ML13281A394

Calvert Cliffs Nuclear Power Plant, Units 1 & 2, Revision 46 to Technical Specification Bases. Sections B 3.1.1-1 to B 3.1.8-5

ML13281A394
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Site:	Calvert Cliffs
Issue date:	09/24/2013
From:	Constellation Energy Nuclear Group
To:	Office of Nuclear Reactor Regulation
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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highest reactivity worth is fully withdrawn.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remains fu lly withdrawn.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with the shutdown CEAs fully withdrawn and the regulating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

transients.

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:

a. LCO 3.1.1; and b. LCO 3.1.6.

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