ML18038B404
| ML18038B404 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 08/29/1995 |
| From: | Hebdon F Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18038B405 | List: |
| References | |
| DPR-33-A-223, DPR-52-A-238, DPR-68-A-197 NUDOCS 9509060334 | |
| Download: ML18038B404 (39) | |
Text
0 CqQ P0 C
~f' O
Ich Cy Yg'+~
~O
++*++
~,p'V UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON' C
2055&0001 TENNESSEE VALLEY AUTHORITY DOCK T NO. 50-259 BROWNS FERRY NUCLEAR PLANT UNIT 1
AMENDMENT TO FACILITY OPERATING LICENSE.
Amendment No. 223 License No.
DPR-33 The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated May 11,
- 1995, and supplemented on June 30,
- 1995, 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 Chapter I; B.
The facility will operate in conformity with the application, the provisions'f 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.
q6090b0334 950829 PDR ADQCK 05000259 P
(
V
~ i~
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No.
DPR-33 is hereby amended to read as follows:
(2)
Technical S ecifications The Technical Specifications contained in Appendices A and B,
as revised through Amendment No. 223, 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 its date of issuance and shall be implemented within 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Project Directorate II-4 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
August 29, 1995
ATTACHMENT TO LIC NSE AMENDMENT NO. 223 CILI Y OPERATI G
IC NS NO.
DPR-33 DOCKET NO. 50-259 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.
The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.
A Spillover page is provided to maintain document completeness.
REMOVE 3.1/4.1-5 3.1/4.1-6 INSERT 3.1/4.1-5 3 1/4 1 6**
NOTES FOR ABLE 3 A
There shall be two OPERABLE or tripped trip systems for each function.
If the minimum number of OPERABLE instrument channels per trip system cannot be met for one trip system, trip the inoperable channels or entire trip system within one hour, or, alternatively, take the below listed action for that trip function. If the minimum number of OPERABLE I
instrument channels cannot be met by either trip system, the appropriate action listed below (refer to right-hand column of Table) shall be taken.
An inoperable channel need not be placed in the tripped condition where this would cause the trip function to occur.
In these
- cases, the inoperable channel shall be restored to OPERABLE status within two hours, or take the action listed below for that trip function.
A. Initiate insertion of OPERABLE rods and complete insertion of all OPERABLE rods within four hours.
In refueling mode, suspend all operations involving core alterations and fully insert all OPERABLE control rods within one hour.
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 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
D.
Reduce power to less than 30 percent of rated.
2.
The scram discharge volume high water level bypass may be used in SHUTDOWN or REFUEL to bypass the scram.discharge volume high-high water level scram signal in order to reset the reactor protection system trip.
A control rod withdraw block is present when this scram signal is bypassed.
3.
Bypassed if reactor pressure is less than. 1055 psig and mode switch not in RUN.
4.
Bypassed when turbine first stage pressure is less than 154 psig.
5.
IRMs are bypassed when APRMs are onscale and the reactor mode switch is in the RUN position.
6.
The design permits closure of any two lines without a scram being initiated.
7.
When the reactor is subcritical and the reactor water temperature is less than 212 F, only the following trip functions need to be OPERABLE:
A.
Mode switch in shutdown B.
Manual scram C.
High flux IRM BFH Unit 1 3.1/4.1-5 ARND11ENT NO. 223
NOTES ROR TABLE 1 A (Oaot'd)
D.
Scram discharge volume high level E.
Not required to be OPERABLE when primary containment integrity i.s not required.
9.
(Deleted) 10.
Not required to be OPERABLE when the reactor pressure vessel head is not bolted to the vessel.
The APRM downscale trip function is only active when the reactor mode switch is in RUH.
12.
The APRM downscale trip is automatically bypassed when the IRM instrumentation is OPERABLE and not high.
13.
Less than 14 OPERABLE LPRMs will cause a trip system trip.
14.
Channel shared by Reactor Protection System and Primary Containment and Reactor Vessel Isolation Control System.
A channel failure may be a channel failure in each system.
15.
The APRM 15 percent scram is bypassed in the RUH Mode.
16 'hannel shared by Reactor Protection System and Reactor Manual Control System (Rod Block Portion).
A channel failure may be a channel failure in each system. if a channel is allowed to be inoperable per Table 3.1.A, the corresponding function in that same channel may be inoperable in the Reactor Manual Control System (Rod Block).
17.
Not required while performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MH(t).
18.
This function must inhibit the automatic bypassing of turbine control valve fast closure or turbine trip scram and turbine stop valve closure scram whenever turbine first state pressure is greater than or equal to 154 psig.
19.
Action 1.A or 1.D shall be taken only if the permissive fails in such a manner to prevent the affected RPS logic from performing its intended function.
Otherwise, no action is required.
20.
(Deleted) 21.
The APRM High Flux and Inoperative Trips do not have to be OPERABLE in the REFUEL Mode if the Source Range Monitors are connected to give a noncoincidence, High Flux scram, at 5 x 10 cps.
The SRMs shall be 5
OPERABLE per Specification 3.10.B.1.
The removal of eight (8) shorting links is required to provide noncoincidence high-flux scram protection from the Source Range Monitors.
BFN Unit 1 3.1/4.1-6 ZKNDIKNTi%a.
223
~ y D,R REGIII Vp0 J I OO I
4p VJ+n
~O
+)t*++
UNITED STATES NUCLEAR REGULATORY COMMISSlON WASHINGTON, D.C. 2055&0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 238 License No.
