ML20086E180
| ML20086E180 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 11/18/1991 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20086E179 | List: |
| References | |
| NUDOCS 9111270080 | |
| Download: ML20086E180 (10) | |
Text
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AHACHMENI.B PROPOSED CHANGES TO APPENDIX A, 1ECHNICAL SPECIFICA110NS Of FACILITY OPERAllNG LICENSES NPF-37, NPF 66, NPF-72 AND NPF-77 BY80ftSTA110N BRA 10H000 STA110N Revised.Pages; 2-5 ReviseLPages:
2-5 3/4 3-26 3,4 3-26 4
ZHLD/1249/8 i 1112 704, :o 9111 i a PliR ADOO: O'.0004'?.4
_ _ _ _ _ _ _ _ _ _._ F CfQ_ __
I
V TABLE 2.2-1 (Continued) g REAC10R TRIP SYSTEM INSTRUMENTATI0rl TRIP SETPOINTS SENSOR c:5 TOTAL ERROR d
FUNCTIONf. UNIT Al'0WANCE (TA)
Z (SE) 1 RIP SETP0 INT F.,l'? ale VALUE 12.
Reactor Coolant iIow-Lov 2.5 1.77 0.6
>90% of loop mini-
.2% of loop mini-
[
Eum measu rd tiew*
measured flow
- 1 13.
Steam Generator Water Level Low-Low n.A.
N.A.
NA 3 3. o 7.
31.0 %
a.
Unit 1 27.1 26-
-Ir&-
>49-8% of narraw
>39-1% of narrow range instrument Fange instrument span span b.
Unit 2 17.0 14.78 1.5
>17% of narrow
>15.3% of narrow range instrument fange instrument span span ey w
14.
Undervoltage - Reactor 12.0 C.7 0
>5268 volts -
14728 volts -
Coolant Pumps each bus each bus 15.
Underfrequency - Reactor 14.4 13.3 0
157.0 Hz 156.5 Hz Coolant Pumps 10.
Iurbine Trip a.
Emergency Trip ileader H.A.
N.A.
N.A.
1540 psig.
1520 psig Pressure b.
Turbine Throttle Valve N.A.
N.A.
H.A.
11% open 11% open Closure l
17.
Safety Injection Input en A.
H.A.
H.A.
N.A.
N.A.
from ESF 18.
Reactor Coolant P:-
N.A.
N.A.
N.A.
N.A.
N.A.
Breaker Position..ep
- Minimum measured flow = 97,600 gpm
~
IABLE 3.3-4 (Continued)
EE ENGINEERED SAFETY FEATURES ACTUATION-SYSTEM INSTRUMENTATION TRIP SETPOINTS TOTAL SENSOR TRIP ALLOWABLE gg Ft8HCTIONAL UNIT ALLOWANCE (fA)
Z ERROR (SE) SETPOINT VALUE Z
vi 5.
Turbine Trip and Feedwater Isolation (continued)
((
c.
Safety Injection See Item 1. above for all Safety Inject' - Trip Setpoints and Allowable Values.
6.
Heaual Initiat ten N.A.
N.A.
N.A.
N.A.
N.A.
b.
Automatic Actuation Logic and Actuation Relays N.A.
N.A.
N.A.
N.A.
N.A.
30 c.
Steam Generator Water us Level-Low-Low-Start E
Motor-Driven Pump and Diesel-Driven Pump g.4 ti.A.
- n. A.
33.o %
- 31. o 9?
- 1) Unit 1
-2 7:1-
- 18. -1:5-140rS% of 139-1% of l
narrow rang:
narrow range instrument instrument span span
- 2) Unit 2 17.0 14:78
- 1. 5 117% of 115.3% of narrow range narrow range inst,rument instrument span span d.
Undervoltage-RCP Bus-H. A.
N.A.
N.A.
15268 valts 14728 volts Start Motor Driven Pump
'and Diesel-Driven Pump e.
Safety' Injection-Start Motor-Driven Pump and see Item 1. above for all Safety Injection Trip Setpoints and Diesel-Driven Pump Allowable Values.
EE TABLE 2.2-1 (Cor tineed) 2:
E!
REACTOR TRIP SYSTEM INSTRUMEN1ATION TRIP SETPOINTS 8
i SENSOR l
TOTAL ERROR E
sIONAL UNIT ALLOWANCE (TA)
Z (SE)
TRIP SETPOINT ALLOWABLE VALUE d
12.
