ML20101L157
| ML20101L157 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 06/29/1992 |
| From: | Barrett R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20101L160 | List: |
| References | |
| NUDOCS 9207060297 | |
| Download: ML20101L157 (20) | |
Text
- -______ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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i UNITED STATES t
i NUCLEAR REGULATORY COMMISSION s,
f W ASHINoTON, D.C. 7E6
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COMMONWEALTH EDISON COMPANY QQLIT NO. 50-232 DRESDEN NUCLEAR POWER STATION. UNIT 2 AMENDMENT TO FACLLITY OPERATLNG LICENEL A:nendment No.115 License No. DPR-19 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Commonwealth Edison Company (the licensee) dated October 14, 1991, as supplemented February 6, 1992, complies with the standards and requirements of the Aton:ic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR 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.
2.
Accordingly, the license is amended by changes to the Technical Specifi-
/ations as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-19 is hereby amended to read as follows:
9207060297 920629 PDR ADOCK 05000237 P
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' (2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 115, are hereby incorporated in the license.
The licensee shall operate the facility in accordance
)
with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Cf W
Richard J. Barrett, Director Project Directorate !!!-2 Division of Reactor Projects - Ill/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
June 29, 1992
~4 ATTACHMENT TO LICENSE AMENDMENT NO. 115 FACILITY OPERATING LICENSE NO. DPR-19 DOCKET NO. 50-237 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the ettached pages.
The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.
REMOVE g1[_RJ r
1/2.1-5 1/2.1-5 B 1/2.1-15 B 1/2.1-15 3/4.1-5 3/4.1-5 3/4.1-7 3/4.1-7 3/4.1-10 3/4.1-10 B 3/4.1-16 B 3/4.1-16 B 3/4.1-20 8 3/4.1-20 i
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DRESDEN 11 DPR-19 Amendment No. 2Z, A3, 82, 115 1.1 SAFETY LIMIT (Cont'd.)
2.1 LIMITING SAFETY SYSTEM SETTING (Cont'd.)
E.
Turbine stop valve scram shall be less than or equal to 10%
valve closure from full open.
F.
Generator Load Rejection Scram 4
which initiates from actuation of the fast acting solenoid valve pressure switches shal', be greater than or equal to 460 psig EHC fluid pressure.
G.
Main Steamline Isolation Valve Closure Scram shall be less than or equal to 10%
valve closure from full open.
H.
Main Steamline Pressure initiation of main steamline isolation valve closure shall be greater than or equal to 850 psig.
I.
Turbine Control Valve fast Closure Scram on loss of control oil pressure shall be set at greater than or equal to 900 psig.
1/2.1-5
ORESCEN 11 DPR-19 A..iendment No. 58, 75, 82, 95, 115 s
4 2.1 LIMITING SAFETY SYSTEM SETTING BASES (Cont'd.)
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.
Tu lower the setpoint for initiation of the ECCS could lead to a loss of effective core cooling.
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 Scram - The turbine stop valve closure scram trip anticipates the pressure, neutron flux and heat flux increase that could result from rapid closure of the turbine stop valves.
With a scram trip setting of 10 percent of valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR remains above the MCPR fuel cladding integrity safety limit, even during the worst case transient that assumes the turbine bypass is closed.
T.
Generator Load Rejection Scram - The generator load rejection scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection and subsequent failure of the bypass; i.e., it prevents MCPR from becoming less than the MCPR fuel cladding integrity safety limit for this transient.
For the load rejection without bypass transient from 100%
power, the peak heat flux (and therefore LHGR) increases on the order of 15% which provides wide margin to the value corresponding to fuel center-line melting and 1% cladding strain.
Tne trip setpoint of greater than or equal to 460 psig EHC fluid pres-sure was developed to ensure that the pressure switch is actuated prior to closure of the turbine control valves (at approximately 400 psig EHC fluid pressure) yet assure that the system is not actuated unnecessarily due to EHC system pressure transients which may cause EHC system pressure to momentarily decrease.
