A06214, Proposed Tech Specs,Providing Limiting Conditions for Operation Per Generic Ltr 83-37

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Proposed Tech Specs,Providing Limiting Conditions for Operation Per Generic Ltr 83-37
ML20150D679
Person / Time
Site: Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date: 07/01/1988
From:
CONNECTICUT YANKEE ATOMIC POWER CO.
To:
Shared Package
ML20150D677 List:
References
RTR-NUREG-0737, RTR-NUREG-737 A06214, A6214, GL-83-37, NUDOCS 8807140103
Download: ML20150D679 (36)
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Text

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't 't Docket No. 50-213' A06214 Attachment 1 Haddam Neck Plant Proposed Changes to Technical Specifications Generic Letter 83 (NUREG-0737)

July 1988

~800714010:3 eso7oi PDR ADOCl; 05000213 P PDC 's

y REACTOR COOLANT SYSl@

Ll.5 REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION 3.3.5.1 At least one Reactor Coolant System (RCS) vent path consisting of at least two valves in series capable of being powered from 125-vol t D.C. buses shall be OPERABLE

  • and closed at each of the following locations:
a. Reactor Vessel Head, and
b. Pressurizer Steam Space.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With the pressurizer vent path inoperable, STARTUP and/or POWER OPERATION may continue provided that: i) the inoperable vent path is maintained closed with power removed from the valve actuator of all the valves in the inoperable vent path and ii) one power operated relief valve (PORV) and its associated block valve is OPERABLE; otherwise, restore either the inoperable vent path or one PORV and its associated block valve to OPERABLE status within 30 days, or submit a special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the path to OPERABLE status,
b. With the reactor vessel head vent path inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed with power removed from the valve actuator of all the valves in the inoperable vent path; restore the inoperable vent path to l OPERABLE status within 30 days, or, submit a Special l Report to the Commission pursuant to Specification 6.9.2  !

within the next 10 days outlining the cause of the malfunction and the plan for restoring the path to OPERABLE status.

l Power to the valves may be removed.

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'. 's SVRVEILLANCE RE0VIREMENTS Each RCS vent path snall be demonstrated OPERABLE at least once per 18 months by;

a. Verifying all manual isolation valves in each path tre locked in the open position,
b. Cycling each valve in the vent path through at least one complete cycle of full travel from the control room during COLD SHUTDOWN or REFUELING, and
c. Verifying flow through the RCS ver;t path during venting during COLD SHUTDOWN or REFUELING.

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3.3 REACTOR COOLANT SYSTEM BASES 3.3.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION 1he 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 1 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 capability 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 capability 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 changas during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.

TherestrictiongonstartinganRCPwithoneormoreRCScoldlegslessthan or equal to 315 F are provided to prever.t RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G to 10 CFR 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 wheg the secondary water temperature of each steam generator is less than 20 F above each of the RCS cold leg temperatures.

The requirement to maintain the boron concentration of an isolated / idled loop greater than or equal to the boron concentration of the operatina loops ensures that no reactivity addition to the core could occur curing 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 bcron concentration in the isolate (/ idled 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 prohibjting isolated / idled loop startup until its temperature is within 20 F of the operating loops. Making the reactor subcritical prior to isolated loop startup prevents any power spike which could otherwise result from this cool water-induced reactivity transient.

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, l provides overpressure relief capability and will prevent RCS overpressurization. In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures.

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REACTOR COOLANT SYSTEM BASES 3.3.2 SAFETY VALVES (continued)

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 assu ning no Reactor trip until the first Reactor Trip System Trip Setpoint ia, 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.

3.3.3 PRESSURIZER The limit on the water level in 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 the plant to control Reactor Coolant System pressure and establish natural circulation.

3.3.4 RELIEF VALVES Operation of the power-operated relief valves (PORVs) minimizes the undesirable 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 inoperable. 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 modo, 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 gPart 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 overpressurization when the transient is limited to either: (1) the start of an idle RCP with the secondary n

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 charging pump (centrifugal) and its injection into a water #911d RCS.

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The Maximum Allowed SLRV Setpoint for the Low Temperature Overpressure Protection System (0PS) is derived by analysis which models the performance of the 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 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 failure. To ensure that mass and heat input transients more severe than those asumed cannot occur, Technical Specifbations require lockout of all but ene charging pump (centrifugal or meterin3) while in MODES 4, 5, and 6 with the reactor vessel head installgd and disallow start of an RCP if secondary temperature is more than 20 F above RCS cold leg temperature.

3.3.5 REACTOR COOLANT SYSTEM VENTS 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.

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INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.23 The accident monitoring instrumentation channels shown in Table 3.231 shall be OPERABLE.

