ML17254A944

Proposed Tech Specs Re Operability & Surveillance Requirements for Certain NUREG-0737 Required Equipment

ML17254A944
Person / Time
Site:	Ginna
Issue date:	09/14/1984
From:	ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17254A941	List: ML17254A940 ML17254A942 ML17254A944
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.1, TASK-2.B.3, TASK-2.D.3, TASK-2.E.1.1, TASK-2.E.1.2, TASK-2.F.1, TASK-2.F.2, TASK-2.G.1, TASK-3.D.3.4, TASK-TM NUDOCS 8409180219
Download: ML17254A944 (41)

Text

3.1.1.5 Pressurizer

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Whenever the reactor's at hot shutdown or critical, the 3.1.1.6 pressurizer shall have at least 100 kw of heaters operable and a water level maintained between 12% and 87% of level span.

If the pressurizer is inoperable due to heaters or water leveli restore the pressurizer to operable status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or have the RHR system in operation within an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Reactor Coolant S stem Vents a 0 When the reactor is at hot shutdown or criticali at least one reactor coolant system vent path con-sisting of two valves in series shall be operable and closed* at each of the following locations:

1.

Reactor Vessel head 2.

Pressurizer steam space

• The PORY block valve is not required to be closed but must be operable if the PORV is capable of being opened.

b.

With one or more vents at the above reactor coolant system vent path locations inoperablei startup may commence and/or power operation may continue provided at least one vent path is operable and the inoperable vent paths are maintained closed with motive power removed from the valve actuator of all the valves in the inoperable vent paths.

If the 8409i80>

000P44 PDR ADOCK 0500 requirements of 3.1.l.6a are not met within 30 days>

be in hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and below 350 F within'he following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

0 l

Change No. gg 3.1-4b.

Amendment No. gP, gP, gP, gg, 61 PROPOSED

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With all of the above reactor coolant system vent paths inoperable; maintain the inoperable vent paths closed with power removed from the valve actuators of all the valves in the inoperable vent paths>

and restore at least one of the vent paths 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 hot 0

shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and below 350 F within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Bases The plant is designed to operate with all reactor coolant loops in operation and maintain the DNBR above the limit value during all normal 3.1-4c PROPOSED

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operations and anticipated transients.

Heat transfer analyses show that reactor heat equivalent to 130 EZ (8.5%) can be removed by natural circulation alone.

Therefore operation with one operating reactor coolant loop while below 130 MWT provides adequate margin.

The specification permits an orderly reduction in power if a reactor coolant pump is lost during operation between 130 t&lT and 50% of rated power:

Above "50% power, an automatic reactor trip will"occur, if either pump is lost.

The power-to-flow ratio will be maintained equal to or less than one which ensures that the minimum DNB ratio increases at lower flow since the maximum enthalpy rise does not increase.

When the reactor coolant system average temperature is above 350 F,,

a single reactor coolant loop provides sufficient heat removal capability for removing decay heat;

however, single failure consideiations require one loop be in operation and the other loop be capable of removing heat via natural circulation.

When Me reactor coolant system average temperature is between 200 F and 350'F or while in cold shutdown, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure.considerations require that at least two loops be operable.

Thus, if the reactor coolant loops are not operable, this specification requires two RHR loops to be operable.

When the boron concentration of the reactor coolant system is to I

be reduced the process must be'niform.Co prevent sudden-reactivity 3.1-4d Amendment No.

PROPOSED

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changes in the reactor.

Mixing of the reactor coolant will be sufficient to prevent a sudden increase in reactivity if at least one reactor coolant pump or one residual heat removal pump is running while the change is taking place.

The residual heat removal pump will circulate the primary system volume in approximately one half hour.

The pressurizer is of no concern because of, the low pressurizer volume and because the pressurizer boron concentration will be higher than that of the rest of the reactor coolant.

When the boron concentration of the reactor coolant system is to be increased, the process must be uniform to prevent sudden reactivity increases in the reactor during subsequent startup of the reactor coolant pumps. 'ixing of the reactor coolant will be suffic'ent to maintain a uniform boron concentration if at least one reactor coolant pump is running while the change is taking place.

