ML20199E414
| ML20199E414 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 01/05/1999 |
| From: | Bill Dean NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20199E422 | List: |
| References | |
| NUDOCS 9901200419 | |
| Download: ML20199E414 (64) | |
Text
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. - maat
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r VERMONT YANKEE NUCLEAR POWER CORPORATION DOCKET NO. 50-271 VERMONT YANKEE NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE l
Amendment No.164 i
l License No. DPR-28
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment filed by the Vermont Yankee Nuclear Power Corporation (the licensee) dated November 3,1998, as supplemented by letter dated December 15, 1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR l
Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
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9901200419 990105 PDR ADOCK 05000271 P
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, 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-28 is hereby amended to read as follows:
(B) Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.164, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effecSva et of its date of issuance, to be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/
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William M. Dean, Director Project Directorate 1-2 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: January 5, 1999 l
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i ATTACHMENT TO LICENSE AMENDMENT NO.164 FACILITY OPERATING LICENSE NO. DPR-28 DOCKET NO. 50-271 Replace the fo!!owing pages of Appendix A Technical Specifications with the attached pages.
The revised pages are identified by amendment number and contain vertical lines indicating the j
areas of change.
i Remove Insert Remove IDIE1 il ii 63 63 71a
)
17 17 20 20 74 74 21 21 Bia 81a a
22 22 88 88 1
23 23 90 90 25 25 92 92 27 27 93 93 29 29 105 105 33 33 106 106 34 34 107 107 35 35 108 108 36 36 118 118 38 38 119 119 39 39 140 140 40 40 142 142 41 41 142a 142a 42 42 166a 166a 43 43 168 168 44 44 193 193 45 45 195 195 46 46 198 198 47 47 202 202
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O s.
O e.
Remove Insert Remove Insert 49 49 206 206 50 50 209 209 51 51 233 233 53 53 237 237 54 54 244 244 55a 249 249 56 56 250 250 57 57 t
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VYNPS l
TABLE OF CONTENTS (Continued)
LIMITING SAFETY Page No.
LYSTEM SETTING SAFETY LIMITS 1.1 FUEL CLADDING INTEGRITY.......................
6 2.1 i
1.2 REACTOR COOLANT SYSTEM........................
18 2.2 l
i LIMITING CONDITIONS OF OPERATION Page No.
SURVEILLANCE-
\\
3.1 REACTOR PROTECTION SYSTEM.....................
20 4.1 j
i i
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 i
and Standby Gas Treatment System Initiation................................
34 C
l 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
J.
(Deleted) 36 J
K.
Degraded Grid Protective System..........
36 K
L.
Reactor Core Isolation Cooling System Actuation.................................
37 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
i BASES 89 3.4 REACTOR STANDBY LIQUID CONTROL SYSTEM.........
92 4.4 i
A.
Normal Operation..........................
92 A
B.
Operation with Inoperable Components......
93 B
C.
Liquid Poison Tank - Boron Concentration.............................
93 C
BASES 97 Amendment No. 64, 46, 160 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 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 j
turbine stop valves. With a scram trip setting of <10% of valve closure l
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.
1 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 j
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 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 safety limit is provided by the IRM high neutron flux scram.
l 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.
l l
l Amendment No. M, M, 44,164 17
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4 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 l
systems required for reactor control systems required for safety.
reactor safety.
Objective:
Obiective:
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 operatien at any power A.
Instrumentation systems level shall be permitted in shall be functionally accordance with Table 3.1.1.
tested and calibrated as The systea response time from indicated in Tables 4.1.1 the opening of the sensor and 4.1.2, respectively contact up to and including the opening of the scram l
solenoid relay shall not i
exceed 50 nd111 seconds.
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.
a i
i Amendment No. 6% 164 20
-[
TABLE 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT REQUIREMENTS Required Modes in Which Functions Must Conditions be Operating Minimum Number When Minimum 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
Shutdown (SA-S1) 2.
X X
1 A
(5A-S3A/B) 3.
IRM ( 7-41 ( A-F) )
High Flux
$120/125 X
X X (11) 2 A
INOP X
X X (11) 2 A
4.
<0. 66 (W-AW) +54 %
X 2
A or B (flow bias)
(4)
High Flux (reduced)
<15%
X X
2 A
INOP X
2(5)
A or B Downscale
>2/125 X
2 A or B 5.
High Reactor 51055 psig X
X X
2 A
i Pressure l
(PT-2-3-55 ( A-D) (M) )
6.
High Drywell
$2.5 psig X
X X
2 A
f Pressure l
l (PT-5-12 ( A-D) (M) )
t 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
Volume High Level (per volume)
I (LT-3-231 ( A-H) (M) )
i Amendment No. 24, 44, 64, 68, %, M, M, 90, 94, 164 21
e-O VYNPS d
1 TABLE 3.1.1.
(Cont'd)
REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT REQUIREMENTS
[
Required Conditions Minimum Number When Minimum
}
l Modes in Which Functions Must Operating Conditions For Trip Settings be Operating Instrument Operation And Allowable Channels Per Are Not l
Trip Function Deviations Refuel (1)
Startup Run Trip System (2)
Satisfied (3)
-l 9.
Main steamline high 3x normal X
X X
2 A or C radiation (7) background at (RM-17-251(A-D))
rated power (8) 10.