APPLICABILITY: As shown in Table 3.23-1 ACTION: As shown in Table 3.23-1 SVRVEILLANCE REQUIREMENTS 3 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION at the frequencies shown in Table 3.23-2 '

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TABLE 3.23-1 ACCIDENT MONITORING INSTRUMENTATION TOTAL MINIMUM NO. OF CHANNELS APPLICABLE INSTRUMENT CHANNELS OPERABLE MODES ACTION

1. Containment Pressure 2 1 1,2,3 4,6
2. Reactor Coolant Cold Leg Temperature - Wide Range 1/ loop 1/ loop 1,2,3 4
3. Reactor Coelant Hot Leg Temperature - Wide Range 1/ loop 1/ loop 1,2,3 4
4. Reactor Coolant Pressure - Wide Range 2 1 1,2,3,4 1,2
5. Containment Water Level - Wide Range 2 1 1,2,3 4,6
6. Core Exit Thermocouples 16/ core 2/ quadrant
  • 1,2,3 4
7. Main Stack Wide Range F sie Gas Monitor 1 1 1,2,3,4*** 3
8. Containment Atmosphere-L "ange Radiation Monitor 2 2 1,2,3,4 7
9. Reactor Vessel Water Level 2** 1** 1,2,3 5
10. Reactor Coolant System Subcooling Margin Monitor 2 1 1,2,3 8 TABLE NOTATIONS
  • Quadrant III may have only one channel OPERABLE.
    • A channel is composed of eight sensors in a probe. A channel is OPERABLE if four or more sensors ~, one or more in the head region (upper two) and three or more in the plenum region (lower six), are OPERABLE.
      • During periods of high steam generator blowdown, the main stack wide range noble gas monitor may be isolated for the duration of blowdown. In these cases, the nonitor must be returned to service for at least 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> at the end of each six day period to demonstrate operability.

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TABLE 3.23-1 (Continued) mil 0N STATEMENTS ACTION 1 -

With the number of OPERABLE accident monitoring instrumentation channels less than the Total Number of Channels shown in Table 3.23-1, restore the inoperable channel (s) to OPERABLE status within 7 days, 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 at least HOT SHUTOOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 2 -

With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Channels OPERABLE requirements of Table 3.23-1, 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 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 at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 3 - With the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirements, return one char:.el to operable status within 7 days, or else prepare and submit Special Report to the Commission pursuant to S)ecification 6.9.2 within the next 10 days outlining:

(le cause of the malfunction, the plans for restoring the nanivel to OPERABLE status, and any alternative methods for estimating s;ack release rates during the interim.

ACTION 4 -

With the number OPERABLE channels less than the Total Number of Channels shown in Table 3.23-1, either restore the inoperable channel (s) to OPERABLE status within 7 days if repairs are feasible without shutting down or prepare and submit a Special report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status and c1 ternate methods in effect for estimating the applicable parameter in the interim.

ACTION 5 -

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 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> if repairs are feasible without shutting down or;

a. Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status; and 3-46a j
b. Restore the system to OPERABLE status at the next scheduled refueling.

ACTION 6 -

With the number of OPERABLE accident monitoring instrumentation channels less than the MINIMUM CHANNELS OPERABLE requirements of Table 3.23-1, 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 submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 drys outlining the cause of the malfunction, the plans f' restoring the channel (s) to OPERABLE status, and any asternate methods in affect for estimating the applicable parameter during the interim.

ACTION 7 - With less than the minimum channel (s) operable, 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 else establish alternate means to determine if significant fuel failure exists. If still inoperable after 7 days, preoare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining: the cause of the inoperability, the plans for restoring operability, and the alternate means established.

ACTION 8 - With the number of channels operable less than the Minimum Channels OPERABLE. Determine the subcooling margin once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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TABLE 3.23-2 .

ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REGUIREMENTS MODES FOR WHICH SURVEILLANCE INSTRUMENT CHANNEL CHECK CHANNEL CALIBRATIC9 IS REQUIRED

1. Containment Pressure D R 1,2,3
2. Reactor Coolant Cold Leg Temperature - M R 1,2,3 Wide Range
3. Reactor Coolant Hot Leg Temperature - Wide Rarge M R 1,2,3

{T' 4. Reactor Coolant Pressure - Wide Range D R 1,2,3,4

5. Containment Water Level - Wide Range M R 1,2,3
6. Core Exit Thermocouples M R** 1,2,3
7. Main Stack Wide Range Noble Gas Monitor D Q 1,2,3,4
8. Containment Atmosphere-High Range Radiation D R* 1,2,3,4 Monitor j l
9. Reactor Vessel Water Level M R** 1,2,3 .j
10. Reactor Coolant System Subcooling Margin M R 1,2,3 Monitor
  • CHANNEL CALIBRATION may consist of an electronic calibration of the channel not including the detector, for I range decades above 10R/hr and a one point calibration check of the detector below 10R/hr with an installed or i portable gamma source.
    • Electronic calibration from the ICC cabinets only.

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INSTRUMENTATION BASES 3,23 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 following an accident. This capability is consistent with the recommendations of degulatory Guide 1.97, Revision 3, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," May 1983 and NUREG-0737, "Clarification of THI Action Plan Requirements," November 1980.

