ML20108F235
| ML20108F235 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 04/29/1996 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20108E782 | List: |
| References | |
| RTR-NUREG-1431 NUDOCS 9605130208 | |
| Download: ML20108F235 (24) | |
Text
_. _ _
... - - ~ _ _
.._m___ _ _ - - - -
1 ATTACHMENT B-1 l
PROPOSED CHANGES TO APPENDIX A, TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-37 AND NPF-66, BYRON STATION UNITS 1 & 2 l
l Revision to:
X 3/4 7-1 3/4 7-2 3/4 7-3 B 3/4 7-1 B 3/4 7-2 l
I i
9605130208 960429 PDR ADOCK 05000454 P
l LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS t
SECTION E8,GE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE 3'a f e ty V a1 ve s............................................
2/4 7-1 TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON l
FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING-FOUR-LOOP 0PERATION...........................................
3/4 7-2 l
TABLE 3.7-2 STEAM LINE SAFETY. VALVES PER L00P.....................
3/4 7-3 1
Auxiliary Feedwater System...............................
3/4 7-4 condensate Storage Tank..................................
3/4 7-6 Specific Activity.......................................'.
3/4 7-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.........................
3/4 7-8 l
1 Main Steam Line Isol ation Va1ves.........................
3/4 7-9 l
3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION..........
3/47-10 l
3/4.7.3 COMPONENT COOLING WATER SYSTEM...........................
3/47-11 l
3/4.7.4 ESSENTIAL SERVICE WATER SYSTEM...........................
3/47-12 3/4.7.5 ULTIMATE HEAT SINK.......................................
3/47-13 3/4.7.6 CONTROL ROOM VENTILATION SYSTEM..........................
3/47-16 3/4.7.7 NON-ACCESSIBLE AREA EXHAUST FILTER PLENUM VENTILATION SYSTEM.......................................
3/47-19 3/4.7.8 SNUBBERS.................................................
3/47-22 FIGURE 4.7-1 THIS FIGURE NOT USED................................
3/47-27 l
TABLE 4.7-2 SNUBBER VISUAL INSPECTION INTERVAL...................
3/47-27a l
3/4.7.9 SEALED SOURCE CONTAMINATION..............................
3/47-28 BYRON - UNITS 1 & 2 X
AMENDMENT N0. 60
[
3/4.7 PLANT SYSTEMS 3/4.7.1 TUR8INE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All r.ain steam line Code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-2.
APPLICABILITY:
MODES 1, 2, and 3.
ACTION:
a.
-With four reactor coolant loops and associated steam generators in operation and with one or more main steam line Code safety valves inoperable, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced
.per Table 3.7-1; otherwise, 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 following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS i
4.7.1.1 No additional requirements other than those required by i
Specification 4.0.5.
1 l
l i
E.YRON - UNITS 1 & 2 3/4 7-1
I INSERT A a.
With up to 3 inoperable main steam line Code safety valves on any one steam generator, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either restore the inoperable valves to OPERABLE status, or reduce the Power Range Neutron Flux High Trip Setpoints per Table 3.7-1; otherwise, 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 HOT SHUiDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With more than 3 inoperable main steam line Code safety valves on any one steam generator, 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
\\
r i
a
l TABLE 3.7-1 l
MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH _ _,, _ _
l INOPERABLE STEAM LINE SAFETY VALVES OURING FOUR LOOP OPERATION MAXIMUM NUMBER OF INOPERABLE MAXIMUM ALLOWABLE POWER RANGE SAFETY VALVES ON ANY NEUTRON FLUX HIGH SETPOINT OPERATING STEAM GENERATOR (PERCENT'0F RATED THERMAL POWER) h 1
87 2
65 3
43. -
e 4
4 d
BYRON - UNITS 1 & 2 3/4 7-2
l TABLE 3.7-2 STEAM LINE SAFETY VALVES PER LOOP (hf?,')
VALVE NUMBER LIFT SETTING (+1th*f
-ORIFICE SIZE MS013(A-D) 1235 psig
-16 in' z
MS014(A-D) 1220 psig 16 in MS015'(A-D) 1205 psig 16 in' MS016(A-D) 1190 psig 16 in' MS017(A-D) 1175 psig 16 in' l
i
- The lift setting pressure shall co respond to ambient conditions of the valve at nominal operating temperature and pressure.
main Steam line Code safety valve lift settings may have a f3% tolerance I
until May 9,1994, by which time the lift settings will be reset to i1%.
