Text

Proposed Tech Specs Re Reactor Trip Sys Instrumentation Trip Setpoints & Allowable Value Units for intermediate-range & source-range Neutron Flux Detectors
	ML20006A285
Person / Time
Site:	Sequoyah
Issue date:	01/12/1990
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
Shared Package
ML20006A281	List: ML20006A280; ML20006A285;
References
	NUDOCS 9001260079
	Download: ML20006A285 (101)
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--.s ENCLOSURE 1 l'

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 W

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~h TABLE 2.2-1 (Continued) wEg REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS' 5

x FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES t-e 3

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-g

14. Steam /Feedwater Flow

< 4'0% of full steam flow at

< 42.5% of full steam flow at

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Mismatch and Low Steam RATED-THERMAL POWER coincident NATED 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

15. Undervoltage-Reactor 1 5022 volts-each bus 1 4739 volts-each bus h89 Coolant Pumps s

y

16. Underfrequency-Reactor

> 56.0 Hz - each bus 1 55.9 Hz - each bus Coolant Pumps e

17. Turbine Trip A.

Low Trip System 1 45 psig

> 43 psig Pressure B.

Turbine Stop Valve

>.1% open

> 1% open Closure

18. Safety Injection Input Notduplicable Not Applicable _

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19. Intermediate Range Neutron l 1 (~10 m

d Flux - (P-6) Enable Block..

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?, g Source Range Reactor Trip '

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20. Power Range Neutron Flux

< 10% of RATED

< 11% of RATED.

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(not P-10) Input to Low Power THERMAL POWER THERMAL POWER

.M o Reactor Trips Block P-7

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BASES-W Range Channels will initiate a reactor trip

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.approximately 25 percent of RATED THERMAL P unless manually blocked when 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.

1 l0vertemperature 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 p

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 detec-tors. With normal axial power distribution, this reactor trip limit is always R.,

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 t

power range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1.

l Operation with a reactor coolant loop out of service below the 4 loop P-8 setpoint does not require reactor protection system setpoint modification because the P-8 setpoint and. associated trip will prevent DNB during 3 loop operation exclusive of the Overtemperature Delta T setpoint. Three loop operation above the 4 loop P-8 setpoint is permissible after resetting the i

K1, K2 and K3 inputs to the Overtemperature Delta T channels and raising the l

P 8 setpoint to.its 3 loop value.

In this mode of' operation, the P-8 inter-lock and trjN functions as a High Neutron Flux trip at the reduced power level.

Overpower Delta T The Ov'erpower Delta T reactor trip provides assurance of fuel integrity, e.g., no melting, under al.1 poss.ih.1 Aourpower, conditions,14mitr the required nnage for.0vertemperature Delta T protection, and provides a backup to the l

High Neutron Flux trip.

The setpoint includes correctioos.fw danges in "d

densitp rA4 twLt. rgv0;...vf m%r dtirtemperature', and dynamic compensation

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for piping delays from the core to the loop temperature detectors.

No credit was taken for operation of this trip in the accident r

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TABLE NOTATION-1 m.

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.

g The channel (s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped condition.

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. g..... x. t;_ e.. _'.... : may be d: r :r;i::d above the P-6 (Block of Source

' Range Reactor Trip) setpoint.

ACTION STATEMENTS i

ACTION 1 -

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable I

channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the reactor trip breakers, i

ACTION 2 -

With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and POWER OPERATION may proceed provided the following conditions are satisfied:

a.

The ineperable channel is placed in the tripped condition

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within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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b.

The Minimum Channels OPERABLE requirement is met; however,

- one additional channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months lR51 for surveillance testing per Specification 4.3.1.1.1.

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c.

Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL and the Power Range, Neutron Flux l

high trip reduced to less than or equal to 85% of RATED

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THcRMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, the QUADRANT-POWER-TILT RATIO is monitored at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

d.

The QUADRANT POWER TIL'T 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when THERMAL POWER is greater than 75% of RATED THERMAL POWER.

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September 17, 1986 SEQUOYAH - UNIT 1 3/4 3-5 Amendment No. A if up

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TABLE 2.2-1 (Continued) c 5

REACTOR TRIP' SYSTEM INSTRUNENTATION TRIP SETPOINTS

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FUNCTIONAL UNIT TRIP SETPOINT

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ALLOWA8LE VALUES E 13. Steam Generator Water Q

Level--Low-Low 1 18% of narrow range instrument 1 17% of narrow range instrument

'm7 span each steam generator.

span each steam generator m

14. Steam /Feedwater Flow

< 40% of full steam flow at i

< 42.5% of full steam flow at:

Mismatch and Low Steam RATED THERMAL POWER coincident RATED THERMAL POWER coincident Generator Water Level i

with steam generator water level with steam generator water level

> 25% of narrow range instru-1 24% of narrow range instru-ment span--each steam generator ment span-each steam generator l.

1 i

15. Undervoltage-Reactor

> 5022 volts-each bus 1 4739 volts each bus R76 i

Coolant Pumps i

16. Underfrequency-Reactor

> 56 Hz - eacle bus

> 55.9 Hz each bus i

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Coolant Pumps

17. Turbine Trip t

'A.

Low Trip System 1 45 psig

> 43 psig I

Pressure B.

Turbine Stop Valve

> 1% open

> 1% open Closure

18. Safety Injection Input Not Applicable Not Applicable f" " ESF

19. Intermediate Range Neutron h @1 1 : IS 28 x to'# */o d RAT"ED 1MENAL.

~ ', x 10 ' % of RATED THEAM Al. -

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24 POWER.

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Flux, P-6, Enable Block Q__

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< 10% of RATED l

4g (not P-10) Input to Low THERMAL POWER

- 11% of RATED THERMAL POWER u+

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, LIMITING SAFETY SYSTEM SETTINGS 1

BASES Intermedia _te and' Source Range. Nuclear Flux (Continued)

Range Channels will initiate a reactor trip t a

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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-sted 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 AT 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 graater than design, as indicated by the difference between top and bottom power range r nuclear detectors, the rer.ctor trip is automatically reduced according to the i notations in Table 2.2-1.

1 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 i

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 Oyartemperature delta T channels and raising the P-8 i

i setpoint to its 3 lopp relue.

and triprfunctions as a,.@. sin this mode of operation, the P-8 interlock L utron Flux trip at the reduced power level, i

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Overpower oT s

IberDverpower delta T reactor trip provides assurance of fuel integrity, e.g., no melting.under all possihlu+em vat romHt40xi Gwite tise required

- iligh Neutron Flux trip.t range..fw Ourtemperature delta T protection, and provides a The setpoint inc]ude.s.correc.tJanssim change; in

' e., j density and heat. r.npar.it

Of mtun.fW.apeMuw, and dynamic compensation.

for, piping delays from the core to the loop temperature detectors.

No credit

...ki!r 'uns taken for operation of this trip in the accidept:analysgs:.bowever, its function 61 capabil.4,ty. at 8.& s;W/hd tHp JeiAiivis required by this L

specificaticy, w enhance the overall reliability of the Reactor Protection

.i System.

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SEQUOYAH - UNIT 2

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' TABLE NOTATION

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. sn 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.

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The channel (s) associated'with the ' protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped condition.

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cification 3 not applicable.

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=:r;;.::d above the P-6 (Block of Source Range Reactor Trip) setpoint.

ACTIONSTATEMEj{T_S a

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 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY

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within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 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:

p

'V, The inoperable channel is placed in the tripped condition a.

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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The Minimum Channels OPERABLE requirement is met; however, 1

L one additional channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 L

l Flux trip setpoint is reduced to less than or equa,l to l

85% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or the 4

QUADRANT POWER TILT RATIO is monitored at leas,t once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when THERMAL POWER is greater than 75% of RATED THERMAL POWER.

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. ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION CHANGE

.>u SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 J'[

(TVA-SQN-TS-88-42)

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DESCRIPTION AND JUSTIFICATION FOR MODIFICATION OF THE TRIP SETPOINT AND. ALLOWABLE

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VALUE UNITS FOR THE INTERMEDIATE RANGE NUCLEAR FLUX DETECTOR AND CHANGES TO THE APPLICABILITY-REQUIREMENTS FOR THE SOURCE RANGE l

NUCLEAR FLUX DETECTOR C

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i ENCLOSURE'2 i

l Description of Change.

Tennessee Valley Authority proposes to modify the Sequoyah Nuclect Plant (SQN) Units 1 and 2 technical specifications (TSs) to revise the trip L

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.-

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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/IP 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 CMny The new Gamma Metrics SR/IR detectors are being installed to achieve L

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 manually 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.

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- The new IR equipment is compatible with the rest of the nuclear

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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 i

from the IR detectors.

Environmental Impact Evaluation L

The proposed revision involves an administrat'ive 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 i

involves no significant increase in the amounts, and no significant change l

in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational L,

radiation exposure. Accordingly, the proposed change meets the h

eligibility criteria for categorical exclusion set forth in L

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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TVA Calculation, " Intermediate Range Neutron i

Flux P-6 Setpoint," Revision 1 l

2.- Safety Evaluation, Revision 3

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Calculation No.

Revision 4

Method of design verification (independent review) used (check method used):

1.

Design Review 2.

Alternato Caleviation 3.

Qualification Test t

Justification (esplein below):

Me t hod,,1:.In the design review method. justify the technical adequacy of the calculation and explain how the ade'quacy was verified (calculation is similar to another, based on accepted handbook methods, appropriate se.asitivity studies included for confidence, etc.).

Le.thod 2: In the alternate calculation mithod' identify the pagen where the anernate calculdtion has been included in the calculttion package

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and o plain why this method is adequate.

Method 3: In the qual'ificallon test method.' identify =the QA documented sourc'e'(s) where testint,'idequitely 60:4cnstrates tlas.t.tuquacy i! tliis i

calc'ulstion and explain.

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CA1.CUI.ATION DESICN VERIFICATION (INDEPENDENT REVIN D

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Calculation No.

Revision Method of design verification (independent review) used (check method useal-3.

Design Review 2.

Alternate Calculation 3.

Qualification Test Justification (esplain below):

?

Method 2:

In the design reglow method, justify the technical adequacy of the i

calculation and esplain how the adequacy was terified (calculation is i

siellar to another, based on accepted handbook nothods, appropriate i

i sensitivity studies included for confidente, etc.).

i-Nethod 2: In the alternate calculation method. Identify the pages where the l

alternate calculation has been included in the calculation package and esplain why this method is adequate.

Method 3:

i In the qualification test method, identify the QA documtated source (s) wher's tcsting adequately 4easonfeces.e1 the adequacy of this calculatton.and esplain.*

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j CAL.CULATICW DESIGN VERIFICATION (INDEPENDENT REVIM TORM l

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Caleviation No.

Revision Method of design verification (independent reflew) used (check method useel:

l 1.

Design Review t

2.

Alternate Calculation-l 3.

Qualification Test Justifiestion (esplain below):

I 4

l Method l' In the design review nothod justify the technical adequacy of the calculation and esplain how tho' adequacy was serifled (calculation is sitellar to another, based on accepted handbook methods, appropriate sensitivity studies included for confidence, etc.).

no t e e.i.

la the alternate caiculation thod, identify the,ates where the

' alternate' calculation has been included in the calculation pactate l

and esplain why thle method is adequate.

Method 3t' In tlie quellTic'at'loti test sletWod.'+16datify the Qt documented

siurders)- tihere* T.es'tieg 'ed> quire %y addraooetestr.co the adeguscy of thi,s calculia'ti~en and explain.

