ML18058A156

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Proposed Tech Specs Surveillance Requirement Re Low Flow Trip Setting for non-operating Pump Combinations & Safety Limits & Limiting Safety Sys Settings
ML18058A156
Person / Time
Site: Palisades Entergy icon.png
Issue date: 02/03/1992
From:
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML18058A155 List:
References
NUDOCS 9202110109
Download: ML18058A156 (25)
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ENCLOSURE I Consumers Power Company Palisades Plant Docket 50-255 TECHNICAL SPECIFICATIONS CHANGE REQUEST Revised Technical Specification Pages February 3, 1992 6 Pages I ,

9202110109 6~8~8~55 PDR ADOCK PDR p

2.0 SAFETY LIMI~AND LIMITING SAFETY SYSTEM SET~GS 2.1 Safety Limit - Reactor Core The Minimum DNBR of the reactor core shall be maintained greater than or equal to the DNB correlation safety limit.

Correlation Safety Limit XNB 1.17 ANFP 1.154 Applicability Safety Limit 2.1 is applicable in HOT STANDBY and POWER OPERATION.

Action 2.1.1 If a Safety Limit is exceeded, comply with the requirements of Specification 6.7 2.2 Safety Limit - Primary Coolant System Pressure (PCS)

The PCS Pressure shall not exceed 2750 psia.

Applicability Safety Limit 2.2 is applicable when there is fuel in the reactor.

Action 2.2.1 If a Safety Limit is exceeded, comply with the requirements of Specification 6.7 2.3 Limiting Safety System Settings - Reactor Protective System (RPS)

The RPS trip setting limits shall be as stated in Table 2.3.1 . .

Applicability Limiting Safety System Settings of Table 2.3.l are applicable when the associated RPS channels are required to be OPERABLE by Specification 3.17.1.

Action 2.3.1 If an RPS instrument setting is not within the allowable settings of Table 2.3.1, immediately declare the instrument inoperable and complete corrective action as directed by Spec ifi cation 3 .17. I.

Amendment No. ~J, i~, ~~' 11~, 1~7, 2-1

TABLE 2.3.1 REACTOR PROTECTIVE SYSTEM TRIP SETTING LIMITS Four Primary Coolant Three Primary Coolant RPS Trip Unit Pumps Operating Pumps Operating

1. Variable High sl5% above core power, sl5% above core ~ower Power wtth a minimum of with a minimum of s30% RATED POWER sl5% RATED POWER and a maximum of and a maximum of sl06.5% RATED POWER. s49% RATED POWER.
2. PCS Flow ~95% Full PCS Flow. ~60% Full PCS Flow.
3. High Pressure s2255 psia. s2255 psia.

Pressurizer

4. Thermal Margin/ (a) (a)

Low Pressure

5. Steam Generator Above the feedwater Above the f eedwater Low Water Level ring center line. ring center line.
6. Steam Generator ~500 psia. ~500 psia.

Low Pressure

7. Containment High s3.70 psig.
  • s3.70 psig.

Pressure (a) The p~essure setpoint for the Thermal Margin/Low ~ressure Trip, Ptrip' is the h1gher of two values, Pmin and Pvar' both rn ps1a:

pmin = 1750 Pvar = 2012(QA)(QR 1 ) + 17.0(Tin) - 9493 where:

QA QA * -0.

= -0.333720t5Il + 1. 028; ASI + 1.067; when -0.628 s ASI < -0.100 when -0.100 s ASI < +0.200 QA = +0.375 ASI + 0.925; when +0.200 s ASI s +0.565 ASI = Measured ASI when Q ~ 0.0625 ASI = 0.0 when Q < 0.0625 QR, = 0 .. 412 (Q) + 0. 588; when Q s 1.0 QR, = Q; when Q > 1.0 Q = Core Power/Rated Power Tin = Maximum primary coolant inlet temperature, in or.

ASI, Tin' and Q are the existing values as measured by the associated instrument channel.

Amendment No. ~J, ~~, JJ~, J~~,

2-2

TABLE 4.1.1 MiniJll.111 Frequencies for Checks, Calibrations and Testing of Reactor Protective System Surveillance Channel Descri~tion Function Fr!l9uenc~ Surveillance Method

1. Power Range Safety Channels a. Check 171 s a. C011'4'8rison of four-power channel readings.
b. Check 131 . D b. Channel adjustment to agree with heat balance calculation.

Repeat whenever flux-AT power C011'4'8rators alarms.

c. Test Ml2l c. Internal test signal.
d. Cal ibrate 161 R d. Channel alignment through measurement/adjustment of internal test points.
2. Wide-Range a. Check s a. C011'4'8rison of channel indications.

Neutron Monitors b. Test p b. Internal test signal.

c. Calibrate R c. Channel alignment through measurement/adjustment of internal test points.
3. Reactor Coolant Flow a. Check s a. C011'4'8rison of four separate total flow indications.
b. Calibrate R b. Known differential pressure applied to sensors.
c. Test Ml21 c. Bistable trip tester.Ill
4. Thermal Margin/Low a. Check: 181 s a. Check:

Pressurizer Pressure (1) T~rature (1) C011'4'8rison of four separate calculated trip pressure input set point indications.

(2) Pressure (2) C011'4'8rison of four pressurizer pressure indications.

input Same as S(a) below.)

b. Calibrate R b. Calibrate:

(1) T~rature (1) Known resistance substituted for RTD coincident with input known pressure and power input.

