ML20215G539
| ML20215G539 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 06/17/1987 |
| From: | VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | |
| Shared Package | |
| ML20215G537 | List: |
| References | |
| NUDOCS 8706230292 | |
| Download: ML20215G539 (16) | |
Text
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i ATTACHMENT 1 Proposed Technical Specifications Changes - Unit 1 I
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i 8706230292 a70617 PDR ADOCK 05000338 PDR P
b REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITINC CO'DITION FOR OPERATION N
3.1.1.4 The moderator temperature coefficient (MTC) shall be:
a.
For the a11 r ds withdrawn, beginning of core life condition
-4 5 0.6 x 10 Ak/k/ F below 70 percent RATED THERMAL power
~4 s 0.0 x 10 ok/k/ F at or above 70 percent RATED THERMAL POWER
-4 b.
Less negative than -5.0 x 10 ok/k/ F for the all rods withdrawn, end of core life at RATED THERMAL POWER.
APPLICABILITY:
Specification 3.1.1.4.a - MODES I and 2* only#
Specification 3.1.1.4.b - MODES 1, 2 and 3 only#
ACTION:
a.
With the MTC more positive than the limit of 3.1.1.4.a above:
1.
Establish and maintain control rod withdrawal limits sufficient to restore the MTC to within its limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
These withdrawal limits j
shall be in addition to the insertion limits of Specification 3.1.3.6.
2.
Maintain the control rods within the withdrawal limits established above until subsequent measurement verifies that the MTC has been restored to within its limit for the all rods withdrawn condition.
l 3.
Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the 1
measured MTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.
4.
With the MTC more negative than the limit of 3.1.1.4b above, be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- With K 2 1.0 ff
- See Special Test Exception 3.10.3 NORTH ANNA - UNIT 1 3/4 1-6
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' REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT SURVEILLANCE REQUIREMENT 4.1.1.4 The MTC shall be determined to be within its limits during each fuel cycle as follows:
q a.
The MTC shall be measured and compared to the BOL Limit of' j
Specification 3.1.1.4a, above, prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading.
b.
The MTC shall be measured at any THERMAL POWER and compared to -4.0
~4 x
10 Ak/k/ F- (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm.
In the egent this comparison indicated the MTC is more negative
~
than -4.0 x 10 Ak/k/ F, the MTC shall be remeasured, and compared to the EOL' MTC limit of specification 3.1.l EFPDduringtheremainderofthefuelcycle.gb,atleastonceper14 1
(1) Once the equilibrium boron concentration (all rods withdrawn, RATED THERMAL POWER condition) is 60 ppm or less, further measurement of the MTC i
in accordance with 4.1.1.4.b may be suspended providing that the measured 1
MTC at an eq boron concentration of s 60 ppm is less negative than -4.7 x 10~gilibrium Ak/k/ F.
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NORTH ANNA - UNIT 1 3/4 1-6a l
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i 2.1 SAFETY LIMITS BASES
,2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the r.ucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been j
related to DNB through a correlation. The DNB correlation has been developed l
to predict the DNB flux and the location of DNB for axially uniform and j
non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR,
]
defined as the ratio of the heat flux that would cause DNB at a particular cora location to the local heat flux, is indicativa of the margin to DNB.
The DNB design basis is as foll.ows:
there must be at least a 95 percent probability that t'e minimum DNBR of the limiting rod during Condition I and II n
events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 correlation in this application). The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit.
i In meeting this design basis, uncertainties in plant operating parameters, f
nuclear and thermal parametern, and fuel fabrication parameters are considered
]
statistically such that there is at least a 95% probability that the minimum
)
DNBR for the limiting rod is greater than or equal to the DNBR lim.it. The
]
uncertainties in the above plant parameters are used to determine the plant
)
DNBR uncertainty.
This DNBR uncertainty, combined with the correlation DNBR
]
limit, establishes a design DNBR vr.lue which must be met in plant safety analyses using values of input parameters without uncertainties.
As an additional criterion, meeting the DNBR limit also ensures that at least 99.9%
of the core avoids the onset of DNB when the plant is at the DNBR limit.
The curves of Figures 2.1-1, 2.1-2, and 2.1-3 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the design limit DNBR, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.
NORTH ANNA - UNIT 1 B 2-1
3 /.4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC) (Continued) conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.
The most negative MTC value was obtained by incrementally correcting the MTC used in the FSAR analyses to nominal operating conditions.
These corrections involved adding the incremental change in the MTC associated with a core condition
.o f Bank D inserted to an all rods withdrawn condition and an incremental change in MTC to account for measurement uncertainty at RATED THERMAL POWEg conditions. ThesecorrectionsresultgnthelimitingMTCvalue of -5.0 x 10- Ak/k/ F.
