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Proposed Tech Specs Removing Table 3.4-1, RCS Pressure Isolation Valves. Justification & NSHC Determination Encl
	ML20155C774
Person / Time
Site:	Sequoyah
Issue date:	04/10/1986
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
Shared Package
ML20155C754	List: ML20155C751; ML20155C774;
References
	TVA-SQN-TS-68, NUDOCS 8604170115
	Download: ML20155C774 (13)
	v • d • e

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o ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION CHANCES SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 (TVA SQN TS 68)

PROPOSED CHANCES TO DELETE TABLE 3.4-1 FROM THE SEQUOYAH TECHNICAL SPECIFICATIONS LIST OF EFFECTIVE PAGES UNIT 1 3/4 4-14 3/4 4-15 3/4 4-15a UNIT 2 3/4 4-18 3/4 4-19 3/4 4-20 fp A P P

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a REACTOR COOLANT SYSTEM -

i OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE, c;

1 GPM total primary-to secondary leakage through all steam generators and 'S '

gallons per day through any one steam generator, d.

10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and e.

40 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 + 20 psig. '

f.

1 GPM leakage at a Reactor Coolant System pressure of 2235 1 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in T_ti; 2. ' 1. appreffkfe gent instrweb%n), .

l APPLICABILITY: MODES 1, 2, 3 and 4 ACTION: -

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD _ SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following c.

With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the af fected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Ccolant System leakages shall be demonstrated to be within each of the above limits by: 1 I

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SEQUOYAH - UNIT I 3/4 4-14 1

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f REACTOR COOLANT SYSTEM

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SURVEILLANCE REQUIREMENTS (Continued) i a.

Monitoring the lower containment atmosphere particulate radioactivity monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

Monitoring the containment pocket sump inventory and discharge at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c.

Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2235 + 20 psig at least once per 31 days with the modulating valve fully open. The provisions of Specification 4.0.4 are not applicable for entry into Mode 3 or 4.

4 d.

Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

e.

Monitoring the reactor head flange leakoff system at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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i 4.4.G.2.2 Each Reactor Coolant System Pr9ssure Isolation Valveby L.4O ~3ll specified h T b h 3.? I shall b.e demonstrated OPERABLE pursuant to Spscification 4.0.5, except that in lieu of any leakage testing requirements required by

, Specification 4.0.5, each valve shall be demonstrated-OPERABLE by verifying leakage to be within its limit:

a. At least once per 18 months.

b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more and if leakage testing has not been performed in the previous 9 months.

c.

Prior to returning the valve to service following maintenance, repair or replacement work on the valve.

d.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve.

The or 4.provisions of Specification 4.0.4 are not applicable foi entry into MODE 3

i SEQUOYAH - UNIT 1 3/4 4-15

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TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER FUNCTION 6 560 Accumulator Discharge 63- 61 Accumulator Discharge 63-5 63-56 Accumulator Discharg 63-622 Accumulator Dischar

/ 63-623 Accumulator Disch ge 63-624 Accumulator Dis arge i Accumulator Of harge

/ 63-625 Accumulator scharge

/ /- 63-551 63-553 Safety Inje ion (Cold Leg)

Safety In ction (Cold Leg)63-557 63-555 Safety I jection (Cold Leg)

/ 63-632 Safety njection (Cold Leg)63-633 - Resi al Heat Removal (Cold Leg)63-634 Res' ual Heat Removal (Cold Leg)

R idual Heat Removal (Cold Leg)63-635 esidual Heat Removal (Cold Leg)63-641 Residual Heat Removal / Safety

-j 63-644 Injection (Hot Leg)

Residual Heat Removal / Safety 63-558 Injection (Hot Leg)63-559 Safety Injection (Hot Leg)63-543 -

Safety Injection (Hot Leg)

Safety Injection (Hot Leg)63-545 63-547 fety Injection (Hot Leg)- - 63-549 Sa ety Injection (Hot Leg)63-640 Saf y Injection (Hot Leg)63-643 Resi al Heat Removal (Hot Leg)87-558 Residu 1 Heat Removal (Hot Leg)

Upper H d Injection 87-559 Upper Hea Injection 87-560 Upper Head njection - 87-561 Upper Head 87-562

'ection Upper Head In ction 87-56 Upper Head Inj tion FCV- -1*

Residual Heat Re val FC 74-2* Residual Heat Rem al

^These v '1ves do not have to be leak tested following manual or aut atic actua ion or flow through the valve.