Main steamline
<10% valve X
4 A or C i
isolation valve closure t
closure (POS-2-80A-A1,B1 POS-2-86A-A1,B1
{
POS-2-80B-A1,B2 POS-2-86B-A1,B2 POS-2-80C-A2,B1 j
POS-2-86C-A2,B1 l
POS-2-80D-A2,B2 i
POS-2-86D-A2,B2) 11.
Turbine control (9) (10)
X 2
A or D valve fast closure l
(PS-(37-40))
{
12.
Turbine stop valve
$10% valve (10)
X 2
A or D closure closure I
(SVOS-5-(1-4))
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t l Amendment No.164 22 f
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VYNPS d
TABLE 3.1.1 NOTES 1.
When the reactor is suberitical and the reactor water temperature is less i
than 212'r, only the following trip functions need to be operable:
l a) 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 systen output relay shall be tripped innediately.
Except for MSIV and j
l 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 l
be tripped.
If the requirements cannot be met for both trip systems, the l
appropriate actions. listed below shall be taken:
a)
Initiate insertion of operable rods and complete insertion of all l
operable rods within four hours.
l b)
Reduce power level to IRM :
s and place mode switch in the i
i "Startup/ Hot Standby" posit _un within eight hours.
l l
c)
Reduce turbine load and close main steam line isolation valves within 8 i
hours.
d)
Reduce reactor power to less 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.
l This difference must be accounted for during single loop operation. AW = 0 1
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 l
additional LPRM input and still be considered operable.
l 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 abnornal radiation levels in prinary coolant.
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l Amendment No.
G+, 24, 68, 44, 94, 164 23
--v--<
. s TABLE 4.1.1 SCRAM INSTRUMENTATION AND LOGIC SYSTEMS FUNCTIONAL TESTS MINIMUM EVNCTIONAL TEST FREQUENCIES FOR SAFETY INSTRUMENTATION, LOGIC SYSTEMS AND CONTROL CIRCUITS Instrument Channel Group"'
Functional Testl73 Minimum Frequencyl8I Mode Switch in Shutdown A
Place Mode Switch in Shutdown Each Refueling Outage Manual Scram A
Trip Channel and Alarm Every 3 Months IRM High Flux C
Trip Channel and Alarm:51 Before Each Startup & Weekly During Refuelingts)
Inoperative C
Trip Channel and Alarm Before Each Startup & Weekly During Refuelingt61 APRM High Flux B
Trip Output Relaystsl Once Each Week High Flux (Reduced)
B Trip Output Relays ts)
Before Each Startup & Weekly During Refuelingl61 Inoperative B
Trip Output Relay 3 Once Each Week Downscale B
Trip Output Relays ts)
Once Each Week Flow Bias B
Trip Output Relays ts) til High Reactor Pressure B
Trip Channel and Alarmis) gil High Drywell Pressure B
Trip Channel and Alarml51 ul Low Reactor Water Levelt21 el B
Trip Channel and Alarml5I til High Water Level in Scram Discharge B
Trip Channel and Alarml53 til volume High Main Steam Line Radiationt23 B
Trip Channel and Alarmt5)
Once Each Week Main Steam Line Iso. Valve Closure A
Trip Channel and Alarm til Turbine Con. Valve Fast Closure A
Trip Channel and Alarm til Turbine stop Valve Closure A
Trip Channel and Alarm til l Scram Test Switch ( SA-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. M, M,
-58, %, 164 25 mi i
n N
t VYNPS t
I TABLE 4.1.2 SCRAM INSTRUMENT CALIBRATION L
MINIMUM CALIBRATION FREQUENCIES EUR 2 ACTOR PROTECTION INSTRUMENT CHANNELS Instrument Channel GroupIII Calibration Standard (4)
Minimum Frequency (2) 1 I
High Flux APRM Output Signal B
Heat Balance Once Every 7 Days l
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 j
Power Hours High Reactor Pressure B
Standard Pressure Source Once/ Operating Cycle l
r Turbine Control Valve Fast Closure A
Standard Pressure Source Every 3 Months
)
t High Drywell Pressure B
Standard Pressure Tource Once/ Operating Cycle High Water Level in Scram Discharge B
Water Level Once/ Operating Cycle
[
Volume i
?
Low Reactor Water Level B
Standard Pressure Source Once/ Operating Cycle Turbine Stop Valve Closure A
(6)
Refueling Outage t
High Main Steam Line Radiation B
Appropriate Radiation Refueling Outage Source (3) j First Stage Turbine Pressure A
Pressure Source Every 6 Months and After l
Permissive (PS-5-14(A-D))
Refueling l
Main Steam Line Isolation Valve A
(6)
Refueling Outage i
Closure i
t
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f i
Amendment No. 44, M,
M,
-58, 64, -76, 16k 27
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VYNPS BASES:
l 3.1 Reactor Protection System 1
The reactor protection system automatically initiates a reactor scram to:
l 1.
preserve the integrity of the fuel barrier; i
2.
preserve the integrity of the primary system barrier; and l
3.
minimize 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 naintenance, testing, or calibration.
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 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 i
produce a reactor seren.
1 The required conditions when the minimum instrument logic conditions are i
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 nornal 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 minimum 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 fcr inputs to each trip system. Additional IRM channels have also been provided to allow bypassing of one such channel.
IRM 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.
