ML20086C521

Proposed Tech Specs Removing Rod Sequence Control Sys from Specs & Reducing Low Power Setpoint for Rod Worth Minimizer from 20% to 10%

ML20086C521
Person / Time
Site:	Cooper
Issue date:	11/15/1991
From:	NEBRASKA PUBLIC POWER DISTRICT
To:
Shared Package
ML20086C519	List: ML20086C521 NLS9100488, Proposed Change 86 to License DPR-46,revising Tech Specs to Remove Rod Sequence Control Sys from Specs & Reduce Low Power Setpoint for Rod Worth Minimizer from 20% to 10% Power
References
NUDOCS 9111220283
Download: ML20086C521 (11)

Text

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2.1 Bases

(Cont'd)

An increase in the APRM scram trip setting would decrease the margin present before the fuel cladding integrity Safety Limit is reached. The APRM scram trip setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation. Reducing this operating . margin would increase the frequency of spurious scrams which have an adverse effect on reactor safety because of the resulting thermal stresses. Thus, the APPJi scram trip setting was selected because it provides adequate margin for the fuel cladding integrity Safety Limit yet allows operating margin that reduces the possibility of unnecessary scrams. ,

The scram trip setting taust be adjusted to ensure that the LHCR transient peak is not increased for any combination of maximura fraction of limiting power density (MFLPD) and reactor core thermal power. The scram setting is adjusted in accordance with the ,

formula in Specification 2.1. A.1.a. when the MFLPD is greater than the fraction of rated power (FRP). This adjustment may be accomplished by increasing the APRM gain and thus reducing the slope and intercept point of the flow referenced APRM Hibh Flux Scram Curve by the reciprocal of the APRM gain change. .;

Analyses of the limiting transients show that no scram adjustment is required to assure MCPR remains above the Safety Limit when the transient is initiated from tne l

- operating MCPR limit specified in the Core Operating Limits Report,

b. APRM Flux Scram Trio Setting (Refuel or Start 6 Hot Standbv Mode') i For cperation in the startup mode while the reactor is at low pressure, the APRM scram setting of 15 percent of rated power provides adequate thermal margin bet'.eaa ,

the setpoint and the Safety Limit, 25 percent of rated. The margin is adequars to l '

accomodate anticipated maneuvers associated .with power plant startup. Ef fects of increasing pressure at zero or low void content are minor, cold water fro sources available -during startup is not much colder than that already in tne system, temperature coefficients are small, and control rod patterns are con nrained to be uniform by operating procedure. backed up by the Rod Worth Minimiter. Worth of l individual rods is very low in a uniform rod pattern. Thus , of. all possible sources .

of reactivity input, uniform control rod withdrawal is the most probable cause of i significant power rise. Because the flux distribution associated with uniform-rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percencage of raced power, the rate of power risa ,

is very slow. Generally, the heat flux is in naar equilibrium with the fission rate.

In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise -is ne more than 5 percent of' rated power per minute, and the APRM system would be more than adequate to assure a scram before the power could exceed . the Safety ,

Limit. .The 15 percent APRM scram remains active until the mode switch is' placed in the RUN position. This change can occur uhen reactor pressure is greater than Specification 2.1.A.6.

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.LTMTTTNG ('ONDTTTnN FOR OPERATTON StfRVEILIANCE RE0tfiRFMFNT 3.3 REACTIVTTY CONTROL

• 4.3 REACTIVITY CONTROL Annlicabilitv: Anolicabilitv:

Applies to the surveillance Applies to the operational status of requ remen s t te ntr rd the control rod system. system.

Obiectlyfl Qbiective To assure the ability of the control To verify the ability of the control rod system to control reactivity. rod system to control reactivity.

Specification: Specifiention:

Peactivity Limitations A. Reactivity Limitations A.