I 1

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7, c3 7 EC w P c.* c. E ee uj e.acs e,rr Instrumentation System (NIS).P-6 is a protection interlock derived e range Nuclear source range NIS reactor trip may be blocked to allow continue i escalation. the source range trip to give the operators time toTherefore, pcVer the same time be above the minimum usable signal in thactuate the block an e cw e intermediate range NIS. establishing th's P-6 setpoint for the intermediate o o cgy of The source range. reactor trip is set at 105 range NIS which corresponds to.about 2 x IQ*9 counts /sec in the soucee rin 10*ge'NIS. 'l amps to 10*). amps with ther'I'ha proe' tss range of the intermed s louer en tcu counts /see in the source ran(ge NIS.'6T* Lbc tw.:n. (merespo 2 about 6 x 10 of 10"11 range (10*10 amps in v'hich to set P-6. This leaves a rangs amos to 2 x 10*9 Midway in this L 1 decade below the source trip to allo #.the operators ti range to achieve a reasonably good signal.and 1 decade c the trip rocess t P-6 setpoint can be calculated by measuring the de e power permissive reactor levels at or below 75% powe.r & interpolating for the pow i urrent for different 10-10 amps, er level 4 T ~}. 0 (* A "*** The Gamma Metrics Design to be installed provides a process ran 200% rated thermal power. This provides an additional two decades of overlap ge of 10*B to with the source range (10*8 to 10-6 % RTP.) be used as the trip setpoint which is functionally equ x 10-5 : RTP shall The values calculated for the P-6 setpoint (using Plant Specifi ent to 10-10 amps, good agreement with the 1 x 10-5 RTP setpoint. c Data) are in adequate margin below the source range trip to give the operatThis se actumre a tripsauc ors time to of the intermediate range drawer (3 decades).and at the same time tw i m ? 9.,._ ..,..-.,--,.,-,-~,,,--n-,,,,---,_-.~,-,.--,
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r t &eC i i i 1 l TENNESSEE VALLEY AUTHORITY i-SEQUOYAH NUCLEAR PLANT { UNIT NUM8tRS 1 AND 2 C PRECAUT20NS, LIMITATIONS AND SETPOINTS 'I FOR i NUCLEAR STEAM $UPPLY SYSTEMS 5 6 .\\ l (... .l I REVISION 9 d MAY,1$81 t..',........,. (fM.S fcised fa$rs 9..... t l EE.A.l - 21..).4 Q "....s. -. il i I V Mk I$) i i ,g-- 4 .mo>to wi cee.teno.s F (,,,,*e = - ernovco var,' ca.:w.:en.' 1.,. WESTINGHOUSE ELECTjt!C CORPORATION ^ N. Nuclear Energy Systems ' ~C I"+'.O E'C' As nr rc. P* 0* 8** 355
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i 17 Cate." 1 2 - Xc 't2-( 3 E I 2., cisma tch . ([( I 4 (FB*l 108 Fl*5205, F8 53CB. FB-540B, 38% of rsted stean FB-! )18 F3 521B, FB 531B, FB.S413) Senerthr$"" I*" I 3, turb1 ne trip stean generator Hi level signal for C feec'witer valve closure, turbi.fie trip and feedwater pump trip (LS.!)?A, LB 527A LB.537A, LB $47A, 75% of level span LB-118A, L3 52SA LB 53BA, L9 64BA, ( LB-!19A,LS$29A,LB539A,LB-BASA) t o %* ',, ~ '+b } X l0 l !!, perptssive ard Interlock e.<cuits / b A. ! P-6(a11cus manual block of source range ! high levelreactortrip)
~~(NC.350, 4C.36D) 10*IO amceres B.,P.7(auto 58tically blocks various "at I pow'4t" trips at low power)~~
' 1. low rrutron flux (see P-10) (( '. 2. 1ow tJrbineload(Seep-13) C. ! P8(alle as one loop loss of flow below !setpoint) '(llc 41N. 'lC 42N, NC 43N, NC,44N) 35% of full' power D.,P-9 lbloc ;s reactor trip on turbine (4 loop cperation) ~ 'tr.ip belcasetpointnuclearpowerlevel) ',(NC 41$ ilC-42$,NC.433.NC-443) 50% of full power *' Ei P10'('U i'i henvas block of power raieve a w ( '(lowsetpaint) trip Intermediaie range Ifr5,and D;18, b1'ecks source range trip
~~and provides a portion of P-7 signal)~~
!(NC-41H, HC 42M, NC 43M NC-44M) F. I 10% of full power P-11(allows :nanual block of safety in-E fjection actuation on low pressurizer pressure. i i($ee!.1.L.4above) L. l -i Pg i4 e Twas t4r 4
-~ -11. 1 t 3b. Low steam line pessure (PB 516A Pl.526A..PB 536A, PB.546A) 600 psig Lead time constant ( I (PY-516B',' PY-5268, PY-5368, PY.546B) 50 secones Lag time constant } (PY-5168, PY-5258, PY 5363, PY.546B) i 5 secor.ds 3c. Low < Low T '.4220 av (78 i 1120. TB , 78 4320, 78 4423) 540'F i 4. Autvistic reset of manual block on high pressurizer pressure (P.11) (P8 d558,PB-4568,P8-4578) 1970 psig 5. Contninment high pressure (PB-9348, PB 9358 PB.9368) 5 1.54 psi; .6. . Time delay on 51 manual reset 1 minute B; Ste6i4'[iteIse14then 1. High steam line. flow (See.!.1.A.3 above) 2. High< high containment pressure ' k6k (PB-1 34A,PB.935A,PB936A,PB-937A) 2.8) psip hi 4 to C. Containme nt Spray Actuation \\ 1. High< high containment pressure,(See !.1.B.2 above) 2. React _or Trips i A., Nuclear ] nstrumentation 1. Source range high-level k. (NC-; 10.NC-320) 5 ( 10 counts /second i 2. Intermediate range high level Current equivalent (NC.: 5F,NC-36F) ' ' '"' P "'" 3. Power range, low range, high level (NC.41P,NC-42P,NC43P,NC-44P) 25% of full power pag.14c, W N t .- L . -... -.. - -.. ~..
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i A n c,, w,,t h M.QtAlldr~b 40,f t::)- -- .,u,, ,g..s. j ..... y A0v. 1 Docu=ent.,. pag., generated from detecting the signal enwa,rd. Where applicsole, this requirement sh:uld be met with all lead, lag, a.d filter time l censtants set to 0FF. 1.14 Controller Tes,sfer functions j WotApplicable ) i 1.i5 Estooints
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IA W t*Mio && /ariable Y ) Rance of Settino j Intermediate Range High Neutron 5 to 30% full power " lux Reactor 7 *ip y source Range H gh' Neutron Flux -10*b to -10~3% of full power 'g ~ [eactorTrip yoO [ntermediateRongo Rod Withdrawal 5 to 25' at full power J, m. 7' >tco (C 11 - 7%g 4g ( 5-6 ~10-5 to ~10'3. of foil cower J I Y \\11 settings v _,-,....~..,,) ,th the exception of time constrats shan b'o e:::ttm:eus'y djustablewith a in their range an'd all time constants shall be ,,, ;ontinuousTy afjustable or adjustable in increments such that any setpoint can bo obtained within t 10% of the setpoint value, t Jer the P 10 si tpoint see Nuclear Power Range Protection (Decument 2). 1.16 - oukements'fTr Test and Cslibration all protection channels.should be supplied with sufficient redundancy "o provide the capability,for channel calibratien and test at p:wer. 'n the case of 1/N logic a bypass'must be.provided to prevent a react:r Srip during tait. ? l. el?3r:1o/;stt37
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t t CN 140. Lfat A L gy 1 .QA Record dj! ,.. V Page 1 of + ag# SAFETY EVALUATION FORM y *E - - m ., met -aa ONM 1. To 3. USQt
~~4. Safety Evaluation Number Seoucych No:! or ?!cnt, Dcy TN o Yes scAstgrg4 2.~~
~~From Fj 6 S!1 P. V No~~
~~9. RIMS Accession Number Ro SQP 65 bil7 503 Rev Tot Date R~~
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~~5. Prepared 6. Reviewed 7. Anoroved 8. Apod iB 2 5 8 81114 5 32 o~~
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~~11. PMP or DCN Number PHP or DCN Revision~~
~~.se uav l un o h I i 't ec.b L d'i t s s e~~
~~12. FCR, SCR..MCR, Cf DCN, or CAQR Number Date of Document (s)~~
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~~13. Other Document Identifier N*u-Date of Document h)/A~~
~~~14. S cla ofV) equirements? See~~
~~15. Potential Tech Spec No Sheet No M Change ( Yes D No Shee5.aM-W See~~
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~~16. References (include systes number and name as appropriate)~~
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~~17. Description of Proposed Activity (Change, Test, or Esperiment) 7Us fw upp da' h. e l u. h a co / c /'e s s 14 fl.e. s neu. a,./~~
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[ c. 4mM,, z. h9erac.s M,n5,ht,2% od 30, gava s.n fluf 3) C " ' ' ' ' [,* cc (Attachments): h RIMS. SL 26 C-K' ,ggg g,4 f. f' dtu g jf.c4.?,., d ; -(-fA 'Y 7 q~.gf ..- ~..,y g n 7; g ~
i e EU.schise, n -Af OF Sheet 2 Safety Evaluation No. ECN L6186 s 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 SQM-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 x 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 and I Wiring" l 11. Tech Specs i 12. Design Criteria SQN-DC-V-26.2 " Environmental Qualification to 10CFR50.49 13. Calculation 1,2-XE-92-1, Rev. 1 RIMS Wo. B25 881117 808. 14. Design Criteria MQN-DC-V-2.3 " Containment Vessel" l 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" l 17. FSAR Tables 7.2.1-1 through 4 and 8.3.1-11, 12, 13, 15, 16 l 18. FSAR chapter 15.0 l 19. Plant Procedure PHYSI-13 " Fire" 20. Westinghouse Test Reports, Nuclear Instrumentation System Isolation Amplifier WCAP-7506-L, NEB 810126303 and WCAP-7819 NEB 8102040314 21. Nuclear Engineering Calculation, " Equipment Required for 10CFR50, Appendix R" SQN-SQ54-127 Rev. 10 (RIMS No. B25800829501) 22. Camma Metrics Neutron Flux Monitoring Instruction Manual No. 72, Rev 4, Contract No. 835545 23. Gama-Metrics Test Report No.135. Rev. O, Isolation Testing of Single Channel Isolators, Contract No. 835545 24. Gamma-Metrics Test Report No. 010. Rev. 1, Neutron Flux Monitoring Qualification Contract No. 835545 25. Gama-Metrics 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-T128-0001 R 3,- 27. Pipe Rupture Calculation SQW-CEB-SCG-4800168 28. Test Report No.12 Rev. O Gama-Metrics RCS series Woutron Flux + Monitoring Seismic Qualification Report and MSLB/LOCA Test Report, Contract No. 835545 29. Gamma-Metrics Test Report No. 31. Test Plan for Qualification of l Gamma-Metrics RCS series of Neutron Flux Monitoring Systems per IEEE STD a e i 323.1974, Contract No. 835545 30. Gamma-Metrics Test Report No. 40, Rev. 3, Test Report for Class it Qualification of Mineral Insulated Cable in the Detector Cable Assembly, Contract No. 835545 1487E n m -t------ t-T-+-* --w-, a rpyh-e i.-e p.-+v w-eqw t--T-4 M'Nw-p ypTTh-**eN swv
1 i 'ECN NO#& C Sheet 3 _ So OF Safety Evaluation No. ECW L6186 WEp 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 Protection Provisions" 34. Design Criteria SQW-DC-V-11.6 "120V A-C Vital Instrument power System" 35. Camma-Metrics Test Report No. 26 Rev. 3 " Optical Isolator, Fiber Optic Transmission System and RCS-211 AmplifLer Qualification. 36. Camma-Metrics Test Report No. 27. Rev. 0, "' seismic Test for optical ) Isolator, Fiber Optic Transmission System and RCS-211 Ampilfier". 37. 120 VAC Vital Inverter Loading Calculation SQN-CPS-021, RIMS Wo. B87 891011 006. I 38. AI-17 Drilling, cutting, Chipping and Excavating, Revision 13. 39. DIM-SQN-DC-V-27.8-3, RIMS No. B37 891101 802. U 40. DIM-SQN-DC-V-27.9-7, RIMS No. 837 891101 801. i 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. 43. HVAC Cooling Load Calculation, Auxiliary Bids., Elev. 714 fL., RIMS No. B25 891102 505. 44. HVAC Cooling Load Calculation. Auxiliary Bldg., General, RIMS No. B25'891102 503. 45. HVAC Cooling Load Calculation, Reactor Bldg., Lower Containment, RIMS No. B25 891102 506 46. VCPS Loading Evaluation, SQN-CPS-022 Rev. 0, RIMS No. B87 891120 002 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 consnitments 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. 1. Replace the non-1E Westinghouse source and intermediate range neutron detectors with Class IE Camma-Metrics source and intermediate range i detectors. l 1487E l
ECNNOt h CS - sheet 4 $ / OF Safety Evaluation No. ECW L6186 NEP 6.6 ADDITIONAL INFORMATION 17. Description of Proposed Activity (Continued) , 2. Reroute cable and conduit, and locato 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 II), 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 Cansna-Metrics Instruction Manual. Reference No. 22. 3. Replace the non-1E Westinghouse pre-amplifier with a 1E Camma-Metrics amplifier assembly. l l 4. Replace the cable from Unit 1 penetrations 43 and 48 and Unit 2 penetrations 23 and 31 to the new emplifier assemblies described in number 3 above. 5. Replace the cables and conduit routed from the amplifiers to the NIS l racks located in the control building. One channel will be routed on elevation 734 above the design basis flood level. L 6. Replace the non-1E Westinghouse intermediate and source range signal processing drawers with Gansna-Metrics intermediate and source range drawers. The new Gamma-Metrics drawers will provide the required electronics to provide qualified signals for compliance with Reference 1. 7. Install a shutdown monitor on each source range neutron monitoring g channel to automatically adjust the high flux at shutdown alarm setpoint l downward for flux decay during shutdown. It will identify and report flux increases that indicat.e a loss of reactor shutdown margin. This will eliminate manual adjustment of this high flux at shutdown setpoint (Reference 22 and 31). l o. 4 1487E ,y-. ,m.. ..c wv.
l ECNfJO LN 4 es Sheet $ Safe!y Evaluation & OF Wo. ECW L6186 NEp 6.6 ADDITIONAL INFORMATION 17. Description of proposed Activity (Continued) 8. Route an isolated tetgorary cable from the Auxiliary Control Room L-10 to Main control Room panel M-19 to provide redundant neutron flux l83 information to the main control room from the Appendix R backup tource range neutron monitor during implementation of items 1-6 above. Only one t channel of the source and intermediate range flux monitors will be worked 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 r will ensure the operator has sufficient neutron flux information during implementation of this ECN. The temporary cable shall be removed atter i n both new channels are declared operable and prior to Mode 3. IG 9. Replace the Appc;ndix R Westinghouse backup source range detector, cabling, and electronics with an optically isolated signal from the new Ctmma-Metrice Amplifier to a source / power range processor and display. This will reduce maintenance, spare parts storage, and enhance the back-up control Room neutron flux readout. This will require a revision-0 l to Referenco 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 qualified penetrations. penetrations 23 (Unit 2) and 48 (Unit 1) have already been replaced under ECN L6490.
~~11. Westinghouse drawings $655D26 sheets 3 and 4 and 1080438 sheets 2, 5 and 43 that are af fected by this ECN will be changed under DCR-3094 for t~~
Units 1 and 2. 12. The new control room indicator and recorder scales required for this ECN are being purchased and their installation coordinated with DCN M01496, Unit 1, and M01497, Unit 2. 13. Westinghouse PLS and setpoint methodology documents that are affected by this ECW will be changed under DCR=3094 for Units 1 and 2. i 14. The Technical Support Center Data System (TSCDS) computer softwarc will i be modifiod to reflect the new range and units (10-8 to 200% RTP) for i: the intermediate range instruments. The calculation of the intermediate and source range start up rates (SUR) will be modified to use time tagged l 1evels rather than assume the stata' is supplied f rom the input i multiplexors every two seconds. l l l l 1487E
WJt40,Lui6r,2S_ Sh30t 6 I 1 ~* gafety Evaluation WS OF Wo. ECN L6186 Ngp 6.6 -ADDITIONAL INFORMATION 17. Description of Proposed Activity (Continued) = 15.. All drilling, cutting and chipping will be performed in accordance with Al-17. Al-17 states that the description of the work shall include all. information required to locate the exact spot for drilling, chipping or cutting by. relating to elevations, distance from column lines and other location references used on plant drawings. Additionally, a review of all effected drawings will be performed to verify embedded piping, electrical conduit, cable troughs, duct work, tunnels or other plant-equipment will not be jeopardized and shall be listed under the drawings reviewed section of the permit. Submittal of an Al-17 permit is to serve a verification that the location of all known embedded piping, electrical conduit, cable troughs, reinforcing steel, duct work, tunnels or other plant equipment will be identified before the drilling, cutting, or shipping is started. Reference No. 38,
~~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 7P 'in to-to) 7.2.1.1.3 8.3.1-13 .2.3 7.2.1.1.8 8.3.1-15 b,* 7.2.2.2.3 8.3.1-16 7.2.4 15.2.4-1 15.2.4.3
~~h Specs (See Question 27)~~
Table-3.3-1 Notes Table -4.3-7 Bases, Limiting Safety System Table 3.3-10 Table 2.2-1 Settings Section 2.2.1 This activity will be performed in the following stages: Steme 1: Will consist of completing items 1 through 8 and 10 by the end of the Unit 1 Cycle 4 outage for Unit 1. Star.e 21 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 modification is not part of the NRC commitments for Cycle 4. 4 k 1487E
ECH NOuLW f* * >- 7 -- SV OF Sheet 7 Safety Evaluation No. ECN L6186 NEP 6.6 ADDITIONAL INFORMATION .I 10. Systems. Structures. Components Affected WM - Neutron Monitoring RPS - Reactor Protection System PAM - Postaccident Monitoring SR - Source Range , CV - Containment Vessel IR - Intermediate Range RTP - Rated Thorval Power ITEM l COMPONENT l SYSTEM AFFECTED l DESCRIPTION OF CHANGE { 1 l Source / Intermediate Neutron 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 ] l l chambers and has a pulse signal out-- l Channel 1. XE-92-5001 lUnit 1 lput proportional to reactor power l 2 l Source / Intermediate Neutron lNM, PAM, RPS, l Replace non-1E detector with IE l l Detector and Cabling
~~Appendix R SR l qualified detector which contains l~~
l l ltwo identical, redundant fission l l l l chambers and has a pulse signal out-1 l Channel II. XE-92-5002 lUnit 2 lout proportional to reactor power l 3 l Unit 2 Primary Containment l Penetration l Replace 75 ohms.triax with l l Penetration No. 23 lFeedthroughs l50 ohms triax l l Channel II l l l 4 l Unit 2 Primary Containment lCV and l Replace complete non-qualified l l Penetration No. 31 'l Penetration lwith IE qualified penetration 1 lChannel I l l l p 5 .l Unit 1 Primary Containment l Penetration l Replace 75 ohm triax with j l Penetration No. 48 lFeedthroughs l50 ohm triax l l l Channel II l l l 6 l Unit 1 Prinary Containment lCV and l Replace complete non-qualified I l Penetration No. 43 l Penetration. lwith IE qualified penetration l L l Channel I l l l 7-l Channel 1 Signal Amplifier lNM, PAM, RPS l Replace non-qualified with IE l l lXM-92-5001A - U1 l Appendix R. SR l qualified amplifier to mate with l IXM-92-5001 - U2 l Unit 1 Inew detector l l f 8 'l Channel II Signal Amplifier lNM, PAM, RPS l Replace non-qualified with IE i lXM-92-5002 - U1 l Appendix R SR l qualified amplifier to mate with l lXM-92-5002A - U2 IUnit 2 Inew detector l 9 l Shutdown Moniter lNM, PAM l Add new device to automatically l l Channel I, XIS-52-5001 l ladjuct high flux at shutdown i L l I lalarm down l l 10 l Shutdown Monitor lNM, PAM l Add new device to automatically l 2 l Channel II, XIS-92-5002 l l adjust high flux at shutdown .I I I lalarm down l l l l' 1487? l