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TABLE 3.9-2 ACCIDENT MONITORING INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS INSTRUMENT OF CHANNELS OPERABLE ACTION

1. Pressurizer Water Level 3 2 1
2. Auxiliary Feedwater Flow Rate 1/S. G. 1/S. G. 1 3- Blank u> 4. PORY Position Indicator 1 1 3

$; Acoustic Flow Monitor er

5. PORV Block Valve Position 1/ valve 1/ valve 3 Indicator
6. Safety Valve Position Indicator 1 1 3 Acoustic Flow Monitor t
  • Item 3 of this table is included in Table 3.23-1 (Item 10).

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TABLE 3.9-2 (CONTINVED)

ACTION 1 -

With the number of OPERABLE channels less than required by Table 3.9-2, either restore the inoperable channel (s) to OPERABLE status within 30 days or be in HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 2 -

Not used ACTION 3 -

With any individual valve position indicator inoperable, obtain quench tank temperature, level and pressure information, and monitor discharge pipe temperature once per shift to determine valve position. This action is not required if the PORY block valve is closed with power removed in accordance with Specification 3.3.C.(6).

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  • Table 4.2-1 (continued)

Channels Action Minimum Freauency

8. Variable Low Calibrate Each refueling Pressure trip set Check Each shift point calculator
9. Rod Position Calibrate Each refueling Digital Voltmeter Check with counters Every six inches of red 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 Hismatch Check Each shift
13. Charging Flow Calibrate Each refueling
14. Residual Heat Calibrate Each refueling Pump Flow
15. Boric Acid Calibrate Each refueling Tirik Level Check Each week
16. Refueling k'ater Calibrate Each refueling Storage Tank Test 90 days Level j 17. Volume Control Calibrate Each refueling Tank Level Test 90 days
18. *Bl ank
19. Radiation Calibrate Each refueling Monitoring Test Each day System
20. Boric Acid Calibrate Each refueling Control
21.
  • Blank
22. Valve Temperature Test Each refueling Interlocks J

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t TABLE 4.2-1 (CONTINVED)

Channel Action Minimum Frecuency

23. Pump Valve Interlock Check Each refueling
24. Reactor Coolant System Calibrate Each refueling OPS Test Each refueling
25. Auxiliary Feedwater Flow Rate Calibrate Each refueling Check Each month
26.
  • Blank
27. PORV Position Indication Calibrate Each refueling (Acoustic Monitor) Check Each month
28. PORV Block Valve Calibrate Each refueling Position Indication
29. Safety Valve Position Calibrate Each refueling Indication (Acoustic Monitor) Check Each month l

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

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@MINISTRATIVE CONTROLS SPECIAL REPORTS 6.9 ?. Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, one copy to the Regional Administrator, Region I, and one copy to the NRC Resident Inspector, within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

a. Inservice Inspection results, Specifications (4.10) and (4.12),
b. Primary Containment Leak Rate Results, Specification (4.4).
c. Reactor Vessel Material Surveillance Specimen Examination, Specification (4.10).
d. Steam Generator Tube Report Following each intervice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission within 15 days.
e. Post-Accident Instrumentation Operability, Specification (3.23 A and B).
f. Fire Protection Systems Operability, Specification (3.22).
g. Reactor Coolant System Vents, Specification (3.3.5.1)
h. Radiological Effluent Reports required by Specifications (7.1.1.2, 7.1.2.2, 7.1.2.3 and 7.1.3) 6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five years:
a. Records and logs of facility operation covering the time interval at each power level,
b. Records and 1 cts of principal maintenar,ce activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.
c. All reportable events,
d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications.

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ADMINISTRATIVE CONTROLS 6.17 RADIOLOGICAL EFFLUENT MONITORING AND OFFSITE DOSE CALCULATION MANUAL (REM 00CM)

Section I, Radiological Effluents Monitoring Manual, shall outline the sampling and analysis programs to determine the concentration of radioactive materials released offsite as well as dose commitments to individuals in those exposure pathways and for those radionuclides released as a result of station operation. It shall also specify operating guidelines for radioactive waste treatment systems and report content.

Changes to Section I shall be submitted to the Commission for approval prior to implementation.

Section II, the Offsite Dose Calculation Manual (0DCM), shall describe the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculations of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints consistent with the applicable LC0's contained in these technical specifications.

Changes to Section II need not be submitted to the Commission for approval prior to implementation, but shall be included in the next Semi-Annual Radioactive Effluent Release Report.

6.18 RADI0 ACTIVE WASTE TREATMENT SYSTEMS Procedures for liquid and gaseous radioactive effluent discharges from the Unit shall be prepared, approved, maintained and adhered to for all operations involving offsite releases of radioactive effluents. These procedures shall specify the use of appropriate waste treatment systems utilizing the guidance provided in the REM 0DCM.

The solid radioactive waste treatment system shall be operated in accordance with the Process Control Program to process wet radioactive wastes to meet shipping and burial ground requirements.

6.19 PASS /Samolino and Analysis of Plant Effluents The licensee shall implement and maintain a program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and contairment atmosphere samples under accident conditions. This program shall include the following:

a. Training of personnel 6-22

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ADMINISTRATIVE CONTROLS

b. Procedures for sv pling and analysis, and
c. Provisions for maintenance of sampling and analysis equipment.

