ML20077A727
| ML20077A727 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 11/18/1994 |
| From: | Bateman W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20077A728 | List: |
| References | |
| NUDOCS 9411250172 | |
| Download: ML20077A727 (24) | |
Text
,93v na%,
e 44 UNITED STATES ye 2
E NUCLEAR REGULATORY COMMISSION
' f WASHINGTON, D.C. 20555-0001 5
sd 9 +****
SOUTH CAROLINA ELECTRIC & GAS COMPANY SOUTH CAROLINA PUBLIC SERVICE AUTHORITY DOCKET NO 50-395 VIRGIL C. SUMMER NUCLEAR STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.120 License No. NPF-12 1.
The iluclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by South Carolina Electric & Gas Company (the licensee), dated
, complies with the standards and requirements of the Atomic f nergy Act of 1954, as amended (the Act), and the commission's rJles and regulations Set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized 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 Commission's regulations; 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; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
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-12 is hereby amended to read as follows:
i 1
l 9411250172 941118 PDR ADOCK 05000395 P
. (2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.120
, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.
South Carolina Electric & Gas Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This amendment is effective as of its date of issuance and shall be implemented within 30 dns of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION App /s[-a'l':?lv.
m liamH.Baeman,yDirector Wi[jectDirectorateII-l Pro Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
November 18, 1994
-s m
ATTACHMENT TO LICENSE AMENDMENT NO.120 TO FACILITY OPERATING LICENSE NO. NPF-12 DOCKET NO. 50-395 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.
The revised pages are indicated by marginal lines.
Remove Paaes Insert P20es 2-4 2-4 2-5 2-5 2-6 2-6 2-7 2-7 2-8 2-8 2-9 2-9 2-10 2-10 B 2-1 B 2-1 B 2-3 B 2-3 B 2-4 B 2-4 B 2-6 B 2-6 3/4 3-15 3/4 3-15 3/4 3-15a 3/4 3-15a 3/4 3-25 3/4 3-25 3/4 3-26 3/4 3-26 3/4 3-27 3/4 3-27 3/4 3-28 3/4 3-28 3/4 3-28a 3/4 3-28a 3/4 3-28b 3/4 3-28b B 3/4 3-1 B 3/4 3-1 B 3/4 3-la B 3/4 3-la
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS RE ACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor trip system instrumentation and interlocks setpoints shall be consistent with the Trip Setpoint values shown in Table 2.2-1.
APPLICA.BILITY: As shown for each channel in Table 3.3-1.
ACTION:
a.
With a reactor trip system instrumentation or interlock setpoint less conservative than the value shown in the Tria Setpoint column of Table 2.2-1 adjust the setpoint consistent wit 1 the Trip Setpoint value.
b.
With the reactor trip system instrumentation or interlock setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement req uirements of Specification 3.3.1 until the channel is restored to OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint Value.
I a
SUMMER - UNIT 1 2-4 Amendment No.120
l TABLE 2.2-1 l
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS v,
Functional Unit Trio Setpoint Allowable Value C
1.
Manual Reactor Trip NA NA
~
2.
Power Range, Neutron Flux High Setpoint
< 109% of RTP
< 111.2% of RTP Low Setpoint
<25% of RTP
< 27.2% of RTP.
i 3.
Power Range, Neutron Flux
<5% of RTP with a time 16.3% ofRTP with a time High Positive Rate constant >2 seconds constant >2 seconds 4.
DELETED 5.
Intermediate Range, Neutron Flux
<25% of RTP
<31% ofRTP 6.
Source Range, Neutron Flux
$105 cps
< 1.4 x 105 cps 7.
Overtemperature AT See note 1 See note 2 8.
Overpower AT See note 3 See note 4 9.
Pressurizer Pressure-Low
>1870 psig
>1859 psig 10.
Pressurizer Pressure-High
$2380 psig
<2391 psig 11.
Pressurizer Water Level-High 192% ofinstrument span 193.8% ofinstrument span 2
12.
Loss ofFlow
.;>_90% ofloop design flow *
>88.9% ofloop design flow
- A l
flow = 94,500 m
Rh desiRAbED THERMAL kOWER
- L l
z C-O
TABLE 2.2-1 (continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 5
Ri t
7 Functional Unit Trio Setpoint Allowable Value f
E 13.