I
. m-4-3 t L,% N sl e s w.
BYRON - UNITS 1 & 2 3/4 7-3 AMENDMENT NO. 61
l f
3/4.7 PLANT SYSTEMS i
_ BASES 3/4.7.1 TURBINE CYCLE i
I 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line Code safety valves ensures that the Secondary Coolant System pressure will be limited to within 110% (1320 psia) of its design pressure of 1200 psia during the most severa anticipated system operational transient.
The maximum relieving capacity is Ossociated with a turbine trip from 102% RATED THERMAL POWER coincident with en assumed loss of condenser heat sink (i.e., no steam dumps to the condenser).
l The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure i
'"~ Code, 1971 Editiin~.~ The total relieving capacity for all valves on all of
~
~
the steam lines is 17.958.x 10 lbs/h which is 119% of the total secondary 8
steam flow of 15.135 x 108 lbs/h at 100% RATED THERMAL POWER.
A minimum of two OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction j
l in Table 3.7-1.
)
STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in Secondary Coolant System steam flow and THERMAL POWER required by the reduced Reactor trip settings of the Power Range Neutron Flux channels.
The Reactor l
l Trip Setpoint reductions are derived on the following bases:
i For four loop operation:
SP = (X) - (Y)(V) x (109).
4 X
l Where:
i SP = Reduced Reactor Trip Setpoint in percent of RATED THERMAL POWER, V = Maximum number of inoperable safety valves per steam line, BYRON - UNITS 1 & 2 B 3/4 7-1
INSERT B The requirement that the main steam line Code safety valves be set to within 1% of the appropriate setpoint is consistent with Section III of the ASME Boiler and Pressure Vessel Code. The allowed operating tolerance of 13% is supported by the Commonwealth Edison Company, Byron /Braidwood Unit 1& 2 Overpressure Protection Report.
INSERT C High@ = 100 ( %f)
Q K
Where:
High(D =
Safety Analysis power range high neutron flux setpoint, in percent.
Nominal NSSS power rating of the plant (including reactor coolant Q
=
pump heat), in Mwt (= 3427.6 MWt).
l Conversion factor = 947.82 (BTU /sec.)/MWt.
K
=
minimum total steam flow rate capability of the operable MSSVs on w,
=
any one steam generator at the highest MSSV opening pressure including tolerance and accumulation, as appropriate, in ibm /sec.
Heat of vaporization for steam at the highest MSSV opening pressure h,,
=
including tolerance and accumulation, as appropriate, in BTU /lbm.
Number of loops in the plant (= 4).
N
=
PLANT SYSTEMS i
_. BASES
-SAFETY VALVES (Continued) l 109 =
Power Range Neutron Flux-High Trip Setpoint for four loop
-operation, Total relieving capacity of all safety valves per steam X =
-line in 1bs/ hour, and Y =
Maximum relieving capacity of any one safety valve in
-lbs/ hour.
j 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM
~
The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Coolant System can be cooled down to less than 350*F from normal operating conditions in the event of a total loss-of-offsite power.-
The motor-driven auxiliary feedwater pump is capable of delivering a j
total feedwater flow of 740 gpm at a pressure of 1450 psig to the entrance of i
the steam generators.
The diesel-driven auxiliary feedwater pump is capable of delivering a total feedwater. flow of 740 gpm at a pressure of 1450 psig to the entrance of the steam generators.
This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and. reduce the Reactor Coolant System temperature to less than 350*F when the RHR System may be placed into operation.
1 4
3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water level of 40% ensures that sufficient water (200,000 gallons) is available to maintain the RCS at HOT STANDBY conditions for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with steam discharge to the atmosphere concurrent with total loss-of-offsite power.
The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.
3/4.7.1.4 SPECIFIC ACTIVITY The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited t.o a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line break.
4 This dose also includes the effects of a coincident 1 gpm reactor to secondary tube leak in the steam generator of the affected steam line.
These values are consistent with the assumptions used in the safety analyses.
