ML20134K377

Proposed Tech Specs Re Requirements for Avoidance & Protection from Thermal Hydraulic Instabilities to Be Consistent w/NEDO-31960 & NEDO-31960,Suppl 1, BWR Owners Group Long-Term Stability Solutions..

ML20134K377
Person / Time
Site:	Cooper
Issue date:	02/10/1997
From:	NEBRASKA PUBLIC POWER DISTRICT
To:
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ML20134K368	List: ML20134K365 ML20134K377 ML20134K383
References
NUDOCS 9702130267
Download: ML20134K377 (24)

Text

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Attachment I to NLS970001 -

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APPENDIX A l l

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. MARKED-UP TECHNICAL SPECIFICATION PAGES 1

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2.1 Bases

The abnormal operational transients applicable to operation of the CNS Unit have been l

analyzed throughout the spectrum of planned operating conditions. The analyses were l l

based upon plant operation in accordance with Reference 3. In addition, 2381 MWt is l

the licensod maximum power level of CNS, - and this represents the maximum steady-state power which shall not knowingly be exceeded.

l The transient analyses performed each reload are given in Reference 1. Models and l model conservatisms are also described in this reference.

Reference 2, As discussed in I the core wide transient analyses for one recirculation pump operation is conservatively bounded by two-loop operation analyses and the flow-dependent rod block and scram setpoint equations are adjusted for one-pump operation.

A. Trio Settinos The bases for individual trip settings are discussed in the following paragraphs.

1. Neutron Flux Trin Settinan

8. APRM Flux Scram Trin Settina (Run Model The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated thermal power (2381 MWt) . Because fission chambers provide the basic input signals, the APRM system responds directly to average neutron flux. During transients, the instantaneous rate of heat l transfer from the fuel (reactor thermal power) is less than the j instantaneous neutron flux due to the time constant of the fuel. j Therefore, during abnormal operational transients, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting. Analyses demonstrate that with a 120% scram trip setting, none of the abnormal operational transients analyzed. violate the fuel Safety Limitw.and

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there is a substantial margin from fuel damage.

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i Amendment No. 133 04/12/90 l

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l LIMITING CONDITIONS POR OPERATION SURVEILLANCE REQUIREMENTS l 3.3.C (Cont'd.) 4.3.C (Cont'd.)

3. The maximum scram insertion time for D. Reactivity Anomalies 90% insertion of any operable control rod shall not exceed 7.00 seconds. During the startup test program and D. Reactivity Anomalies startup following refueling outages, the critical rod configurations will At a specific steady state base cond-be compared to the expected configura-tions at selected operating conditions.

ition of the reactor actual control rod These comparisons will be used as bane inventory will be periodically com- data for reactivity monitoring during pared to a normalized computer pre- subsequent power operation through-diction of the inventory. If the out the fuel cycle. At specific power difference between observed and pre- operating conditions, the critical rod dicted rod inventory reaches the configuration will be compared to the equivalent of 1% ak reactivity, the configuration expected based upon ap-reactor will be shut down until the propriately corrected past data. This cause has been determined and correc- comparison will be made at least every tive actions have been taken as full power month, appropriate.

. Recirculation Pumns E. Restrictions

1. With two recirculation pumps in If Specifications 3.3.A through D operation and with core thermal po er above cannot be met, an orderly reater than the limit specified '

shutdown shall be initiated and the Q'gure3.3.1andtotalcoreflow ess reactor shall be in the Shutdown thyn 45% of rated, establish bayeline condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. APkMandLPRM*neutronfluxnopse leve\ s within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, provi d that F. Recirculation Pumns basel'ne values have not bee pre-viousl established since t last

1. A recirculation pump shall not be core re ueling.

started while the reactor is in natural circulation flow and reactor 2. a) Prior to operation w h one re-power is greater than 1% of rated circul tion pump no in opera-thermal power. tion an core the - 1 power greater han the mit specified With two recirculation pumps in oper - in Figure 3.3.1 otablish tion and with core thermal power baseline A RM a LPRM* neutron greater than the limit specified n flux noise ev 1 Efgure 3.3.1 and total core fl less baseline val @9haves, provided not beenthat than 45% of rated, the APRM a d LPRM* previously e9 ablished since the neutqon flux noise levels a 11 be last core r tu ling. Baseline determined within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and: values sha be established with

