ML20196A069
| ML20196A069 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 12/02/1988 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20196A058 | List: |
| References | |
| NUDOCS 8812050247 | |
| Download: ML20196A069 (80) | |
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ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-88-42)
LIST OF AFFECTED PAGES Unit 1 2-6 B 2-4 3/4 3-2 3/4 3-5 Unit 2 2-6 B 2-4 3/4 3-2 3/4 3-5 e e PDC
I m TABLE 2.2-1 (Continued)
- E g -
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 1
5 . .
t 1
FUNCTIONAL UNIT TRIP SETPOINT c ALLOWABLE VALUES
$ 13. Steam Generator Water 1 18% of narrow range instrument 3 17% of narrow range instrument R20 y Level--Low-Low span each steam generator span-each steam generator
- 14. Steam /Feedwater Flow < 40% of full steam flow at Missatch and Low Steam
< 42.5% of full steam flow at RATED THERMAL POWER coincident RATED THERMAL POWER coincident Generator Water Level with steam generator water level with steam generator water level 1 25% of narrow range instru- 1 24.0% of narrow range instru-ment span--cach steam generator ment span- each steam generator
- 15. Undervoltage-Reactor 2 5022 volts-each bus Coolant Pumps 14739 volts each bus R89 y 16. Underfrequency-Reactor 2 56.0 Hz each bus each bus o Coolant Pumps 1 55.9 Hz
- 17. Turbine Trip A. Low Trip System 1 45 psig Pressure 1 43 psig B. Turbine Stop Valve 1 1% open 1 1% open
. Closure
- 18. Safety Injectian Input Not Applicable Not Applicable frem ESF y
~
-F 4d MTED Ttfstal4L p 3 4 a go-'$ .-f MTEi> Ttf@est4L
- 19. Int?rmediate Range Neutron -10 NM -11 FOWM d -' 1 ' 10 v - " ; 10 c pr Flex - (P-6) Enable Block
%{ Source Range Reactor TripM a a h3 20. Power Range Neutron Flux < 10% of RATED < 11% of RATED
"[ (not P-10) ' Input to Low Pcwer THERMAL POWER THERMAL POWER
.% ?
Reactor Trips Block P-7
~* j dl
g p .4, .
4 ~ ~ & % * " ' f
. u SAFETY LIMXTS BASES WY Range Channels will initiate a reactor trip = ~e"+ 1- ' p cp rt h ,el tv !
approximately 25 percent of RATED THERMAL PO unless manually blocked when P-10 becomes active. No credit was taken for o n of the trips associat with either the Intermediate or Suurce Range Channels e n s; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection System. '
Overtemperature Delta T The Overtemperature Delta T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds),
and pressure is within the range between the High ana Low Pressure reactor trips. This setpoint includes corrections for axial power distribution,
- changes in density and heat capacity of water with temperature and dynamic compensation for piping delays from the core to the loop temperature detec-tors. With normal axial pwer distribution, this reactor trip limit is always below the core safety limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1.
l
,l Operation with a reactor coolant loop out of service below the 4 loop P-8 setpoint does not require reactor protection system setpoint modification because the P-8 setpoint and associated trip will prevent DNB during 3 loop operation exclusive of the Overtemperature Delta T setpoint. Three loop operation above the 4 loop P-8 setpoint is permissible after resetting the K1, K2 and K3 inputs to the Overtemperature Delta T channels and raising the P-8 setpoint to its 3 loop value. In this mode of operation, the P-8 inter-lock and trip functions as a High Neutron Flux trip at the reduced power level.
Overpower Delta T The Overpower Delta T reactor trip provides assurance of fuel integrity, e.g. , no melting, under all po'ssible overpower conditions, lhnits the required range for Overtemperature Delta T protection, and provides a backup to the liigh Neutron Flux trip. The setpoint includes corrections for changes in density and heat capacity of water with temperature,.and. dynamic compensation for piping delays from the core to the loop temperature detectors. No credit was taken for operation of this trip in the accident 0
SEQUOYAH - UNIT 1 , B 2-4 1
. TABLE 3.3-1
. t y REACTOR TRIP SYSTEM INSTRUMENTATION l
t O 1 E :
MINIMUM E TOTAL NO. CilANNELS CilANNELS SUNCTIONAL UNIT, APPLICABLE OF CilAf81ELS 10 TRIP U OPERABLE MODES ACTION r 1. Mancal Reac' tor Trip 2 1 2 1, 2, and
- 1
- 2. Powe'r Range *, Neutron Flux 4 2 3 1, 2 #
4 2
- 3. Power Range, Neutron Flux 4 liigh Positise Rate 2 3 1, 2 2' 4 Power Range, Neutron Flux,
' 4 2 3 1, 2 #
liigh Negative Rate 2
- 5. Intermediate Range, Neutron Flux
{ 2 '
1 2 1, 2, and
- 3 Y
- 6. Source Range, Heatron Flux b
. A. Startup_ 2 1 2
- 8. Jhutdowa 2 and
- 4 0 1 3, 4 and 5 5
}'. Overtemperature Delta T
.cour Loop Operation 4 , 2 3 1, 2 6, S. Overpower Delta T ~
lR45
, Four Loop Operation 4 2 #
3 1, 2 6
- 9. Presstrizer Pressure-Low l R45 4 2 3 1, 2 #
6 10 Pressu:-izer Pressure--liigh 4 #
2 3 1, 2 6 N E" - ,
{n 11. Pressur'izer Wcter Level--liigh 3 2 2 1, 2 7 oa Y. p yy -j h v d ft 5I Y$Y a., ,; J4+ sec10og & d itlit. s 2 9 2 1OwWwa ?
~he-
e , * .
4
_ TABLE 3.3-1 (Continued)
TABLE NOTATION With the reactor trip system breakers in the closed position and the control
==
rod drive system capable of rod withdrawal, and fu11 in the reactor vessel.
The channel (s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped condition.
b pro ions pecif c on 3: . re no plic .
O ;'! "c't:;: te deter +^- = y ha da-aaa T red 2: : th: " O (C'e:L ef 6 c-+
P=;: $r :t:r Trip) ::tp:! S. .
CTI0t nTEMEN ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-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/or open the reactor trip breakers.
ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STAR 70P and POWER OPERATION may proceed provided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. lR$1
- b. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> lR51 for surveillance testing per Specification 4.3.1.1.1.
- c. Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL and the Fower Range, Neutron Flux
. high trip reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,
- d. The QUADRANT POWER TILT RATIO, as indicated by the remaining three detectors is verified consistent with the normalized symmetric power distribution obtained by using the movable incare detectors in the four pairs of symmetric thimble locations at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when THERMAL POWER is
. greater than 75% of RATED THERMAL POWER.
. -s
- September 17, 1986 SEQUOYAH - UNIT 1 ,
3/4 3 5 Amendment No. A
m u, - -
TABLE 2.2-1 (Continued) l! REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SEIPOINTS s
I ii
- l FUNCTIONAL 04 TIT TRIP SETPOINT ALLOWABLE VALUES si 13. Steam Generatar Water 2 18% of narrow range instrument 3 17% of narrow range instrument R7
- Level--!.ow-Low span-each steam generator span-each steam generator r.
