ML18038B877
| ML18038B877 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 05/07/1997 |
| From: | Hebdon F NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML18038B878 | List: |
| References | |
| NUDOCS 9705150077 | |
| Download: ML18038B877 (49) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 247 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 June 21,
- 1996, and supplemented on February 7,
- 1997, 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.
97055.50077 970507 PDR ADQCK 05000259 P
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f
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:
(2)
Technical S ecifications The Technical Specifications contained in Appendices A and B,
as revised through Amendment No. 247, 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 Frederick J.
Hekdon, Director Project Directorate II-3 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
May 7, 1997
ATTACHMENT TO LICENSE AMENDMENT NO..247 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 pages are included to maintain document completeness.
REMOVE
- 1. 1/2. 1-1 1.1/2.1-2 1.1/2.1-8 1.1/2.1-9 1.1/2,1-12 1.1/2.1-13 1.1/2.1-14 1.1/2.1-15 3.3/4.3-17 3.3/4.3-18 INSERT 1.1/2.1-1.
1.1/2.1-2*
1.1/2.1-8 1.1/2.1-9 1.1/2.1-12*
1.1/2.1-13 1.1/2.1-14 1.1/2.1-15 3.3/4.3-17 3.3/4.3-18*
2.1 Applies to the interrelated variables associated with fuel thermal behavior.
Applies to trip settings of the instruments and devices which are provided to prevent the reactor system safety limits from being exceeded.
To establish limits which ensure the integrity of the fuel cladding.
To define the level of the process variables at which automatic protective action is initiated to prevent the fuel cladding integrity safety limit from being exceeded.
The limiting safety system settings shall be as specified below:
1.
Reactor Pressure
>800 psia and Core Flow 10% of Rated.
When the reactor pressure is greater than 800 psia, the existence of a minimum critical power ratio (MCPR) less than 1.10 shall constitute violation of the fuel cladding integrity safety limit.
1.
APRM Flux Scram Trip Setting (RUN Mode)
(Flow Biased) a.
When the Mode Switch is in the RUN
BFN Unit 2
- 1. 1/2. 1-1 Ame'ndmeat No. 247
1.1 2.1 FUEL CL DDING NTEG ITY S FET I
G TY S S
S T N 2.1.A Neut on Flu Tri Settin s
2.1.A.l.a (Cont'd)
S<(0.58W + 62%)
where:
b.
S ~ Setting in percent of rated thermal power (3293 MWt)
W i Loop recirculation flow rate in percent of rated For no combination of loop recirculation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 120% "'of rated thermal power.
BFN Unit 2 1.1/2.1-2 A"iEHDMENNo. P Q P
4
~ 1 The fuel cladding represents one of the physical barriers which separate radioactive materials from environs.
The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.
Although some corrosion or use-related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable.
Fuel cladding perforations,
- however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the protection system setpoints.
While fission product migration from cladding perforation is just as measurable as that from use-related
- cracking, the thermally-caused cladding perforations signal a threshold, beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration.
Therefore, the fuel cladding safety limit is defined in terms of the reactor operating conditions which can result in cladding perforation.
The fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an abnormal operational transient.
Because fuel damage is not directly observable, the Fuel Cladding Safety Limit is defined with margin to the conditions which would produce onset trans~cion boiling (MCPR of 1.0).
Maintaining the MCPR 'greater than the Safety Limit MCPR represents a conservative margin re'lative to the conditions required to maintain fuel cladding integrity.
Onset of transition boiling results in a decrease in heat transfer from the clad and, therefore, elevated clad temperature and the possibility of clad failure.
Since boiling transition is not a directly observable parameter, the margin to boiling transition is calculated from plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution.
The margin for each fuel assembly is characterized by the critical power ratio (CPR) which is the patio of the bundle power which would produce onset of transition boiling divided by the actual bundle power.
The minimum value of this ratio for any bundle in the core is the minimum critical power ratio (MCPR). It is assumed that the plant operation is controlled to the nominal protective setpoints via the instrumented variables, i.e.,
normal plant operation presented on Figure 2. 1-1 by the nominal expected. flow control line.
The Safety Limit has sufficient conservatism to assure that in the event of an abnormal operational transient initiated from a normal operating condition (MCPR limits specified in Specification 3.5.K) more than 99.9 percent of the fuel rods in the core are expected to avoid boiling transition.
The margin between MCPR of 1.0 (onset of transition boiling) and the Safety Limit MCPR is derived from a detailed statistical analysis considering all of the uncertainties in monitoring the core operating state including uncertainty in the boiling transition correlation as described in Reference 1.
The uncertainties employed in deriving the safety limit are provided at the beginning of each fuel cycle.
BFN Unit 2 1.1/2.1-8 Amendment No. 247
Because the boiling transition correlation is based on a large quantity of full scale data there is a very high confidence that operation of a fuel assembly at the condition of MCPR equal to the Safety Limit MCPR would not j
produce boiling transition.
- Thus, although it is not required to establish the safety limit additional margin exists between the safety limit and the actual occurrence of loss of cladding integrity.
However, if boiling transition were to occur, clad perforation would not be expected.
Cladding temperatures would increase to approximately 1,100 F
which is below the perforation temperature of the cladding material.
This has been verified by tests in the General Electric Test Reactor (GETR) where fuel similar in design to BFNP operated above the critical heat flux for a significant period of time (30 minutes) without clad perforation.
If reactor pressure should ever exceed 1,400 psia during normal power operation (the limit of applicability of the boiling transition correlation) it would be assumed that the fuel cladding integrity Safety Limit has been violated.
At pressures below 800 psia, the core elevation pressure drop (0 power, 0 flow) is greater than'.56 psi.
At low powers and flo~s this pressure differential is maintained in the bypass region of the core.
Since the pressure drop in the bypass region is essentially all elevation
- head, the core pressure drop at low power and flows will always be greater than 4.5 psi.
