ML20028C855
| ML20028C855 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 01/10/1983 |
| From: | Heider L VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20028C856 | List: |
| References | |
| FVY-83-1, NUDOCS 8301140235 | |
| Download: ML20028C855 (4) | |
Text
VERMONT YANKEE NUCLEAR POWEll CORPORATION eroposed Change No. 103 2.C.2.1 RD 5, Box 169 Ferry Road, Brattleboro, VT 05301 ncety ro FVY 83-1 y
ENGINEERING OFFICE gw 1671 WORCESTER ROAD FRAMINGHAM, MASSACHUSETTS 01701 T EL E PHONE 8 t F-872-8100 January 10, 1983 United States Nuclear Regulatory Commission Wa shing ton, D. C. 20555 Attention:
Office of Nuclear Reactor Regulation
References:
(a) License No. DPR-28 (Docket No. 50-271)
(b) Letter, VYNPC to USNRC, FVY 81-14, dated January 27, 1981 (c) Letter, VYNPC to USNRC, WVY 80-49, dated March 17, 1980, Proposed Change No. 79 (d) Letter, USNRC to VYNPC, dated December 9,1980, " Generic SP.R for BWR Scram Discharge Volume"
Subject:
Analog Trip System and Scram Discharge System Modifications
Dear Sir:
Pursuant to Section 50.59 of the Commission Rules and Regulations, Vermont Yankee Nuclear Power Corporation hereby proposes the following changes to Appendix A of the operating license.
PROPOSED CHANCE Changes to the Technical Specif L:ations are being proposed to reficct the following modifications:
1.
Pages 19, 22, 25, 28, and 47 provide revised Limiting Conditions for Operation, Surveillance Requirements, and Bases for the scram discharge volume instrumentation inputs to the Reactor Protection System and the Control Rod Block Actuation. Pages 47 and 48 are being revised to clarify / correct Table 3.2.5 notation.
.QD I 2.
Pages 25, 50 through 54, and 57 provide revised Surveillance
\\N/
6,C lq {
Requirements for the drywell pressure instrumentation inputs to the Reactor Protection and Emergency Core Cooling Systems, and reactor pressure instruments for the Emergency Core Cooling, low pressure
,00 permissive.
8301140235 030110 PDR ADOCK 05000271 P
As s
4 h
. S. Nuclear Regulatory Commission '
Page.2
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The proposed Technical Specification chan8es are included 'as an.
attachment to this letter.-
. RE ASON FOR CHANCE Based on our; review of the Brown's Ferryf partial failure to scram event-
- and our scram discharge / instrument volume system, we concluded [ Reference. (b)]
that certain modifications should be implemented at Vermont Yankee to further enhance system availability and reliability. As a result of these.
modifications, we. find that. changes-to Technical Specifications are necessary.
In addition, we are proposing changes to : Technical Specifications to-reflect the replacement of existing pressureiswitches with analog loops;. this upgrade is expected to increase plant.raliability, reduce setpoint drif t, and improve overall plant safety.
BASIS FOR CHANGE Thel scram discharge volume is being modified' to provide two independent instrument volumes. Each volume will be monitored by four~1evel transmitters. These transmitters will be fed into monitoring cabinets wherein analog to digital trip units will provide reactor _ scram _ signals, when the appropriate level is reached in either instrument volume..These analog instrument channels will replace the float switches utilized in the current -
design.. In addition to the scram signals described above, one of the '
transmitters from each instrument volume will' also feed a separate analog to digital trip unit which will provide a _ signal to the rod block actuation system.
The numerical trip point for the high scram discharge volume water level scram will-be lower than previous. This is due to the configuration of the new piping.
However, since only half as many control rod drive units discharge into each volume, the reduced capacity will not create spurious signals. The numerical trip point for the control rod block will remain the same as it has been. This instrumentation has been designed to meet criteria developed by the BWR Owners Group and previously endorsed by the NRC in Reference (d).
The surveillance and calibration requirements have been modified to agree with those which have previously been imposed upon similar analog instrumentation utilized at Vermont Yankee. Replacement of the existing pressure switches-with more reliable instrumentation will improve plant reliability, while at the same time not change the design basis, protective function, redundancy, trip point, or logic of the original system.