Emergency boration without a reactor coolant pump in operation is not prohibited by this specification.

Prohibiting reactor coolant pump starts without a large void in the pressurizer or without a limited RCS temperature differential will prevent RCS overpressurization due to expansion of cooler RCS water as it enters a warmer steam generator.

A 38% level in the pressurizer will accommodate the swell resulting from a reactor coolant pump start with a RCS temperature of 140'F and steam generator secondary side temperature of 340'F, or the maximum temperature which usually exists prior'o cooling the reactor with the"RHR'ystem.

3.1-4e Amendment No. 43 PROPOSED

Temperature requirements for the steam generator correspond with measured NDT for the shell and allowable thermal stresses in the tube sheet.

Each of the pressurizer code safety valves is designed to relieve 288I000 lbs. per hr. of saturated steam at the valve set point.

Below 350,F and 350 psig in the reactor coolant system>

the 0

residual heat removal system can remove decay heat and thereby control system temperature and pressure.

If no residual heat were removed by any of the means available>

the amount of steam which could be generated at safety valve relief pressure would be less than half the valves'apacity.

One valve, thereforei provides adequate defense against overpressurization.

The power operated relief valves (PORVs) operate to relieve RCS pressure 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 inoperable.

The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path.

The requirement that 100 kw of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation at hot shutdown and during cooldown. (3) 3.1-4f.

Amendment No. gg, hagi 43 PROPOSED

Reactor Coolant System Vents Reactor Coolant System Vents are provided to exhaust noncon-densible gases and/or steam from the primary system.that could inhibit natural circulation..core cooling.

The operability of at least one reactor coolant system vent path from the reactor vessel head and one from the pressurizer steam space ensures the capability exists to perform this function.

The valve redundancy of the reactor coolant system vent paths serves to minim'ize 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, capabilitiesi and testing requirements of the reactor coolant system vent systems are consistent with the requirements of Item ZI.B.1 of NUREG-0737, "Clarification of TNI Action-Plan Requirements"

> November 1980.

References (1)

FSAR Section 14.1.6 (2)

PSAR Section 7.2.3 (3)

Letter from L.D. White, Jr.

to D. L. Ziemanni USNRCI dated October 17I 1979 3.1-4g PROPOSED

P yi fi

3.5 Instrumentation S stems Applies to plant instrumentation systems.

To delineate the conditions of the plant instrumentation and safety circuits and to limit the release oz radio-active materials.

S ecification:

3.5.1 Operational Safety Instrumentation 3.5.1.1 The number of Hinimum Operable Channels for instru-mentation shown on Tables 3.5-1 through 3.5-3 shall be OPERABLE for'lant operation at rated power.

3.5.1.2 In the event the number of channels of a particular subsystem in service falls below the limit given in the columns entitled Hinimum Operable Channelsi operation shall be limited according to the requirement shown in 3.5.2 3.5.2.1 the last column of Tables 3.5-1 through 3.5-3.

Accident Honitoring Instrumentation The accident monitoring instrumentation channels shown in Table 3.5-4 shall be operable whenever the reactor is at hot shutdown or is critical.

3.5.2.2 While critical> with the number of operable accident monitoring instrumentation channels less than the Total Required Number of Channels shown in Table 3.5-4I either restore the inoperable channel(s) to operable status within 7 daysi or be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

3 5-1 Amendment No.

57 PROPOSED

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3.5.2.3 3.b.2.4 While critical< with the number of operable accident monitoring instrumentation channels less than the HINIHUN CHANNELS OPERABLE requirements of Table 3.5-4, either restore the inoperable channel(s) to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The radiation accident monitoring instrumentation channels shown in Table 3.5-7 shall be operable whenever the reactor is at hot shutdown or is critical.

With one or more radiation monitoring channels inoperable<

take the action shown in Table 3.5-7.

Startup may commence or continue consistent with the action statement.

