ML20082E328

Proposed Tech Specs,Revising SR 4.0.5 Per Recommendations of Draft NUREG-1482 & NUREG-1431,std TS

ML20082E328
Person / Time
Site:	Sequoyah
Issue date:	04/06/1995
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20082E326	List: ML20082E325 ML20082E328
References
RTR-NUREG-1431, RTR-NUREG-1482 NUDOCS 9504110194
Download: ML20082E328 (17)

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- i s  :. :. ENCLOSURE l

PROPOSED TECHNICAL ~ SPECIFICATION CHANGE, SEQUOYAH NUCLEAR-PLANT UNITS'1 AND~2

' DOCKET NOS. 50-327 AND~50-328' (TVA-SQN-TS-94-18)'

.z-LIST OF AFFECTED.PAGES ,

Unit i VI  ;

3/4 0-2

.3/4 0-3 3/4 4-27 B 3/4 0-5 B 3/4 4-14 Unit 2 i

VI 3/4'0-2 3/4 0-3 .;

3/4 4-32 B 3/4 0-5 B 3/4 4-14 t

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INDEX E

-LIMITING' CONDITIONS FOR OP'ERATION AND SURVEILLANCE REQUIREMENTS

'SECTION~ PAGE=

'3/4.4 -REACTOR COOLANT < SYSTEM-3/4.4.1' REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation............................... 3/4'4-1 Hot Standby................................................ 3/4 4-2 Hot Shutdown..............................................- 3/4 4-3 Cold Shutdown............................................. 3/4'4-5

• 3/4.4.2 SAFETY VALVES - SHUTD0WN..................................- 3/4 4-6 3/4.4.3 SAFETY AND RELIEF VALVES - OPERATING Safety Valves Operating................................... 3/4 4-7 Relief Valves Operating.................................. 3/4.4-8 3/4.4.4 PRESSURIZER........................... 7c................. 3/4 4 -

3/4.4.5 STEAM GENERATORS................ 3/4 4-10 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE- -

Leakage Detection Systems.................................. 3/4~4-17.'

Operational Leakage....................................... 3/4 4 4

i 3/4.4.7 CHEMISTRY................................................. 3/4 4-21 3/4.4.8 SPECIFIC ACTIVITY......................................... 3/4 4-24 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Re a cto r Cool a nt Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.4-28  ;

Pressurizer............................................... 3/4_4-31 R1381, 3/4.4.10 - STRUCTU"A:. I'"EORITY T>Et_eg g C C Cve C1.. 1, 2 end 3 Ce ,esents..................... 3/4 4-32 R138 .

3/45.11 REACTOR COOLANT' SYSTEM HEAD VENTS........................ 3/4 4-33 4, 1

3/4.4.12 OVERPRESSURE PROTECTION SYSTEMS........................... 3/4-4-34' - '

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SEQUOYAH - UNIT 2 VI Amendment No. 106,120,'138 147 N' ]

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W.?jPPLICABILITY:

SURVEILLANCE REQUIREMENTS

• i .4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual Limiting conditions for O unless otherwise stated in an individual Surveillance Requirement. peration

$ 4.0.2 Each Surveillance Requirement shall be performed within the specified surveillance cent of the specified interval with a maximum surveillance allowable extension not to exceed 25 per-interval. R152 4.0.3 Failure to perform a Surveillance Requirement within the: allowed surveillance interval,~ defined by Specification 4.0.2 shall constitute.

noncompliance with the OPERABILITY requirements for a, Limiting Condition for Operation.

The time limits of the ACTION requirements are applicable at the-time it is identified that a Surveillance Requirement has not been performed.

The ACTION requirements may be delayed for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to permit the completion of the surveillance when the allowable outage time limits of the R69 ACTION requirements are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Surveillance Requirements do not-

.have to be performed on inoperable equipment. o 4.0.4 Entry into an OPERATIONAL MODE.or other specified condition shall not bemadeunless-theSurveillanceRequirements)associatedwiththeLimiting.