Steam Generator Water Z
Level Low-Low Barton Transmitter
> 27.0% ofspan
>26.1% ofspan Rosemount Transmitter
>27.0% of span
>25.7% of span 14.
Steam /Feedwater Flo;v Mis-540% offull steam flow at RTP
$42.5% offull steam flow at RTP l
Match Coincident With Steam Generator Water Level Low-Low Barton Transmitter
>27.0% of span 126.1% of span Rosemount Transmitter
>27.0% of span
>25.7% of span 15.
Undervoltage-Reactor
>4830 volts
>4760 volts
'?
Coolant Pump cn 16.
Underfrequency - Reactor
>57.5 Hz
> 57.1 Hz Coolant Pumps 17.
Turbine Trip A. Low Trip System Pressure
>800 psig
> 750 psig B. Turbine Stop Valve Closure
>1% open
> 1%.open E
an 5
RTP - RATED THERMAL POWER e
=
O
_ -..,,. ~. _ -.. -
c, a
1 TABLE 2.2-1 (continued)
RE ACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS m
EM 4;
x Functional Unit Trio Setpoint Allowable Value 18.
Safety Injection Input from ESF NA NA 19.
Reactor Trip System Interlocks
~
A. Intermediate Range Neutron Flux, P-6 27.5 X 10A% indication 24.5 X 10d% indication B. Low Power Reactor Trips i
Block, P-7
- a. P-10 input 110% ofRTP 512.2% of RTP
- b. P-13 input 110% turbine impulse pressure 112.2% of turbine impulse equivalent pressure equivalent O
C. Power Range Neutron Flux P-8 138% ofRTP 140.2% ofRTP t
D. Low Setpoint Power
[
Range Neutron Flux, P-10 210% ofRTP 27.8% ofRTP l
E. Turbine Impulse Chamber
$10% turbine impulse pressure 112.2% turbine pressure Pressure, P-13 equivalent equivalent i
F. Power Range Neutron Flux, P-9
<50% of RTP
<52.2% of RTP 20.
Reactor Trip Breakers NA NA k
21.
Automatic Actuation Logic NA NA.
E F
RTP-RATED TIIERMAL POWER F
i
~
l
-. - ~,.
-- j
t g
TABLE 2.2-1 (continued) j REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS t
7 NOTATION NOTE 1: OVERTEMPERATURE AT
~
~
(1 + I S) g
+ K, ( P - P') - f,(61)
AT s AT K-K T-T o
1 2 (1 + t 8) 2 Measured AT by RTD Instrumentation Where:
AT
=
AT, s
Indicated AT at RATED THERMAL POWER K
s 1.23 l
1 K
2 0.0292/*F l
2 7
1+IS The funtion generated by the lead-lag controller for T
=
1 + t,s dynamic compensation Time constants utilized in lead-lag controller for T,,,,1 2 28 secs.,
=
t,t2 i
1 1 s4 secs.
2 Average temperature, *F T
=
T' s
Indicated T,,, at RATED THERMAL POWER,572.0*F s T's 587.4*F l
g K
2 0.00161/ psi l
3 k
P
=
Pressurizer pressure, psig P'
2 2235 psig, Nominal RCS operating pressure
[
S Laplace transforrr. operator, sec-1
=
b*x m-8 r -- -.
1- --
A TABLE 2.2-1 (continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS h
NOTATION (continued) 5 NOTE 1: (Continued)
Z and f ( AI) is a function of the indicated difference between top and bottom detectors of the power-range
~
t nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:
(i) for q,- q, between -35 percent and + 6 percent f (AI) = 0 wheie q, and q, are percent RATED THERMAL l
1 POWER in the top and bottom halves of the core respectively, and q, + q, is total THERMAL POWER in percent of RATED THERMAL POWER.
(ii) for each percent that the magnitude of q,- q, exceeds -35 percent, the AT trip setpoint shall be l
automatically reduced by 2.46 percent ofits value at RATED THERMAL POWER.
l 7
(iii) for each percent that the magnitude of q,- q, exceeds +6 percent, the AT trip setpoint shall be j
automatically reduced by 3.29 percent ofits value at RATED THERMAL POWER.
l NOTE 2: The channel's maximum trip setpoint shall not exceed its computed trip point by more than 2.2 percent AT Span.