3.5.3 3.5.3.1 Engineered Safety Feature Actuation Instrumentation The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels shown in Tables 3.5-2 and 3.5-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.5-5.

3.5.3.2 With an instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.5-5> declare the channel inoperable and apply the applicable ACTION requirement of Tables 3.5-2 and 3.5-3 until the channel is restored 3.5.3.3 to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint Value.

With an instrumentation channel inoperable>

take the action shown in Tables 3.5-2 and 3.5-3.

3.5-2 Amendment No.

57 PROPOSED

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1 3.5.4 Radioactive Effluent Honitoring Instrumentation 3.5.4.1 The radioactive effluent monitoring instrumentation shown in Table 3.5-6 shall be operable at all times with alarm and/or trip setpoints set to ensure that the

-limits of Specifications 3.9.1.1 and 3.9.2.1 are not 3.5.4.2 exceeded.

Alarm and/or trip setpoints shall be established in accordance with calculational methods set forth in the Offsite Dose Calculation Hanual.

r If the setpoint for a radioactive ef fluent monitor alarm and/or trip is found to be higher than requiredi one of the following three measures shall be taken immediately:

(i) the setpoint shall be immediately corrected without declaring the channel inoperable; or (ii) immediately suspend the release of effluents monitored by the affected channel; or (iii) declare th'e chanrrel inoperable.

3.5.4.3, If the number of channels wtrich are operable is found to be less than required>

take the action shown in Table 3.5-6.

3.5.5 Control Room HVAC Detection S stems*

3.5.5.1 During all modes of plant operation>

detection systems for chlorine gasI ammonia gas and radioactivity in the control room HVAC intake shall be operable with setpoints to isolate air intake adjusted as follows:

5 ppm ammonia I 35 mg/m 3

-8 radioactivity, particulate 1 x 10 uc/cc iodine 9

x 10 uc/cc

-5 noble gas 1

x 10 uc/cc Amendment No.

57 PROPOSED

3.5.5.2 Nith one of the detection systems inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> isolate the control room HVAC air intake.

Naintain the air intake isolated except for short periodsi not to exceed l hour a dayr when fresh air makeup is allowed to improve the working environment in the control room.

• Effective January lI 1985 Basis During plant operationsI the complete instrumentation systems will normally be in service.

Reactor safety is provided by the Reactor Protection System>

which automatically initiates appropriate action to prevent exceeding established limits.

Safety is not compromised>

howeveri by continuing operation with certain instrumentation channels out of service since provisions were made for this in the plant design.

This specification outlines limit-ing conditions for operation necessary to preserve the effectiv'e-ness of the reactor control and protection system when any one or more of the channels is out of service.

3.5-2b Amendment No.

57 PROPOSED

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Almost all reactor protection channels are supplied P

with sufficient redundancy to provide the capability for channel calibration and test at power.

Exceptions are backup channels such as reactor coolant pump breakers.

The removal of one trip channel is accom-plished by placing that channel bistable in a tripped mode; e.g.,

a two-out-of-.three circuit becomes a

one-out-of-two circuit.

Testing does not trip the system unless a trip condition exists in a concurrent channel'.

The operability of the accident monitoring instru-mentation ensures that su ficient information is available on selected plant parameters to-monitor and assess these variables during and following an accident.

This capability is consistent with the recommendations of NUREG-0578,

'iTMI-2 Lessons Learned Task Force Status Report add Short-Term Recommendation".

The radioactive liquid effluent instrumentation is provided to monitor and/or control, as applicable, the releases of radioactive materials in liquid effluents.

The alarm and/or t=ip setpoints for these instruments are calculated in accordance with the ODCM to ensure that alarm and/or trip will occur prior to exceeding the limits of 10 CFR Part 20.

The operability and use

.of this instrumen"ation is consistent with the recuirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

3.5-2c Amendment No. '$7 PROPOSED

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>>~'The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials 'in gaseous effluents.

The alarm and/or trip setpoints for these ir~truments are calculated in accordance with the ODCM to ensure that alarm and/or trip will occur prior to exceeding the limits of 19 CFR Part 20.