1. Reactivity margin core loading 1. Reactivity margin - core loading A sufficient number of control rods Sufficient control rods shall be following a refueling shall be operable so that the core withdrawn could be made suberitical in the outage when core alterations were l most reactive condition during the performed to demonstrate, with a operating cycle with the strongest margin of 0.38% ok/h, that the core control rod fully withdrawn and all can be made suberitical at any time d.e r c,perable control rods fully in the subsequent fuel cycle with inserted. - the analytically determined strongest operable control rod fully withdrawn and all other operable rods ful'y inserted.

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L1MITING CONDITION FOR OPERATION SURVEILIANCE REOUIREMENT 3.3.A (cont'd.) 4.3 (cont'd)

2. Reactivity margin - inoperable 2. Reactivity margin - inoperable .

control rods control rods I

a. Control rods which cannot be moved a. Each partially or fully withdrawn with control rod drive pressure operable control rod shall be shall be considered inoperable. If exercised one notch at least once a partially or fully withdrawn each week, when operating above 30%

control rod drive cannot be moved power. This test shall be performed with driva or scram pressure the at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when reactor shall be brought to a operating above 30% power in the shutdown condition within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> event power operation is continuing unless investigation demonstrates with three or more inoperable control rods or in the event power that the cause of the failure is not

=due to a f ailed control rod drive operation is continuing with one mechanism collet housing. fully or partially withdrawn rod which cannot be moved and for which control rod drive mechanism damage

b. The control rod directional control The valve for inoperable control rods has not been ruled out shall be dissrmed electrically. surveillance need not be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the number of

c. Control rods with scram times inopetable rods has been reduced to greater than those . permitted by less than three and if it has been Specification 3.3.c.3 are demonstrated that control rod drive inoperable, but if they can be mechanism collet housing failure is inserted with control rod drive not the cause of an immovable control rod, pressure they need not be disarmed electrically.

b. Deleted.

d. Control rods with a failed " Full-in" Once per week check the status of or " Full-out" position switch may be c.

considered operable if the actual the pressure and level alarms for l each accumula:or, rod position is known. These rods must be moved in sequence to.their correct _ positions (full in on insertion or rull out on ,

withdrawal).

e. Control rods with inoperable

'  ! accumulators or those whose position cannot be positively determined shall be considered inoperable.

f. Inoperable control rods shall be positioned such that Specification 3.3.A,1 is met, in addition, during reactor power operation, inoperable control-rods shall be separated by at least two control rod cells.

If this Specification cannot be met the reactor shall not be started. or if at power, the reactor shall be brought to a ' shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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. LI;fT rTNo . CONDITION FOR OPERA fTON SURVEILIANCE REOUTREMENT

~ l ,3.3.B (cont' d) 4.3.B (cont'd) l.

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c. During each refueling outage observe that any drive which has been uncoupled from and . subsequently.

recoupled to its control rod does

2. The control rod drive housing not go to the overtravel position.

support systern shall be - in place during reactor power operation or 2. The control rod drive housing when the reactor coolant system is support system shall be inspected pressurized above atmospheric after reassembly and the results of ,

the inspection recorded.

pressure with fuel in the reactor vessel,-unless all control rods are 2

fully- inserted and Specification 3.3.A.1 is met. ,

3,a.- Deleted.

b. Deleted. 3a. Deleted.' l t

c. Whenever the reactor is in the b. Prior to the start of control ' rod

.startup or run modes below 10% rated withdrawal towards criticality and l'-

power the Rod Worth Minimizer shall prior to sttaining 10% rated power l be - operable or a second licensed during rod insertion-while shutting operator or other quhlified employee down, the capability of the Rod shall verify that the operator at Worth Minimizer (RWM) to properly '

the reactor console is following the fulfill its function shall be control rod- program. Reactor verified by the following checks: l

'startup shall not be initiated more frequently than once per calendar year with the RUM inoperable, d Deleted.

