ML18087A273

From kanterella
Jump to navigation Jump to search
Issuance of Technical Specification Amendments for the Sequoyah Nuclear Plant, Units 1 and 2 (TAC Nos. M91217 and M91218) (TS 94-06)
ML18087A273
Person / Time
Site: Sequoyah  Tennessee Valley Authority icon.png
Issue date: 08/02/1995
From: Labarge D
Division of Operating Reactor Licensing
To: Kingsley O
Tennessee Valley Authority
Hon A
References
TAC M91217, TAC M91218
Download: ML18087A273 (24)
v • d • e


Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WAS HI NGTO N. D. C. 20555--0001 August 2, 1995 Mr . Oliver D. Kingsley, Jr.

President, TVA Nuclear and Chief Nuclear Officer Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801

SUBJECT:

ISSUANCE OF TECHNICAL SPECIFICATION AMENDMENTS FOR THE SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 (TAC NOS. M91217 AND M91218) (TS 94-06)

Dear Mr . Kingsley :

The Convnission has issued the enclosed Amendment No. 206 to Facility Operating License No . DPR- 77 and Amendment No . 196 to Facility Operating License No . DPR-79 for the Sequoyah Nuclear Plant , Units 1 and 2, respectively. These amendments are in response to your application dated December 16, 1994, and additional information that was supplied by letter dated July 19, 1995.

The amendments replace the present Auxiliary Feedwater System Specification 3/4 .7.1.2 with new spec i fication s that are modeled after the Westinghouse Standard Technical Specifications .

A copy of the Safety Evaluation is also enclosed . Notice of Issuance will be included in the Convnission's biweekly Federal Register notice.

S~rely,

~id~~Sr.

Project Directorate II- 3 Project Manager Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Docket Nos. 50-327 and 50-328  :

Enclosures : 1. Amendment No . 206 to ...

License No. DPR-77

2. Amendment No. 196 to Licen se No. DPR-79
3. Safety Evaluation cc w/enclosures : See next page

Mr: Oliver D. Kingsley, Jr.

Tennessee Valley Authority SEQUOYAH NUCLEAR PLANT cc:

Mr. O. J. Zeringue, Sr. Vice Pre Nuclear Operations sident TVA Representative Tennessee Valley Authority Tennessee Valley Authority 38 Lookout Place 11921 Rockville Pike 1101 Market St ree t Suite 402 Chattanooga, TN 37402-2801 Rockville, MD 20852 Dr. Mark 0. Medford, Vice Presid Regional Administrator Engineering & Technical Services ent U.S. Nuclear Regulatory Commissi on Tennessee Valley Authority Region II 38 Lookout Place 101 Marietta St ree t, 1101 Market St ree t Atlanta, GA 30323 NW., Suite 2900 Chattanooga, TN 37402-2801 Mr. William E. Holland Mr. D. E. Nunn, Vice President Senior Resident Inspector New Plant Completion Sequoyah Nuclear Plant Tennessee Valley Authority U.S. Nuclear Regulatory Commissi 38 Lookout Place 2600 Igou Ferry Road on 1101 Market St ree t Soddy Daisy, TN 37379 Chattanooga, TN 37402-2801 Mr. Michael H. Mobley, Mr. R. J. Adney, Si te Division of RadiologicalDiHe rector Sequoyah Nuclear Plant Vice President 3rd Floor, L and C Annex alth Tennessee Valley Authority 401 Church St ree t P.O. Box 2000 Nashville, TN 37243-1532

~ Soddy Daisy, TN 37379

~ .

County Judge General Counsel Hamilton County Courthouse Tennessee Valley Authority Chattanooga, TN 37402-2801 ET llH 400 West Summit Hill Knoxville, TN 37902 Drive Mr. P. P. Carier, Manager Corporate Licensing Tennessee Valley Authority 4G Blue Ridge 1101 Market St ree t Chattanooga, TN 37402-2801 Mr. Ralph H. Shell

~site Licensing Manag er  :

Sequoyah Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Sodd~ Daisy, TN 373 79

UNITED STATES NUCLEAR REGULATORY COMMISSIO N WASHINGTON , D.C . 20555--0001 TENNESSEE VALLEY AUTHORITY DOCKET NO . 50-327 SEQUOYAH NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 206 License No. DPR-77

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

A. The application for amendment by Tennessee Valley Authority (the licensee) dated December 16 , 1994, which was supplemented by letter dated July 19, 1995 , compl ies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I;

8. The facility will operate in conformity with the application , the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable ass urance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regu lations ;

D. The issuance of thi s amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-77 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No. 206, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance, to be implemented within 45 days.

