Text

Proposed Tech Specs Sections 2.0 & 3/4.3 Re Safety Limits & Limiting Safety Sys Settings & Instrumentation
	ML20245J470
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	06/23/1989
From:	; CONNECTICUT YANKEE ATOMIC POWER CO.
To:	;
Shared Package
ML20245J455	List: B13269, Proposed Tech Specs Sections 2.0 & 3/4.3 Re Safety Limits & Limiting Safety Sys Settings & Instrumentation; ML20245J450;
References
	B13269, NUDOCS 8907030057
	Download: ML20245J470 (106)
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i Docket No. 50-213 B13269 l

Attachment 1 Haddam Neck Plant Proposed Revised Technical Specifications i

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ni 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop inlet temperature (T -

shown in Figures 2.1-1 and 2.1-2 for four ah8 Yhree loop operation,3 ) shall not excee respectively.-

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop inlet temperature (T 3 and pressurizer pressure has exceeded the appropriate percent 69 Ya)ted thermal power (RTP) line, be in HOT STANDBY within I hour, and comply with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig. I APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

1 MODES I and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.7.1.

MODES 3, 4, and 5:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to be within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1.

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. SAFETY !.IMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2. LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS ,

2.2.1 The Reactor Trip System Instrumentation and Interlocks Setpoints !

shall be set consistent with the Trip Setpoint values shown in Table 2.2-1. i f

APPLICABILITY: As shown for each channel in Table 3.3-1.

ACTION:

4 With a Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its Trip Setpoint adjusted consistent with the Trip Setpoint value.

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, t 2,.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE I The restrictions of this Safely Limit prevent. overheating of the fuel and possible cladding perforation which would result in the release of fission l products to the reactor coolant. Overheating of the fuel cladding is I prevented by restricting fuel operation to within the nucleate boiling .j regime where the heat transfer coefficient is large and the cladding surface 1 temperature is slightly above the coolant saturation temperature.

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Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction it. heat i transfer coefficient. DNB is not a directly measurable parameter during '

operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the W-3 correlation. The W-3 DNB correlation has been developed to predict the DNB flux .ind the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that j would cause DNB at a particular core location to the local heat flux, and is indicative of the margin to DNB.

The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limHed to 1.30. This value bounds a 95% probability at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curves of figures 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, pressurizer pressure and core inlet temperature for which the minimum I DNBR is no less than 1.30, and the core outlet void fraction is no greater than 0.32.

N of 1 These curves are based on total enthalpy hot channel factors, FAna1Yo,wance.60 is and 1.64 for four and three lgop operation, respectively.

included for an increase in F H at reduced power.

These limiting hot channel factors are higher than those calculated at full power for the range from all control rods fully withdrawn to maximum allowable control rod insertion. This insertion limit is described in Specification 3.1.3.6 and shown in Figures 3.1-1 and 3.1-2. The required reduction in power level as dictated by Figures 3.1-1 and 3.1-2 insures that the DNB ratio is always greater than 1.30.

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2.1. 2 - REACTOR C00LANT' SYSTEM PRESSURE- '

The restriction of this Safety Limit protects the integrity of the Reactor d Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment-atmosphere.

The reactor. vessel and pressurizer are' designed to ASME Boiler and Pressure Ves'sel Code,Section VIII; ASME Special Ruling No.1270N and 1273N, and the I

Tentative Structural Design Basis for Reactor Pressure Vessel and directly.

-associated Components - PB151987. A maximum transient pressure of 110%

(2735 psig)' of design pressure is allowed by PB151987. The Reactor Coolant System piping, valves and fittings, are designed to ASA B 31.112H Edition, which permits a maximum transient pressure of 120% (2985 psig) of component design pressure. The Safety Limit of 2735 psig is therefore consistent with

..the design criteria and associated Code requirements.

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. .' 2,2 -LIMITING SAFETY SYSTEM SETTINGS 1

BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 have been selected to ensure that the reactor core and Reactor Coolant System are prevented f l from exceeding their acceptance criteria during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System.in mitigating the consequences of ,

accidents. Operation with a trip set less conservative than its Trip i Setpoint but within its specified Allowable Value is acceptable on the basis i

that the difference between each Trip setpoint and the Allowable Value is equal to or less than a drift allowance accounted for in the design basis t analysis. !

The various Reactor trip circuits automatically open the Reactor trip :

breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level. The Reactor Trip System initiates a Turbine trip signal whenever Reactor trip is initiated. This prevents the reactivity insertion that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System.

Manual Reactor Trio The Reactor Trip System includes manual Reactor trip capability.

Power Ranae. Neutron Flux The Power Range Neutron Flux trip provides core protection against rapid reactivity excursions. In order to provide protection over the entire operating range, the trip function has three (3) different setpoints. The trip is credited in the following design basis events; steam line breck, control rod ejection, excess steam flow, control rod withdrawal and isolated loop startup. In order to reduce the time to trip for certain accidents occuring at low power, the overpower setpoint is lowered to 23 percent when reactor power is below 10 percent. This low overpower is below 10 percent.

This low overpower trip would terminate the postulated large steamline break accident from the hot zero power condition. The lower setting for three ;

loop operation provides protection at the reduced power level equivalent to '

that provided by the setting for four loop operation at full power.

Intermediate and Source Ranae. Neutron Flux i

The Intermediate Range, Neutron Flux, High Positive Rate trip provides core protection during reactor startup. This trip function provides protection for large reactivity insertion events initiated from a subcritical mode of i operation. This trip function is credited in the rod withdrawal from subcritical analysis.

HADDAM NECK B2-3 l

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, c LIMITING SAFETY SYSTEM SETTINGS BASES Pressurizer Pressure The Variable Low Pressurizer Pressure trip protects the core against DNB or excessive core exit quality resulting from either uncontrolled slow reactivity insertions which cause Reactor Coolant System temperature and pressure to increase or a loss of RCS pressure. The formula for the '

Variable Low Pressure Trip Setpoint, which is based on reactor coolant ,

temperature rise (Delta T) and Reactor Coolant System average temperature,. '

defines a minimum allowable pressure for operation which is continually compared to pressurizer pressure. A Reactor trip occurs when the minimum I allowable calculated pressure exceeds the pressurizer pressure. The l Variable Low Pressurizer Pressure trip is credited in the uncontrolled rod withdrawal and steam generator tube rupture analyses. i The Pressurizer High Pressure trip, in conjunction with safety valves, protects the Reactor Coolant System against system overpressure. This trip is credited in the loss of load and turbine trip analyses.

Pressurizer Water level !

The Pressurizer High Water Level trip is provided to prevent water relief through the pressurizer Code safety valves. This trip provides redundant protection to the Pressurizer High Pressure trip in the loss of load, and turbine trip analyses.

Reactor Coolant Flow 1 The Reactor Coolant low Flow trip protects the core against DNB resulting from a reduction in coolant flow while the reactor is at substantial power.

This trip is credited in the partial and total loss of flow, locked rotor, i and sheared shaft analyses. Loss-of-flow protection is also provided by i Reactor Coolant Pump Breaker trip and from Undervoltage trip on a reactor '

coolant pump motor bus. Credit was not taken in the design basis analyses for operation of the latter two trips but their functional capability enhances the overall reliability of the Reactor Trip System.

Steam Flow The High Steam Flow Trip provides protection against a large increase in steam flow by closing the main steam line trip valves and tripping the reactor. This trip is credited in the steam line break and excess steam flow analyses.

Steam /Feedwater Flow Mismatch and Steam Generator Water Level The Steam /Feedwater Flow Mismatch in coincidence with a Steam Generator Water Level - Low trip protects the RCS against an abrupt loss of secondary heat sinks by initiating a reactor trip prior to steam generator dryout.

The trip is credited in the loss of feedwater analyses.

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,- /- LIMITING SAFETY SYSTEM SETTINGS-L . s-Undervoltaae - Reactor Coolant Pumo Buses A Reactor trip is generated on low voltage on either reactor coolant pump bus. . The trip provides protection for certain loss of flow events.

Safety In.iection Input from ESI The ESF automatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a Safety Injection. The ESF instrumentation .

channels which initiate a Safety Injection signal are shown in Table 3.3-2.

This logic acts as a redundant trip to the Pressurizer Pressure low trip.

Reactor Coolant Pumo Breaker A Reactor trip from an opening of the reactor coolant pump breaker provides protection from loss of flow in any reactor coolant loop due to power failure.

Main Steam line Trio Valve Closure A Reactor trip on MSTV closure anticipates the pressure and flux transients which could follow MSTV closure and thereby protects reactor vessel pressure and fuei thermal / hydraulic Safety Limits.

Reactor Trio System Interlocks The Reactor Trip System interlocks perform the following functions:

P-7 On increasing power, P-7 automatically enables Reactor trips on low flow in more than one reactor coolant loop, reactor coolant pump bus undervoltage, more than one reactor coolant pump breaker open, main steam line isolation valve closure, Turbine-trip, and variable low pressure. On decreasing power, the above listed trips are automatically blocked.

P.-7N On increasing oower, P-7N automatically blocks the intermediate range, neutron flux, nigh startup rate trip. On decreasing power, P-7N automatically enables the intermediate range, neutron flux, high startup rate trip.

P-8 On increasing power, P-8 automatically enables Reactor trips on low ,

flow in one or more reactor coolant loops, and one or more reactor I coolant pump breakers open. On decreasing power, the P-8 automatically blocks the above listed trips. '

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. .. . .; s 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION i

.3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table' 3.3-1 shall be OPERABLE.

APPLICABILITY: As shown in Table 3.3-1.

MILO!(: As shown in Table 3.3-1.

1 @ EILLANCE REQUIREMENTS-4.3.1.1 Each Reactor Trip System instrumentation channel and interlock logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements.specified in Table 4.3-1.

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TABLE 3.3-1 (Continued 1 TABLE NOTATION

- With the Reactor Trip-System breakers in the clo' sed position and the Control Rod Drive System. capable of rod withdrawal. !

- The low flow channel associated with trip functions derived.from the out-of-service reactor coolant _ loop shall be in the tripped condition.

'*** With the Reactor Trip System breakers in the open position and the Control Rod Drive System not capable of rod withdrawal.

L # The provisions of Specification 3.0.4 are not applicable.

(a) -THERMAL POWER above 10% of RATED THERMAL POWER.

(b) THERMAL POWER 1 74% of RATED THERMAL POWER.

(c) THERMAL POWER above 10% but below 74% of RATED THERMAL POWER.

ACTION STATEMENTS ACTION 1:

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement', restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

. ACTION 2:

With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , i

I HAoDAn Nea sia s.S

l

\

TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued)

ACTION 2: (Continued)

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.3.1.1.

ACTION 3:

a. With less than the Minimum Number of Channels OPERABLE, within I hour.

determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition or apply Specification 3.0.3.

b. With turbine first stage pressure inoperable, continued power operation may proceed provided the permissive is placed in the more conservative state for existing plant conditions.

ACTION 4:

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes and restore the inoperable channel to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or open/ verify open the Reactor Trip System breakers within the next hour.

ACTION 5:

a. With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, STARTUP and/or POWER OPERATION may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the affected portion of the inoperable channel is placed in the tripped condition within I hour. The inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.3.1.1.

i e

HADDAM NECK 3/4 3-6 4

7 l ,

TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued)

ACTION 6:

With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST prc,ided the inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, the inoperable channel may be bypassed up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for surveillance testing of other channels per Specification 4.3.1.1.

ACTION 7:

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, reduce THERMAL POWER to below 74% of RATED THERMAL POWER (P-8) within I hour and place the inoperable channel in the trip position within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

ACTION 8:

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, reduce THERMAL POWER to below 10% of RATED THERMAL '

POWER (P-7) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 9:

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, STARTUP and/or POWER OPERATION may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided that the inoperable channel is placed in the tripped condition within I hour.

ACTION 10:

With the number cf OPERABLE channels one less than the Minimum Channels OPERABLE requirement for Modes 3, 4, 5, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Reactor Trip System breakers l within the next hour.

ACTION 11:

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement for Modes 3, 4, 5 be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

HADDAM NECK 3/4 3-7

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TABLE 4.3-1 (Continued)

TABLE NOTATIONS

With the Reactor Trip System breakers in the closed position and the Control Rod Drive System capable of rod withdrawal.

(a) THERMAL POWER above 10% of RATED THERMAL POWER.

(b) THERMAL POWER 2 74% of RATED THERMAL POWER.

(c) THERMAL POWER above 10% but below 74% of RATED THERMAL POWER.

(d) THERMAL POWER below 10% of RATED THERMAL POWER.