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Docket No. 50-211 A06214 Attachment 2 Haddam Neck Plant Description of Individual Proposed Changes to Technical Specifications and Discussion on Significant Hazards Consideration Generic Letter 83 (NUREG-0737) 1 I

July 1988

Attachment 2 A06214/Page 1 Generic letter 83-37/NUREG-0737 Related Technical Soecification Chances The format of this attachment follows that of Generic Letter 83-37 and addresses each item listed therein. For each section, the changes proposed by CYAPC0 are compared to those recommended by the NRC Staff, the differences are identified and justification is provided. In addition, in accordance with 10CFR50.92, CYAPC0 has reviewed the proposed changes and has concluded that  ;

they do not involve a significant hazards consideration. The basis for the no significant hazards consideration conclusion is provided below.

It is noted that the proposed Technical Specification changes discussed below are similar to those approved by the NRC on Millstone Unit No. 2 (Docket No. 50-336). (Reference (1)).

(1) Reactor Coolant System (RCS) Vents (Item II.B.1)

NRC Recommendation At least one reactor coolant system (RCS) vent path (consisting of at least two valves in series which are powered from emergency buses), shall i be operable and closed at all times (except for cold shutdown and refuel-ing) at each of the following locations:

a. Reactor Vessel Head
b. Pressurizer steam space
c. Reactor coolant system high point If one of the vent paths becomes inoperable, restore it to service within l 30 days or be in hot standby within 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 />. Two or more vent paths inoperable require restoration of at least 2 of the paths 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 I standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 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 />, i

Surveillance is required every 18 months by verifying all manual valves are locked open, cycling remotely operated valves and verifying flow through the system.

CYAPC0 Position 1

The proposed Technical Specification Section 3.3.5.1 requires that vent i paths be operable from both the reactor vessel head and the pressurizer I a

steam space. The Haddam Neck Plant does not have a vent path that corresponds with "c" Reactor Coolant System High Point Vent as listed in i Generic Letter 83-37. The ACTION statement specifies that if the '

pressurizer vent path is inoperable, remove power from the valves in that I

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Attachment 2 '

A06214/Page 2 path and ensure that at least one PORV path is operable. It also  !

reqdres to restore either the PORV or the pressurizer vent path to operable status within 30 days or submit a special report to the Commission outlining cause and correction. If the reactor vessel head vent path is inoperable, the ACTION statement requires that it be restored to operability within 30 days or submit a special report to the Commission outlining both cause and correction. A proposed change to Section 6.9.2 reflects the requirement for these reports. Surveillance requirements proposed by CYAPC0 are the same as those recommended in '

Generic Letter 83-37.

Justification for Deviation The Haddam Neck Plant RCS vents system provides venting capability from high points of the pressurizer and the reactor vessel head. The noncondensible gases, steem and/or liquids vented from either the pressurizer or the reactor vessel head are discharged to a common sparger located between the No. 2 and No. 3 containment air recirculation fans in the containment outer annulus. The RCS vents from the pressurizer and

. from the reactor vessel head each contain a parallel set of two solenoid-operated valves in series which are remotely controlled from the main control room.

The proposed Technical Specification requires only one train of two valves to be operable. This is acceptable since the RCS vents are not required for any design basis events. Also, this is consistent with NRC recommendations. The ACTION statement ensures that one of the two RCS i vents will be restored to operable status within 30 days or a Special Report is to be submitted to the NRC. The. Special Report ensures NRC cognizance of the malfunction and plans for restoration. CYAPC0 consid-ers that these actions are acceptable. Restricting plant operation due to an inoperable RCS vent is not warranted since the system is not required for any design basis event. l 1

The surveillance interval (18 months) is acceptable since a smaller interval would require valve stroking at power which is not advisable due to the potential for inadvertent actuation. The surveillances are specified to be performed during cold shutdown or refueling, thereby a eliminating the potential of causing an unnecessary plant transient.

The surveillance requirement to verify flow through the RCS vent system serves as a redundant verification of valve position and indication.

Test connections exist to allow f43 Vent System testing while the system is isolated from the RCS. Thus, flow testing the RCS vents is feasible.

The NRC Staff has approved of the RCS Vent System at the Haddam Neck Plant as described h the safety evaluation dated September 6, 1983 (Reference (2)).

... .- j l

l Attachment 2 A06214/Page 3 Sianificant Hazards Consideration )

In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed changes and has concluded that they do not involve a significant hazards consid-eration. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed changes do not involve a significant hazards consideration because the changes would not 1

1. Involve a significant increase in the probability or. consequences of 1 an accident previously evaluated. The RCS vent system is not required for any design basis events. Also, this is consistent with the NRC recommendation 3. The system is intended to increase the plant's ability to deal with beyond design basis events. Since the proposed changes provide greater assurance that this system will be i available to the operators, there is no increase in probability or l consequences of an accident previously analyzed.
2. Create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed changes provide additional controls and restrictions over the system already in service at the plant.
3. Involve a significant reduction in a n'argin of safety. The proposed l changes will ensure the RCS vent system is maintained in an operable. l condition and will require appropriate corrective actions if not.