\s one recir latic pump not in a) if\the APRM and)PRM* neutron operation and cor thermal power flux \ poise levpIs are less than less thp6 or equal to the limit j or eqdal to three times their specif $ed in Figure 3.3.1.

establihhed 6aseline levels, con ~ /

tinue to' termine the noise b) Prior'to operation wi h one re-levels 3 least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> circ 61ation pump not i opera-and al b within 30 minutes after tion and core flow grea er than the mpletion\cf a core thermal 45% of rated, establish aseline pow r increase of at least 5% of core plate AP noise level with r fed core thermal spower while ' core flow less than ur equ 1 to perating in this region of the power / flow map, or 45% of rated, provided that baseline values have not bee

\ previously established with o e I

• DetptorlevelsAandCofoneLPy string per core octant plus detector levels recirculation pump not in l operation since the last core

! A g6d C of one LPRM string in the centsq of refueling.

p e core shall be monitored. Ng Amendment No, 94 09/24/85

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1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS l

") 3.3.F (Cont'd.) 4.3 (Cont'd.)

if the APRM and/or LPRM* neutron noise levels are greater than

[ G. Scram Discharae Volume ee j

timetxtheir established base ne 1. The scram discharge volume (SDV) levels,'immediately initTate correc- l vent and drain valves shall be i tive action'h4r3sto're .the noise eveled and verified open at l levels to withirthe required limits least once every 31 days and within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by inBreasing core crior to reactor start-un.

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flow, /or by initiatingsan orderly r etion of core thermal po'w by 2. The SDV vent and drain valves  !

i nserting control rods.

shall be verified to close (

2f%. The reactor may be started and ope-within 30 seconds after receipt l of a signal for control rod '

rated, or operation may continue with scram once per refuelino evele.

one recirculation loop not in opera-  ;

tion provided that:  !

3. SDV vent and drain valve opera-  !

bility shall be verified follow-

a. with one recirculation pump not ing any maintenance or modifica-l in operation and core thermal i tion to any portion (electrical power greater than the limit l specified in Figure 3.3.1, c re or mechanical) of the SDV which '

may affect the operation of the flow must be greater than o equal to 45% of rated, and vent and drain vavles.

(i) the Surveillance Requir ents of

/

4.3.F.2.a have not bee satisfied, mmediately initiate y tion to r duce core thermal ower to less th or equal to th limit speci-fie in Figure 3.3 1 within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or (ii) the Su eillanc Requirements of 4.3.F.2. hav9/been satisfied, continue o p termine the APRM and LPRM n ytron flux levels at least once er 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and also within 30/mi utes after the completiynof core thermal power increase f at least 5% of ratedj! ore the 1 power while operating in thi region of the powg// flow map. the APRM an47or LPRM* neutro flux noise

, 1 els are greater t an three l p mes their establish d baseline

alues, immediately in tiate l corrective action and r store the l noise levels to wJthin t e required limits wiuhin 2 urs by

• Det ctor levels A and C of one LPRM l str g per core octant plus detector le els

! A d C of one LPRM string in the cente of t core shall be monitored.

Amendment No. 94 -98a- 09/24/85

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Amendment No. 94 -98b. 09/24/85

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LIMITING CONDITIONS FOR OPERATION SURVEILIANCE REQUIREMENTS

3. 3. F (Cont ' d . )

increasing core flow and/or initiating an derly reduction of core thermal power l 1 erting control rods, l

b. With one recirculation pump not n operation and core flow greate than.

i 46% of rated,-and.

