ML20012G544
| ML20012G544 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 02/27/1993 |
| From: | NIAGARA MOHAWK POWER CORP. |
| To: | |
| Shared Package | |
| ML20012G542 | List: |
| References | |
| NUDOCS 9303090036 | |
| Download: ML20012G544 (20) | |
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. ATTACIIMENT A NIAGARA MOIIAWK POWER CORPORATION LICENSE NO. NPF-69 DOCKET NO. 50-410 Proposed Changes to the Technical Specifications Replace existing pages 1-2 and 3/4 3-63 with the attached revised pages. Replace existing Bases pages B2-1, B2-2, B2-3, B2-4, B3/4 2-2, B3/4 2-3 and B3/4 2-4 with the attached revised pages. The pages have been retyped in their entirety and have marginal markings to indicate the changes to the text. !
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DEFINITIONS ;
CHANNEL FUNCTIONAL TEST 1.6 (Continued) ;
i The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps so that the entire channel is tested.
CORE ALTERATION l
1.7 CORE ALTERATION shall be the addition, removal, relocation, or movement of fuel, l sources, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Normal movement of the SRMs, IRMs, TIPS or special movable detectors is not considered a CORE ALTERATION. Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position.
CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY 1.8 The CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY (CMFLPD) shall be the '
highest value of the FLPD which exists in the core.
CRITICAL POWER RATIO 1.9 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly '
which is calculated by application of an approved critical power correlation to cause some point in the assembly to experience boiling transition, divided by the actual fuel l assembly operating power.
- DOSE EQUIVALENT l-131 1.10 DOSE EQUIVALENT l-131 shall be that concentration of I-131, expressed in i microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131,1-132,1-133,1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table ll1 _
of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites." !
E - AVERAGE DISINTEGRATION ENERGY 1.11 E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, expressed in MeV, for isotopes, with half-lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.
EMERGENCY CORE COOLING SYSTEM RESPONSE TIME 1.12 The EMERGENCY CORE COOLING SYSTEM (ECC,S) RESPONSE TIME shall be that time l interval from when the monitored parameter exceeds its ECCS actuation setpoint at the I channel sensor until the ECCS equipment is capable of performing its safety function, )
i.e., the valves travel to their required positions, pump '
NINE MILE POINT - UNIT 2 1-2 Amendment No.
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l 2.1 BASES FOR SAFETY LIMITS 2.
1.0 INTRODUCTION
The fuel cladding, reactor pressure vessel, and primary system piping are the principal barriers to the release of radioactive materials to the environs. 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 so that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit so that the MCPR is not less than 1.07 for two recirculation loop operation and 1.08 for single recirculation loop operation. MCPR greater than 1.07 for two recirculation loop operation and 1.08 for single recirculation loop operation represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers that 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 that occur from reactor operation significantly above design conditions and the Limiting Safety System Settings. Although fission product migration from cladding perforation is just as measurable as that from use-related cracking, the thermally caused cladding perforations signal a threshold 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 that would produce onset of transition 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 The use of critical power correlations is not valid for all critical power calculations performed at reduced 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 x 10'lb/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a'4.5-psi driving head will be greater than 28 x 105 lb/hr. Full-scale ATLAS 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.
NINE MILE POINT - UNIT 2 B2-1 Amendment No. 36
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l l BASES FOR SAFETY LIMITS 2.1.2 THERMAL POWER, Hiah Pressure and Hiah Flow l The fuel cladding integrity Safety Limit is set so that no fuel damage is calculated to occur if the limit is not violated. Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling havo been used to mark the beginning of the region in which fuel damage could occur. Although it is recognized that a departure from nucleate boiling woald 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 CR 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 uncertainties.
The Safety Limit MCPR is determined using a statistical model that combines all of the l
uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using an
, approved critical power correlation. The critical power correlation is valid over the range of conditions used in the tests of the data used to develop the correlation. ,
The required input to the statistical model are the uncertainties listed in Bases Table <
B2.1.2-1 and the nominal values of the core parameters listed in Bases Table 82.1.2-2. l The bases for the uncertainties in the core parameters and the basis for the uncertainty !
l in the critical power correlation are given in Reference 1. The power distribution is I based on a typica! 764 assembly core in which the rod pattern was arbitrarily chosen to produce a skewed power distribution having the greatest number of assemblies et the highest power levels. The worst distribution during any fuel cycle would not be as severe as the distribution used in the analysis.
