ML20095B395
| ML20095B395 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 04/15/1992 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20095B392 | List: |
| References | |
| NUDOCS 9204220166 | |
| Download: ML20095B395 (55) | |
Text
. _ _ _ . _ . _ _ . -- . . _ _ _ . . . ._ _ ._ _ __ . ..
]
i l
l ATTACHMENT B PROPOSED CHANGES TO APPENDIX A-TECHNICAL SPECIFICATIONS FOR FACILITY OPERATING LICENSES NPF-37, -66,-72. AND -77 Revised Pages 2-3 2-4 l 2-5 2-6 2-8 2-10 !
B 2-3 3/4 3-13 3/4 3-23 3/4 3-24 3/4 3-25 3/4 3-26 3/4 3-27 3/4 3-28 B - 3/43 1 B 3/4 3-2 J-4 920i220166 920415 PDR ADOCK 05000454 p PDR
, /scl:lD615:29
SAFETY (! HITS AND LIMITING SAFETY SYSTEM SETTINGS __
- 2. 2 LJS TING SAFETY SYSTEM SETTINGS REACTOR T . SYSTEM INSTRUMENTATION SETPOINij
- 21 The Rcactor Trip System Instrutnentation and Interlock Setpoints shall t e set consistent within the Trip Setpoint values shown in Table 2.2-1.
APPLICADILITY: As shown for each channel in Table 3.3-1.
ACTION:
- a. With a Reactor Trip System Instrumentation or Interlock Setpoint less ~
conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 2.2-1, adjust the Setpoint consistent with the Trip Setpoint value. 3
- b. With the Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value -Nown in the Allowable Values column of Table 2.2-1,.either.
- 1. A dj ust-the-Setpoint-con s i s t e nt-vi t hd he-T ri p 4 e t po i n t-v a l ue-o f-Table-tr2-1 and-determine-withirt-l?-hours- that-Equation a2r2 was-Satisfled-for the -af fected-channel;-or-2r ' beclare the channel inoperable and apply the a.nplicable ACTION statement requirement of Specification 3.3-1 until the channel is restored to OPERABLE status.with its Setpoint adjusted consistent with the Trip Setpoint value.
Equetion-2:2-1 2- +- R E-+-S E-s,-TA ---
y Wheres Z =---The--value-for-Coluren-2-of -Table-2r2-1-f or- the-a f f ec ted channel , -
RE-=-The Eas-measured"-value-(in percent-span)-of-rack-error-for the-
-a f f ected-channeb-SE + Eithee-thel as-measui P-value -(hpercent-span)-of-the sensor-error 7-or-the-velue-for Columr SE (Sensor Error) of Table 2.2-1m
- f or-the-af fected-channe? ,-ona*-
-T A + t he~value-for-Column Tt - (Total - Allowance)-of-Table-2 2-Efor---
the affected thannel BYRON - UNITS 1 & 2 2-3 l
1
r :
j .
, r F
t- l i
co TABLE 2.2-1 5 '
E i e REAC10R TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS C !
i 5 d : TOTAL
$ENSOR--
ERROR- l j w FUNCTIONAL UNIT t
- " -ALLOWAMCE-tT6 y , _ (SE)- TRIP SETPOINT Att0VABLE VALUE i
- 1. Harual Reactor Trip L A.
m H rA. hA. N.A. N.A.
- 2. Power Range Neutron Flux i
- a. High Setpoint AS - ?:55
- .36. !
0- $109% of RTP* 1 M hi% or R1P" .
j b. Low Setpoint -. G . 3 4-56 Y7.h I C 125% of RTP* 15NP-1% of RTP* !
- 3. Power Range, Neutron Flux. -1. E F 0.5 G- 15% or Riv* with liigh Positive Rate 16.3% of RTP! with !
m a time constant a time constant !
- 12. seconds > seconds
_2 t 4 Pov- Range. Neutron Flux, -h 0 0.5 0 <5% of RTP* with t
- i. High Negative Rate <6,3%,of RTP" with !
a time constant a time constant i 12 seconds 12 seconds i
- 5. Intermediate Range, -17.0 w.S 0.? 0- <25% of RTP' ese-9% of I'TP*
Neutron Flux -
t
- 6. Source Range, Heutron Flux 17.0 I.R !
10.0 0 1105 cps $h+ x 105 cps i 4
- 7. Overtemperature AT EM 5.35 Sec- t See Note 1 See Note 2 l
-Mr,te *
- 8. Overpower AT 4.3 1.3 1.2- See Note 3 See Note 4 L
- 9. Pressurizer Pressure-Low 0.0 sto !
2.21 1. 5 -- 11885 psig lie n psig j
- 10. Pressurizer Pressure-!!igh ,13 % i
-3.1 0.71- 1. 5 -- 12385 psig $2396 psig :
6 i 11. Pressurizer Water level-High 43 f 5.0 2.10 1. 5 <92% of instrument <93r6% of instrument span span 1 !
-*RTP = RATED TitERMAL POWER i
m E TABLE _2.2-1 (Continued)
REACTOR TRIP S STEM INSTRUMENTATION TRIP SETPOINTS E
M -sensee s 40TAL
~ FUNCTIONAL UNIT ERROR- 69 3)
-ALLOWAMCFHA) Z (
-_ST+,- TRIP SETPOINT 12.
. _m -
m \ sALLOWABLE VALUF
{ Reactor Coolant Flow-Low ~
-h5 1. 77 - - 0. 5- >90% of loop mini- 89-2%
> of loop mini-sum measured flow
- Eum measured flow"
- 13. Steam Generator Water Level Low-tow 1 I
- a. Unit 1 -Nr A.- N.A.
/
N.A. >33.0% of narrow >31.G% of narrow w
range instrument range instrument '
span span
- b. Unit 2 N.A N.A N.A. LHA i >36.3% of narrow >35r4% of narrow
'? range instrument range instrument
- s. nan span
- 14. Undervoltage - Reactor -E0 0.7 mza 0 >5268 volts -
Coolant Pumps >4R$ volts -
each bus each bus
- 15. Underfrequency - Reactor 4 .%.oS 13.3 0 >57.0 Hz Coolant Pumps E5fr:5 Hz ET 16. Turbine Trip E
i a. Emergency Trip Header SIS
-- N . A . N.A. N.A. >540 psig >520'psig a Pressure 2 b. Turbine Throttle Valve -WA.
? N.A. N. A. ->1% open ~>1% open C1osure
- 17. Safety Injection Input -N.A.
from ESF NA. N.A. N.A. N.A.
- 18. Reactor Coolant Pump M. A . "^
N.A. N.A. H.A.
Breaker Position Trip ..
- Hinimum measured flow = 97,600 gpm
EE E TABLE 2.2-1 (Continued) 2:
. REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETP0lNTS E
~d _truenn v' senavn
~ iOfAt ERROR-FUNCTIONAL UNIT _ (SEl, IRIP SETPOINT 3 -ALLOWANEE_-tf
,m. ,
A), {, . _. Alt 0WABLE VALUE n, 19. Reactor Trip System Interlocks
- a. Intermediate Range tL A.
Neutron Flux, P-6 N. A.- N :e.. - 31 x 10 10 amp >6 x 10 18 amp
- b. Low Power Reactor Trips Block, P-7 m
- 1) D-10 input -N.A. N.A. N.A. $10% of RTP* 27.9% to <12.1% of RTP*
E'
- 2) P-13 input -NrA. N.A. N.A. <10% RTP* Turbine <12.1% RTP* Turbine Impulse Pressure Impulse Pressure Equivalent Equivalent
- c. Power Range Neutron -H.f. N.A. N.A. <30% of RTP* <32.EC of RIP
- Flux, P-8 ~
~
- d. Power Range Neutron M.A. N.A. N.A $10% of RTP*
l Flux, P-10 17.9% to $12.EC of RTP*
! 3I
! @ e. Turbine Impul:e Chamber - N. A. N.A. M.A. <10% RTP* Terbine
! Pressure, P-13 <12.1% RTP* Turbine i @st Impulse Pressure Impulse Pressure r+ Equivalent Equivalent E 20. Reactor Trip Breakers -N.A. N.A. N.A N.A. H.A.
9' 21. Automatic Trip and Interlock -NrA. N.A. h.A. N.A. N.A.
Logic
- 22. Reactor Trip Bypass Breakers " A. N. A - N.A. N.A. N.A.
][
" RIP = RATED IHERMAL POWER l
l TABLE 2.2-1 (Continued)
@ IABLE NOTATIONS (Continued)
NOTE 1: (Continued)
I,t
= Time constant utilized in the measured ! ag lag compensator, tc = 0 s:
~
I' < 588.4 f (riominal T avg at RATED TifERMAL POWER),
P* K3 = 0.00134, PJ P = Pressurizer pressure, psig, P' = 2235 psig (Nominal RCS operating pressure),
S = Laplace transfo m operator, s 8, and f (al) is a functiun of the indicated difference between top and bottom detectors of the power-rangs. neutron ion chambers; with gains to be selected based on measured instrument resptose during plant STARTUP tests such that:
(i) for q - gb between -4 and MM i 4 6 3 cod Du. i. 2 Cyde-f}--and -32% and < 13%
(Unit-1 Cycie 4 'and af ter; Unit-2-Cycle and-af ter), f ,(al) = 0, where q and ab are percent
[ RATED THERMAL POWER in the top and bottom halves of the core respectively, and q +q b i' total THERMAL POWER in percent of RATED illERf".AL POWER; (ii) for cach percent that the magnitude of g t ~9 b m ecds H M W M yc M a W n W 2 W W , - ) emi +13% (Unit-1-Cycle-4-and-af terrUnit 2-Cycle-3-and-af t0r}, the AT Trio Setpoint shall be ) automatically reduced by 2:0%-(Unit-1-Cycle- 3 and-ttnit-2-Cycle-2),1md 1. 74% (ifniti-Cycte (
)
andsfler;-Unit-2-Cycle 3 and af ter) of its value at RATED THERMAL POWER. .
')
3 (iii) 'ar each percent that the magnitude of q t gbexceeds -32%, the ai Trip Setpoint shall be , 2: - g- automatically reduced by 1.67% of its value at RATED THERMAL POWER (Ibit-I-Cycle-4-and-af teer ' g Wt-2-Cycle-3 and-af tert z NOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip 5etpoint by more than g M of AT span.
^$ 41
TABLE 2.2-1 (ContinuedJ TABLE NOTATIONS (Continued) E - 8 NOTE 3: (Continued) g K. = 0.00170/*F for T > T" and K. = 0 for T 5 T", Z
- T = As defined in Hote 1, w
- e- I" = Indicated T avg at RATED TilERMAL POWER (Calibration temperature for AT instrumentation, 5 588.4*F),
5 = As defined in Hote 1, and f 2(AI) = 0 for all AI. NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than N M of AT span. g a.3t
-NOT E-5: The-senser errer for te=perature-is-1-2-end-fee-pressttre-is-1.0.