~ S AFETY CONSIDERATIONS The addition of these Technical Specifications governing scram discharge volume and RPS/ECCS instrumentation provide further assurance of improved operation and reactor safety because:
~~
- U. S. - Nuclear Regulctery Commie icn /
Page 3
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1.
The equipment will: perform its safety. function in. spite of any
-single active failure in the' instrumentation system,t or the' plugging
~of any' single instrument line.-
2._
The equipment lis environmentally and seismically. qualified for the environmental conditions under which it must' ope' rate.
3.
The equipment continuously monitors system operating conditions and provides alarms which annunciate.in the Main Control Room for both l
process and instrumentation malfunctions.-
4.-
The equipment is designed to be functionally tested under power c
operation.
- 15.. Operating experience with similar equipment at Vermont Yankee and -
~
other nuclear power facilities indicates that the calibration drif t-and mechanical problems associated with mechanical float switches are not experienced with this equipment.
6.
The overall accuracy and repeatability of the analog instrumentation are.significantly better 'than those of ~ the equipment it_ will replace.
All changes in the numerical trip levels have been made in-the conservative direction.
The surveillance and calibration requirements described for 'this instrumentation are identical to those previously imposed upon similar equipment currently installed at Vermont Yankee.. See Reference (c).
In light of the above discussion, this change to Technical Specifications does not impair.the safety of the general public in that it does not:
1.
increase the probability of occurrence of a previously evaluated
- accident, 2.
create the possibility of an unreviewed accident, or 3.
reduce the margin of safety as defined in Technical Specifications.
This. proposed change has been analyzed to assure that it does not create any unreviewed safety questions as defined in 10CFR50.59(a)(2). - This proposed -
change has been reviewed by the Vermont Yankee Nuclear Safety Audit Review Committee.
FEE DETERMINATION This proposed change requires an approval that involves a single safety issue and is being made to enhance overall reactor safety. For these reasons, Vermont Yankee Nuclear Power Corporation proposed this change as a Class III Amendment.
A payment of $4,000 is enclosed.-
5
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-., - - -, - - - = - -,
U2fS, Nucirr Regulctcry Cosmiccion.
Page 4 SCHEDULE OF dHANGE The plant modification to _ the Control Rod Drive Scram Discharge and RPS/ECCS Systems will be implemented during the 1983 refueling outage. In order for Vermont Yankee to perform this modification in an orderly manner and incorporate these changes before return to power from the 1983 refueling outage, it is necessary that we receive your approval by Februa'ry 1983.
Very truly yours,
' VERMONT ANKEE NUCLEAR POWER CORPORATION P / 'll J
L. H. Heider Vice President JBS/dd COMMONWEALTH OF MASSACilUSETTS)
)ss MILDLESEX COUNTY
)
Then personally appeared before me, L. II. Heider, who, being duly sworn, did state that he is a.Vice President of Vermont Yankee Nuclear Power Corporation, that he is duly authorized to execute and file the foregoing request in the name and on the behalf of Vermont Yankee Nuclear Power Corporation and that the statements therein are true to the best of his knowledge and belief.
5).8. L,M Cd. B. Sinclair fetary Public My Commission Expires June 1, 1984 l
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VYNPS-TABLE 3.1.1 REACTOR PEOTECTION SYSTEM (SCRAM) INSTRUMENT REQUIREMENTS 85 bO Modes in Which Minimun Number-Required Conditions When Functions Must Be Operating Instrument Minimum Conditions For-Om Operating Channels Per Operation Are Not'
@gTripFunction Trip Settings Refuel W Startup Run Trip System (2)
Satisifed(3)'
o-M 1.. Mode Switch X
X X
1 A
2*
In Shutdown
!2. Manual Scram X
X X
1 A
3.
IRM High Flux
$120/125 X
X X(ll).
2
~ A INOP X
X X(ll) 2 A
'\\
4 APRM High Flux 10.66W+54%C4)
X 2.
- A or B (Flow Bias)
High Flux 115%
X X
.2 A
(Reduced)
INOP X
2(5)
A or B Downscale 22/125 I
2-A or B 5.
High Reactor
$1055 psig X
X X
2 A
Pressure
.l 6.
High Drywell 12.5 psig X
X-X 2-A Pressure 7.
Reactor Low 2127.0 inches (6)
X X
X 2
A Water Level 8.