This instrumentation also includes provisions for monitoring the concentrations of potentially explosive gas mixtures in the waste gas holdup system.

The operability and, use of this instrumentation is consistent with the requirements of General Design Criterion 64 of Appendix A to 10 CFR Part 50.

Reference FSAR - Section'.2.1 3.5-2d Amendment No.

5 t PPOPOSED

V

TABLE 3.5W Accident Honitorin Instrumentation INSTRUMENT KYTAL REQUIRED NO. OF CHANNELS (7)

MINIMUM CHANNELS OPERABLE (7)

l. Pressurizer Water Level (1)

2. Auxiliary Feedwater Flow Rate (2)(3)

3. Steam Generator Water Level Wide Range (3) 2/steam generator 1/steam generator 1/steam generator 1/steam generator

4. Reactor Coolant System Subcooling Margin Monitor (4)

5. Pressurizer PORV Position Indicator (5) 6.

PORV Block Valve Position Indicator (1) 2/Valve 1/Valve 1/Valve 0/Valve

7. Pressurizer Safety Valve Position Indicator (5) 2/Valve 1/Valve

8. Containment Pressure (8)

9. Containment Water Level (Narrow Rangei Sump A)

10. Containment Water Level (Wide Rangei Sump B) 1(6) ll. Core-Exit Thermocouples Notes 4/core quadrant 2/core quadrant (1) Emergency power for pressurizer equipmentI NUREG-0737, item II.G.1.

(2) Auxiliary feedwater system flow indication, NUREG-0737, item II.E.1.2.

(3) Only 2 out of the 3 indications (two steam generator auxiliary feedwater flow and one wide-range steam generator level) are required to be operablei NUREGW737I item ICE.l;2 ~

(4) Instrumentation for detection of inadequate core coolingi NUREG-0737l item item II.F.2.1.

(5) Direct indication of relief and safety valve position>

NUREG-0737I item II.D.3.

Two channels include a primary detector and RTD as the backup detector.

(6) Operation may continue with less than the minimum channels operable provided that the requirements of Technical Specification 3.1.5.3 are met.

(7) See Specification 3.5.2 for required action.

(8) Containment pressure monitori NUREG 0737I item II.F.1.4.

Amendment No.

42 PROPOSED

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'lABLE 3.5-7 Radiation Accident Honitorin Instrumentation Instrument 1.

Containment Area (R-29 and R-30) 2.

Noble Gas Effluent Monitors i.

Plant Vent (R-14A) ii.

A Hain Steam Line (R-31) iii. B Hain Steam Line (R-32) iv.

Containment Purge (R-12A) v.

Air Ejector (R-15A)

Action Statements Minimum Channels

~Oeeable Action Action '1 With the number of operable channels less than required by the Minimum Channels Operable requirements>

either restore the inoperable channel(s) to operable status within 7-days of the eventi or prepare and submit a Special Report to the Commission within 30 days following the event outlining the action takeni the cause of the inoperability and the plans and schedule for restoring the system to operable status.

3.5-18 PROPOSED

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Containment Isolation Valves With one or more of the isolation valve(s) specified in Table 3.6-1 inoperableI maintain at least one isolation valve operable in each af fected penetration that is open and either-a.

Restore the inoperable valve(s) to operable, status within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />si or b.

Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by'se of at least one deactivated automatic'alve secured in the isolation positioni or Co d.

Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual valve or blind flangei or Be in at least hot shutdown within the next 6,hours 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 />.

Isolation valves are inoperable from a leakage stand-point if the leakage is greater than that allowed by 10 CFR 50 Appendix J.

Combustible Gas Control When the reactor is criticali at least two independent containment hydrogen monitors shall be operable.

One. of the monitors may be the Post Accident Sampling System; With only one hydrogen monitor operable, restore a

second monitor to operable status within 30 days or be in at least hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

With no hydrogen monitors operable>

restore at least one monitor 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 shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

3. 6-2 Amendment No.