I the Surveillt.nce Requ rements of (i) 4 3.F.2.b have not b en satisfied immediately initia action to re' duce core flow less than or equh{to45%of ated within '

4 hougs, or-(ii) theSu\rv ill ce Requirements of 4.3.F.2. h e been satisfied, continue determine core plate 1 AP noise t least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> nd so within 30 min es after the completion of a ore the 1 power increase of least 5% o rated thermal ,

j po r. If the co e plate AP n se level is gre er than

. 0 psi and 2 times ts esta-

blished baseline valu ,-imme-  !

diately initiate corre tive action and restore the ise levels to within the requ' red  !

limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by d creas-j j

ing core flow and/or initia 'ng j an orderly reduction of core '

thermal power by inserting control rods.

4 tq. The idle loop is isolated electrically by disconnecting the breaker to the recirculation pump motor generator '

(M/G) set drive motor prior to start-up, or if disabled during reactor operation, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, h(. The recirculation system controls will be placed in the 'inanual flow control mode.

Instri.-n.t;Cb f 43E, l

i Amendment No. 94- -98c- 09/24/85 t

4 e Lscrt e s p re v o w s p ag e.

LIMITING CONDITIONS FOR OPERATION SURVEILIANCE REQUIREMENTS

3. With no recirculation loops in operation, the reactor shall be placed in the hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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l .3.3 and 4.3 BASES: (Cont'd) the control rod motion is estimated to actually begin. However, 200 milliseconds j is conservatively assumed for this time interval in the transient analyses and I this is also included in the allowable scram insertion times of Specification j 3.3.C. The time to deenergize the pilot valve scram solenoid is measured during '

the calibration tests required by Specification 4.1.

D. Reactivity Anomalies During each fuel cycle excess operative reactivity varies as fuel depletes and as any burnable poison in supplementary control is burned. The magnitude of this excess reactivity may be inferred from the critical rod configuration. As fuel burnup progresses, anomalous behavior in the excess reactivity may be detected by comparison of the critical rod pattern at selected base states to the predicted rod inventory at that state. Power operating base conditions provide the most sensitive and directly interpretable data relative to core reactivity. Furthermore, using power operating base conditions permits frequent reactivity comparisons.

Requiring a reactivity comparison at the specified frequency assures that a comparison will be made before the core reactivity change exceeds 1% Ak.

Deviations in core reactivity greater than it Ak are not expected and require thorough evaluation. One percent reactivity limit is considered safe since an insertion of the reactivity into the core would not lead to transients exceeding design conditions of the reactor system.

F. Recirculation Pumns Until analyses are submitted for review and approval by the NRC which prove that recirculation pump startup from natural circulation does not cause a reactivity insertion transient in excess of the most severe coolant flow increase currently l analyzed, Specification 3.3.F.1 prevents starting recirculation pumps while the l reactor is in natural circulation above 1% of rated thermal power. Spe ci ficat4++m+ j q3. 3,Fra-end-a-ere-based upc .- providing-assurance that r.cutten-f4ux limit cycle l o illations, which have a small probability of occurring in the high power /lpw ' i flow - rner of the operating domain, are detected and suppressed. BWR cores typicall erate with neutron flux noise levels of 1%-12% of rated power (peak to peak) due andom boiling and flow noise. These flux nois3 aevels are considered in the ' thermal / mechanical design of GE BWR fuel,,,oc',ur c in a stable mode, and are found to f negligible consequence. How6ver, m. er certain  !

high power / low flow condi thatcouldoccurdfung'a recirculation pump I trip and subsequent Single Loop ration (SLol-where reverse flow occurs in inactive jet pumps, a hydraulic /reac Eic feedback mechanism can be enhanced such that sustained limit cym1 aqillations of flow noise with peak j to peak levels several times no xhibited. Although large j margins to safety limits ajr -Mh,rmal values intained when are these limit cycle oscillations l occur, they are to be p itored for, and suppressed whnn 41ux noise exceeds l the three time baseline value by inserting rods and/or increasing coolant 1 flow. The li e-fn Figure 3.3.1 is based on the 80% rod line belo( which the l probabili of limit cycle oscillations occurring is negligible. Thesthermal I power ore flow, and neutron flux noise level limitations are prescribDe si j ordance with Reference 3. i i

I nse rd: nes t p age  ;

Amendment No. 94 -103- 09/24/85 I

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Thsc,-t og preveous p q e, j operation in natural circulation mode, with no recirculation loops in operation, can place the reactor in a. condition closer to the onset of thenmal-hydraulic instabilities. Based on operating. experience, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is a reasonable time to reach Hot Shutdown from higher power conditions, in an  !

orderly manner and without challenging plant systems.