Condition for Operation have been performed (within the specified surveillanc interval or as otherwise s)ecified. This provision shall not prevent passage through or to OPERATIONAL A0 DES as required to comply with ACTION requirements. R69'

_4.R5ISury Coce Clas llance Requirements for inserv'ce inspect"on and test

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, 2 ant'3 componen shall be pplicable s follows:

>a Inser ce inspect

• n of ASME ode Class 2 and 3 onents- d-ins vice testi of ASME C e Class 1 2,and 3 p s and va es s l be'perfo d in ace dance wit action XI f the AS Boiler d Pressure essel Code nd applic le Addenda s requir by

' 10 CFR-50 action 50, a(g) exc t where sp ific wri en relief  :

Commissi pursuant has been Section , antedi).by)t

.55a(g)(6 10 CFR ,

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6 SEQUOYAH - UNIT 2 3/4 0-2 Amendment No. 69,152 August 13, 1992

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,. APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued) 4.0.5 (Continued)

Survei ance interval specified 6 Section X of the ASME ler and P essure Vessel ode and ap icable Adde da for the in vice ins .ction and te ing activi es required y the ASME Bo er and Pr saure Vessel ode and app 1 cable Adde a shall be ap icable as ollows in the Technical pecificatio s:

A E Boiler and ressure Ves el quired freque les for ode and appli ble Addend erforming ins vice terminology r inservic inspection a testing inspection nd testing ctivities activities eekly At least ce per 7 days Monthly At leas once per 31 day Quart ly or every months At le t once per 92 d s Semi nnually or ery 6 month At 1 st once per 18 days.

Every 9 m ths At east once per 2 days Yearly or nnually least once per 66 days.

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c. Th provisions 9 Specification .0.2 are appli ble to the ab e ,

r quired frequ hcies for perf ming inservice 'nspection and esting ctivities.

Perfo.rman e of the above nservice inspe ion and testi activities shall b in addition to other specified urveillance R quirements.

e. Not ' g in the ASME oiler and Press e Vessel Cod shall be co trued to upersede the r quirements of a Technical Sp cification.

Ls~ %- 1 3/4 0-3 Amendment No. 69 R69 SEQUOYAH - UNIT 2 Augu.n 15, 1088

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.4.0.5 ' . Surveillance Requirements for inservice . inspection an'd i

testing .of ASME ~ Code Class - 1, .2, 'and 3- components shall- be. 'l

-applicable'.as follows:- l l

l Inservice Inspection Program This program provides controls for inservice inspection of ASME Code Class 1, 2, and 3 components, including applicable supports. 3 i

-The program shall: include the followings

a. Provisions that Inservice inspection of ASME Code Class 1, 2, and 3 components shall be performed in accordance-with t Section XI of the ASME Boiler and Pressure-Vessel Code and

.\ appitcable Addenda as required by 10 CFR 50.55a

b. The provisions of SR 0.2 are applicable to the frequencies ,

for perfoming inservice inspection activities; ,

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c. Inspection of each reactor coolant pump flywheel per the recoronendations of Regulation Position c.4.b of Regulatory Guide 1.14. Revision.1, August 1975; and

d. Nothing in the ASHE Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any T$.

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. inservice Testing Program

This program provides'-controls for inservice ' testing of ASME Code  :

class 1, 2, and 3 components including applicable supports. The program shall includ he following:  ;

a. Provisions hatinservicetestingofASHECodeCiass1,2, and 3 pumps alvesg! onuoum shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a;

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.b. Testing frequencies specified in Sectiori XI'of the 'ASME l Boller and Pressure vessel Code and applicable Addenda as - '

follows:

l ASME Boiler and Pressure ,

vessel Code and applicable Addenda terminology for- Required Frequencies inservice testing for performing inservice  ;

testino activities activities Weekly At least once per- 7 days Monthly At least once per 31. days Quarterly or every.

3 months At least once per 92 days Semiannually or-every G rnonths At least once per 184 days .

