Text

Proposed Tech Specs Re Reactor Trip Sys Trip Setpoints & ESFAS Instrumentation
	ML19332E438
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	11/22/1989
From:	; CONNECTICUT YANKEE ATOMIC POWER CO.
To:	;
Shared Package
ML19332E437	List: B13402, Application for Amend to License DPR-61,incorporating NRC Comments on Proposed Revised Tech Specs Re Reactor Protection Sys/Nuclear Instrumentation Sys Upgrade, Electrical Power Sys,Cycle 16 Reload & Fire Protection Sys; ML19332E438;
References
	NUDOCS 8912070225
	Download: ML19332E438 (73)
	v • d • e

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15. Reactor Trip System Interlocks.

q l a. Low Power Block, P-7 195 and siGE of RTP* '

1) ' Power Range, Neutron Flux 275 and 585 of RTP*

i

2) Turbine First' Stage' Pressure 27% and 181 of RTP* at: Turbine First

-25% and $105 of RTP* at Turbine First i- . Stage Pressure Equivalent Stage Pressure Equivalent-b. Low Flow Pe 1stssive P-8 I) Power Range, Neutron ~ Flux 172% of RTP* 1745 of RTP*. l

2) Turbine'First Stage Pressure 172% of RTP* at Turbine First 574% of RTP* at Turbine First_

Stage Pressure Equivalent Stage Pressure.Equivaient y

16. Reactor Trip System Breakers N.A.

N.A. u j

17. Source Range Start-up Rate Rod Stop

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p j 3/4 LIMITING CONDITIONS FOR OPERATION'AND SURVEILLANCE RE0VIREMENTS 3/4.0 APPLICABILITY LIMITING' CONDITION'FOR OPERATION l 3 ~ 0.'I Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met. 3.0.2 Noncompliance with a specification shall exist when the requirement:, of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for 0peration is restored )rior to expiration of the specified time intervals,; completion of t1e Action requirements is not required. 3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within 1 hour action shall-be initiated to place the unit in a MODE in which the specification does not-apply by placing it, as applicable, in: a. AT least HOT STANDBY within the next 6 hours, b. AT_least HOT SHUTDOWN within the following 6 hours, and 'c. AT least COLD SHUTDOWN within the subsequent 24 hours. Where corrective measures are completed that permit' operation under the ' ACTION requirements, the-action may be taken in accordance with the l-specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications. This specification is not applicable in MODE 5 or 6. L 3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall L not-be made when the conditions for the Limiting Conditions for Operation are not met and.the associated ACTION requires a shutdown if they are-not met within a specified time interval. Entry into an OPERATIONAL MODE or L specified condition may be-made in accordance with ACTION requirements when b conformance to them permits continued operation of the facility for an unlimited period of time. This provision shall not prevent passage through l or to OPERATIONAL MODES as required to comply with ACTION requirements. l Exceptions to these requirements are stated in the individual it specifications.- L 4.0.I' Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for l Operation unless otherwise stated in an individual Surveillance Requirement. L 4.0.2 Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25% of the specified surveillance interval. HADDAM NECK 3/4 0-1

4. - APPLICABILITY SURVEILLANCE RE0VIREMENTS 4.0.3 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specification 4.0.2, shall constitute - noncompliance with the OPERABILITY requirements for a Limiting Condition for 0)eration. The time limits of the ACTION requirements are applicable at tie time it is' identified that a Surveillance Requirement has not been - performed. The ACTION requirements may be delayed for up to 24 hours to i permit the completion of the surveillance when the allowable outage time limits of the ACTION requirements are less than 24 hours. Surveillance Requirements'do not have to be performed on inoperable equipment. 4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall - not be made unless the Surveillance Requirement (s) associated with the Limiting Condition for Operation have been performed within the stated surveillance interval or as otherwise specified. This provision shall not prevent passage through or to 0PERATIONAL MODES as required to comply with ACTION requirements. 4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows: a. Inservice inspection of ASME Code Class 1, 2, and 3 equivalent components and inservice testing of ASME Code-Class 1, 2, and 3 equivalent pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code a-applicable Addenda as required by 10 CFR Part 50, Sectior 50.55a(g), except where specific written relief has been. by the Commission pursuant to 10CFR Part 50, Section 50.5i (g)(6)(1). b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications: ASME Boiler and Pressure Vessel Required frequencies for Code and applicable Addenda performing inservice terminology for inservice inspection and testing insoection and testina activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days HADDAM NECK 3/4 0-2

1 w u. 'Y 3/4.0 APPLICABILITY ~ BASES (Con't) .P " Surveillance. requirements are requirements relating to test, calibration, or inspection to ensure that the necessary quality of systems and components is maintained, that facility operation will be within safety. limits, and that the limiting conditions of operation will-be met." ' Snecification 4.0.1-establishes the requirement that surveillances must be -performed during the OPERATIONAL MODES or other conditions for which the -requirements of the Limiting Conditions for_ Operation apply unless otherwise stated in an individual Surveillance Requirement. 'The purpose of this specification is to ensure that surveillances are performed to verify the i operational status of. systems and components and that parameters are within specified: limits to ensure safe operation of the facility when the plant is in a MODE or other specified condition for which the associated Limiting 'f^ Conditions for Operation are applicable. Surveillance Requirements do not have to be performed when the facility is in an OPERATIONAL MODE for which lthe requirements of the-associated Limiting Condition for Operation do not apply unless otherwise specified. The Surveillance Requirements associated with a Special Test Exception _are only applicable when the Special Test Exception'is used as: an allowable. exception to the requirements of a Especification. Soecification 4.0.2' establishes the limit for which the specified time ' interval for Surveillance: Requirements may be extended. It permits an - allowable extension of.the r.ormal surveillance interval to facilitate surveillance. scheduling and' consideration of plant operating conditions that may not be: suitable for conducting the surveillance; e.g., transient L conditions'or other ongoing -surveillance-or maintenance activities. It also provides flexibility to accommodate the length of a fuel cycle for surveil-r lances that are performed at each refueling outage and are specified with an 18-month surveillance interval. It-is not intended that this provision be ) used repeatedly.as a convenience to extend' surveillance intervals beyond n that specified for surveillances that are not performed during refueling E _ outage. The limitation _of Specification 4.0.2_ is based on engineering J K1 judgement and the recognition'that the most probable result of any particular surveillance being performed is the verification of conformance b with the Surveillance Requirements.. This provision is sufficient to ensure .that the reliability' ensured through surveillance activities is not b significantly degraded beyond that obtained from the specified surveillance -interval. Soecification 4.0.3 establishes the failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by the provisions of Specification 4.0.2, as a condition that constitutes a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation. Under the provisions of this specification, systems and components are l assumed to be 0PERABLE when Surveillance Requirements have been l; satisfactorily performed within the specified time interval.

However, l

nothing in this provision is to be construed as implying that systems or components are OPERABLE when they are found or known to be inoperable 1. L -HADDAM NECK B3/4 0-4

j& ,L i 6 3/4.0 APPLICABIllTY t BASES (Con't) v> i although still meeting the Surveillance Requirements. This s)ecification also clarifies that the ACTION requirements are applicable w1en Surveillance Requirements have not been completed within the allowed i~ surveillance interval and that the time limits of ACTION requirements apply from the point in time it is identified that a surveillance has not been t performed and not at the time that the allowed surveillance interval was exceeded. Completion of the Surveillance Requirement within the allowable outage tima limits of the ACTION requirements restores compliance with the requirements of Specification 4.0.3. However, this does not negate the fact that the fatiure to have performed the surveillance within the allowed surveillance interval, defined by the provisions of Specification 4.0.2, was a violation of the OPERABILITY requirements of a Limiting Condition for Operation. L If the allowable outege time limits of the ACTION requirements are less than 24 hours or a shutdown is required to comply with ACTION requirements, e.g., Specification 3.0.3, a 24 hour allewance is provided to permit a delay in implementing the ACTION requirements. This provides an adequate time limit to complete Surveillance Requirements that have not been performed. The >urpose of this allowance is to permit the completion of a surveillance 3efore a shutdown is required to comply with ACTION requirements or before other remedial measures would be required that may preclude completion of_a surveillance. The basis for this allowancu includes consideration for plant conditions, adequate planning, availability of personnel, the time required to perform the surveillance, and the safety significance of the delay in comaleting the required surveillance. If a surveillance is not completed wit sin the 24-hour allowance, the time limits of the ACTION requirements are applicable at that time. When a surveillance is performed within the 24 hour allowance and the Surveillance Requirements are not met, the time limits of the ACTION requireaents are applicable at the time that the surveillance is terminated. Surveillance Requirements do not have to be performed on inoperatie L equipment because the ACTION requirements define the remedial measures that iL aaply.- However, the Surveillance Requirements have to be met to demonstrate t1at inoperable equipment has been restored to OPERABLE status. Specification 4.0.d establishes the requirement that all applicable surveillances must be met before entry into an OPERATIONAL MODE or other L condition of opwation specified in the Applicability statement. The purpose of this specification is to ensure that system and component OPERABILITY requirements or parameter limits are met before entry into a l L MODE or condition for which these systems and components ensure safe operation of the facility. This provision applies to changes in OPERATIONAL L MODES or other specified conditions associated with plant shutdown as well as startup. 1 HADDAM NECK B3/4 0 5 l

REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING SURVEILLANCE REQUIREMENTS (Con't.) L At least once per 31 days by) verifying that each valve (manual, c. power-operated, or automatic in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct l-position; t-l' d. At least once per 18 months during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on a Safety injection Actuation test signal; and e. At least once per 18 months by verifying that the flow path required by Specifications 3.1.2.2.a and 3,1.2.2.b deliver flow to the Reactor Coolant System. r: I. I HADDAM NECK 3/4 1 12 1 ~'y <.-e -.-m - - -- -. - - - - J

o REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN f LIMITING CONDITION FOR OPERATION -3.1.2.3 One charging pump (centrifugal or metering) in the boron injection l flow path required by Specificatior. 3.1.2.1 shall be OPERABLE and capable of l being powered from an OPERABLE emergency power source. APPLICABILITY: MODES 5 and 6. ACTION: l With no charging pump (centrifugal or metering) OPERABLE or capable of being i powered from an OPERABLE emergency power source, suspend all operations L involving CORE ALTERATIONS or positive reactivity changes. SURVEILLANCE REQUIREMENTS i 4.1.2.3.1 The above required charging pump (centrifugal only) shall be . demonstrated OPERABLE as required by Specification 4.5.1.c.3. '4.1.2.3.2 Theaboverequiredchargingpump(metering)shallbe demonstrated OPERABLE at least once every 31 days by running the pump at its maximum speed. 4.1.2.3.3 One centrifugal charging pumi, shall be demonstrated inoperable by verifying that the control switch is in the trip pullout position at least once per 12 hours whenever the Overpressure Protection System of Specification 3.4.9.3a is required. HADDAM NECK 3/4 1-13 1

m y NoCMnp ,3 p4.3 INSTRLMENTAT10N 1/4.3.1 REAtTOR TRIP SYSTEN INETRLatrNTATION LIMITINa c0NDITION FOR OPERATION i 3.3.1 - As a minimum the Reactor Tr -toterlocks of Table $.31 shall be Odi $ stem instrumentatten channels andl RAhlt. APPLICABILITY: As shown in Table 3.3 1. l ' gllg: As shown in Table 3.3 1. { SURVEILLANCE RfDUIREMENTS l l 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock logic shall be demonstrated OPERABLE by the performance of the Reactor Tri System Instrumentation Surveillance Requirements specified in Table 4.3 1.p l l 3 a t 1 i i NADOAM NECK 3/4 3 1

~ .=.. TAIII F 3.3-1 ~ REACTOR TRIP innen inn ___miniium lk RINIIMI TOTAL NO. OWWWIELS OWueELS APPLIO K E ""1 M .y FUNCTIOlWL UlllT OF. DWWeIELS T0_IRIP OPUNLE L z 1 2 1 2' 1, 2. It 2 .I 2 3*,4*,5* 1. Manual Reacter Trip 4 2 3 1,2,3*,4*,5* 2, it l 2. Power Range,Ilectron Flex, a. High Setpoint j b. Mid Setpoint h c. Low Seteeint 3. Interned 1 ate Range, Westran Flex, 2 -1 2 1(d) 2,3*,4*,5*** 4 2 1 2 5* 4 Migh Start Up Rate I *I N I i 4. Pressurizer Pressure-Variable, low 4 2 3 1 Presseriter Pressure--Migh 3 2 2 1, 2 W 5. W 6. Presseriter Water level--Migh 3 2 2 1, 2, 3**** a. Above P-s 4 (1/1eep) 1 (1/1eep) 1(b) 7 I 7. Reacter coolant riew - Low l b. Above P-7 and 4 (1/leep) 2** 4 (1/188P) 1(*I 8' gelow P-8 Gc::. F I to O I. 'aO s W i i

.g ~ i .v. 9 TARIF 3.3-1 [Ceatfneed!. e \\ ItEACicR TRIP SYSTBt INSIMEMRTim i MINNMI i -TOTAL NO. CNlusIELS CfgMELS APPLICfE E FUlICTIONRL UlIIT OF CIWeeIELS To TRIP OPUNLE FE M L .x l 8. Steam Flow-Migh 4 (1/ steam 11ae) 2 1/ steam line

1. ' t.-

W - r t ~,2 W l 9. Steam Generator Water 1/SG 1evel-1/SG 1evel I/SG1evel 1 Level-Low and ceincident and'. i Coincident With -with i SteaWFeedwater Flow 1/steaWfeed-1/steaWfeed-1/steaWfeed-i Nisnatch water flow water flow water flew: mismatch in -mismatch misestch-each SG in same leep in each SG' II *I S

