ML20082A236

Proposed Tech Specs Tables 2.2-1 & 3.3-4 Re Reactor Trip Sys Instrumentation Trip Setpoints & ESF Actuation Sys Instrumentation Trip Setpoints

ML20082A236
Person / Time
Site:	Byron, Braidwood
Issue date:	06/28/1991
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20082A225	List: ML20082A223 ML20082A236
References
NUDOCS 9107100240
Download: ML20082A236 (9)

Text

_ - - .

b q TABLE 2.2-1 (Continued)

m mv) i fjE REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

y. .

gg SENSOR .

65 g TOTAL ERROR

pg'm FUNCTIONAL UNIT ALLOWANCE (TA) Z (SE) TRIP SETPOINT Alt 0WA8tE VALUE oo b',5 ,

[ 12. Reacter Coolant Flow-Low 2.5 1.77 0. 6 - >90% of "leep" Die num measured flow"

>89.2% of loop mini-num measured flow

• o&

'r32,$$ 13. Steam Generator Water -

%, -Level Low-tow I a. Unit 1 27.1 18.28 1.5 >40.8% of narrow >39.1% of narrow range instrument range instrument N.A-9 p NS- ) 4.h.y spag3;3 span 35".4

b. Unit 2 -14r78- -Ir5

>17% of narrow >13-3% of narrow range instrument range instrument 5

y span span w

14. Undervoltage - Reactor 12.0 0.7 0 >5268 volts - >4728 volts -

i Coolant Pumps each bus each bus i

j 15. Underfrequency - Reactor 14.4 13.3 0 >57.0 Hz

>56.5 Hz Coolant Pumps .

1

16. Turbine Trip -

I

a. Emergency Trip Header N.A. N.A. N.A. >540 psig. >520 psig Pressure

b. Turbine Throttle Valve N.A. N.A. N.A. >1% open >1% open Closure 4

17. Safety Injection Input N.A. N.A. N.A. N.A. N.A.

l from ESF

18. Reactor Coolant Pump N.A. N.A. N.A. N.A. N.A.

Breaker Position Trip

• Minimum measured flow = 97,600 gpa i '

_ . . . - - - . . . , _ -.,--.m--

4 -

TABLE 3.3-4 (Continued) '

SE 4

g; ENGINEERED' SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS' z

- -' TOTAL SENSOR TRIP. ALLOWABLE gg FUNCTIONAL UNIT ALLOWAN.E : (TA) 1 ERROR (SE) SETPOINT VALUE

' 11 us 4. Steam Line Isolation-1 p.

! e- a. Manual Initiation N.A. N.A. M.A. N.A. N.A.

b. Automatic Actuation Logic'and Actuation-Relays M.A. .N.A. 'N.A. N.A. M.A.

c. Containment Pressure-High 7. 7 0.71 1. 5 18.2 psig 19 2 psig

d. Steam Line Pressure- 21.2 14.81 1. 5 >640 psig* >617 psig*

Low (Above P-11)

EI e. ' Steam Line Pressure

[, Negative Rate-High 8.0 ' 0. 5 0 1100 psi ** $111.5 psi **

g, (Below P-11)

- en

5. Turbine Trip and Feedwater Isolation

a. Automatic Actuation Logic.and' Actuation Relays -N.A. N.A. N.A. M.A. N.A.

b. Steam Generator Water Level-High-High (P-14) *

1) Unit 1 6.0 4.28 1.5 181.4% of 182.7% of narrow range narrow range instrument instrument

-18.3 3 i2.01 32 spang 80.8 spant92.9-

2) Unit 2 -fr:9 ' t-18 ) 5-} _6 o f 179-9% of narrow range narrow range instrument instrument span span

t TABLE 3.3-4 (Continued)  !

EE L

. pg . ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS )

z 10TAL SENSOR ~ TRIP ALLOWABLE g -FUNCTIONAL UNIT ' ALLOWANCE (TA) Z ERROR (SE) ,SETPOINT VALUE 7

Zo' 5. Turbine Trip ~and  !

d Feedwater Isolation (cont'inued)  !

e. -

t

,, c. . Safety Injection See Item 1. above for all Safety injection Trip Setroints and i Allowable Values. .

i

6. Auxiliary Feedwater i e

a. Manual Initiation N.A. N.A. N.A. M.A. N.A. l

b. Automatic Actuation  !

