Text

Proposed Tech Specs Re Limiting Safety Sys Settings 2.2.1 & Associated Table 2.2-1
	ML20199G543
Person / Time
Site:	Beaver Valley
Issue date:	01/18/1999
From:	; DUQUESNE LIGHT CO.
To:	;
Shared Package
ML20199G537	List: ML20199G534; ML20199G543;
References
	NUDOCS 9901220325
	Download: ML20199G543 (67)
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ATTACHMENT A-1 , ~. Beaver Valley Power Station, Unit No. 1 l License Amendment Request No. 220 The following is a list of the affected pages: i Affected Pages: License page 4 2-5 l 2-6 i 2-7 i 2-7a ! 2-8 2-9 l 2-10 B 2-3 i B 2-5 B 2-6 3/4 3-14 3/4 3-19a 3/4 3-22 3/4 3-22a 3/4 3-23 i 3/4 3-24 l 3/4 3-24a 3/4 3-24b 3/4 3-31a , B 3/4 3-1 i B 3/4 4-1 ; 9901220325 990118 PDR ADOCK 05000334 p PDR

! i l DPR-66 i l (next 9 age is 6) (3) Less Than Three looo Ooeratign _ Duquesne Light Company-sha?1 not Operate the rc :t:.r :t p:w:.v-level-s

Sev- P ' ( : defined-in-Table-3+1-ef-S meificetkn 3.3.1.1 e'-%e Technhl-Spec 4f4c: tion:, App;ndix A) witi less ther three (3) maetor AD ~N s

          '    ceolant 12 cps-in-operation-unt44-safety :n:ly::: fer 4ess-than-t4mee
              -leep Operation-have-been-submitted by the licansaae W annrnval for less then thr c leap cperatien at power levels ebeve P-7 has-been r nted by the Cc--ir:icn by ==d:ent :f thi: license.

(4) Steam Generator Water Rise Rate Deleted per License Amendment No. 24. (5) Fire Protection Procram The license r @shall implement and maintain in effect all-provisions of th , approved fire protection program as described in the Update - . l l Amendment No. 6, M, M, M, M, 33, 39, M6, l c:rruted by letter dated .'+ n :e. = l i (froposed Wod^g ) ,

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.o O SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS I
2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS h Inierl*@ N. 2.2.1 TheYeactor [ rip 8 stem Instrumentation 5tpoints shall set W consistent with the Trip Setpoint values shown in Table 2.2-1."' Aco k APPLICABILITY: AS SHOWN FOR EACH CHANNEL IN TABLE 3.3-1. ACTION: i fith: re::::r trip :y:t-- intrumentatica set?ciat less censer'/:tive M' l th:r th: v:1;: :hr.., in th: "leed!: '! !ues ce!u- ef T:b!: 2.21, i et ;
~
d::1:r: the ch:- el ineper ble ead app 1y the er?i4ceb1, ne m n stat,_ l
=nt r:quirr:nt of Specificatica 3.3.1 1 uat41 the t henael 4 ester:j
OPEM"LE :::te: eith it: trip ::tpef-t adjusted cen:ittent '. ith th:[
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( 2.) Tkt Tr.g tin,k nu d st S c.kp...d. udue sbkd f.e Nskiond11.6 Ad-c6 va\sc. N Jwsk 4he. sekro'd canMh'd "% he tk pi d ve.bc. s w \ ie.w mE ad p dm ht Stk o$"\ f S Tr.g w,h w ewe %sked %ler.uc. LJ .c A. h%
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DELE 7&
AO-06 BUOER VALLEY -- UNIT 1 2-5 L (frymJ wl-Q
Attechacnt A-1
,e EOcycr Vollcy Pow 0r Station, Unit No. 1 License Amendment Reaggst No. 21Q Insert "A" ,
I
a. With a Reactor Trip System Instrumentation or Interlock Setpoint l
outside its calibration tolerance band but more conservative than the value(s) shown in the Allowable Values column of Table , 2.2-1, adjust the Setpoint to within the established calibration l tolerance band of the Nominal Trip Setpoint.(1)(4
b. With a Reactor Trip System Instrumentation or Interlock Setpoint i
less conservative than the value(s) shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION atatement requirement of 1 Specification 3.3.1.1 until the channel is restored to OPERABLE l status with its setpoint adjusted to within the established calibration tolerance band of the Noninal Trip Setpoint. (1)(2) BEAVER VALLEY - UNIT 1 (Proposed Wording)
~
DPR-66 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT //OM /A/AL TRIP SETPOINT ALLOWABLE VALUES l i
1. Manual Reactor Trip Not Applicable Not Applicable

2. Power Range, Neutron Flux Low Setpoint - $ 25* of RATED Low Setpoint - s 27.3% of RATED THERMAL POWER THERMAL POWER f
High Setpoint - [109% of High Setpoint - 5 111.3% of RATED THERMAL RATED THERMAL POWER POWER
3. Power Range, Neutron Flux, % of PATED THERMAL POWER s 6.3% of RATED THERMAL POWER with a time High ?ositive Rate with a time constant 2 2 constant 2 2 seconds roeconds

4. Power Range, Neutron Flux, % of RATED THERMAL POWER s 6.3% of RATED THERMAL POWER with a time High Negative Rate with a time constant 2 2 constant 2 2 seconde seconds

5. Intermediate Range, 5% of RATED T!!ERMAL POWER s 31.1% of RATED THERMAL POWER Neutron Flux
() 8
6. Source Range, Neutron Flux 10' counts per second 5 1.4 x 10' counts per second