Analyses show that with a flow of 28x10 lbs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi.
- Thus, the bundle flow with a 4.56 psi driving head will be greater than 28x103 lbs/hr.
Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt.
With the design peaking factors this corresponds to a core thermal power of more than 50 percent.
- Thus, a
core thermal power limit of 25 percent for reactor pressures below 800 psia is conservative.
For the fuel in the core during periods when the reactor is shut
- down, consideration must also be given to water level requirements due to the effect of decay heat.
If water level should drop below the top of the fuel during this time, the ability to remove decay heat is reduced.
This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation.
As long as the fuel remains covered with water, sufficient cooling is available to prevent fuel clad perforation.
BFN Unit 2 1.1/2.1-9 Amendment No. pqy
?. 1 BlhSES (Cont'd)
The bases for individual setpoints are discussed below:
A.
eut o
lux Scr RM ow-B ased Hi F ux Scram Tri Settin RUN Mode The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady-state conditions, reads in percent of rated power (3,293 MWt).
Because fission chambers provide the basic input signals, the APRM system responds directly to core average neutron flux.
During power increase transients, the instantaneous fuel surface heat flux is less than the instantaneous neutron flux by an amount depending upon the duration of the transient and the fuel time constant.
For this reason, the flow-biased 'scram APRM flux signal is passed through a
filtering network with a time constant which is representative of the fuel time constant.
As a result of this filtering, APRM flow-biased scram will occur only if the neutron flux signal is in excess of the setpoint and of sufficient time duration to overcome the fuel time constant and result in an average fuel surface heat flux which is equivalent to the neutron flux trip setpoint.
This setpoint is variable up to 120 percent of rated power based on recirculation drive flow according to the equations given in Section
- 2. 1.A.1 and the graph in Figure 2.1-2.
For the purpose of licensing transient analysis, neutron flux scram is assumed to occur at 120 percent of rated power.
Therefore, the flow biased scram provides additional margin to the thermal limits for slow transients such as loss of feedwater heating.
No safety credit is taken for flow-biased scrams.
BFN Unit 2 1.1/2.1-12 AM.ENDV,ENT NP. P y g
Analyses of the limiting transients show that no scram adjustment is required to assure NCPR is greater than the Safety Limit NCPR when the transient is initiated from MCPR limits specified in Specification 3.5.k.
For operation in the startup mode while the reactor is at low
- pressure, the APRM scram setting of 15 percent of rated power provides adequate thermal margin. between the setpoint and the safety limit, 25 percent of rated.
The margin is adequate to accommodate anticipated maneuvers associated with power plant startup.
Effects of increasing pressure at zero or low void content are minor, cold water from sources available during st~rtup is not much colder than that already in the system, temperature coefficients are small, and control rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimizer.
Worth of individual rods is very low in a uniform rod pattern..
Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise; Because the flux distribution associated with uniform rod withdrawals does not involve high local
- peaks, and because several rods must be moved to change power by a significant percentage of rated power, the rate of power rise is very slow.
Generally, the heat flux is in near equilibrium with the fission rate.
In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than five percent of rated power per minute, and the APRM system would be more than adequate to assure a scram before the power could exceed the safety limit.
The 15 percent APRM scram remains active until the mode switch is placed in the RUN position.
This switch occurs when reactor pressure is greater than 850 psig.
The IRM System consists of eight chambers, four in each of the reactor protection system logic channels.
The IRM is a five-decade instrument which covers the range of power level between that covered by the SRM and the APRM.
The five decades are covered by the IRM by means of a range switch and the five decades are broken down into 10 ranges, each being one-half of a decade in size.
The IRM scram setting of 120 divisions is active in each range of the IRM.,For example, if the instrument was on range 1, the scram setting would be 120 divisions for that range; likewise if the instrument was on range 5, the scram setting would be 120 divisions for that range.
BFN Unit 2
- 1. 1/2. 1-13 Amendment No. 247
- Thus, as the ZRM is ranged up to accommodate the increase in power
- level, the scram setting is also ranged up.
A scram at 120 divisions on the IRM instruments remains in effect as long as the reactor is in the startup mode.
In addition, the APRM 15 percent scram prevents higher power operation without being in the RUN mode.
The ZRM scram provides protection for changes which occur both locally and over the entire core.
The most significant sources of reactivity change during the power increase are due to control rod withdrawal.
For insequence control rod withdrawal, the rate of change of power is slow enough due to the physical limitation of withdrawing control rods that heat flux is in equilibrium with the neutron flux.
An IRM scram would result in a reactor shutdown well before any SAFETY LIMIT is exceeded.
For the case of a single control rod withdrawal error, a range of rod withdrawal accidents was analyzed.
This analysis included starting the accident at various power levels.
The most severe case involves an initial condition in which the reactor is just subcritical and the ZRM system is not yet on scale.
This condition exists at quarter rod density.
Quarter rod density is discussed in paragraph 7.5.5.4 of the FSAR.
Additional conservatism was taken in this analysis by assuming that the ZRM channel closest to the withdrawn rod is bypassed.
The results of this analysis show that the reactor is scrammed and peak power limited to one percent of rated power, thus maintaining MCPR above the Safety Limit MCPR.
" Based on the above analysis, the ZRM provides protection against local control rod withdrawal errors and continuous withdrawal of control rods in sequence.
The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady-state conditions, reads in percent of rated power (3,293 MWt).
The APRM system responds directly to neutron flux.
Licensing analyses have demonstrated that with a neutron flux scram of 120 percent of rated power, none of the abnormal operational transients analyzed violate the fuel SAFETY LIMIT and there is a substantial margin from fuel damage.
Reactor power level may be varied by moving control rods or by varying the recirculation flow rate.
The APRM system provides a control rod block to prevent rod withdrawal beyond a given point at constant recirculation flow rate and thus prevents scram actuation.