1 J fiCN NO. UrtB(, p Sh;ct 8 ] Safety Evaluation OF No. ECN L6186 NEP 4.6 ADDITIONAL INFORMATION J 18. Systems. Structures. Components Af fected (Continued) NM - Neutron Monitoring RPS - Reactor Protection System .PAM - Postaccident Monitoring SR - Source Range t CV - Containment Vessel IR - Intermediate Range i RTP - Rated Thermal Power i ITEM l COMPONENT I SYSTEM AFFECTED l DESCRIPTION OF CMANCE 'l 11 l Source Range Drawer lNM, PAM, RPS. l Replace Westinghouse with l l Channel I, KK-92-5001 l l Gamma-Metrics ~l l l l } 12 l Source Range Drawer lNM, PAM, RPS l Replace Westinghouse with l l Channel II, XX-92-5002 l l Gamma-Metrics l 1 I l l 13 l Intermediate Range Drawer 101, PAM, RPS l 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 l l l10-8 /2 x 102 % RTP l 1 l l l 14 l Intermediate Range Drawer lNM, PAM, RPS l Replace 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 l10-8 /2 x 102 % RTP 'l l l l l 15 llE Optical Isolator, Unit 1 lNM, PAM, lNew device to isolate Main Control l l Channel I, KM-92-5001B l Appendix R SR l Room from remote shutdown l l IUnit 1 l l 16 l1E Optica1' Isolator, Unit 2 lNM, PAM, lNew device to isolate Main Control l l Channel II, XM-92-5002B -l Appendix R SR l Room from remote shutdown. l l lunit 2 l 1 17 l Appendix R Source Range lNM, Remote l Replace Westinghouse source range l l Channel (detector, pre-amp l Shutdown l channel with optically isolated l l land drawer) XI-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 l } 1487E
~ I L _ [' ECN tJO. LeerNo P3 4 Sh 0t 9 t 9 OF Safety Evaluation-4 No. ECN L6186 ? NEp 6.6 J w i ADDITIONAL INFORMATION 19. Safety Function (s) of System (s) Affe_cted The safety functions of the systems affected by this ECW are described below: 1. POSTACCIDENT MONITORING (PAM) [ The safety function of the postaccident monitoring system is to provide f information on plant variables required by control room operating T i personnel during accident situations to 4 s. 1. permit the operator to take preplanned manual actions to accomplish j safe plant shutdown. 2. determine whether safety systems or systems important to safety are $ performing their intended functions. t 3. determine the potential for causing gross breach of the barriers to I radioactivity release and to determine if a gross breach of a I. barrier has occurred. A L 4. assess the operation of plant systems to make appropriate decisions' as to t. heir use. 5. allow for early indication of release of radioactive materiaF order to initiate action necessary to protect the public and j estimate the magnitude of any impending threat. l-The PAM variable affected by this ECN is. neutron monitoring. Neutron I monitoring provides information for purposes 1 and 2 above. Additional 1 safety function information is available in References 1 and 5.- ? i 3 2. NEUTRON MONITORING SYSTEM - SOURCE AND INTERMEDIATE RANGE G j 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 .. - lR3. > trip setpoint (25% RTP), Tech Specs Table 2.2-1. This trip, which provides protection during reactor startup, can be manually blocked if f 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 bypasped 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 i -e r v
m_ _ 1; - el ECN NO._ Lt> IS(p e s i Safety Evaluation 52_.0F_ _ ~ No. ECN L6186 NEP 6.6 ADDITIONAL INFORMATION 19. Safe 6v 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(o l protection during reactor startup and plant shutdown, can be manually l bypassed when one of the two intermediate range channels reads above the. -+ l -P-6 setpoint value (source range outputs disabled and intermediate' range on scale power level) and shall be automatically reinstated when both intermediate range channels decrease below the P-6 value. This-trip shall be automatically bypassed by two out of four logic from the power range permissive (P-10). This trip function shall also be reinstated below P-10 by an administrative' action requiring manual actuation of two control board mounted switches. Each switch will reinstate the tr%p function in one of the two protection logic trains. The source range trip shall be set l. between the P-6 setpoint and the maximum source range level. The-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, f~~
The source and intermediate range neutron monitoring system also L. comprises a portion of PAM for purposes described above in "Postaccident p' . Monitoring". Additional detailed safety function information is available in References 2, 39, and 40. h ' ) Another safety function of the source range neutron flux monitor is to provide an increasing count rate when RCS boron concentration decreases L during shutdown, which is a condition II fault described in Chapter 15 p of the FSAR. L The intermediate range provides a signal 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 is. designed with sufficient independenco, g3 redundancy, and testability to perfonn 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, 1487E i L. Am--- . m a y va-ewr-w-- r w
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f !iCNI.O.m.tg4 5 Shoct 11 Safety Evaluation 68 op No. ECN L6186 NEp 6.6 ADDITIONAL INFORMATION 19. Safety Function (s) of System (s) Affected (Continued) 3. PRIMARY CONTAINMENT SYSTEM The primary Containment System will'be breached during the replacement of electrical containment penetrations 43 (Ul) 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. I 'The' functional performance requirements of the Reactor Trip System shall include provisions for automatically initiating a reactor trip: L a. -Whenever necessary to prevent fuel damage for an anticipated j transient (Condition II). b. To limit core damage for infrequent faults (Condition III). c.. To keep the energy generated in the core under control to limit s fuel damage such that 10CFR 100 dose-limits are net'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 Source range neutron flux is an instrumentation channel required in the event of a main control room evacuation for safe shutdown. Reference SI 21. 1487E
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~~C:!NOJAJ@ e5 Shr,t 12 gpr OF Safety Evaluation No. ECN L6186 WEp 6.6 SAFETY EVALUATION i~~
20. Effects'on Safety This modification and the Tech Spec change will not affect the safety , functions of the systems' listed in number 19 above or any other systems important to the safety for the following reasons: 1. This ECW upgrades the source and intermediate range neutron monitoring system in order to meet the qualification guidelinen of Reg. Guide 1.97 - The source and intermediate neutron monitors will provide a primary R2. safety function by providing the control room operator information to take preplanned manual actions to accomplish safety shutdown of the plant during accident conditions. As discussed in Referenes 6, Table 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 ^f addition to power range trips. Also, the interlocks described in number 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 room. Tech Spec 3.3.3.5 only requires that the backup source range neutron monitor 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.- l .The existing Westinghouse backup source range electronics has a built-in L isolated output which will be used for the main control room readout. Westinghouse testing shows that no credible fault could damage the g backup source range electronics with-this temporary cable installed. References 20 and 41. This will allow the modification to proceed. g) 7 i - 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 1 .--e,-, s----w.-,n v w-s. e s e n- ,,-,,.w- .,e-e .--p,-m
ECN NO._LCab #5 Sh3ct 13 Safety Evaluation 0"O OF Wo. ECW L6186 ? NEP 6.6 SAFETY EVALUATION 20. Effects on Safety (Continued) q i 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 Camma-Metrics units is documented in calculation SQN-CPS-021 (Reference 37). The effect of this increased load has been evaluated in calculation SQN-CPS-022 (Reference 46) for R3-both the vital AC and DC power systems. The addition of the 1E optical isolator assembly, References 35 and 36,'will allow the use of one of the Main Control Room detector signals yet maintain the separation required for Appendix R. ECN L6186 will, therefore, have no effect electrically on any other system important to safety except for the 120 VAC vital Power boards and 125V vital DC power systems. 4. The upgrade of the source and intermediate range neutron monitoring system.to Reg. Guide 1.97, Rev. 2, category I qualifications assures that 4 '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 L fully redundant and separate from the other in accordance with the requirements of Reference 10. This assures that the primary safety function of the source and intermediate range neutron flux monitors is not compromised by single failuro. 6. The upgrade of the source and intermediate range neutron monitoring system to-1E requires the components of the' system to be' seismically lg3 - mounted. Therefore, components or equipment of other systems important to safety will not be subjected to seismically induced missile damage. 7. This Item moved to item 2 this question for clarity. 8. The effects of pipe rupture (Reference 27) on the new cable routing, junction boxes, and other hardware have been evaluated in accordance with R SQEP-51. The evaluation concluded that the components will not be af fected by pipe rupture, assuring the reliability of the primary safety function of the source and intermediate neutron monitors. (See Special Requirements No. 2). AI t 9. Containment electrical penetra1Lons 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 capable of providing a radioactivity barrier. 1487E
~- ir', %i bcN NO.W Ne #5 Shist 14 Safe y Evaluad on 08 OF No. ECH L6186 NEp 6.6 SAFETY EVALUATION . 20. - Effects'on Safety (Continued) ~ 10. All components affected by this ECN located in harsh or essentially mild environments will be qualified for these environments in accordance with the requirements of Reference 12, ensuring reliability of 4 : instrumentation during Chapter 15 condition III or IV faults. (See gt Special Requirements No. 5).- 3 11. When.the ECN is completed, one channel of the source and intermediate i O neutron monitoring system will provide a' signal through an optical O 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 room will not affect the backup source and intermediate range neutron monitoring system's primary safety function. 12. One channel of the source and intermediate range monitors will be routed 1 above Auxiliary Building floor elevation 734 and the other will be routed above floor elevation 714. The Auxiliary _ Building floor will provide a fire barrier, ensuring that fire in the Auxiliary Building will not affect both neutron monitoring channels and protects one channel from the Design Basis Flood, Reference 33. 13. The new shutdown 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 function is manually performed and addressed in FSAR Section 15.2.4.2. When the count rate achieves a steady value and then eventually increases, the alarm setpoint remains at its lowest value. An alsem will occur when the count rate reaches a value equal to the alarm setpoint 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 be one shutdown monitor for each g3 _l neutron monitoring channel. Each will be electrically separate and fully redundant in accordance with the requirements of Reference 10. Also, visual verification of the setpoint and count rate can be perfomed by the operator any time below 104 counts per second, (Reference 22), when the shutdown monitor is in service to ensure the monitors are perfoming their function. The high flux at shutdown alarm will continue to perform its fpnction as before, with the' alarm setpoint 'being adjusted automatically by the shutdown monitors. The reliability, redundancy, and shutdown monitor tracking verification feature ensures that this high flux at shutdown alarm function will not be affected, hence the consequences of boron dilution, a condition II fault, will not be increased. 1487E
II l HI Il lI .E S e ECNNo.tLth Cs sh3ct'15 61L OF Safety Evaluation' J Wo. ECN L6186 NEp 6.6 SAFETY EVALUATION 20. Effects on Safety (Continued) o , 14..The new Camma Metric source range neutron monitors are equivalent to f the Westinghouse BF-3 detectors with regard to instrument sensitivity. ' Indicated response to neutron flux is not changed-significantly by this modification and will not affect reactor trip interlock setpoints or alarm setpoints with regard to the power.at 4 which they occur. Technical Instruction (TI-) 81, NIS calibration 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 nultiplied 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. In the past, the detector process output was in amps which equated ~to rated thermal power. The new Gamma-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 Gamma-Metrics detectors, an. initial expected full power calibration factor will be supplied by the vendor, A NIS calibration procedure will be prepared which will be a part of o pMT-62, and include instructions for the initial startup after the R3 . installation of the new detectors. In order to help assure that the new source / 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 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 intermediate range 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. 1487E
i ~ .b jCN NO. f a t % D Sh:3L 16 63 OF~ No. ECN L6186 Safety Evaluation NEP 6.6 'i SAFETY EVALUATION - 20. Effects on Safety (Continued) 15.' A fire hazard analysis evaluation, SQN-26-D053-EPM-NHS-022289 has been performed for Unit 2 to ensure that a fire in one location will not affect both instrument channels. See Special Requirement No. 1 for the Unit 1 limitations until this analysis has been performed. 16. There will be minor additional heat loads added to the Reactor, 5 Auxiliary. and Control Buildings, which is documented *in References 42, 43, 44, and 45. This additional heat 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 NRC connitments in the PAM licensing submittal (RIMS L44 881228 808). -Therefore, this safety evaluation.~~
will only refer to the CRFSAR for PAM implementation. See Special E ' Requirement No. 6. l 21.. Would the proposed activity increase the probability of an accident-previously evaluated in the SAR? l$l Yes L5l^ No Justification: The source and intermediate range neutron monitoring system does not provide a function to reduce the probability of Condition III or IV faults. However, the source range monitor provides an alarm for RCS boron dilution t during. shutdown (a condition II fault). The new shutdown monitor described L .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 l.' -correct at all times for'all background conditions, and subsequently reduce j the probability of boron dilution going unnoticed during shutdown. As discussed in Number 20 " Effects on Safety", the lowcred intermediate j. range reactor trip and rod stop setpoint during initial startup after core i i. reload will ensure conservatism in these setpoints. l 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. 1487E er r--,u.--e. r.. w* ,.*,-e.w-.we.vr w ,*er -y-- + m--w
+ ~ ECN NC.u,um,es t. ~ 6M OF Sh30t 17 Safety Evaluation No. ECN L6186 NEP 6.6 .r SAFETY EVALUATION 22. Would the proposed activity increase the consequences of an accident previously evaluated in the SAR? L_/, Yes LK/ No Justification: The source and intermediate range neutron monitor's function in the Reactor Protection System is not creditod for mitigating the consequences of a-Chapter 15 accident, according to Reference 6. However, the source and-intermediate range neutron monitors do provide the operators reactor power level-information af ter a condition II, III, or IV f ault in order to take IR3 preplanned manual actions to. accomplish safe shutdown. As discussed in #20 " Effects on Safety", the new source and intermediate range neutron monitor components and the new shutdown nonitors 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 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, III, or IV f aults so he may accomplish safe shutdown and mitigate the consequences of such faults. The temporary breach of containment to replace the penetration can only be 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-containment breached during Mode 5 has already been analyzed and consequences of an accident cannot be increased. 23. Would the proposed activity increase the probability of a malfunction of equipment important to safety previously evaluated in-the SAR? f[7 Yes L57 No Justification: As discussed in Number 20 " Effects on Safety", the 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. Also, as discussed in Number 20 " Effects on Safety", the source and intermediate range neutron monitors cannot become seismically induced missiles and contribute to the probability of malfunction of other equipment II -important to safety. i 1487E
v nit # K riiCN NO.tMN 25 - Shoot 18 ~ Safety Evaluation (M OF-- a No. ECN L6186 NEP 6.6 4 Safety Evaluation 23. (Continued)- As discussed in Number 20 " Effects on Safety", the addition of the redundant ~ shutdown nonitors will enhance the capability of the high flux at shutdown alarms 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 p3 ratio and the average neutron rate of the last 1200 counts, Reference 22. Survelliance testing will ensure that these alarm setpoints are' operable. As discussed in Reference 20, the operator does not depend entirely on this alarm setpoint but has audible indication of increasing neutron flux from the audible count rate drawar of the NIS system and visual indication from counts per second meters for each channel on the main control board and source range drawer. 24. Would'the-proposed activity increase the consequencen of a malfunction of equipment important to safety previously evaluated in the SAR? -f[7 Yes f57 No Justificationi 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 available 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 f57 No 148?E~~
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I:C!mO.LCeHW O Sh st 19 g,, OF-Safety Evaluation No. ECM L6186 NEp 6.6 ' Safety Evaluation -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 buffer whose input is g3 optically isolated from its output, Reference 22, 26. Would the proposed. activity create a possibility for a malfunction of equipment of a different type than any evaluated in the SAR? (([.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 R3 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, 15, 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.~~
1487E I ~ l