10. Underveltage - Reacter 2 (1/bes)

I 2 (1/bes) R Conlant Pumps = i Y

11. Safety Injection 2

1 2 1, 2 11A s + L i i i i c._ i I F' i N I ) co-1 O ers f I l l 1 w e

1 TABLE 3.3-1 iCentinsedi K MTOR TRIP SYSTEN I N ATIM 1 E NIIIIIENI ' i' x TUTAL 110. OllWWIELS ogueIELS APPLI M E ? FUllCTIGNIL MIT OF QUUNELS J0 TRIP OPERAGLE C M

12. Reacter Coolant Pump

~l i Breaker Positten Trty a. Above P-8 4 (1/ pump) 1 4 (I per pump) 1(b) 7 l j b. Above P-7 and 4(1/ pump) 2** 4(Iperpump) II*I 8 Below P-8 i 5 L

13. Steam Line Isolatten 1/ valve I

1/ valve I*I 8 I Valve Closure I *I 2 I

14. Turbine Trty 3

2 2 ) j

15. Reacter Trip Systen laterlocks t

a. Law Power 81eck, P-7 i 1) Power Range,Ilestron I,I 3a I I Flex 4 2 3 i 2) Turtpine First Stage i j Pressure - 1 1 1 1(a) 3h j { b. Lew Flow Permissive, P-8 i I) Power Range,Ilectron i Flex 4 2 3 IIbI 3a 2) Turbine First Stage I l Pressure 1 I i 1(b) 2

16. Reactor Trip System greakers 2

1 2 1,2 11 1 2 1 2 3*, 4*, 5* 18 17. Source Range Start-up Rate Rod 2 1 2 I (d) 4 i, Stop I I I i' i l l

TABLE 3.3-1 (Continued) j t TABLE.N0TATION With the Reactor Trip System breakers in the closed 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. <---+ 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. (d) TilERMAL POWER below 10% of RATED TilERMAL POWER. ETION STATEMENTS ? ACTION 1: With the number of.0PERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in HOT STANDBY within the next 6 hours, l-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 satisited: a. The inoperable channel is placed in the tripped condition within 6

hours, I

- - - May be bypassed when the reactor is at 1 cast 1.5}A k sube ri tical,

g l HADDAM NECK 3/4 3-5 e

4 m

~ w e- + y ,n-w

e

,,we. om-t--, w w w ,ne----,--, m-w m m w-esa - + - - --e,+~w--.---- -- - *~------ -

w+-

-- ~~- -

T

t 2.2 1 ftenti " )

ACTION ETATDe[NTE f tentinued) t i ACTION 2: (Continued) The Minimum Channels OPERABLE requirement is seti however, the b. inoperable channel any be bypassed for up to 4 hours for surveillance testing of other channels per Specification 4.3.1.1. - NTINI 3: s) With less than the Minimun Number of Channels OPERABLE, within 1 bour i s. determine by observation of the associated permissive annunciator that the interlock is in its required state for the existing window (s)dition or apply Specification 3.0.3. plant con 5 With turbine first stage pressure inoperable, continued power operation b. any proceed provided the permissive is placed in the more conservative 1 stats for existing plant conditions. L ACTION 4: With the number of OPERABLE channels one less than the Minimum Channels OPERA 8LE requirement, suspend all operations involving positive reactivity changes and restore the inoperable channel to OPERABLE status within 8 hours oropen/verifyopen the Reactor Trip System breakers within the next hour. ACTION 5: With the number of OPERABLE channels less than the Minimum l8 a. 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 say be bypassed for up to 4 hours for surveillance testing of other channels per specification 4.3.1.1. e 4 n ~ ( 1%DDAM NECK-3/4 3 6 -.----,egyre.-.,c. ,c..%..e.-.-,--c.,,..,, we,,-,.mm ,,.-wm,mm.- .,,.m.,-w..,~ . -, + -.s---

f TAttf 3.3 1 (Continued) j e ACTION STATEMENTS (Continued) l ACTION 6: With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed until perfomance of i the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 5 hours; however, the taoperable channel may be bypassed up to 8 hours for surveillance testing of other channels per Specification 4.3.1.1. t 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 hours. 1 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 hours. ACTION 9:- With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, STARTUP and/or POWER OPERATION may proceed until

i..

performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided that the inoperable channel is placed in the tripped condition within I hour.- 1 ACTION 10: ' With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement-for Modes 3, 4, 5, restore the inoperable channel to OPERABLE status within 48 hours or open the Reactor Trip System breakers within.the next hour. ACTION ll: With the' number of OPERABLE channels one less than the Minimum Channels - 0PERABLE requirement;be in at least il0T STANDBY within 6 hours. ACTION llA: With the number of OPERABLE channels one less than the minimum channels OPERABLE requirement, be in at least il0T STANDBY within 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1, provided the other channel-is OPERABLE. HADDAM NECK 3/4 3 7 o m h w ---.,_-+_____-.___..u_ _v.m,,..- ,--m.. ..,m_, ,m,-, ..y. _g

+ -. ~ ~ .i TABLE'4.3-1 .x E REACTOR TRIP SYSTEM INSTRWWATION SURVEILLAKE REWIMIEEE n Yt ' TNIP ANALOG ACTMTIIE. M FM ~ CMMEL' OEVICE -IRIICN CHAIRIEL CMMNIEL' - OPERATIONRL OPEMTISML SNRVEILL M i FUNCTIOML UllIT CNECK CALIBRATION TEST -- TEST IS RENIRED { l. Manual Reactor Trtp N.A. N.A. N.A. S/U' ~I,~2 -3*, 4*, 5* y 2. Power Range, Neutron Flux D(2,3)' R(3) B N.A. 11, 2 a. High Setpoint S l b. Mid Setpoint S R(3) - B _ N.A.- IQ l c. Low Setpoint. S R(3) - B N.A. I , 2, 3*, 4*,.5* l 3 3. Intermediate Range, Neutron Flux, Nigh Startup o 2 Rate S R(3) S/U(6) N.A. I(d), 2, 3*,.4*, 5 t i Y a, 4. Pressurizer Pressure-- t Variable Low S R SW N.A. 1(a)- l 5. Pressurizer Pressure-- i High S-R SW N.A. 1, 2 j 1 6. Pressurizer Water Level-- [ High S R SW N.A. 1, 2, 3** 7. Reactor Coolant Flow--Low S R (7) N.A. l(a) [ l i + ( _ 4 l t n,,. r .n. v .,-,,w n, - -, - - =

~ i 1AOLE 4 J-1 (Contissedt [ g E9Cist_1 RIP ST5TEII IIISTMENTATIM SERKILUBEE MEDIMMES h ._ k 1Rh i . E-JuulLOS ACTMTINE= M fM : N amusIEL DEVICE ~ WICN J' UWWEIEL OWWIIEL OPUIATISINIL OPERATIE N. SURUEILIM RINE110lHE MIIT CNECK~ CAllOllATIM TEST TEST 15 EEMIME i I S. Steam Flow--Nigh 5 R(5)- R(5): N.A. I, t 9. Steam Generater Water i Level--Lew Cetacident with SteasqfFeeduster j Flow Mismatch 5 R - R N.A. 1.'2-l

10. Underveltage - Reactor I

Coolant Pumps N.A.- R .N.A. R I,I s

11. Safety Injectlen N.A.

N.A. N.A. R-1, t. w E

12. Ileacter Coolant Peup I

l Breaker Positten Trty N.A. N.A. N. A. - R I*I-

13. Meta Steam Line Trip

{ Valve Closure N.A. N.A. N.A. R-1(a) } j I4. Terttee Trty -N. A. N.A. N.A. R 1(a) [ I i i t n. -t i T k k j i t I i l h

a. =; ~ g

q

'[ TABLE 4.3-1 (Continued) ~ i.- F l ItEACTOR TRIP SYSTER IRSTRINEIRATION SURVEILUEZ llENIMIERIS n TRIP AMLOG -ACTMTIM ' W FM - CHAMEL DEVICE WIGI 1 ~ CHRMEL CHRIBIEL . OPERATIOML OPERATient SMUEILUUKE l IUNCTICIIK UIIIT CHECK CALIORATI0ll TEST TEST IS RE WIRED. _

15. -Reacter Trip System j-Interlocks a.

Low Power Block, P-7 N.A. R R N.A. IbI i

b.. Low Flow Pemissive, N

P-8 N.A. R R N.A. .I i )

16. Reacter Trip System N.A.

N.A. N.A. S/U(1,4) 1, 2, 3*, 4*, 5* j Breakers u l 2 17. Source Range Start--up Rede. 5 N.A. S/U (6) N.A. 1(d) I Rod Stop w l i I l l l t i . - - ~.. -.

.ya E 4.3 1 (Continued) TABLE NOTATIONS J ) With the Reactor Trip System breakers in the closed position and the i 4 Control Rod Drive System capable of rod withdrawal. ] qll THERMAL POWER above 105 of RATED THERMAL POWER. I L THERMAL POWER 1 745 of RATED THERMAL POWER. i f h THERMAL POWER above 105 but below 745 of RATED THERMAL POWER. d b THERMAL POWER below 105 of RATED THERMAL POWER. i (1) If not perfomed in previous 31 days. l (2)- Comparison of calorimetric to excore meer indication above 15% of i RATED THERMAL POWER. Adjust excore etannel gains consistent with calorimetric power if absolute difference is greater than 25. The provisions of Specification 4.0.4. are not applicable for entry into l MODES 1 or 2. This requirement is not applicable when the Power Range Channels have L had their gains adjusted to maintain the 95 trip margin for steady i state conditions at power levels other than 16%, 65%, and 100% RATED THERMAL POWER. When this exception is used,.a heat balance calculation i will-continue to be performed on a daily basis to detemine core power, t and the power range channels will be verified daily to be 95 below the i i - selected overpower trip setpoint. f

p

'(3) Neutron detectors may be excluded from CHANNEL CALIBRATION. f: L (4) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall. independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip System breakers. (5) Following a refueling outage, the calibration is performed subsequent 1 l 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, t

l 1

- May be bypassed when the reactor is at least 1.5% A k subcritical.

l 1 LI HADDAM NECK 3/4 3 11 I .a + .,--....,..--.,,,.,.-,.,w -._,--rr------- -m . ~ - - - - -- -- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' - - ' - ' ' - ~

II ' Mr CJ/Qf e u i

- IMTR13EMAT3gni 1/4.3.2 8'ul:1Nrrarn R&FETY FEATURE E TuhTION SYSTEN IW1TRUMENTAT10W trettitNR P M hlT10N FOR OPERATION 3.3.2 The Engineered Safety Feature c uation System E5FAS) testrumentation channels and interlock shown in Table 3(.3 2 s At NE with their Trip setpoints set consistent with the values shown in

'She Trip Setpoint column of Table 3.3 3. i M It.ARILITY: As shown in Table 3.3 2. l EIlG5: i a. With an t$FAS 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 ACT!0N requirement of Table 3.3 2 until the channel -is restored to OPERABLE status with the Trip setpoint adjusted consistent with the Trip 5etpoint value. I 'b. With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-2. 1 't SURVEILLANCE REQUIREMENTS

,(

_= 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. l l p 9 t 9 8 l MDOAM NECK 3/4 3 12 9 9 --,,----.,oem--.m.,..ms. gm.,,w,-,,-,mmev.w y-.a-N,' rwa,, w-w.reww w e m.,-, - - - w__-__,,,a_ -.me m-e. r ne-w,,-,_ww.eene.-_ -w-

77, [ ~ ~ ~ .~ - .h.* TABLE 3.3-2 i.. =--- smi,,uma s=orrrmEs Acurim mi l g winuass. g -TOTAL NO.- OWINELS - CMfidWELS APPL 10 K E OF CHRIGIELS TO ACTURTE OPERRSLE N M t l E FUNCT10NRL UNIT s 1.- Safety Injection (Reactor l Trip, Start Diesel Generators, a- !~ Containment Isolatten). i i a. Manual In1tiatien 2 1-2 1, 2, 3 ' 23 l b. Centainment. 6(3/ train) 2 in any one 6 (3/trale). I, 2, 3 29 l train Pressere--Migh c. Pressurizer 3-2 2' 1, 2, 3*. 24 t Pressure--Lew' r i 1 2. Steam Line Isolatten j w 1, 2, 3 l w a. Steam Fles in Two l 1 [ Steam Lines--Migh I 1/ steam line 1 in each of I/steamline 21 j

1) roer Leops any 2 steam l

Operating lines U S 22' l

2) Three Leops 1/eperating 1**/any 1/sperating j

l steam line - operating steam line -l i Operating steam 1tne i l i i t . c-- F l N l Co I O cr> W i i i 4 i l 1

I' i T__past 3.3 2 (Csatfamedi~ / ~ e -- i - DEsawaiir 5fetii FEATURES K1MT1W inns in._ =ums f g 10TAL NO. OWWIEL5 OWERIEL5 frFLI N 0F QWWWIEL5 To ACTURIE OPERABLE _IESE5 M e, q E FUNCTI0lWIL Ultif l 'f j-M Auxiliary Feeduster 3. ~ ~ l

e.. wide mange Ste. sen. unter tevel-.