This rod block trip setting, which is automatically varied with r'ecirculation loop flow rate, prevents an increase in the reactor power level to excess values due to control rod withdrawal.
The flow variable trip setting is selected to provide adequate margin to the flow-biased scram setpoint.
BFN Unit 2 1.1/2.1-14 Amendment No. 247
2.1 g~ (Cont'd)
C.
w v
The setpoint for the low level scram is above the bottom of the separator skirt.
This level has been used in transient analyses dealing with coolant inventory decrease.
The results reported in FSAR Subsection 14.5 show that scram and isolation of all process lines (except main steam) at this level a'dequately protects the fuel and the pressure barrier, because MCPR is greater than the Safety Limit MCPR in all cases, and system pressure does
(
not reach the safety valve setti::gs.
The scram setting is sufficiently below normal operating range to avoid spurious scrams.
D.
The turbine stop valve closure trip anticipates the pressure, neutron flux and heat flux increases that would result from closure of the stop valves.
With a trip setting of 10 percent of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained even during the worst case transient that assumes the turbine bypass valves remain closed.
(Reference 2)
Turbine control valve fast closure or turbine trip scram anticipates the
- pressure, neutron flux, and heat flux increase that could result from control valve fast closure due to load rejection or control valve closure due to turbine trip; each without bypass valve capability.
The reactor protection system initiates a scram in less than 30 milliseconds after the start of control valve fast closure due to load rejection or control valve closure due to turbine trip.
This scram is achieved by rapidly reducing hydraulic control oil pressure at the main turbine control valve actuator disc dump" valves.
This loss of pressure is sensed by pressure switches whose contacts form the one-out-of-two-twice logic input to the reactor protection system.
This trip setting, a nominally 50 percent greater closure time and a different valve characteristic from that of the turbine stop valve, combine to produce transients very similar to that for the stop valve.
No significant change in MCPR occurs.
Relevant transient analyses are discussed in References 2 and 3 of the Final Safety Analysis Report.
This scram is bypassed when turbine steam flow is below 30 percent of
- rated, as measured by turbine first state pressure.
BFN Unit 2
'1. 1/2. 1-15
'Amendment No. 247
3.3/4.3 ~ (Cont'd) 5.
The Rod Block Monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power level operation.
Two RBM channels are provided, and one of these may be bypassed from the console for maintenance and/or testing.
Automatic rod withdrawal blocks from one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage.
The specified restrictions with one channel out of service conservatively assure that fuel damage will not occur due to rod withdrawal errors when this condition exists.
The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent fuel damage; i.e., to prevent the MCPR from becoming less than the Safety Limit MCPR.
The limiting power transients are given in Reference 1.
Analysis of these transients shows that the negative reactivity rates resulting from the scram with the average response of all drives as given in the above specifications provide the required protection and MCPR remains greater than the Safety Limit MCPR.
On an early
- BWR, some degradation of control rod scram performance occurred during plant STARTUP and was determined to be caused by particulate material (probably construction debris) plugging an internal control rod drive filter.
The design of the present control rod drive (Model 7RDB144B) is grossly improved by the relocation of the filter to a location out of the scram drive path; i.e., it can no longer interfere with scram performance, even if completely blocked.
The degraded performance of the original drive (CRD7RDB144A) under dirty operating conditions and the insensitivity of the redesigned drive (CRD7RDB144B) has been demonstrated by a series of engineering tests under simulated reactor operating conditions.
The successful performance of the new drive under actual operating conditions has also been demonstrated by consistently good in-service test'esults for plants using the new drive and may be inferred from plants using the older model BFN Unit 2 3.3/4.3-17 Amendment No 247
3.3/4.3 BASES (Cont'd) drive with a modified (larger screen size) internal filter which is less prone to plugging.
Data has been documented by surveillance reports in various operating plants.
These include Oyster Creek, Monticello, Dresden 2,
and Dresden 3-Approximately 5000 drive tests have been recorded to date.
r Following identification of the "plugged filter" problem, very frequent scram tests were necessary to ensure proper performance.
However, the more frequent scram tests are now considered totally unnecessary and unwise for the following reasons:
2 ~
3.
Erratic scram performance has been identified as due to an obstructed drive filter in type "A" drives.
The drives in BFNP are of the new "B" type design whose scram performance is unaffected by filter condition.
The dirt load is primarily released during STARTUP of the reactor when the reactor and its systems are first subjected to flows and pressure and thermal stresses.
Special attention and measures are now being taken to assure cleaner systems.
Reactors with drives identical or similar (shorter stroke, smaller piston areas) have operated through many refueling cycles with no sudden or erratic changes in scram performance.
This preoperational and STARTUP testing is sufficient to detect anomalous drive performance.
The 72-hour outage limit which initiated the start of the frequent scram testing is arbitrary, having no logical basis other than quantifying a "major outage" which might reasonably be caused by an event so severe as to possibly affect drive performance.
This requirement is unwise because it provides an incentive for shortcut actions to hasten returning "on line" to avoid the additional testing due a 72-hour outage.
BFN Unit 2 3.3/4.3-18 TS 370 Letter Dated 11/17/95
<~""50'o Cy 00 I-0 Y/
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055&4001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.207 License No.
DPR-68 The Nuc',ear Regulatory Commission
(+he Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated June 21,
- 1996, and supplemented on February 7,
- 1997, 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-68 is hereby amended to read as follows:
(2)
Techni cal S eci ficati ons The Technical Specifications contained in Appendices A and 8, as revised through Amendment No. 2O7, 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
').,>., 3 c). lJ,.)2.
Frederick J.
Hehdon, Director Project Directorate II-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
May 7,'i997
ATTACHMENT TO LIC NSE AMENDMENT NO.
207 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 pages are included to maintain document completeness.