G.
Reactor Coolant low Pressure Initiates Main Steam Isolation Valve-Closure - The low pressure isolation at 850 psig was provided to give protection against fast reactor depressurization and the resulting rapid cooldown of the vessel.
Advantage was taken of the scram feature which occurs when the main steam line isolation valves are closed to provido for reactor shutdown so that operation at pressures lower than those specified in the thermal hydraulic safety limit does not occur, although operation at a pressure lower than 850 psig would not necessarily constitute an unsafe condition.
H.
Main Steam Line Isolation Valve Closure Scram - The low pressure isolation of the main steam lines at 850 psig was provided to give protection against rapid reactor depressurization and the resulting rapid cooldown of the vessel.
Advantage was taken of the scram
-feature which occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit.
Operation of the B 1/2.1-15
~.
ORESDD II.
DPR-19 Amendment No. N, 2N,115 TABLE 3.1.1 l
REACTOR PROTECTION SYSTEM (SCRAM) INSTRUPENTATION REQUIROENTS i
l
Minimum Number Modes in Which Function Operable last..'
Must be Operable Channeh '~T Mp Startup/ Hot (1) Sysp -
Trip Function Trip level Setting Refuel (6)
-Standby Run Action
- i.
-~
Mode Switch in Shutdown X
X X
A I.
I Manual Scrae X
X
.X A
1 IRM 3
High Flux (LT/E) 120/125 X
X N/A A
of Full Scale 3
Inoperative X
X N/A A
APRM 5
2 High Flux Specification 2.1.A.1 X
X(8)
X A or B 2
Inoperative **
X X(8)
X A or B l
2 High Flux (15% Scram)
Specification 2.1.A.2 X
X N/A
.A 4
2 High Reactor Pressure (LT/E) 1060 psig X(10)
X X
A j
2 High Drywell Pressure (LT/E) 2 psig X(7). X(9)
X(7),(9) X(9)
A 2
Reactor Low Water Level (GT/E) I inch ***
X X
X A
2
. High Water Level in (LT/E) 40 inches above X(2)
X X
A or D (Per Bank)
Scram Discharge Volume-bottom of the Instrument (Thersal and dP Switch)
Volume 2
Turbine Condenser Low (GT/E) 23 in. Hg Vacuum X(3)
X(3)
X A or C l
Vacuum i
2 Main Steam Line High (LT/E) 3 X Normal X
X X(11)
A or C l
Radiation Full Power Background i-4(5)
Main Steam Line-(LT/E) 10% Valve Closure X(3)
X(3)
X A or C' Isolation Valve Closure 2
Generator Load (GT/E) 460 psig****
X(4)
X(4)
X(4)
A or C j
Rejection, turbine 1-control valve tr'p j
systee oil pressure j
low l
2 Turtine Stop Valve (LT/E) 10% Valve Closure X(4)
X(4)
X(4)
A or C j
Closure 2
Turbine Control -
(GT/E) 900 psig X(4)
X(4)
X(4)
A or C Loss of Control Oil Pressure Notes: (LT/E) = Less than or equal to.
i.
(GT/E) = Greater than or equal to.
1 (Notes continue on next tuo pages)
d DRESDEN !!
DPR-19 Amendment No. E2, $$, J0p,115 NOTES:
(For Table 3.1.1 Cont'd) i Required changes in Main Steam Line Radiation Monitor trip setting will be 1
made within 24 h*=
except during controlled power descensions at which time the setpoins.hange will be made prior to going below 20% power.
If due to a recirculation pump trip or other unanticipated power reduction event the reactor is below 20% power without the setpoint change, control rod motion will be suspended until the necessary trip setpoint adjustment is made.
If the first column cannot be met for one of the trip systems, that trip system shall ta tripped.
If the first column cannot be met for both trip systems, the appropriate actions listed below shall be taken:
a.
Initiate insertion of operable rods and complete insertion of all operable rods within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, b.