I l l l I a
. . . . g m
t 2.2 LIMITING SAFE 1Y SYSTEM SETT2NGS BASES 2.2.1 REACT 03 TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal values at which the Reactor trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the core and Reactor Coolant System are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engi-neered Safety Features Actuation System in mitigating the .onsequences of accidents. Thu Setpoint for a Reactor Trip System or interlock function is - considered to be adjusted consistent with the nominel value when the "as ' measured" Setpoint is within the band allowed for calibration accuracy. To accommodate the instrument drif t assumed to occur between operational , tests and the accuracy to which Satpoints can be measured and calibrated, ' Allowable values for the Reactor Trip Setpoints have been specified in Table 2.2-1. Operation with Setpoints-less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error, man-optional-provision-has-b e e n -i nc l uded- fo r-de termi n i ng-t he-OP E RAB I LITY-o f-a - c ha nnel- whe n-i t s-T ri p Se tpo i nt-i s-found -to-exc eed -the- Al l owab l e -Va l ue r--The- methodo l ogy-o f- thi s - option -utilires-the Has-measured"-deviation--f rom-the-specif f ed-calibration-point-for-rack-ind-sensor-components-in-conjunction-with a statistical-combin - 4 t i on -o f-the-othe r-unce rta i n t i e s-o f- the-i ns t rumen ta t i on -to - me a s u re-the- procu s s - v a r i a b l e - a nd - t he-unc erta i nt i e s -i n -c a l i b ra ti ng- the-i ns trumen ta ti on r--I n-Eque = tion 2. 2-Ir-Z-+-RE-+-SE-<-TArthe-interactive-ef fects-of-the-errors-in-the-rack- ' and the-sensory-and-the Ja s-me a s ure d'v al ue s-o f-the-e rro rs - a re-cons i dered r-Zr as specified-in-Table-2r2-Irin-percent-spani-is-the-statistical-summation-of--- 4rrors-assumed 4n-the-analysis-excluding-t50se-associated with-thedensor-and-rack-drif t-and-the-accuracy-of-their-measurementr-TA-or ~ Total-Allowance-is-the differencer-in-percent-spant-between-the-Trip-Setpoint-and-the-value-used-- in the-analysis-for-Reactor-tripr-RE-or-Rack-Error-is-the Has-measured'devie~ t j '
. tionrin-percent--span,-Jor-the-ef fected-channel-f rom-the-specified= Trip-Setpointr
-5E-or-Sensor-Erroe-is-eitherthe nas-measured' deviation-of-the-sensor-from-its ,
-c a l i b r a t i on-point-o r-the-v a l ue-s pec i f i ed -i n-Tabl e-2 r 2-Iri n- pe rc en t- s p a nrfrom- l l the analysis-assumptionsr--Use-of-Equation-2r2-1-allows-for-a-sensor-drif t !
f actor,~an-increased-rack-drif t-factorrand provides-a-threshold-value for- l+ l REPORTABLE-EVENT &r
- ll l
The methodology to derive _the Trip Setpoints is based upon combining all i- of the uncertainties in the channels. Inherent to the determination of the , l i Trip Setpoints are the magnitudes rf these channel uncertainties. Sensors and 9ther instrumentation utilized in-these channels are expected to be-capable of operating within the allowances of these uncertainty. magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has-not met its allowance. Being that there is a small statistical chance that this will happen, an-infrequent excessive drift is expected. Rack or sensor drift, < in excess of the allowance that _is more than occasional, may be indicative of more serious problems and should warrant further investigation. BYRON - UNITS 1 & 2 B 2-3
i
~
l l INSTRUMENTATION
, 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATIO I LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4. '
. }
f APPLICABILITY: As shown in Table 3.3-3. ACTION: a. With an ESFAS Instrumentation or Interlock Trip Setpoint less con-servative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 3.3-4 adjust the Setpoint consistent with the Trip Setpoint value.
- b. With an ESFAS Instrumentation or Interlock Trip Setpoint less con-servative than the value shown in the Allowable Values column of Table
- 3. 3-4,, enhem-fir--Adjust-the--Setpoint-conshtent-with the Trip-Setpoint value of--
\ Table-Sr3-4-and-determine within 12 hours - that-Equation-t-2-t-N
\ wa5-sathf4ed-foe-the-aMeeted-channa1, er-
-h peclarethechannelinoperableandapplytheapplicableACTION statement requirements of Table 3.3-3 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.
Equation-2r2-1 Z -+-RE-+-SE s-TA
-Wheret-
-Z
- -channei The value-free-Coh=n
- 2 cf Teble 3.3 4 fer the-effected-T ,
' --RE M he 8as-eessured" ve4ue-{in percent spen) ef-rack erree-- for the ef fected chennel, i l
-SE-=-E4ther-the 2es-meesruregu-veitte-(4n percent - span)-ef-the-
-+e n se e-e r ro w r-t he-va lue-fe e-Celemn-SE-(S en s o r-E rro e)-o f l
-Ta b l e4ra-4-fo r-the-a Meeted-c hanneFr-an d-TA-*-The-value-free-Column-TA-(-Total-Alievance) of-Table-3-3 for-the-affeeted-channeh- -
; c. .With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3.
BYRON - UNITS 1 & 2 3/4 3-13 AMEN 0 MENT NO. 23
., -- --------,----n-,-,--a , , - , -, w , .-.-, , , -- - - - - - - , , - , , , - - - , - - ----,---n,,~.e,-n e
T Ap' ' 3. 3-4 5 g Et!GINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SE1 POINTS TOTAL ---5ENSOR TRIP ALLOWABLE g FUNCTIONAL UNIT ?.tLOWANCE (TA) Z; - --- ERROR -(SE) SEIPOINT VALUE Z m 1. Safety Injection 9 (Reactor Trip, Feedwater a- Isolation, Start Diesel N Generators, Containment Cooling fans, Control Room Isolation, Phase "A" Isolation, Turbine Trip, Auxiliary Feedwater, Containbent Vent Isolation and Essential Service Water)
- a. Manual Initiation HeA. H A. N r A .-- N.A. N.A.
{ T b. Automatic Actuation H. A. - - -
- N. A. -
N . A .- N.A. N.A. O Logic and Actuation Relays
- c. Containment Pressure- 9L liigh-1 Sr7 Or71 1:5- 5 3.4 psig 3 5-8 psig
- d. Pressurizer-Pressure- it ;
tow (Above P-11) 16.1- 14:41 1:5- 1 1829 psig 3 1823 psig l e. Steam Line Pressure- oN tow (Above P-11) 21:2- 14:81 1:5- 3 640 psig* 3 Gl? psig*
- 2. Containment Spray
- a. Manual Initiation N.- A -. N:A. N: A- N.A. N.A.
I b. Automat.ic Actuation Logic and Actuation l Relays N .-A . H:A. N A-- N.A. N.A.
- c. Containment Pressure- ,a liigh-3 8:0 - - 0; 71 - 1.5- < 20.0 psig < 21.0 psin a
'a m
TABLE 3.3-4 (Continued) N g z ENGINEERED SAFE 1Y FEATURES ACTUATION SYSTEM INSTRUMENTATIOP! TRIP SETPO TOTAL SENSOR-c FUNCTIONAL. UNIT TRIP ALLOWABLE x - AL10WAHEE (TA) Z m ERROR-tSE}- SETPOINT VALUE i d
- 3. Contains.ent Isolation "
.. a. Phase "A" Isolation- t c-
- 1) Hanual Initiation N.A. N-A. N.A N.A.
N N.A.
- 2) Automatic Actuation Logic and Actuation ,
- Relays N
- A. M.A. M-A- N.A. N.A.
- 3) Safety injection See Item 1. above for all Safety Injection Trip Setpolnis and Allowable Values. ,
- b. Phase "B" Isolation R ~
- 1) Hanual Initiation lf. A. ?!. A. H.A. N. A. N.A.
% 2) Autematic Actuation ti.A. M.A.
I Lcgic and Actuation M. A. H.A. N.A. Relays
- 3) Containment Pressure-liigh-3 8.0 C.71 1.5 "A t . A.
5 20.0 psig 1 -M ps ig
- c. Containment Vent Isolation .
- 1) Automatic Actuation '
logic and Actuation , Re1ays N. A. i;. A.
. M.A. N.A. N.A. '
- 2) Hanual Phase "N' N.A. N.A. ii.A. N.A.
Isolation N.A.
- 3) Manual Phase "B" Pt-A. ii. A. ??. A.
N.A. Isolation H.A. e
- 4) Safety Injection See Item 1 above for all Safety Injection Trip Setpoints i and Allowable Values.
- . - ~ . . _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ - _ _ _ _ _ . _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
^
. TABl.E 3.3-4 (Continued)
Y t
@ ENGINEE9ED SAFETY FEAiUPES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOIN!S
- e c- TOTAL- SENSOR- TRIP AtLOWABLE
$ FUNCTIONAL UNIT ALLOWANCE-fTA) Z,- E8R0R (SE)- SETPOINT VALUE '
- y 4. Steam Line Isolation e.
m a. Manual Initiation Na. N.A. N-A:- N.A. N.A.
- b. Automatic Actuation i logic and Actuation Relays N:A. M-A. N Ar N.A. N.A.
- c. Containment Pressure-High-2 9.9
-7 9 091 1. 5 18.2 psig 19-2 psig
- d. Steam Line Pressure- --21:2 14:81 1:5r- >640 psig*
t.4
>617 psig*
w Low (Above P-11) g } e. Steam Line Pressure i W.1 m Negative Rate-High 8:0 0-5 0- 1100 psi ** 1111 4 psi ** 4 w (Be1cw P-11)
- 5. Turbine Trip and I Feedwater Isolation
- a. Automatic Actuation Logic and Actuation g Relays N:A. N- A. -N Ar N.A. N.A. i U b. Steam Generator Water '
N Level-High-High (P-14)
? 1) Unit 1 5^ M1 4-28 1 <81.4% of <82H% of I d iiarrow range iiarrow range I
! g instrument instrument [ span span
- 2) Unit 2 18.1 12-02 3 180.8% of g narrow range 182.8% of narrow range instrument instrument w span span '
m
' i
[ i p
VABLE 3.3-4 (Continued) ' 5 g
, ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS c 40TAL ---- --
- SENSOR z TRIP ALLOWABLE FUNCTIONAL UNIT ALLOWANCE-(TA) - Z
,t y n tRROR-{SE) SETPOINT VALUE tro .- y 5. Turbine Trip and j .i
, Feedwater Isolation (continued)
N
- c. Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and i Allowable Values.
4
- 6. Auxiliary Feedwater
, a. Manual Initiation -N:A. NrA. N.A. N.A. N.A.
- b. Automatic Actuation
- Logic and Actuation m Re1ays -M A. -N-A. NrA-- N.A. N.A.
! D c. Steam Generator Water w Level-tow-Low-Start j g Motor-Driven Pump and Diesel-Driven Pump i
- 1) Unit 1 -N:A. - N. A. N:Ar L 133.0% of 131.0% of i narrow range narrow range instrument instrument span span g 2) Unit 2 -N;A. -- N. A. HAr 349 g 136.3% of 13k4% of g narrow range narrow range j m instrument instrument span span l z d. Undervoltage-RCP Bus- -N.A. -N:A. - NrAr--- 15268 velts
,o Start Motor Driven Pump 14728 volts m and Diesel-Driven Pump
- e. Safety Injection-Start Motor-Driven Pump and See Item 1.. above for all Safety Injection Trip Setp,;nts and Diesel-Driven Pump Allowable Values.