}
BYRON - UNITS 1 & 2 8 3/4 7-2
l ATTACHMENT B-2 PROPOSED CHANGES TO APPENDIX A, TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-72 AND NPF-77, BRAIDWOOD STATION UNITS 1 & 2 Revision to:
X 3/4 7-1 3/4 7-2 3/4 7-3 B 3/4 7-1 B 3/4 7-2 4
s
i LIMITING CONDITIONS FOR OPERATION AND SURVE1LLANCE RE0VIREMENTS PAGE l
SECTION 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Valves............................................
3/4 7-1 TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING-FOUR-LOOP 0PERATION...........................................
3/4 7-2 TABLE 3.7-2 STEAM LINE SAFETY VALVES PER L00P.....................
3/4 7-3 Auxiliary Feedw'ter System...............................
3/4 7-4 a
Condensate Storage Tank..................................
3/4 7-6 Specific Activity........................................
3/4 7-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM........'.................
3/4 7-8 Main Steam Line Isol ation Va1ves.........................
3/4 7-9 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION..........
3/4 7-10 3/4.7.3 COMPONENT COOLING WATER SYSTEM...........................
3/47-11 3/4.7.4 ESSENTIAL SERVICE WATER SYSTEM...........................
3/47-12 3/4.7.5 ULTIMATE HEAT SINK.......................................
3/47-13 3/4.7.6 CONTROL ROOM VENTILATION SYSTEM..........................
3/47-14 3/4.7.7 NON-ACCESSIBLE AREA EXHAUST FILTER PLENUM VENTILATION SYSTEM.......................................
3/47-17 3/47-20 3/4.7.8 SNUBBERS.................................................
7I FIGURE 4.7-1 THIS FIGURE NOT USED................................
3/47-25 TABLE 4.7-2 SNUBBER VISUAL INSPECTION INTERVAL...................
3/47-Es 3/4.7.9 SEALED SOURCE CONTAMINATION..............................
3/47-26 I
a BRAIDWOOD - UNITS 1 & 2 X
AMENDMENT NO. 48
l l
a 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION l
3.7.1.1 All main steam line Code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-2.
APPLICABILITY:
MODES 1, 2, and 3.
ACTION:
a.
With four reactor coolant loops and associated steam generators in operation and with one or more main steam line Code safety valves
)
inoperable, operation in MODES 1, 2, and within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable 1_3.may proceed provided, that
~
alve is restored to OPERABLE status or the Power Range Neutron Frux High. Trip Setpoint is reduced per Table 3.7-1; otherwise, 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 following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
j
.b.
The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE RE0VIREMENTS 4.7.1.1 No additional requirements other than those required by Specification 4.0.5.
The provisions of Specification 4.0.4 are not applicable for Braidwood, Unit 1, Cycle 5,.until the initial-entry--.into HODE 2.
The provisions -of '
Specification 4.0.4 are not applicable-for-Braidwood,-Unit-2,-until the initial entry into Mode 2 following forced outage-A2F27.
_, ~
{
t
'}
f, 6 : -
1 w,.
BRAIDWOOD - UNITS 1 & 2 3/4 7-1 UNIT 2 - AMENDMENT N0. 51
j INSERT A l
l With up to 3 inoperable main steam line Code safety valves on any one steam a.
i generator, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either restore the inoperable valves to OPERABLE status, or l
reduce the Power Range Neutron Flux High Trip Setpoints per Table 3.7-1; otherwise, l
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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With more than 3 inoperable main steam line Code safety valves on any one steam I
generator, 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
1 l
l l
4 I
r-w
s TABLE 3.7-1 MAXfHUM ALLOWA8LE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES OURING'F0UR~ LOOP OPERATION MAXIMUM NUMBER OF I:40PERA8LE MAXIMUM ALLOWA8LE POWER RANGE NEUTRON FLUX HIGH SETPOINT SAFETY VALVES 4N ANY OPERATING STEAM GENERATOR (PERCENT OF RATED THERMAL POWER 1 1.:
2
- :~
3 -
e
~
BRAIDWOOD - UNITS 1 & 2 3/4 7-2
- - _ _ - _ _ _ _ - _ _ _ _ ~
TABLE 3.7-2 STEAM LINE SAFETY VALVES PER LOOP YALVE NUMBER LIFT SETTING ( 1%)*#
ORIFICE SIZE' MS013(A-D) 1235 psig 16 in' MS014(A-D) 1220 psig 16 in J
z MS015(A-D) 1205 psig 16 in' M5016(A-D) 1190 psig 16 in' M5017(A-0) 1175 psig 16 in'
- The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
- Main steam line Coae safety valve lift settings may have a 13% tolerance until May 9, 1994, by which time the lift settings will be reset to fl%.