REMOVE 1.1/2.1-1 1.1/2.1-2 1.1/2.1-8 1.1/2.1-9 1.1/2.1-12 1.1/2.1-13 1.1/2.1-14 1.1/2.1-15 3.3/4.3-17
'3.3/4.3-18 INSERT 1.1/2.1-1 1.1/2.1-2*
1.1/2.1-8 1.1/2.1-9 l,l/2.1-12*
1.1/2.1-13 1.1/2.1-14 1.1/2.1-15 3.3/4.3-17 3.3/4.3-18*
1.1 2.1 Applies to the interrelated variables associated with fuel thermal behavior.
Applies to trip settings of the instruments and devices which are provided to prevent the reactor system safety limits from being exceeded.
To establish limits which ensure the integrity of the fuel cladding.
To define the level of the process variables at which automatic protective action is initiated to prevent the fuel cladding integrity safety limit from being exceeded.
The limiting safety system settings shall be as specified below:
1.
Reactor Pressure
>800 psia and Core Flow 10% of Rated.
When the reactor pressure is greater than 800 psia, the existence of a minimum critical power ratio (MCPR) less than 1.10 shall constitute violation of the fuel cladding integrity safety limit; APRM Flux Scram Trip Setting (Run Mode)
(Flow Biased)
When the Mode Switch is in the RUN
BFN Unit 3
- 1. 1/2. 1-1 Amendment No. 207
1.
2.1 FUEL CLADDING I TEG SAFE L MI ING S
ETY SYSTEM SETTING 2.1.A Neutron Flu.. Tri Settin s
2.1.A.l.a (Cont'd)
S-(
'SSW + 62%)
I where:
S
= Setting in percent of rated thermal power (3293 MWt)
W ~ Loop recirculation flow rate in percent of rated b.
For no combination of loop recirculation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 120% of rated thermal power.
BFN Unit 3 1.1/2.1-2 AMENDMENTg0. I 9 0
1.1 The fuel cladding represents one of the physical barriers which separate radioactive materials from environs.
The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.
Although some corrosion or use-related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable.
Fuel cladding perforations,
- however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the protection system setpoints.
While fission product migration from cladding perforation is just as measurable as that from use-related
- cracking, the thermally-caused cladding perforations signal a threshold, beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration.
Therefore, the fuel cladding safety limit is defined in terms of the reactor operating conditions which can result in cladd'ng perforation.
The fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an abnormal operational transient.
Because fuel damage is not directly observable, the Fuel Cladding Safety Limit is defined with margin to the conditions which would produce onset transition boiling (MCPR of 1.0).
Maintaining the MCPR g eater than the Safety Limit MCPR represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.
Onset of transition boiling results in a decrease in heat transfer from the clad and, therefore, elevated clad temperature and the possibility of clad failure.
Since boiling transition is not a directly observable parameter, the margin to boiling transition is calculated from plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution.
The margin for each fuel assembly is characterized by the critical power ratio (CPR) which is the ratio of the bundle power which would produce onset of transition boiling divided by the actual bundle power.
The minimum value of this ratio for any bundle in the core is the minimum critical power ratio (MCPR). It is assumed that the plant operation is controlled to the nominal protective setpoints via the instrumented variables, i.e.,
normal plant operation presented on Figure 2.1-1 by the nominal expected flow control line.
The Safety Limit has sufficient conservatism to assure that in the event of an abnormal operational transient initiated from a normal operating condition (MCPR limits specified in Specification 3.5.K) more than 99.9 percent of the fuel rods in the core are expected to avoid boiling transition.
The margin between MCPR of,1.0 (onset of transition boiling) and the Safety Limit MCPR is derived from a detailed statistical analysis considering all of the uncertainties in monitoring the core operating state including uncertainty in the boiling transition correlation as described in Reference 1.
The I
uncertainties employed in deriving the safety limit are provided at the beginning of each fuel cycle.
BFN Unit 3
- 1. 1/2. 1-8 Amendment No.
207
Because the boiling transition correlation is based on a large quantity of full scale data there is a very high confidence that operation of a fuel assembly at the condition of MCPR equal to the Safety Limit MCPR would not produce boiling transition.
- Thus, although it is not recp ired to establish the safety limit additional margin exists between the safety limit and the actual occurrence of loss-of-cladding integrity.
k However, if boiling transition were to occur, clad perforation would not be expected.
Cladding temperatures would increase to approximately 1,100 F
which is below the perforation temperature of the cladding material.
This has been verified by tests in the General Electric Test Reactor (GETR) where fuel similar in design to BFNP operated above the critical heat flux for a significant period of time (30 minutes) without clad perforation.
If reactor pressure should ever exceed 1,400 psia during normal power operation (the limit of applicability of the boiling transition correlation) it would be assumed that the fuel cladding integrity Safety Limit has been violated.
At pressures below 800 psia, the core elevation pressure drop (0 power, 0 flow) is greater than 4.56 psi.
At low powers and flows this pressure differential is maintained in the bypass region of the core.
Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi.
Analyses show that with a flow of 28x103 lbs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi.
- Thus, the bundle flow with a 4.56 psi driving head will be greater than 28x10 lbs/hr.
Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt.
With the design peaking 'factors this corresponds to a core thermal power of more than 50 percent.
- Thus, a
core thermal power limit of 25 percent for reactor pressures below 800 psia is conservative.
For the fuel in the core during periods when the reactor is shut down, consideration must also be given to water level requirements due to the effect of decay heat.
If water level should drop below the top of the fuel during this time, the ability to remove decay heat is reduced.
This reduction in cooling capability could lead to elevated cladding temperature's and clad perforation.
As long as the fuel remains covered with water, sufficient cooling is available to prevent fuel clad perforation.
BFN Unit 3 1.1/2.1-9 Amendment No 207
2.1 BASES (Cont'd)
The bases for individual setpoints are discussed below:
A.
