ML20059A954
| ML20059A954 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 12/16/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20059A941 | List: |
| References | |
| NUDOCS 9401030221 | |
| Download: ML20059A954 (17) | |
Text
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UNITED STATES J!
t NUCLEAR REGULATORY COMMISSION l
E WASHINGTON. D. C. 205SS
%.....+
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION' l
-i RELATED TO AMENDMENT NO. 44 TO FACILITY OPERATING LICENSE NO. NPF-72' l
7 AND AMENDNENT NO. 44 TO FACILITY OPERATING LICENSE NO. NPF-77 l
COMMONWEALTH EDISON COMPANY-BRAIDWOOD STATION. UNIT NOS. 1 AND 2 DOCKET NOS. STN 50-456 AND STN 50-457 1
1.0 INTRODUCTION
By letter dated August 5,1992, Commonwealth Edison Company-(CECO) proposed changes to the Braidwood Station, Unit Nos. I and 2, Technical Specifications (TS) for the Reactor Trip System (RTS) and Engineered Safety Features Actuation System (ESFAS) instrumentation (Reference 1). The proposed changes for the RTS and ESFAS provide relief by reducing the frequency of surveillance testing, increasing the allowable time to place an inoperable channel on trip 1
and/or hypassed condition, and increase the allowable time for surveillcnce and maintenance. This change would reduce inadvertent reactor trips, which cause unnecessary plant transients and challenges to the protection system.
Increasing the time to place an inoperable channel in test and/or bypass and increasing the outage time for maintenance should reduce human errors
.i associated with short time restraints.
i
2.0 BACKGROUND
The Westinghouse Owners' Group (WOG) was requested to evaluate-their TS in accordance with the recommendations in NUREG-1024, " Technical Specification-Enhancing the Safety Impact" Task Group on Technical Specifications dated November 1983.
By letter dated February 3,1983, the WOG submitted Topical Report WCAP-10271, " Evaluation of Surveillance Frequencies-and.0ut of Service
- Tiines _for the Reactor Protection Instrumentation System". (Reference 2). On October 4,1983, WOG submitted Supplement 1-to WCAP-10271 to the NRC for review and-included responses to. staff questions submitted (Reference 3).
.The staff completed the review of Topical Report WCAP-10271 and issued a Safety Evaluation-(SE) on February 21, 1985 (Reference 4). 'The' staff concluded in the SE that the Topical: Report was acceptable regarding surveillance intervals for RTS analog ' channel operational tests, time which an inoperable RTS analog channel may be maintained in a non-tripped condition',
and time the inoperable RTS channel may be bypassed.
By_ letters dated-March 20, 1986, and May 12, 1987, WOG submitted Topical Reports WCAP-10271, Supplement 2, and WCAP-10271, Supplement 2, Revision 1, j
i 9401030221 931216 I
. proposing extensions of surveillance test intervals (STI) and test and maintenance allowed outage times (A0T) for the ESFAS (References 5 and 6).
On February 22, 1989, the staff completed review of the Topical Report and issued an SE.
The staff concluded in the SE that the analysis presented in WCAP-10271, Supplement 2, and WCAP-10271, Supplement 2, Revision 1, augmented by a Brookhaven National Laboratory Technical Evaluation Report (TER) are acceptable for resolving the STI and A0T extension issues. However, the staff also required that the licensee confirm the applicability of the generic analysis to their facility and confirne that any increase in instrument drift due to the extended STI is accounted for in the Setpoint Calculation Methodology (Reference 7).
By letter dated April 30, 1990, the staff issued the supplemental SE of Topical Report WCAP-10271, Supplement 2, Revision 1.
The staff concurred with the evaluatica of the Topical Report by the Brookhaven National Laboratory; however, the staff also concluded that STI and A0T extensions proposed in the Topical Report for trip breakers are not acceptable (Reference 8).
3.0 LVALVATION CECO proposed to make the following changes to the Braidwood Station, Unit Nos. I and 2, TS.
Changes are underlined, bold and in brackets [ ).
Clarification are in parentheses ( ):
Table _L3-1 " REACTOR TRIP INSTRUMENTATION" 3.3-1(1)
Note # [The provisions of Specification 3.0.4 are not applicable.)