-l Instrumentation is provided to detect a loss-of-coolant accident and i
initiate the core standby cooling equipment.
This instrumentation is a i
backup to the water level instrumentation which is discussed in l
Specification 3.2.
Amendment No. G4: 164 29
s a
VYNPS BASES:
4.1 REACTOR PROTECTION SYSTEM A.
The scram sensor channels listed in Tables 4.1.1 and 4.1.2 are divided 3
into three groups:
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 1
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.
B.
The ratio of MFLPD 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 adequate.
Amendment No. 64, 64, 164 33
e o
a VYNPS I
3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION 3.2 PROTECTIVE INSTRUMENT SYSTEMS 4.2 PROTECTIVE INSTRUMENT SYSTEMS l
Applicability:
Applic bility:
Applies to the operational status Applies to the surveillance of the plant instrumentation requirements of the systems which initiate and instrumentation systems which control a protective function.
initiate and control a protective i
function.
Objective:
Objective:
To assure the operability of To verify the operability of protective instrumentation protective instrumentation systems.
systens.
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 l
3.5, the instrumentation which indicated in Table 4.2.1.
initiates the emergency core cooling system (s)shall be operable in accordance with Table 3.2.1.
i 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 I
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 Amendment No. 164 34
\\.
VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION 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.
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, l
line isolation valves are automatic isolation and l
open, the mechanical securing of the mechanical vacuum pump she?:' be vacuum pump shall be verified 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 l
be in the cold shutdown l
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
l l
G.
Post-Accident Instrumentation G.
Post-Accident Instrumentation I
i 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.
1 and control the systems used during and following a l
postulated accident or 4
abnormal operating condition shall be operable in accordance with Table 3.2.6.
l Amendment No. 9,.164 35 4
i.
i a
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l VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION H.
Drywell to Torus AP H.
Drywell to Torus AP Instrumentation Instrumentation
- 1. During reactor power The Drywell to Torus AP l
operation, the Drywell to Instrumentation shall be Torus AP Instrumentation calibrated once every six (recorder #1-156-3 and months and an instrument 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 I
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 l
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. M, M, M, M, M4, Me,164 36
L VYNPS TABLE 3.2.1 I
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION 1
1 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 i
2 High Drywell Pressure
<2.5 psig Note 2 i
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 l
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-14-44(A-D))
1 Auxiliary Power Monitor Note 5 l
l (LNPX C/D) 1 Pump Bus Power Monitor Note 5 l
(27/3A/B, 27/4A/B) t 1
Trip System Logic Note 5 i
i Amendment No. 44, 68, MO, 1-M, 440, 442, 164 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 cf 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 $ 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 382.5" above top of Note 2 Level (LT-2-3-72 ( A-D) (S1) )
enriched fuel 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 3100 psig Note 5 Pressure (PS-10-105(A-H))
2 High Drywell Pressure
$2.5 psig Note 2 I
(PT-10-101 ( A-D) (SI) )
Amendment No. 44, 44, 64, 444, M 3, 164 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 T.evel Setting Are Not Satisfied j
1 Time Delay (10A-K45A & B)
$6 minutes Note 5 2
Low Reactor Pressure 300 $ p $ 350 psig Note 2 i
(PT-2-3-56A/B(M) &
I l
PT-2-3-52C/D(M))
1 Auxiliary Power Monitor Note 5 I
(LNPX C/D) j-1 Pump Bus Power Monitor Note 5 I
(27/3A/B, 27/4A/B) s 1
Trip System Logic Note 5 r
t
?
i Amendment No. -14, 44O, 442, 164 40 i
VYNPS I
TABLE 3.2.1 (Cont'd)
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION i
High Pressure Coolant Injection System 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 (Note 3)
Low-Low Reactor Vessel Water Same as LPCI Note 5 l
Level (LT-2-3-72 (A-D) (SI) )
2 (Note 4)
Low Condensate Storage Tank
~> 3%
Note 5 l
Water Level (LSL-107-57)B) 2 (Note 3)
High Drywell Pressure Same as LPCI Note 5 (PT-10-101 (A-D) (M) )
1 (Note 3)
Bus Power Monitor (23A-K41)
Note 5 1 (Note 4)
Trip System Logic Note 5 2 (Note 7)
High Reactor Vessel Water
<177 inches above top of Note 5 l
Level (LT-2-3-72A/B) (S4 )
enriched fuel i
I l
t Amendment No. 68, 86, 90, 16k 41 i
VYNPS TABLE 3.2.1 (Cont'd)
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Automatic Depressurization Minimum Number of Required Action When Operable Instrument Minimum conditions Channels 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 i
l Level (LT-2-3-72 (A-D) (M) )
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 1
Bus Power Monitor (2E-KlA/B)
Note 6 1
Trip System Logic Note 6 i
2 Time Delay
$8 minutes Note 6 (2E-K16A/B, 2E-K17A/B) 2 i
i Amendment No. 44, MG, 16k 42
VYNPS TABLE 3.2.1 i
(Cont'd)
RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION 9
Recirculation Pusp Trip - 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
Low-Low Reactor vessel Water
> 6' 10.5" above top of Note 2 l
Level (LM-2-3-68(A-D))
enriched fuel 2
High Reactor Pressure
$ 1150 psig Note 2 (PM-2-3-54(A-D))
2
' Time Delays (2-3-68 (A-D) (K) )
$ 10 seconds Note 2 1
Trip Systems Logic Note 2
\\
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l t
t I
Amendment No. 68, 68, M, %o IN 43 l
m
VYNPS TABLE 3.2.1 NOTES 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.
l 4.