- 14. Steae/Feedwater Flow < 40% of full steam flow at < 42.5% of full steam flow at Mismatcc aad Lee Steam RATED TilERMAL POWER coincident RATED THERMAL POWER coincident Generator Water Level with steam generator water level with steam generator water level
> 2S% of narrow range instru- 1 24% of narrow range instru-4rnt span- cach steam generator ment span--each steam generator R76
- 15. Uncervoltage-Reactor 1 5022 volts each bus 1 4739 volts-each bus Coolant Pumps
- 16. Uncerfrequency-Reactor 1 56 Itz - each bus 1 55.9 Hz - each bus
? Coolant Purps e
- 17. Turbine Trip A. Low Trip System 1 45 psig 1 43 psig Pressure B. . Turbine Stop Valve 1 1% open > 1% open
. Closure
- 18. Sa#3ty Injection Input Not Applicable Not Applicable
~
>\pJOS$of RATED YttOUltL of RAW W g gy70'
- 19. Intermediate Range Neutron -- I - 10 '" 1.,a-
- Mv
~
"' c-u, Flux, P-6 Enable Block 3g. Source Range Reactor Trip
?y't. 20. Power Range Neutron Flux < 10% of RATED < ILE of RATED THERMAL POWER ,
3 '
(not P-10) Input to Low TilERMAL POWER u r' Power Reactor Trips I'
!*j:
Block P-7 I' 2
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O e
LIMITING SAFETY SYSTEM SETTINGS BASES Intermediate and Source Range, Nuclear Flux (Continued) d '
Range Channels will initiate a reactor trip ct c arrc.t hd prgutbrf-4+-- l approximately 25 percent of RATED THERMAL POWER unless manually blocked whsn P-10 becomes active. No credit was taken for operation of the trips associ-ated with either the Intermediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of tht Reactor Protection System. .
t
'Overtemperature oT The 0vertemperature delta T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping ,
transit delays from the core to the temperature detectors (about 4 seconds),
and trips.
pressure is within the range between the High and Low Pressure reactor This setpoint includes corrections for axial power distribution, changes in density and heat capacity of water with temperature and dynamic compensation for piping delays from the core to the loop temperature detectors.
With normal axial power distribution, this reactor trip limit-is always below the core safety limit as shown in Figure 2.1-1. If axial peaks are greater i than design, as indir.ated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the i
notations in Table 2.2-1. -
I i
Operation with a reactor coolant loop out of service below the 4 loop P-8 setpoint does not require reactor protection system set point modification because the P-8 setpoint and associated trip will prevent DNB during 3 loop operation exclusive of the Overtemperature delta T setpoint. Three loop operation above the 4 loop P-8 setpoint is permissible after resetting the K1, K2, and K3 setpoint inputs to its to the 3 loop Overtemperature delta T channels and raising the P-8 value. In this mode of operation, the P-8 interlock and trip functions as a High Neutron Flux trip at the reduced power level.
Overpower AT '
The Overpower delta T reactor trip provides assurance of fuel integrity,
- e.g., no melting, under all possible overpower conditions, limits the required
' range for Overtemperature delta T protection, and provides a backup to the High heutton Flux trip. 'The setpoint includes corrections for changes in '
density and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop temperature detectors. No credit was taker) for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.
hN SEQUOYAH - UNIT 2 8 2-4
}
s . .
- v. -
TABLE 3.2-1 .
. 4 g . . .
REACTOR TRIP SYSTEM INSTRUMENTATION
- j 5
' I
' MINIMUM TOTAL NO. CHK4NELS CHANNELS APPLICA8LE FUNCTIONAL ilNIT OF CHANNELS
, TO TRIP OPERABLE MODES ACTION
- 1. Manual React 5r Tr'ip 2 1 2 1, 2, and
- 1
- 2. Power Range, ileutron Flux 4 #
2 3 1, 2 2
- 3. Power Range,i;eutron Flux 4 #
2 3 1, 2 2 High Positive Rate
- i
- 4. PowerRange,t[eutronFlux, 4 #
2 3 1, 2
, HighNejativelate 2 I'
{ 5. Intermediate Rpnge, Neutron Flux 2 1 2 1, 2, and
- 3
,Y 6. Source Range, keutron Flux A. Startty 2 1 2 2 , and
- 4
- 8. Sh,'utdowa ., 2 0 1 3, 4 and S 5
- 7. Os.'riesperature AT F* o ur Loop Operatfca 4 2 3 1, 2 6,
- 8. Overpoweb AT
- R33 '
,, Fosr Loop Operation 4 2 3 #
" .y - - 1, 2 6 "s 9. Pressurf ter Pressure-Low 4 y lnn a"
a3 2 3 1, 2 6
!! j
- 10. Pressurizer Pressure--High 4 2 3 1, 2 6-Y u -.
& 11. Pressurfrer Watzr level--High 3 y 2 2 1, 2 O' 7 l
Cu -
l
- 9 l
\ }
- b
4 TABLE 3.3-1 (Continued)
TABLE NOTATION s
With the reactor trip system breakers in the closed position, the control rod drive system capable of rod withdrawal, and fuel in the reactor vessel.
mA The channel (s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped condition. '
, [TheprovisionsofSpecification3.0.4arenotapplicable.
""Migh M t ;e : d;;;;th ::,7 b: d: :::r;t::d :b: : 15: "-S (Sh d eT me- .
":n;: *:::t:r Trip) ::tp: ht.
ACTION STATEMENTS l ACTIuN 1 - With the number of OPERABLE channels one less than required by 4
the Minimum Channels OPERABLE requirement, restore the inoperable i channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-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/or open the reactor trip breakers.
I ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed l provided the following conditions are satisfied:
I
- a. The inoperable channel is placed in the tripped condition ,
.. within 6 he';rs. lR3 ,
- b. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> lR3 i for surveillance testing per Specification 4.3.1.1.1.
- c. Either, THERMAL POWER is restricted to less than or equal ,
to 75% of RATED THERMAL POWER and the Power Range, Neutron Flux trip setooint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 nours; or, the QUADRANT POWER TILT RATIO is monitored at le'..,t once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,
- d. The QUADRANT POWER TILT RATIO, as indicated by the i maining three detectors, is verified consistent with the normalized syr.?etric power distribution obtained by using the movable incere detectors in the four pairs of symmetric thimble locations at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when THERHAL POWER is creater than 75%.of RATED THERMAL POWER.
) -
September 17, 1986 SEQUOYAH UNIT 2 3/4 3-5 Amendment No. 39
4 ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-;28 (TVA-SQN-TS-88-42)
DESCRIPTIOh AND JUSTIFICATION FOR MODIFICATION OF THE TRIP SETPOINT AND ALLOWABLE '
VALUE UNITS FOR THE INTERMEDIATE RANGE NCCLZAR FLUX DETECTOR AND CHANGES TO THE .UPLICABILITY REQUIREMENTS FOR THE SOURCE RANCE NUCLEAR FLUX DETECTOR f
l J
l l
I f
. EiCLOSIAE 2 Description of Change Tennessee Valley Authority proposes to modify t'ae Sequoyah .Vuclear Plant (SQN) units 1 and 2 technical specifications to revise the trip setpoint and allowable value units for the intermediate-range (IR) nuclear flux detector and to revise the applicability requirements for the source range (SR) nuclear flux detector.
Reason for Change TVA is replacing the SR and IR neutron monitors as part of the equipment
- upgrade to comply with Regulatory Guide 1.97 as required by SQN license conditions 2.C.24 (unit 1) and 2.C.14 (unit 2). The new SR/IR monitor is a fission chamber design manufactured by Gamma Metrics. This design does not require high-voltage decaergization as part of the normal SR detector operatioa. Consequently, the applicability table 3.3-1 is being revised to delete an unnecessary note involvinr high-voltage deenergization. The new IR monitor uses a signal that is in units of relative power.
Consequently, the trip setpoint and allowable value are being changed in "
table 2.2-1. The bases to section 2.2 are also being revised to delete references to IR detector current signals that are proportional to power levels. The changes to unit 1 also have appropriate footnotes added to indicate that the changes beccme effective for unit 1 after installation of the new detectors during the unit I cycle 4 refueling outage. The unit 2 detectors will be installed during the unit 2 cycle 3 refueling outage, and the change will be effective at the time of startup following the outage.
Justification for Change The new Camma Metrics SR/IR detectors are being installed to achieve compliance with Regulatory Guide 1.97. The new detectors are class-1E equipment that is seismically and environmentally qualified.
The new SR detector design is compatible with the current systemt however, it includes two improvements over the present detector design. First, the electronic equipment automatically adjusts the high flux at shutdown alarm. Currently, this function is performed manually as described in the Final Safety Analysis Report, cection 15.2.4.2. Second, the new SR detector does not have to be deenergized at higher power levels. Above the P-6 setpoint, the SP, detector output signal is blocked from the reactor trip logic. However, the SR/IR detector assemblies remain energized during the full range of power operation. As a result of this feature, the table notation in table 3.3-1 regarcing high-voltage deenergiration can be deleted iecause it is not applicable to the new design.
