Text

Amend 33 to License NPF-73,modifying Tech Specs to Replace Requirements on Reactor Coolant Resistance Temp Detector (RTD) Manifold Instrumentation W/Requirements for Fast Response thermowell-mounted RTDs
	ML20059N204
Person / Time
Site:	Beaver Valley
Issue date:	09/25/1990
From:	Stolz J; Office of Nuclear Reactor Regulation
To:	; Cleveland Electric Illuminating Co, Duquesne Light Co, Ohio Edison Co, Toledo Edison Co
Shared Package
ML20059N205	List: ML20059N204; ML20059N206;
References
	NPF-73-A-033 NUDOCS 9010100246
	Download: ML20059N204 (17)
	v • d • e

_

.,g UNITEJ STATES -

[.

NUCLEAR REGULATORY COMMISSION r.,

wa:HmoTow. o. c. zones DUQUESNE LIGHT COMPANY i

OHIO EDISON COMPANY i

THE CLEVELAND ELECTRIC ILLUMINATING COMPANY t

i THE TOLEDO EDISON COMPANY l

DOCKET NO. 50-412 l

BEAVER VALLEY POWER STATION, UNIT NO. 2 i

AMENDMENT TO FACILITY OPERATING LICENSE j

Amendment No. 33 License No. NPF-73 i

i 1.

The Nuclear Regulatory Comission (the Comission) has found that:

i A.

The application for amendment by Duquesne Light Company, et al.

(thelicensee)datedApril 16, 1990, complies with the standards and I

requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; i

B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of.

l the Comission, C.

Thereisreasonableassurance(1)that'theactivitiesauthorized by this amendment can be conducted without endangering the health

~

and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; i

D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of.the public; i

and l

I i

E.

The issuance of this amendment is in accorcince with 10 CFR Part l

51 of the Comission's regulations and all applicable requirements.

have been satisfied.

L l

1 i

i I

9010100246 900925 PDR ADOCK 05000412 i'

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1

+. 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-73 is hereby amended to read as follows:

(2).TechnicalSpecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 33, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated in the license. -DLC0 shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment.is effective as of the date of its issuance, to be implemented prior to restart from the 1990 refueling outtge.

FOR THE NUCLEAR REGULATORY COMMISSION Jo n F. Stolz, Dire r

P oject Directorate I-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: September 25, 1990 l

j 1

ATTACHMENT TO LICENSE AMENDMENT h0. 33 FACILITY OPERATING LICENSE NO. NPF-73 DOCKET NO. 50-412 Replace the folicwing pages of the Appendix A Technical Specifications with the enclosed pages as indicated.

The revised pages are identified by amendment nunber and contain vertical lines indicating the areas of change.

Remove Insert 2-4 2-4 2-7 2-7 2-8 2-8 2-9 2-9 2-10 2-10 3/4 2-12 3/4 2-12 3/4 3-6 3/4 3-6 3/4 3-8 3/4 3-3 8 2-4 B 2-4

2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Raactor Trip System Instrumentation and Interlock Setpoints shall be consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY:

As shown for each channel in Table 3.J-1.

ACTION:

With a Reactor Trip System Instrumentation or Interlock Setpoint less con-a.

servative than the value shown in the Trip Setpoint column but more conser-vative than the value shown in the Allowable Value column of Table 2.2-1 adjust the Setpoint consistent with the Trip Setpoint value, b.

With the Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Value column of Table 2.2-1, either:

1.

Adjust the Setpoint consistent with the Trip Setpoint value of Table 2.2-1 and determine within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that Equation 2.2-1.was satisfied for the affected channel or 2.

Declare the channel inoperable and apply the applicable ACTION state-ment requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint value.

EQUATION 2.2-1

' + R + S < TA where:

~

i Z

=

The value for column Z of Table 2.2-1 for the affected channel, R

=

The "as measured" value (in percent span) of rack error for the affeeted channel, S

=

Either the "as measured" value (in percent span) of the sensor error, or the value of Column S (Sensor Error) of Table 2.2-1 for the affected channel, and l

TA

=

The value from Column TA (Total Allowance in % of span) of Table 2.2-1 for the affected channel.

l BEAVER VALLEY - UNIT 2 2-3 AMENDMENT NO. 33 l

TABLE 7.2-1 cn

'S REACTOR TRIP SYSTEN INSTRUMENTATION TRIP SETPOINTS x

FUNCTIONAL UNIT ALLOWANCE (TA)

Z_

S_

TRIP SETPOINT ALLOWA8LE W.LUE-h 1.

Manual Reactor Trip N.A.

N.A.

2.

Power Range, Neutron Flux c3 a.

