ML20238B138

Proposed marked-up Tech Specs,Changing Reactor Fuel Critical Power Ratio & Bundle Heat Generation Rate Info for Second Cycle to Allow More Consistent Ref to GE Std Application for Reload Fuel

ML20238B138
Person / Time
Site:	River Bend
Issue date:	08/14/1987
From:	GULF STATES UTILITIES CO.
To:
Shared Package
ML20238B060	List: ML20238B138 RBG-26444, Application for Amend to License NPF-47,revising Tech Specs to Change Reactor Fuel Critical Power Ratio & Bundle Heat Generation Rate Info for Second Cycle.Fee Paid
References
NUDOCS 8708210183
Download: ML20238B138 (20)

Text

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2.0 SAFETY LIMITS AND LIMfTING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL. POWER, Low Pressure or Low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow. ,

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL. POWER and the reactor vessel steam deme pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTOOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

THERMAL POWER, High Pressure and High Flow 1.07 1 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than amS0'with I the reactor vessel steam dome pressure greater than or equal to 785 psig and core flow greater than or equal to 10% of rated flow.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

1.07 With'MCPR less than ar68"and the reactor vessel steam dome pressure greater than or equal to 785 psig and core flow greater than or equal to 10% of rated flow, ba in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the require-ments of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.

ACTION:

With the reactor coolant system pressure above 1325 psig, as measured in the reactor vessel steam dome, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

RIVER BEND - UNIT 1 2-1

• 8708210183 870014 PDR ADOCK 05000450 P PDR

2.1 SAFETY LIMITS BASES

2.0 INTRODUCTION

The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environ _s.

Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a step-back h approach is used to establish a Sap tv Limit such that the MCPR is not less than e99fr:

to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedoni from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a thres-hold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transi-tion boiling, MCPR of 1.0. These conditions represent a significant departure from the condition intended by design for planned operation.

2.1.1 THERMAL POWER, Low Pressure or Low Flow QCriticalPopThe use of theM correlations QR_ference e li) for all critical power nor, valid calculations at pressures below 785 psig or core flows less than 10% of rated flow. Therefore, the fuel cladding integrity Safety Limit is established by -

other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi. Analyses show that with a bundle flow of i

28,000 lbs/hr, bundle pressure drop is nearly independent of bundle power l and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28,000 lbs/hr. Full scale ATLAS test data taken at I pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER.

Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative. -

l RIVER BEND - UNIT 1 B 2-1 i

\ _ - _ _-_____ A

SAFETY LIMITS BASES 2.1. 2 THERMAL POWER, High Pressure and High Flow The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor opera-~

tion, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertain-ties. _-

Add Insert T afety Limit MCPR a

is determined using the General Electric Therma Arialysis s, GETAB , which is a statistical model that combines all of the uncertainties ~ operating parameters and the procedures used to calculate critical power. probability of the occurrence of boiling transition is I determined using the eral Electric Critical Quality (X) Boiling Length (L), l (GEXL), correlation. Th EXL correlation is valid over the range of condi-tions used in the tests of data used to develop the correlation.

The required input to the stat ical model are the uncertainties listed in Bases Table 82.1.2-1 and the nomina alues of the core parameters listed in Bases Table 82.1.2-2.

'The bases for the uncertainties in the core ameters are given in D

NEDO-20340 and the basis for the uncertainty in the correlation is given a

in NED0-10958-A . The power distribution is based on a typ 1 764 assembly ,

core in which the rod pattern was arbitrarily chosen to produce skewed power '

distribution having the greatest number of assemblies at the highe ower levels. Ine worst distribution during any fuel cycle would not be as *ere as the distribution used in the analysis.

l m elete I a. " Genera ic BWR Thermal Analysis Bases (GETAB) Data, Correlation and Design Appi c 00-10958-A.

b. General Electric " Process Computer Per 4

NE00-20340 and Amendment 1, NEDO-20340-1 un dated. valuation nd December Accuracy" 1974, respectively. x RIVER BEND - UNIT 1 B 2-2

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Insert 1 T!.e Safety Limit MCPR is determined using a statistical model that combines all of the uncertainties in the operating parameters and in the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved General Electric critical power correlation. Details of the fuel claJding integrity safety limit calculation are given in Reference 1. Reference 1 includes a tabulation of the uncertainties used in the determination of the Safety Limit MCPR and of the nominal values of parameters used in the Safety Limit MCPR statistical analysis.

