ML20236B017

Amend 12 to License NPF-47,revising Tech Specs for Cycle 2 Fuel Reload & Operation

ML20236B017
Person / Time
Site:	River Bend
Issue date:	10/19/1987
From:	Calvo J Office of Nuclear Reactor Regulation
To:
Shared Package
ML20236B021	List: ML20236B017 ML20236B093
References
NUDOCS 8710230296
Download: ML20236B017 (28)

Text

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UNITED STATES f

g NUCLEAR REGULATORY COMMISSION r,

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GULF STATES UTILITIES COMPANY DOCKET NO. 50-458 RIVER BEND STATION, UNIT 1 AMEN 0 MENT TO FACILITY OPERATING LICENSE 1

Amendment No.12 I

License No. NPF-47 1.

The Nuclear Regulatory Commission (the Commission or the NRC) has found that:

A.

The application for amendment filed by Gulf States Utilities s

Company, dated August 14, 1987, complies with the standards and l

requirements of the Atomic Energy Act of 1954 asamended(theAct),

anc the Commission's regulations set forth in,10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; C.

There is reasonable assurance:

(i)thattheactivitiesauthorized by this amendment can be conducted without endangering the health and (ii) that such activities will be and safety of the public,ith the Commission's regulations; conducted in compliance w D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and l

E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

i l

2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-47 is hereby amended to read as follows:

1 (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.12 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

GSU shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

8710230296 871019 PDR ADOCK 05000458 P

PDR

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This license amendment is effective as of its date of issuance.

j FOR THE NUCLEAR REGULATORY COMMISSION

&&, 6. A+v c

Jose A. Calvo, Director ProjectDirectorate-IV DivisionofReactorProjects-III, IV, V and Special Projects Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: October 19, 1987 1

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l:l ATTACHelENT TO LICENSE AMEN 0 MENT N0.12 FACILITY OPERATING LICENSE N0. NPF-47 q

DOCKET NO. 50-458-j Replace the following page of the Appendix "A" Technical Specifications with-the enclosed page.

Therevisedpageisidentifiedb9Amendmentnumberand contains' a' vertical.line indicating the area of change.

Overleaf page provided to maintain document completeness.

.1 REMOVE INSERT 2-1 2-1 82-1 B2-1 i

B2-2 B2-2 B2-3 B2-4 3/4 2-1 3/4,2-1 i

3/4 2-2 3/4'2-2 3/4 2-3 3/4 2-3 3/4 2-4 3/4.2-4 3/4 2-5 3/4 2 3/4 2-6 3/4 2-6 3/4 2-6a B3/4 1-2 B3/4 1-2 B3/4 2-1 B3/4 2-1 B3/4 2 B3/4 2-2

.B3/4 2-4 B3/4 2-4 B3/4 2-5 B3/4 2-5 i

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INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.13 PROCESS CONTROL PROGRAM (PCP)................................

6-22

,6.14 0FFSITE DOSE CALCULATION MANUAL (00CM).......................

6-23 1

~6.15 MAJOR CHANGES TO RADIOACTIVE LIQUID, GASEOUS AND SOLID WASTE TREATMENT SYSTEMS............................................

6-23 1

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RIVER BEND - UNIT 1 xxi

,1 f

U INDEX~

r

~~ LIST OF FIGURES l

FIGURE.

TITLE PAGE l

3.1.5-1 Saturation Temperature of Sodium Pentaborate Solution...........................................

3/4 1-21 3.1.5-2 Sodium Pentaborate Solution Volume / Concentration Requirements.......................................

3/4 1-22 L:

3.2.1-1 Maximum Average Planar Linear Heat Generation Rate (BP85RB094)....................................

3/4 2-2

]

3.2.1-2 Maximum Average Planar Linear Heat Generation Rate (BP85RB163)....................................

3/4 2-3 l

3. 2.1-3 Maximum. Average Planar Linear Heat Generation l

Rate (BP85RB248)....................................... 3/4 2-4 l

l 3.2.1-4 Maximum Average Planar Linear Heat Generation l

Rate (BP8 SRB 278).....................................

3/4 2-5 l

i 3.-2.1-5 Maximum Average Planar Linear Heat Generation j

L Rate (BP85RB299).....................................

3/4 2-6 l

1 3.2.1-6 Maximum Average Planar Linear Heat Generation

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Rate (BP8 SRB 305).....................................