(1) If not performed in previous 31 days.

(2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%. The provisions of Specification 4.0.4. are not applicable for entry into MODES 1 or 2.

This requirement is not applicable when the Power Range Channels have had their gains adjusted to maintain the 9% trip margin for steady state conditions at power levels other than 16%, 65%, and 100% RATED THERMAL POWER. When this exception is used,.a heat balance calculation j will continue to be performed on a daily basis to determine core power, i and the power range channels will be verified daily to be 9% below the l selected overpower trip setpoint. 1 (3) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(4) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the 4 Reactor Trip System breakers.

(5) Following a refueling outage, the calibration is performed subsequent to the plant reaching RTP. The provisions of Specification 4.0.4 are not applicable.

(6) If not performed in previous 7 days.

(7) Each scheduled shutdown if not tested or calibrated in preceding 6 months.

HADDAM NECK 3/4 3-11 i

l < <

1 INSTRUMENTATION 3/4.3.7 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels and interlock shown in Table 3.3-2 shall be OPERABLE with their Trip Setroints set consistent with the values shown in the Trip Setpoint column of Table 3.3-3.

APPLICABILITY: As shown in Table 3.3-2.

ACTION:

a. With an ESFAS instrumentation channel or interlock Trip Setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-3, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-2 until the channel !

is restored to OPERABLE status with the Trip Setpoint adjusted '

consistent with the Trip Setpoint value.

b. With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-2.

SURVEILLANCE RE0VIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock logic shall be demonstrated OPERABLE by the performance of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2.

i l

HADDAM NECK 3/4 3-12

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TABLE 3.3-2 (Continued)

TABLE NOTATIONS

The provisions of Specification 3.0.4 are not applicable.

- Trip function may be bypassed in this MODE when RCS pressure is less than 1800 psig.

- - The channel (s) associated with the protective functions derived from the i out-of-service reactor coolant loop shall be placed in the tripped mode, i (a) THERMAL POWER above 10% of RATED THERMAL POWER.

(b) For Surveillance Testing purposes, the train being tested will be placed in " Defeat" function.

ACTION STATEMENTS ACTION 20 -

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 21 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, operation may proceed until performance of the next required ANALOG CHANNEL '

OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within I hour.

ACTION 22 -

With a channel associated with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least H0T STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 23 -

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable I channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

I l

l HADDAM NECK 3/4 3-16

_____ -___ ___ i

4 .

n -

\

TABLE 3.3-2 (Continued)-

ACTION STATEMENTS (Continued)

ACTION 24, -

With the number!of OPERABLE channels one less than the Total Number of Channels, STARTUP'and/or POWER.0PERATION may

. proceed provided the following conditions are satisfied:

a. The inoperable channel .is placed 'in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and

b. The Minimum Channels 10PERABLE_ requirement is met; however, one additional channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing.of other. channels per Specification'4.3.2.1.

' ACTION 25 -

With less than the Minimum Number of Channels ~0PERABLE, within l' hour determine by observation of the associated permissive annunciate.' window (s) that the interlock is in its required state for the existing plant. condition, or apply Specification 3.0.3.

ACTION 26 -

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or reduce the THERMAL POWER to below 10% of RATED THERMAL POWER within the following I hour.

i 4

HADDAM NECK 3/4 3-17 i

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l, ~ INSTRUMENTATION.

3/4 3.3 MONITORING INSTRUMENTATION ~

L'

RADIATION MONITORING FOR PLANT OPERATIONS !

l litilJING CONDITION FOR OPERAT' ION i

'3.3.3.1 The radiation monitoring instrumentation channels for plant ]

operations shown in-Table 3.3-4 shall be OPERABLE. j APPLICABILITY: As shown in' Table 3.3-4.

ACTION:

a.. With .no radiation monitoring channels for plant operations

~ OPERABLE, take the ACTION shown in Table 3.3-4. 1 L

.b.. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. !

SURVEILLANCE RE0VIREMENTS' "4.3.3.1 ' Each radiation monitoring instrumentation channel for plant I operations shall be demonstrated OPERABLE- by the performance' of the CHANNEL' CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST for the MODES and at the frequencies shown in Table 4.3-3.

1 I

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HADDAM NECK 3/4 3-23

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,. INSTRUMENTATION tiOVABLE INCORE DETECTORS LIMITING CONDITION FOR OPERATION 3.3.3.2 The Movable Incore Detector System shall be OPERABLE with:

a. At least 11 of 13 northwest quadrant detector thimbles or at least 10 of 13 northwest quadrant detector thimbles if the inoperable locations are M8, M12, and N8.

b. At least one set of four quadrant symmetric thimbles, and

c. Sufficient movable detectors, drive, and readout equipment to map these thimbles.

APPLICABILITY: When the Movable Incore Detector System is used for:

a. Recalibration of the Excore Neutron Flux Detector System, or

b. Monitoring the QUADRANT POWER TILT RATIO, or

c. Measurement of FAH and the LINEAR HEAT GENERATION RATE.

ACTION:

.a. With the Moveable Incore Detector System inoperable due to less than the minimum required number of detector thimbles as required in 3.3.3.2a or b, penalty factors shall be applied when measuring F N LINEARHEATGENERATIONRATEorQUADRANTPOWERTILTRATIO;ordur$g recalibration of the Movable Incore Detector System, as appropriate.

b. With the Movable Incore Detector System inoperable, due to insufficient movable detectors, drives or readout equipment, do not use the system for the above applicable monitoring or calibration functions. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE0VIREMENTS 4.3.3.2 The Movable Incore Detector System shall be demonstrated OPERABLE by verifying an acceptable voltage plateau for the incore detector (s) at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when used as required by the above Applicability requirements.

HADDAM NECK 3/4 3-26 L

l _--..______J

IN'STRUMENTATION SEISMIC INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.3 The seismic monitoring system instrumentation shown in Table 3.3-5 shall be OPERABLE. l APPLICABILITY: At all times.

AGJ1QH:

a. With the seismic monitoring system inoperable for more than 30 .

days, 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 malfunction and the plans for restoring the system to OPERABLE status.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE0VIREMENTS 4.3.3.3.1 The above seismic monitoring system shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION'and

- ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-4.

4.3.3.3.2 The above required seismic monitoring system actuated during a seismic event shall be restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and a CHANNEL CALIBRATION performed within 10 days following the seismic event.

Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground motion. If it is determined that the magnitude of the event exceeded the Operating Basis Earthquake, then a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 14 days describing the magnitude, frequency spectrum, and resultant effect upon facility features important to safety.

i HADDAM NECK 3/4 3-27 L ___ _ <

o .

.~,.. .

. TABLE 3.3-5 l.

l SEISMIC MONITORING INSTRUMENTATION l ~

MINIMUM INSTRUMENTS L

INSTRUMENTS AND SENSOR LOCATIONS MEASUREMENT' RANGE OPERABLE-1.- Triaxial Servo Accelerometer 0 to 0.59 1

(SSA-302) Basemat-Cable Vault

2. Digital Cassette Accelerograph 5 Volts 1 (DCA-300)**

3. Response' Spectrum Analyzer i 5 Volts -1 (RSA-50)**

4. Playback System 5 Volts 1 (SMR-102)**

5. Seismic Warning Panel N/A l'

'(SWP-300)**

- All located in the Control Room i

llADDAM NECK 3/43-28

c- ,

-o; ,,

s TABLE 4.3-4 3 q

SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE RE0VIREMENTS ANALOG l CHANNEL CHANNEL CHANNEL OPERATIONAL INSTRUMENTS AND SENSOR LOCATIONS CHECK ' CALIBRATION- TEST ,

1.. Triaxial Servo Accelerometer M R SA

~

(SSA-302) Basemat-Cable Vault Digital Cassette Accelerograph

2. M R SA (DCA-300)**

3. Response Spectru:n Analyzer M R SA (RSA-50)**

l- 4. Playback System M R' SA ,

- (SMR-102)** l

5. Seismic Warning Panel M R SA I (SWP-300)**

- All located in the Control Room l

1 i

HADDAM NECK 3/4 3-29

. . j

~

INSTRUMENTATION'

'.. }

METEOROLOGICAL INSTRUMENTATION- 1 I

LIMITING CONDITION FOR' OPERATION 1 l

3.3.3.4 The meteorological monitoring instrumentation channels shown in Table 3.3-6 shall-be OPERABLE. j 1

APPLICABILITY: . At all times. !

l ACTION:

a. With one or more required meteorological monitoring channel's inoperable for more than 7 days, 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' malfunction and the plans for restoring the channel (s) to OPERABLE status.

b. .The' provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE0VIREMENTS 4.3.3.4 Each of the above meteorological monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK and CHANNEL CALIBRATION.at the frequencies shown in Table 4.3-5.

l HADDAM NECK 3/4 3-30 L_ _ _ _ _

v -. : ..

. . TABLE-3.3-6 MEfEOR0 LOGICAL MONITORING INSTRUMENTATION MINIMUM INSTRUMENT / LOCATION OPERABLE

1. Wind Speed

a. Baseline Elev. 33' 1

b. Nominal Elev. 200' .1

'2. Wind Direction

a. Baseline Elev. 33' 1-

b. . Nominal Elev.196' 1

3. Air. Temperature

a. Baseline Elev. 33' - 1

4. Delta T*

a. Nominal Elev.120' 1

b. Nominal Elev. 200' 1

' Delta T is the air temperature of the nominal elevation minus the air temperature at the 33' baseline elevation.

i HADDAM NECK 3/4 3-31 L______________--. !

,'; +: -e

~'

TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION i" . SURVEILLANCE REQUIREMENTS INSTRUMENT / LOCATION -CHANNEL CHECK CHANNEL CALIBRATION ~

l'

,1.~ . Wind Speed'

. a. . Baseline Elev. 33'- D SA.

b. Nominal Elev. 200'- D SA

'2. Wind Direction

a. Baseline Elev. 33' D SA

b. Nominal Elev.196' D' SA

3. Air Temperature

a. Baseline' Elev. 33' D 15A

.4. Delta T* i

a. Nominal ' Elev. 120' D SA

b. Nominal Elev. 200' D SA

Delta T is the air temperature of the nominal elevation minus the air temperature at the 33' baseline elevation.

1 HADDAM NECK 3/4 3-32

, [ INSTRUMENTATION.

ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION l

3.3.3.5 The accident monitoring instrumentation channels shown in Table 3.3-7 shall be OPERABLE.

APPLICABILITY: As 'shown in Table 3.3-7.

ACTION:

I a. As shown in Table '3.3-7.

b. .The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.5' Each accident' monitoring ' instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION at the frequencies shown in Table 4.3-6.

)

HADDAM NECK 3/4 3-33 i

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TABLE 3.3-7 (Continued) f . .

, ACTION STATEMENTS I l I ACTION 35 -

With the number of OPERABLE accident monitoring !

instrumentation channels less than the Total Number of Channels shown in Table 3.3-7, restore the inoperable ;

channel (s) to OPERABLE status within 7 days, or be in at i least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT I SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 36 - With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Channels OPERABLE requirements of Table 3.3-7c restore the inoperable channel (s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 37 -

With the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirements, return one channel to operable status within 7 days, or else 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 malfunction, the plans for restoring the channel to OPERABLE status, and a preplanned alternative method for estimating stack release rates durir.g the interim.

ACTION 38 - With the number of OPERABLE channels less than the Total Number of Channels shown in Table 3.3-7, 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 alternate methods in effect for estimating the applicable parameter in the interim.

ACTION 39 -

With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements, either restore the inoperable channels (s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months j 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

b. Restore the system to OPERABLE status at the next scheduled refueling.

HADDAM NECK 3/4 3-36

1 TABLE 3.3-7 iContinued) l 1j ACTION STATEMENTS ]

')

ACTION 40 - With'the number of channels OPERABLE less than the MINIMUM CHANNELS OPERABLE, determine the subcooling margin once per l 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Restore the system to OPERABLE status at the next -'

scheduled refueling.

ACTION 41 -

With any individual valve position indicator inoperable (a l block valve position indicator or the acoustic flow monitor),

obtain quench tank temperature, level and pressure information and monitor discharge pipe temperature once per l shift to determine valve pcsition. For the case of an [

inoperable block valve position indicator this action is not required if the PORV block valve is closed with power removed !

in accordance with Specification 3.4.4. !

ACTION 42 -

With the number of OPERABLE channels less than the MINIMUM CHANNELS OPERABLE requirements of Table 3.3-7, 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 43 -

With the number of OPERABLE accident monitoring instrumentation channels less than the MINIMUM CHANNELS OPERABLE requirements of Table 3.3-7, restore' the inoperable channel (s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , 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, the plans for restoring the channel (s) to OPERABLE status, and any alternate methods in effect for estimating the applicable parameter during the interim.

