ML20199A110

Proposed Tech Specs,Modifying 6.8.4.a, Primary Coolant Sources Outside Containment to Add Portions of Containment Vacuum Relief Sys & Primary Sampling Sys to Program

ML20199A110
Person / Time
Site:	Waterford
Issue date:	11/13/1997
From:	ENTERGY OPERATIONS, INC.
To:
Shared Package
ML20199A105	List: ML20199A103 ML20199A110
References
NUDOCS 9711170061
Download: ML20199A110 (33)

Text

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AttiINIITRATf VE tolffhfM e 4

l PRottount$ MD ptotRANS (Continued)

• l j. OFFSITE Dost CALCULATION MANUAL laplementation.

i

k. Quality Assurance program for effluent and environmental monitoring, .

l usine and Regulatory the guidance Guidein4.1, Regulatory Revision Guide 1. April1.21,1975. Revision 1, June j 6.8.2 Each procedure of specification 6.8'.1, and changes thereto shall be reviewedandapprovedpriortoimplementationandreviewedperiodIcallyasset forth in administrative procedures. ll

6.8.3 Temporary changes to procedures of specification 6.8.1 may be made provided

a. The intent of the original procedure is not altered;

b. The change is approved by two meers of the plant management staff,

, at least one of whom holds a senior operator license on the unit affected;

c. The change is documented, reviewed and approved as required by

)

administrative procedures within 14 days of tuplementation. l 6.8.4 The following programs shall be established, implemented, and l maintained:

a. Primarv toelant sources Dutside containment A program to reduce leakage from those portions of systems uutside

, conta< nment that could contain highly radioactive fluids during a serious transient or accident to es low as practical levels. The systems include the containment spray safety injection, hydrogen analyzer,andthepost-accidentsamp1Ingsystem. The program shall include the following: ,

l 1. Preventive maintenance and periodic visual inspection i requirements, and i

2. Integrated leak test requirements for each system at refueling cycle intervals or less,

b. Mor. ins  !

A program which will ensure the capability to accurately determine the a'rborne iodine concentration in vital areas under accident .

conditions. This program shall include the following:

liATERFORD - UNIT 3 6-15 AMEN 0 MENT N0. 53, ",109 i

[)t @k 2

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION l The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above 1.20 .

during all normal operations and anticipated transients. In H0 DES 1 and 2 with one reactor coolant loop not in operation, this specification requires that the plant be in at least HOT STANOBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

In H00E 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be OPERABLE.

In MODE 4, and in H0DE 5 with reactor coolant loops filled, a single reactor coolant loop or shutdown cooling train provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops or trains (either shutdown cooling or RCS) be OPERABLE.

In H00E 5 with reactor coolant loops not filled, a single shutdown cooling train provides sufficient heat removal capability f* removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two shutdown cooling trains be OPERABLE.

The operation of one reactor coolant pump or one shutdown cooling (low pressure safety injection) pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reductions will, therefore, be within the capability of operatot-recognition and control. *

• The restrictions on starting a reactor coolant pump in MODES 4 and 5,

with one or more RCS cold legs less than or equal to 285'F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 100'F above each of the RCS cold leg temperatures.

3/4.4.2 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia. Each safety valve is designed to relieve 4.6 x 105 lbs per hour of saturated steam at the valve setpoint. The relief capacity of a single safety valve is adequate to relieve any overpres-l sure condition which could occur during shutdown. In the event that no safety 1

l WATERFORD - UNIT 3 B ?/4 4-1

f,0NTAINMENT SYSTEMit 1

i SABES yd 81.7 CONTAINMENT VENTILATION SYSTEM (Continued) 1 Leskage integrity tests with t . wximum allowabis leakage rete for purge supply and exhaust isolation valves WWI provida My indication of resilient material seal degradation and  !

i wiu allow the opportunity for repair b6, ore gross leakage falluto develops. The 0.60 La leakage limit shsN not be exceeded when the leakage rotes determined by the leakage Integrity tests of these valves are added to the previously determined total for all valves and

!, penetrations subject to Type 8 and C tests.

W4.8.2 DEMESSURlZATION AND COOLING SYSTEMR

, 3M 8.21 and 1/4 8 2.2 CONTAINMENT SPRAY SYSTEM and CONTAINMENT COOLl!gg SYSTEM l

The OPERABILITY of the Cordainment Sprey System and the Containment Coohng System ensures that containment depressurization and cooling capability wlN be available in the event of a LOCA or MSL8 for any double ended break of the largest reactor coolant pipe i or main steam line. Under post accident conditions these systems win maintain the containment pressure below 44 poig and temperatures below 26g.3*F during LOCA 4 conditions or 413.5'F during MSL8 conditions. The systems also reduce the containment pressure by a footor of 2 from its post accident peak within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, resulting in lower

! containment leakage rates and lower offsite dose retos.

! The ContainmerW Spray System also provides a mechanism for removing i

nodine from the containment atmosphere under post-LOCA conditions to maintain doses in accordance with 10 CFR Part 100 limits as described in Section 6.5.2 of the F8AR.

$ in MODE 4 when shutdown cooling is placed in operation, the Containment Spray System is realigned in order to allow isolation of the sprey headers. This is necessary to

, avoid a single failure of the sprey header isolation valve oeusirgReactor Coolant System

depressurtration and inadvertent spraying of the containment To allow for this realiCnment, the Containment Sprey 8 may be taken out of-s8tvios when RC8 pressure is 5 400 ,

is. At this reduced I pressure and the reduced temperature associated with entry into I

4, the and consequences of a LOCA or MSL8 are greatly reduced. The i Containment Syelom is required OPERABLE in MODE 4 and is available to provide

, depressurtsstion cooling capability.

A train of Conteir.mont Cooling consists of two fans (powered from the same esfety l bus) and tho' ***ovi.ned coolers (supplied from the same cooling water loop). Orm Containment Cooling train and Containment Spray trein has sufflaient capsolty to meet post l

socident heat removal requirements.

Operating each containment cooling train fan unit br 15 minutes and vertfying a cooling water flow rete of 625 gpm ensures that all trains are OPERABLE and that all associated controls are functioning property. It also ensures that blockage, fan or motor failure, or 4

excessive vibration con be detected and corrective action taken.

WATERFORD UNIT 3 83/46-3 AMENDMENT NO. 80,131

1 l

e t

! CONTAINMENT SYSTEMS l j

RASES 1

1/4 8 21 and 1/4 8 2 2 CONTAINMENT SPRAY SYSTEM and CONTAINMENT COOLING '

SYSTEM (cont)

The 18 month SurvelNanos Requirement verifles that each containment cooling fan 1 actuates upon receipt of an actual or simulated SlAS actuation signal. The 18 month frequency is based on engineering judgment and has been shown to be soceptable through operating experience. ,

Verifying a cooling water flow rate of 1200 ppm to each cooling unit provides assurance that the design flow rate assumed in the safety analyses wlN be achieved. The safety i analysos assumed a cooung water flow rate of 1100 ppm. The 1200.gpm requirement accounts 'or measurement Instrument 9ncertainties and potential flow degradation. Also

considered in selecting the 18 month frequency were the know reliability of the Cooling Water System, the two train redundancy, and the low probability of a significant degradation of flow .

ocouning between surveillances. The flow measurement for the is month test shan be done  !

, in a configuration equivalent to the socident lineup to ensure that in en socident situation adequate flow wlN be provided to the containment fan coolers for them to perform their safety function, i

Vertfying that each valve actuates to the fun open position movides further assurance

that the valves win travel to their fun open position on a Safety in,ection Actuation Signal.

3/4.8.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment l

stmosphere wit be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurtzation of the containment and .

is consistent with the requirements of 000 54 through GDC 57 of Appendix A to 10 CFR Part 50. Containment isolation within the time limits specified for those isolation volves designed to close autorreri ensures that the release of radioactive material to the i environment wul be consistent with the assumptions used in the analyses for a LOCA.

The opening of looked or sealed closed containment isolation volves on an intermittent basis under administrative control includes the following considerations: (1) stationing an operator, who is in constant communication with control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions wB not preclude socess to close the valves and that this action win I

prevent the release of radioactMty outside the containment.

' Containment teolation Valves", previously Table 3.6 2, have been incorporated into Plard Procedure UNT406026.

3/4.6.4 COMBUSTIBLE GAS CONTROL ,

The OPERABILITY of the equipment and systems required for the detection and L control of hydrogen gas ensures that this equipment win be available to maintain the WATERFORD - UNIT 3 83/464 Amendment No. 4440,131

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+

PLANT SYSTEMS BASES 3/4.7.9 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) Ifmits for plutonium. This limitation will ensure that leakage from baroduct, source, and special nuclear material sources will not exceed allowable intake values.

Sealed sources are classified into three groups according to their use, with Surveillance Requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e. sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shield mechanism.

3/4.7.10 This section deleted.

3/4.7.11 This section deleted.

3/4.7.12 ESSENTIAL SERVICES CHILLED WATER SYSTEM The OPERA 8ILITY of the essential services chilled water system ensures that sufficient chilled water is supplied to those air handling systems which cool spaces containing equipment required for safety-related operations and, during normal plart operation, the nonessential spaces.

t WATERFORD - UNIT 3 8 3/4 7-7 Amendment No. 50

- , , _ , , . - n -, - - --r r r ---<w- - - - - - -

l M CTRICAL POWER SYSTEMS L

RA$f1 l

l A.C. SOURCES. D.C. SOURCES. AND ONSITE POWER DISTRIBUTION SYSTEMS (Continued)

Operation with a battery cell's parameter outs Me the normal limit but within ' i the allowable value specified in Table 4.8-2 is permitted for up to 7 days. During this 7-day period (1) the allowable values for electrolyte level l ensures no physical damage to the plates with an adequate electron transfer  !'

capability; (2) the allowable value for the avera e specific gravity of all the cells, not more than 0.020 below the manufacturer s recommended full charge i specific gravity, ensures-that the decrease in rating will be less than the safety margin provided in sizing; (3) the allowable value for an individual cell's specific gravity, ensures that an individual cell's specific gravity will not be '

more than 0.040 below the manufacturer's full charge specific gravity and that the

• overall capability of the battery will be maintained within an acceptable limit;  :

and (4) the allowable value for an individual cell's float voltage, greater than l 2.07 voitt., ensures the battery's capability to perform its design function.

3/4.8.4 ELECTRICAL EuulPMENT PROTECTIVE DEVICES Containment electrical penetrations and penetration corductors are protected by  ;

either deenergizing circuits not required during reactor operation or by demonstrating the OPERABILITY of primar and backup overcurrent protection circuit-breakers during periodic survei lance.

The Surveillance Requirements applicable to lower voltage circuit breakers and fuses provides assurance of breaker and fuse reliability by testing at least one representative sample of each manufacturers brand of circuit breaker and/or fuse.

Each manufacturer's molded case and metal case circuit breakers and/or fuses are grouped into representative samples which are then tested on a rotating basis to ensure that all breakers and/or fuses are tested. If a wide variety exists within '

any manufacturer's brand of circuit breakers and/or fuses it is necessary to t divide that manufacturer's breakers and/or fuses into groups and treat each group as a separate type of breaker or fuses for surveillance purposes.

The OPERABILITY of the motor-operated valves thermal overload protection and/or  !

bypass devices ensures that these devices will not prevent safety related-valves from performing their function. The Surveillance Requirements for demonstrating the OPERABILITY of these devices are in accordance with Regulatory Guide 1.106,

" Thermal Overload Protection for Electric Motors on Motor Operated Valves,"

Revision 1, March 1977.

' Containment Penetration Conductor overcurrent Protection Devices" and " Motor-Operated Valves Thermal Overload Protection and/or Bypass Devices", previously Tables 3.8-1 and 3.8-2, have been incorporated into Plant Procedure i UNT-005-026.

L WATERFORD - UNIT 3 B 3/4 8-3 - AMENDMENT NO. Mr 92

- - - - _a. = L _ -- .A_ _m_ 4 _ = 4 um _A. -.-. . . .- _ ._

NPF-38-202 ATTACHMENT B I

l

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

J. OFFSITE DOSE CALCULATION MANUAL implementation, k, Quality Assurance Program for effluent and environmental monitoring, using the guidance in

. Regulatory Guide 1.21. Revision 1 June 1974 and Regulatory Guide 4.1, Revision 1, April 1975.

6.8.2 Each procedure of Specification 6.8.1, and changes thereto, shall be reviewed and approved prior to implementsetion and reviewed periodically as set forth in administrative procedures.

6.8.3 Temporary changes to procedures of Specificat!on 6.8.1 may be made provided:

a. The intent of tht, original procedure is not aftered;

b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Operstor license on the unit effected,

c. The change is documented, reviewed and approved as required by administrative procedures within 14 days of implemantation.