Low 1(*I(*I 21 2 channels s l a i) Fear teops in any one operating train 1(*I(*I N t channels s-e l

2) Three to ps in any one l

Operating train l 1(*I N l 1 from each 1/pamp i 1/peup w f . b. Trty of All Main peup i Feedwater Peeps l t - r ,e, 1 i 4.16 W 9es under-3/hes 2/tes 2/tes 1, 2, 3, 4 N 4. Emergency Bus Undervoltage b l i a. l voltage-Level I l 1, 2, 3, 4 N l 3/ bus 2/bes 2/bes b. 4.16 W 9es Underveltage-t Level 2 3/ bus 2/ bus 2/bes 1, 2 N l 4.16 W Sus Underveltage-4 c. Level 3 N l' .h f O n\\ f 2 's 'h { i

y. c .s TABLE 3.3-2 (Contlawedi ~ w ENGINEEltED SAFETY FEATW ES ACI M TION SYST M I m ~ m NImanun ~ g TOTAL uD. OWWWIELS CNUWEELS APPLICABLE FUIICTIOML WIT OF CHMOIELS ID_ACTMTE OPERABLE IMOES..__ glig 5. Containment Isolatten i (Centainment. Air l Recirculation System, Feedwater l Isolatten, Safety Injection) l a. Centainment Pressure- ? High 6(3/Trate) 2 in any one 6 (3/Trafn) I, 2, 3, 4 20 i ,i. train i b. Safety Injectlen See Item 1. above for a11 Safety Injection intitattag functless and i j, requirements. w b8 i e l i t i 1 s I-l I l } 4 i i l _ _. -.... _, _ _ _. - - _ _ - _ _ _. _ _. _. ~ - _ _ _ _ _ _ _ _ _ _ _ _ _.. _. _

4 /l4 C/ t C l 1. Yatt t a.a.f f ront1-H) f

t MBT &Tiant

' 'l RCS pressure is less than l Trip function'ety be bypassed in this NDDE when '8800 psig.

- Thechannel(s)associatedwiththeprotectivefunctionsderivedfr est-of service reactor coolant loop shall be placed in the trippe "8

'(a) THElgEL POWER above 105 of RATED THENEL POWER. I For Surveillance Testing purposes, the train being tested will bl (b): l 'placed in ' Defeat

function.-

5 1 ACTION STAftMfMT1 J With the number of OPERA 8LE channels one les f Minimum Channels OPERABLE requirement, be in at least ACTION 20 STANDBY within 4 hours and in COLD SHUTDOW following 30 hours. f With the number of OPERA 8LE channels one les MinimumChannelsOPERABLErequiremen i ACTION 21 - i ANALOG CHANNEL OPERATIONAL TES l 1 With a channel associated with an operating loop inopera restore the inoperable channel to OPERABLE status with ACTION 22 ' hours or be in at least HOT STAND 8Y within t l and in at least NOT $NUTDOWN within the following 6 hourl 1 With the number of OPERABLE channels one l Minimum Channels OPERABLE requirement, restore ACTION 23 channel to OPERABLE status within 48 hou i NOT STANDBY within the next 6 hours and in within the following 30 hours. L I. 3 3/43-16 HADOAM NECK

l f

F 3.3 2 (Conti g i

ETION ITATDIDffs (Continued) i 1 l ACTION 24 With the number of OPERABLE channels one less than the Total Number o.1 Channels. STARTUP and/or POWER OPERATION may i proceed provided the following conditions are satisfied: L a. The inoperable channel is placed in the tripped condition within l I hour, and I b. The Minimum Channels OPERABLE requirement is est however, one additional channel any be bypassed for up to I hours for surveillance testing of other channels per Specification 4.3.2.1. l - ACT10N tl With less than the Minimum Number of Channels OPERABLE, within I hour detemine by observation of the associated pemissive annunciator 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 i Minimum Channels 0PERABLE requirement, restore the inoperable channel to OPERABLE status within 24 hours or reduce the THERMAL POWER to below 10% of RATED THERMAL POWER within the following I hour. t t j b 4 4 9 6 h R HADOAM NECK 3/4 3-17 4 =

3 TABLE 3.3-3 (Continoedl ~ h ENGINEENED SAFETY FEATURES ACTMTION SYSTUI INSTNWENTATIS 1 RIP 6 o l E i E FUNCTIolWIL WIT TRIP SETFOINT ALLME.E BRM R 4. Emergency Bus Undervoltage. ~ l l a. 4.15 kV Ses Underveltage - 12879** volts < 22784** volts l l Level I b. 4.16 kV 9es Underveltage - 23604 volts 23664 volts i* Level 2 .with a g second with a 8 second l time-delay. to It second time j deiay. c. '4.16 kV Sus underveltage - 24019 volts with 23999 volts with a t i w Level 3 a 9 second time 8 second to le 2 delay. second time delay. l 5 5. Containment Isolatten j (Containment Air Recirculatten System, Feedwater [ j Isolatten, Safety Injection) t i a. Containment Pressere - Nigh 14.7 psig 55.9 psig [ b. Safety Injection See Ites I. above for all Safety Injectlen Trly Setpoints and i allowable values. i TABLE NOTATIONS ] l Rated Thermal Power Setpoint is by tap position. Time delay of device is inverse function of voltage. Device must change i state within 0.95 - 1.05 seconds when the input voltage to the device goes free nennel to 50% of tap setting voltageins+cotaneously. { -l [ l l T l~ ~[ .. ~... ... ~.

1 .; TABLE 4.3-2 fContinued)' b ENGINEERED SAFETY FEATURES ACTMTION SYSTEM INSTRWE NTATION g SURVEILLANCE REQUIREE NTS i5i TRIP ANALOG ACTUATING-MODES FOR' CHANNEL DEVICE WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST IS REOUIRED 3. Auxiliary Feedwater a. Steam Generator Water S R M N.A. 'l l Level-Low i l b. Trip of All Main M.A. N.A. M.A. R 1 i Feedwater Pumps i s 4. Emergency Bus Undervoltage I Y j 5 a. 4.16 kV Bus - N.A. R N.A. M* 1, 2, 3, 4 Undervoltage - level I b. 4.16 kV Bus N.A. R N.A. M* 1, 2, 3, 4 Undervoltage - i l Level 2 c. 4.16 kV Bus N.A. R N.A. M* 1, 2 Undervoltage - Level 3 Trip actuating device operational test is defined as a test of each individual channel. A complete logic test will be performed on a refueling outage basis. On a monthly basis, an undervoltage condition will be initiated at the sensing device to verify the operability of the trip actuating device and verify that the associated alarm relays operate. l ( = w .u w-sa -u wu .xu--- a-..a----- - - - -. - ~,

4.

i ' REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) i 7) Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural l [' integrity in the event of an Operating Basis Earthquake, a loss of-coolant. accident, or a steam line or feedwater line L break as specified in Specification 4.4.5.3c., above; i 8) Tube Insnection means an inspection of the steam generator i tube from the hot leg entry point completely around the f -U bend to the top support of the cold leg; or an inspection from the point of entry (hot leg or cold leg) completely around the U-bend to the opposite end. L i b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or sleeve

all tubes exceeding the plugging limit as defined in Specification 4.4.5.4.a.6 and plug all defective sleeves) required by Table 4.4-2.

4.4.5.5 Reports a. Following the completion of each inservice inspection of steam generator tubes, a Special Report documenting the number of tubes plugged or sleeved in each steam generator shall be reported to the Commission within 15 days pursuant to Specification 6.9.2; b. The complete results of the steam generator tube inservice i inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 90 days following the completion of the inspection. This Special Report shall include: 1) Number and extent of tubes inspected, 2) location and percent of wall-thickness penetration for each indication of an imperfection, and 3) Identification of tubes plugged or sleeved, c. Results of steam generator tube inspections which fall into Category C-3 shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 prior to resumption of plant operation. This report shall provide a description of r L investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence. Tube sleeving shall be performed in accordance with the Connecticut Yankee Steam Generator Sleeving Report, Revision 1, transmitted by letter dated January 7, 1986. 1 HADDAM NECK 3/4 4-26 i

TABLE'4.4-2 STEAM GENERATOR TUBE INSPECTION lst SAMPLE INSPECTION 2nd SAMPLE. INSPECTION 3ni SAMPLE INSPECTION-Sample' Size Result Action Required Result Action Required' ' Result Action Required 4 m R A cinimum of C-1 None N.A. M.A. M.A. M.A. S Tubes per S.G. C-2 Repair defective tubes C-1 None' N.A. N.A. and inspect additional 25 tubes in this S.G.* C-2 Repair defective C-1 None tubes and inspect ~ C-2 Repair defective tubes.* C-2 Repair defective tubes.* -additional'4S C-3 Perform action for C-3 tubes in this. result of first sample. S.G.* i C-3 -Perform action for N.A. N.A. C-3 result of first sample. s 4 C-3 Inspect all All other None N.A. N.A. tubes in S.G.s are M this S.G., repair de-C-1. fective tubes and Some S.G.s Perfone action for N.A. N.A. inspect 2S tubes in C-2 but no C-2 result of each other S.G.* additional second sample. S.G. is C-3. Notification Additional Inspect all to NRC pursuant S.G. is C-3. tubes in each N.A. N.A. to 50.72(b)(2) S.G. and repair of 10 CFR Part 50. defective tubes.* Notification to NRC pursuant to 50.72(b)(2) of 10 CFR Part 50. S=3%x[g] Where N is the number of steam generators in the unit, and n is the number of steam generators' inspected during an inspection. Repair of defective tubes shall be limited to plugging with the exception of those which may be sleeved. Tubes with defective sleeves shall be plugged.

e r.- p REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS r i LIMITING CONDITION FOR OPERATION { p 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:- a. The Containment Atmosphere Gaseous Radioactivity Monitoring System, l and b. The Volume Control Tank Level Monitoring System and the Containment Main Sump Level (Narrow Range) Monitoring System. APPLICABILITY: MODES 1, 2, 3, and 4. E ACTION: a. With the Containment Atmosphere Gaseous Radioactivity Monitoring System Inoperable, operation may continue provided grab samples of the containment atmosphere are obtained and analyzed for gross noble gas activity at least once per 24 hours. Restore the inoperable l monitor to OPERABLE status within 7 days or, prepare and submit a special report to the Commission pursuant to Specification 6.9.2 within 30 days outlining actions taker, cause of inoperability and i plans for restoring the monitor to OPERABLE status. b. With the Volume Control Tank Level Monitoring System inoperable, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours, c. With the Containment Main Sump Level (Narrow Range) Monitoring System inoperable, restore the inoperable level monitoring system to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHVTDOWN within the following 30 hours. d. With both the Containment Main Sump Level (Narrow Range) Monitoring System and the Containment Atmosphere Gaseous Radioactivity Monitoring System inoperable, operation may continue provided grab samples of the containment atmosphere are obtained and analyzed for gross noble gas activity at least once per 4 hours. Restore the Containment Main Sump Level (Narrow Range) Monitoring System to OPERABLE STATUS within 72 hours from time of its initial loss or be L in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. L l L i L HADDAM NECK 3/4 4-29

LIMITING CONDITION FOR OPERATION (Continued) REACTOR COOLANT SYSTEM d. If Residual Heat Removal Pump Seal leakage exceeds its allowable limit, monitor this leakage for at least 36 continuous operating hours. If the leak rate does not decrease or stabilize prior to reaching three liters per hour, the provisions of action statement

b. are applicable.

SURVEILLANCE REOUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by: a. Monitoring the containment atmosphere gaseous radioactivity monitor at least once per 12 hours; b. Monitoring the containment sump inventory at least once per 12 + hours; c. Monitoring the Volume Control Tank level at least once per 4 hours; d. Performance of a Reactor Coolant System water inventory balance at least once per 24 hours; e. Monitoring the Reactor Head Flange Leakoff System at least once per 24 hours; f. Monitoring the RHR Pit, HPSI, and charging pump cubicles for indications of leakage at seals, flanges, and valves at least once per 12 hours, g. Performance of an operational leak rate test, for those portions of the HPSI, charging and RHR system outside containment used for recirculation, at least once per month. The test shall be conducted at a hydrostatic pressure corresponding to the operating pressure under accident conditions. h. Monitoring leakage through cach of the following ECCS valves (SI-CV-862 A, B, C, and D and SI-CV-872 A and B): 1) At least once per 18 months, 2) Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours or more and if leakage testing has not been performed in the previous 9 months, 3) Prior to returning the valve to service following maintenance, repair, or replacement work on the valve, and 4) Within 24 hours following v:1ve actuation due to flow through the valve. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4. HADDAM HECK 3/4 4-32