FOR THE NUCLEAR REGULATORY COMMISSION F~?~ec;a'r~

Project Directorate 11-3 ~

Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: August 2, 1995 r-

\

ATTACHMENT TO LICENSE AMENDMENT NO. 206 FACILITY OPERATING LICENSE NO. DPR-77 DOCKET NO. 50-327 Revise the Appendix A Technical Specifications by removing the pages identifie d below and inserting the enclosed pages. The revised pages are identifie d by the captioned amendment number and contain marginal lines indicating the area of change.

REMOVE INSERT 3/4 7-5 3/4 7-5 3/4 7-6 3/4 7-6 83/4 7-2 83/4 7-2 83/4 7-2a 83/4 7-2a 83/4 7-2b

PLANT SYSJ~;.\;.

AUXILIARY FEEUWAJER <AFWl SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1. 2 Three auxiliary feedwater trains shall be OPERABLE.*

APPLICAf.IUT1: MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat **~:mn\7~ l .

ACTION:

a. With one AFW train inoper&ble in f:'iOD~ ls 2 or 3, restore the inoperable AFW train to OPERABLE status within 72 STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWNhours or be in HOT within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b. With two AFW trains inoperable in MODE 1, 2 or 3, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
c. With three AFW trains inoperable in MODE 1, 2 or 3, innediately initia te corrective action to restore at least one AFW train to OPERABLE status.**
d. With the required AFW train inoperable in MODE 4, inrnediately initia te action to restore the required AFW train to OPERABLE status .

SURVEILLANCE REQUIREMENTS 4.7.1. 2.1 At least once per 31 days, verify each AFW manual, power operated, and automatic valve in each water flow path, and in both steam supply flow paths to the steam turbine driven pump, that is not locked otherwise secured in position, is in the correct position. , sealed, or

  • Only one ARI train, which includes a motor driven pump, is required to be OPERABLE in MODE 4.
    • LCO 3.0.3 and all other LCO ACTIONS requiring MODE changes ar*e suspended until one AFW train is restored to OPERABLE status .

r SEQUOYAH UNIT 1 3/4 7-5 Amendment No. 12, 115, 206

PLANT SYSTEMS

~ SURVEILLANCE REQUIREMENTS Ccontinuedl 4.7.1.2.2 At least once per 92 days, verify the developed head of each AFW pump at the *flow *test point is greater than or equal to the required developed head.*

4.7.1.2.3 Once every 19 months, verify each AFW automatic valve that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.**

4.7.1.2.4 Once every 18 months, verify each AFW pump starts automatically on an actual or simulated actuation signal.***

  • Not required to be completed for the turbine driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after steam supply pressure is greater than or equal to 842 psig.
    • Not applicable in MOOE 4 when steam generators are relied upon for heat removal.
      • Not required to be completed for the turbine driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after steam supply pressure is greater than or equal to 842 psig. Not applicable in MOOE 4 when steam generator(s) are relied upon for heat removal.

SEQUOYAH UNIT 1 3/4 7-6 Amendment No. 12, 77, 114, 206

PL~NT SYSTEMS BASES Q - Nomin2l NSSS power rating of the plant (including reactor coolant pump heat), Mwt K

  • Conversion factor, 947.82 <Bt~fecl w*
  • on any one steam generator at the highest MSSV opening pressure including tolerance and accumulation, as appropriate, in lb/sec.

For example, if the maximum number of inoperable MSSVs on any one steam generator is one, then w should be a su11111ation of the capacity of the operable MSSVs at the highest operable MSSV r.operating pressure, excluding the highest capacity MSSV. If the maximum number of inoperable MSSVs per steam generator is three then w should be a su11111ation of the capacity of the operable MSS~s at the highest operable MSSV operating pressure, excluding the three highest capacity MSSVs.

  • heat of vaporization for steam at the highest MSSV opening pressure including tolerance and accumulation, as appropriate, Btu/lbm N - Number of loops in plant The values calculated from this algorithm must then be adjusted lower to account for instrument and channel uncertainties.

3/4.7.1 .2 AUXILIARY FEEDWATER SYSTEM The AFW System is configured into three trains . The AFW considered OPERABLE when the components and flow paths requireSystem d to is provide redundant AFW flow to the steam generators are OPERABLE. This requires that the two motor-driven AFW pumps be OPERABLE in two diverse paths, each supplying AFW to separate steam generators. The turbine-driven AFW pump is require d to be OPERABLE with redundant steam supplies from each of two main steam lines upstream of the MSIV's, and shall be capable of supplying AFW generator. The piping, valves, instrumentation, and controls into the any steam required flow paths also are required to be OPERABLE.

The AFW System mitigates the consequences of any event with loss of normal feedwater.

The design basis of the AFW System is to supply water to the steam generator to remove decay heat and other residual heat by delive ring at least the minimu~ required flow rdte to the steam generators at a pressu re corresponding to 1085 psig. This pressure is in excess of the maximum expected steam generator pressure with the existing safety valve setpoints.