4 l " '. Y ' (ygggg pj 6~7 _OF__--.-a Safety Evaluation L-- No. ECN L6186 NEP 6.6 Safetp Evaluation 26. (Continued) .l. l l MAS A NEW MALFUNCTION ITEM l COMPONENT l DESCRIPTION OF CRANGE IBEEN CREATED? _ Source / Intermediate Neutron l Replace non-1E qualified with lNo, New component isl l I l Detector and Cabling l detector which has output la sealed, pressure l l Channel I, KE-92-5001 proportional-to reactor power l tight, qualified l l l l device and contains l l l ltwo identical, l l l-l redundant fission. l l l l chambers _l 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 l proportional to reactor power l tight, qualified l l l device and contains l l l ltwo identical, l l l l redundant fission l l l Ichambers l 3 l Unit 2 Primary Containment l Replace 75 ohms triax with lNo, a qualified: l l Penetration No. 23 150 ohms triax to match new l50 ohms feedthrough l lChanoc1 II-l detector cable impedance ,shall replace the l l l 'existina 75 ohm l 4 l Unit 2 Primary Containment l Replace complete non-qualified lNo, an electrically l l Penetration No. 31 lwith IE qualified penetration l qualified one shall l l Channel I l l replace the existing l l l Inon-oualified l I 5 l Unit 1 Primary Containment l Replace 75 ohm triax with 50 lNo, a qualified l l Penetration No. 48 lohms triax.to match new l50 ohms feedthrough l l Channel II l detector cable impedance. lshall replace the l 1 I lexistina 75 ohm 'l 6 l Unit 1 Primary Containment lReplacc 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 1E. 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 l loualified l 8 l Channel II Signal Amplifier l Replace non-qualified with IE lNo, the function is l lKM-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 lnew design is l 'I I loualifled l 1487E t
ECN NOp p ab _ .U Sh r t 21 M OF Safety Evaluation t No. ECN L6186-NEP 6.6 Safety Evaluation .I 26. (Continued) i l l lHAS A NEW MALFUNCTION ITEM l COMPONENT I DESCRIPTION OF CHANGE IBEEN CREATED? i 9 l Shutdown Monitor l Add new device to automatically.lNo, as discussed in l l Channel I . adjust high flux at shutdown lNo.,20, even if this' lX1S-92-5001 l alarm down lnew device were to l l l l fall totally, the l l l l increasing audible l t l l l count rate would l l l lstill.be available l l l l lto alert the oper. l l-I 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 lNo. 20, even'if thisl lXIS-92-5002 l alarm down lnew device were to l l l l fail totally the l l 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 11 l Source Range Drawer l Replace Westinghouse with lNo, new qualified l I l Channel I, l Gamma-Metrics l drawer has the l L-lXX-92-5001 l lsame output.and l l-l -l l function, except l i l l lhigh voltage does l g^ l l lnot have to be l l l ldeenernized l L 12 l Source Range Drawer l Replace Westinghouse with lNo, new qualified . l r l.i l Channel II, l Gamma-Metrics l drawer has.the lXX-92-5002 l lsame' output and l L l l l function,except l l l l lhigh voltage does l l-l l lnot have to be l l. l l Ideenernized-l l' 13 l Intermediate 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 l l functions l D 14 l Intermediate Range Drawer l Replace Westinghouse with lNo, new qualified l ~~ l Channel II, l Gamma; Metrics l drawer has same l lXX-92-5004 l l output and l l l l functions l 15 l1E 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 l -- ~. ~ -. .n,
1 J 1 =. ECN NO. LI*I8/, f4_ Shstt 22
~~I Safety Evaluation 1~~
~~49 OF~~
~~30. ECW L.6186 NEP 6.6 Safety Evaluation 26.~~
(Continued) ,l l lHAS A NEW MALFUNCTION ITEM l COMPONENT l DESCRIPTION OF CHANGE IBEEN CREATED? 16 IE Optical Isolator, Unit 2 New device to isolate Main No, any malfunction Channel II, . Control Room remote shutdown of this will result l lXM-92-5002B l lin loss in detector l l l l signal but there is l i l l lan associated ll l l Iredundant channel L 17 l Appendix R Source Range l Replace Westinghouse source lNo, new source / power l l Channel (detector, pre-amp l range channel with optically l drawer is a l land drawer), lisolated output from l qualified it device U lKI-92-5 l1E-qualified Gamma-Metrics leven though its [ l l amplifier and new source / l signal and power 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 reliability of the l l l lnew fission chamber l l l l detectors and the l l l l routing of all. l p; l l l cables and elec-- l l l ltronic mounted abovel l l lthe Design-Basis l l i l flood level, Ref. 33l l. l lwill offset the l l - l l deletion of the l l l l backup source range l l l l detector and will l l l lnot increase the l l l l susceptibility of. l l' ,1 ITech. Spec. 3.3.3.5.l 18 l Components 1 through 17 l Increased load to the 120 VAC lNo, prior to l l l Vital Instrument Power Board l energizing the equip l I l lthe capacity cale. l l l lfor the Vital AC 1. I 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 lthe capacity calc. l l l lfor the Vital AC l l l Ishall be performed l 1487E
.s N UCN NOiF/IO3 91 Shoot 23 l ,gj_,pp Safety Evaluation No. ECW L6186 NEp 6.6 Safety Evaluation 27. Would the proposed activity reduce any margin of safety as defined in the basis for any technical specificationi f[7 Yes 457 - ' No Justificationt_ 1. The containment integrity and penetration operability requirements are 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" (#18). Testable penetrations Surveillance Instruction (SI-157) will be 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-trip) Setpoint"_is_no longcr required. This was only required when the previous design source range detector was being deenergized. The new detector and amplifier design will be full range and will not be deenergized.' However, in order to reduce operator and site procedures impact, all source range outputs can be disabled above the p-5 setpoint. Therefore Tech Spec table notation for Table 3.3-1 will be revised as g3 shown on Sheets 26 and.27 of this safety evaluation. Sheets 28 and 29 will not be impacted since table notetion is still required. Detector replacement will occur in Stages 1 and 2 described in Block 17, 3. Tech Spec Tables 3.3-10 and 4.3-7 " Accident Monitoring Instrumentation" -will be revised'as shown on SheetsL30 through 33'of this safety evaluation to show the new PAM instrumentation identified in " Systems, j g3 Structures or Components Affected". The monitoring' equipment identified will be installed in Stages 1 and 2. j 4. The intermediate range neutron flux p-6 permissive engineering units will change from amps to percent power to provide the plant operators more 1 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 performed during Stages 1 and 2. i 4 1487E J i i is i i i i i o i i si i
- a ECN NO._4l,/A&, a 3, 8hn#L #4 ?I OF Safety Evaluation / No. ECN L6186 L i NEP 6.6 V Safety Evaluation i 2 7.'- (Continued) j 4 (. 5. A demonstrated loop accuracy analysis (Reference 26) will be performed C 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 i Stages 1 and 2 (see Special Requirements No. 3). -~' Technical Specification Section 2.2.1, " Limiting Safety System p Settings" (Bases), intermediate and source range nuclear flux states "no credit was taken for operation of the trips associated with either i the intermediate or source range channels in the accident analyses; L however, their functional capability at the specified trip settings is 1 required by this specification to enhance the overall reliability of- ~ the Reactor Protection System." ~ 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", e e two neutron monitoring channels will be available at all times during L refueling while this modification is implemented, ensuring changes-in the reactivity condition of the core that may occur will be detected. This will be ensured during Stages 1 and 2. [ 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 g revision is not warranted. 8. Tech Spec Section 2.2.1 " Limiting Safety System Settings, Bases", the I reference to intermediate range current level, shall be revised as ~ shownLon pages 43 and 44, since the intermediate. range will read in i-l- percent power. l .r i L Based on the discussion above, implementing ECW L6186 will not challenge or degrade the margin of safety as defined by any of the Tech ~ Spec bases. 28a. Special Requirements f 1. The Fire Hazard evaluation for the Unit I conduit installation for this R3 ECN has to be completed prior to installation beginning in the outstanding locations. The outstanding locations are inside the Unit 1 annulus and containment vessel, All elevations in the Auxiliary Building have been evaluated. l l r 1487E l, .m
v 9 ECH NO.__ WAG d Sheat. 25 12-OF- . Safety Evaluation-4 No. ECN L6186 ) l NEP 6.6 Safety Evaluation j 28a. (Continued) 4 -, 2. The unverified assumption in pipe rupture calculation associated with each individual channe), 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 Instrument Accuracy calculation (SQN-REB-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-g3 jg - e lengths for that channel. o-5. Installation of detectors, detector cabling, and electrical penetrations cannot begin until EQ Binders SQN EQ'INM001 and SQN f 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 1: Based on the safety evaluation provided above, it is-concluded that no unreviewed cafety question exists as a result of implementing ECN L6186. I 4 S 1 1487E
ECNNOlUAC RE g p g,g g Q9 g ( Q 1@ Sdet Eydud 'e y TABLE 3.3-1 (Continued) Re. ECR L619(o - i TABLE NOTATION 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. 1 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 otovisions mf (na@_3.0Arre not applicable. ) kW~ b $$*t:WIk:1:rmaybehdiNN; above the P-6 (Block of Source R ~i Kange Reaf W TTip D e 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 h6urs, or be in HOT STANDBY within the next 6 hours 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 '1 l-provided the following conditions are satisfied: 3) The inoperable channel is placed in the tripped condition a. within 6 hours. J lR51 - b. The Minimum Channels OPERABLE requirement is met; however, ~ ! s one additional channel may be bypassed for up to 4 hours 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 hours; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours. I d. The QUADRANT POWER TILT RATIO, as indicated by the remaining y i three detectors is verified consistent with the normalized L symmetric power distribution obtained by using the movable incere detectors in the four pairs of symmetric thimble locations at least once per 12 hours when THERMAL POWER is greater than 75% of RATED THERMAL POWER. l l W September 17, 1986 SEQUOYAH - UNIT 1 3/4 3-5 Amendment No. D 9 rm +--e.,.,s-g =,sp- ,.-.,wwe r e--.+- -r w-
gy '1 M t P G.(. Sheet.1t? l [ ECN NOW A r. at y S ASd g,jpghh YN - '4 yy .g, , 3 4. OE ) u N o, E t.N L ES ,' L! s TABLE 3.3-1 (Continued) TABLE NOTATION A 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. AR -The channel (s) associated with the protective f' unctions derived from the out of service Reactor Coolant Loop shall be placed in the tripped i condition.
~~T1e nm d e b e ic on 3.0.4 are not applicable.~~
1
~~NWbb7dt a be UNINN;' :' bove the P-6 (Block of Source. k Range ReYctor Trip~~
~ ACTION STATEMENTS l ACTION 1 - With the number"of OPERA'BdE cha$els one 'less than the Minimum Channels OPERABLE requirement,. restore the inopera L channel to OPERABLE status within 48-hours or be in HOT within the next 6 hours and/or open the reactor trip breakers. ( ACTION 2 - With the number of OPERABLE channels one le Number of Channels, STARTUP and/or POWER OPERATION may pr l provided the following conditions are satisfied: k The inoperable' channel is placed in the tripped. condition a. within 6 hours. IR3L b. The Minimum Channels OPERABLE requirement is met; howeve 'a E one additional channel may be bypassed for up to 4 hours \\R3% i for surveillance testing per Specification 4.3.1.1.1. l c. Either to 75%,of RATED THERMAL POWER and the Neutron Flux trip setpoint is reduced to less than or equa,l to u 85% of RATED THERMAL POWER within 4 hours; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours. L d. The QUADRANT POWER TILT RATIO, as indicated by the three detectors E is verified consistent with the normalized-f symmetric power, distribution obtained by using the movable i L incore detectors in the four pairs of symmetric' thimble locations at least once per 12 hours when THERMAL POWER is greater than 75% of RATED THERMAL POWER. - i september 17, 86 SEQUOYAH - UNIT 2 3/4 3-5 Amendment No. 39 NF-a t 0 gG ', j s.(A, < ~(( ', A{5hf,; ' i; W
I t\\i 3 {' ,N2o rEE S R ~ $f '. aW S 5 5 4 4 R R
~~1. s. <n.e l~~
~~l e N r,m d O I 2 2 2 3 p g 3 1 I T 45 6 6 6 6 s C / 4 = A~~
~~f. 9 <c N*5~~
~~r E~~
C) L d Uau BS n d s AE a n dd CD a nn IO aa LM 2 2 2 2 2 P ,4 2 2 2 2 2 P A _ 1 1 1 1 1 3 1 1 1 1 .SE ~ MLL 3 UEB R MNA 2 3 3 3 2 21 INR NAE N. 3 3 3 2 O. I HP MCO I T A TN .L E S M LP U EI R NR T NT 1 2 2 2 1 1 S N A s3 HO 0 2 2 2 2 2 I CT i 3 M b E E T L S B Y c A S T .S g1 P .L OE I R NN T N 32 . d a' LA 2 4 4 4 2 R AH T C 22 4 4 4 4 3 O
~~c O -~~
T k. C TF yh A O E er c R e .~ x 7s u l F a a .e n h g o x \\ x x x r u h H ~ i h u u u t l u. l l g l u F Tn n w i F F e o o o H l % )o ai N n n n n i L e tt t v p o o o o a i r re re e t l a a e e e yv r er r r r L r t tt tt g u De Te u' ue u ua ua n e p p s s r u e eR eR a N O aO s s e r N N N R e t e~ e t .h o e e rp l p r r a t t n uo eo P P W ,v ,v e epw t o Do t. c e ei ei t guo a g gt gt a ntd r e aL L r r r e n ni na i art er rr z z z t s R a as ag d Rau pu eu i i i e e e . W.
~~R Ro Re e~~
th mo wo r r r S l P N m eSS eF oF u u u L a r r r T A u e eh eh e r ~r c N n w wg wg t u t p s' s s r r O a o oi oi n o.. v v r r r s s s e e e e e I M P PH PH I T SAB O O P P P CNU F 1 2 3 4 5 6 ). ~ 0 1 7 8 9 1 1 oggx* h*w U$E. M E :f s.:, Rs T" %Ea e " t .{j:)ii} 'i, !\\ i (
m - + g-wep ef : 5( k.(2t9 b; Th'a tkany edi Jo4er rep;rd ~ pcr lisiws,"w is' g3 is ~ Sddr ( /y / M s d&y un%+6 n.2
~~1.. ece utor '~~
1 TABLE 3.3-1 v, REACTOR TRIP SYSTEM INSTRUMENTATION i g ' MINIMUM TOTAL NO. CHA' NELS CHANNELS APPLICABLE N FUNCTIONAL UNI 0F CHANNELS TO TRIP OPERABLE MODES ACTION' c .= i U 1. Manual Reactor 2 1 2 1, 2, and
~~1-m I~~
~~2. Power Range, Neutron x 4 2 3 1, 2 2, - 3. Power Range, Neutron Flux 4 2 3 1, 2 2# High Positive Rate i'~~
~~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 I~~
~~Y 6. Source Range, Neutron Flux AI i A. Startup. 2 2 , and~~
~~4 B.~~
Shutdown 2 0 1 3, 4 and 5 5 l 7. Overtemperature AT i Four Loop Operation 4 2 3 . 1, 2 6, 8. Overpower AT R33 Four Loop Operation 4 2 3 ,2 6, u, m,p a 1 l R33 Es 9. Pressurizer Pressure-Low 4 2 .3 1, 6, a& kE 10. Pressurizer Pressure--High 4' 2 3 ' 1, 2 6 T ~ e+
[
~~11. Pressurizer Water Level--High-3 2 2 1, 2 7 m = h 30 uo P. E~~
~~3 o~~
~~3-t Z l i~~
~~(.'~~
f . U S b - - - --]
~ s .n APL* f.i.5 den.f[I JO U! - S< (< ly Evale/<a% E TABLE 3.3-10 A)..Ev2litSCI .m gs ACCIDENT MONITORING ?NSTRUMENTATION I 3 -INSTRUMENTi MINIMUM i 4 -REQUIRED NO. c= CHANNELS 0F CHANNELS Q 1. Reactor Coolant T _0PERABLE gg (Wide Range) 2. Rear. tor Coolant TCold-( ide Range) 2 1 3. Containment Pressure (Wide Hange) 2 1 4. Refueling Wat[er Storage Tank Level 2 1 2 ' l RF 5. Reactor Coolant Pressure (Wide Range) 1 ' Pres'surizer LebeI'(Wide Range) 6. 2 1 k Steam Line Pressure 2 l R5s cn 7. 1 w 8. Stshin Gener'at'o'r Le'v61 - (Wide Range) 2/ steam line 2 1/ steam line N* 9. Steam Gellerator Level -(Narrow Range) 1/ steam generator 1/ steam generator w
~~10. Auxiliary.'Feedwater Flow Rate 1/ steam generator i~~
~~1/ steam generator f 1/ pump . E'~~
11. Reactor Coolant System Subcooling Margin Monitor Jg G/ pump i

12. Pressurizef PORV Positiert Indicator
~~1 0~~
~~1/ valve {~~
13. Pressurizer PORV Block. Valve Position Indicator **
~~2/ valve # -a~~
~~14. Safety~~
~~Valve Position Indicator~~
~~~ 2/ valve p g 3 1/ valve~~
~~15. Containment Water Level (Wide' Range) 2/ valve #~~
~~1/ valve~~
~~16. In Coie Thermocouples 2~~
~~1 1~~
~~17. Reactor Vessel Level instrumentation System *** 4/ core quadrant 2~~
~~~ 2/ core quadrant g.g. if-S~eu.a . L.kes.uf. 4.,y.Nexlem.t s k. g,; f 4 1 a R g,, +: 2. ADD lR50~~
~~Not applicable if the associated block valve is in the~~
~~, g- ~ t g~~
~~Not applicable if the block valve 3s verified in the closed position with p closed position.~~
~~3 es~~
~~This Technical Specification and surveillance requirem~~
. Instructions are developed for the use of this system as committed to in th 2.-. ent will.not be implemented until sequoyah Specifico P* 'NUREG-0737. a ${ e TVA response to Supplement 1 of g50 At least one channel.shall be the acoustic monitors 50' q "( {-I Q jk , :$.e., - l, .g a. .~ -~. - - ~ ~ - -
' n; s '.pgf d ' f.Shaf fl TABLE 3.3-10 'Sok/ylEd.4 M Ab. : g ACCIDENT MONITORING INSTRUMENTATION Et's)~ f fr8( 9 O -<E MINIMUM'., . INSTRUMENT REQUIRED NO. CHANNELS - C OF CHANNELS }
~~1. Reactor Coolant,THot ( de Range).~~
~~OPERABLE 2~~
~~2. Reactor,CoolantJCold(Wdo, Range) to 1~~
2
~~3. Containment. Pressure (Wide Range) 1 2~~
~~~4. Refpp;1ing.WateryStorage Tank Level li-1 2 5..ReactorCool,an(Pressure (WideRange) 1 2~~
~~6. Press,urjzef L, eve (( Wide Range),~~
~~2 1 lI~~
~~7. Steajm, Line,, Press,ure 1~~
~~I 2/ steam line~~
~~8. Steam,Ge,ne,ra, tor Level - (Wide Range) 1/ steam line 1/ steam generator~~
)
~~9. SteamaGenerator Level,(Narrow Range) 1/ steam generator.-~~
~~1/ steam generator T~~
~~10. A,uxiliary 'eedwater Fled R*6'te 1/ steam generator g~~
~~( F u, 1/ pump~~
~~11. Reactor Coolant System Subcooling Margin Monitor 1/ pump~~
~~[ .s 1 O ~~~
12. Pressurizer PORY Position Indicator *
~~~ 0 3 p ....t.. 2/ valve # ~ 1/ valve e~~
~~13. Pressurizer PORV Block Valve Position Indicator **2/ valve O~~
g
~~14. Safety Valve. Position Indicator 1/ valve~~
~~. o.. - 2/ valve # 1' j~~
~~15. Cohtainmenk. Water Level (Wide Range) 1/ valve b'~~
~~16. In' Core ThErmocouples.~~
~~~ 2 1~~
17. Reactor'V 4/ core quadrant 2/ core quadrant g kw es'Eiel l'evel Instrumentation System ***
~~\\ 2 .r u u. m., m. w., w.a z 1 w -Al lR m s [ y@ w, @'E _ - z-~~
Not applicable if the associated block valve is in the closed position l
This' Technical SpecJfication and survefilance requirement will 3l Instructions are developed for the use of this system as committed to in the TVA response to S i
~~$.E NUREG-0737. ,At least'one channel shall be the acoustic monitors upplement 1 of ' R um s".* f. 7 6.. ss~~
~~'iO 't. d-j t s~~
~~9' iJ.~~
~~3. e -~~
~~5 'N,^ #.. st =,.~~
~~7 _'~~~
E- ')".h; -k, I - d, ~~~ 1 QTo a ~ ~ ~' ~ ^ " '
6, NEF t.1 3Lu1 21 [g3, .saf t,. Evalael.., n ' Ecu ttegd. -M _ TABLE 4.3-7 2 o ACCIDENT HONITORING'INSTRUt1ENTATION SURVEILLAt1CE E ~ 5 ~ EISTRUMENT CllANNEL CHANNEL CllECK .E. 1. Reactor Coolant THot (Wide Range) _ CALIBRATION M 2. Reactor Coolant T 4 R 4 Cold (W.ide Range) w H 3. Conta,inment, Pressure (Wide Range) R M 4 Refueling Water Storage Tank Level R M lg$v 5. Reactor Coolant. Pressure (Wide Range) R M 6. Pressurizer Level' R M ja$oi l 7. Steam.Line Pressur.e. ~ R M 8. Steam Generator level - Wide R i R 9. M Steam Ge,nerator Level - Narrow R a i M y,
~~10. Auxiliary Feedwater Flowra' te' R~~
~~2 M 2~~
~~11. -Reactor Coolant System Subcooling Margin Monitor j~~
~~R o i j~~
~~12. Pressurizer-PORV Position Indicator' R~~
~~n e-M 'R { o M 13.: Pressurizer PORV Block Valve Position Indicator M~~
14. Safetg Valve Position Indicator 9-R M

