NLS2014028, License Amendment Request to Move Linear Heat Generation Rate (LHGR) and Single Loop Operation LHGR Limit from Technical Requirements Manual to Technical Specifications

License Amendment Request to Move Linear Heat Generation Rate (LHGR) and Single Loop Operation LHGR Limit from Technical Requirements Manual to Technical Specifications

ML14203A045
Person / Time
Site:	Cooper
Issue date:	07/17/2014
From:	Limpias O Nebraska Public Power District (NPPD)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NLS2014028
Download: ML14203A045 (43)

Text

Nebraska Public Power District "Always there when you need us" 50.90 NLS2014028 July 17, 2014 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001

Subject:

License Amendment Request to Move Linear Heat Generation Rate (LHGR) and Single Loop Operation LHGR Limit from Technical Requirements Manual to Technical Specifications Cooper Nuclear Station Docket No. 50-298, License No. DPR-46

Dear Sir or Madam:

The purpose of this letter is for the Nebraska Public Power District (NPPD) to request an amendment to Facility Operating License DPR-46 in accordance with the provisions of 10 CFR 50.4 and 10 CFR 50.90 to revise the Cooper Nuclear Station (CNS) Technical Specifications (TS). The proposed change would move the Linear Heat Generation Rate (LHGR) and Single Loop Operation LHGR Limit from the Technical Requirements Manual to CNS Technical Specifications.

NPPD requests approval of the proposed amendment by July 17, 2015. Once approved, the amendment will be implemented within 60 days. provides a description of the TS changes, the basis for the amendment, the no significant hazards consideration evaluation pursuant to 10 CFR 50.91(a)(1), and the envirom-nental consideration pursuant to 10 CFR 51.22. Attachunent 2 provides the proposed changes to the current CNS TS in marked up format. Attachment 3 provides the final typed TS pages to be issued with the amendment. Attachment 4 provides conforiming changes to the TS Bases for Nuclear Regulatory Commission (NRC) information. No fornal regulatory commitments are being made by this submittal.

This proposed TS change has been reviewed by the necessary safety review committees (Station Operations Review Committee and Safety Review and Audit Board). Amendments to the CNS Facility Operating License through Amendment 248, issued April 29, 2014, have been incorporated into this request. This request is submitted under oath or affirmation pursuant to 10 CFR 50.30(b).

COOPER NUCLEAR STATION P.O. Box 98 /Brownville, NE 68301-0098 Telephone: (402) 825-3811 /Fax: (402) 825-5211 http://www.nppd.com A -

NLS2014028 Page 2 of 2 By copy of this letter and its attachments, the appropriate State of Nebraska official is notified in accordance with 10 CFR 50.91(b)(1). Copies are also being provided to the NRC Region IV office and the CNS Senior Resident Inspector in accordance with 10 CFR 50.4(b)(1).

Should you have any questions concerning this matter, please contact David Van Der Kamp, Licensing Manager, at (402) 825-2904.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on: 0o /,'7/14-(Date)

ASincry, a .Limpias Vice President-Nucear and Chief Nuclear Officer Jo Attachments: 1. License Amendment Request to Move Linear Heat Generation Rate (LHGR) and Single Loop Operation LHGR Limit from Technical Requirements Manual to Technical Specifications

2. Proposed Technical Specification Revisions (Markup)

3. Proposed Technical Specification Revisions (Re-Typed)

4. Proposed Technical Specification Bases Revisions (Information Only) cc: Regional Administrator w/attachments USNRC - Region IV Cooper Project Manager w/attachlnents USNRC - NRR Project Directorate IV- 1 Senior Resident Inspector w/attachrnents USNRC - CNS Nebraska Health and Human Services w/attachments Department of Regulation and Licensure NPG Distribution w/o attachments CNS Records w/attachments

