ML20082C126

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Proposed Tech Specs Re Elimination of Selected Response Time Testing Requirements from TSs & Associated Base Changes for License NPF-57
ML20082C126
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 03/30/1995
From:
Public Service Enterprise Group
To:
Shared Package
ML20082C121 List:
References
NUDOCS 9504060171
Download: ML20082C126 (7)
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ATTACENENT E TECNNICAL SPECIPICATION PAGES WITN' PEN AMD INE CNANGES

. LICNNSE ANENDNENT APPLICATION 94-36, LR-N95040

. ELIMINATION OF SELECTED RESPONSE TINE TESTING REQUIRENENTS FACILITY OPERATING LICENSE NPP-57 NOPE CREEK GENERATING STATION DOCKET NO. 50-354 The following Technical Specifications have been revised to reflect the proposed changes:

Technical Soecification Page Table 3.3.1-2 3/4 3-6 Table 3.3.2-3 3/4 3-26 3/4 3-27 Table 3.3.3-3 3/4 3-38 Bases 3/4.3.1 B 3/4 3-1 Bases 3/4.3.2 B 3/4 3-2 Bases 3/4.3.3 B 3/4 3-2 i

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9504060171 95033o PDR P ADOCK 05000354 PDR .._..

.c TABLE 3.3.1-2 REACTOR PROTECTION hYSTEM RESPONSE TIES - ~

R m

'FMCTIOML UNIT RESPONSE TIE- -

_ (Seconds)

1. Intermediate Range Monitors:
a. Neutron Flux - High
b. Inoperative m m
2. Average Power Range Monttor*:
a. Neutron Flux - Upscale, Setdown
b. m
c. Flow Blased Simulated Therme1 Peuer - 5 scale < 0.09**

Fixed Neutron Flux - $ scale I 0.09

d. Inoperative R
3. Reactor Vessel Steam Dome Pressure - High 4 4. Reactor Vessel Water Level - Low, Level 3 < 0.55 4 i 1.05 *
5. Main Steam Line Isolation Valve - C1t.sure Y
  • 6. This item intentionally blank -7 0.06
7. Drywell Pressure - High  :
8. Scram Discharge Volume Water Level - High m I
a. Float Switch m
b. Level Transmitter / Trip Unit. m
9. Turbine Stop Valve - Closure m
10. < 0.06 Turbine Centrol Valve Fast = Closure. -

Trip 011 Pressure - Low '

11. Reactor Mode Switch Shutdown Position

< 0.001 %

12. Menue1 Scran R m

M ron detectors are exempt free response time testing. Response time shell be measured .

from the detector output or from the input of the first electronic component in the channel.

    • Not including simulated thersel peuer time constant 6
  • 0.6 seconds.

F Y W sured from start of turbine centrol valve fast closure.

  • 4W is eUmcafect fam response, 6mc testinci for +be RPS cirwits.Tesponse Emc tesbrg cvd ccnforrmnce.

w en administrawye, Urmas forthe arrainigcEnnnel freludin9 rip 4 antt arx1 ruby logic are reguirid, w i re ~h e

+ < , n - - - - _ _ . - . . . _ _ - - _ _ _ - - _ - . - _ _ _ _ _ . _ - . _ -

J TABLE 3.3.2-3 ISOLATION SYSTEN INSTRUNENTATION RESPONSE TIE .

TR., FUNCTION

1. RESPONSE TIE (Seconds k %

PRIMARY CONTAI N NT ISOLATION

a. Reactor Vessel Water Level
1) Low Low. Level 2
2) Low Low Low, Level 1 NA M-
b. Drywell Pressure - High
c. NA -

Reactor Building Exhaust Radiation - Hi1h t NA

d. Manual Initiat'on NA
2. -SECONDARY CONTAINNENT ISOLATION
a. Reactor Vessel Water Level-Low Low, Level 2
b. NA Drywell Pressure - High NA
c. Refueli Floor Exhaust Radiation -

High(b 1 4.0

d. Reactor Building Exhaust 5 4.0

' Radiation - High(b)

e. Manual Initiation NA
3. MAIN STEAM LINE ISOLATION

! a. Reactor Vessel Water Level - Low Low Low, Level 1 b.

c. Main Steam Line Pressure'- Low NA y W*/

i 4

d.

e.

Main Steam Line Flow-High u < 1.0 a i 0.58 g Condenser Vacuum - Low RA f.

i Main Steam Line Tunnel Temperature - Nigh NA 1 g. Manual Initiation M .

4.

REACTOR WATER CLEANUP SYSTEM ISOLATION

a. RWCU A Flow - High
b. NA RWCU A Flow - High, Timer ~

NA

c. RWCU Area Temperature - High
d. NA RWCU Area Ventilation A Temperature - High NA I

' a. SLCS Initiation MA f.. .

