ML20155F742
| ML20155F742 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 11/03/1998 |
| From: | VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | |
| Shared Package | |
| ML20155F740 | List: |
| References | |
| NUDOCS 9811060103 | |
| Download: ML20155F742 (123) | |
Text
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VYNPS l i
TA8LE OF COPrrE?frS (Continued) 1-LIMITING SAFETY Pace No. . SYSTEM SETTING SAFETY LIMITS .
1.1 FUEL CLADDING UfrECRITY....................... 6 ... 2.1 1.2 REACTOR COOLANT SYSTEM........................ 18 ... 2.2 LIMITING CONDITIONS OF OPERATION Pace No. SURVEILLANCE 3.1 REACTOR PROTECTION SYSTEM..................... 20 ... 4.1 BASES 29 3.2 PROTECTIVE INSTRUMENT SYSTEMS................. 34 ... 4.2 A. Emergency Core Cooling System............. 34 ... A B. Primary Containment Isolation............. 34 ... B C. Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation................................ 34 ... C D. Air Ejector of f-cas System Isolation. . . . . . 35 ... D E. Control Rod Block Actuation............... 35 ... E F. Mechanical Vacuum Pump Isolation.......... 35 ... F G. Post-Accident Instrumentation............. 35 ... G H. Drywell to Torus d.P Instrumentation........ 36 ... H I. Recirculation Pump Trip Instrumentation........................... 36 ... I ES 75 l
3.3 CONTROL RCD SYSTEM............................ 81 ... 4.3 A. Reactivity Limitations.................... 81 ... A B. Control Rods.............................. 82 ... B C. Scram Insertion Times..................... 85 ... C D. Control Rod Accumulators.................. 87 ... D ,
E. Reactivity Anomalies...................... 88 ... E !
BASES 89 l.
3.4 REACTOR STANDBY LIQUID CONTROL SYSTDi. . . . . . . . . 92 ... 4.4 ,
1 A. Normal operation.......................... 92 ... A .
B. Operation with Inoperable Components. . . . . . 93 ... B l C. Liquid Poison Tank - Boron Concentration............................. 93 ... C BASES 97 l
v -
1 3G - I I J', (belated) -. -
M. bey &ded and Protease syssaw . . . . . . .% . . . k
- 3Y ~ L L. Rezater C.ve rsekian caeng sy stem A t.ansc.s - .. !
9811060103 981103'7 PDR P
ADOCK 05000271 PDRi W .#
l Amenw. ant No. G, 95 -ii-
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l VYNPS BASES: 2.1 ' Cont'd) f, metal-water reaction to less than it, to assure that core geometry
)
remains intact. l The design of the ECCS components to meet the above criteria was l dependent on three previously set parameters: the maximum break j size, the low water level scram setpoint, and the ECCS initiation setpoint. To lower the ECCS initiation setpoint would nowToprevent the ECCS components from meeting their design criteria. raise the i
ECCS initiation setpoint would be in a safe direction, but it would reduce the margin established to prevent actuation of the ECCS during normal operation or during normally expected transients.
E. Turbine Stoo Valve Closure Scram Trio Settinq The turbine stop valve closure scram trip anticipates the pressure, r
I neutron flux and heat flux increase that could result from rapid
. closure of the turbine stop valves. With a scram trip setting of
<10% of valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR remains above the fuel cladding integrity safety limit even during the worst case transient that assumes the turbine bypass is closed. This scram is bypassed when turbine steam flow is below 30% of rated, as measured by turbine first stage pressure.
F. Turbine control valve Fast Closure Scram The control valve fast closure scram is provided to limit the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection coincident with failure of the bypass system. This transient is less severe than the turbine stop valve closure with failure of the bypass valves and therefore adequate margin exists.
C. Main Steam Line Isolation valve closure Scram The isolation valve closure scram anticiprtes the pressure and flux transients which occur during normal or s.nat.vertent isolation valve
$1ve closure, there is closure. With the scram setpoint at lot c:
no increase in neutron flux. ..
H. Reactor coolant Low Pressure Initiation of Ma.1 Steam Isolation Valve closure l
The low pressure isolation of the main steam lines at 800 psig is provided to give protection against rapid reactor depressurization and the resulting rapid cool,down of the vessel. Advantage is taken of the scram feature which occurs when the main steam line isolation valves are closed, to provide the reactor shutdown so that high power operation at low reactor pressure does not occur. Operation of the reactor at pressures lower than 800 psig requires that the reactor mode switch be in the startup position where protection of the fuel cladding integrity saf ety limit is provided by the IRM high neutron flux scram.
Thus, the combination of main steam line low oressure isolation and isolation valve closure scram assures the(EEril tts?of neutron scram protection over the entire range of applicability of the fuel cladding integrity saf ety limit.
R.wdllJbiM 17 j Amenc.aen t No. 44, 46, 84
VYNPS-3.1 LIMITING CONDITIONS FOR 4.1 SURVEILLANCE REQUIREMENTS OPERATION 3.1 REACTOR PROTEC* ION SYSTEM 4.1 REACWR PROTECTION SYSTEM Aeolicability: *WhD Aeolicability: Castromtau Applies to the 6;:9*4]of Applies to_the surveil 1 ce of plant instrumentation ana the plant O_--------- i--Jand control systems required for control systems required for reactor safety, reactor safety.
Obiective: Obiective:
To specify the limits imposed on To specify the type and plant operation by those frequency of surveillance to be instrument and control systems applied to those instrument and required for reactor safety. control systems required for reactor safety.
Specification: Specification:
A. Plant operation at any power A. Instrumentation systems level shall be permitted in shall be functionally tested accordance with Table 3.1.1. and calibrated as indicated in Tables 4.1.1 and 4.1.2, The system response time from the opening of the respectively.
sensor contact up to and including the opening of the scram solenoid relay shall not exceed 50 milliseconds.
I During operation with the B. Once a day during reactor B.
ratio of MFLPD to FRP power operation the maximum
- greater than 1.0 either: fraction of limiting power density and fraction of
- a. The APRM System gains rated power shall be shall be adjusted by the determined and the APRM ratios given in system gains shall be Technical adjusted by the ratios given Specifications 2.1.A.1 in Technical and 2.1.B or Specifications 2.1.A.1.a and 2.1.B.
20 l Amendment No. 61 -
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(c , VYNPS
{ TABLE 3.1.1 t
REACTOR PROTECTION SYSTEN (SCRAM) INSTRtMENT REQUIREMENTS y
r Required .
Conditions "
Minimum Number When Miniaam
,! Hodes in Which Functions Hust operating Conditions For i be Operating Instrument Operation
'l Channels Per Are Hot
-} Trip Function Trip Settinos Refuel (1) Startup (12) Run Trip System (2) Satisfied (3)
Mode Switch in
. 1. X X X 1 A
! Shutdown {g y[
I 2. Manual Scram 4.- $3A[8 X X X 1 A
- 3. IRH -4t (A-F))
High F;ux $ 125/125 X X X(11)
'7 2 A Mf c INOP X X X(11) 2 A
- 4. APRH hPRM A-Fh i
High Flux 5 066 (W-AWi+54% X 2 A or B (flow bias) (4)
C High Flux (reduced) 515% X X 2 A IHOP X 2(5) A or B Downscale 12/125 X 2 A or B
- 5. High Reactor $1055 psig X X X 2 A I
Pra n = > i ra-6.
((PT-2-3-R(4-6)(MU 5 2 5 psig X X niva an yw.s 4 X 2 A
_ D r a n = = = = --
L ((PT-S-!5(A-b)(NA
- 7. Reactor t.ow to: 2127.0 inches X X X 2 . A 5~ 3 ~ /8(N)
- 8. a. .= uiscnarge $21 gallons X X X 2 A Volume High Level _
(per volume) l[. Qt.T-3-agl (A-t0(M)h ,
I (L1~-9 50.4/9 (M} d Amendment No. H, 44, 64 68, 76, 78, 79, 90, 94 2g A" .
~%
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y l l VYNPS dPi l l
E TAELE 3.1.1 W-Npet(t (Cont *d)
"y SEMc764 PR67Ec7/od Sysyf'M(scssa\ .fAlsTRUMGf EMUME#h Required if Conditions <
h... ;@ When Minimum !
Trip Settings Hodes in Which Functions Must operating Conditions For I And Allowable be operating Instrument operation l Trio Function Deviations Startup Channels Per Are Not Refuel (1) Run Trio System (2) Satisfied (3) h 9. Main steamline high 3x normal X X X f-2 A or C radiation (7) background at ji _ rated power (8)
- 10. Main steamline $10% valve X 4 A or C
[ isolation valve closure p
_ closure h.;;
E ! 11. Turbine control (9)(10) X 2 A or D valve fast closure s ,
"; 1 12. Turbine stop valve $10% valve (10) X 2 f
closure closure A or D (svoS-r-O-4)y 7 (p.s - (n-40))
(po 7 to A-Ali 8I BI po , - p.- gc, A - A 4 ps S %c8 -Ab ##'
po S - a - g48 - A b B
- po 5 - a - 7co c - A A> #I po, - a -gsc -A Di 8 I p o S - p -stop - A */ N#
(, p o S - a. - % 4 D
- A 5 82)
!{
f -f [- 3 l (A ~ 22 b
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TABLE 3.1.1 NOTES
- 1. When the reactor is suberitical and the reactor water temperature is less than 212*F, only the following trip functions need to be operable al mode switch in shutdown b) manual scram c) high flux IRM or high flux SRM in coincidence d) scram discharge volume high water level
- 2. Whenever an instrument system is found to be inoperable, the instrument system output relay shall be tripped immediately. Except for MSIV and >
Turbine Stop Valve Position, .this action shall result in tripping the trip system.
- 3. When the requirements in the column
- Minimum Number of Operating Instrument Channels Per Trip System
- cannot be met for one system, that system shall be tripped. If the requirements cannot be met for both trip systems, the appropriate actions listed below shall be taken a) Initiate insertion of operable rods and complete insertion of all operable raps within four hours.
b)- Reduce power level to* IRM range and place mode switch in the
- Startup/ Hot St position within eight hours, c) Reduce turbine 4eed and close main steam line isolation valves within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. loa
, d) Reduce reactor power to ess than 30% of rated within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
- 4. *W* is percent rated two loop drive flow where 100% rated drive flow is that flow equivalent to 48 x 10* lbs/hr core flow. AW is the difference between the two loop and single loop drive flow at the same core flow.
This difference must be accounted for during single loop operation.
AW = 0 for two recirculation loop operation.
- 5. To be considered operable an APRM must have at least 2 LPRM inputs per level and at least a total of 13 LPRM inputs, except that channels A, C, D, and F may lose all LPRM inputs from the companion APRM Cabinet plus one additional LPRM input and still be considered operable.
6.- The top of the enriched fuel has been designated as 0 inches and prov, ides common reference level for all vessel water level instrumentation
- 7. Channel shared by the Reactor Protection and Primary Containment Isolation Systems. j l
- 8. An alarm setting of 1.5 times normal background at rated power shall be i established to alert the operator to abnormal radiation levels in primary j coolant. )
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_. Amendment No. Et, Ea, 64, 46, 94 23
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VYHPS t
TABLE 4.1.1 SCRAM INSTRUMDRATION AND LOGIC SYSTD(S FUNCTIONAtr' TESTS i
MINIMUM FUNCTIONAL TEST PREOUENCIES FOR SAFETY INSTRUMDRATION, LOGTC SYSTD4S AND CONTROL CIRCUITS ,
Instrument Channel Group I3I Functional Test I7I Minimum FrequenevIII Mode Switch in shutdown A Place Mode Switch in Shutdown Each Refueling Outage Manual Scram A Trip Channel and Alarm Every 3 Months l
IU .
High Flux C Trip Channel and AlarmI6I Before Each Start Weekly During Refuelin '
Inoperative C Trip Channel and Alarm BeforeEachStartu'pWeekly-During Rsfueling APRM High Flux B Trip Output Relays (5) Once Each Week High Flux (Reduced) B Trip output Relays (5)
BeforeEachStartup'pWeekly During Refueling 5 l
Inoperative B Trip Output Relays Once Each Week Downscale B I ~
B Trip Trip Output Output Relays Relays(6I 5)
Once Each Week Flow Bias High Reactor Pressure B Trip Channel and Alarm (5) ;
} High Drywell Pressure B Trip Channel and Alarm (5) {}
Low Reactor Water Level (2)(8) B Trip Channel and AlarmI6I m ,
High Water Level in Scram Discharge B~ Trip Channel and Alarm (5)
Volume l High Main Steam Line Radiation I2I B Trip Channel and Alarm (5) Once Each Week' .
Main Steam Line Iso. Valve Closure A Trip Channel and Alarm (g)
Turbine Con. Valve Fast Closure A Trip Channel and Alam . (g)
Turbine Stop Valve Closure g A Trip Channel and Alam (g)
Scram Test Switch SA -SA (4-6) A Trip Channel and Alarm Each Refueling Outage ,
First Stage Turbine Pressure - A Trip channel and Alarm Every 6 Months :
i Permissive PM-M N-D l Amendment No. 14, 21 58, 76 25
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VYNPS TABLE 4.l.2 SCRAM INSTRUMEtR CALIBRATION MINIMUM CALIBRATION FREOUENCIES FOR REACTOR PROTECTION INSTRUMEtC CHANNELS Instrument Channel ' Group III Calibration StandardI4I Minimum FrequencyI2I High Flux APRM Output Signal B Heat Balance Once Every 7 Days Output Signal (Reduced) B Heat Balance Once Every 7 Days B Standard Pressure and Voltage Refueling Outage Flow Bias source LPRM 4A1 Nb-3-/-/&/(go) B(5) Using TIP System Every 1000 Equivalent Full Power Hours.
B Standard Pressure Source Once/ Operating Cycle-High Reactor Pressure A Standard Pressure Source Every 3 Months Turbine Control Valve Fast Closure High Drywell Pressure B Standard Pressure Source Once/ Operating Cycle l
High Water Level in Scram Discharge B Water Level Once/ Operating Cycle i Volume B Standard Pressure Source Once/ Operating Cycle Low Reactor Water Level Turbine Stop Valve closure A (6) Refueling Outage B AppropgteRadiation Refueling Outage-High Main Steam Line Radiation Source A Pressure Source Every 6 Months and After First Stage Turbine Pressure Refueling Permissive y,pg g.g A (6) Refueling Outage Main Steam Line Isolation Valve Closure 27 Amendment No. 14 E ag, 58, 61, 76
/ i BASES:
3.1 Reactor Protection Svstem The reactor protection system automatically initiates a reactor scram to: )
i
- 1. preserve the integrity of the fuel barrier;
- 2. preserve the integrity of the primary system barrier; and i
- 3. minimize the energy which must be absorbed, and prevent criticality l following a loss of coolant accident. l l
This specification provides the limiting conditions for operation necessary to preserve the ability of the system to tolerate single failures and still perform its intended function even during periods when instrument channels may be out of service because of maintenance, testing, or calibration.
The reactor protection system is of the dual channel type. The system is I made up of two independent logic channels, each having three subsystems of tripping devices. One of the three subsystems has inputs from the j manual scram push buttons and the reactor mode switch. Each of the two i remaining subsystems has an input from at least one independent sensor monitoring each of the critical parameters. The outputs of these subsystems are combined in a 1 out of 2 logic; i.e., an input signal on either one or both of the subsystems will cause a trip system trip. The outputs of the trip systems are arranged so that a trip on both logic channels is required to produce a reactor scram.
The required conditions wnen the minimum instrument logic conditions are not met are chosen so as to bring station operation promptly to such a condition that the particular protection instrument is not required; or the station is placed in the protection or safe condition that the instrument initiates. This is accomplished in a normal manner without subjecting the plant to abnormal operating conditions.
When the minimum requirements for the number of operable or operating trip system and instrumentation channels are satisfied, the effectiveness of the protection system is preserved; i.e., the system can tolerate a single failure and still perform its intended function of scramming the reactor.
Three APRM instrument channels are provided for each protection trip system to provide for high neutron flux protection. APRM's A and E
- operate contacts in a trip subsystem, andD,APRM's C an E operate contacts and F are arranged similarly in the other trip subsystem. APRM's B, in the other protection trip system. Each protection trip system has one more APRM than is r.ecessary to meet the minimum number required. This allows the bypassing of one APRM per protection trip system for maintenance, et.cing, or calibration without changing the minimum number of channels required for inputs to each trip system. Additional IRM channels have also been provided to allow bypassing of one such channel.
IRM assionment to the bypass switches is described on FSAR Figure 7.5-9 and @ T'? " 17 '
a pm ney ,ay 7. $~ g-The bases for the scram settings for the IRM, APRM, high reactor pressure, reactor low water level, turbine control valve fast closure, and turbine stop valve closure are discussed in Specification 2.1.
Instrumentation b r u-~ iG h- h is provided to detect a loss-of-coolant accident and Anattate the core standby cooling equipment.
This instrumentation is a backup to the water level instrumentation which is discussed in Specification 3.2.
29 Amendment No. 21 i .
...c-- m.
VYNPS M8 4.1 REACTOR PROTEC* TON SYSTEM A. The scram sensor channels listed in Tables A, 8 and C.
4.1.1 that Sensors and 4.1.2 make are up Group A l divided into three groups:
are the on-off type and will be tested and calibrated at the
) indicated intervals. Initially the tests are more frequent than Yankee experience indicates necessary. However, by testing more frequently, the confidence level with this instrumentation will i
' increase and testing will provide data to justify extending the test intervals as experience is accrued.
cenun B devices utilize an analog sensor followed by an amplifier and This type of equipment incorporates control bpfABg Mi :td'jortrip circuit. annunciator alarms. A failure in the
! room mounted indicators and sensor amplifier may be detected by an alarm or by an operator who
- observes that one indicator does not track the others in similar The bistable trip circuit failures are detected by the channels.
periodic testing.
Group C devices are active only during a given portion of the operating cycle. For example, the IRM is active during start-up and
' inactive during full-power operation. Testing of these instruments l 1s only meaningful within a reasonable period prior to their use. l l
S. The ratio of MFI.PD to FRP shall be checked once per day to determine if the APRM gains require adjustment. Because few control rod movements or power changes occur, checking these parameters daily is l adequate.
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33 Amendment No. 64, 61
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YYtiPS 3.2 LIMITING CCNDIT:Ctts FOR 4.2 SURVEILLAffCE REQUIRC4CfrS OPEFAT! N 3.2 PROTEC*:VE INSTRUMENT SYSTEMS 4.2 PROTEC"!VE INSTRUNDJT SYSTEMS Acol ica b i l i ty_ : Apolicabiliev:
Applies to the operational Applies to the surveillance requirements of the
- status of the plant instrumentation systems which instrumentation systems which initiate and control a initiate and control a protective function, protective function.
obiective: .obiective:
To assure the operability of To verify the operability of protective instrumentation protective instrumentation systems, systems.
Specifiestion: Specification:
A. Emercenev Core Coolino A. Emercenev Core Coolinc
_ System Svstem When the system (s) it Instrumentation and logic initiates or controls is systems shall be required in accordance with functionally tested and
, Speci f icat ionC . : Q calibrated as indicated in instrumentation wnica Table 4.2.1.
.I.If initiates the emergency core cooling system (s) shall be operable in accordance with Table 3.2.1.
Primarv Containment _ B. Primary Containment 3.
Isolation Isolation When primary containmen? Instrumentation and logic integrity is required, in systems shall be accordance with functionally tested and Specification 3.7, the calibrated as indicated in instrumentation that Table 4.2.2.
initiates primary containment isolation shall be operable in accordance with Table 3.2.2.
C. Reactor Buildino Ventilation C. Reactor Buildino Ventilation 4
Isolation and Standby Gas
- so lat ion and Standbv Cas Treatment System Initiation _
"reatment Svstem Initiation The instrumentation that Instrumentation and logic initiates the isolation of systems shall be functionally tested and the reactor building ventilation system and the calibrated as indicated in actuation of the standby gas Table 4.2.3.
treatment system shall be operable in accordance with Table 3.2.3.
34
_ _ _ _ _ _ _ . _ _ -m__._. ... _ . - . _ . _ . . _ ._. - . _ . - - _ _ _ . . - . . . . _ . - . ,
! l L
'NNPS 3.2 LIMITING CCNCIT!CNS FOR 4.2 SURVEIL .ANCE REQUIREMENTS l
( OPERATION D. Off-Cas 3vstem Isolation D. O f f-cas Syst em tsolation l l
During reactor power Instrumentation and logic ,
operation, the systems shall be !
instrumentation that functionally tested and.
initiates isolation of the calibrated as indicated in Table 4.2.4.
off-gas system shall be operable in accordance with Table 3.2.4.
E. Control Rod Block Actuation E Control Rod Block Actuation j During reactor power Instrumentation and logic operation the systems shall be instrumentation that functionally tested and initiates control rod block calibrated as indicated in Table 4.2.5.
shall be operable in accordance with Table 3.2.5.
Mechanical Vacuum Pump F. Mechanical Vacuum Pump F.
Isolacton Isolation
- 1. Whenever the main steam During each operating cycle, line isolation valves automatic isolation and are open, the mechanical securing of the mechanical vacuum pump shall be vacuum pump shall be capable of being verified while the reactor automatically isolated is shutdown.
and secured by a signal of high radiation in the main steam line tunnel or shall be manually isolated and secured.
- 2. If_ specification 3.2.F.1 is not met following a routine surveillance check, the reactor shall be in the cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
G.- Post-Accident G. Post-Accident Instrumentatton Instrumentatson During reactor power The post-accident operation, the instrumentation shall be instrumentation that functionally tested and displays information in the calibrated in accordance Control Room necessary for with Table 4.2.6.
the operator to initiate and control the systems used
'during and followi postulated accident m
08 abnormal operating con ition shall be operable in accordance with Table 3.2.6.
35
' Amendment No. 9 --
1 i
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- tMPS 3.2 LIMITINC CONDITIONS FCR 4.2 SURVEILLANCE REQUIREHEN!'S OPERATION l
H. Drywell o Torus 3 P. H. Drywell to Torus dP Instrumentation Instrumentation j
- 1. During reactor power 'Ihe Drywell to Torus AP operation. the Drywell Instrumentation shall be to Torus AP calibrated once every'six Instrumentation mor.ths and an instrument (recorder *1-156-3 and check will be made once par instrument DPI-1-158-6) shift.
shall be operable except as specified in 3.2.H.2.