In addition, the AFW System must supply enough makeup steam generator secondary inventory lost as the unit cools water to replace r to MODE conditions. Sufficient AFW flow must also be available to account for*flo losses such as pump recirculation and line breaks.

SEQUOYAH - UNIT 1 4

w B 3/4 7-2 Amendment No. 115, 155, 196, 206

PLANT SYSTEMS

~ BASES

~ .

The limiting Design Basis Accidents (DBAs) and transients for the AFW System are as follows:

a. Feedwater Line Break (FWLB); and
b. Loss of main feedwater (MFW).

In addition, the minimum available AFW flow and system charac teristic s are credited for removing decay heat in the analysis of a small break loss of coolant accident (LOCA).

The AFW System design is such that it can perform its function following a FWLB between the MFW isolation valves and containment, combined with a loss of offsite power following turbine trip, and a single active failure of the steam turbine-driven AFW pump (above 50% power) or one motor-driven AFW pump (below 501 power with steam generator low level reactor trip time delay). For SOI power operation and higher, one motor-driven AFW pump is assumed to deliver to the broken MFW header at the pump run-out flow. Sufficient flow would be delivered to the intact steam generator by the redundant motor-driven AFW pump.

For partial power operation (below 501 power with trip time delay active),

one motor-driven AFW pump is assumed to fail. All flow from AFW pump and the redundant motor-driven AFW pump is assumed tothedelive turbine-driven r to the broken MFW header until the faulted steam generator is isolated by operato action 10 minutes after the break. After isolation of the faulted steam r r generator, suffici ent flow is delivered to the intact stea* generator by the turbine-driven and redundant motor-driven AFW pump.

The Engineered Safety Feature Actuation System (ESFAS) automatically actuates the AFW turbine-driven pump and associated valves and contro ls when required to ensure an adequate feedwater supply to the steam generators during loss of power.

The surveillance requirements (SRs) provide a means of ensuring the AFW system components are capable of supplying required flow to the steam generators, the flow path is aligned correctly, and the automatic functions actuate as designed. The automatic functions are verified through either an actual or simulated actuation signal. The actuation signal associated with SR 4.7.1.2.3 (automatic valve actuation) include the AFW actuation test signal and the low AFW pump suction pressure test signal. The actuation signal associated with SR 4.7.1.2 .4 (automatic pump start) includes only:the AFW actuation test signal.

Each motor-driven auxiliary feedwater pump (one Train and one Train B) supplies flow paths to two steam generators. Each flow path Acontain matic air-operated level control valve (LCV). The LCVs have the sames train an auto-designation as the associated pump and are provided trained air. The turbine-driven auxiliary feedwater pump supplies flow paths to all four steam genera-tors. Each of these flow paths contains an automatic.air-operated LCV, two of SEQUOYAH - UNIT 1 B 3/4 7-2a Amendment No. 115, 155, 196, 206

PLANT SYSTEMS

~ ~BA~S~ES~---------------------------------------------------

wh 1ch are designated as Train A, receive A-train air, and provide flow to the same steam generators that are supplied by the B-train 11e>tor-driven auxiliary feedwater pump. The remaining two LCVs are designated as Train 8, receive 8-trai n air, and provide flow to the same steam generators that are supplied by the A-train 11ator-driven pump. This design provides the required redundancy to ensure that at least two steam generators receive the necessary flow assU11ing any single failure. It can be seen from the description provided above that the loss of a single train of air (A or 8) will not prevent the auxiliary feed-water system from performing its intended safety function and is no more severe than the loss of a single auxiliary feedwater pump. Therefore, the loss of a single train of auxiliary air only affects the capability of a single motor-driven auxiliary feedwater pump because the turbine-driven pump is still capable of providing flow to two steam generators that are separate from the other motor-driven pump.

Two redundant steam sources are required to be operable to ensure that at least one source is available for the steam-driven auxiliary feedwater (AFW) pump operation following a feedwater or main stellD line break. This require-ment ensures that the plant remains within its design basis (i.e., AFW to two intact steam generators) given the event of a loss of the No. 1 steam generator because of a main steam line or feedwater line break and a single failure of the 8-train motor driven AFW pump. The two redundant sources must be aligned such that No. 1 steam generator source is open and operable and the No. 4 stea*

generator source is closed and operable.

~

~,,,

For instances where one train of emergency raw cooling water (ERCW) is declared inoperable in accordance with technical specificatio ns, the AFW turbine-driven pump is considered operable since it is supplied by both trains of ERCW. Similarly, the AFW turbine-driven pump is considered operable when one train of the AFW loss of power start function is declared inoperable in accordance with technical specifications because both 6.9 kilovolt shutdown board logic trains supply this function. This position is consistent with American National Standards Institute/ANS 58.9 requirements (i.e., postulation of the failure of the opposite train is not required while relying on the TS limiting condition for operation).