7. Overtemperature AT See Note 1 See Not e 3

8. Overpower AT See Note 2 See Note 4 ()

9. Pressurizer Pressure--Low 945 peig 2 sig

10. Pressurizer Pressure--High 385 psig s 2394 psig

11. Pressurizer Water 92% of instrument span s 93.9% of instrument span f
* ~
Level--liigh 94
12. Loss of Flow - of design flowa per loop 2 % of design flow
per loop
(* Design flow is 87,200 gpm per loop. 3*p00 AD- M> HEAVER VAI.I.EY - UNIT 1 2-6 Amendment No. W ([fo[03t O M
TABLE 2 2-1 (bntinued) ' E} REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS I A D* OS N
= .ev j < FUNCTIONAL UNIT Atom sp> A L TRIP SETPOINT 19.7 ALLOWABLE VALUES -
; D S~
C j 13. Steam Generator Water Level-Low-Low M span-each h of w range instrumer,t steam genarator
> 4% of narrow range ins rument span-each steam I ' generator E
y 14. Steam /Feedwater Flow 40% of full steam flow at < 43.4% of full steam flow
~ Mismatch and Low Steam RATED THERMAL POWER coincident at RATED THERMAL POWER Generator Water Level with steam generator water level coincident with steam , i g generator water level O 25% of narrow range instru- 2 23.1% of narrow range ment span-each steam generator instrument span- steam o generator o 3 15. Undervoltage-Reactor veFtrs-each bus 2 283 V eii.s-each bus
{yw Coolant Pumps knemm\ bws ve%g h %%) be gab
! 16. Underfrequency-Reactor 57.5 Hz - each bus 2 57.4 Hz - each bus Coolant Pumps
17. Turbine Trip 1 42.
~
b I A. Auto stop oil pressure 45 psig g=5 psig $5 i . k g:. 56 ?$ D D lI E. B. Turbine Stop Valve 2 1% open 2 1% open 3 $ E 5 18. Safety Injection Input Not Applicable Not Applicable 2 from ESF w P f 19. Reactor Coolant Pump Not Applcable Not Applicable Breaker Position Trip
20. Reactor Trip System Interlocks /
? -11 Amps k1 A. Intermediate Range @ 1 x 10-10 Amps 2 6 x 10 Neutron Flux, P-6 1
7 f oc 6 bod u~k be n.8 fAoD
<* to w w. ugm l A0-o5~
A0-0G
-6 +4 U TAlllE 2.2-I $ front i nuedb -DELETE J
s1
s HEACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS y 20. (Continued)
FUNCTIONAL UNIT M p/4 TRIP SETPOINT ALLOWABLE VArJrES c B. Power Rangp 7 30% RATED THERMAL POWER
$ Neutron Flux, P-8 5 32.3% RATED THERMAL POWER a
e C. Power Range 49% RATED THERMAL POWER Neutron Flux, P-9 1 51.3% RATED THERMAL POWER D. Power Range 10% RATED THERMAL POWER Neutros; Plux, P-10 2 7.7% and i 12.3% RATED (Input to P-7) - THERMAL POWER m
$u o
E. Turbine Impulse 10% of RTP Turbine Impulse i 12.3% of RTP Turbine i Chamber Pressure, P-13 4 y (Input to P-7) ressure Equivalent Impulse Pressure Equivalent u E P - 6(M M a A I
o 8
"\
m
- ) AD-06
4 MDO M oW TABLE 2.2-1 Tant l an=,e tv 0 RETE Mg;,qeg l4, h ) y #< REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS cgD NorATION 5PM
, [osere y ol Gf !
j@ k NOTE 1: Overtemperature AT M < ATo K i -K 5 TW Ir -T' +K3 (P-P') -f (AI)
' LWj 2(1+'11+t 2 ) S bTS $ QT = =.heasNdA Indicated AT at RATED THERMAL POWERe. & c where: ATO r l T Average temperature, *F b k-Oh T = F (indicated Tavg at RATED THERMAL POWER)
P = Pressurizer pressure, psig m P' = 2235 psig (indicated RCS nominal operating pressure)
"P 1+t S = The function generated by the lead-lag controller for Tavg dynamic }y4 u sation.
b t1 & v2 = constants utilized in the lead-lag controller for Tavg
$ tg = 30 secs, 12 = 4 secs.
[1 .
= Lag compensator on measured AT 1+t con ant util in t ag compen or for aT,I secs.
g 14
= .\e nf & f asured T /
5 e r 1 = Lag c, sator Ad '
= Time constant utilized in the laq compensator for Tavg F2' secs. l h 15 )
g ( - S = Laplace transform operator. hc\-
~
QPA-66}kOC ~
. I AD-oE i
TABLE 2.2-1 (Continued),ogggg h REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS N # \gj e_ l M W NOTATION (Continued) t( - 0 Op g [ Operat with loops olate oops Op lon with 2 Loops (1 loop o ated) jg J [lc i c- K; = 1.18 Kj
.99 K = 1.1 l 5
K 2
= 0.01655 = .01655- K = 655 2
K 3
= 0. m 01 K' 3 = 0. 01 K = 0.0 01 m # m and f (al) is a function of the indicated difference between top and bottom detectors a of the power-range nuclear ton chambers; with gains to be selected based on measured g instrument response during plant startup tests such that:
b (i) for q between -23 percent and + 11. percent, f (al) = 0 g (wherbq-qIndq are percent RATED THERMAL POWER in the top and bottom o halvesofthechrerespectively,andqt*9b is total THERMAL POWER in g percent of RATED THERMAL. POWER). W (ii) 9 exceeds -23 percent, d for the ATeachtrip perce'nt setpoint shall thatbethe magnitude of automatical (qfy rebu)ced by 1.54 percent of g its value at RATED THERMAL POWER. 2 53 (iii) -q exceeds + 11 percent, 3 foreachpercentthatthemagnitudeof(qfyreNu)cedby1.91percentof the aT trip setpoint shall be automatical 2 its value at RATED THERMAL POWER. P j Ad-OS 1 ~
~ ~
g 88-GG Ah @-d k,, biut i FADLE 2.2-1 es g# i ,gA t ' CC; _ .. . :L"20,, i
$ i O* oas7E- REACTOR TRIP SYSTEM INSTRUMEidATION -TRIP SEiruINTS M--H ---
N, -
$f Ol f oneTE OW - NOTAT. ION ,,
hfMOTE2:.A Overpower3&T ' C ' A T, K4 -K5fiS1 g T-K 6 T -T" -f(aI) ' W4h - sff 7 A00gt 2 %Q GCSA
'NW{
Q.o Q c where: AT g = Indicated a T at RATED THERMAL POWER i [ T = Average temperature, 'F (, ,i T" = Indicated Tayg at. RATED THERMAL POWER S F d K4 = 1.07 K5 = 0.02/*F for increasing average temperature K6 = 0.00128 for T > T"; K6= (0) for T S 7" S = f,, h The ea=ft function generated by the rate lag controller for Tavg dynamic id g o g = d o m.'lon sa bM6 j 06
,3 A gime constant utilized in the rate lag controller for T b 1 avg, t = 10 secs. l - = Lag c nsator on measure 7d ORETE --* * , ime ant uti ed i e lag com sator 1 ' 8 compensa r on sured Tay *f , / /
N z
,o 25 =
Time natant utilized in the lag compensator for Tavg $
\ ,, S =
Laplace transform operator. secs.] f(aI) = gi 0 for all AI f ad P o *'" ' ' ' ' N NOTE - The channel's maximum trip shall not exceed its computed trin @ t b 1o percentQ q ., g, s~r,f G,,4 a c. ,s,,s e, d'g,.,,s2 2,, a, J,e} y more tha n I N : The channel's_ maximum trip
),6 m percent shall not exceed its computed trip ;:_.-i. by more tha n - , g,, a v, T~~ d a r inpdf q /&,udr% Nok2D- - _ _ _ -
jW%
- - _ . _ _ . . . _ . - _ - __ - _ . - - - - - - ~ _ _ -. - ~ + A00 W
l . GLETG 2.2 LIMITING SAFETY SYSTEM SETTINGS - BASES REPLPc5 WrrH TNSEK,T"F" 2.2.1 REACTOR TRIP SET POINTS Th: n;;;;;r T?ip 5:tp; int Limit: :p;; i:d in T:b1: 2.2-1 cre 'the 1 es-at-which-the4eactor4ripsare-set-for-each par:m;ter. The Trip i (-0I 1: y:1u:: h:v: b::n ::10:ted t: :::ure tha.t th: r:::t:r :Or: :nd r::: tor Opera-p
';:01:nt-system-are-prevented 4 rem-exc :di ; their esf:ty 1"it:.
tien aith a Trip-Setpoint4ess-conservative-than it: Setpoint44mit-ht aithin its-speeffied-*14eweMe-Veitte-is-acceptable en the beiis thet 02: 5 A14ewaMe-Vahe-i+-equal-to-ec-4ess-than-the drMt all w:ne: 4 . L::cm:d-to occur for-eech-te4p-used-4n-the-eee4 dent :n:1y:::. Manual Reactor Trip
+ --{The Manual Reactor Trip is a redundant channel to the automatic 3 protective instrumentation channels and provides manual reactor trip .
capability. , Power Range, Neutron Flux 4-i.The Power Range, Neutron Flux channel high setpoint provides reactor
; core protection against reactivity excursions which are The too rapid icw setto be p.
f protected by temperature and pressure protective circuitry. point provides redundant protection in the power range for a powerThe OE trip a l excursion beginning from low power. setpoint may be manually bypassed when P-10 is active (two of the four poyer range channels indicate a power level of above approximately (Fper cent of RATED THERMAL POWER) and is au t below approximate 1yppercent of RATED THERMAL POWER). Power Range Neutron Flux, High Rates I
*---fThe Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of red ejection a from any power level.
Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from partial power. Ad-oG I B 2-3 1 BEAVER VALLEY - UNIT 1
V .
Y(opoSN l
Attcch=;nt A-1 133Gvar Vallcy Powar Station, Unit No. 1 License Amendment Recuest No. 220 Insert "F" The Reactor Trip Setpoints specified in Table 2.2-1 are the nominal values (with the exception of Functional Unit 17.B) at which the l Reactor Trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the reactor 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 Engineered Safety Features Actuation System in mitigating the consequences of accidents. The setpoint for a reactor trip system or interlock function is considered to be adjusted consistent with the nominal value when the "as left" setpoint is within calibration tolerance band. The calibration tolerance band associated with the nominal trip setpoint is specified in plant procedures. 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 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. The Limiting Safety System Settings are . defined as the Trip Setpoints. I l I 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 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. Limiting Safety System Settings 2.2.1 is modified by two notes. Note (1) permits a trip setpoint to be set more conservative than the Nominal Trip Setpoint as necessary in response to plant conditions or revised analysis. The conservative direction is established by the direction of the inequality applied to the Allowable Value. It is consistent with the setpoint methodology for the "as left" trip setpoint to be outside the calibration tolerance band but in the conservative direction with respect to the Nominal Trip Setpoint provided that the Allowable Value(s) is adjusted accordingly. For BEAVER VALLEY - UNIT 1 (Proposed Wording) l l l
=_. _. . _ _ _ . - . . _ .
I l Attach: nt A-1
. Banvar Valley Powar Station, Unit No. 1 Liegnse Amendment Recuest No. 220 l
Insert "F" (continued) example, the Power Range Neutron Flux High trip setpoint may be set to a value less than specified in Table 2.2-1 during a plant i condition where the Heat Flux Hot Channel Factor - Fg(Z) exceeds the technical specification limit. In addition, an increase in the Steam Generator Water Level-Low-Low Setpoint may be required to account for revised environmental and process measurement allowances. The Setpoint would be increased in the conservative direction in order to meet the revised analysis assumptions. Note (2) states that Functional Unit number 17.B in Table 2.2-1 for Turbine Stop Valve position is not a nominal value. Therefore, the trip setpoint is adjusted to be consistent with the Trip Setpoint value specified in Table 2.2-1 in lieu of adjusting the setpoint within the established calibration tolerance band of the Nominal Trip Setpoint. l l 1 l l l l l 1 BEAVER VALLEY - UNIT 1 (Proposed Wording) l
~ @d-6GpA00 l
O LIMITING SAFETY SYSTEM SETTINGS
'ht)-OG EASES f v DELETE !0peration with e resetomolant-4 cop-out-of-service-below-the-3 joloop-P-8-set-point-does-not-require reaetor-protection-system-se mod 444c4 Mon-because the " S Oct point and assee4eted-trip-will-prevent ?
D4&-during-h-loop-operat4sn-exc4u;ive-ef the Overtemperature aT cet O point. 2 loop-operation-above-the-3-loop P 8 set-point-45-permis-
sible-after-resett4ng the K1, K2 and-K3-inputs-to-the-Overtemperature 0 AT-channel; and -raising the P-3 Oct point-to-its 2 loop-value. In I thi; mode-of-operat4eMhc P S interlock-and-trip-funet4 ens-es-a-High ]
4eutron cl= trip at the-reduced-power-leveh J Overpower AT , W The Overpower AT reactor trip provides assurance of fuel integrity, } e.g., no melting, under all possible overpower conditions, limits the required range for Overtemperature AT protection, and provides a backup to the High Neutron Flux trip. The setpoint includes corrections for changes in density and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop tempera-ture detectors. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified j trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System. [g 06 Pressurizer Pressure { M The Pressurizer High and Low Pressure trips are provided to limit , the pressure range in which reactor operation is pennitted. The High Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig). The Low Pressure trip provides protection by tripping the reactor in the event of a loss of reactor coolant pressure. Pressurizer Water Level MThe Pressurizer High Water Level trip ensures protection against Reactor Coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent water relief 1 J BEAVER VALLEY - UNIT 1 B 2-5 g (frope kd*b l l
f hf> A -G6 M
. LIMITING SAFETY SYSTEM SETTINGS BASES
! through the pressurizer safety valves. No credit was taken for l operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System. Loss of Flow 9 o The loss of Flow trips provide core protection to prevent DNB in the $ event of a loss of one or more reactor coolant pumps. { } Above W percent of RATED THERMAL POWER, an automaticfreactor trip will occur if the flow in any two loops drop below 94% of nominal g v3 full loop flow. Above % (P-8) of RATED THERMAL POWER, automatic 9Cd reactor trip will occur if the flow in any single loop drops below ' 94% of nominal full loop flow. /Thi 1 tter trip will prevent th'e '
'Sini :: v: ue c: ne er:P rrer geing belev the de ign DNSR li=it during n:rr:1 eperational trenrients =ad anticipat=d transients " hen 2 leepe ere in operation end the 0"er+==p-reture T trip setreint is AS-
, adjusted to the " slue specified for all leep ir p:rstien. "ith the o4 I overt =perature ; T trip ::tprint djuct:d t; the value specified for t 2 1:0; Oper:tien, the P-S trip at SSt RATED THER"AL POWER with 10:p J 6<d
tcp v lver open--and at ?lt RATED TurgyAt pour? "ith a leep stop T velv: 100:d vill prevent the rinitur value of the DMSR frer geing i balsw the d::ign ONS" li=it during n:r=:1 Operst!rnal transients awu$ i l gnticip:ted trencient: "ith 2 loops-in-operation 4 [ ~
l Steam Generator' Water Level OM 1 The Steam Generator Water Level Low-Low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity. The specified setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays of the auxiliary feedwater system. l Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level The Steam /Feedwater Flow Mismatch in coincidence with a steam Generator Low Water Level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall . l l l 1 l BEAVER VALLEY - UNIT 1 B 2-6 Miendment No.144 (fr.pach LA)orO& I
DPR-66 INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE kCTUATION SYSTEM INSTRUMENTATIO LIMITING CONDITION FOR OPERATION 3.3.2.1 The [ngineered [afety [eature Ictuation Iystem instrumentation channeJs and interlocks shown in Table 3.3-3 shall be I g OPERABLE with their 4 rip sTetpoints set consist with the valuesj Ao-$ s shown in the Trip Setpoint column of Table 3.3-4. O APPLICABILITY: ADD l As shown in Table 3.3-3. { ACTION: 00 MW 4 9 hMdM p W th an engin [a ty dature [ctuation Iystem Q strumentation rip /etpoint less conservative l than the valueLS hown in the Allowable values column Table 3.3-4, declare the channel inoperable and apply the h' s ! of MC* applicable ACTIONS requirements of Table 50 i 3.3-3 until the 9 annel is restored to. OPERABLE 5( l I kS status with 4the-- 4enD oint adjusted %:nciatent with the Trip Satsoint." I m g l i
. With ur%an& ethMo44 cc.Wcsbon %\criste. W A *V 4Nr.W*ovatTo gngineered gafety Eeature Ectuat. ton /rystem ~ . i C instrumentation channel inoperable, take the
n shown AD* N in Table 3.3-3. g .gg SURV'EILLANCE REQUIREMENTS 4.3.2.1.1 Each engineered safety feature actuation system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL l FUNCTIONAL TEST operations during the modes and at the frequencies j shown in Table 4.3-2.
4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by interlock operation. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation. 4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESF function shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESF function as shown in the " Total i No. of Channels" Column of Table 3.3-3.
< h0D TNEg u g, o,o(
BEAVER VALLEY - UNIT 1 3/4 3-14 Amendment No. 240 l l Nep3Eh U t
Attcchtsnt A-1
." Banvar Valley Pow 3r Station, Unit No. 1 License Amendment Reauest No. 220 Insert "B"
a. With an Engineered Safety Feature Actuation System Instrumentation or Interlock Trip Setpoint outsido its j calibration tolerance band but more conservative than the value(s) shown in the Allowable Value column of Table 3.3-4, adjust the Setpoint to within the established calibration tolerance band of the Nominal Trip Setpoint. W j
l l l BEAVER VALLEY - UNIT 1 (Proposed Wording)
._ _ _ _ __ _ m , _ __ . . _ _ _ . _ . , i l
Atte.cPc?nt A-1
. bc0ver Valley Power Station, Unit No. 1 License Amendment Reauest No. 220 Insert "C" i
i (1) A Trip Setpoint may be set more conservative than the Nominal ; Trip Setpoint as necessary in response to plant conditions or . revised analysis. Allowable Value(s) shall be adjusted ! accordingly. 1 1 J l l l i I i l BEAVER VALLEY - UNIT 1 (Proposed Wording)
- TABLE 3.3-3 fContinued)
DPR-66 i ENGINEEprn SAFETY FEATURE ACTUATION SYSTEN INsixtTni.siATION MINIMUM TOTAL NO. CHANNELS, CHANNELS FUNCTIONAL UNIT APPLICABLE OF CHMfMELS TO TRIP OPERABLE MODES
' ACTION
7. AUXILIARY FEEDWATER

a. Steam Gen. Water Level-Low-Lowj(-hoop seep r
{wehresopen) h t*._ gd
i. Start Turbine Driven 3/sta. gen. 2/sta. 2/sta. gen. 1, 2, 3 14 Pump gen. any sta. gen.
ii. Start Motor 3/sta. gen. 2/sta. 2/sta. gen. Driven Pumps 1, 2, 3 14 any 2 sta. gen. any 2 gen. sta. gen.
b. Undervoltage-RCP (Start (3)-1/ bus 2 2 1 Turbine Driven Pump) 14

c. S.I. (Start All Auxiliary Feedwater Pumps) See 1 above (all S.I. initiating functions and requirements) i l f;xyoar;n t w w w ~ )

e. Trip of Main Feedwater 1/ pump V
1 1 1, 2, 3 18 Punpa (Start Motor Driven , ' Pumps) d.(OFLETEO)) : f BEAVER VALLEY - UNIT 1 3/4 3-19a Amendment No. [ O D c -- _ - - - -- _ _ - - - - - - - - - _ - - - - - _ - _ - - _ - - - - - - _ - - - - - _ - - - - - - - - - - - - - - - _ - - - - - _ - - - - _ _ - - - - - - _ _ _ - - - _ - - - - - - - - - _ - - _ _
5 Aon4PA4D A6-*6 TABLE 3.3-4
~
I 1. M 8 '
$ ENGINEERED SAFETY FEATURE ACTUATION SYSTEN INSTRUNENTATION TRIP 5
e FUR lCTIgMAL UNIT M# #MA L- TRIP SETPOINT e , ar.rmars var 23ES N 1. f
- SAFETY INJECTION, TURBINE TRIP AND FEEDNATER ISOIATION
a. Manual Initiation Not Applicable Not Applicable

b. Automatic Actuation Iagic Not Applicable Not Applicable

c. Containment Pressure--High 5 psig
5 sig 7, d. Pressurizer Pressure-Low e.