R? actor Coolant flow-tow 2.5 1.77 0.6
>90% of loop mini-
>89.2% of loop mini-Eum measured flow
- Eu measured flow *
~
o-l 13.
Steam Generator Water I
~
l level Low-Low
- 31. A.
eJ. A -
A!. A.
35.o JI s1.o '4 a
Unit 1
-f7 4fHM -S-
>4G-tt of narrow
>J9--t'e of narrow range instrument range instrument span span b.
Unit 2 17.0 14.78
- 1. 5
>17% of narrow
>15.3% of narrow Fange instrument Fange instrumert 7
span span w
14.
Undervoltage - Reactor 12.0
- 0. 7 0
15268 volts -
34728 volts -
Coolant Pumps each bus each bus 15.
Underfrequency - Reactor 14.4 13.3 0
357.0 Hz 256.5 Hz Coolant Pumps 16.
Turbine Trip a.
Emergency Trip Header N.A.
N.A.
N.A.
3540 psig 1520 psig Pressure b.
Turbine Throttle Valve N.A.
N.A.
N. A.
31% open
>1% open
~
Closure 17.
Safety Injection Input N. A.
N.A.
N.A.
N.A.
N.A.
from ESF 18.
Reactor Coolant Pump N.A.
N. A.
N.A.
N.A.
N.A.
Breaker Position Trip
' Minimum measured flow = 97,600 ppm
TABLE 3.3-4 (Con.inued)
E 2:
ENGINEERED SAFETY FEATURES ACTUATION SYSTEft INSTRUMENTATION TRIP SETPOINTS 2
8 TOTAL SENSOR TRIP ALLOWABLE
]'
FUNCTIONAL UNIT ALLOWANCE (TA)
Z
- ERROR (SE) SETPOINT VALUE E
5.
Turbine Trip and
((
Feedwater Isolation (continued)
~
c.
Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and e.
Allowable Values.
N 6.
Manual Initiation N. A.
N. A.
N.A.
N. A.
N.A.
b.
Automatic Actuation Logic and Actuation Relays N. A.
N. A.
N.A.
N. A.
N. A.
u, S2 c.
Steam Generator Water Level-tow-low-Start s'
Motor-Driven Pump and Diesel-Driven Pump N.A.
At. A.
N.A.
ss.o %
31.o %
~
- 1) Unit 1
- 27.1 -- 29
--l >4 H W of
> M of f
_narrow range narrow range instrument instrument span span
- 2) Unit 2 17.0 14.78
- 1. 5
>17% of
>15.3% of narrow range narrow range instrument instrument span span d.
Undervoltage-RCP Bus-N. A.
N. A.
N. A.
>5268 volts
>4728 volts Start Motor Driven Pump
~
and Diesel-Driven Pump e.
Safety Injection-Start Motor-Driven Pump and See Item 1. above for all safety Injection Trip Setpoints and Diesel-Driven Pump Allowable Values.
i l
l 1
ALIACHMENI_C EVALUAIION_0E3 1GNIB CAfi UlAZARDS_ CONS 101 RAIL 0NS i
Commonwealth Edison has evaluated this proposed amendment and determined t' it involves no significant hazards considerations. According to 10 CFR 50.92(c), a proposed amendment to an operating license involves no significant hazards considerations if operation of the facility in accordance with the proposed amendment would not:
1.
Involve a significant increase in the probability or consequences of an accident previously evaluated; or 2.
Create the possibility of a new or different kind of accident from any accident previously evaluated; or 3.
Involve a significant reduction in a margin of safety.
The proposed changes have been reviewed by Commonwealth Edison Co. and Hestingnouse in accordance with these criteria.
The basis for determining that no significant hazards considerations are introduced is presented below.
The steam generator water level instrumentation is a safety grade system designed to actuate a reactor trip due to a loss of heat sir.k.
The basic function of the reactor protection circuits associated with Low-Low Steam Generator Hater Level is to preserve the Steam Generator heat sink for removal of long term residual heat.
Should a complete loss of feedwater occc, the reactor would be tripped on a Low-Low Steam Generator Water Level.
in addition, an auto-start signal is provided at the same setpoint to two redundant auxiliary feedwater pumps to supply feedwater in order to maintain residual heat removal capability after the tcip.