Neut o
u Scram 1.
PRM Flow-Biased Hi h Flux Scram Tri Settin RUN Mode The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady-state conditions, reads in percent of rated power (3,293 MWt).
Because fission chambers provide the basic input signals, the APRM system responds directly to core average neutron flux.
During power increase transients, the instantaneous fuel I
surface heat flux is less than the instantaneous neutron flux by an amount depending upon the duration of the transient and the fuel time constant.
For this reason, the flow-biased scram APRM flux signal is passed through a
filtering network with a time constant which is representative of the fuel time constant.
As a result of this filtering, APRM flow-biased scram will occur only if the neutron flux signal is in excess of the setpoint and of sufficient time duration to overcome the fuel time constant and result in an average fuel surface heat flux which is equivalent to the neutron flux trip setpoint.
This setpoint is variable up to 120 percent of rated power based on recirculation drive flow according to the equations given in Section 2.1.A.1 and the graph in Figure 2.1-2.
For the purpose of licensing transient analysis, neutron flux scram is assumed to occur at 120 percent of rated power.
Therefore, the flow biased scram provides additional margin to the thermal limits for slow transients such as loss of feedwater heating.
No safety credit is taken for flow-biased scrams.
BFN Unit 3 1.1/2.1-12 AMENDMENT NO. 1 9 0
Analyses of the limiting transients show that no scram adjustment is required to assure MCPR is greater than the Safety Limit MCPR when the transient is initiated from MCPR limits specified in Specification 3.5.k.
For operation in the startup mode while the reactor is at low pressure, the APRM scram setting of 15 percent of rated power provides adequate thermal margin between the setpoint and the safety limit, 25 percent of rated.
The margin is adequate to accommodate anticipated maneuvers associated with power plant startup.
Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup is not much colder than that already in the system, temperature coefficients are small, and control rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimizer.
Worth of individual rods is very low in a uniform rod pattern.
Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise.
Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated
- power, the rate of power rise is very slow.
Generally, the heat flux is in near equilibrium with the fission rate.
In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5 percent of rated power per minute, and the APRM system would be more than adequate to assure a scram before the power could exceed the safety limit.
The 15 percent APRM scram remains active until the mode switch is placed in the RUN position.
This switch occurs when reactor pressure is greater than 850 psig.
The IRM System consists of eight chambers, four in each of the reactor protection system logic channels.
The IRM is a five-decade instrument which covers the range of power level between that covered by the SRM and the APRM.
The five decades are covered by the IRM by means of a range switch and the five decades are broken down into 10 ranges, each being one-half of a decade in size.
The IRM scram setting of 120 divisions is active in each range of the IRM.
For example, if the instrument was on range 1, the scram setting would be 120 divisions for that range; likewise, if the instrument was on range 5, the scram setting would be 120 divisions for that range.
BFN Unit 3
- 1. 1/2. 1" 13 Amendment Np. 207
- Thus, as "the IRM is ranged up to accommodate the increase in power level, the scram setting is also ranged up.
A scram at 120 divisions on the IRM instruments remains in effect as long as the reactor is in the startup mode.
In addition, the APRM 15 percent scram prevents higher power operation without being in the RUN mode.
The IRM scram provides protection for changes which occur both locally and over the entire core.
The most significant sources of reactivity change during the power increase are due to control rod withdrawal.
For insequence control rod withdrawal, the rate of change of power is slow enough due to the physical limitation of withdrawing control rods that heat flux is in equilibrium with the neutron flux.
An ZRM scram would result in a reactor shutdown well before any SAFETY LIMIT is exceeded.
For the case of a single control rod withdrawal error, a range of rod withdrawal accidents was analyzed.
This analysis included starting the accident at various power levels.
The most severe case involves an initial condition in which the reactor is just subcritical and the IRM system is not yet on scale.
This condition exists at quarter rod density.
Quarter rod density is discussed in paragraph 7.5.5.4 of the FSAR.
Additional conservatism was taken in this analysis by assuming that the ZRM channel closest to the withdrawn rod is bypassed.
The results of this analysis show that the reactor is scrammed and peak power limited to one percent of rated power, thus maintaining MCPR above the Safety Limit MCPR.
Based on the above analysis, the ZRM provides protection against local control rod withdrawal errors and continuous withdrawal of control rods in sequence.
Tne average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady-state conditions, reads in percent of rated power (3,293 MNt).
The APRM system responds directly to neutron flux.
Licensing analyses have demonstrated that with a neutron flux scram of 120 percent of -rated power, none of the abnormal operational transients analyzed violate the fuel SAFETY LIMIT and there is a substantial margin from fuel damage.
Reactor power level may be varied by moving control rods or by varying the recirculation flow rate.
The APRM system provides a control rod block to prevent rod withdrawal beyond a given point at constant recirculation flow rate and thus prevents scram actuation.
This rod block trip setting, which is automatically varied with recirculation loop flow rate, prevents an BFN Unit 3
1. 1/2. 1-14 Amendment -.No. 207
0
. ~
increase in the reactor power level to excess values due to control rod withdrawal.
The flow variable trip setting is selected to provide adequate margin to the flow-biased scram setpoint.
The setpoint for the low level scram is above the bottom of the separator skirt.
This level has been used in transient analyses dealing with coolant inventory decrease.
The results reported in FSAR subsection 14.5 show that scram and isolation of all process lines (except main steam) at this adequately protects the fuel and the pressure barrier, because MCPR is greater than the Safety Limit MCPR in all cases, and system pressure does not reach the safety valve settings.
The scram setting is sufficiently below normal operating range to avoid spurious scrams.
D.
The turbine stop valve closure trip anticipates the pressure, neutron flux and heat flux increases that would result from closure of the stop valves.
With a trip setting of 10 percent of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained even during the worst case transient that assumes the turbine bypass valves remain closed.
