ML13281A394: Difference between revisions
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the reactivity balance ensures the core is operating as | the reactivity balance ensures the core is operating as | ||
designed. This Specification does not a pply in MODE 2 because enough operating margin exists to limit the effects of a reactivity anomaly, and THERMAL POWER is low enough | designed. This Specification does not a pply in MODE 2 because enough operating margin exists to limit the effects of a reactivity anomaly, and THERMAL POWER is low enough | ||
( 5% RTP) such that reactivity anomalies are unlikely to occur. This Specification does not apply in MODEs 3, 4, and 5 bec ause the reactor is shut down and the reactivity balance is not changing. | ( 5% RTP) such that reactivity anomalies are unlikely to occur. This Specification does not apply in MODEs 3, 4, and 5 bec ause the reactor is shut down and the reactivity balance is not changing. | ||
In MODE 6, fuel loading results in a continually changing | In MODE 6, fuel loading results in a continually changing | ||
Line 1,986: | Line 1,986: | ||
and experiments, are specified in Reference 1, 10 CFR 50.59 | and experiments, are specified in Reference 1, 10 CFR 50.59 | ||
. The key objectives of a test program (Reference | . The key objectives of a test program (Reference | ||
: 2) are to: a. Ensure that the facility has been adequately designed; | : 2) are to: a. Ensure that the facility has been adequately designed; | ||
: b. Validate the analytical models used in design and | : b. Validate the analytical models used in design and | ||
Line 2,209: | Line 2,209: | ||
and exper iments, are specified in Reference 1, 10 CFR 50.59 | and exper iments, are specified in Reference 1, 10 CFR 50.59 | ||
. The key objectives of a test program (Reference | . The key objectives of a test program (Reference | ||
: 2) are to: a. Ensure that the facility has been adequately designed; | : 2) are to: a. Ensure that the facility has been adequately designed; | ||
: b. Validate the analytical models used in design and | : b. Validate the analytical models used in design and |
Revision as of 15:06, 28 April 2019
ML13281A394 | |
Person / Time | |
---|---|
Site: | Calvert Cliffs |
Issue date: | 09/24/2013 |
From: | Constellation Energy Nuclear Group |
To: | Office of Nuclear Reactor Regulation |
Shared Package | |
ML13281A389 | List: |
References | |
Download: ML13281A394 (50) | |
Text
SDM B 3.1.1 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM)
BASES CALVERT CLIFFS
- UN ITS 1 & 2 B 3.1.1-1 Revision 2 BACKGROUND The reactivity control systems must be redundant and capable of holding the reactor core subcritical when shut down under
cold conditions, in accordance with Reference 1
, Appendix 1C, Criteria 27, 29, and 30
. Maintenance of the SDM ensures that postulated reactivity events will not
damage the fuel.
SHUTDOWN MARGIN requirements provide
sufficient reactivity margin to ensure that acceptable fuel
design limits will not be exceeded for normal shutdown and
anticipated operational occurrences (AOOs).
As such, the
SDM defines the degree of subcriticality that would be obtained immediately following the insertion of all control element assemblies (CEAs), assuming the single CEA of
highest reactivity worth is fully withdrawn.
The system design require s that two independent reactivity
control systems be provided, and that one of these systems
be capable of maintaining the core subcritical under cold
conditions. These requirements are provided by the use of movable CEAs and soluble boric acid in the Rea ctor Coolant System (RCS). The CEA System provides the SDM during power
operation and is capable of making the core subcritical
rapidly enough to prevent exceeding acceptable fuel damage
limits, assuming that the CEA of highest reactivity worth
remains fu lly withdrawn.
The soluble boron system can compensate for fuel depletion
during operation and all xenon burnout reactivity changes, and maintain the reactor subcritical under cold conditions.
During power operation, SDM control is ensured by operating
with the shutdown CEAs fully withdrawn and the regulating
CEAs within the limits of Limiting Condition for Operation (LCO) 3.1.6. When the unit is in the shutdown and refueling MODE s, the SDM requirements are met by means of adjustments
to the RCS boron concentration.
APPLICABLE The minimum required SDM is assumed as an initial condition SAFETY ANALYSIS in safety analysis. The safety analysis (Reference 1, Section 3.4) establishes a SDM that ensures specified acceptable fuel design limits (SAFDLs) are n ot exceeded for SDM B 3.1.1 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.1-2 Revision 43 normal operation and AOOs, with the assumption of the highest worth CEA stuck out following a reactor trip. For
MODE 5, the primary safety analysis that relies on the SDM
limit is the boron dilution analysis.
The acceptance criteria fo r the SDM requirements are that
SAFDLs are maintained. This is done by ensuring that:
- a. The reactor can be made subcritical from all operating
conditions, transients, and Design Basis Events;
- b. The reactivity transients associated with postulated
acci dent conditions are controllable within acceptable
limits (departure from nucleate boiling ratio [DNBR],
fuel centerline temperature limit AOOs, and an
acceptable energy deposition for the CEA ejection
accident [Reference 1, Chapter 14]); and
- c. The react or will be maintained sufficiently subcritical
to preclude inadvertent criticality in the shutdown
condition.
The most limiting accident for the SDM requirements are
based on a main steam line break (MSLB) or an Excess Load event (with failure of an MSIV to close), as described in the accident analysis (Reference 1, Chapter 14). The
increased steam flow causes an increased energy removal from the affected steam generator, and consequently the RCS.
This results in a reduction of the reactor coolant
tempe rature. The resultant coolant shrinkage causes a
reduction in pressure. In the presence of a negative
moderator temperature coefficient (MTC), this cooldown
causes an increase in core reactivity. As RCS temperature
decreases, the severity of the event d ecreases. The most limiting MSLB, with respect to potential fuel damage before
a reactor trip occurs, is a guillotine break of a main steam
line out side containment, init iated at the end of core life.
Following the MSLB or Excess Load event , a post-trip return to power may occur; however, no fuel damage occurs as a
result of the post
-trip return to power, and THERMAL POWER
does not violate the Safety Limit (SL) requirement of
SL 2.1.1. The limiting Excess Load event with respect to potential return
-to-po wer after reactor trip is the opening of all steam dump and bypass valves at full power with failure of an MSIV to close.
SDM B 3.1.1 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.1-3 Revision 43 In addition to the limiting MSLB transient, the SDM
requirement for MODEs 3 and 4 must also protect against an
uncontrolled CEA with drawal from a hot zero power or low
power condition, and a CEA ejection.
In the boron dilution analysis, the required SDM defines the
reactivity difference between an initial subcritical boron
concentration and the corresponding critical boron
concentrat ion. These values, in conjunction with the
configuration of the RCS and the assumed dilution flow rate, directly affect the results of the analysis. This event is
most limiting at the beginning of core life when critical boron concentrations are highest.
The withdrawal of CEAs from hot zero power or low power conditions adds reactivity to the reactor core, causing both
the core power level and heat flux to increase with
corresponding increases in reactor coolant temperatures and
pressure. The withdrawa l of CEAs also produces a time
-dependent redistribution of core power.
The uncontrolled CEA withdrawal transient is terminated by the Variable High Power Trip.
In all cases, power level, RCS pressure, linear heat rate (LHR), and the DNBR do not exceed a llowable limits.
SHUTDOWN MARGIN satisfies 10 CFR 50.36(c)(2)(ii),
Criterion 2. LCO The MSLB (or the Excess Load event) and the boron dilution accidents (Reference 1, Chapter 14) are the most limiting
analyses that establish the SDM value of the LCO. F or MSLB accidents (or the Excess Load event), if the LCO is violated, there is a potential to exceed the DNBR limit and
to exceed the acceptance criteria given in Reference 1,
Chapter 14. For the boron dilution accident, if the LCO is
violated, the minimu m required time assumed for operator
action to terminate dilution may no longer be applicable.
Because both initial RCS level and the dilution flow rate
also significantly impact the boron dilution event in MODE 5
with pressurizer level < 90 inches from t he bottom of the SDM B 3.1.1 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.1-4 Revision 43 pressurizer, the LCO also includes limits for these parameters during these conditions.
SHUTDOWN MARGIN is a core physics design condition that can
be ensured through CEA positioning (regulating and shutdown
CEA) in MODEs 1 and 2 and thr ough the soluble boron
concentration in all other MODEs.
APPLICABILITY In MODEs 3, 4, and 5, the SDM requirements are applicable to
provide sufficient negative reactivity to meet the
assumptions of the safety analyses discussed above. In
MODEs 1 and 2, S DM is ensured by complying with LCOs 3.1.5 and 3.1.6. In MODE 6, the shutdown reactivity requirements
are given in LCO
3.9.1. ACTIONS
A.1, A.2, and A.3 With non-borated water sources of >
88 gpm available, while
the unit is in MODE 5 with the pressurize r level
< 90 inches, the consequences of a boron dilution event may
exceed the analysis results. Therefore, action must be
initiated immediately to reduce the potential for such an
event. To accomplish this, Required Action A.1 requires
immediate suspens ion of positive reactivity additions.
However, since Required Action A.1 only reduces the
potential for the event and does not eliminate it, immediate
action must also be initiated to increase the SDM to
compensate for the non
-borated water sources (Requi red Action A.2). Finally, Required Action A.3 requires periodic
verification, once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, that the SDM increase is
maintained sufficient to compensate for the additional
sources of non
-borated water. Required Action A.1 is
modified by a Note indic ating that the suspension of
positive reactivity additions is not required if SDM has
been sufficiently increased to compensate for the additional
sources of non
-borated water. The immediate Completion Time
reflects the urgency of the corrective actions.
The periodic Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is considered
reasonable, based on other administrative controls available
and operating experience.
SDM B 3.1.1 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.1-5 Revision 27 B.1 and B.2 With the RCS level at or below the bottom of the hot leg nozzles, while the unit is in MODE 5 wit h the pressurizer
level < 90 inches, the consequences of a boron dilution
event may exceed the analysis results. Therefore, action
must be initiated immediately to reduce the potential for
such an event. To accomplish this, Required Action B.1
requires i mmediate suspension of operations involving positive reactivity additions that could result in loss of the required SDM. Suspending positive reactivity additions that could result in failure to meet the minimum SDM limit is required to assure continued sa fe operation.