3/4.7.1.3 CONDENSATE STOBAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures thal sufficient water is available to maintain the RCS at HOT STANDBY conditions for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with steam discharge to:the atmosphere concur-rent with total loss of off-site power. The contained water volume limit includes an allowance for water not useable because of tank discharge line location or other physical characteris tics.

r SEQUOYAH - UNIT 1 B 3/4 7-2b Amendment No. 115, 155, 206f

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2.()55.5-4001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SEQUOYAH NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 196 License No. DPR-79 I. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Tennessee Valley Authority (the licensee) dated December 16, 1994, which was supplemented by letter dated July 19, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the corrunon defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Conunission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-79 is hereby amended to read as follows:

(2) Technical Soecifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No. 196, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance, to be implemented within 45 days.

FOR THE NUCLEAR REGULATORY COMMISSION Fr~~Y~rr-Project Directorate 11-3

(

Attachment:

Changes to the Technical Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation Specifications Date of Issuance: Au~st 2, 1995 r

r ATTACHMENT TO LICENSE AMENDMENT NO. 196 FACILITY OPERATING LICENSE NO. DPR-79 DOCKET NO. 50-328 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

REMOVE INSERT 3/4 7-5 3/4 7-5 3/4 7-6 3/4 7-6 83/4 7-2 83/4 7-2 83/4 7-2a 83/4 7-2a 83/4 7-2b

PLANT SYSTEMS r

\

AUXILIARY FEEPWAJER SYSTEM LIMITING CONQITION FOR OPERATION 3.7.1.2 Three auxiliary feedwater trains shall be OPERABLE.*

APPLICABILITY: MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat removal.

ACTION:

a. With one AFW train inoperable in MODE 1, 2 or 3, restore the inoperable AFW train to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b. With two AFW trains inoperable in MODE 1, 2 or 3, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
c. With three AFW trains inoperable in MODE 1, 2 or 3, innediately initiate corrective action to restore at least one AFW train to OPERABLE status.**
d. With the required AFW train inoperable in MODE 4, innediately initiate action to restore the required AFW train to OPERABLE status.

SURVEILLANCE REQUIREMENTS 4.7.1.2.1 At least once per 31 days, verify each AFW manual, power operated, and automatic valve in each water flow path, and in both steam supply paths to the steam turbine driven pump, that is not locked, sealed, or flow otherwise secured in position, is in the correct position.

  • Only one ARI train, which includes a motor driven pump, is required to be

. OPERABLE in MODE 4.

      • LCO 3.0.3 and all other LCO ACTIONS requiring MOOE changes are suspended until one AFW train is restored to OPERABLE status. :

~

~.

SEQUOYAH - UNIT 2 3/4 7-5 Amendment No. 2, 105, 196

PLANT SYSTEMS

~ SURV~ILL!NCE REPYlBEM~~ lcontlnuedl 4.7.1.2. 2 At least once per 92 days, verify the developed head of each AAI pump at the flow test point is greater than or equal to the required developed head.*

4.7.1.2. 3 Once every 18 months, verify each ARI automatic valve that is not locked, sealed, or otherwise secured in position.actuates to the correct position on an actual or simulated actuation signal.**

4.7.1.2 .4 Once every 18 months, verify each AFV pump starts aut01Rat1cally on an actual or simulated actuation signal.***

  • Not required to be completed for the turbine driven AFV pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after steam supply pressure is greater than or equal to 842 psig.
    • Not applicable in MODE 4 when steam generators are relied upon for heat removal.
      • Not required to be completed for the turbine driven AFV pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after steam supply pressure is greater than or equal to 842 psig. Not applicable in MODE 4 when steam generator(s) are relied upon for heat removal.

SEQUOYAH - UNIT 2 3/4 7-6 Amendment No. 68, 104, 196

PLANT SYSTEMS r

BASES Q

  • Nominal NSSS power rating of the plant (including reactor coolant pump heat), Mwt K - Conversion factor, 947.82 (Bt~ecl w* = Minimum total steam flow rate capability of the operable MSSVs on any one steam generator at the highest MSSV opening pressure including tolerance and accumulation, as appropriate, in lb/sec.

For example, if the maximum number of inoperable MSSVs on any one steam generator is one, then wt should be a sunmation of the capacity of the operable MSSVs at he highest operable MSSV operating pressure, excluding the highest capacity MSSV. If the maximum number of inoperable MSSVs per steam generator is three then w. should be a surrmation of the capacity of the operable MSSVs at the highest operable MSSV operating pressure, excluding the three highest capacity MSSVs.

heat of vaporization for steam at the highest MSSV opening pressure including tolerance and accumulation, as appropriate, Btu/lbm N

  • Number of loops in plant r The values calculated from this algorithm must then be adjusted lower to account for instrument and channel uncertainties.