6.8.4 The following programs shall be established, implemented, and mainteined:

a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the containment spray, safety injection, hydrogen analyzor, and the post accident sampling systgmr and nortions of the sy6temcontainment vacuum relief and nrimarv samnlino svatamr. The program shall include the following:

1, Preventive ma!ntenance and periode visualinspection requirernents, and

2. Integrated leak test requirements, for each system at refueling cycle intervals or less.

b ln Plant Radia_t_i_on Monitorina A program which will ensure the capability to accurately determine the airborne lodine concentration in vital areas under accident conditions. This program shallinclude the following:

WATERFORD UNIT 3- 6 15 AMENDMENT NO. 63,400, ,199 _ l

k 3/4.4 REACTOR CQ.QJ. ANT SYSTEM BASES 3/4.4.1 REACTOR CQOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above 1.20 during all normal operations and anticipated transients. In MODES 1 and 2 w'th one reactor coolant loop not Ir, operation, this specification requires that the plant be in at least HOT STANDBY withm 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations requirn that two loops be OPERABLE.

In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or shutdown cooling train provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops or trains (either shutdown cooling or RCS) be OPERABLE.

In MODE 5 with reactor coolant loops not filled, a single shutdown coolirig train provides sufficient heat removal capability for removing decay heat; but single f ailure considerations, and the unavailability of the st6am generatom as a heat removing component, require that at least two shutdown cooling trains be OPERABLE.

The operation of one reactor coolant pump or one shutdown cooling (low pressure safety injection) pump provides adequate flow to ensure rnixing, prevent stratification and prodtsco gradual reactivity changes during boron concentration reductions in the Reactor Coolant System The reactivity change rate associated with boron reduc' ions will, thorefore, De within the capability of operator recognition and control.

The restrictions on starting a reactor coolard .wmp in MODES 4 and 5, with one or more RCS cold legs less than or equal to 28622 F are provic'ed to prevent RCS pressure transients, l cat' sed by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against oveipressure transients and will not exceed tt e limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volums for the primary coolant to expand into or (2) by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 100'F above each of the RCS cold leg temperatures.

2/4.4.2 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia. Each stifety valve is designed to relieve 4.6 x 106 lbs per hour of saturated steam at tM valve setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur durhg shutdown. In the event that no safety WATERFORD - UNIT 3 8 3/4 4 1 6MENDMENT NO. l

i GQNTAINMENT SYSTEMS BASES  ;

}&,Q,1.7 CONTAINMENT VENTILATION SYSTEM (Continued) l Leakage integrity tests with a maximum allowable leakage rate for purge supply and exhaust isolation valves will provide early indication of resilient material seal degradation and will allow the opportunity for repair before gross leakage failure develops. The 0.60 L. leakage limit shall not be exceeded when the leakage rates determined by the leakage Integrity tests of these valves are added to the previously determined total fer all valves and penetrations subject to Type B and C tests.

Onarahititv concams for nurne tunniv and_ exhaust tantation valves other than those addressed in Actinns "a* and *b* of heificatinn 3.81.7 are mMrmundgh90ecification 3 81

• Containment Isolation Valves.'

3/[61 DsPRESSURIZATION 6ND COOLING SYSTElg 3/4.6 2.1 and 3/4.6.2 2 CONTAINUENT SPRAY SYSTEM and CONTAINMENT COOLING SYSTEM The OPERABILITY of the Containment Spray System and the Containment Cooling System ensures that containment depressurization and cooling capability wil! be available in the event of a LOCA or MSLB for any double-ended break of the largest reactor coolant pipe or main steam line.

Under post accident conditions these systems will maintain the containment pressure below 44 psig and temperatures below 269 3F during LOCA conditions or 413.5'F during MSLB conditions. The systems also reduce the containment pressute by a factor of 2 from its post accident peak within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, resulting in lower containment leakage rates and lower (,ffsite dose rates.

The Containment Spray System also provides a mechanism for remcVing iodine from the containment atmosphere under post LOCA conditions to maintain doses in accordance with 10 CFR Part 100 limits as described in Section 6.5.2 of the FSAR.

In MODE 4 when shutdown cooling is placed in operation, the Containment Spray System i<

realigned in order to allow isolation of the spray headers. This is necessary to avoid a single failure of the spray header isolation valve causing Reactor Coolant System depressurization and inadvertent spraying of the containment. To allow for this realignment, the Containment Spray Syr, tem may be taken out-of service when RCS pressure is s 400 psia. At this reduced RCS pressure and the reduced temperature ast.ociated with entry into MODE 4, the probability and consequences of a LOCA or MSLB are greatly :edaced. The Containment Cooling System is required OPERABLE in MODE 4 aad is available to provide Depressurization and cooling capability.

A train of Containment Cooling consists of two fans (powered from the same safety bus) and their associated coolers (supplied from the same cooling water loop.) One Containment Cooling train and Containment Spray train has sufficient capacity to meet post accident heat removal requirements.

Operating each containment cooling train fan unit for 15 minutes and verifying a cooling water flow rate of 625 gpm ensures that all trains are OPERABLE and that all associeted controls are functioning property. It also ensures that blockage fan or motor failure, or excessive vibration can be detected and corrective action taken.

WATERFORD UNIT 3 8 3/4 6-3 AMENDMENT NO. 89.121l

_ _ _ _ . _ _ . ~ . _ _ _

L CONTAINMENT SYSTEMS BASES 3/4s6 2.1 arid 3/4.6.2.2 CONTAINMENT SPRAY SYSTEM and CONTAINMENT COOLING SYSTEM l Kont'd)

The 18 month Surveillance Requirement verifies that each containment cooling fan actuates upon receipt of an ac'ual or simulated SlAS actuation signal. The 18 month frequency is basad on engincor5g judgment and has been shown to be acceptable through operating experience.

Verifying a cooling water flow rate of 1200 gpm to eacii cooling unit provides assurance that the design flow rate assumed in the safety analyses will be achieved. The safety analyses assumed a cooling water flow rate of 1100 gpm. The 1200 ppm requirement accounts for measurement instrument uncertainties and potential flow degradotion. Also considered in selecting the 18 month frequency were the know reliability of the Cooling Water System, the two train redundancy, and the low probability of a significant degradation of flow occurring between suNeillrices. The flow measurement for the 18 month test shall be done in a configuration equivalent to the accident lineup to ensure that in an accident situation adequate flow will be provided to the containment fan coolcrs for them to perform their safety function.

Verifying that eact valve actuates to the full open posNion provides further assurance that the valves will travel to their full pen position on a Safety injection Actuation Signal.

3/4 6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atrosphere will be isolated from the outside environment in the event of a release of radioactive f.iatorial to the containment atmosphere or pressurization of the containment and is consistent with the requirements of GDC 54 through GDC 57 of Appendix A to 10 CFR Part 50. Co'tainment isolation within the time limits specified for those isolation valves desigaed to close auto.natically ensures that the rslease of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.

The opening of locked or sealed closed containment isolation valves on a intenaittent basis under administrative control includes the following considerations: (1) stationing an operator, who is in constant communication with control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the containment.

" Containment isolation Valvet previously Table 3.6-2, have been incorporated into Plart ProcedureWT-005-026the Technical Reauirements Manoni (TRM).

3/4 6 4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain the WATERFORD - UNIT 3 8 3!< 6-4 AMENDMENT NO. 76,449,121 l

. . .. . _- -.. . - ~ . . . . .. . . . - .- - . . . - -

PLANT SYSTEMS j BASES

- 3/4.7.9 SEAL pri SOURCE CONTAMINAT!ON The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. This limitation will casura that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values.

Sealeu sources are classified into three groups according to their use, with Surveillance Requirements commensurate with the probability of damage to a source in that group, Those sources which are frequently handled are required to be tested more often than those which ur o Sealed sources which are continuously enclosM within a shielded mechanism (i.e. sealed sover within radiation monitoring or boron measuring devices) are considered to be stored and need dat be tested unless they are removed from the shield mechanism.

3/4.7.10 THIS SECTION DELETED 3/4.7.11 THIS SECTION DELETED 3/4.7.12 ESSENTIAL SERVICES CHILLED WATER SYSTEM The OPERABILITY of the essential services chilled water system ensures that sufficient chilled water is supplied to those air handling systems which cool spa;as containing equipment required for safety related operations :^d, deg n0= ! ?!!nt ep^retica, the ncnereM! ! Sp c^ . l A

f WATERFORD - UNIT 3 B 3/4 7-7 AMENDMENT NO. 50 l

y. e v -7,- a v v -

Tr-

= * --n-<+r 7 9 -- - *- -74 b+-'%*

ELECTRICAL POWER SYSTEMS BASES

. A.C. SOURCES. AND ONSITE POWER DISTRIBUTION SYSTEMS (Continued)

Operation with a battery cell's parameter outside the normal limit but within the allowable value specified in Table 4.8 2 is permitted for up to 7 days. During this 7-day period: (1) the allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability., (2) the allowable value for the average specific gravity of all the cells, not more than 0.020 below the manufactarefs recor.imenced full charge specific gravity, ensures that the decrease in rating will be 16ss than the safety margin provided in sizing; (3) the allowable value for an individual cell's specific gravity, ensures that an Individual cell's specific gravity will not be more than 0.040 below the manufacturefs full charge specific gravity and that the overall capability of the battery will be maintained within an acceptable limit , and (4) the allowable value for an individual cell's float voltage, greater than 2.07 volts, ensures the battery's capability to perform its design function.

3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment electrical penetrations and penetration conductors are protected by either deenergizing circuits not required during reactor operation or by demonstrating the OPERABILITY of primary and backup overcurrent protection circuit breakers during periodic surveillance.

The Surveillance Requirements applicable s > lower voltage circuit breakers and fuses provides assurance of breaker and fuse reliability by testing at least one representative sample of each manufacturers brand of circuit breaker and/or fuse. Each manufacturers molded case and metal case circuit breakers and/or fuses are grouped into representalis a samples which are then tested on a rotating basis to ensure that all breakers and/or fuses are tested. If a wide variety exists within any manufacturers brand of circuit breakers and/or fuses it is necessary to divide that manufacturer's breakers and/or fuses into groups and treat each group as a separate type of breaker or fuses for surveillance purposes.

The OPERABILITY of the motor-operated valves thermal overload protection and/or bypass devices ensures that these devices will not prevent safety related val /es from performing their function.

The Surveillance Requirements for demonstrating the OPERABILITY of these devices are in accordance with Regulatory Guide 1.106 " Thermal Overload Protection for Electric Motors on Motor Operated Valves," Revision 1, March 1977.

" Containment Penetration Conductor Overcurrent Protection Devices" and " Motor-Operated Valves Thermal Overload Protection and/or Bypass Devices" , previously Tables 3.81 and 3.8-2, have been incorporated into PlantProcedure-UNT-005-026the Technical Renuirements Manual (TRM). l WATERFORD - UNIT 3 8 3/4 8-3 AMENDMENT NO. 75,.92 l

1 i

l I

l l

l NPF-38-202 ATTACHMENT C

i

~ ADMINISTRATIVE CONTROLS -

PROCEDURES AND PROGRAMS (Continued)

J. OFFSITE DOSE CALCULATION MANUAL implementation.

k. Quality Assurance Program for effluent and environmental monitoring, using the guidancs in Regulatory Guide 1.21, Revision 1 June 1974 and Regulatory Guide 4,1, Revision 1, April

-1975.

- 6.8.2 Each procedure of Specification 6.8.1, and changes thereto, shall be reviewed and approved prior to implementation and reviewed periodically as set forth in administrative procedures, 6.8.3 Temporary changes to procedures of Specification 6.8.1 may be made provided:

a. The intent of the original procedure is not altered; 'P

b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Operator license on the unit affected,

c. The change is documented, reviewed and approved as required by administrative procedures within 14 days of implementation.

6,8.4.The following programs shall be established, implemented, and maintained:

. a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the containment spray, safety injection, hydrogen analyzer, and the post-accident sampling systems; and portions of the containment vacuum relief and primary sampling systems. The program shallinclude the following:

1. Preventive maintenance and periodic visualinspection requirements, and

2. Integrated ,eak test requirements for each s) stem at refueling cycle intervals or less.

b. in-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the airbome iodine concentration in vital areas under accident conditions. This program shall include the following:

- WATERFORD - UNIT 3- 6-15 AMENDMENT NO. 63, M,400 l

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3/4 4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above 1.20 during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation, this specification requires that the plant be in at least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be OPERABLE.

In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or shutdown cooling train provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops or trains (either shutdown cooling or RCS) be OPERABLE.

In MODE 5 with reactor coolant loops not filled, a single shutdown cooling train provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two shutdown cooling trains be OPERABLE.

The operation of one reactor coolant pump or one shutdown cooling (Iow pressure safety injection) pump provides adequate flow to ensure mixing, prevent stratification and produce grcdual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reductions will, therefa s, be within the capability of operator recognition and control.

The restrictions on starting a reactor coolant purnp in MODES 4 and 5, with one or more RCS cold legs less than or equal to 272*7 are provided to prevent RCS pressure transients, l caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 100*F above each of the RCS cold leg temperatures.