(2) Pressure (2) Part of S(b) below.

input

c. Test Ml2l c. Bistable trip tester.Ill
5. High-Pressurizer Pressure a. Check 181 s a. C011'4'8rison of four separate pressure indications.
b. Calibrate R b. Known pressure applied to sensors.
c. Test Ml2l c. Bistable trip tester.Ill 4-3 Amendnent No.J9, U, 118, JJ9, JJ/,,

TABLE 4.1.1 Minill'llll Frequencies for Checks, Calibrations and Testing of Reactor Protective System (continued)

Surveillance Channel Description Function frequency Surveillance Method

6. Steam Generator Level a. Check s a. c~rison of four level indications per generator.
b. Calibrate R b. Known differential pressure applied to sensors.
c. Test Ml2l c. Bistable trip tester.Ill
7. Steam Generator Pressure a. Check s a. C~risons of four pressure indications per generator
b. Calibrate R b. Known pressure applied to sensors.
c. Test Ml2l c. Bistable trip tester.' 11
8. Containment Pressure a. Calibrate a. Known pressure applied to sensors.
b. Test b. Simulate pressure switch action.
9. Loss of Load a. Test p a. Manually trip turbine auto stop oil relays.
10. Manual Trips a. Test p a. Manually test both circuits.
11. Reactor Protection System a. Test a. Internal test circuits.

Logic Units

12. Axial Shape Index CASI) a. Test R a. Known power inputs applied to Thermal Margin Calculator.
13. Ar Power a. Check 171 s a. Same as 1(a).
b. Check 131 D b. Same as 1(b).
c. Test R c. Known temperature inputs applied to Thermal Margin Calculator.

4-4 Amendment No J~, tili, 118, 130 March 23, 1990

TABLE 4.1.1 Mini111J111 Frequencies for Checks, Calibrations and Testing of Reactor Protective System (continued)

Surveillance Channel Description Function Frequency Surveillance Method

14. Thermal Margin Calculator a. Check Q a. Verify constants.

(1) The bistable tr'ip tester injects a signal into the bistable and provides a precision readout of the trip set point.

(2) All monthly tests will be done on only one of four channels at a time to prevent reactor trip.

(3) Adjust the nuclear power or b.T power until readout agrees with heat balance calculations when above 15X of rated power.

(4) Deleted (5) It is not necessary to perform the specified testing during prolonged periods in the refueling shutdown condition If this occurs, omitted testing will be performed prior to returning the plant to service.

(6) Also includes testing variable high power function in the Thermal Margin Calculator.

(7) Required if the reactor is critical.

(8) Required when PCS is >1500 psia.

FREQUENCY Notation Notation Frequency s At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

w At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 6 months.

R At least once per 18 months.

p Prior to each start*up if not done previous week.

NA Not applicable.

4-5 Amendment No J9, iii, iJ9,

ENCLOSURE 2 Consumers Power Company Palisades Plant Docket 50-255 TECHNICAL SPECIFICATIONS BASIS REVISION Revised Limiting Safety System Settings Basis pages February 3, 1992 5 Pages

2.0 BASIS Limits and Limitin S ~m Settin s 2.1 Basis - Reactor Core Safety limit To maintain the inte~rity of the fuel cladding and prevent fission product release, it 1s necessary to prevent overheating of the cladding under normal operating conditions. This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coolant temperature. The upper boundary of the nucleate boiling re~ime is termed "departure from nucleate boiling" (DNB). At this point, there is a sharp reduction of the heat trahsfer coefficient, which would result in high-cladding temperatures and the possibility of cladding failure. Although DNB is not an observable parameter during reactor operation the observable parameters of thermal power, primary coolant fiow, temperature and pressure, can be related to DNB through the use of a DNB Correlation. DNB Correlations have been developed to predict DNB and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB ratio (DNBR},

defined as the ratio of the heat flux that would cause DNS at a particular core location to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to DNB correlation safety limit. A DNBR equal to the DNB correlation safety limit corresponds to a 95% probability at a 95% confidence level that DNB will not occur which is considered an appropriate margin to DNB for all operating conditions.

The reactor protective system is designed to prevent any anticipated combination of transient conditions for primary coolant system temperature, pressure and thermal power level that would result in a DNBR of less than the DNB correlation safety limit. The Palisades safety analyses uses two DNB correlations. The XNB correlation discussed in References 1 and 2 determines the safety limit for those fuel assemblies initially loaded in Cycle 8. The ANFP correlation discussed in References 4 and 5 determines the safety limit for those fuel assemblies initially loaded in Cycle 9 and later. Fuel assemblies initially loaded in Cycle 8 are of a different construction than later assemblies which utilize a High Thermal Performance design.

The minimum DNBR analyses are in accordance with Reference 6.

References 1 XN-NF-62l(P}(A}, Rev 1 2 XN-NF-709 3 Updated FSAR, Section 14.1.

4 ANF-1224 (P)(A}, May 1989 5 ANF-89-19~(~), January 1990 6 XN-NF-82-21(A}, Revision 1 Amendment No. ~1, ~~' 11~, 1~7, B 2-1

2.0 BASIS - Saf~ Limits and Limiting Safety Sy~m Settings 2.2 Basis - Primary Coolant System Safety Limit The primary coolant system< 1> serves as a barrier to prevent radionuclides in the primary coolant from reaching the atmosphere.

In the event of a fuel cladding failure, the primary coolant system is the foremost barrier against the release of fission products.