DPR-52 The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated May 11,
- 1995, and supplemented on June 30,
- 1995, 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 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 Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No.
DPR-52 is hereby amended to read as follows:
3.
(2)
Technical S ecifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No. 238, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Frederick J.
Hebdo
, Director Project Directorate II-4 Division of Reactor Projects I/II Office of.Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
pu~~sg 29, ]995
ATTACHMENT TO LICENSE AMENDMENT NO.
FACILITY OPERATING LICENSE NO.
DPR-52 DOCKET NO. 50-260 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.
The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.
- Overleaf and **spillover pages are provided to maintain document completeness.
REMOVE 3.1/4.1-5 3.1/4.1-6 3.1/4.1-16 3.1/4.1-17 3.1/4.1-18 3.1/4.1-19 3.1/4.1-20 3.1/4.1-21 INSERT 3.1/4.1-5 3.1/4.1-6**
3.1/4.1-16*
3.1/4.1-17 3.1/4.1-18 3.1/4.1-19**
3.1/4.1-20*"
3.1/4.1-21*
NOTES FOR TABLE 1
There shall be two OPERABLE or tripped trip systems for each function.
If the minimum number of OPERABLE instrument channels per trip system cannot be met for one trip system, trip the inoperable channels or entire
)
trip system within one hour, or, alternatively, take the below listed action for that trip function. If the minimum number of OPERABLE instrument channels cannot be met by either trip system, the appropriate action listed below (refer to right-hand column of Table) shall be taken.
An inoperable channel need not be placed in the tripped condition where this would cause the trip function to occur.
In these
- cases, the inoperable channel shall be restored to OPERABLE status within two hours, or take the action listed below for that trip function.
A. Initiate insertion of OPERABLE rods and complete insertion of all OPERABLE rods within four hours.
In refueling mode, suspend all operations involving core alterations and fully insert all OPERABLE control rods within one hour.
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 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
D.
Reduce power to less than 30 percent of rated.
2.
The scram discharge volume high water level bypass may be used in SHUTDOWN or REFUEL to bypass both the scram discharge volume high-high water level and scram pilot air header low pressure scram signals in order to reset the reactor protection system trip.
A control rod withdraw block is present when these scram signals are bypassed.
3.
(Deleted) 4.
Bypassed when turbine first stage pressure is less than 154 psig.
5.
IRMs are bypassed when APRMs are onscale and the reactor mode switch is in the RUN position.
6.
The design permits closure of any two lines without a scram being initiated.
7.
When the reactor is subcritical and the reactor water temperature is less than 212oF, only the following trip functions need to be OPERABLE:
A.
Mode switch in SHUTDOWN B.
Manual scram C.
High flux IRM D.
Scram discharge volume high level BFN Unit 2 3.1/4.1-5 AIM)HEZZ NO 238
NOTES FOR TABLE.l.
(OooO'd)
S E.
Scram pilot air header low pressure 8.
Not required to be OPERABLE when primary containment integrity is not required.
9.
(Deleted) 10.
11.
Not required to be OPERABLE when the reactor pressure vessel head is not bolted to the vessel.
The APRM downscale trip function is only active when the reactor mode switch is in RUN.
12.
The APRM downscale trip is automatically bypassed when the IRM instrumentation is OPERABLE and not high.
13.
Less than 14 OPERABLE LPRMs will cause a trip system trip.
14.
Channel shared by Reactor Protection System and Primary Containment and Reactor Vessel Isolation Control System.
A channel failure may be a channel failure in each system.
15.
The APRM 15 percent scram is bypassed in the RUN Mode.
16.
Channel shared by Reactor Protection System and Reactor Manual Control System (Rod Block Portion).
A channel failure may be a channel failure in each system.
If a channel is allowed to be inoperable per Table 3.1.A, the corresponding function in that same channel may be inoperable in the Reactor Manual Control System (Rod Block).
17.
Not required while performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MH(t).
18.
This function must inhibit the automatic bypassing of turbine control valve fast closure or turbine trip scram and turbine stop valve closure scram whenever turbine first stage pressure is greater than or equal to 154 psig.
19.
Action 1.A or 1.D shall be taken only if the permissive fails in such a
manner to prevent the affected RPS logic from performing its intended function.
Otherwise, no action is required.
20.
(Deleted) 21.
The APRM High Flux and Inoperative Trips do not have to be OPERABLE in the REFUEL Mode if the Source Range Monitors are connected to give a noncoincidence, High Flux scram, at 5 x 10 cps.
The SRMs shall be 5
OPERABLE per Specification 3.10.B.l.
The removal of eight (8) shorting links is required to provide noncoincidence high-flux scram protection from the Source Range Monitors.
BFN Unit 3.1/4.1-6 DKNDHBFX NO. 238
3.1 BASES (Cont'd)
I be accommodated which would result in slow scram times or partial control rod insertion.
To preclude this occurrence, level switches have been provided in the instrument volume which alarm and scram the reactor when the volume of water reaches 50 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 discharge water and precludes the situation in which a scram would be required but not be able to perform its function adequately.
A source range monitor (SRM) system is also provided, to supply additional neutron level information during startup but has no scram functions.
Reference Section 7.5.4 FSAR.
Thus, the IRM is required in the REFUEL and STARTUP modes.
In the power range the APRM system provides required protection.