3/4 2-6A-i

~3.2.3-1 MCPR.............-................................

3/4 2-9

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3.2.3-2 MCPR.............................................

3/4 2-10 p

1 3.4.1.1-1 Thermal Power versus Core Flow.....................

3/4 4-3 3.4.6.1-1 Minimum Temperature Required Versus Reactor

')

Pressure...........................................

3/4 4-24 4

4.7.4-1 Sample Plan for Snubber Functional Test............

3/4 7-15 B 3/4 2.3-1 Power Flow Operating Map...........................

B 3/4 2-6 0 3/4 3 Reactor Vessel Water Leve1.........................

B 3/4 3-8 8 3/4.4.6-1 Fast Neutron Fluence (E>l MeV) at 1/4 T as a Function of Service Life...........................

B 3/4 4-8 5.1.1-1 Exclusion Area.....................................

5-2 5.1.2-1 Low Population Zone................................

5-3 5.1.3-1 Map Defining Unrestricted Areas and Site Boundary for Radioactive Gaseous and Liquid Ef fluents.......

5-4 6.2.1-1 RBNG Organization..................................

6-3 6.2.2-1 River Bend Station Organization....................

6-4 l

RIVER BEND - UNIT 1 xxii Amendment No.12 j

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,T..___________._

INDEX LIST OF TABLES TABLE TITLE PAGE 1.1 Surveillance Frequency Notation..................,...

1-10 l

i 1.2 Operational Conditions..............................

1-11 2.2.1-1 Reactor Protection System Instrumentation Setpoinfs.

2-4 3.3.1-1 Reactor Protection System Instrumentation..........

3/4 3-2

3. 3.1-2 Reactor Protection System Response Times...........

3/4 3-6 4.3.1.1-1 Reactor Protection System Instrumentation Surveillance Requirements......

3/4 3-7 3.3.2-1 Isolation Actuation Instrumentation.................

3/4 3-12 l

3.3.2-2 Isolation Actuation Instrumentation Setpoints.......

3/4 3-19

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3.3.2-3 Isolation System Instrumentation Response Time......

3/4 3-24 i

4.3.2.1-1 Isolation Actuation Instrumentation Surveillance Requirements........................................

3/4 3-26 i

3.3.3-1 Emergency Core Cooling System Actuation Instrumentation......

3/4 3-31 3.3.3-2 Emergency Core Cooling System Actuation Instrumentation Setpoints..........

3/4 3-36

'i 3.3.3-3 Emergency Core Cooling System Response Times.

3/4 3-40 4.3.3.1-1 Emergency Core Cooling System Actuation Instrumentation Surveillance Requirements.

3/4 3-41 3.3.4.1-1 ATWS Recirculation Pump Trip System Instrumentation.

3/4 3-45 3.3.4.1-2 ATWS Recirculation Pump Trip System Instrumentation Setpoints...........

3/4 3-46 1

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RIVER BEND - UNIT 1 xxiii Amendment No.12 i

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v INDEX LIST OF TABLES (Continued)

TABLE TITLE PAGE 4'3.4.1-1 ATWS Recirculation Pump Trip Actuation Instrumentation Surveillance Requirements...........

3/4 3-47

. 3.3.4.2-1 End-of-Cycle Recirculation Pump Trip System Instrumentation.....................................

3/4 3-50

- 3.3.4.2-2 End-of-Cycle Recirculation Pump Trip System Setpoints...........................................

3/4 3-51

. 3.3.4.2-3 End-of-Cycle Recirculation Pump Trip System Response Time................................................

3/4 3-52 4.3.4.2 1-1 End-of-Cycle Recirculation Pump Trip System Surveillance Requirements...........................

3/4 3-53 3.3.5-1 Reactor Core Isolation Cooling System Actuating Instrumentation.....................................

3/4 3-55 3.3.5-2 Reactor Core Isolation Cooling System Actuation Instrumentation tetpoints...........................

3/4 3-57 4.3.5.1-1 Reactor Core Isolation Cooling System Actuation Instrumentation Surveillance Requirements...........

3/4 3-58 i

3.3.6-1 Control Rod Block Instrumentation...................

3/4 3-60 1-3.3.6-2 Control Rod Block Instrumentation Setpoints.........