References:
- 1. General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A (latest approved revision). I NINE MILE POINT - UNIT 2 B2-2 Amendment No.
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BASES TABLE B2.1.2-1 -
UNCERTAINTIES USED IN THE DETERMINATION OF THE FUEL CLADDING SAFETY LIMIT * !
STANDARD !
DEVIATION l OUANTITY (% OF POINT) l t
Feedwater Flow 1.76 i Feedwater Temperature 0.76 t
Reactor Pressure 0.5 _
q Core inlet Temperature 0.2 Core Total Flow 2.5 l j Channel Flow Area 3.0 j Friction Factor Multiplier 10.0 i Channel Friction Factor Multiplier 5.0 !
TIP Readings 8.7 l l R Factor 1.6 Critical Power 3.0 l l
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The uncertainty analysis used to establish the corewide Safety Limit MCPR is based on the assumption of quadrant power symmetry for the reactor core. The-values herein apply to both two recirculation loop operation and single recirculation loop operation. l NINE MILE POINT - UNIT 2 B2-3 Amendment No.)ff
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i BASES TABLE B2.1.2-2 l NOMINAL VALUES OF PARAMETERS' USED IN l THE STATISTICAL ANALYSIS OF FUEL CLADDING INTEGRITY SAFETY LIMIT l t
PARAMETER VALUE -!
THERMAL POWER 3293 MW Core Flow 102.5 Mlb/hr i
Dome Pressure 1005 psig l
Bundle Enrichment 3.0 Wt % U-235 !
i R-Factor. ;
O - 10 GWD/ST O.915 f
10 - 15 GWD/ST 0.954 l
> 15 GWD/ST 0.954 i
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The values in this table are for a representative plant. l NINE MILE POINT - UNIT 2 B2-4 Amendment No.
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z Tahle 3.3.6-2 (Continued 1 z
CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS E
i-n u TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE o
{ 6. Reactor Coolant System Recirculation Flow c- a. Upscale 5111% rated flow 5114% rated flow l Z b. Inoperative NA NA H c. Comparator s10% flow deviation s11% flow deviation 82
- 7. Reactor Mode Switch
- a. Shutdown Mode NA NA
- b. Refuel Mode NA NA ca 5
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Specified in the CORE OPERATING LIMITS REPORT.
9 For fuel loading and startup from refueling the count rate may be less than 3 cps if the following conditions are
@ met: the signal to noise ratio is greater than or equal to 5, and the signal is greater than 1.3 cps.
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POWER DISTRIBUTION LIMITS BASES I
3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady-state operating conditions as specified in .
Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit !
MCPR, and an analysis of abnormal operational transients. For any abnormal 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 2.2. .
To assure that the fuel cladding integrity Safety Limit is not exceeded during any ,
anticipated abnormal operational transient, the most limiting transients have been !
analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO i (CPR). The type of transients evaluated were loss of flow, increase in pressure and l power, positive reactivity insertion, and coolant temperature decrease.
- The evaluation of a given transient begins with the system initial parameters shown in f
, USAR Tables 15.0-3 and A15.0-4 that are input to a GE-core dynamic behavior transient !
computer program. The codes used to evaluate transients are discussed in Reference 2.
The principal result of this evaluation is the reduction in MCPR caused by the transients.