NLS2014028 Page 1 of 12 Attachment 1 License Amendment Request to Move Linear Heat Generation Rate (LHGR) and Single Loop Operation LHGR Limit from Technical Requirements Manual to Technical Specifications Cooper Nuclear Station, NRC Docket No. 50-298, License No. DPR-46 Revised Pages 1.1-3, 3.4-1, 3.7-14 and 5.0-20 New Page 3.2-4 1.0 Summary Description 2.0 Detailed Description 2.1 Proposed Change 2.2 Need for Change 2.3 Bases Changes 3.0 Technical Evaluation 3.1 System Description 3.2 Updated Safety Analysis Report (USAR) Safety Design Basis 3.3 Current TS Bases Safety Analysis 3.4 Technical Justification of Proposed Changes 3.5 USAR Accident Analysis Impact 3.6 Conclusion 4.0 Regulatory Safety Analysis 4.1 Applicable Regulatory Requirements 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusion 5.0 Environmental Consideration

NLS2014028 Attachnment I Page 2 of 12 1.0

SUMMARY

DESCRIPTION This letter is a request to amend Facility Operating License DPR-46 for Cooper Nuclear Station (CNS). The proposed change would move the Linear Heat Generation Rate (LHGR) and Single Loop Operation LHGR Limit from the Technical Requirements Manual (TRM) to Technical Specifications (TS).

CNS requests approval of this Licensing Amendment Request (LAR) by May 14, 2015.

This change is needed to ensure the TS comply with the requirements of 10 CFR 50.36(c)(2)(ii)(B) Criterion 2. Upon receipt of the approved amendment, CNS will implement the change within 60 days.

2.0 DETAILED DESCRIPTION The following revisions are proposed to TS Sections 1.1, 3.4.1, 3.7.7, and 5.6. This LAR also creates new TS Section 3.2.3.

2.1 Proposed Change The proposed change would revise the CNS Operating License to add the definition for LHGR, add the Limiting Condition for Operation (LCO), Actions and Surveillance Requirements for LHGR, revise the Recirculation Loops Operating LCO, revise the Main Turbine Bypass System LCO and revise the requirements for the Core Operating Limits Report (COLR).

2.2 Need for Change 10 CFR 50.36(c)(2)(ii)(B) Criterion 2 requires the inclusion in TS any process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Prior to cycle 23 the Average Planar Linear Heat Generation Rate (APLHGR) thermal limit in TS 3.2.1 was a composite limit of the Emergency Core Cooling System (ECCS) Loss of Coolant Accident (LOCA) limits and the thermal mechanical LHGR limits used to protect the primary fission product barrier (cladding). This allowed the APLHGR TS to meet the requirements for monitoring the ECCS LOCA limits as well as the thermal mechanical limits of the cladding.

Since the implementation of PANAC I1in cycle 23, the application of the APLHGR limits are based solely on the ECCS criteria. Thermal mechanical limits are monitored independently with LHGR. Since the limits are now separate, both limits must reside in the TS and have a unique LCO for each.

LHGR is presently monitored and controlled per the TRM. This change is needed to allow the TS to comply with the above Code of Federal Regulations which will be satisfied by the addition of TS 3.2.3, LHGR.

NLS2014028 Page 3 of 12 TS 3.4.1, Recirculation Loops Operating, and TS 3.7.7, Main Turbine Bypass System, are being revised to be consistent with the Nuclear Regulatory Commission (NRC) approved methodology from GESTAR II.

2.3 Bases Changes Revised TS Bases are provided in Attachment 4 for NRC information. These Bases revisions will be made as an implementing action pursuant to TS 5.5.10, TS Bases Control Program, following issuance of the amendment. The TS Bases for pages B 3.2-1, B 3.2-2, B 3.2-3, B 3.2-4, B 3.2-5, B 3.4-2, B 3.4-3, B 3.4-4, B 3.4-5, B 3.4-8, B 3.7-31, and B 3.7-32 will be revised to conform to the changes proposed for TS 3.2, TS 3.4, and TS 3.7. In addition, TS Bases pages B 3.2-11, B 3.2-12, B 3.2-13, B 3.2-14, have been created to conform to the changes proposed for TS 3.2.

3.0 TECHNICAL EVALUATION

3.1 System Description CNS is a boiling water reactor (BWR) of General Electric (GE) design BWR4, with a Mark 1 containment. The design basis employed for the thermal and hydraulic characteristics incorporated in the core design, in conjunction with the plant equipment characteristics, nuclear instrumentation, and the Reactor Protection System, is to require that no fuel damage occurs during normal operation or during abnormal operational transients.