High Setpoint 7.5 4.56 0

1 109% ef RTP' 1 111.1% of RTP*~

.a m

b.

Low Setpoint 8.3 4.56 0

J25% off RTP

127.1% of RTP*

3.

Power Range,. Neutron Flux,

1. 6 0.50 0

1 5% of RTP* with 1 6.3% of RTP* with High Positive Rate a time constant a time constant

> 2 seconds

> 2 seconds 4.

Power Range, Neutron Flux,

1. 6 0.50 0

1 5% of RTP* with 1 6.3% of RTP* with High Negative Rate a time constant a time constant eu

> 2 seconds

> 2 seconds 1

s.

5.

Intermediate Range, 17.0 8.41 0'

1125% of RTP*

1 30.9% of RTP*

Neutron Flux 6.

Source Range, Neutron Flux

17.0 10.01 0

1 105-cps 1 1.4 x 105 i

cps 7.

Overtemperature AT 7.0 5.10 See Note 5 See Note 1 See Note 2 8.

Overpower AT 4.9

. 1.71 1.49 See Note 3 See Note 4 l

9.

Pressurizer Pressure-Low 3.1

- 0.71 1.67-

> 1945 psig***

> 1935 psig***

Pressurizer Pressure-High

' 6. 2

4. %

0.67

- 1 2375 psig i 2383 psig 10.

G 11.

Pressurizer Water Level-High 8.0 2.18 1.67

< 93.8% of instru-1

-< 92% of instru-

~ ment span ment span E

l 5

ilesign flow **

~> 88.8% of loop 12.

Loss of Flow 2.5 1.39' O.60

> 90% of loop design flow **

U$

RATED THERMAL POWER

=

- Loop design flow ='88,500 gpa

- - Time constants utilized in the'1ead-lag controller for Pressurizer Pressure-Low are 2. seconds for lead and I second for lag.

Channel calibration shall ensure that these time constants are adjusted to those values.

y--

g v'

-+,#

. vs y--w-

.-,ww-,.2,

.,,,.,, -. -, +,

3._w,-

,,,... +.,_.,

..--m m.-.

u.i,m--

-m

~m.-

-___.m.-_-_m__,w

.,_* w

_r,_.__-__..a 6

_,o_m n

em TABLE 2.2-1 (Continued) i E

gj REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS_

m NOTATION I

E NOTE 1:

OVERTEMPERATURE AT 0

Il+TMI ob~2

+r 1

1 b I

3 6

Where:

AT

=

Measured AT; 1+r5 1

Lead-lag compensator on measured AT;

=

1+1 S2 Time constants utilized in lead-lag compensator for AT, t

=8s, In,12

=

y 1 = 3 s; 2

w I

=

Lag compensator on measured aT; 1+T 4

3 Time constants utilized in the lag compensator for AT, 13 = 0 s;

'3

.=

AT,

'=

Indicated AT at RATED THERMAL POWER; K

=

1.28; y

g K.

0.017/'F;

=

4 2

. z' k

'+r5 4

- 2; The. function generated by the lead-lag compensator for T dynamic compensation;

=

1+r z

5 o

Time consta' nts ud Hzed in lead-lag censator for T,yg,14 = 30 s, 15

4 5; r,1

=

4 5

.m.-

.,,,.+-w-w,-,

wmw,_c+-

-.3

.-w v--

r,--y

..w.,-,

iv-wi - --,,

w

-.cce,

,--.,-w-a.e e--w

- * - -#.wr

.-e, wi e

,+-.._--_.-m__

-.w.

TABLE 2.2-1 (Continued) m

~5<

E REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued) h T

Average temperature, "F;

=

1 Lag compensator on measured T,yg;

=

1+1 b

_c 6

3 Time constant utilized in the measured T,yg lag compensator, 16 = 0 s;

-e 1

=

6 i

ro T'

$ 576.2*F (Nominal T,yg at RATED THERMAL POWER);

=

1 K

=

3

0. M 82; P

=

Pressurizer Pressure, psig; P'

2235 psig (Nominal RCS operating pressure);

n 5

Laplace transform operator, s 2;

=

and f (al) is a function of the indicated difference between top and. bottom detectors of the power-range y

nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

T (i)

For gt.~%

ween -3 R and M, f (AI) = 0, W re_'qt y

a are percen RARD TH N b

POWER in the top and bottom halves of the core respectively, and q i

% s total TH N t

POWER in percent of RATED THERMAL POWER;

. (ii)

For each percent that the magnitude of qt

% exce s -3 N, t k af Trip Setpoint shall s.

be automatically reduced by 2.52% of its value at RATED THERMAL POWER; and "z

j g

(iii)

For each percent that the magnitude q g exceeds +9%, the AT Trip Setpoint shall t

be automatically reduced by 1.75% of its value at RATED THERMAL POWER.

zP-NOTE 2:

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 1.6% of AT span.

ww if i

O

+

,--,..-wrw e.#.w-

.,-o..