The purpose of the K, factor specified in the CORE OPERATING LIMITS REPORT is to define operating limits at other than rated core flow conditions. At less than 100% of ,
rated flow, the required MCPR is the product of operating limit MCPR and the K, factor. l The ry factors assure that the Safety Limit MCPR will not be violated. The K, factors are calculated as described in Reference 2. l 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 this point, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerable margin. During initial startup testing of the plant, an MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum recirculation 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 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 completion of s THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shifts, NINE MILE POINT - UNIT 2 B 3/4 2-2 Amendment No.#
BASES TABLE B3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS' PARAMETERS VALUE Plant:
- 1. C ore TH E R MAL PO W E R . . . . . . . . . ... . . . . . . . . . . . . . . . .. .. . . .. . . . . ... . ... . . . .. . .. . . 3461 MWt*
- which corresponds to 105% of rated steam flow
- 2. Ve ssel Ste a m O ut put . . . . . . . . . . . . . . .. .. .. . . . . . . . . .. . . . . . .. . . . . .. . . . . . . . . . . . . . . . . . 15.0 x 10' lbm/hr which corresponds to 105% of rated steam flow
- 3. Vessel Steam Dome Pressure ..... .......................................... 10 5 5 psia
- 4. Design Basis Recirculation Line Break Area for:
- a. La rg e B r e a k s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1 f t 2
- b. S m a 11 B r e a ks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 0. 0 9 f t '
Euel:
- PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE GENERATION RATE PEAKING POWER FUEL TYPE GEOMETRY (kW/ft) FACTOR RATIO' Initial Core 8x8 13.4 1.4 1.20 Reload 8x8 14.4 1.4 1.20 A more detailed listing of input of each model and its source is presented in Section 11 of Reference 1 and subsection 6.3.3 of the USAR. l This power level meets the Appendix K requirement of 102%. The core heatup calculation assumes a bundle power consistent with operation of the highest powered rod at 102% of its Technical Specification LINEAR HEAT GENERATION RATE limit.
' For single recirculation loop operation, loss of nucleate boiling is assumed at 0.1 second after LOCA regardless of initial MCPR.
NINE MILE POINT - UNIT 2 B 3/4 2-3 Amendment No.
POWER DISTRIBUTION LIMITS BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO 3/4.2.3 (Continued) while still allotting time for the power distribution to stabilize. The requirement for calculating MCPR after initially determining that a LIMITING CONTROL ROD PATTERN exists ensures MCPR will be known following a change in THERMAL POWER or power shape, and therefore avoid operation while 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.
The daily requirement for calculating LHGR 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 to calculate LHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion 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 LHGR af ter initially determining a LIMITING CONTROL ROD PATTERN exists ensures that LHGR will be known following a change in THERMAL POWER or power shape that could place operation exceeding a thermal limit.
References
- 1. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566, latest approved revision.
- 2. General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A, latest approved revision.
NINE MILE POINT - UNIT 2 B 3/4 2-4 Amendment No.M
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. t ATTACHMENT B NIAGARA MOIIAWK POWER CORPORATION .
l LICENSE NO. NPF-69 DOCKET NO. 50-410 t
, Suonoriine Information and No Sinnificant Hazards Consideration Analysis
1.0 BACKGROUND
The proposed amendment consists of several changes that will (1) facilitate operation at up to 105% rated core flow, (2) permit the use of NRC approved critical power correlations in Nine Mile Point Unit 2 (NMP2) Reload Analyses, and (3) incorporate changes to references,
- General Electric analytical techniques and the Updated Safety Analysis Report (11SAR) into the NMP2 Technical Specifications Bases. Full power operation at core flows up to 105% of ;
rated has been previously evaluated in Appendix A of the NMP2 USAR. NMP2 is analyzed 6 for and has the capacity to provide core flows of up to 105% of rated and, as such, increased ,
core flow (ICF) provides a valuable method of extending full power capability beyond the ,
" reactivity life" at rated conditions. Currently, because of uncertainties and flow noise, -
operation at greater than 100% rated core flow causes spurious rod block signals and nuisance alarms. Therefore, to allow full utilization of the increased operating domain, L modifications to the high recirculation flow control rod block setpoints are required. The !
proposed amendment would change the rod block instrument setpoints for the reactor coolant .
system recirculation flow upscale condition from 108% to 111% of rated flow and from 111% to 114% of rated flow for the trip setpoint and allowable value, respectively.
I The assumed initial conditions for the current cycle specific transient analyses are provided m [
Appendix A of the NMP2 USAR. These initial conditions include the limiting operating ;
points in the increased core flow region. Therefore, cycle specific limits for the current cycle bound operation in the ICF region. To determine the relative effect of increased core flow ;
operation on operating limits, specific ICF analyses were performed for the first operating cycle. A summary of the results of that analysis is provided in Section 3.0 of this ;
attachment.