For purposes of maintaining adequate thermal margin during normal steady-state operation, the minimum critical power ratio (MCPR) shall not be lower than the limiting values, the APLHGR shall not be greater than the limiting values and the maximum LHGR shall be maintained below the limits for all bundles specified in the COLR. This does not specify the operating power nor does it specify peaking factors; these parameters are determined subject to a number of constraints including the thermal limits noted previously. The core and fuel design bases for steady-state operation, i.e., MCPR, APLHGR, and LHGR limits have been defined to provide sufficient margin between the steady-state operating condition and any fuel damage condition to accommodate uncertainties and to assure that no fuel damage results even during the worst anticipated transient condition at any time in life.

NLS2014028 Page 4 of 12 The transient thermal limits are established such that fuel damage is not expected to occur during the most severe abnormal operating transients.

Fuel damage is defined for design purposes as perforation of the cladding which permits release of fission products. The mechanisms which could cause fuel damage in reactor transients are:

a. Severe overheating of the fuel cladding caused by inadequate cooling.

Fuel damage due to local overheating of the cladding is conservatively defined as the onset of the transition from nucleate to film boiling, although fuel damage is not expected to occur until well into the film boiling regime. If MCPR remains above limiting values, no fuel damage would be calculated to occur as a result of inadequate cooling.

b. Rupture of the fuel cladding due to strain caused by relative expansion of the uranium dioxide pellet and the fuel cladding. A value of 1% plastic strain of Zircaloy cladding is conservatively defined as the one unit below which fuel damage due to overstraining of the fuel cladding is not expected to occur. Available data indicates that the threshold for damage is in excess of this value.

The mechanical overpower is used to evaluate the potential for overstraining of the cladding. The incremental cladding strain during a transient is proportional to the change in fuel volume average temperature, which is proportional to the change in either the fuel rod linear power or the fuel rod surface heat flux at a particular axial location or cross-section of the fuel rod.

Limits are determined such that the fuel cladding is not overstrained due to mechanical overpower.

c. Fuel rod failure can be caused by fuel centerline melting. Thermal overpower is used to evaluate the potential for the fuel entering the molten state at the fuel centerline. Temperature at the fuel centerline is proportional to either the fuel rod linear power or the fuel rod surface heat flux, so the magnitude of these quantities reached during the transients are the parameters of interest.

Limits are determined to protect the fuel centerline temperature from thermal overpower.

The transient limits used in core design require that the peak LHGR remains below that which will cause fuel damage during the worst anticipated transient, that the resulting MCPR does not decrease below the safety limit, and that the thermal-mechanical limits, as defined above, are met while taking into account

NLS2014028 Attachment I Page 5 of 12 special local effects such as control blade history. Demonstration that these limits are not exceeded is sufficient to conclude that no fuel damage occurs. It should be noted also that the steady-state operating limits have been established to assure that sufficient margin exists between the steady-state operating condition and any fuel damage condition to accommodate the worst anticipated transient without experiencing fuel damage at any time in life. MCPR and APLHGR limitations are contained in the TS. LHGR is monitored and its requirements are currently located in the TRM.

3.2 Updated Safety Analysis Report (USAR) Safety Design Basis The safety design basis of the core thermal and hydraulic design per USAR Section III is as follows:

The thermal hydraulic design of the core shall establish limits for use in setting devices of the nuclear safety systems so that no fuel damage occurs as a result of abnormal operational transients.

The thermal hydraulic design of the core shall establish a thermal hydraulic safety limit for use in evaluating the safety margin relating the consequences of fuel barrier failure to public safety.

3.3 Current TS Bases Safety Analysis The APLHGR is a measure of the average LHGR of all the fuel rods in a fuel assembly at any axial location. Limits on the APLHGR are specified to ensure that the fuel design limits are not exceeded during abnormal operational transients and that the peak cladding temperature (PCT) during the postulated design basis LOCA does not exceed the limits specified in 10 CFR 50.46.