w e.
845 psig 2 1830 psig Pd! V Steauline Pressure-law 500 psig steam 4 ine pressure 2 488 psig steam b
*t line pressure i
Gi at f tr a a a
.F x
Ad-c6 4
g d o12-6 h8 @ oG g TABIJI: 3.3-4 icontinnaedt
RgGIggERED SAFETY FE&TURE ACTUATION. SYSTEM INSTRUMENTATION TRIP SETPOINTS I \
, FUNCTIONAL UNIT dM iMA TRIF_SETFOINT ALIANABIE VAIBES '
i n 1.1 SAFETY IN3ECTION-TRANSFER FROM ( H INJECTION TO TNE RECIRCULATION MODE
a. Manual Initiation Not Applicable Not Applicable
, b. Automatic Actuation imgic D:ot Applicable Not Applicable '
o Coincident with Safety i
? u Injection Signal I
c. Refueling Water Storage Tank 18'8-1/2" 2:18'2-1/2" and 519'2-1/2" !
i.v.1-i S ,. U h
d. Refueling Water Storage Tank Level - Auto QS Flow' Reduction 8'6" 18'0" and 59*0"
$s t) l M ll' l ,,
;;' - i T;: ;
ii* Jl !
>a . ~
3- f y )go-o6 , b
TABLE 3.3-4 (Continued) Ao & PR~Gk 1)-@ N;
s ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTHUMENTATION TRIP SET N
P E W 4ONAL UNIT dM 1 AJ
, TRIP SETPOINT Ar_imnLE yArJIES
c 2. CONTAINMENT SPRAY Z ,
y a. Manual Initiation ; Not Applicable Not Applicable
b. Automatic Actuation Imgic Not Applicable Not Applicable

c. Containment Pressure--High-High k w 3. CONTAINNENT ISOLATION
.0 psig sh% l 2 N }. a. Phase "A" Isolation p\
k w" 1. Manual Not Applicable Not Applicable
2. From Safety Injection Not Applicable g
Automatic Actuation Logic ' Not Applicable
;::] i b. Phase "B" Isolation v { '
l !
1. Manual Not Applicable Not Applicable

2. Automatic Actuation Logic Not Applicable Not Applicable '
k 3. Containment Pressure-- .O psig 5%g
! High-High g/
e '
} AC-CE
TABLE 3.3-4 (Continued) DPR-66 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS f FUNCTIONAL UNIT g gjugl} TRIP SETPOINT ALLOWABLE VALUES J
4. STEAM LINE ISOLATION

a. Manual Not Applicable Not Applicable

b. Automatic Actuation Logic Not Applicable Not Applicable

c. Containment Pressure-- h.0psig s g
f Intermediate-High-High 4
j , i
d. Steamline Pressure-Low @ 500 psig steam 2 488 psig steam li pressure line pressure l

e. High Negative Steam Pressure 6/100 psi with a time 5 127 psi with a time f Rate constant 2 50 seconds constant 2 50 seconds l S. TURBINE TRIP AND FEEDUATER ISOLATION Y7 I

a. Steam Generator Water Level (g) of narrow range s % of narrow range !
High-High instrument span each instrument span each j steam generator steam generator j t
6. LOSS OF POWER !

a. 1. 4.16kv Emen Mcy Bus 5% of nominal bus 2 % of nominal bus Undervo29vfo (Loss of Voltage) voltage with a 1 i 0.1 voltage with a 1 i 0.1 l (Trip reen) second time delay jseondtimedelay !

2. 4.16kv Emergency Bus @ 75% of nominal bus 2 74% of nominal bus I (Start Diesel) voltage with a < 0.9 sec- voltage with a < 0.9 sec- ,
93.7 ond time delay (includes and time delay (includes auxiliary relay times) auxiliary relay times)
b. 4.16kv Emergency Bus Undervoltage % of nominal bus 2 % of nominal bus ;
(Degraded Voltage) voltage with a 90 i 5 voltage with a 90 i 5 i 93.7 second time delay econd time delay ;
/
c. 480v Emergency Bus Undervoltage (Degraded Voltage) 7 % of nominal bus voltage with a 90 1 5 2 % of nominal bus '
voltage with a 90 1 5 second time delay second time delay i I i BEAVER VALLEY - UNIT 1 3/4 3-24 Amendment No. 2*2 (MropeJN *
". NM.Y da6 ' h8-@
TAB! LJ_-4 iContinued) Q ENGINEERED SAFETYJEATURE AMTIOD TSTEM IffSTRUMENTATION TRIP SETPOINTS
&' l %m i FUNCTIONAL UNIT //441^/4 TRIP SETPOINT \e .
ALLOWhBLE _VAlllES g 2. i.uxxtzAae rasoWATsa f r. / ,</,3 g,o1
- a. Steam Gerwrator Water Level-low-low i of narrow range > of narrow range instrument span each ._ ins rument span each i
steam generato steam generato j (~)f oYn omh,A l oj, vs omin <r l \
b. Undervoltage - RCP M V 6 RCP bus 3
voltage
> M ve85s RCP bus E Y voltage g~I
c. S.I.
See 1 above (all SI Setpoints) a q h . nderv [ olt M
> -o3 -
g e. Trip of Main Feedwater Not Applicable e Pumps Not Applicable E. B i A. (.O EWTEO) 1 F
+
L TABLE 3.3-4 (Continued) DPR-66 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT er A/A TRIP SETPOINT ALLOWABLE VALUES
8. ESF INTERLOCKS 1

a. P-4 Not Applicable Not Applicable i

b. P-11 -
000 psig 5; 2010 psig
c. P-12 (F 5 4 1 1- 2 539 F ,
f BEhVER VALLEY - UNIT 1 3/4 3-24b Amendment No.4HC' (Next page is 3/4 3-29) m (,fr y esed W sc [
~
TABLE 4.3-2 (Continued) , DPR-66 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS h * (MMD J - 7 CHANNEL MODES IN WHICH-CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REOUIRED
7. AUXILIARY FEEDWATER

a. Steam Generator Water S R Q 1, 2, 3 -
Level-Low-Low i
b. Undervoltage - RCP S R M 1, 2

c. S.I. See 1 above (all SI surveillance requirements)
{*/fM Y / f v 7 J / MA fel*3
e. Trip of Main N.A. N.A. R 1, 2, 3 l Feedwater Pumps

8. ESF INTERLOCKS

a. P-4 N.A. N.A. R 1, 2, 3

b. P-11 N.A. R Q 1, 2, 3

c. P-12 N.A. R Q 1, 2, 3 t
R (,ft Eo BEAVER VALLEY - UNIT 1 3/4 3-31a Ismendment No.-WPI _ _ . - _ - _ _ _ _ . _._.____.._..___.m- _ _ ____.______.__=.___.________.____..____.______.__________._..._.______.m_ _ _ _ _ _ . _ _ . _ . _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ .
_._m _ _ _ _ . . _ . . _ _ _ _ . _ _ _ _ - - _ _ _ _ _ _ . _ . _ . _ . . _ __ ___ DPR-66 3/4.3 INSTRUMENTATION BASES 3 /4. 3.1 AND 3 / 4. 3. 2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESP) INSTRUMENTATION The OPERABILITY of the protective and ESF instrumentation systems and interlocks ensure that 1) the associated ESF action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof exceeds its setpoint, 2) the specified coincidence logic and sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance consistent with maintaining an appropriate level of reliability of l the Reactor Protection and Engineered Safety Features ! instrumentation, and 3) sufficient system functional capability is available for protective and ESF purposes 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 these systems is consistent with the assumptions used in the accident 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 l surveillance tests performed at the minimum trequencies are sufficient to demonstrate this capability. l ! Specified surveillance intervals and surveillance and maintenance l outage times have been determined in accordance wit.h WCAP - 10271,
" Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System," and supplements to that report as approved by the NRC and documented in the SER (letter to J. J. Sheppard from Cecil O. Thomas dated February 21, 1985).
l l Jumpers and lifted leads are not an acceptable method for placing ! equipment in bypass as documented in the NRC safety evaluation l report for this WCAP. l The surveillance requirements for the Manual Trip Function, Reactor Trip Breakers and Reactor Trip Bypass Breakers are provided to reduce the possibility of an Anticipated Transient Without Scram (ATWS) event by ensuring OPERABILITY of the diverse trip features (
Reference:
Generic Letter 85-09). g l The measurement of response time at the specified frequencies j provides assurance that the protective and ESF action function N~0I [ associated with each channel is completed within the time limit assumed in the accident analyses. No credit was taken in the l analyses for those channels with response times indicated as not ! applicable. 00 TNSEAT (frefeb & BEAVER VALLEY - UNIT 1 B 3/4 3-1 Amendment No.HH-
l Attachnsnt A-1 B22v0r Valley Powzr Station, Unit No. 1 Mcense Amendment Reauest No. 220 i Insert "D" l The Engineered Safety Feature Actuation System Instrumentation Trip Setpoints specified in Table 3.3-4 are the nominal values at which , the trip setpoints are set for each functional unit. The trip ' setpoints have been selected to ensure that the response to design basis anticipated operational occurrences as assumed to occur in the safety analysis in mitigating the consequences of accidents due to the action of the Engineered Safety Features Actuation System will occur as predicted. The setpoint for a trip or interlock function is j considered to be adjusted consistent with the nominal value when the "as left" setpoint is within the calibration tolerance band. The calibration tolerance band associated with the nominal trip setpoint is specified in plant procedures. l 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 trip setpoints have been specified in Table 3.3-4. Operation with setpoints less conservative than the Nominal Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error.
LCO 3.3.2.1 is modified by Note (1). This Note permits a trip l setpoint to be set more conservative than the Nominal Trip Setpoint l as necessary in response to plant conditions or revised analysis. ! The conservative direction is established by the direction of the l inequality applied to the Allowable Value. It is consistent with the setpoint methodology for the "as left" trip setpoint to be outside the calibration tolerance band but in the conservative direction with respect to the Nominal Trip Setpoint provided that the Allowable Value(s) is adjusted accordingly. For example, a plant condition that may require a more conservative trip setpoint would be when the 4 kV Emergency Bus Undervoltage Setpoint is set to a value greater than specified in Table 3.3-4 in order to account for voltage drop of l motor leads or relay setpoint inaccuracy. In addition, an increase 3 l in the Steam Generator Water Level-Low-Low Setpoint may be required ; l to account for revised environmental and process measurement allowances. The Setpoint would be increased in the conservative direction in order to meet the revised analysis assumptions. i 4 BEAVER VALLEY - UNIT 1 (Proposed Wording)
DPR-66 3/4.4 REACTOR COOLANT SYSTEM BASES 7 3/4.4.1.1, 2, 3 REACTOR COOLANT LOOPS g in The plartt is designed to operate with all reactor coolant loops Y operation and maintain DNBR above the design DNBR limit during all normal operations and anticipated transients. In Modes 1 and 2, [ fwith ne rea tor pd i ed o es th oo ant lo not a op ratio , MERMAL ,POWE is ) l { e 1 o pe nt Af' RA ED P E un 1 e ve em ra re AT ip ' re 1th a 4 i j, n re that te IBR- wil .
.1 po !
e maintain d a ove e d ig DIGR. (p 1l'mitga)(loss operating above P-7 of flow in two loops will causi a reactor trip if O o
, percent of RATED THERMAL POWER) while,a loss
( flowreentin one loop wi 1 cause a reactor trip if operating above P-8 of RATED TI RMAL POWER). } l 30 no s MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, due to the initial conditions assumed in the analysis for the control rod bank withdrawal from a subcritical condition, two operating coolant loops ) are required to meet the DNB design basis for this Condition II event. In MODES 4 and 5, a single reactor coolant loop or RHR subsystem provides sufficient heat removal capability for removing decay heat; l but single failure considerations require that at least two loops be OPERABLE. Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE. ; l The operation of one Reactor Coolant Pump or one RHR pump ! provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change ; rate associated with boron reduction will, therefore, be within the capability of operator recognition and control. l The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less than or equal to 27 5'F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water level in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than- 25'F above each of the RCS cold leg temperatures. BEAVER VALLEY - UNIT 1 B 3/4 4-1 Amendment No. 4% (,9copech Waba - <
_ . _ . _ _ . ~ . . . _ - ___ - _ _ _ . _ _ ~ _ . . _ . . _ . . _ _ . . _ . _ . . _ _ _ . _ _ . _ _ . _ _ _ _ _ _ . . e ____..__._.-_.__-..._.\
. I j ,
ATTACHMENT A-2 Beaver Valley Power Station, Unit No. 2 j License Amendment Request No. 88 1 ] 1 j The following is a list of the affected pages: i j Affected Pages: 2-3 l 2-4 4 2-5 2-6 l 2-7 2-8 i' 2-9 2-10 i B 2-2 j B 2-3
3/4 3-14 j 3/4 3-23 j 3/4 3-24 i 3/4 3-25 i 3/4 3-26 1
] 3/4 3-27 j j 3/4 3-28 l B 3/4 3-1 l B 3/4 3-la 1 i l
l 1
I I
NPF-73 -A00 j A0-06
2. 2 LIMITING SAFETY SYETEM SETTINGS REACTOR TRIP SYSTEM INSTRUMEhTATION SETPOINTS r
2.2.1 The Reactor Trip System Instrumentation and Interlock Set ts shall 0 Ok b consistent with the Trip Setpoint values shown in Table 2.2-1 g 41 t. 8e. APPLICABILITY: no-% As shown for each channel in Table 3.3-1.
'tCTION: Od3* Et 'N M' O n NWtt b"
a. With a Reactor Trip System Instrumentetion or Interlock Setpoint 44ss-con-

Or;;th- th:n the ..loe she.n in the Trip S:tp: int 00!ur- but more conser-vative than th N'^ - a t the 5e a4++ shown in the Al'owable Value column of Table 2.2-nt h n:j: tent eith D e Trip Setpoint vel = 8 --
s.o,% .h. o%
b. as h.4 ww r=Nw wms.Wa eE Sebwl) (IM2)
With the Reactor Trip System Instrumentation or Interlock setpoint less conservative Table 2.2-1, than 4+' -the value shown in the Allowable Value column of L Adju;t th; S:tpe!M ennc h tent wi+h th: Trip Setpeint valu: ;f-QQg[E =p k ble 2.2-1 :nd d;termine ithin 12 ' curs th=+ Equatier 2.2 1 esj
- ;.si;ed iur tne aii-sted hennel er J -e lare the channel inoperable and apply the applicable ACTION state-ment requirement of Specification 3.3. until the channel is restored to OPERABLE status wi 't.s setpoint adjustedf8:n i;t nt with the
( l Trip Setp . AOC . ggsp %.gM bec i
.x EQUATION 2.2 1- -Z
R

S -: TA
-,,u v.
(t
-2 -
ine valu; for ce! urn 2 ef Tauia 2.2-1 for the afied d chainch
~ %
The "e5 u,wa:,ured" ,elue (in pe.eent spen; vi ra d wirei fur ^Lhe Gf f::ted ehennil,- 40-g 5 2;iner in. "e5 meeLured" v:lue (in p;r:ent spin) Of th- 5= ace" errer, Or th: v:!u c' C0!urr S (Seneer 2rrer) ef Table 2.2-1 for the af' ct:d :h :::1, 2nd TA = Th: '!!!ue #r^- Cc! n TA (T;tel All:u: :- in T e' spin) of Teble 2.2-1 fe. th affe-+-d ch:rn:1, r N-DELEM J. e'0\ (Q A Tc p Sdped g b Sch more. constevokrt. Aun be % %\ Tr.g S 4ps.nk co wutsk m norsatt bel *& c* iht 4 6e troised
%\g t r , Ah e volw e to sbW @ wadgiy, MO (1T % Tr,p saposek vake. sW &c G 6- %,t. nw i,,1, m r%\e 2,'I.-i ts ns k w h ,ne,\ v A v e , h A g b b 3 47 ,,,j AttAk w h % T- + %,a- + %pSQ,4 ab,_m bw.(og A /h BEAVER VALLEY - UNIT 2 2-3 a%w4a3 '
T"'F N o'd F , 5 (Preposd %dh
NPF-73 TABLE 2.2-1 ~ ,AD-0G gggg REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT
+^ L LO* "CE - 'T*' P @oMINA9 ALLOWABLE VALUE l .
[TRIPSETPOINT
1. Manual Reactor Trip -NfAr- -N,As -LA. N.A. N.A.