Reduce power level to IRM range and place mode switch in the Startup/ Hot Standby position within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, c,
Reduce turbine load and close main steam line isolation valves within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
d.
In the refuel mode, when any control rod is withdrawn, suspend all operations involving core alterations and insert all insertable control rods within one hour.
An APRM will be considered inoperable if there are less than 2 LPRM inputs per level or there are lest than 50% of the normal complement of LPRM's to an APRM.
1 inch on the water level instrumentation is greater than nr equal to 504" above vessel zero (see Bases 3.2).
- Trip is indicative of turbine control valve fast closure (due to low EHC fluid pressure) as a result of fast acting solenoid valve actuation.
3/4.1-7
DRESDEN II DPR-19 Amendment No. 82, 84, 115 TABLE 4.1.2 SCRAM INSTRUMENTATION CALIBRATIONS MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Instrument Channel Group (1)
Calibration Test Minimum Frequency (2)
- High Flux IRM C
Comparison to APRM Every shutdown (4) after Heat Balance i
High flux AFRM Output Signal B
Heat Balance Once Every 7 Days Flow Bias B
Standard Pressure Refueling Outage i
and Voltage Source High Reactor Pressure A
Standard Pressure Source Every 3 Months High Drywell Pressure A
Standard Pressure Source Every 3 Months Reactor Low Water B
Water Level (5)
Level Turbine Condenser A
Standard Vacuum Source Every 3 Months low Vacuum Main Steam Line B
Standard Current Every 3 Months High Radiation Source (3)
Turbine Control -
A Pressure Source Every 3 Months Loss of Control Oil Pressure High Water Level in A
Water Level Once per Refueling Scram Discharge Outage Volume (dp only)
Gen m tor Load A
Pressure Source Once per Refueling Rejection Outage NOTES:
(For Table 4.1.2) 1.
A description of the three groups is included in the bases of this Specification.
2.
Calibration tests are not required when the systems are not required to be operable or are tripped.
If tests are missed, they shall be performed prior to returning the systems to an operable status.
3.
The current source provides an instrument channel alignment.
Calibration using u radiation source shall be made during each refueling outage.
- 4.
If reactor startups occur more frequently than once per week, the functional test need not be performed; i.e., the maximum functional test frequency shall be once per week.
S.
Trip units are calibrated monthly concurrently with functional testing (staggered one channel out of 4 every week).
Transmitters are calibrated once per operating cycle.
3/4.1-10
5 DRESDEN 11 DPR-19 Amendment No. S2, 115 4.1 SURVEILLANCE REQUIREMENT BASES (Cont'd.)
of the sensor failure rate and the test interval.
A three-month test interval was planned for group (A) sensors.
This is in keeping with good opcrating practice, and satisfies the design goal for the logic configuration utilized in the Reactor Protection System.
To satisfy the long-term objective of maintaining an adequate level of safety throughout the plant lifetime, a minimum goal of 0.9999 at the 95% confidence level is proposed.
With the (1 ot:t of 2) X (2) logic, this requires that each sensor have an availability of 0.993 at the 95% confidence level.
This level of availability may be maintained by adjusting the test interval as a function of the observed failure history (See Reference 6).
To facilitate the implementation of this technique, Figure 4.1.1 is provided to indicate an appropriate trend ja t.'st interval.
The procedure is as followe,:
1.
Like sensors are pooled into one group for the purpose of data acquisition.
2.
The factor N is the exposure hours and is equal to th<
number of sensors in a group, n, times the elapsed time T (M = nT).
3.
The umulated number of unsafe failures is plotted as an ordinate against M as an abscissa on Figure 4.1.1.
4.
After a trend is established, the appropriate monthly test interval to satisfy the goal will be the test interval to the left of the plotted points.
5.
A test interval of one month will be used initially until a trend is established.
The turbine control valve fast acting solenoid valve pressure switches directly measure the trip oil pressure that causes the turbine control valves to close in a rapid manner.
The reactor scram setpoint was developed in accordance with NEDC-31336, " General Electric Instrument Setpoint Methodology" dated October 1986.
