ML19354D868
| ML19354D868 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 01/12/1990 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML19354D867 | List: |
| References | |
| NUDOCS 9001230056 | |
| Download: ML19354D868 (101) | |
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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 B 2-4 3/4 3-5 Unit 2 2-6 B 2-4 3/4 3-5 1
9001230036 900112 PDR ADOCK 05000327 P
A:r TABLE 2.2-1 (Continued) i E
l g
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS
- t 5
x FUNCTIONAL UNIT TRIP SETPOINT ALLOWA8LE VALUES c
t k
- 13. Steam Generator Water 1 18% of narrow range instrument 1 17% of narrow range instrument R20 Level--Low-low span-each steam generator span each steam generator y
- 14. Steam /Feedwater Flow
< 4'0% of full steam flow at
< 42.5% of full steam flow at Mismatch and Low Steam RATED THERNAL POWER coincident
. RATED THERMAL POWER coincident Generator Water Level with steam generator water level with steam generator water level 2 25% of narrow range instru-1 24.0% of narrow range instru-ment span--each steam generator ment span-each steam generator i
- 15. Undervoltage-Reactor
> 5022 volts-each bus 1 4739 volts each bus h89 i
Coolant Pumps i
I ty
- 16. Underfrequency-Reactor 1 56.0 Hz each bus 1 55.9 Hz each bus Coolant Pumps m
- 17. Turbine Trip i
A.
Low Trip System 1 45 psig 1 43 psig j
s Pressure B.
Turbine Stop Valve 1 1% open 1 1% open Closure i
- 18. Safety Injection Input Not pplicable Not Applicable _
e=
from ESF g
(
- g, g-s o$ ef OTED THE6ndL
-10 Aaq
] g = 79-11 AM
- 19. Intermediate Range Neutron t I d 39 d
Flux - (P-6) Enable Block,
f
%g Source Range Reactor Trip ER lfs
- 20. Power Range Neutron Flux
< 10% of RATED.
< 11% of RATED
"[
(not P-10) Input to Low Power THERMAL POWER THERMAL POWER I
M,o Reactor Trips Block P-7 W
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't SAFETY LIMITS BASES n
Range Channels will initiate a reactor trip
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7rt M L approximately 25 percent of RATED THERMAL P unlessmanuallybiockedwhen P-10 becomes active.
No credit was taken for o n of the trips associat with either the Intermediate or Source Range Channels 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.
~0vertemperature 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 and Low Pressure reactor trips. This setpoint includes corrections for axial power distribution, changes in density and heat capacity of water with temperature and dynamic j
' compensation for piping delays from the core to the loop temperature detec-tors. With normal axial power distribution, this reactor trip limit is always a
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.
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 trWunctions as a High Neutron Flux trip at the reduced power invel.
Overpower Delta T l
The Ov'erpower Delta T reactor trip provides assurance of fuel integrity, e.g., no melting, under al.1 poss.ih.la.nurnower. conditions,14mits the required mege for.0vertemperature Delta T protection, and provides a backup to the High Neutron Flux trip.
The setpoint includes correction fw.1anges in densitneA& Mat revMr af wt?ardtirtemperature', and dynamic compensation for piping de' lays from the core to the loop temperature detectors.
No credit was taken for operation of this trip in the accident 3
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s TABLE 3.3-1 (Continued)
TABLE NOTATION aWith the reactor trip system breakers in the closed position and 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 shs11 be placed in the tripped condition.-
provigions T
..... x.t:;........rmay be 'jification 3 0 g ' not applicable.
': =r; -d above the P-6 (Block of Source Range Reactor Trip) setpoint.
ACTION STATEMENTS 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 th'e number of OPERABLE channels one less than the Total Number of Channels, STARTUP 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 />.
lR51 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 Power-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 incore 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.
e e
September 17, 1986 SEQUOYAH - UNIT 1 3/4 3-5 Amendment No. A
2 TABLE 2.2-1 (Continued) u, E8 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS s
- =
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES E 13. Steam Generator Water 1 18%~of. narrow range instrument 1 17% of narrow range instrument
'a7 Z
Level--Low-Low span-each steam generator-span each steam generator to
- 14. Steam /Feedwater Flow
< 40% of full steam flow at
< 42.5% of full steam flow at Mismatch and Low Steam NATED-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-
> 24% of narrow range instru-ment span--each steam generator ment span--each steam generator-76
- 15. Undervoltage-Reactor 1 5022 volts each bus 1 4739 volts-each bus Coolant Pumps
- 16. Underfrequency-Reactor
> 56 Hz - each bus 1 55.9 Hz - each bus
'?
Coolant Pumps C1
- 17. Turbine Trip
'A.
Low Trip System
> 45 psig
> 43 psig Pressure B.
Turbine Stop Valve
> 1% open
> 1%-open Closure
- 18. Safety Injection Input Not Applicable Not Applicable f" " ESF
> I x o % of RATED THE04AL
~ - --
.;t 6 [lo ' % of 4ATED THERMAL.)
~
~
- 19. Intermediate Range Neutron 1 1 1^ *
- 4 POWEf-1 5 : la 2 m ----
- POWER, l
Flux, P-6. Enable Block enjg Source Range Reactor Trip-
~
- 20. Power Range Neutron Flux
< 10% of RATED
< 11% of RATED THERMAL POWER (not P-10) Input to Low THERMAL POWER u"
Power Reactor Trips
." g Block P-7 J
gy
., =. -
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Q G
d.
'g
', LIMITING SAFETY SYSTEM SETTINGS L
f BASES Intermediate and Source Range, Nuclear Flux (Continued)-
Range Channels will initiate a reactor trip d^ -
-' '-~ ' --~ - -- ' ^^
approximately 25 percent of RATED THERMAL POWER unless manually blocked when 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 the Reactor Protection System.
',0vertemperature 6T 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 and 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 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 than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notatinns 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,
'l K2, and K3 inputs to the Oysrtemperature delta T channels and raising the P-8 setpoint to its 3 lopp relue.
and trjprfunctions as a. @. win this mode of operation, the P-8. interlock L utron Flux trip at the reduced power level.
Overoower AT IhmDverpower delta T reactor trip provides assurance of fuel integrity, e.g., no melting..under all pnssihlumns tm conditAox3 Guitetiie required range.fw betemperature delta T protection, and provides a backup to the sligh Neutron Flux trip.
The setpoint include.corrertJonssim change; in
' p.,j density and hen?, nspadt" *:f tak.er d#rr.'.apehd.vM, and dynamic compensation 4
for piping delays from the c' ore to the loop temperature detectors.
No credit
..e 'uns taken for operation of this trip in the accidept.analyspst;bowever, ite functional capabj.1.ify. at Ge. syWhd tMp JeM.iivis required by this specificatica 'ir enhance the overall reliability of the Reactor Protection System.
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- .}. j SEQUOYAH - UNIT 2 t.N U 2-4 1
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1 TABLE 3.3-1 (Continued)
TABLE NOTATION n
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.
1 The channel (s) associated'with the protective functions derivea from the out of service Reactor Coolant Loop shall be placed in the tripped condition.
pr cification 3.
not applicable.
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... ::.__._ t:
x r ;,n d above the P-6 (Block of Source Range Reactor Trip) setpoint.
9-ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels 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, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
ff V,
The inoperable channel is placed in the tripped condition a.
within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
lR39>
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 /> lR39' 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 setpoint is 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 pe'r 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 incore 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.
D.
September 17, 1986 SEQUOYAH - UNIT 2 3/4 3-5 Amendment No. 39
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i ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-88-42)
DESCRIPTION AND JUSTIFICATION FOR MODIFICATION OF THE TRIP SETPOINT-AND ALLOWABLE
-VALUE UNITS FOR THE INTERMEDIATE RANGE NUCLEAR FLUX DETECTOR AND CHANGES TO THE APPLICABILITY REQUIREMENTS FOR THE SOURCE RANGE NUCLEAR FLUX DETECTOR 9
r ENCLOSURE 2 Description of Change Tennessee Valley Authority proposes to modify the Sequoyah Nuclear Plant (SQN) Units 1 and 2 technical specifications (TSs) 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 2.0.24 (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 deenergization as part of the normal SR detector operation.
Consequently, the footnote (##) for Table 3.3-1 is being revised to change the high-voltage deenergization wording to say that SR outputs may be disabled. The new IR monitor uses a signal that is in units of relative power. Consequently, the trip setpoint and allowable value units are being changed in Table 2.2-1.
Because the new IR detector does not provide output in terms of current, the bases to Section 2.2 are also being revised to delete references to IR detector current signals that are proportional to power levels.
Justification for Change The new Gamma 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 equipment is compatible with the rest of the nuclear
. instrumentation and reactor protection system; however, it includes two improvements over the present. design. First, the electronic equipment automatically decreases the'high flux at shutdown alarm after a reactor trip until the neutron flux stabilizes. Currently, this function is performed annually as described in the Final Safety Analysis Report.
Section 15.2.4.2.
Second, the new SR/IR detector does not have to be deenergized at higher power levels. Above the P-6 setpoint, the SR detector output signal is blocked from the reactor trip logic. The SR/IR detector assemblies'will remain energized during the full range of' power operation. Consequently, the table notation in Table 3.3-1 regarding high-voltage deenergization of the SR detectors has been revised to clarify the wording regarding this feature.
- The new IR equipment is compatible with the rest of the nuclear instrumentation and reactor protection system except that the output signal-is in units of relative power rather than amperes (A). The P-6 setpoint and allowable value listed in Table 2.2-1 are currently listed in units of A.
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 and detector current was established using start-up 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 conservative decade for ease of calculation. A corresponding allowable value was then calculated using the previously
(
established-setpoint and current-power relationship. Finally, the overlap
-between the SR/IR detector ranges was checked to ensure sufficient margin between the P-6 setpoint and the SR trip setpoint.
It is important to note that the actual setpoint is not changed; only the engineering units have changed. A copy of the TVA calculation is included as an attachment to this enclosure.
In summary, two administrative changes are proposed to support.the installation of the Gamma Metrics SR/IR assembly. The first involves the revision of a table notation that is no longer 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.
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 20 and changes to the surveillance requirements. TVA has determined that the proposed change 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. 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 environmental assessment needs to be prepared in connection with the issuance of the amendment.
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Revision n
Mothed of design verification (independent review) used (check method used):
L 1.
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2.
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Cualification Tust t-Just: fica"lon (exploin below):
Methtdj:{ In the dusign review method, justify the technical adequacy of th i
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strellar Lo another, based on accepted handbook methods, appropriate L
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and exp1 min why this method is adequate.
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Method 1:.In the design review method. -justify the technical adequacy of the calculation and explain how the ade'quacy was verified tealculation is similar to another, based on accepted handbook methods, appropriate sensitivity studies included for confidence, etc.).
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Method 2: In the alternate calculation m4thod.' identify the pagen where the alternate calculdtion has been included in the calculetion package and ex' plain wiiy f.his method is adequate.
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cal.CUI.ATION DESIGN VERIFICATION (INDEPENDENT REVIEV)
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Calculation No.
Revision Method of design verification (independent review) used (check method used):
- 1.. Design Review 2.
Alternate' Calculation 3.
Qualification Test
- Justification (esplain below):
Method 1:
In the design review method,. justify the technical adequacy of the calculation and explain how the adequacy was verified (esiculation is similar to another, based on accepted handbook methods, appropriate sensitivity studies included for confidence, etc.).
5,ethod 2: In the alternate calculation method, identify the pages where the alternate calculation has been included in the calculation package
- i and esplain why this method is adequate.
Method 3:
In the qualification test method, identify the QA documented source (s) where testini, adequately 4emon4Yvate1 the adequacy of this calculation.and espIain.'
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'O Calculation No.
Revision Method of design verification (independent review) used (check method used):
1.
Design Review 2.
Alternate calculation 3.
Qualification Test 4
. Justification (explain below):
Method if In the design review method, justify the technical adequacy of the calculation and esplain how tho' adequacy was verified (calculation is silsilar to another, based on accepted handbook methods, appropriate sensitivity studies included for confidence, etc.).
Neth'od 2: 'In the alternate' calculation method, identify the pages where-the
' alternate' calculation has been included in the calculation package and esplain why this method is adequate.
Net' hod 3:' In tlie (ualiTicatTout test sleth'od,Nidentify the QA doeuraanted
stiurceTs)' tihere tes'tieg 'adnquire%yinddeoastent.es the adequacy of this calc @atton and explain.
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g, o TECM 6-! t cAs. s2 Eeq e.atM e:wrr Instrumenestion System (NIS).P-6 is a protection interlock derive
'r ate range Suelear source range NIS reactor trip may be blocked to allow contiWith r escalation.
the source range trip to give the operators time toTherefore, nued power the same tioe be above the minimum usable signal in thactuate the block a y belev e inter =ediate range SIS.
The present Westinghouse Design utill:es the following Meth d l establishing th'a P-6 setpoint for the intermediate range o o ogy of The source range reactor trip is set-at 10 5
range NIS which corresponds to about 2 x 14*9c6unts/seeinthesourec rin 10*ge'NIS.
amps to 10*). amps vi.th therThe pro'sss range of the interme e
I louer end 6T the tw.<
P about 6 x 10 2 of 10"11 amos to 2 x 109 counts /see in the source range NIS.
3 cmiresponding to range (10-10 amps in which to set P-6. This leaves a ranse Hidway in this 1 decade below the source trip to allow.the operators to and 1 decade above the lower end of the intermediate range NISoc range to achieve a reasonably good signal.
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P-6 setpoint can be calculated by eassuring the det 4
e power permissive reactor levels at or below 75% power & interpolating for th current for_different 10-10 amps.
e pouer level *T
{o c.owdE u s so w s The Gamma Metries' Design to be installed provides a process 200% rated thermal power.
with the source range (10-8 go to-6This provides an additional two decade range of 10-8 to
% RTP.)
Therefore, based on the following calculations a'value of 1 be used as the trip setpoint which is functionally equival x 10-5 " RTP sha The values calculated for the P-6 setpoint (using Plant Spe ifient to 10 good agreement with-the 1 x 10~5 c
c Data) are in RTP setpoint.
adequate mergin below-the source range trip to give the operatThis set acto as a trip sus.se ors time to
' of the intermediate range drawer (3 decades).and at the same time em l
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TENNESSEE yALLEY AUTNORITY SEQUOYAH NUCLEAR PLANT UNIT NUMBERS 1 AND 2 C
PRECAUTIONS. LIMITA7!0NS AND SETPOI FOR
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NUCLEAR STIAM SUPPLY SYSTEMS 1
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WESTINGHOUSE ELECTk1C CORPORATION Al MHID i
Nuclear Energy Systems
,,,,".,g3 4
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..~ - a P. O. Box 355
, '..T..,,g":'.a, '.5 W...i!. c3,.. c Ptttsburgh, Pennsylvanta
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- P
! 3.
turb1ne trip r
generator Hi level signal for
)
stean feedwiter valve closure, turbine
\\
L trip and feedwater pump trip (LB-517A LB 527A, LB 537A LB 547A.
75% of level span LB-118A LB 52SA, LB-53BA, LB-64BA,
(
LB-119A, L3 529A, LB 539A LB-549A)
/* yd
+aJX/0' -
!!, pemissive ard interlock Circuits x
?
A A. ! P-6(a11cus manual block of source range
!highlevel reactor trip)
'(NC-35D, 1036D) 10 10 cmaticallyblocksvarious"at amceres B.
P7 aut power" trips at low power)
I 1.
low nutronflux(Seep-10)
' t.
low tarbine load (See P-13)
C.
P8(alic es one loop loss of flow below jsetpoint)
'(NC-41N.10-42N, NC-43N, NC 44N) 35% of full'oower D.
P-9 lblocas reactor trip on turbine (4loopcperation)
,tr.ip belco setnint nuclear power level)
'(NC-415, llc-425,NC435.NC-445) 50% of full p0wer
E. i. p 10'(a' D,G Anual block of power rative
'(lowsetp > int) trip Intermediate range IfrS,and 'C;1g b1'cks source range trip e
'and proviilesaportionofp-7 signal) f(NC-41H, llc 42M, NC 43M NC-44M)
F.
!P-11(allo 10% of full pcwer ws manual block of safety in.
!jection actuation on low pressurizer pressure.
,'(See!.1.t.4above)
(..
j l
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& c." spova.-At-l M
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(
3b. l.ow steam line p' essure (PB-516A. PB 526A,,PB 536A, PB.546A) 600 psig Lead tima constant
{
(PY-il6B',' PY-5268, PY-5368. PY.546B) 50 secenes Lag time constant (PY-516R,PY-5268,PY-5363.PY.546B) 5 seconds 3c. Low <
Low T '.4220 3y
(
(78-1120. TB
.TB4320.TB.442D) 54C'?
4.
Autenatic reset of manual block on high pressurizar pressure (P 11)
(PS-1558,PR-4568,PB-4578) 1970 psig
.i 5.
Contninment high pressure
- (PB-9348, PB.9358, PB.9368) 1.54 psig
((
.6.
,T,ime delay on $1 manual reset 1 minute 8;
Steeni'Lil e Isolation 1.
Hi.gh steam line. flow (See.I.1.A.3 above) 2.
High< high containment pressure
.(D (PB-1 34A, PB 935A, PB 936A, PB-937A) 2.81 psig
- k h
C.
Containmt nt Spray Actuation
'* [
\\
1.
High< high containment pressure,(See 1.1.B.2 above) b*
(
2.
Reactor Trios A.
Nuclear 2 nstrumentation i
1.
Source range high level (NC-: 10. NC-320) 5
(
10 counts /second 2.
Intermediate range high level Current equivalent (HC: 5F,NC-36F) to 25% of fuH power 3.
Power range, low range, high level (NC.4 1P, NC-42P, NC 43P, NC-44P) 25% of full power L.
1 1
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Figu 'e 10.11 Neutron Detectors and Ranges of Operation 10.1 13 Pa s H.9 ws.u i4 er
.~
t(#4hlis.24 M1, _::::> \\ Document 1. Page w
,g.w e>.g.7 g i
4oV. 1 s
./
i
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A
[.
generated frc.m detecting the signal enwa,rd.
Where applicable, this
\\
requirement sh:uld be met with all lead, lag, and filter time constants set to 0FF.
1.14 Controller Traisfer Functions i
T 4ct Applicable l
1.15 Letooints 8/
5*7# I.0 Id a we,% :
/ariable Y
Renee of Settino i
Intermediate Range High Neutren 5 to 30% full power flux Reacter Tip i
$ource Range HghNeutronFlux
~10*b to -10'3%offullpedr
.deacter Trip glo/t; Intermediate Range Rod Withdrawal 5 to 25% at full power J3 4 !
steo (E-!.L
.%q 4, y l
/
t 4-6 0 to ~10'3% of foil oewer l
(.
_ =
~10 m - ---- _
~ s 1
i ill settings y th the exception of time constants shan tio centim:cus'.y djustable wit.h in their range an'd all time constants shall be i
,,, 4 :entinuousTy aijustableoradjustableinincrementssuenthatany setpoint can be obtained within + 10Y. of the setpoint value.
Jer the P 10 so tpoint see Nuclear Power Range Protection (Document 2).
1.16 teouWements 'I: ~r Test and Calibration L
911 protection channels.should be su;: plied with sufficient redundancy
- o provide the capability for channel calibration and test at pcwer.
l-
,n the case of 1/N iogic a bypass must be.provided to prevent a reacter Srip during tai t.