Introduction of coolant inventory must be from sources that have boron concentration greater than tha t required in the RCS for the minimum SDM. This may result in an overall reduction in RCS boron concentration , but provides an acceptable m argin to maintaining subcritical operation. Introduction of temperature changes including temperature increases when operating with a positive MTC must also be evaluated to ensure they do not result in a loss of the required SDM. However, since Required Action B.1 only reduces the potential for the event and does not eliminate
it, immediate action must also be initiated to increase the
RCS level to above the bottom of the hot leg nozzles (Required Action B.2). The immediate Completion Time
reflects the u rgency of the corrective actions.
C.1 If the SDM requirements are not met for reasons other than
addressed in Condition A or B, boration must be initiated
promptly. A Completion Time of immediately is required to
meet the assumptions of the safety anal ysis. It is assumed
that boration will be continued until the SDM requirements
are met.
In the determination of the required combination of boration
flow rate and boron concentration, there is no unique
requirement that must be satisfied. Since it is i mperative
to raise the boron concentration of the RCS as soon as
possible, the boron concentration should be a highly
concentrated solution, such as that normally found in the
boric acid storage tank or the refueling water tank. The SDM B 3.1.1 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.1-6 Revision 27 operator should borate with the best source available for the plant conditions. However, as a minimum, the boration flow rate shall be 40 gpm and the boron concentration shall be 2300 ppm boric acid solution or equivalent.
Assuming that a value of 1%
k/k must be recover ed and a boration flow rate of 40 gpm from the boric acid storage tank, it is possible to increase the boron concentration of
the RCS by 100 ppm in approximately 15 minutes. If an inverse boron worth of 100 ppm/% k/k is assumed, this combination of param eters will increase the SDM by 1%
k/k. These boration parameters of 40 gpm and 100 ppm represent typical values and are provided for the purpose of offering
a specific example.
SURVEILLANCE SR 3.1.1.1 REQUIREMENTS
SHUTDOWN MARGIN is verified by perform ing a reactivity
balance calculation, considering the listed reactivity
effects: a. RCS boron concentration;
- b. CEA positions;
- c. RCS average temperature;
- d. Fuel burnup based on gross thermal energy generation;
- e. Xenon concentration;
- f. Samarium co ncentration; and
- g. Isothermal temperature coefficient.
Using the isothermal temperature coefficient accounts for
Doppler reactivity in this calculation because the reactor
is subcritical and the fuel temperature will be changing at
the same rate as the RCS.
The Frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is based on the generally slow
change in required boron concentration, and also allows
sufficient time for the operator to collect the required
data, which includes performing a boron concentration
analysis, and complete t he calculation.
SDM B 3.1.1 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.1-7 Revision 27 SR 3.1.1.2 and SR 3.1.1.3 These Surveillance Requirements (SRs) periodically verify the significant assumptions of a boron dilution event are maintained. A non
-borated water source of 88 gpm allows for only one charging pump to be cap able of injection during these conditions since each charging pump is capable of an
injection rate of 46 gpm. Each SR is modified by a Note
indicating that it is only required when the unit is in
MODE 5 with the pressurizer level <
90 inches. Since the
a pplicable conditions for the SR may be attained while
already in MODE 5, each SR is provided with a Frequency of
once within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after achieving MODE 5 with pressurizer
level < 90 inches. This provides a short period of time to
verify compliance after the conditions are attained.
Additionally, each SR must be completed once each 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
after the initial verification. The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
is considered reasonable, in view of other administrative
controls available and operating experience.
REFE RENCES 1. Updated Final Safety Analysis Report (UFSAR)
Reactivity Balance B 3.1.2 B 3.1 REACTIVITY CONTROL SYST EMS B 3.1.2 Reactivity Balance
BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.2-1 Revision 2 BACKGROUND According to Reference 1 , Appendix 1C, Criteria 27, 29, and 30 , reactivity shall be controllable, such that, subcriticality is maintained under cold conditions, and acceptable fuel design limits are not exceeded during normal
operation and AOOs. Therefore, reactivity balance is used as a measure of the predicted versus measured core
reactivity during power operation. The periodic
confirmation of core reactivity is necessary to ensure that
Design Basis Accident (DBA) and transient safety analyse s
remain valid. A large reactivity difference could be the
result of unanticipated changes in fuel, CEA worth, or operation at conditions not consistent with those assumed in
the predictions of core reactivity, and could potentially
result in a loss of SD M or violation of acceptable fuel
design limits. Comparing predicted versus measured core
reactivity validates the nuclear methods used in the safety
analysis and supports the SDM demonstrations (LCO 3.1.1) in ensuring the reactor can be brought safely to cold, subcritical conditions.
When the reactor core is critical or in normal power operation, a reactivity balance exists and the net
reactivity is zero. A comparison of predicted and measured
reactivity is convenient under such a balance, since
parame ters are being maintained relatively stable under
steady state power conditions. The positive reactivity
inherent in the core design is balanced by the negative
reactivity of the control components, thermal feedback, neutron leakage, and materials in the core that absorb
neutrons, such as burnable absorbers producing zero net
reactivity. Excess reactivity can be inferred from the
critical boron curve, which provides an indication of the
soluble boron concentration in the RCS versus cycle burnup.
Periodic measurement of the RCS boron concentration for
comparison with the predicted value with other variables
fixed (such as CEA height, temperature, pressure, and power)
provides a convenient method of ensuring that core
reactivity is within design expectation s, and that the
calculational models used to generate the safety analysis
are adequate.
Reactivity Balance B 3.1.2 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.2-2 Revision 2 In order to achieve the required fuel cycle energy output, the uranium enrichment in the new fuel loading and in the
fuel remaining from the previous cycle provides e xcess positive reactivity beyond that required to sustain steady
state operation throughout the cycle. When the reactor is
critical at hot full power, the excess positive reactivity
is compensated by burnable absorbers (if any), CEAs, whatever neutron poi sons (mainly xenon and samarium) are
present in the fuel, and the RCS boron concentration.
When the core is producing THERMAL POWER, the fuel is being
depleted and excess reactivity is decreasing. As the fuel
depletes, the RCS boron concentration is red uced to decrease
negative reactivity and maintain constant THERMAL POWER.
The critical boron curve is based on steady state operation
at RATED THERMAL POWER (RTP). Therefore, deviations from the predicted critical boron curve may indicate deficiencies
in the design analysis, deficiencies in the calculational
models, or abnormal core conditions, and must be evaluated.
APPLICABLE Accurate prediction of core reactivity is either an explicit SAFETY ANALYSES or implicit assumption in the accident analysis eva luations.
Most accident evaluations (Reference 1, Section 14.1) are, therefore, dependent upon accurate evaluation of core
reactivity. In particular, SDM and reactivity transients, such as CEA withdrawal accidents or CEA ejection accidents, are very sens itive to accurate prediction of core
reactivity. These accident analysis evaluations rely on
computer codes that have been qualified against available
test data, operating plant data, and analytical benchmarks.
Monitoring reactivity balance additionally ensures that the
nuclear methods provide an accurate representation of the
core reactivity.
Design calculations and safety analyses are performed for
each fuel cycle for the purpose of predetermining reactivity
behavior and the RCS boron concentration re quirements for
reactivity control during fuel depletion.
The comparison between measured and predicted initial core
reactivity provides a normalization for calculational models
used to predict core reactivity. If the measured and Reactivity Balance B 3.1.2 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.2-3 Revision 2 predicted RCS boron con centrations for identical core conditions at beginning
-of-cycle (BOC) do not agree, the assumptions used in the reload cycle design analysis or the
calculational models used to predict soluble boron
requirements may not be accurate. If reasonable agreemen t
between measured and predicted core reactivity exists at
BOC, the prediction may be normalized to the measured boron concentration. Thereafter, any significant deviations in
the measured boron concentration from the predicted critical
boron curve that d evelop during fuel depletion may be an
indication that the calculational model is not adequate for
core burnups beyond BOC, or that an unexpected change in
core conditions has occurred.
The normalization of predicted RCS boron concentration to
the measur ed value is typically performed after reaching RTP
following startup from a refueling outage, with the CEAs in
their normal positions for power operation. The
normalization is performed at BOC conditions, so that core
reactivity relative to predicted valu es can be continually
monitored and evaluated as core conditions change during the
cycle. The reactivity balance satisfies 10 CFR 50.36(c)(2)(ii),
Criterion 2. LCO The reactivity balance limit is established to ensure plant
operation is maintained withi n the assumptions of the safety
analyses. Large differences between actual and predicted
core reactivity may indicate that the assumptions of the DBA
and transient analyses are no longer valid, or that the
uncertainties in the nuclear design methodology a re larger
than expected. A limit on the reactivity balance of
+/- 1% k/k has been established, based on engineering judgment. A 1% deviation in reactivity from that predicted is larger than expected for normal operation and should
, therefore , be evaluated
. When measured core reactivity is within 1%
k/k of the predicted value at steady state thermal conditions, the core is considered to be operating within acceptable design
limits. Since deviations from the limit are normally
detected by comparing predi cted and measured steady state Reactivity Balance B 3.1.2 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.2-4 Revision 2 RCS critical boron concentrations, the difference between measured and predicted values would be approximately 100 ppm (depending on the boron worth) before the limit is reached.
These values are well within the uncertainty limits for
analysis of boron concentration samples, so that spurious
violations of the limit due to uncertainty in measuring the
RCS boron concentration are unlikely.
APPLICABILITY The limits on core reactivity must be maintained during
MODE 1 because a reactivity balance must exist when the
reactor is critical or producing THERMAL POWER. As the fuel
depletes, core conditions are changing, and confirmation of
the reactivity balance ensures the core is operating as
designed. This Specification does not a pply in MODE 2 because enough operating margin exists to limit the effects of a reactivity anomaly, and THERMAL POWER is low enough
( 5% RTP) such that reactivity anomalies are unlikely to occur. This Specification does not apply in MODEs 3, 4, and 5 bec ause the reactor is shut down and the reactivity balance is not changing.
In MODE 6, fuel loading results in a continually changing
core reactivity. Boron concentration requirements (LCO 3.9.1) ensure that fuel movements are performed within the bounds of the safety analysis. A SDM demonstration is
required during the first startup following operations that
could have altered core reactivity (e.g., fuel movement, or
CEA replacement, or shuffling).