T TABLE 3.3-4 (Continued) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS S
- 10iAL SENSOR- TRIP ALLOWABLE c FUNCTIONAL UNIT Z ALLOWANCE (TA] ERROR (SE)- SETPOINT VALUE CI 6. Auxiliary feedwater (Continued) e e- f. Division 11 for Unit 1 m (Division 21 for Unit 2)
ESF Bus Undervoltage-Start Hotor-Driven Pump NrA. MrA. NrAr 2870 volts 2730 volts
- g. Auxiliary Feedwater Pump Suction Pressure-Low (Transfer to Essential Service I t'
s Water) N.A. N rA. NrAr-- 1.22" lig vac 2" lig vac l
'f 03 7. Automatic Opening of Containment Sump Suction Isulation Valves i 1
- a. Automatic Actuation N;A. - - - - - - N;A; N;Ar- N.A. N.A.
Logic and Actuation Relays
- b. RWST Level-Low-Low NrA. N A. N;A--- 46.7% 44.7%
Coincident with Safety injection See Item 1. above for Safety Injection Trip Setpoints and Allowable Values. s ~s % i _ , a- ,
TABLE 3.3-4 (Continued) g EilGillEERED SAFETY FEATURES ACTUATION SYSfEH INSTRUMENTATION TRIP SETPOINTS e
-TOTAL SENSOR- TRIP ALLOWABLE FUNCTIONAL UNIT ALLOWANCEPA)---{, ERROR-{SE)- SETPOINT VALUE d 8. Loss of Power g .
e- a. ESF Bus Undervoltage N;A. I m NcA. N . A .- 2870 volts 12730 volts j w/1.0s delay w/11.9s delef
- b. Grid Degraded .
Voltage H A.-- N:A. H A:- 3804 volts >3728 volts w/310s delay ad310130s delay
- 9. Engineered Safety feature Actuatfor.
R System Interlocks s
- t. a. Pressurizer Pressure, h P-11 N A. N.A. N- A:- <1930 psIg 51936 psig
- b. Reactor Trip, P-4 N:A. N.A. H-A:- N.A. N.A.
- c. Low-Low T . P-12 N.A. ;;. A. N.A. >550'F 15k F
- d. Steam Generator Water See Item 5.b. above for all Steam Generator Water Level Trip i tevel, P Setpoints and Allowable Values.
I (iiigh-liigh) 1 1 i
. _ _ . _ ~ . _ _ _ _ _ _ _ _ _ . _. . _ _ . _ - _ _ _ . . . _ - __ _______ _ _
,3/4.3
!NSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEEREO SAFETY FEATURES ~'
ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensures that: (1) the associated ACTION and/or Reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is main-tained to permit a channel to be out-of-service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.
~
The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of thest systems is consistent with the assumptions used in the safety analyses. The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability. The Engineered Safety Features Actuation System Instrumentation Trip Setpoints specified in Table 3.3-4 are the nominal va'ues at which the bistables are set for each functional unit. A Setooint is considered to be adjusted consistent with the nominal value when the "as measured" setpoint is within the band allowed for calibration accuracy. To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Setpoints have been specified in Table 3.3-4. Operation with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error. An-opt 4onal-provision-bas-been-4ncluded-for-determining , t h e-O P ERAB R-I TV-o f-a-c ha nnel-whe n-4 t s-T ri p- Se tpo i nt-i s-fo und -to -e x c e e d -t h e - Al-lowable-Value. The-methodology-of-this option-uti-lizes-the "as measured"- dev4*t4en4 rom-t he-spec 444ed-ca44 b r a ti o n-poi nt-fo n-nac k-a nd-sen so n-componenw i 4 n -c o nj u nc t ion-wi th-a-ste tht4c al-comb i n a t4e n-o f-tAe-o t he e-unce rta i n t4 es-of-
-the-4nstrumentat4cn te :csure4he-peac+&&-vae4eble-end4he-unc+rt,a4nt4es-4n, ee44brating-the-4nsttumentet4enr--+n-Equat4en4r3-4, 2 RE- * -SE-t-tar 4he--
interactive-ef fects-of-the-eerort-in-the-rack-and-the-sensor ,-and-the en ,_
. measured'Lvelues-ef-the-eeeees arc- cons 4dered. Z, as specified-in-Tebh 3. 3W i 4n-percent-span, ';=the-s-tet4st4 cal-summation-+f-eerons-essumed-4n4he-analy&46 - '
exc4uding4 hose :ssocieted-wi-th-the-sensor-an'J-cack-defft-and4he-eceveacy-of-their-measurement. TA-oe-Total-Allowance-is-the-dtfference r-4n-percent-spant 4e t we e n -the-Trip-Getpoint-and4he-v e4 u e-us ed-4 n4 he- a n aly s i s-fo e-the-ac tua t4 one ; ilE-orRacKircr is the Eas-measuredu-deviatient-in-peeeent-span r -fon4he- l a f f e c ted-ch anne 4-from-the-specifi ed-Tri p-Se tpoint . SE-or-Sens oHeror-iire4 the* ! BYRON - UNITS 1 & 2 B 3/4 3-1 _ _ _ - - - . ~ _ _ _ _ . _ - _ _ _____ _,_- _ _ ._
INSTRUMENTATION BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUA INSTRUMENTATTUN (N tinued) the "aL4aasured"-deviation-of-the er.;c- f : its-el4beation-pointron-the-yalue-specffied-4n-Teble-373-4,-in percent swh, Use-of-Equation-3-3-1-ailowr-for a sensor tfrift-factor, er,-increased-rackns the-d r44t-facto ra nd-prov4 des-e-t h res hold-veitte-for4 Ef0RTABLE-EVENT S . The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the al.lowanceA of these uncertainty magnitudes. Rack drift e in excess'of met the AT10wable'Value its allowance. ' xhibits the behavior that the rack has n Being that there is a small statisitical chance that this will happen, an infrequent excessive drift is expected. in excess of the allowance that is more than occasional, may be indicative ofRack more serious problems and should warrant further investigation. The measurement of response time at the specified frequencies provides assurance that the Reactor trip and the Engineered Safety Features actuation g associated safety with each channel is completed within the time limit assumed in the analyses. 4 overlapping or total channel test measurements provided that suc strate the total channel response time as defined. Sensor response time veri-fication may be demonstrated by either: (1) in place onsite, or offsite test measurements, or (2) utilizing replacement sensors wIth certified response times. The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents, events, and transients. Once thethose to required logic combination is completed, the system sends actuation signals e ngineered Safety Features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to nitigate the consequences of a steam line break or loss of coolant accident: (1) Safety Injection pumps start and automatic valves position. (2) Reactor trip, (3) feed-water isolation, (4) startup of the emergency diesel generators, (5) containment spray pumps start and automatic valves position, (.6) containment isolation, (7) steam line isolation, (8) Turbine trip, (9) auxiliary feedwater pumps start and automatic valves position, (10) containment cooling fans start and i automatic valves automatic valves position. position, and (11) essential service water pumps start and
.l. ') . , ,
BYRON - UNITS 1 & 2 B 3/4 3-2 AMEN 0 MENT N0. M
SAFETY LIMITS AND LIMITING SAFETY = SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS - REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlock Setpoints shall be set consistent within the Trip Setpoint values shown in Table 2.2-1. APPLICABILITY: As shown for each channel in Table 3.3-1. ACTION:
- a. With a Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 2.2-1, adjust the Setpoint consistent with the Trip Setpoint value.
- b. With the Reactor Trip System Instrumentation or Interlock Setpoint !
less conservative than the value shown in the Allowable Values
-column of Table 2.2)-1. +4%em 2.
. efun-tne-+etpo+nt-conH+tentMM-the-4+4p-Setpofnt-vake-oA
-Teble-2. 2-1 and-determine-within-12--hours-thet-Equeticr. 2. 2 '
was-+at4+f4ed --for the-effec 4ed-eanne4r-oe-
--2,- ,eclare the channel inoperable and apply the applicable ACTION tatement requirement of Specificatior. 3.3.1 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value. ,
fquet4en-2-2-1 2 RE SE i T*-
-Where+-
2- The tulee fer Column-2-of--Table 2. 2-1 foe-the-af4ceted channel,
-AE - The "c: measured"-vake-{h peecent-spanbof-caek-eerer fer the-
--affect +d-channe4r 5E - Either-the "es-measured" +ahe-(4n-percent-spanbef-the-sensor-errer, en-the t>:lue fer-Cohan-GE-(-Sensee-Ecece)-of-Table-fht-1~
4se-the-a44+c-tsd-channeh-snd.
- TA - The sche-foe-Cohen-TA-(-Total-A14ewanceof-Teble-he+4or-4he-af4+eted-channe4-.
a BRAIDWOOD - UNITS-1 & 2 2-3
^ r.
38 TABLE 2.2-1 ' 2: 52 REACTOR TRIP SYSTEM INSTRUMENTITION TRIP SETPOINTS 8 o
' E NSC R-E -T O M I- -- ;R RGR-
] FUNCTIONAL UNIT . A(IOWANCfg " g TRIP SETPOINI All0WABLE VALUE s, 1. Manual Rcstor Trip -M^ N.^ M^ N.A. N.A.
e. s, 2. Power Range, Neutron Flux g gg l36
- a. High Setpoint 7.5 i.55 0 <109% of RTP* $411:i% of RTP*
23.% Low Setpoint
- b. 0.3 4.50 0 <25% of RIP" <ffrft of RTP*
- 3. Power Range, Neutron Flux, 1. 5 0.5 0 <5% of RTP* with <6.3% of RTP* with High Positive Rate 3 time constant a time constant n,o _2 seconds _2 seconds u
- 4. Power Range, Neutron Flux, 1.0 0.5 0 <5% of RTP* with <6.3% of RTP* with l High Negative Rate a time constant a time constant l >2 seconds >2 seconds at #5
- 5. Intermediate Range, -4 7 ^ 0.4 0 $25% of RTP* $3Gr9% of RTP*
Neutron Flux l 1.'17-l 6. Source Range, Neutron Flux 17 ^ 10." 0 -- $105 cps $1:4 x 105 cps l
- 7. Overtemperature AT -2777 Sr38 -- See- See Note 1 See Note 2
'htc 5 l 8. Overp -er AT 4r3 1; 3 1:2 See Note 3 See Note 4 l 9. Pressurizer Pressure-Low 5:0 2:21 1:5 >1885 psig
'M
>teft psig 2.M3
- 10. Pressuri2er Pressure-High 3.1 - - 0. 71- --l . 5 $2385 psig ?2344-psig
! S .55
- 11. Pressurizer Water Level-High 5: 0 - ---2.18 1. 5 $92% of instrument < {~.a of instrument l l span span *
*RTP = RATED THERMAL POWER L .- i
I I u, . s'? I Aill I . 2- 1 (Cont inued) R[ ACTOR IRIP SY:,IIM IN*,TRUMENTATION TRIP SETPOINTS o o o **
-40 int - ERROR- j FUNCTIONAL UNIT
, s 'LLCtlANC1
~ - - - (4A1 2
~ 4W '5E' TRIP SETFOINT ALLOWA8LE VALUE
$12. Reactor Coolant flow-Low 2. 5 .
m 1. 7 ? 0. G- >90% of loop mini- > b of loop ;
~
mum measured flow
- minimum measured i e flow
- I y 13. Steam Generator Water Level Low-Low
- a. Unit 1 N" N ",
N ".
>33.0% of narrow >31.0% of narrow ,
range instrument range instrue nt span 9j span i
- b. Unit 2 -17 0 14-70 iL.3 I' 1.5- >17% (Cycle 3); >15-3% (Cycle 3); !'
