ML20092H543
| ML20092H543 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 09/15/1995 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20092H547 | List: |
| References | |
| NUDOCS 9509210073 | |
| Download: ML20092H543 (6) | |
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EELOSURE 1 SEQUOYAH NUCLEAR PLANT UNIT 1 TECHNICAL SPECIFICATIONS PAGE B3/4 7-5 i
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9509210073 950915 PDR ADOCK 05000327 P
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i PIJ0TT SYSTEMS I
sASIS I
1 3/4.1.s AtrxitrARY BUILDING GAS TREATMENT SYSTIM j
The OPERABILITY of the auxiliary building gas treatment system ensures that radioactive materials leaking from the ICCS equipment following a ICCA are i
i filtered prior to reaching the environment. The operation of this system and l
the resultant effect on offsite dosage calculations was assumed in the accident analyses.
ANSI N510-1975 will be used as a procedural guide for surveillance i
testing.
Cumulative operation of the system with the heaters on for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> over a 31 day period is sufficient to reduce the buildup of moisture on the
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adsorbers and HEPA filters.
l The minimum vacuum relief flow requirement in TS Surveillance Requirement.
BR-6 4.7.8.d.3 is for test purposes only. It is intended to demonstrate an acceptable level of ARGTS performance margin by simulating an ABSCE boundary breach.
The inability to meet the specified minimum test condition under other circumstances does not challenge the operability of the AaGTS.
f 3/4.7.9 SNUBBERS i
Snubbers are designed to prevent unrestrained pipe or component motion under dynamic loads as might occur during an earthquake or severe transient, i
while allowing normal thennal motion during startup and shutdown. The con-sequence of an inoperable snubber is an increase in the probability of structural damage to piping or components as a result of a seismic or other event initiating dynamic loads.
It is therefore required that all snubbers required to protect the primary coolant system or any other safety system or component be operable during reactor operation.
Because the snubber protection is required only during relatively low probability events, a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed to replace or restore the inoperable snubber (s) to operable status and perform an engineering evaluation on the supported component or declare the supported system' inoperable and R16 follow the appropriate limiting condition for operation statement for thag system.
The engineering evaluation is performed to determine whether the mode of failure of the snubber has adversely affected any safety-related component or system.
Safaty-related snubbers are visually inspected for overall integrity and operability.
The inspection will include verification of proper orientation, adequate fluid level if applicable, and attarhment of the snuhber to its anchorage. The removal of insulation or the verification of torque values for threaded fasteners is not required for visual inspections.
The inspection frequency is based upon maintaining a constant level of snubber protection. Thus, the required inspection interval varies inversely with the observed snubber failures. The number of inoperable snuhbors found during a required inspection determines the time interval for the next required inspection.
Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25 percent) may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule.
August 11, 1995 SEQUOYAH - UNIT 1 B 3/4 7-5 n=an h at No. 12 By letter dated 9/15/95
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ENCLOSURE 2 SE000YAH NUCLEAR PLANT UNIT 2 1
IECHNICAL SPECIEICATIONS PAGES B3/4 7-2 AND B3/4 7-5 1
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i PLAhTSYSTEMS BASES R187 Nominal NSSS power rating of the plant (including reactor Q
=
coolant pump neat), Mwt Conversion factor, 947.82 (Btu /sec)
K Mwt
^
Minimum total steam flow rate capability of the operable MSSVs w,
=
on any one steam generator at the highest MSsv epening pressure including tolerance and accumulation, as appropriate, in lb/sec.
For example, if the maximum number of inoperable MSSVs on any 1
one steam generator is one, then w, should be a summation of the l
capability of the operable MSSVs at the highest capacity MSSV operating pressure, excluding the highest capacity MssV.
If the maximum number of inoperable MSSVs per steam generator is three
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then w, should be a summation of the capacity of the operable MSSVs at the highest operable MSSV operating pressure, excluding l
the three highest capacity MSSys.
heat of vaporization for steam at the highest MSSV opening hg pressure including tolerance and accumulation, as appropriate, Btu /lbm Number of loops in plant f
N
=
The valves calculated from this algorichm must then be adjusted lower to account for instrument and channel uncertainties.
3/4.7.1.2 AUXILIARY FTEDWATER SYSTEM The OPERABILITY of the auxiliary feedwater system ansures that the Reactor Coolant System can be cooled down to less then 350*F from normal operating conditions in the event of a total loss of off-site power.
The steam driven auxiliary feedwater pump is capable of delivering 880 gpm (total feedwater flow) and each of the electric driven auxiliary e
feedwater pumps are capable of delivering 440 gym (total feedwater flow) to the l
cntrance of the steam generators at steam generator pressures of 1100 psia. At 1100 psia the open steam generator safety valve (s) are capable of relieving at least lit of nominal steam flow. A total feedwater flow of 440 gym at pressures of 1100 psia is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System j
temperature to less than 350*F where the Residual Heat Removal System may be placed into operation.
The surveillance test values ensure that each pump will provide at least 440 gym plus pump recirculation flow against a steam generator pressure of 1100 psia.
I Each motor-driven auxiliary feedwater pump (one Train A and one Train B)
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cupplies flow paths to two steam generators. Each flow path contains an BR-1, i
eutomatic air-operated level control valve (LCV).