One trip system with initiating instrumentation arranged in a one-out-of-two logic.
5.
If the ndnimum 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, Specifications 3.5.F.3 shall apply.
7.
One trip system arranged in a two-out-of-two logic.
l i
l Amendment No. 33, 164 44
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 runction Trip Setting Not satisfied (Note 2) t r
2 Low-Low Reactor Vessel Water 382.5" 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 (121-124 ) (A-D) )
2/ steam line High Main Steam Line Flow
$140% of rated flow B
l (DPT-2-(ll6-119) (A-D) (' ) )
M 2/(Note 1)
Low Main Steam Line Pressure 2800 psig B
l (PS-2-134(A-D))
2/(Note 6)
High Main Steam Line Flow
$40% of rated flow B
(DPT-2-116A,ll7B, 118C,119D(SI))
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 Jrywell Pressure Same as Reactor A
Protection System 2/(Note 10)
Condenser Low Vacuum
$12" Hg absolute A
1 Trip System Logic A
l Amendment No. 9, 68, 84, 86, 90, 164 45
VYNPS 1
TABLE 3.2.2 (Cont'd) l HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION INSTRUMENTATION I
l r
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
-<212*F Note 3 Temperature (TS-23-(101-104) (B-D) )
1 High Steam Line d/p (Steam
$195 inches of water Note 3 Line Break) (DPIS-23-77/78) 4 (Note 5)
Low HPCI Steam Supply Pressure
>70 psig Note 3 l
-( PS-23-68 (A-D) )
i 2
Main Steam Line Tunnel
<212'F Note 3 1
Temperature (TS-23-(101-104)A) 1 Time Delay (23A-K48)
$35 minutes Note 3 (23A-K49) l 1
Bus Power Monitor (23A-K38) i 1
Trip System Logic
?
Amendment No. 68, 444s IN 46 i
e
'e
' i pg TABLE 3.2.2 (Cont'd)
REACTOR CORE ISOLATION COOLING 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 (Note 2) s
)
2 Main Steam Line Tunnel l
Temperature (TS (79-82) A)
-<212'F Note 3 1
Time Delay (13A-K41)
$35 minutes Note 3 (13A-K42) i 2 per set of 4 High Steam Line Space
-<212*F Note 3 Temperature 6
l (TS (79-82) (B, C, D) )
f 1
High Steam Line d/p (Steam
$195 inches of water Note 3 I
Line Break) (DPIS-13-83/84)
J 4 (Note'5)
Low Steam Supply Pressure
>50 psig Note 3 (PS-13-87(A-D))
f 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 i
i Amendment No. 69, 44-L s 47 l
e VYNPS TABLE 3.2.3 REACTOR BUILDING VENTILATION ISOLATION & STANDBY GAS TREATMENT SYSTEM INITIATION Minimum Number of Required Action When Operable Instrument Minimum Conditions channels per Trip For Operation System Trip Function Trip Setting Are Not Satisfied 2
Low Reactor Vessel Water Level Same as PCIS Note 1 (LT-2-3-57A/B(M),
LT-2-3-5BA/B(M))
2 High Drywell Pressure Same as PCIS Note 1 l
( PT-5-12 (A-D) (M) )
1 Reactor Building Vent
$14 mr/hr Note 1 I
(PM-17-452A/B) 1 Refueling Floor Zone Radiation
$100 mr/hr Note 1 I
(RM-17-453A/B) 1 Reactor Building Vent Trip Note 1 System Logic 1
Standby Gas Treatment Trip Note 1 System Logic 1
Logic Bus Power Monitor Note 1 (16A-K52/53)
Note 1 - If the minimum number of operable instrument channels is not available in either trip system for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor building ventilation system shall be isolated and the standby gas treatment system operated until the instrumentation is repaired.
Amendment No. 164 49
. o e
VYNPS TABLE 3.2.4 OFF-GAS SYSTEM ISOLATION INSTRUMENTATION Minimum Number of Required Action When Operable Instrument Minimum Conditions l
Channels per Trip For Operation i
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 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 continucusly.
I I
i-t l
Amendment No. 9, 83, 16k 50 i
i VYNPS TABLE 3.2.5 CONTROL ROD BLOCK INSTRUMENTATION Minimum Number of Modes in Which Function Operable Instrument Must be Operable Channels per Trip System Trip Function Refuel Startup Run Trip Setting Startup Range Monitor 5
Upscale (Note 2) (7-40 (A-D) )
X X
$5 x 10 cps (Note 3) a.
2 b.
Detector Not Fully Inserted X
X (7-11 (A-D) (LS-4 ) )
g Intermediate Range Monitor (Note 1) 2 a.
Upscale (7-41 ( A-F) )
X X
$108/125 Full 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-ll (E, F, G, H, J, K) (LS-4 ) )
Average Power Range Monitor l
(APRM A-F) 2 a.
Upscale (Flow Bias)
X
$0.66(W-AW)+42% (Note 5) 2 b.