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ECHNO. La&& gy QA Record
-**1 j.
6.6
. 94 *,
Page 1 of + g, 6 SAFETY EVALUATION FORM E.- E, -- - 6 Sheet 1 1.
To 3.
USQt
- 4. Safety Evaluation Number l
Sequoych Nu&:r Nent Da%y, TN O Yes EcA>L(/F4 2.
From El 6 S N P.
V No
- 9. RIMS Accession Number Ro SQ9 65biIT 505 Rev Tot Date R
No.
on
$. Prepared 6. Reviewed 7. Aporoved 8. Anod iB 2 5 8 81114 5 32 l o
s
- 4. A. se s h u/ud;4/w
@/er lB25 881117 570
&hreaf-7d2si "B3789 1127 8 01 1
. m.uu
,,.,.. s,
2 3 *l ~7~ 2 b Ine ILL Is. s,Mt 43 Je(& Mai 3
M/.L di r//.,
F_h _
A N
R t/,13 9 9 5
t 5
1
- 10. Project pad Affected Unit (s)
- 11. PMP or DCN Number PHP or DCN Revision
. fen uo vei U, o /s
/ i *t.
EcA.) L t/&4
DCN, or CAQR Number Date of Document (s)
Dc8 //5 3 ////t %.
- 13. Other Document Identifier N*'n '
Date of Document Alfa
'14. S c a,1 )equirements?
See
- 15. Potential Tech Spec e
No SheetNo.M Change ( Yes a No Shee5 ah b-W See
!R1 No..' r l l
- 16. References (include systes number and name as appropriate)
CMI Sysk.m 11. ~ Ne.aW Ma-, o Vser m'
/(<.Cee< ~c.a :
- f. Itsy Guinbt. /.17 /t't.
/
z., ass j deikis. 5M)-ec-V-17. 9 (c-%nc/e.,Sl<.a.]7)
%& *~" 4"'% Sj.s k "
- 17. Description of Proposed Activity (Change, fest, or Experiment)
- fA 's [M ufyt.de s /w e /ccs e, Ao/ al'a2 s }g /l,.s y na a.,)
l m hen,ada'k e= ~p. ne u lea-n.
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f
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.,,l; a-i Xey a/. /+er Go.dc. /. 9 7, ga.v,.r,- t. Rt.hta c.e. M,as, Af,29, p 3o 3 4
( c. fio,
.r L /
3 )
E*n: 7 U.U.'
cc (Attachruents):
RIMS, SL 26 C-K '
,g{,g
~
~ ^
-f v b
~
a.l.:l.l.5: Jjbh uy, y
- 9... t. }!;+p.p + m.p:::
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Ed.tay4Go es 4 OF Sheet 2 Safety Evaluation No. ECN L6186 NEp 6.6 i
ADDITIONAL INFORMATION 16.
References (Continued) 3.
FSAR Section 7.1.2.1.3 4.
FSAR Sections 7.2.1.1.2, 7.2.1.1.3 5.
Design Criteria SQN-DC-V-19.0 " Post Accident Monitoring" 6.
Design Criteria SQN-DC-V-27.9 " Reactor Protection System" 7.
Design Criteria SQN-DC-V-1.0 " General Civil Design Criteria" 8.
SQN PAM Appendix F " Design Criteria for Qualification of seismic Class I and Seismic Class II Mechanical and Electrical Equipment.
9.
47B601-92 Series I-Tabs 10.
Design Criteria SQN-DC-V-12.2 " Separation of Electrical Equipment und Wiring" 11.
Tech Specs 12.
Design Criteria SQN-DC-V-26.2 " Environmental Qualification to 10CFR50.49 l
13.
Calculation 1,2-XE-92-1, Rev. 1 RIMS No. B25 881117 808.
1 14.
Design Criteria SQN-DC-V-2.3 " Containment Vessel" p
15.
Design Criteria SQN-DC-V-11.3 " Power, Control, and Signal Cable for use in Category I structures 16.
Design Criteria SQN-DC-V-13.10 " Seismic Qualification of Conduit" 17.
FSAR Tables 7.2.1-1 through 4 and 8.3.1-11, 12, 13, 15, 16 18.
FSAR chapter 15.0 19.
Plant Procedure PHYSI-13 " Fire" 20.
Westingh6use Test Reports, Nuclear Instrumentation System Isolation
-Amplifier WCAP-7506-L, NEB 810126303 and WCAP-7819 NEB 8102040314 21.
Nuclear Rngineering Calculation, " Equipment Required for 10CFR50, Appendix R" SQN-SQ54-127 Rev. 10 (RIMS No. B25880829501) 22.
Gamma Metrics Neutron Flux Monitoring Instruction Manual No. 72 Rev. 4, l
Contract No. 835545 23.
Gamma-Metrics Test Report No. 135. Rev. O, Isolation Testing of Single Channel Isolators, Contract No. 835545
- 24.. Gama-Metrics Test Report No. 010, Rev.1. Woutron ' Flux Monitoring Qualification, Contract No. 835545 l
25.
Gamma-Metrico Test Report No. 096, Rev. 1. Source and Intermediate Range Rack Mount Signal Processors Qualification, Contract No. 835545 26.
Demonstrated Accuracy Calculation SQN-EEB-PS-TI28-0001 R3-27.
Pipe Rupture Calculation SQN-CEB-SCG-4E00168 28.
Test Report No. 12. Rev. O. Gamma-Metrics RCS series Woutron Flux Monitoring Seismic Qualification Report and MSLB/LOCA Test Report, Contract No. 835545 l
29.
Ca ms-Metrics Test Report No. 31. Test Plan for Qualification of l
Gamma-Metrics RCS series of Neutron Flux Monitoring Systems per IEEE STD 323.1974, Contract No. 835545 i
30.
Gamma-Metrics Test Report No. 40, Rev. 3, Test Report for Class 1E Qualification of Mineral Insulated Cable in the Detector Cable Assembly, Contract No. 835545 1487E
I i
ECN NO. Lien C She::t 3
_gf_OF Safety 2yaluation l
No. ECW L6186 NEP 6.6 ADDITIONAL INFORMATION 16.
References (Continued) 31.
Camma-Metrics Document No.133. Rev.1 Shutdown Monitor, Contract No. 835545 32.
TI-81 NIS calibration for restart following core load, Rev. 4 33.
Design Criteria SQN-DC-V-12.1 " Flood Prr.tection Provisions" 34.
Design Criteria SQN-DC-V-11.6 "120V A-C Vital Instrument Power System" 35.
Camma-Metrics Test Report No. 26, Rev. 3 " Optical Isolator, Fiber Optic l
Transmission System and RCS-211 Amplifier Qualification.
36.
Camma-Metrics Test Report No. 27. Rev. O, " Seismic Test for Optical Isolator Fiber Optic Transmission System and RCS-211 Amplifier".
37.
120 VAC Vital Inverter Loading Calculation SQN-CPS-021, t
RIMS Wo.. B87 891011 006.
38.
Al-17 Drilling, Cutting, Chipping and Excavating, Revision 13.
39.
DIM-SQN-DC-V-27.8-3, RIMS No. B37 891101 802.
NI'.
40.
DIM-SQN-DC-V-27.9-7, RIMS Wo. B37 891101 801.
41.
Westinghouse Letter - TVA-89-963, RIMS No. B25 891030 010.
42.
HVAC Cooling Load Calculation, Auxiliary Bldg., EGTS Room Elev. 734 ft.,
RIMS No. B25 891102 504.
l 43.
HVAC Cooling Load Calculation, Auxiliary Bldg., Elev. 714 ft.,
RIMS No. B25 891102 505.
44.
HVAC Cooling Load Calculation, Auxiliary Bldg., General, i
RIMS No. B25 891102 503.
45.
HVAC Cooling Load Calculation, Reactor Blds., Lower Containment, RIMS No. B25 891102 506 i
46.
VCPS Loading Evaluation, SQN-CPS-022 Rev. O, RIMS No. B87 891120 002 l
17.
Description of Proposed Activity (Continued)
None of the existing outputs, functions, and interfaces to other systems, such as the Reactor Protection System, will be functionally changed by this upgrado.
This ECW will resolve all outstanding Appendix R commitments for the source range channels.
The specific modifications performed by this ECW are described below and will be performed on both Sequoyah Units 1 and 2.
l l
1.
Replace the non-1E Westinghouse source and intermediate range neutron detectors with Class IE Camma-Metrics source and intermediate range detectors.
+
1487E
,-- m o
ECidido.Lich st Sheet 4 AOF Safety Evaluation Wo. ECN L6186 Nap 6.6 ADDITIONAL IWFORMATION 17.
Description of Proposed Activity (Continued)
, 2.
Reroute cable and conduit, and locate junction boxes and other hardware above containment building flood level from the detectors to containment penetrations 31 (Unit 2 Channel 1), 23 (Unit 2 Channel 11), 43 (Unit 1 Channel 1), and 48 (Unit 1 Channel II). Special pressure tight procut cable from the detector through the containment penetration to the amplifier shall be installed in accordance with Gamma-Metrics Instruction Manual. Reference No. 22, 3.
Replace the non-1E Westinghouse pre-amplifier with a 1E Gamma-Metrics amplifier assembly.
4.
Replace the cable from Unit 1 penetrations 43 and 48 and Unit 2 penetrations 23 and 31 to the new amplifier assemblies described in number 3 above.
5.
Re, place the cables and conduit routed from the amplifiers to the NIS racks located in the control building. One channel will be routed on elevation 734 above the design basis flood level.
6.
Replace the non-1E Westinghouse intermediate and source range signal processing drawers with Gamma-Metrics intermediate and source range drawers. The new Camma-Metrics drawers will pecvide the required electronics to provide qualified signals for compliance with Reference 1.
7.
Install a shutdown monitor on each source range neutron monitoring channel to automatically adjust the high flux at shutdown alarm setpoint downward for flux decay during shutdown.
It will identify and report flux increases that indicate a loss of reactor shutdown margin.
This will eliminate manual adjustment of this high flux at shutdown setpoint (Reference 22 and 31).
lt 4
1487E 1
ECN NO. L6 t hc,. e3 shTt 5 Safety Evaluation O OF-No. ECW L6186 i
l WEp 6.6 ADDITIONAL INFORMATION 17.
Description of proposed Activity (Continued) 8.
Route an isolated temporary cable from the Auxiliary Control Room L-10 to l
Main Control Room panel M-19 to provide redundant neutron flux llt3 infomation to the main control room from the Appendix R backup source j
range neutron monitor during implementation of items 1-6 above. Only one channel of the source and intermediate range flux monitors will be worked i
at one time. The first channel to be replaced, post-mod tested, and documented, shall be declared operable prior to removing the second channel from service. The backup and one operating source range channels will ensure the operator has sufficient neutron flux information during implementation of this ECN. The temporary cable shall be removed after both new channels are declared operable and prior to Modo 3.
IG
]
9.
Replace the Appendix R Westinghouse backup source range detector, j
cabling, and electronics with an optically isolated signal from the new 1
Camma-Metrics Amplifier to a source / power range processor and display.
This will reduce maintenance, spare parts storage, and enhance the l
l back-up control Room neutron flux readout. This will require a revision O
to Reference No. 21 by minicalculations to the UIC4 and U2C4 Appendix R Calculations SQN-SQS2-0094 and -0100, respectively.
]
10.
Replace containment penetrations 31 (Unit 2) and 43 (Unit 1) with IE qualifLed penetrations. penetrations 23 (Unit 2) and 48 (Unit 1) have already been replaced under ECN L6490.
11.
Westinghouse drawings 5655D26 shoots 3 and 4 and 108D438 sheets 2, 5 and 43 that are affected by this ECM will be changed under DCR-3094 for Units 1 and 2.
12.
The new control room indicator and recorder scales required for this ECW are being purchased and their installation coordinated with DCN M01496, O
Unit 1 and M01497 Unit 2.
13.
Westinghouse plS and setpoint methodology documents that are affected by this ECN will be changed under DCR-3094 for Units 1 and 2.
14.
The Technical Support Center Data System (TSCDS) computer software will be modified to reflect the new range and units (10-8 to 200% RTp) for the intermediate range instruments. The calculation of the intermediate and source range start up rates (SUR) will be modified to use time tagged levels rather than assume the Jata'is supplied from the input multiplexers every two seconds.
1487E
, - - - ~. ~
1 ECNfC Lv % es Sheet 6
)
Safety Evaluation 55 OF Wo. ECW L6186 l
NEP 6.6 i
ADDITIONAL INFORMATION 17.
Description of Proposed Activity (Continued) 15.
All drilling, cutting and chipping will be performed in accordance with,N AI-17.
Al-17 states that the description of the work shall include all s information required to locate the exact spot for drilling, chipping or cutting by relating to elevations, distance from colunn lines and other
~
t location references used on plant drawings. Additionally, a review of all effected drawings will be perforned to verify embedded piping, electrical conduit, cable troughs, duct work, tunnels or other plant i
equipment will not be jeopardized and shall be listed under the drawings-reviewed section of the permit. Submittal of an AI-17 permit is to servei a verification that the location of all known embedded piping, electricalf conduit, cable troughs, reinforcing steel, duct work, tunnels or other
~
plant equipment will be identified before the drilling, cutting, or shipping it started. Reference No. 38.
E 16.
Implementing the proposed activity will require revisions to:
FSAR Sections:
4.4.5.3 Tables:
8.3.1-11 and Figure:
7.2.1-1
?
15.2.4.2 7.2.1.1.2.c 8.3.1-12 sheets 3 and 4 7.5 (in to-to) 7.2.1.1.3 8.3.1-13
-r 8.3.1.2.3 7.2.1.1.8 8.3.1-15 g3:
7.2.2.2.3 8.3.1-16 i
7.2.4 15.2.4-1 15.2.4.3 Tech Specs (Sco Question 27)
- 1 Tabic 3.3-1 Notes Table 4.3-7 Bases. Limiting Safety System i
Tabic 3.3-10 Table 2.2-1 Settings Section 2.2.1 This activity will be performed in the following stages:
State 1: Will consist of completing itens 1 through 8 and 10 by the end of the Unit 1 Cycle 4 outage for Unit 1.
State 2: -Will complete items 1 through 8 and 10 by the end of the Unit 2 Cycle 4 outage.
State 3: Will implement item 9 for both units by the end of the Unit 2 Cycle 5 outage. This portion of the modificution is not part of the NRC commitments for Cycle 4.
3 T.:
m
.r u
1487E 7
n
-~.
ECl4 tK1 L4tE6 #S-l SV OF Sh:st 7 Safety Evaluation No. ECN L6186 NgP 6.6 ADDITIONAL INFOKMATION 18.
Systems. Structures. Components Af fected NM - Neutron Monitoring RPS - Reactor Protection System PAM - Postaccident Monitoring SR - Source Range
, CV - Containment Vessel IR - Intermediate Range RTP - Rated Thermal Power ITEM l COMPONENT l SYSTEM AFFECTED l DESCRIPTION OF CMANCE I
1 l Source / Intermediate Woutron lNM, PAM, RPS, l Replace non-1E detector with IE l
l Detector and Cabling l Appendix R SR l qualified detector which contains l
l l
ltwo identical, redundant fission i
l l
l chambers and has a pulse signal out-l l Channel I. XE-92-5001 IUnit I lout proportional to reactor power l
2 l Source / Intermediate Neutron lNM, PAM, RPS, l Replace non-1E detector with IE l
' Detector and Cabling
- Appendix R SR
' qualified detector which contains I
i l
I two identical, redundant fission l
{
l l chambers and has a pulse signal out-1 l Channel II. XE-92-5002 l Unit 2 lout proportional to reactor power l
3 l Unit 2 Primary containment l Penetration
, Replace 75 ohms triax with l
l Penetration No. 23 lFeedthroughs l 50 ohms triax l
IChannel II l
l l
4 l Unit 2 Primary Containment lCV and l Replace complete non-qualified l
l Penetration No. 31 lPeneiration lwith IE qualified penetration l
IChannel I l
l 14' 5
l Unit 1 Primary Containment l Penetration l Replace 75 ohm triax with l Penetration No. 48 lFeedthroughs l50 ohm triax I
l Channel II l
l l
l 6
l Unit 1 Primary Containment lCV and l Replace complete non-qualified l
l l Penetration No. 43-l Penetration lwith IE qualified penetration l
l channel I l
l l
7 l Channel 1 Signal Amplifier lNM, PAM, RPS l Replace non-qualified with IE 1
lKM-92-5001A - U1 l Appendix R. SR l qualified amplifier to mate with IXM-92-5001 - U2 IUnit 1 lnow detector 1
8 l Channel 11 Signal Amplifier lNM, PAM, RPS l Replace non-qualified with IE l
lKM-92-5002 - U1 l Appendix R, SR l qualified amplifier to mate with l
lKM-92-5002A - U2 IUnit 2 Inow detector
_l 9
l Shutdown Monitor lNM, PAM l Add new device to automatically l
l Channel 1, 11S-92-5001 l
l adjust high flux at shutdown l
l l
lalarm down l
10 l Shutdown Monitor lNM, PAM l Add new device to automatically l
l Channel II, XIS-92-5002 l
l adjust high flux at shutdown l
l lalarm down l
i-d T
1487E
l ECl4 IJO. pot % its Shzt 8 Safety Evaluation 6
OF Wo. ECW L6186 NEP 6.6 ADDITIONAL INFORMATION 18.
Systems. Structures. Components Af fected (Continued)
NM - Neutron Monitoring RPS - Reactor Protection System
.PAM - Postaccident Monitoring SR - Source Range CV - Containment Yessel IR - Intermediate Range RTP - Rated Thermal Power ITEM l COMPONENT l SYSTEM AFFECTED l DESCRIPTION OF CHANGE l
11 l Source Range Drawer lNM, PAM, RPS l Replace Westinghouse with l
l Channel 1, XI-92-5001 l
l Gamma-Metrics l
l l
l l
12 l Source Range Drawac lNM, PAM, RPS l Replace Westinghouse with l
l Channel 11. KX-92-5002 l
lGanna-Metrics l
I I
13 l Intermediate Range Drawer
. Replace Westinghouse with l
l Channel 1 XX-92-5003 l
l Gamma-Metrics; range and readout l
l l
l change from 10-11/10-3 amps to l
s l
l l10-8 /2 x 102 % RTP l
l l
l l
14 l Intermediate Range Drawer lNM, PAM, RPS LReplace Westinghouse with l
l Channel II, KX-92-5004 l
l Gamma-Metrics; range and readout l
l l
l change from 10-11/10-3 amps to l
l l
[10-8 /2 x 102 % RTP u
I I
I I
15 l1E Optical Isolator, Unit 1 lNM, PAM, lNew device to isolate Main Control l
l Channel 1 XM-92-5001B l Appendix R SR l Room from remote shutdown l
l l
l Unit I l
l 16 l1E Optical Isolator, Unit 2 lNM, PAM, lNew device to isolate Main Control l
l l Channel II, KH-92-5002B l Appendix R SR l Room fr:m remote shutdown l
l l
IUnit 2 l
l l Appendix R Source Range lNM, Remote l Replace Westinghouse sotace range l
17 l Channel (detector, pre-amp l Shutdown lchenr.el with optically isolated l
land drawer) KI-92-5 l10CFR50 App. R l output from 1E-qualified l
l l
l Gamma-Metrics amplifier and new l
l l
lsource/ power drawer l
18 l Components 1 through 17 l120 VAC Vital l Increased Load l
l l Instrument Power l l
l l Board l
l 19 l Components 1 through 17 l125 VDC Vital l Increased Load l
l l Batteries I
j 1487E
o ECHNO.uctM,es Sheet 9 6W OF Safety Evaluation No. ECW L6186 NEP 6.6 ADDITIONAL INFORMATION 19.