The LCVs have the same train designation as the associated pump and are provided trained air. The turbine-1 driven auxiliary feedwater pump supplies flow paths to all four steam Each of these flow paths contains an automatic opening generators.
BR-7 l
(non-modulating) air-operated LCV, two of S3QUOYAH - UNIT 2 B 3/4 7-2 Amendment No. 105, 187 May 25, 1995 Dy letter dated 9/15/95 i
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s PLANT SYSTEMS BASES 3/4.7.s AtrxrLIARY BUILDING GAS TREATMENT SYSTEM The OPERABILITY of the auxiliary building gas treatment system ensures tha*
radioactive materials leaking from the ECCS equipment following a LOCA are filtered prior to reaching the environment. The operation of this system ana the resultant ef fect on offsite dosager calculations was assumed in the accident analyses. ANSI N510-1975 will be used as a procedural guide for surveillance testing.
Cumulative operation of the system with the heaters on for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> over a 31 day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters.
The minimum vacuum relief flow requirement in TS Surveillance Requirement 4.7.8.d.3 is for test purposes only.
It is intended to demonstrate an BR-8 acceptable level of ABGTS perfonnance margin by simulating an ABSCE boundary breach. The inability to meet the specified minimum test condition under other circumstances does not challenge the operability of the ABGTS.
3/4.7.9 SNUBBERS Snubbers are designed to prevent unrestrained pipe or component motion under dynamic loads as might occur during an earthquake or severe transient, while allowing normal thermal motion during startup and shutdown. The consequence of 1
an inoperable snubber is an increase in the probability of structural damage to i
piping or components as a result of a seismic or other event initiating dynamic loads.
It is therefore required that all snubbers required to protect the primary coolant system or any other safety system or component be operable during reactor operation.
Because the snubber protection is required only during relatively low probability events, a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed to replace or restore the ino;. able snubber (s) to operable status and perform an engineering evaluation on el supported component or declare the supported system inoperable and follow the appropriate limiting condition for operation statement for that system. The engineering evaluation is perfonned to determine whether the ande of failure of the snubber has adversely affected any safety-rslated component l
or system.
R2 l
Safety-related snubbers are visually inspected for overall integrity and l
operability.
The inspection will include verification of proper orientation, adequate fluid level if applicable, and attarh= ant of the snubber to its anchorage. The removal of insulation or the verification of torque values for i
threaded fasteners is not required for visual inspections.
The inspection frequency is based upon maintaining a constant level of snubber j
protection. Thus, the required inspection interval varies inversely with the observed snubber failures. The number of inoperable snubbers found during a required inspection determines the time interval for the next required inspection.
Inspections performed before that interval has elapsed may be used as a aew reference point to determine the next inspection. However, the resulu of such early inspections performed before the original required time interval has elapsed (nominal time less 25 percent) may not be used to lengthen the required inspection interval. Any inspection whose results require e j
shorter inspection interval will override the previous schedule.
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August 11, 1995 Amendment 2 SEQUOYAH - UNIT 2 5 3/4 7-5 l
By letter dated 9/15/95 i
4 Mr. Oliver D. Kingsley, Jr.
SEQUOYAH NUCLEAR PLANT Tennessee Valley Authority cc:
Mr. O. J. Zeringue, Sr. Vice President TVA Representative Nuclear Operations Tennessee Valley Authority Tennessee Valley Authority 11921 Rockville Pike 3B Lookout Place Suite 402 1101 Market Street Rockville, MD 20852 Chattanooga, TN 37402-2801 Regional Administrator Dr. Mark 0. Medford, Vice President U.S. Nuclear Regulatory Commission Engineering & Technical Services Region II Tennessee Valley Authority 101 Mariatta Street, NW., Suite 2900 3B Lookout Place Atlanta, GA 30323 1101 Market Street Chattanooga, TN 37402-2801 Mr. William E. Holland Senior Resident Inspector Mr. D. E. Nunn, Vice President Sequoyah Nuclear Plant New Plant Completion U.S. Nuclear Regulatory Commission Tennessee Valley Authority 2600 Igou Ferry Road 38 Lookout Place Soddy Daisy, TN 37379 1101 Market Street Chattanooga, TN 37402-2801 Mr. Michael H. Mobley, Director Division of Radiological Health Mr. R. J. Adney, Site Vice President 3rd Floor, L and C Annex Sequoyah Nuclear Plant 401 Church Street Tennessee Valley Authority Nashville, TN 37243-1532 P.O. Box 2000 Soddy Daisy, TN 37379 County Judge Hamilton County Courthouse General Counsel Chattanooga, TN 37402-2801 Tennessee Valley Authority ET llH 400 West Summit Hill Drive Knoxville, TN 37902 Mr. P. P. Carier, Manager Corporate Licensing Tennessee Valley Authority l
4G Blue Ridge 1
1101 Market Street Chattanooga, TN 37402-2801 Mr. Ralph H. Shell Site Licensing Manager Sequoyah Nuclear Plant 1
Tennessee Valley Authority P.O. Box 2000 S6ddy Daisy, TN 37379 r
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