NOTE 3: OVERPOWER AT l
(1 S 3
AT s AT K-K T-K T-T o
4 5(1 + t S) 6 3
as defined in Note 1 Where: AT
=
as defined in Note 1 AT
=
a g
K, s
1.078 l
K 2
0.02/*F for increasing average temperature and 0 for decreasing average temperature 3
t S h/
The function generated by the rate-lag controller for T,ydynamic
=
ff 1 + 'a S compensation a
w' T
m w
m TABLE 2.2-1 (continued)
E REACTOR TRIP SYSTEM INSfRUMENTATION TRIP SETPOINTS 3,g NOTATION (continued) g NOTE 3: (continued)
}
Time constant utilized in rate-lag controlbr for T,,,, t 210 secs.
l t
=
a 3
K 2
0.00198/"F for T > T and K, = 0 for T s T i
3 T
as defined in Note 1
=
7 s
Indicated T,,, at RATED THERMAL POWER,572.0*F s Ts 587.4 F 1
as defined in Note 1 y
S
=
~o NOTE 4: The channel's maximum trip setpoint shall not exceed its computed trip point by more than 2.3 percent AT Span.
l SE i
li a
E O
h
4 2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented ay restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could from nucleate boiling (DN$v temperatures because of the onset of departure result in excessive claddin ) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore TIIERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB. This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.
The local DNB heat flux ratio (DNBR) defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.
The DNB design basis is as follows: there must be at least a 95 percent probability that the minimum DNBR of the limiting rod during Concition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used. The correlation DNBR limit is established based on the entire applicable l
experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit.
[
In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95 percent probability with 95 percent confidence level that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit. The uncertainties in the above plant parameters are used to determine the olant DNBR uncertainty. This DNBR uncertainty, combined with the correlation DNBR limit, establishes a design DNBR value which must be met in plant safety analyses using values ofinput parameters without uncertainties.
In addition, margm has been maintained in the design by meeting safety analysis DNBR limits in performing safety analyses.
The curves of Figure 2.1-1 show the loci of points of THERM AL POWER, Reactor Coolant System pressure and average temperature below which the calculated DNBR is no less than the design DNBR value or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid.
SUMMER - UNIT 1 B 2-1 Amendment No. 75., N4,120 l
l
)
2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal values at which the Reactor Trips are set for each functional unit. The Trip Set aoints have been selected to ensure that the reactor core and reactor coo: ant system are prevented from exceeding ideir safety limits during normal operation and design basis anticipated operational occurrences and to assist the En ineered Safety Features Actuation System in mitigating the consequences of accic;ents. The setpoint for a reactor trip system or interlock function is considered to be adjusted consistent with the nominal value when the "as measured" setpoint is within the band allowed for calibration accuracy.
To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which setpoints can be measured and calibrated, Allowable Values for the reactor trip setpoints have been specified in Table 2.2-1. Operation with setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error.
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.
SUMMER - UNIT 1 B 2-3 Amendment No. E. 120
LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)
The various reactor trip circu,its automatically open the reactor trip breakers whenever a condition monitored by the Reactor Protection System reaches a preset or calculated level. In addition to redundant channels and trains, the design approach provides a Reactor Protection System which monitors numerous system variables, therefore, providing protection system functional diversity. The Reactor Protection System initiates a turbine trip signal whenever reactor trip is initiated. This prevents the reactivity insertion that would otherwise result from excessive reactor system cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System.
Manual Reactor Trip The Reactor Protection System includes manual reactor trip capability.
Power Range, Neutron Flux In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a high and low range trip setting. The low setpoint trip ?rovides protection during suberitical and low power operations to mitigate t ie consequences of a power excursion beginning from low power, and the high setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from all power levels.
The low setpoint trip may be manually blocked above P-10 (a power level of approximately 10 percent of RATED THERM AL POWER) and is automatically reinstated below the P-10 setpoint.
Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing.
Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from mid-power.