RMCTOR COOLANT SYSTEM . BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued) The total steam generator tube leakage limit of 0.4 gpm for all steam L_ generators not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of either a steam generator tube rupture or steam line break. The 0.4 gpm limit is consistent with the assumptions a used in the analysis of these accidents. The 150 gpd leakage limit per c steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions. The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems. The 12 hour surveillance for leakage monitoring using the containment atmosphere gaseous radioactivity and containment sump inventory indications is a qualitative check of these indications for obvious signs of an increase in RCS leakage. The 4-hour surveillance for leakage monitoring using the VCT level indication is a quantitative determination of RCS leakage. Since constant RCS conditions are required for a valid calculation, this quantitative determination is required during periods of constant reactor )ower and RCS temperature only. During RCS transients, the VCT level will se monitored at least once per four hours but no quantitative determination is required. The 12-hour surveillance requirement for leakage monitoring for RHR, HPSI, and charging systems involves general inspection of the RHR pit, HPSI, and charging pump cubicles for any obvious signs of leakage. The limitation on the combined leakage from the RHR System HPSI, and charging systems provides an indication of impending seal failures. The radiological consequences associated with the combined leakage is accept-able. This leakage will be considered as a portion of the IDENTlf!ED l LEAKAGE. The specified allowable leakage from the listed ECCS valves is sufficiently low to ensure early detection of possible check valve failure. It is apparent that when pressure isolation is provided by two in-series valves and when failure of one valve in the pair can go undetected for a substan-tial length of time, verification of valve integrity is required. Since these valves are important in )reventing overpressurization and rupture of the ECCS low pressure piping w11ch could result in a LOCA that bypasses containment, these valves should be tested periodically to ensure low probability of gross failure. The Surveillance Requirements for these ECCS valves provide added assurance l of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from these ECCS valves is IDENTI-i FIED LEAKAGE and will be considered as a portion of the allowed limit. HADDAM NECK B3/4 4-6

lpdwp {

o REACTOR COOLANT. SYSTEM l

BASES 3/4.4.7 CHEMISTRY l The limitations on Reactor Coolant System chemistry ensure that corrosion of j the Reactor Coolant System is minimized and reduces the potential for L Reactor Coolant System leakage or failure due to stress corrosion. Main-taining the chemistry within the Steady-State limits provides adequate i corrosion protection to ensure the structural integrity of the Reactor l c Coolant System over the life of the plant. The associated effects of i exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady-State Limits, up to .the Transient Limits, for the specified limited time intervals without t having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued o)eration within the restrictions.of the Transient Limits provides time for ta(ing corrective actions to restore the contaminant concentrations to within the Steady-State Limits. The Surveillance Requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take correc- .tive action. 3/4.4.8 SPECIFIC ACTIVITY [ The limitations on the specific activity of the reactor coolant ensure that l the resulting 2-hour doses at the SITE BOUNDARY will not exceed an appro-I. priately small fraction of 10 CFR Part 100 dose guideline values following a l steam generator tube rupture accident. L The ACTION statement permitting POWER OPERATION to continue for limited time periods with the reactor coolant's specific activity greater than l 1 microcurie / gram DOSE EQUIVALENT I-131, but within the allowable limit ( shown on Figure 3.4-2, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific activity levd s exceeding 1 microcurie / gram DOSE EQUIVALENT I-131 but within the limits shown on Figure 3.4-2 must be restricted to no more than 800 hours per year (approximately 10% of the unit's yearly operating time) since the activity-levels allowed by Figure 3.4-2 increase the 2-hour thyroid dose at the SITE BOUNDARY by a factor of up to 20 following a postulated steam generator tube rupture. The sample analysis for determining the gross specific activity and E can exclude the radioiodines because of the low reactor coolant limit of 1 microcurie / gram DOSE EQUIVALENT I 131, and because, if the limit is i exceeded, the radioiodine level is to be determined every 4 hours. If the i gross specific activity level and radiciodine level in the reactor coolant were at their limits, the radioiodine contribution would be approximately 1%. In a release of reactor coolant with a typical mixture of radioa-ctivity, the actual radioiodine contribution would probably be about 20%. Reducing T to less than 500'F prevents the release of activity should a steamgeneNEortuberupturesincethesaturationpressureofthereactor I: HADDAM NECK B3/4 4-7 1

M y' ..n . REACTOR COOLANT SYSTEM ( .- t BASES 3/4.4.8 ' SPECIFIC ACTIVITY (Continuedt b ' coolant is below the lift pressure of the atmospheric steam relief valves. zThe Surveillance Requirements provide adequate assurance that-excessive specific activity levels in the reactor coolant will be detected in suffi-cient time to take corrective action. A reduction'in frequency of isotopic - analysis following' power changes may be permissible if justified by the data M obtained. 3/4.4.9 PRESSURE / TEMPERATURE LIMITS i Th'e temperature and aressure changes during heatu) and cooldown-are limited l to be consistent-wit 1 the: requirements given in tie ASME Boiler and Pressure ~ . Vessel Code,.Section III, Appendix G: ~ i 1. ..The reactor coolant temperature and pressure and system heatup and 'j cooldown rates (with the exception of the pressurizer) shall be limited in:accordance with Figures 3.4-3, 3.4-4 and 3.4-5 for the service period specified thereon: a '.. - Allowable combinations of pressure and temperature for s)ecific temperature change rates are below and to the right of tle limit l lines shom. Limit lines for cooldown rates between those pre-sented may be obtained by interpolation; and 1 b. Figures 3.4-3, 3.4-4 and 3.4-5 define' limits to assure prevention r? non-ductile failure only. For normal operation, other inhereat i plant characteristics, e.g... pump heat addition and pressurizer L 'aeater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges. 2. These limit lines shall be calculated periodically using methods provided below, ] 3. .The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70'F, f 6 . 4'. The. pressurizer heatup and cooldown rates shall not exceed 100*F/hr and L -?00*F/hr,respectively. l S. System preservice hy/ _1sts and inservice leak and hydrotests shall be performed at presst.as in accordance with the requirements of ASME i Boller and Pressure Vessel Code, Section XI. j The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the NRC Standard Review Plan, ASTM E185-73, and in accordance with additional reactor vessel requirements. .These proparties are then evaluated in accordance with Appendix G of the 11976 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel Code and the calculation methods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves", April 1975. HADDAM NECK. B3/4 4-8 a t

( ) E'om REACTORCOOLANTSYSTEM- -BASES ~ 3/4.'4.9 PRESSUf!E/ TEMPERATURE LIMITS (Continued) Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RT at the end of 22 o effective full power years (EFPY) of service life. YN,22EFPYservicelife period:is chosen.such that the limiting RT at the 1/4T location in the core region is greater than the RT of tNOTlimiting unirradiated material. i TheselectionofsuchalimitingRYDT assures that all components in the ReactorCoolantSystemwillbeoper$Ndconservativelyinaccordancewith applicable Code requirements. The reactor vessel materials have been tested to determine their initial RT . Reactor operation and resultant fast neutron (E greater than 1 MeV) irhNiationcancauseanincreaseintheRT Therefore, an adjusted reference temperature, based upon the fluenUOT.and chemical content of the material in question, can be predicted. The heatup and cooldown limit curves of Figures 3.4-3, 3.4-4, and 3.4.5 include predicted adjustments for this, shift in~RT NDT* Allowable pressure-temperature relaticnships for various heatup and cooldown rates-are calculated using methods darived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50. L The general method for calculating heatup and cooldown limit curves is based upon the 3rinciples of the linear elastic fracture mechanics (LEFM) technol-L ogy. In tie calculation procedures a semielliptical surface defect with a ' depth of one-quarter-of the wall thickness, T, and a length of 3/2T is assumed to exist at-the inside of the vessel wall as well as at the outside of the vessel wall. The dimensions of this postulated crack, referred to in i Appendix G of ASME Section III as' the reference flaw, amply exceed the I current capability of inservice inspection techniques. Therefore, the L reactor operation limit curves developed for this reference crack are I conservative and provide sufficient safety margins for protection against nonductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference temperature, RT is used and this includes the radiation-induced shift, delta RT UNe,spondingtotheend L of the-period for which heatup and cooldown cuYNs, care generated.

The-ASME approach for calculating the allowable limit curves for various i,

heatup and cooldown rates specifies that the total stress intensity factor, .K, for the combined thermal and pressure stresses at any time during heatup of cooldown cannot be greater than the reference stress intensity factor, L K for the metal temperature at that time. K is obtained from the rhe,rence fracture toughness curve, defined in dhpendix G to the ASME Code. The K is given by the equation: IR KIR - 26.78 + 1.223 exp (0.0145(T-RTNDT + 160)) O) Where: K is the reference stress intensity factor as a funttion of the metal temkrature T and the metal nil-ductility reference tercperature RT NDT* L HADDAM NECK B3/4 4-9

^ p [ REACTOR COOLANT SYSTEM-BASES 3/4.4.9 PRESSURE / TEMPERATURE LIMITS (Continued) Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G.of. the ASME Code as follows: )' less'than or equal to K (2) CKgg + kit IR Where: Kjg - the stress intensity factor caused by membrane (pressure): , stress, the stress intensity factor caused by the thermal-gradients, .K It constant provided by the Code as a function of temperature K IR = relative to the RT of the material, NDT C - 2.0 for level A and B service limits, and 1 - C-- 1.5 for inservice hydrostatic.and leak test operations. At any. time during the heatup or cooldown transient, K is determined by the metal temperature at the tip of the postulated flak the appropriate value for RT , and the. reference fracture toughness curve. The thermal stresses resdNing from temperature gradients through tb vessel wall are + calculated and then the corresponding thermal stress intensity factor, K for the reference flaw is computed. From Equation (2) the pressure streN, ~ ' intensity ~ factors are obtained and, from these, the allowable pressures are calculated. C00LDOWN For the calculation of the allowable pressure versus coolant temperature during cooldown, the Code reference flaw is. assumed to exist at the inside c .of the vessel wall. During cooldown, the controlling location of the flaw l 1s always at the inside of the wall because the thermal gradients produce L tensile stresses at the inside, which increase with increasing cooldown L rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, . composite limit curves are. constructed for each cooldown rate of interest. l The use.of the composite curve in the cooldown analysis is necessary because control of the cooldown-procedure is based on measurement of. reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to i the. vessel ID. This condition, of course, is not true for the steady-state L situation. It follows that at any given reactor coolant temperature, the delta T developed during cooldown results in a higher value of X at the 1/4Tlocationforfinitecooldownratesthanforsteady-stateopdEation. Furthermore, if conditions exist such that the increase in K exceeds K J the calculated allowable pressure during cooldown will be grdSter than tO, E steady-state value. r HADDAM NECK B3/4 4-10

,,g N REACTOR C00LANT' SYSTEM w 4 BASES m V '3/4.4.9 PRESSURE / TEMPERATURE LIMITS'(Continued) H .The above procedures. are needed because there is no direct control on ? temperature at the:l/4T location; therefore, allovable pressures may unknow-D ingly be violated if the rate of cooling is decreased at various intervals } along a cooldown ramp. The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire S cooldown period. HEATUP Three-separate calculations are required to determine the limit curvei for W finite heatup rates. As is done in the cooldown analysis, allowable j pressure-temperature relationships are developed for steady-state conditions I as.well as finite.heatup rate conditions assuming the presence of a 1/4T j defect lat the inside of the vessel wall. The thermal gradients during j heatup produce compressive stresses at the inside of the wall that alleviate l L the tensile stresses produced by internal pressure. The metal temperature i at the crack tip lags the coolant temperature; therefore, the K for the j steady-state conditions at the same coolant $m.for the 1/4T cra$ du 1/4T-crack during heatup is lower than the K l perature. During heatup, especially at the end of the transient, conditions may exist such that the i E effects of. compressive thermal stresses and different K 's for steady-state j andfinite.heatupratesdonotoffseteach;otherandthdRpressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for-finite heatup rates when the 1/4T flaw 'is considered.- Therefore' both cases have to be analyzed in order to-assure - that at any coolant temperature the lower value of the allowable pressure 1 calculated for steady-state and. finite heatup rates is obtained. l .The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside j surface flaw is assumed. Unlike the: situation at= the vessel inside surface, the~ thermal gradients established at the outside ' surface during heatup _ produce stresses which are tensile in nature.and thus tend to reinforce any i pressure stresses present. These thermal stresses, of course, are dependent on_both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Furthermore,_ since the thermal stresses, at the outside are ' tensile and increase with increasing heatup rate, a lower bound curve cannot l; be defined. Rather, each heatup rate.of interest must be analyzed on an ' individual basis. Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup -limitations because it is possible for conditions to exist such that over the~ course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion. HADDAM NECK B3/4 4-11

n r . REACTOR COOLANT SYSTEM BASES: PRESSURE)TEMPERATURELIMITS(Continued) i Although t'e pressurizer operates in temperature ranges above those for h which there is reason for concern of nonductile failure, operating limits are provided to. assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements. LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEMS The OPERABILITY of two Spring-loaded relief valves (SLRVs) or an RCS vont opening of greater than 7 square inches ensures that the RCS will be pro-tected from presure transients which could exceed the limits of Apaendix G to-10-CFR Part 50 when one or more of the RCS cold legs are less tlan or equal to-315'F. Either SLRV has adequate relieving capability to protect the .RCS from overpressurization when the transient is limited to either: (1) the start of ar. idle RCP with the secondary water temperature of the steam generator less than-or equal to 20'F above the RCS cold leg temperatures, or (2) the nart of a charging pump (centrifugal) and its injection into a-water-solid RCS. The Maximum Allowed SLRV Setpoint for the Low Temperature Overpressure Protection System (OPS) is derived by analysis which models the performance 4 of the OPS assuming various mass input and heat input transients. Operation with a SLRV Setpoint less than or equal to the maximum Setpoint ensures that Appendix-G criteria wi_ll not be violated with consideration for a maximum ' pressure overshoot beyond the_SLRV Setpoint which can occur as a result of t L time delays in signal processing and: valve opening, instrument uncertain-l

ties,-and single failure. To ensure that mass and heat input transients L

more severe than those assumed cannot occur, Technical Specifications require lockout of all but one charging pump (centrifugal or metering) while in MODES 4, 5, and 6 with the reactor vessel head installed and disallow L start of an RCP if secondary temperature is more than 20*F above RCS cold l leg temperature. The Maximum Allowed SLRV Setpoint for the OPS will be updated based on the results of ' examinations of reactor vessel material irradiation surveillance . specimens performed as required by 10 CFR Part 50, Appendix H, and in accordance with the schedule in Table 4.4-5. 1 3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, and l 3 components ensure that the structural integrity and operational readiness of these components will be mair.tained at an acceptable level throughout the ' life of the-plant. These programs are in accordance with Section XI of the - ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(g)(6)(i). 3/4.4.11 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the RCS that could inhibit natural circulation core L HADDAM NECK B3/4 4-12 =M-=r g + y y w +y-- a -^

f l s g I REACTOR COOLANT SYSTEM ,c BASES 3/4.4.11 -REACTOR COOLANT ' SYSTEM' VENTS (Continued) ] i cooling. The OPERABILITY ofLat least one RCS vent path from the reactor-vessel head and the pressurizer steam space ensures the capability exists to perform this function. ..n The valve redundancy of the RCS vent paths serves to minimize the proba- .bility of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply or control system does not prevent ' isolation of the vent path. l LThe. function, capability and testing requirements of the RCS vents are consistent with the requirements of Item II.B.1 of.NUREG-0737, "Clarifica-tion of TMI Action Plan Requirements," November 1980. L 4 E t b l i l r HADDAM NECK B3/4 4-13

+w m m s t c d' # REACTOR COOLANT SYSTEM _l PRESSURIZER LIMITING CGHDITION'FOR OPERATION- '3.4.9.2 . The. pressurizer temperature. shall be limited to: a.- A maximum heatup of 100'F in any 1-hour period, j ,,x b. A maximum cooldown of 200'F in any 1-hour period, .c. A maximum Reactor Coolant System (RCS) temperature difference of 200'F, and 4 . d.' - Greater than or equal to 70*E whenever pressurizer pressure-is greater than 500 psig. l APPLICABILITY:, At all times. ACTION:. With the pressurizer temperature limits in excess of.any of the above limits: l a. Restore the temperature to within the limits within 30

minutes, b.-

Perform an engineering evaluation to determine the effects of, the out-of-limit condition on the structural integrity of the pressurizer, and c. Determine that the pressurizer.. remains acceptable for continued operation. ' Otherwise..be in at least HOT STANDBY within the next 6 hours .and reduce the pressurizer pressure to less than 500'psig within the following 30 hours. SURVEILLANCE RE0VIREMENTS 4 4'4.9.2.1. The pressurizer temperatures for heatup and cooldown shall be determined to be within the litrits at least once per 30 minutes during system lheatup or.cooldown. ~4.4.9.2.2 The RCS temperature differential shall be determined to be within

the limit at least once per 12 hours.