-ww.-.--g.m

.. ~

W e-v-,gg

,-y

--,e, er e.4, m,

.g 3 4+

--,*.-<.--w

-gy

,_n,-,e-

,.,w*

---=.w

+

-w w

g,--

e..%

e.-*

%..e4,.e.

-m-

- - a

TABLE 2.2-1 (Continued)

S-

_g REACTOR TRIP SYSTEM INSTRUMENT TION TRIP SETPOINTS A

=-

NOTATION (Cor'cinued)

'Np NOTE 3:

OVERPOWER AT Q

O

' 6

+1 5 ~

2( '}l O

7)

II * '6S

+

1+1 5 1 3

g 6

3 ro Measured AT; Where:

AT

=

I+r5 1

Lead-lag compensator on measured AT;

=

1+TS2 Time constants utilized in lead-lag compensator for AT, t

=8s,1

35;

=

r,1 y

2 y

2 Lag compensator on measured AT; 1

=

1+1 b 3

. Time constant utilized in the lag compensator for AT, r = 0 s; r

=

3 3

AT, Indicated AT at RATED THERMAL POWER;

=

1.0781; K

=

4 "0.02M for increasing average twrature and 0 for & creasing average twrature; K

=

g

g 1 5 The function generated by the rate-lag compensator for T dynamic compensation; M

7

=

- U E

1+rS7 a

,5.

7 Time. consta.,t utilized in rate-lag compensator-for T,yg, 17 = 10 s;

=

.r d

l LIMITING SAFETY SYSTEM SETTINGS RASEC specified in Table 2.2-1, in percent span, from the analys'is assumptions.- Use l

of Equation 2.2-1 allows for a sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS.

The methodology to derive the trip setpoints is based upon combining all of the uncertainties in the channels.

Inherent to the determination of the trip setpoints are the magnitudes of these channel uncertainties.

Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes.

Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance.

Being;that there is a small statistical chance that this will happen, an infrequent excessive drift is expected.

Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more seriods problems and should warrant further investigation.

Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

Power Range, Neutron Flux The Power Range, Neutron Flux channel high setpoint provides reactor core protection against reactivity excursions which are too rapid to be protected by temperature ar.d pressure protective circuitry.

The low setpoint provides redund-ant protection in the power range for a power excursion beginning from low power.

The trip associated with the low setpoint may be manually bypassed when P-10 is active (two of the four power range channels indicate a power level of above approximately 10 percent of RATED THERMAL POWER) and is automatically reinstated when P-10 becomes inactive (three of the four channels indicate a power level below approximately 10 percent of RATED THERMAL POWER),

Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of rod ejection events from any power' level.

Specifically, this trip complements the Power Range Neutron Flux High and Low l

trips to ensure that the criteria are met for rod ejection from partial power.

The Pcver Range Negative Rate trip provides protection to ensure that the-minimum DNBR is maintained above 1.30 for control rod drop accidents.

At high power a multiple rod drop accident could cause local flux peaking which, when in conjunction with nuclear power being maintained equivalent to turbine power by action of the automatic rod control system, could cause an unconservative-local DNBR to exist.

The Power Range Negative Rate trip:will prevent this from occurring by tripping the reactor.

No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DNBRs will be greater than 1.30.

t l

BEAVER VALLEY - UNIT 2 B 2-3 AMENDMENT NO. 33 l

2.2 LIMITING SAFETY SYSTEM SETTINGS l

l t

BASEE Intermediate and Source Range, Nuclear Flux f

I The Intermediate and Source Range, Nuclear Flux trips provide reactor core l

protection during reactor startup to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a suberitical condi-tion.

These trips provide redundant protection to the low setpoint trip of the i

Power Range, Neutron Flux channels.

The Source Range Channels will initiate a reactor trip at about 10'5 counts per second unless manually blocked when P-6 becomes active.

The intermediate range channels will initiate a reactor trip at a current level proportional to approximately 25 percent of RATED THERMAL POWER-unless manually blocked when P-10 becomes active. ' Although no explicit t

credit was taken for operation.of the Source Range Channels in the accident i

analyses, operability requirements in the. Technical Specifications will ensure that the Source Range Channels are available to mitigate the consequences of an

[

inadvertent centrol bank withdrawal in MODES 3, 4 and 5 Overtemperature AT The Overtemperature AT trip combinations of pressure, power, provides core protection to prevent DNB for all coolant temperature, and axial power distribu-i tion, provided that the transient is slow with respect to piping transit, ther-i mowell, and RTD response time 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 changes in den-i sity and heat capacity of water with temperature and dynamic compensation for transport, thermowell, and RTD response time delays-from the core to RTD output i

indication. With normal axial power distribution, this reactor' trip limit is i

always below the core safety limit as shown on Figure 2.1-1. If axial peaks are l.

greater than design, as indicated by the difference between, top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1.