2. Power Range, Neutron Flux,

a. High Setpoint -S-- -4r56- 09% of RTP* 5 111.1% of RTP*

b. Low setpoint -&rS-~ .A 54- -O-- 25% RTP* s 27.1% of RTP*

3. Power Range, Neutron Flux, W -Gr50- -G- 5% of RTP* with 5 6.3% of RTP* with M4;h Positive Rate a time constant a time constant 2 2 seconds 2 2 seconde j SP-o)

4. Power Range, Neutron Flux 16- O.50 - <
(.j 5% of RTP* with 5 6.3% of RTP* with t High Negative Rate a time constant a time constant 3p.gy 2 2 seconds 2 2 seconds
5. Intermediate Range, -3 ~J,0-- S 44-- -G--- 25% RTP* s 30.9% of RTP*
Neutron Flux
6. Source Range, Neutron Flux --l-7 1G.Ui --O -. a 08 cps s 1.4 x 105 cps

7. Overtemperature AT -+ 5.40- h "^t^ _S See Note 1 See Note 2 I
Overpow'ar M 4,49-- See Note 4
8. -4v9-- 4,-7 [ See Note 3

9. Pressurizer Pressure-Low 4 +- _ O ? l-- 69-- 1945 psig*** 2 1 psig***

10. Pressurizer Pressure-High 2-- -4,96-- -Gr69- 375 psig s M psig [

11. Pressurizer Water Level-High 8,4- -Grte- -t 2% of s 93.8% of instrument span instrument span

12. Loss of F o .3+- M - G,40- 90% of loop 2 88,9% of loop ,
design flow ** design flow **
I= RATED THERMAL POWER
** Loop design flow = 87,200 gpm ***isTime constants utilized in the lead-lag controller for Pressurizer Pressure-Low are 2 seconds for lead and 1 second for lag. Channel calibration sh.31 ensure that these time constants are adjusted tqthose values. $ 8-O (w Akv4 4kt. cedibghom '+C Amendment No . BEAVER VALLEY - UNIT 2 2-4
([r%p3 ]
.- s.
f - ( t
s..
ADO (WF-7h l .% % TABLE 2.2-1 (Continued) . m1 REACTOR' TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS
$ 19*Y }, ,
QdRwM**** FUNCTIONAL UNIT ALLOW 8MCE (TA) Z S TRIP SETPOINT ALLOWABLE VALUE '
13. Steam Generator Water 1
.8 1.6 of narrow > b of narrow Level-Low-Low ange instrument range instrument i E span-each steam ' span-each' steam -< generator generator '
n
14. DELETED. f {Q

15. Undervoltage - Reactor 2I 1.39 0 h5%of4)US ' >73%ofh0Sfoltage Coolant Pumps Joltage-each bus - each bTds ;
16, Underfrequency-Reactor 10. O. O j0 .5 Hz-each bus' >57.4 Hz-each bus n Coolant Pumps . i fnS 17. Turbine Trip po j ( S N.A. 000 psig >958 psig g a. . Emergency Tri'p Header .A. jt". A. Low Pressure L-
b. Turbine-Stop Valve N' . N.A N.A. >1% open >1% open Closure e g -

18. Safety Injection Input N.A. N.A. N.A. N. AQ+- po N.A. AOM i j
1 from ESF ' g / - 1 { N. A.
. 19. Reactor Coolant Pump W.A. .A. N. A.' . N. A. t Breaker Position T. rip Reactor Trip Breakers N.A. NtA. N.A. N.A. N.A. .
E 20. j
~
u/tk
%r , m ~a m . _ m s n . T y -- A = po-os Q - m . - # po-oG .- u. - . - . . . -:... -..- . ~... _n. . .
, s A0() 9F-7) AD-%
TABLE 2.2-1 (Continued) $ > w g REACTOR. TRIP SY3 TEM INSTRUMENTATION TRIP SETPOIN C 2 i
.i ) J a M o p .4 t.- b[M
Tv l

- FUNCTIONAL UNIT ALLOW 6MCE
, S_ } SETPOINT ALLOWABLE VALUE '
i
-f Q sk , 21. Automatic Trip and Interlock N . < .A. .A. N.A. N.A. <J }
(9 ; y b [4 -4 22. Reactor Trip System l g lI;k m Interlocks 2
a. Intermediate Range . .
2 N.A. 1 x 10 10 amps >6 x 10 11 amps Q Q 3 Neutron Flux, P-6 M M,tu
b. Range Neutron .A. .A. N.A. of RTP* g 132.1% of RTP" b c. Power Range Neutron .A. ' .. N.A. 9% of RTP* 151% of RTP*
' h;,
m4 Flux, P-9 ~
'M e & '
3 { d. Power Range Neutron .A. .A. Q Flux, P-10 (Input to N. . 10% of RTP* >7.9% and 112.1% g l% g of RTP* g.
. P-7) 61.
k 5f g e. Turbine Impulse Chamber LN.A. N . .A. _ 0% of RTP* $12.1%ofRTP* .%, Pressure, P-13 (Input to ( urbine impulse turbine impulse "; u) P-7) pressure equivalent pressure equivalent *
& 2.2lA =
Ten Taunnt A%DAno go_os G
60-6 6 _ TABLE 2.2-1 (Continued) h, 1 E REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION h'
< nom prAL- _ - NOTE 1: OVERT MPERATURE AT F<*'#
1 AT ,
(1 * '3 s 1 2 iS $ + r~ d - 3*E3@~ I~I N l Where: AT = Measured AT 1+r5 1 = lead-lag compensator on measured AT; n I * '2 Al em:r,ah Y'O l
-t> t ,12
o ime constants utilized,in lead-lag compensator for AT, r g =8s, m 1 = 3 s; 2
w 2, . L 1 = Lag compensator on measured AT; E-(+ 1 * '3S 5;"a I b, r3 = /imeconstantsutilizedinthe sg compensator for AT, 13 = 0 s; b AT, = Indicated AT at RATED THERMAL POWER Kg =
&Y - f. 267 K = 0.0183/Op; 2
e a 1+tS 4 The function generated by the lead-lag compensator for T 8 = dynamic compensation; 5 1+rS S M
,,,m,,n, j z
O = T4, 15 .
. Ime c nstants utilized in lead-lag compensator for T,yg,4 r = 30 s, 15"45; S
5 *) a -. - - - -. /_ - n - -- -, n =--- ~~ +www
/ * * \. - ~ t g QQ OQ' TABLE 2.2-1 (Continued) -
k
~
E _ REACTOR TRIP SYSTEM INSTRUENTATION' TRIP SETPOINTS' _ NOTATION (Continued) : E T = : Aver s te.verateN. *F; 5 1 = I
T6S' Lao compensator:on measured I ' ;
M." - 4 1,
=
p me constant ut m zed in tw measured t,,, ug c apensator, r3 = 0 s: 3[M +- T' =
< 576.2*F (Nominal T,,, at RATED THERMAL POWER); ,
AbN K = 0. M ;. 3 , P. = Pressurizer. Pressure,'psig; ' ( , Q~ P' >= 2235 psig (Nominal RCS operating prissure)'; m7 - e's").. S. = Laplacetransform. operator,sN '
- and fy(AI) is a~ function of the indicated difference between f.op and bottion detect 6rs of the power-range k 'tests nuclear.such ionthat:
chambers; with gains to be selected based on mer.sured instrument response;during plant startup (1) For qt ~% tween -3 X and + 2 1, f (AI) = 0, M e qt and % are Percent RATED THERNAL POWER in the top.and bottom halves of the core respectively, and qt
4his total THERNAL POWER in percent of RATED THERMAL POWER; (ii)
For-each percent that the magnitude of'qt ' % exceeds -33%, the AT Trip Setpoint shall be automatically reduced by 2.52% of its value at RATED. THERMAL POWER; and g (iii) For each percent that the magnitude qt ' % exceeds +9%, the AT Trip'Setpoint shall 5- be automatically reduced by 1.75% of its value at RATED THERMAL POWER. 2 . .o- ' NOTE 2: . . T _ The channel's maximum: Trip Setpoint shall not exceed.its computed Trip Setpoint by more tha of AT span s g p y g -f gy j, g . i s' y A0-0 6 ~ ~ w. .
-~:-~.. . -m . ~ , f .m ~ ~41=A - d.*G
. - -- - - - - - - - - - - - ~ - - - - _-_ _ --_- --s =.------------------.---c--. :-~ , . *p . : . - , ? *. : ~. ., . . : ~
t ,.
. 3. .
TABLE 2.2-1 (Con'tinue')- d 9 E 3 nom / nab / ' REACTOR TR!P SYSTEM INSTRUMENTATION TRIP SETPOINTS
. NOTATION (Continued)
NOTE 3: OVE R AT /I' 5 M W ' ' W ,
,3
8) I I - lio5 I(1 g {
I
'I I I )
AT (II *1+Y8)II'+T5) 2 3 HAT. [K4 + 15) 7 .(1 + 165) T-K 6 [T-(I I 1 + r65)
)
I"] - f 2(f }} Where:~ 'Ai = ' Measure 1+tS - l' := . { 1 + Y28' Le~ad-lag compensator on measured AT; 4
- Midh 6 NM j! ,,7 Y,1 g 2- .. = ,
ime constants utilized fri lead-lag compensator for AT, rg = 8 s,12 " 353
*g g , 1 = La ' nsator on measured AT; E, 1+TS3 pe ,,,,w k 1 3 =' /Ime constant utilized in the lag compensator for AT,1 ~
3 = 0 st ys AT, =
~
Indicated AT at RATED THERMAL POWE /F; K = 1.0781; t 4 , t K =. S 0.02/*F for increasing average temperature and 0 for decreasing average temperature; , I g 7
= = -< 1+tS7 The function generated by the rate-lag compensator for T,,g dynamic compensation; p Aln.%
8 -1 7
.=
gime constant utilized l'n rate-lag compensator for T,yg,17 = 10 s; n ' se.' g - [ ** . s *
,sff Q: + *% i '** '"N *'. '*
~
v e. ' h00
hO~Y 9 TABLE 2.2-1 (Continued) g

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS
( NOTATION (Continued) r , 1 = k , 1+1 6 Las compensator on measured T
/Jo ,.,f,,, /
I 1 = 6 . Jime constant utilized in the measured I,yg lag compens'ator, g = 0 s; Z1 K =
~ 6 0.0012/*F for T > T" and K6 = 0 for,T $ T";
T = Average Temperature. *F; o,0( T" = 4Y Indicated T,,g at RATED THERMAL POWER (Calibration temperature for AT instrumentation, 1576.2*F);
.g 2 " S =
Laplace transform operator, s"8 ; and , g f 2(al) = 0 for all al. h NOTE 4: The channel's max' Trip Setpoint shall not exceed its computed Trip Setpoint by more than q y%ofATspan n
o. x - ) e; .- 2,,pu-fr or de p ,ej ;n Nodef} \
p (P M "a&M""- = - wk at % j g>L 3 9 9 9 5 g.os 1
-f NPF-73 SAFETY LIMITS BASES '
2.1.1 REACTOR CORE, continugd
< These limiting heat flux conditions are higher than those I calculatedfortherangeofallcontrolrodsfullywithdrawntothe}A0~ %
maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f(AI) function of the Overtemperature AT trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the overtemperature AT trip will reduce the setpoint to provide protection cons.tstent with core safety limits. 2.1.2 REACTOR COOLANT SYSTEM PRESSURE (c--The restriction of this safety limit protects the integrity of } AN the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere. 4---The reactor pressure vessel, pressurizer, and the RCS piping,lAO-G. ! valves and fittings are designed to Section III of the ASME Code l i for Nuclear Power Plants which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The safety Limit of 2735 psig is therefore consistent with the design criteria and I associated code requirements. i
<E---The entire Reactor Coolant System is hydrotested at 3107 psig l AO #-
j to demonstrate integrity prior to initial operatio . g dhAht M' 4 2.2.1 REACTOR TRIP INSTRUMENTATION SETPOINTS (y 1 ehe M b } g-),Q w OE ) 4-- The Reactor Tri Setpoin specified in Table 2.2-1 are l AO-CC. the nominal valuesCat which the Reactor . Trips are set for each functional unit. The Trip'Setpoints have been selected to ensure that the reactor 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 Engineered Safety Features Actuation System in mitigating the consequences of accidents. The setpoint for a reactor trip system or interlock function is cons ered to be adjusted consistent with the nominal value when the "as setpoint is within the Sp.og M.J. all:"r' f e calib i.conce knD),
d. . To accommodate the instrument drift assumed to occur between IM~05 operational tests and the accuracy to which setpoints 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 th