Establishing a system pressure limit helps to assure the continued integrity of both the primary coolant system and the fuel claddin~.

The Primary Coolant System design pressure is 2500 psia. The max1mum allowable Primary Coolant System transient pressure is limited by the pressure vessel limit (ASME Code,Section III) of 110% of desi~n pressure and by the piping, valve, and fitting limit (ASA Sect1on B31.l) of 120% of design pressure. The initial hydrostatic test was conducted at 125% of design pressure (3125 psia) to verify the integrity of the Rrimary coolant system. Thus, the safety limit of 2750 psia Cll£% of the 2500 psia design pressure) has been established. The settings of the reactor High Pressure Trip, primary safety valves, and secondary safety valves have been established to assure never reaching the primary coolant system safety limit. Additional assurance that the nuclear steam supply system (NSSS) pressure does not exceed the safety limit is provided by the normal setting of the atmospheric steam dump and turbine bypass valves of 900 psia.

References (1) Updated FSAR, Section 4.

(2) Updated FSAR, Section 4.3.

Amendment No i~, JJ~,

B 2-2

BASIS S tltm Settin s 2.3 Basis - Limiting Safety System Settings The reactor protective system consists of four instrument channels to monitor selected plant conditions which will cause a reactor trip if any of these conditions deviate from a preselected operating range to the degree that a safety limit may be reached.

1. Variable High Power - The Variable High Power Trip (VHPT) is incorporated in the reactor protection system to provide a reactor trip for transients exhibiting a core power increase starting from any initial power level (such as the boron dilution transient). The VHPT system provides a trip setpoint no more than a predetermined amount above the indicated core power with a specified upper limit. Operator action is required to increase the setpoint as core power is increased; the setpoint is automatically decreased as core power decreases. Provisions have been made to select different set points for three pump and four pump operations.

During normal plant operation with all primary coolant pumps operating, reactor trip is initiated when the reactor power level reaches 106.5% of indicated rated power. Adding to this the possible variation in trip point due to calibration and instrument errors, the maximum actual steady state power at which a trip would be act~ted is 115%, which was used for the purpose of safety analysis.

2. Primary Coolant System {PCS} Low Flow - A reactor trip is provided to protect the core aga~nst DNB should the coolant flow suddenly decrease significantly. Flow in each of the four coolant loops is determined from pressure drop from inlet to outlet of the steam generators. The total flow through the reactor core is determined, for the RPS flow channels, by summing the loop pressure drops across the steam generators and correlating this pressure sum with the sum of steam generator differential pressures which exists at 100% flow (four pump operation at full power t ve>* The normal flow with three pumfs operatin~ is 74.7% of Ful 1 PCS Flow. Full PCS flow is that f ow which ~xists at RATED POWER, at full power Tave' with four pumps operating.

During four pump operation, the Low Flow Trip setting of 95%

insures that the reactor cannot operate when the flow rate is less thN\ 93% of the nominal value considering instrument errors.

Provisions are made in the reactor protective system to permit operation of the reactor at reduced power if one coolant pump is taken out of service. These low-flow and high-flux settings have been derived in consideration of instrument errors and response times of equipment involved to assure that thermal mar~in and flow stability will be ~~intained during normal operation and anticipated transients. For reactor operation with one coolant pump inoperative, core power must be reduced and then the Variable High Power and Low Flow setpoints must be adjusted to the three pump values before the pump may be stopped.