Definition 1.9, " Critical Power Ratio," defines Critical Power Ratio (CPR) as the ratio of power in an assembly that causes some point in the assembly to experience boiling transition divided by the actual fuel assembly power. Definition 1.9 further states that the power level i at which boiling transition occurs shall be determined by the application of the GEXL ;
correlation, where "GEXL" is a term used to refer to a specific NRC approved critical l power correlation developed by General Electric (GE) to predict the onset of boiling- l transition in boiling water reactors. This application replaces the term "GEXL" with a more generic description of the required correlation.' The change will allow Niagara Mohawk to use any NRC approved GE critical power correlation in the determination of operating limit MCPRs. Bases changes to Section B2.1 are included to account for the change to Definition 1.9.
003414LL Page 1 of 10
Additional changes are proposed to the Bases to (1) incorporate changes to GE's approved analytical techniques, (2) update references in the Bases and (3) reflect changes to Appendix A of the USAR, " Reload Analysis." !
2.0 DESCRIPTION
OF CIIANGE j
, 2.1 Recirculation Flow Uprate Rod Block i
l Table 3.3.6-2. " Control Rod Block Instrumentation Setooints" Item 6.a. Reactor Coolant System Recirculation Flow - Upscale: ;
Present Proposed ;
Setpoint s108% rated flow s111% rated flow Allowable Value s111% rated flow s114% rated flow I
This change will permit operation at greater than 100% rated core flow.
l 2.2 Critical Power Correlation <
Definition 1.9. " Critical Power Ratio" Present Wording: The CRITICAL POWER RATIO (CPR) shall be the ratio of that ;
power in the assembly which is calculated by application of the GEXL correlation to cause some point in the assembly to experience boiling transition, divided by the actual fuel assembly operating power.
1 Proposed Wording: The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of an approved critical power correlation to cause some point in the assembly to experience boiling transition, divided by the actual fuel assembly operating power.
This change will allow the utilization of other NRC approved critical power correlations.
2.3 Bases Changes Bases Section B2.1 and B3/4.2 The changes to Section B2.1 reflect the change to Definition 1.9. Additional changes to Section B2.1, including Tables B2.1.2-1 and B2.1.2-2, reflect changes to the statistical model and uncertainties used in the determination of the fuel cladding safety limit. Changes to B3/4.2 update references to the USAR, reflect design values for newer fuel, and revise references in accordance with changes to GE analytical techniques.
003414LL Page 2 of 10-
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3.0 EVALUATION 3.1 Recirculation Flow Upscale Rod Block Full power operation at core flows up to 105% of rated has been previously evaluated in Appendix A of the NMP2 USAR. The assumptions for the current NMP2 reload analysis include ICF conditions. The MCPR operating limits provided in the current Core Operating Limits Report bound operation in the ICF region and assure that the Safety Limit MCPR will not be violated. Specific analyses, performed for the first fuel cycle, determined the relative '
effect of ICF operation on plant analyses and the MCPR operating limit. A summary of the safety evaluation performed for ICF is provided below:
MCPR Operating Limit Abnormal Operating Transients An analysis of the limiting abnormal operating transients described in the NMP2 USAR evaluated the effect of operation in the ICF region, using nuclear transient data consistent with the original USAR analysis. The most limiting transients are the Load Rejection with Bypass Failure and Feedwater Controller Failure events. The results of the evaluation show that, based on data from the first cycle, the licensing basis point (104.3P,100F) bounds the ACPR values for the Feedwater Controller Failure and the Imad Rejection with Bypass Failure transients in the ICF region. In the ICF region, the intermediate ACPR values for the lead Rejection transient increase only slightly
(<2%). Therefore, operation in the ICF region does not significantly affect the ODYN Option A and Option B Cycle 1 operating limits for bypass and end-of-cycle recirculation pump trip (EOC RPT) operable.
The transients were also examined for equipment out of service options, including turbine bypass and EOC RPT inoperable. The ODYN Option A and Option B Cycle 1 MCPR operating limits in the ICF region for turbine bypass inoperable remain unchanged from the previously established licensing basis values. For EOC RPT inoperable, the Option B operating limit remains unchanged from the licensing basis value, while the Option A operating limit, which is based on the Load Rejection with Bypass Failure transient, increases by 0.01.