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. 1,IMITTNG CONDITION FOR OPERATION st*RVET1,1 ANCF RFnUIRFMENT 3.3.3.3 (cont'd) 4. 3.B. 3.b (cont' d) l e. If Specification 3.3.B.3.c 1) The correctness of the Banked cannot be met, the reactor Position Withdrawal Sequence shall not be started, or if input to the RkM computer the reactor is in the run or shall be verified, startup modes at less than 10% rated power, it shall be 2) The Rkt computer on line l be brought to a shutdown diagnostic test shall condition immediately, successfully performed.

3) Proper annunciation of the selection error of at least one out-of sequence control rod in each fully inserted group shall be verified.

4) The rod block ftmetion of the Rkt shall be verified by withdrawing the first rod as an out of-sequence control rod no more than to the block

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c. Whan required, the presence of a second licensed operator or other qualified employee to verify the following of the correct rod program shall be 4 Control rods shall not be verified.

withdrawn for startup unless at least two source range 4 Prior to control rod channels have an observed withdrawal for startap verify count rate equal to or that at least two source range greater than three counts per channels have an observed second. count rate of at least three counts per second.

5. During operation with limiting control rod 5. When a limiting control rod patterns, as determined by pattern exists an instrument the designated qualified functional test- of the RBM personnel, either: shall be performed prior to withdrawal of the designeted

a. Both RSM channels shall be rod (s).

operable; or

b. Control rod withdrawal shall be blocked: or-

c. The operating power level shall be limited so that the MCPR will remain above the l Safety Limit assuming _a single error that results in complete withdrawal of any single operable control rod.

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1.TMITINC*dONDITION FOR OPEPAT10N SilRVFTT.f ANCE RF011TREMENT 3.3 (cont'd) 4.3 (cont'd)

C. Scram insertion Times C. Eqn Insertion Times

1. The average scram insertion ti me , 1. After each refueling outage all based on the deenergi::ation of 'e  : operable rods shall be scram time '

scram pilot valve solenoids as ..me tested from the fully withdrawn ero, of all operable control rods position with the nuclear system in the reactor power operation pressure above 800 psig. This l condition shall be no greater thani testing shall be completed prior to exceeding 40% power. During all

% Insarted From Avn. Scram Inser- scram time testing below 10% power, Fullv Witbjrawn tion Timey (sec) the Rod Worth Minimizer shall be-5 0.375 operable or a second licensed 20 0.90 operator or c,ther qualified employee 50 2.0 shall verify that the - operator at 90 3.50 the reactor console is following the control rod program.

2. At 16-week intervals, 10% of the operable control rod drives sht. .. be scram timed above 800 psig.

Whenever such scram -time measurements are made, an evaluation shall be made to provide reasonable assurance that proper control rod

.2. The average of the scram insertion drive performance is being maintained, times for the three fastest control rods of all groups of four control rods ~ in a two by-two array shall be i

no greater than:

% Inserted From Ave. Scram Inser-Fullv Withdrawn tion Timer _(sec) 5 0.398  ;

20 0.954 50 2.120 90 3.71 l.

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3.3 and 4.3 BASES (cont'd.)

cannot be moved with drive pressure. If the rod is fully inserted and then disarmed electrically, it is in a safe position

• of maximum contribution to shutdown reactivity. If it is disarmed electrically in a non-fully inserted position, that position shall be consistent with the shutdown reactivity limitation stated in Specification 3.3. A.1., This assures that the core can be shutdown at all times with the remaininB '

control-rods assuming the strongest operable contro_ rod does not insert.

An allowable pattern for control rods valved out of service, wh1ch shall meet.this Specification, will be determined and made available to the operator.

If damage within the control rod drive mechanism and in particular, cracks in drive internal housings, cannot be ruled out, then a generic problem. affecting a number of drites cannot be ruled out.

Circumferential~ cracks resulting from stress assisted intergranular corrosion have occurred in the collet housing of drives at several BWRs.