The new IR detector design is compatible with the current system except that the output signal is in units of relative power rather than amperes (amps). The P-6 setpoint and allowable value listed in table 2.2-1 are currently listed in units of amps. TVA has performed a calculation to determine the relative power values corresponding to the present trip setpoint and allowable value. A relationship between reactor power 3
and detector current was establisaed using startup test data from several power levels between 5- and 90-percent power. This relationship was then used to convert the trip setpoint to a relative power value. The computed [
value was rounded to the next coeservative decade for ease of calibration. A correspond 8.ng allowable value was then calculated using the previously established setpoint and current-power relationship.
Finally the overlap between the SR/IR dstector ranges was checked to ensure sufficient margin betwsen the P-6 setpoint and the SR trip setpoint. It is important to note that the actual setpoint is not changed; only the enginetring units have chenged. A copy of the TVA calculation is incivded as an attachment to this enclosure.
- Footnotes have been added to tables 2.2-1 and 3.3-1 of the unit 1 technical specifications that indicate that the proposed changes do not become effective until installation of the SR/IR detector assemblies during the unit 1 cycle 4 refueling outage scheduled for mid-1990. This method of handling the unit 1 change will allow a single review cf the issue and avoid separate technical specification change reques:s for osch unit.
In stmaary, three administrative changes are proposed to support the installation of the Gamma Metries SR/IR assembly. The first involves the :
deletion of a table note that is not applicable to the design of the new SR detectors. The second involves a change in engineering units for the P-6 setpoint that results from the difference in output signals from the IR detectors. The third involves the addition of certain footnotes to enoble the review and approval of the unit 1 changes to proceed independently of the unit 1 installation schedule.
Environmental Impact Evaluation The proposed revision involves an administrative change to a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes to the surveillance requirements. TVA has determined that the proposed change involves no significant increase in the amounts, and no sig'ificant change in the types, of any effluents that may be released rffsite and that there is no significant increase in individual or cumulatise occupational radiation exposure. Accordingly, the proposed change meets the ;
eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b). no environmental impact statement nor environ ental assessment needs to be prepared in connection with tra issuance of the amendment.
l
ATTACHMENT 1 TVA CALCUI.ATION, "!NTEU1EDIATE RANGE NEUTRON FLUX P-6 SETPOINT," REVISION 1 (B25 881117 808) l
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- 1. : ealga Review
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Just; fica [lon(expl.,ln below):
Mathd1:!Inthedi sign review method, justify the technteal acequacy of the
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In the alternate calculation cethod, identify the pages where the alternata and esp 1., lncalculation has been included in the calculation pa:kage why this nethod is adequate.
Method 3:' source In the(squalification test method, identify the QA docu..ented calculat on where testing adequately demonstrates the adequacy of this and explain.
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~\ @ '45~ % ~\ \
- Calculation No. Revision i
Method of design verification (independent review) used (check method used):
1
- 1. Design Review *
- 2, Alternate Calculation i
- 3. Qualification Test Justific6 tion (esplain below):
Method 1: In the design review method, justify the technical adequacy of the calculation and explain how the adequacy was verified (calculation is
- similar to another, based on accepted handbook methods, appropriate
] sensitivity studies included for confidence, etc.).
] Method 2: In the alternate calculation method, identify the pages where the
- alternate calculation has been included in the calculation package and explain why this method is adequate.
Method 3: In the qualification test method, identify the QA documented source (s) where testing adequately demonstrates the adequacy of this W
calculation and explain.
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page 1 of 1 cat.CULATION DESIGN VERIFICATION (INDEPENDENT REV! N !?R I f 2. - X IJ- 9 a - i )
Calculation No. Revision Method of design verification (independent review) used (chect .iethoc used):
- 1. Design Review
- 2. Alternate Calculation .
- 3. Qualification Test Justification (esplain below):
Method 1:
In the design review reethod. justify the technical adequacy of the calculation and explain how the adequacy was verified (calculation is strellar to another, based on accepted handbook, methods, appropriate sensitivity studies included for confidence, etc.).
Method 2:
In the alternate calculation method. identify the pages where the alternate calculation has been laciuded in the calculation paciste and explain why this method is adequate.
Method 3: In the qualification test teethod. Identify the QA doew entert source (s) where testing adequately demonstrates the scequacy of this calculation and esplain. ,
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l CALCULATIONDESIhNVERIFICATION(INDEPENDENTREVIE FORM it2'XE*42-( O Calculation No. RerAJiou !
Method of design verification (independent review) used (check.ethod used):
- 1. Oc:!&S Review V
?. Alternate Calculation
- 3. Qualificatiss Test Justification (espi,$1n below):
Method 1:
In the des'gn review method, justify the technical adequacy of the calculation and explain how the adequacy was verified (calculation is similar to another, based on accepted handboot methods, appropriate sensitivity studies included for confidence, etc.).
Method 2: In the alternate calculation method, identify the pages where the alternate calculation has been included in the calculation pactate and esplain why this method is adequate.
Method 3: In the qualification test method, identify the QA documented -
source (s) where testing adequately demonstrates the adequacy of this calculation and explain.
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C .
TENNESSEE VAlt.EY AUTHORITY SIGUOYAH NUCLEAR PLANT UNIT NUMBERS 1 AND 2
(' PAECAUT!0NS, LIMITATIONS AND SETPOINTS ~
FOR NVCLEAR STEAM SUPPLY SYSTEMS
( REVISION 9 PAY,1981 . . - - -
1 1 .. .. . . : .t ,
(7 as te.hed p.qrs ) ' ',:. i : ;
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- WESTINGHOUSE El.ECTAIC CORPORATION " ,
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Nuclear Energy Sys m s P* 0* Box 355 .
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(FB.i103, FS.52C3, FS-5303. FS-5403, FB-illB, F3 5213 FB 5313, FB S413) 33% of rated stasn II O* E'" S I'2
generator
! 3. turb'ne trip ,
stear generator Hf level signsi for C fcedwiter valve closure, turbine trip and feedwater pump trip (LS 517A, LS-527A LB-537A, LB 547A,
' 75 of level s; n LB-218A, L! 523A, LB-53SA, LB-548A, (5 519A L3 520A, LB 539A, LB 549A) .
C to f DT h! , Pemissive_ ard intpelock circuits 4b )X lC I A.
i
/
I ' A. ! P-6(a11cus manual bicek of source range
!highlevel reactor trip)
'(NC3!D. 'IC360) . .. . 10 10 ,3;,7,,
B. P 7 (aut: atically blocks various "at pow 6r tripsat1cwpower)
- 1. Iow r'eutronflux(Seep-10)
- 2.
low tarbine load (See P-13)
C.fPB(411c <s one loop loss of flow belew jsetpoint) , -
'(NC41N. 'iC 42N, NC-43N, NC 44N) 35: of full ecoer D. P-9 lblec ;s reacter trip en turbir.e #
, trip bele < set;oint nuclear power level)
- (t:C-415 9C425,NC435,NC-445)
SCt of full p:Wer
E. 'P 10 (411 :ws r.anual bicek of pewer rative
(\. ' '(lewsetpoint) trip, intermediate ~ range ,
- trip, and -
C 11 blecks source range trip,. '
- end preyt l es a portion 'o f 'P-7 signal) '
i I(NC-41H,4C-42M, NC 43H, WC, HM)- ' ~
F. I P-11 (allows eenval block 6f f afety in. '"
10% of full ;cwer
'j
, t 'jection ac tuation en low pressurt:er [
pressure.
I i
'(See!.1.A. 4above)
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(
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3b. Lew stca:n Ifne pessure (P3 516A, P8 526A, .PB 536A, PS.546A)
- 600 psig Leac time constant (PY-il6B',Pf5253,PY5368,PY5463)
Lag 50 secenes time constant IPY-515B,PY5263,PY5363,PY546B) 5 seconds 3c. Lew< t.ew 3T ,9 (TB L120. T3 4220, 73 4320 T8 4423) 540'F .