Intermediate and Source Range, Nuclear Flux The Intermediate and Source Range, Nuclear Flux trips provide reactor core protection during reactor startup to mitigate the consequences of an SUMMER - UNIT 1 B 2-4 Amendment No. 7s, I%,120
LIMITING SAFETY SYSTEM SETTINGS BASES Pressurizer Pressure (Continued)
On decreasing power the low setpoint trip is automatically blocked by P-7 (a power level of approximately 10 percent of RATED THERMAL POWER with turbine impulse chamber pressure at approximately 10 percent of full power equivalent);
and on increasing power, automatically reinstated by P-7.
The high setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure.
Pressurizer Water Level The pressurizer high water level trip is provided to prevent water relief through the pressurizer safety valves. On decreasing power the pressurizer high water level trip is automatically blocked by P-7 (a power level of approximately 10 percent of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10 percent of full equivalent); and on increasing power, automatically reinstated by P-7.
Loss of Flow The Loss of Flow trips provide core protection to prevent DNB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.
On increasing power above P-7 (a power level of approximately 10 percent of RATED THERM AL POWER or a turbine impulse chamber pressure at approximately 10 percent of full power equivalent), an automatic reactor trip will occur if the flow in more than one loop drops below 90% of nominal full loop flow. Above P-8 (a power level of approximately 38 percent of RATED THERM AL POWER) an automatic reactor trip will occur if the flow in any single loop drops below 90 percent of nommal full loop flow. Conversely on decreasing ?ower between P-8 and the P-7 an automatic reactor trip will occur on loss of flow in more than one loop and below P-7 the trip function is automatically blocked.
Steam Generator Water Level The steam generator water level low-low trip protects the reactor from loss of heat sink m the event of a sustained steam /feedwater flow mismatch resulting from loss of normal feedwater. The specified setpoint provides allowances for starting delays of the auxiliary feedwater system.
Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level The steam /feedwater flow mismatch in coincidence with a steam generator low water level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall reliability of the Reactor Protection System. This trip is redundant to the Steam Generator Water Level Low-Low trip. The Steam /Feedwater Flow Mismatch portion of this trip is activated when the steam flow exceeds the feedwater flow by greater than or equal to 40% of full steam flow at RTP. The Steam Generator Low Water level portion of the trip is activated when the water level drops below the low SUMMER - UNIT 1 B 2-6 Amendment No. he,120
INSTRUMENTATION 3/4 3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5.
APPLICABILITY: As shown in Table 3.3-3.
ACTION-a.
With an ESFAS Instrumentation or Interlock Trip Setpoint trip less conservative than the value shown in the Trip Setpoint Column but more conservative than the value shown in the Allowable Value Column of Table 3.3-4, adjust the betpoint consistent with the Trip Setpoint
- value, b.
With an ESFAS Instrumentation or Interlock Trip Setpoint less conseiva-tive than the value shown in the Allowable Value Column of Table 3.3-4, declare the channelinoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel is restored to its OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint value.
c.
With an ESFAS instrumentation channel or interlock inoperable take the ACTION shown in Table 3.3-3.
SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by performance of the engineered safety feature actuation system instrumentation surveillance requirements specified in Table 4.3-2.
4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME ofeach ESFAS function shall be demonstrated to be within the limit at least once per 18 months.
Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once per N times 18 months where N is the total number of redundant channels in a saecific ESFAS function as -Sown in the " Total No. of Channels" Column of Tab e 3.3-3.
SUMMER - UNIT 1 3/4 3-15 Amendment No. N N,120
1 THIS PAGE TO BE DELETED DUE TO REPAGINATION i
SUMMER - UNIT 1 3/4 3-15a Amendment No.1s, IM,120
TABLE 3.3-4 ENGINEEREDSAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E
Functional Unit Trio Setpoint Allowable Value t.
SAFETY INJECTION, REACTOR TRIP, E
FEEDWATER ISOLATION, CONTROL Z
ROOM ISOLATION, START DIESEL GENERATORS, CONTAINMENT
~'
COOLING FANS AND ESSENTIAL SERVICE WATER.
- a. ManualInitiation NA NA
- b. Automatic Actuation Logic NA NA c.
Reactor Building Pressure-fligh 1 53.6 psig s3.86 psig
{
- d. Pressurizer Pressure--Low 21850 psig 21839 psig
'r' e.
Differential Pressure s97 psig
$106 psi U
Between Steamlines-IIigh f.
Steamline Pressure--Low 2675 psig 2635 psigu) 2.