15. Contairinient Water Level (Wide Range)
~~R i M~~
~~16. In Core Thermocouples R~~
~~1 M [R50 I i~~
17. Reactor Vessel' Level Instrumentation **
~~0p R M A 18.r.or -r-f.f&c./,. h-st.ge slador kfromnlafue. m ~ R ~ 7 }. r4 'lR50 ~ g~~
~~This Teghqicdj :.' Specification ard surveillance requirement will not be im y g.~~
Instructions.are developed for the use of this system as committed to in th NUREG-0737., pecific a e TVA response to supplement I of g39 c. Dm. .g { 5," y ~ ~~ _,..,,w, ..y 5! ~ =~ - - ~ ~
, - - = ~ ~ ,:Jp
~~-; a 1~~
~~~ J ER (.(:;J ( 1 03 '.s. G-y s,.f.,a 2 f~~
~~TABLE 4.3-7 A)o. ECA) U/fG IN g~~
SE ACCIDENT MONITORING INSTRUMENTATION' SURVEILLANCE REQUIREMENTS. 2-INSTRUMENT CHANNEL. ' CHANNEL CHECK ~ CALIBRATION E 1. R,eactor Coolant THot ( ide Range) M [ R 2. Reactor Coolant. TCold (Wide Range) H R 3. Containment Pressura (Wide Range) M
~~4..~~
R Refueling Water, Storage Tank Level lR38 ~ .s, M Re' acto,r Coolant Pressure (Wide Range) R. S. 1 M R 6. 'Pressuriz.er Level ht38 4 M 7. SteamEine' Pressure R 1 M R Steam'Ge'neraldr Level -'(Wide) 8. M R s 9. Stiam ' Gen' erat'or L6 vel "(itarrow) 4 M R h
~~10. AuEijiary Fee'dsater Flowrate e.o e~~
~~M R 'r-O E i~~
~~11. ' Reactor Coola'nt System'Subcooling Margin Monitor 3R O~~
~~n M~~
~~12. 'Pressufi'zer PORV Position Indicator g~~
~~o M~~
R'

13. Pressurizsr PORY Block valve Position Indicator M

14. Safety-Valve Position Indicator R
~~l ~ M R~~
~~15. Cqntainment Water Level (Wide Range)~~
~~M R 162 In' Core Thermocouples h38 M _ 17'_Rea R strumentation System~~
~~M R~~
~~~ :Anfe~.<aVn4. An y<. AJ]e.r-1.sf,-o~e fat? ~ H ~ g? j{ E}~~
~~ThisTechnica1'SpecificationandsurveillancerequirementwillnotbeimplementeduttiS n~~
~~Instructions are developed for the use of this~ system as committed to in the TVA response to 8 *g. NUREG-0737.' equoyah Specific-n g33 i of .f-l i.. f-- L Q [j ~ I~~
~~n..-~~
~~h A)U d.4 3 tee.f 34 fQ3 .s g s.n.1,:, AB. EcA1 t,ittg g _ TABLE 2.2-1 (Continued) c8 Y REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS x FUNCTIONAL UNIT C TRIP SETPOINT h~~
~~13. Steam Generator-Water' Att0WA8tE VALUES level--Low-Low.,~~
~~> 18% of narrow range instrument 1 17% of narrow range instrument <20 y span each steam generator~~
~~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 Water level RATED THERMAL POWER coincident RATED THERMAL POWER coincident with steam generator water level <C- > 25% of narrow range instru-with steam generator water level ment span-each steam generator 124.0% of narrow range instro-
~~15. Undervoltage-Reactor ment span-each steam generator~~
~~-> 5022 volts ecch bus Coolant Pumps,... > 4739 volts each bus h.8? ~ ~ l 7 16; Underfrequency-Reactor > 56.0 liz each bus e .Coolang, Pumps. > 55.9 Hz each bus~~
~~17. Turbine Trfp' i~~
~~. A.,. Low Jrip System . Pressure > 45 psig ..o > 43 vsig B. ; Turbine-Step' Valve 91,osurg,, > 1% open > 1% open IS g ! O[1/~~
~~18. -Safety Injection Input~~
~~~ Not Apy.Licable y'? from ESF,,,~~
~~Applicable~~
~~~ f x to -sg o,~~
~~19. Intermediate Range: Neutron~~
~~-10 k d x ro -s 'l / - i 1. 10 ey s-- ?" 11 i,'$ Flux - (P-6) Enable Block s S -p Source Range Reactor Trip~~
^*M