3/4.4.2 SAFETV VALVES The pressurizes code afety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia. Each safety valve is designed to relieve 4.6 x 105 lbs per hour of saturated steam at the valve setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety WATERFORD - UNIT 3 8 3/4 4 1 AMENDMENT NO. l

t CONTAINMENT SYSTEMS BASES 3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM (Continued) - l Leakage integrity tests with a maximum albwable leakage rate for purge supply and exhaust isolation valves will provide early indication of resilient material seal degradation and will allow the opportunity for repair before gross leakage failure develops. The 0.60 L. leakage limit shall not be exceeded when the leakage rates determined by the leakage integrity tests of these valves are added to the previously determined total for all valves and penetrations subject to Type B and C tests.

Operability concerns for purge supply and exhaust isolation valves other than those addressed in Actions "a* and *b" of Specification 3.6.1.7 are addressed under Specification 3.6.3,

" Containment isolation Valves."

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 and 3/4.6.2.2 CONTAINMENT SPRAY SYSTEM and CONTAINMENT COOLING SYSTEM The OPERABILITY of the Containment Spray System and the Containment Cooling System ensures that containment depressurization and cooling capability will be available in the event of a LOCA or MSLB for any double-ended break of the largest reactor coolant pipe or main steam line.

Under post-accident conditions these systems will maintain the containment pressure below 44 psig and temperatures below 269.3F during LOCA conditions or 413.5'F during MSLB conditions. The systems also reduce the containment pressure by a factor of 2 from its post-accident peak within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, resulting in lower containment leakage rates and kwer offsite dose rates.

The Containment P my System also provides a mechanism for removing iodine from the containment atmosphere unw post-LOCA conditions to maintain doses in accordance with 10 CFR Part 100 limits as described in Section 6.5.2 of the FSAh in MODE 4 when shutdown cooling is placed in operation, the Containment Spray System is realigned in order to allow isolation of the spray headers. This is necessary to avoid a single failure of the spray header isolation valve causing Reactor Coolant System depressurization and inadvertent spraying of the containment. .To allow for this realignment, the Containment Spray System may be taken out-of-service when RCS pressure is s 400 psia. At this reduced RCS pressure and the reduced temperature associated with ent y into MODE 4, the probability and consequences of a LOCA or MSLB are greatly reduced. The Containment Cooling System is required OPERABLE in -

MODE 4 and is available to provide Depressurization and cooling capability.

A train of Containment Cooling consists of two fans (powerod from the same safety bus) and their associated coolers (supplied from the same cooling water loop.) One Containment Cooling train and Containment Spray train has sufficient capacity to meet post accident heat removal requirements.

Operating each containment cooling train fan unit for 15 minutes and verifying a cooling water i flow rate of 625 gpm ensures that all trains are OPERABLE and that all associated controls are

functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be j detected and corrective action taken.

WATERFORD- UNIT 3 B 3/4 6-3 AMENDMENT NO. 80,434l

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CONTAINMENT SYSTEMS j BASES 3/4.6.2.1 and 3/4.6.2.2 CONTAINMENT SPRAY SYSTEM and CONTAINMENT COOLING SYSTEM l (cont'd)

The 18 month Surveillance Requirement verifies that each containment cooling fan actuates upon receipt of an actual or simulated SIAS actuation signal. The 18 month frequency is based on engineering judgment and has been shown to be acceptable through operating experience.

Verifying a cookng water flow rate of 1200 gpm to each cooling unit provides assurance that the design flow rate assumed in the safety analyses will be achieved. The safety analyses assumed a cooling water flow rate of 1100 gpm. The 1200 gpm requirement accounts for measurement instrument uncertainties and potential flow degradation. Also considered in selecting the 18 month frequency were the know rehability of the Coohng Water System, the two train redundancy, and the low probability of a significant degradation of flow occurring between surveillances. The flow measurement for the 18 month test shall be done in a configuration equivalent to the accident lineup to ensure that in an accident situation adequate flow will be provided to the containment fan coolers for them to perform their safety function.

Venfying that each valve actuates to the full open position provides further assurance that the valves will travel to their full open position on a Safety injection Actuation Signal.

3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere wil' be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment and is consistent with the requirements of GDC 54 through GDC 57 of Appendix A to 10 CFR Part 50. Containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.

The opening of locked or sealed closed containment isolation valves on a intermittant basis under administrative control includes the following considerations: (1) stationing an operator, whc is in constant communication with control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the containment.

Containment ! solation Valves", previously Table 3.6-2, have been incorporated into the Technical Requirements Manual (TRM).

3/4 6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain the l

! WATERFORD - UNIT 3 B 3/4 6-4 AMENDMENT NO. 4,44,44 l

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PLANT SYSTEMS -

BASES

--3/4.7.9 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values.

Sealed sources are classified into three groups according to their use, with Surveillance Requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e. sealed sources within -

radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shield mechanism.

3/4.7.10 THIS SECTION DELETED 3/4.7.11 THIS SECTION DELETED 3/4.7.12 ESSENTIAL SERVICES CHILLED WATER SYSTEM The OPERABILITY of the essential services chilled water system ensures that sufficient chilled water is supplied to those air handling systems which cool spaces containing equipment required for safety-related operations. l 1-i 2

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WATERFORD UNIT 3 B 3/4 7-7 AMENDMENT NO. 50 l L

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ELECTRICAL POWER SYSTEMS BASES A.C. SOURCES, AND ONSITE POWER DISTRIBUTION SYSTEMS (Continued)

Operation with a battery cell's parameter outside the normal limit but within the allowable value specified in Table 4.8-2 is permitted for up to 7 days. Dunng this 7-day period: (1) the allowablo values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability., (2) the allowable value for the average specific gravity of all the cells, not more than 0.020 below the manufacturers recommended full charge specific gravity, ensures that the decrease in rating will be less than the safety margin provided in sizing; (3) the allowable value for an individual cell's specific gravity, ensures that an individual cell's specific gravity will not be more than 0.040 below the manufacturers full charge specific gravity and that the overall capability of the battery will be maintained within an acceptable limit., and (4) the allowable value for an individual cell's float voltage, greater than 2.07 volts, ensures the battery's capability to perform its design function.

3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICE _S. S Containment electrical penetrations and penetration cor.ductors are protected by either deenergizing circuits not required during reactor operation or by demonstrating the OPERABILITY of primary and backup overcurrent protection circuit breakers during periodic surveillance.

The Surveillance Requirements applicable to lower voltage circuit breakers and fuses provides assurance of breaker and fuse reliability by testing at least one representative sample of each manufacturers brand of circuit breaker and/or fuse. Each manufacturefs molded case and metal case circuit breakers and/or fuses are grouped into representative samples which are then tested on a rotating basis to ensure that all breakers ano/cr fuses are tested. If a wide var.ety exists within any manufacturers brand of circuit breakers and/or fuses it is necessary to divide that manufacturers breakers and/or fuses into groups and treat each group as a separate type of breaker or fuses for surveillance purposes.

The OPERABILITY of the motor-operated valves thermal overload protection and/or bypass devices ensures that these devices will not prevent safety related valves from performing their function.

The Surveillance Requirements for demonstrating the OPERABILITY of these devices are in accordance with Regulatory Guide 1.106 " Thermal Overload Protection for Electric Motors on Motor Operated Valves," Revision 1, March 1977.

" Containment Penetration Conductor Overcurrent Protection Devices" and " Motor-Operated Valves Thermal Overload Protection and/or Bypass Devices" , previously Tables 3.8-1 and 3.8-2, have been incorporated into the Technical Requirements Manual (TRM). l I

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I WATERFORD - UNIT 3 8 3/4 8-3 AMENDMENT NO. 3:5, Q3 l

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NPF-38-202 ATTACHMENT D i

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LIST OF ENCLOSED REFERENCES

1. SER dated May 20,-1997, transmitted by NRC letter from C. P. Patel to C. M.

Dugger dated May 20,1997,

2. . Entergy Letter from M. B. Sellman to U.S. NRC Document Control Desk dated August 21,1996.

3. Waterford 3 LDCR 97-0225 dated June 9,1997.

4. Waterford 3 Technical Specification 3.4.1.3, Amendment 106.

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l REFERENCE '

.. 9 SER dated May 20,1997, transmitted by NRC letter from {

C. P. Patel to C. M. Dugger dated May 20,1997 l 3

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y- *, UNITED STATES -

l j NUCLEAR REGULATORY COMMISSION wa:HINGToN, D.C. 2004M001 REC 9VED

• • • • • May 20, 1997 WAY 28 l'/97 m 91* 010(, *

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Mr. Charles M. Dugger Vice President Operations Entergy Operations, Inc.

P. O. Box B Killona, LA 70066

SUBJECT:

ISSUANCE OF AMENDMENT NO. 128 TO FACILITY OPERATING LICENSE NPF WATERFORD STEAM ELECTRIC STATION, UNIT 3 (TAC NO. M96495)

Dear Mr. Dugger:

The Commission has issued the enclosed Amendment No.128 to Facility Operating License No. NPF-38 for the Waterford Steam Electric Station, Unit 3 (Waterford 3). The amendment consists of changes to the Operating License in response to your application dated August 21, 1996, as supplemented by letter dated March 17, 1997.

The amendment approves revision of Attachment I to Facility Operating License No. NPF-38 concerning design and testing modifications in the Containment Vacuum Relief System (CVR) that penetrates the primary containment at Waterford 3. The penetrations affected are commonly referred to as Penetrations 53 and 65.

A copy of our related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's next biweekly Federal Reaister notice.

Sincerely, 0dwD4 ffb(

Chandu P. Patel, Project Manager Project Directorate IV-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket No. 50-382

Enclosures:

1. Amendment No.128 to NPF-38

2. Safety Evaluation cc w/encls: See next page JqG hhbk h Y l

Mr. Charles M. Dugger Entergy Operations, Inc. Waterford 3 cC:

Administrator Regional Administrator, Region IV Louisiana Radiation Protection Division U.S. Nuc1 car Regulatory Commission Post Office _ Box 82135 611 Ryan Plaza Drive, Suite-1000 i

Baton Rouge, LA 70884-2135 Arlington, TX 76011 Vice President, Operations Resident Inspector /Waterford NPS Support Post Office Box 822 Enterg'/ Operations, Inc. Killona, LA 70066 P. O. Box 31995 Jackson, MS 39286 Parish President Council St. Charles Parish Director P. O. Box 302 Nuclear Safety & Regulatory Affairs Hahnville, LA 70057 Entergy Operations, Inc.

'P. O. Box B Executive Vice-President Killona, LA 70066 and Chief Operating Officer Entergy Operations, Inc.

Wise, Carter, Child & Caraway P. O. Box 31995 P.-0. Box 651 Jackson, MS 39286-1995 Jackson, MS 39205 Chairman General Manager Plant Operations Louisiana Public Service Commission Entergy Operations, Inc. One American Place, Suite 1630 P. O. Box B Baton Rouge, LA 70825-1697 Killona, LA 70066 Licensing Manager Entergy Operations, Inc.

P. O. Box B Killona, LA 70066 Winston & Strawn 1400 L Street, N.W.

Washington, DC 20005-3502

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E NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 30006 4001 b.....o ENTERGY OPERATIONS. INC.

DOCKET NO. 50-382 WATERFORD STEAM ELECTRIC STATION. UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.128 License No. NPF-38

1. The Nuclear Regulatory Comission (the Comission) has found that:

A. The application for amendment by Entergy Operations. Inc. (the licensee) dated August 21, 1996, as supplemented by letter dated March 17, 1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the-Comission's rules and regulations set forth in 10 CFR Chapter I; l B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C. There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the heaith and safety of the public, and (ii) that such activities will br.

conducted in compliance with the Comission's regulations; D. The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the pub.'ic; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

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2. Accordingly, the license is amended by revising Attachment 1 of Facility Operating License No. NPF-38.

3. This license amendment is effective as of its date of issuance to be implemented within 90 days.

FOR THE NUCLEAR REGULATORY CONilSSION C/4mdtt 79[

Chandu P. Patel, Project Manager Project Directorate IV-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation

Attachment:

Attachment 1 of the Operating License Date of Issuance: May 20, 1997 l

1 ATTACHMENT 1 WATERFORD STEAM ELECTRIC STATION OPERATING LICENSE NPF-38 This attachment identifies items which must be completed to the Commission's satisfaction prior to startup-following the refueling outage number 8.

Non-essential Containment Vacuum Relief Sensing Lines:

• Penetration 65 will be modified to reflect a Containment Leak Rate Test connection as indicated in licensee submittal dated August 21, 1996 (Attachment C page 4 of 4).

• Penetration 53 will be modified such that two automatic containment isolation valves will be located outside containment with continuous direct position indication in the control room as indicated in licensee submittal dated August 21, 1996 (Attachment C page 4 of 4).