Safety Function (s) of Systests) Affected The safety functions of the systems affected by this ECN are described below:
1.
POSTACCIDENT MONITOEING (pAM)
The safety function of the postaccident monitoring system is to provide information on plant variables required by control room operating personnel during accident situations tot 1.
Permit the operator to take preplanned manual actions to accomplish safe plant shutdown.
2.
determine Whether safety systems or systems important to safety are performing their intended functions.
3.
determine the potential for causing gross breach of the barriers to radioactivity release and to determine if a gross breach of a barrier has occurred.
4.
assess the operation of plant systems to make appropriate decisions as to their use.
5.
allow 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.
The PAM variable affected by this ECN is neutron monitoring. Neutron monitoring provides information for purposes 1 and 2 above. Additional safety function information is available in References 1 and 5.
2.
NEUTRON MONITORING SYSTEM - SOURCE AND INTERMEDIATE RANCE The source and intermediate range neutron monitoring safety functions are described below Intermediate range high neutron flux trip The intermediate range high neutron flux trip circuit shall trip the reactor when one out of the two intermediate range channels exceed the trip setpoint (25% RTP). Tech Specs Table 2.2-1.
This trip, Which lRs.
Provides protection during reactor'startup, can be manually blocked if two out of four power range channels are above approximately 10 percent Power (p-10).
Three out of the four power range channels below this value automatically reinstates the intermediate range high neutron flux trip. The intermediate range channels (including detectors) shall be separate from the power range channels. The intermediate range channels can be individually bypassed at the nuclear instrumentation racks to Permit channel testing at any time under prescribed administrative procedures and only under the direction of authorized supervision. This bypass action shall be annunciated on the control board.
1487E
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ECH NO._ U> s 60. e s 12__OF Safety Evaluation Wo. ECN L6106 NgP 6.6 1
ADDITIONAL INFORMATION a
19.
Safety Function (s) of System (s) Affected (Continued)
Source range high neutron flux trip The source range high neutron flux trip circuit shall trip the reactor when one of the two source range channels exceeds the trip setpoint (105 CPS) Tech spec Table 2.2-1.
This trip, which provides l(t) protection during reactor startup and plant shutdown, can be manually j
bypassed when one of the two intermediate range channels reads above the j
P-6 setpoint value (source range outputs disabled and intetinodiate range j
on scale power level) and shall be automatically reinstated when both intermediate range channels decrease below the P-6 value. This trip i
shall be automatically bypassed by two out of four logic from the power range permissive (P-10).
l This trip function shall also 'De reinstated below P-10 by an administrative action requiring manual actuation of two control board mounted switches. Each switch will reinstate the trip function in one of the two protection logic trains. The source range trip shall be set between the P-6 setpoint and the maximum source range level. The i
channels can be individually blocked at the nuclear instrumentation racks to permit channel testing at any time under prescribed administrative procedures and only under the direction of authorized supervision. This blocking action shall be annunciated on the control board.
The source and intermediate range neutron monitoring system also comprises a portion of PAM for purposes described above in "Postaccident Monitoring". Additional detailed safety function information is available in References 2, 39, and 40.
l Another safety function of the source range neutron flux monitor is to provide an increasing count rate when RCS boron concentration decreases h
during shutdown, which is a condition 11 fault described in Chapter 15 of the FSAR.
The intermediate range provides a si5nal to block rod withdrawal (C-1) in the event of high neutron flux, Reference 2.
This is a control function only.
- The 120 VAC Vital Instrument Power Boards provides an extremely reliable source of instrument and control power for reactor protection circuits and other critical instruments. It isJ5esigned with sufficient independence, g3 redundancy, and testability to perform its safety function assuming a single failure.
Furthermore, none of the following design basis events shall prevent the vital instrument power system from performing its function:
any single equipment or component failure; any single act, event, component failure, or circuit fault that could cause multiple equipment malfunctions; the safe shutdown earthquake; the postulated accident environments; accident generated missiles; accident generated flooding, sprays or jets; fire; fire protection system operation; loss of off-site power, Reference 34.
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No. ECW 1.6186 NEp 6.6 ADDITIONA1, INFORMATION 19.
Safety Function (s) of System (s) Affected (Continued) 3.
PRIMARY CONTAINMENT SYSTEN The primary Containment System will be breached during the replacement of electrical containment penetrations 43 (U1) and 31 (U2) and during
-the replacement of the feedthroughs on penetrations 23 (U2) and g3 48 (U1). The safety function of the primary containment system is to limit leakage of radioactive material from the containment building under design basis accident conditions. Additional safety function information available in Reference 14.
4.
REACTOR PROTECTION SYSTEM The source and intermediate range neutron monitor input to the reactor trip system high neutron flux trip circuits which are described in #2 above, and will trip the reactor at high source or intermediate flux-levels, respectively, during reactor start-up. The reactor trip system comprises the Reactor Protection System.
The Reactor Protection System is by definition a primary safety system, due to its requirement to shut down the reactor and maintain it in a safe condition whenever a possible dangerous situation exists.
The functional performance requirements of the Reactor Trip System shall include provisions for automatically initiating a reactor tript a.
Whenever necessary to prevent fuel damage for an anticipated transient (Condition 11),
b.
To limit core damage for infrequent faults (Condition III).
c.
To keep the energy generated in the core under control to limit fuel damage such that 10CFR 100 dose limits are met and peak clad temperatures are less than 2200'F.
The Reactor Trip System initiates a turbine trip signal whenever reactor trip is initiated to prevent the reactivity insertion that would otherwise result from excessive reactor system cooldown and to avoid unnecessary actuation of the Engineered Safety Features Actuation System. Additional safety function information is available in Reference 6.
5.
REMOTE SHUTDOWN INSTRUMENTATION Sourco range neutron flux is an instrumentation channel required in the event of a main control room evacuation for safe shutdown. Reference R2 21.
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No. ECN L6186 MEp 6.6 SAFETY EVALUATION 20.
Effects on Safety This modification and the Tech Spec change will not affect the refety functions of the systems listed in number 19 above or any other systems j
important to the safety for the following reecons:
1.
This ECN upgrades the source and intermediate range neutron monitori*ig system in order to meet the qualification guidelines of Reg. Cuide 1.97 R2.
The source and intermediate neutron monitors will provide a primary j
safety function by providing the control room operator information to take preplanned manual actions to accortplish safety shutdown of the plant during accident conditions. As discussed in Keference 6, Table l
3.1.2-1, credit is not taken for source and intermediate range high flux reactor trips in the FSAR Chapter 15 safety analysis since they are in addition to power range trips. Also, the interlocks described in nuniber 19, item 2 are not changed or altered, hence the function of the Reactor Protection System is not changed.
2.
During implementation of this modification, the backup source range neutron monitor which presently outputs to the auxiliary control room will also temporarily output to the main control roo.;.
Tech Spec 3.3.3.5 only requires that the backup source range neutron wonitor be operable only in modes 1, 2, or 3.
Since this modification will be performed in modes 5 and 6 L.C.O. 3.3.3.5 will not be entered.
The existing Westinghouse backup source r&ngo electronics has a built-in isolated output which will be used for the main control room readout.
Westinghouse testing shows that no credible fault could damage tho backup source range electronics with this temporary cable installed.
References 20 and 41.
This will allow the modification to proceed p
during Mode 6 with one permanent source range neutron flux channel
-inoperable without invoking a limiting condition for operation as described in Tech Spec 3/4.9.2 and maintain Appendix R compliance.
The temporary cable from the auxiliary control room to the main control room to connect the backup source range neutron monitor to the main control room (as described in item 2 above) shall be routed such that the qualification values for the isolation amplifier will not be exceeded. Temporary breaches of fire barriers will be administratively controlled as required by Reference 19.
This ensures that the consequences of a fire are not increased during implementation of this ECN.
1487E
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She:t 13 Safety Evaluation G"O OF go. ECW L6186 NEP 6.6 SAFETY EVALUATION 20.
Effects on Safety (Continued) 3.
Electrically, the source and intermediate neutron range monitors affect output only to PAM and RPS as discussed above. No other system receives input from any portion of the source or intermediate neutron monitoring system. Power consumption of the new Gamma-Metrics units is documented in eciculation SQN-CPS-021 (Reference 37).
The effect of this increased load has been evaluated in calculation SQN-CPS-022 (Reference 46) for R1 both the vital AC and DC power systems. The addition of the 1E optical i
isolator assembly, References 35 and 34, will allow the use of one of the i
Main Control Room detector pignals yet maintain the separation required for Appendix R.
ECW L6186 will, therefore, have no effect olectrically on any other system important to safety sxcept for the 120 VAC vital power boards and 125V vital DC power systems.
l 4.
The upgrade of the source and intermediate range neutron monitoring system to Reg Guide 1.97, Rev. 2. Category I qualificatione assures tnat the system will be able to withstand seismic or environmental stresses and remain functional to provide the primary safety function described above.
5.
Each source and intermediate range neutron monitoring channel will be fully redundant and separate from the other in accordsnee with the requirements of Reference 10.
This assures that the primary safety s
function of the source and intermediate range neutron flux monitore is not compromised by singic failure.
6.
The upgrade of the source and intermediate range neutron monitoring system to IE requires the components of the system to be ceismically
{g3 mounted. Therefore, components or equipment of other systems important to asfety will not be subjected to seismically induced missile demage.
7.
This item moved te item 2 thic question for clarity.
8.
The effects of pipe rupture (Reference 27) on the new cable routing, jenetton boxes, and other hardware have been evalustod in accordance with SQEP-51. The evaluation' concluded that the components will not be affected by pipe rupture, assuring the reliability of the primary safety function of the source and intermediate neutron monitors. -(See Special Requirements No. 2).
R3 9.
Contatnment electrical penetrations 43 and 31 will be seismically and environmentally qualified, and leak tested in accordance with Tech spec surveillance requirements 4.6.1.2.
This ensures that containment leak integrity will be within the margin of safety defined in the bases of Tech Specs 3/4.6.1.1 and 3/4.6.1.2 and that the containment will be capabic of providing a radioactivity barrier, t
1487E
U C N N O y,i N, # b_
ghe;t 14 Sa ety Evaluation
&i OF No. ECW L6186 WEP 6.6 SAFETY EVALUATION 20.
Effects on Safety (Continued) 10.
All components affected by this ECW located in harsh or essentially mild environments will be qualified for these environr.ents in accordance with the requirements of Reference 12, ensuring reliability of j
instrumentation during Chapter 15 condition III or IV faults.
(See Special Requirements No. 5).
- 11. WI.en the ECN is completed, one channel of the source and intermediate neutron monitoring system will provide a signal through an optical II -
isolator to the auxiliary control room. The amplifier will receive power directly from an Auxiliary building vital instrument board. This ensures that the effects of an Appendix R fire in the control building or an undesirable habitability condition existing in the main control i
room will not affect the backup source and intermediate range neutron monitoring system's primary safety function.
12.
One chcnnel of th0 acarce and intermediate range monitors will be routed above Auxiliary Building floor elevation 734 and the other will be routed above floor elevation 714. The Auxiliary Building floor will I
provide a fire barrier, ensuring that fire in the Auxiliary. Building will not offect both neutron monitoring channels and protects one channel from the Design Basis Flood, Reference 33.
13.
The new shutdowr. monitors identified in block 17 Item 7, will automatically adjust the high flux at shutdown alarm setpoint downward during plant shutdown as the count rate decreases, presently, this fuaction ir manually performed and addressed in FSAR Section 15.2.4.2.
When the count rate achiever a steady value and then eventually increases, the start setpoint remains Cc its lowest value.
An alarm will occur when the count rate reaches a value equal to the alarm sotpoint which is set at 3 times the average count rate.
The alarm setpoint can be increased only by depressing the alarm setpoint reset at which time a new alarm setpoint will be computed from the current count rate value (Reference 22). There will bs one shutdown monitor for each k3 neutron monitering channel.
Each will be electrically separate and fully redundant in accordance with the requirements of Reference 10.
Alto, visual verification of the setpoint and count rate can 've performed by the operater any time below 104 counts pec second, (Reference 22), when the shutdown monitor is in service to ensure the monitors are performing their function.
The high flux at shutdown alarm will continue to perform its (pnction as before, with the alarm setpoint beinh nojusted automatically by the shutdown monitors.
The reliability, redundancy, and shutdown monitor tracking verification feature ensures that this high flux et shutdown alarm function will not t a affected, hence the consequences of boron dilution, a condition II fault, will not be increaned, i
IAR7E
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Ecno.t1th e3 sh%t 15
,j,tg_op Safety Evaluation No. ECN L6186 NEP 6.6 l
pr SAFETY EVALUATION 20.
Effects on Safety (Continued)
,14.
The now Camma Metric source range neutron monitors are equivalent to the Westinghouse BF-3 detectors with regard to instrument j
sensitivity.
Indicated response to neutron flux is not changed significantly by this modification and will not affect reactor trip 1
interlock setpoints or alarm setpoints with regard to the power at which they occur.
Technical Instruction (TI-) 81, NIS celibration for restart following core loads. Reference 32, is normally used to provide recalibration information to instrumentation for power and intermediate range
{
detectors prior to restart following refueling. TI-81 ratios the new core design to the last core design multiplied by the detector process output measured during the last cycle at full power to obtain the expected process output at full power for the new cycle, l
In the past, the detector process output was in amps which equated to rated thermal power. The new Camma-Metrics detector process output will be in rated thermal power. TI-81 will have to be revised to use these revised engineering units. For the first startup after the installation of the new Camma-Metrics detectors, an initial expected full power calibration factor will be supplied by the vendor.
A NIS calibration procedure will be prepared whien will be a part of pMT-62, and include instructions for the initial startup af ter the R3 i installation of the new detectors.
In order to help assure that the new sourec/ intermediate range detectors do not contribute to any overpower condition or rate of change, the 25 percent intermediate range reactor trip setpoint will be lowered to 12 percent and the 20 l
percent rod stop will be lowered to 9 percent for this initial startup
(
only. Before power is increased above 5 percent, an evaluation of the intermediato rango detector response will be made and_the detector electronics recalibrated if necessary. Once an acceptable calibration has been verified, the trip and rod stop setpoints may be' reset to the 25 and 20 percent values and power increase continued.
precise measurements of reactor power at several plateaus during the first startup after refueling are standard practice.
If necessary, NIS may be recalibrated as a result of any of these measurements.
In addition, TI-81 may be used in the 5 to 25 percent power range to obtain recalibration factors when low leakage loading patterns result in erroneous detector responses.
l l
1487E
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(,3 og safety Evaluation No. XCd 1.6186 NEP 6.6 SAFETY EVALUATION 20.
Effects on Safe _ty (Continued) 15.
A fire hazard analysis avaluation, SQN-26-D053-EPM-MHS-022289 has been performed for Unit 2 to ensure that a fire in one location will not affect both instrument channels.
See Fpecial Requirement No. 1 for the Unit 1 limitstions until this analysis has been performed.
16.
There will be minor additional heat loads addsJ to the Reactor, Auxiliary, and Control Buildings, which is documented in Roferences 42, j
43, 44, and 45.
This additional hesi load is small enough that it will l
not adversely affect these areas.
R3 17.
FSAR Section 7.5 in to-to will be revised by one 10 CFR 50.59 I
evaluation to address the new plant configuration following the modifications necessary to satisfy WRC commitments in the PAM licensing submittal (RIMS L44 881228 808). Therefore, this safety evaluation will only refer to the CRPSAR for PAM ic)ementstion. See Special Requirement No. 6.
21.
Would the proposed activity increase the probability of an accident previously evaluated in the SARf h Yes LU No 5
Justification:
The source and intermediate range neutron monitoring system does not provide a function to reduce the probability of condition IIT or IV faults.
However, the source range monitor provides an alarm for RCS boron dilution during shutdown (a Condition 11 f ault). The new shutdown monitor described in Block 17 (Item 7) will automatically adjust the neutron flux alarm-setpoint down during shutdown. This feature will be verified operable by surveillance testing and eliminate the need for an adjustment to the setpoint during shutdown as described in FSAR section 15.2.4.2.
This will reduce the human involvement in this setpoint, ensure that the setpoint is correct at all times for all background conditions, and subsequently reduce the probability of boron dilution goins unnoticed during shutdown.
As discussed in humber 20 " Effects on Safety", the lowered intermediate range reactor trip and rod stop setpoint during initial startup after core reload will ensure conservatism in,.these setpoints.
Also, the modification is designed so that it will not indirectly affect (by seismically induced missiles, etc.) any other component, equipment, or system necessary for reducing the probability of an accident.
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- 22. -Would the' proposed activity increase the consequences of an accident i
previous 1y evaluated in the SAR?
Yes. LU No i
~
Justificationt The source and intennediate range neutron monitor's function in the Reactor protection System is not credited for mitigating the consequences of a l
Chapter 15 accident, according to Reference 6. -However, the source and intermediato range neutron monitors do provide the operators reactor power Y
level information after a Condition II, III, Or IV fault in order to take lD L
L preplanned manual actions to accomplish safe shutdown. As discussed in #20
" Effects on Safety", the new source and intermediate trMe neutron monitor components and the new shutdown monitors are procured, designed, and will be installed to ensure reliability after being subjected to seismic and
~
environmental stresses.
Also, the system is designed so that the neutron g
monitors will not be rendered inoperable by single failure.
Based on the discussion above, the source and intermediate neutron PAM' parameter will be available to the operator during and after Condition II, l
III, or IV faults so he may accomplish safe shutdown and mitigate the consequences of such faults.
The temporaey breach of containment to replace the penetration can only be l-performed during Mode 5, cold shutdown, due to Tech Spec 3.6.1.1 and 3.9.4 which require containment integrity during Modes 1, 2, 3, 4, and 6.
Having R3 crmtainment breached during Mode 5 has already been analyzed and consequences of an accident cannot bo increased.
i 23.. Would the proposed activity increase the probability of a malfunction of equipment important to safety previously evaluated in the SAR?
L/ ies ^ LK/ No L
Justification:
i l,
As discussed in Number 20 " Effects on Safety", tha source and intermediate
. range neutron monitors are upgraded to safety Class 1E and designed in accordance with the requirements of References 1, 10, and 12.
They will not be susceptible to seismic or environmental stresses, nor will they be
., susceptible to single failure. The probability of failure of the _ neutron monitors is not increased.
1 l.
Also, as discussed in Number 20 " Effects on Safety", the source and h
intennediate range neutron monitors cannot become seismically induced missiles and contribute to the probability of malfunction of other equipment 0
important to safety.
l l
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Safety Evaluation
.pdE OF No. ECN L6186 NEp 6.6 Safety Evpluation 23.
(continued)
As discussed in Number 20 " Effects on Safety", the addition of the redundant shutdown monitors will enhance the capability of the high flux at shutdown alarme since they will be automatically adjusted downward as the background neutron flux level reduces.
This will reduce the human element involved.
~
The alarm setpoint is the lowest previous value of the product of 'the alarm g3 ratio and the average neutron rate of the last 1200 counts, Reference 22.
Surveillance testing will ensure that these alarm setpoints are operable.