. This type of cracking could occur in a number of drives and if the cracks propagated until severance of the collet housing-occurred, scram could be prevented in the affected rods. Limiting the period of operation with a' potentially severed collet housing and requiring increased surveillance

-after detecting one stuck rod will assure that the. reactor will not be operated with a large number of rods with failed collet housings.

B. -Control Rod

1. Control rod drop _ accidents as discussed in the USAR can lead to significant core damage, If coupling integrity is maintained, the

_ possibility of a rod dropout accident is eliminated. The overtravel position feature provides_a_ positive check as only uncoupled drives may reach this pocition. Neutron instrumentation response to rod movement provides a verification that the rod is following its drive. Absence of

-such response _to drive movement could indicate an uncoupled condition.

Rod position indication is required for proper function of the Rod Worth Minimizer (RWM).

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3.3.B and 4.3.B BASES (cont'd.) l

2. The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the extremely remote event of a housing failure. The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage to the primary coolant system. The desi6 n basis is given in Section 1118.2 of the USAR and the safety evaluation is given in Section 111 8.4 This support is not required if the reactor coolant system is at atmospheric pressure since there would then be no driving force to rapidly eject a drive housing. Additionally, the support is not required if all control rods are fully inserted and if an adequate shutdown margin with one control rod withdrawn has- been demonstrated, since the reactor would remain subcritical even in the event of complete ejection of the strongest control rod.

3. The Rod Worth Minimizer (RWM) restricts withdrawals and insorcions of control l rods to prospecified sequences. These sequences are established such that the drop of any in-sequence control rod or control rod segment (i.e. , one or more notches) would not cause the reactor to *ustain a power excursion resulting in a peak fuel enthalpy in excess of 280 cal./6m. An enthalpy of 280 cal./gm. is well below the level at which rapid fuel dispersal could occur (i.e., 425 cal./gm.). Primary system damage in this accident is not possible unless a -

significant amount of fuel is rapidly dispersed. Ref. Sections III 6.6 and XIV-6.2 of the USAR and Reference 1.

In performing the function described above, the RWM is not required to impose any restrictions at core power levels in excess of 10% of raced. Material in the cited references shows that it is impossible to reach 280 calories per gram in the event of a control rod drop occurring at power greater than 10%, l regardless of the rod pattern. This is true for all normal and abnormal patterns including those which maximite the individual control rod worth.

At power levels below 10%' of rated, abnormal control rod patterns could produce l i rod worths high enough to be of concern relative to the 280 calories por gram rod drop. limit. In-this range, the RWM constrains the control rod sequences l and-patterns to those which involve only acceptable rod worths.

The RWM provides automatic supervision to assure that out of sequence control l rods will not be withdrawn or inrsreed; i.e., it limits operator deviations from planned withdrawal sequences. It serves as a backup to procedural control l on control rod sequences, which limits the maximum reactivity worth of control rods. In the event that the - RWM is out ot' service, when required, a second licensed operator or other qualified technical plant employee can manually fulfill the control rod pattern conformance functions of this system.

The function cf the RWM makes it unnecessary to specify a license limit on rod l-worth to preclude unaccepteble ceasequences in the event of a control rod drop, At low powers, below 10%, .his l 101-1

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3 3.3.B and 4;3.B BASES (cont'd.)

l system forces adherence to acceptable _ rod patterns. Above 10% of rated power, l no cons traint - on rod pattern is required to assure that rod drop accident

• consequences are acceptable. Control rod pattern constraints above 10% of l t rated power are imposed by power distribution requirements as defined in Section 3.3.B.5 of these Technical Specifications. Power level for tutomatic cutout of the RWM function is sensed by feedwater and steam flow. l Functional testing of the RWM prior to the start of control rod withdrawal at startup., and prior to attaining 10% rated thermal power during rod insertion l while shutting down, will ensure reliable operation- and minimize- the probability of the rod drop accident.

