ML20134F410

Proposed Tech Specs 2.1.A & 3.10.C to Reflect Change in SLMCPR & Revise Operating CPR Limit for Stability, Respectively

ML20134F410
Person / Time
Site:	Oyster Creek
Issue date:	10/04/1996
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20134F400	List: ML20134F398 ML20134F406 ML20134F410
References
NUDOCS 9611050226
Download: ML20134F410 (7)

Text

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SECTION 2 s

. SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 S AFETY LIMIT - FUEL CLADDING INTEGRITY Applicability: Applies to the interrelated variables associated with fuel thermal behavior.

Obiective: To establish limits on the important thermal hydraulic variables to assure the integrity of the fuel cladding.

Specifications:

A. When the reactor pressure is greater than or equal to 800 psia and the core flow is greater than or equal to 10% of rated, the existence of a minimum CRITICAL POWER RATIO (MCPR) less than 1.09 shall constitute violation of the fuel l cladding integrity safety limit.

B. When the reactor pressure is less than 800 psia or the core flow is less than 10%

of rated, the core thermal power shall not exceed 25% of rated thermal power.

C. In the event that reactor parameters exceed the limiting safety system settings in Specification 2.3 and a reactor scram is not initiated by the associated protective instrumentation, the reactor shall be brought to, and remain in, the COLD SHUTDOWN CONDITION until an analysis is performed to determine whether the safety limit established in Specification 2.1.A and 2.1.B was exceeded.

D. During all modes of reactor operation with irradiated fuel in the reactor vessel, .

the water level shall not be less than 4'8" above the TOP OF ACTIVE FUEL. l Bases:

l 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 l

during reactor operation 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 procedure used to calculate the l

OYSTER CREEK 2.1 1 Amendment No.: 75,135, 9611050226 96100(

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I critical power result in an uncenainty in the value of the critical power. Therefore, the fuel l cladding integrity safety limit is defmed as the CRITICAL. POWER RATIO in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling 4

transition considering the power distribution within the core and all uncensinties.

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The Safety Limit MCPR is determined using the General Electric Thermal Analysis Basis, 3

GETAB"', which is a statistical model that combines all of the uncertamties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence

of boiling transition is determined using the General Electric Critical Quality (X) - Boiling Length (L), GEXL, correlation.

4 l The use of the GEXL correlation is not valid for the critical power calculations at pressures below

] 800 psia or core flows less than 10% of rated. Therefore, the fuel cladding integrity safety limit is

protected by limiting the core thermal power.

At pressures below 800 psia, the core elevation pressure drop (0 power, 0 flow) is greater than l 4.56 psi. At low power and all flows this pressure differential is maintained in the bypass region l

of the core. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and all flows will always be greater than 4.56 psi. Analyses show that

, with a flow of 28 x 10' lbs/hr bundle flow, bundle pressure drop is nearly independent of bundle  ;

power and has a value of 3.5 psi. Thus, bundle flow with a 4.56 psi driving head will be greater j

! than 28 x 10' lbs/hr irrespective of total core flow and independent of bundle power for the range  !

i of bundle powers of concern. Full scale ATLAS test data taken at pressures from 14.7 psia to  !

j 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt.

i With the design peaking factors this corresponds to a core thermal power of more than 50%.

Thus, a core thermal power limit of 25% for reactor pressures below 800 psi or core flow less i than 10% is conservative.

Plant safety analyses have shown that the scrams caused by exceeding any safety setting will assme that the Safety Limit of Specification 2.1.A or 2.1.B will not be exceeded. Scram times are checked periodically to assure the insenion times are adequate. The thermal power transient i resulting when a scram is accomplished other than by the expected scram signal (e.g., scram from

{ neutron flux fohmving clnsure of the main turbine stop valves) does not necessarily cause fuel j damage. Specificatire ? 1.C requires that appropriate analysis be performed to verify that backup protective instrum',n!ation has prevented exceeding the fuel cladding integrity safety limit prior to

resumption of POWER OPERATION. The concept of not approaching a Safety Limit provided j scram signals are OFERABLE is supponed by the extensive plant safety analysis.