ACTION 44 - With less than the minimum channel (s) operable, restore the inoperable channel (s) to operable status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or else establish alternate means to determine if significant fuel failure exists. If still inoperable after 7 days, 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, the plans for restoring operability, and the alternate means established.:

ACTION 45 - With the number of channels operable less than the minimum channels OPERABLE requirement of Table 3.3-7, restore the inoperable channel (s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

1 i

j HADDAM NECK 3/4 3-37 l

i

OHUS MWSI 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

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- N N 3 O N I A

4 T H

A C D M M D M M M M M W M M M M M E T L N B E A M T U R r T o S t N i I n o r G e M o N g t I n e l n i R a g e i n O - R n v g o T - e a e r M I e g w R L a N r e n o M r w O u r a r e r l o o e M t u R r d e e g t l g a t a i t v n a F n T r a e N W a e i c a N e r d W L l i c R E p e i - - e o d i D m p W k n t o n t e I e m l l n o a c I s d C T e - e e a i R b u i C T v v T t u n o W A g e e e e u w S o c e g r l r L L e l o i A l l e u e u g o l m t e L s v s r r a S F e i Y v e d s e s e e r t s R e r l t e L e t t o k r s o O L u o o r r a a t n e y P P s C H P r P W W S a t S r s e T a e d e e t t t t r r r r w t v n t r n n n a o o o e x d n l a a P ae ae a W t t t t i e a a W l g l g l a a a a M e l V e t on on o r r r r W F o v t n oa oa o e e e e d o k l n e CR CR C z n n n g i y C c a e m i e e e n c r o V m n re re r r G G G i A a r l n i od od o u l i o B y i T a ti ti t s m m m e c l t t a N t cW cW c s a a a u i r

i c V e t E n a a a e e e e f x a R f n M o e e e r t t t e o u e O a o U C R R R P S S S R B A R P S C R

T S . . . . .

N . . . . . . . . . 0 1 2 3 4 5 I 1 2 3 4 5 6 7 8 9 1 1 1 1 1 1 gEg R *' T'g

OHUS te MWSI 1 1 1 1 cl el ta

.et ds n

ei N h O .t n I a T g A .nh

' R it

- B di S

I L l uw .

T A cr d

- N C nh e E i/ m M L R r E E t0 o R N o1 f I N *

n r U A *

w e p

Q H * *

l o

- E C R Q R R el R - '

ne 'e nb b E a C hr t N co s A t u L ec m L K he

) I C tt t d E E e s e V H fd e u R C o t n U -

e -

i S L nh l t E ot a n N N i n o O N tf. o C I A ao i

-( T H r t A C M D D M' bn c

- 6 T io n

- N li u 3 E r at f M o ca .

4 U t r y y R i cb l l E T n ii r n L S o nl e o B N M oa t A I - rc r s T n t a t G o ct u e N i en q n I t li i R r a eop a b O o i a T t d n ,

c I i a ae K N n R n C C O o fo E C M M e o H I g .a C T s n t e N a a sd L h E G R i n E t D sa N I e h n N m C l g l or A o C b i e ch . H r A o H v /e C f s N - e yR c e e L a0r y n l e r m1 u l o p g e r o h i u n h e Nes t t o a p t Ov n a c R s a I oa o r o o W Tbm m b m e m Aam i r d t l R a o l e i A e Bsg t a h W s I e c T - t s Lde n k n e Aal o c t c e V Ccb i i i a m ea t n x t n r Ldt i o E S i o E r d r T a t Neo d t N e n t c Ngp a c E r i n a An e M o a o e Har n l U C M C R Cro I E R

T S . . . .

N 7 8 9 0 *

I 1 1 1 2 * *

g3 2{ '

u2 iO

. IN'STRUMENTATION i

FIRE DETECTION INSTRUMENTATION i LIMITING CONDITION FOR OPERATION 3.3.3.6 As a minimum, the fire detection instrumentation for each fire detection zone shown in Table 3.3-8 shall be OPERABLE.

APPLICABILITY: Whenever systems, structures, components, or equipment protected by the fire detection instrumentation are required to be OPERABLE.

ACTION:

a. With any, but no less than the minimum required fire detection instruments shown in Table 3.3-8 inoperable, restore the inoperable instrument (s) to OPERABLE status within 14 days or within the next I hour establish a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, unless the instrument (s) is located inside the containment, then inspect that containment zone at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (or monitor the containment air temperature at least once per hour at the locations listed in Specification 4.6.1.6).

b. With less than the minimum required fire detection instruments in any. fire zone shown in Table 3.3-8 operable, within I hour establish a continuous fire watch to inspect the zone (s) with the inoperable instrument (s), unless the instrument (s) is located inside the containment, then inspect that containment zone at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (or monitor the containment air temperature at least once per hour at the locations listed in Specification 4.6.1.6).

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE0VIREMENTS 4.3.3.6.1 Each of the above required fire detection instruments which are accessible during plant operation shall be demonstrated OPERABLE at least once per 6 months by performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST. Fire detectors which are not accessible during plant operation shall be demonstrated OPERABLE by the performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST during each COLD SHUTDOWN exceeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless performed in the previous 6 months. Detectors which cannot be reset are not l required to be demonstrated OPERABLE by performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST.

HADDAM NECK 3/4 3-40

) ': 4I ._,-i

-j 4

..c'-4 , ,.,

1

. SURVEILLANCE-REQUIREMENTS j

.'4.3.3.'6.2' Supervised circuits' associated with the detec' tor l alarms of each a of the above required fire detection instruments-shall be demonstrated ~i

_ OPERABLE at least once per 6 months.. ~

1

'4.3.3.6.3. Nonsupervised circuits, associated with detector' alarms, between-the instrument and the control-room shall-be demonstrated OPERABLE at least once per 31 days.

1 HADDAM NECK 3/4 3-41

TABLE 3.3-8 FIRE DETECTION SYSTEMS Minimum Number Minimum Number Smoke Detectors Heat Detectors OPERABLE / Detectors OPERABLE / Detectors location Available Available

1. Containment (R-3) 19/22

24/32 **

. (Inaccessible)

2. Cable Spr'eading Area (S-3A) 21/28 *

3. IA Diesel Generator Room (D-1) 4/5

4. IB Diesel Generator Room (D-2) 4/5

5. Switchgear Room (S-2) 35/35

6. Containment Cable Vault (R .1) 3/4

7. Waste Disposal Bldg. (W-I) 2/3

8. Auxiliary Feedwater Pump Room (R-2) 1/2

9. Primary Auxiliary Bldg. ,

Entrance to Corridor, West End 1/2 !

(A-1A) .

Main Corridor and Boric Acid Area 3/4 (A-1A)

East End of Corridor.(A-1A) 1/1 Northeast End of Corridor (A-1A)'

1/1 Drumming Room (A-11) 2/3 Ventilation Equipment Area (A-IN) 2/3 1 Store Room (A-IP) 1/2 RHR Pump Room Cubicles A and B 1/1 (A-lE,A-IF)

10. Control Room (S-1A, Not including 13/18*

kitchen)

11. Screen Well Bldg.

Pump Motor Room (P-1A, P-1B) 8/9***

Hypoclorite Storage Room (P-IC) 1/1

12. Spent Fuel Bldg. (F-1) 5/6

13. PAB Charcoal Filter Bank Heat Detector 7/7 (A-IN) (outlet detectors only)

I l

No two adjacent detectors shall be inoperable at the same time.

- Minimum of 6 heat detectors per reactor coolant pump must be OPERABLE.

- - 0ne detector inoperable cannot be one of the three existing detectors on the upper level of the Screen Well Building.

HADDAM NECK 3/4 3-42 I

4

L- y s*'

.c a

-e

-TABLE 3.3-8 .j FIRE DETECTION SYSTEMS j L' Minimum Number Minimum Number Smoke Detectors . Heat Detectors 10PERABLE/ Detectors OPERABLE / Detectors lq Location Available Available

.14. High pressure turbine deluge-(T-IF) -

4/4 ]

15. Hydrogen real oil reservoir deluge -

1/1 (T-ID)-

16. Turbine lube oil reservoir deluge (T-18) -

'7/7

17. Turbine building mezzanine under -

4/4 generator (T-1F)

18. Turbine building cranewell deluge (T-IC) - 6/6 j

HADDAM NECK 3/4 3-43

..

4, INSTRUMENTATION RADI0 ACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.7 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-9 shall be OPERABLE with applicable Alarm / Trip '

Setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The Alarm / Trip Setpoints shall be determined in accordance with methodology and parameters described in the 0FFSITE DOSE CALCULATION MANUAL (0DCM).

APPLICABILITY: At all times *.

ACTION:

a. Witl. a radioactive liquid effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above specification, without delay suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable, or change the Alarm / Trip Setpoint so it is acceptably conservative.

b. With the number of channels less than the minimum channels operable requirement, take the ACTION shown in Table 3.3-9. Exert best efforts to restore the inoperable monitor to OPERABLE status within 30 days, and, if unsuccessful, explain in the next Semi-annual Effluent Report why the inoperability was not corrected in a timely manner. Releases need not be terminated after 30 days provided the specified actions are continued.

c. The provisions of 3 specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE0VIREMENTS 4.3.3.7.1 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and ANALOG CHANNEL OPERATIONAL TEST operations at the frequencies shown in Table 4.3-7. ;

I

At all times means that channel shall be OPERABLE and in service on a continuous, uninterrupted basis, except that outages are permitted for ,

a maximum of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months each time for the purpose of maintenance and I performance of required tests, checks, calibrations or sampling.

i l

HADDAM NECK 3/4 3-44

d e

s o

p o

r p .

y f_ l o n N o O n N I . . o s O T A. A.

i w I C 0 1 2 3 t o T A 4 4 4 4 N N e l A l f T p N m g E o n M c i U n R n i T o m S p r N u e I t e e G v d N i I t f R c o O e 9 T f e .

- I f s s 3 N e o e O p v 3 M # e r r E m u u

- E T ML o p c L N UB c B E MA e e p A U IR b h m T L NE

t u F IP *

l p F MO 1 1 1 1 *

l r E i o f w f o D .

I tn s e U no e s O ei v u I G mt r L N ea u e I rn c h E D ii t V

I G

N I

V um qr e

v y T I O ee l b C D R rt a A I P e v d 0 V n nc e I OE k TE e k i wi f n D RS n OS n n L ot o i A PA a NA i a da m R E T n E L T n wm e r SL w SL w oo s e RE t o RE t t o lt u t OR s d OR n s d bu e T e w T e e w a e d IF T o IF u T o f h NO l NO l l oe t s O e B O f e B d i MN l MN f T l ni y O c r O E N c r l ov b w YI ye o YI E ye o a io o TT cn t TT r M cn t n tr d l IA ei a IA e E ei a a ap e f VN RL r VN t R RL r C n n II e II a U e io i l TM de n TM W S de n e mt m a CR ng e CR A ng e g r r n AE ar G AE e E ar G r en e a 0T a 0T c M a a to t c I eh m*

ae I i eh m h i e DC tc DC v E t c a c ct d e AI ss en AI r T ss e s ia g RT ai ti RT e A ai t i tc s r A WD SL A S R WD S D ai i a T SM SM mf h N SO SO W oi w c E OT OT O td o s M RU . . RU . L . . . uo l i U GA a b GA a F a b c Am F D R

T S

N . . . *

I 1 2 3 * * *

=

hg wh

. TABLE 3.3-9 (Continued)

ACTION STATEMENTS ACTION 40 - With the number of channels OPERABLE less than required by l the Minimum Channels OPERABLE requirements, effluent releases may continue provided that best efforts are made to repair 1 the instrument and that prior to initiating a release: .

4'

a. At least two independent samples of the tank to be discharged are analyzed in accordance with Specification 4.11.1.1.1, and; l

b. The original release rate calculations and discharge valving are independently verified by a second individual.

i ACTIP -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument and that grab samples are analyzed for gross radioactivity (beta or gamma) at a lower limit of detection of at least 3 x 10-7 microcuries/ml;

1. Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the specific activity of the secondary coolant is greater than 0.01 microcuries/gm DOSE EQUIVALENT I-131.

2. Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of the secondary coolant is less than or equal to 0.01 microcuries/gm DOSE EQUIVALENT I-131.

ACTION 42 - With the numNr of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this rathway e may continue provided that best efforts are made to repair the instrument and that once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months grab samples of the service water effluent are collected and analyzed for gross radioactivity (beta or gamma) at a lower limit of detection of at least 3 x 10'7 microcuries/ml.