4 Autenatic reset of manual block on highpressurtzerpretsure(P11) , ,,
(P3 1558, PR 4563 P5 4578) 1970 Mig
- 5. Contoineent high pressure 5
(PB 1348 FB 9353. P3 9365) 1.!4 pst;
- 6. . Time delay on 5! canual reset
( B.
I minute Steam Lite Isolatten 1
Highsteamline. flow (SetI.l.A.3atye)
- 2. High<high contateent pressure (PS-i 34A, P3 935A, P3 936A P5-937A)
>* D 2.31 ;sig C. Contain.m nt Spray Actuation 1.
\*b*[ p 6[
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(
- 2. Reacter fries A. Nuclear (n.strumentation
- 1. SourTe range high level
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r Figu o 10.11 Neutron Detectors and Ranges of Operation i
+
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) T4,MC nCn, Fry. K r G u [(4 p,y q .c. n 7 6 q j RR.c.o Ga..iFU g cu=ent 1. ga,e 3 gtnerated frc: detecting tu signal enward. 'Where 4;pi ttsble.
equirtrent sh:uld be met with all lead, lag, a,d filter tiec constanti set to 0FF. .
,\.14 Centre 11er Tesisfer Fun:tions yet Appliesble 1.15 Le t:sints .
- 7# I.
Ia 5E 4 toeds 1%s hai lble R4 nce of Setting Y
Intermediate Range High Neutren 5 to 30". full power riux'Reacter I ip -
Bource Range H gh'Neuteen Flux lea:ter Trip '
-10*b t's -10 ". of full ;: er ', g, '
' gl0 V ,
i Interradiate Range Red Withdrawal 5 to 25': a'. full pener . /t i stem (C-1) _ _ = J3 mds t Api
('
4
>6 ,
~10..* to -10.s~% of fu'il ; wer 1 (11 settings w th the exception of time c:nstants shall *.s 7 t t ,'b:w s '.y idjustablewithin their range an'd all time c:nstants shall te (
~
4:ntinu usly a:justable or adjustable in in:rements su:n tha any Let;oint can b ;
e obtained within ?,10% of the set:oint value.
i Jer the P 10 si teeint see Nuclear Pcwer Raage Protection (Oe:
- est 2f.
1.16 locub e ents f: e Test and Calibratien 1 11 protecticr.
(
i hannels.should be su; plied with suff uient redundancy
- c provi,de tF3
- ceMM14 ty,foi' tch, .. ' r.slibrnti:n and tes t at c:wcr.
'n the case of 1/N logic a bypt.
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, . . a 7 , r. , , . g , S .' . . .. *46a
7 *, .QA Record NEP-6.6
- . Attachment 3 v8 Page 1 of @
' M- .
- .;.' .. SAFE 1Y EVALUATION FORM 39 s aatn e
- 1. To - _ Sheet i
- 3. .USQ?
Sequoych Nuc!Oor PIcnt Cab /, TN a Yes 4 Safety Evaluation Number
- 2. Fecm Eca tg/94 F.16 SNP. VN 9. RIMS Accession Number R
Rev No.
Tot DD 5. P' spared Date 0
R SQP B5 of f 7 503
- 6. Reviewed 7. Approved 8. Apod
.' 3 JB25 881114 5 3I O A . A. GNoJ 94 %,Lat(vic t,;.,/VAB //Wl8S' O25 081117 h 7?;
i M~ .
AufLfA 1, t s. sv. v r 2 37 ~7-~M;
. s [#ULL. h.17 7f 3 R 5
4 5
- 10. Project pnd Affected enit(s)
.re.g u a ye l U., , /s I i *t. 11. PMP or DCN Number PMP or DCN Revision
- 13. Other Document Identifier 3////W A) ** <- Date of Docurent
'14 Special Requirements? A)/A See t) Yes (No 15. Potential Tech Spec See SheetNo.MM Ch gr,e ( Yes a No Sheeu No p. ts/
- 16. References (include system number and name as appropriate)
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- 17. Description of Proposed Activity (Change, Test, or Experiment)
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S&fyEvalu/,5 a>. Eco us e s a u t. c AonJ7.toA>Al. JAJroRHA TJoD I F - s* Y e~ tion (s) of System (s) Affected ~ ' e aty functions of the systems affected by this ECN are described POST ACCIDENT MONITORING (PAM) The safety function of the post accident monitoring system is to ' provide information on plant variables required by control roora operating personnel during accident situations tot i
- 1. permit the operator to take preplanned manual actions to ;
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- . determine 't'he potential for causing gross breach of the barriers to radioactivity release and to deternine if a gross breach of a barrier has occurred.
4 assess the'cperation of plant systems to make appropriate decisions as to their use. 5. allo'v for early indication of release of radioactive materials in order to initiate action necessary to protect the public and estimate the magnitude of any impending threat. Tle. ,cA h ya ,, b lfs, r Q & jy f h.s M i.s n e uVo"n
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4 s v .h/ hk u :q <. n e.u f r.* , ~- , o toe o3 sa Ny Intermediate range high neutron flur trip - The intemediate range high neutron flur trip circuit shall trip the reactor the tripwhen one out of the two intermediate range channels exceed setpoint. This trip, which provides protection during - reactor startup, can be manually blocked if two out of four power Three out of the four power range channels below this valueran automatically trip. reinstates the intemediate range high neutron flur be separate from the power range channels.The intemediate shall range channe The intemediate range channels can be individually bypassed at the nuclear 'nstru entation racks to permit channel testing at any time under prescribed administrative procedures and only under the direction of authorized supervision. board. This bypass action shall be annunciated on the control Source range high neutron flur trip - The source range high neutron fluz trip circuit shall trip the reactor when setpoint. one of the two source range channels ezeeeds the trip This trip, which provides protection during reactor startup and plant shutdown, can be manually bypassed when one of the two (source intemediate range range cutoff channels power level)roads above .the P-6 setpoint value and shall be automatically reinstated P-6 value. when both intermediate range channels decrease below the four logic from the power range permissive (P-10).This trip shall be autom This trip function shall also be reinstated below P-10 by *.n board mounted switches. administrative action requiring manual actuation of two cont - fynction in une of the two protection logic trains.Each switch will reinstate the trip The source range trip shall be cet between the P-6 setpoint and the maximun nource range level, the channels can be individually blocked at the nuclear instrumentation racks to permit channel testing at any time un.ter prescribed adainistrative procedures and only under the direction of authorized supervision. This blocking action shall be anauociated on the control botrd. l v7hw**,u- eu n ...,./e..5-,,,,
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The Reactor Protection System la by definition a primaryem safety syst . its requirement to shut down the reactor and maintain it la a s fdue to whenever a possible dangerous situation erists. a e condition ' The functional performance requirements of the Reactora Trip System s includo provisions for automatically initiating a reactor trip: 1. Whenever (Condition II). necessary to prevent fuel damage for an ianticipated s ent t an i
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- 21. evaluatedWould the proposed activity increase the probability of a in the SAR? n accident previously O Yes # No J_us t i fic a t lon:
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- 22. Would the proposed activity increase the c.a y aylu <. ,
Lasokiv . _an ac=ide. f *e ryJM s<ssss**y evaluated in the SAR7 onsequences of an accident previously, 0 Yes il No Justificatlon_: 7~le ' s . u em ., a/ < $ 4e. .e /im /<. ra a)e adv/e**1 e
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- 23. Would the proposed activity increase the probability of a malfuncti on of equipment important to safety previously evaluated in the SARf f) Yes # No Justification:
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- 24. important Would thetoproposed activity evaluated safety previously increase in thethe consequences SARf of a malfunct.lon of egulpm O Yes V No Justification: . . - . -
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- 25. type Would thantheanyproposed activity create evaluated previously a possibility for an accident of a different in the SAR7 0 Yes W No .