If reactor water level should drop below the TOP OF ACTIVE FUEL, the ability to cool the core

, is reduced. This reduction in core

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OYSTER CREEK 2.1-2 Amendment No.: 75, a

cooling capability could lead to elevated cladding temperatures and clad perforation. With a water level above the TOP OF ACTIVE FUEL, adequate cooling is maintained and the decay heat can easily be accommodated. it should be noted that during power generation there is no  ;

clearly defined water level inside the shroud and what actually exists is a mixture level. This l i

mixture begins within the active fuel region and extends up through the moisture separators. For the purpose of this specification water level is defined to include mixture level during power operanons. ,

i The lowest point at which the water level can presently be monitored is 4'8" above the TOP OF j ACTIVE FUEL. Although the lowest reactor water level limit which ensures adequate core cooling is the TOP OF ACTIVE FUEL, the safety limit has been conservatively established at 4'8" above the TOP OF ACTIVE FUEL.

REFERENCES (1) NEDE-24011-P-A-11, General Electric Standard Application for Reactor Fuel and US Supplement NEDE-24011-P-A-11-US.

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2.1-3 Amendment No.: 75,135, OYSTER CREEK

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j 3.10 CORE LIMITS Applicability: Applies to core conditions required to meet the Final Acceptance Criteria for Emergency Core Cooling Performance.

Obiective: To assure conformance to the peak clad temperature limitations during a postulated loss-of-coolant accident as specified in 10 CFR 50.46 (January 4,1974) and to assure conformance to the operating limits for LOCAL LINEAR HEAT GENERATION RATE and minimum CRITICAL POWER RATIO.

Specification:

A. AVERAGE PLANAR LHGR During POWER OPERATION the maximum AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for each fuel type as a function of l l

exposure shall not exceed the limits specified in the CORE OPERATING LIMITS REPORT (COLR).

i If at any time during POWER OPERATION it is determined by normal surveillance )

that the limiting value for APLHGR is being exceeded, action shall be initiated to i restore operation to within the prescribed limits. If the APLHGR is not returned to  !

within the prescribed limits within two (2) hours, action shall be initiated to bring the reactor to the COLD SHUTDOWN CONDITION within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. During this period surveillance and corresponding action shall continue until reactor operation is ,

within the prescribed limits at which time POWER OPERATION may be continued. l B. LOCAL LHGR ,

During POWER OPERATION, the LOCAL LINEAR HEAT GENERATION RATE (LHGR) of any rod in any fuel assembly, at any axial location shall not exceed the maximum allowable LHGR limits specified in the COLR. ,

If at any time during operation it is determined by normal surveillance that the limiting value of LHGR is being exceeded, action shall be initiated to restore operation to within the prescribed limits. If the LHGR is not returned to within the prescribed limits within two (2) hours, action shall be initiated to bring the reactor to 1 the COLD SHUTDOWN CONDITION within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> During this period, surveillance and corresponding action shall continue until reactor operation is within the prescribed limits at which time POWER OPERATION may be continued.

OYSTER CREEK 3.10-1 Amendment No.: 48,75,129,147,

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C. Minimum CRITICAL POWER RATIO (MCPR)

During steady state POWER OPERATION the minimum CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limit as specified in the COLR.

The MCPR limit for each cycle as identified in the COLR shall be greater than or equal to 1.49. l When APRM status changes due to instrument failure (APRM or LPRM input failure), the MCPR requirement for the degraded condmon shall be met within a time interval of eight (8) hours, provided that the control rod block is placed in operation during this interval.

For core flows other than rated, the nominal value for MCPR shall be increased by a factor of kr, where kr is specified in the COLR.

If at any time during POWER OPERATION it is determined by normal sutveillance

, that the limiting value for MCPR is being exceeded for reasons other than instrument failure, action shall be initiated to restore operation to within the prescribed limits. If the steady state MCPR is not returned to within the prescribed limits within two [2]

hours, action shall be initiated to bring the reactor to the COLD SHUTDOWN CONDITION within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. During this period, surveillance and corresponding action shall continue until reactor opention is within the prescribed limit at which time POWER OPERATION may be continued.

Bases:

The Specification for AVERAGE PLAN AR LHGR assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200*F limit specified in 10 CFR 50.46. The analytical methods and assumptions used in evaluating the fuel design limits are presented in FSAR Chapter 4.

LOCA analyses are performed for each fuel design at selected exposure points to determined  !

APLHGR limits that meet the PCT and maximum oxidation limits of 10 CFR 50.46. The analysis l is performed using GE calculational models which are consistent with the requirements of 10 CFR l l

50, Appendix K.