42 PROPOSED

TABLE 4.l-l (CONTINUED) 80 8

4 C

Channel 25.

Containment Pressure 26.

Steam Generator Pressure 27.

Turbine First Stage Pressure 28.

Emergency Plan Radiati.on Instruments Check Calibra te Test S

R M

R Remarks Narrow range containment pressure

(-3.0<

+3 psig) excluded 29.

Environmental Monitors 30.

Loss of Voltage/Degraded Voltage 480 Volt Safeguards Bus M~g OO 8 08 5 08 A

M$ 8 RO 8 Urt 5 Pr +

05 Rl0 1

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QM 35.

36.

Ammonia Detector> Control Room Air Intake*

Radiation Detectors<

Control Room Air Intake*

3l.

Trip of Main Feedwater Pumps 32.

Steam Flow AVG 34.

Chlorine Detector> Control Rocm

, Air Intake*

'NA

4'

TABLE 4.1-3

, Accident Monitorin Instrumentation Surveillance Re uirements Xnstrument Channel Check Channel Calibration Channel Test Pressurizer Water Level (1) see Table 4.1-1 see Table 4.1-1 NA 2.

3.

4 Auxiliary Feedwater Flow Rate (4)

Reactor Coolant System Subcooling Hargin Monitor (2)

Pressurizer PORV Position Xndicator (primary detector)

(3)

Pressurizer PORV Position Indicator (RTD backup detector)

(3) see Section 4.8.1 M

NA 7.

8.

9.

PORV Block Valve Position indicator (1)

Pressurizer Safety Valve Position Indicator (primary detector)

(3)

Presurizer Safety Valve Position Indicator (RTD backup detector)

(3)

Containment Pressure NA 10.

12.

13.

14.

15.

Steam Generator Water Level Wide Range Containment Water Level (Narrow Rangei Sump A)

M Containment Water Level (Wide Range>

Sump B)

M Core Exit Thermocouples Containment Area High Range Radiation (R-29 and R-30) (5)

Plant Vent (R-14A)

NA 16.

17.

18.

19.

A Main Steam Line (R-31)

B Hain Steam Line (R-32)

Containment Purge (R-12A)

Air Ejector (R-15A)

M NA (1)

(2)

(3)

(4)

(5)

Emergency Power Supply Requirements for Pressurizer Level Indicators NVREG 0578 Item 2.1.1 Instrumentation for Detection of. Inadequate Core Cooling NUREG 0578 Item 2.l.l Direct Indication of Power Operated Relief. Valve and Safety Valve Position NUREG

~

0578 item 2.1.3.a Auxiliary Feedwater Flow Indication to Steam Generator NUREG 0578 item 2.1.7.b Acceptable criteria for calibration are provided in Table II.F.1-3 of NUREG 0737 4.l-ll Amendment No.

42 PROPOSED

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a) to demonstrate a reactor coolant loop operable<

the reactor coolant pump(s) I if not in operation> shall be demonstrated operable at least once per 7 days by verifying correct breaker alignments and indicated power availability.

4.3.5.4 4.3.5.5 4.3.5.6 b) to demonstrate a residual heat removal pump is operable>

the surveillance specified in the Inservice Pump and Valve Test Program prepared pursuant to 10 CFR 50.55a shall be performed.

When the reactor is at cold shutdown or when the average coolant temperature is between 200 F and 350 F and fuel 0

0 is in.the reactor> at least one coolant loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

In addition to the. above requirementsi in order to demonstrate that a reactor coolant loop is operablei the steam generator water level shall be greater than or equal to 16% of the narrow range instrument span.

Each reactor coolant system vent path shall be demonstrated operable at least once per 18 months by:

1.

Verifying all manual isolation valves in each vent path are locked in the open position.

2.'erifying flow through the reactor coolant vent system vent paths us'ing either liquid or gas.

Basis:

This material surveillance program monitors changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region of the reactor resulting from exposure to 4.3-4 Amendment No. gg, 43 PROPOSED

neutron irradiation and the thermal environment.