REACTOR BUILDING SPRAY
- a. ManualInitiation NA NA I.
Automatic Actuation Logic NA NA
[
and Actuation Relays h
c.
Reactor Building Pressure-s 12.05 psig s 12.31 psig
}
High 3 (Phase 'A' isolation aligns spray system discharge 2
valves and NaOII tank suction y
valves)
I (1) Time constants utilized in lead lag controller for steamline pressure-low are as follows:
Il 2 50 secs.
I2 s 5 secs.
TABLE 3 3-4 (Continued)
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM I STRUMENTATION TRIP SETPOINTS E
k Functional Unit Trip Setpoint Allowable Value w
3.
CONTAINMENT ISOLATION
[
a.
Phase "A" Isolation 1.
Manual NA NA 2.
Safety Injection See 1 above for all safety injectian setpoints See 1 above for all allowable values 3.
Automatic Actuation Logic NA NA and Actuation Relays b.
Phase "B" Isolation 1.
Automatic Actuation Logic NA NA w1 and Actuation Relays m
L 2.
Reactor Building
< 12.05 psig
< 12.31 psig Pressure-Iligh 3 c.
Purge and Exhaust Isolation 1.
Safety Injection See 1 above for all safety injection setpoints See 1 above for all allowable values 2.
Containment Radioactivity liigh ro 3.
Automatic Actuation Logic NA NA l
and Actuation Relays i
5 1
b
- Trip setpoints shall be set to ensure that the limits of ODCM Specification 1.2.2.1 are not exceeded.
M O
TABLE 3 3-4 (Continued)
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS
- E Functional Unit Trip Setpoint Allowable Value g
4.
STEAM LINE ISOLATION
[
a.
Manual NA NA b.
Automatic Actuation Logic NA NA and Actuation Relays c.
Reactor Building Pressure-
< 6.35
< 6.61 liigh 2 d.
Steam Flow in Two Steamlines-Sa function defined as follows: A Ap na function defined asfollows: A Ap iligh, Concident with corresponding to 40% of full steam flow corresponding to 44% of full steam now between 0% and 20% load and tFen a Ap between 0% and 20% load and then a Ap w
increasing linearly to a Ap corresponding to increasing linearly to a Ap corresponding to Y
110% offull steam flow at fullload 114.0% offull steam flow at fullload U
T vg - Low-Low
> 552.0*F
>548 4*F a
e.
Steamline Pressure-Low 2675 psig
> 635 psig41) 2 Er an E
l
[
(1) Time constants utilized in lead lag controller for steamline pressure low are as follows:
In 2 50 secs.
I2 s 5 secs.
E$
I
TABLE 3.3-4 (Continued)
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E
4 m
m Functional Unit Trio Setpoint Allowable Value E
5.
TURBINE TRIP AND FEEDWATER Z
ISOLATION a.
Steam Generator Water Level-Iligh-liigh Barton Transmitter
< 79.2% of span
< 81.0% of span Rosemount Transmitter
_ 79.2% of span
< 81.0% of span 6.
EMERGENCY FEEDWATER a.
Manual NA NA
{
b.
Automatic Actuation Logic NA NA y
c.
Steam Generator Water S$
Level - Low-Low Barton Transmitter 127.0% of span 226.1% of span Rosemount Transmitter
> 27.0% of span
> 25.7% of span
- d. & f.
Undervoltage-ESF Bus
>5760 Volts with a <0.25 second
>5652 Volts with a <0.275 second time time delay delay
> 6576 Volts with a <3.0 second
> 6511 Volts with a <3.3 second time delay p
time delay i
an O
O
TABLE 3 3-4 (Continued)
ENGINEERED S.* FETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS ExM Functional Unit Trio Setpoint Allowable Value
[
e.
Safety injection See 1 above (all SI Setpoints)
See 1 above (all SI Setpoints)
=
[
g.
Trips of Main Feedwater Pumps NA NA
~
- h. Suet. ion transfer on Low
> 44 2 ft. 4 in. (25
> 441 fL 3 in.
Pressure 7.
LOSS OF POWER a.
7.2 kv Emergency Bus Undervoltage
> 5760 volts with a < 0.25 second 25652 volts with a10.275 second time (Loss of Voltage) time delay delay b.