4.4.9.'2.3 The pressurizer pressure shall be determined to be less than 500 psig at least once per hour whenever the pressurizer is not drained and the i -pressurizer temperature is less than 70*F. l i HADDAM NECK 3/4 4-45

f 5 EMERGENCY CORE COOLING SYSTEMS i SURVEILLANCE REQUIREMENTS k ~ -4.5.1 Each ECCS subsystem shall be demonstrated OPERABLE: .a. At-least-once per 12 hours by: verifying.that the following valves J are in the indicated conditions: m Valve Nu:aber ' Valve Function Valve Position RH-FCV-602 RHR Heat Exchanger Blocked closed. Operator air Bypass Line supply; isolated. RH-FCV-796-RHR Heat Exchanger Blocked open position. Discharge Line-Operator air supply isolated.

RH-M0V 22 Containment Sump C1osed. Manual Operator is Suction Line

-locked. SI-MOV 24 RWST Line Locked open. Operator circuit breaker locked open. i RH-MOV-874 RHR Recirculation Locked closed. Operator Line circuit breaker locked open. SI-MOV-854A-HPSI Pump Open. Manual Operator is L Suction Line locked. l SI-M0V-854B_ HPSI Pump Open. Manual Operator is Suction Line locked. L L SI-MOV-901 RHR/HPSICrosstie Closed. Manual Operator is f

locked, s

SI-MOV-902 RHR/HPSI. Crosstie Closed. Manual Operator is locked. SI-M0V-903. HPSI Miniflow Open. Manual operator is locked. SI-MOV-904 HPSI Miniflow Open. Manual Operator is locked. b. On start-up prior to entering Mode 4: VALVE N0. LOCATION ACTION SI-V-905 HPSI Loop 1 Valve blocked and locked Injection Line in throttled position. SI-V-906 HPSI loop 2 Valve blocked and locked injection line in throttled position SI-V-907 HPSI loop 3 Valve blocked and locked injection line in throttled position. HADDAM NECK 3/4 5-2 1

i a j[ '*; fi ' ' TABLE 4.5-1 SAFETY INJECTION ACTUATED AUTOMATIC' VALVES g 1 h 1 VALVE NUMBER SI POSITION j 8: .SI-MOV-861A Open .SI-MOV-861B Open 4

SI-MOV-8610 Open fSI-MOV-861D'

-Open 1 SI-MOV-871A-Open SI-MOV-871B: Open-1 BA-MOV-373 Open l BA-MOV-32 Open-LD-MOV-200 Closed r .LD-TV-230 Closed CH-MOV-257 Closed ex;' .CH-MOV-2578 Closed 1 .CH-SOV-242-Closed' ? CH-SOV-242B Closed. l I TABLE 4.5-2 g ECCS MANUAL VALVES l Valve Number 9

SI-MOV-24 l

JSI-MOV-873 k SI-MOV-854A SI-MOV-854B SI-MOV-901 ( SI-MOV-902 D SI-MOV-903 ~ SI-MOV-904 i RH-MOV-33A l. RH-MOV-33B \\ l

i; F

L~, I! l!. li -HADDAM NECK 3/4 5-6

g y ~ ~ I / l X ss t . EMERGENCY CORE COOLING SYSTEMS ,ygg SURVEILLANCE REQUIREMENTS I [4 ~.' 5. 2 ; a;; '.The ECCS subsyst5ms: shall be' demonstrated OPERABLE per the ~ applicable Surveillance Requirements of Specification'3.5.1 i y*L with the-exception:that, for valves RH-FCV-602 and~ RH-FCV-796' : restoration of valve: controls be allowed. b.-,0ne: centrifugal: charging pump.and: both High Pressure Safety. Injection' pumps:shall be. demonstrated inoperable at least once per 31 days whenever the-temperature of one or more of ~ the RCg' cold-legs of an unisolated loop-is less than or' equal. to 315 F and the RCS-is not vented by a minimum opening of-3 p. " inches ((nominal diameter)< or its equivalent by verifying:! / el)) That'the High Pressure Safety Injection pump motor 4 circuit breakers are racked out and the cabinets locked, 1' ~ 2)- -That'High Pressure Safety Injection pump discharge. . valves SI-V-855A' and SI-V-855B are closed and-locked, and. ' 3) That the inoperable centrifugal charging pump's control switch is in the trip pullout position.and red tagged, "Do Not Operate."- s \\ .) l .e p i lc l HADDAM. NECK 3/4 5-8 =

1.: ~ +,, n n, y i V CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE ' LIMITING CONDITION FOR OPERATION '[

3.6.1.2, ' Containment leakage ~ rates shall be limited to:

s a.: An overall integrated leakage rate of less than or equal to' La, 5 0.18 % by weight of the containment air per 24 hours at Pa, 39.6 -psig,'or 'b. A combined leakage' rate of less than 0.60 La for all penetrations I and valves subject to Type B and C tests, when pressurized to Pa. APPLICABILITY:L N0 DES 1, 2, 3, and 4. AGl[0][: L. -With either the measured overall integrated containment leakage rate l L exceeding 0.75.La or the measured combined leakage rate for all' penetrations and valves subject to Types B and C tests exceeding 0.60 La, restore the overall integrated leakage rate to less than 0.75 La and the combined 3 leakage rate for all' penetrations subject to_ Type B and C tests to less than E 0.6g La prior to increasing the-Reactor Coolant System temperature above .200 F. p SURVEILLANCE REQUIREMENTS j "4.6.1.2 iThe containment leakage rates:shall be demonstrated at the E following; test schedule, and shall be -determined in conformance with the y criteria specified in Appendix J of 10 CFR Part 50 using the methods and 1 l' provisions of ANSI N45.4-1972: 1 'a. Three Type A tests (0verall Integrated Containment. Leakage Rate) shall-be conducted at 40 10 month intervals during shutdown at a pressure not less than Pa, 39.6 psig during each 10-year service period. ;The. third test of each set shall be conducted during the i l shutdown for the 10-year plant inservice inspection; i i \\?- 'HADDAM NECK 3/4 6-2

. b.l, hD sf ~

SURVEILLANCE REQUIREMENTS fContinuedF s

E l b.- If any periodic; Type A test fails to meet 0.75 La, the test-scheduleifor. subsequent' Type A tests shall be~ reviewed-and-J

~

approved by the Commission.' If two consecutive Type A tests fail to meet 0.75.La,c a Type A test'shall. be performed at least every 18 months until:two consecutive Type A tests meet 0.75 La at which i time the.above test schedule may be resumed or a corrective action -plan may be prepared'and submitted to the NRC that provides an acceptable alternative contingent on_NRC' approval, c. 1The, accuracy of'each Type A test shall be verified using the relationship: o- .(LTM + l f 0.25 L,) M I IlTM + lo + 0.25 L,F o c where:- -L is the percent measured containment. leakage per 24 hours at TM_ pressure--P, g Lg is the percent superimposed leakage, _ L is the percent leakage-obtained from.the ~ supplemental 4 c test result, and L, 'is. replaced with L for reduced pressure tests, t d.< Type B and C tests shall be conducted at intervals no' greater than 24' months and:at-a pressure not less than Pa, 39.6 psig, using halogen gas-detection, soap bubble, pressure decay, or other I . methods of-equivalent sensitivity, except for tests involving: s h 1) Air _ locks, and 2)- Purge supply and exhaust isolation valves with resilient material seals. e. ~ Air locks shall be tested and demonstrated OPERABLE by the L requirements of Specification 4.6.1.3; ~ f.- Purge supply and exhaust isolation valves with resilient material seals shall be tested and demonstrated OPERABLE by the requirements of Specification 4.9.9; g. The provisions of Specification 4.0.2 are not applicable for Specifications 4.6.1.2.a through 4.6.1.2.d. =HADDAM NECK 3/4 6-3 't

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3/4,y PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE-SAFETY VALVES 1

. LIMITING CONDITION FOR OPERATION { ' 3.7.1.1: All main steam line Code safety. valves associated with each steam p^' generator of an unisolated reactor coolant loop shall be OPERABLE with lift settings as specified in Table 3.7-1.

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

ACTION:; e, a. ~ With four reactor coolant loops and associated steam generators in operation and one or more main steam line Code safety valves . inoperable, operation in MODES 1,: 2, and 3 may proceed provided 1 .that.within 4 hours the inoperable valve is restored to OPERABLE status; otherwise, be in HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. b. > With three reactor coolant loops and associated steam generators. .in operation and one or more main steam line Code safety valves associated with an operating loop. inoperable, operation in MODES 1, 2, and 3 may proceed provided that within 4 hours the inoperable valve-is restored to OPERABLE. status;.otherwise, bo in- . HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within ..the-following -30. hours. SURVEILLANCE REQUIREMENTS 4.7.1.1.. In addition to the requirements of Specification 4.0.5, each main -steam line code safety valve associated with ecch steam generator shall be demonstrated OPERABLE by checking its setpoint each refueling..The l provisions of Specification 4.0.4 are not applicable for entry into MODE 3. m j., HADDAM NECK 3/4 7-1

h i C1 ( PLANT SYSTEMS: ' SURVEILLANCE REQUIREMENTS 4.7.6.1.1' The' Fire Water Supply / Distribution System shall. be demonstrated E OPERABLE:. a.- At least once per 31 days.on a STAGGERED TEST BASIS by starting - each pump and. operating it with flow. The diesel pump shall run i

c.

for'30 minutes and the. electric pump for 5 minutes. b. . At least once'per 31 days by verifying that each valve (manual, power; operated or automatic) in the flow path that is not locked, 1 sealed, or otherwise secured is in its correct position, i i At least once per 12 months by cycling each testable valve in the i c. flow path through at least one complete cycle of full travel, d. At least once per 18 months by performing a system functional test which. includes simulated automatic actuation of the system throughout its operating sequence, and -1) Verifying that each pump develops at least 2500 gpm at a system head'of greater than or equal to 100 psig at Rated

Speed, l.

p 2) Cycling each valve in the flow path that is not testable .during plant operation through at least one complete cycle of ! full travel, and 3) Verifying that each high pressure pump starts automatically and sequentially to maintain the Fire Water Supply / Distribution System pressure greater than or equal to 75 psig. L e. At least once per 3 years by performing a flow test of the Distribution System in accordance-with Chapter 5, Section 11 of l l. the Fire Protection Handbook,14th Edition, published by National l Fire Protection Association. 4.7.6.1.2 The fire pump diesel engine shall be demonstrated OPERABLE: i a. At least once per 31 days by verifying: 1) The fuel storage tank contains at least 130 gallons of fuel, r and l 2) The diesel starts from ambient conditions and operates for at L least 30' minutes with flow. L l' L HADDAM NECK 3/4 7-17

c < je '. 7 x,. PLANT SYSTEMS ' SURVEILLANCE REQUIREMENTS 4.7.6.2-Each of the above required Spray and/or Sprinkler Systems shall be demonstrated OPERABLE: a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path, that is not locked.- sealed or otherwise secured in position, is in' its correct-

position, b.