Overpower AT l

The Overpower AT reactor trip provides assurance of. fuel integrity, e.g.,

5 no melting, under all possible overpower conditions, limits the required range 1-for Overtemperature AT protection, and provides a backup to the High Neutron l

Flux trip.

The setpoint includes corrections for changes in density and heat capacity of water with temperature, and dynamic compensation for transport, thermowell, and RTD response time delays from the core to RTD output indication, i

The Overpower AT trip provides protection to mitigate the consequences _of vari-ous size steam line breaks as reported in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Release."

L i

BEAVER VALLEY - UNIT 2 B 2-4 AMENDMENT NO. 33 1.

e 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlock Setpoints shall l

be consistent with the Trip Setpoint values shown in Table 2.2-1.

i APPLICABILITY:

As shown for each channel in Table 3.3-1.

t

,i ACTION:

With a Reactor Trip System Instrumentation or Interlock Setpoint less con-i a.

t L

servative than the value shown in the Trip Setpoint column but more conser-l vative than the value shown in the Allowable Value column of Table 2.2-l' adjust the Setpoint consistent with the Trip Setpoint value, b.

With the Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Value column of Table 2.2-1, either:

1.

Adjust the Setpoint consistent with the Trip Setpoint value of Table 2.2-1 and determine within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that Equation 2.2-1 was i

satisfied for the affected channel or 2.

Declare the channel inoperable and apply the applicsble ACTION state-r I

ment requirement of Specification 3.3.1 until the channel is restored i

l to OPERABLE status with its setpoint adjusted consistent with the l

Trip Setpoint value.

r EQUATION 2.2-1 Z + R + S < TA l

where:

~

)

Z

=

The value for column Z of Table 2.2-1 for the affected channel, i

R

=

The "as measured" value (in percent span) of rack' error for the 6

affected channel, f

S

=

Either the "as measured" value (in percent span) of the sensor 6

error, or the value of Column S (Sensor Error) of Table 2.2-1 for l

the affected channel, and i

TA

=

The value from Column TA (Total Allowance in % of span) of Table 2.2-1 for the affected channel.

'h i

i BEAVER VALLEY - UNIT 2 2-3 AMENDMENT NO. 33

TA8tE 7.2-1 E

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT ALLOWANCE (TA)

Z_

S TRIP SETPOINT ALLOWABLE VALUE h

1.

Manual Reactor Trip N.A.

N.A.

l 2.

Power Range, Neutron Flux cz a.

High Setpoint 7.5 4.56 0

1 109% of RTP*

1 111.1% of RTP*

.-A N

b.

Low Setpoint 8.3 4.56 0

125% of RTP*

$27.1% of Rfr*

3.

Power Range, Neutron Flux, 1.6 0.50 0

's 5% of RTP* with 1 6.3% of RTP* with

~

High Positive Rate

- a time constant a time constant 1 2 seconds 1 2 seconds 4

Power Range, Neutron Flux,

1. 6 0.50 0

1 5% of RTP* with 1 6.3% of RTP* with.

High Negative Rate a time constant a time constant

- ro 1 2 seconds 1 2 seconds a

5.

Intermediate Range, 17.0 8.41

- 0' s 25% of RTP*

1 30.9% of RTP*

~

Neutron Flux 6.

Source Range, Neutron Flux 17.0 10.01 0

1 105 cps 1

5 i

.4 x 10 cps 7.

Overtemperature AT 7.0 5.10 See Note 5 See Note 1 See Note 2 8.

Overpower AT 4.9 1.71 1.49 See Note 3 See Note 4 j '

9.

Pressurizer Pressure-Low 3.1 0.71 1.67 1 1945 psig***

1 1935 psig***

i 10.

Pressurizer Pressure-High 6.2

4. %

0.67 1 2375 psig i 2383 psig

< 93.8% of instru-9 11.

Pressurizer Water Level-High 8.0 2.18 1.67

-< 92% of instru-

~ ment span 3

ment span 4

mz 12.

Loss of Flow 2.5 1.39 0.60

> 90% of loop

> 88.8% of loop 5

design flow **

E

-* = RATED THERMAL POWER

- Loop design flow = 88,500 gpm

- - Time constants utilized in the lead-lag controller for Pressurizer Pressure-Low are 2 seconds for lead and 1 second for lag. -' Channel calibration shall ensure that these time constants are adjusted to those values.