- 2. From and after the date that one of the Drywell to Torus AP instruments is made or found to be i j
inoperable for any reason, reactor operation is permissible only during the succeeding thirty days unless the instrument is sooner made operable.
If both instruments are made or found to be inoperable. and indication cannot be restored within a six hour period an orderly shutdown shall be initiated and the reactor shall be in a hot shutdown condition in six hours and a cold shutdown condition in the following eighteen hours.
I. Recirculation Pumo Trio I. Recirculation Pume Trio Instrumentation Instrumentation During reactor power The Recirculation Pump Trip operation, the Recirculation Instrumentation shall be Pump Trip Instrumentation functionally tested and shall be G n ire) in calibrated in accordance accordance witn Table 3.2.1. with Tabla 4.2.1.
J. Deleted oNygt J. Deleted l l
K. Decraded Crid Pror/ective K. Decraded Grid Protective Svstets System
/
During reactor wer The emergency bus operation, th emergency bus undervoltage instrumentation undervoltage nstrumentation shall be functionally tested shall be G,- n .--D in and calibrated in accordance accordance with Table 3.2.8. with Table 4.2.8.
36 Amendment No. W. W. M. M. +H. 132 _ . . _ _ . _
< o 2 .
VYNPS TABLE 3.2.1
' EMERGENCY CORE COOLING SYSTEM AC1UATION THSTRUMERrATION Core Spray - A & B (Note 1)
Hinimum Number of Required Action When operable Instrument fag Minimum conditions
- Channels per Trip operation system Trio Function Trio Level Settino ee+ Not Satisfied i Hiah nrvum11 pra==ure i 2 <2.5 peig Note 2 CPr-toL ! al(A-b)in)D
~~
N 2 aow - menctor vessel Water 182.5* above top of Note 2 ,
LevelGL7.a.3-7a(4.b)(M)g enriched fuel 1 Low Reactor Pressure- 300 5 P $ 350 peig Note 2 ;
l (PT-2-3-56C/D(H)) !
2 Low Reactor Pressur P T-1.- & 300 1 P $ 350 psig Note 2 l (PT-2-3-56A/B(H) & 52C/D(H)) I 1 Tisme Delay (14A-K16A & B) 510 seconde Hote 2 2 Pump (P-46-1A/B) Discharge 1100 peig Hoce 5 Pressure Q pg_ p q g ,g .
1 A le Q #er Monitor (LuPx c/b Note 5 1 - - _ ) "^rita-r --
Hoce 5 .
1 _(Ta?/3Na> asap 31/JA/a) syss
._ } --
Note 5
. {
. k I
t 8
T l!
s a
! Amanda.nt No. 44, sa, us, n+, we', H2 is l
. I: +
+
- . .. . -. . . . - ~ - . - - . - . . ,_ .
1.s
. t.
3; J.
- TABLE 3.2.1 .
(Cont'd)
- I ENERGENCY CORE COOLING SYSTDI ACTilATION THSTRtAIENTATION q .
1 Low Pressure Coolant Injection System A & B (Note 1)
- Hinimusa Number of Required Action When l
operable Instrument (FT-10-to/(A-b) (M)h Conditione Channele per Trip Systesa Trio Function Trio Level Settina h*Hi operation Hot Setlefied i
1 Low Reactor Pressure 300 $pS 350 pelg Ags Note 2 n]t. l .
(PT-2-3-56C/D(H))
! 2 Minh Drvuell pressure <2.5 peig Note 2 !
- d 10101". 00Q
~~
l ,
Ji 2 Low-Low neoctor veeeel water z 2.5 a above top oe Hote 2 1
3 Level g_y;7pg4,b)(5/))] enriched fuel
} 1 Time Delay (10A-K51A & B) . O seconde Hote 5 f Reactes- - W...i ei. 4 Level 4
' . 1 1 2/3 core height. Note 5 i
((LT .t-3-?JNg(M2u i 7, . j e~
1 Time usasy gava-ns2A,6 B) 160 esconde Hots o 1 Time' Delay (10A-K50A & B) $5 esconde Note 5 I - .
~
t 1 Low Reactor Pressure 100 $ p 1 150 peig Note 2 (PS-2-128A & B) ,,
j f 2 per pump RHR Pump ". ' S Dime ram -
1100 peig Note 5
} Preneure ps-le-for(A-g)Q 2 Hiah nrvwell Pressure ~<2.5 peig Hote 2 !
@ - 10 101". S'*H O , '
?
I f.
P1*--le-IoI (A-b) (SI)) <
j g t t i.
I k I Amendment No. 44, 44, 64, 114 142 M y i t
s
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i VYNPS TABLE 3.2.1 (Cont'd) f05 D4ERGENCY CORE COOLING SYSTD( ACit!ATION INSTRIMDTPATION i
. Low Pressure Coolant Injection System A & B (Note 1) '
f Minimum Number of Requ d Action When l Operable Instrument Hi Conditions Channels per Trip s ration l
System Trip Function Trio Level Settino Not satisfied 1 Time Delay (10A-K45A & B) 16 minutes Note 5 2 Low Reactor Pressure 300 g p 5 350 psig Note 2 -
l (PT-2-3-56A/B(H) & H))
Auxiliam D~~ r Monitor -- Nota 5 .
1
((LkPX c 4g 1 ~"- ----. n i e n" --- -- Note 5
.-- )
1 T(rw 1.---P"=[A .~w - ?/3 A/8.737/44/8))
-- Note 5 pr g Sac /O t
u
+
8'O
~ .
I
'O Amendment No. 4+ 4M,142
. 1 VYNPS TABLE 3.2.1 *
(Cont *d)
D4ERCENCY CORE COOLING SYSTEM ACWATION INSTRtMENTATION o
~
High Pressure Coolant Injection System Hinimum Number of Required Action When Operable Instrument Hin Conditions Channels per Trip fo operation System Trio Function Trip Level Settina Hot Satiefied !
l 2 (Note 3) Low-Low Reactor Vessel Water Same as LPCI gg Note 5 -
Leve((tr.a. ,.,a ga.s)(,,yQ 2 (Note 4) Low condensate Stormaa Tank > 3% t.ote 5 WaterLevelhtSL-/e'f-5A/O 2 (flote 3) Hiah nrvsr.11 pre =<iira Same as LPCI Note 5 1 (Note 3)
RPT-16-/off4-b)(M)0 aus - =" =^ai r --
Note 5 ~
Qi33A -K11 1 (Note 4) Trip sysurm wg c --
Note 5 .
2 (Note 7) High Reactor vessel Water <177 inches above top of Note 5 .
Level --
enriched fuel (LT-a-3-7.1A/s) (sy) :
1 5..
l, o .
I j[ '
.. t .
o ,
b .
Amendment tio . 68, 85, 90 4g -
__ _ _.-._m__ _ _ _ . _ _ _
. ~.
a VYNPS TABLE 3.2.1 (Cont *d) -
D4ERGENCY CORE COOLING SYSTD4 AC'!UATION INSTRtMENPATION Automatic Depressurization Minimum Number of Required Action When Operable Instrument [py-fo-/o/[4 3)(yfp Minimusa Conditions
- Channels per Trip Operation System (Note 4) Trip Function Trio Level Settino 2 = Not Setisfied 2 Low-Low Reactor Vessel Water Same as Core Spray A#sD Note 6 .
Leve1Q_g.3-7p(4 5)(g) '
2 High Drywell Pressure $ 2.5 psig Note 6 M l Time Delay O 2 22 --2 0]
r 1 $120 seconds Note 6 1
G a e.
Bus A~- r PG.-KSA/A --
Hote 6 -
TripUM-k 1 sy-w-!A/sD m ic --
Note 6 2 Time Delav 18 minutes Note 6 '
0, 20 "1 4 - ' 2 Y(120 .CA 2..2 ;
&E-M%&)$, .z f-K17 A/S) t 1
i b
t r;
I F
i
(
0 f Amendment no. u. 1os 42 !',
1
VYNPS~
TABLE 3.2.1 (Cont *d) -
RECTRCtJLATION PtMP TRIP ACTtJATION INSTRtJMEffrATION Recirculation Pump Trip - A & B (Note 1) ;
Minimum Number of Required Action Mien Operable Instrument Hi Conditions Channels per Trip Operation System Trio Function Trip Level Setting Y M eatisfied .
2 Low-Low Reactor vessel Water 1 6* 10.S* above top of 4r Hote 2
- Level entiched fuel y y_3_gg(4,g 2 High naar-enr Pra a nin re 5 1150 psi 9 Note 2 2
TPM Time o2-3-5y palav- (A-b))>3 $ 10 seconda Note 2 1
Cia ss (A-b)(x)))
Trap systarms L.og s u --
Note 2 d
r A
e i
s .
F f
! mend.ent No. so. a. a. >s u !
-~ . - ...- - _ ~ . - . . . .
1 WNPS I* TABLE 3. 2.1 NOTES _
l 1. Each of the two Core Spray, LPCI and RPT. subsystems are initiated and controlled by a trip system. The subsystem *B' is identical to the subsystem 'A*.
- 2. If the minimum number of operable instrument channels are not available, the inoperable channel shall be tripped using test jacks or other permanently installed circuits. If the channel cannot be tripped by the means stated above, that channel shall be made operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or an orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- 3. One trip system with initiating instrumentation arranged in a one-out-of-two taken twice logic. g.,g7, g.,,7pg
- 4. One trip system with initiating instrumentation arranged in a h _a :D
-ewe logic.
I
- 5. If the minimum number of operable channels are not available, the system is considered inoperable and the requirements of Specification 3.5 apply.
I
- 6. Any one of the two trip systems will initiate ADS. If the minimum number of operable channels in one trip system is not available, the If i requirements of Specification 3.5.F.2 and 3.5.F.3 shall apply. the l l
- minimum nwnber of operable channels is not available in both trip systems, specifications 3.5.F.3 shall apply. 1
- 7. One trip system arranged in a two-out-of-two logic.
a 44 Amendment No. 58
- - - , w- - , e 'V$++'-r
VYNPS TABLE 3.2.2 PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION Minimum Number of Operable Instrument go7 Rep ired Action When Hin mum Conditions
[N Channels per Trip .cr Operatio System Trip Function Trip Settiner Not atisfied (Note 2) 2 Low-Low Rea<-ur ha-al Water >82.5* above the top of
_ A Level enriched fuel (tr-a-3 LT 3r74/d((s.t),
-18 4}e sa))
2 of 4 in each of High Main Steam Line euca $212*F B 2 channels Tem,,rature Grg-;n-(12/~/a*/ -)(A ~b))]
2/ steam line Hiu -1n x--- tine r- , ~<1401 cf rated flow B Lo(w(DP rnin T-;1_-Ges m--- -/f9XA-D)(M)))
"- n essute ~>800 psig B 2/(Note 1) '
2/(Note 6) Hi(an m l/4,Ay :nCPS-a2 m emu - -
/t 78,118C, -/3st1b
- / (k~bN)'
<40% rof_-q(51)j, rated flow ~ B 2
(TDPr.m;)im Low a --..-1
- nacer Level , Same as Reactor A (47 .2-3-574/S(.A Protection System 2 High2MainT-2 ~3 -S84/S steam Lw- (M))i .__ ion $3 x background at rated a power (9)
(7) f8) gg_f9_gfjgp 2 High Drywell Pressure Same as Reactor A Protection System 2/(Note 10) Condenser Low Vacuum 112* Hg absolute A l
1 Trip System Logic --
A l
Amendment No. 9, 68, 84, 86, 90 45
s VYNPS TABLE 3.2.2 (Cont'd)
- HTCH PRESSUPE COOLANT TNJECTION SYSTEM TSOLATION INSTRIMErfrATION Minimum Number of
- Required Action When Operable Instrument Minimum Conditions Channels per Trip Operatio ee+ #f system Trio Function Trio Level Settino satisfie 2 per set of 4 High Steam,Line Space 5212*P Note 3 Ta=naratura
({rs-as- (sol -1o4) (6-b)))_
1 Hagn scoam 1Ine afp (steam $195 inches of water Note 3 Line Break)hPrs-M-77hh 4 (Note 5) fnw HPcf Steam Su ply Pressure 370 psi 9 Note 3
((P5 M (A -
2 Haln steam Line_wnnel 1 212*F Note 3 Temperatur((ry-g3 (fog,fov)A 1 Time Delay (23A-K48) 535 minutes Note 3 (23A-K49) 1 Bus Power Honitor gg, --
1 Trip System Logic --
k 2
I i Amendment No. 68, 111 46 l
VYNPS QL TABLE 3.2.2
]
(Cont *d) ,
REAC1T)R CORE ISOLATION COOLING SYSTD4 ISOLATION INSTRUMDTTATION Hinimum Number of Required Action When
, Operable Instrument r M mum conditi 1: Channels per Trip Operation System Trio Function Trip Level Setting _u Satisfied (Note 2) f?
2 Main. Steam Line Tunnei $212*F Note 3 Temperature g _f3_(3.p)A))
.i 1 Time Delay (13A-K41) 535 minutes Note 3
.! (13A-K42) fi 2 per set of 4 High Steam Line Space $ 212*F Note 3 xl. Te=n*r=r"""
'?} G r$-13-(11-82 6, c,b))
team $195 inches of water
~
< 1 Hign steam a.ine asp Note 3 SPts-13-83/94))
l 4 (Note 5) **--- cunniv Pressure - 150 psig Note 3 5-/3 -87(4-bD) 1 Bus rower nonitor --
Note 3 2 1 Trip System Logic --
Note 3 i
i k
i h D K35
[ (15A r >
- r. .
r i
4 xmena.ent No. 69, 111 42 l
1_ - - _ - _
VYNPS gpeyggon AM TABLE 3.2.3 REACTOR BUILDING VEPTPILATION ISOLATION & STANDBY CAS TREA1HENT SYSTD4 INITIATION i
(C T-42 57A/6 (At), Required Action When Minimum Number of Minimum conditions.
Operable Instrument 47,g_y_fg4/g(A,)) S- CpuaiO Channels per Trip Trip Settinq .m f System Trip Function m .- uve Low Reactor Vessel Water Level Same as PCIS Note 1 2
'7 Hi h nrwall Pr***u Same as PCIS Note 1 2
7.~; S-).2 (4-D)fot)
^- -- Note 1 1 Rea i : 1 ^ 1 - ~<14 mr/hr 1
((R?t~lf~
Recue: */Sb- k/4))
1 m1 rionr - "a diation
<100 mr/hr Note 1 Rea @cguttuiuw mPt-ll".YS==uu W ip
$ -- Note 1 1
System Logic Standby Cas Treatment Trip -- Note 1 1
System Logic Logic Bus Power Monitor -- Note 1 1
((Ic. A -M.2/S3 is not available in either trip system for more Note 1 - If the minimum number of operable instrument channels than 24 hours, the reactor building ventilation system shall be isolated and the standby gas treatment system operated until the instrumentation is repaired. 49 e aw
.m.
VYNPS I corabtfl*45 1 oygAkiled ,gg TABLE 3.2.4 g yAf/II#gg OFF-CAS SYSTD4 ISOLATIOtt INSTRIMDTTATION Hinisuun Humber of Required Actinn When Operable Instrument Miniarm Channels per Trip . .. _ , _ . . ; ; e.. System Trip Function Trip settino :.re ;u ;;., i 1 Time Delay (Stack Off-Gas 5 2 minutes Note 1 valve Isolation) (ISTD & 16TD) $ 30 minutes 1 Trip System Logic -- Hote 1 Note 1 - At least one of the radiation monitors between the charcoal bed system and the plant stack shall be operable during operation of the augmented off-gas system. If this condition cannot be met, continued . operation of the augmented off-gas system is permissible for a period of up to 7 days provided that at least one of the stack monitoring systems is operable and off-gas system temperature and pressure are measured continuously. Amendment No. 9, 83 50
VYNPS TABLE 3.2.5 CorTTROL ROD BLOCK INSTRUMENTATION Minimum Number of Modes in Which Function l Operable Instrument Must be Operable Channels per Trip Refuel Startup Run System Trip Function Trip Setting l Startup Range Monitor 2 a. Upscale (Note 2)G-W(4'Olh X X $5 x 105 cps (Note 3) 2 b. Derecent une pullv Inserted X X (19-11(4-0)(L5_-4)D Intermeatace stange zwuiuor ; l (Note 1) 2 a. Upscale h 4/[4*F)D X X $108/125 Full Scale 2 b. Downscale (Note 4) d7-a//[4-#)h X X 35/125 Full Scale 2 c. Detector Met Fully Inserred X X G ?_-si(e,6G,MA K){ar Average m -r -n - mu . Ls -y))]
+
mow n-rD) 2 a. Upscate (fiv. oias X $ 0.66(W-AW)+421 (Note 5) 2 b. Downscale X >2/125 Full Scale l Rod Block Monitor (Note 6) gggg/g}} (Note 9) 1 a. Upscale (Flow Bias) (Note 7) X $ 0.66(W-AW)+N (Note 5) I b. Downscale (Note 7) X 32/125 Full Scale l 1 Scram Discharge Volume X X X $12 Gallons-(Note 8) (per n g g /g (9 volume) 1 Trip System Logic X X X Amendment No . H , 25, 64, 66, M, M, M, 94, 131 51 _ ._________________________m_ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ __ ___m - _ _ _ _ _ _ _ _ _ .
VYNPS 4 TABLE 3.2.6 POST-ACCIDENT INSTRIMENTATION Minimum Number of Operable Instrument Channels Parameter Type of Indication Instrument Rance 2 Drywell Atmospheric Recorder GTR-16-19-45 0-350*F Temperature (Note 1) CC 15 10 30AQ Meter GTI-16-19-30B 0-350*F 2 Containment Pressure (Note 1) Meter GPI-16-19-12A (-15) -(+260) psig Meter NPI-16-19-12B (-15) -(+260) psig 2 Torus Pressure (Note 1) Meter SPI-16-19-36A (-15) -(+65) psig Meter #PI-16-19-36B (-15) -(+65) psig 2 Torus Water Level (Note 3) Meter GLI-16-19-12A 0-25 ft. - - - - Meter SLI-16-19-12B 0-25 ft. h-/C, sy-33 A7 2 Torus Water Temperature Meter ~; 10 - 3 3;r V 0-250*F kK~ #C'~d ' (Note 1) Heter 410-13-33C; 0-250*F 2 Reactor Pressure (Note 1) Meter 8PI-2-3-56A 0-1500 psig Heter GPI-2-3-56B 0-1500 psig 2 Reactor Vessel Water Level Meter '"2 3 OIA? (-200)-0-(+200) *H 2O (Note 1) Heterif2-3-316 (-200)-0-(+200)*H O 2 2 Torus Air Temperature (Note 1) Recorder STR-16-19-45 0-350*F g C'"2 1C l ', , .y Meter BTI-16-19-41 50-300*F 2/ valve Safety / Relief Valve Position Light 2 -1, , 3 Closed - Open From Pressure Switches QA t D)) ~ (Note 4) g F RV-a-71y-c i rWLE- D 9l A \ Fran Ps-a-7/-l/~3)(A -bY ;
& t c. 2-3-9/B}
l Amendment No. GG, 63, 94, 96, 444,145 53 - se +h w mee+- m mN -6 6
- _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _A.
1 VYNPS TABLE 3.2.6 (Cont'd)
. POST-ACCIDEWP INSTRUMEfffATION ,
EI Minimum Number of Operable Instrument Channels Parameter Tvoe o Indication Instrument Rance 1/ valve Safety Valve Position From Mete 2-1A/B Closed - Open Acoustic Monitor (Note 5) 2 Containment Hydrogen / Oxygen SR-VG-6A 0-30% hydrogen Monitor (Note 1) , SR-VG-6B 0-25% oxygen , containment High-Range Meter RM-16-19-1A/B 7 2 i 1 R/hr-10 R/hr : Radiation Monitor (Note 6) { 1 Stack Noble Gas Effluent Meter RM-17-155 0.1 - 107 mR/hr . (Note 7) i Reconen) ; i L I [ t i i- {
..n ._ No . e..._...
t eh- ------_.mausv_a -.
?
VYNPS-TABLE 3.2.7 i t f t : l< (Table 3.2.7 was intenhonally deleted from the Technical Specifications) t . I t i t L [ t i 1 t- I d t t j s r" ; F t I F
- 55a !
F ,
I VYNPS TABLE 3.2.8 ENERGENCY BUS UNDERVOLTAGE INSTRUMENTATION Minimum Number of Operable instruments Parameter Trio Settino Required Action 2 per bus Degraded Bus Voltage - Voltage 3,700 volts i 40 volts Note 1 (27/3Z, 27/3W, 27/4Z, 27/4W) 2 per bus Degraded Bus Voltiage - Time 10 seconds i 1 second Note 2 Delay (62/3W, 62/3Z, 62/4W, 62/4Z) TABLE 3.2 NOTES
- 1. If the minimum number of operable instrument channels are not available, the inoperable channel shall be tripped using test jacks or other permanently installed circuits within one hour.
- 2. If the minimum number of operable instrument channels are not available, reactor power operation is permissible for only 7 successive days unless the system is sooner made operable.
i t i' I . *
\ + .
j. l Amendment No. 98 56
~
VYNPS TABLE 3.2.9 REACTOR CORE IsotATION COOT.ING SYSTEM ACWATION INSTRIMENTATION Hinimum Number of Required Action When Operable Instrument Channels per Trip Hi imum Conditions System r Operation Trio Function Trio Level Settino o&t Satisfied 2 (Note 1) Low-Low Reactor Vessel Water >82.S* Above Top of [ Note 4 Level y ,g.3 -)g-0)(M)) Enriched Fuel 2 (tiote 2) Low Condensate Storage Tank >3% Water Level tiote 4 LT-/07 -12 A/ 2 (Note 3) High Reactor vessel Water $177* Above Top of Level flot e 4 {( g7_,7,3,7=l C/b Enriched Fuel 1 Bus Power Monito (f2) -- 13A -k % tiote 4 1 Trip System Logi -- Note 4 Amendment No. 111 *
$7
.n VYNPS TABLE 4.2.1 (Cont'd) .