7.2 kv Emergency Bus Undervoltage
> 6576 volts with a 13.0 second
> 6511 volts with a <3.3 second time delay R
time delay u
y 8.
AUTOMATIC SWITCIIOVER TO CONTAINMENT SUMP g
a.
RWST Level Low-Low
>18%
1 15 %
b.
Automatic Actuation Logic NA NA l
and Actuation Relays E
S.
(2) Pump suction head at which transfer is initiated is stated in efTective water elevation in the condensate storage tank.
2 M
O
TABLE 3.3-4 (Continued)
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS
$5 Functional Unit Trip Setooint Allowable Value g
9.
ENGINEERED SAFETY FEATURE
[
ACTUATION SYSTEM INTERLOCKS
~
INTERLOCKS a.
Pressurizer Pressure, P-11 1985 psig
> 1974 psig &
11996 psig b.
T ve Low-Low, P-12 552*F
> 548.4*F &
a iS55.G*F c.
Reactor Trip, P-4 NA NA R
u Y
M er a
R 2n O
O
3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Protection System and Engineered Safety Feature Actuation System Instrumentation and interlocks ensure that 1) the associated action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoints,2) the specified coincidence logic and sufTicient redundancy is maintained to permit a channel to be out of service for testing or maintenance consistent with maintaining an appropriate level of reliability of the Reactor Protection and Engineered Safety Features instrumentation and,3) suflicient system functions capability is available from diverse parameters.
The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions.
The mtegrated operation of each of these systems is consistent with the assumptions used in the accident analyses. The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance testa performed at the minimum frec uencies are sufficient to demonstrate this capability. Specified surveillance and surveillance and maintenance outage times have been determined in accordance with WCAP-10271, " Evaluation of Surveillance Frequencies and Out of Service Times for Reactor Protection Instrumentation System," and supplements to that report. Surveillance intervals and out of service times were determined based on maintaining an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation.
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The Engineered Safety Feature Actuation System Instrumentation Trip Setpoints specified in Table 3.3-4 are the nominal values at which the bistables are set for each functional unit. A setpoint is considered to be adjusted consistent with the nominal value when the "as measured" setpoint is within the band allowed for calibration accuracy.
To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which setpoints can be measured and calibrated, Allowable Values for the setpoints have been specified in Table 3.3-4.
Operation with setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error.
l 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. Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertaint,y 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 SUMMER - UNIT 1 B 3/4 3-1 Amendment No. TM, 120
INSTRUMENTATION BASES REACTOR TRIP AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION (continued)
I 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.
The measurement of response time at the specified frequencies provides assurance that the reactor trip and the engineered safety feature actuation associated with each channel is completed within the time limit assumed in the accident analyses. No credit was taken in the analyses for those channels with response times indicated as not applicable. Response time may be demon-strated by anj series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite, or offsite test measurements or 2) utilizing replacement sensors with certified response times.
The Engineered Safety Features response times specified in Table 3.3-5 which include sequential operation of the RWST and VCT valves (Notes 2 and 3) are based on values assumed in the non-LOCA safety analyses. These analyses are for injection of borated water from the RWST. Injection of borated water is assumed not to occur until the VCT charging pump suction isolation valves are closed following opening of the RWST charging pumps suci!on valves. When the sequential o response times (peration of the RWST and VCT valves is not inciuded in the Note 1) the values specified are based on the LOCA analyses.
The LOCA analyses take credit for injection flow regardless of the source.
Verification of the response times specified in Table 3.3-5 will assure that the assumptions used for the LOCA and non-LOCA analyses wRh respect to the operation of the VCT and RWST valves are valid.
The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents, events, and transients. Once the required logic combination is completed, the system sends actuation si pials to those engineered safety features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss of coolant accident 1) safety 1
injection pumps start and automatic valves position,2) reactor trip,3) feed-l water isolation,4) startup of the emergency diesel generators,5) containment spray pumps start and automatic valves position,6) containment isolation,
- 7) steam line isolation,8) turbine trip,9) auxiliary feedwater pumps start and automatic valves position,10) containment cooling fans start and auto-matic valves position,11) essential service, water pumps start and automatic valves position, and 12) control room isolation and ventilation systems start.
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