Amendment No. pl, 11~, 1P7, B 2-3

2.0 BASIS - Saf~ Limits and Limiting Safety Sy~m Settings 2.3 Basis - Limiting Safety System Settings (continued)

3. High Pressurizer Pressure - A reactor trip for high pressurizer pressure is provided in conjunction with the primary and secondary safety valves to prevent primary system overpressure (Specification 3.1.7). In the event of loss of load.without reactor trip, the temperature and pressure of the primary coolant system would increase due to the reduction in the heat removed from the coolant via the steam generators. This setting is consis~en~ >with the trip point assumed in the accident analysis. 8
4. Thermal Margin/Low Pressure (TM/LP) Trip The TM/LP trip system monitors core power, reactor coolant maximum inlet temperature, (Tin), core coolant system pressure and axial shape index. The Low Pressure Trip limit

~~~a~l. is calculated using the equations given in Table The calculated limit (Pv ) is then compared to a fixed Low Pressur,e Trip limit (Pm\:J. The auctioneered highest of these signals becomes "Lne trip limit (Ptr* ) . Ptri is compared to the measured PCS pressure andPa trip ~ignal is generated when the measured pressure for that channel is less than or equal to Ptri . A pre-trip alarm is also ~enerated when P is less than 8r equal to the pre-trip setting Ptrip +

AP.

The TM/LP trip set points are derived from the 4-pump operation core thermal limits through application of appropriate allowances for measurement uncertainties and processing errors. A pressure allowance of 165 psi is assumed to account for instrument drift in both power and inlet temperatures, calorimetric power measurement, inlet temperature measurement, and primary system pressure measurement. Uncertainties accounted for that are not a part of the 165 psi term include allowances for assembly power tilt, fuel pellet manufacturing tolerances, core flow measurement uncertainty and core bypass flow, inlet temperature measurement time delays, and ASI measurement. Each of these allowances and uncertainties are included in the.development of the TM/LP trip set point used in the accident analysis.

Amendment No ~1, ~i, 11~,

B 2-4

2:0 BASIS - Safi Limits and Limiting Safety Sylm Settings 2.3 Basis - Limiting Safety System Settings (continued)

5. Low Steam Generator Water Level - The low steam generator water level reactor trip protects against the loss of feed-water flow accidents and assures that the design pressure of the primary coolant system will not be exceeded. The specified set point assures that there will be sufficient water inventory in the steam generator at the time of trip to allow a safe and orderly plant shutdown and to prevent steam gr~erator dryout assuming minimum auxiliary feedwater capacity.

The setting listed in Table 2.3.1 assures that the heat transfer surface (tubes) is covered with water when the reactor is critical.

6. Low Steam Generator Pressure - A reactor trip on low steam generator secondary pressure is provided to protect against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the primary coolant. The setting of 500 psia is sufficiently below the rated load operating point of 739 psia so-as not to interfere with normal operation, but still high enough to provide the required protection in the event of excessively high sH)am flow. This setting was used in the accident analysis.
7. Containment High Pressure - A reactor trip on containment high pressure is provided to assure that the reactor is shutdown*

before the initiat~Rn of the safety injection system and containment spray. .

References 1 EMF-91-176, Table 15.0.7-1 2 Updated FSAR, Section 7.2.3.3.

3 EMF-91-176, Section 15.0.7-1 4 XN-NF-86-9l(P) 5 ANF-90-078, Section 15.1.5 6 ANF-87-150(NP), Volume 2, Section 15.2.7 7 Updated FSAR, Section 7.2.3.9.

8 ANF-90-078, Section 15.2.1 Amendment No~), ~1, ))~, )~7, B 2-5

ENCLOSURE 3 Consumers Power Company Palisades Plant Docket 50-255 TECHNICAL SPECIFICATIONS CHANGE and BASIS REVISION Existing Pages Marked to Show Changes February 3, 1992 12 Pages

2.0

2. 1 SAFETY LI. AND LIMITING SAFETY SYSTEM saNGS SAFETY LIMITS - REACTOR CORE

-~~

1c* -

Applicability ..

f.l'F'lll. ~!T>L't. ~* ....

s~ee1 1eT"- l.zeR ~~131 i * ~h :~~. .;=ter i ~ in ot standby L I s..... _...., .r;'

Thi$

eo"dition andcaower oper:f1.;,.;s: t  := . . _J rObjectiV,¥ ./ ./ / / . / ~ \-__l.

revent e ~

to the p imary Specifications f'/11 "I I NIV~i\

The11 foNBR of the reactor core shall be maintai.ned greater than or equal to the DNB correlation safety limi~.

~

To maintain the inte~rity of the fuel cladding and prevent fission product release, it 1s necessary to prevent overheating of the .

~pht~ ~0\)~ cladding under normal operating conditions. This is accom~lished by

-.u ....~(? operating within the nucleate boiling regime of* heat transfer, *

~""'orl 2" wherein the heat transfer coefficient is large.enough so that the clad surface temperature is only sl*ightly gteater t~an the c~ola~t fa temperature. The upper boundary of the nucleate bo1 Hng re~1me* 1s :1. *

~ermed Rdeparture.from nucleate bo1l1ng*n (DNB}. At .. this eornt*,, th8'~

1s a sharp reduct 1on of the heat transfer coefficient, wh1 ch would :;;;"

result in high-cladding temperatures and the possibility of cladding .

failure. Although DNB is not an observable parameter during reactor operation the observable parameters of thermal power, primary coolant fiow, temperature and pressure, can be related to DNB through the use of.a DNB Correlation. DNB Correlations have *been developed to predict DNB and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB ratio (DNBR),