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, LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS

3. , Thermal-hydraulic senhild i

(

cors t i power shall not  ;

eicceed 25% o d thermal l

power without force reu-- '

lation.

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. Amendment No. 32 -212a- 11/10/76

m e e Ensed on wious ye LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.11.D. Intentional entry into the Stability Exclusion Region of the power / flow map defined in the Core Operating Limits Report (COLR) is prohibited. If entry into the Stability Exclusion Region does occur, immediately perform one or more of the following until the Stability Exclusion Region has been exited:

a. Insert control rods,

c. Increase the speed of an .

operating recirculation pump.

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l 3.11 Bases: (Cont'd)

! 1 C. Minimum critical Power Ratio (MCPR) l ,

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{ The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.11C are derived from the established fuel cladding integrity Safety Limit and an analysis of abnormal operational transients (References 2 and 11). For any abnormal operating transient analysis lj '

evaluation with the initial condition of the reactor being at the steady state operating limit it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip l setting given in Specification 2.1. '

l To assure that the fuel cladding integrity safety Limit is not exceeded during any anticipated abnormal operational transient, the more limiting transients have been analyzed to determine which result in the largest reduction in critical power ratio (CPR). The models used in the transient analyses are discussed in Reference 1. I Flow-dependent and power-dependent MCPR limits (MCPR, and MCPR ) are used to define the required Operating Limit MCPR (OLCPR) such that the above Safety Limit MCPR requirement is met for all power / flow conditions. MCPR, provides the thermal margin required to protect the fuel from transients resulting from inadvertent core flow increases. MCPRp protects the fuel from the other t limiting abnormal operating transients, including localized events such as a  !

rod withdrawal error. 1 Direct scram on Turbine Stop Valve Closure or Turbine Control Valve fast closure provides the fastest response to an abnormal operating transient such as load rejection, turbine trip, or feedwater controller failure. These direct i scrams are bypassed at low power (Puyp ,,) , to reduce the frequency of scrams during power ascension. For operation at or above Ptyp,,, (30% of rated power),

the required OLNCPR is the larger of MCPR, or MCPE at the existing core pcwer/ flow state; where MCPR, and MCPS are determined in the Core Operating Limits Report by multiplying the scram time dependent MCPR limit for rated 4

power and flow MCPR(100) by the K, factor. Below 30% of rated power, when the i direct scrats are bypassed, a slightly more severe transient response results.

To compensate for the more severe transient response, two power dependent MCPR l limits are established, one for high flow (>50% of rated) conditions and one  !

for low flow (s50% of rated) conditions. These limits are specified in the Core Operating Limits Reporr. Further information on the MCPR operating limits for off-rated conditions is presented in Reference 11. ,

I w serfor References +.Bases ct P 9 e. ( S pe.c. 3.II. N n e. 3.11

1. " General Electric Standard Application for Reactor Fuel, " NEDE-24011-P- A. (The approved revision at the time the reload analyses are performed.) The approved revision number shall be identified in the Core Operating Limits Report.

l

2. " Supplemental Reload Licensing Submittal for Cooper Nuclear Station," (applicable l reload document).

3-8. Deleted

9. Letter (with-attachment), R. H. Buckholz (GE) to P. S. Check (NRC). " Response to l NRC Request for Information on ODYN Computer Model," September 5, 1980.

10. " Cooper Nuclear Station Single-Loop Operation," NEDO 24258.

11.

" Extended Load Line Limit and ARTS Improvement Program Analysis for Cooper Nuclear Station Cycle 14," NEDC-31892P, Revision 1, May 1991.