Allowable Value is -
fN ca.hb.n 4.hti.e. ! g uMITIM 6 gAFETY fV57E M IE-rTIN {j]5'c "y % l M O i y .. w . g .a. c.- a r.) BEAVER VALLEY'- UNIT 2' B 2-2 g"i;; IGC ,lp ter s, ma (Nposa wa f___{-.~Q-w
NPF-73 , b WA' M M 1.h LIMITING SAFETY SYSTEM SETTINGS A 5 are)M- af M I BASES i N
2.2.1 REACTOR TRIP SYM EM INSTRUMENTATION SETPOINTS. continued y accentable since an allowance has been made in the safety analysis to accommodate this error.4j An optional pr ision has been inci ea toF er ing the OPERABILITY o channe when its 1p setpoi is found to ceed the A wable va e. The m hodology f this op 'on utilizes t "as measu " deviat ' from the pecified libration oint for rack nd sensor omponents n conjunc on with a tatistica combination of
~ the other neertain es in cali ating the 'nstrumenta lon. In Equation , 2.2-1, +RtS TA, the in ractive ef ects of th errors in the ck and e senso , and the 'as measur da values of the error are
'o idered. , as spec led in Tab. 2.2-1, i percent span is the atistica summation errors ass ed in the nalysis exclu ng those associat with the ensor and r ek drift d the accura of their measur ient. TA or otal Allowa e is the Afference', in ercentspan,k betw n the trip tpoint and t e value us in the analy s for rea tor tri . R or Rac Error is the 'as measur " deviation, percen pan, f the affect d channel fro the speci ed trip setpo t. S Sensor Drift is ei er the "as easured" viation offt calibratio point or the alue specif'ed in Table 2. - 1, in senso from its rc M an, from the alysis assum'y ions. Us ' f Equation 2. 1 allows for a sensor drift tor, an incr4'ased rack ift factor, and provides a threshold lue for REPORTABLE EVENTS. j wir-- The methodology to derive the trip setpoints is based upon com)Am-%bining 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 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 drif t, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investication ict)ENSEAT 6" M Manual Reactor Trio
't.
The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability. Power Rance. Neutron Flux The Power Range, Neutron Flux channel high setpoint provides reactor core protection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry. The low setpoint provides redundant protection in the power range for a power excursion beginning from low power. The trip BEAVER VALLEY - UNIT 2 B 2-3 Amendment No. 44-(, std bo
Attach::nt A-2 . Baavsr Valley Powar Staticn, Unit No. 2 License Amendment Recuest No. 88 Insert "G" Limiting Safety System Settings 2.2.1 is modified by two notes. Note (1) permits a trip setpoint to be set more conservative than the Nominal Trip Setpoint as necessary in response to plant conditions or revised analysis. The conservative direction is established by the direction of the inequality applied to the Allowable Value. It is consistent with the setpoint methodology for the "as left" trip setpoint to be outside the calibration tolerance band but in the conservative direction with respect to the Nominal Trip Setpoint provided that the Allowable Value(s) is adjusted accordingly. For example, the Power Range Neutron Flux High trip setpoint may be set to a value less than specified in Table 2.2-1 during a plant condition where the Heat Flux Hot Channel Factor - Fg(Z) exceeds the technical specification limit. In addition, an increase in the Steam Generator Water Level-Low-Low Setpoint may be required to account for revised environmental and process measurement allowances. The Setpoint would be increased in the conservative direction in order to meet the revised analysis assumptions. Note (2) states that Functional Unit number 17.b in Table 2.2-1 for Turbine Stop Valve position is not a nominal value. Therefore, the trip setpoint is adjusted to be consistent with the Trip Setpoint value specified in Table 2.2-1 in lieu of adjusting the setpoint within the established calibration tolerance band of the Nominal Trip Setpoint. BEAVER VALLEY - UNIT 2 (Proposed Wording)
NPF-73 INSTRUMENTATION [ 3/4.3.2 l ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTxUMENTATION 1 l LIMITING CONDITION FOR OPERATION. f 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) l .OPERABLE instrumentation withchannels their Tripand interlocks Setpoints setshown consist in Table 3.3-3 shall be l shown in the Trip Setpoint column of Table 3.3-4. with the valuesj ADO APPLICABILITY: As shown in Table 3.3-3. N ACTION: Od3tcht- PI'I C'*\ b f* bod b\ E#'^d b
a.
With an ESFAS
!4cen;;.ietiv: *h=
Instrumentation or Interlock Trip Setpoint less th: value sh;Un in the Tiip C;tpr ht ec1"=n but ! j more columnconservative of Table 3. thanadjust the value shown in the Allowable value Trip Setpoint va4ue. the Setpoint,Yarrict:nt with the L mhe O on at he u%.-MQ cabb% h\mte ! I s b. With an ESFAS Instrumentation or Interlock Trip Setpoint less ~ conservative than the value shown in the Allowable Value column I of Table 3.3-4, eithtre 1.
~ OfA*j"et th; Setyei..t ;;nsist:nt with the Trip C;tp; int value ' f-Table 0. 3- 4 and dwi.auuisse within 12 h:ur \
l 1 2-1 '-?ar estirfi;d for th; off-wied chtnnel, er' hit E~"iti n g _
}
42 clare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the cnannel is restored to OPERABLE status with i" adjusted <enrict;nt ^ ith th; Trip Setpoint vehe.' O Setpoint (i et g., hwb4 , I ggM r^'.'?.TI O:: 2.2-1 1 ^ R
S 5 TA-6' f l l Cc.hkruhto wh;re:
2-l had d The e- el, value for Column ~ cf Tehle 3.2 4 for the rffected MS R- The ":; neesu ed" valu; (in perce".t -"=1 mf va
arrer l fer the aff:;ted whannel,-
E- Eith;r the "&a as;;ured" v:1u: (in p:rc nt = pani nf +he serrer errer, er the v:1u; frs; C:1" n e Inan=n* mim ef TaMa ? ' for the Offected channel, =nd
~
OELETE:P 'n2
- The v:lue frer C lumn T?. (Ta*=1 -2.2-4 f r th; aff;;ted rhin"el All manna) af 'iMe ! c. With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3. < {AOO TNSERT ' EU ) 'M' BEAVER VALLEY - UNIT 2 3/4 3-14 Amendment No. 88- g htof30 $ l
Attrchmnt A-2
.' CO2 var Vallcy Powcr Station, Unit No. 2 License Amendment Reauest No. 88 Insert "E" (1) ' A Trip Setpoint may be set more conservative than the Nominal Trip Setpoint as necessary in response to plant conditions or revised analysis. Allowable Value shall be adjusted accordingly.
t e BEAVER VALLEY - UNIT 2 (Proposed Wording)
- ' ' s. ,
s NO M @ F-73] ;-[) b b O
~
p TABLE 3.3-4 I M ENGINEERED SAFETY' FEATURES ACTUATION SYSTEM INSTRUNENTATION TRIP SETPOINTS-5) m uA& g [. . TOTAL SENSOR } . TRIP p .) FUNCTIONAL UNIT 3l ALLCWANC Z IFT (S) SETPOINT s ALLOWABLE VA'UE f e 1.- SAFETY INJECTION AND FEE 0 WATER ISOLATION
/ [.
= . 4 a. . Manual' Initiation N.A. N.A.
.A. .N.A. H.A.
m . -
~ b. Automatic Actuation Logic and Actuation Relays jN . .
N.A. N. 'A. N. A.- N.A.
~
7 c. Containment Pressuie - High 2.6 4 0 71 1 67 1.5 psig -5 kG sig N g E d. Pressurizer Pressure - k .5 10 - ( / ={ wd e. Low Steamline Pressure -- 1.67 k 3848 psig 1 psig N-oi L4.2 ~ 1.21 1.67 500 psig* 1 psig* U t/f3
/ 1.1 _ SAFETY. INJECTION TRANSFER FROM'.
INJECTION TO THE. RECIRCULATION t f MODE
a. Automatic Actuation Logic,.. l. A. .A. N.A. N.A. N.A.
Coincident with Safety Injection Signal N
b. Refueling Water Stor_ age 2. O.64 1.30 7 feet
> 37 feet Tank Level - Extreme Low inches @ inches
2. CONTAINMENT SPRAY

a. Manual Initiation N.A. .A. N.A. N.A. N.A.
( j - CTime constants uf.ilized in the. lead-lag controllers for Steam Line Pressure-Low are it > 50 seconds and T2<5 sec nds. CHANNEL CALIBRATION shall ensure that these time constants are adjusted to these values. [ 4 .
'~
pedM. L5 .:: : - - - -
~ - - ' = - - n = + ===b A=&~ ~ " =
ADQF F-4 I , b '-0 '*' TABLE 3.3-4 (Continued) l D > ENGINEERED SAFETY FEAT'RES U ACTUATION SYSTEM INSTRUMENTATION T IP SETPOINTS
< ITAL SE . - TRIP
f FUNCTIONAL UNIT ALLOWANCE (TA Z _ DRIFT (ST SETPOIN1 .
ALLOWABLENALUE 1 7 2. CONTAINMENT SPRAY (Continued) E b. Automatic Actuation Logic N.A. .A. N.A. N.A. N.A.
-4 and Actuation Relays g,4
c. Containment Pressure -- ' 2. 9 0.71 f 0 psig i 8/psic]
High-High
3. CONTAINMENT ISOLATION

a. Phase "A" Isolation f)k

1. Manual Initiation N.A. N.A. N.A. N.A. N.A.
3w 2. Automatic Actuation N.A. N.A. N.A. N.A. N.A.
" A> Logic and Actuation e
Relays J' G'f)1 a 3. Safety Injection See Functional Unit 1. above for all Bafety Injection Trip Setpoints and k Allowable Values. &v b. Phase "B" Isolation
/ '
1. Manual Initiation N.A. .A. .A. N.A. N.A.

2. Automatic Actuation .A. N. N.A. N.A. N.A.
Logic and Actuation - If Relays
3. Containment Pressure-- 2.,9 / 0.71 .67 0 psig i psig High-High /
A0-06 i a
^ ~ , . . ;,
I ,' ~ .. ggy
~
pp-Q] f0-CC
*~ .~ , _ TABLE 3.3-4 (Continued) ,
9- '
<) ENGINEERED' SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTATION TRIP SETPOINTS~ , < bTA ## $ FUNCTIONAL UNIT AL E[ _
SES0h(S) IFT . SETPOINT# TRIP ALLOWABLE VALUE
4. STEAM LINE' ISOLATION- M -
E a. Manual Initiation y .
)
m 1. Individual- N. . .A N. Ant? N. . DV p_q
2. System N.A N. .A N.A ,
N.~ . 8-
b. . Automatic Actuation Logic N.A. 3' N. . .N. A. N.A.
and Actuation Relays - G 3. 5 bod 2-- b c. Conta'inment Pressure-- 2 0.[ 1.67 0 psig < a'.? psi db Intermediate-High-High _ _g pgg
d. Steam Line Prensure - Low 14. 1. 2 7 00 psig* .psig*
1
//7
e. .50 0 100 psi < psi Steamline Pressure Rate - { 3.0 -
Qa High Negative w th a time with a time
' constant constant 1 50 seconds 2 50 seconds
5. TURBINE TRIP AND , FEE 0 WATER ISOLATION'