(Reference 2)
Turbine control valve fast closure or turbine trip scram anticipates the
- pressure, neutron flux, and heat flux increase that could result from control valve fast closure due to load rejection or control valve closure due to turbine trip; each without bypass valve capability.
The reactor protection system initiates a scram in less than 30 milliseconds after the start of control valve fast closure due to load rejection or control valve closure due to turbine trip.
This scram is achieved by rapidly reducing hydraulic control oil pressure at the main turbine control valve actuator disc dump valves.
This loss of pressure is sensed by pressure switches whose contacts form the one-out-of-two-twice logic input to the reactor protection system.
This trip setting, a nominally 50 percent greater closure time and a different valve characteristic from that of the turbine stop valve, combine to produce transients very similar to that for the stop valve.
No significant change in MCPR occurs.
Relevant transient analyses are discussed in References 2 and 3 of the Final Safety Analysis Report.
This scram is bypassed when turbine steam flow is below 30 percent of
- rated, as measured by turbine first state pressure.
BFN Unit 3 1.1/2.1-15 Amendment No.
207
3. 3/4. 3 M~ (Cont')
G.
The Rod Block Monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power level operation.
Two RBM channels are
- provided, and one of these may be bypassed from the console for maintenance and/or testing.
Automatic rod withdrawal blocks from one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage.
The specified restrictions with one channel out of service.
conservatively assure that fuel damage will not occur due to rod withdrawal errors when this condition exists.
The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent fuel damage~; i.e., to prevent the MCPR from becoming less than the Safety Limit MCPR.
The limiting power transients are given in Reference 1.
Analysis of these transients shows that the negative reactivity rates resulting from the s ram with the average response of all drives as given in the above specifications provide the required protection and MCPR remains greater than the Safety Limit MCPR.
I On an early
- BWR, some degradation of control rod scram performance occurred during plant STARTUP and was determined to be caused by particulate material (probably construction debris) plugging an internal control rod drive filter.
The design of the present control rod drive (Model 7RDB144B) is grossly improved by the relocation of the filter to a location out of the scram drive path; i.e., it can no longer interfere with scram performance, even if completely blocked.
The degraded performance of the original drive (CRD7RDB144A) under dirty operating conditions and the insensitivity of the redesigned drive (CRD7RDB144B) has been demonstrated by a series of engineering tests under simulated reactor operating conditions.
The successful performance of the new drive under actual operating conditions has also been demonstrated by consistently good in-service test results for plants using the new drive and may be inferred from plants using the older model BFN Unit 3 3.3/4.3-17 Amendment No. 207
3.3/4.3 BASES (Cont'd)
'rive. with a modified (larger screen size) internal filter which is less prone to plugging.
Data has been documented by surveillance reports in various operating plants.
These include Oyster Creek, Monticello, Dresden 2,
and Dresden 3.
Approximately 5000 drive tests have been recorded to date.
Following identification of the "plugged filter" problem, very frequent scram tests were necessary to ensure proper performance.
However, the more frequent scram tests are now considered totally unnecessary and unwise for the following reasons:
l.
Erratic scram performance has been identified as due to an obstructed drive filter in type "A" drives.
The drives in BFNP are of the new "B" type design whose scram performance is unaffected by filter condition.
2.
The dirt load is primarily released during STARTUP of the reactor when the reactor and its systems are first subjected to flows and pressure and thermal stresses.
Special attention and measures are now being taken to assure cleaner systems.
Reactors with'rives identical or similar (shorter stroke, smaller piston areas) have operated through many refueling cycles with no sudden or erratic changes in scram performance.
This preoperational and STARTUP testing is sufficient to detect anomalous drive performance.
3.
The 72-hour outage limit, which initiated the start of the frequent scram testing is arbitrary, having no logical basis other than quantifying a "major outage" which might reasonably be caused by an event so severe as to possibly affect drive performance.
This requirement is unwise because it provides an incentive for shortcut actions to hasten returning "on line" to avoid the additional testing due a 72-hour outage.
BFN Unit 3 3.3/4.3-18 TS 370 Letter Dated 11/17/95
Browns Ferry Nuclear Plant Revised Technical Specification Bases Clarification of Supplemental Spent Fuel Cooling Requirements Unit 1
Unit 2 Unit 3 REMOVE 3.10/4.10-13 3.10/4.10-14 3.10/4.10-13 3.10/4.10-14 3.10/4.10-13 3.10/4.10-14 INSERT 3.10/4.10-13*
3.10/4.10-14 3.10/4.10-13*
3.10/4.10-14 3.10/4.10-13 3.10/4.10-14*
- Denotes overleaf page Enclosure 3
3.10 BASES (Cont'd)
REFERENCES 1.
Refueling interlocks (BFNP FSAR Subsection 7.6)
B.
Core o itorin The SRMs are provided to monitor the core during periods of unit shutdown and to guide the operator during refueling operations and unit startup.
Requiring two OPERABLE SRMs (FLCs) during CORE ALTERATIONS assures adequate monitoring of the fueled region(s) and the core quadrant where CORE ALTERATIONS are being performed.
The fueled region is any set of contiguous (adjacent) control cells which contain one or more fuel assemblies.
An SRM is considered to be in the fueled region when one or more of the four fuel assembly locations surrounding the SRM dry tube contain a fuel assembly.
An FL" is considered to be in the.
fueled region if the FLC is positioned such that it is monitoring the fuel assemblies in its associated core quadrant, even if the actual position of the FLC is outside the fueled region.
Each SRM'(FLC) is not required to read
> 3 cps until after four fuel assemblies have been loaded adjacent to the SRM (FLC) if no other fuel assemblies are in the associated core quadrant.
These four locations are adjacent to the SRM dry tube.
When utilizing FLCs, the FLCs will be located such that the required count rate is achieved without exceeding the SRM upscale setpoint.