Reference

}, " General Electric Standard Application for Reactor Fuel (GESTAR),"

NEDE-24011 ,P-A-8.

e

Delece Bases- Table B2.1.2-l' UNCERTAINTIES USED IN THE DETERMINATION OF THE FUEL CLADDING SAFETY LIMIT

• Standard Deviation Quantity (% of Point)

Feedwater Flow 1.76 Feedwater Temperature 0.76 Reactor Pressure 0.5

~

Core Inlet Temperature .2 Core Total Flow 2.5 Channe.1 Flow Area. 3.0 Friction Factor Multiplier. 10.0 Channel' Friction Factor Multiplier 5.0 TIP Readings . 6.3 R Factor 1.5 Critical Power .6

• T uncertainty analysis used to establish the core wide Safet. Limit MCPR is sed on the assumption of quadrant power symmetry for the reac r core.

RIVER BEND UNIT 1 B 2-3

Bases Table B2.1.2-2 NOMINAL ALUES OF PARAMETERS U 0 IN THESTATISTICALANALYSIS\pFFUELCLADDINQ/INTEGRITYSAFETYLIMIT THERMAL POWER 33 MW Core Flow 08.5 M1b/hr Dome Pressure 1010.4 psig 2

Channel Flow Area 0. 089 ft R-Factor Hig, enrichment - 1.043 Medi enrichment - 1.039 Low en ichment - 1.030 Dc1cte i

RIVER BEND - UNIT 1 B 2-4 l

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3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION Delete 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATESf(APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shalynot exceed the limits shown in Figures 3. 2.1-1, ,3. 2.1-2, 3. 2.1-3, 3. 2.1-4.and= 3. 2.1-5. and 3. 2.1. 6 l APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

pr3.2.1-6 With an APj.HGR exceeding the limits of Figure 3. 2.1-1, 3. 2.1-2, 3. 2.1-3, 3. 2.1-4 Delete ob= en 3.2.1-5'/ initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.1 A11< APLHGRs shall be verified to be equal to or less than the limits determined from Figures 3. 2.1-1, 3. 2.1-2, 3. 2.1-3, 3. 2.1-4 ameh 3. 2.1-5: 4 Dflete hnd 3.2.1-6.

a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and

c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL R00 PATTERN for APLHGR.

d. The provisions of Specification 4.0.4 are not applicable.

RIVER BEND - UNIT 1 3/4 2-1

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R_EACTOR COOLANT SYSTEMS BASES 3/4.1.3 CONTROL RODS The specification.s of this section (1) ensure that the minimura SHUTDOWN MARGIN is maintained and the control rod insertion times are consis'.ent with those used in the safety analyses, and (2) limit the potential effects of the rod drop accident. The ACTION statements permit variations from the basic requirements but impose mdre restrictive criteria for continued operation. A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum. The requirements for the various scram time measurements ensure that any indication of systematic pro-blems with rod drives will be investigated on a timely basis.

Damage within the control rod drive mechanism could be a generic problem.

Therefore, with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period that is long enough to permit determining the cause of the inoperability yet prevent operation with a large number of inoperable control rods.

Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those not fully inserted are consistent with the SHUTOOWN MARGIN requirements.

The number of control rods permitted to be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable rods' could be indicative of a generic problem and the reactor must be shut down for investigation and resolution of the problem.