Reference Section 7.5.7 FSAR.
Thus, the IRM System is not required in the RUN mode.
The APRMs and the IRMs provide adequate coverage in the STARTUP and intermediate range.
The high reactor pressure, high drywell pressure, reactor low water level, low scram pilot air header pressure and scram discharge volume high level scrams are required for STARTUP and RUH modes of plant operation.
They are, therefore, required to be operational for these modes of reactor operation.
The requirement to have the scram functions as indicated in Table 3.1.A OPERABLE in the REFUEL mode is to assure that shifting to the REFUEL mode during reactor power operation does not diminish the need for the reactor protection system.
Because of the APRM downscale limit of y 3 percent when in the RUH mode and high level limit of g15 percent when in the STARTUP Mode, the transition between the STARTUP and RUH Modes must be made with the APRM instrumentation indicating between 3 percent and 15 percent of rated power or a control rod scram will occur.
In addition, the IRM system must be indicating below the High Flux setting (120/125 of scale) or a scram will occur when in the STARTUP Mode.
For normal operating conditions, these limits provide assurance of overlap between the IRM system and APRM system so that there are no "gaps" in the power level indications (i.e., the power level is continuously monitored from beginning of startup to full power and from full power to SHUTDOWH).
When power is being reduced, if a transfer to the STARTUP mode is made and the IRMs have not been fully inserted (a maloperational but not impossible condition) a control rod block immediately occurs so that reactivity insertion by control rod withdrawal cannot occur.
The low scram pilot air header pressure trip performs the same function as the high water level in the scram discharge instrument volume for fast fillevents in which the high level instrument response time may be inadequate.
A fast fillevent is postulated for certain degraded control air events in which the scram outlet valves unseat enough to allow 5 gpm per drive leakage into the scram discharge volume but not enough to cause control rod insertion.
BFH Unit 2 3.1/4.1-16
4.1 BASES The minimum functi.onal testing frequency used in this specification is based on a reliability analysis using the concepts developed in reference (1).
This concept was specifically adapted to the one-out-of-two taken twice logic of the reactor protection system.
The analysis shows that the sensors are primarily responsible for the reliability of the reactor protection system.
This analysis makes use of "unsafe failure" rate experience at conventional and nuclear power plants in a reliability model for the system.
An "unsafe failure" is defined as one which negates channel operability and which, due to its nature, is revealed only when the channel is functionally tested or attempts to respond to a real signal.
Failure such as blown fuses, ruptured bourdon
- tubes, faulted amplifiers, faulted cables, etc., which result in "upscale" or "downscale" readings on the reactor instrumentation are "safe" and will be easily recognized by the operators during operation because they are revealed by an alarm or a scram.
The channels listed in Tables 4.1.A and 4.1.B are divided into three groups for functional testing.
These are:
A.
On-Off sensors that provide a scram trip function.
BE Analog devices coupled with bistable trips that provide a scram function.
C.
Devices which only serve a useful function during some restricted mode of operation, such as STARTUP or SHUTDOWN, or for which the only practical test is one that can be performed at SHUTDOWN.
The sensors that make up group (A) are specifically selected from among the whole family of industrial on-off sensors that have earned an excellent reputation for reliable operation.
During design, a goal of 0.9999 probability of success (at the
- 50. percent confidence level) was adopted to assure that a balanced and adequate design is achieved.
The probability of success is primarily a function 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 practices, and satisfies the design goal for the logic configuration utilized in the Reactor Protection System.
The once per six-month functional test frequency for the scram pilot air header low pressure trip function is acceptable due to:
1.
The functional reliability previously demonstrated by these switches on Unit 2 during Cycles 6 and 7, 2.
The need for minimizing the radiation exposure associated with the functional testing of these switches, and 3.
The increased risk to plant availability while the plant is in a half-scram condition during the performance of the functional testing versus the limited increase in reliability that would be obtained by more frequent functional testing.
BFN Unit 2 3.1/4.1-17 AKNDIKNTHO. 238
4.1 BASES (Cont'd)
I A single failure of one of the scram pilot air header low pressure trip switches would not result in the loss of the trip function. It is highly unlikely that two switches in one channel would experience an undetected failure during the period between six-month functional tests.
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 percent 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 percent confidence level.
This level of availability may be maintained by adjusting the test interval as a function of the observed failure history.
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 generally be used initially until a trend is established.
Group (B) devices utilize an analog sensor followed by an amplifier and a
bistable 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" mid-scale and is not capable of going either up or down in response to an out-of-limits input.
This type of failure for analog devices is a rare occurrence and is detectable by an operator who observes that one signal does not track the other three.
For purpose of analysis, it is assumed that this rare failure will be detected within two hours.
ll 1.
Reliability of Engineered Safety Features as a Function of Testing Frequency, I. M. Jacobs, "Nuclear Safety," Vol. 9, No. 4, July-August,
- 1968, pp. 310-312.
BFN Unit 2 3.1/4.1-18 Al'1'>KNT NO. 238
t 4.1 BASES (Cont'd)
The bistable trip circuit which is a part of the Group (B) devices can sustain unsafe failures which are revealed only on test.
Therefore, it is necessary to test them periodically.
A study was conducted of the instrumentation channels included in the Group (B) devices to calculate their "unsafe" failure rates.
The analog devices (sensors and amplifiers) are predicted to have an unsafe failure rate of less than 20 x 10 failure/hour.