3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The AFW System is configured into three trains. The AFW System is considered OPERABLE when the components and flow paths required to provide redundant AFW flow to the steam generators are OPERABLE. This requires that the two motor-driven AFW pumps be OPERABLE in two diverse paths, each supplying AFW to separate steam generators. The turbine-driven AFW pump is required to be OPERABLE with redundant steam supplies from each of two main steam lines upstream of the MSIV's, and shall be capable of supplying AFW to any steam generator. The piping, valves, instrumentation, and controls in the required flow paths also are required to be OPERABLE.

The AFW System mitigates the consequences of any event with loss of normal feedwater.

The des1gn basis of the AFW System is to supply water to the steam generator to*remove decay heat and other residual heat by delivering at least the minimu~ required flow rate to the steam generators at a pressure corresponding to 1085 psig. This pressure is in excess of the maximum expected steam generator pressure with the existing safety valve setpoints.

In addition, the AFW System must supply enough makeup water to replace steam generator secondary inventory lost as the unit cools to MODE 4 conditions. Sufficient AFW flow must also be available to account for flow losses such as pump recirculation and line breaks.

SEQUOYAH - UNIT 2 B 3/4 7-2 Amendment No. 105, iai 196

. PLANT SYSTEMS r~ BASES The 11m1t1ng Design Basis Accidents (DBAs) and transients for the AFW System are as follows:

a. Feedwater Line Break (FWLB); and
b. Loss of main feedwater (MFW).

In addition, the minimum available AFW flow and system characteristics are credited for removing decay heat in the analysis of a small break loss of coolant accident (LOCA).

The AFW System design is such that it can perform its function following a FWLB between the MFW isolation valves and containment, combined with a loss of offsite power following turbine trip, and a single active failure of the steam turbine-driven AFW pump (above SO% power) or one motor-driven AFW pump (below 50% power with steam generator low level reactor trip time delay). For SOI power operation and higher, one motor-driven AFW pump is assumed to deliver to the broken MFW header at the pump run-out flow. Sufficient flow would be delivered to the intact steam generator by the redundant motor-driven ARI pump.

For partial power operation (below SOI power with trip time delay active),

one motor-driven AFW pump is assumed to fail. All flow from the turbine-driven AFW pump and the redundant motor-driven AFW pump is assumed to deliver to the broken MFW header until the faulted steam generator is isolated by operator r '

action 10 minutes after the break. After isolation of the faulted steam generator, sufficien t flow is delivered to the intact steam generator by the turbine-driven and redundant motor-driven AFW pump.

The Engineered Safety Feature Actuation System (ESFAS) automatically actuates the AFW turbine-driven pump and associated valves and controls when required to ensure an adequate feedwater supply to the steam generators during loss of power.

The surveillance requirements (SRs) provide a means of ensuring the AFW system components are capable of supplying required flow to the steam generators, the flow path is aligned correctly, and the automatic functions actuate as designed. The automatic functions are verified through either an actual or simulated actuation signal. The actuation signal associated with SR 4.7.1.2.3 (automatic valve actuation) include the AFW actuation test signal and the low AFW pump suction pressure test signal. The actuation signal associated with SR 4.7.1.2.4 (automatic pump start) includes only.-the AFW actuation test signal.

Each motor-driven auxiliary feedwater pump (one Train A and one Train 8) supplies flow paths to two steam generators. Each flow path contains an auto-matic air-operated level control valve (LCV). The LCVs have the same train designation as the associated pump and are provided trained air. The turbine-driven auxiliary feedwater pump supplies flow paths to all four steam genera-tors. Each of these flow paths contains an automatic air-operated LCV, two of which are designated as Train A, receive A-train air, and provide flow to the same steam generators that are supplied by the 8-train motor-driven auxiliary feedwater pump. The remaining two LCVs are designated as Train B, receive 8-SEQUOYAH - UNIT 2 B 3/4 7-2a Amendment No. 196

PLANT SYSTEMS

~* ~BA~S~E.s.._..._..._. ________________________________ ________________

\

train air, and provide flow to the same steam generators that are supplied by the A-train motor-driven pump. This design provides the required redundancy to ensure that at least two steam generators receive the necessary flow ass11111ng any single failure. It can be seen from the description provided above that the loss of a single train of air (A or B) will not prevent the auxiliary feecl-water system from performing its intended safety function and is no more severe than the loss of a single auxiliary feedwater pump. Therefore, the loss of a single train of auxiliary air only affects the capability of a single motor-driven auxiliary feedwater pump because the turbine-driven pump is still capable of providing flow to two steam generators that are separate from the other motor-driven pump.