1.07 Thecontrolrodsystemisdesignedtobringthereactorbubcriticalata rate fast enough to prevent the MCPR from becoming less than%46-during the limiting power transient analyzed in Section 15.0 of the FSAR. This analysis shows that the negative reactivity rates, resulting from the scram with the average response of all the drives as given in the specifications, prov.ide the required protection and MCPR remains greater than 4.HiWr#*The occurrence of m i.07 scram times longer then those specified should be viewed as an indication of a systemic problem with the rod drives and, therefore, the surveillance interval is reduced in order to prevent operation of the reactor for long periods of time with a potentially serious problem.

The scram discharge volume is required to be OPERABLE so that it will be available when needed to accept discharge water from the control rods during a reactor scram and will isolate the reactor coolant system from the containment when required.

Control rods with inoperable accumulators are declared inoperable and Specification 3.1.3.1 then applies. This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a scram than has been analyzed even though control rods with inoperable accumulators may still be inserted with normal drive water pressure. Operability of the accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactor.

RIVER BEND - UNIT 1 8 3/4 1-2

1 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temper-ature following the postulated design basis loss-of-coolant accident will not exceed the 2200 F limit specified in 10 CFR 50.46.

3/4.2.1 AVERAGEPLANARL5NEARHEATGENERATIONRATE The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all

• 'e rods of a fuel assembly at any axial location and is dependent only second-

'ly on the rod to rod power distribution within an assembly. The peak clad

~jperature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification. This LHGR times 1.02 is used in the heat'up code along with the exposure-dependent steady state gap conductance and rod-to-rod local peaking factor. The Technical Specification AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR ,

of the highest powered rod divided by its local peaking factor. The limitino value for APLHGR is shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4 W Jeoelete 3.2.1-5.knd3.2.1-6.

The daily requirement fcr calculating APLHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribu-tion shifts are very slow when there have not been significant power or control rod changes. The requirement to calculate APLHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the com-pletion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shifts while still allotting time for the power distribution to stabilize. The requirement for calculating APLHGR after initially determining a LIMITING CONTROL R00 PATTERN i exists ensures that APLHGR will be known following a change in THERMAL POWER or power shape that could place operation into a condition exceeding a thermal limit.

Delete nd 3.2.1-6 {

Thecalculationalprocedureused/toestab'shtheAPLHGRshownonFigures I 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4 + M.3.2.1-5 is based on a loss-of-coolant accident analysis. The analysis was performed using General Electric (GE)

{

calculational models which are consistent with the requirements of Appendix K to 10 CFR 50. A complete discussion of each code employed in the analysis is presented in NEDE-20566(1) . Differences in this analysis compared to previous analyses can be broken down as follows.

a. I,nout Chances

1. Corrected Vaporization Calculation

• Coefficients in the vaporization l correlation used in the REFLOOD code were corrected. l

2. Incorporated more accurate bypass areas - The bypass areas in the top guide were recalculated using a more accurate technique.

RIVER BEND - UNIT 1 8 3/4 2-1

_ _ _ _ _ _ _ _ _ _ _ . _-_ _m

l POWER DISTRIBUTION LIMITS 8ASES AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued)

3. Corrected guide tube thermal resistance.

4. Correct heat capacity of reactor internals heat nodes.

b. Model Change

1. Core CCFL pressure differential - 1 psi - Incorporate the assumption that flow from the bypass to lower plenum must overcome a 1 psi pressure drop in core.

2. Incopo~ rate NRC pressure transfer assumption - The assumption used in the SAFE-REFLOOD pressure transfer when the pressure is increasing was changed.

A few of the changes affect the accident calculation irrespective of CCFL.

These changes are listed below,

a. Input Change

1. Break Areas - The DBA break area was calculated more accurately,

b. Model Chance ,

1. Improved Radiation and Conduction Calculation - Incorporation of CHASTE-05 for heatup calculation.

A list o,f the significant plant input parameters to the loss-of-coolant accident analysis is presented in Bases Table B 3.2.1-1.

3/4.2.2 APRM SETPOINTS The fuel cladding integrity Safety Limits of Specification 2.1 were based on a power distribution which would yield the design LHGR at RATED THERMAL POWER.

