A07433, Proposed Tech Specs Re Leakage Detection Sys

Proposed Tech Specs Re Leakage Detection Sys

ML20247R150
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	05/30/1989
From:	CONNECTICUT YANKEE ATOMIC POWER CO.
To:
Shared Package
ML20247R134	List: A07433, Proposed Tech Specs Re Leakage Detection Sys ML20247R130
References
A07433, A7433, NUDOCS 8906070142
Download: ML20247R150 (9)

Text

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l Docket No. 50-213 A07433 Attachment 2 Haddam Neck Plant Proposed Changes to Technical Specifications STS Format Pages May 1989 8906070142 890530 PDR ADOCK 05000213 P PDC

REACTjgCOOLANTSYSTEM 3.3.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE _RETEC1 ION SYSU RS LIMITING CONDITION FOR OPERATION __ _ ,

3.3.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:

a. The Containment Atmosphere Gaseous Radioactivity Monitoring System, and

b. The Volume Control Tank Level (Narrow Range) Monitoring System and the Containment Main Sump Level Monitoring System.

APPLICABILITY: MODES 1, 2, 3, and 4.

5 ACTION:

a. With the Containment Atmosphere Gaseous Radioactivity Monitoring System inoperable, operation may continue provided grab samples of the containment atmosphere are obtained and analyzed for gross noble gas activity at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Restore the inoperable monitor to OPERABLE status within 7 days or, in lieu of any report required by Specification 6.9.2, prepare and submit a special report to the Commission pursuant to Specification 6.9.3 within 30 days outlining actiens taken, cause of inoperability and plans for restoring the monitor to OPERABLE status.

b. With the Volume Control Tank Level Monitoring System inoperable, be in at led: HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. With the Containment Main Sump Level (Narrow Rhnge) Monitoring System inoperable, restore the inoperable level monitoring system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

d. With both the Containment Main Sump Level (Narrow Range) Monitoring System and the Containment Atmosphere Gaseous Radioactivity Monitor-ing System inoperable, operation may continue provided grab samples of the containment atmosphere are obtained and analyzed for gross noble gas activity at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Restore the Contain-ment Main Sump Level (Narrow Range) Monitoring System to OPERABLE STATUS within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from time of its initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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w . REACTOR COOLANT SYSTEP)

.4 LEbyf_GE DETECTION SYSTEMS SURVEILLANCE REQUIREMENTS The Leakage Detection Systems shall be demonstrated OPERABLE by:

a. Containment Atmosphere Gaseous Radioactivity Monitoring System-performance of CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.2-1, and.

b.- Containment Main Sump'. Level- (Narrow Range) Monitoring System-performance of a CHANNEL CHECK. at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and a CHANNEL CALIBRATION

• at least once per 18 months.

c. Volume Control.' Tank Level Monitoring System-performance of L CHANNEL CHECK *. at.least once per.24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, CHANNEL CALIBRATION at least once per 18 months and ANALOG CHANNEL OPERATIONAL TEST at least once every 90 days.

• Following any seismic event greater than OBE (one half the Safe Shutdown Earthquake) the indicated calibration, or check shall be performed prior to declaring such affected instruments gnerable.

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.3.3 REACTOR COOLANT SYSTE BASES 3.3.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above 1.30 during all normal operations and anticipated transients. In MODES I and 2 with one reactor coolant loop not in operation this specification requires that the plant remain below 65% power. With less than the required reactor coolant loops in operation, the plant shall be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, The loop isolation valves are required to be OPERABLE in the operating loops in order to terminate the primary to secondary leak path in the event of a steam generator tube rupture.

In MODE 3, two reactor coolant loops provide sufficient heat removal capabil-ity for removing core decay heat even in the event of a bank withdrawal accident. Single failure considerations require that three loops be OPERABLE.