defined as the ratio of the heat flux that would cause DNB at a*

particular core location to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated

  • transients is limited to DNB correlation safety limit. A DNBR egual to the DNB correlation safety limit corresponds to a 95% probability at a 95% confidence level that * * *
  • -;~:~~ ~t'*.
)t~f*

2-1 Amendment No. JJ~, -+3r

-fe~FY&i 9 !6. 1991

2.1 SAFETY LIMIT. REACTOR CORE (.Contd)

DNB will not occur which is considered an appropriate margin to DNB for all operating conditions.

The reactor protective system is designed to prevent any anticipated combination of transient conditions for.,primary coolant system temperature, pressure and thermal power level that would result in a DNBR of less than the DNB correlation safety limit. The ONB correlations used in the Palisades safety analysis ~re listed in the following table.

I tlC...L u t>E.'> . Cori~"' ~'""... ,,...l

' i-4

~,.~rt--rf 1,..rMI f

__. - XNB ANFP Safety JJ.mi1 1.17 1.154 The MDNBR analyses are performed in accordance with Reference 6.

References

.. ~* :.

I','**

2-2

2.2 SAFETY LIM~ - PRIMARY COOLANT SYSTEM PRES~E Aopl icabil ity

-~

rl/D VIL- I/

-rD

    • . ~ Afflt(P~ll1" ressure shall not exceed 2750 psia when 1n the reactor vessel.

Basis The primary coolant system 0 >serves as a barrier to prevent radionuclides in the primary coolant from reaching the atmosphere.

In the event of a fuel cladding failure, the primary coolant system is the foremost barrier against the release of fission products.

Establishing a system pressure limit helps to assure* the continued integrity of both the primary coolant system and the fuel cladding.~

The maximum transient pressure allowable in the primary coolant system pressure_ vessel under the ASME Code,Section II I, is 110% oft! .

design pressure~ The maximum transient pr~ssure allowable in the f primary coolant system piping, valves and fittings under ASA Section 831.1 is 120% of design pressure. Thus, the safety limit of 2750 psia (110%. of* the 2500 psia design pressure)-*has been established. <2 >

The setting~ and capacity of the secondary coolant system safety valves (985-1025 psig)< 3 >, the reactor high-pressure trip (~2400 psia) and the primary safety valves (2500-2580' psia) <*>have been established to assure never reaching the primary coolant system pressure safety limit. The. in,itfal hydrostatic. test was conducted at 3125. psia (125%

of design pressure) to-verify the integrity of the* primary coolant system:. Additional assurance that the nuclear steam* supply system (NSSS)... pressure:* does not exceed the safety limit h provided by setting.the.. secondary< coolant system steain dump and bypass** valves at 900 psi~. * *

  • References ~ . \.

c.:~>.J'J-,

( l~::.U~ated: FSA~, Section 4.

( l~.U~ated* FSAR;' Section 4. 3*

(l)l~Ojidated. FSAR,,, Table 4-5 (4f Ui>dated~ FSA~~ Table 4-10 2-3 Amendment No i', -tttt"

-Nevem~eP 16, 1988

2.3 LIMITING SAFI SYSTEM SETTINGS - REACTOR PRO.TIVE SYSTEM Apo] icabil ity {jrc,,w ,o..?rL* (Jl.~ru rf I<:.. !!>e.t.ow J

. pl~ to r~tor eip [etting(

af bypfses ~*

. ~...-----

The TM/LP trip system nioni tors core power, reactor coolant *,

maxtmum inlet temperature, (T1"h core .coolant system . _

U *p pressure and a~.ia l.shape Jndex. The 1ow pr~ssure trfp 111 mi t

(.P,..r) *is ca lcµJ ated us).ng 'the fo 11 owing *equation.

- ~ * !2*012~:('Q

    • var .~:;

..A> (QR

>* .. *. ; i*7 *O;c"T 1 *. + * * ~ - '

-.~ <.: -.::. . . .* . .

l i.>~~- 1 tn .-.-::- *

~ . '

9493,_

w

. ~. *:. ~,

I.

~ *'. ~ . '

where: . ;. *i

' r  :, * *

1*. '

QR1 .. .0.412(Q) + 0.588

  • Q.-~ LO .- Q .. *core power ' *.

.. Q . *.. * .

  • Q >'1.0 rated PQWer ..
  • .r:.-*

.QA ... -0;120'(ASI). + l,:'.Q28 - ~o.:628 -~ ASI < ~o . .100

  • 1 i

I

  • -~0.333(ASI) +.1:-067 .~0:100_~ ASI < +9.200 T0.375(ASI) + 0.925 +0.200 ~ ASI ~ :t-0.565

\ ID : .*

.. ' 1.085 when Q < 0.0625 <

The c~ lctilated l.i.mi t, ~.(-P~.J ~; s -th~n compared to .~ :~f 1xed low

('AOUIJ:-?> pressure tr.i ~ :Jimi t *.(.p.1").* **~The _.auct 1o.neered highest of these IO s i gna'l s becomes th~ ;tr.1 p _Timi t *,(.P trtJ

  • Ptrtp is compared ti>> the*
  • measured ,reactor coo 1ant .pressure ( t') and a trip s1:gna 1 is * . *

. "f!>A--~IS generated when P Js les.s :t~an or equa1 to Ptrtt. ' A pre-trip :a1 arm is also gen~rated ~h~n-P is less than or equa1 to ~he*pre-tr1p

  • setting Ptrtp + AP *.

ofC. r ,,.1!> \.~ . z. '!>. r ~~IL... ~? frLI C R ~ t 'L t.J ~ £..J Attllr-. rzE.a v* ~*t:..r:> '1U 'n.~ Of?ree~~~

2-4 Amendment No. J J~, 4"37-7 Februaf':J 20, 1991

TABLE 2.3.1 Reactor Protective System Trip Setting Limits Four Primary Coolant Three Primary ~ant Pumps Operating pumps Operatin~ a - -

1S"% . .

1. Vari a~-~ High ~~ '~" above core power, ~J:.0% above core power Powe~ with a minimum setpoi"t ~ith a minimum set13oirrt of ~30%\.9-f"' rated power ~..wr-~15% rated power and a maximum of ~106.5% and a maximum of ~49%

..ttf" rated power .f!lf rated power

?c..~ F'Lt>w ru&.L f"C.~ Fto.., h,u.. PL~ Fr..o w

2. lmary ---. c/ ~95% ef Primaty Coolant ~60% ef Pr;ffia1y Cool-

£.eel ant Fl OW>' ~ ew With Fet11 Pumps ~Rt Flew With Four Q13e1 at111g ~~ffif39 9J3eratiRg..

3. High Pressure ~2255 Psia ~2255 Psia Pressurizer (6..)
4. Thermal Margin/ Replaced by ¥at iabie Low Pressur~ ~igh* Powe* Trip--and 1759 Psfa MI 11 Imam tow-

.P1 essur e:. Sett i119