Downscale X
>2/125 Full Scale i
l 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
~
X X
$12 Callons l
(Note 8)
(per (LT-3-231A/G (SI))
volume) 1 Trip System Logic X
X X
i Amendment No. 44, 26, 64, 66, -73, -76, 90, 94, 3% 164 51
VYNPS TABLE 3.2.6 POST-ACCIDENT INSTRUMENTATION Minimum Number of Operable. Instrument Channels Parameter Type of Indication Instrument Range 2
Drywell Atmospheric Recorder STR-16-19-45 0-350*F l
Temperature (Note 1)
(Blue)
Meter STI-16-19-30B 0-350*F 2
Containment Pressure (Note 1)
Meter IPI-16-19-12A
(-15) -(+260) psig Meter IPI-16-19-12B
(-15) -(+260) psig 2
Torus Pressure (Note 1)
Meter GPI-16-19-36A
(-15) - ( + 65 ) psig Meter IPI-16-19-36B
(-15) -(+65) psig 2
Torus Water Level (Note 3)
Meter ILI-16-19-12A 0-25 ft.
Meter SLI-16-19-12B 0-25 ft.
2 Torus Water Temperature Meter ITI-16-19-33A 0-250*F (Note 1)
Meter ITI-16-19-33C 0-250*F 2
Reactor Pressure (Note 1)
Meter IPI-2-3-56A 0-1500 psig Meter IPI-2-3-56B 0-1500 psig j
I 2
Reactor Vessel Water Level Meter ILI-2-3-91A
(-200)-0-(+200) "H O 2
(Note 1)
Meter ILI-2-3-91B
(-200)-0-(+200)"H O 2
2 Torus Air Temperature (Note 1)
Recorder ITR-16-19-45 0-350*F l
(Red)
Meter ITI-16-19-41 50-300*F 2/ valve Safety / Relief Valve Position Lights RV-2-71(A-D)
Closed - Open i
From Pressure Switches From PS-2-71-(1-3) (A-D)
(Note 4)
Amendment No. M, 6-3, M, 96, 4-1-3, 44r,, 164 53
VYNPS TABLE 3.2.6 (Cont'd)
POST-ACCIDENT INSTRUMENTATION Minimum Number of r
Operable Instrument Channels Parameter Type of Indication Instrument Range 1/ valve Safety Valve Position From Meter ZI-2-1A/B Closed - Open I
F Acoustic Monitor (Note 5) 2 Containment Hydrogen / Oxygen Recorder SR-VG-6A (SI) 0-30% hydrogen Monitor (Note 1)
Recorder SR-VG-6B (SII) 0-25% oxygen t
2 Containment High-Range Meter RM-16-19-1A/B 1 R/hr-10' R/hr Radiation Monitor (Note 6) i 1
Stack Noble Gas Effluent Meter RM-17-155 0.1 - 10' mR/hr (Note 7)
{
t I
Amendment No. 64, 90, 94, 98, IN 54 i
t
i
~
VYNPS I
l l
TABLE 3.2.7 l
(Table 3.2.7 was intentionally deleted from the Technical Specifications) i l
I Amendment No.
IN 55a
i l
VYNPS l
TABLE 3.2.8 l
EMERGENCY BUS UNDERVOLTAGE INSTRUMENTATION Minimum Number of Operable 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) 2 per bus Degraded Bus Voltage - Time 10 seconds i 1 second Note 2 Delay (62/3W, 62/3Z, 62/4W, 62/4Z)
TABLE 3.2.8 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
+
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I l
L i
k f
I Amendment No. 98, 164 56 i
F
.'i TABLE 3.2.9 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION Required Action When Minimum Number of' Minimum Conditions Operable Instrument-For Operation Channels per Trip System Trip Function 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 Fuel 2 (Note 2)
Low Condensate Storage Tank
>3%
Note 4 Water Level (LT-107-12A/B(M))
l 2 (Note 3)
High Reactor Vessel Water
<177" Above Top of Note 4 l
Level (LT-2-3-72C/D(S2))
Enriched Fuel Note 4
[
1 Bus Power Monitor (13A-K36) l 1
Trip System Logic Note 4 1
k k
i t
i 57 Amendment No. 44-1, 164
~
J VYNPS
,[
L TABLE 4.2.1 (Cont'd) t 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 l High Reactor Pressure (Notes 1 and 4)
Once/ Operating Cycle Once Each Day I
Trip System Logic Once/ Operating Cycle once/ Operating Cycle
[
s t
L
[
[
c i
f I
L i
t i
i i
i F
i Amendment No. 68, M6, 164 63
---w w
VYNPS TABLE 4.2.7 (Table 4.2.7 was intentionally deleted from the Technical Specifications) 71a l
Amendment No.164
,=
VYNPS 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS OPERATION With the required shutdown margin not met and the mode switch in the " Refuel" position, immediately suspend Alteration of the Reactor Core except for control rod insertion and fuel assembly removal; immediately initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies; within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, initiate action to restore the integrity of the Secondary Containment System.
2.
Reactivity Margin -
2.