APLHGR limits are equivalent to the LHGR limit for each fuel rod divided by the local peaking factor of the fuel assembly. APLHGR limits are developed as a function of exposure and the various operating core flow and power states to ensure adherence to fuel design limits during the limiting abnormal operational transients. Flow dependent APLHGR limits are determined using the three dimensional BWR simulator code to analyze slow flow runout transients. The flow dependent multiplier, MAPFACf, is dependent on the maximum core flow runout capability. The maximum runout flow is dependent on the existing setting of the core flow limiter in the Recirculation Flow Control System. Based on analyses of limiting plant transients (other than core flow increases) over a range of power and flow conditions, power dependent multipliers, MAPFACp, are also generated. Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which turbine stop valve closure and turbine control valve fast closure scram trips are bypassed, both high and low core flow MAPFACp limits are provided for operation at power levels between 25% Rated

NLS2014028 Attachment I Page 6 of 12 Thermal Power (RTP) and the previously mentioned bypass power level. The exposure dependent APLHGR limits are reduced by MAPFACp and MAPFACf, at various operating conditions to ensure that all fuel design criteria are met for normal operation and abnormal operational transients.

LOCA analyses are then performed to ensure that the above determined APLHGR limits are adequate to meet the PCT and maximum oxidation limits of 10 CFR 50.46. The analysis is performed using calculational models that are consistent with the requirements of 10 CFR 50, Appendix K. The PCT following a postulated LOCA is 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. The APLHGR limits specified are equivalent to the LHGR of the highest powered fuel rod assumed in the LOCA analysis divided by its local peaking factor. A conservative multiplier is applied to the LHGR assumed in the LOCA analysis to account for the uncertainty associated with the measurement of the APLHGR.

For single recirculation loop operation, the MAPFAC multiplier is contained in the COLR. This maximum limit is due to the conservative analysis assumption of an earlier departure from nucleate boiling with one recirculation loop available, resulting in a more severe cladding heatup during a LOCA.

The APLHGR satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii)(B).

MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The MCPR Safety Limit (SL) is set such that 99.9% of the fuel rods avoid boiling transition if the limit is not violated. The operating limit MCPR is established to ensure that no fuel damage results during abnormal operational transients. Although fuel damage does not necessarily occur if a fuel rod actually experienced boiling transition, the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion.

The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations have been developed to predict critical bundle power (i.e., the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling). Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.

To ensure that the MCPR SL is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR). The types of

NLS2014028 Attachinent 1 Page 7 of 12 transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest change in CPR (ACPR). When the largest ACPR is added to the MCPR SL, the required operating limit MCPR is obtained.

The MCPR operating limits derived from the transient analysis are dependent on the operating core flow and power state (MCPRf and MCPRp, respectively) to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency. Flow dependent MCPR limits are determined by steady-state thermal hydraulic methods with key physics response inputs benchmarked using the three dimensional BWR simulator code to analyze slow flow runout transients. The operating limit is dependent on the maximum core flow limiter setting in the Recirculation Flow Control System.

Power dependent MCPR limits (MCPRp) are determined mainly by the one dimensional transient code. Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scrams are bypassed, high and low flow MCPRP operating limits are provided for operating between 25% RTP and the previously mentioned bypass power level.

The MCPR satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii)(B).

The specification assures that the LHGR in any rod is less than the design linear heat generation if fuel pellet densification is postulated. The LHGR as a function of core height shall be checked daily during reactor operation at > 25% power to determine if fuel burnup, or control rod movement has caused changes in power distribution. For LHGR to be a limiting value below 25% rated thermal power, the Maximum Total Peaking Factor would have to be greater than 10 which is precluded by a considerable margin when employing any permissible control rod pattern. Pellet densification power spiking in GE fuel has been accounted for in the safety analysis; thus no adjustment to the LHGR limit for densification effects is required.

The LHGR shall be checked daily to determine if fuel burnup, or control rod movement has caused changes in power distribution. Since changes due to burnup are slow, and control rods are moved infrequently, a daily check of power distribution is adequate.

A plant specific LOCA analysis has been performed assuming only one operating recirculation loop. In the event of a LOCA caused by a pipe break in the operating recirculation loop, the ECCS response will provide adequate core cooling, provided the LHGR requirements are modified accordingly.

The LHGR limits for single loop operation are specified in the COLR.

NLS2014028 Page 8 of 12 3.4 Technical Justification of Proposed Changes Prior to cycle 23 the APLHGR thermal limit in TS 3.2.1 was a composite limit of the ECCS LOCA limits and the thermal mechanical LHGR limits used to protect the primary fission product barrier (cladding). This allowed the APLHGR TS to meet the requirements for monitoring the ECCS LOCA limits as well as the thermal mechanical limits of the cladding.