This notation is deleted from the following Functional Unit Action Statements:
g 2
Power Range, Neutron Flux 3
Power Range, Neutron Flux High Positive Rate 4
Power Range, Neutron Flux High Negative Rate 7
Over-Temperature Delta T 8
Overpower Delta T 9
Pressurizer Pressure-Low (Above P-7) 10 Pressurizer Pressure-High 11 Pressurizer Water Level-High (Above P-8) 12 Reactor Coolant Flow-Low 13 Steam Generator Water Level-Low-Low 14 Under-Voltage-Reactor Coolant Pumps (Above P-7) 15 Under-Frequency-Reactor Coolant Pumps (Above P-7) 16 Turbine Trip (Above P-7 or P-8)
The note is deleted because it is no longer applicable.
. 3.3-1(2)
Note ** [The boron dilution flux doublino sionals may be blocked durino reactor startup2] Notation is deleted from the following Functional Unit Applicable Modes:
6 Source Range, Neutron Flux a. Startup The Boron Dilution Protection System (BDPS) is intended to mitigate the consequence of accidents which result in decrease in the reactor coolant system baron concentration. This note was previously deleted as part of Amendment No. 40 to the Braidwood TS dated October 5,1992 (Reference 9).
3.3-1(3)
Note *** [These channels also provide inputs to ESFAS. The Action itatement for the channels in Table 3.3-3 is more conservative artd. therefore. controllinc.) Notation is deleted from the following Functional Unit Action Statements:
9 Pressurizer Pressure-Low (Above P-7) 13 Steam Generator Water level-Low-Low 14 Undervoltage-Reactor Coolant Pumps (Above-P-7)
This note is deleted because the Action Statement for these Functional Units in Table 3.3-3, " Engineered Safety Features Actuation System Instrumentation
- is changed to agree with those in Table 3.3-1; therefore, the note is no longer applicable.
3.3-1(4)
Note **** [A reactor trip on turbine triD is enabled above P-7 (10%1 until the modification is implemented which enables reactor trip on turbine trip above P-8 (30%).] Notation is deleted from the following Functional Unit and [Above P-7) is also deleted:
16 Turbine Trip The Braidwood, Units 1 and 2, Final Safety Analysis Report (FSAR)
Section 10.4.4, " Steam Dump (or Bypass) System" states:
"a reactor power level step change of 10% can be handled by the reactor control system without causing a reactor trip and the condenser steam dump (bypass) can handle a 40% power reduction; therefore, the plant can handle a 50% turbine load rejection without a reactor trip."
This note is deleted because of a modification where the reactor trip on a turbine trip occurs only when power is above P-8 (30%).
On September 30, 1987, the licensee requested an amendment to the facility Operating License for the above modification (Reference 10). The NRC staff requested CECO to address a concern regarding the potential increase in probability of a stuck-open pressurizer power relief valve as a result of implementing the above modification. The licensee responded to the NRC concern by letter dated October 30, 1987 (Reference 11).
The staff accepted the
~
. above modification in the Safety Evaluation for Amendment No.13 issued December 8, 1987 (Reference 12). Since the modification has been completed Note **** is no longer needed and is deleted.
3.3-1(5)
Action Statement 9 is changed and is applicable to the following Functional Units:
17 Safety Injection Input from ESF 21 Automatic Trip and Interlock Logic Action Statement 9:
With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, [ restore the inoperable channel to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. or) be in at least HOT STANDBY within [the next) 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to [4] hours for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE.
This change agrees with the NRC's SE (Reference 4) conclusion.
3.3-1(6)
The following Functional Unit Action Statement 11 is changed:
18 Reactor Coolant Pump Breaker Position Trip Action Statement 11:
With the number of OPERABLE channels less than the Total Number of Channels, operation may continue provu?d the ir. operable channels are placed in the tripped condition within [f] hours.