With four fuel assemblies or fewer loaded around each SRM, even with a control rod withdrawn, the configuration will not be critical.
Under the special condition of removing the full core with all control rods inserted and electrically disarmed, it is permissible to allow SRM count rate to decrease below three counts per second.
All fuel moves during core unloading will reduce reactivity. It is expected that the SRMs will drop below three counts per second before all of the fuel is unloaded.
Since there will be no reactivity additions during this period, the low number of counts will not present a hazard.
When sufficient fuel has been removed to the spent fuel storage pool to drop the SRM count rate below 3 cps, SRMs will no longer be required to be OPERABLE.
Requiring the SRMs to be functionally tested prior to fuel removal assures that the SRMs will be OPERABLE at the start of fuel removal.
The once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verification of the SRM count rate and signal-to-noise ratio ensures their continued OPERABILITY until the count rate diminishes due to fuel removal.
Control rods in cells from which all fuel has been rem=vsd and which are outside the periphery of the then existing fuel matrix may be armed electrically and moved for maintenance purposes during full core removal, provided all rods that control fuel are fully inserted and electrically disarmed.
REFERE CES 1.
Neutron Monitoring System (BFNP FSAR Subsection 7.5)
BFN Unit 1 3.10/4.10-13 TS 348 TVA Letter to NRC Dated 02/23/95
- Morgan, W. R.,
"In-Core Neutron Monitoring System for General Electric Boiling Water Reactors,"
General Electric Company, Atomic Power Equipment Department, November
- 1968, revised April 1969 (APED-5706)
The design of the spent fuel storage pool provides a storage location for approximately 140 percent of the full core load of fuel assemblies in the reactor building which ensures adequate shielding,
- cooling, and reactivity control of irradiated fuel.
An analysis has been performed which shows that a water level at or in excess of eight and one-half feet over the top of the stored assemblies will provide shielding such that the maximum calculated radio agical doses do not exceed the limits of 10 CFR 20.
The normal water level provides 14-1/2 feet of additional water shielding..
The capacity of the skimmer surge tanks is available to maintain the water level at its normal height for three days in the absence of additional water input from the condensate storage tanks.
All penetrations of the fuel pool have been installed at such a height that their presence does not provide a possible drainage route that could lower the normal water level more than one-half foot.
4 The fuel pool cooling system is designed to maintain the pool water temperature less than 125'F during normal heat loads.
If the reactor core is completely unloaded when the pool contains two previous discharge
- batches, the temperature may increase to greater than 125'F.
The RHR system supplemental fuel pool cooling mode can be used under these conditions to maintain the pool temperature to less than 125'F.
The reactor building crane and 125-ton hoist are required to be operable for handling of the spent fuel in the reactor building.
The controls for the 125-ton hoist are located in the crane cab.
The five-ton has both cab and pendant controls.
A visual inspection of the load-bearing hoist wire rope assures detection of signs of distress or wear so that corrections can be promptly made if needed; The testing of the various limits and interlocks assures their proper operation when the crane is used.
The spent fuel cask design incorporates removable lifting trunnions.
The visual inspection of the trunnions and fasteners prior to BFN
~
Unit 1 3.10/4.10-14 TS 377 - TVA letter to NRC Oated 06/21/96
~
~
3.10 JASPERS (Cont'd)
REFERENCES
- 1. Refueling interlocks (BFNP FSAR Subsection 7.6)
The SRMs are provided to monitor the core during periods of unit shutdown and to guide the operator during refueling operations and unit startup.
Requiring two OPERABLE SRMs (FLCs) during CORE ALTERATIONS assures adequate monitoring of the fueled region(s) and the core quadrant where CORE ALTERATIONS are being performed.
The fueled region is any set of contiguous (adjacent) control cells which contain one or more fuel assemblies.
An SRM is considered to be in the fueled region when one or more of the four fuel assembly locations surrounding the SRM dry tube contain a fuel assembly.
An FLC is considered to be in the fueled region if the FLC is positioned such that it is monitoring the fuel assembli.es in its associated core quadrant, even if the actual position of the FLC is outside the fueled region.
Each SRM (FLC) is not required to read
> 3 cps until after four fuel assemblies have been loaded adjacent to the SRM (FLC) if no other fuel assemblies are in the associated core quadrant.
These four locations are adjacent to the SRM dry tube.
When utilizing FLCs, the FLCs will be locatc9 such that the required count rate is achieved without exceeding the SRM upscale setpoint'.
With four fuel assemblies or fewer loaded around each SRM, even with a control rod withdrawn, the configuration will not be critical.
Under the special condition of removing the full core with all control rods inserted and electrically disarmed, it is permissible to allow SRM count rate to decrease below three counts per second.
All fuel moves during core unloading will reduce reactivity. It is expected that the SRMs will drop below three counts per second before all of the fuel is unloaded.
Since there will be no reactivity additions during this period, the low number of counts will not present a hazard.
When sufficient fuel has been removed to the spent fuel storage pool to drop the SRM count rate below 3 cps, SRMs will no longer be required to be OPERABLE.
Requiring the SRMs to be functionally tested prior to fuel removal assures that the SRMs will be OPERABLE at the start of fuel removal.
The once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verification of the SRM count rate and signal-to-noise ratio ensures their continued OPERABILITY until the count rate diminishes due to fuel removal.
Control rods in cells from which all fuel has been removed and which are outside the periphery of the then existing fuel matrix may be armed electrically and moved for maintenance purposes during full core removal, provided all rods that control fuel are fully inserted and electrically disarmed.
REFERENCES
- 1. Neutron Monitoring System (BFNP FSAR Subsection 7.5)
BFN Unit 2 3.10/4.10-13 TS 348 TVA Letter to ViRC Dated 02/23/95
- 3. 10 ~
(Cont 'd)
- Morgan, W. R.,
"ln-Core Neutron Monitoring System for General Electric Boiling Water Reactors,"
General Electric Company, Atomic Power Equipment Department, November
- 1968, revised April 1969 (APED-5706)
The design of the spent fuel storage pool provides a storage location for approximately 140 percen of the full core load of fuel assemblies in the reactor building which ensures adequate shielding,
- cooling, arid reactivity control of irradiated fuel.