Since the implementation of PANAC 1I in cycle 23, the application of the APLHGR limits are based solely on the ECCS criteria. Thermal mechanical limits are monitored independently with LHGR. Since the limits are now separate, both limits must reside in the TS and have a unique LCO for each. This change is needed to allow the Technical Specifications to comply with Code of Federal Regulations 10 CFR 50.36(c)(2)(ii)(B) Criterion 2, which will be satisfied by the addition of TS 3.2.3, LHGR.

The addition of LHGR LCO to the single loop operation section of TS 3.4.1, Recirculation Loops Operating, is consistent with the method allowed in GESTAR II and as is currently specified in the TRM under section T 3.2.1, Linear Heat Generation Rate.

The addition of LCO 3.2.1, APLHGR, and LCO 3.2.3, LHGR, to the Main Turbine Bypass System section TS 3.7.7 is consistent with GESTAR II. The off-rated LHGR and APLHGR limit multipliers are calculated for when all bypass valves are in service and also when one bypass valve is out of service. The results from these two sets of multipliers have always been the same, thus only the results from one bypass valve out of service are used and reported in the COLR.

3.5 USAR Accident Analysis Impact Revising TS 1.1, TS 3.2, TS 3.4, TS 3.7 and TS 5.6 to add LHGR limits has no impact on the USAR accident analysis and ensures assumptions in the USAR accident analyses remain valid.

3.6 Conclusion In summary, the proposed change is technically sound and continues to maintain the same level of safety as the current licensing basis.

NLS2014028 Page 9 of 12 4.0 REGULATORY SAFETY ANALYSIS 4.1 Applicable Regulatory Requirements 4.1.1 10 CFR 50.36, Technical Specifications 10 CFR 50.36(b) requires that each license authorizing operation of a utilization facility will include technical specifications. 10 CFR 50.36(c) specifies the categories that are to be included in Technical Specifications (TS). 10 CFR 50.36(c)(2) identifies Limiting Conditions for Operation (LCOs) as one of the categories to be included in TS.

10 CFR 50.36(c)(2)(ii)(B) Criterion 2 requires a technical specification limiting condition for operation of a nuclear reactor to be established for a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Moving the Linear Heat Generation Rate (LHGR) and Single Loop Operation LHGR Limit from the Technical Requirements Manual (TRM) to TS will ensure that Cooper Nuclear Station (CNS) continues to meet this regulation with the proposed change.

4.1.2 NUREG-1433, Volume 2, Revision 4.0; Section B 3.2 Power Distribution Limits The LHGR is a basic assumption in the fuel design analysis. If any LHGR exceeds its required limit, the assumption is not met. A 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is normally sufficient to restore the LHGR to within its limits, based on the low probability of a transient or Design Basis Accident occurring while the LHGR is out of specification.

If the LHGR cannot be restored to within its required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, thermal power is reduced to less than 25% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, which allows for power reduction in an orderly manner, without challenging plant systems.

4.2 Precedent No applicable precedents have been identified for this proposed license amendment. Nebraska Public Power District (NPPD) requests that this amendment request be reviewed on its own merit.

NLS2014028 Page 10 of 12 4.3 No Significant Hazards Consideration 10 CFR 50.91 (a)(1) requires that licensee requests for operating license amendments be accompanied by an evaluation of no significant hazard posed by issuance of the amendment. Nebraska Public Power District (NPPD) has evaluated this proposed amendment with respect to the criteria given in 10 CFR 50.92(c). The following is the evaluation required by 10 CFR 50.91(a)(1).

NPPD is requesting an amendment of the operating license for the Cooper Nuclear Station (CNS) to move the Linear Heat Generation Rate (LHGR) and Single Loop Operation LHGR Limit from the Technical Requirements Manual (TRM) to Technical Specifications (TS).

10 CFR 50.36(c)(2)(ii)(B) Criterion 2 requires a technical specification limiting condition for operation of a nuclear reactor to be established for a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Moving the LHGR and Single Loop Operation LHGR Limit from the TRM to Technical Specifications TS will ensure that CNS continues to meet this regulation with the proposed change.