The bistable trip circuits are predicted to have unsafe failure rate of less than 2 x 10 ~
failures/hour.
Considering the two hour monitoring interval for the analog devices as assumed
- above, and a weekly test interval for the bistable trip circuits, the design reliability goal of 0.99999 is attained with ample. margin.
The bistable devices are monitored during plant operation to record their failure history and establish a test interval using the curve of Figure 4.1-1.
There are numerous identical bistable devices used throughout the plant's instrumentation system.
Therefore, significant data on the failure rates for the bistable devices should be accumulated rapidly.
The frequency of calibration of the APRM Flow Biasing Network has been established at each refueling outage.
There are several instruments which must be calibrated and it will take several hours to perform the calibration of the entire network.
While the calibration is being performed, a zero flow signal will be sent to half of the APRMs resulting in a half scram and rod block condition.
Thus, if the calibration were performed during operation, flux shaping would not be possible.
Based on experience at other generating stations, drift of instruments, such as those in the Flow Biasing Network, is not significant and therefore, to avoid spurious
- scrams, a calibration frequency of each refueling outage is established.
Group (C) devices are active only during a
given portion of the operational cycle.
For example, the IRM is active during STARTUP and inactive during full-power operation.
- Thus, the only test that is meaningful is the one performed just prior to SHUTDOWN or STARTUP: i.e.,
the tests that are performed just prior to use of the instrument.
Calibration frequency of the instrument channel is divided into two groups.
These are as follows:
1.
Passive type indicating devices that can be compared with like units on a continuous basis.
2.
Vacuum tube or semiconductor devices and detectors that drift or lose sensitivity.
Experience with passive type instruments in generating stations and substations indicates that the specified calibrations are adequate.
For those devices which employ amplifiers, etc., drift specifications call for drift to be less than 0.4 percent/month; i.e., in the period of a month a drift of 4 percent would occur and thus providing for adequate margin.
BFN Unit 2 3.1/4.1-19 AtKiiDEKNT NO. 238
4.1 BASES (Cont'd For the APRM system drift of electronic apparatus is not the only consideration in determining a calibration frequency.
Change in power distribution and loss of chamber sensitivity dictate a calibration every seven days.
Calibration on this frequency assures plant operation at or below thermal limits.
A comparison of Tables 4.1.A and 4.1.B indicates that two instrument channels, have been included in the latter table.
These are:
mode switch in SHUTDOWN and manual scram.
All of the devices or sensors associated with these scram functions are simple on-off switches
- and, hence, calibration during operation is not applicable, i.e., the switch is either on or off.
The sensitivity of LPRM detectors decreases with exposure to neutron flux at a slow and approximately constant rate.
The APRM system, which uses the LPRM readings to detect a change in thermal power, will be calibrated every seven days using a heat balance to compensate for this change in sensitivity.
The RBM system uses the LPRM reading to detect a localized change in thermal power. It applies a correction factor based on the APRM output signal to determine the percent thermal power and therefore any change in LPRM sensitivity is compensated for by the APRM calibration.
The technical specification limits of
These methods use LPRM readings and TIP data to determine the power distribution.
Compensation in the process computer for changes in LPRM sensitivity will be made by performing a full core TIP traverse to update the computer calculated LPRM correction factors every 1000 effective full power hours.
As a minimum the individual LPRM meter readings will be adjusted at the beginning of each operating cycle before reaching 100 percent power.
BFN Unit 2 3.1/4.1-20 ZMRj':KNT NO.
238
~I' 0 1993
.THIS PAGE INTENTIONALLYLEFT BLANK BFN Unit 2 3.1/4.1-21 AMfNOMfItTNO. 2 1 g
gP,R RECIJ P
O~C
>s OO IVl0 Cy V/~
Cy t7 +)t*++
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON> D.C. 20555-0001 T
NN SSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT UNIT 3 AMENDMENT TO FACI ITY OPERATING LICENSE Amendment No. 197 License No.
DPR-68 The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated Hay 11,
- 1995, and supplemented on June 30,
- 1995, 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 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 i'n 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 ED 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 Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No.
DPR-68 is hereby amended to read as follows:
3.
(2)
Technical S ecifications The Technical Specifications contained in Appendices A and B,
as revised through Amendment No. 197, are hereby incorporated in the license.
The licensee. shall operate the facility in accordance with the Technical Specifications.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Frederick J. Hebdo, Director Project Directorate 11-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
August 29, 1995
ATTACHMENT TO LICENSE AMENDMENT NO. lo7 FACILITY OPERATING LICENSE NO.
DPR-68 DOCKET NO. 50-296 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.
The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.
Overleaf and **spillover pages are provided to maintain document completeness.
REMOVE 3.1/4.1-2 3.1/4 '-3 3.1/4.1-4 3.1/4.1-5 3.1/4.1-7 3.1/4.1-8 3.1/4.1-10 3.1/4.1-11
.3.1/4.1-15 3.1/4.1-16 3.1/4.1-17 3.1/4.1-18 3.1/4.1-19 3.1/4.1-20 INSERT 3.1/4.1-2*
3.1/4.1-3 3.1/4.1-4 3.1/4.1-5 3.1/4.1-7*
3.1/4.1-8 3.1/4.1-10 3.1/4.1-11*
3.1/4.1-15 3.1/4.1-16 3.1/4.1-17 3.1/4.1-18**
3.1/4.1-19"'
3.1/4.1-20*
TABLE 3.1.A REACTOR PROTECTION SYSTEH (SCRAH)
INSTRUHENTATION RE(}UIREHENTS Hin. No. of Operable Instr.