AMENDMENT NO. 128

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UNITED STATES

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.128 TO FACILITY OPERATING LICENSE NO. NPF-38 ENTERGY OPERATIONS. INC.

WATERFORD STEAM ELECTRIC STATION. UNIT 3 DOCKET NO. 50-382

1.0 INTRODUCTION

By applicatin. Jated August 21. 1996, as supplemented by letter dated March 17, 1997, Entergy Operations, Inc. (the licensee), submitted a request for changes to the License No. NPF-38 issued for Waterford Steam Electric Station, Unit 3 (Waterford 3). The requested changes involve the design and 8

testing modifications at Waterford 3. These modifications require the Nuclear Regulatory Commission (NRC) staff approval fJr the adequacy of both the containment isolation arrangement and the Type C leak testing of the barriers of two instrument sensing lines at Waterford 3. These lines are in the Containment Vacuum Relief System (CVR) that penetrate the primary containment.

The penetrations are commonly referred to as Penetrations 53 and 65. These modifications were found necessary because it was discovered that the current plant configuration did not agree with information provided to the NRC st-ff during the licensing process for Waterford 3. In addition the licensee proposed to revise a License Condition (Attachment 1) to the Operating License No. NPF-38 for Waterford 3 to reflect the modifications in Penetration 53 and 65.

The March 17, 1997, letter provided additional information that did not change the initial proposed no siginficant hazards consideration determination.

2.0 DllCUSSION AND EVALUATION During the licensing process for Waterford 3, the staff requested additional information concerning the isolation arrangements and testing commitments for Penetrations 53 and 65 at Waterford 3.

The licensee's response described the arrangement by indicating that penetrations 53 and 65 each contain two instrument lines. One is considered an essential line sensing differential pressure across the containment vessel and provides a signal to actuate the vacuum relief system, the other monitors this differential pressure and provides an input to the plant computer. This other signal is considered to be non-essential. Whether or not the line is essential is important since the isolation requirements differ between essential and non-essential-lines.

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The response indicated that the essential line contains an excess flow check valve. The non-essential line also contains an excess flow check valve but a commitment was made to also add a solenoid operated valve during the first refueling outage. This added valve would be automatically closed on a containment isolation signal. This commitment to add the solenoid valves was included as a license condition in the Waterford 3 license.

In addition to the excess flow check valves and the solenoid valves, the licensee indicated that both lines formed a closed system outside containment, are seismically qualified, and terminate in an area exhausted by the filters of the Controlled Ventilation Area System. Based on this information, the staff concluded that the closed system was an acceptable barrier.

Therefore, the essential line was assumed to have two barriers; the excess flow check valve and the closed system. The non-essential line also had the same two barriers and a third barrier would be added during the first refueling outage. This barrier would be a solenoid operated automatic valve.

The licensee further indicated that with the above isolation arrangements for both lines, Type C leak testing was not required to be performed. The NRC in the Waterford 3 SER accepted the design configuration and testing requirements for the CVR instrument lines in penetrations 53 and 65. The commitment to install the solenoid valves was satisfied on January 19, 1987, however, the license condition was not removed.

l The above situation continued until it was discovered by the licensee that the l conditions for a closed system were not consistent with the as built situation. This required a reassessment of both the isolation arrangements as well as the testing requirements for both penetrations.

The reassessment separated the essential and non-essential lines, since tha requirements are different. The discussion will begin with the essential lines in penetrations 53 and 65.

The essential lines do qualify for the criteria of Regulatory Guide (RG) 1.141. The significance of this criteria is to not require automatic containment isolation. The next criteria that was identified as applicable t was General Design Criterion (GDC) 56. This GDC addresses the isolation requirements of penetrations that are connected to the. containment.

l Generally, this GDC requires specific isolation valve types that would be acceptable as containment isolation barriers. However, instrument lines that are considered to be vital or essential to the overall safety of the power plant have further relaxations. Excess flow check valves, for example, are normally not considered ~ as acceptable containment isolation barriers.

However, GDC 56/57 allow specific relaxations for signals that are considered to be important enough that interruption of the signal represents a reduction to the plant safety. This line class is identified as essential.

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l This consideration allowed the use of an excess flow check valve as an acceptable containment isolation barrier. The next consideration was the determination of whether or not the system represented a closed system beyond the excess flow check valve. Normally, the system should be designed in accordance with Quality Group B standards as defirad by RG 1.26. This means ASME Section III, C1;ss 2. However, instrument lines are not covered by RG 1.26. Therefore, the licensee criteria classified these lines as ISA-67.02. The licensee considered that this classification was consistent with the endorsement by the NRC. The staff agrees that the above criteria is consistent with the requirements of Quality Group C (i.e., ASME,Section III, Class 3) which is consistent with the staff's interpretation of the criteria governing instrument lines. The staff conclusion is that the design meets all of the criteria of a closed system. Therefore, the staff concurs that the

, system is a closed system.

The staff concludes that the existing containment isolation provisions of the essential lines of penetrations 53 and 65 are acceptable without any hardware modifications.

Consideration of the non-essential lines is sign ffrantly different from the essential lines. Each line has a solenoid gloh vaha that will automatically close on a containment isolation signal. Howewar, the second containment isolation barrier was in question. The remaining tt.Mng beyond the snienoid valve is non-safety. In addition, the monitoring lines downstream of the isolation valves are not classified as seismic Category I. These combined variations cause the staff to conclude that a closed system is not present and therefore, the existing hardware arrangement does not meet the containment isolation requirements.

An important consideration is the acceptability of the excess flow check valve as a containment isolation barrier. The staff concludes that this barrier is unacceptable since it does not meet the criteria of RG 1.11. The acceptably of this type of barrier can only be justified for the essential signals. In order to comply with the criteria of Safety Guide ll, one must be able to show that the importance of the line signal is safety significant. For a non-safety line, this condittun cannot be met. The relaxation of the use of an excess flow check valve as an acceptable barrier is therefore unacceptable.

Therefore, an acceptable barrier in addition to the solenoid automatic valve is required.

The licensee recognized this limitation and proposed an additional valve. The licensee proposes to add a second solenoid automatic valve. This is acceptable'to the staff.

'There were two penetrations for the non-safety function. The licensee proposes that one of the lines be closed via closing the penetration with seal welding. For the other penetration, the license will add the automatic solenoid valve from the seal' closed line as-the second valve meeting the acceptance criteria of GDC 56. Based on the above, the staff finds the isolation criteria of penetrations 53 and 65 acceptable.

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For the essential sensing instrument lines, the licensee indicated that the lines will be pressurized and leak tested at refueling intervals. This satisfies the testing of the closed syste'n. The testing pressure will be 48 psig and the measured leakage will be added to the bypass leakage total. In addition, in light of the failures of the excess flow check valves, functional testing will be performed at refueling intervals.

For the non-essential lines, the licensee has indicated that local leak rate testing (LLRT) and inservice testing (IST) program testing will be performed on both automatic solenoid containment isolation valves. LLRT means Type C testing under the Appendix J program.

Based on the above information, the staff finds that the testing requirements have been satisfied and, therefore, are acceptable for both the essential and non-essential instrument lines in penetrations 53 and 65.

The licensee proposed to revise a License Condition (Attachment I to the License) to reflect the proposed changes. The proposed changes in the License Condition are acceptable to the staff.

Based on the above findings, the staff finds the proposed modifications to the lines in penetrations 53 and 65 acceptable. In addition, the testing criteria proposed for the barriers are also acceptable. The proposed changes in the License Condition are also acceptable.

3.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Louisiana State official was notified of the proposed issuance of the amendment. The State official had no com:nents.

4.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requf rement with respect to installation or use of a facility component located within the restrict ' trea as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and'that-there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a pro-posed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (61 FR 57484).

, Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

5.0 CONCLUSlQN The Commission has concluded, based on the considerations discussed above.

that: (1) there is reasonable assurance that the health and safety of the _

public will not be endangered by operation in the proposed manner,-(2) such activitiet will be conducted in compliance with the Commission's regulations, .

and (3) the issuance of the amendment will not be inimical to the common def ense and security or to the health and safety of the public.

Principal Contributor: J. Kudrick Dati: May 20, 1997 4 ,_, * '

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REFERENCE 2-l Entergy Letter from M. 8, Sellman to U.S. NRC Document  ;

Control Desk dated August 21,1996 ,

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- August 21,1996 U.S. Nuclear Regulatory Commission Attn: Document Control Desk ,

Washington, D.C. 20555

Subject:

Waterford 3 SES Docket No. 50-382 Ucense No. NPF-38 License Amendment Request NPF-38-181 Discrepancy Regarding the Design and Testing of Instrument Sensing Lines Penetrating the Primary Containment Gentlemen:

, On July 23,1996, a Condition Report was generated because it appeared that the carrent plent configuration did not agree with information provided to the NRC review starf during the initial licensing phase at Waterford 3. The information was provided to justify the leak testing method that would be performed, on the Containment Vacuum Relief instrument sensing lines penetrating the primary containment, pursuant to 10 CFR 50 Appendix J. The justification provided by Waterford 3 was subsequently documented by the NRC staff in the Waterford 3 SER, issued in July 9,1981, as the bases for accepting the design configuration and testing requirements for these instrument lines. Upon confirmation of the above discrepancy, Waterford 3 notified and described the condition to the NRC, in that discussion Wate:1ord 3 indicated that an evaluation was being conducted under-the provisions of 10 CFR 50.59 to dotermine thG impact of the current configuration on plant safety. This evaluation analyzed the current plant configuration under postulated accident conditions and dstormined that the results vere within the existing licensing bases limits. Additional provisions were l then taken to provide for def9nse in depth approach. These provisions were atso l

evaluated and determined acceptable under the provisions of 10 CFR 50.59.

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~ License Amendment Request NPF-38-181 Discrepancy Regarding the Design and Tesiing of Instrument Sensing Lines Penetrating the Primary Containment W3F1-96-0144 '

Page 2 August 21,1996 At this point Waterford 3 reviewed and evaluated a variety of solutions to resolve the identified discrepancy. The proposed solution as described herein was chosen ,

because 1) the proposed change can be performed in a timely fashion 2) it can be performed on line without risk to the plant or plant personnel 3) the proposed fix will provide for a level of protection commensurate with the safety function to be performed.

Waterford 3 believes the proposed modification is an equivalent method for complying with the recommended position contained in Regulatory Guide 1.11 and seeks the

' staffs approval, in addition, the proposed change describes a modification to the plant that will remove a containment isolation valve and cap the line. The containment isolation valve is an automatic valve and therefore, is provided with direct position l indication in the control room pursuant to Regulatory Guide 1.97 Post Accident L

Monitoring Instrumentation . The requirement for position indication on the penetration that will be affected by the modification, was included in the Operating License as a condition to be met prior to startup following first refuel. The license condition concerning position indication does not contribute to the discrepancy associated with the affected penetrations. However, with the proposed modification implemented no position indication will be required because no valve will be installed. Therefore, the attached description and safety analysis supports a license amendment request pursuant to 10 CFR 50.90 to remove statements in the license that will not be applicable upon approval and implementation of the proposed change. The proposed change to the Operating License does not remove or reduce any regulatory roouirement.

Watofford 3 intends to implement the proposed solution in a timely fashion as described herein.

1 The proposed change has been evaluated in accordance with 10CFR50.91(a)(1) using criteria in 10CFR50.92(c) and it has been determined that this change involves no significant hazards considerations. The bases for these determinations are included in the attached submittal.'

I

_ __ _ __ __. _ . , _ ~ ~ . - - _ __ - - , _ _ _

i l

License Amendment Request NPF-38-181 Discrepancy Regarding the Design and Testing of instrument Sensing Lines Penetrating the Primuy Containment W3F1-96-0144

~ Page 3 August 21,1996 Waterford 3 intends to install the modification in an expeditious maroer and requests that the implementation date for this change be within 90 days of NRC issuance to allow for FSAR and procedure revisions necessary to implement this change. Although this request is neither exigent nor emergency, your prompt review is requested.

Should you have any questions or comments concerning this request, please contact Mr. James Fisicaro at (504) 739-6242.

Very truly yours, r

pmm M.B. Sellman Vice President, Operations Waterford 3 MBS/PLC/ssf

Attachment:

Affidavit NPF-38-181 cc: L.J. Callan, NRC Region IV C.P. Patel, NRC-NRR R.B. McGehee N.S. Reynoida NRC Resident inspectors Office Administrator Radiation Protection Division (State of Louisiana)

American Nuclear Insurers

l UNITED STATES OF AMERICA -

NUCLEAR REGULATORY COMMISSION

• ' In the matter of - )

)

Entergy Operations, incorporated ) Docket No. 50-382 Waterford 3 Steam Electric Station )

AFFIDAVIT F.J. Drummond, being duly sworn, hereby deposes and says that he is Director l Site Support - Waterford 3 of Entergy Operations, incorporated; that he is duly authorized to sign and file with the Nuclear Regulatory Commission the attached License Amend;nent NPF-38-181; that he is familiar with the content thereof; and that the .

matters set forth therein are true and correct to the best of his knowledge, information l

_ and belief.