Logic and Actuation i

,, Relays N.A. N.A. N.A. N.A. N.A. [

32 c. Steam Generator Water g> Level-Low-Low-Start.

[

g> Mator-Driven Pump and i Diesel-Driven Pump 2

1) Unit 1 27.1 18.28 1.5 >40.8% of >39.It of narrow range narrow range  ;

instrument instrument i N-8 I4 A- 11. 4 . span y spang 35.+

2) Unit 2 -lhe-} ?8]- -14 } >M o f >15--3% of l

- Harrow range narrow range instrument instrument i span' span i

d. Undervoltage-RCP Bus- N.A. N.A. N.A. >S268 volts >4728 volts

• Start Motor Driven Pump and Diesel-Driver: Pump

e. Safety Injection-i Start Motor- i Driven Pump and See Item 1. above for all Safety Injection Trip Setpoints and i l

Diesel-Driven Pump Allowable Values.  !

r i

l i

b

%. ~. , - . , , .- , _

m_ _. . . . . .

s

(

E TABLE 2.2-1 (Ceritinued)

E E REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 8

O

' SENSOR TOTAL ERROR E FUNCTIONAL UNIT ALLOWANCE (TA) Z (SE) TRIP SETPOINT ALLOWA8LE VALUE d 12. Reactor Coolant Flow-Low 2. 5 1.77 0.6 >90% of loop mini- >89.2% of loop mini-

~

mum measured flow

• Eum measured flow *

. e-m 13. Steam Generator Water Level Low-low

a. Unit 1 27.1 18.28 1.5 >40.8% of narrow >39.1% of narrow Fange instrument _ iange instrumen u m ' "

5

b. Unit 2 17.0 (cid 14.78 1.5 g,s) spar} g,3g y , span k1); 354 r*d)

>17%,of narrow >15. ,of narrow

%. rs. A . (Cs & 4 *-J J A.(+) st.4.(^l range instrument range instrument

• .g ) span span 14 Undervoltage - Reactor .2.0

$0.7 0 >5268 volts - >4728 volts -

Coolant Pumps each bus each bus i

15. Underfrequency - Reactor 14.4 13.3 0 >57.0 Hz >56.5 Hz Coolant Pumps -

16. Turbine Trip

a. Emergency Trip Header N.A. N.A. N.A. >540 psig >520 psig Pressure

b. Turbine Throttle Valve N.A. N. A. N.A. ->1% open ->1% open Closure

17. Safety Injection Input N.A. N.A. N.A. N.A. N. .A.

from ESF t

18. Reactor Coolant Pucp N.A. N.A. N. P.. N.A. N.A.

Breaker Position Trip

• Minimum measured flow = 97,600 gpm

N O

4 i

=

. TABLE 3.3-4 (Continued)-

! 3:

[ .c{ ENGINEERED SAFETY FEATURES ACTUATION' SYSTEM INSTRUMENTATION TRIP SETPOINTS j

8 TOTAL SENSOR TRIP ALLOWA8LE

] . FUNCTIONAL UNIT: ' ALLOWANCE (TA): Zi ERROR (SE) SETPOINT. VALUE E

~

4. Steam Line Isolation

-o

[ 'a. Manual Initiation N.A. N.A. N.A. N. A. M.A.

e. b. Automatic Actuation l-  % Logic.and Actuation

[ . Rela,s N.A. N.A. N.A. N.A. M.A.

i

c. Containment Pressure-High-2 -

7.7 0.71. 1. 5 <8.2 psig 19 2 psig i l~

d. Steam Line Pressure- 21.2 14.81 1.5 >640 psig* >617 psig*

i

, w Low (Above P-11) i 1 e. Steam Line Pressure

i. w Negative Rate-High 8.0 0.5 0 -<100 psi'* <111.5 psi **

j E (Below P-11)-

i

"'\

i 5. Turbine Trip and (qcteNiie87*

(c,,g,5)'-Sir.9%

~a h) )