The PCT following a postulated LOCA is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is not strongly influenced by the rod to rod power distribution within an assembly. Since expected location variations in power OYSTER CREEK 3.10-2 Amendment No.: 48,75,111,129,147, 176,

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! *. distribution within a fuel assembly affect the calculated peak clad temperature by less than 20*F l

' relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation ra:e is sufficient to assure that calculated temperatures are below the limits specified in l 10 CFR 50.46.

l l The maximum AVERAGE PLANAR LHGR limits for the various fuel types currently being used

are provided in the COLR. The MAPLHGR limits for both five-loop and four-loop operation with the idle loop unisolated are shown. Four-loop operation with the idle loop isolated (suction,

discharge and discharge bypass valves closed) requires that a MAPLHGR multiplier of 0.98 be i applied to all fuel types. Additional requirements for isolated loop operation are given in

. Specificadon 3.3.F.2.

j Fuel design evaluations are performed to demonstrate that the cladding 1% plastic strain and

- other fuel design limits are not exceeded during anticipated operational occurrences for operation with LHGRs up to the operating limit LHGR.

The analytical methods and assumptions used in evaluating the anticipated operational occurrences to establish the operating limit MCPR are presented in the FSAR, Chapters 4, 6 and 15 and in Technical Specification 6.9.1.f. To assure that the Safety Limit MCPR is not exceeded during any moderate frequency transient event, limiting transients have been analyzed to determine the largest reduction in CRITICAL POWER RATIO (CPR). The types of transients evaluated are pressurization, positive reactivity insertion and coolant temperature decrease. The operational MCPR limit is selected to provide margin to eccanmadate transients and uncertainties in monitoring the core operating state, manufacturing, and in the critical power correlation itself. This limit is derived by addition of the MCPR for the most limiting transient to the safety limit MCPR designated in Specification 2.1.

A lower bound of 1.49 has been established for the operating limit MCPR value to provide l sufficient margin to the MCPR safety limit in the event of reactor thermal-hydraulic instability.

The 1.49 limit will be considered against the minimum operating MCPR limit based on reload l transient and accident analysis. The higher of core stability or reactor transient and accident determined MCPR will be used to determine the cycle operating limit. l

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The APRM response is used to predict when the rod block occurs in the analysis of the rod l withdrawal error transient. The transient rod position at the rod block and corresponding MCPR  !

can be determined. The MCPR has been evaluated for different APRM responses which would result from changes in the APRM status as a consequence of biTM APRM channel and/or failed / bypassed LPRM inputs. The steady state MCPR required to protect the minimum transient MCPR for the worst case APRM status condition (APRM Status 1) is determined in the rod l

withdrawal error transient analysis. The steady state MCPR values for APRM status conditions 1, 2, and 3 will be evaluated each cycle. For those cycles where the rod withdrawal error transient is not the most severe transient the MCPR value for APRM status conditions 1,2, and 3 will be the same and be equal to the limiting transient MCPR value.

3.10-3 Amendment No.: 75,129,147,176, OYSTER CREEK

The time interval of Eight (8) hours to adjust the steady state of MCPR to account for a degradation in the APRM status isjustified on the basis ofinstituting a control rod block which precludes the possibility of experiencing a rod withdrawal error transient since rod withdrawal is physically prevented This time interval is adequate to allow the operator to either increaw the MCPR to the appropriate value or to upgrade the status of the APRM system while in a condition which prevents the possibility of this transient occurring.

Transients analyzed each fuel cycle will be evaluated with respect to the operational MCPR limit specified in the COLR.

The purpose of the kr facteds to define operating limits at other than rated flow conditions. At

less than 100% flow the required MCPR is the product of the operating limit MCPR and the kr

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factor. Specifically, the kr factor provides the required thermal margin to protect agamst a flow j increase transient.  ;

i The kr factor curves, as shown in the COLR, were developed generically using the flow control l line corresponding to rated thermal power at rated core flow. For the manual flow control mode,

< the kr factors were calculated such that at the maximum flow state (as limited by the pump scoop i tube set point) and the corresponding core power (along the rated flow control line), the limiting l

I bundle's relative power was adjusted until the MCPR was slightly above the Safety Limit. Using this relative bundle power, the MCPR's were calculated at different points along the rated flow control line corresponding to different core flows. The ratio of the MCPR calculated at a given l

point of core flow, divided by the operating limit MCPR determines the value of kr.

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4 3.10-4 Amendment No.: 75,129,140,147, OYSTER CREEK