For the current cycle (Cycle 3), the limiting pressurization transient for normal operation is Load Rejection with Bypass Failure, with an Option A limit of 1.35 and an Option B limit of 1.31. The limiting MCPR value for normal operation is a function of the rod withdrawal error event, with a MCPR operating limit of 1.37. !
With Turbine Bypass or EOC RPT out of service, the Option A MCPR operating limit l is 1.39 and the Option B MCPR operating limit is 1.35. All the Cycle 3 MCPR :
operating limits are bounding for operation in the ICF region. '
003414LL Page 3 of 10 l
Rod Withdrawal Error The current licensing basis Rod Withdrawal Error transient analysis is based on a rod block monitor (RBM) setpoint of 110% at 100% core flow. The rod block monitor setpoint is a function of power and flow and the RBM setpoint increases as core flow is increased. For operation between 100% and 105% of rated core flow, the RBM upscale setpoint is " clamped" at the 100% core flow value of 110%. Therefore, the ACPR values detennined at 100% core flow bound those at higher core flows.
End-of-Cycle Power Coastdown During end-of-cycle power coastdown, decreasing the power from the 100% rated condition alorg the 105% core flow limit line will result in an increase in the transient ACPR values for some events. However, this increase is less than the increase in operating CPR due to the power decrease, and therefore, such operation will not result in postulated transient events violating the Safety Limit MCPR along this increased flow line.
LOCA and ECCS Performance The effect of increased core flow on loss-of-coolant accident analyses is not significant because the parameters that most strongly affect the calculated peak cladding temperature (PCT), high power node boiling transition and core reflooding time, are relatively insensitive to increased core flow conditions. Results of the LOCA analysis performed show that the PCT for ICF conditions increases by less than 5"F throughout the break spectrum when compared to rated flow conditions. The calculated PCT for operation at the 100P/100F point is 1921'F, so substantial margin is maintained below the limit of 2200"F.
Overoressurization The limiting transient for ASME code overpressurization analysis (MSIV closure with flux scram) was evaluated for EOC conditions in the ICF region. The analysis for the ICF region produced a peak pressure of 1272 psig, which is slightly higher than the 1268 psig for the standard operating domain, but well below the upset code limit of 1375 psig and is, therefore, acceptable.
Containment The effect ofincreased core flow operation on the containment LOCA response was evaluated qualitatively for BWR Mark II containments. For a given operating power level, the equivalent power level within the ICF region will have a lower reactor coolant subcooling t because of the lower rod line operation. The lower initial coolant subcooling will lead to a less severe containment response during a LOCA. This trend is also applicable to NMP2.
This qualitative assessment leads to the conclusion that the containment LOCA response for !
ICF operation is bounded by the corresponding USAR results.
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Reactor internals The internal pressure differences on the reactor internal components and core support structures were evaluated to define a set of bounding loads that would apply to increased core !
flow operation. The reactor internal components most affected by ICF operation were i
evaluated using these bounding loads under normal, upset, emergency and faulted loads. The evaluation determined that the stresses produced in these and other components are within the allowable design limits given in the USAR, including the ASME Code,Section III, Subsection NG.
Fuel Assemblics The fuel assemblies, including fuel bundles and channels, were evaluated under normal, t upset, faulted and fatigue load combinations for increased core flow operation considering the effects ofloads discussed above. Results of the evaluation demonstrate that the fuel assemblies are adequate to withstand effects to 105% rated flow.
Flow-Induced Vibration To assure that the flow-induced vibration response of the reactor internals is acceptable, a single reactor of each product line and size undergoes an extensive vibration test during the initial plant startup. After analyzing the results of those tests and assuring that all responses fall within acceptable limits of the established criteria, the reactor is classified as a valid prototype in accordance with Regulatory Guide 1.20. All other reactors of the same product line and size undergo a less rigorous confirmation test to assure similarity to the base test.
The increased core flow vibration analysis was performed by analyzing the startup test vibration data for the prototype BWR/5-251 plant applicable to NMP2. Based on the analysis results from the prototype plant,105% increased core flow operation is feasible for NMP2. Evaluation of the startup test data for NMP2 confirmed this conclusion.