This change agrees with the NRC's SE (Reference 4) conclusion.
l Table 4.3-1 " REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RE0UIREMENTS" 4.3-1(1)
Note ** [These channels also provide lnouts to ESFAS. The Operational Test Frecuency for these channels in Table 4.3-2 is more conservative and. therefore. controllino].- Notation is deleted from the Analog Channel Operational Test of the following Functional Units:
9 Pressurizer Pressure-Low (Above P-7) 13 Steam Generator Water level-Low-Low The Note ** is also removed from the Trip Actuating Device Operational Test for the following Functional Unit:
14 Undervoltage-Reactor Coolant Pumps (Above-P-7) i
This note is deleted because the operational test frequency for these functional Units is being changed in Table 4.3-2 from Monthly to Quarterly and now agree with Table 4.3-1; therefore, it is no longer applicable.
4.3-1(2)
Note # [The specified 18 month interval may be extended to 32 months for cycle 1.] Notation is deleted from Channel Calibration of the following Functional Units:
2 Power Range, Neutron Flux a.
High Setpoint; b. Low Setpoint 3
Power Range, Neutron Flux High Positive Rate 4
Power Range, Neutron Flux High Negative Rate 5
Intermediate Range, Neutron Flux 6
Source Range, Neutron Flux 7
Overtemperature Delta T 8
Overpower Delta T t
9 Pressurizer Pressure-Low (Above P-7) 10 Pressurizer Pressure-High 11 Pressurizer Water Level-High (Above P-7) 12 Reactor Coolant Flow-Low 13 Steam Generator Water Level-Low-Low 19 Reactor Trip System Interlocks a.
Intermediate Range Neutron Flux, P-6 b.
Low Power Reactor Trip Block, P-7 c.
Power Range Neutron Flux, P-8 d.
Low Setpoint Power Range Neutron Flux, P-10 e.
Turbine Impulse Chamber Pressure, P-13 This note is deleted because it reflects a cycle-specific relief that is no longer applicable.
4.3-1(3)
Note 1 is changed for Trip Actuating Device Operational Test for the following Functional Units:
16 Turbine Trip a.
Emergency Trip Header Pressure b.
Turbine Throttle Valve Closure Note 1:
If not performed in previous [11] days.
This change agrees with the NRC's SE (Reference 4) conclusion.
4.3-1(4)
Note 3 for Monthly Channel Calibration for the following Functional Unit:
2 Power Range, Neutron Flux "a. High Setpoint"
l
. Note 3:
[The initial sinale point comparison'of incore to excore AXIAL FLUX DIFFERENCE following a refueline outace shall be Derformed prior to exceedina 75% of RATED THERMAL POWER. Otherwise the]
[1) ingle point comparison of incore to excore AXIAL FLUX DIFFERENCE [shall be performed) above 15% of RATED THERMAL POWER.
Recalibrate if the absolute difference is greater than or equal to.
3%. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1. [For the purpose of this surveillance.
ponthly shall mean at least once per 31 EFPD (effective full power.
cay)..The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> completion time Drovisions of SDecification 4.0.3 are not aDDlicable.]
The first change allows comparison to be made prior to exceeding 75% RTP. This measurement is generally taken with the core at or near steady state conditions. The definition of monthly, meaning 31 EFPD is consistent with the NRC's SER, (Ref.4).
4.3-1(5)
Note 6 is changed for Quarterly Channel Calibration test for Functional Unit:
2 Power Range, Neutron Flux a. High Setpoint Note 6:
Incore - Excore Calibration, above 75% of RATED THERMAL POWER.
The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
[For the purpose of this surveillance.
cuarterly shall mean to least once per 92 EFPD.)-
The definition of Quarterly meaning 92 EFPD is. consistent with the.
NRC's SER, (Ref 4.)
4.3-1(6)'
Note 8 has been deleted from the Analog Channel Operational Quarterly Test for the following Functional Unit:
j 19 Reactor Trip System Interlocks t
b.
Low Power Reactor Trip Block, P-7 c.
Power Range Neutron Flux, P-8 d.
Low Setpoint Power Range Neutron Flux, P-10 e.
Turbine Impulse Chamber Pressure, P-13 Note 8-
[With Dower-creater than or Gaual to the interlock SetDoint the reauired ANALOG CHANNEL OPERATIONAL TEST shall consist of verifyina that the interlock is in the recuired state by observino the permissive annunciator window.)
'i
. This note has been removed because it is no longer used.