Scram Discharge 1 21 gallons X
X X
2~
A
-[
Volume High (per volume)
.l' Level 19'
VYNPS TABLE 4.1.1 SCRAM INSTRUMENTATION AND LOGIC SYSTEMS FUNCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTRUMENTATION, LOGIC SYSTEMS AND CONTROL' CIRCUITS ak Instrument Channel Group (3)
Functional Test (7)
Minimum Frequency (4)
Mode Switch in Shutdown A
Place Mode Switch in Shutdown Each Refueling Outage '
Trip Channel and Alarm Every 3 Months IRM High Flux C
Trip Channel and Alarm (5)
Before Each Startup & Weekly During Refueling (6)
Inoperative C
Trip Channel and Alarm Before Each Startup(6)&Weekk During Refueling APRM High Flux B
Trip Output Relsys(5)
Once Each Week High Flux (Reduced)
B Trip Output Relays (5)
Before Each Startup & Weekly -
.During Refueling (6)
Inoperative B
Trip Output Relays Once Each Week Downscale B
Trip Output Relays (5)
Once Each Week Flow Bias B
Trip Output Relays (5)
(1)
High Reactor' Pressure B
Trip Channel and Alarm (5)
(1)
High Drywell Pressure B
Trip Channel and Alarm (5)
(3) l 4
1 Low Reactor Water Level (2)(8)
B Trip Channel and Alarm (5)
- (1)
High Water Level in Scram Discharge B
Trip Channel and Alarm (5)
(3) j' Volume High Main Steam Line Radiation (2)
B Trip Channel and Alarm (5)
Once Each Week'
~
Main Steam Line Iso. Valve Closure A
Trip Channel and Alarm (1)
Turbine Con. Valve Fast Closure A
Trip Channel and Alarm (1)
Turbine Stop Valve Closure A
Trip Channel and Alarm
. (1) 22 _-
1 VYNPS TABLE 4.1.2 SCRAM INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Instrument Channel Group (I)
Calibration Standard (4)
Minimus Frequency (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 Source Refueling Outage LPRM 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 B
Water Level Once/ Operating Cycle Discharge 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 Source (3)
Refueling Outage-First Stage Turbine Pressure A
Pressure Source Every 6 Months and Pe rmis sive After Refueling.
Main Steam Line Isolation A
(6)
Refueling Outage Valve Closure 25
VYNPS 3.1 (continued)
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 (pressure switches) is provided to detect a loss-of-coolant accident and initiate the core standby cooling equipment. This instrumentation is a backup to the water level instrumantation which is discussed in Specification 3.2.
The Control Rod Drive Scram System is designed so that all of the water that is discharged from the reactor by the scram can be accommodated in the discharge piping. This discharge piping is divided into two sections. One section services the control rod drives on the north side of the reactor, the other serves the control rod drives of the south side. A part of the piping in each section is an instrument volume which accommodates in exrass of 21 gallons of water and is at the low point in the piping. No credit was taken for this volume.in the
~
design of the discharge piping as concerns the amount of water which must be accommodated during a scram.
During normal operation, the discharge volume is empty; however, should it fill with water, the water discharged to the piping from the reactor could not be accommodated, which would result in slow scram times or partial or ro control rod insertion. To preclude this occurrence, level instrumentation has been provided for the instrument volume which scram the reactor when the volume of water reaches 21 gallons. As indicated above, there is sufficient volume in the piping to accommodate the scram without impairment of the scram times or amount of insertion of the control rods. This function shuts the reactor down while sufficient volume remains to accommodate the discharged water, and precludes the situation in which a scram would be required but.not be able to perform its function adequately. The present design af the Scram Discharge System is in concert with the BWR Owner's Group criteria, which have previously been endorsed by the NRC in their generic " Safety Evaluation Report (SER) for Scram Discharge Systems", dated December 1, 1980.
Loss of condenser vacuum occurs when the condenser can no longer handle the heat input. Loss of condenser vacuum initiates a closure of the turbine stop valves and turbine bypass valves which eliminates the heat input to the condenser. Closure of the turbine stop and bypass valves causes a pressure transient, neutron flux rise, and an increase in surface heat flux. To prevent the clad safety limit from being exceeded if this occurs, a reactor scram occurs on turbine stop valve closure. The turbine stop valve closure scram function alone is adequate to prevent the clad safety limit from being exceeded in the event of a turbine trip transient without bypass.