As discussed in Reference 20, the operator does not depand entirely on this alarm setpoint but has audible indication of increasing neutron flux from the audible count rate drawer of the NIS system and visual indication from counts per second meters for each channel on the main control board and I
source range drawer.
24.
Would the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR?
/[7 Yes 457 No Justification:
As discussed in the~ Question 23 Justification, the probability of failure of the neutron monitors is not increased. Hence, the consequences of a reactor power excursion at low-power operation will not be increased since the neutron monitors will be availabic to initiate a reactor trip.
Also, after a Condition II, III, Or IV fault, the operator will be able to rely on the neutron flux PAM parameter to determine whether certain equipment (such as reactor rods or safety injection, etc.) responded to the fault as required. This will allow the operator to take the necessary action to mitigate the consequences if that equipment did not respond as
. required.
- 25. Would the proposed activity create a possibility for an accident of a different type than any evaluated previously in the SAR?
((7 Yes fj7 No i
1487E m
F
e-I LC!J NO.1Cew, AS Sheet 19 G4 op Safety Evaluation No. ECN L6186 NEP 6.6 Safety Evaluation et Justification:
The impacted instrumentation will perform the identical function' following the modification as prior to the modification with higher availability and reliability. The intermediate range high neutron flux trip and the source-range high neutron flux trip will co.ntinue to function as described in FSAR Sections 7.2.1.1.2b and c, respectively. PAM instrumentation will be installed as described in #20 " Effects on Safety" and will not be susceptible to single failure. As a result, the operator will have access
-to reactor power level information and will be able to make decisions based on that information to avoid any possibility of a type of_ accident not previously evaluated in the FSAR.
Also, since the safety functions of other systems and structures are not affected, no new accident can be created by this modification.
In the event of a failure of the shutdown monitor alarm, the audible count-rate and visual indication is still available to the operators.
The neutron flux signal to the shutdown monitor is through a pulse buf fer whose input is gs.
optically isolated from its output, Reference 22.
26.
Would the proposed activity create a possibility for a malfunction of equipment of a different type ti.6n any evaluated in the SAR?
L[7 Yes 457 No Justification:
Implementing the proposed activity is necessary to comply with the USNRC's Reg Guide 1.97 Rev. 2.
Following this modification, the affected E3 instrumentation will perform its safety function and comply with the design requirements as described in Reference 1, 2, 5, 6, 7, 8, 10, 12, 14. IS, and 16*,
Therefore, the proposed activity will not create a possibility for a malfunction of equipment of a different type than any evaluated previously in the FSAR. Additional information for each component affected is provided in the following table:
g3
- PMT 62 will be successfully completed prior to declaring the new instrumentation operable.
t 1487E e
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Safety Evaluation g
No. ECN L6186
.l NEP 6.6 Safety Evaluation 26.
(Continued)
.l l
lHAS A NEW MALFUNCTION ITEM l COMPONENT l DESCRIPTION OF CHANGE IBEEN CREATED?
1 l Source / Intermediate Neutron l Replace non-1E qualified with lNo, New component isl
.l Detector and Cabling l detector which has output a sealed, pressure
^
l Channel I, KE-92-5001 l proportional to reactor power
. tight, qualified l
l l device and contains li
)
l l
ltwo identical, ll l
l l redundant fission l j l
l l chambers l
1 2
l Source / Intermediate Neutron l Replace non-1E qualified with lNo, New component isl l Detector and Cabling l detector which has output la sealed, pressure l l Channel II, KE-92-5002 jproportional to reactor power
. tight, qualified l
l device and contains,
i l~
l ltwo identical, l
l l
l redundant fission l
l' l
l chambers l
3 l Unit 2 Primary Containment l Replace 75 ohms triax with jNo, a qualified l
l Penetration No. 23 l50 ohms triax to match new l50 ohms feedthrough l l Channel II l detector cable impedance lshall replace the l
1 l
lexisting 75 ohm l
4 l Unit 2 Primary Containment l Replace complete non-qualified iNo, an electrically l l Penetration No. 31 lwith 1E qualified penetration l qualified one shall l l Channel I l
l replace the existing l t
l l
Inon-aualified l fi 5
l Unit 1 Primary Containment l Replace 75 ohm triax with 50 lNo, a qualified l
l Penetration No. 48 l ohms triax to match new l50 ohms feedthrough l l Channel II l detector cable impedance.
lshall replace the l
l l
lexisting 75 ohm l
6 jUnit 1 Primary Containment l Replace complete non-qualified lNo, an electrically l l Penetration No. 43 lwith IE qualified penetration-l qualified one shall l l Channel I l
l replace the existing l l
Inon-oualified l
7 l Channel I Signal Amplifier l Replace non-qualified with IE lNo, the function is l lKM-92-5001A - U1 l qualified amplifier to mate lthe same as the l
lKM-92-5001 U2 lwith new detector l existing design and l l
l lnew design is-l l
1 loualified l
8 l Channel II Signal Amplifier l Replace non-qualified with IE lNo, the function is l.'
IXM-92-5002 U1 l qualified amplifier to mate lthe same as the l
lKM-92-5002A - U2 lwith new detector l existing design and l l
l jnew design is l
l l
laualified l
1487E
_.,w.
a.
9,-.
y
.-m m
m
.m m
e
_,,c
-w r
y-ge
.e-w-qyww--r-g w
-g
.- ~.
l I?
4 a,
ECN NOdfd M f k Sh:,t 21 C6. OF_.
safety Evaluation No. ECW L6186 NEP 6.6 Safety Evaluation l
L I
26.
(Continued) l l
l lHAS A NEW MALFUNCTION
' ITEM l COMPONENT l DESCRIPTION OF CHANGE IBEEN CREATED?
9 l Shutdown Monitor l Add new device to automatically lNo, as discussed in l l Channel I l adjust high' flux at shutdown i No. 20, even if this lKis-92-5001 l alarm down Inew device were to l.
l l fail totally, the l
I l
l increasing audible l
l l
l count rate would l
l l
lstill be available l
-l l
lto alert the oper.- l l
l lto the event l
10 l Shutdown Monitor l Add new device to automatically lNo..as discussed in l l Channel II l adjust high flux at shutdown No. 20, even if this l lXIS-92-5002 l alarm down new device were to j l
l l fail totally the l
l l
l increasing audible l
l l
l count rate would l
l lstill be available l l
l lto alert the oper.
l l
l lto the event l
11 l Source Range Drawer l Replace Westinghouse with lNo, new qualified i
l Channel I, l Gamma-Metrics l drawer has the l
(XX-92-5001 l
lsame output and l
l l
l function, except l
l l-l lhigh voltage does l K l
l lnot have to be l
l' l
ldeeneraired l
12
.l Source Range Drawer l Replace Westinghouse with lNo, new qualified
-l l Channel II,
)Camma-Metrics l drawer has the l
IXX-92-5002 l
lsame output and l
l l
l function, except l
-l l
lhigh voltage does l
l l
lnot have to be l
1-l ldeeneralred l
13 lIntetuediate Range Drawer l Replace Westinghouse with lNo, new qualified l
l Channel I, l Gamma-Metrics l drawer has same l
lXX-92--5003 l
l output and l
l' I
Ifunctions l
- 14. l Intermediate Range Drawer l Replace Westinghouse with lNo, new qualified l
l Channel'II, lGamm4-Metrics l drawer has same l
lXX-92-5004 l
l output and l
l' I
! functions l
15-llE Optical Isolator, Unit 1 lNew device to isolate Main lNo, any malfunction l l Channel I, l Control Room remote shutdown lof this will result l lXM-92-5001B l
lin loss in detector l l
l l signal but there is l l
l lan associated l
l l
Iredundant channel l
1487E
ECN NO. $No LS-Eh*L 22
~
Safety Evaluation
_49_OF No. ECW L6186 NEP 6.6 Safety Evaluation
.I 26.
(Continued)
,l l
lHAS A NEW MALFUNCTION ITEM l COMPONENT l DESCRIPTION OF CHANGE lBEEN CREATEDt 16 11E Optical Isolator, Unit 2 ; New device to isolate Main Wo, any malfunction l Channel II, l Control Room remote shutdown
. of this will result lXM-92-5002B l
lin loss in detector l l
l l signal but there'is ll l
l.
jan associated I;
s 1
I tredundant channel
[
17 l Appendix R Source Range l Replace Westinghouse source lNo, new source /poweri 1
l Channel (detector, pre-amp l range channel with optically l drawer is a j
land drawer),
l isolated output from l qualified lE device l lXI-92-5
. 1E-qualifled Gama-Metrics even though its L l
'l
- amplifier and new source /
isignal and powerf l
l l intermediate drawer l cabling is not.,
l l
l
-l designed IE.
The.
l l
l l power source is the l l
l lsame as previous.
l l
l l design vital instr. l l
l l power. The increased l L
l l
l reliability of the l
l l
l lnew fission chamber l l
l l
l detectors and th'e l
l l
l l routing of al1~
l R l
l l
l cables and elec'-
l L
l l
ltronic mounted abovel l
l l
lthe Design Basis:
l l
l l flood level, Ref. 33l l
l lwill offset the,
[
l l
l deletion of the l
3 l
l l backup source range l c
l l
l l
l detector and will l
L l
l lnot increase the.-
l l
-l l
l susceptibility of l
l l
l l Tech. Spec. 3.3.S'.5.l 18-l Components 1 through 17 l Increased load to the 120 VAC lNo, prior to.
l lt l
l Vital Instrument Power Board l energizing the equip l l
l lthe capacity cale.- l l
l lfor the Vital AC l.'
l l
lshall be performed l 19 l Components 1 through 17 l Increased load to the 125 VDC lNo, prior to l
l l Vital Batteries l energizing the equip l l
l l
lthe capacity esic.
l l
l~
lfor the Vital AC.
l l
l lshall be performed l
1487E
bCN NO.L7//A4 g3 Shzt 23 ACF Safety Evaluation No. ECN L6186 NEP 6.6 i
Safety Evaluation l
l 27.
Would the proposed activity reduce any margin of safety as defined in the 5
basis for any technical specification?
' [7 Yes 457 No f
Justificationt L
1.
The containment integrity and penetration operability requirements are I
addressed in Tech Spec Section 3/4.6.1.
Implementing the proposed activity (Stages 2 and 3) will require upgrading the two electrical penetrations identified in " Systems Structures, or Components Affected" 1
-(#18).. Testable penetrations Surveillance Instruction (SI-157) will be L
implemented following the modification to assure the above Tech Spec requirements are met.
2.
The note in Table 3.3-1 of the Tech Spec stating "High Voltage To Detector May Be Deenergized Above the p-6 (block of source range reactor 1
trip) Setpoint" is no longer required. This was only required when the previous design source range detector was being deenergized. The new L
detector and amplifier design will be full range and will not be deenergized. However, in order to reduce operator and site procedures e
I impact, all source range outputs can be disabled above the p-6 setpoint.
Therefore, Tech Spec table notation for Table 3.3-1 will be revised as p3 l
shown on Sheets 26 and 27 of this safety evaluation. Sheets 28-and 29
. will not be impacted since table notation is still-required.
Detector replacement will occur in Stages 1 and 2 described in Block 17, 1
3.
Tech Spec Tables 3.3-10 and 4.3-7 " Accident Monitoring Instrumentation" will be revised as shown on Sheets 30 through 33 of this safety evaluation to show the new PAM-instrumentation identified in " Systems, Jg Structures or Components Affected". The monitoring equipment identified will be installed in Stages 1 and 2.
4.
The intermediate range neutron flux p-6 permissive engineering units will change from amps to percent power to provide the plant operators more
-meaningful information in the main control room. This will result in a q
revision to Tech Spec Table 2.2-1 as shown on Sheets 34 and 35 of the BI-safety evaluation. Reference 13 justifies this scale change. The scale change will be perforwed during Stages 1 and 2.
4 l
l t.-
1487E
t E C H N O.l.(,/ h a 3 l
Sht t 24 7 / OF Safety Evaluation No. ECW L6186 l
NEp 6.6 r
Safety Evaluation 27.
(Continued) 5.
A demonstrated loop accuracy analysis (Reference 26) will be performed to prove that the allowable values for these reactor trips specified in Table 2.2-1 are acceptable. The~new monitors will be installed during Stages 1 and 2 (see Special Requirements No. 3).
Technical Specification Section 2.2.1, " Limiting Safety System p3 :
Settings" (Bases), intermediate and source range nuclear flux states "no credit was taken for operation of the trips associated with either L
.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 i
the Reactor protection System."
l 6.
The margin of safety as specified in the bases of Tech Spec 3/4.9.2 L
will not be reduced. As mentioned in Block #20 " Effects on Safety",
j two neutron monitoring channels will be available at all times during H
refueling while this modification is implemented, ensuring changes in the reactivity condition of the core that may occur will be detected.
L This.will be ensured during Stages 1 and 2.
l-7.
Contrary to Revision 2 of this safety evaluation.. Tech Spec.3/4.3.3.5 l
Table 3.3-9'and 4.3-6 Sheets 39 through 42 will not be revised. Source range instrumentation is still available for remote shutdown.
Intermediate range and decades per minute indication will be available, however, is not required for remote shutdown. Therefore, a Tech Spec gg s
revision is not warranted.
8.
Tech Spec Section 2.2.1 "Limitins Safety System Settings, Bases", the reference to intermediate range current level, shall be revised as L
shown on pages 43 and 44, since the intermediate range will read in L
percent power.
Based on the discussion above, implementing ECU L6186 will not challenge or degrade the margin of safety as defined by any of the Tech p
Spec bases.
L 28a. Special Requirements 1.
The Fire Hazard evaluation for the Unit 1 conduit installation for this ECN has to be completed prior to installation beginning in the R3 outstanding locations. The outstanding locations are inside the Unit'l annulus and containment vessel All elevations in the Auxiliary Building have been evaluated.
l u
j, 1487E P
i
2..
j ECN NO. Uo/Ar. # E She0L 25 12L.0F Safety Evaluation No. RCN L6186 NEP 6.6 Safety Evaluation-I l28a. (Continued) l 2.
The unverified assumption in pipe rupture calculation associated with each individual channel, SQN CEB-SCG-4E00168, will have to be resolved before that channel can be declared operable.
3.
Prior to the. Unit 1 and 2 Camma-Metrics equipment being installed, the Instrunent Accuracy calculation (SQW-EEB-PS-TI-28-0001) shall be completed.
4.
Equipment / channel cannot be considered operable until voltage drop calculation SQN-VD-VAC-016 can be verified with as-constructed cable gy lengths for that channel.
5.
Installation of detectors, detector cabling, and electrical penetrations cannot begin until EQ Binders SQN EQ INM001 and SQN I
EQ PENE005 are issued for this ECN.
6.
Prior to declaring-any of the system operable, this safety evaluation will be revised to reference the CRFSAR for the PAM implementation-(FSAR Section 7.5 in to-to).
28b and 29 Based on the safety evaluation provided above, it is concluded that no unreviewed safety question exists as a result of implementing ECN L6186.
1
?
e.
4 1487E
.-~
ECN NO.Lc E R.E
- NE9 4.(, SQt 4 l R3 P
Ol_.0F sde E.ydodde TABLE 3.3-1 (Continued)
Re. EC9 L41%
TABLE NOTATION a
With the reactor trip system breakers in the closed position and 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.
,#The otorisionA rf haeffication_3.0.8 ** not applicable.
- {Av'WA*AO@A4_U_,,'y'be di bh above the P-6 (Block of Source R3 Tange eEN/
ACTION STATEMENTS 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 h0drs, 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, STARTUP and POWER OPERATION may proceed provided the following conditions are satisfied:
3)
The inoperable channel is placed in the tripped condition a.
within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
lR51
)
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 Power 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
-incore 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.
.i; O'
September 17, 1986 SEQUOYAH - UNIT 1 3/4 3-5 Amendment No. D
,,_____,.-m__
. - - - - - - - - - - - - - ~ - - " - - - - - - - - - ' - - " ' " " " - ' - - - - - - - - - ~ ~ ^ ^ - - ^ " '
1 ICNNO.Wa s
%P GJ. sheet 17 i K3:
SASd EynivatN as y
n-
) M OL Ho,ELN L.M%
_ TABLE 3.3-1 (Continued)
_ TABLE NOTATION-a With the reactor trip system breakers.in the closed position control rod drive system capable of rod withdrawal, and fuel,in the
'l the reactor vessel.
j an The channel (s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped
' condition.
T ie L nenvi tian, ic on_3.0.4 are not applicable.
'Y1'Y2'i:'r a be if-YN;;' :d bove the P-6 (Block of Source b
Range ReTctor Trip se n.
~
ACTION STATEMENTS ACTION 1-With'thenumber'ofOPERA'BNEchan$e'Isone'lessthanrequir the Minimum Channels OPERABLE requirement, restore the inoperab 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 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 les Number of Channels, STARTUP and/or POWER OPERATION may proc provided the following conditions are satisfied:
l 1
The inoperable channel is placed in the tripped. condition a.
within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />..
. lR3h
- 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 />,
lR30 for surveillance testing per Specification 4.3.1.1.1.
c.-
Either to 75%,of RATED THERMAL POWER and the.
Neutron Flux trip setpoint is reduced to less than or equa,l 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 PCWER TILT RATIO is monitored at leas,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 re three detectors symmetric power, distribution obtained by using incore detectors in the four pairs of symmetric thimble locations at least once par 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.
i september 17, 1 86-SEQUOYAH - UNIT 2 3/4 3-5 Amendment No. 39
.W-
,. ;i._ y. :%:. ;55}yijf
. n.
2.
- r
.% i 0 T
J. y '
,,-,v,-------
.a-n--.-
n a
n-...,',_
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.--.-.-.,.e wn.e,--
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. pef * (c ' [.S eA1.1ht R3 ~
re. oIeeo per.d,53 c u35 o^i ial'
. Ss.9 Ev.fsdu l{I:
'Tks. thndge aib }ce3v S ai.1 c.y A} omhh S e th}* **
t A
g.
g;, gjfes -
TABLE 3.3-1_
x28s REACTOR TRIP SYSTEM INSTRUMENTATION
^
h FUNCTIONAL U TOTAL NO.
CHANNELS CHANNELS APPLICABLE MINIMUM.
-4 OF CHANNELS TO TRIP OPERABLE MODES ACTION-w 1.
Manual Reactor rip 2
1 2
1, 2, and
- 1 2.
Power Range, Neutron lux 4
2 3
1, 2 3.
Power Range, Neutron Flux 2
High Positive Rate 4
2
'3 1, 2 2
4.
Power Range, Neutron Flux, 1
High Negative Rate 4
2 3
1, 2 S
2 R
5.
Intermediate Range, Neutron Flux 2
w 1
2 1, 2, and
- 3 6.
Source Range, Neutron ~ Flux A.
Startup AE.L GTE B.
Shutdown 2
1 2
2
, and
- 4 7.
Overtemperature Delta T 0
1 3, 4 and 5 5
Four Loop Operation 4
4 2
8.
Overpower Delta T 1, 2 6,
i Four Loop Operation l R45 4
2 3
2 6,
9.
Pressurizer Pressure-Low i
~
4 g,g3 g
2 10.
3 1, 2 NE Pressurizer Pressure--High ah 6
4 2
11.
3 1, 2 vt r
EG Pressurizer Water Level--High 6
E 3
?
Y 2
M,i 2
1, 2 y
,+
E
- 5. ?
1 Q-
~
e
~
t.
x i
4
.\\
Thi 5 thAN K 13 gh fodgeri Pe p rd ffe Jisicd 5sh h
^
.acp C.C."f[se.f Q4.