Reactivity Margin -
Ynoperable Control Rods Inoperable Control Rods l
Control rod drives which Each partially or fully cannot be moved with withdrawn operable control rod drive control rod shall be pressure shall be exercised one notch at considered inoperable.
least 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> unless power operation is investigation.
continuing with one fully demonstrates that the or partially withdrawn
.cause of the failure is rod which car.not 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the number control rods shall be I
Amendment No. 444, 164 Bla t....
a VYNPS 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS 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 during configurations will be power operation shall not enmpared to the expected confj urations at selected exceed 14 Ak/k. If this g
limit is exceeded, tha operating conditions. These reactor will be shut down comparisons will be used an until the ceuse has been base data for reactivity 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
be nade at least every equivalent full power month.
I Amendment No. M, 444,164 88
=
=
- -. -... _ ~
~
o 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 of 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 witndrawal 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 it subcritical and all other rods are fully inserted. Above 20% power, the RWM is not needed since even with a single error an aperator 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 hus 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, l
therefore, two operable SRM's are specified for added conservatism.
l 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 develop due to the occurrence of inoperable control rods.
a d
i
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Amendment No. &&, 49, 64, go,164 90
m VYNPC e
3.4 LIMZTING CONDITIONS FOR 4.4 SURVEZLLANCE 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:
Obiective:
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 Control System shall be operable by:
operable during periods when fuel is in the reactor unlesti 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 I
35 gpm at 1275 psig 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 operable by:
3.
Testingthatthesettingl of the pressure relief I
valves is between 1400 l-and 1490 psig.
l 4.
Initiating one of the l
loops, excluding the l
primer chamber and inlet fitting, and verifying that a flow path from a pump to the reactor s
a Amendment No. 4GG, 434,164 92
VYNPS 3.4 LIMITING CONDITIONS FOR 4.4 SURVEILLANCE REQUIREMENTS 4
OPERATION i
vessel is available by l
pumping demineralized water into the reactor vessel.
Both loops shall be tested over the course of two operating cycles.
5.
Testing the new trigger l
assemblies by installing 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 demonstratvid to be operation is permissible operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
during the succeeding seven days unless such component is sooner made operable.
C.
Liquid Po,ison Tank - Boron C.
Liquid Poison Tank - Boron Concentration Concentration At all times when the i
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 cuct. ton solution in the standby piping shall be t:hecied liquid control tank at least daily, shall meet the requirements of Figure 3.4.1.
i Amendment No. W, 444, 164 93
)
~
_ -. _... _. _ =
VYNPS 1
3.5 LIMITING CONDITION FOR 4.5 SURVEILLANCE REQUIREMENT OPERATION j
3.
From and after the date 3.
When the Alternate that the Alternate Cooling Subsystem or both Cooling Tower Subsystem Station Service Water 8
or both Station Service Subsystems are made or Water Subsystems are made found to be inoperable, or found inoperable for the operable subsystem (s) l any reason, reactor shall have been or shall
}
operation is permissible be demonstrated to be only during the operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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 l
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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
E.
High Pressure Cooling E.
High Pressure Coolant Iniection (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 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 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.
Amendment No. M, M4, ag,164 105
\\
VYNPS 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 or found to be inoperable, the Automatic inoperable for any Depressurization System reason, reactor operation shall have been or shall is permissible only be demonstrated to be during the succeeding operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
seven days unless such subsystem is sooner made NOTE: Automatic operable, provided that Depressurization during such seven days System operability all active components of shall be the Automatic demonstrated by Depressurization performing a Subsystems, the Core functional test of Spray Subsystems, the the trip system LPCI Subsystems, and the logic.
3.
If the requirements of Specification 3.5.E cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall be reduced l
to 5 120 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
F.
Automatic Depressurization F.
Automatic Depressurization System System Surveillance of the Automatic l
Depressurization System shall i
be performed as follows:
l 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 System shall be operable at any time the reactor i
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 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, valve when the Amendment No. G4, 444, 448, 164 106
VYNPS 3.5 LIMITING CONDITION FOR 4.5 SURVEILLANCE REQUIREMENT 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 l
3.
If the requirements of Specification 3.5.F l
cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall be reduced l
to 5 100 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
G.
Reactor Core Isolation G.
Reactor Core Isolation Cooling System (RCIC)
Cooling System (RCIC)
Surveillance of the RCIC l
System shall be performed as follows:
1.
Except as specified in 1.
Testing Sp9cification 3.5.G.2 Item Frequency below, the RCIC System j
shall be operable whenever the reactor Simulated Each re-automatic fueling pressure is greater than 150 psig and irradiated actuation outage fuel is in the reactor test vessel.
(testing valve l
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. 44, 34, 444, 4G4 e 164 107
VYNPS 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 l
to s'120 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
H.
Minimum Core and Containment H.
Minimum Core and Containment Cooling System Availarility Cooling System Availability 1.
During any period when 1.
When one of the emergency l
l one of the emergency diesel generators is made diesel generators is or found to be i
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 days, provided that all operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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 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. G7, 444, 164 108
e VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION e.
With the radioiodine concentration in the reactor coolant greater than 1.1 microcuries/
gram dose equivalent I-131, a sample of reactor coolant shall be taken every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and analyzed 4
for radioactive 1
iodines of I-131 through I-135, until the specific j
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 Samho/cm l
Chloride ion 0.1 ppm 3.
For reactor startups the 3.
a.
With steaming rates maximum value for greater than or 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 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> and reactor coolant water for when the continuous the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after conductivity placing the reactor in monitors indicate the power operating abnormal condition, conductivity (other than short-term spikes), and analyzed for conductivity and chloride ion content.