ACTIONS A.1 and A.2 Should an anomaly develop between measured and predicted
core reactivity, an evaluation of the core design and safety
analysis must be performed. Core conditions are evaluated
to determine their consistency with input to design
calculations. Measured core and process parameters are
evalu ated to determine that they are within the bounds of
the safety analysis, and safety analysis calculational
models are reviewed to verify that they are adequate for representation of the core conditions. The required Completion Time of 7 days is based on the low probability of
a DBA occurring during this period, and allows sufficient Reactivity Balance B 3.1.2 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.2-5 Revision 2 time to assess the physical condition of the reactor and to complete the evaluation of the core design and safety analysis. Following evaluations of the core design and safe ty analysis, the cause of the reactivity anomaly may be
resolved. If the cause of the reactivity anomaly is a
mismatch in core conditions at the time of RCS boron
concentration sampling, a recalculation of the RCS boron concentration requirements may be p erformed to demonstrate
that core reactivity is behaving as expected. If an
unexpected physical change in the condition of the core has
occurred, it must be evaluated and corrected, if possible.
If the cause of the reactivity anomaly is in the calculatio n
technique, the calculational models must be revised to provide more accurate predictions. If any of these results
are demonstrated, and it is concluded that the reactor core
is acceptable for continued operation, the boron letdown curve may be renormali zed, and power operation may continue.
If operational restrictions or additional SRs are necessary
to ensure the reactor core is acceptable for continued
operation, they must be defined.
The required Completion Time of 7 days is adequate for
preparing w hatever operating restrictions or SR s may be required to allow continued reactor operation.
B.1 If the core reactivity cannot be restored to within the 1% k/k limit, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The
allowed Completion Time is reasonable, based on operating
experience, for reaching MODE 2 from full power conditions
in an orderly manner and without challenging plant systems.
SURV EILLANCE SR 3.1.2.1 REQUIREMENTS Core reactivity is verified by periodic comparisons of
measured and predicted RCS boron concentrations. The
comparison is made considering that other core conditions
are fixed or stable
, including CEA position, moderator Reactivity Balance B 3.1.2 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.2-6 Revision 3 temperature, fuel temperature, fuel depletion, xenon concentration, and samarium concentration. The SR is
performed prior to entering MODE 1 as an initial check on
core conditions and design calculations at BOC and every
31 days after 60 effective full power days (EFPD). The SR
is modified by two Notes. The Note in the SR column
indicates that the normalization of predicted core
reactivity to the measured value may take place within the first 60 EFPD after each fuel loading. This allows
sufficient tim e for core conditions to reach steady state, but prevents operation for a large fraction of the fuel
cycle without establishing a benchmark for the design
calculations. The required subsequent Frequency of 31 EFPD following the initial 60 EFPD, after ente ring MODE 1, is acceptable, based on the slow rate of core changes due to
fuel depletion and the presence of other indicators (e.g., quadrant power tilt ratio, etc.) for prompt indication of an anomaly. The Frequency Note, "only required after 60 EFPD aft er each fuel loading," is added to
the Frequency column to allow this.
REFERENCES
- 1. UFSAR MTC B 3.1.3 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.3 Moderator Temperature Coeff icient (MTC)
BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.3-1 Revision 29 BACKGROUND The MTC relates a change in core reactivity to a change in reactor coolant temperature. A positive MTC means that
reactivity increases with increasing moderator temperature;
conversely, a negative MTC means that reactivity decreases
with increa sing moderator temperature. The reactor is
designed to operate with a negative MTC over a large range
of fuel cycle operation. Therefore, a coolant temperature
increase will cause a reactivity decrease, so that the
coolant temperature tends to return tow ard its initial
value. Reactivity increases that cause a coolant
temperature increase will thus be self limiting, and stable
power operation will result.
Moderator temperature coefficient values are predicted at
selected burnups during the safety evalua tion analysis and
are confirmed to be acceptable by measurements.
R eload cores are designed so that the MTC is less positive than that allowed by the LCO. The actual value of the MTC is dependent on core characteristics, such as fuel loading and reactor coolant soluble boron concentration. The core
design may require additional fixed distributed poisons (burnable poison) to yield an MTC at the BOC within the
range analyzed in the plant accident analysis. The end
-of-cycle (EOC) MTC is also limited by the requirements of the
accident analysis. Fuel cycles that are designed to achieve
high burnups or that have changes to other characteristics
are evaluated to ensure that the MTC does not exceed the EOC
limit.
APPLICABLE The acceptance criteria for the spe cified MTC are:
SAFETY ANALYSES
- a. The MTC values must remain within the bounds of those
used in the accident analysis (Reference 1, Section 14.2.2); and
- b. The MTC must be such that inherently stable power
operations result during normal operation and d uring accidents, such as overheating and overcooling events.
MTC B 3.1.3 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.3-2 Revision 43 Reference 1 , Section 14.1.2 contains analyses of accidents that result in both overheating and overcooling of the
reactor core. Moderator temperature coefficient is one of
the controlling par ameters for core reactivity in these
accidents. Both the most positive value and most negative
value of the MTC are important to safety, and both values
must be bounded. Values used in the analyses consider
worst-case conditions, such as very large solub le boron concentrations, to ensure the accident results are bounding.
Accidents that cause core overheating, either by decreased
heat removal or increased power production, must be
evaluated for results when the MTC is positive. Reactivity
accidents tha t cause increased power production include the
CEA withdrawal and CEA ejection transient s from either zero or full THERMAL POWER. The limiting overheating event
relative to plant response is based on the maximum
difference between core power and steam gen erator heat
removal during a transient. The most limiting event with
respect to a positive MTC is a CEA ejection accident from full power (Reference 1, Section 14.
13).
Accidents that cause core overcooling must be evaluated for
results when the MTC is m ost negative. The event that
produces the most rapid cooldown of the RCS, and is
therefore the most limiting event with respect to the
negative MTC, is a steam line break (SLB) event. Following
the reactor trip for the postulated EOC SLB event, the large
moderator temperature reduction combined with the large
negative MTC may produce reactivity increases that are as
much as the shutdown reactivity. When this occurs, a
substantial fraction of core power is produced with all CEAs
inserted, except the most reactive one, which is assumed
withdrawn. Even if the reactivity increase produces
slightly subcritical conditions, a large fraction of core
power may be produced through the effects of subcritical
neutron multiplication.
Moderator temperature coefficie nt values are bounded in
reload safety evaluations assuming steady state conditions
at BOC , peak RCS boron, and EOC. A 2/3 core burnup MTC
measurement is conducted and the measured value may be MTC B 3.1.3 BASES CALVERT CL IFFS - UNITS 1 & 2 B 3.1.3-3 Revision 29 extrapolated to project the EOC value, in order to confirm reload design predictions.
The MTC satisfies 10 CFR 50.36(c)(2)(ii), Criterion
- 2.
LCO Limiting Condition for Operation
3.1.3 requires
the MTC to
be within specified limits of the Core Operating Limits
Report (COLR), with the maximum positive limit speci fied in Figure 3.1.3
-1, to ensure the core operates within the
assumptions of the accident analysis. During the reload
core safety evaluation, the MTC is analyzed to determine
that its values remain within the bounds of the original
accident analysis duri ng operation. The limit on a positive
MTC ensures that core overheating accidents will not violate
the accident analysis assumptions. The negative MTC limit for EOC specified in the COLR ensures that core overcooling accidents will not violate the accide nt analysis
assumptions.
Moderator temperature coefficient is a core physics
parameter determined by the fuel and fuel cycle design and
cannot be easily controlled once the core design is fixed.
During operation, therefore, the LCO can only be ensured
t hrough measurement. The surveillance checks at BOC and
2/3 core burnup provide confirmation that the MTC is behaving as anticipated, so that the acceptance criteria are
met. APPLICABILITY In MODE 1, the limits on the MTC must be maintained to
ensure that any accident initiated from THERMAL POWER
operation will not violate the design assumptions of the
accident analysis. In MODE 2, the limits must also be
maintained to ensure startup accidents, such as the
uncontrolled CEA or group withdrawal, will not vi olate the
assumptions of the accident analysis. In MODEs 3, 4, 5, and 6, this LCO is not applicable, since no DBAs using the
MTC as an analysis assumption are initiated from these
MODEs. However, the variation of the MTC, with temperature
in MODEs 3, 4, and 5 for DBAs initiated in MODEs 1 and 2, is accounted for in the accident analysis. The variation of the MTC, with temperature assumed in the safety analysis, is
accepted as valid once the BOC and 2/3 core burnup measurements are used for normalization.
MTC B 3.1.3 BASES CALVERT CL IFFS - UNITS 1 & 2 B 3.1.3-4 Revision 29 ACTIONS A.1 Moderator temperature coefficient is a function of the fuel and fuel cycle designs, and cannot be controlled directly
once the designs have been implemented in the core. If MTC
exceeds its limits, the reactor must be placed in MODE
- 3.
This eliminates the potential for violation of the accident
analysis bounds. The associated Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, considering the probability of an accident
occurring during the time period that would require an MTC
value within the LCO li mits, and the time for reaching
MODE 3 from full power conditions in an orderly manner and
without challenging plant systems.
SURVEILLANCE SR 3.1.3.1 and SR 3.1.3.2 REQUIREMENTS The SRs for measurement of the MTC at the beginning and 2/3 core burnup of e ach fuel cycle provide for confirmation of the limiting MTC values. The MTC varies with boron concentration during fuel cycle operation. The MTC becomes more negative as the RCS boron concentration is reduced. The requirement for measurement prior to en tering MODE 1 after each fuel loading satisfies the confirmatory check on the most positive (least negative) MTC value. The
requirement for measurement, within 7 EFPD of initially reaching an equilibrium condition with THERMAL POWER 90% RTP, and within 7 EFPD of reaching 2/3 core burnup, satisfies the confirmatory check of the most negative MTC value. The 2/3 core burnup measurement is performed at any THERMAL POWER, so that the projected EOC MTC may be
evaluated before the reactor actually reaches the EOC condition. Moderator temperature coefficient values may be
extrapolated and compensated to permit direct comparison to
the specified MTC limits.
Surveillance Requirement 3.1.3.2 is modified by a Note, which indicates that if the extrapolated MTC is more negative than the EOC COLR limit, the SR may be repeated, and that shutdown must occur prior to exceeding the minimum
allowable boron concentration at which MTC is projected to
exceed the lower limit. An engineering evaluation is MTC B 3.1.3 BASES CALVERT CL IFFS - UNITS 1 & 2 B 3.1.3-5 Revision 29 performed if the ext rapolated value of MTC exceeds the Specification limits.
REFERENCES
- 1. UFSAR CEA Alignment B 3.1.4 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.4 Control E lement Assembly (CEA) Alignment
BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.4-1 Revision 2 BACKGROUND The OPERABILITY (e.g., trippability) of the shutdown and regulating CEAs is an initial assumption in all safety
analyses that assume CEA insertion upon reactor trip.