S*/e fr (v4/**/J 5ddy. u ab nt fy. 37, y 9
c i
p.. sw -c ew TABLE 3.3-1 mj REACTOR TRIP SYSTEM INSTRUMENTATION o5
[
MINIMUM TOTAL NO.
CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNI 0F CHANNELS TO TRIP OPERABLE MODES ACTION
.J c
U 1.
Manual Reactor 2
1 2
1, 2, and
- 1 2.
Power Range, Neutron x
4 2
3 1, 2 2#
. 3.
Power Range, Neutron Flux 4
2 3
1, 2 2W High Positive Rate
't -
4.
Power Range,. Neutron Flux, 4
2 3
1, 2 2#
High Negative Rate
{
5.
Intermediate Range, Neutron Flux 2
1 2
1, 2, and
- 3 Y
6.
Source Range, Neutron Flux Af A.
Startup.
2 2
, and
- 4 8.
Shutdown 2
0 1
3, 4 and 5 5
7.
Overtemperature AT Four Loop Operation 4
2 3
. 1, 2 6,
8.
Overpower AT R33 y
Four Loop Operation 4
2 3
,2 6,
{E'e [
9.
Pressurizer Pressure-Low 4
2 3
1, 6
R33 y
tr M E
'10.
Pressurizer Pressure--High 4
2 3 '
1, 2 6
f c*
g,E 1
m 11.
Pressurizer Water Level--High 3
2 2
1, 2 7#
h 30 39 E
e 0
a*
L/
G
'ML*t.fJla.jo}t3 . Ss(e.ly Eval.4,4. 2*. Ecs) U(8C- .m -TABLE 3.3-10
- c
-O ACCIDENT HONITORING INSTRUMENTATION >x I_NSTRUMENT., MINIMUM g REQUIRED NO. CHANNELS OF CHANNELS Z 1. Reactor Coolant T OPERABLE itot ( e Range) 2. Reactor Coolant T 2 s Cold (Wide Range) 1 3. '2 Containment Pressure (Wide Range) 1 4. Refdeling Water 'S'torage Tank Level 2 1 ~ 2 l R *,0 5. Reactor Coolant Pressure (Wide Range)- 1 i
- 6. ~ Pres'surizer LebeE (Wide Range) 2 i
1 .ma Steam Line Pressure 2 lR$d 7. 1 w 8. SteIm Generatfr Le'vs1 - (Wide Range) 2/ steam line 1/ steam line s* 9. Steam Ge*nerator Level:- (Narrow Range) 1/ steam generator 1/ steam generator x ls3 di
- 10. Auxiliar'.'Feedwater Flow Rate' 1/ steam generator 1/ steam generator i
on "' \\ e* 1/ pump
- 11. Reactor Coolant System Subcooling Margin Monitor G/ pump
.g ' i
- 12. Pressprizer"PORV Positiort. Indicator
- 0 Jg 1
- 13. Pressurizer PORV Block Valve Position Indicator **
p 2/ valve # .i 1/ valve
- 14. Safety ialve Position Indicator 2/ valve g-g 1/ valve
- 15. Containment Water Level (Wide Range) 2/ valve #
1/ valve
- 16. In Coie Thermocouples 2
1
- 17. Reactor Vessel Level Instrumentation System ***
4/ core quadrant 2/ core quadrant .Qm Rg,, -_/f- $**ru ^.'L. fee ~<dae fe. Q Noefear-J&^r. 2' 'l 1 ,,.y,L 1 ADD lR50 ' =-
- Not applicable if the block valve is verified in th"Not applicable if the as f
g 3 g
- This Technical Specification and surveillance requirement will not be imp e closed position.
s ., Instructions are developed for the use of this system as committed Pg z NUREG-0737. emented until Sequoyah Specific - SG e TVA response to Supplement -1 of. R50 - ,'-At least one channel shall be the acoustic monitors >S ( .n c. - r. Q
- l'^
- ," ' ~~
Af:.-;I-c.g, * ~ ~,
I '~ _ TABLE 3.3-10 ~ g/' t'.[ _ funf0 N S*NI E 'L*f f og ACCIDENT MONITORING INSTRUMENTATION Ecs) ~ L (18[ E MINIMUM. _ INSTRUMENT REQUIRED NO. e CHANNELS-OF CHANNELS {
- 1. Reactor Coolant T OPERABLE Hot ( de Range) 2
- 2. Reactor, Coolant JCold (WId* Range) m 1-2
- 3. Containment. Pres {sure (Wide. Range) 1 2
- 4. Refpgl ling. Water, Storage Tank Level
[ 1 2 5..Reacter.Cgol,an(Pressure (WideRange) 1 2 6.Presspr,f.zef.Leve[(WideRange), 2 'h 1
- 7. Stea,m,Line,, Pressure 1
2/ steam line 4
- 8. Steam, Gene,ra, tor Level 7 (Wide Range) 1/ steam line 1/ steam generator
}
- 9. SteamaGenerator Level,(Narrow Range) 1/ steam generator.
~ T
- 10. Audliary 'Feedw5ter F10w R8te 1/ steam generator.
1/ steam generator m vi ~i 6 1/ pump
- 11. Reactor Coolant System Subcooling Margin Monitor 1/ pump
[ 1 0 -3 '12.Pressuriz'eIPORVPositionIndicator* o p ....;_..i.,, 2/ valve # ^ 1/ valve E i tp
- 13. Pressurizer PORV Block Valve Position Indicator **
2/ valve O g
- 14. Safety Valv's PosTtion Indicator 1/ valve
.n. 2/ valve # '15. Con'tainmenf Water Level (Wide Range) 1/ valve 2 16.InCoreThNuocouples. 1 i 4/ core quadrant
- 17. Reactbr'i/enel l'evel Instrumentation System ***
2/ core quadrant 2 g a,
- 8. Souree-i.74er_dedale.' Aang[Neeleaa-1.,,jnda,1,.
1 2 R- = ~ x-
- Not applicable if'the associated block valve is in the closed position
[ &E
- This Technical Spec.ification and surveillance requirement **Not applicable if the 3l e valve operator removed.
will not be implemented until Sequoyah Specific - Instructions are, developed for the use~ of' this system as committed to in the TVA respo i F.E. g NUREG-0737.. R At 1 east'one channel shall be~the acoustic monitors. w u$ E s; s.-
- 5 "..
. - f' ' 'it:ff d} \\. s' ~ _, i. O .P + a .s.;
- 3.. i.
~. .:y.., y. c ~ l~.]. 5*' m. W j '}' l s ^ ' g.. L... ~ ~~ ~ ~ ~ ~ " ' ~~
-y =- w a :.uw Wje .ra 4 eu s..i..,_ .M*. _ TABLE 4. 3-7 EcAsLCrSS. sn . E8 ACCIDENT HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENT s ,7 _ INSTRUMENT CilANNEL CifANNEL CllECK .f8 CALIBRATION E 1. Reactor Coolant THot ( ide Range) M ] R 2. Reactor Coolan't TCold (pdeRange) M 3. Conta.inment, Pressure (Wide Range) R M R 4. Refueling Water Storage Tank. Level luso M R 5. Reactor Coolant;. Pressure '(Wide Range) M 6. Pressurizer Level R lg30 M 7. Steam.Line Pressure R M R 8. Steam Generator Level _- Wide M R 9. Stears Generator Level - Narrow R s M R z y,
- 10. Auxiliary Feedwater Flowratel z
M O R
- 11. Reactor Coolant System Subcooling Margin Monitor J
'r-M
- R e
- 12. Pressurizer PORY Position Indicator'
'o iii; H .R 1 e
- 13. : Pressurizer PORV Block Valve Position Indicator M
R
- 14. Safet4 Valve Position Indicator M
R
- 15. Containnient Water Level (Wide Range)
M R
- 16. In Core Thermocouples
-lR50 M R
- 17. Reactor Vessel Tevel Instrumentation **-
gp M g /t R .y }, _ .Teoran. {.T &.e. leek Asye Akxlens Zafr omhhh ~ _ M ~ ~ lRSo.
- This TecjhdicdI-5pecification and surveillance requi
$ g. Instructions,are developed for the use of this system as committed to in the TVA response to S g g NUREG-0737., R50 ement 1 of c. g ;;:..
- W SE
. =. . * *El. ,it - ' l[} g; -i".. . Y; I ~ 'N ? . L- -. "^~ ~ ~~
~ Ejo $.( b.$ Esa.f)V .s. G. Ao. y s d. w E _TABtE 4.3-7 Ecn L t/ff o8 ACCIDENT' MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMEN s f INSTRUMEf;T CHANNEL CHANNEL CHECK CALIBRATION _E 1. Reactor Coolint Tifot (Wide Range) M [ 2. Reactor CoolanLTCold (Wide Range) R H R 3. Containment Pressure (Wide Range) 4 M
- 4. " Re'fue' ling' Water' Storage Tank Level la38 R
i M R React 6r Coo' larit'Pressdre '(Wide Range) 5. a e M R 6. Pressuriz'er Level b38- ~ M-7. Steam L'ine' Pressure R M R ~ 8. Steam 'Generatdr Level - (Wide) ~ ~ M R 9. StEsm ' Gen' erat'oi LEve1 "(flarrow) R I M R
- 10. 5ukijiary FeeMater Flowrate w
-M 3R [o E R
- 11. ' Reactor Coola'nt System'Subcooling Margin Monitor t
M
- 12. 'Pressuffzer PORV Position' Indicator E
o M s
- R
- 13. Pressurizsh PORY Block Valve Position Indicator
.M R
- 14. Safety-Vilve Position Indicator-
~ M R
- 15. Containment Water Level (Wide Range)
M. R 16; Ir' Core Thermocouples . h38 ~ M 'R _ 17* Reactor _Ve g l Instrumentation System
- M R
/8 b "'<=. :Anfer 4 A 4. Ra ye. No g? e lasfee -lat,an M- ~ =-
- =
_ w /Z 38 @)
- ThisTechnical"Specificationandsurveillancerequirementwillnotbeimplemented E g.
Instructions are developed for the use of this system as committed to in the TVA response to untti Sequoyah Specific- .$. "a NUREG-0737.' ent 1 of 138 a. 3* ,f ' i{; g, i
g m J Alt l' d. 4 Jtea.f 3 4' ~ 3afefy fue/osf$,. [g3 ^). ccn weg g _ TABLE'2.2-1 (Continued) .o8 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPO W i x FUNCTIONAL UNIT E TRIP SETPOINT ~ ALLOWABLE VALUES g
- 13. Steam Generator Vater Level--low-Low,,
1 18% of narrow range instrument span each steam geneiator > 17% of narrow range instrument R20
- 14. Steam /Feedwater' Flow
~ span each steam generator Mismatch and low Steam < 40% of full steam flow at < 42.5% of full steam flow at Generator Wate'r Level RATED THERMAL POWER coincidentRATED THERMAL POWER coincident with steam generator water level ~'0' t 25% of narrow range instru. with steam generator water level ment span each steam generator > 24.0% of narrow range instro-
- 15. Undervoltage-Reactor ment span-each steam generator 1 5022 volts each bus Coolant Pumps.,,.
t 4739 volts each bus fa89 I 'y 16; Underfrequency-Reactor i .Coolan{fumps. > 56.0 Hz each bus m > 55.9 Hz each bus
- 17. Turbine ' Trip
' N. + ..A.,. Low-Trip System . Pressure > 45 psig .. o > 43 vsig 'B. F Turbine-St6p' Valve ,yl,osurg,, - > 1% open --> 1% open r. 2 \\/N g
- 18. Safe Injectibn Input from',SF Not Apy.Licable Applicable
> I x io-s 2o -4 g' g,
- 19. Intermediate Range: Neutron
-10 k & x to g 1 1.c 10 as5- -11 Flux - (P-6). Enable Block 15 10 0p Source Range Reactor Trtp g ,, mg, 7gggp
- 20. Power' Range Neutron Flux 7::
\\ < 10% of RATED (not P-10) Input to' Low Power THERMAL POWER < 11% of RATED Uf Reactor Trips Block P-7 THERMAL POWER z '0B 9 E g.. h ~ E = g.h f 4 ~ [9
. s m ~ g. .i p tf C.C S Q )5j%3 Safe.Q Evrlue b-~~- u, f5 Ma, 8 TABLE 2.2-1 (Continued) sw & ttf f-2 f TUNCTIONAL UNIT - _ REACTOR TRIP SYSTEM INSTRUMENTATION TRIP ,~. TRIP SEIPOINT E
- 13. Steam Generator Water Att0WA8tE VALUES A
Level--tow-Low 1 18% of narrow range instrument span each steam generator 1 17% of narrow range instrument m
- 14. Steam /Feedwater Flow span each steam generator tt Mismat'ch hnd low Steam
< 40% of full steam flow at i Generator.. Vater Level RATED THERMAL POWER' coincident < 42.5% of full steam flow at with steam generator water level RATED THERMAL POWER coincide 125% of narrow range instru-wit $ steam generator water level ment span-each steam generator 3 24% of narrow range instru-
- 15. Undervoltage-Reactor Coolant Pumps' 3 5022 volts each bus ment span-each steam generator i
3 4739 volts each bus g..
- 16. Underfrequency-Reactor
,? Coolant Pdinps 1 56 Hz each bus 1 1 55.9 HI each bus
- 17. Turbine Tr.ip A. 'l.ow Tri,p System Pressure 1 45 psig B.
Turbi'ne^Stop Valve 1 43;psig. ClosbFe '> 1% open > 1% open
- 18. Safety Injection Input -
frontiESF - Not Applicable qvl6 b 2 i sc to g.4 Not Applicable [ j
- 19. Intermediate Range Neutron p
Flux, P-6,' Enable.' Block " 10-10 2 c. X to ,4) a I oy Source Range Reactor Trip og ggeo rue = % Ruen. ^ 5 x 10 ' e '9 ~
- h.
- 20. Power Range Neutron Flux y go %m g m
I g (not P-10) Input to Low < 10% of RATED Power Reactor Trips THERMAL POWER < 11% of RATED THERMAL POWER E u" m l ." g ~ Block'P-7 l G 6 \\ 8*~. r-r ~ U \\
- [
l S l o
- ..~.
l$ s .t 5 q e ,v-
UNREVIEWED SAFETY QUESTION DETERMINATION Sksd 3 6 [ , TvA 10b$1 (CN 005 2 81) SHEET #1 ECN NO. L f= la h M To: Secuovah NnClear Plant, Daisy, TN 60 / $6R.O ' " " ~ M OF lOEN TiriE si (Cg"'"g't '~ R fM l enErAnED nEviEwCD AreRovED
- SQP '85 0117 503 RtSE A 2.aR4 0C AI tark.k &.,9,d o-a i
1 \\ R b 2 at1 Y' S s a % As y vy A S >R INITI AL I s PROJECT SQN AFFECT' NIT (S) ld g ECN NO. I bI E ECN DATE-A l ** I , u n, m nf YES/NO SHEET NO. Meft/DCR NO. I I6 G DATE-UlREMENT(S) M0 OTHER DATE POTENT CHANGE /E3 O r-t1 % Le n e t ~n Fa r e a Shee t - Y strtRtNCtS, CnnfrmNoI] wi+IA derrufH . SGEP elec ( t I #A 0 \\ N,.,,,/ A A V N $ DESCRIPTION OF CHANGE Neute M t'
- rd m
? t i o n ra de. +ke ennee Y+#$eleke4 ton < n u r ee. f &n 0' meek Cla s s e n v. iE ancj seih mi[ e I = nru b ceo vin, en 'h u Ree Guide I.'lY.'Yh is mociihehohl$ kcW her +o n A h roche reRuble co<:fn ec eclent m onhe i n[*inMk enhnfa+ia,t. M3.. 6 ee viomen+ ud iI c.o vir e 120-V ClassIEN & bd wil I clea b ahnroximate3 ar on n. Tk co vioEen+MI I also l l' 6 Environme nbilu T, ml.m m C u a.,9 ' 7 A a .c ..a A .m., _e .1
- l..
\\ '1$;?. *L. JC (ATTACHMCNTS): NO.YE5 CHIEF. ARCHIT[CTUR AL DESIGN DR ANCH.W4Cl26 C.K CHIEF NUCLEAR ENQiNEER.W10Cl26 C CHIEF. CIVIL ENQlNEERING DR ANCH,W90224 C.K CHIEF. MECHANICAL EN DES DRANCH 302 PT.K 'si 5 CHIEF. CIVIL EN OCS DR ANCH. W3C:26 C.K CHIEF. QUALITY ASSUR ANCE BR ANCH W))Cl26 C.K ' [. CHIEF. COST PLANNING AND CONTROL STAFF.W12 MAN AGER OF CONST RUCTION, E7B24 C.K , CHIEF. ELECTRICAL ENGINEEntNO BRANCH, WBC126 C K CHIEF. ELECTRICAL EN DES DRANCH,W20224 C.K 4 CHIEF. MECHANICAL ENCINEERINO BRANCH.W100225 C.K PLANT 3UPERINTENDENT ' m*/.- I MEOS.E4037 C.K OIRECTOR. NUCLEAR POWER OlvislON. 716 ES<, " ^ ^ ^ - - " ' ^ - ~ ' ^ '
IV A lvdbl A (LN DCS I SO) Sheet % UNREVIEWCO SAFETY GUES? ION DETERMINATION I LG I Z6 R.o ECN NO. LC I A s. el IDENTIFIER Project _ b Unreviewed Safety Questions u 1.15 the probability of occurrence or te onsequences of an accident or rnalfunction of eculpment or tc fety previously evaluated in the Safety Anal) Report inc ed? .......................Yes No_.)c Justificatlont Ne Aeuff n To rl n at 4t.sYeen doe 3 r%ct oeekorm ,.c,ma:A me,JMA LR +.;n +k. m m W i< m corstdeh[
- Tke, balman unaraded O ensure es A
e ev e k ',ti d f o ri m 1 [ d e c ; d O +. ' N J +h us ' orovide. _b e, aner'aYOfs ' ri d ek I t n_ orsma [OA fe et n e ell eL ' + b _co rNd t [e A ek ike (" C.. fan sh - O AY +ks. ca uJ CarvInmenY u; t t e x cee d +h e- ~ dh n reodri adnis of +Ae in.ev;nlent _ bt. I n % f P O $ct r e e$ Yhe ra ha l A sa occ urren t e t= O t. _6663 o vence 3 o an n [Cle nI n* val l uaf ed
- A "i~k.4, m e %, mn+ :n a a. \\
Io - / \\ \\ IIIi t 1 \\ "YX S .k
- 2. 15 the possibility' for'an accide$t br r5alf l' of a fferent D
~~ than any evaluated previously in the S ety aly Repor t crea di............. - Y No1 J_uttificationt _De rs'e u) e[hierneki jl 5[tasm i c ca L _Clau ' 16. ' "Ih e r c#e rS. ' +/ sus +~.h ll as eced - + g o at __ mmt d S$4n Ireme nIs. and C er e f.> of bl % _ _ [o r ~ tL e en*f e r-nA _rnal ursch n[ di h. r\\of creoY'e_h. a. ere ce is _a d d i dio A.. d 6 - c V$ fu a ht'o w'll a d e. w' 6 ensure N a t Y b e. add l'o n c lead C s I ' on ra e_ su steen uji l l nd n _b4Ve n[/ Pere a((cets. nru + o \\ ~ -.--
- 3. Is tHe margin of safcty as defined in the basis for any technical specifi, cation. -.. -
t hd uc ed ?.......................~..................... ; ........Yes - No d Thl < rnodlAen.. Hon... . server +o imarove +Le ac elele -- 1mo ni to r ;n a en on6; ti+w o f +b e. olan h Thu s. +Le Mr 3 /4, 3, 3,.7 pp c,,;g,,@ln. o#Aggjjg; 3afe FE a9 de[le.ed inO 4,cg ,. 7,,, c IV%n 9nr.;- Incfr ame d Son.f i> nert redoeed.' Pee.