At least once'per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel, c. . At least once per 18 months: 1) By performing a system functional test which includes simulated automatic actuation of the system, and: a) LVerifying that the system alarm check / main actuation valves in the flow path actuate to their correct positions on a simulated test signal, and b) Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel. '2) By visual inspection of the spray headers to verify their m integrity,'and f r -3) By visual inspection of each nozzle to verify that there is 1 L no blockage'and that each. spray area is not obstructed. 4 d. At least once per 3 years by performing an air flow test through - each open: head system and verifying the supply piping and nozzles l are unobstructed. p u l.- L i (-l HADDAM NECK 3/4 7-20

p:' L-L 4 2 2 2 \\ y b' ' PLANT SYSTEMS J ~

CO2. SYSTEMS'

] LIMITING CONDITION FOR OPERATION' 3.7.6.3-The following High Pressure C0 Systems shall be-0PERABLE with the l minimum number of storage bottles shown $ach having a net weight of not less J J than 90% charge weight: a. Cable Vault - 18 bottles, and j b. Primary Auxiliary-Building Charcoal Filters - 8 bottles. APPLICABILITY: Whenever systems, structures, components, or equipment protected by the High Pressure ~C0 -Systems are required to be OPERABLE. 2 ACTION:- a. With one or more of the above required High Pressure-C0 Systems inoperable, within I hour establish a continuous fire w$tch with backup fire suppression equipment for the unprotected equipment and/orarea, b. The provisions of Specification 3.0.3 are not applicable. ' SURVEILLANCE RE0VIREMENTS 4.7.6.3 Each of the above required High Pressure C0 Systems shall be 2 - demonstrated OPERABLE: At least once per 6 months by verifying CO, dstorage bottle weight a. to be~at least 90% of full charge weight, an b. At'least once per 18 months by: 1) Verifying that all system components operate properly when manual devices are manually operated and automatic devices are actuated by a simulated signal, and L-2) Performance of an air flow test through system headers and H nozzles to assure no blockage. HADDAM NECK 3/4 7-21 ~

F . 7;, j , y'O t ,,y,,, JUL2S1989 MALON SYSTEMS l([

- j I

LINITING CONDITIONS FOR OPERATION l i 3.7.6.4. The following Halon 1301-Systems shat 1 be OPERABLE with the t weight of not sinimum number of storage containers shown each having a no k less than 951r, and:a pressure of not less than 324 psig: e) p - a. Switchgeat Room'- 8 containers, 4 -.b. Control Room 3 containers. Whenever systems, structures, components, or equipment APPLICABILITY: = protected by the Halon System are required'to be OPERABLE. R EllQN: With-one or more of the abeve required Halon 1301 Systems. inoserable, within 1 hour establish a continuous fire watch with la. bac tup fire suppression equipment for the unprotected equipment-1 and/or area, The provisions of Specification 3.0.3 are not applicable, b. L SURVEILLANCE REQUIREMENTS-p 4.7.6.4' Each of the-above required Halon 1301 Systems shall be ' demonstrated OPERABLE: l At least once per 6' months by verifying each Halon storage 1 container net' weight'to be at least 95% of full charge weight and - a '. -with a pressure of not less than 324 psig (adjusted for temperature).and . 'At least once per 18 months by: b. 1)' Verifying that all system components operate properly when manual devises are manually operated and autotratic devices are actuated by a simulated signal, and ' 2) Performance of an air flow test through headers and nozzles to assure no blockage. HADDAM NECK 3/4 7-22 '1 -t = + w me w -,%.s ..w. --w-

7,; } = w PLANT SYSTEMS 3/4.7.7 FIRE RATED ASSEMBLIES . LIMITING CONDITION FOR OPERATION et 3 '. 7 e 7 All; fire rated assemblies (walls, floor / ceiling, cable tray H enclosures, and other fire barriers) separating safety-related fire areas or n parating tertions of redundant systems required OPERABLE for safe. shutdown within a tn area and all sealing device.c in fire rated assembly penetrations (fire doors, fire dampers, cable, piping, and ventilation duct penetration seals) shall be OPERABLE. APPLICABILITY:' At all times unless otherwise determined that the separation of safety-related fire areas or separating portions of redundant systems important to safe shutdown within a fire area is not required based on the MODE of operation. ACTION:- I a._ With one or more of the above required fire rated assemblies and/or penetration sealing devices inoperable, within 1 hour: 1. Determine that the fire areas / zones on both sides of the affected. fire rated assembly and/or penetration sealing device are monitored by either an OPERABLE fire detection or automatic suppression system at the fire barrier and establish a fire watch patrol that inspects both areas at least once per hour,.qt 2. Establish a continuous fire watch on at least one side of the affected fire rated assembly and/or penetration seal. b. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE RE0VIREMENTS 4.7.7.1_ At least once per 18 months, the above required fire rated assemblies and penetration sealing devices shall be verified OPERABLE by performing a visual-inspection of: a. The exposed surfaces of each fire rated assembly, HADDAM NECK 3/4 7-28

o-es t PLANT SYSTEMS s BASES L In the event that portions of the Fire Suppression Systems are inoperable, alternate backup fire-fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service. When the inoperable fire-fighting equipment _ is intended for use as a backup means of fire suppression, a longer period of time is allowed to provide an 0 alternate means of fire fighting than if the inoperable equipment is the primary means of fire suppression. The Surveillance Requirements provide assurance that the minimum OPERABILITY-requirements of the Fire Suppression Systems are met. In the event-the Fire Suppression Water System becomes inoperable, immediate corrective measures must be taken since this system provides the major fire ' suppression capability of'the plant. 3/4.7.7 FIRE RATED ASSEMBLIES The functional integrity.of the fire rated assemblies and barrier penetrations ensures that fires will be confined (for a time of at least L

equal to the minimum time rating of the associated fire barrier) from spreading to adjacent portions of the facility.

These design features 1 minimize the possibility of-a single fire rapidly involving several areas of the facility prior to detection and extinguishing of the fire. The fire l barrier penetrations are a passive element in the facility Fire Protection L . Program and are subject to periodic inspections. i Fire barrier penetrations, including cable penetration barriers, fire doors and dampers are considered functional when the visually observed condition is the same as the as-designed condition. l During periods of time when a barrier is not functional, alternate measures are taken to prevent the possible spread of fire. These measures include verifying-the operability of fire. detection or suppression systems on both sides of the affected barrier and establishing a fire watch patrol, or . posting a continuous fire watch in the vicinity of the affected barrier. 3/4.7.8 FLAMMABLE LIOUIDS CONTROL The control of flammable liquids in the control room substantially reduces any fire loadings in the control room. Specification 3/4.7.8 also satisfies a NRC condition of an Appendix R control room exemption issued November 14, 1984. HADDAM NECK B3/4 7-5

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ELAlff XYSTElis

2/4.L11 PRIMARY AUXILIARY BUILDING AIR CLEANUP SYSTEM

~I-1 h LIMITING' o-rITION FOR OPERATION' 1 [ 43.7.111 The-Primary Auxiliary Building Air Cleanup' System shall be OPERABLE y. andzin operation. APPLICABIL T": MODES 1, 2, 3,.and.4. Also, during perations involving .) k ~ sovement e M uel assemblies or control rods within the containment. i u. e EllM: With-the Primary Auxiliary Building Air Cleanu System inoperable l during MODES 1, 2, 3 and 4 restore the inoperagle system to ('~ a.- LOPERABLE status within 7 days or be in at least NOT STANDBY o within the next'6 hours and in COLD: SHUTDOWN within the ) ,:following 30 hout:. With the: Primary Auxiliary Building ' Air Cleanup System inoperable .b.- or not'in operation'during operations involving movement of fuel y assemblies or control rods within the containment, suspend all . operations involving movement of fuel assembliss or control rods within the containment. Theprovi5ionsofSpecifications3.0.3and3.0.4arenot .c. applicable. S a.cu

SURVEILL ANCE REQUIREMENTS l

The' Primary Auxiliary Building Air Cleanup System shall be 4.7.11 demonstrated OPERABLE and in operation: At least once per.18 months or (1) after any structural a. maintenance on the HEPA filter or charcoal adsorber housings, or (2) following major painting, fire, or chemical release in any ventilation zone communicating with the system by: Verifying that"the cleanup system satisfies the in-place .1) penetration and bypass leakage testing acceptance criteria of less-than 15 and uses the test procedure guidance in Regulatory Positions C.5.a. C.S.c, and C.5.d of Regulatory Gu< de 1.52 Revision 2, March 1g78, and the system flow rate is' greater than 20,000 cfe; Verifying, within 31 days after removal, that a laboratory 2) analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2,-March Ig78,-meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1g78, for a methyl iodide penetration 7 of'less than 10% at test conditions of 86'F, g5% relative humidity, atmospheric pressure, and 40 feet / min. face velocity.in accordance with ASTM D3803; and W HADOAM NECK 3/47-34

f g; ~'-

PLANT SYSTEMS

. SURVEILLANCE REQUIREMENTS (Continued) 3)- Verifying a system flow rate of greater than.20,000 cfm 'during system operation when-tested in accordance with ANSI 6 N510-1980. 4)' Verifying that the pressure drop across the combined HEPA ^

filters and charcoal adsorber banks is less than 1.6 inches Water Gauge while operating the system at a flow rate of-20,000 cfm i10%.

7 b. After each com)1ete or partial replacement of a HEPA filter bank, g ' by verifying t1at the' cleanup system satisfies the in-place J penetration and bypass leakage testing acceptance criteria of less than 1% in accordance with ANSI N510-1980 for a D0P test aerosol while operating the system at.a flow-rate of greater than 20,000 cfm; and c. After each complete or partial replacement of a charcoal adsorber w . bank, by verifying that the cleanup system satisfies the in-place penetration.and bypass leakage testing acceptance criteria of less-than 1% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of greater than 20,000 cfm. d. JAfter every 720 hours of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a< representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision . 2, March 1978, meets the laboratory testing criteria of Regulatory V Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 10% at test conditions of 860F, 95% relative humidity,. atmospheric pressure l and 40 feet / min. face. velocity in accordance with ASTM D3803. e.- At.least once per 18 months, verify a system flowrate of greater than 20,000 cfm during system operation aligned for Containment Purge. l l HADDAM NECK 3/4 7-35

g p w ] PLANT' SYSTEMS BASES r 3/4.7.9 FEEDWATER ISOLATION VAtVES J The accident analysis for a main steam line break assumes that the main feedwater isolation valves will close on a containment isolation actuation- . signal- (CIAS). Also, the closure of these valves based on a CIAS is j credited in determining the Pressure / Temperature limits for the purpose of environmental qualification. The feedwater isolation valves act as a backup-l to-the feedwater regulation valves in.the event a feedwater regulation valve fails open during a Main Steam Line Break. 3/4.7.10 EXTERNAL FLOOD PROTECTION I The thresholds regarding flood protection ensure that facility protective actions will be taken-(and the orderly shutdown of the plant to MODE 3 will=

be made) in the event of flood conditions. The estimated Connecticut River probable maximum flood (PMF) level, including wave effects (i.e., still water level),-is 39.5 feet mean sea level. Normal flood control measures provide protection to safety-related equipment to El. 30 feet mean sea level. Normal flood protection to this elevation is based on a low probability of exceedance and structural capacity limitations.

Based on the one to two day rise period of-the PMF, alternative means of providing decay L. heat removal-for flooding events up to the PMF is provided in A0P 3.2-24. l 3/4.7.11 PRIMARY AUXILIARY BUILDING (PAB) AIR CLEANUP SYSTEM / PAB Air Cle'anup System consists of two exhaust fans, two prefilters, a 'HEPA-HECA filter assembly, and interconnecting ductwork. Air cleanup is accomplished using one exhaust fan, one prefilter, the HEPA-HECA filter, and interconnected ductwork. l L The radiological consequences analyses for loss-of-coolant accidents assume Primary Auxiliary Building efficiencies which are ensured by this Technical Specification. Also, in consideration of a fuel handling accident inside containment, (i.e., when the. containment is being purged) the purge discharge would be directed through the Primary Auxiliary Building charcoal filters. Credit-is again taken for these filters in reducing the radiological consequences. L 1 (3 HADDAM NECK B3/4 7-6

p~

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F, - h. [ REFUELING OPERATIONS

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS

~ LIMITING CONDITION FOR OPERATION r -3.9.4 The containment-building penetrations shall be in the following "p status:- [ a. -The equipment hatch installed and secured in place, k b.- A minimum of ~one door in the airlock.is closed, and ..c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be, either: 1) Closed by an. isolation valve, blind flange, manual valve, or special device, or-2) Be capable of being closed'by an OPERABLE containment purge supply, purge exhaust, or purge exhaust bypass isolation valve. APPLICABILITY: During CORE ALTERATIONS or movement of. irradiated fuel within the containment.. m-ACTION: With the requirements of the above specification not satisfied, immediately ~ . suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment building, other than placing an irradiated fuel assembly into a safe storage location. g SVRVEILLANCE RE0VIREMENTS 4.9.4 Each of the above-required containment building penetrations shall be determined to be either in its closed / isolated condition or' capable of 'being closed by an OPERABLE containment purge supply, purge exhaust, or . purge exhaust bypass isolation valve or capable of being. closed by an

0PERABLE. airlock door within 100 hours prior to the start of and at least

> once per 7 days.during' CORE ALTERATIONS or movement of irradiated fuel in 1: the' containment building by: a. Verifying the penetrations referred to in Specification 3.9.4c are in-their closed / isolated condition, b. Testing the containment purge supply, purge exhaust, and purge ) cL exhaust bypass isolation valves in accordance with Specification 4.9.9, and c. Varifying that at least one of the doors in the airlock is capable of being closed. L L HADDAM NECK 3/4 9-4

v6 ll h? REFUELING OPERATIONS " '1 3/4.9.7 CRANE' TRAVEL'- SPENT FUEL STORAGE POOL BUILDING ' LIMITING CONDITION FOR OPERATION' / 3.9.7 . Loads'in excess' of _1650 pounds shall be prohibited from travel over l fuel. assemblies in the storage _ pool. -r ! APPLICABILITY: With: fuel assemblies in the storage pool.. ACTION: ia; With the requirements-of the above specification not satisfied, ' place theLcrane load in a safe condition, b.. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE RE0UIREMENTS <4.9.7- . Administrative controls that 3revent the travel of loads ~in excess - of 1650 pounds over fuel assemblies 'slall be_ in place prior to lifting a load in excess of 1650_ pounds, a I L 4 s. k h . i I L .-'t (.' r i I- -HADDAM NECK 3/4 9-7

a '3 3 J. 4 REFUELING OPERATIONS 3/4.9.9 CONTAINMENT PURGE SUPPLY. PURGE EXHAUST. AND PURGE EXHAUST BYPASS. q-ISOLATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9.9~ The containment purge supply, purge exhaust, and purge exhaust bypass isolation valves shall be OPERABLE. APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment. ACTION:, a. With the requirements of the above specification not satisfied, k 'immediately. suspend all operations involving CORE ALTERATIONS or movement of' irradiated fuel in the containment building, other e~' ' than placing an irradiated fuel assembly into a safe storage location. b. The provisions of Soecification 3.0.3 are not applicable. SU'RVEILLANCE REQUIREMENTS f i