The MTC value of -4.0 x 10 Ak/k/ F represents a
~
conservative value (with corrections for burnup and soluble boron) at a core conditionof300ppmequilibriumboronconcentrationandgsobtainedby making these corrections to the limiting MTC value of -5.0 x 10- Ak/k/ F.
Once the equilibrium boron concentration falls below about 60 ppm, dilution operations take an extended amount of time and reliable MTC measurements become more difficult to obtain due to the potential for fluctuating core conditions over the test interval.
For this reason, MTC measurements may be suspended provided the measured MTC value at an equilibrium gli power boron concentration s 60 ppm is less negative than
-4.7 x 10 Ak/k/ F.
The cifference between this value and the limiting MTC value of -5.0 x 10 Ak/k/ F
~
conservatively bounds the maximum credible change in MTC between the 60 ppm equilibrium boron concentration (all rods withdrawn, RATED THERMAL POWER conditions) and the licensed end-of-cycle, including the effect of boron concentration, burnup, and end-of-cycle coastdown.
The surveillance requirements for measurement of the MTC at the beginning
{
and near the end of each fuel cycle are adequate to confirm that the MTC j
remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.
)
3/4.1.1.5 MINIMUM TEMPERATURE FOR CRITICALITY 4
This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 541 F.
This limitation is required to ensure 1) the moderator temperature coefficient is within its analyzed temperature range, 2) the protective instrumentation is within its normal operating range, and 3) the P-12 interlock is above its setpoint.
3/4.1.2 B0 RATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow paths, 4) boric acid transfer pumps, 5) associated heat tracing systems, and 6) an emergency power supply from OPERABLE diesel generators.
NORTH ANNA - UNIT 1 B 3/4 1-2
POWER DISTRIBUTION LIMITS
_ BASES 1
a.
abnormal perturbations in the radial power shape, such as from rod N
misalignment, effect F m re directly than F,
AH q
b.
although rod movement has a direct influence upon limiting F to j
Ibit w thin its limit, such control is not readily available to AH, and c.
errors in prediction for control power shape detected during startup physics tests can be compensated for in F by restricting axial flux Nq distributions. This compensation for F is ess rea y availaMe.
l 6H Fuel rod bowing reduces the value of the DNB ratio. Credit is available to offset this reduction in the margin available between the safety analysis design DNBR value (1.46 for Virginia Electric and Power Company statistical methods) and the limiting design DNBR value (1.26 for Virginia Electric and Power Company statistical methods). A discussion of the rod bow penalty is presented in the FSAR.
j 3/4.2.4 QUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis.
Radial power distribution measurements are made during startup testing and periodically during power operation.
The limit of 1.02 at which corrective action is required provides DNB and linear heat generation rate protection with x-y plane power tilts.
The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misa11gned rod.
In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing the power by 3 1
percentforeachpercentoftiltinekcessof1.0.
For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that
)
the normalized symmetric power distribution is consistent with the QUADPANT POWER TILT RATIO.
The incore detector monitoring is done with a full incore j
flux map or two sets of 4 symmetric thimbles. The two sets of 4 symmetric thimbles is a unique set of 8 detector locations. These locations are C-8, j
E-5, E-11, H-3, H-13, L-5, L-11, and N-8.
)
NORTH ANNA - UNIT 1 B 3/4 2-5 1
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POWER DISTRIBUTION LIMITS BAS'ES 3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters I
f are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses.
The limits are censistent with the initial FSAR assumptions and have been analytically demonstrated adequate to
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maintain a minimum DNBR greater than the design limit throughout each analyzed transient. Measurement uncertainties are accounted for in the DNB design g
margin.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters thru instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The 18 month periodic measurement of the RCS total flow rate is adequate to detect, flow. degradation and ensure correlation of the flow indica, tion channels with measured flow such that the indicated percent flow will provide safficient verification of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.
3/4.2.6 AXIAL POWER DISTRIBUTION The limit on axial power distribution ensures that F will be controlled o
and monitored on a more exact basis through use of the APDMS when operating above P% of RATED THERMAL POWER.
This additional limitation on F is necessar?inordertoprovideassurancethatpeakcladtemperatureswillreOain below ti
~CCS acceptance criteria limit of 2200 F in the event of a LOCA. The value for P is based on the cycle dependent potential violation of the F xK(Z) limit, wheEe K(Z) is the graph shown in Figure 3.2-2.