On this bases, the margin of safety of any Technical Specification is not reduced.

l (2) Post Accident Samoling (Item II.B.3) )

l HRC Recommendation Licensees should ensure that their plant has the capability to obtain and analyze reactor coolant and containment atmosphere samples under accident conditions. An administrative program should be established, impleanted and maintained ts ensure this capability. The program should incluue: j a) training of personnel b) procedures for sampling and analysis, and c) provisions for maintenance of sampling and analysis equipment It is acceptable to the Staff, if the licensee elects to reference this program in the administrative controls section of the Technical Specifi-cations and include a detailed description of the program in the plant operation manuals. A copy of the program should be easily available to the operating staff during accident and transient conditions.

  • ' .o Attachment 2.

A06214/Page 4 CYAPC0 Position Proposed Technical Specification 6.19 provides a requirement to establish and maintain a program to sample the Rcactor Coolant System and the containnent atmosphere in accordance with the guidance provided in Generic Letter 83-37. CYAPC0 has proposed Technical Specifications that conform to Generic Letter 83-37. The post accident sampling system for the Haddam Neck Plant was approved by the NRC as described in a safety evaluation dated June 5, 1984 (Reference (3)).

Sionificant Hazards Consideration In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed changes and has concluded that they do not involve a significant hazards consid-eration. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed changes do not involve a significant hazards consideration because the changes would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated. Since the proposed changes provide greater assurance that the post accident sampling system will be available to the operators, there is no increase in proba-bility or consequences of an accident previously analyzed.
2. Create the possibility of a new or different kind of accident from any accident previously evaluated. Presently, the reactor coolant and containment air post accident sampling system are operationally tested periodically, semi-annually at a minimum to ensure system availability. Therefore, the proposed changes formalize existing CYAPC0 practices and do not impose any new or previously unevaluated requirements.
3. Involve a significant reduction in a margin of safety. The proposed changes help ensure that the post accident sampling system is available for use by operating personnel. These proposed changes represent additions to the Technical Specifications. On this basis, the margin of safety of any Technical Specification is not reduced.

(3) Lona Term Auxiliary Feedwater System Evaluation (item II.E.1.1)

NRC Recommendation The objective of this item is to improve the reliability and performance l of the auxiliary feedwater (AFW) system. Technical Specifications depend on the results of the licensee's evaluation and staff review of each plant. The limiting conditions of operation (LCOs) and surveillance requirements for the AFW system should be similar to safety-related systems. Typical generic Technical Specifications are provided in the

e Attachment 2 A06214/Page 5 enclosure (i.e., to Generic Letter 83-37). These specifications are for a plant which has three auxiliary feedwater pumps. Plant specific Technical Specifications could be established by using the generic Technical Specifications for the AFW system.

CYAPC0 Position The existing Technical Specifications (Section 4.8) conform to the recommendation of Generic letter 83-37 considering two steam driven auxiliary feedwater pumps. One difference between Generic letter 83-37 and the existing Technical Specifications for the Haddam Neck Plant is in the area of surveillance requirements. Generic Letter 83-37 specifies that when tested, the auxiliary feedwater pumps should be verified to deliver a particular flow at a particular discharge pressure. In the existing Technical Specifications, the discharge pressure is verified on minimum recirculation rather than looking at a specific pressure and fl ow.

Justification for Deviation CYAPC0 has determined that the existing Technical Specifications are appropriate and adequate. The lack of a specified recirculation flow rate is acceptable since the auxiliary feedwater pumps are incorporated in the Inservice Test (IST) program. The pumps are tested quarterly and any significant decrease in pump performance would be corrected. Accord-ingly, the above deviation is acceptable.

(4) Noble Gas Effluent Monitors (II.F.1.1)

!!RC Recommendation Noble gas effluent monitors provide information, during and following an accident, which are considered helpfe! to the operator in aasessing the '

plant condition. It is desired that these monitors be operable at all times during plant operation, but they are not required for safe shtedown of the plant. In case of fail ire of the monitor, appropriate actions should be taken to restore it; operational capability in a reasonable period of time. Considering the importance of the availability of the equipment and possible delays involved in administrative controls, 7 days is considered to be the appropria % lime period to restore the operabili-ty of the monitor. An alterna..? method for monitoring the effluent should be initiated as soon as practical, but no later than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the identification of the failure of the monitor. If the monitor is not restored to operable conditions within 7 days after the failure a special report should be submitted to the NRC within 14 days following the event, outlining the cause of inoperability, actions taken and the planned schedule for restoring the system to operable status.

Attachment 2 A06214/Page 6 CYAPC0 Position The proposed Technical Specification Table 3.23-1 requires the main stack high range radiation monitor to be operable during Modes 1, -2, 3, and 4 and the ACTION statement proposed substantially conforms to the guidance of Generic Letter 83-37. A note to Table 3.23-1 has been added to clarify that the monitors may be isolated, but still maintained operable, during periods of high steam generator blowdown. The proposed Technical Specification Table 3.23-1 does not include an alarm trip setpoint requirement presently included in the existing Technical Specification Section 3.23. This is included in the plant's surveillance procedure and is controlled administrative 1y. Surveillance requirements call for a daily channel check and a channel calibration quarterly. A proposed change to Technical Specification Section 6.9.2 reflects the requirement for a Special Report included in the ACTION statement.