- [,
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t vA 10b51.u (EN DESd 80) ,1 y UNREVIEWED SAFETY QUESTION DETERMINATION h8, s . Sheet 0 A ECNNOJ.reits e3 gy $5._ _OF _ "*'E" 1 Special equirement(s) or Precaution (s).. ( , Marks Subject to Which Sa fety Evd ation.......... 7.. -..... This Page Applies) Preparer 8A 6 Additional in rmation......... ....U Reviewer _ 8% (initial) iJ '.. 1 4 ~i N. 90 ten Yic l 7 e E ' Sric' M vm w e x g \\ "dT . tdUC PR showlel kevietJ eitclee M +ech se.>ces are ; ele nf on ms+r _ h +,o E I a A c a e en;+e ;,o i .F - ch o.es L re, A s. A VI 1 o C-ore u n
- g..... _..... y _
y .... p y j .., x-....... -.r...; j 3- ;\\ Vf 3 / \\ \\ /AA f I \\ 'IA ( q p,\\ fy. a L \\ fjf e y p y,- X. A y 4 j z.._ _. y jy gr.... .y,z x 7 pg.. .g g.a.. mf. .g. m. .y. _.., _ g r f. ( .....y.........f......x y._ ~ j - \\. y \\ ./ g r g p L g \\,
- t..
SNhi.Ii .f..s... 'i e.. D ggma g ,,gg,,,ggap .N i -w
Q m .01& ? f)L41s OpaucGis S A**#'" } ,f (N. fit;f - 1980 h'S . Srps.ry Evss.esroou ' ~ w 9.m ->. s. v. ,,a s w i n era w W M ' ECh' W O h ippg.rt EVA' varied ABLE 3.3-9 'S i-E.. .S1 i} gp REMOTE SHUT 00WN MONITORING INSTRUMENTATICP y y-7- 78 ' c. INSTRUMENT READOUT MINIMUM ,i z MEASUREMENT-CHANNELS ~~ LOCATION --4 f uTF_C /f Z o/AT RANGE OPERABLE A i\\ T - w 1. Source Ran Mux 6 r1 U J ~ i NOTE 1 1 t6 1 x 106 cp 1 2. Reactor Trip Breaker ndication. i sig wzoo% CTP '{ i at trip switchgear ' OPEN-CLOSE ke' actor Coolant Temperatu 1/ trip breaker 3. NOTE 1 , Hot Leg 3 0-6 0*F 1/ loop lR80 4. Pressurize'r Pressure OTE 1 0-3000 psig 1 g 5. Pressurizer Level l g NOT 0-10.0% 1 lR80 y 6. . Steam Generator Pressure., l NOTE 1 t I m. s 0-1200 psig 1/ steam generator 7. 'SteamGeneratorLedel I s NOTE 2 or i I near Auxilary F. W. -100% i I -Pump 1/ste'am generator 8. Full Length Control' Rod Position i Auxilary; Instrument Limit Switches 1 insertion limit 4 5. - Room: Racks R41-44 On off switc.Vrod e 9. RHR Flow Rate: , NOTE 1 , 0-4500 gpm 1 10. RHR Te~mperature NOTE 1 50-400'F 1 (, 11. !fk Auxiliary Feedwater. Flow Rate NOTE 1 0-440 gpm 1/ steam nerator R80 n I i , i OR ?+ d 2
- m is i
I0t> 9 i . g,o ap >b - P -)g w ikt L e y e ~- . 's *$'E n lr$ w w. ,p-4.
4 -g ..' ] ~. g.,. p,3 f gucerIS $M A 5: W Qpgo }ll cat.;rus Di.seuss?. k.
- uttCaf,[GutaTj$[Jt3f y
,i' ' Srpory ; gyrwarou ^ ~ ~ ' ,f' .f Sgz yeo o ~g 7,7 O F W
- ' ' Q y,77 g yAtvA. 7/*^2 TABLE 3.3-9
~ e h Phh:. REf10TE SHUTDOWN MONITORING INSTRUMEN J h x L 'J \\ c IfiSTRUMEl READOUT HINIMUM gte,.medhht. - LOCATION ttEASUREMENT -i
- CHANNELS
- 1. 'Sourco,.
~ ~ RAN21 N an Nuclear Flux' ~ OPERABLE NOTE 1 1 to 1 x 106 2. Reactof Trip Br cp /%'A@7. RTP er Indication at trip switchgear -OPEN-CLOSE t ' 3. 'Reactcr Coolant-Tempe 1/ trip breaker-Hot 8.eg ture - fl'3TE 1 0-650*F ,4 1/ loop R67 4. Pressurizer Pressure I:0TE 1 5. Pressurizer Level 0-3000 psig 1 E1 , 0-100% (f 6. Steam Generator Pressure 1 w i NOTE 1 R67 Steam Generator level 0-1200 psig 7. 1/ steam generator 5 NOTE 2 or near Auxilary. W. 0-100% ~ Pump 8. Full' Length Control Rod.' 1/ steam generator Position 'timit Switches ' Auxilary Instrument -1'- Room: Racks R41-44 On of 1 insertion limit .. i. 9. RHR' Flow Rate ' switch / rod NOTE 1
- 10. RHR, Temperature 0-4500 gpm 1
NOTE 1 50-400*F 11. Auxiliary Feedwater Flow' Rate 1 h! NOTE 1 0-440 gpm N *R 1/ steam generator R67 h t m "s S 2 N ~ e .p M g ~ g p g gn B g 9- -d ' _, _,.. + ~ _p.mu*
a - d, g f.m. t ^ _ 7'Mrsi l?xwC**'/s Asa so$ced. YWN ~ 'l s}j, M, M 7# r AScufS/s4/ 3 .UGP G*E $CTbl l?3; /A> Se-cr/ *As 76.2' bavy EVAcas,7@ - 5 (:f.N,i' --Of
- y}{s- $dB ff EVALDe *T'*V
+ Moo. E5C.H : Lf 1SG p( p _ TABLE 4.3-6' m '@.-6; '~'r. 1
- s...
t ' t :t.- REMOTE SHUTDOWN MONITORING INSTRUMENTATION SE SURVEILLANCE REQUIREMENTS ~ i: I!b 5 ~ INSTRUMENT CHANNEL CHANNEL INTansce47G. _ CHECK . Z h fl. S$urce, Rang CALIBRATION' itu ear Flux H.. Reactor Trip Breaker dicition " 2. g g 3 M 3. Reactor Coolant Temperature N.A. Hot Leg P'ressurizer Pressure M 4. R g 5. Pressurizer Level' g M 6. R Steam Generator Pressure R M 7. Steam Generator Level R w e M 5 8. Full' Length control Rod' Position Limit Switches R M 9. RHR E10w Rate R y
- 10. RHR Tempera'.ture-
.g ^
- 11. Auxilidrv Feedwater Flow Rate M
R M
- 12. Pressurizer Relief Tank Pressure R
i gg
- 13. Containment Pressure M
i e YE M R n q ." 3 1 wh ~ - rn O -5 O' N -1 l& 9 ~.<. O' 't 1 O .d F-
- .a s
.w.. .. ~. ~ 3 s. v -,-.v, y 7 ,..#v c,.,
m-rm . l' ; ~ypts CMaHCC /S NO A"CSA' "90'#E;O F 4 g., t 4: pg ype Disco Ssica !O M'* 27 7 ' . ] hs? QSuwQ } g3 (; u S m unwsiw.- ~' pp yyg-fygg sy-Gv'A wA T/R \\ . qf, y, y ggg
- l E
REMOTE ff!UT00WM HONITORING INSTRtRIEN[ATION 5 SURVEILLANCE REQUIREMENIS-q 'b c N 3 INSTRUhF L'- ~ . CilANNEL CHANNEL i Igyec ue,orafg CllECK Call 8 RATION 1. Source Ran f40ciear F M " H R 2. Reactor Trip Br er Indication H N.A. 3. Reactor Coolant Tempe ture - Hot Leg.. . H k 4. Pressurizer, Pressure H R S. Pressurizer Level H R. i A 6. Steam Generator Pressure y M R g 7. Steam Gener,ator 8,evel M -R 8. Full Length, Control Rod Position Limit Switches R20 M R* 9. RHR Flos Rate R 10. RHR. Temperature .M R 11. Auxiliary Feedwater Flow Rati k M R \\- P .t gg a q 12. Pressurizer Relicf Tank Pressur?. g H y a r B,".
- 13. Containment Pressure
?. "w 4 [o H R 1 r-u e o.g g c?. t
- For cycle 1, this surveillance is to be completed before the next cooldowin or by A6 gust 3,1983 whicheyer is earlier.
g, 's'- p 'i
- 4
.n 3 N 5 - k)[ .P' % i
NE P 4.lo. Shtd 4 3 =* ECN NO.L(.I efe R3 SAFETY LIMITS
- N EVAL 6.4 No.Y 40 OF Ec.w L 4 tS(o
['t BASES i ( Range Channels will initiate a reactor trip at e e rrer.t !:v:1 re ertieral t: approximately 25 percent of RATED THERMAL POWE ss nua y o ed whe P-10 becomes active. No credit was taken for operation of the trips associated 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 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),- l and pressure is within the range between the High and Low Pressure reactor y h trips. This setpoint includes corrections for axial power distribution, changes in density and heat capacity of water with temperature.and dynamic R3 l compensation for piping delays from the core to the loop temperature detec-tors. 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 than design, as indicat d by the difference between top and bottom e power range nuclear detectors, the reactor trip is automatically reduced j(g'l. l according to the notations in Table 2.2-1. 't Operation with a reactor coolant loop out of service below the 4 loop P-S 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 l-K1, K2 and K3 inputs to the Overtemperature Delta T channels and rais.ing 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 Dcita T 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 td the High 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 cred,it was taken for operation of this trip in the accident SEQUOYAH - UNIT 1 B 2-4 ~ c
~ u NEP M %d M Lr ' EON NO. L u a6 a.g
- gg LIMITING SkFETY SYSTEM SETTINGS 9I OF Sd Evnbd*p.4
'N" v BASES u, _ Intermediate and Source Range, Nuclear Flux'(Continued) ~' approximately 25 percent of RATED THERMAL POWERange 'i unless~manualTyNKea wh L P-10 becomes active. ated with either the Intermediate or Source Range Channels analyses; however, their functional capability at the specified trip setting o is required by this specification to enhance the overall reliability of the Reactor' Protection System. 'Overtemperature AT for all combinations of. pressure, powerThe Overtemperature delta ~ coolant temperature and axial power distribution, provided that the transien,t is slow with respec,t to piping transit delays from the core to the temperature detectors (about 4 seconds) and trip. pressure is within the range between the High and Low Pressure reactor i This setpoint includes corrections for axial power distribution, s. changes in density and heat capacity of water with temperature and dynamic! t With normal axial power distribution, this reactor trip lim N i the core safety limit as shown in Figure 2.1-1. If axial peaks are greater p ~i L than design, as indicated by the difference between top and bottom powe 'i notations in Table 2.2 1. nuclear detectors, the reactor $ rip is automatically re I L setpoint does not require reactor protection system set p operation exclusive-of the Overtemperature delta T setpoi operation above the 4 loop P-8 setpoint,is permissible after resetting the K1 Three loop 1 K2, and K3 inputs to the Overtemperature delta T charnels and raising the P-setpoint to its 3 loop value. and trip functions as a High Neutron Flux trip at the reduced powe j J Overpower AT L The Overpower delta T reactor trip provides assurance of fuel integrity, L e.g., no melting, under all possible overpower conditions, limits the ' required range for Overtemperature delta T protection, and provides a backup to the ~y; High. Neutron Flux trip. The setpoint includes corrections for changes in density and heat capacity of water with temperature, and dynamic compensatio for piping delays from the cora to the loop temperature detectors. l was taken for operation of this trip in the accident analyses; however, its No credit l functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection l~ System. f _. { l y .') SEQUOYAH - UNIT 2 B 2-4
- a ~
- -?
..s. . s.. ,.t.j wi' v I;.Gi i
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[?r 30 93.-e. ....~....o 100-c. ~ L
F-Y3 wp M ,Ml I a. c. ~ \\ .3 ~ Sad (,6ve /oo //ow A-Eco ( //S( REVISION LO! wei sk.d (,3 ) g;} Revision w.. DESCRIPTION OF REVISION O#" "I } t or_- O fn Ur All f sS o 4 -__l ) I s { ge v a dCU L$'ES i O.SQb f /{Z/73 SQp 8 Col / 7 Ch \\ fa mE*ete*rel4 NEf S, f itQusere 4 /s. 19 0 *f il uso,o,,adealaA N..S ~ 4a nly *s sLuJs 2:
- r, lt 1g
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- a*%A lY r sl NV/ $4.
$am& p <v.,/ do k.s he.k <c a e d 1 l. t 7y h pJ g g e pr k.( N hk b ffY 8W4f NN A b & JN tows ks, socc ue re9e oh hs %hlel sh*9e ed p 'i ve r'i w s duib6. 4 edd<d sewial reguiremdr +c -i .4 mk snt Nu a.m Tw m/,s a claJric*1 load changes Wh'icA "c5cIttd h ,g,jg u t g,,,,, A, to nAR ch@.c T. neets y r M r. q b q, 7,, end to dic.l ud e. %t FSAR t h age,3 e e g, n-j, i Y i g. .T l 1 } s* m TVA tel34 (CN OCS.4 70) k . ' 2,,4 % ' " - ^ ' '
s f_ tm % ' + m.. r( s:J4 EvolchA. EutAlr6 ecimO. u,& p.y y y M. 0L m L As discussed in Lebparagraph 7.2.1.1.2,. factors included in establishing the Overtemoerature aT ana Overpower ai trip setpoints incluces the L reactor coolant temperature in each loop and the axial distribution of core power through the use of the two section ex-core neutron detectors. 4.4 1 3_ Instrumentatien to Limit Maximum power Outout The w&V AwW~ w ~ s 4pt of the three ranges (source, Intermediate, and power) +/-. de4M or;. ;!!' 05; th+4e*44+ of l+e. nuclear instruments, are used to limit the maximum power output of the reactor within their respective ranges. I There are sixEadial locationt. mining a 1,,i 31 ' "4 C neg n sa flu Idetecorsinstalleoarouncthereactor.inthe rimary shiel two (Issh. chah asnd p::; ordo:.d
- untm for the source /6Eg'e'Tn talled on oppo ite " flat" portions of the core containing the primary startup sources.at-*tt p ig i.
737+y g 9 ge agpg q g.m m y Aer."..**e :h: ::r; for : '-t:n d;;; 7; ;;, lo;;;;d ' th; ;w.e-a .,f g 3 _g. yg p = _g e, c e :::' tier.ed 24-e :6 :t!:r :Or- -d!u :: er: 'd' of S: : Ore bWJT51Tr~ ciTalTeetIon M::::TYat1on TffiEtPa's shroTies for c o tnelower range installed vertically at the four corners of the core and located equidistant from thu reactor vessel at all points and, to minimize neutron flux pattern distortions, within one foot of the reactor vessel. Each power range detector provides two signals corresponding to tlie neutron flux in the upper nJA lower sections of a core quadrant. The three ranges o _ d i" are used as inputs to monitor C neutron flux from a completely s wn condition to 120 percent of full power with the capability of recording overpower excursions up to 200 -g percent of ful power. The difference in neutron flux between the upper and lower sections of the ecwer range detectors are used to limit the Overtemperature aT and Overpower ai trip setpoints and to provide the operator with an inoication of the core power axial offset. In addition, the output of the power range channels are used for: + l 1. the rod speed control function, 2. to alert the operator to an excessive power unbalance between the [ quadrants, 3. protecting the core against the cc.nsequences of rod ejection' l accidents, and 4 pre >tecting the core against the consequences of adverse power distributions resulting from dropped rods. l Details of the neutron detectors and nuclear instrumentation design and the control and trip logic are given in Chapter 7. The limits on neutron fluk operation and trip setpoints are given in the SQN Technical Speelfi-cLtions. 4.4-34 0048F/COC4 6 O e
m. r_ th holvkh'a EcN gg %*8 hh ECN NO. Lcu n gg SQN-6 34 OF a. Power range high neutron flux trip. The power range high neutron flux trip circuit trips the reactor when two of the four power range channels exceed the trip setpoint. There are two independent Distables each with their own trip setting (a high and a, low setting) per channel-(four channels 6 i total). The high trip setting provides protection during normal power operation and is always active. The low trip setting, which provides protection during startup, can be manually bypassed when two out of the four power range channels read above apr'oximately 10 percent power (P-10). Three out of the four channels below 10 percent automatically reinstates the trip function. Refer to Table 7.2.1-2 for a listing of all protection system interlocks. t:. Intermediate range high neutron fluu trip The intermediate range high neutron flux trip circuit trips the reactor when one out of the twe intermediate range channels exceed the trip setpoint. This trip. which provides protection during reactor startup, can be manually blocked if two out of four power range channels are above approximately 10 percent power (P-10). Three out of the four power range channels below this value automatically reinstates the intermediate range high neutron flux trip. The Intermediate range thannels (including detectors) are separate from the power range channels. The ) intermediate range channels can be individually bypassed at the nuclear instrumentation racks to permit channel testing at any time under prescribed administrative procedures and only under the direction of authorized supervision. This bypass action is . g.M s annunciated c*e the contrcl board. U' E U 2
- c.