a. Automatic. Actuation ~ Logic i N.A. N.A. N.A. N.A. N.A.
I and Actuation Relays- 1V,6
b. Steam Generator Water Level 5 2.18 1.67 of narrow < W 8% of narrow
-- High-High, P-14
_b range instru- range instrument)
. ment span span ~ '
c. Safety Injection See Functional Unit 1. above for all Safety Injection Trip Setpoints
~ and Allowable Values.
CTime constants utilized in the lead-lag controllers for Steam Line Pressure-Low are ti > 50 seconds and 12 '< 5 s:c:nds. CHANNEL CALIBRATION shall ensure that.these time constants are adjusted to these values.
- a At)-06 . ., , 2 ., , , - , . , s ., a +.g._ , c ,,a,
. - - . . _ _ . _ ._.m _._. _ _ . _ . , _ .
( t .
~
y AhQvp-g AEoE g * - TABLE 3.3-4 (Continued) .
' ~ ! ENGINEEREDSAFETYFEATURESACTUATIONSYSTEMINSTRUMENTATION.TRIPSkTP0INTS FUNCTIONAL UNIT ) FT (S) E P0 OWABLE VALUE 6 LOSS OF POWER.
7 E a. Bus . l 4 - 4.16.kV 1.- Undervoltage- Emergency (Trip Feed) 15.0 1.39 - 0.0 75% of nominal > 73% of nominal m u's Voltage Bus Voltage with a 1 1 0.1 with a.1 i 0.1~ second time second time n dela .. delay g 1 ,. 0.0 5% of nominal > 73% of nominal f
2. Undervoltage (Start. '15 1.39 Diesel) us Voltage, Hus Voltage, b 20 cycles i 20 cycles. p gg 2 cycles .. 2 cycles ,g6j w3- b. 4.16 kV Emergency Bus 15'.0' s . 1.39 0.0 of nominal >
us. Voltage
' . of ' nomina'l Bus. Voltage W
k( (Degraded Voltage)
with a 90 i with a 90 i -
[} 5 second time 5 second ti 9
, 8- ggy "' "' 'Y 92,2 8 $
d of nominal # 0.0
~
c. 480. Volt Emergency Bus . 15.0 1.39 of nominal'>.
Y. ~(Degraded Voltage) i~ , us Voltage Bus Voltage with a 90 1 - with a 90 1 5 second time 5 second time
~
( .. delay
delay
}
7. AUXILIARY FEEDWATER* ,
. ( . a. Automatic Act'uation Logic lN. . N. . N.A. N.A. N.A. ~
and Actuation Relays
~
k ;
Manual initiation is included in sr ecification 3.7.1.2 ;
b O--O. .E I . p .g. . _ . _, . ,: _ . y . __,;..
.. . ,s .-**~t-~-
2.1 s -..; .e. w. .u.,.~._..~- ....~..%. .
, .4 ,,
^
Thar.m 3.3-4 fcontinuM1 NPF-73
~
ENCImunDun mayETY FEATnDER ACTUATION SYSTEK Tu-udh-RATION TRTD SETPOX is NW FUNCTIONAL UNIT
. TOT SM - . TRIP . I Ar A __ fTA) /
g FT fSi SETPOINT jitt.inuant.n VaLUE
7. AUXILIARY FEEDWATER (Continued) '

b. Steam Generator Water / 7.2.
Level--Low-Low
/7,7 h
4f I
1. Start Turbine .5 .18 1.67 4 W5% of .1 % of Driven Pump arrow range narrow range j instrume t instrument 1 R /2,2. .
span 77,7 ' fS ,' ,
2. Start Motor '
Driven Pumps 'if. 5 , 10.1 1.6 1 4 of arrow range' 2- % of narrow range I instrument instrument r span span !
c. Undervoltage - RCP 27.7 1 9 75% of 2 73% of (Start Turbine Driven ominal bus nominal bus ;
Pump) T,_ voltage voltage g g g.
d. Safety Injection See Item 1. above for all Safety Injection Trip Setpoints (Start All Auxiliary and Allowable Values.
Feedwater Pumps)
^
c. Trip of Nain Feedwater N.A. N.A.
Pumps (Start Notor- I ' Driven Pumps) DDEETE j BEAVER VALLEY - UNIT 2 ,
";/4 3-27 Amendment No. 89A t. %N.7 - -.a - - .x:.-.~; .w.u
_ , _ ~ . . s ,=.'. TABLE 3.3-4 (Continued) ' NPF-73 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS i
-] AJd H iA/A L ,
FUNCTIONAL UNIT TOA ELOWAN d (TA). Z S NSOR IFT ( S )' k TRIP SETPOINT
, ALLOWABLE VALUE,
8. ENGINEERED SAFETY FEATURE INTERIDCKS

a. Reactor Trip, P-4 N.A. N.A. .A. N.A. N.A.

b. Pressurizer Pressure, N. . N . N.A. 000 psig s 2010 psig ,
P-11 , {
c. Low-Low Tavg, P-12 4.0 0.82 0.87 541"F 2 538.5'F
< y.
C@ad) 4 j 4eV BEAVER VALLEY - UNIT 2 3/4 3-28 AmendmentNo.[ (Next page is 3/4 3-33) .
7_ _ _ _ _ T i 2 . %e. .b.p s Apnis h.os bem sa\ etbl b saswe WJ-khe. ec3fawre. b duign bab3 # dtCip* N ofM 566* ocqgef,,ngt3 ,,3 ,ggu ,,,) 4 , ,et,g 3,gggg,\ ? NPF-73 3/4.3 INSTRUMENTATION N '
*d kh' C*^1T'^'D d i acttcle45 dwt ek( be. cathon of Iks. %ict Sc.M Fed M Ad M stM S syh w .\\ O CC #
BASES g jedsth), 3/4.3.1 and 3/4.3.2 REACTOR TRIP SY!LTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATI,QH The OPERABILITY of the Reactor Protection System and Engineered Safety Feature Actuation System Instrumentation and interlocks ensure 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 and sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance consistent with maintaining an appropriate level of reliability of the Reactor Protection and Engineered Safety Features instrumentation, and 3) sufficient system functional capability is available from diverse parameters. The OPERABILITY of these systems is required to providt 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 these systems is consistent with the assumptions used in the accident 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. l Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with WCAP
- 10271, " Evaluation of Surveillance Frequencies and out of Service Times for the Reactor Protection Instrumentation System," and supplements to that report as approved by the NRC and documented in the SER (letter to J. J. Sheppard from Cecil O. Thomas dated February 21, 1985). Jumpers and lifted leads are not an acceptable method for placing equipment in bypass as documented in the NRC safety evaluation report for this WCAP. .
The Engineered Safety Feature Actuation System Instrumentation Trip setpoints specified in Table 3.3-4 are the nominal values at - which the bistables are set for each functional unit.4A setpoint is lo-ol m naidered to be adjusted consistent with the nominal value when the N "as %easpeed" setpoint is within the 6 F.d alle.;;c rc:;) calibration y.o) I - b lerco c o. L c) DE LETE- J 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 accaotable f c.h e.% h\smse lo.Ad associcM wJrw ht. w=*^-\ W 30 f*'^ v% c \w. .a i- e.n.an.. . - BEAVER VALLEY - UNIT 2 B 3/4 3-1 Amendment NoJrt-
@tefoStd Y
) NPF-73 1 S~ O 3/4.3 INSTRUMENTATION t 6- I BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) since an allowance has been made in the safety anal _ysis to .
accommodate this errorJ an opuunat provision as been includedis for *i 4
; -tietEYmining tMrr.MABILITY of a channel w n its trip setpoint i found to exc the Allowa alue. The ethodology of this option utilizes "as measur deviation om the specified calibration point f rack and sensor comp ents in conjunctio ith a statis cal combi ion of t other uncertaint of the inst mentation measare the p cess variable and th uncertainties ;
l in alibrating e instrument ion. In Equation 2 - 1, Z+R+S$ '
, the inte ctive effects the errors in the ck and the sen or, and the " measured" va es of the errors e considered o,I as specifie in Table 3.3 4, in percent sp , is the s istical summdt n of errors ssumed in the alysis excl ing those as ated with the n: sor and rack dri and the acc acy of their aa rement. TA or otal Allowance the differ ce, in percent n, between the rip setpoint and he value us in the anal sis for the actuatio . R or Rack Epror the "as me red" devia n, in percent span, or the affe6ted channel fro the speci ed trip setpoint. S Sensor Drift'i ither the " measured" viation of the sensor, om its calibr point or t value sp ied in Table 3.3-4, id percent span om the a sis ass ions. Use of I Equation 2.2-1 allows fo a sensor dr t factor, n increased rack {
drift factor, and provides a threshold value for REPORTABLE EVENTS. A 1
' D El.ET E @-O\ &(002N56(Wh BEAVER VALLEY - UNIT 2 B 3/4 3-la Amendment No44 Yt%fs3cb O P
.* \
l i Attsch2snt A-2 l B3svar Vallcy Powar Station, Unit No. 2 l License Amendment Reauest No. 88 l l Inser'; "E" l I LCO O.3.2.1 f.s scJified by Note (1). This Note permits a trip ' setpoint to bs net. more conservative than the Nominal Trip Setpoint as necessary 20 .Tssponse to plant conditions or revised analysis. The conec*vetiv6 direction is established by the direction of the inequality .rN iad to the Allowable Value. It is consistent with the setpoint me? a- ; ology for the "as left" trip setpoint to be outside the calibraf;f .1 tolerance band but in the conservative direction with respect iLc s Nominal Trip Setpoint provided that the Allowable Value is ad? ste.1 ecer dingly. For example, a plant condition that may require. a more conservative trip setpoint would be when the 4 kV Emergency '.u<* Undervoltage Setpoint is set to a value greater than specified in Table 3.3-4 in order to account for voltage drop of motor lee.ds or relay setpoint inaccuracy. In addition, an increase in the Steam Generator Water Level-Low-Low Setpoint may be required to account for revised environmental and process measurement allowances. The Setpoint would be increased in the conservative direction in order to meet the revised analysis assumptions. ~ i
)
4 5 I l BEAVER VALLEY - UNIT 2 (Proposed Wording) l l
. 1
.- l l
. ATTACHMENT B Beaver Valley Power Station, Unit Nos. 1 and 2 License Amendment Request Nos. 220 and 88 REVISION OF REACTOR TRIP AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION REQUIRED SETPOINTS i AND ASSOCIATED DASES SECTIONS M)
A. DESCRIPTION OF AMENDME!!T REQUEST l The proposed changes include the following: I For Beaver Vallev Power Station (BVPS) Unit No. 1 only:
Revise the values for " Trip Setpoint" contained in Table 2.2-1 titled "Mactor Trip System Instrumentation Trip Setpoints,"
for Tabla item numbers 12, 13, and 15.
Revise the values for " Allowable Value" contained in Table 2.2-1 for Table item numbers 9, 12, 13, 15 and 17.A.

Revise the values for " Trip Setpoint" contained in Table 3.3-4 titled " Engineered Safety Feature Actuation System i Instrumentation Trip Setpoints" for Table item numbers 5.a, l 6.b, 6.c, 7.a and 7.b.

Revise the values for " Allowable Value" contained in Table l 3.3-4 for Table item numbers 1.c, 2.c, 3.b.3, 4.c, 5.a, 6.a.1, l 6.b, 6.c, 7.a, and 7.b. l l

The term AT in the overtemperature and overpower setpoint equations, contained in Notes 1 and 2 of Table 2.2-1, would be defined. The value for T' specified in Note 1 of Table 2.2-1 for the Overtemperature AT eqcation would be revised. Four equation variables for the overtemperature and overpower setpoint equations that are specified in Notes 1 and 2 would be deleted. The words "AT Trip Setpoint" would be added to the equation titlen contained in hotes 1 and 2. Additionally, the value for T" specified in Ncte 2 of Table 2.2-1 for the Overpower AT equation would be evised. Table 2.2-1 Notes 3 and 4 would also be revised. 1. 0 editorial change would be made to the T3 value specified in NG'e 2 of Table 2.2-1.

Limiting Safety System Setting (LSSS) 2.2.1 titled " Reactor Trip System Instrumentation Setpoints" would be modified by the addition of two footnotes and modification to the current action statement. The current action statement would be revised into two separate Actions a. and b. to be more consistent with the current BVPS Unit No. 2 actions. The two
! proposed action statements would be modified to reflect the proposed changes to Table 2.2-1. The term "and Interlock" would be added to LSS 2.2.1. The first letter in certain l words would also be capitalized. l
. - _ - . .- ..- - .... _ _.-. - . - - _- . . = . . . . - - . - . - . . - - _ . . -. ..- . .
s' , ATTACHMENT B, continusd I, Licens3 Ausndmant Request Nos. 220 and 88 l Page 2 1
Limiting Condition for Operation (LCO) 3.3.2.1 would be I revised by the addition of one footnote pertaining to trip l setpoint value adjustment. The current Action a. would be !
modified to reflect the proposed changes to Table 3.3-4. The ; modified Action a. wording would be designated as Action b. A ' new Action a. would be added which provides guidance when a trip setpoint is less conservative than the nominal trip I setpoint. The current Action b. would be designated as i Action c. The term "or Interlock" would be added to the proposed Action b and c. The term " statement" would also be added to proposed Action b. The first letter in certain words of LCO 3.3.2.1 would be capitalized.
The following items would be revised to delete references
corresponding to two loop operation, or loop stop valves being open, Facility Operating License Condition 2.C.(3), Table 2.2-1 Note 1, Bases section for "Overtemperature AT" on page l B 2-5, Bases Section for " Loss of Flow" on page B 2-6, Table 3.3-3 titled " Engineered Safety Feature Actuation System Instrumentation" item 7.a, and Bases section for " Reactor Coolant Loops" on page B 3/4 4-1. In addition, a l typographical error would be corrected in License Condition !
2.C.(5). l l
The Bases for " Reactor Coolant Loops" on page B 3/4 4-1 would be revised to reflect the current trip setpoints for P-7 and j P-8.

The Bases section for LSSS 2.2.1 titled " Reactor Trip Set Points" would be revised to be consistent with the proposed changes to LSSS 2.2.1 and Table 2.2-1.

Bases page B 2-3 for " Power Range, Neutron Flux" would be revised to reflect the current trip setpoint for P-10.

Bases page B 2-6 for " Loss of Flow" would be revised to reflect the proposed Loss of Flow trip setpoint of 90.2% and the current P-8 trip setpoint of 30%. In addition, the Bases section would be revised to reflect the current trip setpoint for P-7.

The Bases section titled "3/4.3.1 and 3/4.3.2 Protective and Engineered Safety Features (ESF) Instrumentation" would be revised to be consistent with the proposed changes to LCO 3.3.2.1 and Table 3.3-4.