The applicable criteria for these reactivity and power
distribution design re quirements are found in Reference 1 , Appendix 1C, Criteria 6, 27, 29, and 30
, and Reference 2. Mechanical or electrical failures may cause a CEA to become
inoperable or to become misaligned from its group.
Control element assembly inoperability or misal ignment may cause increased power peaking, due to the asymmetric reactivity
distribution and a reduction in the total available CEA
worth for reactor shutdown. Therefore, CEA alignment and
OPERABILITY are related to core operation in design power
peaking limits and the core design requirement of a minimum
SDM. Limits on CEA alignment and OPERABILITY have been established, and all CEA positions are monitored and
controlled during power operation to ensure that the power
distribution and reactivity limits defined by the design
power peaking and SDM limits are preserved.
Control element assemblies are moved by their control element drive mechanisms (CEDMs). Each CEDM moves its CEA
one step (approximately 3/4
-inch) at a time.
The CEAs are arranged into g roups that are radially
symmetric. Therefore, movement of the CEA groups do not
introduce radial asymmetries in the core power distribution.
The shutdown and regulating CEAs provide the required
reactivity worth for immediate reactor shutdown upon a
reac tor trip. The regulating CEAs also provide reactivity (power level) control during normal operation and
The axial position of shutdown and regulating CEAs is
indicated by two separate and independent systems, which are
CEA Alignment B 3.1.4 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.4-2 Revision 2 the Plant Computer CEA Position Indication System and the Reed Switch Position Indication System.
The Plant Computer CEA Position Indication System counts the
commands sent to the CEA gripper coils from the CEDM Control
System that moves the CEAs. There is a one step counter for each CEA. Individual CEAs in a group all receive the same
signal to move and should, therefore, all be at the same
position indicated by the group step counter for that group.
Plant Computer CEA Position Indication System is considered
highly precis e (+/- 1 step or +/- 3/4
-inch). If a CEA does not move one step for each command signal, the step counter will
still count the command and incorrectly reflect the position
of the CEA.
The Reed Switch Position Indication System provides a highly
accurate ind ication of actual CEA position, but at a lower
precision than the step counters. This system is based on
inductive analog signals from a series of reed switches
spaced along a tube with a center
-to-center distance of 1.5 inches, which is two steps. To in crease the reliability
of the system, there are redundant reed switches at each
position. APPLICABLE Control element assembly misalignment accidents are SAFETY ANALYSES analyzed in the safety analysis (Reference 1 , Section s 14.2, 14.11, and 14.13
). The a ccident analysis defines CEA misoperation as any event, with the exception of sequential
group withdraws, which could result from a single
malfunction in the reactivity control systems. For example, CEA misalignment may be caused by a malfunction of the C
- EDM, CEDM Control System , or by operator error. A stuck CEA may be caused by mechanical jamming of the CEA fingers or of the
gripper. A dropped CEA could be caused by an electrical
failure in the CEA coil power programmers.
The acceptance criteria for addressing CEA inoperability/
misalignment are that:
- a. There shall be no violations of:
- 1. S AFDLs , or 2. RCS pressure boundary integrity; and
CEA Alignment B 3.1.4 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.4-3 Revision 2
- b. The core must remain subcritical after accidents or transients.
Two types of misalignment are distingu ished in the safety
analysis (Reference 1 , Appendix 1C
). The first type of misalignment occurs if one CEA fails to insert upon a
reactor trip and remains stuck fully withdrawn. This
condition requires an evaluation to determine that
sufficient reactivity worth is held in the remaining CEAs to
meet the SDM requirement with the maximum worth CEA stuck
fully withdrawn. If a CEA is stuck in the fully withdrawn
position, its worth is added to the SDM requirement, since
the safety analysis does not take two st uck CEAs into
account. The second type of misalignment occurs when one
CEA drops partially or fully into the reactor core. This
event causes an initial power reduction followed by a return
toward the original power, due to positive reactivity feedback fr om the negative MTC. Increased peaking during the power increase may result in excessive local LHRs (Reference 1, Section 14.14
). None of the above CEA misoperations will result in an
automatic reactor trip. In the case of the full
-length CEA drop, a p rompt decrease in core average power and a
distortion in radial power are initially produced, which, when conservatively coupled, result in a local power and
heat flux increase, and a decrease in DNBR parameters.
The results of the CEA misoperation analy sis show that
, during the most limiting misoperation events, no violations
of the SAFDLs, fuel centerline temperature, or RCS pressure
occur. Control element assembly alignment satisfies 10 CFR 50.36(c)(2)(ii), Criteria 2 and 3.
LCO The limits on shutdo wn and regulating CEA alignments ensure
that the assumptions in the safety analysis will remain
valid. The requirements on OPERABILITY ensure that upon
reactor trip, the CEAs will be available and will be
inserted to provide enough negative reactivity to shut down
the reactor. The OPERABILITY requirements also ensure that CEA Alignment B 3.1.4 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.4-4 Revision 43 the CEA banks maintain the correct power distribution and CEA alignment.
The requirement is to maintain the CEA alignment to within
7.5 inches
between any CEA and its group.
Failur e to meet the requirements of this LCO may produce
unacceptable power peaking factors and LHRs, or unacceptable
SDMs, all of which may constitute initial conditions
inconsistent with the safety analysis.
APPLICABILITY The requirements on CEA OPERABILITY a nd alignment are
applicable in MODEs 1 and 2 because these are the only MODEs
in which neutron (or fission) power is generated, and the
OPERABILITY (e.g., trippability) and alignment of CEAs have the potential to affect the safety of the plant. In MODEs 3 , 4, 5, and 6, the alignment limits do not apply
because the CEAs are bottomed, and the reactor is shut down
and not producing fission power. In the shutdown MODEs, the
OPERABILITY of the shutdown and regulating CEAs has the
potential to affect the requir ed SDM, but this effect can be
compensated for by an increase in the boron concentration of
the RCS. See LCO 3.1.1 for SDM in MODEs 3, 4, and 5, and LCO 3.9.1 for boron concentration requirements during
refueling.
ACTIONS A.1 and B.1 A CEA may become mi saligned, yet remain trippable. In this
condition, the CEA can still perform its required function
of adding negative reactivity should a reactor trip be
necessary.
If one or more regulating or shutdown CEAs are misaligned by
> 7.5 inches and 15 inche s but trippable, or one CEA is misaligned by >
15 inches but trippable, continued operation in MODEs 1 and 2 may continue, provided CEA alignment is restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for CEAs misaligned 15 inches and within the time specified in the COLR for CEAs m isaligned 15 inches. (The maximum time provided in the COLR is 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> s.)
CEA Alignment B 3.1.4 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.4-5 Revision 43 Regulating and shutdown CEA alignment is restored by either aligning the misaligned CEA(s) to within 7.5 inches of its
group or aligning the misaligned CEAs group to within
7.5 i nches of the misaligned CEA.
Xenon redistribution in the core starts to occur as soon as
a CEA becomes misaligned. Restoring CEA alignment ensures
acceptable power distributions are maintained. For small misalignments ( 15 inches) of the CEAs, there i s: a. A small effect on the time
-dependent, long
-term power distributions relative to those used in generating LCOs
and limiting safety system settings setpoints;
- b. A negligible effect on the available SDM; and
- c. A small effect on the ejected CEA wort h used in the
accident analysis.
With a large CEA misalignment ( 15 inches), however, this misalignment would cause distortion of the core power distribution. This distortion may, in turn, have a
significant effect on the time
-dependent, long
-term powe r distributions relative to those used in generating LCOs and
limiting safety system settings setpoints.
The effect on the available SDM and the ejected CEA worth
used in the accident analysis remains small.
Therefore, this condition is limited to a si ngle CEA misalignment, while still allowing time for recovery.
In both cases, the allowed time period is sufficient to:
- a. Identify cause of a misaligned CEA;
- b. Take appropriate corrective action to realign the CEAs;
and c. Minimize the effects of xe non redistribution.
If a CEA is untrippable, it is not available for reactivity
insertion during a reactor trip. With an untrippable CEA, meeting the insertion limits of LCOs 3.1.5 and 3.1.6 does
not ensure that adequate SDM exists.
Condition F must be entered.
CEA Alignment B 3.1.4 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.4-6 Revision 37 C.1 and C.2 If any CEA is not restored to within its alignment limits within the Completion Time provided in Required Action A.1
or B.1, an additional 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is allowed to restore CEA alignment, provided THERMAL POWER is reduced 70% RTP.
Prompt action must be taken to reduce THERMAL POWER, and the reduction must be completed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Reducing THERMAL
POWER ensures acceptable power distributions are maintained
during the additional time provided to restore alignment.
The Completion Times are acceptable based on the reasons
provided in the Bases for Required Actions A.1 and B.1.
D.1, D.2.1, and D.2.2 The CEA motion inhibit permits CEA motion within the
requirements of LCO 3.1.6, and prevents regulating CEAs from
being misaligned fr om other CEAs in the group.
Performing SR 3.1.4.1 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter is considered acceptable, in view of other
information continuously available to the operator in the
Control Room.
With the CEA motion inhibit inoperable, a Co mpletion Time of
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is allowed for restoring the CEA motion inhibit to
OPERABLE status, or fully withdrawing the CEAs in groups 3
and 4, and withdrawing all CEAs in group 5 to < 5%
insertion.
Withdrawal of the CEAs to the positions required in Requi red Action D.2.2 provides additional assurance that core
perturbations in local burnup, peaking factors, and SDM will
not be more adverse than the Conditions assumed in the
safety analyses and LCO setpoint determination (Reference 1,
Chapter 14).
The 6-h our Completion Time takes into account Required
Action D.1, the protection afforded by the CEA deviation
circuits, and other information continuously available to
the operator in the Control Room, so that during actual CEA
motion, deviations can be detect ed.
CEA Alignment B 3.1.4 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.4-7 Revision 37 Required Action D.2.2 is modified by a Note indicating that performing this Required Action is not required when in
conflict with Required Actions A.1, B.1, C.2, or E.1.
E.1 When the CEA deviation circuit is inoperable, performing
SR 3.1.4.1 withi n 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter
ensures improper CEA alignments are identified before
unacceptable flux distributions occur. The specified
Completion Times take into account other information
continuously available to the operator in the Control Roo m,
so that during CEA movement, deviations can be detected, and
the protection provided by the CEA inhibit and deviation
circuit is not required.