3/4 3-62 l

4.3.6-1 Control Rod Block Instrumentation Surveillance Requirements........................................

3/4 3-63 3.3.7.1-1 Radiation Monitoring Instrumentation................

3/4 3-66 i

4.3.7.1-1 Radiation Monitoring Instrumentation Surveillance l

Requirements........................................

3/4 3-69 3.3.7.2-1 Seismic Monitoring Instrumentation..................

3/4 3-71 j

4.3.7.2-1 Seismic Monitoring Instrumentation Surveillance Requirements........................................

3/4 3-72 3.3.7.3-1 Meteorological Monitoring Instrumentation...........

3/4 3-74 4.3.7.3-1 Meteorological Monitoring Instrumentation Surveillance Requirements...........................

3/4 3-75 RIVER BEND - UNIT 1 xxiv

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OPERTlu)NAL CONDITIONS 1 and 2.

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'With THERMAL POWER ' exceed 3g 25% of. RATED THERMAL' POWER and the reactor vessel steam' dome pressure'less twn 785 psig or. core flow less than 10% of frated flow,

'be.in'atleastHOTSHUTDOWQithin2_hoursandcomplywith-the-requirementsof Specif kEtion-.7.1 Y

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2b The MINIM 0iICRITICAL POWER-RATIO (MCPR) shall not less than 1.07.with

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'the-reactorvessel.steamdome'dressuregreater-thinorequalto:785psigand' w

~c' ore: flow: greater than or equal.tuL10% of rated flow.

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M '(ICABILIT_YQ, OPERATIONAL CONDITIONS 1 and 2.

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Wth MCPR less hun 1.07 and-the reactor vessel. steam dome pressure greater l

than oriequal t O 85 psig and core flow greater than'or equal to 10% of rated flow,-- be 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.

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-REACTOR COOLANT-SYSTEM PRESSURE gsb.i 4,

7 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steak domec shall not exceed 1325 ppfg

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.APPLICQILITY:

OPERATIONALCONDIT{0NS1,2,3and4.

3.

' ACTION:

'S With the reactor coolant 9systie.

ressure above 1325 psig, as measured in,the reactor vessel steam domd be b at least HOT SHUTDOWN with reactor coolant system pressure less than'or equal to 1325. pig' 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 Specid cation 6.7.1.

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.4 RIVER BEND - UNIT 1 2-1 Amend 3ent No. 12 s *+

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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SAFETY LIMITS (Continued)

REACTOR VESSEL WATER LEVEL 2.1.4 The reactor vessel water level shall be above the top of the active irradiated fuel.

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APPLICABILITY: OPERATIONAL CONDITIONS 3, 4 and 5 ACTION:

With the reactor vessel water level at or below the top of the active irradiated fuel, manually initiate the ECCS to restore the water level, af ter depressurizing the reactor vessel, if required.

Comply with the requirements of Specification 6.7.1.

W RIVER BEND - UNIT 1 2-2

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2.1 SAFETY LIMITS BASES I

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 environs.

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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 l

approach is used to establish a Safety Limit such that the MCPR is not less than 1.07.

MCPR greater than 1.07 represents a conservative margin relative l

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 freedom 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 wnich 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.

I 2.1.1 THERMAL POWER, Low Pressure or Low Flow The use of the GE Critical Power correlation (Reference 1) is not valid l

for all critical power 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 28,000 lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi.

Thus, the bundle flow s

with a 4.5 psi driving head will be greater than 28,000 lbs/hr.

Full scale Al'LAS test data taken at 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.

RIVER BEND - UNIT 1 B 2-1 Amendment No. 12 i

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'SAFETV 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 boillng 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.

The 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 cladding integrity safety limit calculation are given in Reference 1.

Reference 1 includes a tabulation of the uncertain-ties 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 1.

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

NEDE-24011-P-A-8.

RIVER BEND - UNIT 1 L 2-2 Amendment No.12

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RIVER BEND - UNIT 1 B 2-4 Amendment No. 12 e

t 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR' LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE-PLANAR LINEAR HEAT GENERATION RATES (APLHGRs).for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown.in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 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:

With an APLHGR exceeding the limits of. Figure 3.2.1-1, 3.2.1-2, 3.2.1-3,

~3.2.1-4, 3.2.1-5 or-3.2.1-6, 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 All 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, 3.2.1-5 and 3.2.1-6:

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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 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 c.

with a LIMITING CONTROL ROD PATTERN for APLHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

l RIVER BEND - UNIT 1 3/4 2-1 Amendment No. 12

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VERSUS AVERAGE PLANAR EXPOSURE BP8 SRB 094 RIVER BEND - UNIT 1 3/42-2 Amendment No. 12

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3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 SHUT 00WN MARGIN A sufficient SHUT 00WN MARGIN ensures that 1) the reactor can be made sub-critical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition.