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! REACTOR COOLANT SYSTEM j ,

OPERATIONAL LEAKAGE _

LIMITING CONDITION FOR OPERATION 3.4.5.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. 1 GPM total primary-to-secondary leakage through all steam generators and 500 gallons per day through any one steam generator,

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. 40 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of -

2235 ! 20 psig..

f. 1 GPM leakage at a Reactor Coolant System pressure of 2235 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in TuM e 3.1-1. l affrcftide ft:.nt instr 4chens.

APPLICABILITY: MODES 1, 2, 3 and 4

}

_ ACTION: -

With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY a.

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . .-

b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . -

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REOUIREMENTS 4.4.5.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

SEQUOYAH - UNIT 2 3/4 4-18

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j i REACTOR COOLANT SYSTEM 1

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<! SURVEILLANCE REQUIREMENTS (Continued) -

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a. Monitoring the lower containment atmosphere particulate radioactivity monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

b. Monitoring the containment pocket sump inventory and discharge at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pumo seals when the Reactor Coolant System pressure is 2235 2 20 psig at least once per 31 days with the modulating valve fully open. The provisions of Specification 4.0.4 are not applicable for entry into Mode 3 or 4.

d. Performance of a. Reactor Coolant System water inventory balance at

'least once per /2 hours.

e. Monitoring the reactor head flange leakoff system.at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . m by L.C.O. M 6, des 4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified-4e-tam e 3.? I shall be demonstrated OPERABLE pursuant to Specification 4.0.5, except that in lieu of any leakage testing requirements required by Specifica-tion 4.0.5, each valve shall be demonstrated OPERABLE by verifying leakage to

- be within its limit: _

a. At least once per 18 months. .

b. Prior to entering MODE 2 whenever the plant har been in COLD SHUT 00'aN for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more and if leakage testing has not been performed in the previous 9 months.

c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve.

d. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

SEQUOYAH - UNIT 2 3/4 4-19

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3 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER FUNCTION 63-560 Accumulator Dischar 3-561 Accumulator Discn ge 6 -562 Accumulator Disc arge 63 63 Accumulator 01 harge 63- 2 Accumulator D'scharge 63-62 ,

Accumulator ischarge 63-624 Accumulato Discharge 63-625 Accumula r Discharge 63-551 Safety njection (Cold Leg)63-553 Safet Injection (Cold Leg)63-557 Safe y Injection (Cold Leg)

J 63-555 Sa ty Injection (Cold Leg)

N 63-632 R sidual Heat Removal (Cold Leg)63-633 esidual Heat Removal (Cold Leg)

'fy 63-634 Residual Heat Removal (Cold Leg)63-635 Residual Heat Removal (Cold Leg)

N 63-641 Residual Heat Removal / Safety Injection (Hot Leg)63-644 Residual Heat Removal / Safety Injection (Hot Leg) ;63-558 Safety Injection (Hot Leg) ,63-559 Safety Injection (Hotleg) ' 63-543 - - Safety Injection (Hot Leg)63-545 Safety Injection ~(Hot Leg)63-547 Safety Injection (Hot Leg)63-549 Safety-Injection (Hot Leg)63-640 Residual Heat Removal (Hot Leg)63-643 sidual Heat Removal (Hot Leg)87-558 U er Head Injection 87-559 Up r Head Injection 87-560 Uppe Head Injection 87-561 Upper ead Injection

87-56 Upper ad Injection 87-5' Upper He d Injection FC' 74-1" Residual at Removal F /-74-2* Residual H t Removal j

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4 ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION CHANGES SEQUOYAH NUCLEAR PLANT' UNITS 1 AND 2 (TVA SQN TS 68)

JUSTIFICATION FOR DELETION OF TABLE 3,4-1 FROM SEQUOYAH TECHNICAL SPECIFICATIONS I

Description of Change This proposed change would delete Table 3.4-1 of LCO 3.4.6.2, listing pressure isolation valves, from the technical specifications for both units 1 and 2. This table would be placed in the appropriate plant instructions to ensure proper control.