F.1 If any Required Action and associated Completion Time of
Condition C, Condition D, or Condition E is no t met, one or
more regulating or shutdown CEAs are untrippable, two or
more CEAs are misaligned by >
15 inches, the unit is
required to be brought to MODE
- 3. By being brought to
MODE 3, the unit is brought outside the MODE of
applicability. Continued ope ration is not allowed in the
case of more than one CEA misaligned from any other CEA in
its group by >
15 inches, or one or more CEAs untrippable.
This is because these cases could result in a loss of SDM
and power distribution and a loss of safety functi on, respectively.
When a Required Action cannot be completed within the
required Completion Time, a controlled shutdown should be
commenced. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, for reaching
MODE 3 from fu ll power conditions in an orderly manner and
without challenging plant systems.
SURVEILLANCE SR 3.1.4.1 REQUIREMENTS
Verification that individual CEA positions are within
7.5 inches
(indicated reed switch positions) of all other
CEAs in the group are per formed at Frequencies of within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of any CEA movement of 7.5 inches and every CEA Alignment B 3.1.4 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.4-8 Revision 37 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The CEA position verification after each movement of 7.5 inches ensure that the CEAs in that group are properly aligned at the time when CEA misalignments are most likely to have occurred. The 12
-hour Frequency allows the
operator to detect a CEA that is beginning to deviate from
its expected position. The specified Frequency takes into
account other CEA position information that is continuously
available to the operator in the Control Room, so that
during CEA movement, deviations can be detected, and
protection can be provided by the CEA motion inhibit and
deviation circuits.
SR 3.1.4.2 Demonstrating the CEA motion inhibit OPERABLE verifies that
the CEA mo tion inhibit is functional, even if it is not
regularly operated. The verification shall ensure that the
motion inhibit circuit maintains the CEA group overlap and
sequencing requirements of LCO 3.1.6, and prevents any
regulating CEA from being misaligned from all other CEAs in its group by 7.5 inches (indicated position). The 31
-day Frequency takes into account other information continuously available to the operator in the Control Room, so that
during CEA movement, deviations can be detected, and
prot ection can be provided by the CEA deviation circuits.
SR 3.1.4.3 Demonstrating the CEA deviation circuit is OPERABLE verifies
the circuit is functional. The 31
-day Frequency takes into
account other information continuously available to the
operator in the Control Room, so that during CEA movement, deviations can be detected, and protection can be provided
by the CEA motion inhibit.
SR 3.1.4.4 Verifying each CEA is trippable would require that each CEA
be tripped. In MODEs 1 and 2, tripping each CEA would result in radial or axial power tilts or oscillations.
Therefore, individual CEAs are exercised every 92 days to provide increased confidence that all CEAs continue to be
trippable, even if they are not regularly tripped. A
movement of 7.5 inches is adequate to demonstrate motion
without exceeding the alignment limit when only one CEA is CEA Alignment B 3.1.4 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.4-9 Revision 37 being moved. For the purposes of performing the CEA operability test, if the CEA has an inoperable position
indicator channel, the alternate indication system (pu lse counter or voltage dividing network) will be used to monitor
position. The 92
-day Frequency takes into consideration
other information available to the operator in the Control
Room and other SRs being performed more frequently, which
add to the determ ination of OPERABILITY of the CEAs.
Between required performances of SR 3.1.4.5, if a CEA(s)is
discovered to be immovable, but remains trippable and
aligned, the CEA is considered to be OPERABLE. At any time, if a CEA(s) is immovable, a determination of the trippability (OPERABILITY) of the CEA(s) must be made, and
appropriate action taken.
SR 3.1.4.5 Performance of a CHANNEL FUNCTIONAL TEST of each reed switch
position transmitter channel ensures the channel is OPERABLE
and capable of indicating CEA p osition over the entire
length of the CEA's travel.
A successful test of the
required contact(s) of a channel relay may be performed by
the verification of the change of state of a single contact
of the relay. This clarifies what is an acceptable CHANNEL
FUNCTIONAL TEST of a relay. This is acceptable because all
of the other required contacts of the relay are verified by
other Technical Specification tests at least once per
re fueling interval with applicable extensions.
Since this
SR must be performed w hen the reactor is shut down, a
24-month Frequency to be coincident with refueling outages
was selected. Operating experience has shown that these
components usually pass this SR when performed at a
Frequency of once every 24 months. Furthermore, the
Fre quency takes into account other SRs being performed at
shorter Frequencies, which determine the OPERABILITY of the
CEA Reed Switch Indication System.
SR 3.1.4.6 Verification of CEA drop times determined that the maximum
CEA drop time permitted is consis tent with the assumed drop
time used in that safety analysis (Reference 1, Chapter 14).
Control element assembly drop time is measured from the time
when electrical power is interrupted to the CEDM until the CEA Alignment B 3.1.4 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.4-10 Revision 37 CEA reaches its 90% insertion position, from a fully withdrawn position, with T ave 515 F and all reactor coolant pumps operating. Measuring drop times prior to reactor criticality, after reactor vessel head removal, ensures that
reactor internals and CEDM will not interfere with CEA
motion or drop t ime, and that no degradation in these
systems has occurred that would adversely affect CEA motion
or drop time. Individual CEAs whose drop times are greater
than safety analysis assumptions are not OPERABLE. This SR
is performed prior to criticality, bas ed on the need to
perform this SR under the conditions that apply during a
unit outage and because of the potential for an unplanned
unit transient if the SR were performed with the reactor at
power.
REFERENCES
- 1. UFSAR 2. 10 CFR 50.46, "Acceptance Crite ria for Emergency Core
Cooling Systems for Light Water Nuclear Power Plants" Shutdown CEA Insertion Limits B 3.1.5 B 3.1 REACTIVITY CONTROL S YSTEMS B 3.1.5 Shutdown Control Element Assembly (CEA) Insertion Limits
BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.5-1 Revision 2 BACKGROUND The insertion limits of the shutdown CEAs are initial assumptions in all safety analyses that assume CEA insertion
upon reactor trip. The insertion limits directly affect
core power distributions and assumptions of available SDM, ejected CEA wo rth, and initial reactivity insertion rate.
The applicable criteria for these reactivity and power
distribution design requirements are in Reference 1, Appendix 1C, Criteria 6, 27, 28, 29, and 30
, and Reference 2. Limits on shutdown CEA insertion have b een established, and all CEA positions are monitored and
controlled during power operation to ensure that the
reactivity limits, ejected CEA worth, and SDM limits are
preserved.
The shutdown CEAs are arranged into groups that are radially
symmetric. The refore, movement of the shutdown CEAs does
not introduce radial asymmetries in the core power distribution. The shutdown and regulating CEAs provide the required reactivity worth for immediate reactor shutdown
upon a reactor trip.
The design calculation s are performed with the assumption
that the shutdown CEAs are withdrawn prior to the regulating
CEAs. The shutdown CEAs can be fully withdrawn without the
core going critical. The shutdown CEAs are controlled
manually by the Control Room operator. Duri ng normal unit
operation, the shutdown CEAs are fully withdrawn. The
shutdown CEAs must be completely withdrawn from the core
prior to withdrawing any regulating CEAs during an approach
to criticality. The shutdown CEAs are left in this position until th e reactor is shut down. They affect core power, burnup distribution, and add negative reactivity to shut
down the reactor upon receipt of a reactor trip signal.
Shutdown CEA Insertion Limits B 3.1.5 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.5-2 Revision 38 APPLICABLE Accident analysis assumes that the shutdown CEAs are fully SAFETY ANALYSES withd rawn any time the reactor is critical. This ensures that: a. The minimum SDM is maintained; and
- b. The potential effects of a CEA ejection accident are
limited to acceptable limits.
Control element assemblies are considered fully withdrawn at
129 inch es.
On a reactor trip, all CEAs (shutdown and regulating),
except the most reactive CEA, are assumed to insert into the
core. The shutdown and regulating CEAs shall be at or above
their insertion limits and available to insert the required
amount of ne gative reactivity on a reactor trip signal. The
regulating CEAs may be partially inserted in the core as
allowed by LCO 3.1.6. The shutdown CEA insertion limit is
established to ensure that a sufficient amount of negative
reactivity is available to shut down the reactor and
maintain the required SDM (see LCO 3.1.1) following a
reactor trip from full power. The combination of regulating
CEAs and shutdown CEAs (less the most reactive CEA, which is
assumed to be fully withdrawn) is sufficient to take the
re actor from full power conditions at rated temperature to
zero power, and to maintain the required SDM at rated no
load temperature (Reference 1, Sections 3.2 and 3.4). The
shutdown CEA insertion limit also limits the reactivity
worth of an ejected shutdow n CEA.
The acceptance criteria for addressing shutdown CEA, as well
as regulating CEA insertion limits and inoperability or
misalignment, are that:
- a. There be no violation of:
- 1. SAFDLs, or
- 2. RCS pressure boundary damage; and
- b. The core remains subcritical after accident transients.
As such, the shutdown CEA insertion limits affect safety
analyses involving core reactivity, ejected CEA worth, and
SDM (Reference 1, Section 14.1.2).
Shutdown CEA Insertion Limits B 3.1.5 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.5-3 Revision 38 The shutdown CEA insertion limits satisfy 10 CFR 50.36(c)(2)(i i), Criterion
- 2. LCO The shutdown CEAs must be within their insertion limits any
time the reactor is critical or approaching criticality.
This ensures that a sufficient amount of negative reactivity
is available to shut down the reactor and maintain the
required SDM following a reactor trip.
APPLICABILITY The shutdown CEAs must be within their insertion limits, with the reactor in MODEs 1 and 2. The Applicability in
MODE 2 begins anytime any regulating CEA is not fully
inserted. This ensures that a suf ficient amount of negative
reactivity is available to shut down the reactor and
maintain the required SDM following a reactor trip. In
MODE 3, 4, 5, or 6, the shutdown CEAs are fully inserted in
the core and contribute to the SDM. Refer to LCO 3.1.1 for
SDM requirements in MODEs 3, 4, and 5. Limiting Condition
for Operation
3.9.1 ensures
adequate SDM in MODE
- 6. This LCO has been modified by a Note indicating the LCO
requirement is suspended during SR 3.1.4.4. This SR
verifies the freedom of the CEAs t o move, and requires the
shutdown CEAs to move below the LCO limits, which would
normally violate the LCO.
