ML20087G969
| ML20087G969 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 04/07/1995 |
| From: | OMAHA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML20087G963 | List: |
| References | |
| NUDOCS 9504180083 | |
| Download: ML20087G969 (9) | |
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l TECRVICAL SPECIFICATIONS - FIGURES
- TABLE OF CONTENTS l
I PAGE WEDCH FIGURE DESCRIPTION FIGURE FOLLOWS 11 TMLP Safety Limits 4 Pu~~ Operati .............. ... .......13
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-;-2 . .il .";;;; O. ,.22 2 LSSS f;:4 N g ^icxxcr .......... 14-2.! A RCS Pressure-Temperature Limits for Heatup . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 1
21B RCS Pressure-Tempera:ure Limits for Cooldown . . . . . . . . . . . . . . . . . . . . . . . . . 24 I t
2-3 Predicted Radiation Induced NDIT Shift . . . . . . . . . .................... 24 2 11 MIN BAST Level vs Stored BAST Concentration . . . . . . . . .. .. . . . . . 2 19 2 12 Boric Acid Solubility tn Water ...................- .... .. ... 2 2 10 Spent Fuel Pool Region 2 Storage Criteria .........
..................238 28 Flux Peaking Augmentation Factors ....... ...................,.,.,,2.S3
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_g 1 10 '- SAFETY LIMITS 'AND LIMITING SAFETY SYSTEM SunnNGS m
f, t 1.1 Safety Limits - Reactor Core (continued) ~
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((f would cause DNB at a particular core location to the actual heat flux at that location, is indicative of the margin to DNB. The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.18. A DNBR of 1.18 corresponds to a 95% probability at a 95% confidence level that DNB
. will not occur, which is considered an appropriate margin to DNB for all operating.
conditions.m The curves of Figure 1-1 represent the loci of points or reactor thermal power (either y neutron flux instruments or AT instruments), reactor coolant system pressure, and cold 1 c
leg temperature for which the DNBR is 1.18. The area of safe operation is below these '
lines.
The reactor core safety limits are based on radial peaks limited by the CEA insertion .
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% i ection 2.10 and axial shapes within the axial power distribution trip limits in the Cot.R F: re !M nd a total unrodded planar radial peak (F,y T) as specified in the COLR. 'Ihe 3
ermal Margin / Low Pressure trip requirements shall be within the limits provided la
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the COLR. The Thermal Margin / Low Pressure trip is based on an unrodded integrated i
total radial peak (F,T) that is provided in the COLR.
i Flow maldistribution effects for operation under less than full reactor coolant flow have been evaluated via model test.m The flow model data established the maldistribution factors and hot channel inlet temperature for the thermal analyses that were used to- '
establish the safe operating envelopes presented in Figure 1-1. The reactor protective system is designed to prevent any anticipated combination of transient conditions for reactor coolant system temperature, pressure, and thermal power level that would result in a DNBR of less than 1.18.m References (1) USAR, Section 3.6.6 i
(2) USAR, Section 1.4.6
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l-2 Amendment No. 8,32,43,47 70,77,92,i17,125,Mt !
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AXIAL SHAPE INDEX, Y 3
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AxialPowerDistributionLSSS OmahaPublicPowerDistrict figu're 1 for4PumpOperation FortCalhounStation-UnitNo.i
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~'. :120 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 1.3 H Limiting Safety' System Settings, Reactor Protective System
- (Continued),
'(7) Containment High Pressure - A reactor. trip on containment-high ;
pressure is provided to assure that the reactor is shut down-simultaneously with the' initiation of the safety injection system. The setting of this trip'is identical-to that'of the containment high pressure' signal which indicates safety.injec-tion system operation.
(8) Axial Power Distribution - The axial power trip _islprovided'to ensure that excessive axial peaking will not.cause fuel damage.
The Axial Shape Index is determined from the axially 4 -
excore detectors. The set point functions shown_i Figurc ' 2 the COLR ensure that neither a DNBR of less than 1.18 nor a m 1 +
linear heat rate of more. than 22 kW/ft (Ceposited in the fuel)-
will exist as a consequence of axial power maldistributions.
Allowances have been made for instrumentation inaccuracies and ,
uncertainties associated with the excore symmetric offset'-
incore axial peaking relationship. l (9) Steam Generator Differential Pressure - The Asymmetric Steam >
Generator Transient Protection Trip Function'(ASGTPTF) utilizes.
a trip on steam generator differential pressure to ensure that 4
neither a DNBR of less than 1.18 nor a peak _ linear heat ~ rate;of {
more than 22 kW/f t occurs as a result of the. loss of load to one steam generator.