BRAIDWOOD - UNITS 1 & 2 3/4 7-3 AMEN 0 MENT NO. 49
s 3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY val.VES The OPERASILITY of the main steam line Code safety valves ensures that the Secondary Coolant System pressure will be Ifmited to within 110% (1320 psia) of its design pressure of 1200 psia during the most severe anticipated system operational transient.
The maximum relieving capacity is associated with a turbine trip from 102% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam dumps to the condenser).
The specified valve lift settings ar.d relieving capacities are in accordance with the_ requirements of Section III of the ASME Boiler and Pressure Code,1971 Edition.) The total relieving capacity for all valves on all of the steam lines is 17.958 x 108 lbs/h which is 119% of the total secondary steam flow of 15.135 x 108 lbs/h at 100% RATED THERMAL POWER. A minimum of two OPERA 8LE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-1.-
- ~
~ ~ '
STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on thelasts of the reduction in Secondary Coolant System steam flow and THERMAL POWER required by the reduced Reactor trip settings of the Power Range Neutron Flux channels.
The Reactor Trip Setpoint reductions are derived on the following bases:
For four loop operation:
r.
SP = UI ~ UI I x (109).
X f-
'x
/
Where:
./
N SP = Reduced Reactor Trip Setpoint in percent 'of RATED THERMAL POWER, V = Maximum number. of 1.) operable safety valves per steam line, 1
t BRAIDWOOD - UNITS 1 & 2 8 3/4 7-1 4.4
m PLANT SYSTEMS BASES SAFETY VALVES (Continued)
Power Range Neutron Flux-High Trip Setpoint for four loop 109 =
operation, Total relieving capacity of all safety valves per steam X =
line in Ibs/ hour, and Maximum relieving capacity of any one safety valve in Y =
1bs/ hour.
3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERA 8ILITY of the Auxiliary Feedwater System ensures that the Reactor Coolant System can be cooled down to less than 350*F from normal operating conditions in the event of a total loss-of-offsite power.
The motor-driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 740 gpa at a pressure of 1450 psig to the entrance of the steam generators. The diesel-driven auxiliary. feedwater. pump is capable of delivering a' total feedwater flow of 740 gpm'at a~ pressure of 1450 psig to the entrance of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reouce the Reactor Coolant System temperature to less than 350*F when the RHR System may be placed into operation.
3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water level of 40% ensures that sufficient water (200,000 gallons) is available to maintain the RCS at HOT STAN08Y conditions for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with steam discharge to the atmosphere concurrent with total loss-of-offsite power.
The contained water volume limit includes an allowance for water not usable because of tank 2
discharge line location or other physical characteristics.
3/4.7.1.4 SPECIFIC ACTIVITY The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line break.
This dose also includes the effects of a coincident 1 gpm reactor to secondary tube leak in the steam generator of the affected steam line.
These values are consistent with the assumptions used in the safety analyses.
i 8RAIDWOOD - UNITS 1 & 2 8 3/4 7-2
)
1 INSERT B i
The requirement that the main steam line Code safety valves be set to within 11% of the appropriate setpoint is consistent with Section III of the ASME Boiler and Pressure Vessel Code. The allowed operating tolerance of 3% is supported by the Commonwealth Edison Company, Byron /Braidwood Unit 1& 2 Overpressure Protection Report.
INSERT C High4 = 100 ( W,V)
Q X
Where:
Highs =
Safety Analysis power range high neutron flux setpoint, in percent.
Nominal NSSS power rating of the plant (including reactor coolant Q
=
pump heat), in Mwt (= 3427.6 MWt).
Conversion factor = 947.82 (BTU /sec.)/MWt.
K
=
minimum total steam flow rate capability of the operable MSSVs on w,
=
any one steam generator at the highest MSSV opening pressure including tolerance and accumulation, as appropriate, in Ibm /sec.