MINIMUM TEST AND CALIBRATION FREOUENCIES EMERGENCY CORE COOLING AC111ATION INSTRIMEffrATION Recirculai. ion Pump Trip Actuation System Trio Function Functional Test (8) Calibration (8) Instrument Check Low-Low React Water Level Vessel ( Ni a.Ih once/ Operating cycle Once Each Day Reactor Pressure y nr 1 Once/ Operating Cycle Ooce Each Day l Trip System Logic Once/ Operating Cycle Once/ Operating Cycle -- )
~
(notes / Anib 'l - t 1 1 w 1 e i
'I l
Amendment No. 58, 106 63
I Il Illl i!Il 1i i
~
)
s n o it a m ic f ic ep S la ic n . h c e T e ht 7 m S 2 o r f P 4 N E d e Y L t e V B l A e T d ly l a n io t n t e in s a w 7 2 4 le b a T (
VYNPS TABLE 1.2 NOTES
- 1. Initially once per month; thereafter, a longer interval as determined by test results on this type of instrumentation.
- 2. During each refueling outage, simulated automatic actuation which opens all pilot valves shall be performed such that each trip system logic can be verified independent of its redundant counterpart..
- 3. Trip system logic calibration ahall include only time delay relays ~and timers necessary for proper functioning of the trip system.
- 4. This instrumentation isfi5:..i 4)from functional test definition. The functional test will consist of ujecting a simulated electrical signal into the measurement channel.
( 5. Deleted. FKC#PI
- 6. Functional tests, calibrations, and instrument checks are not required when these instruments are not required to be operable or are tripped.
Functional tests shall be performed before each startup with a required frequency not to exceed once per week. Calibration shall be performed prior to or during each startup or controlled shutdown with a required frequency not to exceed once per week. Instrument checks shall be performed at least once per day during those periods when instruments are required to be or.reble.
- 7. This instrumentation is excepted from the functional test definitions and shall be calibrated using simulated electrical signals once every three months.
- 8. Functional tests and calibrations are not required when systems are not required to be operable.
- 9. The therwecouples associated with safety / relief valves and safety valve position, that may be used for back-up position indication, shall be verified to be operable every operating cycle.
- 10. Separate functional tests are not required for this instrumentation. The calibration and integrated ECCS tests which are performed once per operating cycle will adequately demonstrate proper equipment operation.
- 11. Trip system logic functional tests will include verification of operation of all automatic initiation inhibit switches by monitoring relay contact movement. Verification that the manual inhibit switches prevent opening all relief valves will be accomplished in conjunction with section 4.5.F.1.
Amendment No. 64, 96, 146, 145 74
l.. ) l
. 7TMFS ; 'I *~ - 3.3 LIMITZM3 CONDITIONS FCR 4.3 SORyutf.f. m -T REQUIREMENTS j OF= m'Pf0N with the required i shutdown margin not mac l . and the mode switch in the
- Refuel
- Position, i- immediately suspend
- t Alteration of.the .-
Reactor Core except for I control rod insertion and fuel assembly removal *: immediately - - i initiate action to fully insert all insertable - l control rods in core cells containing one or more fuel assembliess within 1 hour, initiate
. action to restore the integrity of the l
Secondary conta4n==nc System.
- 2. Reactivity Marvin - 2. Reactivity Marvin -
Inocerable Control Rods Inocerable Control Rods DNIN M i r M which Each parcially or fully cannot be movec with withdrawn operable control rod drive control rod shall be pressure shall be exercised one notch at ~ considered inoperable.
- 1 east once each week.
If a partially or fully This test shall be withdrawn control rod performed at least once drive cannot be moved per 24 hours in the with drive or scram event power operation is pressure, the reactor continuing with two or shall be brought to a more inoperable control shutdown condition rods or in the event withis 48 hours unless power operation is
- , investigation continuing with one
' demonstrates that the fully or partially cause of the ' failure is withdrawn rod which not due to a failed cannot be moved and for control rod drive which control rod drive w h=a4sm damage has no:
w h=nism collet housing. The control been ruled out. 'The rod directional control surveillance need not be completed wi.hin valves for inoperable l control rods shall be 24 hours if the number L i i ! sia [ amendment me.148
- ... . . . . . . _ . _ . _ _ ._. _ _ _ . . . _ . . - -Q
. VYNPS
~
- - 323 LIMITING CONDITIONS FOR - - ~
+. 3 - - SURYBHAANCE -REQUMBfENTS---- --- '-- I
} OPERATION
- E. Reactivity Anomalies E. Reactivity Anomalies The reactivity equivalent of During the startup test the difference between the program and startups actual critical rod following refueling outages, configuration and the the critical rod expected configuration configurations will be during power operation shall compared to the expected l not exceed it Ak/k. If this configurations at selected c
limit is ex'eeded, the ' operating conditions. These reactor will be shut down comparisons will be used as until the cause has been base data.for reactivity determined and corrective monitoring during subsequent actions have been taken if power operation throughout such actions are the fuel cycle. At specific appropriate. power operating conditions, the critical rod l F. If(treificcti . 2.;;J configuration will be throu abo, e are not met, compared to the an r y sh pdown shall be configuration expected based tiated and] ' the reactor upon appropriately corrected shall be in ti e esid past data. This comparison shutdown condi tion within will be made at least every 1 4 3, D 24 hours. equivalent full power month. p ( 5pe.mcaEws 3.3.] 4 4
. l l
l l l l l l l l l l l 1 Amendment No. M,140 88 j
. .- - - . . - - - . - . . - _- -__ . - ~ - - - ~ .--_- - - ,_~ .
i i ! VYNPS .p gggg: 3.3 t. 4.3 (Cont'd)
- 2. The control rod housing support restricts the outward movement of a control rod to less chan 3 inches in the extremely remote event of a housing failure. The amount of reactivity which could be j added by this small amount of rod withdrawal, which is less than
- a normal single withdrawal increment, will not contribute to any ,
damage of the primary coolant system. The design basis is given in subsection 3.5.2 of the PSAR, and the design evaluation'is given in subsection 3.5.4. This support is not required if the reactor coolant system is at acaospheric pressure since there would then be no driving force to rapidly eject a drive housing.
- 3. In the course of performing normal startup and shutdown t procedures, a pre-specified sequence for the withdrawal or insertion of control rods is followed. Control red dropout accidents which might lead to significant core damage, cannot
- occur if this sequence of rod withdrawals or insertions is followed. The Rod Worth Minimi
- er restricts withdrawals and insertions to those listed in the pre-specified sequence and provides an additional check that the reactor operator is following prescribed sequence. Although beginning a reactor 4
startup without having the RWM operable would entail unnecessary risk, continuing to withdraw rods if the RWM fails subsequently is acceptable if a second licensed operator verifies the - withdrawal sequence. - Continuing the startup increases core i power, reduces the rod worth and reduces the consequences of i dropping any rod. Withdrawal of rods for testing is permitted ' with the RWM inoperable, if the reactor is subcritical and all other rods are fully inserted. Above 201 power, the RWM is not I needed since even with a single error an operator cannot withdraw a rod v,ich sufficient worth, which if dropped, would result in i anyt;ning but minor consequences, j l 4. Refer to the Vermont Yankee Core Performance Analysis port. 1
! 5. The Source Range Monitor (SRM) system has no scram functions. It i
does provide the operator with a visual indication of neutron
; level. The consequences of reactivity accidents are a function of the initial neutron flux. The requirement of at least three should it occur, ' counts begins per at or secondaboveassures thatvalue the initial any transiong,of of 10~ rated power used in the analyses of transients from cold conditions. One operable
- sRM channel is adequate to monitor the approach to criticality, therefore, two operable SRM's are specified for added j conservatism.
ii 6. The Rod Block Monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power level operation. During reactor operation with certsin limiting control rod patterns, the withdrawal of a designated single control rod could result in one or more fuel rods During with MCPR use of less than the fuel cladding integrity safety li. nit. such patterns, it is judged that testing of the RBM system prior to withdrawal of such rods will provide added assurance that isproper withdrawal does not occur. It is the responsibility of the Nuclear Engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence of inoperable control rods. No. 36, 49, Gir, 7 0 90 a===A==at - . - , . _ _ . -
VntPS 1 n. 3.4 LIMITING CONDITIONS FOR 4.4 SURVETTlm REQUIRDGNPS OPERATION 3.4 a m roR STANDgY LIQUTD CoffrROL 4.4 RE W R STANDSY LICUID CONTRQQ SYSTEM EXIEti Aeolicabiliev_: Acolicabiliev: , Applies to the operating status Applies to the periodic testing requirement for the Reactor of the Reactor Standby Liquid Control System. Standby Liquid control Systems. Obiective: Obiective: To assure the availability of an To verify the operability of the independent reactivity control Standby Liquid control System. mechanism. So eifteaeion_: So*ei f teaeion_: A. Not nal operation A. Normal oeeration Except as specified in 3.4.8 The Standby Liquid control below, the Standby Liquid System shall be verified control System shall be operable by: operable during periods when fuel is in the reactor unless:
, 1. The reactor is in cold 1. Testing pumps and valves i shutdown in accordance with specification 4.6.E. A and minimum flow rate of 35 gym at 1275 peig shall be verified for each pump by recirculating domineraliced water to the test tank. .
Control rods are fully 2. Verifying the continuity 2. inserted and of the explosive charges Specification 3.3.A is at least monthly, met. . In addition, at least once during each operating cycle, the Standby Liquid Control System shall be verified operable by: Testing that the setting 3, of the pressure relief valves is between 1400 and 1490 psig. Initiating one of the standby liquid control 4, loops, ,exeluding the primer chamber and inlet fitting, and verifying that a flow path from a pump to the reactor 92
. W No. MG, 128
8 VYNPS r [ ' LIMITING CONDITIONS FOR 4.4 SURVEILLANCE REQUIRDE2TPS
- 3.4 OPERATICN vessel is available by I pumping domineralized water into the reactor vessel. Both loops shall be tested over the course of two operating cycles.
Testing the new trigger V assemblies by installing I one of the assemblies in the test block and firing it using the installed circuitry. Install the unfired
- assemblies, taken from the same batch as the fired one, into the explosion valves.
Recirculating the 6' borated solution. Coeration with Inocerable B. Ooeration with Inocerable ! B. Comoonents Comoonenes_ When a component becomes From and after the date that inoperable, its redundant a redundant component is component shall be or shall made or found to be have been demonstrated to be inoperable, reactor operable within 24 hours. operation is permissible during the succeeding seven days unless such component is sooner made operable. Liould Poison Tank - Boron C. Licuid Poison Tank - Boron C. concentration Concentration . At all times when the standby Liquid control system is required to be operable, the following conditions shall be met: The not volume versus 1. The solution volume in
- 1. the tank and temperature concentration of the in the tank and suction sodium pentaborate piping shall be checked solution in the standby at least daily.
liquid control tank shall meet the requirements of Figure 3.4.1. 93 MO- -
. . . . \
l
\
1 VYNPS 3.5 I,IMITING CCNDITICN FOR 4.5 SURVEIIMNCE REQUIRDG2rr OPERATION j 3. From and after the date 3. When the Alternate i that the Alternate cooling subsystem or ! Cooling Tower Subsystem both Station Service or both Station Service Water Subsystems are Water Subsystems are made or found to be made or found inoperable inaa-rable, the operable for any reason, reactor C ,.. W shall have operation is periaissible been or shall be only during the demonstrated to be succeeding seven davs operable within unless such * -- W 24 hours. ge made offerable, provided that during - such seven days all Cfv3yyg r(M(5) other active conneenants ::r'085/5T6*T(.T) of the other h J are operable. g
- 4. If the requirements of Specification 3.5.D cannot be met, an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition within 24 hours.
E. Mich Pressure Coolinc E. Hich Pressure Coolant l 2 In1ection ( H PCI ) System In7ection (HPCI) System Surveillance of HPCI System shall be performed as follows:
- 1. Except as specified in 1. Testine Specification 3.5.E.2, whenever irradiated fuel Item Frequencv !
is in the reactor vessel ,. and reactor pressure is Simulated Each re-greater than 150 psig Automatic fueling , t end prior to reactor Actuation outage startup from a cold Test coridition : Operability testing of
- a. The HPCI System the pump and valves shall be operable. shall be in accordance with
- b. The condensate Specification 4.6.E.
storage tank shall The HPCI System shall contain at least deliver at least 75,000 gallons of 4250 gpm at normal condensate water. reactor operating pressure when recirculating to the Condensate Storage Tank. Amendment No. M, H4, 128 105
-- . . - . - . - . . . . - . ~ . - - . . - . . . ~ - - . - . - . - . . - - . . ~ . .-. -.._, - - - . ~ -
V NPS 3.5 LIMITING CONDITION FOR 4.5 SURVEILLANCE REQUIREMENT OPERATION , 2. From and after the date 2. When the HPCI Subsystem that the HPCI Subsystem is made or found to be is made er found to be inoperable, the inoperable for any Automatic . reason, reactor Depressurization System
, operation is permissible shall have been or shall only during the . be demonstrated to be succeeding seven days operable within unless such subsystem is 24 hours.
sooner made operable, provided that during NOTE: Automatic such seven days all Depressurization active components of the System , Automatic operability shall Depressurization be demonstrated i Subsystems, the Core by performing a Spray Subsystems, the functional test . LPCI Subsystems, and the of the trip i RCIC System are ' system logic. operable. 2
- 3. If the requirements of Specification 3.5.E cannot be met, an l orderly shutdown shall 4
be initiated and the reactor pressure shall be reduced t 120 psig within 24 h s. F. Automatic Deore surization F. Automatie Depressurization System System I , Surveillance of the g6; Automatic Depressurization System shall be performed as follows: i
- 1. Except as specified in 1. Operability testing of Specification 3.5.F.2 the relief valves shall
. below, the entire be in accordance with 4 Automatic Specification 4.6.E. Depressurization Relief System shall be operable at any time the reactor pressure is above . 100 psig and irradiated fuel is in the reactor vessel.
- 2. From and after the date 2. When one relief valve of 4
that one of the four the Automatic Pressure
- relief valves of the Relief Subsystem is made i
Automatic or found to be Depressurization inoperable, the HPCI i subsystem are made or Subsystem shall have
< found to be inoperable been or shall be due to malfunction of demonstrated to be the electrical portion operable within j- of the valve when the 24 hours.
' Amendment No. e4, 444, 128 106
_.m -. , ,. __ .. _ _ _ _ _ , , _ . . ._ . . . , _ , _ , . . _ . . _ . . , . . ,_, . . . _ . _ _ , . _ . . . , . . . . _ _ . .
- . .- - - . . . - -.. - ~
t I i L VYNPS i 3.5 LIMITING CONDITION FOR 4.5 SURVEILLANCE REQUIRDiENT l OPERATION l reactor is pressurized above 100 psig with irradiated fuel in the reactor vessel, continued reactor operation is permissible only during the succeeding seven days unless such a valve is sooner made operable, provided that during such seven days both the remaining Automatic Relief System valves and the HPCI System are operable.
- 3. If the requirements of Specification 3.5.F cannot be amt, an orderly shutdown shall be initiated and the reactor pressure shall be reduced to 100 psig within 24 ho s.
G. Reactor Core Is ation G. Reactor Core Isolation Coolino system (RCIC) Coolino system (RCIC) Surveillance of the RCIC
.d.
System shall be performed as
~~~
follows:
- 1. Except as specified in 1. Testino Specification 3.5.G.2 Frecuency below, the RCIC System Item shall be operable l whenever the reactor Simulated Each re-pressure is greater tha,, automatic fueling 150 psig and irradiated actuation outage test fuel is in the reactor vessel. (testing valve
- 2. From and after the date operability) that the RCIC System is made or found to be Operability testing of inoperable for any the pump and valves reason, reactor shall be in accordance operation is permissible with specification 4.6.E.
only during the succeeding 7 days unless The RCIC System shall such system is sooner deliver at least 400 gpm made operable, provided at normal operating that'during such 7 days pressure hen all active components of recircul ing to the the HPCI System are Condensa e Storage Tank. operable. 107 Amendment No. +8, 24, +F4, 128 l.
l
. l VYNPS o -4. .,
l w' ) 3 '. 5 LIMITING CONDITION FOR 4.5 SURVEILLANCE REQUIREMENT l 1 OPERATION .
- l
- 3. If the requirements of l Specification 3.5.c l cannot be met, an 4 orderly shutdown shall { '
be initiated and the reactor pressure shall OY , 4 be reduced tq,120 psig I within 24 hours. H. Minimum core and containment H. Minimum core a containment Coolinc System Availability Coolino system XVAilability
- 1. During any period w 1. When one of t Il_:2..cyj; ;
one of thd[;;-..dL- diesel generators is diesel generaccra is made or found to be inoperable, continued inoperable, the reactor operation is remaining diesel permissible only during. generator shall have' ; the succeeding seven been or shall be 1 days, provided that all demonstrated to be ,
' operable within (
of thefE:ncontainment (fceelin;fano Pr:: ur: C:::'5 24 hours. l Cooling Subsystems connecting to the operable diesel 4pc generator shall be / c*"CJ5pedy I operable. If this 7, requirement cannot be met, an orderly shutdown shall be initiated and I the reactor shall be in the cold shutdown condition within 24 hours.
- 2. Any combination of inoperable components in '
the Core and Containment l Cooling Systems shall not defeat the capability of the remaining operable components to fulfill the core and containment cooling functions.
- 3. When irradiated fuel is in the reactor vessel and the reactor is in the cold shutdown condition, all Core and Containment Cooling Subsystems may be inoperable provided no work is permitted which has the potential for draining the reactor vessel.
l l Amendment No. B4, 114 108 l
. .~ . . . . .- -- - . . ~. _. . ._ - .- . - . - . _ . - -
L ~ - . - - .. . . ,,,, _, _ l r VYNPS n. i ' -) 4.4 SURVEII.I.ANCE REQUIRDENTS l 3.6 - LIMITING CONDITIONS FOR i CPERATION
- e. With the radiciodine i
concentration in - ' the reactor coolant
- greater than 1.1 microcuries/
l gram dose equivalent I-131, a sample of reactor coolant shall be taken every 4 hours and analyzed for radioactive iodines of I-131 through I-135, until the
- specific activity of the reactor coolant is restored below 1.1 microcuries/
gram dose equivalent I-131.
- 2. During startups and at
- 2. The reactor coolant steaming rates below
! water shall not exceed 100,000 pounds per hour, the following limits a sample of reactor j with steaming rates less coolant shall be taken . than 100,000 pounds per every four hours and hour except as specified analyzed for in specification conductivity and 3.6.3.3: chloride content. Conductivity c/cm chloride ion . ppm 1 With steaming rates OQol 3. For reactor startups the 3. a. greater than or maximum value for equal to y conductivity shall not 100,000 pounds per i exceed 10 unho/cm and hour, a reactor the = w i - "= value for coolant sample chloride ion shall be taken at concentration shall not least every exceed 0.1 ppa, in the 96 hours and when reactor coolant water the continuous 1 for the first 24 hours conductivity j after placing the monitors indicate > reactor in the power operating condition, abnormal conductivity (other than short-term 4 I spikes), and l analyzed for l conductivity and chloride ion content. 4 I i l I j I 118 L NA -m
VYNPS t 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIRDfENTS OPERATION
- b. When the continuous conductivity
- monitor is ;
inoperable, a I reactor coolant sample shall'be taken every four hours and analyzed for conductivity and chloride' ion content. I
- 4. Except as specified in Specification 3.6.B.3 above, the reactor !
coolant water shall not exceed the following limits with steaming rates greater than or l equal to 100,000 pounds l per hours. Conductivity 5 uhmo/cm chloride ion 0.5 ppm 8,$. A
- 5. If SpecificationC . : . t 4 is not met, an orderly i shutdown shall be l initiated and the reactor shall be in the cold shutdown condition within 24 hours.
C. Coolant Leakace C. Coolant Leakace Any time irradiated 1. Reactor coolant system l 1. a. leakage, for the purpose fuel is in the reactor vessel and of satisfying - reactor coolant Specification 3.6.C.1, temperature is shall be checked and above 212*F, logged once per shift, reactor coolant not to exceed 12 hours. leakage into the primary containment from unidentified sources shall not exceed 5 gpm. In addition, the total reactor coolant system leakage into . the primary containment shall not exceed 25 gpm. .
- b. While in the run mode, reactor coolant leakage into the primary containment from
+ unidentified' sources shall not 1 1 I l 119 ) Amendment No. 139 .
. - - ... ~ ~ - -_ .
i I VYNPS i (~ gggy,: 3.6 and 4.6 (Cont'd) The actual shift in RTNDT of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASD( E185 reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor l vessel in the core area. Since the neutron spectra at, the . irradiation samples and vessel inside radius are essentially I identical, the measured transition shif t for a sample can be . applied I with confidence to the adjacent section of the reactor vessel. satte11e Columbus Laboratory Report BCL-585-84-3, dated May 15, 1984, provides this information for the ten-year surveillance capsule. In i order to estimate the material properties at the 1/4 and 3/4 T positions in the vessel plate, the shift in RT g is determined in accordance with Regulatory cuide 1.99, Revision . The heatup and cooldown curves must be recalculated when the dRT determined from the surveillance capsule is different fremthecablateddRTNDT fCf
. the equivalent capsule radiation exposure.