4.3-1(7)
Analog Channel Operational Test frequency is changed from Quarterly to [Refuelino Outace] for the following Functional Units:
19 Reactor Trip System Interlocks Intermediate Range Neutron Flux, P-6 b.
Low Power Reactor Trip Block, P-7 c.
Power Range Neutron Flux, P-8 d.
Low Setpoint Power Range Neutron Flux, P-10 e.
Turbine Impulse Chamber Pressure, P-13 This Note is changed because it has been generically approved in NRC's SE (Reference 4).
4.3-1(8)
Note 9 is changed for Analog Channel Operational Test, for Functional Unit:
6 Source Range, Neutron Flux Note 9:
Surveillance in MODES 3*, 4*, and 5* shall also include verification that permissive P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive window. Delete the following:
[ Surveillance shall include verification of the Boron Dilution Alarm Setooint of less than or eaual to an increase of twice the count rate within a 10-Einute period.)
This note was previously changed as part of Amendment No. 51 to the Braidwood TS dated October 5, 1992.
4.3-1(9)
Note 12 is deleted for Channel Calibration of Functional Unit:
6 Source Range, Neutron Flux Note 12:
[At least once per 18 months durino shutdown verify that on a simulated Boron Dilution Doublino test sional CVCS valves ll2D and E open and 1128 and C close within 30 seconds.]
This note was previously deleted as part of Amendment No. 51 to the Braidwood TS dated October 5, 1992.
4.3-1(10)
Note 14 is changed for Trip Actuating Device Operational Test of Functional Unit:
l i
t L t
J l
Manual Reactor Trip Note-14:
q Verify that the appropriate signals reach the Undervoltage and Shunt Trip Relays, for both the Reactor Trip _ and Bypass Breakers
(
from the manual Trip Switches. Delete the following:
[ Initial performance of this surveillance recuirement for Reactor Tr1D gypass Breakers is to be completed prior to the startuD folloWinG
~
the Unit 1. Cycle 1 Refuel Outaae.]
The Note is changed because the cycle-specific relief is no longer applicable.
4.3-1(11)
Note 16 is changed for Trip Actuating Device Operational Test of Functional Unit:
22 Reactor Trip Bypass Breakers Note 16:
l Automatic Undervoltage Trip.
Delete the following:
[ Initial performance of this surveillance reauirement is to be completed 4
prior to the startup followino the Unit 1. Cycle 1 Refuelino Outace2]
The Note is changed because the cycle-specific relief is no longer applicable.
l Table 3.3-3 " ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION" 3.3-3(1)
Functional Unit 9d, Steam Generator Water Level.-High-High, is i
deleted because it is a duplication of Functional Unit 5b.
3.3-3(2)
Note * [The provisions of Specification 3.0.4 are not applicable.)
Notation has been deleted from the Action Statement of the
)
following Functional Units:
Ic Containment Pressure-High-1_
Id Pressurizer Pressure-Low (above P-11) le Steam Line Pressure-Low (above P-11) 4c Containment Pressure-High-2 4d Steam Line Pressure-Low (above P-11) 4e Steam Line Pressure-Negative-P. ate-High (below P-II) l Sb Steam Generator Water Level-High-High (P-14) j 6g Auxiliary Feedwater Pump Suction Pressure-Low l
8a ESF Bus Undervoltage 8b Grid Degraded Voltage t
-r d
i cf
t
_g_
The Pete is deleted because it does not agree with the NRC's SER, (Ref. 8.)
3.3-3(3)
The following Functional Units Action Statement is changad from 15 to 19:
Ic Containment Pressure-High I le Steam Line Pressure-Low (above P-II) 4c Containment Pressure-High 2 4d Steam Line Pressure-Low (above P-11) 4e Steam Line Pressure-Negative Rate-High (below P-II)
Action Statement 15:
With the number of OPERABLE channels one less than the Total Number of Channels, operation may precede until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the trip condition within I hour.
Action Statement 19:
With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are met:
a.
The inoperable channel is placed in the tripped condition within [ ! ) hour [s] and b.
The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to [ & ] hours for surveillance testing of other channels per Specification 4.3.2.1.
This change is made because Action Statement 19 is in agreement with NRC's SE (Reference 8).