1

-E.-..

..ww.-..,e

w w

v..----

+=w

,e

w.,

,.--w+

.,.,ec---%..

,...w.,--

w..v-

+ % vw,

.,.-,-.-,,ew,,+-

--ev. m --

y._.ew-.

' w o

LIMITING SAFETY SYSTEM SETTINGS RASES specified in Table 2.2-1, in,,srcent span, from the analysis assumptions.

Use of Equation 2.2-1 allows for a sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS.

The methodology to derive the trip setpoints is based upon combining all of the uncertainties in the channels.

Inherent to the determination of the trip setpoints are the magnitudes of these channel uncertainties.

Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes.

Rack drift in excess.of the Allowable Value exhibits the behavior that the rack has not met its allowance.

Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected.

Rack or sensor drift, in excess of the allowance that is more than. occasional.may be indicative of more serious problems and should warrant further investigation..

Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

Power Ranos, Neutron Flux The Power Range, Neutron Flux channel =high setpoint provides reactor core protection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry.

The low setpoint provides redund-I ant protection in the power range for a power excursion beginning from low powsr.

The trip associated with the low setpoint may be manually bypassed when P-10 is active (two of the four power range channels indicate a power level of above approximately 10 percent of RATED THERMAL POWER) and is automatically reinstated when P-10 becomes inactive (three of the four channels indicate a power. level below approximately 10 percent of RATED THERMAL POWER),

Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of rod ejection events from any power level.

Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from partial power.

The Power Range Neptive Rate trip provides protection to ensure that the minimum DNBR is maintained above 1.30 for control rod drop accidents.

At high' power a multiple rod drop accident could cause local flux peaking which, when in conjunction with nuclear power being maintained equivalent to turbine power by action of the automatic rod control system, could cause an unconservative local DNBR to exist.

The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor.

No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DNBRs will be greater than 1.30.

BEAVER VALLEY - UNIT 2 B 2-3 AMENDMENT.NO. 33

2. 2 LIMITING SAFETY SYSTEM SETTINGS I

BASES l

Intermediate and Source Range, Nuclear Flux

(

1 l

The Intermediate and Source Range, Nuclear Flux trips provide reactor core i

protection during reactor startup to mitigate the consequences of an uncon-t trolled rod cluster control assembly bank withdrawal from a suberitical condi-

-l tion.

l These trips provide redundant protection to the low setpoint trip of the l

Power Range, Neutron Flux channels.

The Source Range Channels will initiate a i

l reactor trip at about 10'5 counts per second unless manually blocked when P-6 becomes active.

The intermediate range channels will initiate a reactor trip i

at a currant level proportional to approximately 25 percent of RATED THERMAL.

- i POWER unless manually blocked when P-10 becomes active.

Although no explicit credit was tat * 'or operation of the Source Range Channels in the accident-analyses, oper e ;fty requirements in the Technical Specifications will ensure j

that the Source Range Channels are available to mitigata the consequences of an t

inadvertent control bank withdrawal in MODES 3, 4 and S.

Overtemperature aT The Overtemperature AT trip provides core protection to prevent DN8 for all i

combinations of pressure, power, coolant temperature, and axial power distribu-tion, provided.that the transient is slow with respect to piping transit, ther-mowell, and RTO response time 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 changes in den-i sity and heat capacity of water with temperature and dynamic compensation for transport, thermowell,-and RTD response time delays from the core to RTD output i

indication. With normal axial power distribution this reactor trip limit is alwaysbelowthecoresafetylimitasshownonFlgure2.1-1.Ifaxialpeaksare greater than design, as indicated by the difference between too and bottom power l

range nuclear detectors, the reactor trip is automatically reduced according j

to the notations in Table 2.2-1.

4 Overpower AT The Overpower AT reactor trip provides assurance of fuel integrity, e.g.,

i no melting, under all possible overpower conditions, limits the required range l

for Overtemperature AT protection, and provides a backup to the High Neutron i

Flux trip.

The setpoint includes corrections for changes in density and heat capacity of water with temperature, and dynamic compensation for transport, thermowell, and RTD response time delays from the core to RTD output indication.

The Overpower AT trip provides protection to mitigate the consequences of vari-ous size steam line breaks as reported in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Release."

i r

I l

~;

-BEAVER VALLEY - UNIT 2 B 2-4 AMENDMENT N0. 33 l

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E

NOTATION s

r-

{

NOTE 1:

OVERTEMPERATURE AT 1

AT

{

(I * '3 5

I * '65 3

)~I 5

o 1

2 1

l Where:

AT Measured AT;

=

1+r5 1

Lead-lag compensator on measured AT;

=

1+T S 2

Time constants utilized in lead-lag compensator for AT, 11=8s, In. Y2

=

m 1 = 3 s; 2

9 1

Lag compensator on measured AT;-

=

1+T b3 3

Time constants utilized in the lag compensator for AT, 13 = 0 s; r

=

AT, Indicated AT at RATED THERMAL POWER;

=

K

=

3 1.28; 3

K 0.017/*F;

=

2 E.

,5,,

1+rS 2;

4 The function generated by the lead-lag compensator for T dynamic compensation;

=

1+Y5 M

z g

?