F.J. Orunwnond Director, Site Support STATE OF LOUISlANA )  ;

) as PARISH OF ST. CHARLES )

. Subscribed and swom to before me, a Notary Public in and for the Parish and State above named this . day of (cv , d 1996. I

J

> <-e e Notary Public Jh

. My Commission expires d4 .

I l ~.

i DESCRIPTION AND SAFETY ANALYSIS l OF PROPOSED CHANGE NPF 384181  !

i The proposed change will eliminate Attachment i to the Operating License. The f

attachment contains a requirement that was completed in January 1987.  !

Existina Operatina License NPF-38 l

Attachment A provides that portion of the Waterford 3 Operating License NPF49 that i v/ill be affected by the proposed change.

Proposed Revision to the Operatina License NPF-34 t Attachment B indicates the requented changes to the Operating License NPF 38 -

Prouem Deecription During the pre licensing phase at Waterford 3, the NRC submitted a request for additional information (RAl) concerning the Type C leak testing of two instrument sensing lines in the i Containment Vacuum Relief System (CVR) that ponerate the primary containment (Penetrations 53 & 65). The request in part stated the following:

'The Justification given in Table 6.2-43 for not including Penetrations 53 and 65 in Type C leak tests is inadequate. Show that containment isolation valves associated with these penetrations do not constitute potential containment atmosphere leak paths following a loss 4-coolant accident.'

Waterford 3 resp 4 to the request above as follows:

l ' Penetrations 33 and 65 each contain two instrument lines. One senses differential l

pressure across the containment vessel and provides a signal to actuate the vacuum relief system; the other monitors this differential pressure and provides an ,

, input to the plant computer. The actuation (i.e. Essential) line contains an excess I

flow check valve outside containment; the monitoring (i.e., Non-Essential) line has l an esones flow check valve and will be orovided with a solenoid operated valve.

closed on a containment isolation signal. The excess flow check valve is designed to close on excess flow and reopen when conditions retum to a spec 4ied normal state. Both of these lines formardoned_ system outside. containment. are seismically cualified and terminate in an area exhausted by the filters of Controlled Ventilation Area System. A Type C test is, therefore, not required or perfonwl on these lines."

L

.1 l

i l

The justification provided by Waterfore13 (underlined above) was used and documented by the NRC in the Waterford 3 SER, as the bases for accepting the design configuration l and testing requirements for the CVR instrument lines in penetrations 53 and 65.  ;

l Cor:trary to the above, it was recently discovered that neither of these lines terminate in

an area exhausted by the filters of the Controlled Venti

ation Area System (CVAS). In l addition, the monitoring line is not in compliance with the design critona for crediting a closed system as a leakage boundary to preclude bypass leakage.

4 Criteria Currently Appiled to the CVR instrument Linee General Design Criteria (GDC) 56 states the following:

'Each line that connects directly to the containment atmosphere and penetrates primary - ,

reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions fa a specific  !

class of lines such as instrument lines, are acceptable m some other definerf bas!..' ,

Regulatory Guide 1.11, Instrument Lines Penetrating Primary Containment, provides the NRC ondorsed method of implementing GDC 56 requirements for instrument lines.

The CVR instrument line penetrations are GDC 56 penetrations because they communicate directly with the containment atmosphere. Waterford 3 applied the l recommendations of R.G.1.11 to the CVR instrument lines to comply with the containment isolation requirements as follows:

a) All CVR instrument lines were equipped with a manually operated l valve installed as close to the containment as practicable and, b) A self actuated excess flow check valve installed as close to the

containment as practicable downstream of the manually operated valve.

I As a result of an apparent NRC concem with isolating the Non-Essential lines, i Waterford 3 committed (as indicated in the response to the RAl) to replace the manual valves with automatic solenoid operated isolation valves.

Docketed information appears to indicate that the design of these instrument lines, up to and including the outboard isolation valve was in full compliance with R.G 1.11 and therefore, acceptab!e to ths NRC, in so far as containment isolation and GDC 56 was concemed. >

i i 2 l

a - .- - - _ _ .--.---_-_.-.--_-----_-x---

Attachment i to Operating Llconse NPF.38 Initially, all four CVR instrument lines were equipped with a normally open manual valve  !

upstream from the excess flow check valves. As a result of installing the automatic isolation valves (CVR 401 A & CVR 4018) in each of the monitoring lines, it became apparent during the implementation of R.G.1.97 (Post Accident Monitoring Instrumentation), that these valves would require direct position indication in the control room. Waterford 3 committed to install this feature prior in startup following the first refueling outage. The staff documented this commitment in Supplement 8 to the SER and incorporated this requirement as a license condition in Attachment i to the Operating License. This Condition of the Operating License was completed January 19, 1987. The proposed change describes a plant modification that will remove the (

containment isolation valve (CVR 4018) from the non-essential monitoring line in penetration 65. A welded pipe cap will then be installed on the line and the line will be abandoned in place. As such valve position indication for penetration 65 will no longer be applicable and the license condition in Attachment 1 to NPF-38 should be modified.

The modification to the license will eliminate Attachment 1. This will not remove or +

reduce any regulatory requirement at Waterford 3. The license condition was a method to assure that Waterford 3 fully complied with the provisions of R.G.1.97. Subsequent to this proposed change, Waterford 3 will continue to be committed to R.G.1.97 and in full compilance. The modification will include an additional automatic isolation valve in penetration 53 and position !ndication in accordance with R.G.1.97 will be provided.

Leak Testing Criteria i

Pursuant to Branch Technical Position (BTP) CSB 6-3, Determination of Bypass Leakage Paths in Dual Containment Plants, 'If a closed system is proposed as a leakage boundary to preclude bypass leakage, then the system should:

a. Either (1) not directly communicate with the containment atmosphere, or (2) not directly communicate witn the environment, following a loss-of-coolant accident.

b. Be designed in accordance with Quality Group B standards, as oefined by Regulatory Guide 1.26. (Systems designed to Quality Group C or D standards that qualify as closed systems to preclude bypass leakage will be considered on a case-by .,ase basis.)

c. Meet seismic Category I design requirements,

d. Be designed to at least

• nrimary containment pressure and temperature design conditions.

3

f 1  !

l i

e. Be designed for protection against pipe whip, missiles, and jet forces in a  !

4 manner similar to that for engineered safety features.

l l

f. Se tested for leakage, unless it can be shown that during normal plant  :

) operations the system integrity is maintained.'

! This preceding criteria was inappropriately used to justify not ;erforming Type C leak testing in accordance with 10 CFR Appendix J on the CVR instrument sensing lines.

Current Configuration Ceft Eseontial/ Sensing instrument Line The CVR protects the containment vessel by maintaining the pressure differential across the vessel lower than the design value. If the containment atmosphere is .

rapidly cooled lowering containment pressure a differential pressure between the containment and the annulus that is higher than the design pressure difference could be created. The CVR penetrations 53 and 65 each contain an essential instrument i ano. The sensing line senses differential pressure across the containment vessel and provides a signal to actuate tN CVR. Due to its primary safety function, these redundant instrument sensing lines of the CVR are designated as ' essential

• in .

accordance with R.G.1.141, thus, not requiring automatic containment isolation.

i The CVR instrument sensing lines meet the specified regulatory positions of 6..G.1.1' paragraphs C.1.a, C,1.b, C.1.c, C.1.d and C.1.e.

The CVR instrument sensing lines meet the criteria of BTP CSB 6-3 (previously listed) for crediting a closed system as a leakage boundary to preclude bypass leakage with the exception of item (b). Item (b) requires the system to be designed in accordance with Quality Group 8 standards, as defined by R.G.1.26 ( l. e., ASME Section lil, Class 2). However, instrument lines are not covered by R.G.1.26. Section B of R.G.1.26 indicates that instrument lines should be designed, fabricated, erected, and tested to standards commensurate with the safety function to be performed. Standard Review ,

Plan 3.2.2, System Quality Group Classification, Section 111 provides examples of

! systems (which includes instrument systems) that are not covered by R.G.1.26, and according to the staff thould be classified as Quality Group C (i.e., ASME, Section lil, Class 3). This le consistent with the design criteria applied at Waterford 3 (i.e., ISA-67.02 ' Nuclear-Safety-Related Instrument Sensing Line Piping and Tubing Standards ,

for Use in Nuclear Power Plants' as endorsed by NRC Draft R.G. entitled instrument Sensing Lines, Position C.4.) The CVR instrument sensing lines form a seismically r qualified, closed system outside containment.

i 4

= _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ - ~ _ . . . _ _ _ _ _ _

d-These lines are ASME Section lil, Class 3 stainless tubing designed and built to meet i seismic Category I criteria of LOU 1564 B-430 which defines support spans for weight, i seismic, and thermal expansion and movements. The sont,ing line tubing is rated for a ,

j design pressure in excess of 4,500 psig at 300'F in accordance with the manufacturers f data. The system design for this section of tubing is 44 psig at 300'F. Therefore, the i tubing design conditions are approximately 1% of its rated capacity. Even though the i j expected MSLB temperature inside the containment peaks above 300*F, it should be t 2

noted that the containment will be above 300*F for a very short period of time, and the instrument tubing is not expected to reach this temperature due to delayed heat ,

transfer. The tubing support / span criteria used at Waterford 3, limits bending stresses

~

in the tubing to less than 50% of the allowable. This fact, combined with the extremely low internal pressure stress gives a combined stress well below the break exclusion '

criteria.  !

The instrument cabinets C 3A(B) are seismic Category I and safety related. The

, instruments are Safety Class 1E and have a static pressure rating of 1000 psig. The  ;

sensing line and valves downstream of the excess flow check valve (CVR-302A&B) are - '

designed to ASME Section lil, Class 3. In light of these facts, the CVR instrument sensing lines may be credited as a pressure barrier, as described under Class D penetrations in the Waterford 3 FSAR, subsection 6.2.4.1.2. A pressure barrier is said to exist when the line is part of a closed system outside containment which is designed for pressure equal to or greater than containment.

Failure Analyste i

Due to redundancy, a single active failure in the CVR instrument sensing line will not prevent the system from performing its safety related function.

A gross passive failure in the CVR instrument sensing line concurrent with a LOCA is not considered credible due to the design qualification described above. Branch Technical Position ASB 3-1, Plant Design for Protection Against Postulated Piping i Failures in Fluid Systems Outside Containment, Section b.3 a, states tot piping failures should be postulated in accordance with BTP ME8 3-1, Section 3.a of BTP ME8 31 refers to R.G.1,11 for piping 1 inch or less or tubing. R.G.1.11 limits the occurrence of a postulated passive failure, to periods during normal plant operations.

A gross failure in the Essential CVR instrument sensing lines during normal plant l

operations will not impact public health and safety because there is no radioactive source term. Based on the above bypass leakage due to passive failure in the CVR instrument sensing lines will be limited to leakage from failed valve packing or mechanical seal rather than the complete severance of the line.

i This is consistent with ANSI N658 Single Failure Criteria for PWR Fluid Systems, Section 3.6. Leakage of this nature will be precluded by testing and inspection at periodic intervals. .

5

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l Proposed Solution for the CVR Essential / Sensing instrument Line f

The proposed solution for the CVR sensing line will not require any plant modification.  !

The following discussion provides the regulatory licensing bases for the design and [

acceptance of these safety related essentialinstrument lines. l Based on the above, the Waterford 3 design meets Reg. Guide 1,11 requirements up i i to the downstream isolation valves and meets the requirements of ANSI /ANS-  !

56.2/N271 1976, ISA 67.02-1980 and the Draft Reg. Guide Instrument Sensing Lines. I These lines meet the critoria of BTP CSB 6 3 for crediting a closed system as a l l leakage bop dary to preclude bypass leakage by being designed, fabricated, erected, i and tested to standards commensurate with the safety function to be performed. The l proposed change will apply the appropriate testing and acceptance criteria to ensure ,

that any leakage associated with these potential bl pass leakage paths, will not exceed the limits used in the Waterford 3 safety analysis or result in a significant increase in analyzed dose consequences. Therefore, no plant modification will be pursued.  !

Licensing Bases l

, Upon approval of the proposed change, a revision to the FSAR will be processed to specifically describe the bases for meeting BTP CSB 6-3 Section g.b as it applies to the CVR instrument sensing lines. In addition these lines will be pressurized and leak

tested at refueling intervals in accordance with Plant Operating procedure OP-903-110.

Leakage within the acceptance limits of this procedure will ensure compliance with .

. technical specification requirements and other requirements associated with  !

[ containment integrity. The normally closed valves in the instrument cabinets will be included in existing administrative controls that periodically verify valve position.