Every 9 months 'At least once per 276 days Yearly or annually At least once per 366 days:

Biennially or overy 2 years ' ' At least once per 731 days

c. The provisions of SR J.D.2 are applicable to the above required Frequencies for performing inservice testing activitiest -

d. TheprovisionsofSR/.0.3areapplicabletoinservice I testing activities; and i e. Nothing in the ASME Boiler ami Pressure Vessel Code shall be construed to supersede the requirements of any TS.

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REACT 0R'Cf0LANT SYSTEM

- 3/4.4.1 STRUCT L INTEGRITY

-ASME CODE C 1. 2 and 3 C NENTS-

~LIMITI DITION FOR TION /. ,

1 3.4 0 The struct al integrity of E Code Class , 2 and 3 components- all b maintained in ccordance with S cification 4.4 0.

APPLICABILITY 4 All MODES ACTION:

6 With the st ctural integrit of any ASME Code C1 s I component (s) not confo ing to the abov requirements, rest the structural

, /. integri of the affecte component (s) to wi n its limit or i late the 3 ected componen ) prior to increas g the Reactor Co nt '

Sy tem temperature e than 50*F above e minimum temper ure quired by NDT c siderations.

b With the stru ural integrity of ASME Code Class component (s) not conform g to the above req ements, restore e structural integrity f the affected com nent(s) to withi ts limit or is .te the aff ted component (s) p or to increasing e Reactor Cool Syste temperature above 0*F.

c. W the structural egrity of any A Code Class 3 onent(s) ot conforming to e above require s' restore the tructural integrity of the ffected componen s) to within i limit or isolate the affected c ponent(s) from s vi ce.- ,

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d. The provis ns of Specifica on 3.0.4 are n applicable.

SURVEILLANCE UIREMENTS l R168 4.4.10 addition to thef uirements of pecification 4.0 each Reactor Coolan ump flywheel s il be inspect 4er the recomme ions of Regula ry Posi on C.4.b of Regy)ptory Guide 1.)p, Revision st 1975. 1, Au,

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SEQUOYAH - UNIT 2 3/4 4-32 Amendment No. 138, 168 March 15, 1994 4

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BASES'-

4.0.4 This specification establishes the requirement that all applicable surveillances must be met before entry into an OPERATIONAL MODE or other co dition of operation specified in the Applicability statement. The purpose of this specification is to ensure that system and component OPERABILITY requirements or parameter limits are met before entry into a MODE or condition for which these systems and components ensure safe operation of the facility.

This provision applies to changes in OPERATIONAL MODES or other specified conditions associated with plant shutdown as well as startup. R69 under the provisions of this specification, the applicable Surveillance Requirements must be performed within the specified surveillance interval to ensure that the Limiting Conditions for Operation are met during initial plant startup or following a plant outage.

When a shutdown is required to comply with ACTION requirements, the provisions of Specification 4.0.4 do not apply because this would delay placing the facility in a lower MODE of operation.

4.0.5 This specification ensures that inservice inspection of ASME Code Class 1, 2 and 3 components and inservice testing of ASME Code Class 1, 2 and 3 pumps and valves will-be performed in accordance with a periodically updated denda version of Section as_ required XI of 50.55a) by 10 CFR the A5M4.Bo{ler m and Press, p Vessel

.........,......m requir a:n.., hc F bB prodes-in-wR4nd-by-the-Gc=ission and i: not-a part of-the:c Technica]ll

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j This specification includes a clarification of the frequencies for performing the inservice inspection and testing activities required by Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda. This clarification is provided to ensure consistency in surveillance intervals throughout these Technical Specifications and to remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing activities.

Under the terms of this specification, the more restrictive requirements-of the Technical Specifications take precedence over the ASME Boiler and Pressure Vessel Code and applicable Addenda. For example,' the requirements of Specification 4.0.4 to perform surveillance activities prior to entry into an OPERATIONAL MODE or other specified applicability condition takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows pumps to be tested up to one week after return to normal operation. And for example, the Technical Specification definition of OPERABLE does not grant a grace period before a device that is not capable of performing its specified function is declared inoperable and takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows a valve to be incapable of performing its specified function for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before being declared inoperable.