Table 3.3-3 titled " Engineered Safety Feature Actuation System i Instrumentation," Table 3.3-4 and Table 4.3-2 titled
" Engineered Safety Feature Actuation System Instrumentation Surveillance Requirements," Item 7.d, pertaining to Auxiliary Feedwater Emergency Bus Undervoltage would be deleted.
B-2
, ATTACHMENT B, continu:d 6 Lic ngs Amsndm:nt R quest Nos. 220 and 88 Page 3 For BVPS Unit No. 2 only:
j 1
Ravise the value for " Trip Setpoint" contained in Table 2.2-1 l titled " Reactor Trip System Instrumentation Trip Setpoints,"
for Table item number 13. i e Revise the values for " Allowable Value" contained in Table 1 2.2-1 for Table item numbers 9, 10, 12, and 13. )
Revise the values for " Trip Setpoint" contained in Table 3.3-4 titled " Engineered Safety Features Actuation System i Instrumentation Trip Setpoints," for Table item numbers 1.d, 1.1.b, 5.b, 6.b, 6.c, 7.b.1, and 7.b.2. 1

Revise the values for " Allowable Value" contained in Table 3.3-4 for Table item numbers 1.c, 1.d, i.e, 1.1.b, 2.c, 3.b.3, 4.c, 4.d, 4.e, 5.b, 6.b, 6.c, 7.b.1, and 7.b.2.

The values and column headings for " Total Allowance (TA)",
"Z", and "S", specified in Tables 2.2-1 and 3.3-4, would be deleted. This deletion includes Equation 2.2-1 in LSSS 2.2.1 and Note 5 specified in Table 2.2-1.
The Bases section titled "2.2.1 Reactor Trip System Instrumentation Setpoints" would be revised to reflect the
[ deletion of the equation 2.2-1 specified in LSSS 2.2.1. l
The Bases section titled "3/4.3.1 and 3/4.3.2 Reactor Trip System and Engineered Safety Features Actuation System l Instrumentation" would be revised to reflect the deletion of the equation 2.2-1 specified in LCO 3.3.2.1.
i
For Table 2.2-1, the value for "K1 " specified in Note 1 for l the overtemperature AT equation would be revised.
The wording contained in Note 2 referenced in the overtemperature AT Allowable value column of Table 2.2-1, would be revised. The wording contained in Note 4 referenced in the overpower AT . I Allowable value column of Table 2.2-1, would be revised. A typographical error would be corrected. In the term: l r 3 1 l 1+TS l l r i (1+TS) 2 the number one designation, for the equations specified in Notes 1 and 3 of Table 2.2-1, should be a subscript instead of a superscript. The words " Trip Setpoint" would be added to the equation titles contained in Notes 1 and 3.
LSSS 2.2.1 titled " Reactor Trip System Instrumentation Setpoints" would be modified by the addition of two footnotes and modification to the current action statements. Action b.1 B-3
f
-ATTACHMENT B, cantinutd a License Amendment Request Nos. 220 and 88 Page 4 would be deleted. Action b.2 would be incorporated into Action b. Equation 2.2-1 would be deleted. The term " nominal" would be added to Actions a. and b. in reference to the trip setpoint value. The two proposed action statements would be modified to reflect the proposed changes to Table 2.2-1. 'The term " Set" would also be e.dded to LSSS 2.2.1. The reference to Specification 3.3.1 would be changed to 3.3.1.1.
LCO 3.3.2.1 would be revised ' by the addition of one footnote pertaining to trip setpoint value adjustment. Action b.1 and Equation 2.2-1 would be deleted. Action b.2 would be incorporated into Action b. Actions a. and b. would be modified to reflect the proposed changes to Table 3.3-1.

The footnote on Table 2.2-1, page 2-4, designated by three asterisks, would be modified by the addition of the term
" Nominal" and the words pertaining to calibration tolerance bond.
A new footnote would be added to Table 2.2-1 on page 2-6 which defines the term "RTP". This footnote is consistent with a similar footnote for Table 2.2-1 on page 2-4.
For BVPS Unit Nos. 1 and 2:
Removal of the inequalities applied to the " Trip Setpoint" column of Table 2.2-1 (includes Table Notes numbers 1 and 2 setpoint equations for BVPS Unit No. 1 and Table Notes numbers 1 and 3 setpoint equations for BVPS Unit No. 2) with the exception of table item 17.B " Turbine Stop Valve" for BVPS
! Unit No. 1 and item 17.b " Turbine Stop Valve Closure" for BVPS Unit No. 2. A new footnote would be added to Table 2.2-1 which denotes this exception for item 17.B (Unit 1)/17.b (Unit 2).
Removal of the inequalities applied to the " Trip Setpoint" l
column of Table 3.3-4.
The addition of the term " nominal" to the Trip Setpoint column l in Tables 2.2-1 (includes Table Notes numbers 1 and 2 equations and certain defined terms for BVPS Unit No. 1 and Table Notes numbers 1 and 3 equations and certain defined terms for BVPS Unit No. 2), and Table 3.3-4.