Channels Per Trip Tri L v 1
e tin Shut-down H d in Mhi h Func ion Startup/
~Rfu 1 '7
~H~t>~ndb Run A
ion 1
Hode Switch in Shutdown Hanual Scram IRH (16)
High Flux Inoperative X
X X
X
<120/125 Indica ted X(22)
X(22)
X on scale X
X 1.A 1.A (5) 1.A (5) 1.A APRH (16)(24)(25)
High Flux (Fixed Trip)
High Flux (Flow Biased)
High Flux Inoperative Downscale
< 120K See Spec. 2.1.A.l 15% rated power (13)
> 3 Indicated on Scale High Reactor Pressure
< 1055 psig (PI 5-3-22AA, BB,C, 0)
X(21)
X(17)
X(21)
X(17)
(11)
(11)
X(10)
X X
1.A or 1.B X
1.A or 1.8 (15) 1.A X
1.A X(12) 1.A or 1.B 1.A High Drywell Pressure (14)
< 2.5 psig (PIS-64-56 A-D)
Reactor Low Mater Level (14)
> 538" above (LIS-3-203 A-D) vessel zero X(8)
X(8) 1.A 1.A CO Ch
TABLE 3.1.A REACTOR PROTECTION SYSTEM (SCRAH)
INSTRUHENTATION REQUIREMENTS Hin. No.
Operable Instr.
Channels Per Trip IA 2
c I4l 4
of High Water Level in West Scram Discharge Tank (LS-85-45A-D)
High Water Level in East Scram Discharge Tank (LS-85-45E-H)
Main Steam Line Isolation Valve Closure Turbine Control Valve Fast Closure or Turbine Trip Turbine Stop Valve Closure Turbine First Stage Pressure Permissive (P IS-1-81ALB)
(P IS-1-91AC 8)
Low Scram Pilot Air Header Pressure Tri L v 1
< 50 Gallons
< 50 Gallons
<10K Valve Closure
>550 psig
<10K Valve Closure not >154 psig
> 50 psig Hd inWhihFn in Shut-Startup/
X(2)
X(2)
X(2)
X(2)
X X(18)
X(18)
X(2)
X(2)
~Ai ~nl 1.A X
1.A X(6) 1.A or 1.C X(4) 1.A or 1.D X(4) 1.A or 1.D
'X(18) 1.A or 1.0 (19) 1.A
NOTES FOR TABLE 1
There shall be two OPERABLE or tripped trip systems for each function.
If the minimum number of OPERABLE instrument channels per trip system cannot be met for one trip system, trip the inoperable channels or entire trip system within one hour, or, alternatively, take the below listed action for that trip function. If the minimum number of OPERABLE instrument channels cannot be met by either trip system, the appropriate action listed below (refer to right-hand column of Table) shall be taken.
An inoperable channel need not be placed in the tripped condition where this would cause the trip function to occur.
In these
- cases, the inoperable channel shall be restored to OPERABLE status within two hours, or take the action listed below for that trip function.
A.
Initiate insertion of OPERABLE rods and complete insertion of all OPERABLE rods within four hours.
In refueling mode, suspend all operations involving core alterations and fully insert all OPERABLE control rods within one hour.
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 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
~
D.
Reduce power to less than 30 percent of rated.
2.
The scram discharge volume high water level bypass may be used in SHUTDOWN or REFUEL to bypass both the scram discharge volume high-high water level and scram pilot air header low pressure scram signals in order to reset the reactor protection system trip.
A control rod withdraw block is present when these scram signals are bypassed.
3.
DELETED 4.
Bypassed when turbine first stage pressure is less than 154 psig.
5.
IRMs are bypassed when APRMs are onscale and the reactor mode switch is in the RUN position.
6.
The design permits closure of any two lines without a scram being initiated.
7.
When the reactor is subcritical and the reactor water temperature is less than 212 F, only the following trip functions need to be OPERABLE:
A.
Mode switch in shutdown B.
Manual scram C.
High flux IRM D.
Scram discharge volume high level BFN Unit 3 3.1/4.1-4 AIKND>KNTIIO. 197
NOTES FOR ABLE E.
Scram pilot air header lov pressure 8.
Not required to be OPERABLE when primary containment integrity is not required.
9.
(Deleted) 10.
Not required to be OPERABLE vhen the reactor pressure vessel head is not bolted to the vessel.
The APRM dovnscale trip function is only active when the reactor mode switch is in RUN.
12.
The APRM downscale trip is automatically bypassed when the IRM instrumentation is OPERABLE and not high.
13.
Less than 14 OPERABLE LPRMs vill cause a trip system trip.
14.
Channel shared by Reactor Protection System and Primary Containment and Reactor Vessel Isolation Control System.
A channel failure may be a channel failure in each system.
15.
The APRM 15 percent scram is bypassed in the RUN Mode.
16.
17.
Channel shared by Reactor Protection System and Reactor Manual Control System (Rod Block Portion).
A channel failure may be a channel failure in each system. If a channel is allowed to be inoperable per Table 3.1.A, the corresponding function in that same channel may be inoperable in the Reactor Manual Control System (Rod Block).