""#N 3
J Y:r 20, Power' Range Neutron, Flux =~ n, 0,o[ (not. P-10) Input to Low Power THERMAL POWER < 11% of RATED g < 10% of RATED Reactor Trips Block P-7 THERMAL POWER z 'ge 9 p"s Q 3g. h ~ E = = e t a: . n'
4.
~~.s Gk/ }$fR3 AJU t./ SafsG Gala~~
~~44 -~~
m JJo. fo) c D8C TABLE 2.2-1 (Continued) x>8 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS s = FUNCTIONAL UNIT TRIP SETFOINT E 13. Steam Generator Water Att0WA8tE VAIUES Level--Low-tow > 18% of narrow range instrument span each steam generator 1 17% of narrow range instrument n I s.
~~14. Steam /Feedwater Flow span each steam generator Mismatch and, Low Steam~~
< 40% of full steam flow at i Generator Water. Level NATED THERMAL POWER coincident< 42.5% of full steam flow at with steam generator water level RATED THERMAL POWER coincident > 25% of narrow range instru-with steam generator water level ment span-each steam generator 124% of narrow range instru-
~~15. Undervoltage-Reactor ment span-each steam generator 1~~
~~~> 5022 volts each bus Coolant Pumps > 4739 volts each bus~~
~~16. Underfrequency-Reactor~~
~~,? Coolant Pdingis' ~ > 56 Hz each bus a ao > 55.9 Hz each bus~~
~~17. Turbine Trip 8~~
~~A. ^ tow Trip System Pressure > 45 psig > 43 psig. B. TurbineStop Valve ,Closb[e > 1% open ~ > 1% open~~
~~18. Safety Injection Input froni.ESF -~~
~~Not Applicable go.iS b Not Applicable [ >t t x so_ s.4~~
~~19. Intermediate Range Neutron~~
~~?_ c p: so ,4) p Flux, P-6. Enable'Blocit 1 1 - In 1o g ,3-Source Range Reactor Trio ano me= % Eben.~~
~~5 2 ;; nn L,~~
~~og hE ggo %, g ~~~
l
~~20. Power Range Neutron Flux (not P-10) Input to Low~~
~~< 10% of RATED en < 11% of RhTCD THERMAL POWERE ~ Power Reactor Trips THERMAL POWER r' u ".g Blocit' P-7 6~~
r
-;:.~ U h . 4 - g a S g .C C 'I g -e => e - ;'*s
UNREVIEWED SAFETY QUESTION DETERMINATION Sktd 3 6 l TVA 10%) (EN 0E$ 2 81) ECN NO. U= 58t-R2 TO. Secuovch Nocicer Plant. Doisv TN 66 / F6" O ROM: F16 SNR 83 or notNuriEn m,,w3,oy no, &Tg' PREPARED R T REVIEWED APPROVED
~~SQP '85 0117 503 R~~
o, E. /24M4 % AA tML,la., M,tQt/ a 1 R h 2 L1 Y S 3 a T Ae rya '.n iNm AL 1 PROJECT-S S N AFFECT' JNIT(S)_ld 'l NO x ECH NO. I bib ECN DATE A A l'" b , s n,m nf YES/NO SHEET NO. heeftfDCR NO. -I 15 G DATE-UlREMENT(5' >J o s OTHER-DAT E _ POTENT CHANGE [f3 O Y[ 'tEFE RENCESI - UM N io n r-Y-bbI* nre -s Cnn#[rsaNeA I ulid derr 8Hadd3r. SGEP e lee ( i I #A 0 - V\\ N,# A 'A
~~#N Pf~~
[ DESCRIPTION OF CHANGL _ c ukr-IY) r' ndl ' ^ h $a & n tn ele Y h e sn ur re. c a rw e_ ife Ne chinn N.Ya en u ~N crom is na me ;LJ e,4L,,~ A,.. v L'M < &- 685L btllCC l. b t',5 Me I t'a o i NC Farm YD M E mA reLUe cex+neeWent -onb n.[inMbh+ L. 48 / A 5 ..v; -,+ m, n ' e., er, e,20 v eiAu,& * % min a X b _ahoreximnfel l~a m n h e'nvir enme nf llu %umi,r,w m.#- aem w. The co vioEen+_Ml l als0 a,,9 ' 7 A o ..o .x. .a 5 zg. ]. ~ g CC (ATTACHMENT $b NO YC$ .e... ; CHIEF. ARCHITECTUR AL DESIGN BR ANCH.W4Cif 6 C K CHIEF NUCLEAR ENQlNEER.W10Cl26 C s.^**l
~~2 l-CHIEF. CIVIL ENGINEERING BR ANCH.W90224 C.K CHtEF. MECHANICAL EN DES BRANCH.102 TK~~
~ CHIEF. CIVIL EN DE$ BR ANCH.W3 Cit 6 C K CHIEF. QUALITY ASSUR ANCE BR ANCH.WllCit6 C K I.Y CHIEF. ELECTRICAL ENGINEERING SR ANCH.WOCII6 C K MANAGER Of* CONSTRUCTION. E?t24 C K CHIEF. ELECTRICAL. EN DES OR ANCH.W20224 C K CHIEF COST PLANNING AND CONTROL STAFF.WitC74 C K '). ~. CHIEF MECHANggAL gNggNggglNQ 3 RANCH W100f ts C K PLANT SUPERINTENDENT MEDS. E4837 C K DIRECTOR. NUCLEAR POWER DIVI $lON. 736 ESC 1
IV A Iml A (L N ObS4 80) - sheet %, j.- 1 ~
~~UNREVICW ED SArETY QUESTION DCTERMINATION LG I 8(, n.O C~~
~~projegg gQ ECHNO. Le t A d, as IDENTirlCR Un'eviewed Safetv Oucatient~~
~~34. y 1.15 the probability of occurrence or t' onsecuences of an accident or malfunction of ecusprnent or is $fety previousiy evaluated in the Safety Anats Report inc~~
,ed?.. .......................Yes. NoS aun,siestion, Th e neu&r A;orina sus +e does not oerfa>rm senat. A n.,,L M A o_a +.:m + x. m nW is sM, comeliM1 The O m_ *, bein. oa.ca de d to ensues. n A n~rd tidfer fdf heeidO+ _ '.CoJ A us orovide. .% e. herafors % N saA k n):Jarma$ low reeneel!n + La. condi+/em of d e k rc. ( [ni;d [ I n a+ +A 9.w c."uin.at W h E re / Itdais of +Ae m; t t n e,t d +h e. N ea baLYt S u $ e~eeureenee.. v;el,'er he t m re cla re d - +ka Ne ' +k e. _c enNo u,m'ee s o8 an d ident dl , val o.+ e d i s tA.n SAR ks nn+ el /0 \\ /Zls ( 1 \\ "y% ( -W ~f ( ik
~~2. is the posilbility'voran accidest 67 e6ait ic of a iferent 6I"~~
~~~~ than any evaiusted previovoy in ine s ety any neporicre..d7............ Y No.)<_ Juomemon, The ncw eiCiement~~
~~Il'~ k se'Amic~~
.nh ~ e _Cl a u 16. TLec#er9 f +/ 6 1 11 seced + __. n.1 dt ' re me nYs and Ne ce$iki k C., as $s! tn:1 se-s ee nf Y d1 Week & S e is %.+ crea+ i uneti ma 9-ddifi.e 4~. ovalva+/ow will-M( 3 ode q ensure + A n t n e. a dd i frond Tacl ~ o', @ cff'lE %aweA yusbe will es hase an+ odverre a#fe E ~ o ~ \\u\\ g V
~~3. Is the*rnargin of Safety as defined in the basis for any technital Specification r ed uc ed ?.............................................. i........ Y Thi s mad 10 e n. F. ipa se e ve r~~
+o im a ro v e. +Le aee;de -- _mo ni to r i n. en om bi li+w oF ue o f nf. Thu s. +Ae 3/4,3, 3, f pp e,;g,,+ c.;~.# afeA,S de[isd ino 4,,g 0 , c,, c ".qn.g. Un, tm.rnefe aAed a+ ton.)'y nb eedveed.~ Pe e +ke n o+a. .(V % : l thh_Q4r_4 +na.-en o , 0 a. +te, +, e t_, pet _,osU_he;..c,y.iya +o en elm d e os R :;q fi:'" i P s 'I
!VA 10bbt.O (EN 1305 7 80) . Sheet # UNREVIEWED SAFETY QUESTION DETERMINATION \\ lbl0S R* O ECNNOAsess es i (, Project SAN $5 _OF ' DENT 'r' E n L Special equirement(s) or Precaution (s).. ( b, Marks Subject to Which Safety Eva at i o n............ ".'........ Preparer - 84 6 Additional in emation......... Reviewer-8% (initial) = j _ '.. ,,1._pt en f io I 1.h Scie K#Fva A e \\ AT r _ tslUC PR showld \\evietJ aSo IIM+ec h.roces accide nf on _m onif er inn i m+r umh f ate's +[ eledenst%e *# .a n w chanses R ccb a. \\ M U
~~g.... _. _..... c/ )~~
p,,r _ \\- - / // ., x............r..; j y Y \\ V/ i / \\ \\ /m A f I \\ 'IA L f, A ' /Y. ( fjf , A. 7 A f[ y \\ a j f. ff.. f g._ _ g. _ _ __. .y 7.ex. y f .x 3.g. ..f .y. r;g/ 4 ..,.\\ ... _.. g ..... f.... g 7._ ,f -y y \\ \\ \\ \\ < t.'... :.s ' f:
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~~g y y g vs t va rise)~~
TABLE 3.3-9 m 'S a - 1-REMOTE SifUTDOWN MONITORING INSTRUMENTATION .Si gp [ .3 - v-c- E' 3a INSTRUMENT READOUT MINIMUM MEASUREMENT CHANNELS ~~ LOCATION -4 fuTEC /fs oATE f T H 1. Source Rang ~ RANGE OPERABLE lear Flux; !;0TE 1 1 td 1 x 10 j 3 I J 6 y 2. Reactor Trip Breake ndication } to' vs zoo % CTP at trip switchgear 'OPEN-CLOSE ~. - 1/ trip breaker 3. Reactor Coolant Temperatu i NOTE 1 0-650*F Hot leg 1/ loop lR83 4. Pressurize'r Pressure t 'OTE 1 0-3000 psig 1 y 5. Pressurizer level 3 NOT 0-10.0% 1 lR80 y 6. . Steam Generator Pressure.i NOTE 1 5-0-1200 psig 1/ steam generator 7. ' Steam Generator Level 1 NOTE 2 or ~ l near Auxilary F. W. 100% i Pump 1/steas generator 8. Full Length Control' Rod Position 'Auxilary Instrument Limit S. witches ~ Room: Racks R41-44 On off switch / rod 1 insertion limit 9. RHR Flow Rate: I . NOTE 1 0 1500 gpm 1 10. RHR Temperature NOTE 1 50-400*F 1 11. ,a Auxiliary Feedwater. Flow Rate NOTE 1 0-440 gpm 1/ steam nerator Eso xg 7
?,
u 2 % *z y ' ;& g. .1' t. p& < = - y jf; i 3 r- .rr .i ia; \\ .~.jk i o .:E T32 M
TNis O MA M IS h4 A* f*'# ~,. E ' ^ 'Margepaav AEdt TM OrJcuiJ# k '.5l...' '*4 0 - (' i,., j$ M $* *I E V^ ** T'*" '.'~ /v $ircreoo_ c7,7 OF 7N~ "* EC" (.C ( 6E m TABLE 3.3-9 $4 fg7f E vat.v A *7/**A E REf10TE SHUTDOWN MONITORING INSTRUMENTATION I(i> )> I E i fad), t e b IfiSTRUffEr READOUT ffEASUREMENT cA 5 te eraMde. L CATION RANGE w OPERABLE ~
~~1.,Sourc6 N~~
an Nuclear Flux NOTE 1 ~ 6 1 to 1 x 10 cp 2. ReactoE Trip Br er Indication at trip switchgear OPEN-CLOSE 1/trpbrEaker 3. Reactcr Coolant Tempe ture - NOTE 1 Ho,t 8.eg 0-650'F 1/ loop R67 4 Pressurizer Pressure N01E 1 0-3000 psig 1 g - 5. Pressurizer Level .E1 0-100% u 1 t y 6. Steam Generator Pressure 'R67 NOTE 1 m 0-1200 psig 7. Steam Generator Level 1/ steam generator ? NOTE 2 or near Auxilary. W. D-100% Pump 8. Full Length Control Rod
~~1/ steam generator Position Timit Switches '~~
~~Auxilary Instrument 1 insertion limit i 1 '- Room: Racks R41-44 On-of switch / rod 9. RHR' Flow Rate NOTE 1 0-4500 gpm 1~~
10. RHR Temperature FOTE I 50-400*F 1