A single reactor coolant loop provides sufficient heat removal capability for

decay heat if a bank withdrawal accident can be prevented (i.e., by opening

' the reactor trip system breakers or de-energizing the control rod drive lift coils). Single failure considerations require that two loops be OPERABLE.

In MODE 4, two reactor coolant loops provide sufficient heat removal capabil-ity for removing decay heat even in the event of a bank withdrawal accident.

Single failure considerations require that three loops be OPERABLE. A single reactor coolant or RHR loop provides sufficient heat removal capability for decay heat if a bank withdrawal accident can be prevented, i.e., by opening the reactor trip system breakers or de-energizing the control rod drive lift coils. Single failure considerations require that two loops be OPERABLE.

In MODE 5 with reactor coolant loops filled, a single RHR loop provides sufficient heat removal capability for removing decay heat. A bank withdrawal accident is prevented by opening the reactor trip system breakers or de-energizing the control rod drive lift coils. Single failure considerations require that at least two RHR loops be OPERABLE.

In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat. A bank withdrawal accident is prevented by opening the reactor trip system breakers or de-energizing the control rod drive lift coils. Single failure considerations and the unavailability of the steam generators as a heat removing component require that at least two RHR loops be OPERABLE.

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'3 . 3 .1 " REACTOR COOLANT SYSTEM BASES 3.3.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION (continued)

The operation of one Reactor Coolant Pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant Systein. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.

The restrictions on starting an RCP with one or more RCS cold legs less than or equal to 3150F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G to 10CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 200F above each of the RCS cold leg tempera-tures.

The requirement to maintain the boron concentration of an isolated / idled loop greater than or equal to the boron concentration of the operating loops or the boron concentration required to meet SHUTDOWN MARGIN requirements ensures that no unacceptable reactivity addition to the core could occur during startup of an isolated / idled loop. Verification of the boron concentration in an isolated / idled loop prior to opening the stop valves provides a reassurance of the adequacy of the boron concentration in the isolated / idle loop.

Startup of an isolated / idled loop could inject cool water from the loop into the core. The reactivity transient resulting from this cool water injection is minimized by prohibiting isolated / idled loop startup until its temperature is within 200F of the operating loops.

3.3.2 SAFETY 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 293,300 lbs. per hour of saturated steam at 2485 psig.

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 OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressure-zation. In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures.

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l l REACTOR C00LAJT SYSTEM BASES l

3.3.2 SAFETY VALVES (continued)

During operation, all pressurizer Code safety valves must be OPERABLE to l 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 l maximum surge rate resulting from a complete loss-of-load assuming no Reactor I trip until the first Reactor Trip 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 valve's 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.

3.3.3 PRESSURIZER j The limit on the water level on the pressurizer assures that the parameter is maintained within that assumed in the safety analyses. The 12-hour periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The requirement that a

, minimum number of pressurizer heaters be OPERABLE enhances the capability of I

the plant to control Reactor Coolant System pressure and establish natural l circulation.

l 3.3.4 RELIEF VALVES Operation of the power-operated relief valves (PORVs) minimizes the undesir-I able opening of the spring-loaded pressurizer Code safety valves and provides an alternate means of core cooling. Each PORV has a remotely operated block valve to provide a positive shutoff capability should a PORV become inoper-able. One of two redundant PORV relief trains must be OPERABLE to adequately cool the core in the event that the steam generators are not available to remove core decay heat. When in automatic mode, all PORVs and block valves open automatically on high pressurizer pressure. The PORVs and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump.

The OPERABILITY of two spring-loaded relief valves (SLRVs) or an RCS vent opening of greater than 7 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 315'F.