Not required while performing low power physics tests at atmospheric pressure during or after refueling at pover levels not to exceed 5 MWt.
18.
This function must inhibit the automatic bypassing of turbine control valve fast closure or turbine trip scram and turbine stop valve closure scram whenever turbine first stage pressure is greater than or equal to 154 psig.
19.
Action 1.A or 1.D shall be taken only if the permissive fails in such a
manner'o prevent the affected RPS logic from performing its intended function.
Othervise, no action is required.
20.
(Deleted) 21.
The APRM High Flux and Inoperative Trips do not have to be OPERABLE in the REFUEL Mode if the Source Range Monitors are connected to give a noncoincidence, High Flux scram, at 5 x 10 cps.
The SRMs shall be OPERABLE per Specification 3.10.B.1.
The removal of eight (8) shorting links is required to provide noncoincidence high-flux scram protection from the Source Range Monitors.
BFN Unit 3 3.1/4.1-5 AMEND>IENT NO. 197
TABLE 4.1.A REACTOR PROTECTION SYSTEH (SCRAH)
INSTRUHENTATION FUNCTIONAL TESTS HINIHUH FUNCTIONAL.TEST FRE()UENCIES FOR SAFETY INSTR.
AND CONTROL CIRCUITS Hode Switch in Shutdown Hanual Scram
]~rou i~2 F n ion 1
T Place Hode Switch in Shutdown Trip Channel and Alarm Hinimum Fr u
n Each Refueling Outage Every 3 Honths IRH High Flux Inoperative Trip Channel 'and Alarm (4)
Trip Channel and Alarm (4)
Once Per Week During Refueling and Before Each Startup Once Per Week During Refueling and Before Each Startup APRH High Flux (15% Scram)
High Flux (Flow Biased)
High Flux (Fixed Trip)
Inoperative Downscale Flow Bias High Reactor Pressure (PIS-3-22AA,BB,C,D)
High Drywell Pressure (PIS-64-56 A-D)
Reactor Low Mater Level (LIS-3-203 A-D)
Trip Output Relays (4)
Trip Output Relays (4)
Trip Output Relays (4)
Trip Output Relays (4)
Trip Output Relays (4)
(6)
Trip Channel and Alarm (7)
Trip Channel and Alarm (7)
Trip Channel and Alarm (7)
Before Each Startup and Meekly Mhen Required to be Operable Once/Week Once/Week Once/Week Once/Meek (6)
Once/Honth Once/Honth Once/Honth
4J I00
~r)~2 High Water Level in Scram Discharge Tank Float Switches (LS-85-45C-F)
A Electronic Level Switches (LS-85-45A, 8, G, H)
Hain Steam Line Isolation Valve Closure Turbine Control Valve Fast Closure or turbine trip Turbine First Stage Pressure Permissive (PIS-1-81A and 8, PIS-1-91A and 8)
Turbine Stop Valve Closure Low Scram Pilot Air Header Pressure (PS 85-35 Al, A2, Bl, and 82)
TABLE 4.1.A (Continued)
Fn i nl T Trip Channel and Alarm Trip Channel and Alarm (7)
Trip Channel and Alarm Trip Channel and Alarm Trip Channel and Alarm (7)
Trip Channel and Alarm Trip Channel and Alarm Hinim m Fr n
Once/Honth Once/Honth Once/3 Honths (8)
Once/Honth (1)
Every three months Once/Honth (1)
Once/6 Honths
TABLE 4.1.8 REACTOR PROTECTION SYSTEH (SCRAH)
INSTRUHENT CALIBRATION HINIHUH CALIBRATION FRY)UENCIES FOR REACTOR PROTECTION INSTRUHENT CHANNELS n
rmn hnnl IRH High Flux APRH High Flux Output Signal Flow Bias Signal LPRH Signal High Reactor Pressure (PIS-3-22AA,BB,C,D)
High Drywell Pressure (PIS-64-56 A-D)
Reactor Low Water Level (LIS-3-203 A-0)
Qrrg>~1 8
Qlibr~in Comparison to APRH on Controlled Startups (6)
Heat Balance Calibrate Flow Bias Signal (7)
TIP System Traverse (8)
Standard Pressure Source Standard Pressure Source Pressure Standard Hinim m Fr n
2 Note (4)
Once Every 7 Days Once/Operating Cycle Every 1000 Effective Full Power Hours Once/6 Honths(9)
Once/18 Honths(9)
Once/18 Honths(9)
High Water Level in Scram Discharge Volume Float Switches (LS-85-45C-F)
Electronic Lvl Switches
.(LS-85-45-A, 8, G, H)
Turbine First Stage Pressure Permissive (PIS-1-81ALB, P IS-1-91ASB)
Turbine Stop Valve Closure l4 Turbine Control Valve Fast Closure or Turbine Trip Low Scram Pilot Air Header Pressure (PS 85-35 Al, A2, Bl and 82)
Hain Steam Line Isolation Valve Closure A
Calibrated Water Column (5)
Calibrated Water Column Note (5)
Standard Pressure Source Standard Pressure Source Note (5)
Standard Pressure Source Note (5)
Once/Operating Cycle (9)
Note (5)
Once/18 Honths(9)
Once/Operating Cycle Note (5)
Once/18 Honths
NOTES FOR TABLE 4.
A description of three groups is included in the Bases of this specification.
2.
Calibrations are not required when the systems are not required to be OPERABLE or are tripped. If calibrations are missed, they shall be performed prior to returning the system to an OPERABLE status.