~t ns ed seg t pacje. ( Ae f. Fu h se s 3. / / . .d')

Amendment No. 151 -214a- 11/29/91

%9h 61v 7~4Vl0 LLC R b fL424 Y 3,II D D. Thermal-Hydraulic Stability The reactor is designed such that thermal-hydraulic oscillations are prevented or can be readily detected and suppressed without exceeding specified fuel design limits. To minimize the likelihood of a thermal-hydraulic instability, a Stability Exclusion Region ,

to be avoided during normal power operation, is calculated using the approved methodology of References 12 and 13. Since the Etability Exclusion Region may change 4

each fuel cycle, the Exclusion Region is contained in the Core Operating Limits Report j (COLR). Specific directions are provided to avoid operation in this region and to 1 immediately exit upon entry. Entries into the Stability Exclusion Region are not part of normal operation. An entry may occur as the result of an abnormal event, such as a single recirculation pump trip. In these events, operation in the Stability Exclusion Region may be needed to prevent equipment damage, but actual time spent inside the Region is minimized. Although operator action can prevent the occurrence of and protect the reactor from an instability, the APRM flow biased scram function will suppress ,

oscillations prior to exceeding the Safety Limit MCPR. While core-wide reactor l instability is the predominate mode and the regional mode oscillations are not expected to occur, the reactor is protected from regional mode oscillations through avoidance of the Stability Exclusion Region and administrative controls on reactor conditions which are primary factors affecting reactor stability.  ;

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12. "BWR Owner's Group Long-Term Stability Solution Licensing Methodology," NEDO-31960 (the approved revision at the time the reload analyses are performed). i

13. "BWR Owner's Group Long-Term Stability Solutions Licensing Methodology," NEDO-31960, Supplement 1.

1

Corp j k atina Limits Renort (Continued)

b. ;u Linear Heat Generation Rate for Specification 3.11.B.

e The "f : : fler MCPE 2djustrent f acter fer Specifiertien 2.11.C. )

h The inimum )[ritical y wer patio (MCPR) for Specification 3.11.C.

t The 1 pkock jifbnitor upscale setpoint for Table 7d 3.2.C of Specification 3.2.C.

St4[Sy 3, g g,y ,

t. The rut +/4toa The analytical methods%p u,de.Nin he Exclusion s Reglow b Spec.ibdlen  : S. be %. l sed to etermine the core operating limits shall I those previously reviewed and approved by the NRC in NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel." (The approved revision at the time the reload analyses are performed.) The approved revision number shall i be identified in the Core Operating Limits Report. '

.The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS i limits, nuclear limits such as shutdown margin, transient analysis limits, and '

accident analysis limits) of the safety analysis are met. l The Core Operating Limits Report, including any mid-cycle revisions or supplements thereto, shall be provided, upon issuance for each reload cycle, j to the NRC Document Control Desk with copies to the Regional Administrator and a Resident Inspector. '

6.5.2 Reoortable Events  !

A Reportable Event shall be any of those conditions specified in Section 50.73 l to 10CFR Part 50. The NRC shall be notified and a report submitted pursuant to the requirements of Section 50.73. Each Reportable Event'shall be reviewed by SORC and the results of this review shall be submitted to SRAB and the Nuclear Power Group Manager. ..

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I Amendment No. 142 -234- 05/22/91 j I

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Attachment I to NLS970001 Page 5 of 5 1

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i APPENDIX B REVISED TECIINICAL SPECIFICATIONS PAGES l

2.1 Bases.

The abnomal operational transients applicable to operation of the CNS Unit have been analyzed throughout the spectrum of planned operating conditions. The analyses were based upon plant operation in accordance with Reference 3. In addition, 2381 MWt is the licensed maximum power level of CNS, and this represents the maximum steady-state power which shall not knowingly be exceeded.

The transient analyses performed each reload are given in Reference 1. Models and model conservatisms are also described in this reference. As discussed in Reference 2, the core wide transient analyses for one recirculation pump operation is conservatively bounded by two-loop operation analyses and the flow-dependent rod block and scram setpoint equations are adjusted for one-pump operation.