(10) Physics Testing at Low-Power - During physics testing.at' power-levels less than 10-'% of rated' power, the tests may' require that the reactor be critical. For-these tests only'the low-reactor coolant-flow gnd. thermal margin / low pressure trips'may be bypassed below 10-'% of rated power. Written test procedures.
which are approved by the Plant ' Review Comittee will be in effget during these tests. At reactor power levels less than 10*'% of rated power-the. low reactor coolant flow and the thermal margin / low pressure trips are not requir.' to prevent fuel element thermal limits being exceeded. Both of these trips are bypassed using the same bypass switch. The low steam generator pressure trip is not required because the low steam ,
generator pressure will not allow a severe reactor cooldown if a steam line break were to occur during the tests.
References (1) USAR, Section 14.1 1 (2) USAR, Section 7.2.3.3 (3) USAR, Section 7.2.3.2 (4) USAR, Section 3.6.6 (5) USAR, Section 14.6.2.2, 14.6.4 (6) USAR, Section 14.7 *
(7) USAR, Section 7.2.3.1 (8) USAR, Section 3.6 (9) USAR, Section 14.10 Amendment No. 7,37,70,77,g2j ,e 1-9 l a N
TABLE 1-1 RPS LIMITING SAFETY SYSTEM SETTINGS NL_ ' Reactor Trio Trio Setoolnts 1 High Power Level (A) 1107.0% of Rated Power-4-Pump Operation 2 Low Reactor Coolant Flow (B)(F) -,.
4-Pump Operation >
__.95% of 4 Pump Flow 3 Low Steam Generator Water Level 31.2% of Scale (Top of feedwater ring; 4'10" below normal water level) 4 Low Steam Generator Pressure (C) >
_._500 psia 5 High Pressurizer Pressure 12400 psia 6 Thermal Margin / Low Pressure (B)(F) 1750 psia to 2400 psia (depending on the reactor coolant temperature as shown in the Thermal Margin / Low Pressure 4 Pump Operation Figure provided in the COLR) 7 High Containment Pressure (D) 15 psig 8 Axial Power Distribution (E) ,(Figure 1-2)
V 9 Steam Generator Differential Pressure 1135 psid
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(as shown in.the A7ial Power Distribu. tion for + Pump Opera &n Floure orovided in th'e CO _R) 1-10 Amendment No. 7,32,47,7h92,MT 1
U.S. Nuclear Regulatory Commission LIC-95-0083 ATTACHMENT B i
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DISCUSSION, JUSTIFICATION AND NO SIGNIFICANT HAZARDS CONSIDERATION DISCUSSION AND JUSTIFICATION The Omaha Public Power District (0 PPD) proposes to revise the Fort Calhoun Station (FCS) Unit No.1 Technical Specifications (TS) to relocate the Axial Power Distribution (APD) limits (Figure 1-2) to the Core Operating Limits Report (COLR). The COLR is contained in the Technical Data Book of the FCS Operating Manual.
Relocation of cycle specific parameters to the COLR was approved in TS Amendment 141 for FCS. The existing APD limits contained in the TS were revised prior to operating Cycle 8. The APD wts originally requested to be included in the COLR by 0 PPD, but after discussion with NRC staff it was mutually agreed that the APD did not meet the definition of cycle-specific parameters of Generic Letter 88-16, " Removal of Cycle-Specific Parameter Limits from Technical Specifications." This was due to the fact that the limits did not have to be revised very frequently. Recent operations difficulties have identified the need to modify the APD limits on a more frequent basis.
During a recent power reduction to locate a leaking fuel pin at FCS, it was noted that power could not be rapidly reduced, (approximately 10-20% per hour) without violating the Limiting Conditions for Operations (LCO) Axial Shape Index (ASI) restrictions or the APD limits. The loss of rapid shutdown ability is magnified by the use of an extreme low radial leakage fuel management which places low power fuel assemblies in the location of the lead regulating control rod bank. The subsequent reduction in rod worth makes ASI control within the current APD limits very difficult during a rapid power reduction without tripping the plant.
Additional examination determined that the ASI requirements will need to be optimized on a cycle-specific basis in order to avoid this situation. To provide the operators with more ASI margin, the APD figure contained in the TS, and LC0 figures contained in the COLR, will require revisions on a cycle by cycle basis. Operation with overly conservative setpoints, which would allow the use of the APD figure for future operating cycles, would increase the probability of unnecessary reactor trips resulting in a situation which may decrease the overall safety of the plant.