ACTIONS A.1 When one shutdown CEA is withdrawn 121.5 inches and 129 inches, the accumulated times the shutdown CEAs have been withdrawn within this range must be verified. The
Completion Time for this action is once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter. Operation is allowed for 7 consecutive
days and a total of 14 days per 365 days. The peaking
factors may not be outside required limits when one shutdown
CEA is misaligned; therefore, continued operation is
allowed. Since the power distribution limits are being
maintained via the LCOs of Technical Specification Section 3.2, any out
-of-limit peaking factor conditions will
require entry into th e Actions of the appropriate
Section 3.2 LCO(s). The limits on consecutive days and
total days in this condition reflect that the core may be
approaching the acceptable limits placed on operation with Shutdown CEA Insertion Limits B 3.1.5 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.5-4 Revision 38 flux patterns outside those assumed in the long
-term b urnup assumptions. Therefore, operation in this condition cannot
continue and the CEA is required to be restored per Action
B. The accumulated times are required to be verified once
within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to determine which accumulated time limit is
more limitin
- g. The periodic Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the initial completion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is adequate to
ensure that the accumulated time limits are not exceeded.
B.1 Prior to entering this condition, the shutdown CEAs were
fully withdrawn or all but one shutdown CEA was withdrawn 129 inches. If one shutdown CEA is withdrawn 121.5 inches and 129 inches for 7 days per occurrence or 14 days per 365 days, or one shutdown CEA withdrawn 121.5 inches, or two or more shutdown CEAs withdrawn 129 in ches, the out
-of-limit CEAs must be restored to within limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The Completion Time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> reflects that the power distribution limits may be
outside required limits and that the core may be approaching
the acceptable limits placed on op eration within flux
patterns outside those assumed in the long
-term burnup
assumptions.
The CEA(s) must be restored to within limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
The 2-hour total Completion Time allows the operator
adequate time to adjust the CEA(s) in an orderly ma nner. C.1 When Required Action A.1 or B.1 cannot be met or completed
within the required Completion Time, a controlled shutdown
should be commenced. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, for reaching
MODE 3 from full power conditions in an orderly manner and
without challenging plant systems.
SURVEILLANCE SR 3.1.5.1 REQUIREMENTS
Verification that the shutdown CEAs are within their
insertion limits prior to an approach to criticality ensures
that when the reactor is critical, or being taken critical, the shutdown CEAs will be available to shut down the Shutdown CEA Insertion Limits B 3.1.5 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.5-5 Revision 38 reactor, and the required SDM will be maintained following a reactor trip. This SR and Frequency ensure that the
shutdown CEAs are withdrawn before the regu lating CEAs are
withdrawn during a unit startup.
Since the shutdown CEAs are positioned manually by the
Control Room operator, verification of shutdown CEA position
at a Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is adequate to ensure that the
shutdown CEAs are within their insertion limits. Also, the
12-hour Frequency takes into account other information
available to the operator in the Control Room for the
purpose of monitoring the status of the shutdown CEAs.
REFERENCES
- 1. UFSAR 2. 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plants" Regulating CEA Insertion Limits B 3.1.6 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.6 Regulating Control Element Assembly (CEA) Insertion Limits
BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.6-1 Revision 43 BACKGROUND The insertion limits of the regulating CEAs are initial assumptions in all safety analyses that assume CEA insertion
upon reactor trip. The insertion limits directly affect
core power distributions, assumptions of available SDM, and
initial rea ctivity insertion rate. The applicable criteria
for these reactivity and power distribution design
requirements are Reference 1, Appendix 1C, Criteria 27, 29, 30, and 31, and Reference
- 2.
Limits on regulating CEA insertion have been established, and all CEA positions are monitored and controlled during
power operation to ensure that the power distribution and
reactivity limits defined by the design power peaking, ejected CEA worth, reactivity insertion rate, and SDM limits
are preserved.
The regulating CEA groups operate with a predetermined
amount of position overlap, in order to approximate a linear relation between CEA worth and CEA position (integral CEA worth). The regulating CEA groups are withdrawn and operate
in a predetermined sequence. The g roup sequence and overlap
limits are specified in the COLR. Regulating CEAs are
considered to be fully withdrawn when withdrawn to at least
129.0 inches.
The regulating CEAs are used for precise reactivity control
of the reactor. The positions of the r egulating CEAs are
manually controlled. They are capable of adding reactivity
very quickly (compared to borating or diluting).
The power density at any point in the core must be limited
to maintain SAFDLs, including limits that preserve the
criteria spe cified in Reference
- 2. Together, LCOs 3.1.6, 3.2.4, and LCO
3.2.5 provide
limits on control component
operation and on monitored process variables to ensure the
core operates within the LHR (LCO 3.2.1); and Total Integrated Radial Peaking Factor (r T F) (LCO 3.2.3) limits in the COLR. Operation within the LHR limits given in the COLR
prevents power peaks that would exceed the loss of coolant
accident (LOCA) limits derived by the Emergency Core Cooling
Regulating CEA Insertion Limits B 3.1.6 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.6-2 Revision 43 System analysis. Operation within the r T F limit given in the COLR prevents departure from nucleate boiling (DNB) during a loss of forced reactor coolant flow accident. In addition to the LHR and r T F limits, certain reactivity limits are preserved by regulatin g CEA insertion limits.
The regulating CEA insertion limits also restrict the
ejected CEA worth to the values assumed in the safety
analysis and preserve the minimum required SDM in MODEs 1
and 2.
The regulating CEA insertion and alignment limits are
pr ocess variables that together characterize and control the
three-dimensional power distribution of the reactor core.
Additionally, the regulating bank insertion limits control the reactivity that could be added in the event of a CEA
ejection accident, and the shutdown and regulating bank
insertion limits ensure the required SDM is maintained.
Operation within the subject LCO limits will prevent fuel
cladding failures that would breach the primary fission
product barrier and release fission products to th e reactor
coolant in the event of a LOCA, loss of flow, ejected CEA, or other accident requiring termination by a Reactor
Protective System trip function.
APPLICABLE The fuel cladding must not sustain damage as a result of SAFETY ANALYSES normal operation (Condition I) and AOOs (Condition II). The
acceptance criteria for the regulating CEA insertion, ASI, r T F , LHR, and AZIMUTHAL POWER TILT (T q) LCOs are such as to preclude core power distributions from occurring that would
violate the fol lowing fuel design criteria:
- a. During a large break LOCA, the peak cladding
temperature must not exceed a limit of 2200°F (Reference 2); b. During a loss of forced reactor coolant flow accident, there must be at least a 95% probability at a 95%
confiden ce level (the 95/95 DNB criterion) that the hot
fuel rod in the core does not experience a DNB
condition;
Regulating CEA Insertion Limits B 3.1.6 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.6-3 Revision 43
- c. During an ejected CEA accident, the energy input to the fuel must not exceed accepted limits (Reference 1,
Section 14.3); and
- d. The CEAs must be capable of shutting down the reactor
with a minimum required SDM, with the highest worth CEA
stuck fully withdrawn, Reference 1, Appendix 1C, Criterion 29.
Regulating CEA position, ASI, r T F , LHR, and T q are process variables that togeth er characterize and control the three
-dimensional power distribution of the reactor core.
Fuel cladding damage does not normally occur when the core
is operated outside these LCOs during normal operation.
However, fuel cladding damage could result if an accident or
AOO occurs with simultaneous violation of one or more of
these LCOs. Changes in the power distribution can cause
increased power peaking and corresponding increased local
LHRs.
The SDM requirement is ensured by limiting the regulating
and s hutdown CEA insertion limits, so that the allowable
inserted worth of the CEAs is such that sufficient
reactivity is available to shut down the reactor to hot zero
power. SHUTDOWN MARGIN assumes the maximum worth CEA
remains fully withdrawn upon trip (Ref erence 1, Section 3.4). The most limiting SDM requirements for MODEs 1 and 2
conditions at BOC are determined by the requirements of
several transients, e.g., Loss of Flow, Seized Rotor, Boron Dilution, etc. However, the most limiting SDM requirements f or MODEs 1 and 2 at EOC come from the SLB and Excess Load transient s. The requirements of the SLB and Excess Load event s at EOC for both the full power and no load conditions are significantly larger than those of any other event at
that time in cycle. T o verify that adequate SDMs are available throughout the cycle to satisfy the changing requirements, calculations are
performed at both BOC and EOC. It has been determined that
calculations at these two times in cycle a are sufficient
since the differen ces between available SDMs and the Regulating CEA Insertion Limits B 3.1.6 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.6-4 Revision 43 limiting SDM requirements are the smallest at these times in a cycle. The measurement of CEA bank worth performed as
part of the Startup Testing Program demonstrates that the
core has the expected shutdown capability. C onsequently, adherence to LCOs 3.1.5 and 3.1.6 provides assurance that
the available SDM at any time in a cycle will exceed the
limiting SDM requirements at that time in a cycle.
Operation at the insertion limits or ASI limits may approach
the maximum al lowable linear heat generation rate or peaking
factor, with the allowed T q present. Operation at the insertion limit may also indicate the maximum ejected CEA
worth could be equal to the limiting value in fuel cycles
that have sufficiently high ejected CE A worths.
The regulating and shutdown CEA insertion limits ensure that
safety analyses assumptions for reactivity insertion rate, SDM, ejected CEA worth, and power distribution peaking
factors are preserved (Reference 1, Section 3.4).
The regulating CE A insertion limits satisfy
10 CFR 50.36(c)(2)(ii), Criterion
- 2.
LCO The limits on regulating CEAs sequence, overlap, and
physical insertion, as defined in the COLR, must be
maintained because they serve the function of preserving
power distribution, ensur ing that the SDM is maintained, ensuring that ejected CEA worth is maintained, and ensuring
adequate negative reactivity insertion on trip. The overlap
between regulating banks provides more uniform rates of
reactivity insertion and withdrawal and is impo sed to maintain acceptable power peaking during regulating CEA
motion. The power-dependent insertion limit (PDIL) alarm circuit is
required to be OPERABLE for notification that the CEAs are
outside the required insertion limits. The PDIL alarm
circuit r equired to be OPERABLE receives its signal from the
reed switch position indication system. When the PDIL alarm
circuit is inoperable, the verification of CEA positions is
increased to ensure improper CEA alignment is identified
before unacceptable flux d istribution occurs.
Regulating CEA Insertion Limits B 3.1.6 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.6-5 Revision 43 APPLICABILITY The regulating CEA sequence, overlap, and physical insertion limits shall be maintained with the reactor in MODEs 1
and 2. These limits must be maintained, since they preserve
the assumed power distribution, ejected CEA worth, SDM, and
reactivity rate insertion assumptions. Applicability in
MODEs 3, 4, and 5 is not required, since neither the power
distribution nor ejected CEA worth assumptions would be
exceeded in these MODEs. SHUTDOWN MARGIN is preserved in
MODEs 3, 4 , and 5 by adjustments to the soluble boron
concentration.