ACTION 43 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument and that the flow rate is estimated once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump performance curves generated insitu may be used to estimate fl ow.

HADDAM NECK 3/4 3-46 1

_______ _ D

~

d L

e s

A o N p LO o EI r S NT p T NCT ) ) ) . . .

N ANS 3 3 3 a E HUE ( ( ( A. A. A.

M CFT Q Q Q N N N f E o R

I n

U o O N i E O t R I e T l E LA p C ER m N NB o A NI ) ) ) . . c L AL 2 2 2 L HA ( ( ( A. A. n I CC R R R R N N o E p V u R

U e S v i

N t O c I

T E e

f A CK f T RC . . .- e N UE E OH A. A. A. e 7 M SC P M M N N N m

- U o 3 R c T e 4 S b N

E I l

L l B G i A N L w T I E R NK t O NC ) ) ) ) ) ) n T AE 1 1 1 1 5 4 e I HH ( ( ( ( ( ( m N CC 0 D D D 0 D e O r M ii T

G N

u q

N I I e E D D r U I I L V VC n F O OI w F RN RT k o E PO PA n d D ST I

n SO M a n

w RA T o I

U w RT w lt ON t o OU t t o b O TI s d TA n s d I IM e w I e e w f L NR Te o NG u T o o E

OE n l ONE l l MT ei B MIS f T e B n V lL DA f N l o I YC c r YIE E E c r l i T TI ye o TVL M ye o a t C IT cg t IOE r E cn t n a A VA er a VRR e R ei a a n 0 IM Ra r IP t U RL r C i I TO h e T F a S e m D CT dc n CTO W A d e n e r A AU ns e AO E ng e g e R 0A ai G 0NN e M ar G r t I E D I O c a a DDS ANA e m*

ae DTI i E eh m h c tk AUT ve T t c ae c i RAE sn en RBA rn A ss en s t L aa ti N ei R ai ti i a T SME WT SL SMI SL WD SL D m N SRR SRH W o E OA OAP O t M RLF . . RLE . L . . . u U GAO a b GAT a F a B c A R

T S

N . . .

I I 2 3

5Eg wS w "

j -.

l, .;. .

, TABLE 4.3-7 (Continued)

TABLE NOTATION g

(1) CHANNEL CHECK need only be performed daily when discharges are made L from this pathway. The CHANNEL CHECK should be done when the discharge-is in process.

(2) CHANNEL CALIBRATION shall be performed using a known radioactive liquid or solid source whose strength is determined by a detector which has l been calibrated to an NBS source. The radioactive source shall be in a l known, reproducible geometry.

(3) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:-

1. Instrument indicates measured levels above the alarm / trip setpoint*.

2. Instrument indicates a downscale failure or circuit failure.

3. Instrument controls not set in operate mode.

(4) Pump status should be checked at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for the purpose of determining flow rate.

(5) Blowdown throttle valve position should be checked daily when.

discharges are being made via this pathway.

Automatic isolation shall also be demonstrated annually for the test tank discharge monitor line and steam generator blowdown line.

l l

HADDAM NECK 3/4 3-48

_--__ _ J

a .

INSTRUMENTATION RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE with applicable Alarm / Trip Setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded. The setpoints shall be determined in accordance with the methodology and parameters as described in the ODCM.

APPLICABILITY: At all times *.

ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above Specification, without delay suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the Alarm / Trip Setpoint so it is acceptably conservative.

b. With the number of channels less than the minimum channels operable requirement, take the ACTION shown in Table 3.3-10.

Exert best efforts to restore the inoperable monitor to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semi-annual Effluent Report why the inoperability was not corrected in a timely manner. Releases need not be terminated after 30 days provided the specified actions are continued.

1

c. The provisions er Specifications 3.0.3 and 3.0.4 are not ,

applicable. '

SURVEILLANCE RE0VIREMENTS 4.3.3.8.1 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and ANAL 0G CHANNEL OPERATIONAL TEST operations at the frequencies shown in Table 4.3-8.

At all times means that the channel shall be OPERABLE and in service on ;

a continuous basis, except that outages are permitted for a maximum of '

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months each time for the purpose of maintenance and performance of required tests, checks, calibrations.

HADDAM NECK 3/4 3-49

N O

.I T

C 5 6 6 7 7 N A 4 4 4 4 4 O '

I T

A T

N E

M U

R T

S : ' .

N I

G N

I R

O T

0 I 1 N SE

- O MLL 3 M UEB MNA 3 T INR N NAE E E IHP .

L U MCO 1 1 1 1 1 B L A F T F E s e

S s U a O e E l S e A R G

E m

e V ct I i s T t y C ras A om r O tos o I it a t D nuG i A oA n R M e o yns dt r M e

_ taa l e i W p t vm r a m e a

_ irf t R tao e S a cl l R w AAn p e o o m t w l s gi a a o F

_ ant S l l Gia u Fr r

_ dn e c o e

_ K eii n i kt l

_ C l vm i t ci p A b or d r an m T ore o a t o a S NPT P T

N E

N I

I SM S M A . . . . .

_ U M a b c d e R

T S

N .

I 1

_ gE$ 2Rn t' { d, O jl

.. g

. TABLE 3.3-10 (Continu'ed)-

ACTION STATEMENTS ACTION 45 - With the number of channels OPERABLE'1ess than required by the Minimum Channels OPERABLE requirement, releases via the-Waste Gas Holdup System may continue'provided that best efforts are made to repair-the' instrument and'that prior to initiating the release:

-(a)- For the tank to be discharged, at least two independent samples-'of. the tank's contents are analyzed; and, (b) The release rate calculations and discharge valve-lineups are _ independently verified by a second individual.

Otherwise, suspend releases from the Waste Gas Holdup System.:

Releases from all pathways other than the Waste Gas Holdup System may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for gross radioactivity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

-ACTION 46 - With the number of channels OPERABLE less than required by.

the Minimum Channels OPERABLE requirement, effluent releases via this pathway may' continue provided that best efforts are made to repair the instrument and that samples are continuously collected with auxiliary sampling equipment for- .

periods of seven (7) days and analyzed for principal gamma l emitters with half lives greater.than 8 days within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after the end of the sampling period. Auxiliary sampling shall be established within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of declaring the channel INOPERABLE.

ACTION 47 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that the best efforts are made to repair the instrument and that the flow rate is estimated once per 4' hours.

1 4

HADDAM NECK 3/4 3-51

S L T A N N E LO M EIT E NTS .

R NCE ) . . . .

I ANT 3 U HU ( A. A. A. A.

Q CF Q N N N N E

R E

C N N O A I L T L LA I ER E NB .

V NI ) . .

R AL 2 U HA ( A. A. -

_ . S CC R N N R' R N

O I

T A

T N

E M E 8 U CK

- R RC . . . .

3 T UE S OH A. A. A. A.

4 N I

SC M N N N N E

L G B N A I T R

~

O T

I L N E O NK M NC ) )

AE 1 1 T HH ( (

N CC D W W D D E -

U L

F F

E S

U O

E r S o r A t o G i r t n o i E o t n V M i o I r n M T y e o C t l p

M e A i t 0 v m e a I i r a t R D t e S a A c l R w R A p e o m t w l s a a o F a S l l G u F r K e c e C e n i k l A l i t c p T b d r a m S o o a t a T N I P S S N N E I M A . . . . .

U M a b c d e R

T S

N .

I 1

_5h$

m5 Y$

,! l r Ll

L .

l l- ,

TABLE 4.3-8'(Continued)-

TABLE NOTATION (I) CHANNEL CHECK daily when releases exist via this pathway. I (2) Calibration shall be performed using a known source whose strength is determined by a detector which has been calibrated to an NBS source.

These sources shall be in a known, reproducible geometry.

1 (3) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room 1 alarm annunciation occurs if any of the following conditions exist: j

a. Instrument indicates measured levels above the Alarm / Trip Setpoint*.

b. Instrument indicates a downscale failure or circuit failure.

c. Instrument controls not set in operate mode.

l Automatic isolation of the waste gas releases by the noble gas activity monitor should also be demonstrated.

HADDAM NECK 3/4 3-53

, ( .w

- - . - z.

Y- 2 <

IN'STRUMENTATION

/i?

'3/4'3;4 INTERNAL ~ FLOOD PROTECTION'

~ . LIMITING CONDITION FOR OPERATION c3.3.4; The. liquid level instrumentation ~ channels for flooding protection-

'shown in: Table 3.3-11'shall be OPERABLE.

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

ACTION:

a. WithlessthantheMinimumChannelsOPERABLE-foranyL11guidlevel-instrumentation Functional Unit, .take the ACTION shown in Table

!3.3-11.

b. ,The provisions of Specifications '3.0.3 and 3.0.4 are not -

l applicable. -

i SURVEILLANCE RE0VIREMENTS 4.3.4 Each l'i quid level instrumentation channel for flood. protection shall

'be demonstrated OPERABLE:

a. ' At;1 east once. per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by performance of.the ' CHANNEL- CHECK,

b. At least once per 6' months by performance of an' ANALOG CHANNEL.

OPERATIONAL' TEST and visually verifyingLno obstruction.-

-l HADDAM NECK 3/4 3-54 l= q l

I

T A

O LN FO

/I MT RC

_ AA -_ ' ~

3

.L 8 8 8 8 8 8 8 8 9 8 8 0 0 0 A 4 4 4 4 4 4 4 4 4 4 4 5 5 5 N

O I

T C

E T

O R

P G

N _SE I MLL D UEB O MNA O INR L NAE F IHP =

~

MCO l 1 1 1 1 1 1 1 l' 1 l 2 2 2 1 R ' -

1 O

- F 3

N 3 O I

E T L A B T "

A N T E M M U R R A T L S A N

I O -

~

T 1 1 1 1 1 1 1 1 1 1 1 I 1 1 L . .

. E V

E L

D I

U Q ) ) ) )

I " "

L ) 6 "6 ~

"6 6

) ' '

"6 '5 '5 a 1 1

' 1 1 e 2 2 5 ( ( r ) ( (

1 -

A

) ) ) ( e e .

"6 m m

' " " l l p p .

o o 9 3 2 e c c p m m '8 ) o o 1 ' l i i m u u ( R R

- 3 '4 c b b u P P '8

( 1 1 i u u P p ( B A

( ( b C C n n m e ) ) u n r o u e r r l t t C A B o) u i S s o o c "6 "6 s s i" t)t u t t i ' ' a e p p p t6 e" c m o a a

) b 1 1 E W m m m c R6e o H r r n u 2 2 u u u e '5 j o e e To Ii C ( ( e s

e s

P P P j1 n(

e '8 n t(I R l l

n e

Nt p t t a a g g 'g I a g e G G Ua m s s h h n n n e sey n w v u a e C C i i i yl nl t i n l l Le P E W r g g t c ece m e e e Al e e e r r ei dif m e s s NE R B B p p t a a fb nba u r e e O H A A i i e h h au ouS r c i i I r R P P P P M C C SC CC( D S D D To Co Nl . . . . .

UF . . . . . . . . . 0 1 2 3 4 F( 1 2 3 4 5 6 7 8 9 1 1 1 1 1 g@j gu "

l- # l p n .

l ..

., TABLE 3.3-11 (Continued)- J

-ACTION STATEMENTS-

? ,

q ACTION 48 -

' With no channels OPERABLE, restore an. inoperable channel (s) to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or within'the next I hour establish a liquid level watch patrol to inspect the zone (s) without an OPERABLE channel at least once.per hour.

. ACTION 49 - With no channels OPERABLE for Functional Unit 9, Safety Injection Pump Cubicle, the Minimum Channels OPERABLE requirement.for Functional Unit 10, Condensate Return Pump Cubicle, must be met or take the action specified in ACTION 48, above.

ACTION 50 -

With the. number of OPERABLE channels one less than.the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 7 days or within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months establish a liquid level watch patrol to inspect the zone (s) with the inoperable channel at least once per hour.

With no channels OPERABLE, take the action specified in ACTION 48, above.

1 i

HADDAM NECK 3/4 3-56

L 3 ,

. 1 3/4.3' INSTRUMENTATION e

BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION i The OPERABILITY of the Reactor Trip System and Engineered Safety Features Actuation System instrumentation and interlocks ensure that: (1) the associated ACTION and/or Reactor trip will be initiated when the parameter ,

monitored by each channel or combination thereof reaches its Setpoint, .(2) ;

the specified coincidence logic is maintained, (3) sufficient redundancy._is i maintained to permit a channel to be out-of-service for testing or >

maintenance, and (4) sufficient system functional capability'is available from diverse parameters.