The pressure-temperature limit lines, shown on Figure 3.6.1, for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10CFRSO for reactor criticality and for inservice leak and hydrostatic testing. The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are evided to assure compliance with the requirements of Appendix H to ePR-Part 50. Gw 1 u} CooLMTCAEA'Wh Qocpg A steady-state radiciodine concentration limit of 1.1 Ci of I-131 __ dose equivalent per gram of water in the Reactor Coolant System can y'g'g*l> be reached if the gross radioactivity in the gaseous ef fluents 1 near the limit, as set forth in SpecificationC .t . . k_72Ir there is a failure or prolonged shutdown of the cleanup domineraliser. In the event of a steam line rupture outside the drywell, the NRC staff calculations show the resultant radiological dose at the site boundary to be less than 30 Ram to the thyroid. This dose was calculated on the basis of the radiciodine concentration limit of 1.1 pCi of I-131 dose equivalent per gram of water, atmospheric dif fusion from an equivalent elevated release of 10 meters3 at the nearest site boundary (190 m) for a X/Q = 3.9 x 10~3 sec/m (Pasquill D and 0.33 m/sec equivalent), and a steam line isolation. valve closure time of fiva seconds with a steam / water mass release of 30,000 pounds. The iodine spike limit of four (4) microcuries of I-131 dose equivalent per gram of water provides an iodine peak or spike limit for the reactor coolant eo centration to assure that the radiological consequences of a@ r 1 n LOCA are within 10CFR Part 100 dose guidelines, y gg y The reactor coolant sample will be used to assure that the limit of
- Specification 3.6.a.1 is not exceeded. The radiciodine concentration would not be expected to change rapidly during steady-state operation over a period of 96 hours. In addition, the trend of the radioactive gaseous effluents, which is continuously monitored, is a good indicator of the trend of the radioiodine concentration in the reactor coolant. When a significant increase in radioactive gaseous i af fluents is indicated, as specified, an additional reactor coolant
' sample shall be taken and analyzed for radioactive iodine. 1
- - - - - . . J~'^ 'R hfDdSe 4 9 3 _
140 , ,_ _,,
l l VntPS
*b' ' '
ggggi 3.6 and 4.6 (Cont'd) impurities will also be within their ormal ranges. The reactor cooling samples will also be used a determine the chlorides. Therefore, the sampling frequency a considered adequate to detect long-term changes in the chlorid ion content. Isotopic analyses required by Specification G. ;.;. Omay be performed by a gamma scan and gross beta ani alpha determnation. i The conductivity of the feedwater is continuously monitored and alarm set points consistent with Regulatory requirements given in Regulatory Guide 1.56, " Maintenance of Water Purity in Boiling Water Reactors,a have been determined. The results from the conductivity monitors on the feedwater can be correlated with the results frem the conductivity monitors on the reactor coolant water to indicate demineralizer breakthrough and subsequent conductivity levels in the reactor vessel water. C. coolane taakaae > The 5 gym limit for unidentified leaks was established assuming such leakage was coming from the reactor cool : system. Tests have been conducted which demonstrate that a rela ionship exists between the size of a crack and the probability e the crack will propagate. These tests suggest that for leaka e omewhat greater than the limit specified for unidentified leakag .he probability is small that imperfections or cracks associated with such leakage would grow rapidly. I,eakage less than the limit specified can be detected within a few hours utilizing the available leakage detection systems. If the limit is exceeded and the origin cannot be determined in a i i reasonably short time the plant should be shutdown to allow further l investigation and corrective action. The 2 gym increase limit in any 24 hour period for unidentified leaks I was established as an additional requirement to the 5 gym limit by Generic Letter 88-01, "NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping.
- The removal capacity from .he drywell floor drain sump and the equivalent drain sump is 50 gpu each. Removal of 50 gym from either of these sumps can be accouplished with censiderable margin.
D. Safety and Relief Valves Safety analyses have shown that only three of the four relief valves are required to provide the reconsnanded pressure margin of 25 psi below the safety valve actuation settings as well as maintaining the fuel cladding integrity safety limit for the limiting anticipated'* overpressure transient. For the purposes of this limiting condition, a relief valve that is unable to actuate within tolerance of its set l pressure is considered to be as inoperable as a mechanically l malfunctioning valve. The setpoint tolerance value for as-lef t or refurbished valves is i specified in section III of the ASME Boiler and Pressure Vessel Code ; as tit of set pressure. However, the code allows a larger tolerance j value for the as-found condition if the supporting design analyses demonstrate that the applicable acceptance criteria are met. Safety analysis has been performed drich shows that with all safety and safety relief valves within 23% of the specified set pressures in Table 2.2.1 and with one inoperable safety relief valve, the reactor , coolant pressure safety limit of 1375 psig and the MCPR safety limit l
~
are not exceeded during the limiting overpressure transient. I l l R M, W. 4.W, W,160 142 _ _ _ _ . ,
. . . _ _ - . . - _. - . - . . ...__ .. - - _ . . . - . ~ .
l
.eb ,
l l 3.6 and 4.6 (cont'd) d.2 -3 l M: l E. Structural Inteority and operability Testinc , A pre-service inspection of the components listed in Tabl 4.0 ^ of I the FSAR was maand ed after site erection to assure freedoen defects greater than code allowances in addition, this. serves as a reference base for further inspections. Prior to operation, the reactor prima n syntesa was free of gross defects. In addition,' the facility has been designed such that gross defects should not occur l 4 0
+
I l l ' _ ., _ ,- ,+ +.......,.-, n a-w 142a
g V BASES: . (Cont'd) The interiors of the drywell and suppression chamber are painted to prevent rusting. The inspection of the paint during each major i refueling outage, approximately once per year, assures the paint is intact. Experience with this type of paint at fossil fueled generating stations indicates that the inspection interval is adequate. , 1 Because of the large volume and thermal capacity of the suppression
, pool, the volume and temperature normally changes very slowly and I monitorir>g these parameters daily is sufficient to establish any l temperathre trends. By requiring the suppression pool temperature to <
be continually monitored and frequently logged during periods of l significant heat addition, the temperature trends will be closely ! followed so that appropriate action can be taken. The requirement I for an external visual examination following any event where potentially high loadings could occur provides assurance that no significant da==ge was encountered. Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to be the points of highest stress. Visual inspection of the suppression chamber including water line regions each refueling outage is adequate to detect any changes I in the suppression chamber structures, i
\
1 l l s Amendment No, 143 1"*
l
. .. .... .. . l I
l l VYNPS l ggg 4.7 (Cont'd) WM l l l The -w4_-- al owable cose leak race at the peak accident pressure of l 44 psig (La) s 0.80 weight 1 per day. The =wi == allowable test leak race the retest pressure of 24 psig (Lc) has been conserv=H ely determined to be 0.59 weight percent per day. This valueC.115:jverified to be conservative by actual primary containment Aeak race measurements at both 44 psig and 24 psig upon completion of the contabmane structure. I ! As most leakage and deterioration of integrity is expected to occur l l through penetrations, especially those with resilient seals, a l periodic leak race test program of such penetration is conducted at l l the peak accident pressure of 44 psig to insure not only that the l leakage remains acceptably low but also that the sealing materials can withstand the accident pressure. I The Primary Containment Leak Race Testing Program is based on option B to 10CFR50, Appendix J, for development of leak race testing and surveillance schedules for reactor containment vessels. l Surveillance of the suppression Chamber-Reactor Building vacuum breakers consists of operability checks and leakage tests (conducted ! as part of the containment leak-tightness tests) . These vacuum I breakers are normally in the closed position and open only during l-tests or an accident condition. Operability testing is performed in conjunction with Specification 4.6.E. Inspections and calibrations ! are performed during the refueling outages; this frequency being based on equipment quality, experience, and engineering judgment. The ten (10) drywell-suppression vacuum relief valves are designed to
- open to the full open position (the position that curtain area is equivalent to valve bore) with a force equivalent to a 0.5 psi i
differential acting on the suppression chamber face of the valve disk. This opening specification assures that the design limit of 2.0 psid between the drywell and external environment is not exceeded. Once each refueling outage each valve is tested to assure that it will open fully in response to a force less than that l specified. Also it is inspected to assure that it closes freely and j operates properly. The containment design has been examined to establish the allowable2 bypass area between the drywell and suppression chamber as 0.12 f t , This is equivalent to one vacuum breaker open by three-eighths of an inch (3/8*) as measured at all points around the circumference of the disk or three-fourths of an inch (3/4") as measuredSince at thethese bottom of the disk when the top of the disk is on the seat. valves open in a manner that is purely neither mode, a conservative allowance of one-half inch (1/2*) has been selected as the maximum l permissible valve opening. Assuming that permissible valve opening could be evenly divided among all ten vacuum breakers at once, valve open position assumed to indication for an individual valve must be activated less chan fif ty-thousandths of an inch (0.050 * ) at all points along the seal surface of the disk. Valve closure within this limit may be determined by light indication from two independent position detection and indication systems. Either system provides a control room alarm for a nonseated valve. t 168
- Amendment No. 9 , He, 152
. _ _ _ .__ . . . _ _ _ _ _ . _ , . ~ _ . _ , _ -.
_~ _ - - - - . - VYtiPS TABLE 3.9.1 RADI0 ACTIVE LIOUID EFFLUEttr MONITORItC INSTRUMENTATION l Minimum Channels Operable Notes
- 1. Cross Radioactivity Monitors not Providing Automatic Termination of Release
- a. Liquid Radwaste Discharge Monitor l' 1,4,5
- b. prvice Water Discharge Monitor ,
1 2,4,5
- 2. Fo ate Measurement Devices
. Liquid Radwaste Discharge Flow 1* 3,4 tate Monitor g
- Du ing releases via this p thway.
t (nr-ae -as s/ea) gam- t+ - 5SQ (Rm-17-554) Amendment tio . 83 193
(Rm - n-I50 AloQ , VYNPS l 4 TABLE 3.9.2 GASEOUS EFFLUENT MONITORING INSTRUMENTATION l Minimum
\1.
Instrument team Jet Air Ejector (SJAE) Channels Operable Notes
- a. Noble Gas Activity Monitor 1 7, 8, 9
- 2. Augmented Off-Gas System
- a. Noble cas Activity Monitor Between 1 2, 5, 6, 7 the Charcoal Bed System and the Plant Stack (Providing Alarm and Automatic Termination of Release)
M g b. Flow Rate Monitor 1 1, 5, 6 e ydrogen Monitor 1 3, 5, 6
- 3. P t Stack i
l a. Noble cas Activity Monitor 1 5, 7, 10 l
/ b Iodine Sampler Cartridge 1 4, 5 l
- c. Particulate Sampler Filter 1 4, 5
- d. Sampler Flow Integrator 1 1, 5 g e. tack Flow Rate Monitor 1 1, 5 (fr-/$$-as[
Q r-Iv-/r4//s g l M - / f - / sG, 8M - / 7 -/F k kM24d- cG -a t a/ 4/d Ihn A4-0G -99aa 4/6) (pr - c a - s44a, p.r- o n - a ddy, f.r-o a -a 44e) { SAN - QG - 3/ Q y, Red *CG -3/3 8 Amendment No. H, 103 195
VYNPS TABLE 3.9.3 (Cont'd)
- RADIOLOGTCAL ENVIROt94EIRAL HONITORING PROGRAM Exposure Pathway Number of Sayle Sampling and collection Type and Frequency Locationa '
and/or Sample Frequency of Analysis
- 2. DIRECT RADIATIONb 40 routine monitoring Quarterly. Gamma does, at least once.
stations as follows: per quarter. 16 incident response Incident response TLDs in - stations one in each the outer monitoring {=::::!g.nlyector) locations, de-dose only watnan a . . of 0 to quarterly unless gaseous 4 kmS; release 140 was exceeded in period. Mgg4od;/ cal. 16 incident response stat i== i== in each (seeeeeneg4eeCaector) within a range of 2 to 8 km9s the balance of the stations to be placed in special interest areas and i control station areas. r a 1 <. d ,
- 1 Amendment No. 83 ggg i L _- - - - - - - - - - - - _ _ - - - - - -- -- - - - - - _ - _ _ _ - - - - - - - - - - - - - _ _ _ - - - - -
T VYNPS TABLE 3.9.4 I REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRO ENTAL SAMPLES "I Reporting Levels Airborne Particulate or Fish Hilk Vegetation Sediment Analysis Water (pCi/1) Gases (pCi/m3) (pCi/Kg, wet) (pci/1) (pci/Kg, wet) (pC1/Kg, dry) H-3 2 x 104(b) 1 x 10 3 3 x 10 4 Mn-54 4 x 10 2 1 x 10 4 Fe-59 Co-58 1 x 10 3 3 x 10 4 Co-60 3 x 102 1 x 10 4 3 x 103(c) 4 l Zn-65 3 x 10 2 2 x 10 Zr-Nb-95 4 x 10 2 I-131 0.9 3 1 x 102 3 Cs-134 30 10 1 x 10 60 1 x 103 3 Cs-137 50 20 2 x 10 70 2 x 103 Ba-La-140 2 x 10 2 3 x 102 (a) Reporting levels may be averaged over a calendar quarter. When more than one of the radionuclides in Table 3.9.4 are detected in the sampling medium, the unique reporting requirements are not exercised if the following condition holds: concentration (1) , concentration (2) ,*** . . m 4 f,[ reporting level (1) reporting level (2) When radionuclides other than those in Table 3.9.4 are detected and are the result of plant affluents, the potential annual dose to a member of the public must be less than or equal to the calendar year limits of - Specifications 3.8.B, 3.8.E and 3.8.P. (b) Reporting level for drinking water pathways. For nondrinking water pathways, a value of 3 x 104 pCi/l may I be used. [
\ (c) Reporting level for individual grab samples taken at North Storm Drain Outfall only. !
Amendment No. 83, 103 202
. n o.a.e .u _,_ - __
. a.au .. . = . ; a. _
l i
- v. ,
. VYNPs .. . b l e TABLE 4 9.2 NOTATION l j (1) The Instrument Functional Test shall also demonstrace that automatic isolation of this pathway and the control Room alarm annunciation occurs if any of the following condicions exists: (a) Instrumene M measured levels above che alarm setpoint. (b) Circuit failure. , XA/b /c473rg'] (c) Instrument indicates a downscale failure. (d) Instrument controls not set in operate mode. (2) The Instrument Functional Test shall also demonstrate that control Room alarm annunciation occurs when any of the following conditions exists (a) Instrument indicates measured levels above the alarm seepoint. (b) Circuit failure. (c) Inst. w e indicates a downscale failure. (d) Instru:nent controls are not set in operate mode. (3) The Instrument calibracion for radioactivity measurement instrumentation shall include the use of a known (traceable to Nacional Institute for standards and Technology) radioactive source positioned in a reproducible geomecry with respect to the sensor. These standards should permit calibrating the system over its normal operating range of race capabilities. - (4) The Inst.wt Calibration shall include the use of standard gas samples (high range and low range) containing suitable concentrations, hydrogen balance nitrogen, for the detection range of interest per specification 3.8.J.1. 4 G en S
.a=.a,6-n e No . g, 151 206
. _ _ _ . - _ _ _ . _ . . _ _ _ _ _ _ . - . . - . _ . . _ . _ _ . _ _ . _.m _ .. . . _ _
l VYNPS (' Ms 3.9 RADIQACTIVE EFFWENT MON!"CRING SYSTEMS - A. Licuid Ef fluent Instrumentation l The radioactive liquid effluent instrumentation is provided to l monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm seepoints for these instruments are to ! ensure that the alarm will occur prior to exceeding 10 times the concentration limits of Appendix 3 to 10CFR20.1001-20.2401. Table 2, Column 2, values. l Automatic isolation function is not provided on the liquid radwaste discharge line due to the infrequent nature of batch, discrete volume, liquid discharges (on the order of once per year or less), and the administrative controls provided to ensure that conservative discharge flow races / dilution flows are set such that the probability l 'of exceeding the above concentration limits are low, and the potencial off-site dose consequences are also low. l
- 3. Gaseous Effluent Instrumentation The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous offluents during actual or potencial releases of gaseous effluents. The alarm / trip setpoints for these instruments are provided to ensure that tha alarm / trip will occur prior to l exceeding design bases dose rates identified in 3.8.E.1. This instrumentation also includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the waste gas holdup system.
C. Radiolecical Envir a.. ental Monitorine Procram l The radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures of member (s) of the public ' resulting from the station operation. This monitoring program e implements sectica IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the radiological offluent monitoring program by verifying . that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure. pathways. Ten years of plant operation, including the years prior to the implementation of the Augmanced off-Gas System, have amply l demonstrated via routine affluent and environmental reports that I plant effluent measurements and.modeling of environmental pathways ! are adequately conservative. In all cases, environmental sample results have been two to three orders of magnitude less- than expected by the model employed, thereby representing small percentages of the ALARA and environmental reporting levels. This radiological environmental monitoring program has therefore been sienificantly I modified as provided for by Regulatory Guide t. 2 M?' --4]4.1 - ! (C.2.b), Revision 1, April 1975. Specifically, the air particulate i and radioiodine air sampling periods have been increased to i semimonthly. based on planc effluent and environmental air sampling data for the previous con years of operation. An I-131 release race l trigger value of f x 10-1 uCi/sec from the plant stack will require i- that air sample collection be increased to weekly. The amad==at No. n,151 209
I
. VTEFS i
~~
l - 3.12 I,IMITING CONDITICNS FCR -
- t 12- StfRTEM*aer*-ItEQtrtstEbtERTS-- --
l l i OPE 3tATION
- g. Ext-d =d Cors Maintenance I .' Breended Core Maintenance l
one or more control rods may Prior.co control rod be withdrawn or removed freut withdrawal for extended core 4a*==a-a, that control. the reactor core proMded
- the following conditions are control cell shall be i satisfied ed to contain no fuer
******LL**-
rod's
- 1. The reactor mode switch - '1. . This sury.411.nre shall be locked in the requirement is the same as that given in
- Refuel' position. The refueling incariock Specification 4.12.A.
which prevents more than one control rod from being wittuirawn may be bypassed on a withdrawn control rod after the l fuel assemblies in the q cell conta4" %g ; (controlled by) that { control rod have been removed from the reactor i core. All other refueling interlocks shall be operable. l
- 2. SRMs shall be operable in the core quadrant 2. This surveillance where fuel or control requirement is the same !
rods are being moved,
- as that given in ,
and in an adjacent Specification 4.12.B. l quadrant. The l requirements for an SRM j co be considered i operable are given in - - Specification 3.12.B. . . 4
- 3. If the spiral -
unload / reload method of core alteration is to be used, the following , conditions shall be mets
- a. Prior to spiral unload and reload, the SRMs shall be proven operable as stated in Specification 3.12.31 and 3.12.82. Bowever, during spiral unloading, the count race may drop below 3 cps.
I 233 w + rt No, M. M, 43.148
~
VYNPS m 3.3 3.12 t. 4.12 REWEL Dic A. During refueling operations, the reactivity pot cial of the core is being altered. It is necessary to require cor in interlocks and restrict certain refueling procedures such t there is assurance that inadvertent criticality does not occur. To minimize the possibility of loading fuel into a cell containing no control rod, it is required that all ce rol rods are fully l inserted when fuel is being loaded into e reactor core. This requirement assures that durfng refueling the refueling interlocks, as. designed, will prevent inadvertent c i icality. The core I reactivity limitation of Specification limits the core alterations to assure that the resulting core loading can be controlled with the Reactivity Control System and interlocks at any i
, time during shutdown or the following operating cycle.
The addition of large amounts of reactivity to the core is prevented by operating procedures, which are in turn backed up by refueling interlocks on rod withdrawal and movement of the refueling platform. When the mode switch is in the ' Refuel
- position, interlocks prevent the refueling platform from being moved over the core if a control rod is withdrawn and fuel is on a hoist. l Likewise, if the refueling platform is over the core with fuel on a i hoist, control rod motion is blocked by the interlocks. With the l mode switch in the refuel position, only one control red can be withdrawn. .
1 B. The SRMs are provided to monitor th'e core during periods of station shutdown and to guide the operator during refueling operations and station startup. Requiring two operable SRMs in or adjacent to any core quadrant where fuel or control rods are being moved assures adequate acnie.oring of that quadrant during such alterations. The requirement of 3 counts per second provides assurance that neutron flux is being monitored. Under the special condition of complete spiral core unloading, it is expected that the count rate of the SRMs will drop below 3 cps before all the fuel is unloaded. Since
.there will be no reactivity additions, a lower number of counts' will not present a hatard. When all of the fuel has been removed to the spent fuel storage pool, the SRMs will no longer be required.
Requiring the SRMs to be operational prior to fuel removal assures that the SRMs are operable and can be relied on even when the count rate may go below 3 cps. 1 Prior to spiral reload, two diagonally adjacent fuel assemblies, which have previously accumulated exposure in the reactor, will be loaded into their designated core positions next to each of the 4 SRMs to obtain the required 3 eps. Exposed fuel continuously produces neutrons by spontaneous fission of certain plutonium isotopes, photo fission, and photo disintegration of deuterium in the moderator. This neutron production is normally great enough to meet the 3 cps minimum SRM requirement, thereby providing a means by which SRM response may be demonstrated before the spiral reload begins. During the spiral reload, the fuel will be loaded in the reverse sequence that it was unloaded with the exception of the initial eight (8) fuel assemblies which are loaded next to the SRMs to provide a means of SRM response, l . l l l l Amendment No. W. M, 77 237 j
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l l l l l VYNPS' . I 3.13 LIMITING CONDITICNS FCR 4.13 SURVEILLANCE REQU'.REMENTS 1 l OPERATION. l l 1) The batteries, cell plates and battery , racks show no ! visual indication of physical damage or abnormal deterioration,
. and
- 2) The battery-co-battery and terminal
- connections are clean, tight, free of corrosion and coated with anti-corrosion material.