As part of the calculation, a specific calibration period is utilized to achieve a nominal trip point and an allcwable setpoint (Technical Specification value).
The nominal setpoint is procedurally controlled.
Based on the calculation input, the calibration period is defined to be every Refueling Outage.
Group (B) devices utilize an analog sensor followed by an amplifier and a bi-stable trip circuit.
The sensor and amplifier are active components and a failure is almost always accompanied by an alarm and an indication of the source of trouble.
In the event of failure, repair or substitution can start immediately.
An "as-is" failure is one that " sticks" Reference 6:
Reliability of Engineered Safety Features as a Function of Testing Frequency, I.M. Jacobs, Nuclear Safety, Vol. 9, No. 4, July-Aug. 1968, pp. 310-312.
B 3/4.1-16
DRESDEN 11 DPR-19 Amendment No. $5, 104, 115 4.1 SURVEILLANCE REQUIREMENT BASES (Cont'd.)
A comparison of Tables 4.1.1 and 4.1.2 indicates that six instrument channels have not been included in the latter Table.
These are:
Mode Switch in Shutdown, Manual Scram, High Water Level in Scram Dischargo-Volume dP and Thermal Switches, Main Steam Line Isolation Valve Closure, and Turbine Stop Valve Closure.
All of the devices or sensors assoc-l iated with these scram functions are simple on-off switches and, hence, calibration is not applicable; i.e., the switch 16 either on or off.
Further, these switches are mounted solidly to the device and have a very low probability of moving; e.g., the switches in the scram discharge volume tank.
Dased on the above, no calibration is required for these six instrument channels.
B.
The FDLRC shall be checked once per day to determine if the APRM gains i
or scram requires adjustment.
This may normally be done by checking the LPRM readings, TIP traces, or process computer calculations.
Only a small number of control rods are moved daily and thus the l
peaking factors are not expected to change significantly and thus a daiiy check of the FDLRC is adequate.
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B 3/4.1-20
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UNITED STATES c
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- ,I NUCLEAR REGULATORY COMMISSION o
WASHINo1ON. D.C. 20h66 i
COMMONWEALIlLED.1 SON COMPANY DOCKET.t(Q. 50-249 QBI1 DEN NUCLEAR POWER STATION. UNIT 3 AMENDMENT TO FACILITY OPERATING LICENS.E Amendment'No. 112 License No. DPR-25 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Commonwealth Edison Company (the licensee) dated October 14, 1991, as supplemented February 6, 1992, complies with the standards and requirements of the Atomic rules.and regulations set forth(the Act), and the Commission's Energy Act of 1954, as amended in 10 CFR Chapter I; B.
The facility will operate in conformity with the applicatie2, the provisions of the Act and the rules and regulations of the Commission; t
C.
There is reasonable assurance (i) that the activities authorszed 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 thc 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.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph-3.B. of Facility Operating License No. DPR-25 is hereby amended to read as follows:
i
., B.
Technical Sntstfications The Technical Specifications contained in Appendix A, as revised through Amendment No.112, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specificat:ois.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
.b k/)fM/V Richa d J. Barrett, Director Project Directorate 111-2 Division of Reactor Projects - Ill/IV/V Office of Nuclear Reactor Regulation Attac'nment:
Changes to the Technit. 1 Specifications Date of Issuance:
June 29, 1992 y
m-,
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-r2-y,---
6TTACHMENT TO LICENSE AMENDMENT NO. 112 FACILITY OPERATING LICENSE NO. DPR-21 QQ[KET NO. 50-249 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.
The revised pages are identified by the captioned amendment number and contain.aarginal lines indicating the area of change.
REMOVE LN11R1 1/2,1-5 1/2,1-5 B 1/2.1-15 B 1/2 1-15 3/4.1-5 3/4.1-5 3/4,1-7 3/4,1-7 3/4.1-10 3/4.1-10 B 3/4.1-16 B 3/4.1-16 8 3/4.1-20 B 3/4.1-20 l
1
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DRESDEN III DPR-25 Amendment No. #2, 52, 63, 75 112 1.1 SAFETY LIMIT (Cont'd.)