Theamountofpothntial violation is determined by subtracting 1 from the maximum ratio of the predicted F (Z) analysic (flyspeck) results for a particular fuel cycle to the 0
F xK(Z) limit. This amount of potential violation, in percent, is subtracted g
from 100% to determine the value for P.
If P is equal to 100%, no axial power distribution surveillance is required" P, wi7l not exceed 100%.
I NORTH ANNA - UNIT 1 B 3/4 2-6 l
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ATTACHMENT 2 Proposed Technical Specifications Changes - Unit 2
If REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING. CONDITION FOR OPERATION 3.1.1.4 The mcderator temperature coefficient (MTC) shall be:
a.
For the all rods withdrawn, beginning of core life condition
-4 s 0.6 x 10-4 Ak/k/gF below 70 percent RATED THERMAL POWER s 0.0 x 10 ok/k/ F at or above 70 percent RATED THERMAL POWER
~
b.
Less negative than -5.0 x 10 ok/k/ F for the all rods withdrawn, end of core life at RATED THERMAL POWER.
APPLICABILITY:
Specification 3.1.1.4.a - MODES 1 and 2* only#
Specification 3.1.1.4.b - MODES 1, 2 and 3 only#
ACTION:
a.
With the MTC more positive than the limit of 3.1.1.4.a above, operations in MODES 1 and 2 may proceed provided:
1.
Control rod withdrawal limits are established and maintained sufficient to restore the MTC to within its limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> l or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6.
2.
The control rods are maintained within the withdrawal limits established above until subsequent measurement verifies that the MTC has been restored to within its limit for the all rods withdrawn condition.
3.
Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.
b.
With the MTC more negative than the limit of 3.1.1.4.b above, be in i
HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- With K,gg 2 1.0
- See Special Test Exception 3.10.3 NORTH ANNA - UNIT 2 3/41-5
f-i 1
REACTIVITY CONTROL SYSTEMS
' MODERATOR TEMPERATURE COEFFICIENT SURVEILLANCE REQUIREMENTS 4.1.1.4 The-MTC shall be determined to be within its limits during each fuel cycle as follows:
a.
The MTC shall be measured and compared to the BOL Limit of Specification 3.1.1.4.a above, prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading..
b.
TheMTCshaglbemeasuredatanyTHERMAL'POWERandcomparedto-4.0 x
4 10 Ak/k/ F (all rods. withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm.
In the egent this comparison indicated the MTC is more negative than -4.0 x 10 Ak/k/ F,.the MTC shall be remeasured, and compared to l
~
-the EOL MTC limit of specification 3.1.l EFPDduringtheremainderofthefuelcycle.gb,atleast_onceper14 1
(1) Once the equilibrium boron concentration (all rods withdrawn, RATED THERMAL POWER condition) is 60 ppm or less, further measurement of the MTC in accordance with 4.1.1.4.b may be suspended providing that the measured oggilibrigm boron concentration of s 60 ppm is less negative MTC at an than -4.7 x 10 Ak/k/ F.
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NORTH ANNA - UNIT 2 3/4 1-6
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l 2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and i
possible cladding perforation which would result in the release of fission i
products to the reactor coolant.
Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excessive claddd ig temperatures because of the onset of departure from nucleate boiling (DNB) nd the resultant sharp reduction in heat transfer
{
coefficient.
DNB is not a directly measurable parameter during operation and i
therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through a correlation.
The DNB correlation has been developed l
to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.
The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.
The DNB design basis is as follows:
there must be at least a 95 percent probability that the minimum DNBR of the limiting rod during Condition I and II q
events is greater than or equal to the DNBR limit of the DNB correlation befug l
used (the WRB-1 correlation in this application)._ The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit.
In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95% probability that the minimum i
DNBR for the limiting rod is greater than or equal to the DNBR limit.
The l
uncertainties in the above plant parameters are used to determine the plant l
DNBR uncertainty. This DNBR uncertainty, combined with the correlation DNBR limit, establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties.
As an l
additional criterion, meeting the DNBR limit also ensures that at least 99.9%
of the core avoids the onset of DNB when the plant is at the DNBR limit.
1 The curves of Figures ?.1-1, 2.1-2, and 2.1-3 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for I
which the minimum DNBR is no less than the design limit DNBR, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.
NORTH ANNA - UNIT 2 B 2-1
3/4.1 REACTIVITY CONTROL SYSTEMS BASES i
3/4.1.1.<4 MODERATOR TEMPERATURE COEFFICIENT (MTC)
The limitations on MTC are provided to ensure that the value of this coefficient remains within the limiting conditions assumed for this parameter in the FSAR accident and transient analyses.