This LCO has been modified by a Note indicating the LCO
requirement is suspended during SR 3.1.4.4. This SR
verifies the freedom of the CEAs to move, and requires the
regulating CEAs to move bel ow the LCO limits, which would
normally violate the LCO.
ACTIONS A.1 and A.2 Operation beyond the transient insertion limit may result in
a loss of SDM and excessive peaking factors. The transient
insertion limit should not be violated during normal
ope ration; this violation, however, may occur during
transients when the operator is manually controlling the
CEAs in response to changing plant conditions. When the
regulating groups are inserted beyond the transient
insertion limits, actions must be taken to either withdraw
the regulating groups beyond the limits or to reduce THERMAL
POWER to less than or equal to that allowed for the actual
CEA insertion limit. Two hours provides a reasonable time
to accomplish this, allowing the operator to deal with
cur rent plant conditions while limiting peaking factors to
acceptable levels.
B.1 and B.2 If the CEAs are inserted between the long
-term steady state
insertion limits and the transient insertion limits for
intervals >
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, and the sh ort-term steady state insertions are exceeded, peaking factors can
develop that are of immediate concern (Reference 1,
Chapter 14).
Regulating CEA Insertion Limits B 3.1.6 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.6-6 Revision 43 Verifying the short
-term steady state insertion limits are not exceeded ensures that the peaking factors that do
develop a re within those allowed for continued operation.
Fifteen minutes provides adequate time for the operator to
verify if the short
-term steady state insertion limits are
exceeded.
Experience has shown that rapid power increases in areas of
the core, in whi ch the flux has been depressed, can result
in fuel damage, as the LHR in those areas rapidly increases.
Restricting the rate of THERMAL POWER increases to 5% RTP per hour, following CEA insertion beyond the long
-term steady-state insertion limits, ensure s the power transients
experienced by the fuel will not result in fuel failure.
C.1 With the regulating CEAs inserted between the long
-term steady state insertion limit and the transient insertion
limit, and with the core approaching the 5 EFPD per 30 E FPD or 14 EFPD per 365 EFPD limits, the CEAs must be returned to
within the long
-term steady state insertion limits, or the
core must be placed in a condition in which the abnormal
fuel burnup cannot continue. A Completion Time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is allotted to r eturn the CEAs to within the long
-term steady state insertion limits.
The required Completion Time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> from initial
discovery of a regulating CEA group outside the limits until
its restoration to within the long
-term steady state limits, shown on the figures in the COLR, allows sufficient time for
borated water to enter the RCS from the chemical addition
and makeup systems, and to cause the regulating CEAs to
withdraw to the acceptable region. It is reasonable to
continue operation for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> aft er it is discovered that
the 5-day or 14-day EFPD limit has been exceeded. This
Completion Time is based on limiting the potential xenon
redistribution, the low probability of an accident, and the
steps required to complete the action.
D.1 When the PDI L alarm circuit is inoperable, performing
SR 3.1.6.1 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter Regulating CEA Insertion Limits B 3.1.6 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.6-7 Revision 43 ensures improper CEA alignments are identified before unacceptable flux distributions occur.
E.1 When a Required Action cannot be completed within the
req uired Completion Time, a controlled shutdown should be
commenced. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, for reaching
MODE 3 from full power conditions in an orderly manner and
without challenging plant system
- s. SURVEILLANCE SR 3.1.6.1 REQUIREMENTS
With the PDIL alarm circuit OPERABLE, verification of each
regulating CEA group position every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient to detect CEA positions that may approach the acceptable limits, and to provide the operator wit h time to undertake
the Required Action(s) should the sequence or insertion
limits be found to be exceeded. The 12
-hour Frequency also
takes into account the indication provided by the PDIL alarm
circuit and other information about CEA group positions
ava ilable to the operator in the Control Room.
SR 3.1.6.2 Verification of the accumulated time of CEA group insertion
between the long
-term steady state insertion limits and the
transient insertion limits ensures the cumulative time
limits are not exceeded. The 24-hour Frequency ensures the
operator identifies a time limit that is being approached
before it is reached.
SR 3.1.6.3 Demonstrating the PDIL alarm circuit OPERABLE verifies that
the PDIL alarm circuit is functional. The 31
-day Frequency
takes into account other SRs being performed at shorter
Frequencies that identify improper CEA alignments.
REFERENCES
- 1. UFSAR 2. 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plants" 10 CFR 50.46 STE-SDM B 3.1.7 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.7 Special Test Exception (STE)
-SHUTDOWN MARGIN (SDM)
BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.7-1 Revision 2 BACKGROUND The primary purpose of the SDM STE is to permit relaxation of existing LCOs to allow the performance of certain PHYSICS TESTS. These tests are constructed to determine the CEA worth. R eference 1 , Appendix B,Section XI requires that a test p rogram be established to ensure that structures, systems, and components will perform satisfactorily in service. All
functions necessary to ensure that specified design
conditions are not exceeded during normal operation and AOOs must be tested. Testing is required as an integral part of
the design, fabrication, construction, and operation of the
power plant. Requirements for notification of the Nuclear
Regulatory Commission, for the purpose of conducting tests
and experiments, are specified in Reference 1, 10 CFR 50.59
. The key objectives of a test program (Reference
- 2) are to: a. Ensure that the facility has been adequately designed;
- b. Validate the analytical models used in design and
analysis; c. Verify assumptions used for predicting plant respo nse; d. Ensure that installation of equipment in the facility
has been accomplished in accordance with the design;
and e. Verify that operating and emergency procedures are
adequate. To accomplish these objectives, testing is required prior to
initial criticality, after each refueling shutdown, and
during startup, low power operation, power ascension, and at
power operation. The PHYSICS TESTS requirements for reload
fuel cycles ensure that the operating characteristics of the
core are consistent with t he design predictions, and that
the core can be operated as designed (Reference 3, Section 13.4).
PHYSICS TESTS' procedures are written and approved in
accordance with an established process. The procedures STE-SDM B 3.1.7 BAS ES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.7-2 Revision 2 include all information necessary to permit a detailed execution of testing required to ensure that the design
intent is met. PHYSICS TESTS are performed in accordance
with these procedures, and test results are independently
reviewed prior to continued power escalation and long
- term power operation. Examples of PHYSICS TESTS include
determination of critical boron concentration, CEA group
worths, reactivity coefficients, flux symmetry, and core
power distribution.
APPLICABLE It is acceptable to suspend certain LCOs for PHYSICS TESTS SAFETY ANALYSE S because fuel damage criteria are not exceeded. Even if an
accident occurs during PHYSICS TESTS with one or more LCOs
suspended, fuel damage criteria are preserved because
adequate limits on power distribution and shutdown capability are maintained durin g PHYSICS TESTS.
Reference 2 defines the requirements for initial testing of the facility, including PHYSICS TESTS. Requirements for reload fuel cycle PHYSICS TESTS are defined in the UFSAR Reference 3, Section 13.4. Although these PHYSICS TESTS are ge nerally accomplished within the limits of all LCOs, conditions may occur when one or more LCOs must be suspended
to make completion of PHYSICS TESTS possible or practical.
This is acceptable as long as the fuel design criteria are
not violated. As long a s the LHR remains within its limit, fuel design criteria are preserved.
In this test, the following LCOs are suspended:
Therefore, this LCO places limits on the minimum amount of
CEA worth required to be available for reactivity control
when CEA worth measurements are performed.
The individual LCOs cited above govern SDM CEA group height, insertion, and alignment. Additionally, the LCOs governing RCS flow, reactor inlet temperature, and pressurizer pressure contribut e to maintaining DNB parameter limits.
The initial condition criteria for accidents sensitive to
core power distribution are preserved by the LHR and DNB STE-SDM B 3.1.7 BAS ES CALVERT CLI FFS - UNITS 1 & 2 B 3.1.7-3 Revision 43 parameter limits. The criteria for the LOCA are specified in Reference 2, 10 CFR 50.46. The crite ria for the loss of
forced reactor coolant flow accident are specified in
Reference 3, Chapter 14. Operation within the LHR limit
preserves the LOCA criteria; operation within the DNB
parameter limits preserves the loss of flow criteria.
Surveillance te sts are conducted as necessary to ensure that
LHR and DNB parameters remain within limits during PHYSICS
TESTS. Performance of these SRs allows PHYSICS TESTS to be
conducted without decreasing the margin of safety.
Requiring that shutdown reactivity equ ivalent to at least
the highest estimated CEA worth (of those CEAs actually
withdrawn) be available for trip insertion from the OPERABLE
CEA provides a high degree of assurance that shutdown
capability is maintained for the most challenging postulated
acci dent, a stuck CEA. When LCO 3.1.1 is suspended, there
is not the same degree of assurance during this test that
the reactor would always be shut down if the highest worth
CEA was stuck out and calculational uncertainties or the
estimated highest CEA worth was not as expected (the single
failure criterion is not met). This situation is judged
acceptable, however, because SAFDLs are still met. The risk
of experiencing a stuck CEA and subsequent criticality is
reduced during this PHYSICS TESTS exception by the Surveillance Requirements; and by ensuring that shutdown
reactivity is available, equivalent to the reactivity worth
of the estimated highest worth withdrawn CEA (Reference 3,
Chapter 3). PHYSICS TESTS include measurement of core parameters or
exerci se of control components that affect process
variables. Among the process variables involved are total integrated radial peaking factor, T q and ASI, which represent initial condition input (power peaking) to the
accident analysis. Also involved are the s hutdown and
regulating CEAs, which affect power peaking and are required
for shut down of the reactor. The limits for these
variables are specified for each fuel cycle in the COLR.
As described in LCO 3.0.7, compliance with STE LCOs is optional and , the refore , no criteria of 10 CFR STE-SDM B 3.1.7 BAS ES CALVERT CLI FFS - UNITS 1 & 2 B 3.1.7-4 Revision 43 50.36(c)(2)(ii) apply. Special Test Exception LCOs provide flexibility to perform certain operations by appropriately
modifying requirements of other LCOs. A discussion of the
criteria satisfied for the other LCOs is provide d in their
respective Bases.