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- 26. Would the proposed activity create a possibility for a malfunction of equipment of a different type than any evaluated previously in the SAR7 0 Yes (( No Justification:
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p u d. 4 S l a .-f 1 2. ,j s J a4.fy Eve / val,.a TABLE 3.3-1 (Continued) #- # TABLE NOTATION tor trip system breakers ~in the closed position tea capable of rod withdrawal, and fuel and in the r the control eactor vessel.
) associated with the protective functions e deriv d f t
actor Coolant Loop shall be placed in the rom tripped the out on, conditi
.0.4 ar'e not applicable, to Jr!aldetector atPoiat may,._ be de energized aboveock the P-6 (Bl of Source
~
ACTION STATEMENTS Oth the number of channels OPERABLEy one less than required b ne Minimum Chunnels OPERABLE requirement, - restore the inoperabl 1annel to OPERABLE status within 48 hours e ithin the next 6 hours and/or open or be the in HOT reactor STANDBYtrip br eakers, th the number aber -of Channels STARTUP of OPERABLE and POWER OPERATION channels oa may proceed one less than the T evided the follow,ing cond',tions are satisfied: ,, The 6inoperable within hours. channel is placed in the tripped condition ~ ' lR51 The Minimum Channels OPERABLE requirement is met; howev er to 4 hours , one additional channel may be bypassed for up for surveillance testing per Specification . 4.3 1 1 1 lR51 to 75% of RATED THERMAL and the Power RangeEither,
, :.eutron Flux high trip POWER THsRMAL reduced to less within than 4 hours; or or equal to 85% of RATED RnTIO is monitored at least once,per 12 hoursthe QUADRANT POWER TILT
- The QUADRANT POWER "ILT RATIO, nas ng indicated by the remaimi three detectors symmetric is serifiedobtained power dist.aibution consistent byzed using with thethe normali movable incore detectors in ti,e four pairs s of 'ymmetric thimble a; locatiens greater than 75%at ofleast RATED once per POWER.
THERMAL 12 hours whens THERMAL POWER i '"
,. . EE211LE l'
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, S,* 0,*.e wb e r 1/ , 19 8 6 3/4 3;5
.' Amendment No. 47 -
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TABLE NOTATION
=
With the reactor trip system breakers in the closed position, the control rod drive system capable of rod withdrawal, and fuel in the reactor vessel. The channel (s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped condition. he N revisions of Specification 3.0.4 are not applicable.
~
High voltage to detector may be de-energize ~ Range Reactor Trip) setpoint. 6 (81'ock of Source ACTION STATEMENTS f&lETG ACTION 1 - With the number of OPERABLE channels one less than re utred by the Minimum Channels OPERABLE requirem nt, restore th inoperable channel to OPERABLE status within 48 hours or be in HOT STANOBY
' within the next 6 hours and/or open the reactor trip breakers.
ACTION 2 - With the number of OPERABLE channels one less than the Tot Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
- a. The inoperable channel is placed in the trioped condition within 6 hours.
lR35 b. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 4 hours lR39 for surveillance testing per Specification 4.3.1.1.1. c. Either, !!iERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range, Neustron - Flux trip setpoint is reduced to less than or e 85% of RATED THERMAL .e0WER within 4 hours; the or, qual to . i QUADRANT POWER TILT RATIO is monitored at least once per 12 hours. d. The QUADRANT POWER TILT RATIO, as indicated by the remaining three detectors is verified consistent with i.he normalized symmetric power, distribution obtained by using the movable incore detectors in the four pairs of symmetric thimbic - locations at least once per 12 hours when THERMAL POWER is
- greator than 75% of RATED THERMAL POWER. '
.e
- 4 l
September 17. 1986 ' SEQUOYAH - UNIT 2 3/4 3-5 Amendment No. 3? - e. i , t b ' ,
,g.
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w,
- a. sce t.//W
- i E ,
TABLE 3.3-1 . o
) $ REACTOR TRIP SYSTEM INSTRUNENTAM -
i
# * "i * ^-
FUNCTIONAL UtsIT TOTAL NO. MINIMUM CHANNELS CHANNELS
- OF CHANNELS
-TO TRIP APPLICA8tE
- 1. _
OPERA 8tE Manual Re.setor Trip MODES _ ACTION
" 2 _
- 2. 1 2 Poe.er Range, Neutron Flux 1, 2, and
- 1 4
- 3. 2 3 Power Ran;e, Neutron Flux 1, 2 R 2
High Positive Eate 4 2 3 1. 2 # 4 ~ 2 Power Range, Neutron Flux, High Negative Rate 4 2 3 1, 2 # R 5. 2 g.
~
Inter. mediate Range, deutron Flux 2
- Y 6. 1 2
"* Sourc,e Range, Neutron Flux 1, 2, and a 3 A. Startep
- 8. .ShutdoJn~ 2 ~ j)gcETE 1
2 2 , and *
- 7. 0 1 4 Overtemperatere Delta T 3, 4 and 5 5 Four Loo ~p Operation 4
2 i,.
- 8. Overpower Delta T 3 1, 2 6, 3,
.- Four Loop Operation :
4 lR45
$. 9. 2 3 1, 2 6, h Pressu izer Pressure-Low 4 D. 2 3 l R45
- 10. 1, 2 #
?
II" Pressu.-izer Pressere .'Ifgh 4 6
$% 2 3
- 11. 1, 2 9 Pressurizer Weter Leve!--High 3 6
i
.E.E.
De 2 2 f* M
- 1, 2 7,
= ! T ba ! y- .:
ii .G. e.* e '} _ . 2,-
}4
*O.
- v. . -
*1~
Q .
'* ,rp , em >e *$~*>**' N;
,.s>
*'o*' *
- \ ~
nGf Li S$sef.25 l Safedy too/w.bl , p.. Eco c cret
- v. TABLE 3.3-1 M
g .
- g. ,
REACTOR TRIP SYSTEM INSTRUMENTATION
~
^
MINIMUM FUNCTIONAL (; NIT TOTAL tiO. CHANNELS CHANNELS APPLICAB'.E
- .- OF CHANNELS TO TRIP OPERABLE MODES ACTION U 1. Manual Reactor Trip
, u 2 1 2
' 1, 2, and
- 1
- 2. Power Range, Neutron Flux 4 2 3
,2 2
- 3. Power Range, Neutron Flux 4 2 #
High Positive Rate 3 1, 2 2
- i. . .
- 4. Power Range, Neutron Flux,
- 4 2 3 High Negative Rate 1, 2 2 u
g 5. Intermediate Range, Neutron Flux 2 1 '~ 2 1, 2, and
- 3.
4 6. Source Range, Neutron Flux A. , Startup 2 , d B. 1 2 , and
- Shutdown 2 4
'.- 0 1 3, 4 and 5 5
- 7. Overtemperature AT
', ' Four Loop Operation 4 2 5 6,
. 1, 2
- 3. Overpswer si R33
- c. . Four Loop Operation 4
).
4g. e 2 3 1, 2 6 M i
?
- aE "s ?. Pressurizer *ressure-Low 4 l R33 2 u C. aa 3 1, 2 6
[, , 4 $ 10. Presstrizer Pressure--High 4 2
':- u 2 3 1, 2 6 o 11. Pressu-izer .'ater Level--High 3 y
- , 2 2 1, 2 7
- G.
c
%Je y
, f f ..-
h - . y
. l ** :
m[_. V d 5
Alu 1./ JSea f 1$ 'f A% Safe.ly Evabal$ , n.. e<.a atai
- m _ TABLE 3.3-10 ACCIDENT MONITORING INSTRt#tENTATION MINIM 0M JNSTRUPENT g
REQUIRED NO. OF CllANNELS CHANi4ELS
., 1. OPERABLE Reactor C:olsac Tg (Wide Range)
2
- 2. Reactor Cao'iang T 1 Cold * '"9'}
- 3. Containzer.'. Pre.ssure (Wide Range) .