2.1 LIMITING SAFETY SYSTEM SETTING (Cont'd.)
E.
Turbine stop valve scram shall be less than or equal to 10%
valve closure from full open.
F.
Generator Load Rejection Scram which initiates from actua-tion of the fast acting solenoid valve pressure switches shall be greater than or equal to 460 psig CHC fluid pressure.
G.
Main Steamline Isolation Valve Closure Scram shall be less than or equal to 10%
valve closure from full open.
H.
Main Steamline Pressure initiation of main steamline isolation valve closure shall be greater than or equal to 850 psig.
I.
Turbine Control Valve Fast Closure Scram on loss of control oil pressure shall be set at greater than or equal to 900 psig.
1 l
1/2.1-5 t
l
DRESDEN ZII DPR-25 Amendment No. 75, A7, 112 2.1 LIMITING SAFETY SYSTEM SETTING BASES (Cont'd.)
The design of the ECCS components to meet the above criteria was dependent on three previously set parameters:
the maximum break cize, the low water level scram setpoint and the ECCS initiation setpoint.
To lower the setpoint for initiation of the ECCS could lead to a 1 css of effective core cooling.
To raise the ECCS initiation setpoint would be in a safe direction, but it would reduce the margin estab-lished to prevent actuation of the ECCS during normal operation or dering normally expected transients.
E.
Turbine Stop Valve Scram - The turbine stop valve closure scram trip anticipates 'he pressure, neutron flux and heat flux increase that could result from rapid closure of the turbine stop valves.
With a scram trip setting of 10 percent of valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR remains above the MCPR fuel cladding integrity safety limit, even during the worst case transient that assumes the turbine bypass is closed.
F.
Generator Load Rejection Scram - The generator load rejection scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection and subsequent failure of the bypass; i.e., it prevents MCPR from becoming less than the MCPR fuel cladding integrity safety limit for this transient.
For the load rejection without bypass transient from 100% power, the peak heat flux (and therefore LHGR) increases on the order of 15% which provides wide margin to the value corresponding to fuel centerline melting and 1% cladding strain.
The trip setpoint of greater than or equal to 460 psig EHC fluid pres-sure was developed to ensure that the pressure switch is actuated prior to closure of the turbine control valves (at approximately 400 psig EHC fluid pressure) yet assure that the system is not actuated unnecessarily due to EHC system pressure transients which may cause EHC system pres-sure to momentarily decrease.
G.
Reactor Coolant low Pressure Initiates Main Steam Isolation Valve Closure - The low pressure isolation at 850 psig was provided to give protection against fast reactor depressurization and the resulting rapid cooldowr, of the vessel.
Advantage was taken of the scram feature which occurs when the main steam line isolation valves are closed to provide for reactor shutdown so that operation at pressures iower than those specified in the thermal hydraulic safety limit does not occur, although operation at a pressure lower than 850 psig would not necessarily constitute an unsafe condition.
H.
Main Steam Line Isolation Valve Closure Scram - The low pressure isolation of the main steam lines at 850 psig was provided to give protection against rapid reactor depressurization and the resulting rapid cooldown of the vessel.
Advantage was taken of the scram feature which occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit.
Operation of the B 1/2.1-15
DRE5 DEN III DPR-25 Amendment No. C3.76. 112 TABLE 3.1.1 REACTOR PROTECTION 5l STEM (SCRAM) INSTRUMENTATION REQUIREMENTS Minimum Number Modes 17 Which Function Operable Inst.