Heat of vaporization for steam at the highest MSSV opening pressure h
=
g including tolerance and accumulation, as appropriate, in BTU /lbm.
Number of loops in the plant (= 4).
N
=
l l
ATI'ACHMENT C I
EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS Commonwealth Edison has evaluated the proposed amendment and determined that it involves j
no significant hazards considerations. According to 10 CFR 50.92 (c), a proposed amendm to an operating license involves no significant hazards considerations if operation of the facility in accordance with the proposed amendment would not:
Involve a significant increase in the probability or consequences of an accident 1.
previously evaluated; or i
Create the possibility of a new or different kind of accident from any accident 2.
previously evaluated; or 3.
Involve a significtint reduction in a margin of safety.
l Commonwealth Edison (Comed) proposes to revise Technical Specification (TS) 3/4.7.1, Turbine Cycle Safety Valves, and the associated Bases. The proposed revisions include:
l Revising Technical Specification (TS) 3.7.1.1. Action a., to require the unit to be in 1.
hot shutdown, rather than cold shutdown, for consistency with NUREG 1431,
" Standard Technical Specifications for Westinghouse Plants," and adding a new Action I
- b. to clarify the shutdown requirements when there are more than 3 inoperable main steam line Code safety valves on any one steam generator.
Revising Technical Specification Surveillance Requirement (TSSR) 4.7.1.1 to clarif 2.
that Specification 4.0.4 does not apply for entry into Mode 3 for Byron and Braidwood, and, for Braidwood only, deleting the one-time requirements for Unit 1, Cycle 5 and Unit 2 after outage A2F27, Revising the maximum allowable power range neutron flux high trip setpoints in T 3.
3.7-1.
Revising Table 3.7-2 to increase the as-found main steam safety valve (MSSV) lift 4.
l setpoint tolerance to 3%, provide an as-left setpoint tolerance of 1%, and change a table notation.
5.
Deleting the orifice size column from Table 3.7-2.
i Revising the Bases for TS 3.7.1.1 to be consistent with the proposed changes to 6.
3.7.1.1.
I The pmposed change does not involve a significant increase in the pmbability or A.
consequences of an accident previously evaluated.
The text describing reactor coolant loops and steam generators is redundant. TS 3.4.1.1, " Reactor Coolant Loops and Coolant Circulation - Startup and Power Operation," and 3.4.1.2, " Reactor Coolant Loops and Coolant Circulation - Hot Standby," provide restrictions on the number of operating reactor coolant loops and steam generators. Therefore, deleting the text that requires having four reactor coolant loops and associated steam generators in operation from TS 3.7.1.1, Action a., has no impact on any analyzed accident.
The proposed change to TS 3.7.1.1, Action a., to require the final mode to be hot shutdown rather than cold shutdown is consistent with the Applicability section of the specification, which does not require the MSSVs to be operable in hot shutdown.
There are no credible transients requiring the MSSVs in modes 4 and 5. The steam generators are not normally used for heat removal in modes 5 and 6, and thus cannot be overpressurized. The change also eliminates the unnecessary transient that had been imposed on the unit by forcing entry into cold shutdown.
The new Action b. for TS 3.7.1.1 and text changes to Action a. clarify the shutdown requirement times based on the number of inoperable valves. Tkre are no changes to these times.
Changing TSSR 4.7.1.1 to delete the one-time requirements imposed by previous amendments and allow entry into Mode 3 prior to performing the requirements of TSSR 4.0.5 has no impact on any accident. The change permits testing the MSSVs in accordance with the applicable codes and allows a reasonable amount of time for completion of the surveillance. The conditions requiring the one time requirements have been corrected, so the one-time requirements are no longer required.
The proposed setpoints in Table 3.7-1 are more limiting than those currently allowed in Specification 3.7.1.1. Westinghouse determined that the current setpoints are non-conservative for some combinations of reduced MSSV availability and reactor power levels. By reducing the setpoints, the original design margins for safety will be met.
l Reduced reactor trip setpoints due to reduced availability of the MSSVs are not precursors to any accidents, but are used in the safety analysis to establish that plant response will be within required margins for accidents of concern.