28 t
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%.' G VYNPS TABLE 3.2.5 i
CONTROL ROD BLOCK INSTRUMENTATION Minimum Number of Operable Instrument Modes in Which Function Channels Per Trip Must 1: Operable System Trip Function Refuel Startup Run Trip Setting Startup Range Monitor 2
a.
Upscale (Note 2)
X X
< 5 x 105 2
b.
Detector Not Fully Inserted X
X cp, 3ng, 3)
Intermediate Range Monitor (Note 1) 2 a.
Upscale X
X
. <108/125 Full Scale 2
b.
Downscale (Note 4)
X X
75/125 Full Scale 2
c.
Detector Not Fully Inserted I
X
~
Average Power Range Monitor 2
a.
Upscale (Flow Bias)
I
<0.66W + 42% (Note 5)(Note 8)
T2 2
b.
Downscale X
_ /125 Full Scale
~
Rod Block Monitor (Note 6)
(N:te 10) 1 a.
Upscale (Flow Bias) (Note 7)
I
< 0.66W + N (Note 5) 1 b.
Downscale (Note 7)
X
> 2/125 Full Scale J
(Note 9) 1 Scram Discharge Volume X
X X-
<12 Callons (per volume) 1 Trip System Logic X
X X
47 1
VYNPS TABLE 3.2.5 NOTES 1.
There shall be two operable or tripped trip systems for each function in the required operating mode.. If; the -
minimum number of operable instruments are not.available for one of the two trip systems, this condition may exist.for up to seven days provided that during the_ time the operable system is functionally tested immediately and daily thereafter; if the condition lasts longer than seven days, the system shall be tripped. If.the minimum number of instrument channels are not available for both trip systems, the systems shallibe tripped.
2.
One of these trips may be bypassed. The SRM function may be bypassed in the higher IRM ranges when the INN upscale rod block is operable.
3.
This function may be bypassed when count rate is >100 cp, or when all IRM range switches. are above. Position 2.
~
i
. 4 IRM downscale may be bypassed when it is on its lowest scale.
5.
"W" is percent rated drive flow where 100% rated drive flow is that flow equivalent-to 48 x 106 lbs/hr core finw. Refer to LOO 3.11.C for acceptable values for N.
E 6.
The minimum number of operable instrument channels may-be reduced by one for. maintenance.and/or testing:for
i' periods not in excess of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 30-day period.
- 7.
The. trip may be bypassed when the reactor power is < 30%' of rated. An RBM channel will be considered inoperable
't if there are less than half the total number of normal inputs from any LPRM level..
/
8.
For special stability tests, the APRM rod ' block shall' be < 0.66W + 75% for the duration ^ of - testing..
~
19.
With the number. of operable channels less than required by the minimum operable channels per trip function.
requirement, place the inoperable channel in the tripped condition within one hour.
10.
With one RBM channel inoperable:
~
'i-Verify that the reactor is not operating on a limiting. control rod pattern, and a.
b.
Restore the inoperable RBM channel to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
_.n Otherwise, place the inoperable rod block monitor channel in the tripped condition within 'the.next hour.
^
48 o
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4 4
viv
W fl o;
VYNPS TABLE 4.2.1 MINIMUM TEST AND CALIBRATION FREQUENCIES EMERGENCY CORE COOLING ACTUATION INSTRUMENTATION Core Spray System Trip Function Functional Test (8)
Calibration (8)
Instrument Check High Drywell Pressure (Note 1)
Once/ Operating Cycle Once Each Day l
Low-Low Reactor Vessel (Note 1)
Once/ Operating Cycle Once Each Day Water Level low Reactor Pressure (Note 1)
Once/ Operating Cycle
'l Pump 14-1A, Discharge Press (Note 1)
Every 3 Months Auxiliary Power Monitor (Note 1)
Every Refueling Once Each' Day Pump Bus Power Monitor (Note 1)
None Once Each Dsy High Sparger Pressure (Note 1)
Every 3 Months.
Trip System Logic Except Every 6 Months Every 6 Months Relays 14A-K11A (Note 2)
(Note 3) 14A-K11B 14A-K19A 14A-K19B
'50
j VYNPS TABLE 4.2.1.(continued) 1 Low Pressure Coolant Injection System
. Trip Function Functional Test (8)
Calibration (8)
Instrument Check' Low Reactor Pressure #1 (Note 1)
Once/ Operating Cycle High Drywell Pressure #1 (Note 1)
Once/ Operating Cycle Once Each Day.