- 4. 2 1 Refyling Water Storage Tank Level lR5
- 5. 2 1 Rear.ror Cc,olant Pressure (Wide Range) 2 1
' 6. ' Pressurizer, Level (Wide Range)
- l RSI
. 7. Steaa Line Pressure 2 1 u 8. 2/ steam line 1/ steam line N
- Steam Generator Level - (Ulde Range) ,
- 9. S tear.: 1/ steam generator 1/ steam generator Generator Level - (Narrow Range) m 10. Aux 111ary feedwa.ter Flow Rate 1/ steam generator 1/ steam generator ,
1/ pump 1/ pump
- 11. Reactor Cooltnt System Subcooling Margin Monitor 1
- 12. Pressurizer PORV , Position. Indicator
- 0 2/ valve # 1/ valve
- 13. Pressurizer p RV. Block Valve Position Indicator **
- 2/ valve 1/ valve
- 14. Safety Valve ~Posi, tion. Indicator
-
- 2/ valve # 1/ valve
*15. Conta(neent hyter Leve! (Wide Range) 2 1 > 16. In Cora Therm > couples 4/ core quadrant 2/ core quadrant
- 17. Reacts Vesstl level Instrumentation System *** 2
& 1
- . .N 5 _: ~- - r '
&g faas L lR50
/
TNot applicable if tt'e associate;$ olock valve is in th e closed1 - 2 cr position. 34 *"Not applicable ff the block valve is verified in the closed position with power to tne valve cperator removed. ~ z-P* **"This Technical 5pecification and surveillance
. Instructbns are developed for the use of this system as committed to in the TVA res
PUREG-0737.
requirement will not be impl R50 } -$U # . o Supplement 1 of ? ? . .At least ene channel shat! be the acoustic monitors. ( .'.
.v .
L-w .- . .> ~ .
w n ~ . _ _ _ _ _ p n G-Algf f.f S Uf TABLE 3.3-10 O* AJT E"'!/# N Ecs) t. h8 C
-2 o ACCIDENT MONITORING INSTRUMENTATION
^*
-E MINIMUM .
REQUIRED NO. CHANNELS INSTRUMENT e
, OF CHANNELS OPERA 8LE L
{ 1. Reactor Coolaret THot ( Ide Range) 2 1
. e. '2. Reactor Coolant T I Cold I'"9')
- 3. Containsect Pressure (Wide Range) 2 1
- 4. Refueling Water Storage Tank Level l1 2 1
- 5. Reactor Coolant Pressure (Wide Range) 2 1 1
- 6. Pressurizer tevel (Wide Range) l3 2 1
- 7. Steam Line Pressure 2/ steam line 3/ steam ifne
, 8. Steam Generator Level - (Wide Range) 1/ steam generator 1/ steam generator 1 9. Steam Generator Level - (Narrow Range) 1/ steam generator 1/ steam generator
~; Y 10. Auxiliary Feedwater Flow Rate
.- u. 1/ pump 1/ pump
" ,q 11. Reactor Coolant System Secooling Margin Monitor 1 0
- 12. Pressurizer PORV Positio.~.14dicator* 2/ valve # 1/ valve
- 13. Pressurizer PORV a, 1 Vaive Position Indicator **
2/ valve 1/vaiv2
.- 14. Safety Valve Po,- im Indicator 2/ valve # 1/ valve
*15. Cantafrurent Wa4 + i.s A (Wide Range) 2 1
, 16. In Core Thernoct . .. a .
4/ core quadrant 2/ core quadrz .
- 17. Reactor 'lessel-.
Level Instrumentation Systema "a 2 1 R
/8 Sos **.Rl 7-% m- cof.aJ Aa g_ N.oc1w. j,,,j,,,4 1,;., x _
F .m - U
+
?
$3
- Net applicable if the associated block valve is in the closed position. l
. @$ **Not applicable if the block valve is verified in the closed position with power to the valve operator removed.
3 7 ***This Techrical Specification and surveillance req. irement will not be implemented until Sequoyah Specific Instructio?s are deve*oped for the use of this system as committed to in the TVA response to Supple. ment 1 of R 2 . ~ f .~. NUREG-0737. 4
- ,At l east ont channel shall be the, acoustic monitors.
~
s 4- p ,
. . S*,
y .;. ' f. c .; . u . .:
- A )E P l. t J lss i ;L t l .t L~
.raSQ evale:/4 4
Et*-Al lfild M . TABLE 4.3-7 E . 8 '. {
^ ACCIDENT MONIT0 KING IMSTRUMENTATION SURVEILLANCE
.g -INSTRUAENT -
CilANNEL CilANNEL 'U CilECK E 1. CALIBRATION
' Rea,ctor Cqolar.t T,gg (Wide Range)
- 2. M
] Reactor Cqolact T R
^
Cold d* N'"9*)
- 3. R l Containeer:t Pressure (Wide Range) -
4 M R Refjeling 'Jate- Storage Tank Level M lRT J. R Ree: tor Coolant Pressure (Wide Range) t. M R Prefsurizer Level 7 M lR2 Stease Lire, Pressure R
- 8. M R Steam Generator Level - Wide
- 9. M L
{ Swam Generator Level - Narrow . M R g., 10. Aux 1,11ary Feedwater Flowrate d '.1., M R Reac*cr Cochat System Subcooling Margin Monitor M
=
R
- 32. Pressurizer PORV Position Indicator M _'
R
- 13. ;Fressurizer PORV Block Valve Position Indicator K d
~ i4. Safety Valve Pos1 tion Indicator n R
{.S. Containment'WaterLevel(WideRange) i M x s 26.InCe[eTrer;nocouples R
*t M IR50 3 0p R
g !. 17. Reactbr e Vessel L2 vel Instrumentation"* M
- t. . ~ R g, 3. .taara- {f k edenk-y- - ~ sieye JJac.$w hsfro* <nla/* 14
~
lg$o y 3 } "*"r.'s i S s. Tecanical Specification and surveillance requirement will not be inga~nented until S
< $= Irntr:;tions Ni'4EO-0737. are developed for the use of this system as concsitted to in *he TVA res pecific g3o pcase in Suppiement 1 of
~
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~ AE/ t.( 5f Y. 2E
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TA8tE 4.3-7 A.se . sc.n unc : w - ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS h INSTRUMENT
' CHANNEL CHANNEL
. CHECK CALIBRATION
_E 1. Oesetcr Coolant Tg (Wide Range) M R
] 2. Keactor Cool,snt T
*
- Cold e av) M R
- 3. Q ntainment Pressure (W!de Range) M R
- 4. Refueling Water Stc. age Tank Level ht38 M R
- 5. Reactor Coolant Pressum (Wide Range) M '
R
- 6. Pressurizer Level ht38 M Q
- 7. Steam Line Pressure M R
, J. Steam Generator Level - (Wide) M R Y
s 9-- Steam Generator Level - (Marrow) M R
<a 10,. Auxf 7.f ary Feedweter Flowr-te M R
E 11. Reactor Coolant. System Subcooling Margin Monitor M .R
- 12. Presssrizer PORV Position Indicator M R
- 13. Press'arizer PORY Block Ys1ve Position Indicator M R
- 14. Safety Valve Position Ind!cator M R
- 15. Contajnsent Water Level (Wide Range) M R 16.,In Co;e ^re-mocouples h38
~ M R [ ,, r _171 Reacto 1 Instrumentation System
- M R 1**ns : f-t<~ l n-Q /9 A4. % nd : s4,m f.1,an n _n p3s
-~ ~ e n -g y y? w _- w -
f .
$3 i d g. "This Techr.ical Specification and survefilance requirement will not be impienented until Sequoyah Specific
<3g 3a Instructions are developed for the use of this system as committed to in the TVA response to Supplement I of NUREG-0727 i a 'S ,
o . e , b 5 .:; . .
- ,/. - - . - .. ; ,1 ..s. . .. . .
I A)U J.4 JLea.] 30"[$1.