Reactor Yessel Water Level - Low Low, Level 2 NA

g. Manual Initiation NA 5.

REACTOR CORE ISOLATION C00 LING SYSTEN ISOLATION

a. RCIC Steam Line A Pressure (Flow) - High
b. NA RCIC Steam Line A Pressure (Flow) - High, Timer NA

! c. RCIC Steam Supply Pressure - Low NA

d. 'RCIC Turbim, Exhaust Diaphragm Pressure - High

, NA L

i l

HOPE CREEK 3/4 3-26 Amendment No. 53

~_ _ _ _ _ _ _ _ _ , _ _ _ _ -

Y

  • e TABLE 3.3.2 3 (Continued)

ISOLATION SYSTEM-INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)\ X

REACTOR CORE ISOLATION COOLING SYSTEM. ISOLATION 4 . e. RCIC Pump Room Temperature - High NA
f. RCIC P ;, Room Ventilation Ducts A Temperature *

- HiL NA

g. RCIC l- Aouting Area Temperature - High NA
h. RCIC Torus Compartment Temperature - High NA
1. Drywell Pressure - High NA
j. Manual Initiation NA
6. HIGH PRESSURE COOLANT. INJECTION SYSTEM ISOLATION
a. HPCI 5 team Line A Pressure (Flow) - High NA
b. HPCI Steam Line A Pressure (Flow) - High, Timer NA
c. HPCI Steam Supply Pressure - Low NA
d. HPCI Turbine Exhaust Diaphragm Pressure - High NA
e. HPCI Pump Room Temperature - High NA
f. HPCI Pump Room Ventilation Ducts a Temperature - High NA
g. HPCI Pipe Routing Area Temperature - High NA
h. HPCI Torus Compartment Temperature - High NA
1. Drywell Pressure - High NA

. J. Manual Initiation NA

7. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION
a. Reactor Vessel Water Level - Low, Level 3 NA
b. Reactor Vessel (RHR Cut-in Permissive)

Pressure - High NA

c. Manual Initiation NA (e) R:!:tien .y.t ;n.i. ntetien ree;;;n;; ;; specifiea inciudes div>=i
r
ter s+=" % and g en:: hediaii dehys.

(b) Radiation detectors are exempt from response time testing. Response time shall be measured from detector output or the input of the first electronic component in the channel.

  • Isolation system instrumentation response time for MSIVs only. No diesel generator delays assumed for MSIVs.

l "I.eletivo y te;, in;tr"- ntati:n re: pen;; th fer :::: ieted JEhe,-

l  :::;;pt i451Vs7

'/. M solation system instrumentation response time specified for the Trip Function actuating each valve group shall be added to isolation time l shown in. Table 3.6.3-1 and 3.6.5.2-1 for valves in each valve group to t

obtain ISOLATION SYSTEM RESPONSE TIME for each valve.

4

  • Sensor is eliminaJed. From respmse, time., tettfrg for 4he MSW cMm

.. (cfjic, cirettttS.TeSponse, ttmo testing cted cmforrynrce to 4he odrninidn:L+ive, limtli for 4hc rerrcuntng Chctnnel McILAciang trip LLntts oncL rtby (Egic Ctre.

l _

qWred .

i l

HOPE CREEK 3/4 3-27 i

TABLE 3.3.3-3 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES ECCS RESPONSE TIME (Seconds)

1. CORE SPRAY SYSTEM $ 27 & p,.
2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM i 40 L&
3. AUTOMATIC DEPRESSURIZATION SYSTEM NA
4. HIGH PRESSURE COOLANT INJECTION SYSTEM 5 35 L b
5. LOSS OF POWER NA l

s ponte- b'

  • a u m h *C#'

'"S S h ECCS oc M'* iO" l

' testirg.

.a r 1

HOPE CREEK 3/4 3-38 Amendment No. 24

o ,

3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:

a. Preserve the integrity of the fuel cladding,
b. Preserve the integrity of the reactor coolant system.
c. Minimize the energy which must be adsorbed following a loss-of-coolant accident, and l- d. Prevent inadvertent criticality.

, This specification pr Dies the limiting conditions for operation necessary to preserve the ability of a system to perfore its intended function even during periods when instrument channels may be out of service because of main-tenance. When necessary, one channel may be made inoperable for brief intervals to conduct required surveillance.

The reactor protection system is made up of two independent trip systems.

There are usually four channels to monitor each parameter with two channels in

, each trip system. The outputs of the channels in a trip systes are combined in a logic so that either channel will trip that trip system. The tripping of both trip systems will produce a reactor scram. The system meets the intent of IEEE-279 for nuclear power plant protection systems. Specified surveillance  ;

intervals and surveillance and maintenance outage times have been determined 1 in accordance with NEDC-30851P, " Technical Specification Improvement Analyses for BWR Reactor Protection Systee," as approved by the NRC and documented in i the SER (letter to T. A. Pickens from A. Thadani dated July 15, 1987). The i bases for the trip settings of the RPS are discussed in the bases for Specifi-cation 2.2.1.