3.
(Deleted) 4 ~
Required frequency is initial startup following each refueling outage.
Physical inspection and actuation of these position switches will be performed once per operating cycle.
6.
On controlled startups, overlap between the IRMs and APRMs will be verified.
7.
The Flow Bias Signal Calibration will consist of calibrating the sensors, flow converters, and signal offset networks during each operating cycle.
The instrumentation is an analog type with redundant flow signals that can be compared.
The flow comparator trip and upscale will be functionally tested according to Table 4.2.C to ensure the proper operation during the operating cycle.
Refer to 4.1 Bases for further explanation of calibration frequency.
8.
A complete TIP system traverse calibrates the LPRM signals to the process computer.
The individual LPRM meter readings will be adjusted as a
minimum at the beginning of each operating cycle before reaching 100 percent power.
9 ~
Calibration consists of the adjustment of the primary sensor and associated components so that they correspond within acceptable range and accuracy to known values of the parameter which the channel monitors, including adjustment of the electronic trip circuitry, so that its output relay changes state at or more conservatively than the analog equivalent of the trip level setting.
BFH Unit 3 3.1/4.1-11 AMENDMENTM.I85
3.1 BASES (Cont'd) be accommodated which would result in slow scram times or partial control rod insertion.
To preclude this occurrence, level switches have been provided in the instrument volume which alarm and scram the reactor when the volume of water reaches 50 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 discharge water and precludes the situation in which a scram would be required but not be able to perform its function adequately.
A source range monitor (SRM) system is also provided to supply additional neutron level information during startup but has no scram functions.
'Reference Section 7.5.4 FSAR.
Thus, the IRM is required in the REFUEL and STARTUP modes.
In the power range the APRM system provides required protection.
Reference Section 7.5.7 FSAR.
Thus, the IRM System is not required in the RUN mode.
The APRMs and the IRMs provide adequate coverage in the,STARTUP and intermediate range.
The high reactor pressure, high drywell pressure, reactor low water level, lov scram pilot air header pressure aud scram discharge volume high level scrams are required for STARTUP and RUH modes of plant operation.
They are, therefore, required to be operational for these modes of reactor operation.
The requirement to have the scram functions as indicated in Table 3.1.1 OPERABLE in the REFUEL mode is to assure that shifting to the REFUEL mode during reactor power operation does not diminish the need for the reactor protection system.
Because of the APRM downscale limit of g 3 percent when in the RUH mode and high level limit of g15 percent when in the STARTUP Mode, the transition between the STARTUP and RUN Modes must be made with the APRM instrumentation indicating between 3 percent and 15 percent of rated power or a control rod scram will occur.
In addition, the IRM system must be indicating below the High Flux setting (120/125 of scale) or a scram will occur when in the STARTUP Mode.
For normal operating conditions, these limits provide assurance of overlap between the IRM system and APRM system so that there are no "gapseg in the power level indications (i.e., the power level is continuously monitored from beginning of startup to full power and from full power to shutdown).
When power is being reduced, if a transfer to the STARTUP mode is made and the IRMs have not been fully inserted (a maloperational but not impossible condition) a control rod block immediately occurs so that reactivity insertion by control rod withdrawal cannot occur.
The low scram pilot air header pressure trip performs the same function as the high water level in the scram discharge instrument volume for fast fillevents in which the high level instrument response time may be inadequate.
A fast fillevent is postulated for certain degraded control air events in which the scram outlet valves unseat enough to allow 5 gpm per drive leakage into the scram discharge volume but not enough to cause control rod insertion.
BFH Unit 3 3.1/4.1-15 AtIENDIlENT ITO. 197
'I I
4.1 BASES The minimum functional testing frequency used in this specification is based on a reliability analysis using the concepts developed in reference (1).
This concept was specifically adapted to the one-out-of-two taken twice logic of the reactor protection system.
The analysis shows that the sensors are primarily responsible for the reliability of the reactor protection system.
This analysis makes use of "unsafe failure" rate experience at conventional and nuclear power plants in a reliability model for the system.
An "unsafe failure" is defined as one which negates channel operability and which, due to its nature, is revealed only when the channel is functionally tested or attempts to respond to a real signal.
Failure such as blown fuses, ruptured bourdon
- tubes, faulted amplifiers, faulted cables, etc., which result in "upscale" or "downscale" readings on the reactor instrumentation are "safe" and will be easily recognized by the operators during operation because they are revealed by an alarm or a scram.
The channels listed in Tables 4.1.A and 4.1.B are divided into three groups for functional testing.
These are:
A.
On-Off sensors that provide a scram trip function.
B.
Analog devices coupled with bistable trips that provide a scram function.
C.
Devices which only serve a useful function during some restricted mode of operation, such as STARTUP or SHUTDOWN, or for which the only practical test is one that can be performed at shutdown.
The sensors that make up group (A) are specifically selected from among the whole family of industrial on-off sensors that have earned an excellent reputation for reliable operation.
During design, a goal of 0.99999 probability of success (at the 50 percent confidence level) was adopted to assure that a balanced and adequate design is achieved.
The probability of success is primarily a function 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 practices, and satisfies the design goal for the logic configuration utilized in the Reactor Protection System.
The once per six-month functional test frequency for the scram pilot air header low pressure trip function is acceptable due to:
1.