.sm% s./,sa, ,. .
g - Ala . ECAIt.//8f' TABLE 2.2-1 (Continued) 5 . REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOIMS _FUNCsIONAL JNIT g . . _ TRIP SETPOINT
; 13. Eteam Generator Water All0WABLE VALUES y
level--Low-Low 118% of narrow range instrument span each steam generator 1 17% of narrow range instrument R20
. span each steam generator
- 14. Steam /feedwater Flow Mismatch.and Low Steam < 40% of full steam flow at Generator Water level RATED TitERR'.i. POWER coincider.t < 42.5% of full steam flow at with steam generator water level RATED Tt!ERMAL POWER coincident 1 25% of narrow range instru- with steam generator water level ment span- each steam generator 1 24.0% of narrow range instru-
- 15. Ur.dervoltsge-Reactor ment span- each steam generator Coolant P xps 1 5022 volts each bus 14739 volts each bus y 16; Underfregeency-Reactor ha9 I
e Coalant Pumps 1 56.0 Itz each bus 1 55.9 Hz each bus
- 17. Ttc.bine Trip A.- Low Tr;p System Pressure 1 45 psig B. .Turbir.e Step Valve 1 43 psig Closure -> 1% open
-~> 1% open i3
,. 18. Safety Injection Input from ESF Not Ap ic3hle h{
> txto -s ,I*
pplicable '9
- 19. Inte.rmediai,3 Raage Neutron it 4 x to -s %
" 1m 10' v3 5 'S: 10 re'
",, e Rang a r r W^ * " " ## N e, # ##
g' s,, # M 20, Power Range Neutron Flux [ "[ < 10% of RATED (not ?-10) Input to Low Power < 11% of RATED
.% ,o Reactor Trips Block P-7 THERMAL POWER N
THERMAL POWER qD M ,' i
- co O3 *.
tis'. P * =
; e ., o f 4'8
,o 4, , ((A,.N4 -M#- O
, _ . _ __. _ - - , - - _ _ _ __ - . _ _ _ - - - , - - - _ , - _ _ . - - - - - _ _._ . -__ , - - %:1 _ _ _
/JE9 C S ' S W
- 3. f '-
p Satf faLa 4-:. , g t a.. m au
.o TABLE 2.2-1 (Continued)
_FUNs.T5ONALUNIT REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS TRIP SETPOINI All0WABLE VALUES L, OE 13. Steam Generator Water Level--Lov-Loi 118% of narrow range instrument 0*J span each steam generator 1 17% of narrow range instrument '
- 14. Stram/Feedwater Flow span each steam generator i, t
< 40% of full ster.2 flew at Mi watch aed low Steam Generater .iater Level RATED THERMAL POWER coincident < 42.5% of fu11 steam flow at with steam generator water level RATED THERMAL POWER ceincidut
- 1 25% of narrow range instru- with steam generator water level ment span- each steam generator 124% of narrow range instru-
- 15. Under: voltage-Reactor ment span- each steam generator l
Coolaat Pumps 1 5022 volts each bus K to 1 4739 volts each bus i
- 16. Underfrequency-Reactor 4 7
os Coolant. Pumps 1 56 Hz each bus 1 55.9 Hz each bus -
- 17. -Turbine Trip A. Lov Trip System Pressure 1 45 psig E. 1 43 psig Turpine Stop Valve
- C1csure 3 1% open
> 1% open
,; 18. Safety 'njection Input Not Applicable from E57 - Not Applicable 2 i x. e c g u.
- 19. Intersec.iate Range Neutron M-x-10 ' O m 2 c, s io %
Flux, P-5 Enable Block ~
' g:
."y Source Penge deactor Trip og 2g eo n.e = % We u. ' S x 10 ow.
p n, , Pr ;
* 'g- 20. Power Range Neutron Flux
& Jr Aw me. Re !
e , t < 10% of RATED (not P- N) Input to tow u F- < 11% of RATED THERMAL POWER Pc.*er Reactor Trips THERMAL POWER i, '- "
. y Bisck P-7 4
.M ll' .
. J m
)) * . b .
"5 y
! y 1 %. ?-
~
& ye lpo qp %. . , - * * * * ,
1 _ . . _ _ _ . . _ . _ _ . _ , _ _ . _ , _ . ~ , - . _ _ _ . . . _-. _
, . _ _ . _ _ _ _ . _ . _ m . _ _ . . _ . _ _ .
- i g y(,{,_
O NeP(.,4 Cwser # 31 At * -
, - r:
[ '. . . . .
%ssvy Evat.unsou
..y. ., . .; . .y g* "
MO ECn LC18(a '~ r'.'
,- m 3, ;- s' TABLE 3.3-9 4
I o i. ~ - -
-> { '(
~
.g. .
hh, %- i REMOTE SHUTDOWN MONITORING INSTRUMENTATION ) 1 T. .
$. 78 -
MINIMUM READOUT INSTRUMENT _ LOCATION MEASUREMENT RANGE CHANNELS f ji T w 1. 7 eoveerrscATE OPERABLE Tl U > Soucce Ran d uclear Flux -- 2.
, y - NOTE 1 1 to 1 x 106 7 Reactor Trip Breakr Indication at trip switchgear ' OPEN-CLOSE
/si' w zoo % CTP
- 3. 1/ trip D aker R'eactor Coo! ant Temperature -*
NOTE 1 0-650'F
, Hot. Leg .
1/ loop lR80
,4
, , Pressurizer Pressure NOTE 1 0-3000 psig 1 g
- 5. Pressurizer Level ' NOTE 1 0-100% '
1 y 3 6 Steam Generator. Pressure lR80
- w. NOTE 1 0-1200 psig
- 7. 1/ steam generator "Steam Generator Level NOTE 2 or near Auxilary F. W. 0-100%
Pump 1/ste'an generator
- 8. Full Length Control Rod Position Limit Switches Auxilary Instrument .
Room: Racks R41-44 On-off 1 insertion limit 9.. RHR Flow Rate switch / rod
. .N0TE 1 0-4500 gpm 1
- 10. RHR Temperature NOTE 1 50-400*F 1
, 11. Acxiliary Feedsater Flow Rate n c h, NOTE 1 0-440 gpm 3 . 1/ steam generator R80 '!
U -
- =
p
. *~
25
^2 .
- 2. .
g .: a 1 y
. .~.l f, ,
3,. .,a u. .. 3'. -
,, ;3
- .cb [j ,
h MEP Cc6 Susur$f- gf-
~ . ,)_ . '%* . . . '.p;:ni-
- y - .z s
EW U"T'*# p oa EC.H LLibG jf. .,I I'
.>.- TABLE 4.3-6
., m . ,,.
, s<. - ..---
- ~5f 5 .,'
c .
-< .i.i .
-.-.vnn..! HONITORING INSTRUMENTATION
. g SURVEILLANCE REQUIREMENIS IJ -
~ INSTRUMENT '.
CHANNEL CHANNEL
@ T h.T& M2ca1G"
- CHECK CALIBRATION
-e nL. 1. e Soerce Rang
' Juclear Flux
.~. s H
~
R
- 2. Reactor Trip Breaker Indication u H H.A.
- 3. Reactor Coolant Temperature - Hot Leg H R
- 4. P'ressurizer Pressure H R
- 5. Pressurizar Level H .
R
- 6. Steara Generator. Pressure M
H R -
- 7. Steam Generator Level T
H R
$ 8. Fulh length Con.rol Rod Position Li iit Switches M R
- 9. RHR J10w Rn e H R
- 10. RHR icepera<ure H R
- 11. Auxi?iary Ferdwer Flow Rate H R
, 12. Press .rrizer .telf 2f Tank Pressure - H R yg
- e. .
17 ntainment Pressure : H C 3 R et 3 l . e.
"w u,
. I
.- y
- a. .
e 1, . .. . . ;. h , . a . . . . . .~
.'-4 ,
r
.=~ ,. .
' ,* { .
;* ~
l Ntt Ce(o SHEET 3b $f
,q..
- .i. .