Must be Operable Channels per Trip Startup/ Hot (1) System Trip Function Trip Level Setting Refuel (6)
Standby Run Action
- 1 Mode Switch in Shutdown X
X X
A 1
X X
A IRM 3
High flux (LT/E) 120/125 X
X N/A A
of Full Scale 3
Inoperative X
X N/A A
APRM 2
High Flux Specification 2.1.A.1 X
X(8)
X A or B 2
Inoperative **
X X(8)
X A or B 2
High Flux (ISK Scram)
Specification 2.1.A.2 X
X N/A A
2 High Reactor Pressure (LT/E) 1060 psig X(10)
X X
A 2
High Drywell Pressure (LT/E) 2 psig X(7),X(9)
X(7),(9)
X(9)
A 2
Reactor Low Water Level (GT/E) I inch ***
X X
X A
2 High Water Level in (LT/E) 37.25 inches above X(2)
X X
A or D (Per Bank)
Scram Discharge Volume bottom of the Instrument (rloat and dp Switch)
Volume 2
Turbine Condenser Low (Gi/E) 23 in. Hg Vacuum X(3)
X(3)
X A or C Vacuus 2
Main Steam Line High (LT/E) 3 X Normal X
X X
A or C Radiation full Power Background 4(5)
Main Steam Line (LT/E) 13% Valve Closure X(3)
X(3)
X A or C Isolation Valve Closure 2
Generator Load (GT/E) 460 psig****
X(4)
X(4)
X(4)
A or C Rejection, turbine control valve trip system oil pressure low 2
Turbine Stop Valve (LT/E) 10% Valve Closure X(4)
X(4)
X(4)
A or C i
Closure 2
Turbine Control -
(GT/E) 900 psig X(4)
X(4)
X(4)
A or C Loss of Control Oil Pressure Notes: (LT/E) = Less than or equal to.
(GT/E) = Greater than or equal to.
(Notes continue on next two pages) 3/4.1-5 i
ORESDEN III
- DPR-25 Amendment No. $$, $4,112 NOTES:
(For Table 3.1.1 Cont'd)
If the first column cannot be met for one of the trip systems, that trip system shall be tripped.
If the first column cannot be met for both trip systems, the appropriate actions listed below shall be taken:
a.
Initiate insertion of operable rods and complete insertion of all operable rods within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, b.
Reduce power level to IRM range and place mode switch in the Startup/ Hot Standby position within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
c.
Reduce turbine load and close main steam line isolation valves within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
d.
In the refuel mode, when any control rod is withdrawa, suspend all operations involving core alterations and insert all insertable control rods within one hour.
An APRM will be considered inoperable if there are less than 2 LPRM inputs per level or there are less than 50% of the normal complement of LPRM's to an APRM.
1 inch on the water level instrumentation is greater than or equal to 504" above vessel zero (see Bases 3.2).
- Trip is indicative of turbine control valve fast closure (due to low EHC fluid pressure) as a result of fast acting solenoid valve actuation.
3/4.1-7
- W-e
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DRESDEN III DPR-25 Amendment No. 75, 89, 112 TABLE 4.1.2 SCRAM INSTRUMENTATION CALIBRATIONS MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Instrument Cha'inci Group (1)
Calibration Test Minimum Frequency (2)
- High Flux IRM C
Comparison to APRM Every Shutdown (4) after Heat Balance High Flux APRM Output Signal B
Heat Balance Once Every 7 Days Flow Bias B
Standard Pressure Refueling Outage and Voltage Source High Reactor Pressure A
Standard Pressure Source Every 3 Months High Drywell Pressure A
Standard Pressure Source Every 3 Months Reactor Low Water B
Water Level (5)
Level Turbine Condenser A
Standard Vacuum Source Every 3 Months Low Vacuum Main Steam Line B
Standard Current Every 3 Months High Radiation Source (3)
Turbine Control -
A Pressure So'urce Every 3 Months Loss of Control Oil Pressure High Water Level in A
Water Level Once per Refueling i ram Discharga Outage volume (dp only)
Generator Load A
Pressure Source Once per Refueling Rejection Outage NOTES:
(For Table 4.1.2) 1.
A description of the three groups is included in the bases of this Specification.
2.
Calibration tests are not required when the systems are not required to be operable or are tripped.