3.3-3(4)
Functional Unit 6g, Auxiliary Feedwater Pump Suction' Pressure-Low, (Transfer to Essential Service Water) Total Number of Channels, Channels to Trip, and Minimum Channels Operable is changed from 2 to [1/ Train). Also the Action Statement is changed from 15 to-15a.
Action Statement 15:
With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed until performance of i
the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the trip condition within I hour.
i Action Statement 15a:
. [With the number of OPERABLE channels one less than the Total No.
of Channels. declare the associated Dumo INOPERABLE and take the ACTION reauired by Specification 3.7.1.2.)
Functional Unit 6g number of channels has been changed to agree with the as-built plant configuration of one pressure transmitter installed at the AF pump suction.
Action Statement 15 requires the inoperable channel be placed in a trip condition until the next required ANALOG CHANNEL OPERATION TEST. Although, a ESFAS signal is also required to cause the AF pump suction to transfer to the Essential Service Water System (SX) this armed condition could be present for up to a month.
Therefore, should the AF pump receive a ESFAS signal there would -
be an inadvertent injection of SX water into the steam generators (SGs). The untreated SX water injection into the SGs would have a potential for causing future tube leaks or rupture. Therefore, Action Statement 15a requires that pump be declared inoperable with a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> A0T.
3.3-3(5)
Action Statement for the following Functional Unit is changed from 16 to 15:
7b RWST Level-Low-Low Coincident with Safety Injection (Auto opening of Containment Sump Suction Isolation Valves)
Action Statement 16:
With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is met. One additional channel may be bypassed for up to [4] hours for surveillance testing per Specification 4.3.2.1.
Action Statement 15:
With the number of OPERABLE channels one less than the Total Number of Channels, operation may precede until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the trip condition within [ f ]
hour [ 1 ].
1 This change to Functional Unit 7b, RWST-Low-Low Coincident with Safety Injection, Action Statement from 16 to 15 is proposed because the current design does not permit bypass of an inoperable l
channel which is required by Action Statement 16. The current design would require the use of jumpers which is not endorsed in the NRC's SE (Reference 4) as stated:
" Testing of RTS Analog Signals in the bypassed conditions by use of temporary jumpers or
s i
by lifting' leads is not acceptable." Therefore, the licensee proposes to place the inoperable channel in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> as required by Action Statement 15.
Further, the time to place.the inoperable channel in the test condition is e
changed from I hour to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> which is in accordance with the-conclusions of the SE (Reference 8). Although'the WOG STS states:
place the inoperable channel in bypass within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, the Action Statement 15, at,ove, is for the trip condition which is more restricted.
l 3.3-3(6)
Action Statement 14 for the following Functional Units is changed:
Ib Automatic Actuation Logic And Actuation Relays (Safety Injection) 2b Automatic Actuation Logic and Actuation Relays (Containment Spray) 7a Automatic Actuation Logic and Actuation Relays (Recirculation)
Action Statement 14:
With the number of OPERABLE channels one less than the minimum Channels OPERABLE requirements, [ restore the inoperable channel to=
OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. or) be in at least HOT STANDBY within [the next) 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the l,
following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />; however, one channel may be bypassed for up to i
[ 1 ) hours for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE.
~!
This change agrees with the NRC's SE (Reference 8) conclusion.
l 3.3-3(7)
Action Statement 16 for the following Functional Unit is changed:
2c Containment Pressure-High-3' Action Statement 16:
With the number of OPERABLE channels one less-than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is met.. One additional channel may be bypassed for up to [ 1 ] hours for surveillance testing per q
Specification 4.3.2.1.
.i This change agrees with the NRC's SE-(Reference 8)' conclusion.
3.3-3(8)
Action Statement 21 for the following Functional Unit is changed:-
4b Automatic Action Logic and Actuation Relays (Steam Line Isolation)
-.. 'I
+
. Action Statement 21:
With the number of OPERABLE Channels one less than the Minimum Channels OPERABLE requirement, [ restore the inoperable channel to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. or] be in at least HOT STANDBY within [the next] 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at lest H0T SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to
[ & ) hours for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE.
This change agrees with the NRC's SE (Reference 8) conclusion.