The functional reliability previously demonstrated by these switches on Unit 2 during Cycles 6 and 7, 2.
The need for minimizing the radiation exposure associated with the functional testing of these switches, and 3.
The increased risk to plant availability while the plant is in a half-scram condition during the performance of the functional testing versus the limited increase in reliability that would be obtained by more frequent functional testing.
BFN Unit 3 3.1/4.1-16 AtKNDIKb~i NO. 197
4.1 BASES (Cont'd)
A single failure of one of the scram pilot air header low pressure trip switches would not result in the loss of the trip function. It is highly unlikely that two switches in one channel would experience an undetected failure during the period between six-month functional tests.
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-percent 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 percent confidence level.
This level of availability may be maintained by adjusting the test interval as a function of the observed failure history.
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 generally be used initially until a trend is established.
Group (B) devices utilize an analog sensor followed by an amplifier and a
bistable 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" mid-scale and is not capable of going either up or down in response to an out-of-limits input.
This type of failure for analog devices is a rare occurrence and is detectable by an operator who observes that one signal does not track the other three.
For purpose of analysis, it is assumed that this rare failure will be detected within two hours.
1.
Reliability of Engineered Safety Features as a Function of Testing Frequency, I. M. Jacobs, "Nuclear Safety," Vol. 9, No. 4, July-August,
- 1968, pp. 310-312.
BFN Unit 3 3.1.4.1-17 AiKNDI1EVZi'. 197
4.1 BASES (Cont'd)
The bistable trip circuit which is a part of the Group (B) devices can sustain unsafe failures which are revealed only on test.
Therefore, it is necessary to test them periodically.
A study was,conducted of the instrumentation channels included in the Group (B) devices to calculate their "unsafe" failure rates.
The analog devices (sensors and amplifiers) are predicted to have an unsafe failure rate of less than 20 x 10 failure/hour.
The bistable trip circuits are predicted to have unsafe failure rate of less than 2
x 10 failures/hour.
Considering the two hour monitoring interval for the analog devices as assumed
- above, and a weekly test interval for the bistable trip circuits, the design reliability goal of 0.99999 is attained with ample margin.
The bistable devices are monitored during plant operation to record their failure history and establish a test interval using the curve of Figure 4.1-1.
There are numerous identical bistable devices used throughout the plant's instrumentation system.
Therefore, significant data on the failure rates for the bistable devices should be accumulated rapidly.
The frequency of calibration of the APRM Flow Biasing Network has been established at each refueling outage.
There are several instruments which must be calibrated and it will take several hours to perform the calibration of the entire network.
While the calibration is being performed, a zero flow signal will be sent to half of the APRMs resulting in a half scram and rod block condition.
Thus, if the calibration were performed during operation, flux shaping would not be possible.
Based on experience at other generating stations, drift of instruments, such as those in the Flow Biasing Network, is not significant and therefore, to avoid spurious
- scrams, a calibration frequency of each refueling outage is established.
Group (C) devices are active only during a
given portion of the operational cycle.
For example, the IRM is active during STARTUP and inactive during full-power operation.
- Thus, the only test that is meaningful is the one performed just prior to SHUTDOWN or STARTUP; i.e the tests that are performed just prior to use of the instrument.
Calibration frequency of the instrument channel is divided into two groups.
These are as follows:
1.
Passive type indicating devices that can be compared with like units on a continuous basis.
2.
Vacuum tube or semiconductor devices and detectors that drift or lose sensitivity.
BFN Unit 3 AJKNDI'~ EJO.
197
4.1 BASES (Cont'd) a l
Experience with passive type instruments in generating stations and substations indicates that the specified calibrations are adequate.
For those devices which employ amplifiers, etc., drift specifications call for drift to be less than 0.4 percent/month; i.e., in the period of a month a drift of.4-percent would occur and thus providing for adequate margin.
For the APRM system drift of electronic apparatus is not the only consideration in determining a calibration frequency.
Change in power distribution and loss of chamber sensitivity dictate a calibration every seven days.
Calibration on this frequency assures plant operation at or below thermal limits.
A comparison of Tables 4.1.A and 4.1.B indicates that two instrument channels have been included in the latter table.
These are:
mode switch in SHUTDOWN and manual scram.
All of the devices or sensors associated with these scram functions are simple on-off switches
- and, hence, calibration during operation is not applicable, i.e., the switch is either on or off.
The sensitivity of LPRM detectors decreases with exposure to neutron flux at a slow and approximately constant rate.
The APRM system, which uses the LPRM readings to detect a change in thermal power, will be calibrated every'even days using a heat balance to compensate for this change in sensitivity.
The RBM system uses the LPRM reading to detect a localized change in thermal power. It applies a correction factor based on the APRM output signal to determine the percent thermal power and therefore any change in LPRM sensitivity is compensated for by the APRM calibration.
The technical specification limits of
These methods use LPRM readings and TIP data to determine the power distribution.
Compensation in the process computer for changes in LPRM sensitivity will be made by performing a full core TIP traverse to update the computer calculated LPRM correction factors every 1000 effective full power hours.
As a minimum the individual LPRM meter readings will be adjusted at the beginning of each operating cycle before reaching 100 percent power.
BFN Unit 3
> 1(4 1-19 AIIEKMZHNQ. 197
NN 2 0 1993 THIS PAGE INTENTIONALLYLEFT BLANK BFN Unit 3 3.1/4.1-20 AMENDMENTNO. T VO