CYAPC0 has proposed Technical Specifications that substantially conform to Generic letter 83-37.

Sianificant Hazards Consideration Please refer to item (10).

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

See SeP 3n (2), Post Accident Sampling System (PASS). The Technical Specifications being proposed by CYAPC0 to address sampling and analysis of plant effluents are the same as those proposed to address the PASS.

Therefore, the writeup for Item (2) fully applies here. This is in conformance with the guidance of Generic Letter 83-37.

(6) .C_q.ntainment Hiah-Ranae Radiation Monitor (II.F.1.3)

NRC Recommendation A minimum og two in-conpainment radiation-level monitors with a maximum range of 10 rad /hr (10 R/hr for photon only) should be operable at all times except for cold shutdown and refueling outages. In case of failure of the ionitor, appropriate actions should be taken to restore its operational capability as soon as possible. If the monitor is not restored to operable condition within 7 days after the failure, a special report should be submitted to the NRC within 14 days following the event, outlining the cause of inoperability, actions taken and the planned schedule for restoring the equipment to operable status.

Typical surveillance requirements are shown in Enclosure 3 of Generic Letter 83-37. The setpoint for the high radiation level alarm should be determined such that spurious alarms will be precluded. Note that the l

Attachment 2 A06214/Page 7 .)

acceptable calibration techniques for these monitors are discussed in NUREG-0737.

CYAPC0 Position The pronosed Technical Specification Table 3.23-1 requires containment high-rt .ge radiation monitors to be operable during Modes 1, 2, 3, and 4 and the ACTION statement proposed therein is more conservative than recommended in Generic Letter 83-37. The ACTION statement requires that in the event one (or both) channels are inoperable, alternate means be established within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to determine if significant fuel failure exists. Further, if the channel (s) are still inoperable after 7 days, prepare and submit a Special Report to the Commission pursuant to Section 6.9.2 within the next 10 days. The proposed Tech'iical Specification Table 3.23-1 de not include an alarm trip setpoint requirement present-ly included in the existing Technical Specification Section 3.23. This is included in the plant's surveillance procedure and is controlled administratively. Surveillance requirements call for a daily channel check and a channel calibration during refueling outages. There is table notation that says higher ranges are calibrated electronically.

CYAPC0 has proposed Technical Specifications that substantially conform to Generic Letter 83-37.

Sianificant Hazards Consideration In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed changes and has concluded that they do not involve a significant hazards consid-eration. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed changes do not involve a significant hazards consideration because they would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The containment high range L radiation monitor is credited in the Emergency Operating Procedures.

These monitors are needed to decide whether to use high-head recir-culation via HPSI or the charging fill header. The proposed changes require that in the event one (or both) channels are inoperable, alternate means be established within 48-hours to determine if significant fuel failure exists. This is in lieu of the present administrative control for this instrumentation (ACTION requirement contained in Administrative Technical Specifications 3.23). There are alternate means that could be implemented that would give the operator sufficient information to choose HPSI or charging high-head recirculation. Since no other alternative is given in the Action Statement, inability to establish the alternative means would require a shutdown. Therefore, the proposed changes are equivalent to the current (administrative) requirements for these monitors and

,.o ,.

Attachment 2 A06214/Page 8 there would be no impact on the consequences of any design basis accident.

2. Create the possibility of a new or different kind of accident from any previously analysed. Again, due to the nature of these changes, no new accidents are created and also these proposed changes would not affect the plant response to the point where a different accident results.
3. Involve a significant reduction in a margin of safety. The changes-have no impact on the consequences of an accident or any of the protective boundaries. Therefore, there is no reduction in any margin of safety.

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

NRC Recommendation Containment pressure should be continuously indicated in the control room of each operating reactor during Power Operation, Start-up and Hot Standby modes of operation. Two channels should be operable at all times when the reactor is operating in any of the above mentioned modes.

Technical Specifications for these monitors should be included with other accident monitoring instrumentation in the present Technical Specifica-tions. Limiting conditions for operation (including the required Actions) for the containment pressure monitor should be similar to other tecident monitoring instrumentation included in the present Technical Specifications. Typical acceptable LC0 and surveillance requirements for accident monitoring instrumentation are included in Enclosure 3 of Generic Letter 83-37.

(8) C_pntainment Water Level Monitor (II.F.1.5)

NRC Recommendation A continuous indication of containment water level should be provided in the control room of each reactor during Power Operation, Start-up and Hot Standby modes of operation. At least one channel for narrow range and two channels for wide range instruments should be operable at all times when the reactor is operating in any of the above modes. Narrow Range instruments should cover the range from the bottom to the top of the containment sump. Wide range instruments should cover the range from the bottom of the containment to the elevation equivalent to a 600,000 gallon (or less if justified) capacity.