Current Configuration for the CVR Noneesential/ Monitoring Instrument Line The CVR monitoring instrument lines communicate directly with the containment atmosphere. Each redundant line runs from containment to a solenoid globe valve that  !

closes automatically on a CIA 8. An excess flow check valve is located downstream of i the automatic valves. The tubing for the monitoring lines up to and including the i excess flow check valves is ASME Section ill, Class 2, seismic Category I. The remaining portion of the lines are non-safety tubing and although seismically supported, the monitoring lines downstream of the isolation valves are not classified as seismic Category 1. These lines terminate at the C-4 cabinet that is located outside the area exhausted or filtered by the CVAS. .

The instrumentation for the monitoring lines messe.e differential pressure between containment and annulus, and provide signal to the plant computer and containment purge, These lines do not assist in mitigating the effects of an accident nor are they necessary for safe plant shutdown, therefore, they are categorized as ' nonessential" pursuant to R.G.1.141, and require automatic isolation. j 6

i An excess flow check valve with flow limiting orifice, is essentially a containment isolation valve with a post accident position that is open. This fact combined with (1) only one automatic isolation valve capable of closing in the line and (2) the design of the lines downstream of the excess flow check valve (i.e., non safety, non seismic),

4 does not meet the criteria for containment isolation following a LOCA and single active failure. Waterford 3 has evaluated the consequences of an accident concurrent with a single active failure in these lines and obtained acceptable results (i.e., dose consequences within GDC 19 and 10 CFR 100 limits). However, to provide for defense in depth during this interim period, CVR 401 A and B are currently de-activated and secured in the closed position. CVR 401 A may be opened for various reasons such as, whenever the Containment Purge is placed in service, for the conduct of technical .

specification surveillance testing, or when the PMC out of se vice. The Wate:Tord 3  !

Technical Speciflettions limit Containment Purge to less than 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> por 365 days.

This 90 hour0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> limit is currently imposed on the opening of CVR 401 A.

Proposed solution for the NoneesentiallMor.itorire instrument Line Waterford 3 proposes to modify the CVR instrument monitoring lines at power by abandoning one line and providing the remaining line with two automatic isolation valves in series located outside containment. We believe this configuration to be an improvement over that allowed by R.G.1.11 regulatory position C.2.a, and therefore, it is proposed as an acceptable alternative method for complying with the NRC recommended position.

Attachment C provides a stop by step visual presentation of the proposed plant modification. Waterford 3 intends to perform this modification while at power and in strict compliance with all applicable license / regulatory requirements.

While this modification is performed, administrative controls will require containment integrity to be maintained by a seismic Category 1, ASME Section lil, Class 2, passive containment isolation device. The criteria associated with containment isolation defines a passive isolation device as a closed manual valve, a de-activeted automatic valve secured in the closed position, a blind flange, or closed system. During the performance of this modification no single creditsle failure or malfunction of an active component will result in a loss of containment isolation or containment leakage that exceeds the limits assumed in the Waterford 3 safety analysis.

The proposed modification involves the following steps:

1. The monitoring line from penetration 65 will be cut and capped in the C-4 instrument cabinet. instruments CVR-IDPT-50178 and CVR-IDPT-5017C will be tied in to the monitoring line from penetration 53.

l l 7

2. A containment entry will be made and a threaded ASME Section Ill, Class 2 pipe cap will be installed on the Penetrations 53 and 65 monitonng lines. A compression tubing cap will also be installed on the penetration 65 line terminating in the instrument cabinet. A local leak rate test (LLRT) on penetration 65 will be performed by pressurizing the line from the test connection located between CVR 401B and containment with CVR 4018 closed to ensure the leak tight integrity of the cap installed inside containment. This test will again be performed with CVR 401B open to ensure the leak tight integrity of the cap installed outside the CVAS boundary. Penetration 53 will be pressurized between the pipe cap and the installed instrumentetion. These caps will provide for a passive containment /CVAS barrier while the line outside containment is being cut as described in the following steps.

3. CVR 4018 and CVR 402B will be cut out of the system and line end caps will be installed. The line in penetration 65 will have a welded pipe cap installed outside containment. The remaining portion of this line will have a welded pipe cap on the line end located within the CVAS boundary.

4. A solenoid valve will be installed upstream of CVR 401 A and LLRT test connections will be installed. The tubing and valves from containment up to and including the outer most containment isolation valve will meet ASME Section Ill. Class 2, seismic Category I criteria. The excess flow check valve CVR 402A will be removed from the penetration 53 line and replaced with LRT 402 manual valve.

5. A LLRT test will be performed on penetrations 53 and 65,

6. The power to the new series isolation valve will be attached and appropriate operability testing will be performed. The cap inside containment on the monitoring line in penetration 53 will be removed.

Upon completion of this plant modification an operability test of the CVAS boundary will be performed in accordance with the requirements in the tectinical specifications.

Failure Analysis Two automatic valves in series ensure that a single containment isolation barrier will be available following an accident with a single active failure.

Currently the CVR monitoring line from penetration 65 foods differential pressure transmitters CVR IDPT-5017B and CVR IDP 5071C (containment-annulus). The CVR monitoring instrument line from penetration 53 feeds differential pressure transmitters CVR-IDPT-5017A (containment-annulus) CAP-IDPT-5171 (containment-ambient). The l 8

t i

t

proposed change will re-route tubing within cabinet C-4 such that the above l instruments will be fed from penetration 53. CAP IDPT-5171 foods PMC point A51000 and measures the differential pressure between the containment and ambient i atmosphere. There is also an interlock which prevents the opening of Containment Atmospheric Purge isolation valves when differential pressure is greater than (more negative) -8.4 inches of water. CVR IDPT-5017A, B, and C provide inputs to the PMC (t/O points A51400, A51401, and A51402) that measures the differential pressure

• between the Containment and the Annulus. The instruments in cabinet C 4 are non-safety. Alternate instruments that measure containment annulus differenti.I pressure are available in case of a failure of penetration 53's non-essential instrument line. The

following is a discussion on the use of altemate instrumentation: ,

LocalIridications Available:

CVRIOPISS220A, B CVRIOPISS221A, B Range: 0 to 15 IN.

Lacedon: RB +21 C 3A /C 38 Computer Indications Available:

CAPIOPT5258A, B Containment

Atmosphen OP

., Range: 10 to 20 IN.

PION: A51001, A51002 i SBVIOPTS054A,8 Annulus: Atmosphen CW Range: +2 to 10 IN.

PIOC: A51702, A51703 ANPIOPT5075 Annuius Nepenve Pressum Renpe: 0 to 20 IN.

PIOC: A51000 Atmospheric Pressum @om Met. Tower)

P900c C40510 l

• The local indications can be read directly.

. The computer indications (Containment : Atmosphere D/P and Annulus : Ambient D/P)_can be used to determine Containment : Atmospheric D/P. This is accomplished as follows:

• CNTMT Annulus = (CNTMT- Atmosphere)-(Annulus - Ambient) i therefore:  ;

CNTMT Annulus = A51001 A51702 or CNTMT Annuius = A51002 - A51703 9

f I

• Another method of determining CNTMT : Annulus D/P is available.  !

- CNTMT - Annulus = (CNTMT Atmosohere) + (Atmosohere) . (Annulus)

CNTNT- Annulus = A51001 + C48516 (conneted to 9 . A51600 or CNTNT Annulus = A51002 + C48516(connitedto 9 A51600 Licensing Bases Upon approval and implementation the FSAR will be updated to reflect the proposed ,

configuration. Isolation valves will be added to the LLRT program and IST program as  !

appropriate.

The License Condition as it is currently documented in Attachment i to NPF-38 will no i longer represent the plant configuration. As .$escribed above, the CVR monitoring instrument line in penetration 65 will be capped and of the two lines, only penetration 53 will have valve position indication. Therefore, upon approval of the proposed change Attachment i to NPF.38 will be deleted.

Conclusion Waterford 3 seeks to correct previously dodoted infomW.lon that was in error. This information was material in nature and used as the bases for accepting the design configuration and testing requirements for these instrument lines by the NRC if the correct plant configuration had been known at the time, Waterford 3 believes that a l design change would have been pursued. The proposed change described herein is

ntended to resolve this riericc,>LUriing condition in the most expedient manner known. ,

The change pmposes an altemotive method for complying with the NRC recommended position in R.G.1.11, Waterford 3 believes that the level of protection

, provided by tha proposed chunge is commensurate with the safety function to be l performed.

Safety Analysis  ;

The proposed change described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas:

l' 1. , Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No 10

l 1

The proposed change will in not increase the probability of previously analyzed accidents. The proposed change seeks to clearly document the  !

design and licensMg bases for acceptance of the CVR sensing instrument lines.

The proposed change to the monitoring lines will provide greater assurance that ,

containment integrity will be maintained following a LOCA concurrent with a single active failure. The design change to the non-essential monitoring line will  ;

reduce the potential bypass leakage from penetrations 53 and 65 by adding a redundant automatic containment isolation valve on penetration 53 and isolating the non essential instrument line on penetration 65. This design change can be 4

performed at power without violating any !lconse/ regulatory requirements that ensure containment integrity is maintained.

3 There is no change in the function of the instrumentation. The only difference is that CVR-IDPT-50178 and C non safety differential transmitters that monitor the CVR system will be sensing containment pressure from penetration 53. If the non-essential line coming from penetration 53 becomes inoperable, containment to annulus differential pressure can be obtained from alternate instnJmontation.

The essential sensing line that actuates the CVR system to protect containment within design vacuum pressure is not affected by the design change.

Adding a redundant automatic containment isolation valve in penetration 53's non-essential instrument line instead of the excess flow check valve and  ;

isolating the non essential line in penetration 65's will significantly reduce the potential bypass leakage. The proposed change will credit the essential instrument lines as a closed system outside containment. T . appropriate testing and acceptance criteria will be applied to ensure that any leakage associated with these potential bypass leakage paths, will not exceed the limits used in the Waterford 3 safety analysis or result in a significant increase in analyzed dose consequences. Therefore, the proposed change will not involve significant increase in the probability or consequences of any accident .

previously evaluated.

2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different type of accident from any accident previously evaluated?

r Response: No.

The proposed change will credit the essential sensing lines outside containment as a closed system and will not affer* the plant or the manner in which the plant i in operated.

11

I The failure modes associated with containment isolation remain unchanged as a result of the design change to the non-essential monitoring lines. The function of the non-safety instrumentation is not affected. The only difference is that all of the non-safety inse montation will be sensing containment pressure from penetration 53. Ho /ever, if the non essentialline coming from penetration 53

• becomes inoperable, containment pressure can be obtained from altemate instrumentation. Adding a redundant automatic containment isolation valve in series with CVR 401 A in the non-essential instrument line ensures containment isolation following a LOCA with a concurrent a single active failure. Therefore, .

the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

l

3. Will operation of the facility in accordance with this proposed change  !

involve a significant reduction in a margin of safety? ,

Response: No The addition of a redundant automatic containment isolation volve in series with CVR 401 A in the non essential instrument line breaching penetration 53 onwres l

containment isolation postulating a single active failure on a Containment l Isolation Actuation Signal (CIAS). While this modification is performed, administrative controls will require containment integrity to be maintained by a seismic Category 1, ASME Section Ill, Class 2, passive containment isolation device.

The essential CVR instrument sensing lines form a seismically qualified, closed '

, system outside containment which is designed for pressure equal to or greater than containment. The instrument cabinets C-3A(B) are seismic Category I ano safety related. The instruments are Safety Class 1E and have a static pressure ra;ing of 1000 psig. These lines most the criteria of BTP CSB 6-3 for crediting a  ;

closed system as a leakage boundary to preclude bypass leakage by being designed, fabricated, erected, and tested to standards commensurate with the seroty function to be performed. The proposed change will apply the appropriate testing and acceptance criteria to ensure that any b9=;= associated with these potential bypass leakage paths, will not exceed the limits used in the Waterford ,

- 3 safety analysis or result in a significant increase in analyzed dose -

consequences. Therefore, the proposed change will not involve a significant '

reduction in a margin of safety.

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Safety and significant Hazards Determination Based on the above safety analysis, it is concluded that: (1) the proposed change does not constitute a significant hazards consideration as defined by 10CFR50.92; and (2) I there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as desenbod in the NRC final environmental statement. ,

13

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Reference List Paos 1 of 3

1. ANSI /ANS 51.71976 (ANSI N6581976), Sinole Failure Criteria for PWR Fluid Systems.

2. ANSI /ANS 56.21976 (ANSI N271 1976), Containment isolation Provisions for Fluid Systems.

3. NUREG 0787, July 9,1981, Safety Evaluation Rooort Related to the Ooeration of Waterford Steam Electric Station Unit No. 3 Docket No. 50 382

4. NUREG-0787, Sucolement No. 8 December 1984. Evaluation Rooort Related to the Operation of Waterford Steam Electric Station. Unit No. 3 Docket No. 50 382

5. NUREG-0800, Standard Review Plan. Section 3.2.2, " System Quality Group Classification",

i 6. NUREG 0800, Standard Review Plan Section 3.6.1, ar'lant Design For Protection Against Postulated Piping Failures in Fluid Systems Outsioe l Containment". .