C. Fire Mose Stations C. Fire Hose Stations
- 1. rvrage as specified in 1. Each fire hose station 3 . ;; . c ._ijbe low, all shall be verified to be nos. scations inside operable:
the Reactor Building, Turbine Building, and a. At least monthly by those inside the visual inspection 3 , 13 . C gg Aaministracion muilding oc th. seation to which provided coverage assure all of the Control Room equipment is Building shall be available, operable whenever . equipment in the areas b. At least once each protected by the fire 18 months by - home stations is removing the hose required to be for inspection and operable. replacing degraded coupling gaskets
- 2. With one or more of the and raracking.
fire hose stations specified in 3.13.C.1 c. At least once each above inoperable, route year by an additional hydro-statically equivalent capacity testing each fire hose to the outside hose at unprotected area (s) 250 lbs. from an operable hose station within one d. At least once per hour. 3 years by hydro-statically testing inside hose at 150 lbs. G 244 Amendment No . +4, 67
'- _ me _
. . VYNPS ,
- f. 3.13 1.INITING CONDITICHS FOR 4.13 SURVER. LANCE REQUIRIDGNrs OPERATION At least once per 3 years by C. Performing an air flow test through the Recirculation M.C. Set foaan header and '
verifying each foam nozzle is unobstructed. 249 Annendment No. 67
- - -.J , . . . , , _a . --m_--'
. . - . _ . . _ - . _ . = - ,
l VYNPS . f.. TABLE 3.13.A.1 FZRE DETECTION SENSCRS Mini === No. of Sensora I . Required to Be Operable i sensor Location g h .g ! 1. Cable Spreading Room f. Station Battery Room - - 23
)
Switchaear Room @E45 7D - - _2 . - g e s. ~ ~ o n m , w a,,.n, usemeA veneracor m.~- m iv 2 l l g FC Diesel Generator Room (B) - - 2 Intake Structure (Service Water) 1 1 1
- hh Recirc Motor Generator Set Area 3 -
8 14 l g@ Control Room Zone 1 (Control Room Ceiling)
- - 18 lhM Control Room Zone 2 (Control Room Panels) - -
25 j h6 Control Room Zone 3 (Control Rocal Panels) ' - - 10
* @ Control Room Zone 4 (Control Room Panels) 2 j@M Control Room Zone 5 (Exhaust t. Supply - -
Ducts) ih* @ Rx Bldg. Corner Rm NW 232 - - 1 1 7 h* @ Rx Bldg. Corner Rm NW 213 (RCIC) - - !he @ Rx Bldg. Corner Rm NE 232 - - 1 hQ Rx Bldg. Corner Rm NE 213 - - 1 Rx Bldg. Corner Rm SE 232 1 h - - Ih> Q Rx Bldg. Corner Rm SE 213 - - I hM Rx Bldg. Corner Rm SW 232 - - 1-g --Q HPCT " :
- - B 16 12
[pe@ Torus area 7 h - - Q Rx Bldg. Cable Penetration Area 13 h* h Refuel Floor
- 1* 1
' hup h Diesel Oil Day Tank Room (A) 1* 1 [ _ @ Diesel 011 Day Tank Room (B) 3 e M Turbine Leading Bay (vehicles) r. l
- NOTE: The Diesel Day Tank Rooms require only one detector operable 2 (1 glame gy,1 smoke).
l i Y Amendment No. +ir, 67 250 4 m ,.
l l l l l ATTACHMENT F l REVISED TECHNICAL SPECIFICATIONS PAGES TECHNICAL SPECIFICATIONS PROPOSED CHANGE NO. 205 VERMONT YANKEE NUCLEAR POWER STATION l LICENSE NO. DPR-28, DOCKET NO. 50-271 l (BVY 98-118) l 1 i
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.VYNPS~
TABLE OF CONTENTS (Continued) LIMITING SAFETY Page No. SYSTEM SETTING , SAFETY LIMITS 1.1 FUEL CLADDING INTEGRITY....................... 6 ... 2.1 1.2' REACTOR COOLANT SYSTEM........................ 18 ... 2.2 LIMITING CONDITIONS OF OPERATION Page No. SURVEILLANCE-t 3.1- REACTOR PROTECTION SYSTEM..................... 20 ... 4.1 J , . BASES ~ '29 3.2' PROTECTIVE INSTRUMENT SYSTEMS . . . . . . . . . . . . . . . . . - 34 ... 4.2 A. Emergency Core Cooling System............. 34 ... A B. Primary Containment Isolation............. 34 ... B C. Reactor. Building Ventilation Isolation l'i and Standby Gas Treatment System Initiation................................ 34 ... C D. Air Ejector Off-Gas System Isolation...... 35; ... -D E. Control Rod Block Actuation............... 35 ... E F. . Mechanical Vacuum Pump Isolation.......... 35 ... F
'G. Post-Accident Instrumentation............. 35 ... G H. Drywell to Torus AP Instrumentation........ 36 ... H l I. Recirculation Pump Trip.
Instrumentation........................... 36 ... I l J. (Deleted) ................................ 36 ... J K. Degraded Grid Protective System .......... 36 ... K L.: ~ Reactor Core Isolation Cooling System _ Actuation..............'................... 37- ... L ; l BASES 75 3.3 CONTROL ROD SYSTEM............................ 81 ... 4.3
. A. Reactivity Limitations.................... 81 ... A B. Control Rods.............................. 82 ... B C.. Scram Insertion Times..................... 85 ... C D. Control Rod Accumulators.................. 87 ... D E. Reactivity Anomalies...................... 88 ... E BASES 89 3.4 -REACTOR STANDBY LIQUID CONTROL SYSTEM......... 92 ... 4.4 A. Normal Operation.......................... 92 ... A
-B. Operation with Inoperable' Components...... .
93 ... B
'C. Liquid Poison Tank - Boron Concentration............................. 93 ... C 4
BASES 97. Amendment.No. 64, 96 l l 1
. s . ,- . . ~ . , . . . . , - - , . - . . - . , , - - . . , - , ,I
VYNPS BASES: 2.1 (Cont'd) metal-water reaction to less than 1%, to assure that core geometry remains intact. The design of the ECCS components to meet the above criteria was dependent on three previously set parameters: the maximum break size, the low water level scram setpoint, and the ECCS initiation setpoint. To lower the ECCS initiation setpoint would now prevent the ECCS components from meeting their design criteria. To raise the ECCS initiation setpoint would be in a safe direction, but it would reduce the margin established to prevent actuation I of the ECCS during normal operation or during normally expected transients. E. Turbine Stop Valve Closure Scram Trip Setting The turbine stop valve closure scram trip anticipates the pressure, neutron flux and heat flux increase that could result from rapid closure of the turbine stop valves. With a scram trip setting of <10% of valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR remains above the fuel cladding integrity safety limit even during the worst case transient that assumes the turbine bypass is closed. This , scram is bypassed when turbine steam flow is below 30% of rated, as measured I by turbine first stage pressure. F. Turbine Control Valve Fast Closure Scram The control valve fast closure scram is provided to limit the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection coincident with failure of the bypass system. This transient is less severe than the turbine stop valve closure with failure of the bypass valves and therefore adequate margin exists. G. Main Steam Line Isolation Valve Closure Scram The isolation valve closure scram anticipates the pressure and flux transients which occur during normal or inadvertent isolation valve closure. With the scram setpoint at 10% of valve closure, there is no increase in neutron flux. H. Reactor Coolant Low Pressure Initiation of Main Steam Isolation Valve closure The low pressure isolation of the main steam lines at 800 psig is provided to give protection against rapid reactor depressurization and the resulting rapid cooldown of the vessel. Advantage is taken of the scram feature which occurs when the main steam line isolation valves are closed, to provide the reactor shutdown so that high power operation at low reactor pressure does not occu.r. Operation of the reactor at pressures lower than 800 psig requires that the reactor mode switch be in the startup position where protection of the fuel cladding integrity safety limit is provided by the IRM high neutron flux scram. Thus, the combination of main steam line low pressure isolation and l isolation valve closure scram assures the availability of neutron scram protection over the entire range of applicability of the fuel cladding integrity safety limit. Amendment No. 44, B&, B4 17
VYNPS 3.1 LIMITING CONDITIONS FOR 4.1 SURVEILLANCE REQUIREMENTS OPERATION 3.1 REACTOR PROTECTION SYSTEM 4.1 REACTOR PROTECTION SYSTEM Applicability: Applicability: l Applies to the operability of ~ Applies to the surveillance of plant instrumentation and control the plant instrumentation and systenu required for reactor control systems required for safety. reactor safety. Objective Objective l To specify the limits imposed on To specify the type and frequency plant operation by those of surveillance to be applied to instrument and control systems those instrument and control required for reactor safety. systems. required for reactor safety. Specification: Specification: A. Plant' operation at any power A. Instrumentation systems level shall be permitted in shall be functionally ; accordance with Table 3.1.1. tested and calibrated as I The system response time from indicated in Tables 4.1.1 l the opening of the sensor and 4.1.2, respectively-contact up to and including the opening of the scram solenoid relay shall not ; exceed 50 milliseconds. ' B. During operation with the B. Once a day during reactor ratio of MFLPD to FRP greater power operation the than 1.0 either maximum fraction of limiting power density and
- a. The APRM System gains fraction of rated power shall be adjusted by the shall be determined and ratios given in Technical the APRM system gains Specifications 2.1.A.1 shall be adjusted by the and 2.1.B or ratios given in Technical Specifications 2.1.A.1.a
- b. The power distribution and 2.1.B.
shall be changed to reduce the ratio of MFLPD to FRP. l l 1 I Amendment No. n 20
'VYNPS TABLE 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT REQUIREMENTS Required '
Modes in Which Functions Must conditions be Operating Minimum Number When Mininnrm , Operating Conditions For ! Instrument Operation Channels Per Are Not Trip Function Trip Settings Refuel (1) Startup (12) Run Trip System (2) ~ Satisfied (3)
- 1. Mode Switch in X X X 1- A l Shutdown (5A-S1) ;
- 2. Manual Scram X X X 1 A (5A-S3A/B)
- 3. IRM (7-41 ( A-F) ) !
High Flux ~<120/125 X X X(11) 2 A- ' I INOP X X X(ll) 2 -A
- 4. APRM (APRM A-F)
High Flux <0.66 (W-AW)+54% X 2 A or B (flow bias) T4) High Flux (reduced)
<15% X X 2 A i X 2 (5)
INOP A or B Downscale >2/125 X 2 A or B ,
- 5. High Reactor $1055 psig X X X 2 A Pressure !
l (PT-2-3-55 (A-D) (M) ) : t
- 6. High Drywell $2.5 psig X X X 2 A ,
Pressure l l (PT-5-12 (A-D) (M) ) {
- 7. Reactor Low (6) . >127.0 inches X X X 2 A Water Level ,.
(LT-2-3-57A/B(M)) ! (LT-2-3-58A/B(M)) [
- 8. Scram Discharge $21 gallons X X X 2 A f Volume High Level (per volume)-
I (LT-3-231 (A-H) (M) ) i Amendment No. G-1, 44, 64, 66, %, M, M, GG, G4 21 i; _ . . _ . _ . _ _ _ . _ _ . . _ _ _ _ _ _ _.____.._m__________._ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ = _ . _ _ . _ _ . _ _ _ . _ _ . _ _ _ _ _ _ _ _ . _ _ _ . . _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _
7 i a r VYNPS -i TABLE 3.1.1 , (Cont'd) i
' REACTOR' PROTECTION SYSTEM (SCRAM) INSTRUMENT REQUIREMENTS i
' Required l i
Conditions ,
.l.I Minimum Number- When Minimum f
! Modes in Which Functions Must ' Operating Conditions For '. Trip Settings be Operating Instrument Operation ; And Allowable Channels Per Are Not ! Trip Function Deviations Refuel (1) Startup Run . Trip System (2) Satisfied (3) [ , 9. Main steamline high 3x normal X X -X 2 A or C I radiation (7) background at (RM-17-251(A-D)) rated power (8)
- 10. Main steamline $10%. valve X 4- A or C isolation valve closure e closure f (POS-2-80A-A1,B1 '
POS-2-86A-A1,B1 ' < POS-2-80B-A1,B2 [ i POS-2-86B-A1,B2 ! POS-2-80C-A2,B1 t POS-2-86C-A2,B1 3 POS-2-80D-A2,B2 t POS-2-86D-A2,B2) !
- 11. Turbine control (9) (10) .X 2 A or D L valve fast closure i l (PS-(37-40)) j
- 12. Turbine-stop valve $10% valve (10) X 2 A or D [
closure closure l (SVOS-5-(1-4)) -l L i 4 .i ?
=
l Amendment.No. i L I Y
VYNPS TABLE 3.1.1 NOTES i
- 1. When the reactor is suberitical and the reactor water temperature is less than 212*F, only the following trip functions need to be operable: '
a) mode switch in shutdown b) manual scram l c) high flux IRM or high flux SRM in coincidence d) scram discharge volume high water level
- 2. Whenever an instrument system is found to be inoperable, the instrument j system output relay shall be tripped immediately. Except for MSIV and j Turbine Stop Valve Position, this action shall result in tripping the trip i system.
- 3. When the requirements in the column " Minimum Number of Operating Instrument Channels Per Trip System" cannot be met for one system, that system shall be tripped. If the requirements cannot be met for both trip systems, the appropriate actions listed below shall be taken:
l a) Initiate insertion of operable rods and complete insertion of all I operable rods within four hours. b) Reduce power level to IRM range and place mode switch in the ;
"Startup/ Hot Standby" position within eight hours. l l c) Reduce turbine load and close main steam line isolation valves within 8 '
hours. d) Reduce reactor power to less than 30% of rated within 8 hours.
- 4. "W" is percent rated two loop drive flow where 100% rated drive f_ a is that flow equivalent to 48 x 10' lbs/hr core flow. AW is the difference between the two loop and single loop drive flow at the same core flow.
This difference must be accounted for during single loop operation. AW = 0 for two recirculation loop operation.
- 5. To be considered operable an APRM must have at least 2 LPRM inputs per level and at least a total of 13 LPRM inputs, except that channels A, C, D, and F may lose all LPRM inputs from the companion APRM Cabinet plus one additional LPRM input and still be considered operable.
- 6. The top of the enriched fuel has been designated as 0 inches and provides common reference level for all vessel water level instrumentation.
- 7. Channel shared by the Reactor Protection and Primary Containment Isolation Systems.
- 8. An alarm setting of 1.5 times normal background at rated power shall be established to alert the operator to abnormal radiation levels in primary coolant.
Amendment No. 34, GG, 64, 74, 94 23
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_ . - - ~ - - _ . . . - .
.p VYNPS i
TABLE 4.1.1 ; SCRAM INSTRUMENTATION AND LOGIC SYSTEMS FUNCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTRUMENTATION, LOGIC SYSTEMS AND CONTROL CIRCUITS Instrument Channel Group"3 Functional TestI7) Minimum FrequencyNI Mode Switch in Shutdown A Place Mode Switch in Shutdown Each Refueling Outage
- Manual Scram A Trip Channel and Alarm Every 3 Months IRM i
- High Flux C Trip Channel and Alarm (5) Before Each Startup & Weekly !
! During Refueling (6) j , Inoperative C Trip Channel and Alarm Before Each Startup & Weekly During Refueling (6) ; ! + APRM I High Flux .B Trip Output Relays (5) Once Each Week [ High Flux (Reduced) B Trip Output . Relays (5) Before Each Startup & Weekly During Refueling (6) l' Inoperative B Trip' Output Relays Once Each Week Downscale B Trip Output Relays (51 Once'Each Week Flow Bias B (1)
- Trip Output Relays (5) l f
{ s High Reactor Pressure B Trip Channel and Alarm (s) til High Drywell Pressure B Trip Channel and Alarm (5) til -l Low Reactor Water Level (2)(83 B Trip Channel and Alarm (s) gt)
- High Water Level in Scram Discharge B Trip Channel and Alarm (53 til Volume High Main Steam Line Radiation (2) B Trip Channel and Alarm (51 Once Each Week t
! l Main Steam Line Iso. Valve Closure A Trip Channel and Alarm (11 . Turbine Con. Valve Fast Closure A Trip Channel and Alarm til
. Turbine Stop Valve closure A Trip Channel and Alarm (1) r
- l Scram Test Switch ( 5A-S2 ( A-D) ) A Trip Channel and Alarm Each Refueling Outage First Stage Turbine Pressure - A Trip Channel and Alarm Every.6 Months Permissive (PS-5-14 (A-D) )
Amendment No. 44, 2-1, . 68, % 25
VYNPS TABLE 4.1.2 SCRnit INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Instrument Channel GroupIII Calibration StandardI4) Minimum Frequencv(2) High Flux APRM Output Signal B Heat Balance Once Every 7 Days Output Signal (Reduced) B Heat Balance Once Every 7 Days Flow Bias B Standard Pressure and Voltage Refueling Outage Source LPRM (LPRM ND-2-1-104(80)) B(5) Using TIP System Every 1000 Equivalent Full Power Hours High Reactor Pressure B Standard Pressure Source Once/ Operating Cycle Turbine Control Valve Fast Closure- A Standard Pressure Source Every 3 Months High Drywell Pressure B Standard Pressure Source Once/ Operating Cycle High Water Level in Scram Discharge B Water Level Once/ Operating Cycle Volume , Low Reactor Water Level B Standard Pressure Source Once/ Operating Cycle Turbine Stop Valve Closure A (6) Refueling Outage High Main Steam Line Radiation B Appropriate Radiation Refueling Outage Source (3) First Stage Turbine Pressure A Pressure Source Every 6 Months and After Permissive (PS-5-14 (A-D) ) Refueling Main Steam Line Isolation valve A (6) Refueling Outage closure Amendment No. -14, M, M, GB, M, M 27 I
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I I VYNPS j BASES: I 3.1 Reactor Protection System The reactor protection system automatically initiates a reactor scram to: l 1. preserve the integrity of the fuel barrier;
- 2. preserve the integrity of the primary system barrier; and
- 3. minitedze the energy which must be absorbed, and prevent criticality following a loss of coolant accident.
This specification provides the limiting conditions for operation necessary to preserve the ability of the system to tolerate single failures and still perform its intended function even during periods when instrument channels may be out of service because of maintenance, testing, or calibration. l The reactor protection system is of the dual channel type. The system is made up of two independent logic channels, each having three subsystems of tripping devices. One of the three subsystems has inputs from the manual scram push buttons and the reactor mode switch. Each of the two remaining l subsystems has an input from at least one independent sensor monitoring each of the critical parameters. The outputs of these subsystems are combined in a 1 out of 2 logic; i.e., an input signal on either one or both I of the subsystems will cause a trip system trip. The outputs of the trip systems are arranged so that a trip on both logic channels is required to produce a reactor scram. The required conditions when the minimum instrument logic conditions are not met are chosen so as to bring station operation promptly to such a condition that the particular protection instrument is not required; or the station is placed in the protection or safe condition that the instrument initiates. This is accomplished in a normal manner without subjecting the plant to abnormal operating conditions. When the minimum requirements for the number of operable or operating trip system and instrumentation channels are satisfied, the effectiveness of the protection system is preserved; i.e., the system can tolerate a single failure and still perform its intended function of scramming the reactor. Three APRM instrument channels are provided for each protection trip system to provide for high neutron flux protection. APRM's A and E operate contacts in a trip subsystem, and APRM's C an E operate contacts in the other trip subsystem. APRM's B, D, and F are arranged sindlarly in the other protection trip system. Each protection trip system has one more APRM than is necessary to meet the ndnimum number required. This allows the bypassing of one APRM per protection trip system for maintenance, testing, or calibration without changing the minimum number of channels required for inputs to each trip system. Additional IRM channels have also been provided to allow bypassing of one such channel. IM1 assignment to the bypass switches is described on FSAR Figure 7.5-9 and in FSAR Section 7.5.5.4. The bases for the scram settings for the IRM, APRM, high reactor pressure, reactor low water level, turbine control valve fast closure, and turbine stop valve closure are discussed in Specification 2.1. Instrumentation is provided to detect a loss-of-coolant accident and l initiate the core standby cooling equipment. This instrumentation is a backup to the water level instrumentation which is discussed in Specification 3.2. Amendment No. 31- 29
VYNPS BASES: 4.1 REACTOR PROTECTION SYSTEM A. The scram sensor channels listed in Tables 4.1.1 and 4.1.2 are divided into three groupst A, B and C. Sensors that make up Group A are the on-off type and will be tested and calibrated at the indicated intervals. Initially the tests are more frequent than Yankee experience indicates necessary. However, by testing more frequently, the confidence level with this instrumentation will increase and testing will provide data to justify extending the test intervals as experience is accrued. Group B devices utilize an analog sensor followed by an amplifier and bistable trip circuit. This type of equipment incorporates control room mounted indicators and annunciator alarms. A failure in the sensor or amplifier may be detected by an alarm or by an operator who observes that one indicator does not track the others in similar channels. The bistable trip circuit failures are detected by the periodic testing. Group C devices are active only during a given portion of the operating cycle. For example, the IRM is active during start-up and inactive during full-power operation. Testing of these instruments is only meaningful within a reasonable period prior to their use. l l B. The ratio of MFLPD to FRP shall be checked once per day to determine if l the APRM gains require adjustment. Because few control rod movements or power changes occur, checking these parameters daily is adequate. i I f i l l 1 Amendment No. M, 41 33
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I I VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS l OPERATION ' l 3.2 PROTECTIVE INSTRUMENT SYSTEMS 4 PROTECTIVE INSTRUMENT SYSTEMS Applicability: Applicability: l Applies to the operational status Applies to the surveillance of the plant instrumentation requirements of the systems which initiate and instrumentation systems which I control a protective function. initiate and control a protective ! function. Objective: Objective: , To assure the operability of To verify the operability of protective instrumentation protective instrumentation systems, systems. Specification: Specification: A. Emergency Core Cooling System A. Emergency Core Cooling System When the system (s) it initiates Instrumentation and logic or controls is required in systems shall be functionally accordance with specification tested and calibrated as 3.5, the instrumentation which indicated in Table 4.2.1. l initiates the emergency core . cooling system (s)shall be operable in accordance with Table 3.2.1. B. Primary Containment Isolation B. Primary Containment Isolation When primary containment Instrumentation and logic integrity is required, in systems shall be functionally accordance with tested and calibrated as Specification 3.7, the indicated in Table 4.2.2. instrumentation that initiates primary containment isolation shall be operable in accordance with Table 3.2.2. C. Reactor Building Ventilation C. Reactor Building Ventilation Isolation and Standby Gas Isolation and Standby Gas Treatment System Initiation Treatment System Initiation The instrumentation that initiates Instrumentation and logic the isolation of the reactor systems shall be functionally building ventilation system and tested and calibrated as the actuation of the standby gas indicated in Table 4.2.3. treatment system shall be operable in accordance with Table 3.2.3. l l Amendment No. 34
. .-m. _ _ . - _ _ . _ ...- ._ _ _ _ - . _ _ . . _ _ _ _ - _ _ . _ . _ _ _ _ _ . . - _ .... . _ . . _ .