The Reduced Notch Worth Procedure for control rod withdrawal allows CNS to take l-advantage of the Banked Position Withdrawal Sequence (BPWS) (Ref. 4).. The BPWS has the advantage cf having been proven statistically to have such - low individual control rod worths that the possibility of a control rod drop accident (CRDA), which exceeds the 280 cal /gm peek ft.el enthalpy limit, is precluded (Ref. 1).

The Reduced Notch Worth Procedure is programmed into the RWM. In the l precheckerboard pattern (100% to 50% control rod density), the RWM will enforce the Reduced Notch Worth Procedure.

4 The Source Range Monitor (SRM) system performs no automatic safety system function; i.e., it has no scran function. It does provide the operator with a visual indication of neutron level. The. consequences of reactivity accidents are functions of the initial neutron flux. The requirements of at least 3 counts per second assures that any transient, should it occur, begins at or above the initial value of 10 g% of rated power used in the analyses of transients cold conditions. One operable SRM channel would be adequate to monitor the approach to criticality using homogeneous patterns of scattered control rod withdrawal. A minimum of two operable SRM's are provided as an

-added conservatism.

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LIMITTNG CONDITTONS FOR OPFRATTON SURVETT.T.ANCF RFOUTRFMFNT9 3.22 SPECIAL TESTS /EXCEPTTON9 (CONT'D) . 4.27 SPECTAL TFSTR/EXCFPTIONS (CONT'D_)_

l 2. Leleted.

3. PJiP, System The RHR system may be aligned in the shutdown cooling mode with at least one shutdown cooling mode loop OPERABLE while performing the Shutdown

-Margin Demonstration.

4. Containment Systems Primary--containment is not required while performing the Shutdown Margin Demonstration when reactor water '

temperature is equal to or less than 212*F.

'B. Trainine Startuo B. Trainine Startun

1. LPCI Mode of RHR I The reactor vnssel shall be verified to be unpressurized and the thermal The LPCI mode ~ is required to be power verified to be less than 1% of operable with the exception that the rated thermel power at least once RHR system may be aligned in the per hour during training startups.

shutdown - cooling modo rather than the LPCI mode while per forming training startups at atmo pheric pressure at power levels less than 1% of rated thermal power.

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3.22 6 4.22 BASES A. Shutdown Martin Demonstration Performance of shutdown margin demonstrations requires additional restrictions in order to ensure that criticality does not occur. Additional surveillance l requirements ensure that shutdown margin requirercents and individual rod worths do not exceed values assumed in the safety analysis. 31nce power levels attained during the demonstration are kept below the level of significant heat addition, the residual heat removal system can remain aligned in the shutdown cooling mode.

B. Traininc Startup Specification 3.22.B provides for the performance of training startups without realigning the residual heat removal system from the shucdown cooling mode to the LPC1 mode. Power levels during training startups are kept below the level of significant heat addition.

I This exception is made in order to minimize contaminated water discharge to the '

radioactive waste disposal systera. -

C. Physics Tests Ar. exception is made to primary containment integrity during initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required. There will be no pressure on the system at this time, thun greatly reoucing the chances of a pipe break. The reactor ms.y be taken critical during this period; however, restrictive operating procedures will be in effect again to minimize the probability of an accident occurring. Procedures and the rod worth minimizer would limit control worth such that a rod drop would not result in any fuel damage. In addition, in the unlikely event that an excursion did occur, the reactor building and standby gas treatment system, which shall be operational during this time, offer a sufficient barrier to keep off site dc 3es well below 10CFR100 limits.

D. Startun Test Procram Relief from the oxygen concentration specif'. cations is necessary in order to provide access to the primary containment during the init. 1 startup and testing phase of operation. Without this access the startup and tot program could be restricted and dela"ed.

The recirculation flow exception permits reactor criticality under nmflow conditions and is required to perform certain startup and physics tests while at low-thermal power levels.

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