~
SA.amTy E.VAtoATswo t - f kb
~
_ TABLE 3.3-9 E m LG(6s i ;- REMOTE $HUTDOWN MONITORING INSTRUMENTATION
~~
C hbb ' e . z . READOUT INSTR EENT ',' MINIMUM
-4 MEASUREMENT J
. LOCATION
- ti e tMd t. _. RANGE CHANNELS
. 1. ~four-f ange Nuclear Flux OPERABLE NOTE I
- 2. 1 to 1 x 106 cp ,
React / Trip Breaker Indication 304i.3@7 3T gP\ at trip switchgear OPEN-CLOSE
- 3. ' Reactor' Coolant Temperature - 1/ trip breaker Fpt. Leg NOTE 1
. 0-650'F 1/ loop 067 4.
Pressurizer Fressure NOTE 1 0-3000 psig g - 5. Pressurizer lavel 1 NOTE 1 y 0-100%
- 6. 1 -
Steam Generator Pressure NOTE 1 a67
- 7. 0-1200 psig Steam Generator level 1/ steam generator
- NOTE 2 or
; near Auxilary F. W. 0-100%
Pop 8 Full Length Control Rod
- 1/ steam generatcr
. ,Positipn Limit Switches Auxilary Instrument Room: Racks R41-44 On off 1 insertion limit 9' .
RHR Flow Rate switch / rod NOTE 1 I, 10. xHR Temperature 0-4500 Spa 1 1 -
*- NOTE 1 50-400*F u 11. 7.uxilitry Feedwater Flow h.te 1 c$ NOTE 1 G L*. 0-440 gpa 1/ steam generator a67
,- n .G. -
*3 w ",
. .O %
0 . y.. V V '
) ._
.t
- e
- i. .. . ',i~ :q -: . m c <. su r. g .n yf_r
> i
- 4~
t } ...d,. ~ ~ s.=w e
,,so.-
I '" 5 ~ I,
. G. .
TABLE 4.3-6 *
"*IW ,.'
Ip
~
RDm)TE SHUT 00WN MONITORING INSTRUMENT"ATION
-< 6 .
5 SURVElttANCE REQUIRENEN15
. \~
i
~
e
.g , . . n s .. ; c -
i CHANNEL ClANNEL I- 3 INSTRUMENT ) .. CilECK
- Call 8AAT10N -
w m ue**sra
- m. .
. 1. Source Aas* uclear Fium M R
\; ; *
. 2. Reactor Tr' Breaker Indication M N.A.
. 3. Ce.rcter Cor,] ant , Temperature - Het Leg . M~ R
~
~
- 4. Pressurizer,. Pressure ; M R
' ~
, 5. Preswrizer Leve) -
M R
- u N
A.
- 6. .
Steae,Genert, tor Pressure M R u - e ~
. =
s
, 7. Steam.Gener , tor Level -
M
. .- R
~
- 8. Full Length i:ontrol Rod Posttien Limit Switches M R*
l l
- 9. RHR Flow Rate -
M R
- 10. RHR.Tegeratar e : -
M A . ,g li. Auxili.ary fet,Asat:r F*ow Rati M A
~.
.t c ,f gp __.
~ ?- N di' 10 Presse.-izer R 11e* Tad Pressure
- a. M -
R ,~ n -
- - ,a, ? 13. Contal.ieent Pressure . M R a e .
!6 o **
;_ ~
- Fo: cycle 1, this surveillance is to be completed before the next cooIJwn or by A6 gust 5.1983 whichever is earlier. R20 i[I
! . \ ,~ 4 lh j
.N R, '
- k. i. . i. . .. . ... u . *U = -
~~
, TV.A 10551 (EN DES 2 81) UNHEVIEWED SAFETY QUF.STION DETERMINATION Skui 3 6 gi SHEET #1 TO: Secuevch Nuclear Plant. DOisy, TN 061%Ro
- nou *. El6 SNP. """'"
MEDS ACCESSION NO. oT D^TE DP pp PREPARED R EVIEYvEO APPROVED APPO q E S@ 'M 0117 503 M AP3 l i[, .
<l 1
\ R k b '
\ Y '. "
\ \ #JJj : .
5
\ FA R
'R 'INITI AL l$
PAOJECT__6 G N AFFECT NIT (5) ld ".1 ECN NO. Md.b ECN DATE -} Al' b 7g ggg gj , YE S/NO SHEET NO. McM1DCR NO 1 15 c. _.DATE t. 2VIREMENT(5) Mo OTHER DATE _ i POTENT CHANGE [f5 O
/" 1ErERENCES Llb S Sup nrI -6bre bbt CnnjfirmNe n oi+ L GEP e le c derr[H x - n N
DES (.RIPTION OF CHANGE Cib N e r* 1 _ __
$ .LL rade. f.'ume < < e . ut e el b e41c_~ en v, t
$ me Chu_.1F tab re t e. e ceb:.bt o e_<j o
$ g Guide i.95,brd hie d.h bond & F ' meme
' ~
M .mer.e re U dle pmfuc;clent men. c
- n. 'n.R J etm+ie. d. _ [
I
$ Vi rnt u>i'l l e ui r .. 130-V ClamM
_ d u;e l l dra
.d[Pf_O MR }
4 Mp M # (dfRfpf 4 50 b f_rdiPorsmEnf4ll
. Q t!
ftr ff ta k t h e r " l
. . . . . . . . . . - ~ . . . .
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UN AEVIEWED SAFETY QUESTION DETE AMINATION LGIS&R.o ) (. Unreviewed Safety Questiont lDENTIFIER I
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ENCLCSURE 3 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 ; DOCKET NOS. 50-327 AND 50-328 ' (IVA-SQN-TS-L8-42) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS r i O
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e ENCLOSURE 3 Significant Hazards Evaluation TVA has evaluated the preposed technical specification 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 SQN in accordance with the proposed amendment will nots (1) Involve a significant increase in the probability or consequence of an accident previously evaluated. The three administrative changes - are proposed to support the installation of the Camma Metrics source range (SR) and ir,termediate range (IR) detector assemblies. The first involven the deletion of a table note that is not applicable to the design of the new SR detectors. The second involves a change in engineering units for the P-6 setpoint that results from the difference in output signals from the IR detectors. The third involves the addition of a certain footnote to enable the review and approval of the unit 1 change to proceed independently of the unit 1 installation schedule. The new SR/IR detectors are class-1E equipment that is seismically and environmentally qualified and compatible with the present design requirements. Because the new hardware is compatible with the present design requirements and the proposed technical specification changes are administrative in nature, the proposed amendment will not involve a significant increase in the probability or consequences of an accider t previous 1/ evaluated. (2) Create the possibility of a new or different kind of accident from any previously analyzed. The three administrative changes are proposed to support the installation of the Gamma Metrics SR and IR detector assemblies. The first involves the deletion of a table note that is not appilcable to the design of the new SR detectors. The second involves a change in engineering units for the P-6 setpoint that results from the difference in output algnals from the IR de tec tors. The third involves the addition of a certain footnote to i enable the review and approval of the unit 1 change to proceed independently of the unit 1 installation schedule. The new SR/IR detectors are class-1E equipment that is seismically and environmentally qualified and compatible with the present design l requirements. Because the new hardware is compatible with the present design requirements and the proposed technical specification I changes are administrative in nature, the proposed amendment will not create the possibility of a new or different hinJ of accident from 3 4' any.previously cnalyzed. (3) Involve a significant reduction in a margin of safety. The three administentive changes arv proposed to support the installation of the Carna Metrics SR and IR detector assemblies. The first involves , the deletion of a table note that is not applicable to the design ,
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of the new SR detectors. The second involves a change in engineering units for the P-6 setpoint that results from the difference in output signals from the IR detectors. The third involves the addition of a certain footnote to enable the review and approval of the unit I change to proceed independently of the unit 1 installation schedule. The new SR/IR detactors are class-1E equipment that is seismically and environmentally qualified and compatible with the present design requirements. Because the new hardware is compatible with the present design requirements and the proposed technical specification changes are administrative in nature, the proposed amendmcnt will not involve a significant reduction in a margin of safety. ' 4 t
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