4. 9. 9 '

The-containment purge supply, purge exhaust, and purge exhaust bypass: isolation valves shall be verified OPERABLE within 100 hours prior to the start of.and at least once per 7 days during CORE ALTERATIONS and prior to movement of irradiated fuel within containment by verifying that i containment purge, bypass or exhaust isolation valves are accessible for l manual operation or verifying that the associated penetrations are blind j flanged. j i HADDAM NECK 3/4 9-10

l i 1 3/4.5' EMERGENCY CORE COOLING SYSTEMS ] BASES 1 3/4.5.1 and 3/4.5.2 ECCS SUBSYSTEMS The'0PERABILITY of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration. 'Either subsystem is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break . sizes. ranging from the' double ended break of the largest RCS cold leg pipe downward. In addition, each ECCS subsystem provides long-term core cooling capability in the recirculation mode during the accident recovery period. With the RCS temperature below 350 F, one OPERABLE ECCS subsystem is accept-able witnout single failure consideration on the basis of the stable reac-tivity condition of the reactor and the limited core cooling requirements. In Mode 4, the RHR heat exchanger (s) may be considered OPERABLE while aligned to i the Component Cooling System. .The limitation for a maximum of one centrifugal charging pump and one metering pump to be OPERABLE and the Surveillance Requirement to verify the remaining centrifugal charging pump and high pressure safety injection pumps to be g inoperable below 315 F provides assurance that a mass addition pressure

transient can be relieved by the operation of a single low temperature overpressurization relief. valve. -

In order to use the HPSI pumps to provide high pressure recirculation follow-ing a small break loss of coolant accident (LOCA) coincident with a single. active failure, the following' modifications to the emergency core cooling system have been made. A piping crosstie between each HPSI pump suction and the RHR sump discharge has been installed. Two valves, SI-M0V-901 and SI-MOV-902 1 ave been installed in this crosstie. The two manual HPSI. pump suction valves have been replaced with motor-operated valves, SI-M0V-854A and B, to prevent contaminated water from entering the RWST when using the HPSI pumps to provide flow to the core during recirculation. Two MOVs (SI-MOV-903 and 904)- in series have been added to the HPSI miniflow line to provide for redundant isolation of the RWST during sump recirculation. The manual core deluge-isolation valve has been replaced with a de-energized motor-operated valve, SI-M0V-873. This valve is locked open with the breaker ' locked open to ensure that adequate flow is available to the core deluge system. -The valve may be energized and closed if a core deluge valve failed to close during transfer to sump recirculation. The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that as a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance requirements for throttle valve position and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA. HADDAM NECK B3/4 5-1

N pp

a..

m-PLANT SYSTEMS" BASES-AUXILIARY FEEDWATER SUPPLY (Continued)

.In addition, the auxiliary feedwater system can be initiated manually. In this-case,- feedwater is available from the DWST by gravity feed to the auxiliary feedwater pump. The-specified 50,000 gallons of water in-the DWST I is adequate' for decay heat removal for a period of at least 2 hours. Within this period, decay heat removal demands are reduced to approximately 150 gpm. Makeup water is available during this period from the PWST which contains a minimum volume of 80,000 gallons. The PWST transfer pumps can transfer 200 gpm from the PWST to the DWST. An alternate supply can be i provided from the 100,000 gallons Recycled Primary Water Storage Tank. _j '3/4.7.1.4 SPECIFIC ACTIVITY I The limitations on! Secondary Coolant System specific activity ensure that ~ the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line rupture. This dose also includes the effects of a coincident 0.4 gpm reactor-to secondary tube leak in the steam generator of the affected steam i line. These values are consistent with the assumptions used in the safety

analyses, 3/4.7.1.5 MAIN STEAM LINE TRIP VALVES The OPERABILITY of the main. steam line trip valves ensures that no more than one steam generator will' blowdown in the event of a steam line rupture.

This: restriction is required to: (1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, andL(2) limit the pressure rise within containment in the event the steam-line rupture occurs within containment. The OPERABILITY of the main steam s line trip valves within the closure' times of the Surveillance Requirements 1 are consistent with the assumptions used in the safety analyses. i 3/4~7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The' limitation on steam generator pressure and temperature ensures that the pressure-inducedstressesinthesteamgeneratorsdonotexceedthgmaximum allowable fracture toughness stress limits. Thglimitationsof70Fand200 psig are based on a steam generator RTNDT of 10 F and are sufffcient to prevent brittle fracture. The heatup and cooldown rate of 100 F/hr for the steam generators are specified to ensure that stresses in these vessels are maintained within acceptable design limits. 3/4.7.3 SERVICE WATER SYSTEM The OPERABILITY of the Service Water System ensures that sufficient cooling ce.pacity is available for continued operation of safety-related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analysis. A service water header is comprised of the two service water pumps associated with each diesel generator and the HADDAM NECK B3/4 7-2

w s' h j PLANT SYSTEMS BASES

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h . safety-related piping and components. The. service water headers may be tied together by an open service water header cross-connect in the intake structure.. 3/4.7.4 SNUBBERS ' All snubbers are required to be OPERABLE to ensure that the structural integrity of the reactor coolant system and all other-safety related systems { is maintained during and following a seismic or other event initiating dynamic' loads. Snubbers excluded from this inspection program are those i installed on nonsafety-related systems and then only if their failure, or failure of.the system on which they are installed, would have no adverse effect on any safety-related system. The visual inspection frequency is based upon maintaining a constant. level of snubber protecti.on to systems. Therefore,~the required inspection interval varies inversely with the observed snubber failures and is deter-4 mined by the-number of inoperable snubbers found during an inspection. ~I Inspections performed before that interval has elapsed may be used as a new -reference point to determine the next inspection. However, the results of l such early-inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspec-l y tion _ interval will override the previous schedule. When the cause of the rejection of.a snubber is clearly established and i remedied for that snubber and for any other snubbers that may be generically I susceptible-and verified by inservice functional testing, that snubber may be exempted from being counted as inoperable. Generically susceptible snubbers are those which are of a specific make or model and have the same . design features directly related to rejection of the snubber by visual t ' inspection or are similarly located or exposed to the same environmental -l conditions, such as temperature,' radiation, and vibration, j i - When a snubber is found inoperable, an engineering evaluation is performed, i in addition to the determination of the snubber mode of failure, in order to 4 ' determine if any safety-related component or system has been adversely

affected by the inoperability of the snubber.

The engineering evaluation shall determine whether or not the snubber :aode of failure has imparted a o significant effect or degradation on the-supported component or system. .To provide-assurance of snubber functional reliability, a representative . sample of the installed snubbers will be functionally tested during plant i shutdowns at 18-month intervals. These tests will include. stroking of i snubbers to verify freedom of movement over the full stroke, restraining characteristics, and drag force (if applicable). Ten percent (10%) of the total of each type of snubber represents an adequate sampling for these tests. Observed failures on these samples require testing of additional units. Hydraulic snubbers and mechanical snubbers may each be treated as a different entity for the above surveillance program. i .HADDAM NECK. B3/4 7-3 c

~ s 9 5.0 DESIGN FEATbRES 5.1 SITE I E LUSION AREA e 5;1.1 The Exclusion Area shall' be as shown in F19are 5.1-1. The minimum distance-to the boundary of the exclusion ares at defined in 10CFR100.3 'shall be 1740 feet. 10W POPULATION ZONE 5.1.2 The Low Population Zone shall be as shown in Figure 5 l-2. L' .HADDAM NECK 51

q m i i i O DESIGN FEATURES L2 CONTAINMENT l CONFJGURATION

5. 2. l ' - The containment building is a steel-lined, reinforced concrete i-building of cylindrical shape, with a hemispherical dome roof and a flat base, and having the following design features:

f a. Nominal inside diameter of cylinder = 134 feet - 11.25 inches. I b. Nominal inside height of cylinder = 119.5 feet (Not including dome). L c. Nominal inside height of containment - 188 feet - 11.5 inches (From containment floor to inside top of dome). d. Minimum thickness of concrete walls = 4.5 feet, e. Minimum thickness of concrete dome = 2.5 feet. f. Minimum thickness of concrete bottom mat = 9 feet. g. Nominal thicus of steel liner = 1/4 to 1/2 inch (bottom = 1/4 inch, spheric 6 dome = 1/2 inch and side wall = 3/8 inch). 6 h. Net free volume = 2.232 x 10 cubic feet (nominal). 1 DESIGN PRESSURE AND TEMPERATURE 5.2.2 - Thecontainmentbuildingisdesignedandshallbemaigtainedfora maximum internal pressure of 00 psig and a temperature of 260 F. 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The core shall contain 157 fuel assemblies with each fuel assembly l containing 204 fuel rod locations'. Fuel rod locations may at any time l during plant life consist of one of the following: (1) fuel rods clad with l Type 304 stainless steel, (2) filler rods fabricated from Type 304 stainless steel, or (3) vacancies as evaluated by the cycle-specific reload analysis. Each fuel assembly shall have a nominal active fuel length of 120.5 inches 1. and contain a typical weight of 411.5 kilograms uranium. The core loading . shall have a maximum enrichment of 4.00 weight percent U-235. L l L HADDAM NECK 5-4 A. m 9'% I 7 v' -m-

r p. a. ~>., e. . g ww-- r ~~ w w1 6 3 gg... 7. o. m e,em . n. ,^ Rt ~ + >g 4% .] c N 3: ~ n l 'J; '; a Docket No.'$0 213 r m, B13402 e P' i j ' 14 ', .y. a. Mi y s t if L 't, -t ) 'i.. e r ( .,, y. p- ,/i .-( 1 o p. -4 ., p - Qy,:p,l a T I + :; 4 o ~ M" . Haddam Neck Plant e i,; Description;of the Proposed Technical Specification Changes MW ' 4

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i { w i L U.S. Nuclear Regulatory Commission B13402/ Attachment 2/Page 1 November 22, 1989 l L Haddam Neck Plant Description of The Proposed Technical Specification Changes i and Sionificant Hazards Consideration Discussion j i-Descriotion of the Prooosed Chances f References (1) through (11) provided a complete set of the proposed revised Technical Specifications (RTS) for the Haddam Neck Plant. All sections of the existing Technical Specifications for the Haddam Neck-Plant have been refor-matted and upgraded except for Sections 3.6, 3.12, and 4.5 [ References (1) through (6)]. These sections, 3.6, 3.12, and 4.5, have been reformatted and upgraded and are included in References (8) and (11). In addition, the changes included in References (8) and (11) reflect the modification completed in the 1989 refueling outage. The Technical Specification changes included in Reference'(7) reflect the upgrade of the reactor protection system and the nuclear instrumentation. The Technical Specification changes included in Reference (9) reflect Ctcle 16 reload changes and, lastly, the Technical Specification changes included in Reference (10) reflect the changes associ-L ated with the fire detection / suppression system upgrades in support of the new l L switchgear building. The following changes are proposed to the RTS primarily to incorporate the NRC's comments to the proposec

RTS, l.-

lables 2.2-1. 3.3-1. and 4.3-1. Item 17. Source Ranoe Start-Up Rate Rod E9.2 The existing ' Technical Specifications (TS) Table 3.9-1 requirement for the source range start-up rate rod stop is reinstated in the proposed

RTS, Corresponding changes are incorporated in Table 3.3-1, Reactor Trip

. System Instrumentation and Table 4.31, Reactor Trip System Instrumenta-tion Surveillance Requirements. The changes.will make the proposed RTS consistent with the existing TS. However, the changes proposed in Reference (7) will supersede this change. The proposed changes provided herein are only for record purposes. No design basis accident analysis takes credit for the source range start-up rate rod stop. As such, no I design basis accidents are adversely affected due to the changes. 2. Section 4.0.2 and Correspondina Bases Section The proposed change to the requirements of Specifications 4.0.2 will remove any unnecessary restriction on extending the surveillance require-ment and will result in a benefit to safety when plant conditions are not conducive to the safe conduct of the surveillance requirement. The removal of the 3.25 limit will provide greater flexibility in the use of the provision for extending surveillance intervals, reduce the adminis-trative burden associated with its use, and have a positive effect on i L safety. This change is consistent with Generic Letter 89-14. No design basis accidents are adversely impacted due to the change. I 1

p 1 k e U.S. Nuclear Regulatory Commission B13402/ Attachment 2/Page 2 j November 22, 1989 j i 3. Section 4.1.2.2. Paae 3/4 1-12 L The change corrects a typographical error and requires verification of L flow paths from the boric acid tank via the boric acid pumps and centrif-ugal charging pumps to the reactor coolant system by referencing Specifi- [ cation 3.1.2.2.b. Thus, the surveillance requirement of Section 4.1.2.2 l L will become restrictive by the change. No design basis accidents are adversely affected due to the change. 4. Section 3.1.2.3. Daae 3/4 1-11 [ The change corrects a typographical error. No design basis accidents are l affected due to the change. 5. Tables 3.3 1 and 4.3-1. Item 3. Intermediate Ranoe. Neutron. Flux. Hiah Start-Vo Rate The proposed change will also require the trip to be operable in Mode 1 I below 10 percent power and increase the number of minimum channels e L operable from 1 to 2 in Table 3.3-1, Reactor Trip System Instrumentation. Corresponding changes to Table 4.3-1, Reactor Trip System Instrumentation Surveillance Requirement, are also proposed. As such, the proposed i ~ hange provides more stringent requirements compared with the proposed c RTS and is equivalent to the existing TS. However, the changes proposed in Reference (7) will supersede this change. The proposed change pro-vided herein is only for record purposes.- No design basis accident is adversely affected due to the change. 6. Tables 3.3-1 and 4.3-1. Item 6. Pressurizer Water level-Hiah l l The change will require the trip to be operable in Modes 1, 2, and 3 instead of Mode 1 only. In addition, the change will allow the high pressurizer event trip to be bypassed when the reactor is at least l.5 percent AK suberitical, it is noted that the original safety evalua-tions performed are still applicable since the change will be consistent-with the existing TS. No design basis accident is adversely affected due to the change. 7. Table 3.3-1. Item 15a. Low Power B.ock. P-7 1 The proposed change will require the trip to be operable at any power level above 10 percent of rated thermal power. The proposed RTS in the previous submittal [ Reference (4)) require the trip to be operable for power above 10 percent but below 74 percent of rated thermal power. The change is consistent with the existing TS. No design basis accident is adversely affected due to the change.