A. Trio Settin32 The bases for individual trip settings are discussed in the following paragraphs.

1. Neutron Flux Tric Settinas

a. APRM Flux Scram Trio Settina (Run Mode)

The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated thermal power (2381 MWt). Because fisolon chambers provide the basic input signals, the APRM system responds directly to average neutron flux. During transients, the instantaneoua rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel.

Therefore, during abnormal operational transients, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting. Analyses demonstrate that with a 120% scram trip setting, none of the abnormal operational transients analyzed violate the fuel Safety Limit and there is a substantial margin from fuel damage. Also, the flow biased neutron flux scram provides protection to the Safety Limit MCPR in the unlikely event of a thermal-hydraulic instability.

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e e LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.3.C (Cont'd.) 4.3.C (Cont'd.)

3. The maximum scram insertion time for D. Reactivity Anomalies 90% insertion of any operable control rod shall not exceed 7.00 seconds. During the startup test program and startup following refueling outages, D. Reactivity Anomalies the critical rod configurations will be compared to the expected configura-At a specific steady state base cond- tions at selected operating conditions.

ition of the reactor actual control rod These comparisons will be used as base inventory will be periodically com- data for reactivity monitoring during pared to a normalized computer pre- subsequent power operation through-diction of the inventory. If the out the fuel cycle. At specific power difference between observed and pre- operating conditions, the critical rod dicted rod inventory reaches the configuration will be compared to the equivalent of 1% k reactivity, the configuration expected based upon ap-reactor will be shut down until the propriately corrected past data. This cause has been determined and correc- comparison will be made at least every tive actions have been taken as full power month. ,

G. Scram Discharae Volume E. Restrictions

1. The scram discharge volume (SDV)

If Specifications 3.3.A through D vent and drain valves shall be above cannot be met, an orderly eveled and verified open at  ;

shutdown shall be initiated and the least once every 31 days and i reactor shall be in the Shutdown orier to reactor start-uo.  ;

condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. /

2. The SDV vont and drain valves Recirculation Pumos shall be verified to close j within 30 seconds after receipt ,

1. A recirculation pump shall not be of a signal for control rod '

started while the reactor is in scram once per refuelina cvele, natural circulation flow and reactor power is greater than 1% of rated 3. SDV vent and drain valve opera- '

thermal power. bility shall be verified follow- l ing any n intenance or modifica-

2. The reactor may be started and ope- tion to hay portion (electrical rated, or operation may continue with or mechanical) of the SDV which one recirculation loop not in opera- may affect the operation of the tion provided that; vent and drain valves.

a. The idle loop is isolated electrically by disconnecting the breaker to the recirculation pump motor generator (M/G) set drive motor prior to start-up, or if disabled during reactor operation, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. The recirculation system controls j will be placed in the manual flow control mode.

3. With no recirculation loops in operation, the reactor shall be placed in the hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

s l

- .o 3.3 and 4.3 BASES: (Cont'd) the control rod motion is estimated to actually begin. However, 200 milliseconds

! is conservatively assumed for this time interval in the transient analyses and this is also included in the allowable scram insertion times of Specification l

3.3.C. The time to deenergize the pilot valve scram solenoid is measured during l the calibration tests required by Specification 4.1.

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D. Reactivity Anomalies

! During each fuel cycle excess operative reactivity varies as fuel depletes and as any burnable poison in supplementary control is buened. The magnitude of this excess reactivity may be inferred from the critic, rod configuration. As fuel burnup progresses, anomalous behavior in the excess reactivity may be detected by comparison of the critical rod pattern at selected base states to the predicted rod inventory at that state. Power operating base conditions provide the most i sensitive and directly interpretable data relative to core reactivity. Furthermore, j using power operating base conditions permits frequent reactivity comparisons. '

Requiring a reactivity comparison at the specified frequency assures that a  ;

comparison will be made before the core reactivity change exceeds 1% k.

Deviations in core reactivity greater than 1% k are not expected and require i thorough evaluation. One percent reactivity limit is considered safe since an j insertion of the reactivity into the core would not lead to transients exceeding l design conditions of the reactor system.

l F. Recirculation Pumns i

Until analyses are submitted for review and approval by the NRC which prove that I recirculation pump startup from natural circulation does not cause a reactivity I insertion transient in excess of the most severe coolant flow increase currently analyzed, Specification 3.3.F.1 prevents starting recirculation pumps while the reactor is in natural circulation above 1% of rated thermal power. Operation in natural circulation mode, with no recirculation loops in operation, can place the reactor in a condition closer to the onset of thermal-hydraulic instabilities.