Since core reactivity values will vary through core life as a function of fuel depletion and poison burnup, the demonstration of SHUTOOWN MARGIN will be performed in the cold, xenon-fret condition and shall show the core to be subcritical by at least R + 0.38% delta k/k or R + 0.28% delta k/k, as appro-priate.

The value of R in units of % delta k/k is the difference between -

the calculated value of maximum core reactivity during the operating cycle and the calculated beginning-of-life core reactivity.

The value of R must be positive or zero and must be determined for each fuel loading cycle.

Two different values are supplied in the Limiting Condition for Operation to provide for the different methods of demonstration of the SHUTDOWN MARGIN.

The highest worth rod may be determined analytically or by test.

The SHUTDOWN MARGIN is demonstrated by an in-sequence control rod withdrawal at the beginning-of-life fuel cycle conditions and, if necessary, at any future time in the cycle if the first demonstration indicates that the required margin could be reduced as a function of exposure. Observation of subcriticality in this condition assures suberiticality with the most reactive control rod fully withdrawn.

This reactivity characteristic has been a basic assumption in the analysis of plant performance and can be best demonstrated at the time of fuel loading, but the margin must also be determined any time a control rod is incapable of insertion.

3/4.1.2 REACTIVITY ANOMALIES Since the SHUTDOWN MARGIN requirement for the reactor is small, a careful comparison of actual conditions to the predicted conditions is necessary, and the changes in reactivity can be inferred from these comparisons of rod patterns.

Since the comparisons are easily done, frequent checks are not an imposition on normal operations.

A 1% change is larger than is expected for normal operation so a change of this magnitude should be thoroughly evaluated.

A change as large as 1% would not exceed the design conditions of the reactor and is on the safe side of the postulated transients.

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l REACTOR COOLANT SYSTEMS BASES 3/4.1.3 CONTROL RODS The specifications of this section (1) ensure that the minimum SHUT 00WN MARGIN is maintained and the control rod insertion times are consistent with those used in the safety analyses, and (2) limit the potential effects of the rod drop accident.

The ACTION stateraents permit variations from the basic requirements but impose more restrictive criteria for continued operation.

A limitation on inoperable rods !s 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 SHUTDOWN MARGIN requirements.

The number of control rods permitted to be inoperable could be more than the eight allowed by tha " w ification, but the occurrence of eight inoperable rods could be indicauve of a generic problem and the reactor must be shut down for investigation and resolution of the problem.

The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from becoming less than 1.07 during the l

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 dverage response of all the drives as given in the specifications, provide the 1

required protection and MCPR remains greater than 1.07.

The occurrence of l

l l

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 1

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 l

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 l

I 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 Amendment No. 12 1

._________J

4 <

3/4.2 POWER DISTRIBUTION LIMITS i

BASES The' specifications of.this section assure.that the peak cladding temper -

-l

' ature following the postulated design basis. loss-of-coolant accident will. not exceed the 2200 F limit specified in-10 CFR.50.46.

i

(

3/4.2.1 ' AVERAGE PLANAR' LINEAR HEAT GENERATION RATE

. The peak cladding temperature:(PCT) following 'a ' postulated loss-of-coolant j

-accident is primarily a function of the average heat generation rate of all the. rods of.a fuel assembly at any axial location and is dependent only second-rily on the ' rod to ' rod power distribution within an assembly. The peak clad a

temperature.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'-

1

. times 1.02 is used'in the heatup: code along with the. exposure-dependent steady j

state gap conductance and rod-to-rod local peaking'factar.

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 limiting

.value:for APLHGR is'shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5.

.and 3.2.1-6.

The daily requirement for 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-i 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 R0D PATTERN 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.

The calculational procedure used to establish the APLHGR shown on Figures

- 3.2.1-1,3.2.1-2,3.2.1-3,3.2.1-4,3.2.1-5and3.2.1-6'isbasedonaloss-of-coolant l 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.

l a.