If tests are missed, they shall be performed prior to returning the systems to an operable status.
3.
The current source provides an instrument channel alignment.
Calibration using a radiation source shall be made during each refueling outage.
- 4.
If reactor startups occur more frequently than once per week, the functional test need not be performed; i.e., the maximum functional test frequency shall be once per week.
5.
Trip units are calibrated monthly concurrently with functional testing (staggered one channel out of 4 every week).
Transmitters are calibrated once per operating cycle.
3/4.1-10
DRESDEN III DPR-25 Amendment No. 59, 75, 112 4.1 -SURVEILLANCE REQUIREMENT BASES (Cont'd.)
of the sensor failure rate and the test interval.
A three-month test interval was planned for group (A) sensors._ This is in keeping with good operating practice, and satisfies the design goal for the logic configuration utilized in the Reactor Protection System.
To satisfy the long-term-objective of maintaining an adequate level of safety throughout the plant lifetime, a minimum goal of 0.9999 at the 95% confidence level is proposed.
With the (1 out of 2) X (2) logic, this requires that each sensor have an availability of 0.993 at the 95% confidence level.
This level of availability may be maintained by adjusting the test interval as a function of the observed failure history (See Reference 6).
To facilitate the implementation of this technique, figure 4.1.1 is provided to indicate an appropriate trend in test interval.
The procedure is as follows:
1.
Like sensors are pooled into one group for the purpose of data acquisition.
2.
The factor M is the exposure hours and is equal to the number of sensors in a group, n, times the elapsed time T (M = nT).
3.
The accumulated number of unsafe failures is plotted as an ordinate against M as an abscissa on Figure 4.1.1.
4.
After a trend is established, the appropriate monthly test interval to satisfy the goal will be the test interval to the left of the plotted points.
5.
A test-interval of one month will be used initial?y until a trend is established.
The turbine control valve fast acting solenoid valve pressure switches directly measure the trip oil pressure that causes the turbine control valves to close in a rapid manner.
The reactor scram setpoint was developed in accordance with NEDC-31336, " General Electric Instrument Setpoint Methodology" dated October 1986, As part of the calculation, a_ specific cali-bration period is utilized to achieve a nominal trip point and an allowable setpoint (Technical Specification value).
The nominal setpoint is procedurally controlled.
Based on the calculation input, the calibration period is defined to be every Refueling Outage.
Group (B) devices utilize an analog sensor followed by an amplifier and a bi-stable trip circuit.
The sensor and amplifier are active components and a failure is almost always accompanied by an alarm and an indication of the source of trouble.
In the event of failure, repair or substitution can start immediately.
An "as-is failure is one that " sticks" a
Reference 6:
Reliability of Engineered Safety Features as a Function of Testing Frequency, j
I. M. Jacobs, Nuclear Safety, Vol. 9, No. 4, July-Aug.1968, pp. 310-312.
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i DRESDEN III DPR-25 Amendment No. 89, 94, 112 4.1 SURVEILLANCE REQUIREMENT BASES (Cont'd.)
A comparison of Tables 4.1.1 and 4.1.2 indicates that six instrument channels have not been included in the latter Table.
These are:
Mode Switch in Shutdown, Manual Scram, High Water Level in Scram Discharge Volume Float Switches, Main Steam Line Isolation Valve Closure, and Turbine Stop Valve Closure.
All of the devices or sensors associated I
with these scram functions are simple on-off switches and, hence, calibration is not applicable; i.e., the switch is either on or off.
Further, these switches are mounted solidly to the device and have a very low probability of moving; e.g., the switches in the scram discharge volume tank.
Based on the above, no calibration is required for these six instrument channels.
B.
The FDLRC shall be checked once per day to determine if the APRM gains or scram requires adjustment.
This may normally be done by checking in the LPRM readings, TIP traces, or process computer calculations.
Only a small number of control rods are moved daily and thus the peaking factors are not expected to change significantly and thus a daily check of the FDLRC is adequate.
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