K yc!- 3) (Cycic 3) (Cycic 3)- 536.3% (Cycle 4 IF 4%
" N^ (Cyc!a A
*! ^ (Cyc1 A u ^
( CyM - and after) of
! and aft d narrow range Y b ~range fter
) o(Cycle
-wfafM;-) f narrow4 and !
md-after) instrument !
!nstrument span 14 Undervoltage - Reactor G sgg l P_7 0 >5268 volts -
Coolant Pumps >474R- vol ts - l each bus '
- 15. Underfrequency - Reactor each bus
-14.4 13.3 0 OL.c5 I l Coolant Pumps >57.0 Hz >S6.5 Hz i
{ & l 6. Turbine Trip I g ; p a. Emergency Trip Header -# A . loco 8i5 , g N.A. N.A. >54& psig Prersure >E29 psig ;
- b. Turbina Throttle Valve "^
N.A. N.A.- >1% open 9 Closure >1% open 4'17. I Safety Injection Input N A. '. A i i from ESF N . A .-- N. A. N.A. ;
- 18. Reactor Coolant Pump N^ i "A "^
N.A. Breaker Position Trip N. A.
*Ninimum measured flow = 97,600 gpm
s a - E TABLE 2.2-1 (Continued) b y REACTOR TRIP SYSTEM INSTRiNENTAi!ON TRIP SETPOINTS 8 o
-5ENSOR- '
l E -TOTAL ERRG4-FUNCTIONAL bNIT 7 _fSE)-,, TRIP SEIPOINT A&LOWDLC VALUE Z. v -4EiOWRICE y w (gA - v
~ 19. Reactor Trip Systers
- o. Interlocks m
- a. Intermediate Range F.A N. A .H.A_ >l x 10 3C arp ~>6 x 20 28 amp
)
Neutron Flux, P-6
~
l j b. Low Power Reactor Trips ! , Block, P-7 l l ; i
- 1) P-10 input -N _ A . -- N _ A M_ A -
<10% of RIP * ->7.9% to <12.1% of RIP
- g N -
J. 2) P-13 input -H A N-A. " I,- <10% 2TP* Turbine <12.1% RTP* Turbine Impulse Pressure Impulse Pressure Equivalent Equivalect
- c. Power Range Neutron -N. A. u a u. A- <30% of RTP*
<32.1% of RTP*
Flux, P-8
- d. Power Range Neutron 4L A. u g, - u^
-<10% of RTP* >7.9% to <12.1% of RTP*
Flux, P-10 - -
- e. Turbine impulse Chamber "^
"A ".A. <10% RTP* Turbine <12.1% RTP* Turbine Pressure, P-13 Tepulse Pressure impulse Pressure Equivalent Equivalent
{ m
- 20. Reactor Trip Breakers N. A. N.A. N.A N.A. N.A.
@ 21. Automatic Trip and Interlock N.A.- H:A. N A: N.A. N.A.
{ Logic l 22. Reactor Trip Bypass Breakers N:A. - N.A. -N.A. N.A. N.A. t *RTP = RATED THERMAL POWER
m l - 1 CD TAB 1E 2.2-1 (Continued)
*~' I o T ABLE N01ATIONS (Contintei) 6 o
e NOTE 1: (Continued) lag compensator, t ,= 0 s. e T.
= Time constant utilized in the measured Tavg 55.,
u < 588.4*F (Mosinal T ,,g at RAM MM M), T* w e- = 0.00134
% K3
) P
= Pressurizer pressure, psig,
= 2235 psiQ (Mosinal RCS operating pressure).
P'
- S
= laplace transform operator, s 3, top muf bottom detectors of the d instrument
'? and f (aI) is a forv. tion of the indicated difference between
- power-range neutron ion chambers; with gains to be selected base ( on measure response during plant STARTUP tests such that: ,
d c4 +1TC, (Mt-1-Eyele 2 r.d Unit 2 Cycic U. c.d -32'E and +131 (Wit 4 (1) for q g g between h g and ba *** P'*l ;
-Cycle- 3 a.d efter; 'b.it-4-Cyc4 2 :M ef4er)-f (AI) = 0, w ere g #%*f RATED THERMAL POWER in the top and bottom halves of the t. ore respectively, and qt total THEFMAL POWER in percent of RATED THUm&L POWER; ,; ,
exceeds +1GE,-(WitMyele2 e : Uedt Z Oycle 1}, for each percent that the magnit.wh of g .,. - 't-e=4-afted the AT Trip Setpoint (ii) -and 1314Mt I rycle 3 --- et t=cAit 1) end 1.74 (htt -
/
shall be automatically reduced by-2AE-(Mt 1 Cy-?: 2 xd-Wit-2-Cycle-R 4yc" ' '" -fteet-48mit ? Cple 2 c.d-efted of its value at RATED TIEBfEL POWER. l be E s{ g exceeds -32%, the AT trip setpoint shal EI for each percent that the magnitude of q, %1e
- ef tar +4-av.amatically reduced by 1.67% of its vaTue at RATED THE1 MAL PohER (M E (iii)
Unit-2-Cych 2 .dsftee)- y than The channel's maximum Trip Setpoint shell not exceed its computed Trip Setpoint by imore g O NOTE 2: 3:-9% of AT span. 33I
m
> TABLE 2.2-1 (Continued) o 6
o TABLE NOTATIONS (Cintinued) NOTE 3: (Continued)
$ 'Ks =
O 0.00170/*F for T > T" and ( = 0 for T T", w T = As defined in Note 1, m T" = Int cated T avg at RATED THIRMAL POWER (Calibration temperature for AT ins trumenta tion,1 588. 4*F), 5 = As defined in Note 1, and f 2(AI) = 0 for all al. 7 NOTE 4: ' g Theofchannel's e-6% AT span. maximum Trip Setpoint shall stot excred its computed Trip 5etpoint by more
'I. 3 \
M TE 5: The-wnsac--eecer fer terptratur i: 1.2 crd fer p c;;ur M-s-1 , i i i
- ___ _ _ _ _ _ - - - - - ~ ~ '
l l i
- 2. 2 LIMITING SAFETY SYS?2H SETTINGS BASES 2.2.1 REACTOR TRIP SYSlEM INSTRUMENTATION SETPOINTS 1
values at which the Reactor trips are set for each functional u Setpoints have been selected to ensure that the core and Reactor CoolantThe Trip System are prevented from exceeding their Safety Limits during normal operationli and design basis anticipated operational occurrences and to assist the Engi-neered Safety Features Actuation System in mitigating the consequences of accidents. The Setootnt for a Reactor Trip System or interlock function is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy. To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated Allowable Table 2.2-1.Values for the Reactor Trip Setpoints have been specified in but within the Allowable Value is acceptable since an allowan in
=w the safety analysis to accommodate this error. -An-+pt4+nal-prov454on-has '
- Setpcint wuw.44c
!! casrenv.9-t.ns-oruuwM44-os-a-enanne4- wnen-4 tsH e+p --
'c und-t e - exceed - the - A14cwaW44 ue,--The-ee t hodo l ogy-o f-th is - -
(_ wtjem ut!14aes -the ":s-measured"-dev4444on4 rom-the4pecif 4ed-ealibrat4er. rniatJordact Stica ud tensor-components '^ onjunc44on-Wth-e-stethtleebcombin -.- aariahleaM-the er.cect44*t4es-4n~<al4 beat 4ng-the-instrumentati -
--t44n-L24 2 - ME+SE.-4 -7Ar4hed"teroG44V6-+ff+6t*-Of-the-effort i r -- the-pac -Ink-Qus -
---anMt.e4ensey-end--the Eas-measured" v0 kes of--the-ereers--ere-considered. 2,
- n specified b T:bh-G+1, !> percent-spant-is-the-statisticabsummation of-~~-
-+r4er+-466umed-i n- the-a na ly 6 4 6-e x chd i ng 4ho s e-a s s oc4 a ted -w i t h-t he -s e n s o r- a nd
-+ac-A-de44t-4+the -
- .c*cacy-of-the4+-measurementr--TA-oe-TotabAHowance-is
-44.e-di44er+nce , ' percent-+paa r -beteter the-454p-Setpoint-4+the-valve-used----
-P the 2^21ys t; f:- Re:: tee-4e4p. RE c- Rech-Error i, the "as n asured" dc.ic-r
--14en -in-pe+0ent-spany-f+w4Ae-af4ested-channe4-f+om4he4 pee 4f4ed-4*ip-Geti+4e SE-or-Genson-f+ree-44-e4thee-the- "as-eessured"-deviat4on-of-the-sen
-cat 4 bra t4on-point-or-the-va ke- spec ifi ed-i n -Ta bl e-2 ra-1 ;-4 n-percent-spa n the $na1#sassumptions-Use4f-Quauon-24-1---a11ews-fce-a-$ensoe-dei frfrom-t
---feetee r-en-4ecreaeed-eask-dni4-t-4acteer-and-prov4 des-e-threshold-va4ue-fan
-REMRTABLE-EVENT 6v l
The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in excess met of the Allowable its allowance. Being that Value there isexhibits a small the behavior that the rack has not statistical will happen, an infrequent excessive drif t is expected. chance that this in excess of the allowance that is more than occasional, may be indicative ofRack or sen more serious problems and should warrant further investigation. BRAIDWOOD - UNITS 1 & 2 B 2-3
l l - INSTRUMENTAT10_N 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safet) Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4. c' J APPLICABILITY: As shown in Table 3.3-3. ACTION: a. With an ESFAS Instrumentation or Interlock Trip Setpoint less con-servative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 3.3-4 adjust the Setpoint consistent with the Trip Setpoint value, b. With an ESFAS Instrumentation or Interlock Trip Setpoint less con-servative 3.3-4 +-Rhen than the value shown in the Allowable Values column of Table
-1 Adjett-4he-4etpointmnsistent with-theirip-Setpoint vslue-et-J4W_ 2. 3 ^ nd determine + Rhin-12-houes-thet-44:st-ica 2.2 +4+-+a44+f4ed-feMhe affest+d-<han^^1, c*-
(. 2,
/eclarethechannelinoperableandapplytheapplicableACTION statement requirements of Table 3.3-3 until the channel is
.estored to OPERABLE status with its Setpoint adjested consistent with the Trip Setpoint value.
ft;uetica 2.-2--1 - Z RE SE i T A--- dhcrt:
-Z
--ehenneiy The-value-4rc- Cc'omn-4-of-Table 3. 3
- foe-4Aea ffect+4 --
Af- - The "es-eeetwee#-velve-(-4n9ereentepen)-of-cec Weretw
-foMhe f fected4hanaals
-!E - E ithee-the "c; ::asueee"-value-f4n-peeeent-spen)-of-the~
-gen so n-e epo nge- t he-va lue-JeMo lw.n-W-(f e wo e4eeee}-e f-
.Iable 2.3=4-f4.n-4he-af4ec4ed-cM=t 1, ond-JA Tha-value froc-Column TA-(Jotal-Allowance.)-of tor-thaf4eeted-ehenne4,- -Teb4+-4r4+
c. With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3.
, ff m^3 atantroLacom-isolation-noir-eequ4+ed-peice-t+-4n444al cri ticeU y ca Cycle-1.