1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

LHGR limits have been defined to provide sufficient margin between the steady-state operating condition and any fuel damage condition to accommodate uncertainties and to assure that no fuel damage results even during the worst anticipated transient condition at any time.

The proposed change to move the LHGR limits from the TRM to TS, including the change to TS 3.4.1, Recirculation Loops Operating, and TS 3.7.7, Main Turbine Bypass System, does not modify the limits, change assumptions for the accident analysis, or change operation of the station.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

NLS2014028 Attachment I Page 11 of 12 Response: No.

The proposed change does not modify the limits, change assumptions for the accident analysis, or change operation of the station.

The proposed change does move LHGR limits that have been defined to provide sufficient margin between the steady-state operating condition and any fuel damage condition to accommodate uncertainties and to assure that no fuel damage results even during the worst anticipated transient condition at any time from the TRM to TS.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Do the proposed changes involve a significant reduction in a margin of safety?

Response: No.

The proposed change to move the LHGR limits from the TRM to TS, including the change to TS 3.4.1 and TS 3.7.7, does not modify the limits, change assumptions for the accident analysis, or change operation of the station.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the responses to the above questions, NPPD concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

10 CFR 51.22 provides criteria for, and identification of, licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment or

NLS2014028 Attachment I Page 12 of 12 environmental impact statement. 10 CFR 51.22(c)(9) identifies an amendment to an operating license for a reactor which changes a requirement as a categorical exclusion provided that operation of the facility in accordance with the proposed amendment would not: (i) involve a significant hazards consideration, (ii) result in a significant change in the types or significant increase in the amount of any effluents that may be released off site, or (iii) result in a significant increase in individual or cumulative occupational radiation exposure.

CNS review has determined that the proposed amendment, which would add a requirement, does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that might be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

NLS2014028 Page 1 of 6 Attachment 2 Proposed Technical Specification Revisions (Markup)

Cooper Nuclear Station, Docket No. 50-298, DPR-46 Revised Pages 1.1-3, 3.4-1, 3.7-14 and 5.0-20 New Page 3.2-4

Definitions 1.1 1.1 Definitions DOSE EQUIVALENT 1-131 1-133,1-134, and 1-135 actually present. The DOSE (continued) EQUIVALENT 1-131 concentration Is calculated as follows:.

POSE EQUIVALENT 1-131 = (1-131) + 0.0060 (1-132) + 0.17 (1-133) + 0.0010 (1.134) + 0.029 (1-135). The dose qonversion factors used for this calculation are those listed in Federal Guidance Report (FGR) 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989.

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or

2. LEAKAGE into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;

b. Unidentified LEAKAGE Ail LEAKAGE into the drywell that is not identified LEAKAGE:

The LHGR shall be the heat generation rate per unit length of fuel Total LEAKAGE rod. It is the integral of the heat flux over the heat transfer area associated CC. Sum of the identified and unidentified LEAKAGE:

with the unit length.

d. Pressure Boundary LEAKAGE LINEAR HEAT GENERATION RATE (LHGR) LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test TEST of all logic components required for OPERABILITY of a logic circuit, (continued)

Cooper 1.1-3 Amendment 234

LHGR 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)

LCO 3.2.3 All LHGRs shall be less than or equal to the limits specified in the COLR.

APPLICABILITY: THERMAL POWER a 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any LHGR not within limits. A.1 Restore LHGR(s) to within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> limits.

B. Required Action and B.1 Reduce THERMAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion Time to < 25% RTP.

not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify all LHGRs are less than or equal to the limits Once within 12 specified in the COLR. hours after > 25%

RTP AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter Cooper 3.2-4 Amendment No.

Recirculation Loops Operating 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched flows shall be in operation outside of the Stability Exclusion Region of the power/flow map specified in the COLR.

OR One recirculation loop shall be in operation outside of the Stability Exclusion Region of the power/flow map specified in the COLR with the following limits applied when the associated LCO is applicable:

c. LCO 3.2.3, "LINEAR HEAT GENERATION a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE RATE (LHGR)," single (APLHGR)," single loop operation limits specified in the loop operation limits COLR; specified in the COLR; LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," single and loop operation limits specified in the COLR; ed LCO 3.3.1.1, "Reactor Protection System (RPS)

d. Instrumentation," Function 2.b (Average Power Range Monitor Neutron Flux-High (Flow Biased)), Allowable Value of Table 3.3.1.1-1 is reset for single loop

_2 operation.