The OPERABILITY of these systems is required to provide the overall' ,

reliability, redundancy, and diversity assumed available in the facility l design for the protection and n;itigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions.used in the safety analyses. The Surveillance ;

Requirements specified for these systems ensure that the overall system ;

functional capability is maintained comparable to the original design {

standards. The periodic surveillance tests performed at the minimum l frequencies are sufficient to demonstrate this capability. l The Engineered Safety Feature Actuation System Instrumentation Trip Setpoints specified in Table 3.3-3 are the nominal values at which the bistables are set for each functional unit. A Setpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band' allowed for calibration accuracy.

To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Setpoints have been specified in Table 3.3-3.

Operation with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error.

The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents, events, e d transients. Once the required logic combination is completed, l

HADDAM NECK B3/4 3-1

-......)

m

. s. ., !

. INSTRUMENTATION BASES l

\

I REACTOR TRIP SYSTEM INSTRUMENTATION AND ENGINEERED SAFETY FEATURES ACTUATTON 1 SYSTEM INSTRUMENTATION (Continued) the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the i condition. As an example, the following actions may be initiated by the i Engineered Safety Features Actuation System to mitigate the consequences of I a steam line break or loss-of-coolant accident: (1) Safety Injection pumps ;

start and automatic valves position, (2) Reactor trip, (3) startup of the emergency diesel generators, (4) containment isolation, (5) Turbine trip, )

(6) auxiliary feedwater pumps start and automatic valves position, {

(7) containment cooling fans start and automatic valves position, and (8) '

essential service water pumps start and automatic valves position.

3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING FOR PLANT OPERATIONS The OPERABILITY of the Containment Atmosphere Gaseous Radioactivity Monitoring System ensures that Gaseous Radioactivity Monitoring System will ;

monitor inside containment as a means to detect RCS leakage.

3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of 1 this system accurately represent the spatial neutron flux distribution of l the core. The OPERABILITY of this system is demonstrated by irradiating {

each detector used and determining the acceptability of its voltage curve. l 3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient I capability is available to determine the magnitude of a seismic event and j evaluate the response of those features important to safety. This l capability is required to permit comparison of the measured response to that I used in the design basis for the facility. '

3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that j sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release ,

of radioactive materials to the atmosphere. This capability is required to I evaluate the need for initiating protective measures to protect the health j and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972.

i HADDAM NECK B3/4 3-2 1

. INSTRUMENTATION BASES I

3/4.3.3.5 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 recoinmendations of Regulatory 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 TMI Action Plan Requirements," November 1980.

f f

1 r

HADDAM NECK B3/4 3-3

, INSTRUMENTATION BASES 3/4.3.3.6 FIRE DETECTION INSTRUMENTATION j i

The OPERABILITY of the fire detection instrumentation ensures that both adequate warning capability is available for prompt detection of fires and that Fire Suppression Systems, that are actuated by fire detectors, will i discharge extinguishing agents in a timely manner. Prompt detection and l suppression of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility Fire Protection Program.

Fire detectors that are used to actuate Fire Suppression Systems represent a more critically important component of a plant's Fire Protection Program than detectors that are installed solely for early fire warning and i notification. Consequently, the minimum number of OPERABLE fire detectors '

must be greater.

The loss of detection capability for Fire Suppression Systems, actuated by fire detectors, represents a significant degradation of fire protection for any area. As a result, the establishment of a fire watch patrol must be initiated at an earlier stage than would be warranted for the loss of detectors that provide only early fire warning. The establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

I 3/4.3.3.7 RADI0 ACTIVE LIOVID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and j control, as applicable, the releases of radioactive materials in liquid l effluents during actual or potential releases of liquid effluents. The '

Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that j the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. l The OPERABILITY and use of this instrumentation is consistent with the l requirements of General Design Criteria 60, 63, and 64 of Appendix A to '

10 CFR Part 50.

3/4.3.3.8 RADI0 ACTIVE GASE0VS EFFLUENT MONITORING INSTRUMENTATION i The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous ;

effluents during actual or potential releases of gaseous effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the REMODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part S0.

HADDAM NECK B3/4 3-4

i INSTRUMENTATION i BASES 1 3/4.3.4 FLOODING PROTECTION The liquid level instrumentation is provided to monitor liquid levels in areas of potential flooding caused by local pipe ruptures. The system ensures that early warning will occur so that protective action can be taken ,

l l in the event of a localized flooding condition in areas of the plant that house safety-related equipment. The loss of detection capability represents !

a degradation of flooding protection for any area. As a result, the !

establishment of a liquid level watch patrol must be initiated at an early j stage. The establishment of frequent liquid level watch patrols in the '

affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

i l

4 9

HADDAM NECK B3/4 3-5

e .

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h. '

Docket No. 50-213 l B13269 Attachment 2 Haddam Neck Plant Description of Individual Proposed Changes to the Technical Specifications and Discuss 4.on on the Significant Hazards Consideration Section 2.0 - Safety Limits and Limiting Safety System Settir.gs Section 3/4.3 - instrumentation June 1989 l _ _ - - _ _ _ _ _ _ _ - _ - -

^

Docket No. 50-213 B13269 I l 1

Attachment 2 Haddam Neck Plant Technical Specification Section 2.0, Safety Limits and Limiting Safety System Settings j l

4 June 1989 i

k i

I

_ _ _ _ _ _ - _ _ _ - _ - - - - _ l

I' '

l . :s

' Attachment 2

.- %. Section 2.0 m B13269/Page'l Technical Specification Section 2.0 Safety Limits and Limitina Safety System Settinas Section 2.1. Safety Limits The proposed revised Technical Specification (RTS) Section 2.1, Safety Limits, has been prepared by converting the existing Technical Specification Sec-l tion 2.1, Introduction, Section 2.2 Safety . Limits - Reactor Core, and Sec-p tion 2.3, Safety Limits - Reactor Coolant System Pressure' to a format consis-tent with the Westinghouse Standard Technical Specifications (H STS). The content of the proposed RTS sections is the same as the existing . Technical Specifications. No changes have occurred to these sections other than the renumbering to achieve consistency with the H STS. The proposed RTS is an enhancement to the existing Technical Specifications. It provides clear applicability, action, and surveillance requirements modeled after the H STS.

The proposed changes are compared to the existing Technical Specifications and the H STS. A matrix summarizing this comparison is included in Attachment 3.

Significant Hazards Consideration 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 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. The content of the RTS is the same as the previously approved version of the existing Technical Specification Section 3.17. No changes have occurred to this section other than renumbering to cJieve consistency with the H STS. Therefore, there is no increase in the probability or consequences of an accident previously analyzed.

2. Create the possibility of a new or different kind of accident from any previously evaluated. The proposed changes do not impact the operation of any component or system. The proposed changes do not introduce any new failuras. Therefore, the proposed changes do not create the possi- .

1 bility of a new or different kind of ac::ident from those previously analyzed.

3. Involve a significant reduction in a margin of safety. Since the pro-posed changes do not affect the consequences of an accident previously analyzed, there is no reduction in the margin of safety.

Section 2.2. Limitina Safety System Settinas The proposed revised Technical Specification Section 2.2, Limiting Safety System Settings, has been prepared by converting the existing Technical Speci-fication Section 2.4, Protective Instrumentation to a format consistent with the H STS. For additional discussion, refer to Section 3/4.3, Instrumen-tation.

_ -- ___a

1 Docket No. 50-213 B13269 Attachment 2 Haddam Neck Plant Technical Specification Section 3/4.3 Instrumentation l

l l

1 l

I June 1989 l

Attachment 2

, . , , Section 3/4.3 B13269/Page 1 Technical Specification Section-3/4.3. Instrumentation The proposed . revised Technical Specification (RTS) Section 3/4.3 has been prepared by converting the existing Technical Specification Section 2.4, Protective Instrumentation, Section 3.8, Turbine Cycle, Section 3.9, . 0pera-tional Safety Instrumentation and Control Systems,. Section 3.11, Containment, Section ' 3.21, Safety-Related Equipment Flood Protection, Section 3.22 Fire Protection Systems Section, 3.23, Post Accident Instrumentation, Section 4.2, Operational Safety Items, Section 4.3, Core Cooling System - Periodic Testing, Section 4.8, Auxiliary Steam Generator Feed Pump, Section 4.14, Flood Protec-tion Annunciation, Section 4.15 Fire Protection Systems, and Section 7/8.2, .

- Instrumentation, to a format consistent with the Westinghouse' Standard Tech- I nical Specifications (1(p STS). In addition, applicable Administrative Tech-nical Specifications at the Haddam Neck Plant have also been included in the proposed RTS. The proposed changes are compared to the existing Technical Specifications and the H STS. A matrix summarizing this comparison is includ-ed in Attachment 3.

Section 2.2.1 - Reactor Trio System Instrumentation and Section 3/4.3.1 -

! Reactor Trio System Instrumentation The above two sections are tied closely together, therefore they are discussed under a common heading. The reactor trip setpoint limits specified in Table 2.2-1 have been selected to ensure that the reactor core and reactor coolant system (RCS) are prevented from exceeding their acceptance criteria during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System (ESFAS) in mitigating the consequences of accidents. The proposed RTS sections 2.2.1 and 3/4.3.1 have been prepared by converting existing Technical Specification Section 2.4, 3.9 and 4.2 to a format consistent with the W STS. The following is a descrip-tion of the changes between the existing Technical Specifications and the proposed RTS. New and additional requirements are a conservative change and do not require further justification. Clarifications are considered to be a i more detailed definition of the requirement or action and do not require justification. All relaxed or deleted requirements are justified below.

New/ Additional Requirements

1. The trip settings given in the existing Technical Specification Sec-tion 2.4 corresponds to the allowable values in Table 2.2-1 of the proposed RTS. The proposed RTS also gives the actual trip setpoints, which are not in the existing Technical Specifications.

2. The applicability section for the proposed RTS Section 3.3.1 specifies mode requirements rather than just requiring them to be operable at full power as specified in the existing Technical Specifications.

(1) Administrative Technical Specifications at the Haddam Neck Plant are administrative procedures that were implemented as an interim measure prior to converting the Technical Specifications to the W STS format. !

_ _ _ _ _ i

Q e Attachment 2 Section 3/4.3 s B13269/Page 2

3. The action for the proposed RTS Section 2.2.1 requires that channels that i are out of calibration be declared inoperable. '

4. The following reactor trip system instrumentation given in Table 2.2-1, !

Table 3.3-1 and Table 4.3-1 do not have any corresponding requirements in {

the existing Technical Specifications: 1 Item 10 - Undervoltage - Reactor Coolant Pump Item 11 - Safety Injection ;

Item 12 - Reactor Coolant Pump Breaker Position Trip- )

Item 13 - Main Steam Line Trip Valve Closure (Except for Table j 3.3-1)

Item 14 - Turbine Trip Item 15 - Reactor Trip System Interlocks Item 16 - Reactor Trip System Breakers

5. The high steam flow reactor trip requirements given as Item 8 of Table !

4.3-1 do not have any corresponding requirements in the existing Tech-nical Specifications.

6. The st &- generator low level coincident with steam /feedwater mismatch trip requirements given as Item 9 of the proposed Table 2.2-1 do not have any corresponding requirements in the existing Technical Specifications.

7. Item 1 of the proposed Table 3.3-1 requires two manual trip channels instead of the one channel required by the existing Technical Specifica-tions.

8. The requirements for a manual reactor trip are also included in the proposed RTS Table 2.2-1 and the surveillance requirements on the trip are added to the proposed Table 4.3-1.

9. Item 2 of the proposed Table 3.3-1 only allows 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months with 3 operable high power (neutron flux) trip channels. The existing Technical Specifi-cation allows indefinite operation under these conditions.

10. Item 9 of the proposed Table 3.3-1 requires an inoperable channel for low steam generator level coincident with steam /feedwater mismatch to be placed in the tripped condition within one hour. The existing Technical Specification allows operation to continue with continuous operator surveillance.

11. The following surveillance requirements are added to the proposed Table 4.3-1. They do not have any corresponding requirements in the existing Technical Specifications.

Item 2 - The channel calibration at refueling interval was added to the power range neutron flux instrumentation.

Item 3 - The channel calibration at the refueling interval was added to the intermediate range neutron flux instrumentation.

Item 4 - The analog channel operational test at a six week interval was added for the variable low pressure trip.

l

Attachment 2

, .;,- Section 3/4.3 B13269/Page 3 Item 9 - The analog channel operational test was added at the refuel-ing interval for the low steam generator level coincident with steam /feedwater mismatch trip.