SEQUOYAH - UNIT 2 B 3/4 0-5 Amendment No. 69 August 15, 1988 e -- _ _ __

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. REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve is constructed based on a point-by point .o comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration.

The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over-the course of the heatup ramp the controlling condition switches from the inside -

to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

The leak test limit curve shown in Figure 3.4-2 represents the minimum ,

temperature requirements at the leak test pressure specified by applicable '

codes. The leak test limit curve was determined by methods of Branch Technical '

Position MTE8 5-2 and 10 CFR 50, Appendix G.

The criticality limit curve shown in Figure 3.4-2 specifies pressure-temperature limits for core operation to provide additional margin during a148 ,

actual power production. The pressure-temperature limits for core operation (except for low power physics tests) require the reactor vessel to be at a -

temperature equal to or higher than the minimum temperature required for the in-service hydrostatic test, and at least 40 degrees F higher than the minimum ,

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pressure-temperature curve for heatup and cooldown. The maximum temperature for the in-service hydrostatic test for the SQN Unit 2 reactor vessel is 274 degrees F. A vertical line at 274 degrees F on the pressure-temperature ,

curve, intersecting a curve 40 degrees F higher than the pressure-temperature i limit curve, constitutes the limit for core operation for the reactor vessel.

Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the. respective curves.  ;

Although the pressurizer operates in temperature ranges above those for which there is reason for concern of non-ductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

g7g 3/4.4.10 STRUCTURAL INTEGRITY The inse ice inspect n and testing ograms for A Code Class 1, 2 and 3 compo nts ensure at the structu 1 integrity a operational rea ness of these nonts wi be maintained t an acceptahl level throughout he ,

life of he plant. ese programs a in accordancy ith Section XI o the ASME oiler and P sure Vessel Co and applicabi Addenda as requi d by ,

10 R Part 50.5 (g) except who specific writ n relief has bee ranted by e Commission ursuant to 10 R Part 50.55a ) (6) (1).

Compo ts of the reac r coolant syst were designed pr r to the issuance Section XI of he ASME Boiler nd Pressure Vesse Code. These

compone s will be tes to the extent rectical within t limitations of ...,

l the o ginal plant de gn, geometry a materials of con ruction.

f SEQUOYAH - UNIT 2 B 3/4 4-14 Amendment No.148 ,

March 31, 1992 l 1

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= ENCLOSURE 2 PROPOSED. TECHNICAL SPECIFICATION (TS) CHANGE' SEQUOYAH. NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328-(TVA-SQN-TS-94-18)'

DESCRIPTION AND JUSTIFICATION FOR REVISION TO SURVEILLANCE REQUIREMENT 4.0.5 AND THE RELOCATION OF TS 3/4.4.10.

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} TVA proposes'to modify the Sequoyah Nuclear Plant (SQN) Units'1;and 2:

technical. specifications (TSs) to incorporate the recommendations of

, . draft'NUREG-1482 and to. reflect the intent of'the standard TSs (STS)

(NUREG-1431)-for the' inservice inspection (ISI) and inservice test -(IST);

. programs. Specifically, the proposed' changes revise Surveillance Requirement-(SR) 4.0.5 and TS 3/4.4.10 as described below:.

a.- Remove SQN's current SR'4.0.5 language and replace it with the STS- l

-(NUREG-1431) language-for ISI and IST programs. >

b. Remove TS 3/4.4.10, Structural Integrity, and associated bases for '

relocation under'the newly proposed SR 4.0.5..