Based on operating experience, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is a reasonable time to reach Hot Shutdown from higher power conditions, in an orderly manner and without challenging plant systems.

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{ LIMITING CONDITIOWS FOR OPERATION SURVEILIANCE REQUIREMENTS

[ 3.11.D. Thermal-hydraulic stability 1

, Intentional entry into the j Stability Exclusion Region of y the power / flow map defined in the' Core Operating Limits Report (COLR) is prohibited.

If entry into the Stability Exclusion Region does occur, immediately perform one or

, more of the following until l 1 the-Stability Exclusion )

Region has been exited

l i

l a.-Insert control rods,

b. Increase the speed of an

, operating recirculation l pump.

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! 3.11 Bases- (Cont'd) l C. Minimum critical Power Ratio (MCPR)

The required operating limit MCPRs at steady state operating conditions as i specified in Specification 3.11C are derived from the established fuel cladding integrity Safety Limit and an analysis of abnormal operational transients (References 2 and 11). For any alnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.1.

To assure that the fuel cladding integrity Safety Limit is not exceeded during l any anticipated abnormal operational transient, the more limiting transients I have been analyzed to determine which result in the largest reduction in critical power ratio (CPR). The models used in the transient analyses are discussed in Reference 1.

l l Flow-dependent and power-dependent MCPR limits (MCPRr and MCPS ) are used to l define the required Operating Limit MCPR (OLCPR) such that the above Safety i Limit MCPR requirement is met for all power / flow conditions. MCPR, provides i the thermal margin required to protect the fuel from transients resulting from l inadvertent core flow increases. MCPR, protects tne fuel from the other limiting abnormal operating transients, including localized events such as a rod withdrawal error.

l Direct scram on Turbine Stop Valve Closure or Turbine Control Valve fast closure provides the fastest response to an abnormal operating transient such as load rejection, turbine trip, or feedwater controller failure. These direct i

scrams are bypassed at low power (Pqyo), to reduce the frequency of scrams during power ascension. For operaticn at or above P wn. (30% of rated power),

the required OLNCPR is the larger of MCPRr or MCPS at the existing core power / flow state; where MCPRr and MCPS are determined in the Core Operating l Limits Report by multiplying the scram time dependent MCPR limit for rated l power and flow MCPR(100) by the K, f actor. Below 30% of rated power, when the l direct scrams are bypassed, a slightly more severe transient response results.

To compensate for the more severe transient response, two power dependent MCPR limits are established, one for high flow (>50% of rated) conditions and one for low flow ( 50% of rated) conditions. These limits are specified in the Core Operating Limits Report. Further information on the MCPR operating limits

[ for off-rated conditions is presented in Reference 11. i D. Thermal-Hvdraulic Stability l The reactor is designed such that thermal-hydraulic oscillations are prevented or can be readily detected and suppressed without exceeding specified fuel design limits. To minimize the likelihood of a thermal-hydraulic instability, l a Stability Exclusion Region, to be avoided during normal power operation, is j calculated using the approved methodology of References 12 and 13. Since the l Stability Exclusion Region may change each fuel cycle, the Exclusion Region is contained in the Core Operating Limits Report (COLR). Specific directions are provided to avoid operation in this region and to immediately exit upon entry.

Entries into the Stability Exclusion Region are not part of normal operation.

An entry may occur as the result of an abnormal event, such as a single recirculation pump trip. In these events, operation in the Stability Exclusion Region may be needed to prevent equipment damage, but actual time spent inside the Region is minimized. Although operator action can prevent the occurrence i

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3.11 Bases: (Cont'd) of and protect the reactor from an instability, the APRM flow biased scram function will suppress oscillations prior to exceeding the Safety Limit MCPR.