LCO This LCO provides that a minimum amount of CEA worth is
immediately available for reactivity control when CEA worth
measurement tests are performed. The STE is required to
permit the periodic verification of the actual ve rsus predicted worth of the regulating and shutdown CEAs. The
SDM requirements of LCO 3.1.1, the shutdown CEA insertion
limits of LCO 3.1.5, and the regulating CEA insertion limits
of LCO 3.1.6 may be suspended.
APPLICABILITY This LCO is applicable in MO DEs 2 and 3. Although CEA worth
testing is conducted in MODE 2, sufficient negative
reactivity is inserted during the performance of these tests
to result in temporary entry into MODE
- 3. Because the
intent is to immediately return to MODE 2 to continue C EA worth measurements, the STE allows limited operation to
6 consecutive hours in MODE 3, as indicated by the Note, without having to borate to meet the SDM requirements of
LCO 3.1.1. ACTIONS A.1 With any CEA not fully inserted and less than the minimum
required reactivity equivalent available for insertion, or
with all CEAs inserted and the reactor subcritical by less
than the reactivity equivalent of the highest worth CEA, restoration of the minimum SDM requirements must be
accomplished by increasing th e RCS boron concentration. The boration flow rate shall be 40 gpm and the boron concentration shall be 2300 ppm boric acid solution or equivalent. The required Completion Time of immediately is required to meet the assumptions of the safety analysis.
It is assumed that boration will be continued until the SDM
requirements are met.
STE-SDM B 3.1.7 BAS ES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.7-5 Revision 11 SURVEILLANCE SR 3.1.7.1 REQUIREMENTS Verification of the position of each partially or fully
withdrawn full
-length or part
-length CEA is necessary to
ensure that the min imum negative reactivity requirements for
insertion on a trip are preserved. A 2
-hour Frequency is
sufficient for the operator to verify that each CEA position
is within the acceptance criteria.
SR 3.1.7.2 Prior demonstration that each CEA to be withdr awn from the
core during PHYSICS TESTS is capable of full insertion, when
tripped from at least a 50% withdrawn position, ensures that
the CEA will insert on a trip signal. The Frequency ensures
that the CEAs are OPERABLE prior to reducing SDM to less
tha n the limits of LCO 3.1.1. The SR is modified by a Note that allows the SR to not be performed during initial power escalation following a refueling outage if SR 3.1.4.6 has been met during that refueling outage. This allows the CEA drop time test, whic h also proves the CEAs are trippable, to be credited for this SR.
REFERENCES
- 1. 10 CFR Part 50 2. Regulatory Guide 1.68, Revision 2, "Initial Test
Programs for Water
-Cooled Nuclear Power Plants,"
August 1978 3. UFSAR STE-MODEs 1 and 2 B 3.1.8 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.8 Special Test Exceptions (STE)
-MODEs 1 and 2 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.8-1 Revision 2 BACKGROUND The primary purpose of these MODEs 1 and 2 STEs is to permit relaxation of existing LCOs to allow the performance of certain PHYSICS TESTS. These tests are conducted to
determine specific reactor core characteristics.
R eference 1 , Appendix B,Section XI requires that a test program be established to ensure that structures, systems, and components will perform satisfactorily in service. All
functions necessary to ensure that specified design
conditions are not exceeded during normal operation and AOOs must be tested. Testing is required as an integral part of
the design, fabrication, construction, and operation of the
power plant. Requirements for notification of the Nuclear
Regulatory Commission, for the purpose of conducting tests
and exper iments, are specified in Reference 1, 10 CFR 50.59
. The key objectives of a test program (Reference
- 2) are to: a. Ensure that the facility has been adequately designed;
- b. Validate the analytical models used in design and
analysis; c. Verify assumptio ns used for predicting plant response;
- d. Ensure that installation of equipment in the facility
has been accomplished in accordance with design; and
- e. Verify that operating and emergency procedures are
adequate. To accomplish these objectives, testing is required prior to
initial criticality, after each refueling shutdown, and
during startup, low power operation, power ascension, and at
power operation. The PHYSICS TESTS requirements for reload
fuel cycles ensure that the operating characteristics of the core are consistent with the design predictions, and that
the core can be operated as designed (Reference 3, Section 13.4).
PHYSICS TESTS procedures are written and approved in
accordance with established formats. The procedures include
all informat ion necessary to permit a detailed execution of STE-MODEs 1 and 2 B 3.1.8 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.8-2 Revision 43 testing required to ensure that design intent is met.
PHYSICS TESTS are performed in accordance with these
procedures and test results are approved prior to continued
power escalation and long
-term power o peration. Examples of PHYSICS TESTS include determination of critical
boron concentration, CEA group worths, reactivity
coefficients, flux symmetry, and core power distribution.
APPLICABLE It is acceptable to suspend certain LCOs for PHYSICS TESTS SAFET Y ANALYSES because fuel damage criteria are not exceeded. Even if an
accident occurs during a PHYSICS TESTS with one or more LCOs
suspended, fuel damage criteria are preserved because the
limits on power distribution and shutdown capability are
maintained during PHYSICS TESTS.
Reference 3, Section 13.4 defines the requirements for
initial testing of the facility, including PHYSICS TESTS.
Although these PHYSICS TESTS are generally accomplished
within the limits of all LCOs, conditions may occur when one
or more LCO must be suspended to make completion of PHYSICS
TESTS possible or practical. This is acceptable as long as
the fuel design criteria are not violated. As long as the
LHR remains within its limit, fuel design criteria are
preserved.
In this t est, the following LCOs are suspended: LCO 3.1.3; LCO 3.1.4; LCO 3.1.5; LCO 3.1.6; LCO 3.2.3; and LCO 3.2.4.
The safety analysis (Reference 3, Section 13.4) places
limits on allowable THERMAL POWER during PHYSICS TESTS and
requires the LHR and the DNB p arameter to be maintained
within limits.
The individual LCOs governing CEA group height, insertion and alignment, ASI, r T F , and T q preserve the LHR limits.
Additionally, the LCOs governing RCS flow, reactor inlet temperature (T c), and p ressurizer pressure contribute to maintaining DNB parameter limits. The initial condition
criteria for accidents sensitive to core power distribution
are preserved by the LHR and DNB parameter limits. The
criteria for the LOCA are specified in Reference 1,
10 CFR 50.46. The criteria for the loss of forced reactor STE-MODEs 1 and 2 B 3.1.8 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.8-3 Revision 43 coolant flow accident are specified in Reference 1, 10 CFR 50.46. Operation within the LHR limit preserves the
LOCA criteria; operation within the DNB parameter limits
preserves the loss of flo w criteria.
During PHYSICS TESTS, one or more of the LCOs that normally
preserve the LHR and DNB parameter limits may be suspended.
The results of the accident analysis are not adversely
impacted, however, if LHR and DNB parameters are verified to
be wi thin their limits while the LCOs are suspended.
Therefore, SRs are placed as necessary to ensure that LHR
and DNB parameters remain within limits during PHYSICS
TESTS. Performance of these SRs allows PHYSICS TESTS to be
conducted without decreasing the m argin of safety.
PHYSICS TESTS include measurement of core parameters or
exercise of control components that affect process variables. Among the process variables involved are r T F , T q , and ASI, which represent initial condition input (p ower peaking) to the accident analysis. Also involved are the
shutdown and regulating CEAs, which affect power peaking and
are required for shut down of the reactor. The limits for
these variables are specified for each fuel cycle in the
COLR. As descr ibed in LCO 3.0.7, compliance with STE LCOs is
optional and, therefore, no criteria of 10 CFR 50.36(c)(2)(ii) apply. Special Test Exception LCOs provide
flexibility to perform certain operations by appropriately
modifying requirements of other LCOs. A di scussion of the
criteria satisfied for the other LCOs is provided in their respective Bases.
LCO This LCO permits individual CEAs to be positioned outside of
their normal group heights and insertion limits during the
performance of PHYSICS TESTS, such as those required to:
- a. Measure CEA worth;
- b. Determine the reactor stability index and damping
factor under xenon oscillation conditions;
- c. Determine power distributions for nonnormal CEA
configurations;
STE-MODEs 1 and 2 B 3.1.8 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.8-4 Revision 43
- d. Measure rod shadowing factors; and
- e. Measur e temperature and power coefficients.
The requirements of LCO 3.1.3, L CO 3.1.4, LCO 3.1.5, LCO 3.1.6, LCO 3.2.3, and LCO 3.2.4 may be suspended during the performance of PHYSICS TESTS, provided THERMAL POWER is
restricted to test power plateau, which sha ll not exceed
85% RTP. APPLICABILITY This LCO is applicable in MODEs 1 and 2 because the reactor
must be critical at various THERMAL POWER levels to perform
the PHYSICS TESTS described in the LCO section. Limiting
the test power plateau to <
85% RTP ensu res that LHRs are
maintained within acceptable limits.
ACTIONS A.1 If THERMAL POWER exceeds the test power plateau, THERMAL
POWER must be reduced to restore the additional thermal
margin provided by the reduction. The 15
-minute Completion
Time ensures t hat prompt action shall be taken to reduce
THERMAL POWER to within acceptable limits.
B.1 and B.2 If Required Action A.1 cannot be completed within the
required Completion Time, PHYSICS TESTS must be suspended
within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and the reactor must be brou ght to MODE
- 3.
Allowing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for suspending PHYSICS TESTS allows the
operator sufficient time to change any abnormal CEA
configuration back to within the limits of LCO 3.1.4, LCO 3.1.5, and LCO
3.1.6. Bringing
the reactor to MODE 3
within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> incre ases thermal margin and is consistent
with the Required Actions of the power distribution LCOs.
The required Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is adequate for
performing a controlled shutdown from full power conditions
in an orderly manner and without challengin g plant systems, and is consistent with power distribution LCO Completion
Times.
STE-MODEs 1 and 2 B 3.1.8 BASES CALVERT CLIFFS
- UNITS 1 & 2 B 3.1.8-5 Revision 2 SURVEILLANCE SR 3.1.8.1 REQUIREMENTS Verifying that THERMAL POWER is equal to or less than that allowed by the test power plateau, as specified in the
PHYSICS TESTS procedu re and required by the safety analysis, ensures that adequate LHR and DNB parameter margins are
maintained while LCOs are suspended. The 1
- hour Frequency is sufficient, based on the slow rate of power change and
increased operational controls in place du ring PHYSICS
TESTS. REFERENCES
- 1. 10 CFR Part 50 2. Regulatory Guide 1.68, Revision 2, "Initial Test Programs for Water
-Cooled Nuclear Power Plants
," August 1978 3. UFSAR