Ai tiny Builm ng Ventuation actuat4on*,ot-cequired-peier-to-initie4-epera-
- Lion-at ; 20*. 90te4--The :1 Pc e e -{4He)-en-Eyele-1--
BRAIDWOOD UN1151 & 2 3/4 3-13 AMENDMENT N0.-l+ 1
,A
/ '
/
03 TABLE 3.3 -* o , ENGINEERED SAFETY FEATURES ACTUATION SYSTEi! INSTRUMENTATION TRIP SETPOINTS 6 c)
-T&TAL SENSOM-FUNCT10NAL UH1T TRIP Alt 0WABLE , -pt-tO'iniCE 'TA) Z (AR34-65E-}- SETPOINT v v vv - v m - v v , VALUE E 1. Safety Injection 23 (Reactor Trip, Feedwater Isolation, Start Diesel Generators, Containment ** Cooling fans, Control Room Isolation, Phase "A" Isolation, Turbine Trip,
' Auxiliary Feedwater, Containment Vent Isolation and Essential Service Water) h; a. Manual Initiation ~u _ A. "^ "^-.- w - N.A. N.A.
', b. Automatic Actuation "* -F^ M. ." "A Logic and Actuation N. A. N.A.
4 Relays
- c. Containment Pressure-High-1 -5.' J4.L. 8 O.71 1. 5 - < 3.4 psig < -Es& psig
- d. Pressurizer Pressure-j tow (Above P-11) 16.1 $$
l'.il- 1. 5 - 3 1829 psig > 4623 psig
- e. Steam Line Pressure-Low (Above P-11) S- I' . 81- 1:5-- Ldk I 1 640 psig* 3-617 psig*
- 2. Containment Spray
- a. Manual Initiation "A ;
".A. " A. N.A. N.A. ;
4
- b. Automatic Actuation logic and Actuation i
Relays 4L A 4
".A. ".A. N.A. N.A.
- c. Containment Pressure-High-3 8.0 1Lt."2.
0.71 1. 5 < 20.0 psig < P+-A- ns i n
r
.R -
TARLE 3.3-4 (Continued) 55' 2: c3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS k 40fAL SsNSOR- TRIP ALLOWABLE FUNCTIONAL UNIT ?,L ' 0"* "CE--{T '} _CERO: 15E), stTPOINT s vw - u J vw- y VALUE E 3. Containment Isolation il a. Phase "A" Isolation [] 1) Manual Initiatien -,
"^
. ".^ N^ N.A. N.A.
- 2) Automatic Actuation Logic and Actuation Relays "^ "^ "
- a. N.A. N.A. l
- 3) Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and l i
Allowable Values. u, b. Phase "B" Isolation D u, 1) Manual Initiation N^ N^ u ^ N.A. N.A.
- 2) Autom tic Actuation N ^. "^ u 8 N.A. N. A.
Logic anu Actuation Relays
- 3) Containment SUt il Pressure-High-3 S.0 0.71 1.5 $ 20.0 psig 5-fire-psig
- c. Containment Vent Isolation
- 1) Automatic Actestion
= Logic and Actuation Re: lays "^ N ". M . A .- N.A. M.A.
- 2) Manual Phase "A" -H-A-- N^ N ?. N.A. N.A.
Isolation
- 3) Hanual Phase "B" ""
"3 N.A. N.A.
Isolation
- 4) Safety Injection See Itee 1 above for all Safety Injection Trip Setpoints and Allowable Values.
^ m I TABLE 3.3-4 (Continued)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SCTPOINTS C -TOfAL SEN50fE TRIP ALLOWABLE l l ;;
-ALLO'?Ji;C (TA) Z CRB0" (SCl- SETPOINT VALUt FUNCTIONAL UNIT--- .- -- - ,s v -- - -\
% 4. Steam Line Isolation I
- E A. M.A. LAr. N.A. N.A.
E.-. a. Manual Initiation . W w b. Automatic Actuction ! w logic and Actuation H.A. ! J. A. N.A. N. A- N.A. Relays o-
- c. Containment Pressure-9.4 High-2 7.7 0.71 -h6 - 182 psig 19-2 psig
- d. Steam Line Pressure-Lt4
--E 2 - 14.81 - L 5- >640 psig* >fd7- psig*
Low (Above P-11) _
- e. Steam Line Pressure US.3 Negative Rate-High 1Hh5r psi'* I 9.9 01 5 0 1100 psi ** i g (Below P-11) l, y 5. Turbine Trip and !'
m w Feedwater Isolation
- a. Automatic Actuation Logic and Actuation N.A.
Relays -WA. & A. WAt- N.A.
- b. Steam Generator Water level-High-High (P-14) 85.4 ;
I
-fr.-C 4.28 L6- <81.4% of <B2-7% of
- 1) Unit 1 Harrow range Harrow range j E
~ instrument instrument l
-f m
span s S.0 2.15 1. 5 - <l8.1% (Cycle < dgn (Cycle 3); ' 3 2) Unit 2
-(Cyc4: 3) (Cycle (C W <80.8% 782.8% (cycle 4 and E 3);
(CycTe 4 and lif ter) of narrow / E -18.9 3) 12-02 3) 3.2 range instrument E 4 Cycle ' (Cycle---(Eycic 4 after) of ,
* --andaftec} ' and- :,d after)- narrow range span 5 -e f t**}--- instrtment span T,
TABLE 3.3-4 (Continued) m ENGINEERED SAFETY FEATURES _ACTUA1104 SYSTEM INSTRUMENTATION TRIP SETPOINTS 5 SEN50E TRIP Al.LOWASLE g TOTAL -- g FUNCT10NAL UNIT AttOVANM-(t&}--4 r wE --R3R-fSE)-
- . SETPOINT VALUE 1
, o c-S. Turbine Trip and c Feedwater Isolation (continued) !
$ c. Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and f m Allowable Values. -
[ 6. Auxiliary Feedwater f
- a. Manual Initiation N A. N.A. N . A .- N.A. N.A. i
- b. Automatic Actuation l Logic and Actuation !
Relays _N.A._-- -- N A. -- N-A-- N.A. N.n.
- c. Steam Generator Water Level-Low-Low-Start Motor-Driven Pump and
{ Diesel-Driven Pump
- 1) Unit 1 -HA M Ar M.A7 >33.0% of >31.0% of narrow range narrow range instrument instrument span span W .3
- 2) Unit 2 -17,0 14.78 - 125- >17% (Cycle >15-3% Cycie 3);
e (Cycle-3) (Cyc le-3)-(Cycle-3)- 3); >36.3% 5 % (Cycle 4 5 -NrA. N; A.- - N. A z- (CycIe 4 and after) of f i; (Cycle-4 ef+efter) (Cycle - -(Cycle 4 4 end and site after) of @ narrow rang F , narrow range in trument
- after)- instrument span = span F oftZO dg d. Undervoltage-RCP Bus- J- A . N . A . --- - N A- >5268 volts >4728 volts Start Motor Driven Pump i
and Diesel-Driven Pump i
- e. Safety Injection-Start Motor-Driven Pump and bee item 1. above ft,r all Safety Injection Trip Setpoints and Diesel-Driven Pump A1!cuable Values.
N n , FABLE 3.3-4 (Cont inued) 3 o ENGINEERED SAFETY FEA1URES ACTUATION SYSTD! INSTRUMENTATION TRIP SETPOIP65 g h -TOTAL SfMSOP TRIP ALLOWABLE (RF0t($f-}- SETPOINT VALUE FUNCTIONAL UNIT ;AltitWANtiMTjd h - -- w r E 6. Auxiliary Feedwater (Contir.ued) Z
- f. Division 11 for Unit 1 )
(Division 21 for Unit 2) ESF Bus Undervoltage-
" Start Motor-Driven Pump ikA. N A. LAr- 2870 volts 2730 volts l
- g. Auxiliary feedwater Pump Suction Pressure-i Low (Transfer to l w Essential Service 1 Water) 44rA. N. A'. .u. ^ 1.22" Hg vac 2" Hg vac T
~ " Automatic Opening of r 7.
l Containment Sump Suction isolation Valves f
" N.A. N.A.
- a. Automatic Actuation M A. ?' A At Logic and Actuation Relays b, RWST Level-Low-Low -M-A. "A.
u 2- 46.7% 44.7% Coincident with Safety Injection See Item 1. above for Safe ty Injection Trip Setpoints and Allowable Values. 1
^
O ' TABLE 3.3-4 (Coqtinued). O 6 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP S
- 8 TOTAL - = - -
-SENSOR-FUNCTIONAL. UNIT TRIP ALLOVABLE pl10/ANGEdlA) Z- ERM9-{6G- !ETPOINT s m - - VALUE '
g 8. Loss of Power 3-d a. ESF Bus Undervoltage 9 **--- M. A. ---NA 2670 volts >2730 volts c- b. Grid Degraded w/1.8s delay m U/<1.9s delay
~
Voltage -LA M.A. H-A- 3804 volts >3728 volts w/310s delay U/310 1 30s delay
- 9. Engine. red Safety Feature Actuation System Interlocks R
- a. Pressurizer Pressure, P-11 - N . A . - --N:A. --N .- A .- <1930 psig <1936 psig
- b. Reactor Trip, P-4 -N.A. -
-- N . A r- N,A- N.A. N.A.
- c. Low-Low T3yg, P-12 -H.A. Sd .1 MA -MA 1550*F 1h47,4*F
- d. Steam Generator Water level P-14 See Item 5.b.
Setpoints above for Values. and Allowable all Steam Generator Water Level Trip (High-High)
l 3/4.3 INSTRUMENTATION , l [ BASES _ -.. i 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES {TfiXT10N SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensures that: (1) the assot.f ated ACTION and/or Reactor trip will be inititted when ~the parameter monitored by each channel or combination thereof reaches its Setpoint., (2) the specified coincidence logic is maintained, (3) sufficient redundancy is main-tained to permit a channel to be out of-service for testing or maintenance, and (4) sufficient system functional capability is availablo from diverse parameters. The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and niitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses. The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to l demonstrate this capability. The Engineered Safety Features Actuation System Instrumentation Trip C ~ -Setpoints specified in T&ble 3.3-4 are the nominal values at which the bistables are set for each functional unit. A Setpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy. To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Seipoints have been specified in Table 3.3 a. Operation with Setpoints less conservative than the Trip Setpoint but withi.) the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error.
- Opti^a:1 p r icier W beer N !uded fer--dete + %g-
--the 0? EDA 941-7-Y-# : chamel when its-Teip-Setpoint i; found-to-exceed-th
-A14ewable Value.-The- methodclogy-of-this-option-uti442es-the 8as-measured" ht4en f c t h e-speeM ied- cal i b ra ti on-point-fo r-ra c k-and-s e ns o r-compo nent s
-4n-confueet4cn-with-antatistical-ecmMnat4on-of-the-other uacertetnt4es-of
-4he-instrumentation-to-meastwe-the process-verieMe-and4.he-uneertaint4es-4n
-eaMbrating the insttumentation. In-Equation-3+-1, Z ^ RC ^ SE 1 TArthe l -fateract4vc cf fcct+-of-the-errors-4n-the+-rack-and-the-sensort-and-the Ha;
-measured" values-of-4he-eercr: cre-consVdered. Z, as-spee H4ed-in-Table-3 + 4r l
40-peec ent-spa n;-i s-the-s ta t i s t i cal-s ummation-o f-errors- assumed-i-n-the-ana lys i s-
-ext 4uding-those-a s s oc i a te d - w i tt - the-sens o e-e nd- rac k-&Mt-end-t he-acc u racy- o f-
-theie-measucement.--TA-or-Total- Allowance-is-the-dif ference r -in-percent-spent-
--between-the-Tr ip-Setpoin t-end-the-v a l ue-us ed-i n- the--an alys i s-f or-th e-actua ti e nt
-RE-oe4eck-Eeroo-ts-the "es-measured"-deviationrin-percent-span;--for-the----
-affeeted-channel from the specified-Trip-Setpointe -SE~or-Sensor-Error-is-eithc.+-
l i l . BRAIDWOOD - UNITS 1 & 2 B 3/4 3-1 T
($ INSTRUMENTAT!g BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYS INSTRUMENTATION (Continued)
-the Eas-meessted" deviet40.t-of-th; ;ca ;r frc; it: ::l f bration-point-or-the. '
wak; spcified in Tebia 3.3-t, 'n percent ;pa, im t.% :n: lysis-essumptions.