While core-wide reactor instability is the predominate mode and the regional mode oscillations are not expected to occur, the reactor is protected from regional mode oscillations through avoidance of the Stability Exclusion Region and administrative controls on reactor conditions which are primary factors affecting reactor stability.

References for Bases 3.11

1. " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A. (The approved revision at the time the reload analyses are performed.) The approved revision number shall be identified in the Core Operating Limits Report.

2. "Supnlemental Reload Licensing Submittal for Cooper Nuclear Station," (applicable reloau document).

3-8. Deleted

9. Letter (with attachment), R. H. Buckholz (GE) to P. S. Check (NRC). " Response to NRC Request for Information on ODYN Computer Model," September 5, 1980.

10. " Cooper Nuclear Station Single-Loop Operation," NEDO 24258.

11. " Extended Load Line Limit and ARTS Improvement Program Analysis for Cooper Nuclear Station Cycle 14," NEDC-31892P, Revision 1, May 1991.

12. "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," NEDO-31960 (the approved revision at the time the reload analyses are perforned). i

13. "BWR Owners' Group Long-Tem Stability Solutions Licensing Methodology, " NEDO-31960, )

Supplement 1.

4.11 pases i l

A&B. s.eraae aul Local LHGR The LHGR shall be checked daily to determine if fuel burnup, or control rod movement has caused changes in power distribution. Since changes due to burnup are slow, and only a few control rods are moved daily, a daily check of power distribution is adequate.

C. Minimum Critical Power Ratio (MCPR) - (Surveillance Recuirement)

I At core thermal power levels less than or equal to 25%, the reactor will be operating I at less than or equal to minimum recirculation pump speed and the moderator void l content will be very small. For all designated control rod patterns which may be l employed at this point, operating plant experience indicated that the resulting MCPR 1 value is in excess of requirements by a considerable margin. With this low void

~

content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. During initial start-up testing of the plant, a MCPR evaluation was made at 25% themal power level with minimum recirculation pump speed. The MCPR margin was thus demonstrated such that subsequent MCPR evaluation below this power level was shown to be unnecessary. The daily requirement for calculating MCPR above 25% rated thermal power is sufficient since power distribution shifts are very slow when there have not been significant power or control rod

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! changes. The regirement for calculating MCPR when an operating limit - MCPR 10 approached ensures that MCPR will be known following a change in power or power shape (regardless of magnitude) that could place operation at a thermal limit.

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Core Oneratina Limits Renort (Continued)

b. The Linear Heat Generation Rate for Specification 3.11.B.

c. The Minimum Critical Power Ratio (MCPR) for Specification 3.11.C.

d. The Rod Block Monitor upscale setpoint for Table 3.2.C of Specification 3.2.C.

e. The power / flow nap, defining the Stability Exclusion Region for Specification ?.11.D.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel." (The approved revision at the time the reload analyses are performed. ) The approved revision number shall be identified in the Core Operating Limits Report.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, transient analysis limits, and accident analysis limits) of the safety analysis are met.

The Core Operating Limits Report, including any mid-cycle revisions or supplements thereto, shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk with copies to the Regiona4 Administrator and Resident Inspector.

6.5.2 Renortable Events A Reportable Event shall be any of those conditions specified in Section 50.73 to 10CFR Part 50. The NRC shall be notified and a report submitted pursuant to the requirements of Section 50.73. Each Reportable Event shall be reviewed by SORC and the results of this review shall be submitted to SRAB and the Nuclear Power Group Manager.

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l l ATTACHMENT 3 LIST OF MRC COMMITMENTS l Correspondence No: NLS970001 1

The following table identifies those actions committed to by the District in this document. Any other actions discussed in the submittal represent intended or planned actions by the District. They are described to the NRC for the NRC's ,

information and are not regulatory commitments. Please notify the Licensing Manager l at Cooper Nuclear Station of any questions regarding this document or any associated '

regulatory commitments.

COMMITTED DATE COMMITMENT OR OUTAGE None j I

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I l PROCEDURE NUMBER 0.42 l REVISION NUMBER 1.2 l PAGE 8 OF 10 l