VYNPS r 4 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION 1
, D. Off-Gas System Isolation D. Off-Gas System Isolation During reactor power Instrumentation and logic operation, the systems shall be functionally.
instrumentation that tested and calibrated as
. initiates isolation of the indicated in Table 4.2.4.
off-gas _ system'shall be operable in accordance with Table 3.2.4. i E. Control' Rod Block Actuation: E. Control Rod Block Actuation During reactor power Instrumentation and logic operation the instrumentation systems shall be functionally that initiates control rod tested and calibrated as
-block shall be operable indicated in Table 4.2.5.
in accordance with Table 3.2.5. F. Mechanical Vacuum Pump F. Mechanical Vacuum Pump Isolation Isolation
- 1. Whenever the main steam During each operating cycle, line isolation valves are automatic isolation and open, the mechanical securing of the mechanical
. vacuum pump shall be vacuum pump shall be verified 4
capable of being while the reactor is ' automatically isolated and shutdown, secured by a signal of high' radiation in the main steam line tunnel or shall be manually isolated and secured.
- 2. If Specification 3.2.F.1 is not met following a routine surveillance check, the reactor shall be in the cold shutdown within 24 hours.
G. Post-Accident Instrumentation G. Post-Accident Instrumentation During reactor power The post-accident operation,the instrumentation instrumentation shall be that displays information in functionally tested and the Control Room necessary calibrated in accordance with . for the operator to initiate Table 4.2.6. and control the systems used during and following a postulated accident or l
, abnormal operating condition shall be operable in accordance with Table 3.2.6.
Amendment No. 9- 35 2, , _ . _ _ _ _ . ;_ _ _ _ .- _ _ _ _ _ , . .. - __ - . . . ,
1 l VYNPS 1 1 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION ' i l H. Drywell to Torus AP H. Drywell to Torus AP i Instrumentation Instrumentation
- 1. During reactor power The Drywell to Torus AP operation, the Drywell to Instrumentation shall be Torus AP Instrumentation calibrated once every six
- (recorder #1-156-3 and months and an instrument I
instrument DPI-1-158-6) check will be made once per shall be operable except shift. as specified in 3.2.H.2.
- 2. From and after the date that one of the Drywell to Torus AP instruments is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding thirty days unless the instrument is sooner made operable. If both instruments are made or found to be inoperable, and indication cannot be restored within a six hour period, an orderly shutdown shall be initiated and the reactor shall be in a hot shutdown condition in six hours and a cold shutdown condition in the following eighteen hours.
I. Recirculation Pump Trip ~ I. Recirculation Pump Trip Instrumentation Instrumentation During reactor power The Recirculation Pump Trip operation, the Recirculation Instrumentation shall be Pump Trip Instrumentation functionally tested and
. shall be operable calibrated in accordance with in accordance with Table 4.2.1.
Table 3.2.1. J. Deleted J. Deleted K. Degraded Grid Protective K. Degraded Grid Protective System System During reactor power The emergency bus operation, the emergency bus undervoltage instrumentation undervoltage instrumentation shall be functionally tested l shall be operable in and calibrated in accordance accordance with Table 3.2.8. with Table 4.2.8. Amendment No. 60, 68, 96, 96, 144, 133- 36
VYNPS TABLE 3.2.1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Core Spray - A & B (Note 1) Minimum Number of Required Action When Operable Instrument Minimum Conditions Channels per Trip For Operation System Trip Function Trip Level Setting Are Not Satisfied 2 High Drywell Pressure 52.5 psig Note 2 l (PT-10-101 (A-D) (M) ) 2 Low-Low Reactor Vessel Water >82.5" above top of Note 2 l Level (LT-2-3-72 (A-D) (M) ) enriched fuel 1 Low Reactor Pressure 300 $ P $ 350 psig Note 2 (PT-2-3-56C/D(M)) 2 Low Reactor Pressure 300 $ P $ 350 psig Note 2 (PT-2-3-56A/B(M) & I PT-2-3-52C/D(M)) 1 Time Delay (14A-K16A & B) $10 seconds Note 2 2 Pump (P-46-1A/B) Discharge >100 psig Note 5 l Pressure ( PS-14-4 4 (A-D) ) 1 Auxiliary Power Monitor -- Note 5 (LNPX C/D) l 1 Pump Bus Power Monitor -- Note 5 l (27/3A/B, 27/4A/S) 1 Trip System Logic -- Note 5 1 i Amendment No. 44, 68, 444, 4-M, 4-44, 443 38
VYNPS TABLE 3.2.1 (Cont'd) EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Low Pressure Coolant Injection System A & B (Note 1) Minimum Number of Required Action When Operable Instrument Minimum Conditions Channels per Trip For Operation System Trip Function Trip Level Setting Are Not Satisfied 1 Low Reactor Pressure 300 $ p 5 350 psig Note 2 (PT-2-3-56C/D(M)) 2 High Drywell Pressure $2.5 psig Note 2 l [PT-10-101(A-D) (M) ) 2 Low-Low Reactor Vessel Water >82.5" above top of Note 2 Level (LT-2-3-72 (A-D) (SI) ) enriched fuel l 1 Time Delay (10A-K51A & B) O seconds Note 5 1 Reactor Vessel Shroud Level >2/3 core height Note 5 l (LT-2-3-73A/B(M)) 1 Time Delay (10A-K72A & B) $60 seconds Note 5 1 Time Delay (10A-K50A & B) $5 seconds Note 5 1 Low Reactor Pressure 100 $ p $ 150 psig Note 2 (PS-2-128A & B) 2 per pump RHR Pump (A-D) Discharge >100 psig Note 5 Pressure (PS-10-105(A-H)) 2 High Drywell Pressure $2.5 psig Note 2 I ( PT-10-101 (A-D) (S1) ) Amendment No. 44, ~ 4, 68, 444, 442 39
VYNPS TABLE 3.2.1 (Cont'd) EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Low Pressure Coolant Injection System A & B (Note 1) Minimum Number of Required Action When Operable Instrument Minimum conditions Channels per Trip For Operation System Trip Function Trip Level Setting Are Not Satisfied 1 Time Delay (10A-K45A & B) 56 minutes Note 5 2 Low Reactcr Pressure 300 $ p $ 350 psig Note 2 (PT-2-3-56A/B(M) & l PT-2-3-52C/D(M)) 1 Auxiliary Power Monitor -- Note 5 I (LNPX C/D) 1 Pun:p Bus Power Monitor -- Note 5 I (27/3A/B, 27/4A/B) 1 Trip System Logic -- Note 5 i Amendment No. 44, 440, 442 40
.. . .. - . - . - . . . , ~. - . . . = .. . .- .
2 1 VYNPS
- t TABLE 3.2.1 (Cont'd)
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMEh7ATION High Pressure Coolant Injection System Minimum. Number of Required Action.When l' Cperable Instrument Minimum Conditions Channels per Trip, For Operation j System Trip Function Trip Level Setting Are Not Satisfied 4 2 (Note 3) Low-Low Reactor Vessel Water Same as LPCI Note 5 ! l Level (LT-2-3-72 (A-D) (SI) ) 'f !' 2 (Note 4) Low Ccndensate Storage Tank. > 3% Note 5 l Water Level (LSL-107-5A/B) , 2 (Note 3) High Drywell Pressure Same as LPCI- Note 5 ( PT-10-101 ( A-D) (M) ) t 1 (Note 3) ' Bus Power Monitor (23A-K41) - Note 5 t t 1 (Note 4) Trip System Logic -- Note 5 2 (Note 7) High Reactor Vessel Water <177 inches al.ove top of Note 5 t 1 l Level (LT-2-3-72A/B) (S4 ) enriched fue)
- I
, 'I 4 i
- I
! .i ! o
.L t !
a , a I t Amendment No. 68, M, M 41 i
.. - _ _ . _ _. _. .. ._ . . . _ .._ ._ . . . . _ . . . . . . . _ _ . .,. s
+ .j t i VYNPS L TABLE 3.2.1 . (Cont'd) ,! i EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ; I t Automatic Depressurization Minimum Number of Required Action When .; Operable Instrument Mininnnn Conditions- ! Channele per Trip For Operation ! System (Note 4) . Trip Function Trip Level Setting Are Not Satisfied 2 Low-Low Reactor Vessel Water Same as Core Spray Note 6 j 3 l Level (LT-2-3-72 (A-D) (M) ) ' l 2 High Drywell Pressure $2.5 psig Note 6 l (PT-10-101 (A-D) (SI) )
- 1 Time Delay (2E-KSA/B) $120 seconds Note 6 ;
i . 1 Sus Power Monitor - (2E-K1A/B) -- Note 6- , 1 Trip System Logic -- Note 6 { 2 Time Delay <8 minutes- Note 6 (2E-K16A/B, 2E-K17A/B)
.r t
l
- I
\
f . l l t t - F l 1 e L i l Amendment No. 44, MG 42 i
w-f I VYNPS } TABLE 3.2.1 + (Cc,nt 'd) - l RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION Recirculation Pump Trip - A & B (Note 1) { Minimum Number of Required Action When l Operable Instrument Minimum Conditions l Channels per Trip For Operation' i System Trip Function Trip Level Setting Are Not Satisfied . 2 Low-Low Reactor Vessel Water > 6' 10.5" above top of Note 2 ; l l Level (LM-2-3-68(A-D)) enriched fuel ; l 2 High Reactor Pressure < 1150 psig Note 2 [ (PM-2-3-54(A-D)) ' 2 Time Delays (2-3-68 (A-D) (X) ) < 10 seconds Note 2 j 1 Trip Systems Logic -- Note 2 i i l ; i I b i 4, I I i
?
t i , i 3 i 1 ; i Amendment No. 68, 48, M, M 43 l l
- l. !
1 l l VYNPS l TABLE 3.2.1 NOTES
- 1. Each of the two Core Spray, LPCI and RPT, subsystems are initiated and l controlled by a trip system. The subsystem "B" is identical to the i subsystem "A".
)
i l 2. If the udnimum number of operable instrument channels are not available, the inoperable channel shall be tripped using test jacks or other permanently installed circuits. If the channel cannot be tripped by the ! means stated above, that channel shall be made operable within 24 hours or ! an orderly shutdown shall be initiated and the reactor shall be in the cold l shutdown condition within 24 hours. l l
- 3. One trip system with initiating instrumentation arranged in a j one-out-of-two taken twice logic.
{
- 4. One trip system with initiating instrumentation arranged in a one-out-of- I two logic.
- 5. If the minimum number of operable channels are not available, the system is considered inoperable and the requirements of Specification 3.5 apply. '
- 6. Any one of the two trip systems will initiate ADS. If the minimum number of operable channels in one trip system is not available, the requirements of Specification 3.5.F.2 and 3.5.F.3 shall apply. If the minimum number of operable channels is not available in both trip systems, i Specifications 3.5.F.3 shall apply.
- 7. One trip system arranged in a two-out-of-two logic.
I 1 I l l l Amendment No. M 44 l l
VYNPS TABLE 3.2.2 PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION Minimum Number of Required Action When Operable Instrument Minimum Conditions Channels per Trip For Operation Are System Trip Function Trip Setting Not Satisfied (Note 2) 2 Low-Low Reactor Vessel Water >82.5" above the top of A Level (LT-2-3-57A/B(S2), enriched fuel LT-2-3-58A/B(S2)) 2 of 4 in each of High Main Steam Line Area ~<212*F B 2 channels Temperature l (TS-2-(121-124 ) (A-D) ) 2/ steam line High Main Steam Line Flow $140% of rated flow B l (DPT (116-119) (A-D) (M) ) 2/(Note 1) Low Main Steam Line Pressure >800 psig B l (PS-2-134(A-D)) 2/(Note 6) High Main Steam Line Flow $40% of rated flow B (DPT-2-116A,117B, ll8C,119D(S1)) 2 Low Reactor Vessel Water Level Same as Reactor A (LT-2-3-57A/B(M), Protection System LT-2-3-58A/B(M)) 2 High Main Steam Line Radiation $3 x background at rated B (7) (8) (RM-17-251(A-D)) power (9) 2 High Drywell Pressure Same as Reactor A Protection System 2/(Note 10) Condenser Low Vacuum $12" Hg absolute A i 1 Trip System Logic -- A Amendment No. 4, 68, 44, 46, GO 45
VYNPS TABLE 3.2.2 (Cont'd) HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION INSTRUMENTATION Minimum Number of Required Action When Operable Instrument Minimum Conditions Channels per Trip For Operation Are System Trip Function Trip Level Setting Not Satisfied 2 per set of 4 High Steam Line Space Note 3
$212*F Temperature (TS (101-104 ) (B-D) )
1 High Steam Line d/p (Steam $195 inches of water Note 3 Line Break) (DPIS-23-77/78) l 4 (Note 5) Low HPCI Steam Supply Pressure >70 psig Note 3 l -(PS-23-68 (A-D) ) 2 Main Steam Line Tunnel <212*F Note 3
~
I Temperature (TS-23-(101-104)A) 1 Time Delay (23A-K48) $35 minutes Note 3 . (23A-K49) l 1 Bus Power Monitor (23A-K38) -- -- 1 Trip System Logic -- -- o Amendment No. 68, 4+1 46
VYNPS TABLE 3.2.2 (Cont'd) RFACTOR CORE ISOLATION COOLING SYSTEM ISOLATION INSTRUMENTATION Minimum Number of Re. quired Action When Operable Instrument Minimum Conditions Channels par Trip For Operation Are System Trip Function Trip Level Setting Not Satisfied (Note 2) 2 Main Steam Line Tunnel Note 3
~<212*F l Temperature (TS-13-(79-82)A) a 1 Time Delay (13A-K41) $35 minutes Note 3 (13A-K42) 2 per set of 4 High Steam Line Space -<212*F Note 3 Temperature (TS-13-(79-82)(B,C,D)) '
l 1 High Steam Line d/p (Steam $195 inches of water Note 3 l l Line Break) (DPIS-13-83/84) 4 (Note 5) Low Steam Supply Pressure >50 psig Note 3 (PS-13-87(A-D)) 1 Bus Power Monitor (13A-K33) -- Note 3 1 Trip System Logic -- Note 3 1 Time Delay (13A-K7) 35 t $7 seconds Note 3 (13A-K31) i Amendment No. 64, 44 47 t
. m . - . . . _ ._ . , - - . ;__. _ _ . .
t [ 4 4 VYNE3 . t
' TABLE 3.2.3 [
1-REACTOR BUILDING VENTILATION ISOLATION & STANDBY GAS TREATMENT SYSTEM INITIATION , }- i t . Minimum Number of Required Action When ; Operable Instrument Minimum Conditions Channels per Trip _ For Operation -I
^
System Trip Function Trip Setting Are Not Satisfied j 2 Low Reactor Vessel Water Level Same as PCIS Note 1 l (LT-2-3-57A/B(M), '- LT-2-3-58A/B(M)) ! - 2 High Drywell Pressure Same as PCIS Note 1 , l (PT-5-12 (A-D) (M) ) { l 1 Reactor Building Vent $14 mr/hr Note 1 l (RM-17-452A/B) 4 j g 1 Refueling Floor Zone Radiation $100 mr/hr Note 1 [ l 1 (RM-17-453A/B) { j 1 Reactor Building Vent Trip -- Note 1 [
- System Logic j 1 Standby Gas Treatment Trip --
Note 1 j System Logic ! j 1 Logic Bus Power Monitor -- Note 1 { (16A-K52/53) e f I l.
- Note 1 - If the minimum number of operable instrument channels is not available in either trip system for more }
than 24 hours, the reactor building ventilation system shall be isolated and the standby gas I treatment system operated until the instrumentation is repaired. ! i i I l I I
- i
, l Amendment No. 49 [ E I
VYNPS TABLE 3.2.4 OFF-GAS SYSTEM ISOLATION INSTRUMENTATION Minimum Number of Required Action When Operable Instrument Minimum Conditions 1 Channels per Trip For Operation System Trip Function Trip Setting Are Not Satisfied i 1 Time Delay (Stack Off-Gas $ 2 minutes Note 1 Valve Isolation) (15TD & 16TD) $ 30 minutes 1 Trip System Logic -- Note 1 Note 1 - At least one of the radiation monitors between the charcoal bed system and the plant stack shall be operable during operation of the augmented off gas system. If this condition cannot be met, continued i I operation of the augmented off-gas system is permissible for a period of up to 7 days provided that at least one of the stack monitoring systems is operable and off gas system temperature and pressure are measured ; continuously. ; l l Amendment No. 9, && SO
VYNPS TABLE 3.2.5 CONTROL ROD BLOCK INSTRUMENTATION Minimum Number of Operable Instrument Modes in Which Function Channels per Trip Must be operable System Trip Function Refuel Startup Run Trip Setting Startup Range Monitor 3 a. Upscale (Note 2) (7-40 (A-D) ) X X $5 x 105 cps (Note 3) 2 b. Detector Not Fully Inserted X X (7-11 (A-D) (LS-4 ) ) Intermediate Range Monitor (Note 1) 2 a. Upscale (7-41 ( A-F) ) X X $108/125 Fun Scale 2 b. Downscale (Note 4) X X >5/125 Full Scale 2 (7-41 ( A-F) )
- c. Detector Not Fully Inserted X X (7-11 (E, F, G, H, J, K) (LS-4 ) )
Average Power Range Monitor (APRM A-F) l 2 a. Upscale (Flow Bias) X
$0.66(W-AW)+42% (Note 5) 2 b. Downscale X >2/125 Full Scale i
_ Rod Block Monitor (Note 6) l (RBM A/B) (Note 9) 1 a. Upscale (Flow Bias) (Note 7) X
$0.66(W-AW)+N (Note 5) ,
1 b. Downscale (Note 7) X >2/125 Full Scale 1 Scram Discharge Volume X X X $12 Gallons l (Note 8) (per (LT-3-231A/G (SI)) volume) 1 Trip System Logic X X X L Amendment No. M, 24, 64, 66, M, M, 94, 94, M1 51 - _ - _ _ _ = - - _ _ _ = _ --
-=-:--=_.==_-_ _-- -
=a2ar__-==='e==.--===.-_=e= -
-emd_K. ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - . _ - _ _ _ - _ . _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - - - - - - _ - - - _ - _ _ _ - - - - - - _ - - _ _ - - - - _ _ - - -
3 VYNPS
. TABLE 3.2.6 POST-ACCIDENT INSTRUMENTATION i
, -Minimum Number of i Operable Instrument Channels Parameter Type of Indication Instrument Range 2 Drywell Atmospheric Recorder #TR-16-19-45 0-350*F
- l Temperature-(Note 1) (Blue)
Meter #TI-16-19-30B 0-350*F i i 2 Containment Pressure (Note 1) Meter iPI-16-19-12A (-15) -(+260) psig ,
?
Meter #PI-16-19-12B (-15) -(+260) psig. 2 Torus Pressure (Note 1) Meter #PI-16-19-36A (-15) -(+65) psig i Meter #PI-16-19-36B (-15) -(+65) psig . 2 Torus Water Level (Note 3) Meter #LI-16-19-12A 0-25 ft. l Meter #LI-16-19-12B 0-25 ft. l 2 Torus Water Temperature Meter #TI-16-19-33A 0-250*F (Note 1) Meter #TI-16-19-33C 0-250*F 2 Reactor Pressure (Note 1) . Meter #PI-2-3-56A 0-1500 psig Meter IPI-2-3-56B 0-1500 psig 2 Reactor Vessel Water Level Meter #LI 7 2-3-91A (-200)-0-(+200)"H 2O (Note 1) Meter ILI-2-3-91B (-200)-0-(+200)"H 2O 2 Torus Air Temperature (Note 1) Recorder ITR-16-19-45 0-350*F (Red) I Meter #TI-16-19-41 50-300*F 2/ valve Safety / Relief Valve Position Lights RV-2-71(A-D) Closed - Open From Pressure Switches From PS-2-71-(1-3) (A-D) ; (Note 4) f Amendment No. M, 63, M, %, -M-3, 445 53 [
. . . . . ~ .
.w. .. .
VYNPS-' l TABLE 3.2.6 - (Cont'd) , POST-ACCIDENT INSTRUMENTATION i Minimum Number of Operable-Instrument Channels Parameter Type of Indication Instrument Range
]
1/ valve Safety Valve Position From. . Meter ZI-2-1A/B. Closed 'Open I Acoustic Monitor (Note 5) 2 ! 2 Containment Hydrogen / Oxygen Recorder SR-VG-6A-(SI)' . 0-30% hydrogen Monitor (Note 1) Recorder SR-VG-6B -(SII) 0-25% oxygen 3 r 7 2 Containment High-Range Mete.r RM-16-19-1A/B 1 R/hr-10 R/hr Radiation Monitor (Note 6) 1 Stack Noble Gas Effluent Meter'RM-17-155 0.1 - 10' mR/hr ' (Note 7) t I i i i
+
t 1 1 t Amendment No. 63, 90, 96, 48 54 e
.-.-__.___.___.__m._.__.._m_.-_ _______m .__ m._______.. __ m____ _ _ , _ _ _ . . __.__.____.-.______m _ _ _ . _ _ . . _ _ _ _ _ _ _ . - . _ _ _ _m._. -..______
VYNPS TABLE 3.2.7 (Table 3.2.7 was intentionally deleted from the Technical Specifications) L i i t I Amendment No. 55a m - . _ _ _ _ ._. . - - _ _ _ _ _ ___._______..__.__._.______-..___.;._.___ _ _ _ _ . _ _ _ _ _ . _ _ . . _ . _ _ _ _ . _ . _ _ _
,_ _ .- . _ _ . _ . . ..-.._.m , _ _ ._. _ - _ _ _ . . . . . _ . . _ . .
VYNPS- ' f i TABLE 3.2.8
.i i EMERGENCY-BUS UNDE JOLTAGE INSTRUMENTATION i Minimum Number of i Operable l
-Instruments Parameter Trip Setting Required Action 2 per bus Degraded Bus Voltage - Voltage 3,700 volts i 40 volts Note 1 ~!
(27/3Z,'27/3W, 27/4Z, 27/4W) i !. 2 per bus Degraded Bus Voltage - Time 10 seconds'i 1 second Note 2- ' l Delay (62/3W, 62/3Z, 62/4W, 62/4Z) , TABLE 3.2.8 NOTES [ ] ! I
- 1. If the minimum number of operable instrument channels are not available, . the inoperable channel shall be :
tripped us?ng test jacks or other permanently installed circuits within one hour. !