O

'4' '5 Time constants utilized in lead-lag compensator for T,yg, r4 = 30 s, 15

4 *;

=

~.

TABLE 2.2-1 (Continued)

.9

- g REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

- =

NOTATION (Continued)

$p T

Average temperature, 'F;

=

E

- 1 Lag compensator on measured T,yg;

=

1+1 c_

6 3

Time constant utilized in the measured T,yg lag compensator, 1

-4 1

=

0 s;

=

6 6

T'

=

< 576.2*F (Nominal T,yg at RATED THERMAL POWER);

K 0 2 2;

=

3 P

Pressurizer Pressure, psig;

=

P' 2235 psig (Nominal RCS operating pressure);

=

c'o S

Laplace transform operator, s-1;

=

i and f (AI) is a function of the indicated difference between top and bottom detectors of the power-range y

nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup i

ter,ts such that:

(i)

For g A

ween -3 3 and M, f (AI) = 0, h e q and q are percent RATED THERMAL-t b

y t

b POWER in the top and bottom halves of the core respectively, and qt

9 is o al TH N b

POWER in percent of RATED. THERMAL POWER; (iii)

For each percent that the magnitude of q A exceeds -33%, the AT Trip Setpoint shall t

b 3

be automatically reduced by 2.52% of.its value at RATED THERMAL POWER; and

. "z g

(iii)

For each percent that the magnitude q g exceeds'+9%, the AT Trip Setpoint shall t

.b

- 5 be automatically reduced by 1.75% of its value at RATED THERMAL POWER.

)~

NOTE 2:.

The channe'1's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 1.6% of l

AT span.

ww S

v=i m - ww-re.c -

,-e+ym,

.,.e 3-

,r 3

e-,,-.<,._g y,

y,g 3.-r,-

v

---,-,i,

,s-e--,,-

+%

,w.

u,-

,wr

..-c-w w

c_

..--wwe...w..._-~,.we+w g-

,.,--w ym+y 9

1 TABLE 2.2-1 (Continued) 1 9g REACTOR TRIP SYSTEM INSTRUMENTATION, TRIP SETPOINTS w

NOTATION-(Continued)

$p NOTE 3:

OVERPOWER AT

-Q AT (II *

b) I I

) < AT {K

-K T-K I

)

I I

)

6 [T (I

} - I"] - f (ai)}

I 2 ) II * '3 ) -

4 5 (1 + 1 1+1 b

5 i.

g 7)

II * '6 )

b 5

6) 2 1+1 5

"w M

Where:

AT

=

Measured AT; 1

1+rS 1

Lead-lag compensator on measured AT;

=

1+T b2 ty,12 Time constants utilized in lead-lag compensator for AT, 11 = 8 s,12 = 3s;

=

1 Lag compensator on measured AT;

=

1+T b 3

3 zed in W 1ag compensatoi-for. AT,1 = 0 s; 1

=

me c ns an u 3

I AT, Indicated AT at RATED THERMAL POWER;

=

K

=

1.0781; 4

i g

0.02 M for increasing' average temperature and 0 for & creasing a m age temperature; K

=

$9 1 S i

'E 7

=

The function generated by the rate-lag compensator for T dynamic compensation; l

9 1+1S

- 9 w.

7 5

1

=

Time constant utilized in rate-lag compensator for T,yg, 17 = 10 s; 7

U I

.~.

TABLE 2.2-1 (Continued) 9.

X REACTOR TRIP SYSTEM INSTRUMENIATION TRIP SETPOINTS NOTATION (Continued)

-g.

p 1

=

Lag compensator on measured T Q

l+1 6 Time constant utilized in the measured T lag compensator, 1 0 s; T

=

6 gy 6

U K

=

0.0012/ F for T > T" and K6 = 0 for T $ T";

6 ro T

=

Average Temperature, F; Indicated T,yg at RATED THERMAL POWER (Calibration temperature for AT T"

=

instrumentation, $576.2'F);

Laplace transfort. operator, s-8; and 5

=

7 g

f (AI) 0 for all al.

=

2 NOTE 4:

The channel's maximum Trip Setpoint shall not exceed its computed Trip-56tpoint by more than 2.5% of AT span.