Low-Low Reactor Vessel (Note 1)
Once/ Operating Cycle' Once Each Day Water Level Reactor Vessel Shroud Level (Note 1)
Every 3 Months Low Reactor Pressure #2 (Note 1)
Once/ Operating Cycle
~l RRR Pump Discharge Pressure (Note 1)
Every 3 Months High Drywell Pressure #2 (Note 1)
Every 3 Mon'ths Low Reactor Pressure #3 (Note 1)
Every 3 Months Auxiliary Power Monitor (Note 1)
Every Refueling Outage Once Each Day Pump Bus Power Monitor (Note 1)
None Once Each Day.
LPCI Crosstie Monitor None None' Once Each Day Trip System Logic Every 6 Months Every 6 Months-(Note 2)
(Note 3) 51 l
MWPS TABE 4.2.1 (continued) t High Pressure Coolant Injection System Trip Function Functional Test (8)
Calibration (8)
. Instrument Check Low-Low Reactor Vessel (Note 1)
Once/ Operating Cycle Once Each Day Water Level Low Condensate Storage Tank (Note 1)
Every 3 Months Water Level High Drywell Pressure (Note 1)
Once/ Operating Cycle Once Each Day High Suppression Chamber (Note 1)
Every 3 Months Water Level Bus Power Monitor (Note 1)
None Once Each Day Trip System Logic Every 6 Months Every 6 Months (Note 2).
(Note 3)-
52-
- p i
VYNPS TABLE 4.2.1 (continued)
Automatic Depressurization System Trip Function Functional Test (8)
Calibration (8)
Instrument Check-low-Low Reactor Vessel (Note 1)
Once/ Operating Cycle.
Once Each Day
. Water Level l
High Drywell Pressure (Note 1)
Once/ Operating Cycle Once Each Day.
l Bus Power Monitor (Note 1)
None
.Once Each. Day Trip System Logic Every 6 Months Every 6 Months 4
(Except Solenoids of (Note 2)
(Note 3)
Valves) i i
1
. 53
.. - ~ _ _ _.
VYNPS TABLE 4.2.2 MINIMUM TEST AND CALIBRATION FREQUENCIES PRIMARY CONTAINMENT ISOIJLTION INSTRUMENTATION i.
l l
l Trip Function Functional Test (8)
Calibration (0)
Instrument Check Low-Low Reactor Vessel (Note 1)
Once/ Operating Cycle-Once Each. Day Water Level g
High Steam Line Area (Note 1)
Each Refueling Outage 1
Temperature
- High Steam Line Flow (Note 1)
Every 3 Months.
Once Each Day Low Main Steam Line Pressure (Note 1)
Every 3 Months Iow Reactor Vessel Water (Note 1)
Once/ Operating Cycle Level High Main Steam Line Radiation (Notes 1 & 7)
Each Refueling Outage Once Each Day High Drywell Pressure (Note 1)
Once/ Operating Cycle OncelEach Day l
Condenser Low Vacuum (Note 1)
Every 3 Months.
Trip System Logic Every 6 Months Every 6 Montha Except Relays 16A-K13 (Note 2)
(Note 3) 16A-K14 16A-K15 16A-K16 16A-K26 16A-K27
'54
TABLE 4.2.3 MINIMUM TEST AND CALIBRATION FREQUENCIES REACTOR BUILDING VENTILATION AND STANDBY GAS TREATMENT SYSTEM ISOLATION Trip Function Functional Test (8)
Calibration (8)
Instrument Check Low Reactor Vessel Water (Note 1)
Once/ Operating Cycle Level High Dryvell Pressure (Note 1)
Once/ Operating Cycle
'l Reactor Building Vent Monthly Every 3 Months Once.Each Day Exhaust Radiation Refueling Floor Zone Monthly Every 3 Months Once Each Day Radiation During Refueling Reactor Building Vent Every 6 Months Every 6 Months Trip System Logic (Note 2).
(Note 3) i Standby Gas Treatment Every 6 Months Every 6 Months Trip System Logic (Note 2)
(Note 3)
Iogic Bus Power Monitor (Note 1)
None Once Each Day i
1 7
57 l
a