Increasing the as-found valve setpoint tolerance from 1 % to 3% does not have a significant impact on any accident. The peak primary and secondary pressures remain below 110% of design at all times. The departure from nucleate boiling ratio and peak cladding temperature values remain within the specified limits of the licensing basis. All of the applicable loss-of-coolant accident (LOCA) and non-LOCA design basis acceptance criteria remain valid.
1 The MSSVs are actuated after accident initiation to protect the secondary systems from overpressurization. Increasing the as-found setpoint tolerance will not result in any hardware modification to the MSSVs. Therefore, there is not an increase in the l
l t
I probability of the spurious opening of a MSSV. Sufficient margin exists between the normal steam system operating pressure and the valve setpoint with the increased tolerance to preclude an increase in the probability of actuating the valves. The MSSVs also remain capable of relieving any unlikely system overpressure during all i
applicable operating modes.
Although increasing the as-found valve setpoint tolerance may increase the steam l
l release from the ruptured steam generator above the Updated Final Safety Analysis l
Review (UFSAR) value by approximately 2%, the steam generator tube rupture analysis indicates that the calculated break flow is still less than the value reported in the UFSAR. Therefore, the radiological analysis indicates that the slight increase in the steam release is offset by the decrease in the break flow such that the offsite 1
l radiation doses are less than those reported in the UFSAR The evaluation also concluded that the existing mass releases used in the offsite dose calculation for the remaining transients (i.e., steam line break, rod ejection) are still applicable.
Therefore, based on the above, there is no increase in the dose releases.
Neither the mass and energy release to the containment following a postulated LOCA, l
nor the analysis of containment response following the LOCA credit the MSSVs in mitigating the consequences of an accident. Therefore, changing the MSSV lift l
setpoint tolerances would have no impact on the containment integrity analysis. In addition, based on the conclusion of the transient analysis, the change to the MSSV l
tolerance will not affect the calculated steam line break mass and energy releases l
inside containment.
Deleting the orifice size column from Table 3.7.1-2 has no impact on previously evaluated accidents. There is no change to the orifice size, which is stated in the j
UFSAR and incorporated as needed in the accident analyses.
l The proposed changes do not introduce any new equipment, equipment modifica or any new or different modes of plant operation. The MSSVs are not precursors to j
any analyzed accident. The proposed changes will not affect the operational characteristics of any equipment or systems.
Therefore, these proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
l f
The pmposed change does not create the possibility of a new or different kind of H.
accident fmm any accident pivviously evaluated.
Deleting the text describing reactor coolant loops and steam generators from TS 3.7.1.1 Action a. has no impact on plant operation since the specific restrictions on the number of operating reactor coolant loops and steam generators are provided in TS 3.4.1.1 and 3.4.1.2.
i The proposed change to TS 3.7.1.1, Action a., to require the final mode to be hot shutdown rather than cold shutdown is consistent with the Applicability section of the
specification, which does not require the MSSVs to be operable in hot shutdow There are no credible transients requiring the MSSVs in Modes 4 and 5. The steam generators are not normally used for heat removal in Modes 5 and 6, and thus be overpressurized. NUREG-1431 does not include requirements for the MSS operable in these modes. The change will also eliminate the unnecessary had been imposed on the unit by forcing entry into cold shutdown.
The new Action b. for TS 3.7.1.1 and text changer to Action a. clarify the shutdown l
requirement times based on the number of inoperable valves. There P.re no c l
the times.
The proposed change to TSSR 4.7.1.1 to clarify that TSSR 4.0.4 does not ap entry into Mode 3 will allow Comed to continue to perform MSSV testing a operating pressure and temperature as required by the applicable codes. Th precludes having to enter an action statement to perform the testing and eli severe time restrictions o,n the valve testing and conflicts with other plant startup requirements.
The proposed recalculated setpoints of Table 3.7-1 are more limiting than thos currently allowed in the Specification and ensure that the original design ma safety are met. The secondary system pressure remains within design limits.
l 3% will not increase the l
lucreasing the as-found tolerance on the MSSV setpoint to l
challenge to the MSSVs or result in increased actuation of the valves. The c the Bases document the method for calculating the reduced reactor trip setp l
on reduced availability of MSSVs.
Deleting the orifice size column from Table 3.7-2 and the obsolete one-time requirements in TSSR 4.7.1.1 are administrative changes only.