- 2. If the minimum number of operable instrument channels are not available, reactor power operation is. ;
permissible for only 7 successive days unless the system is sooner made operable. , L i i L 4 = 4
?
Amendment No. G8 56 l I i
.I' a..
VYNPS TABLE 3.2.9 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION Minimum Number of Required Action When. Operable Instrument Minimum Conditions !
. Channels per Trip For Operation System Trip Panction Trip Level Setting Are Not Satisfied
- 2 (Note 1) Low-Low Reactor Vessel Water >82.5" Above Top of Note 4 -
l Level (LT-2-3-72 (A-D) (M) ) Enriched Fue1 I 2 (Note 2) Low Condensate Storage Tank >3% Note 4 j l Water Level (LT-107-12A/B(M)) t i 2 (Note 3) High Reactor Vessel Water $177" Above Top of Note 4 l Level (LT-2-3-72C/D(S2)) Enriched Fuel 1 1 Bus Power Monitor (13A-K36) -- Note 4 l
-t 1 Trip System Logic -- Note 4 i
i l i i 1 1 1 ! I i [ Amendment No. 444 57
t VYNPS . TABLE 4.2.1
- .. (Cont'd) ~ .j MINIMUM TEST AND CALIBRATION FREQUENCIES EMERGENCY CORE COOLING ACTUATION INSTRUMENTATION .
Recirculation Pump Trip Actuation System , Trip Function Functional Test (8) Calibration (8) Instrument Check l Low-Low Reactor Vessel- (Notes 1 and 4) 'Once/ Operating Cycle .Once Each Day Water Level High Reactor Pressure (Notes 1 and 4) Once/ Operating Cycle Once Each Day i Trip System Logic Once/ Operating Cycle once/ Operating Cycle -- 1 r
.l i
.i 5
i ! 4 t k i I Amendment No. M, 4% 63 [ t
VYNPS TABLE 4.2.7 (Table 4.2.7 was intentionally deleted from the Technical Specifications) i 2 l Amendment No. 71a
_ . - . . ._-_ . _ - - _ . _ . _ . _ _ _ . . _._m _ . _ . _ . _ . - _ _ _ - _ _ _ . _ _ . . _ _ VYNPS TABLE 4.2 NOTES
-1. Initially once per month; thereafter, a longer interval as determined by test results on this type of instrumentation.
- 2. During each refueling outage, simulated automatic actuation whi'ch opens all pilot valves shall be performed such that each trip system logic can be verified independent of its redundant counterpart.
- 3. Trip system logic calibration shall include only time delay relays and i
1 timers necessary for proper functioning of the trip system. I l l 4. This instrumentation is excepted from functional test definition. The ! functional test will consist of injecting a simulated electrical signal into the measurement channel. j l S. Deleted. t l 6. Functional tests, calibrations, and instrument checks are not required when these instruments are not required to be operable or are tripped. Functional tests shall be performed before each startup with a required frequency not to exceed once per week. Calibration shall be performed prior to or during each startup or controlled shutdown with a required frequency not to exceed once per week. Instrument checks shall be performed at least once per day during those periods when instruments are required to be operable.
- 7. This instrumentation is excepted from the functional test definitions and l shall be calibrated using simulated electrical signals once every three I months.
- 8. Functional tests and calibrations are not required when systems are not required to be operable.
- 9. The thermocouples associated with safety / relief valves and safety valve position, that may be used for back-up position indication, shall be verified to be operable every operating cycle.
- 10. Separate functional tests are not required for this instrumentation. The calibration and integrated ECCS tests which are performed once per operating cycle will adequately demonstrate proper equipment operation.
- 11. Trip system logic functional tests will include verification of operation of all automatic initiation inhibit switches by monitoring relay contact movement. Verification that the manual inhibit switches prevent opening all relief valves will be accomplished in conjunction with Section 4.5.F.1.
Amendment No. 63, GB, 446, 146 74
. m , - _ . . _ . . . . _ . . . _ _ _ . _ _ . _
_ .m ._m
^
l f- 1 VYNPS 1 L 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS-OPERATION l With the required shutdown margin not met and the mode switch in - I the " Refuel" position,, I
.immediately'auspend Alteration of the Reactor Core except for control rod insertion and fuel I assembly removal; I immediately initiate .
action to fully insert ! all insertable control rods in core cells ) l containing one or more l fuel assemblies; within 1 hour, initiate action i to restore the integrity of the Secondary Containment System.
- 2. Reactivity Margin - 2. Reactivity Margin -
Inoperable Control Rods Inoperable Control Rods Control rod drives which { cannot be moved with Each partially-or fully withdrawn operable control rod drive control rod shall be pressure shall be exercised one notch at , considered inoperable. least once each week. < l -If a partially or fully This test shall be l withdrawn control rod performed at least once drive cannot be moved per 24 hours in the event with drive or scram power operation is pressure, the reactor continuing with two or shall be brought to a more inoperable control shutdown condition within rods or in the event 48 hours unless power operation is I investigation continuing with one fully demonstrates that-the or partially withdrawn cause of the failure is rod which cannot be moved not due to a failed and for which control rod control rod drive drive mechanism damage mechanism collet housing, has not been ruled out. The control rod The surveillance need not directional control be completed within-valves for inoperable 24 hours if the number control rods shall be t> l l Amendment No. -148 Bla i,
l VYNPS ( 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS OPERATION l E. Reactivity Anomalies E. Reactivity Anomalies 1 The reactivity equivalent of During the startup test the difference between the program and startups actual critical rod following refueling outagcs, configuration and the the critical rod ; expected configuration during configurations will be l power operation shall not compared to the expected l exceed 1% Ak/k. If this configurations at selected limit is exceeded, the Operating conditions. These reactor will be shut down comparisons will be used as until the cause has been base data for reactivity l determined and corrective monitoring during subsequent ' actions have been taken if power operation throughout such actions are appropriate, the fuel cycle. At specific power operating conditions, F. If Specifications 3.3.B the critical rod through 3.3.D above are not configuration will be met, an orderly shutdown compared to the configuration shall be initiated and the expected based upon reactor shall be in the cold appropriately corrected past shutdown condition within data. This comparison will 24 hours. be made at least every equivalent full power month. Amendment No. 49, -148 88
q VYNPS BASES: 3.3 & 4.3 (Cont'd)
- 2. The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the extremely remote event of a housing failure. The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage ef the primary coolant system. The design basis is given in Subsection 3.5.2 of the FSAR, and the design evaluation is given in Subsection 3.5.4. This support is not required if the reactor coolant system is at atmospheric pressure since there would then be no driving force to rapidly eject a drive housing.
- 3. In the course of performing normal startup and shutdown procedures, a pre-specified sequence for the withdrawal or insertion of control rods is followed. Control rod dropout accidents which might lead to significant core damage, cannot occur if this sequence of rod withdrawals or insertions is followed. The Rod Worth Minimizer restricts withdrawals and insertions to those listed in the pre-specified sequence and provides an additional check that the reactor operator is following prescribed sequence. Although beginning a reactor startup without having the RWM operable would entail unnecessary risk, continuing to withdraw rods if the RWM fails subsequently is acceptable if a second licensed operator verifies the withdrawal sequence. Continuing the startup increases core power, reduces the rod worth and reduces the consequences of dropping any rod. Withdrawal of rods for testing is permitted with the RWM inoperable, if the reactor is subcritical and all other rods are fully inserted. Above 20% power, the RWM is not needed since even with a single error an operator cannot withdraw a rod with sufficient worth, which if dropped, would result in anything but minor consequences.
l 4. Refer to the Vermont Yankee Core Performance Analysis Report.
- 5. The Source Range Monitor (SRM) system has no scram functions. It does provide the operator with a visual indication of neutron level. The consequences of reactivity accidents are a function of the initial neutron flux. The requirement of at least three counts per second assures that any transient, should it occur, begins at or above the initial value of 10" of rated power used in the analyses of transients from cold conditions. One operable SRM channel is adequate to monitor the approach to criticality, therefore, two operable SRM's are specified for added conservatism.
- 6. The Rod Block Monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power level operation. During reactor operation with certain limiting control rod patterns, the withdrawal of a designated single control rod could result in one or more fuel rods with MCPR less than the fuel cladding integrity safety limit. During use of such patterns, it is judged that testing of the RBM system prior to withdrawal of such rods will provide added assurance that improper withdrawal does not occur.
It is the responsibility of the Nuclear Engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as they dnvelop due to the occurrence of inoperable control rods. Amendment No. 36, 39, 6+, 4G 90
VYNPC - 3. 4 ' LIMITING CONDITIONS FOR 4.4 SURVZILLANCE REQUIREMENTS OPERATION 3.4 REACTOR STANDBY LIQUID CONTROL 4'4 REACTOR STANDBY LIQUID CONTROL SYSTEM SYSTEM Applicability: Applicability: Applies to the operating status Applies to the periodic testing of the Reactor Standby Liquid requirement for the Reactor < Control System. Standby Liquid Control System. Objective: Objective: To assure the availability of an To verify the operability of the
-independent reactivity control Standby Liquid Control System.
mechanism. Specification: Specification: A. Normal Operation A. Normal Operation Except as specified in 3.4.B The Standby Liquid Control below, the Standby Liquid System shall be verified Contro1' System shall be operable by: operable during periods when fuel is in the reactor unless: I
- 1. The reactor is in cold 1. Testing pumps and valves shutdown in accordance with Specification 4.6.E. A and Minimum flow rate of 35 gpm at 1275 psig 1 shall be verified for each pump by recirculating demineralized water to the test tank.
- 2. Control rods are fully 2. Verifying the continuity inserted and of the explosive charges Specification 3.3.A is at least monthly.
met. In addition, at least once during each operating cycle, the Standby Liquid Control System shall be verified 1 operable by: l
- 3. Testingthatthesettingl of the pressure relief valves is between 1400 and 1490 psig.
- 4. Initiating one of the l standby liquid control loops, excluding the primer chamber and inlet fitting, and verifying j that a flow path from a i pump to the reactor i l
I Amendment No. M3, MB 92 l 1
. . , . - . ~ - .~ .- . .- -. . - . - _ . - . - . .. - - . .
VVNPS 3.4 LIMITING CONDITIONS FOR 4.4 SURVEILLANCS REQUIREMENTS OPERATION " vessel is available by pumping demineralized water into the reactor vessel. Both loops shall be tested over the course of two operating cycles.
- 5. Testing the new trigger assemblies by installing l one of the assemblies in the test block and firing it using the installed circuitry.
Install the unfired assemblies, taken from the same batch as the fired one, into the explosion valves.
- 6. Recirculating the l borated solution.
B. Operation with Inoperable B. Operation with Inoperable Components Components From and after the date that When a component becomes
.a redundant component is inoperable, its redundant made or found to be component shall be or shall inoperable, reactor- have been demonstrated to be operation is permissible operable within 24 hours.
during the succeeding seven , days unless such component-is sooner made operable. C. Liquid Poison Tank - Boron C. Liquid Poison Tank - Boron Concentration Concentration At a11' times when the Standby Liquid Control System is required to be operable, the following conditions shall be met:
- 1. The net volume versus 1. The solution volume in concentration of the the tank and temperature sodium pentaborate in the tank and suction solution in the standby piping shall be checked liquid control tank at least daily, shall meet the requirements of Figure 3.4.1.
Amendment No, M3, -144 93
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l i VYNPS 3.5 LIMITING CONDITION FOR 4.5 SURVEILLANCE REQUIREMENT l OPERATION 1
- 3. From and after the date 3. When the Alternate !
- that the Alternate Cooling Subsystem or both i
Cooling Tower Subsystem Station Service Water or both Station Service Subsystems are made or Water-Subsystems are made found to be inoperable, or found inoperable for the operable subsystem (s) any reason,-reactor l shall have been or shall
- operation is permissible be demonstrated to be
, only during the operable within 24 hours, i succeeding seven days l unless such subsystem (s) ! are made' operable, provided that during such ,. seven days all other active components of the l other subsystem (s) are operable. 4. If the requirements of Specification 3.5.D-r cannot be met,Lan orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition within 24 hours. E. High Pressure Cooling E. High Pressure Coolant j Injection (HPCI) System Injection (HPCI) System Surveillance of HPCI System shall be performed as follows:
- 1. Except as specified in 1. Testing Specification 3.5.E.2, whenever irradiated fuel Item Frequency is in the reactor vessel 4-and reactor pressure is Simulated Each re-greater than 150 psig and Automatic fueling prior to reactor startup Actuation outage from a cold condition: Test i a. The HPCI System Operability testing of
. shall be operable. the pump and valves shall be in accordance with
- b. The condensate Specification 4.6.E. The storage tank shall HPCI System shall deliver contain at least at least 4250 gpm at 75,000 gallons of normal reactor operating condensate water. pressure when recirculating to the Condensate Storage Tank.
4 Amendment No. 34, 444, 448 105
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i VYNPS 3.5 LIMITING CONDITION FOR 4.5 SURVEILLANCE REQUIREMENT OPERATION i 2. From and after the date 2. When the HPCI Subsystem that the HPCI Subsystem is made or found to be is made or found to be inoperable, the Automatic i inoperable for any Depressurization System , j reason, reactor operation shall have been or shall l 1s permissible only be demonstrated to be ' during the succeeding operable within 24 hours. seven days unless.such subsystem is sooner made NOTE: Automatic operable, provided that Depressurization during such seven days System operability 1 all active components of shall be the Automatic demonstrated by Depressurization performing a Subsystems, the Core functional test of 4 Spray Subsystems, the the trip system LPCI Subsystems, and the logic. RCIC System are operable.
- 3. If the requirements of Specification 3.5.E cannot be met, an orderly l
, shutdown shall be initiated and the reactor I pressure shall be reduced l to s 120 psig within j l 24 hours. l F. Automatic Depressurization F. Automatic Depressurization System- System Surveillance of the Automatic Depressurization System shall ! be performed as follows:
- 1. Except as specified in 1. Operability testing of Specification 3.5.F.2 the relief valves shall below, the entire be in accordance with Automatic Specification 4.6.E.
Depressurization Relief i System shall be operable l ) at any time the reactor l pressure is above a 100 psig and irradiated fuel is in the reactor vessel.
- 2. From and after the date 2. When one relief valve of that one of the four the Automatic Pressure relief valves of the Relief Subsystem is made
' Automatic or found to be Depressurization inoperable, the HPCI Subsystem are made or Subsystem shall have been found to be inoperable or shall be demonstrated due to malfunction of the to be operable within electrical portion of the 24 hours.
valve when the Amendment No. M, 4-14, 4*B 106
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VYNPS 3.5 LIMITING CONDITION FOR 4.5 SURVEILLANCE hE;OIREMENT OPERATION
^
-reactor is pressurized above 100 psig with irradiated fuel in the
., reactor vessel, continued reactor operation is permissible only during the succeeding seven days unless such a valve is sooner made operable, provided'that during such seven days both the remaining Automatic
< Relief System valves and the HPCI System are operable.
S
- 3. If the requirements of Specification 3.5.F cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall be reduced l to s 100.psig within, 24 hours..
G. Reactor Core Isolation G. Reactor Core Isolation Cooling System (RCIC) Cooling System (RCIC) Surveillance of the RCIC System shall be performed as follows:
- 1. Except as specified in 1. Testing Specification 3.5.G.2 below, the RCIC System Item Frequency shall be operable whenever the reactor Simulated Each re-pressure is greater than automatic fueling 150 psig and irradiated actuation outage fuel is in the reactor test vessel.- (testing valve
- 2. From and after the date operability) that the RCIC System is made or found to be Operability testing of inoperable for any the pump and valves shall reason, reactor operation be in accordance with is permissible only Specification 4.6.E. The during the succeeding RCIC System shall deliver 7 days unless such system at least 400 gpm at is sooner made operable, normal reactor operating l
provided that during such pressure when 7 days all active recirculating to the components of the HPCI Condensate Storage Tank. System are operable.
' Amendment No. 43,5 M, 444, MB 107
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VYNPS 4 3.5 LIMITING CONDITION FOR 4.5 SURVEILLANCE REQUIREMENT
, OPERATION
- 3. If the requirements of Specification 3.5.G cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall be reduced to 5 120 psig within l
24 hours. H. Minimum Core and Containment H. Minimum Core and Containment Cooling System Availability Cooling System Availability
- 1. - During any period when 1. When one of the emergency l one of the emergency diesel generators is made l or found to be diesel generators is inoperable, continued inoperable, the remaining reactor operation is diesel generator shall permissible only during have been or shall be the succeeding seven demonstrated to be i days, provided that all operable within 24 hours.
of the LPCI, Core Spray and Containment Cooling Subsystems connecting to the operable diesel generator shall be operable. If this requirement cannot be met,.an orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours.
- 2. Any combination of inoperable components in the Core and Containment Cooling Systems shall not defeat the capability of the remaining operable components to fulfill the core and containment cooling functions.
- 3. When irradiated fuel is in the reactor vessel and i the reactor is in the
! cold shutdown condition, all Core and Containment Cooling Subsystems may be inoperable provided no
- work is permitted which
, has the potential for draining the reactor vessel. Amendment No. N, 444 108
k VYNPS 3.6. LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION >
- e. With the radiciodine concentra^ ion in the reactor ciolant greater tlan 1.1 micr(curies /-
gram dos, equivalent I-131, a sample of , reactor coolant shall be taken every 4 hours and analyzed ' for radioactive iodines of I-131 through I-135, until the specific activity of the reactor coolant is restored below 1.1 microcuries/ gram dose equivalent I-131.
- 2. The reactor coolant. water 2. During startups and at shall not exceed the steaming rates below following limits with 100,000 pounds per hour, steaming rates less than a sample of reactor 100,000 pounds per hour coolant shall be taken except as specified in every four hours and ,
Specification 3.6.B.3: analyzed for conductivity and chloride content. Conductivity Sumho/cm l Chloride ion . 0.1 ppm c 3. For reactor startups the 3. a. With steaming rates maximum value for greater than or 4 conductivity shall not equal to exceed 10 umho/cm and the 100,000 pounds per maximum value for hour, a reactor chloride ion coolant sample shall concentration shall not be taken at least exceed 0.1 ppm, in the every 96 hours and reactor coolant water for when the continuous the first 24 hours after conductivity placing the reactor-in monitors indicate i the power operating abnormal ,- condition. conductivity (other than short-term spikes), and analyzed for conductivity and , chloride ion content. I Amendment No. M 118
, - . - . . . . ' . , , - - , . , , -_ . . . _ _ _ ..-.__.,,__..m.r .em_.. -__.__,_,__.__.,_.,,#m..__, ,,--...,,~,_,.,____,..,,ic. - - , , ,_, . , _ , , - . ,
.- - . .- -- .~. _. .
VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION
- b. When the continuous conductivity monitor is inoperable, a reactor coolant sample shall be taken every four hours and analyzed
- 4. Except as specified in for conductivity and Specification 3.6.B.3 chloride ion above, the reactor content, coolant water shall not exceed the following limits with steaming rates greater than or equal to 100,000 pounds per hours.
Conductivity 5 uhmo/cm Chloride ion 0.5 ppm
- 5. If Specification 3.6.B is not met, an orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours.
- c. Coolant Leakage C. Coolant Leakage 1.a. Any time 1. Reactor coolant system irradiated fuel is leakage, for the in the reactor purpose of satisfying vessel and reactor Specification 3.6.C.1, coolant shall be checked and temperature is logged once per shift, above 212'r, not to exceed reactor coolant 12 hours, leakage into the primary containment from unidentified .
sources shall not exceed 5 gpm. In addition, the total reactor coolant system leakage into the primary containment shall not exceed 25 gpm.