Two redundant steam sources are required to be operable to ensure that at least one source is available for the steam-driven auxiliary feedwater (ARI) pump operation following a feedwater or main steam line break. This require-ment ensures that the plant remains within its design basis (i.e., AFW to two intact steam generators) given the event of a loss of the No. 1 steam generator because of a main steam line or feedwater line break and a single failure of the 8-train motor driven AFW pump. The two redundant sources must be aligned such that No. 1 steam generator source is open and operable and the No. 4 steam generator source is closed and operable.

For instances where one train of emergency raw cooling water (ERCW) is declared inoperable in accordance with technical specifications, the ARI turbine-driven pump is considered operable since it is supplied by both trains r of ERCW. Similarly, the AFW turbine-driven pump is considered operable when one train of the AFW loss of power start function is declared inoperable in accordance with technical specifications because both 6.9 kilovolt shutdown board logic trains supply this function. This position is consistent with American National Standards Institute/ANS 58.9 requirements (i.e., postulation of the failure of the opposite train is not required while relying on the TS limiting condition for operation).

3/4.7.1.3 COMPENSATE STOBAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with steam discharge to the atmosphere concur-rent with total loss of off-site power. The contained water volume limit includes an allow111ce for water not useable because of tank discharge line location or other physical characteristics.

SEQUOYAH - UNIT 2 B 3/4 7-2b Amendment No. 105, 187, 1961

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATIQN BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 206 TO FACILITY OPEBATING LICENSE NO. DPR-77 AND AMENDMENT NO. 196 TO FACILITY OPERATING LICENSE NO. DPR-79 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT. UNITS 1 AND 2 DQCKET NOS. 50-327 AND 50-328

1.0 INTRODUCTION

By application dated December 16, 1994, the Tennessee Valley Authority (TVA or the licensee) proposed an amendment to the Technical Specifications (TS) for Sequoyah Nuclear Plant (SQN) Units 1 and 2. The requested changes would resolve a nonconservatism in the TS regarding the minimum turbine-driven (TD) auxiliary feedwater (AFW) system differential pressure (dp) requirements by replacing the current AFW TS 3/4.7.1.2 with requirements modeled after the Westinghouse Standard TS (STS), NUREG-1431. The licensee also proposed related changes to the Bases section of the TS.

r The need for these TS changes stems from a licensee event report (LER 50-327/93024) that involves an accident scenario assumption associated with the main feedwater line break (MFWLB) analysis. The analysis indicated that the plant needed to establish new minimum AFW flow requirements.

The current MFWLB analysis assumes the worst-case single active failure to be loss of the TDAFW pump. However, the licensee noted that Westinghouse had considered a single active failure of one of the motor-driven AFW pumps for the Eagle 21 reanalysis in 1989. This analysis assumes that 1070 gpm flow is available from two pumps (410 gpm from one motor-driven AFW pump and 660 gpm from the TDAFW pump). The design flow capacity of the TD pump is 880 gpm, but the TS surveillance test acceptance criteria only tests for a pump flow capacity of 440 gpm. The TDAFW test data indicates that its flow capacity is greater than the 660 gpm required for the accident scenario. The cause of the condition was a failure of the engineering design revie~s by Westinghouse and TVA to ensure that valid assumptions were used in the analysis.

Because of this discrepancy, the licensee requested that Westinghouse conduct a review of the accident analyses addressed in the Final Safety Evaluation Report (FSAR) in 1994 to verify other design basis assumptions. The reevaluation established that the minimum required flow from one TDAFW pump is 660 gpm. The minimum available AFW flow and system characteris tics have accounted for flow losses in a MFWLB and are credited .for removing decay heat ENCLOSURE 3

in the analysis of a small loss of coolant accident {LOCA). Based on the reevaluation, the licensee has proposed to revise the TS and the FSAR to reflect the new minimum AFW flow requirements and to identify the design basis assumptions.

The licensee also noted that the pump dp test values are not required as part of the surveillance requirements in the STS. Therefore, the licensee proposed to replace current AFW Specification 3/4.7.1.2 and the associated bases with the TS that are modeled after the Westinghouse STS and eliminate the dp test values from the TS.

By letter dated July 19, 1995, the licensee supplied additional information that co11111itted to evaluate testing of the AFW pumps on a staggered test basis by October 15, 1995. It did not change the no significa nt hazards consideration.

2.0 EVALUATION The AFW system at SQN, Units 1 and 2, is an engineered safety feature {ESF) system that is relied upon to aid in preventing core damage in the event of transients such as loss of normal feedwater, a secondary system pipe rupture, or a small-break LOCA. Except for the coRlllOn supply line from the condensate tanks and some shared support facilitie s such as the condensate storage tanks and parts of the control system, the two reactor units have separate and r identical AFW systems as shown in Figure 10.4.7-12 of the FSAR.