Renann for change Draft'NUREG-1482 discusses the concern that. situations could arise, which -

would put s' licensee in a condition that is not in strict compliance with TS SR 4.0.5, Land'could lead to a plant shutdown to perform testing.- This yl concern is contained in SQN's current SR 4.0.5, which prohibits the plant from implementing relief from impractical American Society of Mechanical Engineers (ASME) code requirements before obtaining approval from.NRC. An-example of a condition where literal compliance with SR 4.0.5 would result in a plant shutdown la described below:-

The plant is operating at full power and it is identified that a-  !

noncompliance condition exits (i.e., a safety-related valve is installed on a system required by TSs and is not included for periodic testing in- t

' the IST program). If testing the valve while at power or during cold-shutdown is impractical, relief from the'ASME code would be required to  !

perform testing during refueling' outage conditions. Under the current-SR 4.0.5, written relief must be' granted by NRC before the nonconformance ,

condition can be' cleared and the valve considered to be operable. Under .'

these circumstances, NRC approval may not be obtained within the TS {

allowed outage time'for the applicable system. Accordingly, the TS limiting condition'for operation-(LCO) would not be met and a forced shutdown solely to comply with code test requirements would be .

I necessary. This condition would not be in the best interest of safety-because it could unnecessarily challenge plant systems. To rectify the >

labove condition, NRC recommends in draf t NUREG-1482, " Guidelines for Inservice Testing at Nuclear Power Plants," that licensees > incorporate the STSs for IST programs.

In conjunction with TVA's proposed change to SR 4.0.5, relocation of TS 3/4.4.10, Structural Integrity, is needed to eliminate redundant structural integrity requirements within TSs. TS:3/4.4.10 provides requirements for ASME Code Class 1, 2,'and 3. components to ensure that ,

structural integrity of these components is maintained at all times. At SQN, and other pressurized water reactor Westinghouse Electric-Corporation plants, structural integrity of ASME Class 1, 2, and 3  ;

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Q Q,7 components- is controlled arul maintained by Specification 4.0.5 under the ,

ISI program. . Since.' structural integrity.is required by the ASME code as a  ;

programmatic ISI. requirement, TVA proposes to relocate SQN's current --

structural integrity requirements under SQN.TS SR 4.0.5.

JilElif.i.CAliQn for ChaDER I

'IVA's proposed license amendment revises SQN SR 4.0.5 in accordance with - J the recommendations of draft NUREG-1482 and incorporates the programmatic requirements'of the STS (NUREG-1431) for the-ISI and IST programs. Unlike the STS, TVA elected to incorporate the STS language under SQN's-SR 4.0.5 rather than delete SR 4.0.5 and relocate the ISI and IST program. '

requirements under-the administrative controls section of TSs. Retaining SR 4.0.5 in its current location is preferred by TVA since it precludes revising approximately 15-20 other references to SR 4.0.5 that are located throughout the SQN TSs.

Draft NUREG-1482, entitled " Guidelines for Inservice Testing at Nuclear .

Power Plants," specifically addresses the situation in which TSs maybe in conflict with the regulations of 10 CFR 50.55a. As discussed in the draft NUREG-1482, situations may arise, which would put the licensee in a condition that is not in strict comp *1ance with the TS 4.0.5 requirement to

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comply with ASME Section XI "except where specific written relief has been 1 granted by the Commision pursuant to 10 CFR 50, Section 50.55a(g)(6)(1)."  !

According to the draft NUREG, if TS 4.0.5 was interpreted literally, in the case of the IST program, it would require the licensee to address these situations by shutting down the plant solely to perform inservice testing.

In addition, although the draft NUREG addresses this concern-for IST [

programs, it is TVA's understanding that the NRC staff considers this issue

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i to be generic to both the ISI and IST programs.

As stated in the draf t NUREG-1482, NRC recommends that licensees revise the .

TSs to include the recommendations from the revised standard TSs (NUREG-1431) for the ISI and IST programs. With the revised STS, upon ,

finding an ASME code requirement impractical because of prohibitive dose j rates or limitations in the design construction, or system configuration, l the licensee can implement the relief request at that time provided-it:has' ' '

been (1) acceptably reviewed pursuant to 10 CFR 50.59; (2) approved by the  ;

plant staff in accordance with the administrative process described in the- i ISI and IST programs administrative procedures; and (3) reviewed and l approved by the Plant Nuclear Safety Committee. Based on the {

recommendations of draft NUREG-1482, TVA is incorporating the STS language l into SR 4.0.5 and revising the associated bases to require that ISI and IST programs be implemented in accordance with 10 CFR 50.55a.