The results of the safety evaluation performed for ICF demonstrated that the current Technical Specifications are adequate to preclude the violation of any Safety Limits during ICF operation, as documented in the USAR. To maintain adequate operating margin and account for flow noise, the control rod block instrument setpoints for the reactor coolant l system recirculation flow upscale condition, contained in Table 3.3.6-2, " Control Rod Block l
Instrumentation Setpoints," are being changed from 108% to 111% of rated flow and from j 111% to 114% of rated flow for the trip setpoint and allowable value, respectively. These I setpoints are being increased to maintain an adequate trip setpoint margin allowing for uncertainties and noise, so as not to impact plant high flow operations. The operational design basis for this rod block is to alert the operator to unexpected high core flow operation. The flow unit upscale trip performs no safety function and no design basis transient or accident analysis takes credit for it. Whereas the licensed core flow range was increased by 5% (from 100% to 105%), the high flow trip setpoint is being increased only 3% (from 108% to 111%); therefore, the function of the trip as an indication / alarm of unintended high flow operation is actually enhanced.
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3.2 C~ritical Power Correlation The Safety Limit MCPR is established to protect the integrity of the fuel cladding, one of the principal barriers to the release of radioactive materials to the environment. This fuel cladding Safety Limh is set so that no fuel damage is expected to occur if the limit is not viola *ed. The required Operating Limit MCPRs are derived from the Safety Limit MCPR and an analysis of abnormal operational transients, where the resulting 6CPR does not decrease MCPR below the Safety Limit MCPR. ,
i The NMP2 Technical Specifications, in Definition 1.9, " Critical Power Ratio," defines Critical Power Ratio as the ratio of power in an assembly that causes some point in the assembly to experience boiling transition divided by the actual fuel assembly power. l Definition 1.9 further states that the power level at which boiling transition occurs shall be determined by the application of the GEXL correlation, where "GEXL" is a term used to refer to a specific NRC approved critical power correlation developed by GE to predict the onset of boiling transition in boiling water reactors. GE is continually improving its ,
analytical techniques and as better correlations are developed and approved by the Staff, they are incorporated into plant analyses. Therefore, to accommodate utilization of these improved correlations, this application replaces the term "GEXL" with a more generic description of the required correlation. The change will allow Niagara Mohawk to use any NRC approved GE critical power correlation in the determination of operating limit MCPRs. Revising the definition of critical power ratio is a potential modification to analytical techniques; however, the transient analyses will continue to be performed using the methodologies approved for the Core Operating Limits Report. The resultant operating limit MCPRs will still assure that the Safety Limit MCPR is not violated during abnormal operating transients. Associated changes are proposed to Bases Section B2.1 to reflect the definition change.
3.3 Bases Changes Changes are proposed to Section B2.1.2, " Thermal Power," to update the references. The current references, NEDO-10958-A and NEDO-20340, are no longer maintained current by GE. Changes to GE methodology are implemented only through NEDE-24011-P-A-US,
" General Electric Standard Application for Reactor Fuel (GESTAR-II)," therefore it is a more appropriate reference.
Changes are also proposed to Table B2.1.2-1, " Uncertainties Used in the Determination of the Fuel Cladding Safety Limit," and Table B2.1.2-2, " Nominal Values of Parameters Used in the Statistical Analysis of Fuel Cladding Safety Limit." The changes to Table B2.1.2-1 incorporate values for newer fuel designs, including GE9, which is now used in NMP2. The new values bound the older fuel designs still in use. The changes to Table B2.1.2-2 correct an error in the original table. The statistical analysis for fuel cladding integrity uses nominal values from a representative plant. The actual values for NMP2 were inadvertently provided when the table was developed.
1 Additional changes are proposed to Section B3/4.2, " Power Distribution Limits." In Section B3/4.2.3, " Minimum Critical Power Ratio," USAR refcrences are revised to reflect the reload analysis section added to the USAR. References which are no longer maintained current by GE are removed. The methodologies and information contained in these I
003414LL Page 6 of 10
references is also contained in GESTAR-II, which remains as a reference. Finally, Table !
B3.2.1-1, "Significant Input Parameters to the loss-of-Coolant Accident Analysis," is revised to incorporate design parameters for GE9 fuel, which is now utilized at NMP2. .