The LC0 figures are currently optimized on a cycle-specific basis for inclusion into the COLR, This would not change. To optimize the APD limits on a cycle-specific basis without this proposed change will require NRC review and approval prior to implementation each cycle. Therefore, the proposed change would relocate the APD figure to the COLR such that NRC review and approval of the specific limits would not be required each cycle.
The safety limits (Minimum DNBR, RCS Pressure, Peak Linear Heat Rate, and 10 CFR 100 releases) for each accident considered on a cycle-specific basis would not change. These safety limits are the same as those accepted by the NRC in previous cycles and will remain unchanged unless prior approval by the NRC is obtained.
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9 The Limiting Safety Settings (LSSSs) are developed to maintain acceptable margins to those limits. For FCS these limiting settings are presented in Section 1 of the TS. One of the key elements of the LSSS, designed specifically to protect against violation of the safety limits on DNBR and fuel centerline melt, is the APD trip. The approved methodology used to define this trip utilizes cycle dependent power distributions as opposed to '
cycle independent power distributions used at many other nuclear plants.
Historically, the limits of the APD have been set very conservatively with respect to the actual limits calculated in the setpoint analysis. The ,
methodology utilized to develop the APD limits is reviewed and approved by the
- NRC, as required by TS 5.9.5.
Other criteria in Section 1 of the TS are developed and implemented to assure that the plant conditions during off-normal situations will not exceed the defined safety limits. These criteria include both cycle-specific reactor protective system setpoints and acceptable fuel design limits. In no case do the cycle-specific parameters contained in Section 1 conflict with or exceed the defined safety limits. By operating within the LCOs and thus maintaining the existing safety limits, FCS is assured that the limits in tha COLR remain within the safety analysis assumptions.
The use of NRC approved reload analysis methodology topical reports, as required by TS 5.9.5, does not permit substantial discretion on the part of OPPD in calculating the LSSS and LC0 values nor does the NRC approved methodology allow substantial engineering jtdgement on the part of the analyst preparing the reload evaluation. Therefore, assurance that the changes to the COLR are consistent with previously reviewed and approved methodology is maintained. ,
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. BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION: j The proposed changes do not involve significant hazards consideration because operation of Fort Calhoun Station Unit No.1 in accordance with these changes would not: 1 (1) Involve a significant increase in the probability or consequences of an accident previously evaluated. '
The proposed change relocates the cycle-specific Axial Power Distribution (APD) limits contained in Figure 1-2 of the Technical Specifications (TS), to the Core Operating Limits Report (COLR). This change is consistent with the NRC recommendations of Generic letter 88-16, and will not modify the methodology used in generating the limits nor the manner in which they are implemented. The methodology used to determine the APD limits is reviewed and approved by the NRC in accordance with TS 5.9.5. The APD limits will continue to be determined by analyzing the same postulated events as previously analyzed. The plant will continue to ogerate within the limits specified in the COLR and will take the same remedial actions if the APD limit is exceeded as required by the current TS. Therefore, the proposed change would not increase the probability or consequences of an accident previously evaluated.
(2) Create the possibility of a new or different kind of accident from any accident previously evaluated.
There will be no physical alterations to the plant configuration, changes to setpoint values, or changes to the implementation of setpoints or limits as a result of this proposed change. The proposed ,
change only relocates the APD figure from the TS to the COLR consistent l with NRC Generic Letter 88-16. Therefore, the proposed change does not l create the possibility of a new or different kind of accident from any previously evaluated.
(3) Involve a significant reduction in a margin of safety.
As indicated above, the implementation of the APD into the COLR, consistent with the guidance of NRC Generic letter 88-16, makes use of i the existing safety analysis methodologies and the resulting limits and setpoints for plant operation. Additionally, the safety analysis acceptance criteria for operations with the proposed change have not ;
changed from that used in the current reload analysis. Therefore, the ;
margin of safety is not reduced due to the relocation of the APD from the TS and implementation in the COLR.
Therefore based on the above considerations, it is OPPD's position that this proposed amendment does not involve significant hazards considerations as defined by 10 CFR 50.92 and the proposed changes will not result in a condition which significantly alters the impact of the Station on the environment. Thus, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(3) and pursuant to 10 CFR 51.22(b) no environmental assessment need be prepared.
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