NOTE 5:

The sensor error for' temperature is 1.49% and 0.73% of span for pressure.

i s9-e w

v-e-re-evs w

,,*-+,-r+--

ww w

w

-*r--

e---

'w--

. w

-W 1wwr--ee-'-s,a=o=

wew---

---e,,

w>w e

w i-,

-w'--**~+r w w

g%

w e-

we-e, we yw+w e *-,

,-~w e-

=

vs em

')

.i i

j POWER DISTRIBUTf0N LIMITS DNS PARAMETERS LIMfTINo CONDITION FOR OPERATION 3.2.5 The following DN8 related parameters sh.;11 be maintained within the limits shown on Table 3.2-1:

(

Reactor Coolant System T,yg a.

b.

Pressurizer Pressure

'i c.

Reactor Coolant System Total Flow Rate APPLICABILITY: FODE 1*

ACTION:

+

With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5 percent of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

i t

SURVEILLANCE REQUIREMENTS 4.2.5.1.1 Each of the~ parameters of Table 3.2-1 shall be verified to be indi-cating within their limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.2.5.1.2 The provisions of Specification 4.0.3 and 4.0.4 are not applicable for the reactor startups following the initial fueling for Reactor Coolant System total flow rate to allow a calorimetric flow measurement and the cali-bration of the Reactor Coolant System total flow rate indicators, 4.2.5.2 The Reactor Coolant System total flow rate shall be determined to be i

within its limit by measurement at least once per 18 months.

l l

"The provisions of Specification 3.0.2 are not applicable for the reactor startup following the initial fueling for Reactor Coolant System total flow rate to allow a calorimetric flow measurement and the calibration of the Reactor Coolant System total flow rate indicators.

i BEAVER VALLEY - UNIT 2 3/4 2'11 Amendment No. g, 33 l

-APR 26 g t

t i

TABLE 3.2-1

[

DNB PARAMETERS 3 Loops in PARAMETER Operation l

Reactor Coolant System T,yg

< 580.3'F Pressurizer Pressure 1 2220 psia

Reactor Coolant System 1 274,800 gpm**

Total Flow Rate l

i I

i l

e i

t i

i t

1

" Limit not applicable during either a THERMAL POWER ramp increase in excess of

+

5 percent RATED THERMAL POWER per minute or a THERMAL POWER step increase in excess of 10% RATED THERMAL POWER.

- Includes a 3.5% flow measurement uncertainty.

BUUER '. VALLEY. - UNIT 2 3/4 2-12

-Amendment No. M, 33

[

TABLE 3.3-1 (Continued)

TABLE NOTATION

)

With the reactor trip system breakers in the closed position and the control rod drive system capable of rod withdrawal.

(1) Trip function may be manually bypassed in this' MODE above P-10.

i (2) Trip function may be manually bypassed in this MODE above P-6.

ACTION STATEMENTS ACTION 1 -

With the number of OPERABLE channels one less than the Minimum l

Channels OPERABLE requirement, be in at least HOT STAN08Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE.

ACTION 2 -

With the number of OPERABLE channels one less than the Total Number of Channels and with the THERMAL POWER level:

a.

Less than or equal to 5% of. RATED THERMAL POWER, place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after increasing THERMAL POWER above 5% of RATED THERMAL POWER; otherwise reduce thermal tower to less than 5% RATED THERMAL POWER within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , b.

Above 5% of RATED THERMAL POWER, operation may continue provided all of the following conditions are satisfied:

1.

The inoperable channel is placed in the tripped condi-tion within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.

The Minimum Channels OPERABLE reautrement is met; how-ever, the inoperable channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing of other channels per i

Specification 4.3.1.1.

3.

Either, THERMAL POWER is restricted to <75% of RATED THERMAL and the Power Range, Neutron Flux trip setpoint is reduced to <85% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, tiie 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 per Specification 4.2.4.C.

l ACTION 3 -

With the number of channels OPERABLE one less than required by l

the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a.

Below P-6, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 setpoint.

BEAVER VALLEY - UNIT 2 3/4 3-5 AMENDMENT 'NO. : 33

t TABLE 3.3-1 (Continued) b.

Above P-L but below 5% of RATED THERMAL POWER, restore the I

inoperable channel-to OPERABLE status prior to increasing THERMAL POWER above 5% of RATED THERMAL POWER.

t c.

Above 5% of RATED THERMAL POWER, POWER OPERATION may continue.

l ACTION 4 -

With the number of channels OPERABLE one less than required by

+

the Minimum Channels OPERABLE requirement and with the THERMAL l

POWER level:

6 a.

Below P-6, restore the inoperable channel to OPERABLE status prior to, increasing THERMAL POWER above P-6 setpoint and 1

suspend positive reactivity operations.

i b.