Feedwater Isolation (94J W*

a. Automatic Actuation

.' Logic and. Actuation i Relays N.A. N.A. N.A. N.A.- N.A.

b. Steam Generator Water

Level-High-High (P-14) 4

1) Unit 1 6.0 4.28 1. 5 -<81.4% of <82.7% of narrow range narrow range r instrument instrument N' (% span span i; 2) Unit 2 5.0 (qck 3) 2.18 1.5 .178.1%,of $79.9%,of l k4 - st.oA (q<1Mk4w 3i2.A(gel 9 narm range 7ar m range

! -gg.9 cc ,g u._4,. ] ._2 ,) instrument instrument

span span ll 4,

1

. . . . , - - _ - - --....-,,,-,n,> -, ~ - , ~ -- -- . - - - - - - - - - - - - - -

,- -r ,

j .i 07 TABLE 3.3-4 (Con +.inued) b ENGINEERED SAFETY FEATURES ACTUATION SYSTEH INSTRUMENTATION TRIP SETPOINTS  !

TOTAL SENSOR TRIP ALLOWA8LE C

FUNCTIONAL UNIT ALLOWANCE (TA) VALUE i

Z_ ERROR (SE) SETPOIIE

, E 5. Turbine Trip and -

j. Z Feedwater Isolation (continued) i c. Safety Injection See Item 1.- above for all Safety Injection Trip Setpoints and

i Allowable Values.

N l 6. Auxiliary Feedwater

(

a. Manual Initiation N.A. M.A. N.A. N.A. N.A.

4

b. A* stomatic Actuation Logic and Actuation w Relays N.A. N.A. M.A. N.A. M.A.

A c. Steam Generator Water '--~~

T Level-toy-Low-Start 3}, n.n M A4 d *

g Motor-Driven Pump and [ Ag.[s.4 7, (c gele- 4 Diesel-Driven Pump

,c4.A.

t l 1) Unit 1 27.1 18.28 1.5 >40.8% of

' >39.1% of A M  !

' narrow range narrow range instrument instrument  :

(C AN (%AN span span j 2) Unit 2 17.0 @ci d 14.78 4 1. 5 , >17%hf >15.3% of

! c.A.(cg a4 4 &4 d.A-(C T A 4 narrow range narrow range

.I w i ,s.g. ca cc3 .49 a.JM) instrument instrument '

i fi span span

d. Undervoltage-PCP Bus- N.A. N.A. N.A. >5268 volts >4728 volts Start Motor Driven Pump

and Diesel-Driven Pump

e. Safe'y injection-

/

Start Motor-  !

Driven Pump and See Item 1. above for all safety Injection Trip Setpoints and Diesel-Driven Pump Allowable Values.

~i

ATTACHMENT _O I

EVAL.UATION OF_SIGNIElCANT. HAZARDS CONSIDERATIONS Commonwealth Edison has evaluated this proposed amendment and determined that it involves no significant hazards considerations. According to 10CFR50.92(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 sl0nificant increase in the probability or consequences of an accident provinusly 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 stearn generator water levelinstrumentation is a safety grade system designed to actuate a reactor trip due to a loss of heat sink. The basic function of the reactor protection circuits associated with Low Low Steam Generator Water Levelis to preserve the Steam Generator heat sink for removal of long term residual heat. Should a complete loss of feodwater occur, the reactor would be tripped on a Low Low Steam Generator Water Level. In addition, an auto start signalis provided at the same sotpoint to two redundant auxillary feedwater pumps to sup aly feedwater in order to maintain residual heat removal after the trip. The reactor tr p acts before the Steam Generators are dry. This n,euces the required auxiliary feedwater capacity, and minimizes the thermal transient on the Steam Generate :.nd Reactor Coolant System.

The auto-start of the auxiliary feedwater pumps at the same setpoint as the trio ensures a se cnoory heat sink is continually available after a trip coincident wIth a loss of normal feedwater. The low low water level trip serves as input to the following accidents:1) Loss of Normal Feedwater Flow, ll) Feedwater System Pipe Break, and lii) l Steambreak. The high high level steam generator trip is an equipment protective trip aroventing excessive moisture carryover which could damage turbine blading The ligh-high evel trip is relied on in the Feedwater System Malfunction Causing an increase in Feedwater Flow Analysis.