7. NUREG 0800, Standard Review Plan. Section 3.6.2, " Determination Of Rupture Locations And Dynamic Effects Associated With The Postulated Rupture Of Piping".

l 8. NUREG-0800, Standard Review Plan. Section 3,6.4, "Conta!nment isolation System".

9. NUREG 0000, Standard Review Plan. Brand Technical Position CSB 6-3, " Determination of Bypass Leakage Paths in Dual Containment Plants".

l l 10. WATERFORD STEAM ELECTRIC STATION, UNIT 3 l

6-26. Dated December 18,1964.

l l 11. WATERFORD STEAM ELECTRIC STATION, UNIT 3 '

Facility Oooratina License NPF-38. Dated March 16,1965.

l 12. NUREG-1117, Technical Soecifications Waterford Steam Electric Station Unit No. 3 Docket No. GO-382. Section 3.6.1.1, " Containment Systems; Primary l

l Containment, Containment Integrity".

, , , . . , - - _ _ , . . . , , . . . , . , _ , , _ , . , -r,--y.-. , - m__ ...... . ..- y - , --

i Reference List Continued Page 2 of 3 f

13. NUREG 1117. Technical Soecifications Waterford Steam Electric Station Unit No. 3 Docket No. 50-382. Section 3.6.1.2, " Containment Systems; Containment Leakage",

t

14. NUREG 1117, Technical Soecifications Waterford Steam Electric Station Unit No. 3 Docket No. 50 382. Section 3.6.1.4, " Containment Systems; Internal Pressure".

15. NUREG 1117 Technical Soecifications Waterford Steam Electric Station Unit NoJ D.ggk8LNgJ0-202, Section 3.6.1.7, " Containment Systems; Containment Ventilation Systems".

4

16. Safety Guide 11. February 10,1971, Instrument Lines Penetratino Primary Reactor Containment.

17. Regulatory Guide 1.26, Revision 3 February 1976, Quality Group Classifications And Standards For Water. Steam. And Radioactive Wagte Containing Components Of Nuclear Power Plants.

18. Regulatory Guide 1.141 Revision 1, May 1980, Containment isolation Provisions For Fluid Systems.

19. DRAFT Regulatory Guide IC 126-5, March 1982, Instrument Sensino  :

UDea.

20. Waterford 3 SES Updated Final Safety Analysis Report Docket 50-382 daaratino Licanae NPF-38 . Section 1.8.1,11, " Regulatory Guide 1.11, instrument Lines Penetrating Primary Reactor Containment", Revision 0, March 1971, pg.1.8-3.

21. Waterford 3 SES Undated Final Safety Analysis Report Docket 50-382 Operappq Licanae NPF-38. Section 6.2.4.1.2, " Criteria For Isolation of Fluid Systems Penetrating The Containment", pg. 6.243.

22. Waterford 3 SES Updatqq Final Safety Analysis Recort Docket 50-382 Operating License NPF-38. Section 6.2.4.1.3, " Criteria For isolation of Fluid l Instrument Lines Penetrating The Containment", pg. 6.2 63.

t

, _ . _ ..- _ _.._. _ ._,_.,_... . . _ . - , . , , . . - . _ . . , . , _ . , _ = . . _ _ . . _ . _ , _ , _ . - . . _ . , _ . . . , . , _ _ _ . . , . . . . . . . _ . . . . . . . _ , , , . . , _ , _

Reference List Continued Page 3 of 3

~

23. Waterford 3 SES Updated Final Safety Analysis Rooort Docket 50 382 Oooratina License NPF-38. Section 6.2.4.2.2, " Instrument Lines", pg 6.2 64.

- 24. Waterford 3 SES Updated Final Safety Analysis Report Docket 50-382 Operatino License NPF-38. Section 7.1.2.7 "Comperison of Design with NRC Regulatory Guides", R.G.1.11 Instrument Lines Penetrating Primary Reactor -

Containment (3/10/97), pg. 7.1-8.

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, t NPF-36181 ATTACHMENT A Attachment A Provides the Waterford 3 Operating License NPF-38 That Will Be Affected by the Proposed Change.

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who ray acquire an interest unde.' this transaction (s) are i prchibited from exercitirg direst'y or inotractly any control cver (i) the facility, (ii) power or energy prnduce:

bythefacility,or(iii)thelicenseeofthefacility. l Further, any rights acquired under this authorization say

te exercised only in ecogliarce with ano subject to the reevirements and restrictions of this operating ifcerse, the Atomic Energy Act of 1994, as atended an regulations. For purposes of this cor.dttlon,d the the limitatics MC's of 10 CFR 50.81, as now in effect and as they may be subsequently amor.ded, are fully applicable to the equity investors and any successors in interest to the equity investors, as long as the license for the facility remains in effect.,

(b) prior LP&L, (or its desiynee) to any change to notify in (t ) the terms or conettions the NRCofinany writing' le agreements saecuted as part of the above authert:6d financial transactions (11) any facility operating asreement involving a I1censee that is in effect now or will be in effect in the future, er (iii) the existing property insurance coverages for the facility that would, set materiallyaltertherepresentationsandcondItions,he forth in the staff's Safety Evaluation enclosed to t NRC letter dated September 18,194g. In addition LP&L or its designee is required to notify the NRC of any, action by equity investers or successors in interest to LP&L that r.ay have an effect on the ope.ation of the facility.

C. This Itcense shall be deemed to contain and is subject to the conditions specified in the Casmission's regulations set forth in 10 CFR Chapter 1 i and is subject to all applicable provistens of the Act and to the rules, and regulattens is subject toand theorders of theconditions additional Commission new orer specified hereafter incorporated in effect; be low:

1. Maiam Peuer Level 801 is authorized to operate the facility at reactor core power leeels not in excess of 3390 megawatts therus) (1001 power) in ateerdases with the conditions specified herein and in Attachment 1 to this license. The items identified in Attachment 1 to this l license shall be completed as specified. Attachment 1 is hereby I

incorporated into this license.

2. Technica),,$p,egj,f,1,cLt,1, ens and,,fgj,re,n,gJnJ Prot,ec,tjo,n,,Pla,n The Technical Specifications contained in Appendix A, as revised through Amendment No. 58, and the Environmental Protection Plan centalped in Appendix B, are herehy incorporated in the license.

E01 shall operate the facility in ectordance with the Technical Specifications and the Environmental Protection Plan.

. _ _ _ _ i

9 H.

This license midnight is ef fective on December as the date of issuance and shall expire at 18, 2024 FOR THE NUCl, EAR REGULATORY CO*!$5!0N Harold R. Denton, Director Office of Nuclear Reactor Regulation

Enclosures:

1. Attachamt 1

2. Attachment ?

3. AppendixA(TechnicalSpecifications)(NURES.1117) 4 Appendix B L Environmental Protection Plan)

5. Appendix C ll Antitrust Conditions)

Oate of !ssuance: March 16,1985 s a

4 '

9 L_ _ _ . _ , _ -- - - - . -

-- - ^- ' "

(

r FARli ATTACHMENf 1 WATERFORD STEAM ELECTRIC $TATION OPERATING LICENSE NPF.38 This attachment identifies items which must be completed to the Comission's satisfaction prior to startup following the first refueling outage.

• Continuous, direct position indication in the control room for the containment isolation valves for instrument line penetrations 53 and

65. ,

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s ATTACHMENT B Attachment 8 indicates the requested changes to the Operating License NPF-38 l i

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who may acquire an interest under this transaction (s) are prohibited from exercising directly or indirectly any control over (i) the facility, (ii) power or energy produced i by the facility, or (iii) the licensee of the facility, rurther, any rights acquired under this authorization may be exercised only in compliance with and subject to the requirements and restrictions of this operating license, the Atomic Energy Act of 1954, as amended, and the NRC's regulations, for purposes of this condition, the limitations of 10 CFR 50.01, as now in effect and as they may be subsequently amended, are fully applicable to the equity investors and any successors in interest to the equity investors, as long as the license for the facility remains in effect.

(b) LP&L,(or its designee) to notify the NRC in writing prior to any change in (i) the terms or conditions of any lease agreements execut,ed as part of the above authorized financial transactions, (ii) any facility operating agreement involving a licenses that is in effect now or will be in effect in the future, or (iii) the existing property insurance coverages for the facility, that would materially alter the representations and conditions, set forth in the staff's Safety Evaluation enclosed to the NRC letter dated September 18, 1989. In tddition, LP&L or its designee is required to notify the NRC of any action by equity investors or successors in interest to LP&L that may have an effect on the operation of the facility.

C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set fotth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

1 Maximum Power Level EOI is authorized to operate the facility at reactor core power levels not in excess of 3390 megawatts thermal (100% power) in accordance with the conditions specified herein. :nd in Att:ch::nt &

t thi: lic:n::. The it::: id:ntifi;d in Att::Ex:nt i t: thi:

licen::: ch:11 b; :: ;1:0:d :: :p::ifi:d. Att::h::nt 1 i: h:::by in:: p:::t:d int; 2; li;;n::.

3. Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment 56 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

l E0I shall operate the facility in accordance with the Technical specifications and the Environmental Protection Plan.

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9

- H. This license is effective as the date of issuance and shall expire at midnight on' December 18, 2024.

FOR THE NUCLEAR REGULATORY COMMISSION H. R.-Denton /sd/

Harold R. Denton, Director office of Nuclear Reactor Regulation Enclosures

1. '

. tt:;hr.:r.t i This Attachment has been deleted. l I

2. Attachment 2 .

3. Appendix A (Technical Specifications) (NUREG-1117) 4.- Appendix B_(Environmental Protection Planh  :

S.. Appendix C.(Antitrust Conditions)

Date of Issuance:- March 16, 1985 6

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f NPF-38-181 ATTACHMENT C Visual Representation of the Proposed Plant Modification f

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REFERENCE 3 Waterford 3 LDCR 97-0225 dated June 9,1997

UCENSING DOCUMENT CHANGE REQUEST FORM Page 1 of 1-LICENSING DOCUMENT CHANGE REQUEST (LDCR) NUMBER: O - OM (To Be Assigned By Licensing)

DESCRIPTION OF PROPOSED CHANGE:

This LDCR revises FSAR Section 7.1.2.7 to identify NRC acceptance of the essential lines for penetrations 53 and 65 as meeting the criteria for a closed system. ,

UFSAR SECTIONS, TABLES, OR FIGURES REQUIRING CHANGES:

FSAR 7.1.2.7 ,_

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ATTACH ALL UFSAR PAGES APFECTED BY THE PROPOSED CHANGE. MARK UP THE UF8AR

! PAGES SO THAT THE PROPOSED CHANGE CAN BE CLEARLY UNDERSTOOD. IF THE PROPOSED CHANGE REQUM PORC REVIEW, THEN FORWARD A COPY OF THE PRE SCREENING OR SAFETY EVALUATION WITH AN LDCR FORM TO LICENSING AEIER PORC REVIEW. IT IS THE RESPON8488LITY OF THE PREPARhR TO ENSURE THAT UCENSING RECENES A COPY OF THE LDCR FORM.

Preparer: E.Lemke _. b T"T7 NAME (PRINT) Signature /Date 1

W2.302, Rev. 3 Attachment X Page 1 of 1 l

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i* WSES FSAR UNIT S -  ;

t R G.1.11 Instrument Lines Penetratina Primary Reactor Containment (3/10/71)

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• Containment isolation provisions are complied with by the following design *

' a) A manually operated valve rhall be installed as close to the containment as $56 N

- practicable and, 9Mt5 b) Essential system instrument lines shall have a A-self actuated excess flow check va;ve4hei64e installed as close to the containment as practicable downstream of the manually opeisted valve.

r--- "r 2:2 ;f;: 9 ;:r:r =  ::":t; :' :. 'M  :-":' ; :' : r."" 5:;: :

nrf': vf; M"' t !: t:; ::?:;: ;:n S: 2:2 ';f;: -N S: "r :2!: " ..".- M ---- e auseident, Uoon hiah oressure in containment. the excess flow check valve will close Since the excess flow check valys is onficed. the sensino line will caualize causino the excess flow

@eck valve to open and permit normal operation of the actuatina system. The instrument lines downstream of the excess flow check valves are desianed to Safety Class 3. Seismic Category

1. therefoes niest a closed system for instrument lines outside containment (See Licensina l Amendment 3GfG399i(8 der),

c) Non essential system instrument lines will be fit couinced 'with redundant solenoid valves located outside containment in lieu of manual valves. The solenoid valves will close on SIAS fegg ::: er" M r:;: .:: m: " ";'  : leeskentro6 station. This containment isolation '

provision was accroved Der Licensino Amendment 499996(later-), -

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d) The containment extreme w6de range pressure instrumentation for post accioent monitoring consists of a sealed liquid filled system with bellows, following the guidelines in ANS- '

56.2/N271 1976.