Input Changes _

1.

Corrected Vaporization Calculation - Coefficients in the vaporization correlation used in the REFLOOD code were corrected.

2.

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

RIVER BEND - UNIT 1 B 3/4 2-1 Amendment No. 12

- " - - - - - - ~ -

l POWER DISTRIBUTION LIMITS BASES 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.

1 2.

Incoporate NRC pressure transfer assumption - The assumption used in l

the SAFE-REFLOOD pressure transfer when the pressure is increasing I

was changed.

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

l These changes are listed below.

l a.

Input Change j

1 1.

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

b.

Model Change i

1.

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

A list of 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 sir.mlated 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 1.07 l

or that > Et 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 a peak power distribution, to ensure that an LHGR transient would not be increased in degraded conditions.

RIVER BEND - UNIT 1 8 3/4 2-2 Amendment No. 12

E' POWER DISTRIBUTION LIMITS Bases Table B 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS Plant Parameters; Core THERMAL POWER....................

3015 Mwt* which corresponds to 105% of rated steam flow Vessel Steam Output...................

13.08 x 10 lbm/hr which

[

6 corresponds to 105% of rated steam flow 1

Vessel Steam Dome Pressure.............

1060 psia Design Basis Recirculation Line Break Area for:

2 2

3 a.

Large Breaks 2.2 ft,

2 b.

Small Breaks 0.09 ft.

Fuel.'arameters:

l PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL ASSEMBLY GENERATION RATE PEAKING POWER j

FUEL TYPE GEOMETRY (kw/ft)

FACTOR RATIO j

3 Initial Core 8x8 13.4 1.4 1.17 f

A more detailed listing of input of each model and its source is presented in Section II of NEDE 20566(1) and subsection 6.3.3 of the FSAR.

• This power level meets the Appendix K requirement of 102%.

The core heatup calculation assumes an assembly power consistent with operation of-the highest powered rod at 102% of its Technical Specification LINEAR

-HEAT GENERATION RATE limit.

l L

l l

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

1 l

l

POWER DISTRIBUTION LIMITS BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO l

The required operating limit MCPRs at steady state operating conditions I

as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR of 1.07 and an analysis of abnormal operational transients.

For 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 'he largest de'.ta MCPR.

When added to the Safety Limit MCPR of 1.07, the required minimum operating limit MCPR of Specification 3.2.3 is obtained and is presented in Fi.gure 3.2.3-1.

The power-flow map of Figure B 3/4 2.3-1 shows typical regions of plant operation.

The evaluation of a given transient begins with the system initial param-eters identified in Reference 2 that are input to a GE core dynamic behavior transient computer program.

The codes used to evaluate transients are described in Reference 2.

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

The pu. pose of the MCPR and MCPR of Figu.es 3.2.3-1 and 3.2.3-2 is to f

p 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 and MCPR at the existing core flow and power state.

The MCPR s f

p f

are established to protect the core from inadvertent core flow increases such that the 99.9% MCPR limit requirement can be assured.

The MCPR s were calculated such that, for the maximum core flow rate and f

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.

Uting 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 is defined as MCPR.

f RIVER BEND - UNIT 1 B 3/4 2-4 Amendment No. 12

POWER DISTR.IBUTION LIMITS BASES-MINIMUM CRITICAL POWER RAT 1_0 (Continued) l The MCPR s are established to protect the core from plant transients other p

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 P

given core power level ~.

L At THERMAL POWER levels lless 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 l

evaluation will be made at 25% of RATED THERMAL POWER level with minimum recir-i culation pump speed.

The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary.

The I

daily requirement for calculating MCPR, when THERMAL POWER is greater than or

(

equal to 25% of RATED THERMAL POWER, is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.

The requirement for calculating MCPR 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 di tribution shifts while still allotting time for the power distribution is : tabilize.

The requirement for calculating MCPR after initially determining a LIMITING CONTROL R00 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 L

limit.

3/4,2.4 ' LINEAR HEAT GENERATION RATE i

i This specification assures that the Linear Heat Generation Rate (LHGR) in j

any rod is less than the design linear heat generation rate even if fuel pellet J

densification is postulated.

References:

l 1.

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

2.

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

l 1

RIVER BEND - UNIT 1 B 3/4 2-5 Amendment No.12

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