The modification to the D5 steam generator tap location will not increase the probability of an accident previously evaluated in the UFSAR. Evaluations and reanalysis have determined that the modification of the steam Denerator level setpoints will not affect any equipment or circumstances for initiation of the UFSAR accidents. The replacement transmitters are the same model as the current transmitters but they will have a different range. The replacements will have identical characteristics compared to the current transmitters such that a feedwater malfunction event resulting from transmitter error will be no more probable than already assumed in the UFSAR.

i i

/scl:1019:26

e co ATTACHMENT _C (continued)

EVALUATION.0F_SIGNIEICANI HAZARDS CONSIDERATIONS The consequences of an accident are not increased. All the applicable non LOCA acceptance criteria are still met for bcth the transients evaluated and the two events analyzed (Feedwater System Malfunction Causing an increase in Feedwater Flow and Loss of Normal Feodwater Flow). The SGTR Analysis of Record and the NFS SGTR Analvsis have been concluded to remain valid. Small perturbations in steady state steam generator level do not affect large break LOC A, long term core cooling, hot let switchover, or LOCA hydraulic forces. The norninal steam generator level and water mass at 100% power will not be changed as a result of the implementation of this modification. This being the case, and since only a nominal mass value is analyzed, the small break LOCA analysis is not adversely affected by the change in level tap location. The impact on ATWS was also considered. The ATWS setpoint will be changed such that the assumption for ATWS initiation, specified as 3% of narrow range span below the low low SG level Reactor Trip / Auxiliary Feedwater initiation Setpoint, is I catisfied. Finally, no new limiting single failure is introduced by the proposed change.

Therefore, there is no potential for an increase in the dose releases and the l consequences of an accident previously evaluated in the UFSAR are not increased. .

1

. he level instrumentation will utilize lower taps; otherwise, its function remains the same. No new modes of operation have been introduced by this modification. Since the accident analysis conclusions as presented in Chapter 15 of the UFSAR are bounding and remain valid, and no new failure mechanism has been identified, the possibility of a new or different hind of accident from any previously evaluated is not created.

New Safety Analysis Limits (SAL) have been incorporated for the steam generator water low low level und high high level trip functions. The Technical Specification trip setpoints have accounted for instrument uncertaintles to ensure that the SAL is not exceeded. The consequences of previously evaluated accidents have either been re-evaluated or re-analyzed assuming the revised SAL. The results are either bounded by previous analyses or insignificantly changed. In all cases, the results are within the applicable design and safety criteria, and the conclusions of the UFSAR remain valid.

Furthermore, the modification will improve safety since the modified system will be less susceptible to feedwater transients, thus reducing the potential for reactor and turbine trips and avoiding unnecessary transients on the primary and secondary systems.

Therefore, this modification does not result in a significant reduction in tae margin to safety.

Therefore, based on the above evaluation, Commonwealth Edison has concluded these changes do not involve significant hazards considerations.

/scl:1019:27

o1 ATTACHMENT D ENVIRONMENTALASSESSMENT Commonwealth Edison has evaluated the proposed amendment against the criteria for and identification of licensing and regulatory actions requiring environmental assessment in accordance with 10CFR51.21. It has been determined that the proposed change meets the criteria for a categorical exclusion as orovided for under 10CFR51.22(c)(9). This determination it based on the fact that th s change is being proposed as an amendment to a license issued pursuant to 10CFR50, it involves changes to the installation or use of facility components located with the restricted area as defined in 10CFR Part 20 and the amendment meets the following specific criteria:

(i) the amendment involves no significant hazards consideration.

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 effluents that may be released offsite.

As documented in Attachment A, there will be no change in types or increase in the amounts of any effluents released offsite.

(iii) 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 controls or methodology used for processing of radioactive effluents of 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.

l l

/scl:1019:28 l

.. . . - . . . , - - . - . .