The reactor trip acts prior to Steam Generator tube uncovery.
This reduces the required auxiliary feedwater capacity, increases the time interval before the auxiliary feedwater pumps are reluired, and minimizes the thermal transient on the Steam Generator and Reactor toolant System.
The auto-start of the auxiliary feedwater pumps at the same setpoint as the trip ensures a secondary heat sink is continual;y available after a trip coincident with a loss of normal feedwater.
The reactor trip function generated at the low-low Steam Generator Water Level trip setpoint is assumed to provide primary protection for the loss of Normal Feedwater/ Loss of All Non-Emergency AC Power events, Feedline Break event.
Loss of Load / Turbine Trip event and certain cases of the superheated Steam Line Break Mass and Energy Release Calculations outside containment.
All of these analyses assume a " Safety Analysis Limit" value for the reactor trip setpoint lower (more conservative) than the nominal Technical Specification trip setpoint value because it must account for all applicable errors and uncertainties associated with the low-Low Steam Generator Level trip function.
The " Safety Analysis limit" value assumed for these accidents for the low-Low Steam Generator Hater Level Reactor Trip /Aurillary Feedwater initiation was 13.7% of span and remains at that value for the proposed ZNLD/1249/9
AITACHMENI C EVALU A110tLOLS I GN1 f lCAN UlAl6RD LCONS I DERAIIONS l
(continued) changes.
The difference between the original Technical Specification setpoint for Trip / Auxiliary feedwater Initiation (40.8%) and the " Safety Analysis Limit" value of 13.7% contained an excess margin of 10.3% of span.
This excess margin was in excess of that required to account for instrument error and uncertainties identified in the statistical setpoint study.
This change will permit reduction of the Unit 1 Low-low Steam Generator Water Level Reactor Trip / Auxiliary feedwe+er Initiation Setpoint from 40.8% to 33.01, reducing the excess margin from 10.3% of span to 2.5% of span. However, as stated previously, the margin of safety is unaffected since the new Reactor Trip / Auxiliary Feedwater Initiation setpoint is bounded by the original
" Safety Analyses Limit" value used in the applicable safety analyses.
The use of a revised Steam Generator nominal Trip / Auxiliary feedwater Initiation setpoint does not involve a significant increase in the probability or consequences of any accident previously evaluated.
The non-LOCA and LOCA accidents were reviewed verifying that the applicable regulatory or design limit was satisfled fe each case defined in the UFSAR.
The small and large break analyses which calculate peak cladding temperatures presented in the UFSAR remain valid.
In addition, the safety analysis limit of 13.7% of span for the steam generator water level low-low trip setpoint remains unchanged.
The use of a revised Steam Generator nominal Trip / Auxiliary feedwater Initiation setpoint does not create the possibility for a new or different Lind of accident from any accident previously evaluated.
Since the accident analysis conclusions as presented in Chapters 3, 6 and 15 of the UFSAR are bounding and remain valid, and no new failure mechanism has been identified, the possibility of a different accident being created does not axist.
The effect of the revised Reactor Trip / Auxillary feedwater Initiation setpoint does not involve a significant reduction in a margin of safety.
The investigation of the affect of these changes on non-LOCA and LOCA transients has verified that plant operation will remain within the bounds of safe, analyzed conditions as defined in the UFSAR.
Conclusions presented in the UFSAR remain valid.
As such, no reduction in the margin of safety between the UFSAR safety analyses limit and the Technical Specifications safety limits (such as DNBR or pressur:) has taken place for operation with the revised Steam Generator nominal Reactor Trip / Auxiliary Feedwater Initiation setpoint.
Based on the above, this change will not increase the probability or consequences of a previously analyzed accident, introduce the possibility of an accident not previously evaluated, or significantly decrease the margin of safety.
Therefore, this change does not involve a significant hazards consideration.
ZNLD/1249/10
AIJACf!MENL.D LNVIRONMEMIAL_ASSESSMERI Commonwealth Edison has evaluated the proposed amendment against the criteria for and identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21.
It has been determined that the proposed change meets the criteria for a categorical exclusion as provided for under 10CFR51.22(c)(9).
This determination is based on the fact that this change is being proposed as an amendment to a Itcense issued pursuant to 10CFR50, it involves changes to the installation or use of facility componants located with the restricted area as defined in 10CFR 20 and the amendment meets the following specific criteria:
(l) the amendment involves no significant hazards consideration.