Technical Specifications for containment water level monitors should be included with other accident monitoring instrumentation in the present Technical Specifications. LCOs (including the required Actions) for wide range monitors should be similar to other accident monitoring

1 l

l Attachment 2  !

A06214/Page 9 l l

instrumentation included in the present Technical Specifications. LCOs for narrow range monitor should include the requirement that the inopera-ble channel will be restored to operable status within 30 days or the plant will be brought to Hot Shutdown condition as required for other accident monitoring instrumentation. Typical acceptable LC0 and surveil- ,

l lance requirements for accident monitoring instrumentation are included i in Enclosure 3. l CYAPC0 Position i The proposed Technical Specification Table 3.23-1 ("Accident Monitoring l Instrumentation") includes the following:

Total 1 Required Minimum  !

Channels Operable Containment Pressure 2 1 ,

Containment Water Level, Wide Range 2 1

)

The ACTION statement for Containment Pressure and Containment Water Level (wide range) specifies that if the total channels operable are less than l the total number of channels shown in Table 3.23-1, either restore the i inoperable channel within 7 days or submit a Special Report to the j

. Commission outlining the cause, plans for restoration and any alternate '

l methods to be employed in the interim for estimating these parameters.

If the total channels operable are less t'lan the minimum channels opera-ble, either restore the inoperable channels within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or submit a l special report to the Commission outlining the cause, plans for restora-tion and any alternate methods to be employed in the interim for estima-ting these parameters. A proposed change to Section 6.9.2 reflects the requirement for this report. The Containment Water Level (narrow range) is not specified in the Technical Specifications. The surveillance requirements call for a monthly channel check and a calibration every refueling outage for the containment water level. The surveillance l requirements call for a daily check and a calibration every refueling j outage for the containment pressure. These requarements are consistent i with the existing Technical Specification Table 4.2-1. It is noted that  !

item 5 of the proposed Table 3.23-1 will replace item 21 of the existing Technical Specification Table 4.2-1 and item 1 of the proposed Table 3.23-1 will replace item 18 of the existing Technical Specification 4.2-1.

Justification for Deviation l

Both the basic LCOs and the surveillance requirements conform to those recommended by Generic Letter 83-37. However the ac?. ion statements

, differ. It 's the position of CYADC0 that the action statements proposed are much with the actual function of the more commensurate l

Attachment 2 M6214/Page 10 instrumentation involved. Neither the containment pressure nor the containment water level monitoring systems are required for the safe shutdown of the Haddam Neck Plant for any design basis accident.

Therefore, a shutdown requirement is deemed by CYAPC0 to be inappropri-ate. However, since these instruments, as with others covered in this amendment request, do enhance the ability of the operator to monitor and assess these variables during and following an accident, actions in response to inoperable instrumentation have been proposed.

The Haddam Neck Plant has installed a narrow range containment water level monitor system in the sump in containment. This sump water level sensor is not environmentally qualified. CYAPC0 does not plan to install a new qualified sump water level sensor. The staff accepted this position in Reference (4).

Sianificant Hazards Consideration Please refer to item (10)

(9) Containment Hydroaen Monitor (Item II.F.1.6)

The need for continuous Hydrogen monitors is being evaluated i r. the Haddam Neck Plant Integrated Safety Assessment Program. No specifica-tions are being proposed for this item.

(10) Instrumentation for Detection of Inadeauate Core Coolina (Item II.F.2)

NRC Recommendation Subcooling margin monitors, core exit thermocouples, and a reactor coolant inventory tracking system (e.g., differential pressure measure-ment system designed by Westinghouse, Heated Junction Thermocouple System designed by Combustion Engineering, etc.) may be used to provide indica-tion of the approach to, existence of, and recovery from inadequate core cooling (ICC). These instrumentation should be operable during Power Operation, Start-up, and Hot Standby modes of operation for each reactor.

Subcooling margin monitors should have already been included in the present Technical Specifications. Technical Specifications for core exit Thermocouples and the reactor coolant inventory tracking system should be included with other accident monitoring instrumentation in the present Technical Specifications. Four core-exit thermocouples in each core  ;

quadrant and two channels in the reactor coolant tracking system are required to be operable when the reactor is operating in any of the above i mentioned modes. Minimum of two core-exit thermocouples in each quadrant  !

and one channel in the reactor coolant tracking system should be operable  !

at all times when the reactor is operating in any of the above mentioned i modes. Typical acceptable LC0 and surveillance requirements for accident j

i

. * .- l 1

Attachment 2 A06214/Page 11

- monitoring instrumentation are provided in Enclosure 3 of Generic Letter 83-37.

CYAPC0 Position j Subcooling margin monitors, core exit thermocouples and reactor vessel water level instrumentation included in the proposed Technical Specifica-tions Table 3.23-1 conform to Generic Letter 83-37. This instrumentation is required to be operable during Modes 1, 2, and 3, as required by Generic Letter 83-37.

ACTION Statement 4 for the core exit thermocouples specifies that with the number of operable channels less than the total number of channels, restore the inoperable channel (s) to operable status within 7 days. If repairs are feasible without shutting down, or prepare a special report within the next 10 days outlining the cause of the inoperability and the plans and schedule for remtoring the system to operable status and alternate methods in effect for estimating this parameter in the interim.