Relaxed / Deleted Requirements

1. A footnote to Item 2 of the existing . Technical Specification Section 2.4 l allows the high pressurizer level trip to be bypassed when the reactor is at least 1.5% delta k subcritical. This may be interpreted to imply that it must be available in Modes 1, 2, and part of 3. The proposed RTS only requires the trip to be operable in Mode I above 10% power. The reduc-tion in the mode requirement is acceptable since this trip is not cred-ited in any safety analysis below 10% power.

2. A footnote to Item 8 of the existing Technical Specification Section 2.4 allows the high startup rate trip to be bypassed above 10% of rated power. This may.be interpreted to imply that it is required below 10%

power. The proposed RTS only requires the trip below 5% power. 'This relaxation is acceptable since the high startup rate trip-is not required between 5% and 10% power for any design basis accidents.

3. Item 6 of the existing Technical Specification Section 2.4 gives the

' requirements for the reactor coolant loop valve temperature interlocks.

These requirements are included in the proposed RTS surveillance 4.4.1.6.2. There is no corresponding limiting safety system settings or limiting condition for operation in the proposed RTS. The reactor coolant loop valve temperature interlock is considered a control grade rather than a safety grade. Therefore it is inappropriate for this interlock to appear in the proposed Table 2.2-1. The ACTION statement of the proposed RTS Section 3.4.1.6 duplicates the existing requirements of the specification 2.4.

4. The existing Technical Specification Section 2.4 gives a description of '

r the bases of the various trip functions. Appropriate corresponding descriptions are moved to the bases of the proposed RTS.

5. Item C of the existing Technical Specification 3.9 gives specific opera-bility requirements for neutron monitoring equipment. All reactor safety requirements are included in the proposed RTS Sections 2.2.1 and 3.3.1.

Any other requirements are related to plant reliability and are not appropriate for inclusion in the Technical Specifications. Therefore they are not included in the proposed RTS.

6. The existing Technical Specification Table 3.9-1 requirement for the source range start-up rate rod stop is not included in the proposed RTS.

This requirement is not credited in any of the plant's design basis safety analysis. i

7. The requirements in the existing Technical Specification Table 3.9-1 for the shutdown high neutron level alarm is being deleted from the proposed RTS. Instead it will be included in the proposed RTS 3/4.9.2. This l

l l

Attachment 2 Section 3/4.3

..- 8,13269/Page 4

. section was submitted to the NRC(2) and will be revised accordingly and submitted to the NRC at a later date.

Section 3/4.3.2 - Enaineered Safety Features Actuation System Instrumentation The proposed RTS Section 3.3.2 has been prepared by converting 'the existing Technical Specification Sections 3.8, 3.9, 3.11, 4.2, and 4.8 to a format consistent 'with the M STS. The following is a description of the changes between the existing Technical Specifications and the proposed RTS. New and additional requirements are a conservative change and do not require further justification. Clarifications are considered to be a more detailed definition of the requirement or action and do not require justification. All relaxed or deleted requirements are justified below.

New/ Additional Requirements

1. The following items are added to the proposed Table 3.3-2 and these items do not have any corresponding requirements in the existing Technical Specifications.

Item 1(a) - Manual Initiation of Safety Injection

- Item 4 - Emergency Bus Undervoltage Item 5 - Containment Isolation

2. Item 1(b) of the proposed Table 3.3-2 requires two trains of high con-tainment pressure signals to be used for safety injection. The existing Technical Specifications only require one train.

3. Item 2(a)(2) of the proposed Table 3.3-2 requires one out of three logic for high steam flow steam line isolation. The existing Technical Speci-fications allow two out of three logic.

4. The trip setpoints from the proposed Table 3.3-3 do not have any corres-ponding items in the existing Technical Specifications for the following:

Item 1 - Safety Injection Item 2 - Steam Line Isolation Item 3 - Auxiliary Feedwater Item 4 - Emergency Bus Undervoltage

5. Item 1(a) of the proposed Table 4.3-2 (manual initiation of safety injection) is not required in the existing Technical Specifications.

6. Item 4 (Emergency Bus Undervoltage) and Item 2 (Steam Line Isolation) included in the proposed Table 4.3-2 do not have any corresponding requirements in the existing Technical Specifications. !

(2) E. J. Mroczka letter to U.S. NRC, Revised Technical Specifications, dated i

October 26, 1988.

1 L____________

l Attachment'2 I

. . t: Section 3/4.3 -

. B13269/Page~5

7. The ^ following surveillance requirements . are included in the' proposed Table 4.3-2. There are no corresponding requirements in the existing Technical Specifications.

L -

Item 3(a)- -

. A shift channel check and refueling channel calibra-

tion for auxiliary feedwater initiation on low steam generator level.

l -

Item 1(b) -

A refueling analog channel operational check for safety injection on high containment pressure.

Item 5(a) -

A refueling analog channel operational check. for containment isolation on high containment pressure.

Clarifications

.1. The action statement of Item 1(b) of the proposed Table 3.3-2 allows six hours in two out of two logic prior to shutdown. The existing Technical Specification does not give an allowable time frame to execute the action statement and applies for one operable train.

2. The action statement for Item 1(c) of the proposed Table 3.3-2 allows one hour in two out of two logic before p1 acing an inoperable channel in trip. The existing Technical Specification does not specify a time frame.

3. The action statement for Item 2(a)(1) of the proposed Table 3.3-2 re-quires an inoperable channel to be placed in the trip mode in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

The proposed RTS allows one out of three operation at 100% power. The existing Technical Specification allows two- out of three operation, but only with three loop operation.

4. The action statement for Item 3(a)(1) of the proposed Table 3.3-2 allows one hour to place an inoperable channel in trip for four' loop auxiliary feedwater initiation on low steam generator level. The existing Techni-cal Specification Table 3.8-1 does not specify the time requirements.

5. The action statement for Items 3(a)(2) of the proposed Table 3.3-2 allows 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to fix an inoperable channel or the power must be reduced below 10%. The plant may run in a two out of two mode in one train during this time. The existing Technical Specification requires placement of an inoperable channel in the tripped mode, but does not require reduction in power.

Relaxed / Deleted Requirements

1. .The requirements for manual initiation of auxiliary feedwater in Item a of the existing Technical Specification Table 3.8-1 are not included in Item 3 of the proposed Table 3.3-2. The requirement of manual initiation of the auxiliary feedwater is functionally met by manually starting the auxiliary feedwater pumps and opening the bypass valve. The existing auxiliary feedwater system does not have a separate manual initiation.

i

q

' Attachment 2

[~ . u Section 3/4.3 1

. B13269/Page 6

2. The action statement for Item 3(b) of the proposed Table 3.3-2 allows 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to restore an inoperable auxiliary feedwater initiation channel or power must be reduced below 10%. The existing Technical . Specification Table 3.8-1 does not give a time frame but does require the plant to L remain in Modes 3 or 4. This change is acceptable since the decay heat L loads below 10% power are small and would allow more than adequate time l for the operator to manually initiate the - auxiliary feedwater pumps.

Therefore, this reduction in the action statement requirements is accept-able.

3. Item 14 (Residual Heat Pump Flow Instrumentation) of the existing Tech- 3 nical Specification Table 4.2-1 is not included in the RTS. This exclu- 1 sion does not have a significant impact on plant safety since it is not credited in the design basis analysis. In addition, the W STS format does not require this parameter to be included in Technical Specifica-tions.

Section 3/4.3.3.1 - Radiation Monitorina for Plant Operations The proposed RTS provides limiting conditions of operation (LCO), action statements, surveillance requirements, and associated bases for radiation monitoring instrumentation during plant operation. The existing Technical Specifications do not address radiation monitoring' instrumentation operability except for an operability test (Item 19 of Table 4.2-1). This proposed RTS clarifies that. testing requirement by making it clear that the only radiation monitor with safety activity monitor andsignificance is consistentiswith the the containment atmospjare recently issued gaseous radio-3y amendment. The safety significance of this monitor is its capability to detect primary system leakage as required by the proposed RTS Section 3.4.6.1. This proposed RTS will help ensure the operability of this monitor for that purpose.

Section 3/4.3.3.2 - Movable Incore Detectors The proposed RTS section provides LCOs, action statements, surveillance requirements and corresponding bases regarding movable incore detectors. The operability of the movable incore detectors with the specified minimum comple-ment of equipment ensures that measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the core. The existing Technical Specifications do not address the movable incore detectors operability.

Section 3/4.3.3.3 - Seismic Instrumentation The proposed RTS section provides LCOs, action statements, surveillance requirements and corresponding bases regarding seismic instrumentation. The operability of the seismic instrumentation ensures that sufficient capability is available to determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required (3) A. B. Wang letter to E. J. Mroczka, Issuance of Amendment #116, dated ;

May 31, 1989.

, Attachment 2

"" Section 3/4.3' B13269/Page 7 to permit' comparison of the measured response to that used in the design basis for the facility. The. existing Technical Specifications do not address the seismic instruments. ,

Section 3/4.3.3.4 - Meteorological Instrumentation The proposed RTS section provides LCOs, action' statements, surveillance requirements, and corresponding bases regarding meteorological instruments-tion. The operability of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation dose to the public as a result of routine or accidental release of radioactive materials to the atmosphere. The existing Technical Specifications do not

. address the meteorological instrumentation.

'Section 3/4.3.3.5 - Accident Monitorina Instrumentaij_on q The proposed RTS Section 3.3.3.5 has been prepared by converting the existing Technical Specification sections 3.9 and 3.23 to a format . consistent with.the W STS. The following is a description of the changes between the existing Technical Specifications and the proposed RTS. New and additional require-ments are a conservative change and do not' require further justification. The proposed Table 3.3-7 is equivalent to the H STS in most cases. However, in many cases, a number of action statements are less restrictive than those of theWSTSbuttheproposedgSisconsistentwithGenericLetter83-37andthe recently issued amendment. Applicability modes are either consistent with or more restrictive than those of the W STS.

New/ Additional Requirements l

1. The following items on the proposed Table 3.3-7 are required for Modes 1 through 3. These are only required in the existing Technical Specifica-l tions when the reactor is critical.

l l

Item 5 - Pressurizer Water Level l -

Item 11 - Auxiliary Feedwater Flow Rate Item 13 - PORV Block Valve Position Indicator

- Item 14 - Safety Valve and PORV Acoustic Flow Monitor

2. The following items are added to the proposed Table 3.3-7. They do not have any corresponding requirements in the existing Technical Specifica-l tions.

Item 6 - Steam Generator Pressure Item 7 - Narrow Range Steam Generator Water Level

- Item 8 - Wide Range Steam Generator Water Level Item 9 - Refueling Water Storage Tank Level Item 10 - Boric Acid Solution Tank Level i

(4) A. B. Wang letter to E. J. Mroczka, Issuance of Amendment #113, dated j April 24, 1989.

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-Attachment 2 Section 3/4.3 i B13269/Page 8

3. The following items on the proposed Table 4.3-6 have a monthly check required that is not required in the existing Technical Specifications.

Item 9 - Refueling Water Storage Tank Level .

Item 13 - PORV Block Valve Position Indication j

4. The following items are added to the proposed Table 4.3-6. They do not have any corresponding requirements in the existing Technical Specifica-tions.

Item 6 - Steam Generator Pressure Item 8 - Wide Range Steam Generator Water Level Section 3/4.3.3.6 - Fire Detection Instrumentation l This proposed RTS section provides a Limiting Condition for Operation (LC0) requirement for the minimum number of OPERABLE fire detectors whenever sys-tems, structures, components or equipment protected by the fire detection I instrumentation are required to be OPERABLE. These requirements are equiv-alent to the existing Technical Specification with the following exceptions: t

1) Instrumentation in additional fire zones are now included in the specification. This is an enhancement of the existing specifica-tion.

2) Additional requirements have been added when detectors have failed but the number operable still meet the minimum requirements. This is an enhancement of the existing specification. !

3) A continuous fire watch is now :equired instead of a fire watch patrol when the number of operable detectors is not met. This is an enhancement of the existing specification. !

4) Surveillance requirements are now included for nonsupervised cir-cuits. This is an enhancement of the existing specification.

5) Surveillance requirements for detectors which cannot be reset have been deleted. These devices have links which melt to alarm and would have to be replaced each time they were tested if they were '

demonstrated operable by a Trip Actuation Device Operational Test.

However, the circuits for these devices are still tested consistent with the remaining detector circuits to assure circuit integrity.

This relaxation of surveillance requirements has been shown not to ,

degrade the 'overall system reliability and is considered to provide equivalent protection.

6) The requirement for submitting a Special Report has been deleted.