Reason for Change NRC pointed out in the referenced safety evaluation report (SER) the need for inclusion of valves FCV-87-7 and FCV-87-8 to both units 1 and 2 technical specifications. This SER was a written response to TVA submittals describing Sequoyah's Inservice Test Program (IST). These valves are located in a line connecting the liquid waste disposal system to the upper head injection system (UHI) . The purpose of this line is leakage testing of the UHI check valves. The valves in Table 3.4-1 are required to be operable in order to prevent leakage f rom the reactor coolant system (RCS) into a lower pressure system. The inoperability of these valves creates the potential of an intersystem loss of coolant accident (LOCA) .

Justification for Change By the referenced SER, TVA became awa re of the need to add the two UHI valves (FCV-87-7 and FCV-87-8) to the list of pressure isolation valves.

In order to add these valves to this list as currently configured, a change to technical specifications is required. Sequoyah has recently !

received a change to this table for unit I technical specifications (Amendment 39 to license No. DPk-77, June 20, 1985).

j

r TVA believes that removing this list f rom Technical Specifications, and making it a part of plant-controlled instructions, will remove an unnecessary licensing burden from both TVA and NRC.

10CFR50.59 states that a change to a licensed facility or procedures described in the safety analysis may be made without prior Commission approval. However, the change must be evaluated by the criteria of 10CFR50.59 (a)(2)(i),(ii), and (iii) to constitute neither an unreviewed safety question nor a change in technical specifications. Moving the list into plant procedures would allow use of the 10CFR50.59 process in order to make timely changes to the plant. Yet this process would still require an analysis determining the effects of said change upon safe plant opera tion. Changes to this procedure (other than editorial or clarifying changes) made prior to Commission approval using the criteria of 10CFR50.59(a)(2) will be reported to the Commission at least once annually.

This proposed change will allow changes to the facility and procedures (i.e., adding or deleting valves to or from the list) tha t do no t have an adverse effect on safety without prior Commission approval. Changes which do not satisfactorily meet the criteria of the USQD will require prior Commission approval similar to that received for changes to technical specifications.

TVA believes that this change will fully meet the requirements of safety while reducing the administrative burden of future changes of technical specifications to both TVA and NRC.

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The list will be included (as currently configured in Technical Specifications) in the appropriate plant instructions. The two valves referenced previous 13 (FCV-87-7 and FCV-87-8) will be added to this list. These valves are currently being tested by TVA-Sequoyah, as part of our ASME Section XI testing program. With approval of this change.

their test interval will be brought into concurrence with the remaining technical specification valves.

A 10CFR50.59 evaluation will be performed addressing the addition of FCV-87-7 and FCV-87-8 to this list. This evaluation will be performed l as part of the revision to the appropriate plant instruction adding the l list of valves.

References T. M. Novak's letter to H. G. Pa rris da ted April 5, 1985, " Safety l Evalua tion Report on Sequoyah Inservice Test Program for Pumps and Valves (IST)" (A02 850415 008)

ENCLOSURE 3 TVA SQN TS 68 4

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS FOR DELETION OF TABLE 3.4-1 FROM THE SEQUOYAH TECHNICAL SDECIFICATIONS i

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SIGNIFICANT HAZARDS CONSIDERATIONS 1.

Is the probability of an occurrence or the consequences of an accident previously evaluated in the safety analysis report significantly increased?

No--This proposed change is primarily a change to the method of reporting amendments to the list of pressure isolation valves. This proposed change to technical specifications does not reflect a change to plant design, configuration, or testing requirements. Any changes to this list will be subjected to an unreviewed safety question determination (USQD) per the criterion of 10CFR50.59. Changes potentially adverse to plant safety will be determined as unreviewed safety questions (USQ) and

, require Commission review and approval prior to implementation. Changes sa tisfactorily meeting the criterion of the USQD will be made and reported to the Commission in an annual report.

2. , Is the possibility for an accident of a new or different type than evaluated previously in the safety analysis report created?

, No--This proposed change does not create a possibility for accidents not previously evaluated in the safety analysis because it does not change the current plant configuration of existing euqipment. The possibility of adding or deleting equipment from this list causing a different type of accident would be identified by performance of a USQD for each specific change.

3 Is the margin of safety significantly reduced?

No--The margin of saf ety remains unaffected solely due to this proposed change since no equipment parameter or test described in the Technical Specifications or their bases are added or deleted by this cha nge . Specific changes to this equipment list will be subject to USQD as they are implemented.

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ML20155C774

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