Source range high neutron flux trip The source range high neutron flux trip circuit trips the reactor when one of the two source range channels exceeds th trip setpoint'. This trip, which provides protection during reactor startup and plant shutdown, can b6 sanually bypassed when one of the te utarma Mate range channels reads above the P-6 setpoint alue (source rang i power level) and is automatically
- j P. g g. g e
ae wi h ermediate range channels decrease below M A interweluh the P-6 value. This trip is also automatically bypassed by two eg, a scale, - out of four logic from the power range permissive (P-10). This trip function can sIso be reinstated below P-10 by an administrative action requiring manual actut.tlon of two control board nounted switches. Each switch will reinstate the trip function in one of the two protection logic trains. The source range trip is set between the P-6 setpoint and the maximum source i range level. The channels can be individually blocked at the nuclear instrumentation racks to permit channel testing at any I l-7.2-3 0068F/COC4 l l a - -, _. -.. _. .v.. ...,y
n.,_,
SM Eubta, Eugm ? M g,,g j shdy 47 ECNNO.Lt eat at n,. SQN-6 35 OF \\ The auto stop oil pressure signal also dumps the stop emergency trip fluid, closing all of the turbine steam stop valves. When all stop i valves are closed, a reactor trip signal will be initiated if the reactor is above P-9 setpoint. This trip signal is generated by 6 redundant (two each) limit switches on the stop valves. 7. Safety injection Signal Actuation Trip A reactor trip occurs when the Safety Injection System is actuated. [ The means of actuating the Safety Injection System are described in Section 7.3. This trip protects the core against a loss of primary 'b or secondary coolant. Figure 7.2.1-1. Sheet 8. shows the logic for this trip. A detailed functional description of the process equipment associated with this trip function is provided in Reference 1. ,t 8. Manual Trip The manual trip consists of two switches with two outputs on each-switch. One output is used to actuate the train A trip breaker. the other output actuates the train B trip breaker. Operating a manual trip switch removes the voltage from the undervoltage trip coil and energizes the shunt reactor trip breaker trip coll. 6 There are no interlocks which can block this trip. Figure 7.2.1-1, Sheet 3. shows the manual trip logic. 7.2.1.1.3 Reartor TrtD System Interlocks 1. Power Escalation Permissives The overpower protection provided by the out of core nuclear instrumentation consists of three discrete, but overlapping, levels. Continuation of startup operation or power increase requires a permissive signal from the higher range instrumentation channels before the lower range level trips can be manually blocked by the ~ operator. A one out of two intermediate range permissive signal -) is 6 u ri r to source range level trip blocking and ':t::ta M;h source up = 9t:;t :S =_. rte rJange level trips are automatic y Je= MvSala .;h M t restored when both intermediate range m-fg 9 - / N h M i.1 6 m - e perm ssive (P-6) level. There is a manual if reset switch fo_r adminhtratively reactivating the source range level trip ana@5;W M;n ;P 9 P-10 lever,77 req 9fV. *M* hen between the permissive P-6 and. v
== Source range level trip block and -'eds lh AO out re'" : nte" are always maintained when above the permissive eve. i 7.2-11 006BF/COC4 6 ,.h. _ .m y ,,,4.- -,m r..
f so4g EvdvaM VCN t Of( n 0 Q,,,, 12 : -- ^ %. t M% SQN ' ECN NO, w a 6 Ex M op_ o 7.2.1.1.5 Pressurizer Water Level Reference Leo Arranaement The design of the pressurizer water level instrumentation includes a slight modification of the usual tank level arrangement using differential pressure between an upper and a lower tap. The modification shown in Figure 7.2.1-3, consists of the use of a sealed reference leg instead of the conventional open column of water. Refer to 7.2.2.3.4 for an analysis of this arrangement, 7.2.1.1.6 Analoo System L 7he process analog system is described in Reference 1. 7.2.1.1.7 Solid State Logic Protection System The solid state logic protection system takes binary inputs (voltage /no voltage) from the process and nuclear instrument channels corresponding to conditions (normal / abnormal) of plant parameters. The system combines these signals in the required logic combination and generates a trip signal (no voltage) to the undervoltage coils of the reactor trip circuit breakers when the necessary combination of signals occur. The system also provides annunciator, status light and computer input signals which indicate the condition of bistable input signals, partial trip and full trip functions and the status of the various blocking, permissive and l actuation functions. In addition the system includes means for semi-automatic testing of the logic circuits. A detalled description of l this system is given in Reference 3. 7.2.1.1.8 Isolation Ampilfiers In certain appilcations, Westinghouse considers it advantageous to employ control signals derived from individual protection channels through isolation amplifiers contained in the protection c.hannel, as permitted by IEEE-279. In all of these cases, analog signals derived from protection channels for non-protective functions are obtained through isolation amplifiers located in the analog protection racks. By definition, non-protective functions include those signals used for control, remote process indication, and computer monitoring. ' D Isolation amplifier qualification tests are described in References 4 441 3 591 n 7.2.1.1.9 Energy Supply and Environmental Variations ' The energy supply for the Reactor Trip System, including the volt' age and frequency variations, is described in Section 7.6. The environmental variations, throughout which the system will perform, are given in Section 3.11. L 7.2-13 0068F/COC4
hM Y-volnh'cw ECN L41% u-t y (& sl M 1. b. u ) " c5 partial trip alarm and channel status light actuation in the control l C room. Each channel contains those switches, test points, etc. necessary to test the channel. See Reference I for additional i information. .j The power range channels of the Nuclear Instrumentation System are j tested by superimposing a test signal on the actual detector signal being received by the channel at the time of testing. The output of the bistables is not placed in a tripped condition prior to testing. Also, since the power range channel logic is two out of four, bypass of this reactor trip fu*ction is not required. To test a power range channel, a " TEST-OPERATE" switch is provided to require deliberate operator action and operation ~of which will i initiate the " CHANNEL TEST" annunciator in the control room. l Bistable operation is tested by increasing the test signal level up to its trip setpoint and verifying bistable relay operation by 1 control board annunciator and trip status lights. j It should be noted that a valid trip signal would cause the channel under test to trip at a lower actual reactor power level. A reactor trip would occur when a second bistable trips. No provision has been made in the channel test circuit for reducing the channel signal 1evel below that signal being received from the Nuclear , Instrumentation System detector. ( ) A Nuclear Instrumentation System channel which can cause a reactor (, trip through one of two protection logic (source or intermediate range) is provided with a bypass function which prevents the initiation of a reactor trip from that particular channel during the short period that it is undergoing test. These bypasses initiate an alarm in the control room. For a detalle escription of the Nuclear Instrumentation System see Reference g l The logic of the Reactor Trip System are designed to be capable of complete testing at power, except for those trips listed L in Subsection 7.2.3. Annunciation is provided in the control room to indicate when a train is in test, when a reactor trip is bypassed and when a reactor trip breaker is bypassed. Details of the logic system testing are given in Reference 3. The reactor coolant pump breakers cannot be tripped at power without causing a plant upset by loss of power to a coolant pump. H,owever, the reactor coolant pump breaker open trip logic and continuity through the shunt trip coil can be tested at power. Manual -trip cannot be tested at power without causing a reactor trip since operation of either manual trip switch actuates both Train A and Train B. Note, however, that manual trip could also be initiated from outside the control room by manually tripping one of the reactor 7.2-23 0068F/COC4 r .m
'*Nd Evolustdn EcN Lbin j;
- g 7
ECN NO. u.,at n, 5%.ct 50 E Wk is '). *t. 9 ) SQN-6 = a
- 11. The Institute of Electrical and Electronic Engineers Inc., "!EEE j
Trial-Use Standard; General Guide for Qualifying Class ! Electric Equipment for Nuclear Power Generating Stations " IEEE Std. 323-1971. i
- 12. The Institute of Electrical and Electronte. Inc., "IEEE Trial-Use Guide for Type Tests of Continuous-Duty Class ! Motor Installed Inside the Containment of Nuclear Power Generating Stations," IEEE Std. 334-1971.
i
- 13. The Institute of Electrical and Electronic Engineers, Inc., "!EEE Trial-Use Guide for Seismic Qualification of Class ! Electric Class !
Electric Equipment for Nuclear Power Generating Stations," IEEE Std. 344-1971.
- 14. " General Design Criteria for Nuclear Power Plants," Appendir A to "
3 ittie 10 CFR 50, July 7,1971. ? of
- 15. E. P. Rahe, " Evaluation of Surveillance Frequencies and Out,d 4
i Service Times for Reactor Protection System," NCAP 10271 an Supplement 1 (Westinghouse NES Proprietary).
- 16. W. H. Moomau, " Westinghouse Setpoint Methodology for Protection O
Systems. Sequoyah Units 1 and 2." HCAP 11239 Rev. 3, October 1987 (Westinghouse Proprietary Class 2). l'1. T V A hv\\ronmedAl Qonht.di$w B;8 der @N EQ-NM ool, %. fews, N ev% a nox %%%. l 9 'S e G 4 7.2 37 006BF/COC4 L -ns ..----r,..--
SQ h EVAleef s ECA L61t4! k,s !.y'g3,l.a.3) I
- g W
shut 51 i SQN-6 Electrical Penetration Assemblies rwess+ s Thegelectrical penetration assemblies have been tested to TVA specification requirements which conform to IEEE-317, 1971, "IEEE \\, ~ Standard for Eltetrical Penetration Assemblies in Containment Structures t for Nuclear Power Generating gtations." The. W x g,hctch y g uete 4 ;y pgg't,Mlb life c'u atTo o C' reports of all tests required and listed in the specifications and quality assurance appendtr, and applicable TVA inspector's reports. Each electrical penetration assembly furnished has been shop Inspected by a commissioned representative of the National Board of Boller and Pressure Vessel Inspectors. Each assembly has been Code stamped, in accordance with the 1971 Edition ASME Boller and Pressure Vessel Code. Section !!!. The dose rate at which TVA has conducted 100 hour tests on materials and equipment is 10' Rad /hr dose rate that may occur during the first hour l6 of a LOCA. It is the TVA position that a factor of 5 in dose rate is not significant in this region. There is no mechanism that TVA is aware of that would tend to produce significant increases in degradation in the r region between 10' and 10' Rad /hr. However, radiation-induced l6' oxidation of materials can become an important damage mechanism at lower exposure rates and consequent longer exposure times. Therefore, IEEE 278, " Guide For Classifying Electrical Insulating Materials Exposeo to i L Neutron and Gamma Radiation," recommend using exposure rates above 10' j Rad /hr. It is the TVA position that 10' Rad /hr for 100 hours 6-represents a reasonable and conservative combination of dose rate and exposure time for radiation testing. l l' Cable terminations to low voltage power, control, and Indication penetration assemblier are generally made in all metal spilce boxes. However, in a number of instances on the outboard side of containment electrical penetrations, field cables were spliced to the penetration t l pigtalls in cable trays. In these cases, a special enclosure was used to ( act as a qua11 fled fire stop (refer to Figure 8.3.1-37, -38, and -39). l6 l These particular splices are located within the last 5-foot section of the cable tray. The trays in the annulus area of containment containing f l these splices are fitted with solid top and bottom covers in the immediate area of these splices. A qualtfled fire barrier made of silicone foam and'cerafors/kaowool fiberboard was installed on the side of the splice opposite to the penetration as shown on Figure 8.3.1-37, g -38, and -39. On the other side of the spilce in the tray (end of tray runs toward the electrical penetration), knowool materials were inserted In the volds between conductors, and, all'the exposed conductors to the electrical penetration were covered with flamemastic material. This configuration constitutes a quallfled fire barrier which in the unlikely event of a fire in the splice area, will contain and isolate the fire from adjacent trays of electrical equipment. 8.3-40 0078F/COC4 L
c4*u %. 5 d 15VAlp4i,$p N LblIb h W sedda 16.p.4.2) 5 hu+ $ % 7 NO.,(g,LA6 23 u M _OF_ cperation are considered in this analysts. Table 15.2-1 contains tne time sequence of events for this accident. Dilution Durino Refuelino An uncontrolled boron dilution accident cannot occu RCS from the potential source of unborated water. '\\ .Various valve combinations' that are required to be locked closed during 1 . refueling operations are specified in technical specification 3.9.1. These valves will block the flow paths which could allow unborated makeuj to reach'the RCS. Any makeup which is reautred during refueling will be borated water supplied either from the refueling water storage tank by the low head safety injection pumps or the centrifugal charging pumps, or from the boric acid tanks via a boric acid transfer pump and a i centrifugal charging pump. 6 Dilution Durino Startun j Prior to startup the RCS is filled with borated (approximately 2000 ppm) water from the refueling water storage tank, i Nuclear instrumentation is monitored closely in anticipation of an i '6 unplanned reactivity rate of change. accomplished by operation of the reactor coolant pumps. Mixing of the reacto range flux level and all reactor trip alarms are effective.High source In the analysis, a maximum dilution flow of 300 gpm limited by the l capacity of the two primary water makeup pumps is considered. of the reactor coolant is approximately 9967 f t*, which is the activeThe volume volume of the Reactor Coolant System excluding the pressurizer. Dilution Followino Reactor Shutdown lowing reactor shutdown, when in hot standby, hot shutdown, and g subsequent cold shutdown condition, and once belo setpoint, h h ve!'h :e Met e:t!c-CO2, rui;3 qg3 ::r-e e "' t' S; :ye M;;'::r "l =t 5 ' t; ' =d 2 M ' 'g'** gje:tmm.t art., put gm-s, the high flux at shutdown alarm setting will bef adjustedyo no higher than 4!2 ';:19 S'^"e3 the^Ay..wJcount rate M -' ;t: fous 3 e ihr
- tpe'at rett b ::t er :raM: m.f t:r ;hN*butder.r%lvess.
as gn aSally dbwward h ':d :::ry 20 & ct:: f:r th: f'r:t-o, ! '^" 'c'^H a;mr +s..ph-t t-!;, ev: y 2 ' cure fer t% aert 6 %; 5, i {'...m.,We'e~ve[ill[n.iee. test Iw G 1Yi el s' ore '+5. +' tad ._r... ,,-.ii .u. s....s t.. .uis..a 'lTTtion at ower l l With the unit at power and the RCS at* pressure, the dilution rate is the charging pump. limited by the capacity of the primary water makeup p f A conservatively high value for the expected boron rate of 300 gpm was used. concentration (1575 ppm) at power and a conserva g Ye of t e Aker-clot.s ge-f deptild *NMr
- N
'3 * ** ""
- l
.x s ca. n a c.,w.% ;;,%.y % * * ~~ ge;7f Mf* nrs e saa w \\,wa u %.a tr,n u,.n, o c s u o w a,,a w, z,,,, a v E*NN d *d bt. Mai.I C,eded ),ene) ANd f60cte rawy demer. g ~ m A
e n.m n 5k4 gvyy;;th ECN NO. L o a s e 5 SQN-6 & t.H LM % soo op 15.2.4.3 Conc'usions 56.{. 53 For dilution during refueling: Dilution during refueling cannot occur due to administrative controls .a M (see Section 15.2.4.2). 4., g 4q g sovru ca,og Theoperatorhaspromptanddefinitehndicationofanyborondilution ga from the audible count rate instrgntation3 Migh count rate is 6d rdvei U e414-44 reactor containment and the$ control room. In addition, a high source range flux level is alarmed in the control room. The count rate increase is proportional to the subtritical multipiltation factor, w - -- M .'s % render For dilution during s ar up: codhlete For dilution during startup, there is adequate time (-52 minutes) f r m g transient initiation for the operator to recognize the high count rate signal and terminate manually the source of dilution flow. The Operator is alerted to the uncontrolled reactivity insertion during startup via the increasing count rate on the Source Range Nuclear Instrumentation. Recorders on the control board continuously provide a time history of the nuclear flux level. This increase in flux level is very slow, based on the reactivity insertion rate for the startup case, it takes approximately 42 minutes for the flux level to increase by a factor of 2. This is adequate time for the operator to recognize from / the recorders the need for action. Thus there would still be 48 minutes y for the operator to ascertain and isolate the source of the reactivity insertion. Also on the source range channel is the high f "x at shutdown alarm. The setpoint for this alarm is normally placed at Y) times the source level. Even assuming that the operator doesn't rec gnize the increasing count rate, the alarm will occur at approximately 70 minutes into the transient. Thus there would still be 21 minutes for the operator to stop the dilution. For dilution during full power operation: 1. With the reactor in automatic control, the power and temperature increase from boron dilution results in insertion of the rod cluster control assemblies and a decrease in the shutdown margin. The operator is alerted to an uncontrolled reactivity insertion by the rod insertion limit alarms. Two insertion limit alarms are available: The first occurs when the rods are 10 steps above the insertion limit (Lo Insertion Limit Alarm) and the second occurs at the insertion limit (Lo-Lo Insertion Limit Alarm). The analysis assumed that the operator is algrted to the need for action by the Lo-lo Alarm although action would be taken when the first alarm occurs. Thus the analysis already assumes a 10 step allowance for / 15.2-15 COC4/0115F
f. ^ ~ a: - - ~. Tdd bld.m _ b li,e n y SON Etp4 t4 g4 ECN NO.m a 6 al El _ __OF l Shut 54 TABLE 15.2.4-1 SE0VENCE OF EVENTS [cu111brium Xe Case Time (sec) Reactor Trip 0 Reactor Power = 7.57, of nom tee reactor neriod Intermediate NIS reads J 4mps. % Pow e st 10 Source Range NIS Available 930 Source Range NIS no longer decreasing (without l f dilution event, flux would stabilize at this point - an 18 day half Ilfe decay of flux i would be normal'). 1,250 d 3 % t.s 0;e eter set High Flux 44 Shutdown Alarm 0.5 dende .abas-stabli tzed flux level. For this example, the value is I 700 cps. 1,250 High pressurtzer Level Trip and Alarm 1,800 Source Range High Flux at Shutdown Alarm 7,400 K.,e = 1.0 12,960 ' Source Range Count Rate would change from 200 cps to 197 cps i 9-i 0726F/COC4
a ^ ^~ : =, iu 2 $;dg Evs3od, w tpo n.u.n l ELA t 4\\1% E;N No._q_ SCN-6 SN N l o t,, a l a TABLE 8.3.1-11 (Sheet 1) 120V AC VITAL INSTRUMENT POWER BOARO 1-f i (Batterv Boare !) j BKR SAFETY CCNNECTED !!O., LOAD REL4TfD LOAO-VA 1 SSPS (A) Ch ! Input Relays (Pnl 1-R-48) Yes 1,080 2 SSPS (B) Ch ! Input Relays (Pnl l-R-49) Yes 60 3 NIS Instr Power Ch ! (JB 3398) Yes ce9-6 't 4 NIS Control Power Ch ! (Pnl 1-M-13) Yes -44A to l 5 Process Protection Set I (Pnl 1-R-1) Yes s4 l 6 UNI Accumulator Ch ! Isolation Valve (Pnl 1-M-23A) Yes 62 7 Cnmt. Bldg. Rad. Monitor (1-RE-90-106 & 0-RE-90-133) Yes 949 8 Instrumentation Bus A (Pnl 0-M-278) Yes 50 9 SSPS Aux Relays (Pnl 1-R-73) Yes 110 10 Reactor Building Isolation Valve (JB2670) Yes 109 11 Aux Comoressor A Aux Bldg 1501 Valve (FCV-32-82) Yes 52 12 Radiation Rate Meters (PNL 0-M-12) Yes 1.406 13 Radiation M6nitors (0-RE-90-125) Yes 360 la Instruments (125V Vital Battery Board !) Yes 20 15 Incore TC Monitoring (PNL l-R-59) Yes 308 16 Chlorine Detector (Pnl 0-L-450) Yes 24 17 PASF Solenold Valves (PNL l-L-572/C) Yes 619. i 18 Aux Relays PCO-65-81 & PCO-65-86 Yes 84 19 Tollet & Locker & Spread Room Isol Dampers Yes 168 6 20 80P Process Instr Control Rack Yes 48B 21 Aux Bldg Instr Bus A (PNP l-L-26) Yes 150 22 Aux Bldg Stm Isol Viv FCV-12-82 (Pn1 1-M-9) Yes 62 23 Containment Purge Air Exhaust Rad Monitor Yes 36') 24 Reac. Vessel Hd. Vent Throttle Valve FCV-68-397 Yes 60 25 NSSS Aux Relay Rack A Bus (Pn1 1-R-54) Yes 462 26 Sep & Aux Relays (Pnl 1-R-73) Yes 532 3
- 27. A Relay Bus (Pnl 1-L-IIA)
Yes 266 28 A Instrument Bus (Pnl 1-L-11A) Yes 140 -29 RVLIS (Pnl 1-R-148) Yes 748 30 Aux Oryer Train A Yes 981 31 Sep & Aux Relays (Pnl 1-R-74) Yes 238 32 Borid Acid Tank A Htr A-A Control (1-L-303) Yes 31 33 - Aux Bldg Cas Treat Fan A-A Mod Dmpr (0-L-429) Yes 142 34 Boric Acid Tank C Hrt A-A Control (0-L-306) Yes 31 35 Radiation Monitor (0-RE-90-205) Yes 360 36 RCPI UV & UF Relays Yes 30 37 Process Control Group ! (Pnl 1 R-14) No 715 38 Instrument Bus 1 (Pnl 0-M-278) No 56 39 Plugmold Instrument Bus 1 (Pnl 1-M-5) No 225 40 Plugmold Instrument Bus I (Pnl 1-M-6). No 280 41 Instrument Bus I (Pnl 1-M 4) No 60 0431F/COC4 e o m w-m.r .- +- - ,+,w a - 4
~ Sa (8y bdo:h,d, E Cid N O.L. r. u s g5 gg SQN-6 gt $4 ,LQ},_,op_ TABLE 8.3.1-11 (Sheet 2) (Continued) l 120V AC VITAL INSTRUMENT POWER BOARD 1-! (Battery Board I) j BKR SAFETY CONNECTED I .NQ_ M RELATED LOAD-VA 42 Fire Pump 2A-A Sep Relays No 84 -- u$s KYs 0 i = 43 ' UNI IitTt i n"^^^--% ( 46 PR-30-310 (Installation on Hold) Yes 30 '47 5th Vital Battery Instrumentation Yes 26 l 48 """* Spa re "" * " TOTAL 14,100 i 5.9111 1. Each inverter ls capable of supplying 15 kva continuously, g 2. No automatic load stripping or load sequencing is employed. 3. Power Boards 1-I,1-!!, I-I!!, and 1-IV also supply common plant loads and are more heavily loaded than respective unit 2 boards. ( :. e 0431F/COC4 .m.