Each potential reportable event will be reviewed in accordance with ,

the requirements of 10CFR50.73 as stated in proposed RTS Section !

6.6.1.

e r .

Attachment 2

"' Section 3/4.3 B'13269/Page 9 3/4.3.3.7 - Radioactive Liouid Effluent Monitorina Instrumentation t

{

The proposed RTS Section 3.3.3.7 has been prepared by converting the existing i Technical Specification Section 7/8.2.1 to a format consistent with the W STS. l The proposed RTS 3.3.3.7 is the same as the existing Technical Specification Section 7/8.2.1 except for an additional footnote added; on the applicability statement. This footnote clarifies that outages are permitted for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for maintenance, tests, etc. This is considered as a restrictive requirement when compared to the existing Technical Specification which allows outages within bounds permitted by the action statements.

1 3/4.3.5.8 - Radioactive Gaseous Effluent Monitorina Instrumentation The proposed RTS Section 3.3.3.8 has been prepared by converting existing Technical Specification Section 7/8.2.2 to a format consistent with the W STS.

The proposed RTS is the same as the existing sections except as discussed below:

1. There is a footnote on the applicability statement that restricts outages up to .12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for maintenance, tests, etc. This is more restrictive than the existing Technical Specification which allows outages within the bounds permitted by the action statements.

2. The action statement for Item 1(b) and 1(c) of the proposed Table 3.3-10 requires that auxiliary sampling be initiated within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of a channel being declared inoperable. This requirement is not in the existing Technical Specifications.

Section 3/4.3.4 - Internal Flood Protection The proposed RTS Section 3.3.4 has been prepared by converting the existing Technical Specification Sections 3.21 and 4.14 to a format consistent with the H STS and is same as the existing sections except as discussed below:

1. The-proposed surveillance requirement (4.3.4(a)) requires a 12-hour channel check. There is no corresponding requirement in the existing Technical Specification.

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

1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The determination of whether or not a j proposed change is equivalent, more restrictive (or a new requirement),

or less restrictive is based on the Limiting Condition for Operation and Applicability Requirements since it is these requirements which will ,

impact the design basis accidents. In general, the conversion to the H (

STS yields more extensive and/or restrictive Action and Surveillance Requirements. As described foove, most of the changes are more i l

Attachment 2

. 4 Section 3/4.3 !

B13269/Page 10 !

l restrictive in that they are a conservative change and there are no l comparable requirements in the existing Technical Specifications. This I will help ensure the operability and reliability of the systems covered under the proposed RTS. For the few changes that are less restrictive, justification is provided for the changes. Based upon the above discus- f sion, the proposed RTS >ill not increase the probability or consequences of any accident previously analyzed.

2. Create the possibility of a new or different kind of accident from any J previously evaluated. Since there are no hardware modifications asso-ciated with the proposed changes, the performance of safety-related systems remains unaffected during operations. The operability require-ments are increased over the current requirements thus enhancing the performance of safety systems. Therefore, the proposed RTS will not modify the plant response to the point where it can be considered a new accident nor are any credible failure modes created.

3. Involve a significant reduction in a margin of safety. Because the changes proposed herein provide acceptable results for the design basis accident, no additional burden will be placed on the protective bound-aries for postulated accidents. In addition, there are no plant modifi- 1 cations associated with these changes and hence, there is no direct impact on the protective boundaries. The proposed RTS do not affect the i safety limits of the protective boundaries and the bases of the proposed RTS have been modified to reflect the proposed changes.

l l

l 1

. .. s Docket No. 50-213 813269 Attachment 3 Technical Specification Comparison Matrix j

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1 1

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l June 1989

.-____-__D

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, Attachment 3 r

. .- Page 1

]

TECHNICAL SPECIFICATION COMPARISON MATRIX l

Introduction The Technical Specif' cation Comparison Matrix (TSCM) was prepared to l facilitate the revision of the existing Haddam Neck Technical Specifications .

(T.S.). The TSCM is set up denoting the proposed Technical Specification .

section numbers in the left hand column followed by a short description. 1 The next column lists the corresponding existing T.S. section number. The final two columns compare the requirements contained in the proposed section with the existing T.S. and the Westinghouse STS, respectively. The key at the bottom of each page provides an explanation for the symbols located in the two comparison columns. The equivalent notation "E" may either denote that exact wording has been transposed from the existing T.S. or different verbage conveying equivalent requirements has been used. In many cases, there was not a one-for-one relationship, but rather multi-section relationships, whereas the requirements in a given T.S. section may be divided between several different sections in the proposed Technical Specification. The additional requirement notation "++" denotes that the proposed Technical Specification is more restrictive because it is an entirely new requirement as compared to the existing T.S. or it is more restrictive in the sense that the existing T.S. requirements have been changed such that they are more restrictive. This matrix is provided in a i summary fashion and highlights the more significant changes. A detailed ,

comparison in terms of additional requirements and/or less restrictive j requirements is provided in Attachment 2 of this submittal. '

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l

i- ..

Attachment 3- ,

.o: Matrix 3/4.3- ,

. 'B13269/Page_1 1 2.0 Safety Limits and Limitina Safety Settinas ;

'and 1 3/4.3 INSTRUMENTATION-

~ TECHNICAL SPECIFICATION COMPARIS0N MATRIX Comparison Comparison ;

Existing With Existing With ;

T.S.# Description T.S. # T.S. W STS '

2.0 . Safety Limits and Limiting Safety System Settings

.2.1 Safety Limits 2.2 E E 2.1.1 Reactor Core 2.2-1 E E Applicability 2.2-1 E 'E- ,

Action 2.2-1 E E 2.1.2 ' Reactor Coolant System 2.3 E' E.

Pressure Applicability 2.3 E E Actions 2.3 E E 2.2 Limiting Safety System Setting 2.2.1 Reactor Trip System 2.4 ++ E Instrumentation Setpoints Applicability 2.4 ++(11) E Action 2.4 ++ E Table 2.2-1 Reactor Trio System Instrumentation Setooints ++(35) !

Item 1 Manual Reactor Trip - ++ E Item 2 Power Range, Section 2.4 E E Neutron Flux Item 4 Item 3 Intermediate Range Section 2.4 E E Neutron Flux Item 8 Item 4 Pressurizer Section 2.4 E E Pressure - Item 3 Variable low I

1

)

W Attachment 3 l 4

'O.: Matrix 3/4.3

~c B13269/Page2-Comparison Comparison- j Existing -With Existing With j T.S.# Description T.S. # T.S. W STS Item 5- Pressurizer Section 2.4 i Pressure - High Item 1 E E

]

l

,. Item l6 Pressurizer Water Section 2.4 E E Level.- High Item 2 ,

Item 7 Reactor Coolant Section 2.4 E E Flow - Low- Item 5 Item 8 Steam Flow'- High Section 2.4 E ++

Item 7 Item 9 ' Steam Generator -

++ ++

Water level low Item 10 Undervoltage .

++ E Reactor Coolant Pump Item 11 Safety Injection -

++ E Item 12. RCP-Breaker -

++

Position - Open Item 13 Main Steam Line- -

++ ++

Trip Valve Closure Item 14 Turbine Trip -

++ E

' Item 15 Reactor Trip System -

++ E Interlocks Item 16 Reactor Trip System -

++ E-Breakers 3.3.1 Reactor Trip System Instrumentation *(13)-

Table 3.3-1 Item 1 Manual Reactor Trip Table 3.9-1 ++(1) E Item 8 Item 2 Power Range, Neutron Table 3.9-1 ++(2),*(36)- E Flux Item 1 Item'3 Intermediate Range, Tabl e . 3.9-1 ++,*(36) E Neutron Flux Intermediate Range SUR Reactor Trip Item 4 Pressurizer Pressure - Table 3.9-1 E E Variable Low Item 2 J

Attachment 3

.- Matrix 3/4.3

. B13269/Page 3 Comparison Comparison Existing With Existing With 1 T.S.# Description T . S . # _. T.S. W STS j Item 5 Pressurizer Table 3.9-1 (3) E Pressure - high Item 3 l Item 6 Pressurizer Water Table 3.9-1 ++(1) and *(36) E level - high Item 4 Item 7 Reactor Coolant Flow Table 3.9-1 E *(4)

Item 5 Item 8 Steam Flow High Table 3.9-1 ++ ++(6)

Item 10 i Item 9 Steam Generator Table 3.9-1 ++ *(8) i Water Level - Item 9 Low Coincident with steam / feed-water flow mismatch Item 10 Undervoltage - -

++(7) (9)

Reactor Coolant Pump Item 11 Safety Injection E E Item 12 Reactor Coolant Pump -

++(7) ++(6)

Breaker Position Trip Item 13 Steam Line Isolation Table 3.9-1 E ++(6)

Valve Closure Item 10 Item 14 Turbine Trip -

++(7) E Item 15 Reactor Trip System -

++(7) *(10)

Interlocks Item 16 Reactor Trip System -

++(7) E Breakers Applicability Table 3.9-1 ++(11) E Action Table 3-9-1 ++(12) *(24) 4.3.1.1 Demonstrated Operable Table 4.3-1 Item 1 Hanual Reactor Trip -

++ E 1

'i

' Attachment 3 l

. . Matrix 3/4.3 ?

B13269/Page 4 i Comparison Comparison Existing With Existing With T.S.# Description T.S. # T.S. W STS Item 2 Power Range, Neutron Table 4.2-1 ++(14) ++(15)

Flux Item 1

. Item 3 Intermediate Range Table 4.2-1 ++(16) E Neutron Flux Item 2 Item 4 Pressurizer Pressure Table 4.2-1 ++(17) .*(18)

Variable low Item 8 Item 5. Pressurizer Table 4.2-1 E *(18)

Pressure'- High Item 7 g Item 6 . Pressurizer Water Table 4.2-1 E *(18)

Level - High Item 6 Item 7 Reactor Coolant Table 4.2-1 E *(19)

Flow - Low Item 5 Item 8 Steam Flow - High ++(7)- ++(6)

Item 9 Steam Generator Table 4.2-1 ++(17) *(19)

Water Level - Item 12 Flow mismatch Item 10 Undervoltage - -

++(7) *(19)

Reactor Coolant Pump Item 11 Safety Injection -

++(7) E Item 12 Reactor Coolant Pump -

++(7) ++(6)

, Breaker Position Trip Item 13 Main Steam Line Trip -

++(7) ++(6)

Valve Closure Item 14 Turbine Trip -

++(7) *(20)

Item 15 Reactor Trip System -

++(7) *(19)

Interlocks Item 16 Reactor Trip System -

++(7) *(21) ,

Breaker !

3.3.2 Engineered Safety Features Actuation *(13)

System Instrumentation Tab *.e 3.3-2 4

i s - 4 .

Attachment 3 '

. .O ~ Matrix 3/4'.3-

.. B13269/Page 5' Comparison -Comparison.

Existing With Existing -With

-T.S.# Description- T.S. # T.S. W STS i

Item 1' Safety Injection. Table 3.9-1 ++ E !

Items 6 and 11 3.11 H 'i

' Item 2- Steam Line Isolation Table 3.9-1 ++. E Item-10 Item'3 Auxiliary Feedwater 3.8.B *(36) E-Item 4 Emergency Bus -

++ E Undervoltage Item 5 Containment' Isolation -

++ E Applicability Table 3.8-1 and ++(11) *(23)

Table 3.9-1 Action Table 3.8-1 .. ++(12) *(25) and Table 3.9-1 Table 3.3.3 Engineered Safety Features Actuation System (ESFAS)

Instrumentation Trip Setpoints

Item 1 Safety Injection -

++ E Item 2 Steam Line Isolation -

++ E Item 3 Auxiliary Feedwater -

++ E Item 4 Emergency Bus --

++ E Undervoltage

-Item 5 Containment Isolation 3.11.H ++ E

~ Table 4.3-2 ESFAS Instrumentation Surveillance Requirements

. Item 1 Safety-Injection Table 4.2-1 E++ *(26)

Item 7 .

Item 2 Steam Line Isolation -

++ *(27) !