Either SLRV has adequate relieving capability to protect the RCS from over-pressurization when the transient is limited to either: (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 20*F above the RCS cold leg temperatures, or (2) the start of a i charging pump (centrifugal) and its injection into a water-solid RCS. j l

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3.3 REACTOR COOLANT SYSTEM BASES I

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'The Maximum Allowed SLRV Setpoint for the Low Temperature Overpressure Protec- ]

tion System (OPS) is derived by analysis which models the performance of the 1 l OPS assuming various mass input and heat input transients. Operation with a SLRV Setpoint less than or equal to the maximum Setpoint ensures that Appendix o G criteria will not be violated with consideration for a' maximum pressure overshoot beyond the SLRV Setpoint which can occur as a result of time delays in signal processing and valve opening, instrument uncertainties, and single I

failure. To ensure that mass and heat input transients more severe than those I assumed cannot occur, Technical Specifications require lockout of all but one l charging pump (centrifugal or metering) while in MODES 4, 5, and 6 with the reactor vessel head installed and disallow start of an RCP if secondary temperature is more than 200F above RCS cold leg temperature.

l 3.3.5 REACTOR COOLANT SYSTEM VENTS l

Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the RCS that could inhibit natural circulation core cooling.

The OPERABILITY of at least one RCS vent path from the reactor vessel head and the pressurizer steam space ensures the capability exists to perform this ,

function.

The valve redundancy of the RCS vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply or control system does not prevent isolation of the vent path.

The function, capabilities, and testing requirements of the RCS vents are consistent with the requirements of Item II.B.1 of NUREG-0737, " Clarification of TMI Action Plan Requirements", November 1980.

3.3.6 LEAKAGE DETECTION SYSTEMS The RCS Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.

This technical specification ensures a reliable means of detecting unidenti-fied leakage in tha reactor coolant system which potentially could be due to a circumferential ttrough-wall flaw in primary system piping. The required instrument sensitivity is 1 gallon per minute in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> as stated in Condi-tion (2) of Generic Letter 84-04. Because the Volume Control Tank Level Monitoring System and the Containment Main Sump Level (Narrow Range) Monitor-ing System are not seismically qualified, surveillance requirements for a seismic event greater than OBE (one half of the Safety Shutdown Earthquake) are imposed. ,

The RCS Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.

These Detection System are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

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, ., - c, Table 4.2-1 (contin'ued)

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Channell Action Minimum Frecuency I

8. Variable Low Calibrate- Each refueling Pressure trip set Check Each Shift

9. Rod. Position Calibrate Each refueling Digital Voltmeter Check with cour:ters Every six inches of rod motion when data logger .is _

out of service.

10. Rod Position Test Each refueling Counters Check with Digital Every six inches of rod Voltmeter motion when data logger is out of service.

11. Steam Generator Calibrate Each refueling Level Check .Each shift

12. Steam Generator . Calibrate Each refueling Flow Mismatch Check Each shift

13. Charging Flow Calibrate Each refueling

14. Residual-Heat Calibrate Each refueling Pump Flow

. 15. Boric Acid Calibrate Each refueling Tank Level Check Each week

16. Refueling Water Calibrate Each refueling Storage Tank Test 90 days Level

17. Blank **

18. Blank *

19. Radiation Calibrate Each refueling Monitoring Test Each Month System Channel Check Each Shift

20. Boric Acid Calibrate Each refueling Control

21. B1ank*

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Table 4.2-1 (continued)

Channels astion Hjnimum Freauency

22. Valve Temperature Test Each refueling Interlocks

23. Pump Valve Check Each refueling Interlock

24. Reactor Coolant Calibrate Each refueling System OPS Check Each refueling

25. Auxiliary Feedwater Calibrate Each refueling Flow Rate Check Each month

26. Blank *

27. PORV. Position Calibrate Each refueling Indication Check Each month (Acou tic Monitor)

28. PORV Block Valve Calibrate Each refueling Position Indication

29. Safety Valve Calibrate Each refueling Position Check Each month Indication (Acoustic Monitor)

• Items 18, 21, and 26 of this Table are included in Table 3.23-2 (Items 1, 5, and 10).

• • Item 17 of this Table is included in Specification 3.3.6.1.

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