The test data obtained from this program will be used 'to determine the conditions under which the reactor vessel can be operated with adequate margins of safety against fracture throughout its service life.

The surveillance requirements on pressurizer equipment will assure proper performance of the pressurizer function and give early indication of malfunctions.

4. 3-5.

PROPOSED

4.

the tendon containing 6 broken wires) shall be inspected.

The acceptance criterion then shall be no more than 4 broken wires in any of the additional 4

tendons.

If this criterion is not sa'tisfied, all of the tendons shall be inspected and if more than 5% of-.the total wires are broken>

the reactor shall be shut down and depressurized.

4.4.4.2 Pre-Stress Confirmation Test a

0 b.

Lift-off tests shall be performed on the 14 tendons identified in 4.4.4.1a aboveI at the intervals specified in 4.4.4.lb.

If the average stress in the 14 tendons checked is less than 144,000 psi (60% of ultimate stress), all tendons shall be checked for stress and retensionedr,if necessary>

to a stress of 144I000 psi.

Before reseating a tendonI additional stress (6't) shall be imposed to verify the ability of the tendon to sustain the added stress applied during accident conditions.

4.4.5 4.4.5.1 Containment Isolation Valves Each isolation valve specified in Table 3.6-1 shall be demonstrated to be operable in accordance with the Ginna Station Pump and Valve Test Program submitted in accordance with 10 CFR 50.55a.

4.4.6 4.4.6.1 Containment Isolation Res onse Each containment isolation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations for the NODES and at the frequencies shown in Table 4.1-1.

4.4.6.2 The

RESPONSE

TINE of each containment isolation function shall be demonstrated to be within the limit at least once per 18 months.

Each test shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific function as shown in the "Total No. of Channels" Column of Table 3.5-3.

The response time limit shown on Table 3.6-1 does not include diesel generator starting times but does include valve travel times for all valves that change position.

The times determined in independent testsI such as electronic response of portions of the initiating circuitry and valve travel times>

may be combined to determine the total function response time.

4.4-8 Amendment No.

42 PROPOSED

4.4.7 4.4.7.1 4.4.7.2 Basis:

Containment H dro en Monitors Demonstrate that two hydrogen monitors are operable at least daily by verifying that the unit is on or in standby.

At least once per quarter perform a channel calibration using two sample gases containing known concentrations of hydrogen.

The cyy)ainment is designed for an accident pressure of 60 psig While the reactor is operating I the internal environment of the containment will be air at approximately atmospheric pressure and a

maximum temperature of the steam-air mixture at the 0

~

peak accident pressure of 60 psig is calculated to be 286 F.

4.4-8a

.PROPOSED

Attachment B

Licensees were requested to propose additional THI Technical Specifications by Generic Letter 83-37.

Enclosure 1 to that letter detailed eleven topics that were to be addressed in the plant technical specifications..

Each of these topics has been considered as outlined below.

Reactor Coolant System Vents (II.B.l)

New specifications 3.1.1.6 and 4.3.5.6 are proposed which require operability and surveillance testing of this equipment.

Consistent with our current approved practice i valve cycling requirements will be included in the Ginna ISI program in Appendix C of the Quality Assurance Nanual.

(2) Post Accident Sampling (II.B.3)

(3)

The post accident sampling capability required by NUREG-0737 has been provided.

Existing technical specifica-tion 6.8.1.a requires that applicable procedures recommended by Regulatory Guide 1.33 be'established<

implemented and maintained.

This requires that procedures exist to respond to emergency conditions and significant eventsi for control of 'radioactivity includ-ing sampling and monitoring and for control of measuring "and test equipment.

In additioni technical specifica-,

tion 6.8.1.e requires that procedures be established for implementation of the Emergency Plan.

Additional specifications are unnecessary and will only add to the already sizable volume of specifications and place an additional burden on operators and those required to know the technical specifications.

Plant readiness to cope with emergency conditions is periodically audited by NRC Inspection and Enforcement.

Long-Term Auxiliary Feedwater System Evaluation (II.E.l.l)

Post TNI auxiliary feedwater system specifications were included in Amendment 42 and are found in sections 3.4 and 4. 8.