The following provides a description and justification for each proposed j change associated with SR 4.0.5:  ;

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a. The newly proposed SR 4.0.5 divides the IS into the ISI program l requirements and the IST program requirements. This change is editorial in nature and is consistent with the STS and 10 CFR 50.55a.

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b.;[Inspectionrequirementsassociateditith'thereactoricoolantpump

' flywheel, which are currently partiof.SQN SR 4.4.10, have_been relocated as a programmatic requirement under'the newly proposed SR 4.0.5.. This change in TS formattisEconsistent1with.STS and;is

, ' justified becauseithe flywhee1' inspection. requirements remain .  ;

tunchanged. ,

c. The newly proposed SR 4.0.5. incorporates a,TS' format change only and  ;

defines'a new frequency (i.e., " Biennially or.every 2 years" - at- ,

least once per 731 days) that is not' currently defined in SQN's:

current SR 4.0.5. This change is considered to be a.TS improvement that provides consistency between TS' frequencies and-the~ASME Section.

XI code frequencies.

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d. TheLSTS IST program requirements incorporate the' provisions of SR 4.0.3 that address failure to perform a SR within the allowed.

surveillance interval. This requirement-is not contained.in SQN's current SR 4.0.5. 'This change is considered to be.a TS improvement that provides guidance within the TS for failure to perform inservice ,

tests within the allowed surveillance interval.

e. The STS IST program requirements include ASME Code Class 1, 2, and 3 pumps, valves, and snubbers. TVA's proposed change does not include snubbers as part of SR 4.0.5. TVA has elected to retain snubber test and inspection requirements in SQN's current' TS 3/4.7.9, " Snubbers."

f. TVA's proposed change to SR 4.0.5 includes a change to the associated l basen. The current TS 4.0.5 bases contains a sentence.that states:

" Relief from any of the.above requirements has.been provided in writing by the Commission and is not a part of these Technical .i Specifications." Since this language would imply that any relief request from ASME code (10 CFR 50.55a) requirements have aiready been provided in writing by the Commission, the issue of literal ^

compliance with TS, as discussed in NUREG-1482 would remain in SQN TSs. Accordingly, TVA is removing this language from the bases.

In conjunction with the proposed changes to SR 4.0.5 described above. TVA proposes to relocate SQN.TS-3/4.4.10,'" Structural Integrity," and the associated bases. This proposed change is based on the Commission's final policy statement for relocation of current TS that do not meet any of the screening criteria for retention. During the development of the'

- STSs the existing structural integrity TS was evaluated and screened against the interim policy statement criteria. Based on this evaluation, it was determined that the structural integrity requirements could be relocated to another controlled document. For SQN, the structural i integrity requirements of TS 3/4.4.10 are maintained under SR'4.0.5 and l controlled by SQN's ISI program documentation. Accordingly, TVA's i proposed change relocates the structural integrity requirements under the provisions of SR 4.0.5 l

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., , The proposed change request does not involve an unreviewed environmental

. question because operation of SQN Units l'and 2.in.accordance with this change would not ,

1. ' Result in 'a significant increase in' any adverse environmental impact previously evaluated-in the Final Environmental Statement (FES) as q l

modified by the staff's testimony to the Atomic Safety.and Licensing Board, supplements to the FES, environmental impact appraisals, or decisions of the Atomic Safety and Licensing Board.

2. ' Result in a significant change in effluents or power levels.

3. Result in matters not previously reviewed in the licensing basis-for SQN that may have a significant environmental impact.

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Enclosure 3 ]

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' PROPOSED TECHNICAL SPECIFICATION CHANGE F

SEQUOYAH NUCLEAR PLANT UNITS 1'AND 2.

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DOCKET NOS. 50-327 - AND 50-328

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DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION,

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- Significant. Hazards EvaluationL

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TVAlhas evaluated the proposed. technical specification (TS) change and-has determined that it does;not: represent.a significant hazards-consideration based'on criteria established in 10 CFR 50.92(c).

Operation;of Sequoyah' Nuclear-Plant-(SQN).2in accordance with the proposed.