Increasing the lift setpoint tolerance on the MSSVs does not introduce a initiator mechanism. The proposed change does not introduce any new equipm l
equipment modifications, or any new or different modes of plant operati failure modes have been defined for any system or componerlimportant to sa l
has any new limiting single failure been identified. This change will not l
operational characteristics of any equipment or systems. Thus, ther L
the margin for safety.
Therefore, these proposed changes will not create the possibility of a new type of accident from any accident previously evaluated.
The pmposed change does not involve a significant mduction in a margin l
C.
Deleting the text describing reactor coolant loops and steam generators h 1-f on plant operation since the specific restrictions on the number of opera coolant loops and steam generators are provided in TS 3.4.1.1 and 3.4.1.2.
(
I
_~ - - - _. - - - - - - -. - -._. - - - --. -
The change requiring hot shutdown instead of cold shutdown entry is more appropr than the existing specification since the action statement places the plant in a mode l
where operability of the MSSVs is not required. The Technical Specification is applicable in Modes I,2, and 3, therefore, entering Mode 4 places the plant in a condition where the MSSVs are not required to be operable. There are no credible transients requiring the MSSVs in Modes 4 and 5. The steam generators are not normally used for heat removal in Modes 5 and 6, and thus cannot be overpressu NUREG-1431 does not include requirements for the MSSVs to be operable in these modes.
Changing the mode in which the MSSVs are tested will not change the operational characteristics of the MSSVs.. Comed will continue to test the MSSVs a operating pressure and temperature as required by the applicable codes.
l The proposed reactor trip setpoints in Table 3.7-1 are more limiting than the current setpoints in the Specification. Reactor trip settings were calculated using a revise methodology to account for the non-linear relationship of reactor trip setpoints and reduced MSSV availability. The revised setpoints ensure the original design margin o safety is maintained. The proposed changes to the Bases include the revised eq l
used to calculate the reduced reactor trip setpoints.
Increasing the as-found lift setpoint tolerance on the MSSVs will not adversely affe the operation of the reactor protection system, any of the protection setpoints, or an l
other device required for accident mitigation. The proposed increase in the setpoint tolerance does not invalidate the LOCA and non-LOCA conclusions presented in the l
UFSAR accident analyses. In letter CAE-91-209/CAE 91-219, Westinghouse concluded that the new loss of load / turbine trip analysis satisfied all applicable acceptance criteria and demonstrated that the conclusion presented in the UFS remains valid. For all the UFSAR non-LOCA transients, the departure from nucleate boiling design basis, primary and secondary pressure limits, and dose release li Peak cladding temperatures remain well below the limits specified l
continue to be met.
in the 10 CFR 50.46.
Deleting the orifice size column from Table 3.7-2 and the obsolete one-time
-l requirements in TSSR 4.7.1.1 are administrative changes.
The proposed changes do not introduce any new equipment, equipment modi l-or any new or different modes of plant operation. These chac;es will not affect th operational characteristics of any equipment or systems. Therefore, no red margin of safety will occur as a result of changes.
L
\\
Therefore, based upon the above evaluation, Commonwealth Edison has concluded tha f
.:hanges involve no significant hazards considerations.
i i
i
~
1 ATTACHMENT D ENVIRONMENTAL ASSESSMENT-l Commonwealth Edison Company (Comed) has evaluated this proposed license amendment I
request against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. Comed has determined t proposed license amendment request meets the criteria for a categorical exclusion set forth in 10 CFR St.22(c)(9). This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50 that changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or that changes an inspection or a surveillance requirement, and the amendment meets the following specific criteria:
l l
(i) the amendment involves no significant hazards considerations As demonstrated in Attachment C, this proposed amendment does not involve any significant hazards considerations.
(ii) there is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite As documented in Attachment A, there will be no change in the types or significant increase in the amounts of any effluents released offsite, f
(iii) there is no significant increase in individual or cumulative occupational radiation exposure.
The proposed changes will not result in changes in the operation or configuration of the facility. There will be no change in the level of controls or methodology used for processing of radioactive effluents or handling of solid radioactive waste, nor will the proposal result in any change in the normal radiation levels within the plant. Therefore, there will be no increase in individual or cumulative occupational radiation exposure resulting from this change.
I i