  • 5~* Steam Generator Not bewer T~a~ tbe Net Lower Tha" the Ce&ter Low Water Level CeRter Li 11e of bi Re ef FeeEI WateF Ri'ag-FeeEI Watet Ring
  • l'-TJo\J~ '"tt-4~ Fifa..'>._.,-.~

~t)&"IZ- , ...~ ~"Xoll!A'1!.~ .. ~,,.,G- e~tt. tr,.,,L-*..::~

1(111~ ~,...,.~tr.. \.t..S'!- .

6. Steam Generator ~soo Psia
  • 4':500:: Psi a, Low Pressur~
7. Containment High ~3*. 70 Psig ~3.70' Ps1g Pressure e

power.

(2) May b ypassed below cir itry is operab . For low p al margin/l w essure, primar. coolant flow d low steam gener pressure tri may be bypassed unt thei'r react rnts are reached roximately 175 sia 1and

  • psia~ res tivelyy, pro <fed automatic byP, remova'1 circui at 10* %

rated~ P,- . er is operab

  • Mini ..-*':~rip set ti shall be 1750 p *
  • Ope a~.fo~,with t ee pumps for a m imum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> permitted to pr l i 'ted-'t1me: for* repair/pump resta t, to provide. for n orderly shutd n

.provide__ for the: conduct of react.or internals noise or:iitoring test measurements.

Amendment No. ~J, ~~' JJ~. ~

  • Febt tia1' 22, 1991

~-. 3- LIMITING SAFETY .TEM SETTINGS - REACTOR PROTEC. -SYSTEM {C~ntd)

Basis _

The reactor protective system consists of four instrument channels to monitor selected plant conditions which will cause a reactor trip if any of these conditions deviate from a preselected operating range to the degree that a safety limit may be reached.

1. Variable High Power - The variable high power trip {VHPT) is incorporated in the reactor protection system to provide a reactor trip for transients exhibiting a core power increase starting from any initial power level {such as the boron dilution transient). The VHPT system provides a trip setpoint flO more than a predetermined amount above the indicated c~rJ!t::power.

Operator action is required to increase the setpoifit*as core

-power is increased; the setpoint is automatically de:creased as core power decreases. Provisions have been made to select different set points for three pump and four pump.operations.

During normal plant operation with all primary coolant .pumps

-operating, reactor trip is initiated when the reactor power level reaches 106.5% of indicated rated power. Adding to this the

.possible variation in trip point due to calibration and instrument errors, the maximum actual steady state power at which a trip would be actuated is 115%, which was used for the purpose

  • I of safety analysis. <1> . *- _
2. Primary Coolant System Low Flow - A reactor trip is provided to protect the core against DNB should the coolant flow suddenly decrease significantly.<~ Flow in each of the four coolant loops 1s determined from a measurement of pressure drop from -tnlet to outlet of the steam generators. The total flow throu~h the:

reactor core is measured by summing the loop pressure drops .

.across the steam generators and correlating this pressure sum with the pump calibration flow curves. The percent .of .normal core flow i-s _shown in the following table:

4 Pumps 100.0%

3 Pumps 74.7%

During four-pump operation, the low-flow trip setting of 95%

insures that the reactor cannot operate ~hen the flow rate is Less than 93% of the nominal value considering instrument errors.<*> -

2-6 Amendment No. ~J, JJ~, 137 February 20, 1991

2.3 e

LIMITING SAFETY SYSTEM SETTINGS - REACTOR PROTECTIVE SYSTEM (Contd) e Basis (Contd) (

Provisions are made in the reactor protective system to permit I operation of the reactor at reduced power if one coolant pump is I taken out of service. These low-flow and high-flux settings have I been derived in consideration of instrument errors and response I times of equipment involved to assure that thermal margin and I flow stability will be maintained during normal operation and I anticipated transients. (S) For reactor operation with one coolant I pump inoperative, the low-flow.trip points and the overpower trip I points must be manually changed to the specified values for the I selected pump condition by means of set point selector switches. I The trip points. are shown in Table 2.3.1. I

3. High Pressurizer Pressure - A reactor.trip for high pressu~izer I pressure- is provided. in conjunction with the primary and secondary safety valves to.-prev~nt primary system.:overpr~ssure (Specification 3.1.7). In the event of loss of load without reactor trip, the- temperature and pressure of th.e primary coolant system would increase due to. the reductio~ .. in the. heat removed.

from the coolan~ via the. steS!D generator's... This. ~et ting is, I consistent, witlt* the t.~ip point assumed. in. the. ac.cident (11) . .. *~.

analys.is-. .<* - ~...c.- o,__  ! . * .!:* * ..... *.: .

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2-7 Amendment No J~; 118 November 15, 1988 TSP1088-0181-HL04 .

2.3 LIMITING SAFETY SYSTEM SETTINGS - REACTOR PROTECTIVE SYSTEM (Continued)*

Basis (Continued)

4. Thermal .*Margin/Low""'.Pressure Trip The TM/LP trip .set .. points are derived from the 4:..pump operation I core thermal limits *through ~pplication of ~ppropriate allowances I

.for **measurement* uncertainties and processing errors. A pressure I allowance of 165 psi is assumed to account for: instrument I drift in both power and inlet temperatures; calorimetric power I measurement; inlet temperature measurement; *:and *primary system I pressure measurement. Uncertainties accounted *for-;~.~~ _are not I a part of the 165 psi terin include allowances *for: *~jliembly

  • I power tilt; fuel pellet" manufacturing *tolerances; core'- flow I
  • measurement uncertainty and core bypass flow; .;inlet .t.elnperature I measurement time ~elays*;. and ASI .measurem~.nt.