3.3-3(9)
Action Statement 24 for the following Functional Unit is changed:
Sa Automatic Actuation Logic and Actuation Relays (Turbine Trip & Feedwater Isolation)
Action Statement 24:
With the number of OPERABLE Channels one less than the Minimum Channels OPERABLE requirements, [ restore the inoperable channel to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. or) be in at least HOT STANDBY within [the next] 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to [ 4 ) hours for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE.
This change agrees with the NRC's SE (Reference 8) conclusion.
Table 4.3-2 *ESFAS INSTRUMENTATION SURVEILLANCE REQUIREMENTS" 4.3-2(1)
Analog Channel Operational Test" (ACOT) interval is changed from Honthly to I Quarter 1v 1 for the folle':ing Functional Units:
Ic Containment Pressure-High-1 Id Pressurizer Pressure-Low (above P-ll) le Steam Line Pressure-Low (above P-11) 2c Containment Pressure-High-3 3b(3) Containment Pressure-High-3 4c Containment Pressure-High 2 4d Steam Line Pressure-Low (above P-ll) 4e Steam Line Pressure-Negative Rate-High (below P-11)
Sb Steam Generator Water Level-High-High (P-14) 6c Steam Generator Water Level-Low-Low 7b RWST Level-Low-Low Coincident with Safety Injection 9a Pressurizer Pressure, P-ll 9c Low-Low T,y, P-12 This change agrees with the NRC's SE (Reference 8).
e
4 l
t>
. 4.3-2(2)
Functional Unit 9d " Steam Generator Water Level-High-High" is removed from the table because it is a duplication with Functional Unit 5b.
4.3-2(3)
Note # [The specific 18 month interval may be extended to 32 months for Cycle 1 only.] Notation is deleted from the following Channel Calibration functional Units:
Ic Containment Pressure-High-1 Id Pressurizer Pressure-Low (above P-11) le Steam Line Pressure-Low (above P-11) 2c Containment Pressure-High-3 I
3b(3) Containment Pressure-High-3 l
4c Containment Pressure-High 2 4d Steam Line Pressure-Low (above P-II) 4e Steam Line Pressure-Negative Rate-High (below P-11) 5b Steam Generator Water Level-High-High (P-14) 6c Steam Generator Water Level-Low-Low 7b RWST Level-Low-Low Coincident with Safety Injection 9a Pressurizer Pressure, P-ll 9c Low-Low T
, P-12 The Note is deleted because the cycle-specific relief is no longer applicable.
l 4.3-2(4)
Functional Unit Sb, Steam Generator Water Level-High-High (P-14),
Actuation Logic Test and Master Relay Test is changed from Not Applicable to [ Monthly Note 1) and the Slave Relay Test is changed from Not Applicable to [0uarter1Y).
Note 1: "Each train shall be tested at least every 62 days on a i
STAGGERED TEST BASIS."
This change agrees with the NRC's SE (Reference 8).
4.3-2(5)
Functional Unit 6d, Undervoltage-RCP Bus, Trip Actuating Device Operational Test, is changed from Monthly to [ Quarterly).
This change agrees with the NRC's SE (Reference 8).
'f The staff's SE (Reference 4) also required that the licensee confirm the applicability of the generic analysis to their facility and confirm that any i
increase in instrument drift due to the extended STI is accounted for in the Setpoint Calculation Methodology.
Ceco has stated in their August 5,1992, submittal that the generic analyses used in WCAP-10271 and Supplcments is applicable to the Braidwood Station.
l The Braidwood Station uses the Westinghouse 7300 Process Control system and the Westinghouse Solid State Protection System (SSPS) for both the ESFAS and i
r
. i RTS. Both of these systems were specifically modelled in the generic analyses.
The ESFAS Functional Units implemented are all addressed by the generic analysis except, Functional Unit 7.b., Automatic Opening of Containment Sump Suction Valves - RWST Level-Low-Low Coincident with Safety Injection. This functional Unit is addressed on a plant specific basis in the " Technical Specification Optimization Program, RWST Switchover Justification for Braidwood Nuclear Station" and it has been determined that the Functional Unit has a decrease in availability of less than 12%.