Y - amenoment will'not:-

P  : 1.;sInvolve a significant increase.in the probability or consequences of an accident:previously evaluated.

1 Operation 'of the : facility in accordance with the proposed amendment

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would not. involve any increase in the probability of occurrence or' consequences of an accident previously evaluated.. The Inservice

Inspection and Testing Programs, pursuant to 10 CFR'50.55a are-described in the TSs...The proposed amendment, in accordance with NUREG-1431 and' draft NUREG-1482 permits relief-from an American-Society of Mechanical Engineers (ASME) code requirement in'the-interim between the time of submittal of a relief request and NRC approval of the relief. The changes being proposed do'not affect' assumptions contained in plant safety analyses or change the physical.

design and/or operation of the plant, nor do they' affect TSs that~

preserve safety. analysis assumptions. Any relief from the approved ASME Section XI code requirements.that'is implemented prior to NRC review.and approval will require evaluation under the 10 CFR 50.59

process to determine that no TS changes or unreviewed safety.

questions exist. This evaluation process will ensure that the impact of any code relief is thoroughly evaluated and that the structures, systems, and components remain in conformance with assumptions made.

in the safety analysis. The proposed change to delete SQN TS-3/4.4.10, Structural Integrity,'does not affect: plant safety analyses or. change the physical design or operation of the plant.

The proposed amendment relocates the structural integrity requirements under the existing TS Survelliance Requirement (SR) 4.0.5 to allow these requirements to be governed and controlled within the inservice inspection _(ISI) program. . Therefore.. operation of the facility"in accordance with.the proposed amendment would not.

affect the probability or. consequences of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any previously analyzed.

. Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different. kind of accident from any accident previously evaluated. The Inservice Inspection and Testing Programs, pursuant to 10 CFR 50.55a are described in the TSs. The proposed amendment, in accordance with NUREG-1433 and draft NUREG-1482, permits relief from an ASME code-requirement in the' interim between the time of-submittal of a relief

~.

request and NRC approval of the relief. The changes being proposed will not change the physical plant or the modes of operation defined in the Facility License. The changes do not involve the addition or f vt t *w $ew, av as-+--

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modification of' equipment.nor do they alter the design or operation of plant systems.- Any. relief from the approved ASME Section XI code requirements that is implemented prior to NRC review and' approval will require evaluation 'under the 10 CFR 50.59 process - to determine that-no TS changes or unreviewed safety questions exist. This evaluation process will ensure that the impact of the code relief.is thoroughly evaluated and that the structures, systems,' and components-remain in conformance with assumptions made in the safety analysis.

The proposed change to delete SQN TS 3/4.4.10 does not affect plant safety analyses or change the physical design or operation of the plant.- The proposed amendment relocates the structural integrity requirements under the existing TS SR 4.0.5 to allow these requirements to be governed and controlled within the-ISI program.

Therefore, operation of the facility in accordance with the proposed ,

amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Involve a significant reduction in a margin of safety.

Operation of.the facility in accordan'ce with the proposed amendment would not involve any reduction in a margin of safety. The Inservice Inspection and Testing Programs, pursuant.to 10 CFR 50.55a, are required by the SQN TSs. The-proposed amendment, in accordance with NUREG-1431 and draft NUREG-1482 permits' relief from an'ASME code requirement ~ in the interim between the time of submittal of. a relief request'and NRC approval of the relief.. Any relief from the ASME Section XI code is required to be evaluated under the 10 CFR 50.59 '

process to determine that no TS changes or unreviewed safety questions exist. This evaluation process will ensure that code' relief does not affect the ability of structures, systems, or components to perform their design function, affect compliance with any TS requirements or reduce the margin of safety. The proposed-change to delete SQN TS 3/4.4.10 does not affect plant safety analyses or change the physical' design or operation of the plant.

The proposed amendment relocates the structural integrity-requirements under the existing TS SR 4.0.5 to allow these requirements to be governed and controlled within SQN's ISI program.

Therefore, operation of the facility in accordance with the-proposed amendment would not involve a reduction in the margin of safety.

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