4.0 CONCLUSION
l The proposed amendment consists of several changes that will (1) permit unencumbered operation at up to 105% rated core flow, (2) permit the use of NRC approved critical power correlations in NMP2 reload analyses, and (3) incorporate changes to references, GE analytical techniques and the USAR into the NMP2 Technical Specifications Bases. !
r The proposed amendment would change the instrument setpoints for the reactor coolant j system recirculation flow upscale condition from 108% to 111% of rated flow and 111% to '
114% of rated flow for the trip setpoint and allowable value, respectively. This change ,
would alleviate spurious rod block actuations and nuisance alarms currently associated with operation above 100% of rated core flow. The operational design basis for high flow rod block is to provide an operator warning signal in response to unexpected high core flow
- operation. The flow unit upscale trip performs no safety function and no design basis ,
transient or accident analysis takes credit for it.
3 The assumptions for the current NMP2 reload analysis include ICF conditions. Therefore, f j the MCPR operating limits provided in the current Core Operating Limits Report bound i j operation in the ICF region and assure that the Safety Limit MCPR will not be violated. The li consequences of anticipated operational occurrences and the design basis LOCA have been 1
evaluated using NRC approved methods. PCT has been evaluated and substantial margin ,
still exists below the limit of 220CTF. Loads on reactor internals, containment, and piping systems have been evaluated and stresses remain within design limits. .
3 Assurance of the continued operability of components in the increased core flow region has !
been previously demonstrated. Sufficient margin exists in the design basis for each system to i accommodate operation at up to 105% core flow. The aggregate affect of operation at up to !
105% of core flow has been evaluated and found to have no resulting impact on system reliability or performance. The proposed changes provide necessary operating margin such that the plant can operate unencumbered at 105% core flow.
"GEXL" is a term used to refer to the various NRC approved critical power correlations
, developed by GE to predict the onset of boiling transition in boiling water reactors. This application replaces the term GEXL with a more generic description of the required correlation. Revising the definition of critical power ratio is a modification to analytical techniques; however, the transient analyses will continue to be performed using critical power correlations approved by the Staff. The resultant operating limit MCPRs will still assure that the Safety Limit MCPR is not violated during abnormal operating transients.
4 I
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. .- . _ _ ._ _ - _ _ _ . _ . _ . _ ~ . _ ..
)
The changes to Bases Section B2.1 and B3/4.2 are proposed to reflect the change to Definition 1.9 and update the references. In addition, analytical values and fuel design data I are being revised to reflect the NMP2 reload values. ;
Therefore, based on the above analysis, there is reasonable assurance that operation of ;
NMP2 in the proposed manner will not endanger the public health and safety and that ;
issuance of the proposed amendment will not be inimical to the common defense and l security. ;
10 CFR f 50.91 requires that at the time a licensee requests an amendment, it must provide ;
to the Commission its analysis using the standards in 10 CFR f 50.92 concerning the issue of no significant hazards consideration. Therefore, in accordance with 10 CFR S 50.91, the following analysis has been performed:
The operation of Nine Mile Point Unit 2, in accordance with the proposed amendment, l will not involve a significant increase in the probability or consequences of an accident t previously evaluated.
Recirculation Flow Rod Block ,
The proposed change would alleviate spurious rod block actuations and nuisance alarms currently associated with operation above 100% of rated core flow by revising i the high recirculation flow rod block instrument setpoints from 108% to 111% of rated l
- flow and 111% to 114% of rated flow for the trip setpoint and allowable value, j respectively. The operational design basis for high flow rod block is to provide an i operator warning signal in response to unexpected high core flow operation. This flow unit upscale trip performs no safety function and the design basis transient and i accident analyses do not take any credit for it. The probability of any accident is not ;
significantly increased by operating at 105% core flow because the APRM flow biased
~
scram and the Rod Block Monitor are " clamped" at their 100% core flow values and j the 100% core flow values are not changing. The effect ofICF operation on MCPR operating limits has been previously evaluated and found not significant and MCPR 3
operating limits for the current cycle are bounding for operation in the ICF region. ;
PCT has been evaluated and substantial margin still exists below the limit of 2200"F.