As stated in Section 10.4.7.2 of the FSAR, the AFW system has two 440 gpm motor-driven {MD) pumps and one 880:gpm TD pump. Each of the two MD pumps can supply two steam generators and the TD pump can supply all four steam generators. All three pumps automatically deliver the minimum required flow of 440 gpm within 60 seconds upon loss of offsite power, loss of both main feedwater pumps, or actuation of a safety injection signal. The FSAR indicates that sufficien t feedwater flow can be provided over the required pressure ranges for the design basis accident/transient, even assuming the worst single failure.

Due to the nonconservative assumptions found in the current TS with respect to the minimum dp requirements of the MDAFW pumps, the licensee proposed to revise TS 3.7.1.2 limiting conditions for operation {LCO) as follows:

I. Replace the paragraph "At least three independent steam generator AFW pumps and associated flow paths shall be operable: with---" with "three AFW trains shall be operable." The term "train" would denote the combination of components and flow path required to ensure the performance of the AFW function. This change conforms to the STS.

2.: Extend the current applicab ility in Modes 1, 2, and 3 to Mode 4. Mode 4 would then require that one MDAFW pump be maintained operable when steam generators are relied upon for heat removal, because the AFW system must supply enough makeup water to replace steam generator secondary inventory lost as the unit cools down to Mode 4. This change conforms to the STS.

~

~ 3. Rewrite the LCO action requirements:

Action *a* reads: "With one AFW train inoperable in Mode 1, 2, or 3, restore the inoperable AFW train to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in hot standby (Mode 3) within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in hot shutdown (Mode 4) within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.* In Action "a,* the licensee proposed adding "in Mode 1, 2, or 3," changing "AFW pump* to *AfW train," and extending the hot shutdown time from 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The extension of hot shutdown (Mode 4) time is within the STS time-limit of 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.

Action "b" reads: "With two AFW trains inoperable in Mode 1, 2, or 3, be in at least hot standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in hot shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />." In action "b," the licensee proposed adding "in Mode 1, 2, or 3" and changing "AFW pumps" to "AFW trains." The changes conform to the STS.

Action "c" reads: "With three AFW trains inoperable in Mode 1, 2, or 3, invnediately initiate corrective action to restore at least one AFW train to operable status." In action "c," the licensee proposed adding "in

. Mode 1, 2, or 3" and changing "AFW pumps" to "AFW trains." The changes conform to the STS.

4. Add a note for Action "c" to state: "LCO 3.0.3 and all other LCO actions requiring Mode changes are suspended until one AFW train is restored to

~ operable status.* This note would suspend entry into LCO 3.0.3 or other

~* LCO actions requiring mode changes until one AFW train is restored to operable status. The note is adopted from the STS.

5. Add Action "d" to state: "With the required AFW train inoperable in Mode 4, irmnediately initiate action to restore the required AFW train to operable status." The term "required AFW train" would apply to the designated motor-driven AFW pump that is being maintained operable in Mode 4. This action is adopted from the STS and is not required in the current TS.

The current surveillance requirement (SR) 4.7.1.2 is subdivided into requirements a, b, and c. The licensee proposed to change the format and rewrite SR 4.7.1.2 as follows:

1. SR 4.7.1.2.1 would state: "At least once per 31 days, verify each AFW manual, power operated, and automatic valve in eath water flow path, and in both steam supply flow paths to the steam turbine driven pump, that is not:locked, sealed, or otherwise secured in position, is in the correct position." This statement is adopted from the STS.

The current SR 4.7.1.2.c requires the licensee to verify each non-automatic valve flowpath in its correct position every 7 days. This requirement is deleted because the STS require the licensee to verify correct valve position every 31 days for manual, power operated, and r '

3 automatic valves. The licensee stated that the relaxation from 7 days to 31 days is based on existing procedural controls for valve configuration.

The current method of surveillance for these valves involves valve manipulation. The licensee stated that the intent of this change is not to require testing or manipulation of these valves, rather it is to verify that these valves that are capable of being mispositioned, are in the correct position. The proposed SR ensures that the automatic valves will actuate to their required position upon receipt of an AFW actuation signal. The staff concurs with the change.

2. SR 4.7.1.2.2 would state: "At least once per 92 days, verify the developed head of each AFW pump at the flow test point is greater than or equal to the required developed head."

The proposed SR will replace current SR 4.7.1.2.a, which requires the operability of each AFW pump to be verified by measuring the dp across each pump while the pump is on recirculation flow. These dp values correspond to the minimum required AFW flow for each pump. The proposed SR 4.7.1.2.2 will require the licensee to verify the developed head of each AFW pump at the flow test point that is greater than or equal to the required developed head and eliminates the numerical dp test values r provided in the current TS for each MD pump. The removal of the dp test values from the TS will allow testing of these pumps in the normal operation alignment and allow administrative control of these values within the design basis of the plant.