The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are com-plated within the time limit assumed in the safety analyses. No credit was

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taken for these channels with response times indicated as not applicable.

i Response time eey be demonstrated by any series of sequential, overlapping or total channel test measurement, provided such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either (1) inplace, onsite or offsite test measurements, or j (2) utilizing replacement sensors with certified response times. Selech ri sersor' re.spc<se. tinw testieg requirements tuere., cUrninciacci. tnsed u-ptn j , ' NGCC -M2c)l , " .3gstern ArthfSc5 .for Glirn irnilm of Sejected hpcnse. TGne; l

TeWrcJ 'Reguirernerris,' as cy:prwed EN the. htR0 arti &comeracti uhresee

((e&.r 63 g. A. Pinetu fnern GrucL A. Ocger cicecci. December 2B,icS4.).

HOPE CREEK B 3/4 3-1 Amendment No. 26 JUN 51989

  • %3eleckd sensor respmse 4tme, lestirg ecq irements were, cumincetect baseci.
  • Mto NE.OO-32.2 Al r Sgshcrn Anolyses for E.itrrviccd(en o9 Sciec.+ed Respens; Th

. Testirg 6eq Jremenbga as approved W ahe Nu.ard dccu.mented v5 +he. SER (tener-

. fo R.A. Pinctlt f rem 6ruce. A. 6eger ciaded ()"cember-16,1994-).

INSTRUMENTATION BASES 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the OPERABILITY trip setpoints and response times for isolation of the reactor systems. Specified surveillance intervals and surveillance and maintenance outage times have been determined in kccordance with NEDC-30851P-A, Supplement 2, " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation Common to RPS and ECCS Instrumentation," and NEDC-31677P-A, "Tochnical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation."

The safety evaluation reports documenting NRC approval of NEDC-308FlP-A, Supplement 2 and NEDC-31677P-A are contained in letters to D. N. Grace from C.E.

Rossi dated January 6, 1989 and to S. D. Floyd from C. E. Rossi dated June 18, 1990. When necessary, one channel may be inoperable for brief intervals to conduct required surveillance. Some of the trip settings may have tolerances explicitly stated where both the high and low values are critical and may have a substantial effect on safety. The setpoints of other instrumentation, where only the high or low end of the setting have a direct bearing on safety, are established at a level away from the nortral operating range to prevent inadvertent actuation of the systems involved.

Except for the MSIVs, the safety analysis does not address individual sensor response times or the response times of the logic systems to which the sensors are connecte [ For M , ey sased valve., . 3 .ecend J lay i; es:u . d --

befnem the vulva .* mete 4c egy.. Tw. h . C, . vy.geted valve , it-4310ssumec Enat the a . f* . nnume .upply le leet end--le--Tester;d by Otettup of - the e&esgeiicy diesel-generators -I*-this-eventy-44Lre of 13 esconde i; essum;d befere Lhle v4Lua_atarts to move- In adottlan +n *h= pipe-break, the fa&-1me af tha arc.

l operat=d velie 4e-ee.mmed; thu: the :ign:.1 deiey (.snuor respor.:: i=

concurrent "ith the 10 eecend diesel at=rtup. The eafety an:ly;is ceneiders7h allcwebi: inventory lose-in-each : ee % ich la *"*- date*=inee th= valva = peed

.in-centmet4cr. with the 13 ..cei d deley. It fclicw; that chachisig the selvir-speeds-and *ke 13 eecead t!ma far emeroency power estah14.hmant will establish the re.yvnse time for the leeletion function:.

Operation with a trip set less conservative than its Trip Satpoint but within its specified Allowable Value is acceptable on the basis that the l

difference between each Trip Setpoint and the Allowable Value is an allowance

' for instrument drift specifically allocated for each trip in the safety l

)

analyses.

i 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION l

l

' The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control. This specification provides the OPERABILITY requirements, trip setpoints and response times that will ensure effectiveness of the systems to provide the design protection.#Specified 3

  • ECOS octucrHon (ngtrumentedien i.s ettmteaaeci. erun reseense, ume.
  • cat <n:,ept ce rernmt ueremer*s sf vi<n <

I ECr>-32.2.9t, "aguem Arolgscs for ElimUnHcn of Scicckd 'Vc5Fmemm.Te.m ro m Bruce A. ,

MRc, ctrd decurrented. A the sen ( teuer- +c R. A.Pinciti l e appw q +N gr ciakd Decernbcc 'l.8 (99b.

B 3/4 3-2 Amendment No. 70 l l HOPE CREEK 1 .

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