ACTION Statement 5 for the reactor vessel coolant level is based on the sample ICC Technical Specifications transmitted to the NRC Staff in a letter from the Combustion Engineering Owner's Group, dated February 19, 1985 (Reference (5)). The ACTION statement proposed by CYAPC0 specifies that with the number of operable channels less than the minimum channels operable, either restore the inoperable channel? within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or:

1) submit a special report to the Commission within 30 days of the event outlining the cause and plans for restoration and
2) restore the system to operable status during the next refueling outage. 1 The only difference between the CYAPC0 proposal and those in Reference (5) is the statement regarding "initiating an alternate method of monitoring the reactor vessel inventory." CYAPC0 did not include this in the action statement since no specific procedure could be written to give the operators clear "step-by-step" guidance on how to meet this step. Though there are parameters by which vessel level can be inferred generally, this could not be realistically proceduralized. A proposed change to section 6.9.2 reflects the requirement to submit special I

reports to the NRC Staff.

4 ACTION Statement 8 for the subcooling margin n.onitor specifies that with the number of channels operable less than the minimum channels operable, determine the subcooling margin once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The surveillance requirement specified for the subcooling margin monitors are consistent with Generic Letter 83-37. It should be noted that item 10 of the proposed Table 3.23-1 will replace item 3 of Table 3.9-2 of the existing

n.

Attachment 2 A06214/Page 12 Technical Specifications and Item 5 of Table 3.23-2 will replace item 26 of Table 4.2-1 of the existing Technical Specifications.

Surveillance requirements proposed in Table 3.23-2 include a monthly channel check and a channel calibration during refueling outages. This calibration is an electronic calibration from the ICC cabinets only, since authentic stimulation of the Core Exit Thermocouples or the Reactor Vessel Coolant Level Monitoring System cannot be conducted due to their physical location and range.

Justification for Deviation The ACTION . statement proposed by CYAPC0 for the reactor vessel coolant level is very similar to that proposed by CE Owner's Group as previously described and is appropriate since this system is not required for safe shutdown. CYAPC0 did not propose an action statement for subcooling margin monitor and core exit thermocouple monitors to require a shutdown as proposed by Generic Letter 83-37. CYAPC0 believes that a shutdown requirement is inappropriate f:r this instrumentation since it is not required for the safe shutdown of the unit.

Sionificant Hazards Consideration In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed changes and has concluded that they do not involve a significant hazards consid-eration. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed changes do not involve a significant hazards consideration because the changas would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated. Since the proposed changes provide additional reporting requirements and greater assurance that instrumentation will be available to the operators, there is no increase in the probability or consequences of an accident previous-ly analyzed.
2. Create the possibility of a new or differert kind of accident from any previously analyzed. Again, due to the nature of these changes as described above, no new accidents are created. These proposed changes provide additional controls and restrictions over systems already in service at the plant.
3. Involve a significant reduction in a margin of safety. The proposed changes have no impact on the consequences of an accident or any of the protective boundaries. These prooosed changes will ensure the accident monitoring instrumentation is maintained in an operable condition and will require appropriate corrective actions if not.

Therefore, there is no reduction in any margin of safety.

Attachment 2 A06214/Page 13 It is noted that the above write-up regarding determination of signifi-cant hazards consideration is also applicable to items 4 (Noble Gas Effluent Monitors), 7 (Containment Pressure Monitors) and 8 (Containment Water Level Monitors).

(11) Control Room Habit =bility (111.D.3.4)

Control Room Habitability requirements are being evaluated in the Haddam Neck Plant Integrated Safety Assessment Program. No specifications are being proposed for this item.

(12) Proposed Chanaes to Ter.hnical Sogcification Table 3 9-2 Editorial changes have been proposed to line items 4 and 6 of Technical Specification Table 3.9-2. These changes will clarify the actual total number of channels and minimum channels operable for the PORV position indicator (Acoustic flow monitor) and safety valve pocition indicator (Acoustic flow monitor). As these changes are editorial in nature, they do not involve a significant hazards consideration.

..b-Attachment 2 A06214/Page 14

References:

(1) D. H..Jaffe letter to E. J. Mroczka, Issuance of Amendment (TAC #s 54546 and 54399), dated September 28, 1987.

(2) D . fl. Crutchfield letter to W. G. Counsil, NUREG-0737, Item II.B.1, Reactor Coolant System Vent, dated September 6, 1983.

(3) D. M. Crutchfield. letter to W. G. Counsil, Post-Accident Sampling System (NUREG-0737, Item II.B.3.), dated June 5, 1984.

(4) J. A. Zwolinski letter to J. F. Opeka, Containment Pressure and Water Level Monitors. TMI Action Items II.F.1.4 and II.F.1.5, July 9, 1985.

(5) R. W. Wells letter to Hugh L. Thompson, dated February 19, 1985, "Technical Specifications for the Reactor Vessel level Monitoring System."

1

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