-Vee-of-Eqtrat4erHk3-1-allows-for-e-s+nsor-dr+ft-factor r-en-increased-rack--
drif t faster, :nd prov4 des,-a-threshold v;12: f:r "E*09Tf"LE-P!ENTS.- The methodology to derivo the Trip Setpoints is based upon combining all of the uncertainties in the channels. Inherent to the detemination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensor and rack instrumentation utilized in these channels.are expected to be capable of operating within the allowance:, of these uncertainty magnitudes. Rack drift in excess met of the Allowable Value axhibits the behavior that the rack has not its allowance. Being that there is a small statisitical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drif t, in excess of the ailowance that is more than occasional, may be indicative of more serious problems and should warrant furthat investigation. The measurement of response time at the specified frequencies provides assurance that the Reactor trip and the Engineered Safety Features actuation associated with each channel is completed within the time limit assumed in the safaty analyses. - Response time may be demonstrated by any series of sequential,
-overlapping or total channel test acasurements previded that such tests demon-i strate the total channel response time as defined. Sensor response time veri-fication may be demonstrated by either: (1) in place, onsite, or offsite test measurements, or (2) utilizing rephcement sensors with certified response times.
The. Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetemined limits are being exceeded. If. they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents, events, and transients. Once the required logic combination is completed, the system sends actuation signals to those Engineered Safety Features compenents whose aggregate function best serves the requirements of the condition. As an example -the following actions l may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss of coolant accident: (1) Safety Injection pumps start and automatic valves position, (2) Reactor trip (3) feed-L water isolation, (4) startup of the emergency diesel generators, (5) containment L L spray pumps start ano automatic valves position, (6) containment isolation, t- (7)' steam lina' isolation, (8) Turbine trip (9) auxiliary feedwater pumps start and automatic valves position, (10) cuntainreent cooling fans start and automatic valves. position
. automatic valves position., and (11) essential service wrter p aps start and BRAIDWOOD - UNITS 1 & 2 B 3/4 3-2 AMENDHENT NO.42-
ATTACHMENT C EVAL.UATION DE SIGNIFICANLHAZARDS CONSIDERATIONS CECO has evaluated this aroposed amendment and determined that it involves no significant hazards consicerations. According to 10 CFR 50.92(c), a proposed amendment to an operating license involves no significant hazards considerations if operation of the f acility in accordance with the proposed amendment would not:
- 1. Involve a significant increase in the probability or consequencae of an accident previously evaluated; or
- 2. Create the possibility of a new or different kind of accident from any accident pieviously evaluated; or
- 3. Involve a significant reduction in a margin of safety.
The basis for this determination of no significant hazards considerations is presented below. Unit (s): 1, 2 Applicable Mode (s): 1,2,3,4 O' hor Relevant Plant Conditions: None System (s) affected: EF,RP Equipment Name(s): Allowable Values (AVs) for various Engineered Safety Features Actuation System (ESFAS) Instrumentation AVs for various Reactor Trip System (RTS) Instrumentation RTS Turbine Trip Emergency Trip Header Pressure Trip Seipoint
/ sci:lDG15:30
Describe the proposod change and the reason for the change. As a result of the Setpoint Study, CECO is requesting changes in AVs for the RTS and ESFAS instrumentation listed in Table 1. NED calculations support the requested changes for RTS and ESFAS instrumentation AVs. The methodology used is the same that used by Westinghouse and documented in WCAP 12583, Westinghouse Setpoint Methodology for Protection Systems, May 1990. NED reviewed the assumptions of calibration tolerances and MTE and revised them to reflect those that will actually be used at the stations. No changes have been made that affect the ability of RTS or ESFAS instrumentation to perform their intended design functions. CECO is requesting that the Turbine Trip Emergency Trip Header Pressure Trip Setpoint and AV be revised from >540 psig to >1000 psig and >520 psig to >815 psig, respectively. The requested changes to the Turbine Trip Emergency Tnp Header Pressure Trip Setpoint and AV are conservative with respect to the current Technical Specificaibn s 11ues. This change would result in a reactor trip sooner than the existing setpoint followirg a decrease in the main turbine emergency trip header pressure. The purpose of the reactor trip on turbine trip above 30% power (P-8) due to decreasing electro-hydraulic fluid in the emergency trip header is for equipment protection and is anticipatory in nature. This reactor trip is not taken credit for in any accident analyses. CECO is requesting that the TA, Z, and SE values currently in Technical Specification Tables 2.2-1 and 3.3-4 be deleted from those tables. As a result of this proposed change, Equation 2.2-1 would also be deleted from Technical Specifications 2.2.1 and 3.3.2, the corresponding action statements would be changed to reflect the deletion of Equation 2.2-1, and the corresponding Technical Specification Bases would also be changed to reflect the deletion of Equation 2.2-1. The rec uested change to delete the TA. Z, and SE values from Technical Specification Tab es 2.2-1 and 3.3-4 will not affect plant operation and is conservative with respect to determining channel operability. The TA, Z, and SE values are only used to evalute the operability of an instrument channel that has been determined to be in excess of the AV. The instrument channel would still be operable ii Technical Specification Equation 2.2-1 was satisfied: Z + RE + SE <; TA where RE is the "as reaasured" value of Rack Error for the affected channel
/scl:lD615:31
The TA, Z, and SE values are only applicable if there is excess margin between the associated channel's AV and the safety analysis limit. If excess margin is not available, then the values of TA, Z, and SE are listed as N.A. (Not Applicable)in Technical Specification Tables 2.2-1 and 3.3-4. The Setpoint Study would require revision to the TA, Z, and SE value for 12 of the 22 RTS Functional Units and 7 of the 9 ESFAS Functional Units. As a result, ap3roximately three fourths of the TA, Z, and SE values listed in Technical Specification " ables 2.2-1 and 3.3-4 would be N.A. since the study used varying amounts of excess margin to maintain the safety analyses setpoints at their current values. As a practical matter, Equation 2.21 is not used to determine operability because the channel is declared inoperable and the appropriate action statements are followed during surveillance testing and the channel must be restored to within its associated AV prior to returning the channel to service. CECO is requesting that the cyde specific requirnments contained on page 2-8, Note 1, Parts (i), (ii), and (iii) be deleted since they are no longer applicable. CECO is requesting that the cycle specific relief for BRNPS Technical Specification 3.3.2 be deleted since it is no longer applicable. List the applicable Safety Analysis Report (SAR) sections which describe the affected eystems, structures, or components (SSCs) or activities. Also list the SAR accident analysis sections which (5iscuss the affected SSCs or their operation. List any other
, Trtle 'l0 Code of ,Federal controlling documerits such as Safety Evaluation Reports Regulations (10CFR), Regulatory Guides, previous modi (SERs)fications or Safety Evaluations, etc...
SAR Chapter 7 Sections 2 and 3 SAR Chapter 15
/scl:lD615:32
- . -- - -.- - - . .__-=_- ._. _ _~ Desenbe how the change will affect plant operation wtwn changed SSCs function as intended (i.e., focus on system operation / interactions in the absence of equipment failures).~ Consider all applicable operating modes. Include a discussion of any changed interactions with other SSC's. Review the following areas for interactions: Mechanical, Electrical, I&C, Structural, Fire Protection, Environmental Qualifications, Site / Environmental impacts, Radiological Concerns, Security Concerns and Flooding, and discuss adverse affects. The proposed chances to the AVs for RTS and ESFAS Instrumentation will continue to ensure that the associated RTS or ESFAS actuation signals will be generated when required within the bounds of the plant's safety analyses. The proposed changes to the RTS Turbina Trip Emergency Trip Header Pressure Trip Setpoint and AV wiligenerate a reactor trip signal quicher than the current Technical Specification values due to the conservative nature of the change.
- The proposed removal of the TA, Z, and SE values from the Technical Specifications will have no effect on plant operations.
The proposed removal of cycle specific relief no longer applicable from the Technical Specifications will have no effect on plant operations. Desenbo how the change will affect reactivity management. Some of the proposed changes to the AVs for RTS and ESFAS instrumentation are in the conservative direction with respect to the current Technical Specification value. This will cause the associated RTS or ESFAS actuation to occur sooner when the monitored parameter exceeds its AV. This will have an overall positive affect on reactivity management. The proposed change to the Turbine Trip Emergency Trip Header Pressure Trip Setpoint and AV will initiate a reactor trip sooner than the currently allowed Technical Specification values thus providing a positive affect on reactivity management. Many of the proposed changes to the AVs for RTS and ESFAS instrumentation are in the nonconservative direction with respect to the current Technical Specification value. This will cause a delay in the associated RTS or ESFAS actuation when the monitored parameter exceeds its AV. This will have an overall negative affect on reactivity management. However, all of these changes in the nonconservative direction are bounded by the plant's safety analyses and therefore, the negative affect on reactivity management is acceptable. The proposed removal of the TA. Z, and SE values from the Technical Specifications will have no effect on reactivity management. The proposed removal of cycle specific roliof no longer applicable from the Technical Specifications will have no effect on reactivity management.