[ Tdth tram.k M '~N L " %,,g tot wu
- D.14m lAr. a3_ l g gg SCN-6 IQR_
j_ TABLE 8.3.1-12 (Sheet 1) 120V AC VITAt INSTRUwfNT POWER BOARD 1-!! J Ba tterv 6carc !!) BKR SAFETY CONNECTED !!O,. 1.0A0 RELATED LCAO-VA 1 $$PS (A) Ch I! Relays (Pnl 1-R-46) Yes 600 2 SSPS (B) Ch 11 Input Relays (Pnl 1-R-51) Yes 1 O!Le - 3 NIS Instr Power Ch II (JB 3399) -440 4y1]j
- ' 340 Yes 4 NIS Control Power Ch !! (Pnl 1-M-13)
Yes 5 Process Protection Set !! (Pnl 1-R-5) Yes 1,1): 6 UHI Accumulator Ch II Isolation Valve (Pnl 1-M-238) Yes 62 7 ERCH & Containment Rad. Monitor (Pnl 1-RE-90-112) Yes 949 8 Instrumentation Bus 8 (Pnl 0-M-278) Yes 30 9 SSPS Aux Relays (Pnl 1-R-78) Yes 103 10 Reactor Building Isolation Valve (FCV-32-102A) Yes 62 11 Aux Compressor 8 Aux Bldg 15o1 Valve (FCV-32-85) Yes 48 12 Radiation Rate Meters (PNL 0-M-12) Yes 1,054 13 Radiatton Nonitors (0-RE-90-126) Yes 360. 14 Instruments (125V Vital Battery Board !!) Yes 26 15 Incore TC Monitoring (PNL l-R-60) Yes 308 16 Chlorine Detector (Pnl 0-L-451) Yes 24 17 PASF Solenoid Valves (PNL 1-M-10) Yes 619 18-Aux Relays PCO-65-83 & PCO-65-87 Yes 84 19 Toilet & Locker & Spread Room Isol Dampers (Pn1 1-R-78) Yes 168 20 BOP Process Instr Control Rack (Pnl 1-R-131) Yes 203 6 21 Aux Bldg Instr Bus 8 (PNP l-L-26) Yes 150 22 Aux Boller Stm Isol Viv FCV-12-79 (Pnl 1-M-9) Yes 62 23 Containment Purge Air Exhaust Rad Monitor (1-RE-90-131) Yes 360 24 Reac. Vessel Nd. Vent Throttle Valve FCV-68-396 Yes 60 25 NSSS Aux Relay Rack 8 Bus (Pnl 1-R-55) Yes 308 26 Aux Rly Rack Sep & Aux Relays (Pnl 1-R-78) Yes $46 27 Aux Cent Pnl 8 Relay Bus (1-L-llB) Yes 168
- 28. Aux Cont Pnl 8 Instrument Bus (1-L-Il8)
Yes 184 29 RVLIS (Pn1 1-R-148) Yes 768 l-30 Aux Dryer Train 8 Yes 981 1 31 Aux Rly Rack Sep & Aus Relays (Pn1 1-R-77) Yes 210 32 Borld Acid Tank A Ntr 8-8 Control (Pnl 1-L-304) Yes 31 33 Aux Bldg Gas Treat Fan 8-8 Mod Dmpr (0-L-428) Yes 142 34 Boric Acid Tank C Nrt 8-8 Control (Pnl 0-L-305) Yes 31 35 Radiation Monitor (0-RE.90-206) Yes 360 36 RCP2 UV & UF Relays Yes 30 i 37 Process Control Group 2 (Pn1 1-R-17) No 743 i l 38 Instrument Bus 2 (Pnl 0-M-278) No 38 39 Plugmold Instrument Bus 2 (Pnl 1-M-3) No 600 40 ~Plugmold Instrument Bus 2 (Pnl 1-M-6), No 166 41 Accous tic Flow Monitor (Pnl 1-M-2'7A) No 117 l i 0431F/COC4 l-i
h p rw % % =r'a u SLQ qM4 y gcu no, e c,, u as ttA LO % 16 5 OF SON-6 h -4. g TABLE 8.3.1-13 (Sheet 1) 120V AC VITAL INSTRUMENT POWER BOARD 1-!!! (Battery Board !!!) BKR SAFETY CONNECTED l (QAQ RELATED LOAD-VA i 1 SSPS (A) Ch !!! Input Relays (Pn1 1-R-46) Yes 600 2 SSPS (8) Ch III Input Relays (Pn1 1-R-49) Yes 600 3 NIS Instr Power Ch !!! (JB 3400) Yes 88 4 NIS Control Power Ch !!! (Pn1 1-N-13) Yes 240 5 Process Protection Set !!! (Pnl 1-R-9) Yes 456 6 UNI Accumulator Ch !!! !sciation Valve (Pn1 1-M-23A) Yes 65 7 RCP 3 UV & UF Relays Yes 30 8 Aux Feed lurb Controller (Pn1 1 L-381) Yes 222 9 Turb Dr Aux FW PMP St Gen (Pn1 1-L-361) Yes 62 10 Tur. n.b.D.r Aux FW PMP St Gen (Pn1 1-L.11 A) Yes 80 jj
- Spare " * ""
12 ***** Spare ****"* 13 '""" Spare """* 14 Instrument Bus & Xfor Pwr (Pn1 1-M-3) No 30 15 Aux Cont Pn1 A Inst Bus (PNL i-L-11A) No 475 16 Process Cont Group 3 (Pn1 1-R-20) No 282 17 BOP Process Instr Control Rack (Pn1 1-R-126) No 3,170 18 Control Room Doors Security Lock No 100 b 19 Emergency Gas Treat Filter Train A (Pn1 0-L-25) No 20 20 BOP Process Instr Control Rack (Pn1 1-R-128) No 845 21 "'" Spar e """
- 22 Aux Bldg Inst A Bus 1 (Pn1 1 L-57)
No 20 I 23 Aux Relay Rack A Bus (Pn1 1-R-76) No 364 l 24 Aux Relay Rack C Bus (Pn1 1-R-76) No 196 25 NSSS Aux Relay Rack A Bus (Pn1 1-R-58) No 84 26 Aux Control Panel A Bus (Pn1 1-L-10) No 168 27 Aux Relay Rack A Bus (Pn1 1-R-75) No 434 28 SSPS Control Room Demod (Pn1 1-M-22) No 480 29 Aux Control Panel C Relay Bu: (Pn1 1-L-10) No 42 - - - (Wh 30 Aux Control Panel A Instr Bus (Pn1 1-L-10) No 31 Control Air Ndr A Holsture Als (JB281) No 13 - 1 32 Aux Relay Rack A Bus (Pn1 1-R-72) No 182 l 33 """
- Spa re " " * "
34 Post Accident Monitoring (Pn1 1-M-5) No 48 35 "" * " Spare " * "" 36 LOCA H2 Cntant flow Monttor (Pn1 1-M-10) No 106 37 CO2 Fire Protection Computer Room No 38 CO2 Fire Protect Olesel Gen & Lube Oil Rm No 39 Feed To Bkrs 37 & 38 No 1,500 40 Loose Parts Monitor Equipment Panel (Pn1 0 R-139) No 240 41 Reactor Vessel Level Instr System No 110 r 4 0431F/COC4
(L- "hh %% %i 4 W4 L 415te SC'i-6 bd 51 EcNuggg,g3_ l >0 LL%.c.=_ ; TABLE 8.3.1-15 (Sheet I) IIOV AC VITAL INSTRL'wfNT POWER BOARD 2-T (Batterv toare !) BKR SAFETY CONNECTED NO,, LOAD RELATED LCAD VA 1 SSPS (A) Ch ! Input Relays (Pnl 2-R a6) Yes 600 2 SSPS (B) Ch ! Input Relays (Pnl 2-R-49) Yes 600_ m 3 NIS Instr Power Ch ! Yes rec.,%e 5 4 N!S Control Power Ch ! (Pnl 2-M-13) Yes 431 ~J 5 Process Protection Set I (Pnl 2-R-1) Yes 1,07 6 UH! Accumulator Ch ! ! solation Valve (Pnl 2-N-283) Yes 62 l 7 RCP IUV L MF Relays Yes 30 8 Aux Feed Pump Turb Flow Cont Yes 222 9 Turb Dr Aux FW PMP St 3&4 Gen (Pn1 2-L-381) Yes 75 10 Turb Dr Aux FW PHP St 3&4 Gen (Pnl 2-L-llA) Yes 80 11 RVLIS (Pnl 2-R-148) Yes 745 i 12 Incore TC Honitoring (Pnl 2-R-60) Yes 308 l JJ eeessee gpggg eeeeeee j 14 Instr 6us (Pnl 2-M-4) No 60 - l 15 Plugmold Inst Bus 1 (PNL 2-M-5) No 194 16 Plugmold Inst Bus 1 (Pnl 2-M-6) No 280 17 Process Cont. Group (Pnl 2-R-14) No 749 18 BOP Process Instr Cont Rack (Pnl 2-R-126) No 3.100 6 19 Aux Cont Pnl A Instr Bus (Pnl 2-L-llA) No 455 20 BOP Process Instr Cent Rack (Pnl 2-R-128) No 600 21 ***'*** Spare ******* 22 Aux Bldg Inst A Bus (Pnl 2-L-57) No 50 23 Aux Relay Rack A Bus (Pnl 2-R-76) No 182 24 Aux Relay Rack C Bus (Pnl 2-R-76) No 140 25 NSSS Aux Relay Rack A Bus (Pnl 2-R-58) No 84 26 Aux Cont Pnl A Relay Bus (Pnl 2-L-10) No 154 27 Aux Relay Rack A Bus (Pnl 2-R-75) No 266 28 SSPS Cont Rm Demod (Pnl 2-M-22) No 480 29 Aux cont Pnl A Inst Bus (Pnl 2-L-10) No 210 30 "" * " Spa r e '""" t 31 "" *" Spa r e "" * " 32 Aux Relay Rack A Bus (Pnl 2-R-32) No 56 33 * '"" Spa r e '""" 34 Post Accident Monitoring 1 (Pnl 2-M-5) No 48 35 * * * * * " Spa r e '""" 36 LOCA H2 Cntant flow Monitor (Pnl 2-M-10) No 106 37 e.e.see Spare ******* 33 e s s e s" Spare ""*" 39 " * "" Spa re * """ - 40 """* Spa re " * "" 41 '""" Spa re '""" 0431F/COC4 1 1 e-r- m o -*p. --y-y g-e-
c u _m ,, e s%N E glvats ' "p,uf t y y EONNO.Lr.isk 23 SQN-6 EW LM% Skd 4D E> l_OF TABLE 8.3.1-15 (Sheet 2) (Continued) 2-I 120V AC VITAL INSTRUMENT POWER BOAR $14.- (Battery Board I) BKR' SAFETY CONNECTED g g RELATED LOAD-VA 42 ******* spare ******* 43 * " " *
- Spa r e * " * * *
- 44 * """ Spare """
- 45 UNI Instrument Bus I (Pn1 2-N-23A)
No 32 .46 * "" *
- Spa r e """
- 6 47 * """ Spare * * ""
- 48 """
- Spa re "* ""
TOTAL 11.723 NOTES 1. Each inverter is capable of supplying 15 kva continuously. 2. No automatic load stripping or load sequenc1_ng is employed. 3. Power Boards 1-!, 1-!!, 1-!!!, and 1-IV also supply common plant loads and are more heavily loaded than respective unit 2 boards. u l; l l l L L t l l \\ 0431F/COC4 )
C i-m -,- ~ M EVnluda,g " n *>l y t,g g ECN NO. L Q a 6 a's 1 SQN-6 ha,P fol j_ .L.1 0F-TABLE 8.3.1-16 (Sheet 1) 120V AC VITAL INSTRUMENT POWER BOARD 2-!! (Battery Boare !!) BKR SAFETY CONNECTED _NQ. igAQ RELATED LOAD-VA i SSPS (A) Ch !! Input Relays (Pnl 2-R-46) Yes 600 2 SSPS (B) Ch !! Input Relays (Pn12-R-49) Yes 600 3 NIS Instr Power Ch !! (JB 3403) Yes %3 4 NIS Control Power Ch II (Pnl 2 M-13) Yes -440-y.39 5 Process Protection Set !! (Pnl 2-R-5) Yes 6 UN! Accumulator Ch !! Isolation Valve (Pnl 2-N-23B) Yes 62 7 RCP 2VV & UF Relays Yes 30 l 8 Aux Feed Pump Turb Flow Cont (2-L-381) Yes 222 9 Turb Dr Aux FW PHP St Gen 182 LIC-3-173,174 Yes 60 1 10 Turb Dr Aux FM PMP St Gen 1&2 Instr Loop Yes 244 11 RVLIS (Pn1 2-R-148) Yes 745 12 Incore TC Monitoring (Pnl 2-R-60) Yes 308 13 " * "" Spa r e * """
- 14. Process Cont Group 2 (Pnl 2-R-17)
No 743 15 Plugnold Inst Bus 2 (PNL 2-M-3) No 509 16 Aux Cont Pnl B Inst Bus (Pnl 2-L-llB) No 535 17 Plu9 mold Instr Bus 2 (Pn1 2-M-6) No 136 6 18 B.O.P Process Instr Cont Rack (Pn1 2-R-122) No 629 19 .."
- Spa re " * ""
l 20 BOP Process Instr Cont Rack (Pnl 2-R-130) No 454 1 21 * * * "" Spa r e * """ 22 Aux Bldg Inst B Bus (Pnl 2-L-299) No 40 23 Aux Relay Rack B Bus (Pn1 2-R-76) No 168 24 NSSS Aux Relay Rack C Bus (Pn12-R-58) No 322 25 NSSS Aux Relay Rack 8 Bus (Pnl 2-R-58) No 84 l 26 Aux Relay Rack 8 Bus (Pnl 2-R-75) No 294 27 Aux Relay Rack C Bus (Pnl 2-R-75) No 168 28 Aux Cont Pnl B Relay Bus (Pnl 2-L-10) No 84 l' '29 Aux Cont Pn1 C Relay Bus (Pnl 2-L-10) No 1 30 Aux Cont Pn1 B Instr Bus (Pnl 2-L-10) No 3M l 31 Emerg VHF Radio (Pnl G) -- Equipment Removed -- l 32 Aux Relay Rack B Bus (Pn1 2-R-72) No 112 33 Aux Relay Rack C Bus (Pal 2-R-72) No 56 34 Post Accident Monitoring 2 (Pnl 2-N-4) No 64 35 "" *" Spa re " * "" 36 LOCA H2 Cntmnt Flow Monitor (Pnl 2-N-10)' No 106 37 RVLIS (Pnl 2-R-148) Yes 729 38 """
- Spa re "" * "
39 " * "" Spare * """ 40 '""" Spare *""" Al " * "" Spare "" * " I l i 0431F/COC4 .. ~
~ ( '9 'sp -.au a cu f-SM Evdud.'y _,{ ik si 't y SON-6 EW L4t% IECNictggg.g g i bd bl. C TABLE 8.3.1-16 (Sheet 2) 109 CF_ (Continued) 120V AC VITAL INSTRUMENT POWER BOARD 2-!! ~ (Battery Board II) BKR SAFETY CONNECTED g g RELATED LOAD-VA 42' """* Spare *""" 43 ""* " Spare * """ 44 Inplant VHF Radfo Repeater F1 No 1,068 45 UHI Instrument Bus 1 (Pnl 2-M-238) No 32 46 On $lte Paging Radio 4o 1,068 6 7 " ? " " ( n a r3. * " " "- m 48.
w --- " HT. s asis ( m.u $ooma roomeN NYu N 2461 T0fEL ~
~12.240 NOTES i l 1. Each inverter is capable of supplying 15 kva continuously. 6 2. No automatic load stripping or load sequencing is employed. 3. Power Boards 1-1,1-!!,1-III, and 1-IV also supply common plant loads and are more heavily loaded than respective unit 2 boards. 4 4 4 I l-l l L 0431F/COC4 1 l' s.
'~ f-h u. L f ) i ENCLOSURE 3 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (WA-SQN-TS-88-42 ) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS f i l r t t It l' l l-f, i ~
4 'C. i. l ENCLOSURE 3 Significant Hazards Evaluation IVA has evaluated the proposed TS change and has determined that it does not represent a significant hazards consideration based on criteria established in 10 CFR 50.92(c). Operation of SQN in accordance with the proposed amendment will nots (1) Involve a significant increase in the probability or consequence of an accident previously evaluated. Two administrative changes are proposed to support the installation of the new Gamma Metrics SR and IR detector assemblies. The first involves a revision to the notation contained in TS Table 3.3-1 regarding the high-voltage deenergisation that will no longer occur for 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 new SR/IR detectors are Class-1E equipment that is 1 seismically and environmentally qualified and compatible with the 2 present design requirements. Because the new hardware is compatible with the present design requirements and the proposed TS changes are administrative in nature, the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated. t (2) Create the possibility of a new or different kind of accident from i any previously analyzed. Two administrative changes are proposed to l support the installation of the Causna Metrics SR and IR detector assemblies. The first involves a revision to the notation contained I in TS Table 3.3-1 that is no longer applicable to the design of the j new SR detectors. The second involves a change in engineering units t [ for the P-6 setpoint that results from the difference in output ( signals from the IR detectors. 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 TS changes are administrative in nature, the proposed amendment will not create the possibility of a new or different kind I of accident from any previously analysed. (3) Involve a significant reduction in a margin of safety. Two administrative changes are proposed to support the installation of the Gamma Metrics SR and IR detector assemblies. The first involves a revision to the notation contained in TS Table 3.3-1 that is no longer applicable to the design of the new SR detectors. The second L involves a change in engineering units for the P-6 setpoint that l results from the difference in output signals from the IR detectors. 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 TS changes are administrative in nature, the proposed amendment will not involve a significant reduction in the margin of safety. .}}