Item 3' Auxiliary Feedwater 4.8.1.b, 4.8.3.C ++ E Table 4.2-1 Item 11 Item 4 Emergency Bus -

++ E Undervoltage

i . . ;

Attachment 3 l Matrix 3/4.3 J B13269/Page 6 j l

I Comparison Comparison Existing With Existing With W I T.S.# Description T.S. # T.S. STS s l l Item 5 Containment Isolation Table 4.2-1 ++ *(26) 1 Item 18 3.3.3.1 Radiation Monitoring -

++ E I System LCO i l

I Applicabili ty -

++ E 1

Action -

++ E {

4.3.3.1 Demonstrated Operable Table 4.2-1 E E Item 19 3.3.3.2 Movable Incore -

++ E(29) 1 Detectors - LC0 j

'1 Applicability -

++ E Action -

++ E 4.3.3.2 Demonstrated Operable -

++ E 3.3.3.3 Seismic Instrumentation - -

++ E(29)

Low Applicability -

++ E Action -

++ E 4.3.3.3.1 Demonstrated Operable -

++ E !

and j 4.3.3.3.2 j i

3.3.3.3.4 Meteorological -

++ E ,

Instrumentation - LC0 ,

i Applicability -

++ E

)

Action -

++ E j 4.3.3.4 Demonstrated Operable -

++ E 3.3.3.5 Accident Monitoring Instrumentation ++ *(30)

Table 3.3-7 Item 1 Containment Pressure Table 3.23-1 E E i Item 1 l

, sp < c ..e -

g' . , ?

Attachment:

3?

vo .

= * . Matrix 3/4.3' to 'B13269/Page 7 ,

y ,

Comparison . -Comparison

[fl ,

Existing.- With Existing With W-T.S.# Description. -T.S. #- T.S. STS p

' Item 2. RCS Cold Leg-Temp. - Tabl e ' 3.23-1; .Ec E Wide, Range Item 2 .

p

Item 3 .RCS Hot LegLTemp. Table 3.23-1 E E'

. Item 3 L Item 4 RCS. Pressure Wide' Range . Table'3-23-1 E E

-Item 4 >

Item 5: -Pressurizer Water Level Table 3.9-2 ++- E

' Item 6 Steam' Generator Pressure .- ' ++ E Item 7- , Steam' Generator Water -

++

Level - Narrow l Range

' Item 8- ' Steam. Generator-Water -

++ E.

j. Level -Wide Range

~

. Item 9: 'RWST Level -

++ E 3

. Item 10 -Boric. Acid Tank-Solution

++- E Level Item.11 Auxiliary Feedwater Table 3.9-2 ++ E' Flow rate Item 2

~

Item 12' RCS Subcooling Margin- Table 3.23-1 E E-Monitor Item 10 Item 13 - PORV Block Valve Table 3.9-2 ++ E Position Item 5

. Item 14 Safety Valve and PORV Table 3.9-2 ++ E

-(Acoustic Flow Monitor) Items 4 and 6 Item 15~ Containment Water Level- Table 3.23-1 E E Wide Range Item 5

< Item 16 Containment Water Level -

++ E Narrow Range Item 17 Core Exit Thermocouple Table 3.23-1 E E Item 6

. Item l'8 Main Stack - Wide Range Table 3.23-1 E E Noble Gas Monitor Item 7 L

s. , ..

Attachment 3

+

Matrix 3/4.3

,- B13269/Page 8 Item'19 Contaiment Atmosphere- Table 3.23-1 E. E High Range Radiation Item 8' Monitor

. Item 20 Reactor Vessel Water- Table 3.23-1 E- E level Item 9

' Applicability refer to Note 30 Action refer to Note 30 4.3.3.5' Demonstrated Operable Table.

4.3-6 Accident Monitoring' Instrumentation ++(31) ++(31)~'

Surveillance' Requirements Item 1 Containment Pressure Table 3.23-2 E E Item 1 Item 2. .RCS Cold Leg' Temp - Table 3.23-2 E E.

Wide Range . Item 2 y

Item 3 : RCS Hot- Leg Temp - Table 3.23-2 E E Wide Range Item 3 Item 4 RCS Pressure -- Item 3.23-2 E- E Wide Range Item 4

' Item 5 Pressurizer Water .

++ E Level Item 6 -Steam Generator. Pressure -

++ E Item 7-- Steam Generator Water -

++. E Level - Wide Range l- Item 8 Steam-Generator Water - ++ E j Level - Narrow Range Item 9 RWST Level Table 4.2-1 ++ E Item 16 Item 10 Boric Acid Tank Table 4.2-1 E E Solution Level Item 15 l~ Item 11 Auxiliary Feedwater Table 4.2-1 E E Flow Rate Item 25

. Attachment 3

.o Matrix 3/4.3

, B13269/Page 9 Comparison Comparison Existing With Existing With W T.S.# Description T.S. # T.S. STS Item 12 RCS Subcooling Margin Table 3.23-2 E E Monitor Item 10 Item 13 PORV Block Valve Table 4.2-1 ++ E .

Position Indicator Item 28 .!

Item 14 Safety Valve and PORV Table 4.2-1 E E Acoustic Flow Monitor Items 27 and 29 ;

Item 15 Containment Water Level Table 3.23-2 E E Wide Range Item 5 Item 16 Containment Water Level -

++ E Narrow Range Item 17 Core Exit Thermocouple Table 3.23-2 E E Item 6 '

-Item 18 Main Stack-Wide Range Table 3.23-2 E E Noble Gas Monitor Item 7 Item 19 Containment Atmosphere Table 3.23-2 E E High Range Radiation Item 8 Monitor '

Item 20 Reactor Vessel Water Table 3.23-2 E E Level Item 9 3.3.3.6 Fire Detection i Containment 3.22.E E E Cable Spreading Area 3.22.E E E 1A Diesel Generator Room 3.22.E E E 18 Diesel Generator Room 3.22.E E E Switchgear Room 3.22.E E E Containment Cable Vault 3.22.E E ++

Waste Disposal Building 3.22.E E ++

Aux. Feedwater Pump Room 3.22.E E ++ ;

Primary Aux. Building 3.22.E E ++

Control Room 3.22.E E E Screen Well Building 3.22.E E ++

Spent Fuel Building 3.22.E E E PAB Charcoal Filter Bank -

++ ++ ,

Applicability 3.22 E E 1' Action 3.22.E.2 ++(33) E 4.3.3.6.1 Demonstrated Operable 4.15.E ++(34)++(32) E

(36) i 4.3.3.6.3 Demonstrated Operable -

++ E i

' Attachment 3

- .* . Matrix 3/4.3

-t_ B13269/Page 10 Comparison. Comparison Existing With Existing With W T.S.# Description T.S. # T.S. STS

-3.3.3.7 Radioactive Liquid' Effluent Monitoring. Instrumentation ,

Table 3.3-9 Item 1 Gross Radioactivity -Table 7.2-1 E E Monitors (Auto) Item 1 Item 2 Gross Radioactivity Tabl e . 7.2-1 E E Monitors.(Non-Auto) Item 2 Item 3 Flow Rate Measurement Table 7.2-1 E *(29)

, Item 3 Applicability - Table 7.2-1 ++ E Action Table 7.2-1 E E 4.3.3.7.1 Demonstrated Operable -8.2.1.1 E E 3.3.3.8 Radioactive Gaseous. Effluent Monitoring Instrumentation Table 3.3-10 Main Stack Item la Noble Gas Activity Table 7.2-2 E E Monitor 3.23 Item lb Iodine Sampler Table 7.2-2 ++ E Item Ic Particulate Sampler Table 7.2-2 & ++ E 3.11.F.1 Item Id Stack Flow Rate Monitor Table 7.2-2 E E 1 Item le Sampler Flow Rate Table 7.2-2 E E Monitor Applicability Table 7.2-2 ++ E l I

Action Table 7.2-2 E E i 4.3.3.8.1 Demonstrated Operable Table 8.2-2 E E 3.3.4 Internal Flood Protection l

Table 3.3-11 Liquid Level Table 3.21-1 E ++

Instrumentation 1

-?

', ;a : ,' :L .

Attachment 31 l' JC Matrix 3/4.3 ')

1. iB13269/Page 11 -j '

Comparison : Comparison

> Existing With Existing With W j

'T,S.# . Description T.S. # T.S. STS- q Applicability 3.21 E ++ ]

Action 3.21 E ++

l 4.3.4' Demonstrated Operable 4.14' ++- ++ )

{

Notes .. i E - Equivalent Requirements i

= Less restrictive-requirements

++- Additional Requirements -

i i

_ . _ . _ . . _ _-_____________....m._______ _-_ _ ___ _ _

, .~ h - .p

. Attachment ~3 fr Matrix 3/4.3

(- B13269/Page 12

~.

Section 3/4.3 Notes

[ (1) The' proposed revised _ Technical -Specification -(RTS)- requires . 2. minimum channels operable instead of 1.

- (2): The proposed RTS requires 3 minimum channels operable ,instead of 2.

(3) The proposed RTS requires 2 minimum channels operable instead of 1.

(4) The proposed RTS requires 1 minimum channel operable per loop instead of 2.

(5)~ Not used (6) The W STS does not include this instrumentation in the reactor trip system.

(7) Although not included in the existing Technical Specifications, many of these instruments are included under current plant surveillance proce-dures.

(8) The proposed RTS requires 1 steam /feedwater flow mismatch instrumentation in each steam generator instead of 2.

(9) The proposed ~RTS requires 1 minimum channel per bus whereas the H STS requires 3 minimum out 4 total channels available.

(10) The W STS includes additional interlocks.

(11) The existing Technical Specification is applicable to full power opera-tion only, while the proposed RTS includes additional power levels and modes.

(12) The proposed RTS has more detailed action statements and provides time contraints for subsequent action, while the existing Technical Specifica-tion does not.

(13) The M STS includes response time requirements for these instruments, whereas the proposed RTS does not. It is noted that the response time table will be provided after all the RPS modifications are complete.

(14) The proposed RTS adds the requirement to perform a channel calibration of the high, mid and low setpoints at each refueling.

(15) The W STS requires operational testing at least once per 31 days, while the proposed RTS requires testing once per 14 days.

. (16) The channel calibration is added to the proposed RTS.

(17) The operational test is added to the proposed RTS.

(18) The M STS requires operational testing at least once per 31 days, while the proposed RTS requires testing once per 42 days.

{'+ . .

.l 4 Attachment 3-

c. A Matrix 3/4.3-

^

. B13269/Page 13 I (19):The M STS requires operational testing at least once per 31. days, while the proposed RTS requires testing at least once per 18 months.

n (20) The' W STS requires operational testing at each criticality, while the ]

l proposed RTS requires testing at least once per 18 months. [

(21) The M STS requires operational testing once per 31 days, while the {

proposed RTS requires testing at each criticality. ;

(22) Not used. ,

I (23) Auxiliary feedwater initiation is less restrictive than the W STS. 1 (24) For the majority of instruments, the action statements are equivalent; hcwever, for the Power Range Neutron Flux and all Pressurizer Pressure instrumentation, the proposed RTS is slightly less restrictive .than the W STS.

(25) For the majority of instruments, the Action Statements are equivalent; however, the - steam flow isolation (3 loops operating) and auxiliary feedwater initiation on trip of all main feedwater pumps are less restrictive, while the high containment pressure isolation is more restrictive than the H STS.

(26) The W STS requires channel checks and more frequent operational testing 1 '

- for the high containment pressure safety injection. Also, more frequent operational testing of the low pressurizer pressure on safety injection is required per the W STS.

(27) More frequent operational testing is required per the W STS.

(28) Not used.

(29) The H STS is not directly comparable due to significant differences between the Haddam Neck Plant system and the H design.

(30) The proposed RTS Table 3.3-7 is shown equivalent to the H STS, however, in many cases the number of channels and action statements are less i restrictive, but the proposed RTS' Table is consistent with the Generic Letter 83-37 and CYAPCO's submittals dated July 1, 1988 and March 1, 1 1989 and recently issued amendment #115. Applicability Modes are either consistent with or more restrictive than the H STS.

(31) The surveillance requirement included in the proposed RTS Table 4.3-6 are either consistent with or more restrictive than the W STS.

(32) The proposed RTS has required action if any detectors are inoperable, whereas the current T.S. requires action only if a minimur number of detectors are not available.

i

T- Attachment 3 Matrix 3/4.3 c- .i-B13269/Page 14 I

(33) Alth'ough the current T.S. requires a' Channel Functional Test, the pro-L, posed RTS requires a Trip Actuating Device Operational Test, which is an i

improvement becciuse the Trip Actuating Device- Operational Test verifies the device along with the channel circuit.

(34) There is an additional surveillance requirement in the proposed RTS which covers non-supervised circuits.

(35) The trip setpoints given in the existing Technical Specification-corresponds to the allowable values in Table 2.2-1. The proposed RTS also gives the actual trip setpoints which are not in the existing Technical Specifications.

(36) Refer to the discussicn under relaxed requirement in Attachment 2.

-__---_--a

B13269, Proposed Tech Specs Sections 2.0 & 3/4.3 Re Safety Limits & Limiting Safety Sys Settings & Instrumentation

Text

Attachment:

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