/scl:lD615:33
i Describe how the change will affect equipment failures. In particular, desciibe any now failure modes and their impact during all applicable Operating modes. The proposed changes to the AVs for RTS and ESFAS instrumentation will only affect the setpoint at which a piece of equipment is actuated. No physical equipment changes are being made, and therefore, no new equipment failure modes are being introcuced as a result of these proposed changes. . l The proposed changes to the RTS Turbine Trip Emergency Trip Header Pressure l Trip Setpoint and AV will will affect the setpoint at which a reactor trip signal is generated as a result of decreasing pressure in the main turbine emergency trip header. No physica! equipment changes are being made, and therefore no new equipment failure modes are being introduced as a result of theso proposed changes. The proposed removal of the TA, Z, and SE values from the Technical Specifications will have no effect on plant equipment, and therefore, no new equipment failure modes are being introduced as a result of these proposed changes. The proposed removal of cycle specific relief no longer applicable from the Technical Specifications will have no effect on plant equipment, and therefore, no new equipment failure modes are being introduced as a result of these proposed changes. i l l l
/ sci:ID615:34
Identify each accident or anticipated transient (i.e., large/small break LOCA, loss of load) described in the SAR where any of the following is true: The change alters the initial conditions used in the SAR analysis The changed SSC is explicitly or implicitly assumed to function during or after the accident Operation or failure of the changed SSC could lead to the accident Other ACCIDENT: High-High D-4 SG Water Level Turbine Trip and Feedwater Isolation Actuation SAR SECTION: 15.1.2 ACCIDENT: Loss of Reactor Coolant System (RCS) Flow Reactor Trip SAR SECTION: 15.3.1 4 List each Technical Specification (Safety Umit, Limiting Safety System Setting or Limiting Condition for Operation) where the requirement, associated action items, associated surveillances, or bases may be affected. TECHNICAL SPECIFICATION SECTION(S): 2.2.1 Table 2.2-1 3.3.2 Table 3.3-4
/scl:lD615:35
To determine if the probability or the consequences of an accident or malfunction of equipment important to safoty previously ovaluated in the SAR may be increased, answer the following questions for each accident listed. Provide the rationale for all NO answers. Answer separately for each accident that is affocted in a different manner. Affected Accident: High-High D-4 SG Water Level Turbine Trip and Feedwater Isolation Actuation SAR Section: 15.1.2 May the probability of the accident be increased? The probability of this accident occurring will not increase. Sufficent redundancy of equipment exists to ensure that the appropriate actuation signals are generated when the monitored parameters exceed their associated trip setpoints. May the consequences of an accident (offsite dose) be increased? The consequences of this accident will not be increased. The increase in the AV for the D-4 SG High-High Water Levs Turbine Trip and Feodwater Isolation Actuation will delay the generation of those actuation signals approximately 2.5 secords from the time they would have occurred using the current Technical Specification AV. Increasing the Safety Analysis Limit (SAL) to accommodate the proposed AV will have no appreciable effect on the Departure from Nucleate Boiling Ratio (DNBR) since it is effectively constant at the time of signal actuation and remains well above the Departure from Nucleate Boiling (DNB) limit throughout the entire transient. May the probability of a malfunction of equipment important to safety increase? The probability of a malfuntion of equipment important to safety will not increase. There will be no change in plant equipment as a result of these proposed changes. Sufficent redundancy of equipment currently exists to ensure that the appropriate actuation signals are generated when the monitored parameters oxceed their ae sociated trip setpoints. l.
/scl:lD615:36
. . . . - . - - . - - . - . - ~ _ . . . - . . - . - _ . _ . . . -. ..
May the consequences of a maNunction of equipment important to safety increase? The consecuences of a malfuntion of equipment important to safety will not increase. There will be no change in plant equipment as a result of these proposed changer. Sufficent redundancy of equipment cunently exists to ensure that the appropriate actuation signals are generatad whan the monitored parameters exceed their associated trip setpoints. Affected Accident: Loss of RCS Flow Reactor Trip SAR Section: -15.3.1 May the probability of the accident be increaseoi The probability of this accident occurring will not increase. Sufficent redundancy of equipment exists to ensure that the appropriate actuation signals are generated when the monitored , parameters exceed their associated trip setpoints. May the consequences of an accident (offsite dose) be increased? The consequences of this accident will not be increased. The decrease in the AV for RCS Flow Low can be accounted for by decreasing the SAL without affecting the outcome of the safety. analysis For the Locked Rotor / Shaft Break event, there will be no change in the tirae in which a reactor trip is initiated since the reduction in RCS flow is so rapid. For the Partial Loss of Forced Flow event, the change in the SAL will result in a delay of the initiation of the reactor trip signal by less than 0,1 second. This change will not significantly affect the DNBR transient. May the probability of a malfunction of equipment important to safety j increase? - The probability of a malfuntion of equipment important to safety will not increase. There will be no change in plant equipment as a result of these proposed changes. Sufficent redundancy of equipment currently exists to ensure that the appropriate actuation signals are
- generated when the monitored parameters exceed their associated l- trip setpoints, i.
l l l
/ sci:lD615:37
l . May the consequences of a ma!! unction of equipment important to safety increase? The consequences of a malfuntion of equipment important to safety will not increase. There will be no change in plant equipment as a result of these proposed changes. Sufficent redundancy of equipment currently ex sts to ensure that the appropriate actuation signals are generated when the monitored parameters exceed their associated trip setpoints. Based on the answers above, does the change adversely impact systems or functions so as to create the possibility of an accident or malfunction of a type different imm those in the SAR? Describe the rationale for this answer. The possibility of a new or different type of accident will not be created as a result of these proposed changes. Except for the two types of accidents previously discussed, these proposed changes were already bounded by the existing safety analyses. For the two accidents previously discussed, the correspending SALs were changed to bound the proposed changes without affecting the outcomes of the corresponding safety analyses. Determine if parameters used to establish the Technical Specification requirements are changed. If no Technical Specifications are impacted, then no reduction in margin of safety exists in the context of this question.
, There is no reduction in the margin of safety from these proposed changes. Except for the two types of accidents previously discussed, these proposed changes were already bounded by the existing safety analyses. For the two accidents previously discussed, the corresponding SALs were changed to bound the proposed changes without affecting the outcomes of the corresponding safety analyses.
I
/ sci:lD615:38
1 -
SUMMARY
_OF_THELSIGNIElGNRHAZARDS CONSIDERATIONS
=
Commonwealth Edison Company (CECO) has evaluated this proposed amendment and determined that 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 pr( oability or consequences of an accident previcusly 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 basis for this determination of no significant hazards considerations is presented below: CECO is requestina numerous changes in the Allowable Values (AVs) for the Reactor Trip System (FITS) and Engineered Safety Features Actuation System (ESFAS) instrumentation listed in Technical Specification Tables 2.2-1 and 3.3-4. CECO Nuclear Engineering Department (NED) calculations support the requested changes for RTS and ESFAS instrumentation AVs. The methodology used is the same that used by Westinghouse and documented in WCAP 12583 Westinghouse Setpoint Methodology for Protection Systems, May 1990. NED reviewed the assumptions of calibration tolerances and measurement and test equipment (MTE) and revised them to reflect those that will actually be used at the stations. CECO is requesting that the Turbine Trip Emergency Trip Header Pressure Trip Setpoint ani AV be revised from 2540 psig to 21000 psig and 2520 psig to ;>815 psig, respectively. The requested changes to the Turbine Trip Emergency Tnp Header Pressure Trip Setpoint and AV are conservative with respect to the current Technical Specification values. This change would result in a reactor trip sooner than the existing setpoint following a decrease in the main turbine emergency trip header pressure. The purpose of the reactor trip on turbine trip above 30% power (P-8) due to decreasing electro-hydraulic fluid in the emergency trip header is for equipment protection and is anticipatory in nature. This reactor trip is not taken credit for in any accident analyses.
/scl:lD615:39
~. . . = - - = - - - - - -
l CECO is requesting that the Total Allowance (TA), Z, and Sensor Error (SE) values currently in Technical Specification ^1 ables 2.21 and 3.3-4 be deleted from those tables. As a result of this proposed change, Equation 2.2-1 would also be deleted from Technical Specifications 2.2.1 and 3.3.2, the corresponding action statements would be changed to reflect the deletion of Equauon 2.2-1, and the corresponding Technical Specification Bases would also be changed to reflect the deletion of Equation 2.21. The requested change to delete the TA, Z, and SE values from Technical Specification Tables 2.2-1 and 3.3-4 will not affect plant operation and is conservative with respect to determining channel operability. The TA, Z, and SE values are only used to evalute the operability of an instrument channel that has been determined to be in excess of the AV. The instrument channel would still be operable if Technical Specifica: ion Equation 2.2-1 was satisfieo: Z + RE + SE <; TA where RE is the 'as measured" value of Rack Error for the affected channel The TA, Z, and SE values are only applicable if there is excess margin between the associated channel's AV and the safety analysis limit. If excess margin is not available, then the values of TA, Z, and SE are listed as N A. (Not Applicable) in Technical Specification Tables 2.2-1 and 3.3-4. The Setpoint Study would require revision to the TA, Z, and SE value for 12 of the 22 RTS Fur.ctional Units and 7 of the 9 ESFAS Functional Units. As a result, approximately three fourths of the TA, Z, and SE values listed in Technical Specification Tables 2.2-1 and 3.3-4 would be N.A. since the study
- used varying amounts of excess margin to maintain the safety analyses setpoints at their currei., values. As a practical matter, Equation 2.2-1 is not used to determine operability because the channel is declared inoperable and the appropriate action statements are followed during surveillance testing and the channel must be restored to within its associated AV prior to returning the channel to service.
CECO is r9 questing that the cycle specific requirements contained on page 2-8, Note 1, Parts (i), (ii), and (iii) be deleted since they are no longer applicable. CECO is requesting that the cycle specific relief for BRNPS Technical Specification 3.3.2 be deleted since it is no longer applicable. The proposed changes to the AVs for RTS and ESFAS Instrumentation will continue to ensure that the associated RTS or ESFAS actuation signals will be generated when required within the bounds of the plant's safety analyses. The proposed changes to the RTS Turbine Trip Emergency Trip Header Pressure Trip Setpoint and AV will generate a reactor trip signal quicker than the current Technical Specification values due to the conservative nature of the change. The proposed removal of the TA, Z, and SE values from the Technical Specifications will have no effect on plant operations. The proposed removal of cycle specific relief no longer applicable from the Technical Specifications will have no effect on plant operations.
/scl:lD615:40
Some of the proposed changes to the AVs for RTS and ESFAS instrumentation are in the conservative direction with respect to the current Technical Specification value, This will cause the associated RTS or ESFAS actuation to occur sooner when the monitored parameter exceeds its AV. This will have an overall positive affect on reactivity management. Many of the proposed changes to the AVs for RTS and ESFAS instrumentation are in the nonconservative direction with respect to the current Technical Specification value. This will cause a delay in the associated RTS or ESFAS actuation when the monitored parameter exceeds its AV. This will have an overall negative affect on reactivity management. However, all but two of the proposed changes were already bounded by the existing safety analyses. For those two proposed changes that were not, the Safety Analysis Limit (SAL) for the corresponding accidents were changed to bound the proposed changes without affecting the outcomes of the accident analyses. Therefore, the negative affect on reactivity management is acceptable. The proposed change to the Turbine Trip Emergency Trip Header Pressure Trip Setpoint and AV will initiate a reactor trip sooner tnan the currently allowed Technical Specification values thus providing a positive affect on reactivity management. The proposed removal of the TA, Z, and SE values from the Technical Specifications will have no effect on reactivity management. The proposed removal of cycle specific relief no longer applicable from the Technical Specifications will have no effect on reactivity management. The proposed changes to the AVs for RTS and ESFAS instrumentation will only affect the setpoint at which a piece of equipment is actuated. No physical equipment changes are being made, and therefore, no new equipment failure modes are being introduced as a result of these proposed changes. i The proposed changes to the RTS Turbine Trip Emergency Trip Header Pressure Trip Setpoint and AV will will affect the setpoint at which a reactor trip signal is generated as a result of decreasing pressure in the main turbine emergency trip header. No physical equipment changes are being made, and therefore, no new equipment f ailure modes are being introduced as a result of these proposed changes. The proposed removal of the TA, Z, and SE values from the Technical Specifications will have no effect on plant equipment, and therefore, no new equipment failure modes are being introduced as a result of these proposed changes. The proposed removal of cycle specific relief no lonaer applicable from the Technical Specifications will have no effect on plant equipment, and therefore, rn new [ equipment failure modes are being introduced as a result of these proposed changes. The possibility of a new or different type of accident will not be created as a result of these pro 30 sed changes. Except for the two types of accidents previously discussed.11ese proposed changes were already bounded by the existing safety analyses. For the two accidents previously discussed, the corresponding SALs were changed to bound the proposed changes without affecting tho outcomes of the corresponding safety analyses.