As demonstrated in Attachment C, this proposed amendment does not involve.
any significanc hazards considerations.
(11) there is no significant change in the types or significant increase in the amount of any effluents that may be released offsite.
As documented in Attachments A and C there will be no change in types or increase in the amounts of any effluents relea.ed offsite since the original safety analyses remain valid for the new values of trip setpoint and allowable value.
(111) there is no significant increase in individual or cumulative occupational radiation exposure.
This proposed change will not result in changes in the operation or configuration of-the facility, there will be no change in the level of L
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 rautation exposure resulting from this change.
ZNLD/1249/ll-1
AllACllMLNL_E DESCRIRT10R 0[ _ CHANGES MADL _TO ORIGINAL PROPOSAL CECO's previous amendment request proposed a reduction in the excess margin for the steam generator low-low level trip from 7.8 percent of span to 1.8 percent of span.
This margin was that which Westinghouse calculated to be in excess of that required to account for instrument error and ether uncertainties, this change would have lowered the Unit I steam generator low-low reactor trip and ESF actuation setpoints from 40.8 percent of span to 34.8 percent of span.
In 1990, CECO began a setpoint reconciliation program to do a recalculation of all Westinghouse supplied Reactor Trip and Engineered Safety Features setpoints.
This program was a comprehensive review of the design assumptions made by Westinghouse comparing current plant sp9tific ~,
to the original Westinghouse assumptions.
This effort was rece0tly templeted.
The results of this plant and unit specific evaluation revealod that additional margin was available in the steam generator low-low trip setpoint.
In addition, new values were calculated for Total Allowance, Sensor Error, Z, and the Allowable Value.
Based on these calculations, CECO has revised the values that were previously submitted.
The new calculated excess margin is 10.3 rercent of span.
CECO proposes to reduce the steam generator low-low trip setpoint by 7.8 percent, from 40.8 to 33.0 percent of span.
This leaves 2.5 percent of span excess margin, slightly greater than the
'.8 percent proposed in our previous submittal.
The new Allowable Value is 31.0 percent of span.
The new Total Allowance, Z ar.d Sensor trror are Not Applicable (N.A.).
The following table summarizes the previous and currently proposed changes.
The first column identifies the parameter, the second column lists the value for the currently approved Technical Specifications, the third column lists the values requested in our previous submittal, the fourth column lists the values supported by the setpoint reconcillation program and the firth column lists the values requested in this letter based on the review of the values in column four.
All values are given in percent of instrument span.
Currently Previously Revised Proposed Approved Requested Study Requested Parameter _
Values Values Values Values Total Allowance 27.1 21.1 27.1 H.A. (19.3)
Z 18.28 18.28 15.1 N.A. (15.1)
Sensor Error 1.5 1.5 2.5 N.A. ( 2.5)
Allowable Value 39.1 33.1 38.8 31.0 Trip Setpoint 40.8 34.8 40.8 33.0 Margin 7.8 1.8 10.3 2.5 Technical Specification 2.2.1 Action b allows additional options if the trip setpoint is less conservative than the Allowable Value but Equation 2.2-1
((Z 4 RE + SE)<TA, where RE is the Rack Error) is satisfied.
Due to the reduction in the margin of the Trip Setpoint and the other limilit g error ZNLD/1249/12
i l
v'alues, this equation would never be satisfied.
For example, if the as found setpoint were below the Allowable Value (31.0), the Rack Error (Trip Setpoint l
minus the actual as found value) would be greater than 2.0 (33.0 - 31.0).
The resulting sum of Z + RE + SE, 19.6 (15.1 4 2.0 + 2.5), is greater than the Total Allowance (19.3).
Therefore, the options provided by satisfying this equation are not available for the steam generator low-low setpoint.
Therefore, the values of Total Allowance, Z, and Sensor Error are listed as Not Applicable (N.A.).
This position is consistent with that taken for the Unit 2 steam generator low-low level setpoint amendment application submitted for the Byron /Braidwood Unit 2 0-5 Steam Generators.
Changes to Total Allowance, Z, Sensor Error and the Allowable Value would have been required independent of our request to lower the steam generator low-low setpoint.
These revised values are a result of the setpoint reconciliation program.
ZNLD/1249/13
/