Amendment No. G4,164 118
o o
l VYNPS I
3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION l
b.
When the continuous I
conductivity monitor is inoperable, a reactor coolant sample shall be l
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 j
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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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'F, not to exceed j
reactor coolant 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
leakage into the primary containment from unidentified sources shall not exceed 5 gpm.
In addition, the total reactor coolant system leakage into the primary i
containment shall not exceed 25 gpm.
l
- b. While in the run mode, reactor coolant leakage Into the primary containment from unidentified sources shall not Amendment No. 449, 164 lig
]
a VYNPS i
BASES:
3.6 and 4.6 (Cont'd)
The actual shift in RTug of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185 reactor vessel naterial irradiation surveillance specimens installed near the inside wall of the reactor vessel in the Since the neutron spectra at the irradiation samples and core area.
vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the 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 shif t in RTng is determined in accordance with Regulatory Guide 1.99, Revision 2.
The beatup and cooldown curves must be recalculated when the ARTng decermined from the surveillance capsule is different from the calculated ARTmv for the equivalent capsule radiation exposure.
The pressure-temperature lindt 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 radiciodine concentration limit of 1.1 pCi 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 l
lindt, as set forth in Specification 3.8.E.1, or there is a failure or 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 Ci 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 10~3 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 nass release of 30,000 pounds.
I i
l 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 l
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 radiciodine concentration would not be expected to change rapidly during steady-state operation over a period of 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.
In addition, the trend of the radioactive gaseous effluents, which is continuously monitored, is a good indicator of the trend of the radiciodine 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.
Amendment No. 44, 6B, 91, 93e 164 140
y VYNPS BASES:
3.6 and 4.6 (Cont'd) impurities will also be within their normal ranges. The reactor 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.l.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,"
have been determined. The results from the conductivity monitors on the feedwater can be correlated with the results from the conductivity monitors on the reactor coolant water to indicate demineralizer breakthrough and subsequent conductivity levels in the reactor vessel water.
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 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 compliance with the MCPR cafety 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 specified in Section III of the ASME Boiler and Pressure Vessel Code as 11% of set pressure.
However, the code allows a larger tolerance value for the as-found condition if the supporting design analyses demonstrate that the applicable acceptance criteria are met.
Safety analysis has been performed which shows that with all safety and safety relief valves within i3% 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 are not exceeded during the limiting overpressure transient.
Ch:.g: 16/" = M, 10H, M, M, MB, MG, MG,Mo,164 142 I
o VYNPS BASES:
3.6 and 4.6 (Cont'd)
E.
Structural 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
142a Amendment No. 444, 164
,4 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. 444, 164 166a
4 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 I
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.
As most leakage and deterioration of integrity is expected to occur through penetrations, especially those with resilient seals, a periodic leak rate test program of such penetration is conducted at the peak 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.
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 accident condition.
Operability testing is performed in conjunction I
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 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 specified. Also it is inspected to assure that it closes freely and operates properly.
The containment design has been examined to establish the allowable z
bypass area between the drywell and suppression chamber as 0.12 ft,
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 measured at the bottom of 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 by light indication from two independent position detection and indication systems.
Either system provides a control room alarm for a nonseated valve.
1 Amendment No, M, Me, -1Ms 160 168 1
i d
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) i 2.
Flow Rate Measurement Devices a.
Liquid Radwaste Discharge Flow Rate 1*
3,4 1
Monitor j
(FIT-20-485/442) i l
During releases via this pathway.
J I
Amendment No. 83s 164 193
4
,8 i
VYNPS TABLE 3.9.2 GASEOUS EFFLUENT MONITORING INSTRUMENTATION Minimum Channels Instrument Operable Notes 1.
Steam Jet Air Ejector (SJAE) a.
Noble Gas Activity Monitor 1
7, 8, 9 I
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 i
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-2008) c.
Hydrogen Monitor 1
3, 5, 6
)
(H2AN-OG-2921A/B, H2AN-OG-2922A/B) 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 c.
Particulate Sampler Filter 1
4, 5 d.
Sampler Flow Integrator 1
1, 5 (FI-17-156/157) e.
Stack Flow Rate Monitor 1
1, 5 (FI-108-22)
I Amendment No. 84, 44a, 164 195
VYNPS TABLE 3.9.3 (Cont'd)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGR.'W Exposure Pathway Numbec of Sample Sampling and Collection Type and Frequency and/or Sample Locationsa Frequency of Analysis b
40 routine monitoring Quarterly.
Gamma dose, at least once 2.
DIRECT RADIATION 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 release LCO was exceeded 4 km9; 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. 83, 164 198
VYNPS O
TABLE 3.9.4 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES'*8 Reporting Levels Airborne Particulate or Fish hilk Vegetation Sediment Analysis Water (pC1/1)
Gases (pCi/m3)
(pCi/Kg, wet)
(pC1/l)
(pCi/Kg, wet)
(pCi/Kg, dry)
H-3 2 x 10*
- 3 3 x 10*
Mn-54 1 x 10 2
1 x 10' Fe-59 4 x 10 3
3 x 10' Co-58 1 x 10 3#
2 1 x 10' 3 x 10 Co-60 3 x 10 2
2 x 10' Zn-65 3 x 10 2
Zr-Nb-95 4 x 10 2
I-131 0.9 3
1 x 10 3
3 Cs-134 30 10 1 x 10 60 1 x 10 3
3 Cs-137 50 20 2 x 10 70 2 x 10 2
2 Ba-La-140 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 condition 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 Sp7cifications 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.