/scl:lD615:41 i
.. . . - . . - . _ - = . . . - . . - . _ - . . - .-_. . ..-.-. __. _-. .. .-
There is no reduction in the margin of safety from these proposed changes. Except for the two types of accidents previously discussed, these proposed changes were already bounded by the existing safety analyses. For the two accidents previously discussed, the corresponding SALs were changed to bound the proposed changes without affecting the outcomes of the corresponding safety analyses. l t l
/scl:lD615:42
ATTACHMENT D ENVIRONMENTAL ASSESSM"NT , 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 provided for under 10CFR51.22(c) (9). . The proposed change involves revisions to RTS and ESF Allowable Values and one ESFAS trip setpoint. No setpoints for containment isolation or other systems that could impact radiation releases are affected. All accident assump*. ions are preserved, or have been successfully re-evaluated. Also, none of the proposed changes involve . Irreversible concequences. The proposed change does not involve a significant hazards consideration as discussed in Attachment C to this letter. Also, this proposed amendment will not involve significant changes in the types or amounts of any radioactive effluents nor does it affect any of the permitted release paths. In addition, this change does not involve a significant increase in individual or cumulative occupational exposure. Therefore, this change meets the categorical exclusion permitted by 10CFRS1.22(c)(9). T l
/scl:lD615:43
ATTACHMENT E Revised Values for TA, Z and SE The following RTS and ESFAS Instrumental n values for TA, Z and SE were calculated as part of the Setpoint Study. The calculated values are included here for information only. "(N.A.)"is listed after the values for cases where the TA, Z and SE for this instrumentation cannot meet the Equation 2.2-1 if the setpoint is found outside the Allowable Value. This information will be maintained in CECO administrative programs and maintained through the 50.59 process. 0
/ sci:lD615:44
BEACIOELTBIP_SXSIEMJ N STRUM E NTAIlON_ TRIP _.SETEQlNIS TOTAL SENSOR EUNCIIRNALUMI ALLOWANCEJTA) 2. ERBOB.fSE)
- 1. Manual Reactor Trip N.A N.A. N.A.
-2. Power Range, Neutron Flux
- a. High Setpoint 7.5 4.56 0
- b. Low Setpoint 8.33 4.56 0
- 3. Power Range, Neutron Flux, 1.58 0.5 0 High Positive Rate
- 4. Power Range, Neutron Flux, 1.58 0.5 0 High Negative Rate
- 5. Intermediate Range, 17.0 8.41 0 Neutron Flux
- 6. Source Range, Neutron Flux 17.0 10.01 0
- 7. Overtemperature AT 9.65 (N. A.) 5.26 (N.A.) (P)1.79(N.A.)
(T)1.33(N.A.)
- 8. Overpower AT 4.75 (N. A.) 1.54 (N. A.) 1.88(N.A.)
- 9. Pressurizer Pressure-Low 5.0 (N.A.) 1.0 (N.A.) 2.5 (N. A.)
- 10. Pressurizer Pressure-High 6.9 (N.A.) 5.0 (N.A.) 1.5 (N.A.)
- 11. Pressurizer Water Level-High 5.0 (N.A.) 2.18 (N. A.) 2.75(N.A.)
- 12. Reactor Coolant Flow Low 4.45 (N.A.) 3.6 (N.A.) 1.1 (N.A.)
- 13. Steam Generator Water Level Low-Low
- a. Unit 1 19.3 (N. A.) 15.1 (N.A.) 2.5 (N.A.)
b Unit 2 17.7 (N. A.) 15.08 (N.A.) 2.51(N.A.)
- 14. Undervoltage - Reactor Coolant Pumps 8.35 0.27 0
- 15. Underfrequency - Reactor 44.4 13.3 0 Coolant Pumps l
l l l-l ( /scl:lD615:45 L 1 -
l BE ACIOR_T RI P_ SYSTEM INST BUMENT AllON IRIP_SETPOINIS TOTAL SENSOR FUNCTIONAL UNII ALLOWANCE (TA) Z ERROR (SE)
- 10. Turbine Trip
- a. Emer0ency Trip Header N.A. N.A N.A.
Pressure
- b. Turbine Throttle Valve N.A. N.A. N.A.
Closure
- 17. Safety ection Input N.A. N.A. N.A.
from E
- 18. Roactor Coolant Pump N.A. N.A. N.A.
Breaker Position Trip i 19. Reactor Trip System Interlocks
- a. Intermediate Range N.A. N.A. N.A.
Neutron Flux, P-6
- b. Low Power Reactor Trips Block, P-7
- 1) P-10 input N.A. N.A. N.A.
- 2) P-13 input N.A. N.A. N.A.
- c. Power Range Neutron N.A. N.A. N.A.
Flux, P-8 , d. Power Range Neutron N.A. N.A. N.A.
- Flux, P-10
- e. Turbine impulse Chamber N.A. N.A. N.
Pressure, P-13
- 20. ReactorTrip Breakers N.A. N.A. N.A.
- 21. Automatic Trip and Interlocks N.A. N.A. N.A.
Logic
- 22. Reactor Trip Bypass Breakers N.A. N.A. N.A.
s
/scl:lD615:46
- . . . ~ .- -.. . . - - . . - - - . -. .. - -. - - . .
ENGlNEEBED_SAEETX_EEATURES. ACTUATION SYSIEM. INSTRUMENTATION TRIP.SETPOINIS TOTAL SENSOR EUNG_TJONAkUEI ALLOWANCE (TA) 1 EBBOR(SE)
- 1. Safety injection (Reactor Trip, Feedwater Isolation, Start Diesel Generators, Containment Cooling Fans, Control Room isolation, Phase"A" isolation, Turbine Trip, Auxiliary Feedwater, Containment Vent Isolation and Essential Service Water)
- a. Manual Initiation N.A. N.A.- N.A.
- b. Automatic Actuation N.A. N.A. N.A.
Logic and Actuation Relays
- c. Containment Pressure-High-1 5.7 0.71 2.55
- d. Pressurizer Pressure-Low (Above P-11) 16.1 (N. A.) 11.01 (N.A.) 2.5 (N. A.)
L
- e. Steam Line Pressure-Low (Above P-11) 21.23 11.01 2.57 i 2. Containment Spray
- a. Manual Initiation N.A. N.A. N.A.
- b. Automatic Actuation Logic and Actuation Relays N.A. N.A. N.A.
- c. Containment Pressure-High-3 8.0 0.71 2.55 i
i
/scl:lD615:47.
. . . - . , , . - . . . - . . . . . - , _ - - - - . - - . . - . - - . . . ~ .
ENGINEEBED. SAFETY EEAIU_RES. ACIVATION_ SYSTEM INSIRUMENTAIION TRIE SEIEOJNIS TOTAL SENSOR FUNCTIONAL UNIT ALLOWANCE (TA) Z ERROR (SE)
- 3. Containment isolation
- a. Phase "A" Isolation
- 1) ManualInitiation N.A. N.A. N.A.
- 2) Automatic Actuation Logic and Actuation Relays N.A. N.A. N.A.
- 3) Safety injection See item 1.
- b. Phase "B" Isolation
- 1) ManualInitiation N.A. N.A. N.A.
- 2) Automatic Actuation N.A. N.A. N.A.
Logic and Actuation Relays
- 3) Containment Pressure-High-3 8.0 0.71 2.55
- c. Containment Vent isolation
- 1) Automatic Actuation Logic and Actuation Relays N.A. N.A. N.A. _
- 2) Manual Phase "A" N.A. N.A. N.A.
Isolation
- 3) Manual Phase "B" N.A. N.A. N.A.
Isolation
- 4) Safety injection See item 1
/scl:lD615:48
ENGINEEBED_SAEETXEEATURESAGIVAT.lON_ :XSIEMJNS.TBUMENTATION IBIP_SEIEOI R IS TOTAL ' SENSOR F_UNCIJQNAkV. Nil ALLOWANCElTA) 1 EBROR.(SE)
- 4. Steam Line isolation
- a. Manual Initiation N.A. N.A. N.A.
- b. Automatic Actuation Loalc and Actuation Relays N.A. N.A. N.A.
- c. Containment Pressure- -
High-2 7.7 0.71 2.55
- d. Steam Line Pressure- 21.23 11.01 2.57 )
Low (Above P-11)
- e. Steam Line Pressure Negative Rate-High (Below P-11) 5.92 0.5 0
- 5. Turbine Trip and Feedwater isolation
- a. Automatic Actuation Logic and Actuation Relays N.A. N.A. N.A.
- b. Steam Generator Water Level-High-High (P 14)
- 1) Unit i 18.6 4.8 2.5
- 2) Unit 2 18.9 12.02 2.51 _
- c. Safety injection See item 1
/scl:lD615:49
ENGINEERED SAFETXFEATURES ACTUATION. SYSTEM INSTRUMENTATION IRIP_SETP_OJNIS TOTAL SENSOR EUNGILQNALMNlI ALLOWAN. GELT _6) 1 EBROB.ISE)
- 6. Auxiliary Feedwater
- a. Manual Initiation N.A. N.A. N.A.
- b. Automatic Actuation Logic and Actuation Relays N.A. N.A. N.A.
- c. Steam Generator Water Level-Low-Low-Stari Motor-Driven Pump and Diesel-Driven Purnp
- 1) Unit 1 19.3 (N. A.) 1.51 (N.A.) 2.5 (N. A.)
- 2) Unit 2 17.7 (N. A.) 15.08 (N.A.) 2.51(N. A.)
- d. Undervoltage-RCP Bus- N.A. N.A. N.A.
Start Motor Driven Pump and Diesel-Driven Pump
. e. Safety injection.
Start Motor-Driven Pump and See item 1. Diesel Driven Pump
- 1. Division 11 for Unit 1 (Division 21 for Unit 2)
ESF Bus undervoltage - Start Motor Driven Pump N.A. N.A. N.A. l g. Auxiliary Feedwater Pump l Suction Pressure-Low (Transfer to Essential Service Water) N.A. N.A. N.A. ! 7. Automatic Opening of
- Containment Sump Suction Isolation Valves-
- a. Automatic Actuation- N.A. N.A. N.A.
l Logic and Actuation Relays l l b. RWST Level-Low-Low l Coincident with l Safety injection l
. /scl:lD615:50 '
l l EN GIN E E B ED_SAEETIEEATU R ES ACT U ATION.SYSTE M_IN STRUM ENI AIlON IRIP_SETfQlNTS TOTAL SENSOR EVNGIlONAL.UNlT ALLOWANCE _(IA) .Z_ EBROR_(SE)
- 8. Loss of Power
- a. ESF Bus Undervoltage N.A. N.A. N.A.
- b. Grid Degraded Voltage N.A. N.A. N.A. .
- 9. Engineered Safety Feature Actuation System Interlocks
- a. Pressurizer Pressure, P-11 N.A. N.A. N.A.
- b. Reactor Trip, P-4 N.A. N.A. N.A.
- c. Low-Low Tavg, P-12 N.A. N.A. N.A.
- d. Steam Generator Water See item 5.b Level, P-14 (High-High)
. /scl:lD615:51
.. ,}}