11. Auxiliary eedwater Flow Rate NOTE 1 E.2 0-440 gpm 1/ steam generator R67 "a
~,,, C2 ~ ~ ? h g 2 E. 69 e ~ a r ~ ~. 'h 1 ,. m. g m
3 gj S.:-?.'. ki.). '/N'S CL( w C a* /5 A/# /ow(ed.' k s o m e w w,yl l h NEP 4 6 Suest g) 9.g.%), ' ~'Q).N *i' }pj..krue-asewsrus n se-er~ 27 ? g_ -Or
~~yye-$are rf EVAwa ~r'*^'~~
.. por Ecu g.c gsc f TABLE 4.3-6 ? e,,y.d' '3 't -< y-[ REMOTE SHUTDOWN HONITORING INSTRt#fENTATION t SURVEILLANCE REQUIREHENTS INSTRUMENT CHANNEL CHANNEL _E y,il. CHECK CALIBRATION /p.y cc. ns ca q - -* e'l Source Rang qu ear Flux -g H g - 2. Reactor Trip Breaker dication g N.A. R"eactor Coolant Temperature 3. Hot leg g g P'ressurizei Pressure 4. ~ g g 5. Pressurizer Level' M R 6. Steam Generator Pressure R M R 7. Steam Generator' Level y M R S 8. Full' Length Control Rod' Position Liaiit Switches M R ~ 9. RHR Flow Rate M
~~10. RHR Temper'atu.' e, R~~
~~r y~~
~~11. Auxilidrv Feedwater Flow Rate M~~
R
~~12. Pressurizer Relief Tank Pressure M~~
gg
~~13. Conteinment Pressure e~~
~~g g EE %E .u E j .E o z h "e~~
h,
!st u, ) .:3 s. .:.g. e " n
(. ~7HIS CMAM /$ Y O C W' ) {: pa rua oisu s m &.scr'=. 27 7 R %ET C G $Uset (q{R) 3 s s s of 72T $4fr7Y N W # *' e mawaw.. gg7g 4_3_g Ye WW sW y - s,,
~~} - O.~~
REMOTESHUTDOWNHONITORINGINSTRi#fENThTION x SURVElttANCE REQUIREtlENis ~ c i I-2". INSTRUHE l' . CilANNEL CilAl#4EL i i~ tvienueorste _C11ECK _ CALIBRATION 1. Source Ran fluclear Flux M A 2. Reactor Tri er Indication M N.A. 3. Reactor Coolant Tempe ture - Hot leg. H^ R 4. Pressurizer, Pressure M R ~ 5. Pressurizer level ~ i M R c i 1 6. Steam Generator Pressure i y M A , 7. Steam Genergtor Level H R 8. Full Length Control Rod Position Limit Switches M R* R20 9. RHR Flow Rate. 1 R 10. RHR. Temperature ~ M R 11. Auxiliary Feedwater Flow Rati i 3 j M ~R g g 1;;- 12. Pressurizer Relief Tank Pressure ^ M ~ 5 s - l,. y?
~~13. Containment Pressure -~~
~~4 l M R p9 o ~ e 4 r o g g e.p l ~~~
~~For cycle 1, this surveillance is to be completed before the next cooldown w-l or by A6 gust 5,1983 whicheyer is earlier.,~~
~~g \\ T'- g .m ) \\~~
1
)
~~3., \\~~
~~~ ,,n.~~
'f
^ ' ._e4as,-e' F
NE P 4.lo, $\\td M S [ ECN NO.L&s 66 e3 SAFETY LIMITS 84d Evabb, J l f No.yE c.g L 4t$(o l BASES N j s Range Channels will initiate a reactor trip at t errent 61:b 5 trk A te I approximately 25 percent of RATED THERMAL POWE s nua y octea who 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), 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 g3 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 indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced 5 } according to the notations in Table 2.2-1. I 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 K1, K2 and K3 inputs to the Overtemperature Delta T channels and raising the P-8 setpoint to its 3 loop value, in this mode of operation, the P-8' inter-lock and trip functions as a High Neutron Flux trip at the reduced power level. Overpower Delta T The Overpower Delta T reactor trip provides assurance of fuel integrity, e.g., no melting, under all possible overpower conditions, limits the required range for Overtemperature Delta T protection, and provides a backup to 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 I
E CN N O.L u ta R.s N E P 66. SQ g ' LIMITING SAFETY SYSTEM SETTINGS 9I OF N O Ev6sh'm 'J W y th.E.c.W LQ% BASES 9 Intermediate and Source Range. Nuclear Flux (Continued) ~ ~ ' Range Channels will initiate a reactor trip at approximately 25 percent of RATED THERMAL POWERTnless manua P-10 becomes active. ~ ated with either the Intermediate or Source Range Channels in the accid 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 AT The Overtemperature delta T trip pr for all combinations of pressure, power,ovides core protection to prevent DNB coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping and pressure is within the range between the High and Low Pre 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. D 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 (g s nuclear detectors, the reactor frip is automatically reduced according to the notations in Table 2.2-1. setpoint does not require reactor protection system set point mod because the P-8 setpoint and associated trip will prevent DN8 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 interlock and trip functions as a High Neutron Flux trip at the reduced power level. Overpower AT The Overpower delta T reactor trip provides assurance of fuel integrity, e.g., no melting, under all possible overpower conditions, limits the ' required range for Overtemperature delta T protection, and provides a backup to the High. Neutron Flux trip. The setpoint includes corrections for changes in density and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop temperature detectors. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System. f# 2 SEQUOYAH - UNIT 2 8 2-4 4R 4 .,,,n. a -fj ,,.. m.. ilc., .v,.. w.. x. m;. n.tcW3 71 in.? .'*.I. 'i, 7,*p, 8 t[ ~
u a num D gw "M ,Mp..,' ~ n. "*
~~Sa A-h G ve faa -h,N L gy (,,gg REVISION LO a e visio,.~~
~~l R: N o. DESCRIPTION OF REVISION ENOM' M ay'~~
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In Nid $ fss o e_. Ma op ~ l /fev* Y & L' OJ;Q6 ( / erns sop eron7r4 -t...,a. w 4. nw ?c u s a.c n a h M S p ; n /-s. a no.f 4tk -L I,a n ly m sluJ.s 2:, : r, l: -g ~3 4, 3 7, (38 z-In a... p. e s 4-p /* n i li<.E-sin) <-.,-a., s, .f.,, ei A t.-.-./,/ubs.L -4 h a sak G ev./9, L.../ - / go,.es/a k y s><.2 }* K LT t Am,s lo I%fltme.nfI J h ts o f f c n t. //GS. y b "*" J bk O ad /*/, e ) pxvf $4, 34Y < ve /v a fu n.s h aja < ee, d a /y, y %ggf gf h k W N t.owor}t soocc e r e e4 g od uts dl sable) change Aad 3 p ve rbos c3 acibbs.-4 edd<d stwial nguiremd ..a mk seu. suu h t, Tw mA.,A 'c - 3 ,1,,g g g elaJrscM kon) c h ANgts WkM Of-50!k4 2 ('el its & e.$k q,u,h b Fr A R c.haph e T, N ets g,,, nakd to i ic.t v4 %t FSAR c.h g tyir' e d. o i-4 7 e' 5^ ion. ic~ oc..,,, -y a n. eg-p. = % #4L. 4M8 h ___.___.__.__.._____._________._______m.
w_ e - @ a.w es soyY E v a boh.',, E th1 L 41 M EcNto. w a m h et 45 SON 4 ~ 9 5..c;: l As discussed in Subparagraph 7.2.1.1.2,. factors included in establishing ' { the Overtemperature aT and Overpower ai trip setpoints includes the reactor coolant temperature in each loop and the axial distribution of core power through the use of the two section ex-core neutron detectors, 4.4.13_ instrumentation to limit Maximum Power Outout The w+e5 odi pt of the three ranges (source, intermediate, and power) +S-d: :: tor:, x!!' :5; :!::tr:r t:; of We-nuclear instruments, are used to limit the maximum power output of the reactor within their respective ranges. There are six1 radial locations. r-id aia; a :;t31 cT e.m .wNn fl u [cetectors installeo arouna the reactor. in theprimary shielQ two Cul.w c.hak assed pro;;r: ion:,1 : n at:r: for the source /6Eg"e"Mstalled on opposite " flat" portions of the core containing the primary startup sources.H-en M"it':- 2:: cr' :! 'r : : ;uirt c' +5
~~ore ' !;5t~~
~~%c ::*:earited !:nt:: +!-~~
~~r::::r; for 5: '-t:rm;d!;t: r;ng;, lo::::d '- t5: ::::~~
~~i M --r-',ma^:+:nd 1:':: ::.;;:::1!:: ;; t': :=r:: r:n;- d:::: tor, are me i + 1 s sm 1-m,+im. ,m..-,-- at-- am --- k t e or~~
wm
,m,a de M 5 [fiouI o u5iie[tTbNniIliihhs$5Ii6'id515r)NN$ sh mbilN, n r the power range Installed vertically at the four corners of the core and located equidistant from the reactor vessel at all points and, to minlml:e neutron flux pattern distortions, within one foot of the reactor vessel. Each power range detector provider two signals corresponding to the neutron flux in the upper fn +5p iower sections of a core quadrant. The three ranges o ?!f:Y3are used as inputs to monitor ( neutron flux from a completely s utcown condition to 120 percent of full power with the capability of recording overpower excursions up to 200 percent of full power. The difference in neutron flux between the upper and lower sections of the power range detectors are used to limit the Overtemperature aT and Overcower AT 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: 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 consequences of rod ejection' accidents, and 3 protecting the core against the consequences of adverse power distributions resulting from' dropped rods. Details of the neutron detectors and nuclear instrumentation design and the control and trip logic are given in Chapter 7. The limits on neutron flux operation and trip setpoints are given in the SQN Technical Specif t-cations. 4.4-34 0048F/COC4 )
Dq
=
t hMy EvabtA E (- N av6 Shed N4 ECN No. Lce nt g3 l 'e i $%6 94 _OF ~ 4. 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 bistables each with their own trip setting (a high and a. low setting) per channel (four channels 6 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 approximately 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. b. Intermediate range high neutron flux trip 5 The intermediate range high neutron flux trip circuit trips the reactor when one out of the two intermediate range channels l 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 i this value automatically reinstates the intermediate range high neutron flux trip. The intermediate range channels (including detectors) are separate from the power range channels. The intermediate range channels can be individually bypassed at the t l nuclear instrumentation racks to permit channel testing at any time under prescribed administrative procedures and only under l the direction of authorized supervision. This bypass action is My annunciated on the control board. IEU 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 the trip setpoint. This trip, which provides protection during reactor startup and plant shutdown, can be manually bypassed when one of - % _the twn p+ arm M ate range channels reads above the P-6 setpoint value (source power level) and is automatically 0jP4pg e ated wh n ermediate range channels decrease below ul Mit*thah the P-6 value. This trip is also automatically bypassed by two l r u y on semk - out of four logic from the power range permissive (P-10). This trip function can also be 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 is set between the P-6 setpoint and the maximum source range level. The chanaels can be individually blocked at the nuclear instrumentation racks to permit channel testing at any 7.2-3 0068F/COC4 ~ w., o _m,- r .,--m
5 04* EMIvebb Etgt(,)% %,g dd qq ECNNO.Lkeak am SON-6 9 5 op 'w The auto stop oil pressure signal also dumps the stop emergency trip fluid, closing all of the turbine steam stop valves. When all stop f 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. J 7. Safety injection Signal Actuation Trip 7 A reactor trip occurs when the Safety Injection System is actuated. ki 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 -t functional description of the process equipment associated with this 4-trip function is provided in Reference 1. 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 1 other output actuates the train B trip breaker. Operating a manual trip switch removes the voltage from the undervoltage trip coil and energtzes the shunt reactor trip breaker trip coll. 6*, 4 There are no interlocks which can block this trip. Figure 7.2.1-1, Sheet 3 shows the manual trip logic. k
~~7. 2.1.1. 3 Reactor Trto System Interlocks 1.~~
Power Escalation Permissives l5 The overpower protection provided by the out of core nuclear 4 instrumentation consists of three discrete, but overlapping, levels, i Continuation of startup operation or power increase requires a ^: permissive signal from the higher range instrumentation channels -.4 before the lower range level trips can be manually blocked by the ., / operator. 4 A one out of two intermediate range permissive signal -) is 6 h Is-M e *A S ad a_ t r ve: 2_ restored when both intermediate range ern 1 e t ca I
~~gfuys / %=inirs.m ur rv.~~
e perm ssive (P-6) level. There is a manual reset switch for adminh_tritively reactivating the source range level 2 trip and(;:M:tr' i ::'M P-101evei ir requrff. *WI"* hen between the permissive P-6 and ~ Source range level trip block and b' ds lh AW o ~-.
~~r' r m" are always maintained when above the permissive~~
' eve. + 'l J 7.2-11 0068F/COC4 / a A O
5*4t E volvat.4, h g g nep Cr^^ SQN E % W as an Wa 9% l N OL 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 The process analog system is described in Reference 1. 7.2.1.1.7 Solid State Loaic Protection System The solid state logic protection system takes binary inputs (voltage /no voltage) from the process and nuclear instrument channels correspending 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 actuation functions. In addition the system includes means for semi-automatic testing of the logic circuits. A detailed description of this system is given in Reference 3. 7.2.1.1.8 Isolation Amplifiers In certain applications, Westinghou te considers it advantageous to employ control signals derived from indivicual protection channels through isolation amplifiers contained in the prctcction channel, as permitted by IEEE-279. In all of these cases, analog signals derived from protection channels for non-protective functions are obtained through isolation amplifters located in the analog protection racks. By definition, non-protective - functions include those signals used for control, remote process indication, and computer monitoring. Isolation amplifier qualification tests are described in References 4 y Sul n 7.2.1.1.9 Enerav 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. 7.2-13 0068F/COC4 ---~-----A---____ m-m-___
hM Eva\\ud[.w ELN Ljl% i-y h SQN- -( & Se,d yrgn.p.u) ,e partial trip alarm and channel status light actuation in the control C room. Each channel contains those switches, test points, etc. necessary to test the channel. See Reference 1 for additional information. The power range channels of the Nuclear Instrumentation System are 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 b1 stables 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 function 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 initiate the " CHANNEL. TEST" annunciator in the control room. Blstable operation is tested by increasing the test signal level up to its trip setpoint and verifying bistable relay operation by control board annunciator and trip status lights. 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 level 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 detallt escription of the Nuclear Instrumentation System see Reference g The logic s of the Reactor Trip System are designed to be capable of complete testing at power, except for those trips listed 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 man 9al trip could also be initiated from outside the control room by manually tripping one of the reactor L 7.2-23 0068F/C0C4
SNd f.volv:S EC.N LMD '~r.
~~c 7~~
~~N "'u aj SMt So L A h (Wikig '), g,4 ) SQN-6 i~~
~~11. The Institute of Electrical and Electronic Engineers, Inc., "!EEE~~
~~) 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.

13. The Institute of Electrical and Electronic Engineers, Inc., "!EEE Trial-Use Guide for Seismic Qualtftcation of Class ! Electric Class !
~~Electric Equipment for Nuclear Power Generating Stations," IEEE Std. 344-1971. i~~
14. " General Design Criteria for Nuclear Power Plants," Appendix A to ittie 10 CFR 50, July 7, 1971.

15. E. P. Rahe, " Evaluation of Surveillance Frequencies and Out' of 4
~~Service Times for Reactor Protection System," WCAP 10271 anh Supplement 1 (Westinghouse NES Proprietary).~~
~~16. W. N. Moomau, " Westinghouse Setpoint Methodology for Protection 6~~
Systems Sequoyah Units 1 and 2." WCAP 11239 Rev. 3, October 1987 (Westinghouse Proprietary Class 2). F1, 7 V A hviroNMdA k Qonl'it di Au B'.nder 59 N EQ-N M ool, c, . mws, thh n o x %bU. S A e. i e D. 4 7.2-37 0068F/COC4 e. .m..
jg SQM haleds ECA L6tt4 y km 1.3,l A.3) shur 51 W SQN-6 Electrical Penetration Assembites wnwv.4 Thegelectrical penetration assemblies have been tested to TVA specification requirements which conform to IEEE-317, 1971, "!EEE '\\ Standard for Electrical Penetration Assemblies in Containment Structures ca.mx t,\\t deM) g.ictr 4
~~y prs't,@'m for Nuclear Power Generating tations." Tbt Nut the. J'13 vtesim o EEe.- m.~~
= e locumentatIon of J reports of all tests required and listed in the specifications and quality assurance appendix, and appitcable 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 Bot.ler and Pressure Vessel Code. Section III. The dose rate at which TVA has conducted 100 hour tests on materials and equipment 1s 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 tena to produce significant increases in degradation in the region between 10' and 10' Rad /hr. However, radiation-induced l6 oxidttlon 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 Exposed to Neutron and Gamma Radiation," recommend using exposure rates above 10' Rad /hr. It is the TVA position that 10' Rad /hr for 100 hours b represents a reasonable and conservative combination of dose rate and exposure time for radiation testing. Cable terminations to low voltage power, control, and Indication penetration assemblies are generally made in all metal spilte boxes. However, in a number of instances on the outboard side of containment electrical penetrations, fleid cables were spliced to the penetration pigtalls in cable trays. In these cases, a special enclosure was used to act as a qualtfled fire stop (refer to Figure 8.3.1-37. -38, and -39). l6 These particular sp1tces are located within the last 5-foot section of the cable tray. The trays in the annulus area of containment containing these splices are fitted with solid top and bottom covers in the immediate area of these spitees. A qualifted fire barrier made of -silicone foam and'ceraform/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 splice 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 e',ectrical penetration were covered with Flamemastic material. This configuratloa constitutes a qua11 fled 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