U.S. Nuclear Regulatory Commission B13402/ Attachment 2/Page 3 November 22, 1989 8. Table 3.3-1. Item 11. Safety Iniection The proposed change adds a new action statement to be required for the safety injection system. The new action statement (Action 11A) reads as follows: "with the number of OPERABLE channels one less than the minimum channels Operable requirement, be in at least HOT STANDBY within 6 hours. However, one channel may be bypassed for u) to 2 hours for surveillance testing per Specification 4.3.1.1, providoc the other channel is OPERA-BLE." This added restriction eliminates the 48 hour period for restoring the channel prior to applying the 6 hour hot standby requirement. The change will make the proposed R1S consistent with the existing TS and with the Westinghouse STS. The change will not impact the consequences -of any design basis accidents. 9. Table 3.3-1. Action 11 The change corrects a typographical error. No design basis accident is affected by the change.

10. Table 3.3-2.

Items 1.b and 5.a. Containment Pressure - Hiah: Item 3.a. Wide Ranae SG Water Level - Low: Item 3.b. Trio of All Main Feedwater Pumps The proposed change will increase the number of minimum channels operable from 4 to-6 for Containment pressure-high in Table 3.3-2, Engineered Safety Features Actuation System Instrumentation. The change will also change the Action statement for item 5.a to be consistent with the existing TS. For items 3.a and 3.b, the change requires the allowable restoration time for an inoperable channel to be reduced from 48 hours to 24 hours. The 24 hour time is based on engineering judgment to take actions to repair the inoperable channel. The change provides more stringent requirements relative to system's surveillance and operation. No design basis accidents are adversely impacted due to the change. It is noted that the proposed change to Table 3.3-2, Items 1.b and 5.a, will 4 be superseded by the changes proposed in Reference (7). 11. Table 3.3-3. Footnote to Item 4.a. 4.16 kV Bus Undervoltaae - Level 1: Table 4.3-2. Footnote to items 4 b and 4.c. 4.16 kV Bus Undervoltaae - Level ? and Level 3 Table 3.3-3, Engineered Safety Features Actuation System Instrumentation Trip Setpoints of the proposed RTS in the previous submittal (Refer-ence 8)] contains the description of test parameters for the 4.16-kV Bus Level 1 undervoltage setpoint and minimum allowable value. This descrip-tion states that the device must change state within.95 - 1.05 seconds when the input voltage to the device goes from normal to zero volts instantaneously. l

s a p o } 'U.S. Nuclear Regulatory Commission B13402/ Attachment 2/Page 4 November 22, 1989 The proposed change requires that the relays actuate when the input voltage decreases instantaneously from normal to 50 percent of the tap setting voltage. The 4.16-kV Bus Leve'. I undervoltage allowable value is the minimum voltage.to which the safety bus - may degrade before all required loads must be transferred from offsite to onsite power sources. .The table notation for the allowable value further defines the operation t of the undervoltage relays and its actuation parameters. By requiring the device to change state within one second i 5 percent when the input voltage to the device reduces from normal to 50 percent of tap setting voltage instantaneously, the relay is being challenged to operate in a o real degraded voltage situation. Previously, allowing the input voltage to the device to drop to zero did not fully test the time-voltage charac-teristics of the induction coil in the degraded voltage range because loss of all voltage just causes the relay to return to its de energized state. The change represents a more conservative test and is consistent with the plant's standard method of testing undervoltage relays of this type. The proposed change also adds the footnote (*) to the frequency at which i the 4.16 kV bus undervoltage Level 2 and Level 3 trip actuating device operational test is specified. Adding the footnote defines the monthly surveillance as a test of each individual channel rather than a test of the logic, the Level 1 trip actuating device operational test currently 4 has this footnote, l In order to test the complete logic associated with these undervoltage trip relays,. two channels would need to be placed out of service. This would be in direct conflict with Action Statement 24 of Table 3.3-2, which allows only one channel to be bypcssed for up to 2 hours for surveillance testing with one channel inoperable, it is noted that a complete undervoltage relay calibration and logic check with still be performed during refueling and is not affected due to the change. -Based on the above, the changes do not impact the consequences of any design basis accidents and the probability of failure of a safety system associated with these changes is not increased. L 12. Sectico 4.4.5.5. Reports, and Table 4.4.2. Steam Generator Tube insoec- .11911 The proposed change adds words "or sleeved" to Section 4.4.5.5.a and deletas a note to Table 4.4-2 to make the proposed RTS consistent with the existing TS. This eliminates the ability to take exception to a requirement and is thus more restrictive. The change does not impact the consequence of any design basis accidents, l-l l

~ .O 'l [ U.S. Nuclear > Regulatory Commission l B13402/ Attachment 2/Page 5 November 22, 1989

13. Section 3.4.6.1.b and Action a The change corrects a typographical error.

No design basis accident is affected due.to the change. l ~14. Section 3.4.6.2. Operational Leakaae L The proposed ' change revised the Surveillance 4.4.6.2.1.c (VCT level monitoring) frequency from "at least once per 12 hours" to "at least once L per 4 hours." In addition, the bases section is changed to clarify L " monitoring" term. As such, the proposed change provides a more restric-tive surveillance requirement. No design basis accidents are adversely affected due to the change.

15. Jiection 3.4' 9.2. Pressurizer The proposed change deletes the surveillance requirements for the auxil-iary spray (Section4.4.9.2.2) as there is no limiting condition for operation for this surveillance.

This was inadvertently included in the previous submittal [ Reference (3)]. 16. Sections 4.5.1 and 4.5.2. ECCS. Table 4.5-1 and Table 4.5-2 i F The change corrects typographical errors. No design basis accidents are affected due to the change. 17. Section 3.6.1.2. Footnote on Pace 3/4 6-2, and Section 4.6.1.2.a The proposed change deletes the footnote regarding the exemption for certain containment penetrations since CYAPC0 has not received a perma-nent exemption to Appendix J, paragraph III.C.2.(a) for these penetra-tions. The change provides a more restrictive requirement. No design basis accidents are adversely affected due to the change. Sec-1 ' tion 4.6.1.2.g is revised to clarify that the exception to Specifica-tion 4.0.2 is only applicable to Sections 4.6.1.2.a through 4.6.1.2.d. l No design basis accidents are affected due to this change. 18. Section 4.7.1.1. Turbine Cycle Safety Valves Surveillance The proposed change will allow the plant to enter in Mode 3 to perform SG + code safety testing following an outage. The existing TS requires the safety valves to be operable when critical. This is consistent with the plant's current practices and the existing TS. No design basis accidents are adversely-affected due to the change. L l

j t 'O i L I U.S. Nuclear Regulatory Commission j p .B13402/ Attachment 2/Page 6 November 22, 1989 I-19. Section 4.7.6.1.1. Item e. Fire Water Supply / Distributions Surveillance 'Reauirement- .The proposed change specifies the reference. document that provides a criterion to perform ' a system flow test. The reference document is Chapter 5, Section 11, of the Fire Protection Handbook, 14th Edition, i published by National Fire Protection Association. No design basis -accidents are affected due to the change. -20. Section 4.7.6.2. Item d. Sorav and/or Sorinkler Systems The proposed change removes the option to use water to perform a flow ' test through each open head system. The change is consistent with the current plant practices. No design basis accidents are adversely affect-ed.due to the change. 21. Section 3.7.6.3, Action item (a), C0 System 2 The proposed change clarifies the location and the equipment the ACTION affects. No design basis accidents are affected due to the change. 22. Section 4.7.6.3,-Item b.1, CO System Surveillance Requirements 2 -The proposed change revises the wording of the surveillance to clarify that all system components will be verified. The change is' editorial in nature. No design basis accidents are affected due to the change. t 23. Section 3.7.6.4. Action item (a). Halon System The proposed change clarifies the location and the equipment the ACTION affects. No design basis accidents are affected due to the change. 24. Section 4.7.6.4. Item b.1. Halon System Surveillance Reauirements The proposed change revises the wording of the surveillance to clarify i that all syster components will be verified. The change is editorial in i nature. No design basis accidents are affected due to the change. 25. Section 3.7.7. Action 3. Fire Rated Assemblies, and Bases Section 3/4.7.7 l .The proposed change deletes the option to temporarily repair inoperable ' fire-rated assemblies. The change is consistent with the existing TS. .This change will also revise the bases section by deleting the reference to the installation of a temporary fire stop. No design basis accidents j are adversely affected due to the change.

p, 2 g 4 ? i U.S. Nuclear Regulatory Commission B13402/ Attachment 2/Page 7 November 22, 1989 26. Section 3/4.7.11. Primary Auxiliary Buildina Air Cleanuo System Surveil-lance and Bases lhe change will achieve consistency among the proposed RTS by changing t.ae documentation referenced from ANSI N510-1975 to ANSI N510-1980. The change will also revise the bases section to indicate number of fans and/or filtration system required to be operable to meet LCO of Section 3.7.11. The change is administrative in nature. No design basis . accidents are affected by the change. 27. Section 3.9.4. Containment Buildina Penetration. LCO (b) The proposed change will require one door in the airlock CLOSED rather than OPERABLE. This change is consistent with the current plant prac-tices and is consistent with the Westinghouse STS. The other change proposed herein recognizes the current name of the system (i.e., Contain-ment Purge Supply, Purge Exhaust, and Purge Exhaust Bypass System). No design-basis accidents are adversely affected due to the change. 28. Section 3/4.9.7. Crane Travel-Soent Fuel Storace Pool Buildina The proposed change will clarify the movement of load in excess of 1650 pounds over spent fuel assemblies in the storage pool. The change is an administrative change. No design basis accidents are affected due to the change. 29. Section 3/4.9.9. Containment Purae Suppiv. Purae Exhaust, and Purag Exhaust Bypass System .The proposed change recognizes the current name of the system.- No design basis accidents are affected due to the change. 30. Bases Sections 3/4.5.1 and 3/4.5.2. ECCS Subsystems: 3/4.7.3. Service Water System ' The proposed change adds a statement to clarify the operability of RHR heat exchangers in Mode 4 (Section 3/4.5.1 and 3/4.5.2). In addition, the change adds a statement to Bases Section 3/4.7.3, Service Water System, to indicate that the service water headers may be tied together by an-open service water header cross-connect in the intake structure. As such,-the change in the bases section will not affect any design basis accidents. 31. Section 5.1.1. Exclusion Area .1 The change adds an explicit statement to indicate the minimum distance to the boundary of the exclusion area. The change will make the proposed TS consistent with the existing TS. No design basis accidents are affected due to-the change.

i-

. g U.S. Nuclear Regulatory Commission B13402/ Attachment 2/Page 8 November 22, 1989 32.

Section 5.3.1. Fuel-Assemblies inspection of fuel assemblies during the current refueling outage has revealed pins with the fuel cladding failures. This problem is being addressed by fuel reconstitution which in many cases results in failed fuel pins being. replaced with " filler rods" fabricated from Type 304 -stainless steel or even left as vacancies. This fuel reconstitution will be conducted as demonstrated to be acceptable by a cycle-specific reload ~ analysis. The change is proposed to reflect the options of having " filler rods" or " vacancies" as well as actual fuel rods. Since changes due to reconstitution will be fully evaluated prior to implementation, the extra flexibility allowed by the change to the design features section has no impact on any design basis accidents. In general, most of the. above changes are either administrative or will make i the proposed RTS consistent with the existing TS. Essentially, the changes provide more restrictive requirements than the proposed RTS in the previous submittals. -Sionificant Hazards Consideration it should be noted that the significant hazards consideration (SHC) provided below supplements the SHC discussion previously included in References (1) i through (11). -In accordance with 10CFR50.92, CYAPC0 has reviewed the changes to the base document and has concluded that they do not involve a significant hazards s 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 r ^ significant hazards consideration because the changes would-not: -1. Involve a significant increase in the probability of occurrence or consequences of an accident previously analyzed. The proposed changes provide some additional requirements from the' proposed RTS (previous submittals) to bring the specifications more consistent with the existing TS requirements. Also, the proposed changes correct typographical errors and clarify some ambiguous wording and reference documentation. These changes do not have any impact on the design basis accidents. Sec-tion. 5.3.1 for fuel assemblies will now allow fuel rod locations to be filled with alternate rods as used in reconstitution. Since the changes due to reconstitution will be fully evaluated prior to implementation, the additional flexibility allowed by the changes to the Design features Section of the RTS has no impact on the design basis accidents. 2. Create the possibility of a new or different kind of accident from any previously analyzed. Since there are no changes in the way the plant is operated, the potential for an unanalyzed accident is not created. No new failure modes are introduced.

R F< 3. i i s W R, j j 'U.S. Nuclear Reg 01 story Commission n: .B13402/ Attachment 2/Page.9 1 (- - November. 22, 1989 j j o i 3. Involveia _significant-reduction.in a margin: of safety.. The changes do l Since the I x not have' any; adverse _ impact on > the protective boundaries. i changes' also do not affect the consequences of any accident previously ,~ analyzed, there is no reduction in a margin of safety.- j i k.. i l s [,' 'l i-.. i pc l ?- Y r: t i g' 5 i: 1: 1 1 f i ) f 5 1 .t.j;? >}}

ML19332E438

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