R.G.1.22 Periodic Testina of Protection Systems Actuation Function (2/17/72)

Testing of the RPS and ESFAS in compliance with Ragulatory Guide 1.22 is described in sections 7.2 and 7.3.'

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- R.G.1.29 Seismic Desian (6/7/12)

The instrumentation and control of safety-related systems and safety-related portions of systems comply with Regulatory Guide 1.29.

R.GJ1.30 Quaitr Amaurance Raouirements For the Installation. Inspection and Testina of Insammentation and Electric Eautoment (08/11/72) 4 References discussing comportson of the design with the recommendations of Regulatory Guide 1.30 is l provioed in Table 8.1-3.

i i-  : R.G.1.40 Qualification Test of Continuous Duty Motors installed Inside the Containment 4 l- of Water Cooled Nuclear Power Plants (3/10/73 A comparison of the design with the recommendations of Regulatory Guide 1.40 is provided in section 3.11.-

7.1-8 i-l

-. i

't Proposed change to FSAR Section 7.1.2.7:  ;

1

... manually operated valve. The CVR essential instrument line(s) is considered a closed system outside containment. These lines do not meet the requirements of Quality Group B standards defined in RG 1.26. However, the lines do meet the requirements defined in ISA-67.02. The NRC staff agreed (in License Amendment 128) that the criteria of ISA-67.02 are consistent with Quality .

Group C (i.e., ASME, Section Ill, Class 3) which is consistent with the staff's interpretation of the criteria goveming instrument lines.

Therefore, the staff concluded that the system is a closed j system. Upon high pressure ..."

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10CFR50.59 REVIEW PRE SCREENING FORM PAREI OF i Sinnatures Preparer: 4-98?7 NAME (PRINT) Signature /Date i Reviewer: DO lItr SYb '77Y+h 6- ?- 9 7 NAME (PRINT) S ate

/

/Ag3 % $~/d'fl Supervisor: hL l fdD/447 /M .

NAME (PFORT) ' Signature /Date /

DOCUMENT EVALUATED:

LDCR 97 0225 DESCRIPTION OF THE PROPOSED CHANGE:

This i DCR revises FSAR Section 7.1.2.7 in accordance with License Amendment 128. Among

, other things, this Amendment documents NRC acceptance of the essential lines for penetrations i 53 and 65 as being a closed system.

CHECK THE APPUCABLE BOXES SELOW. IP THE PROPOSED CHANGE Is COMPumR.Y mesn BY 00GE OR MORE OF TM POLLOWBee BONES, NErfhER A sCREEDWee REVEW NOR A sAPE"(Y EVALUAT10N Is MCEssARY. PROVioE SUPPORTING DOCUMWrfATION OR REPERENCEs As APPROPRIATE.

O The change is editorial or typographicel as dehned in 86cdon 4.2.1. Identify the applicable condition (s) by number:

O The change is controlled in its entiMy by 10CFR50.54 as denned in Section 6.E I O - A Screening covering all aspects of the change already exists per Section C.2.4. If taking credit for an existing screening, it must be attached.

O An SE covering all aspects of the change already exists per Section 6.2.4. If taking credit for an existing SE, it must either be attached, referenced by qumber, or the documentation associated with the first use of the SE clearly identined:

X The change, in its entirety, has been approved by the NRC. Reference NRC document:

License Amendment 128 O The change resolves conflicts between the SAR and actual plant design per Section 6.2.4.

O The chang sesquirse a change to the TS or Operating License per Section 6.2.7.

! O The charW involves the implementst'on of an approved Commitment Change Evaluation L Form (CCEP) por Section 4.2.8. Reference CCEF Tracking Number:

l ALL UPsAR CHANGES tsueT BE DOCUMBrTED ON AN LDCR FORM stME.AR TO ATTACHMENT X AND ATTACHED TO THIT. ,

PRE eCREmaNG. A COPY OP THE PRE eCRESENG AND LDCR FORM MUsT BE FORWARDED TO UCENelNG.

W2.302, Rev. 3 Attachment ll Page 1 of 1

BRANCH TECHNICAL POSITION CS8 6-3 DETERMINATION OF SYPASS LEAKAGE PATHS IN DUAL CONTADMENT PLANTS t

A. BACKGROUND The purpose of this branch position is to provide guidance-in the determination of that portion of the primary containment leakage that will not be collected and processed by the secondary containment. Bypass leakage is defined as that  ;

leakage from the primary containment which can circumvent the secondary contain-ment boundary and escape directly to the environment, i.e., bypasses the leakage collection and filtration systems of the secondary containment. This leakage component r.ust be considered in the radiological analysis of- a loss-of-coolant accident.

The secondary contairnment consists of a structure which completely encloses the primary containment and can be maintained at a pressure lower than atmospheric

so that primary containment leakage can be collected or processed before release to the environment. The secondary containment may include an enclosure building which forms an annular volume around the primary containment, the auxiliary building where it completely encloses the primary containment, and other regions of the plant that are provided with leakage collection and filtration systems.

Depressurization systems are provided as part of the secondary containment to decrease or maintain the secondary containment volume at a negative pressure.

All primary conteinment leakage may not be collected because (1) direct primary containment leakage can occur while the secondary containment is being depres-surized and (2) primary containment leakage can bypass the secondary containment through containment penetrations and seals which do not terminate in the second-ary containment.

Direct leakage. from the secondary containment to the environment can occur when-ever an outward positive differential pressure exists across the secondary con-tainment boundary. The secondary containment can experience a positive pressure transient following a postulated loss-of-coolant accident in the primary contain-ment as a result of thermal loading and infiltration from the environment and the primary containment that will occur until the depressurization systems become effective. An outward positive differential on the secondary containment wall can also be created by wind loads. In this regard, a " positive" pressure is defined as any pressure greater than -0.25 in, w.g. (water gauge), to account for wind loads and the uncertainty in the pressure measurements. Whenever the i pressure in the secondary containment volume exceeds -0.25 in, w.g., the leakage-prevention function of the secondary containment is assumed to be negated.

Since leakage from the secondary containment during positive-pressure periods cannot be determined, the conservative assumption is made that, all primary containment leanage is released directly to the environment during these time periods. Therefore, it b?comes necessary to determine the time periods during which these threshcid conditions exist.

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The existence and duration of periods of positive pressure within the secondary l containment should be based on analyses of the secondary containment pressure L response to postulated loss-of-coolant accidents within the primary containment and tne effectiveness of_the depressurization systems.

6.2.3-9 hv. 2 - July 1981 L __. .

e The evaluation of bypass leakage involves both the identification of bypass leakage paths and the determination of leakage rates. Potential bypass leakage paths are formed by penetrations which pass through both the primary and secondary containment boundaries. Penetrations that pass through both the primary and secondary containment may include a number of barriers to leakage (e.g., isolation valves, seals, gaskets, and welded joints). While each of these barriers aid in the reduction of leakage, they do nut necessarily eliminate leakage. Therefore, in identifying potential leakage paths, each of these penetrations should be considered, together with the capability to test them for leakage in a manner similar to the containment leakage tests required by Appendix J to 10 CFR Part 50.

B. BRANCH TECHNICAL POSITION

1. A secondary containment structure should completely enclose the primary containment structure, with the exception of those parts of the primary containment that are imbedded in the soil, such as the base mat of the containment structure. For partial dual containment concepts, leak rates less than the design leak rate of the primary containment should not be used in the calculation of the radiological consequences of a loss-of-coolant accident, unless the magnitude of unprocessed leakage can be adequately demonstrated. Quantitative credit for leakage col-lection in a partial-dual containment will be reviewed on a case-by-cai,e basis.

2. Direct leakage from the primary containment to the environment, equivalent to the design leak rate of the primary containment, should be assumed to occur following a postulated loss-of-coolant accident whenever the secondary containment volume is at a " positive" pressure; i.e., a pressure greater than -0.25 in, w.g. Positive pressure periods should be determined by a pressure response analysis of the secondary containment volume that includes thermal loads from the primary con-tainment and infiltration leakage.

3. The secondary containment depressurization and filtration systems should be designed in accordance with Regulatory Guide 1.52, " Design, i Testing, and Maintenance Criteria for Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water Cooled Nuclear Power

! Plants." Preoperational and periodic inservice inspection and test programs should be proposed for these systems and should include means for determining the secedary containment infiltration" rate, and the capability of the systems to draw down the secondary contain-i ment to the prescribed negative pressure in a prescribed time.

) 4. For secondary containments with design leakage rates greater than 100 i volume percent per day, an exfiltration analysis should be provided.

5. The following leakage barriers in paths which do not terminate within the secundary containment should be considered potential bypass leakage paths around the leakage collection and filtration systems of the secondary containment:

a. Isolation valves in piping which penetrates both the primary and secondary containment barriers.

6.2.3-10 Rev. 2 - July 1981

b. Seals ano gaskets on penetrations which pass througn both the primary and secondary containment barriers.

c, Welded , joints on penetrations (e.g. , guard pipes) which pass through both the primary and secondary containment barriers.

6. The total leakage rate for all potential bypass leakage paths, as ideritified in item 5 above, shonid be determined in a realistic manner, considering equipment design limitations and test sensitivities.

This value should be used in calculating the offsite radiological consequences of postulated loss-of-coolant accidents and in setting technical specification limits with margin for bypass leakage.

7. Provisions should be made to permit preoperational and periodic 'sakage rate testing in a manner similar to the Type B or C tests of Ap N . dix J to 10 CFR Part 50 for each bypass leakage path listed in item 5 above.

An acceptable alternative for local leakage rate testing for welded joints would be to conduct a soap bubble test of the welds concurrently with the integrated (Type A) leakage test of the primary containment required by Appendix J. Any detectable leakage determined in this manner would require repair of the joint.

8. If air or water sealing systems or leakage control systems are pro-posed to process or eliminate leakage through valves, these systenis should be designed, to the extent practical, using the guidelines for leakage control systems given in Regulatory Guide 1.96 (Ref. 4).

9. ' If a closed cystem is proposed as a leakage boundary to preclude bypass leakage, then the system should:

a. Either (1) not directly communicate with the containment atmosphere, of (2) not directly communicate with the environment, following a loss-of-coolant accident.

b) Be designed in accordance with Quality Group B standards, as defined by Regulatory Guide 1.26f (Systems designed to Quality Group C or D standards that qualify as closed systems to preclude bypass leakage will be considered on a case-by-case basis.)

c. Meet seismic Category I design requirements.

d. Be designed to at least the primary containment pressure and temperature design conditions.

e. Be designed for protection against pipe whip, missiles, and jet forces in a manner similar to that for engineered safety features.

f. Be tested for leakage, unless it can be shown that during normal plant operations the system integrity is maintained.

C. REFERENCES

1. 10 CFR Part 50, Appendix J, " Primary Reactor Cont:f r nent Leakage Testing for Water-cooled Power Reactors."

l 6.2.3-11 Rev. 2 - July 1981 l

o

2. Regulatory Guide 1.26, " Quality Group Classification and Str tards for Water , Steam , and Radioactive-Waste-Containing Compone of Nuclear Power Plants."

3. Regulatory Guide 1.52, " Design, Testing, and Maintenance Criteri for Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants."

4. Regulatory Guide 1.96, " Design of Main Steam Isolation Valve Leakage Control Systems for Boiling Water Reactor Nuclear Power Plants."

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6.2.3-12 Rev. 2 - July 1981 A

REFERENCE 4 Waterford 3 Technical Specification 3.4.1.3, Amendment 106.

REACTOR C00LANT SYSTEM HDT SHUIDOW LIMIIIELCQE11.10N T FOR OPERATION 3.4.1.3 At least two of the loop (s)/ train (s) listed below shall be OPERABLE and at least one reactor coolant and/or shutdown cooling loops shall be in operation.*

a. Reactor Coolant Loop ; and its associated steam generator and at least one associated reactor coolant pump,**

b. Reactor Coolant Loop 2 and its associated steam generator and at least one associated reactor coolant pump,**

c. Shutdown Cooling Train A,

d. Shutdown Cooling Train 8.

APPLICABILITY: MODE 4 AGilDH:

a. With less than the above required reactor coolant and/or shutdown cooling loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; if the remaining OPERABLE loop is a shutdown cooling loop, be in COLD SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. With no reactor coolant or shutdown roolino loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

l

• All reactor coolant pumps and shutdown cooling pumps (LPSI pumps) may be ,

deenergized for up to I hour provided (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and

' (2) core outlet temperature is maintained at least 10*F below saturation t

)

temperature. 1

• • A reactor coolant pump shall not be started with one or more of the Reactor  !

Coolant System cold leg temperatures less than or equal to 272*F unless l '

(1) the pressurizer water volume is less than 900 cubic feet or (2) the secondary water temperature of each steam generator is less than 100*F above each of the Reactor Coolant System cold leg temperatures.

3/4 4-3 Amendment No.106 .

WATERFORD - UNIT 3 L m )