The licensee proposed adding a 92-day frequency in the SR for testing each AFW pump. This test frequency deviates from that stated in the STS of every 31 days on a staggered test basis. The licensee stated that the 92-day frequency is consistent with the test frequency required by the plant in-service test program. Also, the "staggered test basis" defined in the STS is different from the current plant TS definition.

Therefore, incorporation of the STS definition for staggered test basis would impact a number of other plant specifications and would significantly expand the scope of the change. By letter dated July 19, 1995, the licensee co11111itted to evaluate testing of the AFW pumps on a staggered test basis by October 15, 1995. The staff finds this position satisfactory.

The licensee proposed adding a note for SR 4.7.1.2.2 to state: "Not required to be completed for the turbine-driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after steam supply pressure is greater than or equal to 842 psig." This note will establish the necessary conditions for performing the surveillance and provide a suitable timeframe to complete the testing.

This note is adopted from the STS.

3. SR 4.7.1.2.3 would state: 0nce every 18 months, verify each AFW 11 automatic valve that is not locked, sealed, or otherwise secured in

r position, actuates to the correct position on an actual or simulated actuation signal.* This SR conforms to the STS.

The proposed SR 4.7.1.2.3 would replace current SR 4.7.1.2.b.l, which requires each automatic valve actuation to be verified to its correct position upon receipt of an AFW actuation test signal or a low AFW pump suction pressure test signal every 18 month. The licensee stated that this testing is performed under Surveillance Instruction (Sl-OPS-003-118.0), which retains both the AFW actuation test signal and the low pump suction pressure test signal. Therefore, the proposed SR 4.7.1.2.3 will include both test signals and ensure the surveillance is performed under administrative control to ensure that the valves are in the required position.

The licensee also proposed adding a note to SR 4.7.1.2.3 to state: "Not applicable in Mode 4 when steam generators are relied upon for heat removal." In Mode 4, the AFW actuation signals are not required to be operable. This note is adopted from the STS.

4. SR 4.7.1.2.4 would state: "Once every 18 months, verify each AFW pump starts automatically on an actual or simulated actuation signal." The proposed SR 4.7.1.2.4 would retain the same requirements as current SR 4.7.1.2.b.2 except two additional provisions adapted from the STS would be added as follows:

(a) Current SR 4.7.1.2.b.2 verifies AFW pump automatically starts upon receipt of an AFW actuation test signal, which is a simulated signal. The proposed SR 4.7.1.2.4 would use an actual or a simulated signal. An actual signal could be used for a non-planned test to satisfy the SR.

(b) A note would be added to SR 4.7.1.2.4 to indicate that the surveillance is not required to be completed for the turbine driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after steam supply pressure is greater than or equal to 842 psig, and the surveillance is not applicable in Mode 4 when steam generators are relied upon for heat removal.

These notes are adopted from the STS and will establish the conditions and timeframe to complete the testing.

The licensee also proposed revisions to the Bases for TS 3/4.7.1.2 to justify these changes and explain the assumptions for the limiting design basis accidents and transients for the AFW system. These Bases would clarify the intent of the TS. The staff finds the revised bases are adequate to support the proposed TS changes.

The staff finds the proposed changes to the Sequoyah 1/2 TS concerning the auxiliary feedwater system acceptable based on the following considerations:

  • The proposed TS do not alter the staff's previous review conclusion stated in Section 11.E.l.l of the Sequoyah Safety Evaluation Report regarding AFW system reliability evaluation.
  • The TS changes are needed to correct the errors in the requirements of dp testing for the motor-driven pumps resulting from previous assumptions in the main feedwater-line-break analysis.
  • The TS changes are consistent with the Westinghouse STS, which provide a TS improvement that has been approved by the *NRC.
  • The proposed changes are justified by the licensee with technical bases that have been reviewed and found to be satisfactory by the staff.

3.0 STATE CONSULTATION

In accordance with the Co11111ission's regulations, the Tennessee State official was notified of the proposed issuance of the amendment. The State official had no comments.

4.0 ENVIRONMENTAL CONSIPERATION The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR r Part 20 and 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 Co11111ission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public connent on such finding (60 FR 6309). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR Sl.22(c)(9). Pursuant to 10 CFR Sl.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

5.0 CONCLUSION

The Co111nission 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 activities will be conducted in compliance with the Co11111ission's regulations, and (3) the issuance of the amendment will not be inimical to the conman defense and security or to the health and safety of the public.

Principal Contributor: Jin-Sien Guo Dated: August 2, 1995

Retrieved from "https://kanterella.com/w/index.php?title=ML18087A273&oldid=2533502"