The flow biased simulated thermal power-high scram trip setpoint and the flow biased neutron flux-upscale control rod block trip setpoints of the APRM instru-ments must be adjusted to ensure that the MCPR does not become less than hG6Jd.07 or that > 1% plastic strain does not occur in the degraded situation. The scram settings _and rod block settings are adjusted in accordance with the formula in this specification, when the combination of THERMAL POWER and CMFLPD indicates j

.a peak power distribution, to ensure that an LHGR transient would not be j increased in degraded conditions. i I

RIVER BEND - UNIT 1 8 3/4 2-2

POWER DISTRIBUTION LIMITS 8ASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO 1.07 The required operating limit MCPRs' at steady state operating conditions as specified in Specification 3.2.3 are erived from the established fuel cladding integrity Safety Limit MCPR of and an analysis of abnormal -

operational transients. Eor any abnormal operating transient analysis, 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 settings given in Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest delta MCPR;  :

When added to the Safety Limit MCPR of 1.CS D e required minimum operating - 1.07 j limit MCPR of Specification 3.2.3 is obtained and is presented in Figure 3.2.3-1.

The power-flow map of Figure B 3/4 2.3-1 shows typical regions of plant operation. The codes used to evaluate transients ifentified in Reference 2 are described in reference 2. j

.Th evaluation of a given transi egins with the system initial para-meters.g' w , ' ~ 4 tam e Q.O p at are input to a GE core dynamic behavior transient computer program. l a...,.a.a , . . uem u ,r d M . a su. . . . . , . . . . .. a r . . .... r.... ,, , , . . ,

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. The principal result of this evaluation is the reduction in MCPR caused by the transient.

The purpose of the MCPR f and MCPR p of Figures 3.2.3-1 and 3.2.3-2 is to define operating limits at other than rated core flow and power conditions.

At less than 100% of rated flow and power the required MCPR is the larger value of the MCPR f and MCPR p at the existing core flow and power state. The MCPRfs are established te protect the core from inadvertent core flow increases such that the 99.9% MCPR limit requirement can be assured.

The MCPR f s were calculated such that, for the maximum core flow rate and the corresponding THERMAL POWER along the 105%-of-rated steam flow control line, the limiting bundle's relative power was adjusted until the MCPR was slightly above the Safety Limit. Using this relative bundle power, the MCPRs were calcu-lated at different points along the 105%-of-rated steam flow control line corresponding to different core flows. The calculated MCPR at a given point of core flow i,s defined as MCPR7, RIVER BEND - UNIT 1 8 3/4 2-4

1 POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)

The MCPR p s are established to protect the core from plant transients other than core flow increases, including localized events such as rod withdrawal error. The MCPR s were calculated based upon the most limiting transient at the given core powerP level.

At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at these low power levels, operating plant ex-perience indicates that the resulting MCPR value is in excess of requirements by a considerable margin. During initial start-up testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum recir-culation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will bei shown to be unnecessary. The daily requirement for calculating MCPR, when TlHERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER, is sufficient l since powe: distribution shifts are very slow when there have not been Significant power or control rod changes. The requirement for calculating MCPFf within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the com-pletion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shifts while still a11otting time for the power distribution to stabilize. The requirement for calculating MCPR after initially determining a LIMITING CONTROL ROD PATTERN exists ensures that MCPR will be known following a change in THERMAL POWER or power shape that could place operation into a condition exceeding a thermal limit.

3/4.2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than the design linear heat generation rate even if fuel pellet densification is postulated.

References:

1. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566, November 1975. '

2. R. O. Linford, Analytical Metheds of Plent Trensient Evaluations for the OE OUR, ,1[00-10000, Tebruery 1973r

0. belificatien of th: On: Oincasional Cor: Transient Medel Fcr Bei' %g Weter Reactors, NE00 24154, October 1970.

4. TASC 01-A Ovepster Progros Tei The Tran;ient An ly; h Of 0 S hg h '

Channel, Technical Descriptici., NEDE 25149, Jenvery 1000.

2. General Electric Standard Application for Reactor Fuel,NEDE-24011-P-A.

RIVER BEND - UNIT 1 B 3/4 2-5

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