The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly. stated will require extrapolation to those conditions in order to permit an accurate comparison.
The most negative MTC value was obtained by incrementally correcting the MTC used in the FSAR analyses to nominal operating conditions.
These corrections involved adding the incremental change in the MTC associated with a core condition of Bank D inserted to an all rods withdrawn condition and an incremental change in MTC to account for measurement uncertainty at RATED THERMAL POWER cogditions. These corrections result in thg limiting MTC value
~
of
-5.0 x 10 dk/k/ F.
The MTC value of -4.0 x 10 Ak/k/ F represents a conservative value (with corrections fos burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and 1s btained h mahng
~4 these corrections to the limiting MTC value of -5.0 x 10 Ak/k/ F.
Once the equilibrium boron concentration falls below about 60 ppm, dilution operations take an extended amount of time and reliable MTC measurements become more difficult to obtain due to the potential for fluctuating core conditions over the test interval.
For this reason, MTC measurements may be suspended provided the measured MTC value at an equilibrium full ~4 power boron concentration s 60 ppm is less negative than
-4.7 x 10 Ak/kfF. The i
~
difference between this value and the limiting MTC value of -5.0 x 10 Ak/k/ F conservatively bounds the maximum credible change in MTC between the 60 ppm equilibrium boron concentration (all rods withdrawn, RATED THERMAL POWER conditions) and the licensed end-of-cycle, including the effect of boron concentration, burnup, and end-of-cycle coastdown.
The surveillance requirements for measurement of the MTC at the beginning and near the end of each fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.
3/4.1.1.5 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 541 F.
This limitation is required to ensure 1) the moderator temperature coefficient is within its analyzed temperature range, 2) the protective instrumentation is within its normal operating range, and 3) the P-12 interlock is above its setpoint, 4) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and 5) the reactor pressure vessel is above its minimum RTNDT temperature.
NORTH ANNA - UNIT 2 B 3/4 1-2
)
i POWER DISTRIBUTION LIMITS I
BASES i
When F is measured, 4% is the appropriate experimental error allowance j
fulfy for a core map taken with the incore detection system. The specified limit for F contains a 4% error allowance. Normal operation will result in measuredh less than or equal to 1.49.
The 4% allowance is based on the a
followingconsk$erations:
abnormal perturbations in the radial power shape, such as from tod a.
N misalignment, effect F m re directly than F,
AH q
b.
although rod novement has a direct influence upon limiting F to n
wfthinitslimit, such control is npt readily available to limit j
F gg, and f
c.
errors in prediction for control power shape detected during startup physics tests can be compensated for in F by restricting axial flux Nq
]
distributions. This compensation for F is less readily available, a
AH Fuel rod bowing reduces the value of the DNB ratio. Credit is available to offset this reduction in the margin available between the safety analysis j
design DNBR value (1.46 for Virginia Electric and Power Company statistical methods) and the limiting decign DNBR value (1.26 for Virginia Electric and Power Company statiotical methods). A discussion of the rod bow penalty is presented in the FSAR.
The hot channel factor F M(Z) is measured periodically and increased by a g
cycle and height dependent power factor, N(Z), to provide assurance that the limit on the hot channel factor, F (Z),
is met.
N(Z) accounts for the n
non-equilibrium effects of normal opdration transients and was determined from expected power control maneuvers over the full range of burnup conditions in the core.
The N(Z) function for normal operation is provided in the Core Surveillance Report per Specification 6.9.1.7.
3/4.2.4 QUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis.
Radial power distribution measurements are made during startup testing and periodically during power operation.
The limit of 1.02 at which corrective action is required provides DNB and linear heat generation rate protection with x-y plane power tilts.
The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned rod.
In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing the power by 3 n
percent for each percent of tilt in eReess of 1.0.
NORTH ANNA - UNIT 2 B 3/4 2-5 l
l I
I l
POWER DISTRIBUTION LIMITS BASES 1
)
For purposes of monitoring QUADRANT POWER ' TILT RATIO when one excore detector is inoperable, the movable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO.
The incore detector monitoring is done with a full incore flux map or two sets of 4 symmetric thimbles. The two sets of 4 symmetric j
thimbles is a unique set of 8 detector locations. These locations are C-8, E-5. E-11, H-3, H-13, L-5, L-11, and N-8.
3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters maintained within the normal steady state envelope of operation assumed in are the transient and accident analyses.
The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR greater than the design limit throughout each analyzed transient.
Measurement uncertainties are accounted for in the DNB design
]
margin.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters thru instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.
The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis, s
NORTH ANNA - UNIT 2 B 3/4 2-6 4
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