Loads on reactor internals, containment, and piping systems have been evaluated and stresses remain within design limits. Therefore, there is no increase in the consequences of any accident.
Critical Power Correlation Revising the definition of critical power ratio is a modification to analytical techniques and has no effect on the probability of any transient or accident. The transient analyses will continue to be performed using NRC approved critical power correlations and the methodologies approved for the Core Operating Limits Report. The resultant operating limit MCPRs will still assure that the Safety Limit MCPR is not violated during abnormal operating transients. The Safety Limit MCPR provides assurance that one of the principal barriers to the release of radioactive materials, the fuel cladding, is not degraded. Therefore, revising the definition of critical power ratio will not increase the consequences of any transient or accident.
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_ _ _ _ _ _ _ _ __ . . _ . ~ _ _ _ _ _ . . . _ _ . . __
The ope' ration of Nine Mile Point Unit 2, in accordance with the proposed amendment, will not create the possibility of a new or different kind of accident from any accident ;
previously evaluated. j i
Recirculation Flow Rod Block l
)
Operation with ICF has been evaluated in the USAR and does not provide any new l accident modes. The control rod block instrument setpoints for the reactor coolant :
system recirculation flow upscale condition change from 108% to 111% of rated flow and 111% to 114% of rated flow for the trip setpoint and allowable value, respectively. The operational design basis for high flow rod block is to provide an ;
- operator warning signal in response to unexpected high core flow opcration. The flow
~
unit upscale trip performs no safety function and the design basis transient and :
accident analyses take no credit for it. i Critical Power Correlation ,
The change to the definition of Critical Power Ratio does not alter the intent or !
interpretation of the Technical Specifications. The existing operability requirements for MCPR will remain intact. Thus, the proposed change will not alter the plant i configuration or any mode of operation. i The response to previously evaluated accidents remains within previously assessed limits of temperature and pressure, assuring the environmental qualification of plant equipment is not adversely affected by this proposed amendment. Further, all safety-related systems and components remain within their applicable design limits. Thus, system and component ,
performance is not adversely affected by this change, thereby assuring that the design j capabilities of those systems and components are not challenged in a manner not previously assessed so as to create the possibility of a new or different kind of accident. Therefore, ,
operation of Nine Mile Point Unit 2, in accordance with the proposed change, will not create ;
the possibility of a r.ew or different kind of accident from any previously assessed.
The operation of Nine Mile Point Unit 2, in accordance with the proposed amendment, !
will not involve a significant reduction in a margin of safety. ,
Recirculation Flow Rod Block i'
The results of the reload analysis, as documented in Appendix A of the USAR, assure that the current Technical Specifications, including the limits specified in the Core l Operating Limits Report, are adequate to preclude the violation of any Safety Limits during ICF operation. To maintain adequate operating margin, the control rod block i instrument setpoints for the reactor coolant system recirculation flow upscale condition are being changed from 108% to 111% of rated flow and 111% to 114% of rated flow
{
for the trip setpoint and allowable value, respectively. These setpoints are being increased to maintain an adequate trip setpoint margin allowing for uncertainties and noise, so as not to impact plant high flow operations. i l
t 003414LL Page 9 of 10 l
1 I
. Whereas the analyzed core flow range was previously increased by 5% (from 100% to 105%), the high flow trip setpoint is being increased only 3% (from 108% to 111%); i therefore, the function of the trip as an indication / alarm of unintended high flow l operation is actually enhanced. Further, the RBM upscale setpoint is " clamped" at its 100% core flow value of 110%.
Critical Power Correlation The proposed change revises the definition of Critical Power Ratio to provide a more generic designation for the critical power correlation. The proposed change provides clarification and will not alter the Limiting Condition for Operation or Surveillance Requirements associated with MCPR. The safety and operating limits for MCPR will still be determined in accordance with an approved NRC methodology in accordance with the Core Operating Limits Report.
The proposed changes will not cause existing Technical Specification operational limits or system performance criteria to be exceeded. The proposed changes assure that the system response to postulated accidents remains within accepted limits. Thus, the margins of safety established by the Technical Specifications are not altered by this amendment. Therefore, operation of Nine Mile Point Unit 2, in accordance with the proposed amendment, will not result in a significant reduction in a margin of safety.
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