Above P-6, operation may continue.

l ACTION 5 -

With the number of OPERABLE channels one less than the Minimum-

-1 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 open the Reactor Trip System breakers, suspend all operations involving positive reactivity 1

changes and verify Valve 2CHS-91 is closed and secured in i

position within the next hour.

i ACTION 6 -

This Action is not used.

ACTION 7 -

With the number of OPERA 8LE channels

one less than the Total I

Number of Channels and with the THERMAL POWER level:

I

-t a.

Less than or equal to 5% of RATED THERMAL POWER, place the inoperable channel in the tripped condition within 1 hour-t restore the inoperable-channel to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after increasing THERMAL POWER'above 5% of RATED THERMAL POWER; otherwise reduce THERMAL POWER to less than i

5% of RATED-THERMAL POWER within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , j

b.

Above 5% of RATED THERMAL POWER, place the inoperable

[

channel in the tripped conaition within I hour; operation may continue until performance of the next required i

CHANNEL FUNCTIONAL TEST.

ACTION 8 -

With the number of OPERABLE channels one less than the Total Number of Channels and with the THERMAL POWER level above P-9, place the inoperable channel in the tripped condition within i

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; operation may continue until performance of the next required CHANNEL FUNCTIONAL TEST.

j ACTION 9 -

This ACTION is not used.

[

An OPERABLE hot leg channel consists of: 1) three RTD's per hot leg, or i

2) two RTD's per hot leg with the failed RTD disconnected and,the required l

bias applied.

I BEAVER VALLEY - UNIT 2 3/4-3-6 AMENDMENT NO. 33 i

~ -..

=..

'l A

i TABLE 3.3-l'(Continued) l ACTION 10 -

This Action is not used.

i ACTION 11 -

With less than the Minimum Number of Channels OPERABLE, operation may continue provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

}

i ACTION 12 -

With the number of channels OPERABLE one less than required by..

the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in H0T STAND 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 39 -

With the number of OPERABLE charat3s one less than the Minimu 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 open the reactor trip breakers within the next hour.

j ACTION 40 -

'a.

With one of the diverse trip features (undervoltage or. shunt trip attachment).of a-reactor trip breaker inoperable, i

restore the diverse trip feature to OPERA 3LE ~ status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Neither breaker shall be i

bypassed while one of the diverse trip features is inoperable except for the time required for perfoming maintenance to restore the breaker to OPERASLE status.

b.

With one reactor trip breaker inoperable as a result of i

something other than an inoperable diverse trip feature be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one, channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per specification 4.3.1.1, provided the other l

channel is OPERABLE.

ACTION 44 -

l With less than the Minimum Number of channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation of the associated permissive l

annunciator window (s) that the interlock is in its required state for the existing plant condition, or app 1v Specification 3.0.3.

(

l l

BEAYER VALLEY - UNIT 2 3/4 3-7 AmendmentNo.JA,33 Ott i s 1988

~

j i

f TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RE'SPONSE TIME 1.

Manual Reactor Trip NOT APPLICABLE

_-0.5. seconds

2.

Power Range, S utron' Flux 3.

Power Range, Neutron Flux, High' Positive Rate.

NOT APPLICABLE i

4.

Power Range, Neutron Flux,.

[

High Negative Rate-

<;0.5 seconds *

.l 5.

Intermediate Range, Neutron' Flux-NOT APPLICf8LE-6.

Sourco hW;g NeutronEFlux NOT APPLICABLE' (Below M f5 7.

Overtemperature AT.-

1 6.0 seconds

8.

Overp.wer AT

< 6.0 seconds *i l

(Above P-7)-

-< 2.0 seconds 9.

Pressurizer Prer,sure--Low'

.i

10. Pressurizer Pressure--High 1 2.0' seconds v

I t

11. Pressurizer Water Level--High NOT APPLICABLE L

(Above P-7) o

12. Less of Flow - Single-Loop

-l (Above P-8) 1 1.0 seconds" j

13. Loss of Flow - Two Loop

< 1.0 seconds l

~

(Above P-7 and below P-8) i

14. Steam Generator Water Level--Low-Low

<:2.0Tseconds

~

(Loop Stop Valves Open)

15. DELETEL.

16. Undervoltage-Reactor Coolant Pumps 5 1.5 seconds (Above P-7)-

17. Underfrequency-Reactor Coolant Pumps'

$.0.9-seconds (Above P-7)

I

Neutron detectors are exempt from response time-testing.

Response

time shall be measured from detector output or input of first electronic component in channel.

h BEAVER VALLEY - UNIT;2 3/4 3-8 Amendment No.

24, 27, 33

.

ML20059N204

Text

Attachment:

Response

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