Page format would be revised to permit incorporation of additional text and updating to current page format. The ;
format changes include the addition of new Technical i I i Specification pages. !
License number DPR-66 (for BVPS Unit No. 1) or NPF-73 (for BVPS Unit No. 2) would be added to each page which does not l currently contain this designation. l B-4 l
i
P' ! ATTACHMENT B, centinued I Licsnce Amand32nt Rsquest Nos. 220 and 88 Page 5 l
The page footer would be rotated on Table pages (as l applicable) to be consistent with the text format. )
l B. DESIGN BASES ; The purpose of the Reactor Trip System (RTS) and Engineered Safety Feature Actuation System (ESFAS) is to automatically initiate protective actions to aid in the mitigation of certain design basis accidents and to provide a backup to operator actions in the form of permissive interlocks that ensure bypassable functions are in operation under the conditions assumed in the safety analyses. In order for the RTS and ESFAS to function correctly, the trip setpoints must be adjusted to values that ensure that the trip bistable will actuate at or before the value assumed in the safety analysis when all known uncertainties in the channel are accounted for. A methodology has been developed that properly accounts for channel uncertainties and was used to calculate the nominal trip setpoint and allowable values which are proposed in this license amendment request. The methodology is documented in the following reports listed below. Westinghouse reports WCAP 11419 Rev. 1, Westinghouse Setpoint Methodology for Protection Systems Beaver Valley Unit 1, May 1994 and WCAP 11366, Rev. 3, Westinghouse Setpoint Methodology for Protection Systems Beaver Valley Unit 2, May 1994 provide revised Reactor Trip and Engineered Safety Feature Actuation System setpoints and allowable values applicable to the current BVPS Unit No. 1 and No. 2 protection systems. The allowable values for the overtemperature and overpower delta T reactor trips identified in these WCAPs were later revised. The revisions were based on calculations performed to address a Westinghouse l technical bulletin, (ESBU-TB-96-07-RO) which discusses the l effects of observed temperature deviations in the hot and cold legs due to non-uniform temperature distributions. Additionally, Duquesne Light Company performed analysis to support several trip setpoints and allowable values that are proposed in this license i amendment request. For example, the ESFAS trip setpoints and allowable values associated with the 4.16 kV Emergency Bus Undervoltage - Degraded Voltage and 480 volt Emergency Bus Undervoltage calculations were recalculated by Duquesne Light Co. l and are also included. The revised setpoints, allowable values, and their supporting calculations have been developed by using the setpoint methodology as documented in the aforementioned WCAPs. ! The methodology used is the statistical " square root sum of the i squares" which has been utilized in other Westinghouse reports ! and is consistent with the current licensing basis methodology. l The methodology enacted explicitly accounts for the calibration
tolerance band of the final proposed setpoint nominal value.
B-5
I i ATTACHMENT B, continurd Licznca Ausndn3nt Raquest Nos. 220 and 88 ( Page 6 l The proposed amendment for the RTS and ESFAS trip setpoints and l allowable values will eliminate the need for the current l l administrative controls to ensure that non-conservative trip i setpoints are maintained consistent with safety analysis : assumptions. j The amendment also proposes changes with regard to the treatment of time constant nominal values. Westinghouse has determined that nominal time constants with settings that are within 10% of the value assumed in the safety analysis will provide adequate ! dynamic compensation for the impacted trip function. This ! conclusion is stated in Westinghouse letter DLC-96-310 of l December 20, 1996, to the Duquesne Light Company. In the BVPS Unit No. 1 Technical Specifications, the ESFAS Trip Function for Auxiliary Feedwater Pump Start on Emergency Bus Undervoltage is deleted since this function is not initiated directly by bus undervoltage. The undervoltage start of the motor-driven AFW pumps is accomplished indirectly via a combination of 1) Emergency Bus feed breaker opening 2) valid l Technical Specification required start signal from (Steam Generator low-low level, Safety Injection (SI) or trip of all operating main feedwater pumps), and 3) . diesel generator sequencer actuation. During the above occurrences, electrical power to the motor-driven AFW pumps is shed. The emergency diesel generator is started. The diesel generator sequencer is then activated to add predetermined safety loads which include the automatic start of the motor-driven AFW pumps. Diesel start j and sequencer operation on loss of emergency bus voltage and auxiliary feedwater pump start on the three actuation signals denoted above are included in the ESFAS related Technical Specifications, i l The current 5 column format of the BVPS Unit No. 2 trip setpoints ! is replaced with the 2 column format currently used in the BVPS l Unit No. 1 Technical Specifications and in the Improved Standard Technical Specifications in NUREG 1431 Revision 1. Both the 2 column and 5 column formats provide the same setpoint allowable value for use in operability determinations. The changes to the bases are necessary to preserve consistency with the changes to the Technical Specifications and to provide supporting information to clarify the intent of the proposed Technical Specifications. I l 4 B-6
/ , ATTACHMENT B, continutd
. 'Liczn32 Amsndment Request Nos. 220 and 88 Page 7 C. JUSTIFICATION Note: The marked up Technical Specifications contain a justification key in the right side of the page to classify the type of change. The keys are defined in the following paragraphs. The key prefixes are SP for setpoint, MR for more restrictive, AD for administrative changes, and B for Bases change. SP- The proposed changes to the RTS and ESFAS trip setpoints in 01 Technical Specification setpoints will be consistent with the current safety analysis limits, and the need to administrative 1y control these setpoints will be eliminated. Allowing nominal trip setpoints and nominal time constants (as defined with an equality) to include a calibration tolerance band will permit the trip setpoints to be set within the tolerance band assumed in the setpoint methodology WCAPs and the time constants to be set within the tolerance assumed in the safety analysis. Westinghouse has determined that the time constant calibration tolerance band allowed by the proposed change is acceptable and within safety analysis input assumptions. The action statements for LSSS 2.2.1 and LCO 3.3.2.1 would be revised to address the calibration tolerance band. This proposed change will provide a clear delineation'of the design analysis requirements as they are translated into the plant Technical Specifications. The deletion of the inequalities from the trip setpoints stated in Table 2.2-1 (with the exception of the turbine stop valve trip setpoint) and Table 3.3-4 along with the change in terminology from " Trip Setpoint" - to " Nominal Trip Setpoint" is consistent with the setpoint methodology used to develop the required trip setpoints. The proposed amendment will facilitate a two-sided calibration tolerance around the nominal trip setpoint. The uncertainty calculations, utilized in the development of the required trip setpoint, assume that the "as left" tolerance (both in the conservative and non-conservative , direction) is satisfied on a reasonable, statistical basis. It ' is acceptable for the "as left" condition, immediately after calibration of process rack modules or the bistable, to be in the , non-conservative direction provided that the magnitude is within I the calibration tolerance band specified in plant procedure (s) . The specified calibration tolerance band will consist of rack calibration accuracy in conjunction with the rack comparator settine, a:: curacy as required. These terms are defined in the setpoint WCAPs for each Unit. A note is added to each setpoint LSSS and LCO to allow setpoints to be set more conservatively than the nominal value as necessary in response to plant conditions or revised analysis. The proposed addition of this note is consistent with the existing Technical Specification requirements. The current Technical Specifications, with the inequalities in place, permit the trip setpoint to be set in the conservative direction of the specified B-7
0
, ATTACHMENT B, continutd License Amendment Request Nos. 220 and 88 Page 8 trip setpoint. It is consistent with the setpoint methodology for the "as left" trip setpoint to be outside the calibration tolerance band but in the conservative direction with respect.to the Nominal Trip Setpoint provided the allowable value(s) is adjusted accordingly. The conservative direction is established by the direction of the inequality applied to the allowable value. This note will facilitate, for example, the Power Range Neutron Flux High trip setpoint being set to a value less than specified in Table 2.2-1 during a condition where the Heat Flux Hot Channel Factor -
Fg(Z) exceeds the Technical Specification limit. In addition, an increase in the Steam Generator Water Level-Low-Low Setpoint may be required to account for revised environmental and process measurement allowances. The setpoint would be increased in the conservative direction in order to meet the revised analysis assumptions. l The format for defining the setpoint as a nominal trip setpoint l has been submitted and approved by the NRC for another Westinghouse plant (Vogtle) that uses the same setpoint methodology as Beaver Valley Power Station. Vogtle Technical Specification Amendments 101 (Unit 1 TAC No. MA0194) and 79 (Unit 2 TAC No. MA0195) implemented this format change for defining the setpoint as a nominal trip catpoint. The proposed values for the RTS and ESFAS trip setpoints are conservative in nature since there would be a reduction to the current trip setpoints with respect to plant safety analysis limits. A revision to the trip setpoints is being proposed to ensure that the current safety analysis limits continue to be met. Typically, the setpoint changes were due to the previously referenced plant-specific analysis. The amount of measurement error calculated was increased which results in the trip setpoint being required to be moved closer to the nominal plant operating values. l l The allowable values typically are being modified due to the l plant-specific analysis which resulted in changes to the uncertainties used in the determination of the allowable values. The net results of these analyses were changes to allowable values both in the conservative and non-conservative directions; I hcwever, this amendment request contains changes to only allowable values that are more restrictive and are currently under administrative control. The allowable value is used to
' determine instrumentation channel ' operability. The difference between the bistable trip setpoint and the associated allowable value is compiled by the addition of several uncertainties
associated with the process racks or the final output device, in l the cases of a remote actuation device. The original l calculations were completed using generic numbers for these I uncertainties. The latest revision of the WCAPs utilized plant-
! specific information for these variables. In several cases, i discrepancies were noted and centered around instrumentation uncertainties associated with the plant process racks. These B-8
, ATTACHMENT B, continusd i License Amendment Request Nos. 220 and 88 Page 9 items included rack comparator settings accuracy and rack measurement and test equipment accuracy.
The proposed change to the K1 value from 1.28 to 1.265 is conservative in nature in that the overtemperature AT setpoint would be reduced. This change is necessary due primarily to increased process measurement accuracy errors which were not included in the previous calculation of this setpoint. The increased process measurement accuracy errors were due to the effect of cold leg streaming errors in the RCS and their effects on temperatures and additional uncertainties addressed from Westinghouse technical bulletin (ESBU-TB-96-07-RO). SP- The change in the BVPS Unit No. 2 table format for trip 02 setpoints from a 5 column to a 2 column format is made because the flexibility allowed by the 5 column format, in making an operability determination to serve as the basis for making a determination relative to potential reportable events, is no i longer necessary. In 1983, 10 CFR 50.73 (as documented in NUREG ; 1022) changed the reporting requirements from a single inoperable channel to loss of functionality for a trip; i.e., reportability occurs when two or more channels are inoperable for a trip ; function. Therefore, the advantages of the 5 column format are i no longer present. Changing to the 2 column format is consistent l with the current format contained in BVPS Unit No. 1 Technical Specification, and in NUREG 1431, Revision 1 titled " Standard i Technical Specifications Westinghouse Plants." l MR- (Not used) ! 01 ! 4 MR- The values of T' in the overtemperature AT equation and the 02 value of T" in the overpower AT equation would be revised for BVPS Unit No. 1 only. The proposed change for T' and T" from 576.3*F to 576.2*F is conservative in nature and results in a reduced overtemperature and overpower setpoint. This change is necessary to make the values for T' and T" stated in the plant's Technical Specifications consistent with the nominal Reactor Coolant System average temperature assumed in the safety analysis. AD- The deletion of time constants T4 and T5 in BVPS Unit No. 1 Table 01 2.2-1 Notes 1 and 2 is being proposed so that the table correctly reflects the hardware configuration of the Unit 1 Trip System. The deleted time constants have been removed from the analog channel hardware by plant hardware change Design Change Package (DCP) 698. The current requirement is to set the time constants to a value equal to or less than 2 seconds. A deleted time constant is implicitly set to zero, which is consistent with the required setpoint value. The proposed change is made to remove the implication that the time constants exist in the hardware. B-9
v ATTACHMENT B, continutd J, Licence Amendment Rsquest Nos. 220 and 88 Page 10 AD- Several editorial changes are proposed in the BVPS Unit No. 1 02 Technical Specification action requirements to address "as found" setpoints that are between the nominal and allowable value, and to improve consistency with the BVPS Unit No. 2 wording. AD- In the BVPS Unit No. 1 ESFAS tables associated with Technical 03 Specification 3.3.2.1, the function for Auxiliary Feedwater (AFW) Pump Start or. emergency bus undervoltage is deleted because the undervoltage relay does not directly start the motor-driven AFW pumps. The motor-driven AFW pumps will start following emergency bus undervoltage after the emergency diesels start and the load sequencer picks up the bus loads provided the loss of power also resulted in a trip of all operating main feedwater pumps, SI, or Steam Generator low-low level. Diesel generator start and sequencer operation on a loss of emergency bus voltage and motor-driven AFW pump start on the three actuation signals denoted above are included in the ESFAS related Technical Specifications. NUREG 0800, " Standard Review Plan" (SRP), Appendix 8A, BTP PSB-1,
" Adequacy of Station Electric Distribution System Voltages,"
requires that when an undervoltage condition'has been detected, Class 1E loads are to be shed from the bus, and at BVPS Unit No. 1, the diesel generators are started. Loading of the emergency bus commences after the correct logic has been sent to the sequencer. AD- The proposed deletion of the various references to two loop 04 operation and the loop stop valves being open for BVPS Unit No.-1 is administrative in nature. Plant operation in Modes 1 and 2 with less than three reactor coolant system loops in service is not permitted by LCO 3.4.1.1. In addition, LCO 3.4.1.4.1 requires all reactor coolant loop isolation valves to be open during plant operation in Modes 1 through 4. Based on the requirements for loop operation and loop isolation valve position contained in LCOs denoted above, the license condition which prohibits two loop operation along with the other references to two loop operation and loop isolation valve position are no longer necessary and can be deleted. AD- For the turbine stop valve position trip channel, the current 05 setpoint and allowable value settings and inequalities are retained. This channel is an exception to the other channels listed in . Table 2.2-1 because the sensing element is a limit switch. This limit switch is only subject to relatively course adjustment. This device is not subject to electronic drift between functional tests like true analog channels which include electronic signal conditioning which must be periodically recalibrated on line. Also, the safety analysis does not require a specific setting for this trip function. Therefore, the current setpoint is retained and the nominal value as defined for the other channels does not apply. B-10
f ATTACHMENT B, continusd j 9 Licanza Am2nd snt Request Nos. 220 and 88 Page 11 AD- . Editorial changes including typographical corrections and format 06 changes were.made as required. These type of changes do not affect plant safety. B- The bases changes are made to ensure consistency between the 01 proposed Technical Specifications and the bases. The additional information is intended to make the Technical Specifications less ambiguous. D. SAFETY ANALYSIS By incorporating the setpoints in WCAPs 11366, Rev. 3 and 11419, Rev. 1, and additional calculations using the same setpoint methodology from Westinghouse and Duquesne Light Co., into the Technical Specifications, the assumptions in the safety analysis relative to initiation of RTS and ESFAS are preserved. The changes in format relative to the presentation of the nominal trip-setpoints and allowable values and the proposed revisions to LSSS 2.2.1 and LCO 3.3.2.1 have no impact on the safety analysis. These changes are consistent with the assumptions in the setpoint methodology WCAPs and the assumptions for time constants modeled in dynamic transfer functions that are credited in the safety analysis. < The retention of the inequality in the setpoint for the turbine stop valve position trip channel has no safety analysis impact because a trip setpoint which is anywhere in the range permitted by the inequality will be conservative. Furthermore, the trip function is not credited as a primary trip in the safety analysis. The deletion of the trip function for undervoltage start of the
. motor-driven AFW pumps (for BVPS Unit No. 1 only) does not have any impact on any safety analysis result. The proposed deletion of the lag compensation for AT and Tavg in the OTAT and OPAT reactor trip setpoint equations and associated time constants (for BVPS Unit No. 1 only) reflects the current hardware configuration of the BVPS Unit No. 1 plant protection system hardware and does not have any impact on any safety analysis result.
For BVPS Unit No. 1 only, the proposed changes for T' and T" from 576.3 F to 57 6. 2*F results in a reduced overtemperature and overpower trip setpoint. At the current 'value of 576.3*F for T' and-T", adequate margin still exists such that the plant's safety analyses are not affected. However, these values are being revised to maintain the setpoint channel margins specified in WCAP-11419 Revision 1. The change in format to a 2 column format for the BVPS Unit No. 2 setpoints has no safety analysis impact because this change is consistent with the Ossumptions in the methodology used to calculate the trip setpcints. B-11
_ _ _ _ . _ ._ _ .._. _._ _ _.- ..__ _._ _ ___._ _ _ _ _ _ _ . _ _ _ _ . ~ . . _ _ . _ ATTACHMENT B, continued ! o License Amendment Request Nos. 220 and 88 f Page 12 The proposed deletion of the references to two loop operation and i the loop stop valves being open will not affect plant safety. I LCO 3.4.1.1 and associated surveillance requirement ensure that ; the plant is not operated with less than three loops in operation i in Modes 1 and 2. LCO 3.4.1.4.1 and associated surveillance ' requirement ensure that the loop stop valves remain open. The remaining proposed changes are administrative in nature and do not affect plant safety. For the above reasons, the proposed Technical Specification ' changes have no adverse impact on any safety analysis result. E. NO SIGNIFICANT HAZARDS EVALUATION . The no significant hazard considerations involved with the proposed amendment have been evaluated. The evaluation focusing ' on the three standards set forth in 10 CFR 50.92(c) are as quoted below: The Commission may make a final determination, pursuant to the procedures in paragraph 50.91, that a proposed amendment to an operating license for a facility _ licensed under paragraph 50.21(b) or paragraph 50.22 or for a testing facility involves no significant hazards consideration, if l operation of the facility in accordance with the proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety. The following evaluation is provided for the no significant hazards consideration standards.
1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
This proposed amendment includes changes to nominal Reactor Trip System (RTS) and Engineered Safety Feature Actuation System (ESFAS) trip setpoints and allowable values that have been determined with the use of an approved methodology. The new values ensure that all automatic protective actions will be initiated at or before the condition assumed in the safety analysis. This change, which includes modification of the requirements stated in Limiting Safety System Setting (LSSS) 2.2.1 and Limiting Condition for Operation (LCO) 3.3.2.1, will allow the nominal trip setpoints to be adjusted within B-12
'o ATTACHMENT B, continued License Amendment Request Nos. 220 and 88 Page 13 the calibration tolerance band allowed by the setpoint methodology. There will be no adverse effect on the ability of the channels to perform their safety functions as assumed in the safety analyses. Since there will be no adverse effect on the trip setpoints or the instrumentation associated with the trip setpoints, there will be no significant increase in the probability of any accident previously evaluated. Other changes in trip system function, content and format are proposed based on the current configuration of the trip , system hardware at Beaver Valley Power Station (BVPS) Unit No. 1. Similarly, since the ability of the instrumentation to perform its safety function is not adversely affected, there will be no significant increase in the consequences of any accident previously evaluated. Since the safety analysis is unaffected by this change there is no change in the consequences of any previously evaluated accident. ! The editorial changes do not affect plant safety. The l administrative change, for BVPS Unit 1 only, pertaining to l two loop operation and Reactor Coolant System isolation valve position, does not affect plant safety. The Technical Specification requirements in LCOs 3.4.1.1 and 3.4.1.4.1 will continue to ensure safe plant operation by properly controlling the operation and position of the reactor coolant loops and Reactor Coolant System isolation valves. l Therefore, this change does not involve any significant increase in the probability of occurrence of any accident previously evaluated. !
2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed amendment includes changes to the format and magnitudes of nominal trip setpoints and allowable values that preserve all safety analysis assumptions related to ! accidr mitigation. The protection system will continue to initit the protective actions as assumed in the safety l analysAd. The proposed changes to LSSS 2.2.1 and LCO 3.3.2.1 will continue to ensure that the trip setpoints are . maintained consistent with the setpoint methodology and the plant safety analysis. This proposed amendment does not involve additional hcrdware changes. Plant operation will not be changed. Other proposed changes are made so that the Technical Specifications more accurately reflect the plant-specific trip system hardware in BVPS Unit No. 1. B-13
P-. p ATTACHMENT B, continued , License Amendment Request Nos. 220 and 88 Page 14 I Furthermore, the proposed changes do not alter the functioning of the RTS and ESFAS. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
3. Does the change involve a significant reduction in a margin of safety?
The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The proposed RTS and ESFAS trip setpoints are calculated with an approved , methodology. The proposed changes to LSSS 2.2.1 and LCO l 3.3.2.1 will continue to ensure that the trip setpoints are maintained consistent with the setpoint methodology and the plant safety analysis. Therefore, the response of the RTS l and ESFAS to accident transients reported in the Updated 1 Final Safety Analysis Report is unaffected by this change. No additional hardware changes are involved. Therefore, accident analysis acceptance criteria are not affected. Other proposed changes are made so that the protection system Technical Specifications more accurately reflect the plant-specific trip system hardware in BVPS Unit No. 1. The editorial changes do not affect plant safety. The administrative change, for BVPS Unit 1 only, pertaining to j two loop operation and Reactor Coolant System isolation valve position, does not affect plant safety. The Technical Specification requirements in LCOs 3.4.1.1 and 3.4.1.4.1 will continue to ensure safe plant operation by properly controlling the operation and position of the reactor coolant loops and Reactor Coolant System isolation valve. Therefore, operation of the facility in accordance with the proposed amendment will not involve a significant reduction in a margin of safety. F. NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the considerations expressed above, it is concluded that the activities associated with this license amendment request satisfy the no significant hazards consideration standards of 10 CFR 50.92(c) and, accordingly, a no significant hazards consideration finding is justified. I l l I l B-14
-. ... . . - . - - - .- . . . . . . _ _. . . - - .. - . - . - - - - . ~. . .
o ATTACHMENT B, continutd i License Amendment Request Nos. 220 and 88 ' Page 15 G. ENVIRONMENTAL CONSIDERATION The proposed amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. It has been determined that the proposed amendment involves no significant increase in the amounts, and no significant change in the types of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22 (c) (9) . Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this proposed amendment. H. UFSAR CHANGES No UFSAR changes have been determined to be necessary as a result of this proposed amendment. B-15}}

ML20199G543

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