Additional correspondence related to this system is found in references 4 through 20 of a letter from John E. Naier to Dennis N. Crutchfieldi USNRCi dated January 19>

1982.

Alignment of safeguards systems was also addressed by RGSE letter dated Nay 22, 1980 from L. D. White> Jr.

to Dennis H. Crutchfield and was approved by a letter from Hr. Crutchfield dated June 16>

1982.

(4) Noble Gas Effluent Monitors (II;F.l)

Operability of the post accident noble gas effluent monitors is required by proposed specification 3.5.2.4.

Surveillance of these monitors is required by an addi-tion to existing technical specification Table 4.1-3.

(5) Sampling and Analysis of Plant Effluents (II.F.1.2)

Additional specifications are unnecessary as described in (2) above.

The existing technical specifications along with the specifications proposed in (4) above assure that the required capability exists.

Containment High Range Radiation Monitor (II.F.1.3)

Containment area high range radiation monitor operability and surveillance requirements are given in proposed specification 3.5.2.4 and Table 4.1-3.

(7) Containment Pressure Monitor (II.F.1.4)

Containment pressure monitor operability and surveil-lance requirements have been added in proposed revisions to Tables 3.5-4 and.4.1-3.

Containment Water Level Monitor (II.F.1.5)

Containment pressure monitor operability and surveil-lance requirements have been added in proposed revisions to Tables 3.5-4 and 4.1-3.

(9) Containment Hydrogen Monitor (II.F.1.6)

Proposed specifications 3.6.4 and 4.4.7 include operability and surveillance requirements for the con-tainment hydrogen monitors.

(lO)

(>>)

Instrumentation for Detection of Inadequate Core Cooling (II.F.2)

P Operability and surveillance requirements for'he subcooling margin monitors and core exit thermocou'ples are included in revised Tables 3.5-4'and 4.1-3.

RGGE has'ommitted to install a differential pressure type reactor vessel water level instrument by 1986.

Appropriate technical specifications will be.proposed for the water level instrument after its installation and testing is complete.

Control Room Habitability Requirements (III.D.3.4)

Control room HVAC detection systems operability and surveillance requirements are included in proposed specification 3.5.5 and revised Table 4.1-1.

Approval

A P

of the existing plant conf iguration to meet control room habitability requirements was issued in an NRC letter from Dennis M. Crutchfield to John E. Maier dated April ll> 1983.

Operability and surveillance requirements for the control room emergency air treatment system are given in existing technical specifications 3.3.5 and 4.5.2.3.6 through 4.5.2.3.9.

In accordance with 10 CFR 50.91<

the proposed changes to the Technical Specifications have been evaluated against three criteria to determine if the operation of the facility in accordance with the proposed amendment would:

l.

involve a significant increase in the probability or consequences of an accident previously evaluated; or 2.

create the possibility of a new or different kind of accident from any accident previously evaluated; or 3.

involve a significant reduction in a margin of safety.

None of the proposed changes will have an adverse impact as judged against these criteria.

Each of the proposed changes noted above constitute an additional limitation> restriction or control not presently included in the technical specifications or incorporates an additional surveillance requirement.

These proposed changes thus conform to the Commission's example (ii) of changes that do not involve a significant hazards consideration.

In addition to the proposed changes addressed above>

minor revisions have also been made which are administrative in nature.

Table 3.5-4 has been revised to reference the requirements of NUREG-0737 instead of NVREG-0578> to correctly describe pressurizer discharge pipe temperature elements as RTDs instead of thermocouples<

and to reflect that auxiliary feedwater flow indication is required for each steam generator instead of each pump.

A note has been deleted from Table 4.1-1 which required that the loss of voltage/degraded voltage surveillance specifica-tion be effective no later than the Spring 1982 refueling outage.

These administrative changes conform to the Commission's example (i) of changes that do not involve a significant hazards consideration.

Therefore<

a no significant hazards consideration is warranted for the proposed technical specification changes.

V