- b. While in the run mode, reactor coolant leakage Into the primary containment from unidentified sources shall not Amendment No. 449 119
VYNPS BASES: 3.6 and 4.6 (Cont'd) l The actual shift in RTun of the vessel material will be established I periodically during operation by removing and evaluating, in accordance ' with ASTM E185 reactor vessel material irradiation surveillance l specimens installed near the inside wall of the reactor vessel in the ' core area. Since the neutron spectra at the irradiaticn samples and vessel inside radius are essentially identical, the mersured transition shift for a sample can be applied with confidence to tha adjacent section of the reactor vessel. Battelle Columbus Laboratory Report BCL-585-84-3, dated May 15, 1984, provides this information for the ten-year surveillance capsule. In order to estimate the material properties at the 1/4 and 3/4 T positions in the vessel plate, the shift in RTug is determined in accordance with Regulatory Guide 1.99, Revision 2. The heatup and cooldown curves must be recalculated when the ARTue determined from the surveillance capsule is different from the calculated ARTun for the equivalent capsule radiation exposure, The pressure-temperature limit lines, shown on Figure 3.6.1, for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10CFR50 for reactor criticality and for inservice leak and hydrostatic testing. The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided to j assure compliance with the requirements of Appendix H to 10CFR Part 50. l B. Coolant Chemistry A steady-state radioiodine concentration limit of 1.1 Ci of I-131 dose equivalent per gram of water in the Reactor Coolant System can be reached if the gross radioactivity in the gaseous effluents is near the limit, as set forth in Specification 3.8.E.1, or there is a failure or l prolonged shutdown of the cleanup demineralizer. In the event of a steam line rupture outside the drywell, the NRC staff calculations show the resultant radiological dose at the site boundary to be less than 30 Rem to the thyroid. This dose was calculated on the basis of the radiciodine concentration limit of 1.1 pCi of I-131 dose equivalent per gram of water, atmospheric diffusion from an equivalent elevated release of 10 meters at the nearest site boundary (190 m) for a X/Q = 3.9 x 10" sec/m' (Pasquill D and 0.33 m/sec equivalent), and a steam line isolation valve closure time of five seconds with a steam / water mass release of 30,000 pounds. The iodine spike limit of four (4) microcuries of I-131 dose equivalent per gram of water provides an iodine peak or spike limit for the reactor coolant concentration to assure that the radiological j consequences of a postulated LOCA are within 10CFR Part 100 dose guidelines. The reactor coolant sample will be used to assure that the limit of Specification 3.6.B.1 is not exceeded. The radioiodine concentration would not be expected to change rapidly during steady-stato operation over a period of 96 hours. In addition, the trend of the radioactive gaseous effluents, which is continuously monitored, is a good indicator of the trend of the radioiodine concentration in the reactor coolant. When a significant increase in radioactive gaseous effluents is indicated, as specified, an additional reactor coolant sample shall be taken and analyzed for radioactive iodine. Arendment No. 33, 63, 94, 43 140
l VYNPS BASES: 3.6 and 4.6 (Cont'd) impurities will also be within their normal ranges. The reactor l cooling samples will also be used to determine the chlorides. Therefore, the sampling frequency is considered adequate to detect long-term changes in the chloride ion content. Isotopic analyses l required by Specification 4.6.B.1.b may be performed by a gamma scan and gross beta and alpha determination. The conductivity of the feedwater is continuously monitored and alarm set points consistent with Regulatory requirements given in Regulatory Guide 1.56, " Maintenance of Water Purity in Boiling Water Reactors," l have been determined. The results from the conductivity monitors on I the feedwater can be correlated with the results from the conductivity monitors on the reactor coolant water to indicate demineralizer i breakthrough and subsequent conductivity levels in the reactor vessel water. i C. Coolant Leakage The 5 gpm limit for unidentified leaks was established assuming such leakage was coming from the reactor coolant system. Tests have been conducted which demonstrate that a relationship exists between the size of a crack and the probability that the crack will propagate. These tests suggest that for leakage somewhat greater than the limit l specified for unidentified leakage, the probability is small that imperfections or cracks associated with such leakage would grow rapidly. Leakage less than the limit specified can be detected within a few hours utilizing the available leakage detection systems. If the limit is exceeded and the origin cannot be determined in a reasonably short time the plant should be shutdown to allow further investigation and corrective action. The 2 gpm increase limit in any 24 hour period for unidentified leaks was established as an additional requirement to the 5 gpm limit by Generic Letter 88-01, "NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping." The removal capacity from the drywell floor drain sump and the equivalent drain sump is 50 gpm each. Removal of 50 gpm from either of these sumps can be accomplished with considerable margin. D. Safety and Relief Valves safety analyses have shown that only three of the four relief valves are required to provide the recommended pressure margin of 25 psi below the safety valve actuation settings as well as maintaining the fuel cladding integrity safety limit for the limiting anticipated overpressure transient. For the purposes of this limiting condition, a relief valve that is unable to actuate within tolerance of its set pressure is considered to be as inoperable as a mechanically malfunctioning valve. The setpoint tolerance value for as-left or refurbished valves is l specified in Section III of the ASME Boiler and Pressure vessel Code as i1% of set pressure. However, the code allows a larger tolerance value i for the as-found condition if the supporting design analyses l demonstrate that the applicable acceptance criteria are met. Safety analysis has been performed which shows that with all safety and safety relief valves within 13% of the specified set pressures in Table 2.2.1 f and with one inoperable safety relief valve, the reactor coolant pressure safety limit of 1375 psig and the MCPR safety limit are not exceeded during the limiting overpressure transient. Ghengc 16/ March 29, 1971, M, M, MG, M9, MO,MO 142
VYNPS BASES: 3.6 and 4.6 (Cont'd) E. Structura? Integrity and Operability Testing A pre-service inspection of the components listed in Table 4.2-3 of the FSAR was conducted after site erection to assure freedom from defects greater than code allowance; in addition, this serves as a reference base for further inspections. Prior to operation, the reactor primary system was free of gross defects. In addition, the facility has been designed such that gross defects should not occur i n i Amendment No. 434 142a
VYNPS BASES: 4.7 (Cont'd) The interiors of the drywell and suppression chamber are painted to prevent rusting. The inspection of the paint during each major refueling outage, approximately once per year, assures the paint is intact. Experience with this type of paint at fossil fueled generating stations indicates that the inspection interval is adequate. Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends. By requiring the suppression pool temperature to be continually monitored and frequently logged during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken. The requirement for an external visual examination following any event where potentially high loadings could occur provides assurance that no significant damage was encountered. Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to be the points of highest stress. Visual inspection of the suppression chamber including water line regions each refueling outage is adequate to detect any changes in the suppression chamber structures. Amendment No. 44-3 166a
VYNPS BASES: 4.7 (Cont'd) The maximum allowable test leak rate at the peak accident pressure of 44 psig (La) is 0.80 weight % per day. The maximum allowable test leak rate at the retest pressure of 24 psig (Lt) has been conservatively determined to be 0.59 weight percent per day. This value was verified to be conservative by actual primary containment leak rate measurements at both 44 psig and 24 psig upon completion of the containment structure. 1 As most leakage and deterioration of integrity is expected to occur 1 through penetrations, especially those with resilient seals, a periodic l leak rate test program of such penetration is conducted at the peak i accident pressure of 44 psig to insure not only that the leakage remains acceptably low but also that the sealing materials can withstand the accident pressure. i The Primary Containment Leak Rate Testing Program is based on Option B to 10CFR50, Appendix J, for development of leak rate testing and surveillance schedules for reactor containment vessels. Surveillance of the suppression Chamber-Reactor Building vacuum breakers consists of operability checks and leakage tests (conducted as ' part of the containment leak-tightness tests). These vacuum breakers are normally in the closed position and open only during tests or an i accident condition. Operability testing is performed in conjunction I with Specification 4.6.E. Inspections and calibration 3 are performed ' during the refueling outages; this frequency being based on equipment ) quality, experience, and engineering judgment. The ten (10) drywell-suppression vacuum relief valves are designed to open to the full open position (the position that curtain area is l equivalent to valve bore) with a force equivalent to a 0.5 psi ' differential acting on the suppression chamber face of the valve disk. This opening specification assures that the design limit of 2.0 psid between the drywell and external environment is not exceeded. Once i each refueling outage each valve is tested to assure that it will open , fully in response to a force less than that specified. Also it is inspected to assure that it closes freely and operates properly. )' 1 The containment design has been examined to establish the allowable I bypass area between the drywell and suppression chamber as 0.12 f ta , This is equivalent to one vacuum breaker open by three-eighths of an i inch (3/8") as measured at all points around the circumference of the i disk or three-fourths of an inch (3/4") as measured at the bottom of I the disk when the top of the disk is on the seat. Since these valves open in a manner that is purely neither mode, a conservative allowance of one-half inch (1/2") has been selected as the maximum permissible valve opening. Assuming that permissible valve opening could be evenly divided among all ten vacuum breakers at once, valve open position assumed to indication for an individual valve must be activated less than fifty-thousandths of an inch (0.050") at all points along the seal , surface of the disk. Valve closure within this limit may be determined l by light indication from two independent position detection and indication systems. Either system provides a control room alarm for a i nonseated valve. Amendment No. M, MB, MB 168
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I VYNPS TABLE 3.9.1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Minimum Channels Operable. Notes
- 1. Gross Radioactivity Monitors not Providing Automatic Termination of Release
- a. Liquid Radwaste Discharge Monitor -1* 1,4,5 (RM-17-350)
- b. Service Water Discharge Monitor 1 2,4,5 (RM-17-351)
- 2. Flow Rate Measurement Devices
- a. Liquid Radwaste Discharge Flow Rate 1* 3,4 Monitor I
(FIT-20-485/442)
- During releases via this pathway.
- t f
i f ( Amendment No. 44 193
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1 VYNPS TABLE 3.9.2 GASEOUS EFFLUENT MONITORING INSTRUMENTATION a Minimum Channels Instrument Operable Notes
- 1. Steam Jet Air Ejector (SJAE)
- a. -Noble Gas Activity Monitor 1 7, 8, 9 l (RM-17-150A/B)
- 2. Augmented Off-Gas System
- a. Noble. Gas Activity Monitor Between 1 2, 5, 6, 7 the Charcoal Bed System and the l Plant Stack (Providing Alarm and '
Automatic Termination of Release) (RAN-OG-3127, RAN-OG-3128)
- b. Flow Rate Monitor ,1 1, 5, 6
, (FI-OG-2002, FI-OG-2004, FI-OG- 1 2008)
- c. Hydrogen Monitor 1 3, 5, 6 3
)
(H2AN-OG-2921A/B, H2AN-OG-2922A/B) l 3. Plant Stack
- a. Noble Gas Activity Monitor 1 5, 7, 10 (RM-17-156, RM-17-157)
- b. Iodine Sampler Cartridge 1 4, 5 J. c. Particulate Sampler Filter 1 4, 5 1
- d. Sampler Flow Integrator 1 1, 5 (FI-17-156/157)
- e. Stack Flow Rate Monitor 1 1, 5
, (FI-108-22) _ 1 i 1 i 5 l I t I 1 l Amendment No. 83, 443 195
VYNPS TABLE'3.9.3 (Cont'd) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Sample Sampling and Collection Type and Frequency and/or Sample Locationsa Frequency of Analysis
- 2. DIRECT RADIATIONb 40 routine monitoring Quarterly. Gamma dose, at least once stations as follows: per quarter.
16 incident response Incident response TLDs in stations (one in each the outer monitoring meteorological sector) locations, de-dose only within a range of 0 to quarterly unless gaseous 4 km9; release LCO was exceeded in period. 16 incident response stations (one in each meteorological sector) within a range of 2 to 8 km9; the balance of the stations to be placed in special interest areas and control station areas. Amendment No. 84 198
UYNPS TABLE 3.9.4 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES ('l Reporting Levels Airborne i Particulate or Fish Milk Vegetation Sediment Analysis Water (pCi/1) Gases (pCi/m3) (pCi/Kg, wet) (pCi/1) (pCi/Kg, wet) (pCi/Kg,. dry) .. H-3 2 x 10*
- 3 Mn-54 1 x 10 3 x 10' r 2
Fe-59 4 x 10 1 x 10' Co-58 3 8 1 x 10 3 x 10 3 x 10 D Co-60 2 4 3 x 10 1 x 10 2 Zn-65 3 x 10 2 x 10' Zr-Nb-95 2 4 x 10 I-131 0.9 3 2 1 x 10 Cs-134 30 10 3 60 3 1 x 10 1 x 10 Cs-137 50 20 3 70 3 2 x 10 2 x 10 Ba-La-140 2 2 2 x 10 3 x 10 (a) Reporting levels may be averaged over a calendar quarter. When more than one of the radionuclides in Table 3.9.4 are detected in the sampling medium, the unique reporting requirements are not exercised if the following ccndition holds: concentration (1) concentration (2)
+... < 1. 0 reporting level (1) reporting level (2)
When radionuclides other than those in Table 3.9.4 are detected and are the result of plant effluents, the potential annual dose to a member of the public must be less than or equal to the calendar year limits of Specifications 3.8.B, 3.8.E and 3.8.F. (b) Reporting level for drinking water pathways. For nondrinking water pathways, a value of 3 x 10' pCi/1 may be used. l (c) Reporting level for individual grab samples taken at North Storm Drain Outfall only. Amendment No. 83, M3 202 N_---. - -_ _ - _ -. -- __- _. - - - _ _ _ - _ _ - _ _ _ - - - _ _ _ - _ _ _ _ - _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ -- ._ - - - - - - - - - _
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-~~.-_.~.~ _._.~
VYNPS 4 TABLE 4.9.2 NOTATION (1): The' Instrument Functional Test shall also demonstrate that automatic isolation of this pathway and the Control Room alarm annunciation occurs if i any.of the following conditions exists: l- (a) Instrument indicates measured levels above the alarm setpoint. (b) Circuit failure. (c) Instrument indicates a downscale failure.
'(d) Instrument controls not set in operate mode.
J (2)' The Instrument Functional Test shall also demonstrate that Control Rocm alarm annunciation occurs when any of the following conditions exist: (a) Instrument indicates measured levels above the alarm setpoint. ' (b). Circuit failure. (c) . Instrument indicate: awnscale failure.
,(d) Instrument controls not set in operate mode.
(3) The'Instrumes.c Calibration for radioactivity measurement instrumentation ' 'shall include the use of a known (traceable to National Institute for Standards and Technology) radioactive source positioned in a reproducible '. geometry with respect to the sensor. These standards should permit j~ calibrating the system over its normal operating range of rate capabilities.
- ( 4 )' The Instrument Calibration shall include the u*e of standard gas samples 4
(high range and: low range) containing suitable concentrations, hydrogen
. balance nitrogen, for the detection range of interest per i Specification 3.8.J.l.
l l 1 I l 4 . 1 1 ' e 4
. Amendment No. S3,,464 206 l
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VYNPS BASES: 3.9 RACIOACTIVE EFFLUENT MONITORING SYSTEMS A. Liquid Effluent Instrumentation The radioactive liquid effluent instrumentation is_provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm setpoints for these instruments are to ensure that the alarm will occur prior to exceeding 10 times the concentration limits of Appendix B to 10CFR20.1001-20.2401, Table 2, Column 2, values. Automatic isolation function is not provided on the liquid radwaste discharge line due to the infrequent nature of batch, discrete volume, liquid discharges (on the order of once per year or less), and the administrative controls provided to ensure that conservative discharge-flow rates / dilution flows are set such that the probability of exceeding the above concentration limits are low, and the potential off-site dose consequences are also low. B. Gaseous Effluent Instrumentation The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm / trip setpoints for these instruments are provided to ensure that the alarm / trip will occur prior to exceeding design bases dose rates identified in 3.8.E.1. This instrumentation also includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the waste gas holdup system. C. Radiological Environmental Monitoring Program The radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures of member (s) of the public resulting from the station operation. This monitoring program implements Section IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels ] j of radiation are not higher than expected on the basis of the effluent i measurements and modeling of the environmental exposure pathways. Ten years of plant operation, including the years prior to the implementation of the Augmented Off-Gas System, have amply demonstrated via routine effluent and environmental reports that plant effluent measurements and modeling of environmental pathways are adequately conservative. In all cases, environmental sample results have been two to three orders of magnitude less than expected by the model employed, thereby representing small percentages of the ALARA and environmental reporting levels. This radiological environmental monitoring program has therefore been significantly modified as provided for by Regulatory l Guide 4.1 (C.2.b), Revision 1, April 1975. Specifically, the air 1 particulate and radioiodine air sampling periods have been increased to l semimonthly, based on plant effluent and environmental air sampling data for the previous ten years of operation. An I-131 release rate trigger value of 1 x 10 4 uCi/sec from the plant stack will require that air sample collection be increased to weekly. The Amendment No. 64, 164 209 < m
VYNPS 3.12 LIMITING CONDITIONS FOR 4.12 SURVEILLANCE REQUIREMENTS OPERATION E. Extended Core Maintenance E. Extended Core Maintenance One or more control rods may Prior to control rod be withdrawn or removed fror. withdrawal for extended core l the reactor core provided the maintenance, that control following conditions are rod's control cell shall be l satisfied: verified to contain no fuel assemblies.
- 1. The reactor mode switch 1. This surveillance shall be locked in the requirement is the same l
" Refuel" position. The as that given in refueling interlock which Specification 4.12.A.
prevents more than one control rod from being withdrawn may be bypassed on a withdrawn control rod after the fuel assemblies in the cell containing (controlled by) that control rod have been removed from the reactor core. All other refueling interlocks j shall be operable, j l
- 2. SRMs shall be operable in the core quadrant where 2. This surveillance fuel or control rods are requirement is the same being moved, and in an as that given in adjacent quadrant. The Specification 4.12.B.
requirements for an SRM to be considered operable are given in Specification 3.12.B.
- 3. If the spiral unload / reload method of
" ore alteration is to be used, the following conditions shall be met:
- a. Prior to spiral unload and reload, the SRMs shall be proven operable as stated in Specification 3.12.B1 and 3.12.B2.
However, during spiral unloading, the count rate may drop below 3 cps. Amendment No. 44, 69, 44, 146 233
VYNPS BASES: 3.12 & 4.12 REFUELING A. During refueling operations, the reactivity potential of the core is being altered. It is necessary to require certain interlocks and restrict certain refueling procedures such that there is assurance that inadvertent criticality does not occur. To minimize the possibility of loading fuel into a cell containing no control rod, it is required that all control rods are fully inserted when fuel is being loaded into the reactor core. This requirement assures that during refueling the refueling interlocks, as designed, will prevent inadvertent criticality. The core reactivity lindtation l of Specification 3.3 limits the core alterations to assure that the resulting core loading can be controlled with the Reactivity Control System and interlocks at any time during shutdown or the following operating cycle. The addition of large amounts of reactivity to the core is prevented by operating procedures, which are in turn backed up by refueling interlocks on rod withdrawal and movement of the refueling platform. When the mode switch is in the " Refuel" position, interlocks prevent the refueling platform from being moved over the core li a control rod is withdrawn and fuel is on a hoist. Likewise, if the refueling platform is over the core with fuel on a hoist, control rod motion is blocked by the interlocks. With the mode switch in the refuel position, only one control rod can be withdrawn. B. The SRMs are provided to monitor the core during periods of station shutdown and to guide the operator during refueling operations and station startup. Requiring two operable SRMs it or adjacent to any core quadrant where fuel or control rod, are being moved assures adequate monitoring of that quadrant during such alterations. The
- equirement of 3 counts per second provides assurance that neutron flux is being monitored. Under the special condition of complete spiral core unloading, it is expected that the count rate of the SRMs will drop below 3 cps before all the fuel is unloaded. Since there will be no reactivity additions, a lower number of counts will not present a hazard. When all of the fuel has been removed to the spent fuel storage pool, the SRMs will no longer be required.
Requiring the SRMs j to be operational prior to fuel removal assures that the SRMs are < operable and can be relied on even when the count rate may go below 3 cps. Prior to spiral reload, two diagonally adjacent fuel assemblies, which have previously accumulated exposure in the reactor, will be loaded into their designated core positions next to each of the 4 SRMs to obtain the required 3 cps. Exposed fuel continuously produces neutrons by spontaneous fission of certain plutonium isotopes, photo fission, and photo disintegration of deuterium in the moderator. This neutron production is normally great enough to meet the 3 cps minimum SRM requirement, thereby providing a means by which SRM response may be demonstrated before the spiral reload begins. During the spiral reload, the fuel will be loaded in the reverse sequence that it was unloaded with the exception of the initial eight (8) fuel assemblies which are loaded next to the SRMs to provide a means of SRM response. Amendment No. M, M, -74 237
i VYNPS i 3.13 LIMITING CONDITIONS FOR 4.13 SURVEILLANCE REQUIREMENTS OPERATION
- 1) The batteries, l cell plates and l battery racks l show no visual indication of physical damage or abnormal I deterioration, and
- 2) The battery-to-battery and terminal connections are clean, tight, l free of corrosion and l I
coated with anti-corrosion material. C. Fire Hose Stations C. Fire Hose Stations
- 1. Except as specified in 1. Each fire hose station 3.13.C.2 below, all hose shall be verified to be stations inside the operable Reactor Building, Turbine Building, and a. At least monthly by those inside the visual inspection of Administration Building the station to which provided coverage assure all equipment of the Control Room is available.
Building shall be operable whenever b. At least once each equipment in the areas 18 months by protected by the fire removing the hose hose stations is for inspection and required to be operable. replacing degraded coupling gaskets and
- 2. With one or more of the reracking.
fire hose stations specified in 3.13.C.1 c. At least once each above inoperable, route year by an additional equivalent hydro-statically capacity fire hose to testing each outside the unprotected area (s) hose at 250 lbs. from an operable hose station within one hour. d. At least once per 3 years by hydro-statically testing inside hose at 150 lbs. Amendment No. 43, 9 244
V'INPS 4 3.13 LIMITING CONDITIONS FOR 4.13 SURVEILLANCE REQUIREMENTS OPERATION
- c. At least once per 3 years by performing an air flow test through the Recirculation M.G. Set foam header and verifying each foam nozzle is unobstructed.
a e Amendment No. g 249 4
, _ .. . . - . _ _ _ _ . _ _ . . . _ _ _ _ . . - - - . _ . _ - _ - . _ . . _ . - . . . ~ . - - - - _ . . - . _ _ _ . . _ . . _
4 VYNPS TABLE 3.13.A.1 FIRE DETECTION SENSORS e Minimum No. of Sensors Required to Be Operable Sensor Location Heat Flame. Smoke
- 1. Cable Spreading Room & Station. Battery Room - -
23
- 2. Switchgear Room (East) - -
10
- 3. Switchgear Room-(West) - -
10
- 4. " Diesel Generator Room (A) - -
2
'5. Diesel Generator Room (B)- - -
2 2
- 6. Intake Structure (Service Water) 1 1 1
. 7. Recire Motor Generator Set Area 3 -
8
. 8.a Control Room Zone 1 (Control Room Ceiling) - -
14 9.b Control Room Zone 2 (Control Room Panels) - - 18 8.c Control Room Zone 3 (Control Room Panels). - - 25 8.'d -Control Room Zone 4 (Control Room Panels) - - 10 8.e Control Room Zone 5 (Exhaust & Supply - - 2 Ducts) 9.a Rx Bldg. Corner Rm NW 232 - - 1 L' 9.b . Rx' Bldg.: Corner Rm NW 213 (RCIC) - - 1 9.c Rx. Bldg. Corner Rm NE 232 - - 1 9.d ' .Rx Bldg. Corner Rm NE 213 - - 1 9.e Rx Bldg. Corner Rm SE 232 - - 1 9.f Rx Bldg. Corner Rm SE 213 - - 1
,' 9.g Rx_ Bldg. Corner Rm SW 232 - -
1 J
- 10. HPCI Room - -
8
- 11. . Torus area 12 -
16 l i
, 12. Rx Bldg. Cable Penetration Area - -
7 l
- 13. Refuel' Floor -~ -
13 ; i 14 . - Diesel Oil Day Tank Room (A) - 1* 1*
- 15. Diesel Oil Day Tank Room (B) -
1* 1*
-16. Turbine Loading Bay (vehicles) -
3 - i4 4
- NOTE: The Diesel Day Tank Rooms require only one detector operable (1 flame i or 1 smoke).
i 4 j g Amendment No.-44, 47. 250 4 # - + - , ,+ , . r w .-,e ., - , - -,--,e _ , - - , _ . - - - - -.----.-r . ---.---}}