(c)
Reporting level for individual grab samples taken at North Storm Drain Outfall only.
202 Amendment No. 83, M3, 164
4 a
VYNPS 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 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.
(2)
The Instrument Functional Test shall also demonstrate that Control Room alarm annunciation occurs when any of the following conditions exist:
(a) Instrument indicates measured levels above the alarm setpoint.
(b) Circuit failure.
(c) Instrument indicates a downscale failure.
(d) Instrument controls are not set in operate mode.
(3) The Instrument 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 calibrating the system over its normal operating range of rate capabilities.
(4)
The Instrument 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.l.
Amendment No. 44, 46b 164 206 i
l
e VYNPS BASES:
3.9 RADIOACTIVE EFFLUENT MONITORING SYSTEMS A.
Liquid Effluent Instrumentation The radioactive liquid efiluent 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 rauAcactive 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 radionuciddes which lead to the highest potential radiation exposures of member (s) of the public resulting from the station operation.
This monitoring program implementsSection 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 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 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 3, April 1975.
Specifically, the air particulate and radiciodine air sampling periods have been increased to 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 uCi/see from the plant stack will require that 4
air sample collection be increased to weekly. The Amendment No. 84, 464, 164 209
~
VYNPS I
i 3.12 LIMITING CONDITIONS FOR 4.12 SURVEILLANCE REQUIREMENTS OPERATION E.
Extended Core Maintenance E.
Extended Core Maintenance l
One or more control rods may Prior to control rod be withdrawn or removed from withdrawal for extended core l
the reactor core provided the maintenance, that control.
following conditions are rod's control cell shall be satisfied:
verified to contain no fuel l
assemblies.
1 1.
The reactor mode switch 1.
This surveillance shall be locked in the requirement is the same
" Refuel" position. The as that given in i
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 interlocke shall be.:pte ble.
1 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 core alteration is to be used, the following conditions shall be met:
a.
Prior to spiral unload and reload, l
the SRMs shall be i
proven operable as l
stated in Specification 3.12.B1 and 3.12.B2.
However, during l
spiral unloading, the count rate may drop below 3 eps.
d lbk 233 Amendment No. 44, 69, 44, 4448
e VYNPS BASES:
3.12 & 4.12 RETUELING 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 leading 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 limitation 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 if 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 block'4 by the interlocks. With the mode switch in the refuel position, only or..i 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 in or adjacent to any core quadrant where feel or control rods are being moved assures adequate monitoring 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 eps 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 fue:. has been removed to the spent fuel storage pool, the SRMs will r.o longer be required.
Requiring the SRMs to be operational prior to fuel removal assures that the SRMs are operable and can be relied on eren when the ecunt rate may go below 3 eps.
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 eps 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. 44, 69, 44, 164 237
- 3. l*!
LIMITING CONDITIONS FOR 4.13 SURVEILIANCE REQUIREMENTS OPERATION 1)
The batteries, cell plates and battery racks show no visual indication of physical damage or abnornal deterioration, and 2)
The battery-to-battery and terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material.
J C.
Fire Hose Stations C.
Fire Hose Stations i
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 Buildi.ng, and a.
At least monthly by those inside the visual inspection of Administration bcilding the station to which provided coverage assure all equipment of the Control Room is available.
l Building shall be operable whenever b.
At least once each i
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. 44, 64* 164 244
VYNPS 3.13 LIMITING CONDITIONS FOR 4.13 9URVEILLANCE 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.
Amendment No. W, 164
p RTYNPS l
TABLE 3.13.A.1 1
FIRE DETECTION SENSORS 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 10 3.
Switchgear Room (West) l 4.
Diesel Generator Room (A) 2 5.
Diesel Generator Room (B) 2 6.
Intake Structure (Service Water) 1 1
1 8
7.
Recirc Motor Generator Set Area 3
14 8.a Control Room Zone 1 (Control Room Ceiling) 8.b Control Room Zone 2 (Control Room Panels) 18 25 8.c Control Room Zone 3 (Control Room Panels) 10 8.d control Room Zone 4 (Control Room Panels) 2 l
8.e Control Room Zone 5 (Exhaust & Supply Ducts) 9.a Rx Bldg. Corner Rm NW 232 1
1 l
9.b Rx Bldg. Corner Rm NW 213 (RCIC) 9.c Rx Bldg. Corner Rm NE 232 1
1 9.d Rx Bldg. Cornet Rm NE 213 9.e Rx Bldg. Corner Rm SE 232 1
9.f Rx Bldg. Corner Rm SE 213 1
1 9.g Rx Bldg. Corner Rm SW 232 8
10.
HPCI Room l
11.
Torus area 12 16 l
12.
Rx Bldg. Cable Penetration Area 7
l 13.
Refuel Floor 13 l
1*
1*
14.
Diesel 011 Day Tank Room (A)
[
15.
Diesel 011 Day Tank Roca (B) 1*
1*
3 16.
Turbine Loading Bay (vehicles)
- NOTE:
The Diesel Day Tank Roons require only one detector operable (1 flame or 1 smoke).
Amendment No. 4a, sa,164 250