An analysis has been performed which shows that a water level at or in excess of eight and one-half feet over the top of the stored assemblies will provide shielding such that the maximum calculated radiological doses do not exceed the limits of 10 CFR 20.
The normal water level provides 14-1/2 feet of additional water shielding.
The capacity of the skimmer surge tanks is available to maintain the water level at its normal height for three days in the absence of additional water input from the condensate storage tanks.
All penetrations of the fuel pool have been installed at such a height that their presence does not provide a possible drainage route that could lower the normal water level more than one-half foot.
The fuel pool cooling system is designed to maintain the pool water temperature less than 125'F during normal heat loads.,
Zf the reactor core is completely unloaded when the pool contains two previous discharge
- batches, the temperature may increase to greater than 125'F.
The RHR system supplemental fuel pool cooling mode can be used under these conditions to maintain the pool temperature to less than 125'F.
D.
The reactor building crane and 125-ton hoist are required to be operable for handling of the spent fuel in the reactor building.
The controls for the 125-ton hoist are located in the crane cab.
The five-ton has both cab and pendant controls.
A visual inspection of the load-bearing hoist wire rope assures detection of signs of distress or wear so that corrections can be promptly made if needed.
The testing of the various limits and interlocks assures their proper operation when the crane is used.
The spent fuel cask design incorporates removable lifting trunnions.
The visual inspection of the trunnions and fasteners prior to attachment to the cask assures that no visual damage has occurred during prior handling.
The trunnions must be properly attached to the cask for lifting of the cask and the visual inspection assures correct installation.
BFN Unit 2
- 3. 10/4. 10-14 TS 377 - TVA letter to MK Oated 06/21/96
0
3.. 10 ~
(Cont 'd)
- Morgan, W. R.,
"In-Core Neutron Monitoring System for General Electric Boiling Water Reactors,"
General Electric Company, Atomic Power Equipment Department, November
- 1968, revised April 1969 (APED-5706)
The design of the spent'uel storage pool provides a storage location for approximately 140 percent of the full core load of fuel assemblies in the reactor building which ensures adequate shielding,
- cooling, avd reactivity control of irradiated fuel.
An analysis has been performed which shows that a water level at or in excess of eight and one-half feet over the top of the stored assemblies will provide shielding such that the maximum calculated radiological doses do not exceed the limits of 10 CFR 20.
The normal water level provides 14-1/2 feet of additional water shielding.
The capacity of the skimmer surge tanks is available to maintain the water level at its normal height for three days in the absence of additi'on'al water input from the condensate storage tanks.
All penetrations of the fuel pool have been installed at such a height that their presence does not provide a possible drainage route that could lower the normal water level more than one-half foot.
The fuel pool cooling system is designed to maintain the pool water temperature less than 125'F during normal heat loads.
If the reactor core is completely unloaded when the pool contains two previous discharge
- batches, the temperatures may increase to greater than 125'F.
The RHR system supplemental fuel pool cooling mode can be used under these conditions to maintain the pool temperature to less than 125'F.
D.
The reactor building crane and 125-ton hoist are required to be OPERABLE for handling of the spent fuel in the reactor building.
The controls for the 125-ton hoist are located in the crane cab.
The five-ton has both cab and pendant controls.
A visual inspection of the load-bearing hoist wire rope assures detection of signs of distress or wear so that corrections can be promptly made if needed.
The testing of the various limits and interlocks assures their proper operation when the crane is used.
The spent fuel cask design incorporates removable lifting trunnions.
The visual inspection of the trunnions and fasteners prior to BFN Unit 3
3.10/4.10-13 TS 377 - NA letter to NRC Oated 06/n/96
3.10 BASES (0BBB'd) attachment to the cask assures that no visual damage has occurred during prior handling.
The trunnions must be properly attached to the cask for lifting of the cask and the visual inspection assures correct installation.
3.10.F S ent Fuel Cask Handlin
- Refuelin Floor Although single failure protection has been provided in the design of the 125-ton hoist drum shaft, wire ropes, hook and lower block assembly on the reactor building crane, the limiting of lift height of a spent fuel cask controls the amount of energy available in a dropped cask accident when the cask is over the refueling floor.
An. analysis has been made which shows that the floor and support members in the area of cask entry into the decontamination facility can satisfactorily sustain a dropped cask from a height of three feet.
The yoke safety links provide single failure protection for the hook and lower block assembly and limit cask rotation.
Cask rotation is necessary for decontamination and the safety links are removed during decontamination.
4.10 BASES A.
Refuelin Interlocks Complete functional testing of all required refueling equipment interlocks before any refueling outage will provide positive indication that the interlocks operate in the situations for which they were designed.
By loading each hoist with a weight equal to the fuel
- assembly, positioning the refueli.ng platform, and withdrawing control
- rods, the interlocks can be sub)ected to valid operational tests.
Where redundancy is, provided in the logic circuitry, tests can be performed to assure that each redundant logic element can independently perform its function.'.
Core Monitori Requiring the SRMs to be functionally tested prior to any CORE ALTERATION assures that the SRMs will be OPERABLE at the start of that alteration.
The once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verification of the SRM count rate and signal-to-noise ratio ensures their continued OPERABILITY.
REFERENCES 1.
Fuel Pool Cooling and Cleanup System (BFNP FSAR Subsection 10.5) 2.
Spent Fuel Storage (BFNP FSAR Subsection 10.3)
BFN Unit 3
- 3. 10/4. 10-14 TS 348 TVA Letter to NRC Dated 02/23/95