Evaluation of the Functional Units in the generic program and the plant-specific evaluation determined that an increase in unavailability of less than 12% is acceptable.
7 CECO has implemented programs at the Braidwood Station to evaluate setpoint drift in accordance with WOG position given in the WOG Guidelines for Preparing Submittal Requesting Revisions of RPS TSs. The licensee has l
determined that the values used in the Braidwood Station setpoint methodology properly account for drift due to extended STIs.
The staff finds the proposed changes to Braidwood, Units 1 and 2, TS 3/4.3.1 and 3/4.3.2 acceptable because they are either deletions due to plant modifications, reflect a cycle-specific relief, are no longer applicable, or are in conformance with the WOG Topical Reports approved by the NRC (References 4 and 8) and reflected in the Westinghouse Standard Technical Specifications.
The generic analysis of WOG Topical Reports which formed the basis of increase of STI or A0T is applicable to the Braidwood Station and the licensee has adequately addressed any increase in instrument drift due to the extended STI.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Illinois State official was notified of the proposed issuance of the amendments.
The State official had no comments.
5.0
[L4VIRONMENTAL CONSIDERATION The amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase'in the amounts, and no significant change in the types, of any effluents that may be released off-site, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (57 FR 48816). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR r
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'. Sl.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
F. Paulitz Date:
December 16, 1993 l
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7.0 REFERENCES
1.
Letter from T.W. Simpkin (Ceco) to T.E. Murley (NRC), "Braidwood Station Units 1 and 2 - Braidwood Station Units 1 and 2 Application for Amendment to Facility Operating License," August 5, 1992.
2.
Westinghouse Letter CAW-83-7, submitting Topical Report WACP-10271,
" Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System," February 3, 1983.
3.
Letter from J.J. Sheppard (WOG) to C. Thomas (NRC), " Supplement I to WCAP-10271," April 4, 1983.
4.
Letter from Cecil 0. Thomas (NRC) to J.J. Sheppard (WOG),
Subject:
" Acceptance for Referencing of Licensing Topical Report WCAP-10271,
' Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System,'" February 21, 1985.
5.
Westinghouse March 20, 1986, letter submitting G.R. Andre, R.C. Howard, R.L. Jansen, and K. Leonelli, " Evaluation of Surveillance Frequencies and Out of Service Times for the Engineered Safety Features Actuation System," WCAP-10271, Supplement 2, February 1986.
6.
Westinghouse May 12, 1987, letter submitting G.R. Andre, R.C. Howard, R.L. Jansen, and K. Leonelli, " Evaluation of Surveillance Frequencies and Out of Service Times #or the Engineered Safety Features Actuation System," WCAP-10271, Supplement 2, Revision 1, March 1987.
7.
Letter from Charles E. Rossi (NRC) to Roger A. Newton (WOG),
Subject:
" Westinghouse Topical Reports WCAP-10271, Supplement 2, and WCAP-10271, Supplement 2, Revision 1, ' Evaluation of Surveillance Frequencies and Out of Service Times for the Engineered Safety Features Actuation System,'" February 22, 1989.
8.
Letter from Charles E. Rossi (NRC) to Gerard T. Goering (WOG),
Subject:
" Westinghouse Topical Reports WCAP-10271, Supplement 2, Revision 1,
' Evaluation of Surveillance Frequencies and Out of Service Times for the Engineered Safety features Actuation System,'" April 30, 1990.
9.
Letter from J.B. Hickman (NRC) to T.J. Kovach (CECO), " Amendment No. 40 to facility Operating Licenses No. NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2," October 5, 1992.
10.
Letter from K.A. Ainger (CECO) to NRC " Application for Amendment to Braidwood Unit 1 and 2 Facility Operating License," September 30, 1987.
11.
Letter from K.A. Ainger (CECO) to NRC " Anticipatory Reactor Trip Upon Turbine Trip," October 30, 1987.
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o 12.
Letter from L.N. Olshan (NRC) to L.D. Butterfield, Jr. (CECO),
" Amendment No. 3 to Allow Deletion of Reactor Trip on Turbine Trip Below 30% Power," to Facility Operating License No. NFP-72 for Braidwood Station Unit 1, dated December 8, 1987.
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