  • 1~ach :of these I allowances and uncertainties :*are *included~ 'in .the :d*velopment I of the TM/LP trip set point used* in. the- :accident '&nal'Ysis. I r . . ~~. ~ : .; "

. For three-pump *operation, :contl.nued* power oper-ation: *is restricted. 1.

.,During* *this mode-'*of* operatian,"'-'tbe .*hi'8h*>po¥er .:.level *trip in I conjunction wit:h *the TM/LP*~tnp* '(minilmim*ta&t):.point ** ~ 750 psia) and *the. secondary '&yst'na :Safety valves ('8et at':approximately 1000 psia) nsure that ade<<1uate DNB margin* 7is maintained. <5 ) I I

I I

I

.5. Low Steam Generator Water Level - The low steam generator water level reactor trip protects against the loss of feed-water flow accidents and assures that the design pressure of the primary coolant system will not be exceeded. The specified set point assures that there will be sufficient water inventory in the steam generator at the time of trip to allow a safe and orderly l plant shutdown and to prevent steam generator dryout assuming I minimum auxiliary feedwater capacity. <9 > . I The setting listed in Table 2.3.1 assures that the heat transfer surf ace (tubes) is covered with water when the reactor is critical.

2-8 Amendment No Jt, 8t, 118 November 15, 1988 TSP1088-0181-NL04

.;. 2 .3 LIMITING SAFE~SYSTEM SETTINGS - REACTOR PRO~IVE SYSTEM (Contd)

Basis (Contd)

6. Low Steam Generator Pressure - A reactor trip on low steam generator secondary pressure is provided to protect against an excessive rate of heat extraction from the steam generators and

~ubsequent cooldown of the primary coolant. The setting of 500 psia is sufficiently below the rated load operating point of 739 psia so as not to interfere with normal operation, but still high enough to provide the required protection in the event of excessively high steam flow. This setting was used in the accident analysis. <9 >

7. Containment High Pressure - A reactor trip *on containment high pressure is provided to assure that the reactor is shutdown before the initiation of the safety injection system and containment spray~ <10 > * *
8. Low Power Physics Testing - For low power physics tests, certain tests will require the reactor to be critical at low temperature

- (~260°F) and low pressure (~415 psia). For these certain tests only, the thermal margin/low pressure, primary coolant flow and low steam generator pressure trips may be bypassed in order that reactor power can be increased for improved data acquisition.

Special operating precautions will be in effect during these ~

tests in accordance with approved written testing procedures~ At reactor power 1eve ls be 1ow .10- 1% of rated power, the thermal :;i margin/low-pressure trip and low flow trip are nQt required to

  • prevent fuel rod thermal limits from being exceeded. The low steam generator pressure trip is not required because the low steam generator pressure will ~ot allow a severe reactor cooldown, should a steam line break occur during these tests.

References (1) ANF-90-078, Table 15.0.7-1 (2) deleted (3) Updated FSAR; Section 7.2.3.3.

(4) ANF-90-078, Section 15.0.7-1 (5) XN-NF-86-91(P)

(6) deleted (7) deleted ._.

(8) .ANF-90*078*; Section 15. I. 5 (9) :.~~'ANf-87-lSO(NP), Volume 2, Section 15.2.7

( 10) 2~;updated FSAR, Section 7. 2. 3. 9. *

.(11) fc~A8f.~90-078, Section 15.2.1

~-- .. c=-

2-9 Amendment No. ~J, JJ~, 137 February 20, 1991

TABLE 4.1.1 HiriillLm Frequencies for Checks, Calibrations and Testing of Reactor Protective System(S)

,_; :y.: ;.,j,:)~~ '

Surveillanee Chamel pescripticin Function. Frequency Surveillance Method

1. Power Range Safety Channels a. Ch~ki*i* s a. Coq>&rison of four-power chamel readings. *
b. Check 1' 1. D b. Channel adjustment to agree wl th heat balarlc:e calculation~
c. Test . ..,.,. Repeat Whenever flux-AT power coq>arators alarms *
c. Internal test signal. * .
d. Cal tbrate 111 it d. Chamel al tgnnent thrOugh measurement/adjwtment of. Internal test )>Oints *

,1_.1 .~.

I .,, f*' . " ' *

's ' fi}

' ... r.

Wtde*Ran9e 111. Check ~.;..' a* coq,artion of chamel lridicatlons. *:

Neutron Monitors b. Test p b~.,.lnternail tHt* signal~ .. : .. , .t, .

c. Calibrate ~* '~ R c.:-:Chamel al lgnnent thrc)ugh measurement/adji.istment of Internal t..t

, 1, test points. ..

V, *"

'*  !; ~ *t * , ** I

  • f i

.3. Reactor Coolant Flow 111. Check ,.

.:~

':, *~~ a *. Coq:ia'rtilon of. four separate total flow' ti'idtcatlohs.

b. Cal tbrate
c. Test* * ,.J .! ~!

. R

  • ...,., b.* 'icno.in differential press~e appl led to sensors.
c. Btstilble tdp tester
  • 1' 1>"' * * . .
  • 1 (J *-

101

4. Thermal Margin/Low a. Check: a:; cfieck: . .

Pressurizer Pressure (1) T~rature*.  : '(1) .Coq)artson of'.foUt separate calculated trip pressure Input \l set *point. lndfC:atlons. , .* ..

(2) PressJre . <2) 'C~rtson of four pressurizer pressure Indications.

Input I ,'., :Same as 5(a) below.)

b. Calibrate 'R b *., ciilitir~te:* . "l . . . ' .

(1) Teq>erilture .(1'). Known reifstance substitutea for*RTD coinelClent with Input ... kn6wli' pressure aild power i riput.

(2) Pressure . .:.-, <2> Part* of 5(b) below.

c. Test Input ;
  • ..,., c;*eist~l~ t'rip tester.';,

' ' i' Lr;

  • 1 ~
5. iiiiih*Pressuriter Pressure a~ Check ,,,  :*~:'.;' ' "S '
b. Calibrate

. c. Test

... \

, <ii

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. " I ..... , .

I,* (~ ~

i*.

+*

4-3 Amendnent No.JS, Iii, 118, 1JS, ~

. Ftlsl IMI l~!i; "1991=-

J . ,  ! ~ . ..

rABLE 4.1.1

. ;'; ., Surveillance ,. :: *,

Oulnnel Punction '1'