,.5 E - ' u sdt, Evaluat'..N Min Eolmo.Ls.ias," %. j ( id sedda 15.p '4.Q 5 hea.+ 5 A en ' ; u dii ot_ operation are considered in this analysis. Table 15.2-1 contains the time sequence of everts for this accident. Dilution Durino Re'velino An uncontrolled baron dilution accident cannot occu RCS from the pot ntial source of unborated water, i Various valve c Anbinations' that are required to be locked closed durin , refueling opert.ttons are specified in technical specification 3.9.1 These valves will block the flow paths which could allow unborated make to reach'the 4CS. Any makeup which is required during refueling will be borated water supplied either f rom the refueling water storage tank by the low hear' safety injection pumps or the centrifugal charging pumps, or from the bc/ic acid tanks via a boric acid transfer pump and a s centrifugal charging pump. 6 ) Dilution Durino Startup 1 Prior to startup the RCS is filled with borated (approximately 2000 ppm) water from the refueling water storage tank. Huclear instrumentation is monitored closely in anticipation of an 6 l 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 f In the analysis, a maximum dilution flow of 300 gpm limited by the capacity of the two primary water makeup pumps is considered. of the reactor coolant is approximately 9967 ft*, which is the active The volume volume of the Reactor Coolant System excluding the pressurizer, i i Dilution Following Reactor Shutdown v lowing reactor shutdown, when in hot standby, hot shutdown, and ( subsequent cold thutdown condition, and once below the P-6 Interlock setpoint, la 3::0-e : % rve! tact Init ::t h- $02, " t;h r hu t' S;;;ye t: h :r " hat Unit: i :nd AM 'dp hdj :t - t ^#te- $%t & ", the high flux at shutdown alarm setting will be l !!! trie ib e3the count rate 3^ -'=t:: f.adjustedgo no higher than 3 bu 'ter ;het th:tder.rsJoce.s. /AyowJ aMma&nity 41wwwmed as " ehr ::tpe!-t er:t 5: ::t :: ::rt' ud :::ry 20 ! :t:: ':: th: ' k ; t- !M- # ' Mrtia; pNt t-!;. :> ry ? % r: '~ t% aert 5 Me :, rad I (e : riane. ~ iesti g s.,cytil there:fter urt th: 'h. h :? 5:: :: 541!::d. 9 will ensure that the nlarm setp.;wt is ereceW 11ution at N With the unit at power and the RCS at* pressure, the dilution rate is the charging pump. limited by the capacity of the primary water makeup pu / A conservatively high value for the expected boron rate of 300 gpm was used, concentration (1575 ppm) at power and a conserva i % o{ee c Afor- :l,e 3 ud,4*1 "d. h 'h[4$,"4ro 0
~~d t u b a t., a 4 iw5rt*88 av~~
$152-14 hv 5+9 i 4tswer-h44 v'inu d 't4d h dii+3 4'cw we,ds pte se.c. owl 94tes4., each COC4/0115F aman e\\.a 4t. mai,i c.edc.) t.ack and swne <.Se Arme. xA ~ -~^ .-.i
h[Nti SW6 EydVdEN ECN NO._ L ru a 6 85 E uJ LUM too op 15.2.4.3 Conc'usions 564 53 For dilution during refueling: Dilution during refueling cannot occur due to administrative controls ~ ~~ g 4.,, 4%g g seo m n.oge, M (see Section 15.2.4.2). The operator has prompt and definite indication of any boron dilution g& from the audible count rate instrgntation> @ count rate is 44 seed edutg U..r e414-4Ae. reactor containment and the$ control room. In addition, a high i source range flux level is alarmed in the control room. The count rate increase is proportional to the subtritical multiplication factor. M .'a % re d se For dilution during s at up: ud ald ataa For dilution during startup, there is adequate time (-52 minutes) f r g transient initiation for the operator to recognize the high count rate signal and terminate manually the source of dilution flow. The eperator is alerted to the uncontrolled reactivity insertion during 1 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 M imes 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. l For dilution during full power operation: l 1. With the reattor 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 I assumed that the operator is aigrted 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/Oll5F
i m;m - La a SaId told m _ L tb E n y j Etra L6i16 ECN NO. w a 6 m3 Sqq shad 64 163. OF f TABLE 15.2.4-1 SEQUENCE OF EVENTS tout 11brium Xe Case Time (sec) Reactor Trip 0 Reactor Power. 7.5% of nomi tee reactne neriod r Intermediate NIS reads f emps. % Pewt A 10 l Source Range NIS Available 930 i Source Range NIS no longer decreasing (without dilution event, flux would stabilize at this point - an 18 day half life decay of flux would be normal'). 1,250 d 3 +; m t.s Operete* tet High Flux to Shutdown Alarm 0.5 e :: e h stabilized flux level. For this example, the value is I 700 cps. 1,250 High pressurizer Level Trip and Alarm 1,800 Source Range High Flux at Shutdown Alarm 7,400 i K.,, = 1.0 12,960 ' Source Range Count Ra'te would change from 200 cps to 197 cps k i 4 0726F/COC4
~~2 :n N :O sokyI.vdud,w go 9.w n s E4 LMW E;;Nic.Me18 6 E3 3~~
b SCN-6 Sktd N l l o t.. r TABLE 8.3.1-11 (Sheet 1) 120V AC VITAL INSTRUMENT POWER BOARD 1-1 JBatterv Beare 1) EKR SAFETY CONNECT [D !LO., 1.0A0 RELaTED LOAD-Va 1 $$PS (A) Ch ! Input Relays (Pn1 1-R 48) Yes 1,080 2 SSPS (B) Ch ! Input Relays (Pnl 1-R-49) Yes 60 3 N!$ Instr Power Ch ! (JB 3398) Yes se9-4 4 NIS Control Power Ch ! (Pnl 1-M-13) Yes -44% ggi ) 5 Pf'ocess Protection Set ! (Pnl 1-R-1) Yes ,4 6 UH! Accumulator Ch ! Isolation Valve (Pn1 1-M-23A) Yes 62 7 Cnet. Bldg. Rad. Monitor (1-RE-90-106 & 0-RE-90-133) Yes 949 8 Instrumentation Bus A (Pn10-M-278) Yes 50 9 SSPS Aux Relays (Pn1 1-R-73) Yes 110 10 Reactor Bu11dino Isolation Valve (JB2670) Yes 109 11 Aux ComDressor A Aux Bldg Isol Valve (FCV-32-82) Yes 52 12 Radiation Rate Meters (PNL 0-M-12) Yes 1,406 13 Raciation Monitors (0-RE-90-125) Yes 360 14 Instruments (125V Vital Battery Board !) Yes 20 15 Incore TC Monitoring (PNL I-R-59) Yes -308 16 Chlorine Detector (Pnl 0-L-450) Yes 24 17 PASF Solenoid Valves (PNL l-L-572/C) Yes 619 18 Aux Relays PCO-65-81 & PCO-65-86 Yes 84 19_ Toilet & Locker & Spread Room ! sol Dampers Yes 168 6 20 BOP Process Instr Control Rack Yes 488 24 Aux Bldg Instr Bus A (PNP l-L-26) Yes 150 22 Aux Bldg Stm 1501 Viv FCV-12-82 (Pnl 1-M-9) Yes 62 23 Containment Purge Air Exhaust Rad Monitor Yes 360 24 ReaC. Vessel Hd Vent Throttle Valve FCV-68-397 Yes 60 25 NSSS Aux Relay Rack A Bus (Pnl 1-R-54) Yes 462 26 Sep & Aux Relays (Pnl 1-R-73) Yes 532 27 A Relay Bus (Pnl 1 L-IIA) Yes 266 28 A Instrument Bus (Pnl 1-L-11A) Yes la0 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 Gas Treat Fan A-A Mod Dmpr (0-L-429) Yes 142 3a 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 1 (Pnl 1-M-6) No .80 41 Instrument Bus 1 (Pn1 1-M-4) No 60 0431F/COC4
2
^^ Sa ($y Ecled,J,, gy w9y ECfM O. W.t h E3 l 4 SON-6 gd g4 Qop_ TABLE 8.3.1-11 (Sheet 2) (Continued) 120V AC VITAL INSTRUMENT POWER BOARD 1-! (Battery Board I) BKR SAFETY CONNECTED i ML LQA_Q RELATED LOAD-VA 42 Fire Pump 2A-A Sep Relays No 84 l de P.an Exhault fan 1A] ].Dw fnntr -L- ) 43 ~URI I___

~~wu Lsis (rs.h-toom %.sooi tw vYes N 2 6 2.~~

6 -irf ' w -X 46 PR-30-310 (Installation on Hold) Yes 30 5..th..V.e*al Battery Instrumentation Yes 26 '47 it gg Spare *""*
~~TOTAL 14,100 j~~
.NQ][j 1. Each inverter is capable of supplying 15 kva continuously, g 2. No automatic load stripping or' load sequencing is employed. 3. Power Boards 1-!,1-!!,1-!!!, and 1-IV also supply common plant loads and are more heavily loadad than respective unit 2 boards. 6 9 4 e e 0431F/COC4 4 .,-n,. ,--..-.,,a-.u ,.n...
.i 5dthtval M " {S,g tot tus 6 L M 57 C3fmJat. a3 l SCN-6 _IkN u TABLE 8.3.1-12 (Sheet 1) 120V AC VITAL INSTRUNENT POWER BOARD 1-f! (Batterv Beare !!) BKR SAFETY CONNECTED 10,, LCAO Rtl&TED .LCAD-VA I S$PS (A) Ch !! Relays (Pnl 1-R-46) Yes 600 2 SSPS (B) Ch !! Input Relays (Pal 1-R-51) Yes 1 3 NIS Instr Power Ch !! (JB 3399) Yes ' $40 4 NIS Control Power Ch !! (Pnl 1-M-13) Yes -446 4g1 5 Process Protection Set !! (Pnl 1-R-5) Yes 1,666 6 UN! Accumulator Ch !! ! solation Valve (Pnl 1-M-238) Yes 62 7 ERCW & Containment Rad. Monitor (Pnl 1-RE-90-Il2) Yes 949 8 Instrumentation Bus B (Pnl 0-M-278) Yes 30 9 SSPS Aux Relays (Pn1 1-R-78) Yes 103 10 Reactor Bullding Isolation Valve (FCV-32-102A) Yes 62 11 Aut Compressor B Aux Bldg Isol Valve (FCV-32-BS) Yes 48 12 Radiation Rate Meters (PNL 0-M-12) Yes 1,054 13 Radiation Monttors (0-RE-90-126) Yes 360. 14 Instruments (125V Vital Battery Board !!) Yes 26 15 Incore TC Monitoring (PNL I-R-60) Yes 308 16 Chlorine Detector (Pnl 0-L-451) Yes 24 17 PASF Solenoid Valves (PNL l-M-10) Yes 619 la Aux Relays PCO-65-83 & PCO-65-87 Yes 84 19 follet & Locker & Spread Room Isol Dampers (Pnl 1-R-78) Yes 168 20 BCP Process Instr Control Rack (Pnl 1-R-131) Yes 203 6 21 Aux Bldg Instr Bus B (PNP l-L-26) Yes 150 22 Aux Boller Stm Isol Vlv 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 Hd. Vent Throttle Valve FCV-68-396 Yes 60 25 NSSS Aux Relay Rack B Bus (Pn1 1-R-55) Yes 30B 26 Aux Rly Rack Sep & Aux Relays (Pnl 1-R-78) Yes 546 27 Aux Cent Pn1 B Relay Bus (1 L-IIB) Yes 168 28 Aux Cont Pnl B Instrument Bus (1-L-llB) Yes 184 29 RVLIS (Pn1 1-R-148) Yes 768 30 Aux Dryer Train B Yes 981 31 Aux Rly Rack Sep & Aux Relays (Pnl 1-R-77) Yes 210 32 Borld Acid Tank A Ntr B-B Control (Pnl 1-L-304) Yes 31 33 Aux Bldg Gas Treat Fan B-B Hod Dmpr (0-L-428) Yes 142 34 Beric Acid Tank C Hrt B-B Cont.rol (Pnl 0-L-305) Yes 31 35 Radiation Monitor (0-RE-90-206) Yes 360 36 RCP2 UV & UF Relays Yes 30 37 Process Control Group 2 (Pnl 1-R-17) No 743 38 Instrument Bus 2 (Pnl 0-M-278) No 38 39 Plugmold Instrument Bus 2 (Pnl 1-H-3) No 600 40 Plugmold Instrument Bus 2 (Pnl 1-M-6), No 166 41 Accoustic Flow Monitor (Pnl 1-M-2*7A) No 117 0431F/COC4 I
W g, - r-i p,. + S N tvdo M y ECN NO. L c.% E3 tu tM % 165 oP l 50N-6 h-{. g j TABLE 4.3.1-13 (Sheet 1) 120V AC V!TAL INSTRUMENT POWER B0aRD 1-!!! (Battery Boare !!!) BKR SAFETY C0hNECTED 89., LQAQ RELATED LOAD-vA 1 $$PS (A) Ch !!! Input Atlays (Pn1 1-R-46) Yes 600 2 SSPS (B) Ch !!! Input Relays (Pn1 1-R-49) Yes 600 3 NIS Instr Power Ch !!! (JB 3400) Yes 88 4 NIS Control Power Ch !!! (Pn1 1-M-13) Yes 240 5 Process Protection Set !!! (Pn1 1-R-9) Yes 456 6 UHI Accumuli, tor Ch !!! Isolation Valve (Pn1 1-M-23A) Yes 65 7 RCP 3 UV & UF Relays Yes 30 8 Aux Feed Turb Controller (Pn1 1-L-381) Yes 222 9 Turb Dr Aux FW PMP St Gen (Pn1 1-L-361) Yes 62 10 Tu.rb..Or Aux FW PMP St Gen (Pn1 1-L-11 A) Yes 80 t ij " Spa r e * * * * * *
~~ig neeeee Spg7, u n su 13 '""" Spa re " * ""~~
14 Instrument Bus & Xfer Pwr (Pn1 1-M-3) No 30 l 15 Aux Cont Pn1 A Inst Bus (PNL I-L-11 A) No 475 l 16 Process Cont Group 3 (Pn1 1-R-20) No 282 e 17 90P Process Instr Control Rack (Pn1 1-R-126) No 3.170 ( 18 Control Room Doors Security Lock No 100 b 19 Emergency Cas Treat Filter train A (Pn10-L-25) No 20 20 BOP Process Instr Control Rack (Pn1 1-R-128) No 845 21 "" Spare ******* 22 Aux B1dg Inst A Bus 1 (Pn1 1-L-57) No 20 l 23 Aux Relay Rack A Bus (Pn1 1-R-76) No 364 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 Bus (Pn1 1-L-10) No 42 (44 h 30 Aus Control Panel A Instr Bus (Pn1 1-L-10) No 31 Control Air Ndr A Motsture Alm (JB281) No 13 - 32 Aux Relay Rack A Bus (Pn1 1-R-72) No 182 33 " * " " S pa r e " " * " 34 Post Accident Monitoring (Pn1 1-M-5) No 48 35 * "" *
Spa re "" "

36 LOCA H2 Cntent flow Monitor (Pn1 1-M-10)
No 106 37 CO2 Fire Protection Computer Room No l 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 0431F/COC4 D +.,. c v s.
th dah p kW L t.itle N"I Si "D% Q. I g.i-6 1%ct o -_ j TABLE 8.3.1-15 (Sheet 1) 120V AC VITAL INSTpyw(NT POWER BOARD 2-1 (Batterv toarc !) $KR SAFETY CONNECit0 No LOAD REL Af t0 _LCAO-VA 1 SSPS (A) Ch ! Input Relays (Pnl 2-R a6) Yes 600 2 SSPS (B) Ch ! Input Relays (Pn12-R-49) Yes
~~600, 3 NIS Instr Power Ch !~~
Yes 4.M4 4 NIS Control Power Ch ! (Pnl 2-M-13) Yes 1,@F@ 44-5 Process Protection Set ! (Pnl 2-R-1) Yes 6 UHI Accumulator Ch ! ! solation Valve (Pnl 2-N-283) Yes 62 7 RCP IUV & MF Relays Yes 30 8 Aux Feed Pump Turb Flow Cent Yes 222 9 Turb Dr Aux FW PMP St 3&4 Gen (Pnl 2-1.-381) Yes 75 10 Turb Dr Aux FW PHP St 3&4 Gen (Pnl 2-L-11A) Yes 80 11 RVLIS (Pnl 2-R-148) Yes '7a5 i 12 Incore TC Monitoring (Pnl 2-R-60) Yes 208 13 '""" Spa re "'"" la Instr Bus (Pnl 2-M-4) No 60* l 15 Plugmold Inst Bus 1 (PNL 2-M-5) No 194 16 Plugmold Inst Bus I (Pnl 2-M-6) No 280 17 Process Cont. Group (Pnl 2-R-14) No 7a9 18 BOP Process Instr Cont Rack (Pnl 2-R-126) ho 3.100 6 19 Aux Cont Pnl A Instr Bus (Pnl 2-L-11 A) No ASS 20 BOP Process Instr Cont Rack (Pnl 2-R-128) No 600 21 '""" Spa re "'"" 22 Aux Sldg Inst A Bus (Pnl 2-L-57) No 50 23 Aux Relay Rack A Bus (Pnl 2-4 76) No 162 24 Aux Relay Rack C Bus (Pnl 2-R-76) No 140 25 NSSS Aus 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 $$PS Cent Rm Demod (Pnl 2-M-22)- No 480 29 Aux Cont Pnl A Inst But (Pnl 2-l,-10) No 210 30 '""" S pa r e '""" 31 " * "" Spa r e "'"" 32 Aux Relay Rack A Bus (Pnl 2-R-32) No 56 33 eeee*" Spa re *""" 34 Post Accident Monitoring 1 (Pnl 2-M-5) No 48 35 * * * * * * ' Spa r e "" * " 36 LOCA H2 Cntmnt Flow Monitor (Pnl 2-N-10) No 106 37 eee.eee Spare ******* 33 e e e e ee e Spare '' 39 eeeeeee Spare ******* 40 '""" Spa re "* "" 41 """' Spa re "'"" + 0431F/COC4 ..m -==,.i n.
\\ Syd; Enivd g -M ~%PyW pu rt EON NO. L c.s h L t2 SON-6 W W% b.4 Gb !I;L1-.0F TABLE 8.3.1-15 (Sheet 2) (Continued) 2-1 120V AC VITAL INSTRUMENT POWER BOARD +=l% LBattery Board I) BKR SAFETY CONNECTED - E M RELATED LOAD-VA i 42 ******* Spare ******* 43 * "" " Spa re "" "
44 " " * " Spa r e " " "

45 UNI Instrument Bus 1 (Pn12-N-23A)
No 32 46 " " * " Spa r e " * " " 6 Spare * *" *" .'.".."* *
Spa r e " " "

47 4g TOTAL 11,723 NOTES 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-II, 1-III, and 1-IV also supply common plant loads and j are more heavily loaded than respective unit 2 boards. ? l-t 9 4 ? \\ 0431F/COC4 i _-s ,..-,y ..,r,,, ._y,_
L 2!- > Su W -- h I.Vdud, p *Pl y, o,a - Eu W% _ ECN NO. L r.s a d. ra SON-6' LE hEl~ lol l 0 & OF_. u TABLE 8.3.1-16 (Sheet 1) 4 120V AC V',TAL INSTRUMENT POWER BOARD 2-!! (Batterv Boaro II) h 8KR SAFETY CONNECTED WAD RELATED LOAD-VA t 'i. S$PS (A) Ch II Input Relays (Pn12-R-46) Yes 600 T 2 SSPS (B) Ch II' Input Relays (Pn12-R-49) Yes 600 3 NIS Instr Power Ch II (JB 3403) Yes %3 i~ 4 NIS Control Power Ch II (Pn1 2-M-13) Yes -440- u n 5 Process Protection Set II'(Pn1 2-R-5) Yes . 6 UNI-Accumulator Ch II Isolation Valve (Pn12-N-238) Yes ^62 L 7 RCP_2VV & UF Relays Yes 30 l 8 Aux Feed Pump Tarb Flow Cont (2-L-381) Yes 222 9 Turb Dr Aux FW PHP St Gen 1&2 LIC-3-173,174 Yes 60 1 10~ Turb Dr Aux FW PMP St Gen 1&2 Instr Loop Yes 244 g L 11 RVLIS (Pnl-2 ft-148) Yes 745 L 12 Incore TC Horitoring (Pn12-R-60) Yes 308 1 13 " * "" S pa r'a "" * " 14 Process Cont Group 2 (Pn1 2-R-17) No 743 i 15-Plugmold inst Bus 2 (PNL 2-M-3) No 509 I 16 Aux Cont Pnl B Inst Bus (Pnl 2-L-11B) No 535 17' Plugmold Instr Bus 2 (Pn1 2-M-6)- No 136 6 h 18 B.OP Process Instr Cont Rack (Pn1 2-R-122) .No 629 m - jg
~~" * " Epare * " * * "~~
20 BOP Process Instr Cont Rack-(Pnl-2-R-130) No 454 21 " * " * */ Spa re * """ 22 o Aux Bldg Inst B Bus (Pn12-L-299) No 40 p - 23 Aux Relay Rack B Bus (Pn1 2-R-76) No if* 24 NSSS Aux Relay Rack C Bus (Pnl 2-R-58) No 322 25 NSSS Aux Relay Rick B Bus (Pn1 2-R-58) No 84 26 Aux F.elay Rack B Bus (Pn1 2-R-75) No ~294 27 Aux Relay Rack C Bus (Pnl 2-R-75) No 168 . 28 Aux Cont Pn1 B Relay Bus _(Pn1 2-L-10)- No 84 u 29 Aux Cont'Pnl C Relay Bus (Pnl 2-L-10) No l 30 Aux Cont Pn1' B Instr Bus (Pn12-L-10) No 3M L 31 Emerg VHF Radio (Pn1 G) ---Equipment Removed -- 32 Aux Relay Rack B Bus (Pn1 2-R-72). No 112. 23-Aux Relay Rack C Bus (Pn1 2-R-72) No 56 34 Post Accident Monitoring 2 (Pn1 2-N-4) No 64 l 35 * """ Spare *""" - 36 LOCA H2 Cntmnt Flow Monitor (Pnl 2-N-10) No 106 L 37 RVLIS (Pnl 2-R-148) Yes 729 p 38 *""" Spa re "" * " l-39 * * * "" Spar e ' * """ 40 " * "" Spare *""" - 41: "" * " Spar e " * "" = Li 0431F/COC4 1 i -.?4. m a#-m'm e p t w=* w --M'-9WNWM 7 ~ * - "
/ Ethuakioq ~ '744 ~eum e+ A s) tg SQN-6 Di L(el% E6fIidQg *l Sh*'t R 109 __cp ']. TABLE 8.3.1-16 (Sheet 2) (Continued) 120V AC VITAL INSTRUMENT POWER BOARD 2-!! (Battery Board !!) BKR-SAFETY CONNECTED LOAD RELATED LOAD-VA - 42' * * * * * *
~~Spa r e * * * * * "~~
~~43 - * * * *"~~
~~Spare *"""~~
~~44 Inplant VHF Radio Repeater F1 No 1,068~~
~~45. UHI Instrument Bus 1 (Pn12-N-238)~~
No 32 46 On Site Paging Radio 40 1,068 6 7 *""" 48 ' - - -- - w$na rg""" HT. S IWsi s ( m % Soona 4 sooa t) % NWs-h _er - Nan- ^ TOTAI 1 q NOTES ~ 1. Each inverter is capable of supplying 15 kva continuously. 6 2. No automatic load stripping or load sequencing is employed. i 3. Power Boards 1-1, I-II, 1-!!I, and 1-!V also supply common plant loads and are more heavily loaded than respective unit 2 boards. = E e 4 e 0431F/COC4 $e
f.e ENCLOSURE 3 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-88-42) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS 1
4 L ENCLOSURE 3 i Significant Hazards Evaluation TVA has evaluated the proposed TS change and has determined that it does not represent a significant hasards 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 deenergization 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 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 increase in the probability or consequences of an accident previously evaluated. (2) Create the possibility of a new or different kind of accident from any previously analyzed. Two administrativa changea are proposed to support the installation of the Gamma Metrica-BR 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 involves a change in engineering unite 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 of accident from any previously analyzed. (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 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 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. .}}

ML20006A285

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