affected systems, b. Conditions leading to operation in a degraded mode permitted 4 by a limiting condition for operation or plant shutdown required by a limiting ccndition for operation. c. Observed inadequacies in the implementation of admin-trative or procedural controls which threaten to cause reduct en of degree of redundancy provided in reactor protection s stems or engineered safety feature systems. i i d. Abnormal degradation of systems other than those specitled in 6.9.1.8.c :above designed to contain radioactive material resulting from the fission process. SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Director of the Office of Inspection und Enforcement Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirenents of the applicable reference specification: a. Inoperable seismic Monitoring Instrumentation, Speci-fication 3.3.3.3. 1 b. Inoperable Meteorological Monitoring Instrumentation, Specification 3.3.3.4. c. Inservice Inspection Program Reviews, Specifications 4.4.10.1 and 4.4.10.2. d. ECCS Actuation, Specifications 3.5.2 and 3.5.3. e. Sealed Source Leakage in excess of limits, Specifica-tion 4.7.7.1.3. f. Seismic Event Analysis, Specification 4.3.3.3.2. g. Fire Detection Instrumentation, Specification 3.3.3.7. h. Fire Suppression Systems, Specifications 3.7.8.1 and 3.7.8.2.

i. - Accident Monitoring Instrumentation, Specification 3.3.3.9.

j. -Control Building Modification Connection Bolts, Specifications 3.7.11 1

'and 4.7.11.1. i TROJ AN-UNIT ' 1 6-18 i _ -}}

ML19339A561

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