~~~~~Y~.* Surveillance Method Olec:Jt jl r/v.* ,)l)(" s .:. f:' a~ Comparison of four level. indications per generator.*.

'i/ ).u ~  ;**!;t'

. Calibrate ,.,. b*. ICnown differential pressure applied to sensors *

    • .i: M(2) c. Bistable trip.tester.Cl) teat *'*

r ,\;\*** .....

s a. Callparisons of four pressure indications

'\'~. l, .~. ~...... ~) . *.:,;per generator.

b. Calibrate R b. Known .pressure applied to sensors.

~ *~; ... .. ~ c:. ' *'fA!St. i '.~ - HC2) c. Bistable trip tester.Cl)

  • a~ Containment Pressure a. Calibrate R a. ICnown pressure applied to sensors.
b. rest H(2) b. *. Simulate pre11sure switch action.
9. Loss of Load

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.:11

.l;est

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p . <a * . Han1J81lY tr~p turbine auto stop oil relays.

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10. Manual trip~*

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a. teat"'  ;=-i:~;, . .. *:.*;  :;i  !*'***
a. Han.ually test both circuits.

Reactor Pl"~~~ti(in Sys~*:*

. .j.* *

u. *' * * ** ... "(2)

~le lh)~tf; * 'i, **,;: .*l:*,;u i)l.J A ...* ;~ . :.~; ,

  • . r  ;
  • Az*~1 'fiba~ i~. (~~) .

1 ' . ', 1 ~ ' .. ,,v * ....

J2. **. re11t R a. ICnown power inputs applied to lbermal

.', {. ..:. .:: .~ ~('. t-,- :,, i.~1} f~i.~ . , .,.:.:.... . ,,. '~ . ~rs~~ C.lc;ulator.

..~ .. :---.. *'( ~. 1:"*" ; . ,.

~~-i;*, { .. '.f :1~. . ~. 'Lj .
a. *Cb9c:Jt (7 f  ::T-~!.4.i;,;;.:;.. a. Same as l(a). I
b. aie~tiq) ,*;.* D ' . b*;1.., Saiae es l(b).
  • .fJ,~,J..b~io I'*£' * ,,*:' ':. *
c. Teat R * * . *c;.* ;; t(nown ~perature impute applied to

. ~**  ; -;.- . '~. :i .

    • Dlermal Hilrgin Calculator.
  • J **

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4-4 Amendment No Jf, * tt*, 130 March 23, 1990 TSP1088-0181-NL04-NL02

Minimum Frequencies for Oieclta, Calibrations and Testing of Reactor Protective Syatem(5) (Contd)

su~1iH1nC
e aiarine 1 Description . i'tinction Frequency Surveillance Method
14. lberilal Margin Calculator a. Dleclt Q a. Verify constants.

\'  :{*  ;, '( '1 ' *, . ' l .,,* '* I;! ' >I.  ;,

(l)l'he ~latable trip teeter injects a signal into th~ bistable.and provides a preclslon readout of the trip &et point.

11 (l)!Clj~t' tbi msclear '°""

(2)Ail 9onthly testa will be d0ne on only one of tour channels at ,a tlile* to prevent reactor trip.

bi ta po1'er untti reado\l~-*ireea "1.tb liiit'b~t~'nce '~h~~lattona when above 15% of rated nation for er tlia~oper ng~Ullp c binatlona must* be and within hours a reBU11tng ~tion ~

nat10n h88:n0t **en te&te.(! within 'the pre~d~a montp- .

(5)It ls not neceaa*ary to perform the apeclfled teatlng during prolonpd petioda in the refueling shutdown condition If thia occurs, omitted testing will be perforeed prior to returnlni the* plant to service.

(6)Alao,lncludea testing variable hip paver function in the lberaal Margin Calculator.

J. ~  ;  ; r (7)itequirid U the reactor la critical. I

(&)Required when PCS ta > 1500 pals. '. *. **:  ;*,.. I

' .*, ,,., t 'i. " ". \

.' ~ * ' ~ ! ..

Rotsi:lon Freqaency

. ~~ ._

, .. i '.:

s .

At least one~' per 12 boors.

D At least one~ per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

~ ' ** ~1 * . :.

.I ~ .. w At lea at once per 7 days.

  • ,'.-\I M .*.At least once J>!!lr i 31 days.

" Q . At leaat

':Ii..:

once per 92 days.

SA , At least once per 6 months.

R *;:

At least once per 18 months.

p *l>ri0r to ~ch' at~it-up u not ci0ne previous welt~~

NA Not applicable.

4-5 Amendment No If, tt*, 130 March 23, 1990 TST *0181-NL04-NL02

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