APPLICABILITY: MODES I and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or two A.1 Initiate action to Immediately recirculation loops in exit the Stability operation with core Exclusion Region.

flow as a function of core THERMAL POWER in the Stability Exclusion Region of the power/flow map.

(continued)

)_j Cooper 3.4-1 Amendment No. +M

Main Turbine Bypass System 3.7.7 3.7 PLANT SYSTEMS K-3.7.7 The Main Turbine Bypass System LCO 3.7.7 The Main Turbine Bypass System shall be OPERABLE.

0_LCO 3.2.1, "AVERAGE PLANAR LINEARI HEAT GENERATION RATE (APLHGR)," I LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR), limits for one inoperable main turbine bypass valve, as spelified in the COLR, are made applicable.

and LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)."

APPLICABILITY: THERMAL POWER > 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Satisfy the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> LCO not met. requirements of the LCO.

-1)

B. Required Action and B.1 Reduce THERMAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion to < 25% RTP.

Time not met.

OR Two or more main turbine bypass valves inoperable.

KU Cooper 3.7-14 Amendment No. 4--8

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5 63 Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit shall be sub-mitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODAM and the Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.

564 (Deleted) 5.6,5 CORE OPERATING LIMITS REPORT (COLR)

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

The Average Planar Linear Heat Generation Rates for SpecificationS 2. 1.*--- and 3.7.7.

2 The Minimum Critical Power Ratio for Specifications 3.2.2 and 3.7.7

3. The Linear Heat Generation Rates for Specifications 3.2.3 and 3.7.7.

he three Rod Block Monitor Upscale Allowable Values for Specification 3.3.2.1.

[5]_.*\4- The power/flow map defining the Stability Exclusion Region for Specification 3.4.1.

b The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1 NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel" (Revision specified in the COLR).

(continued)

Cooper 5.0-20 Amendment No. 2-34 1

NLS2014028 Page 1 of 6 Attachment 3 Proposed Technical Specification Revisions (Re-Typed)

Cooper Nuclear Station, Docket No. 50-298, DPR-46 Revised Pages 1.1-3, 3.4-1, 3.7-14 and 5.0-20 New Page 3.2-4

Definitions 1.1 1.1 Definitions DOSE EQUIVALENT 1-131 1-133, 1-134, and 1-135 actually present. The DOSE (continued) EQUIVALENT 1-131 concentration is calculated as follows:

DOSE EQUIVALENT 1-131 = (1-131) + 0.0060 (1-132) + 0.17 (1-133) + 0.0010 (1-134) + 0.029 (1-135). The dose conversion factors used for this calculation are those listed in Federal Guidance Report (FGR) 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989.

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or

2. LEAKAGE into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;

b. Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE;

c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE;

d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per unit length of RATE (LGHR) fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all TEST logic components required for OPERABILITY of a logic circuit, (continued)

Cooper 1.1-3 Amendment No.

LHGR 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)

LCO 3.2.3 All LHGRs shall be less than or equal to the limits specified in the COLR.

APPLICABILITY: THERMAL POWER ->25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any LHGR not within limits. A.1 Restore LHGR(s) to within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> limits.

B. Required Action and B.1 Reduce THERMAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion Time to < 25% RTP.

not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify all LHGRs are less than or equal to the limits Once within 12 specified in the COLR. hours after > 25%

RTP AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter Cooper 3.2-4 Amendment No.

Recirculation Loops Operating 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched flows shall be in operation outside of the Stability Exclusion Region of the power/flow map specified in the COLR.

OR One recirculation loop shall be in operation outside of the Stability Exclusion Region of the power/flow map specified in the COLR with the following limits applied when the associated LCO is applicable:

a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits specified in the COLR;

b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," single loop operation limits specified in the COLR;

c. LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," single loop operation limits specified in the COLR; and

d. LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation,"

Function 2.b (Average Power Range Monitor Neutron Flux - High (Flow Biased)), Allowable Value of Table 3.3.1.1-1 is reset for single loop operation.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or two recirculation A.1 Initiate action to exit the Immediately loops in operation with core Stability Exclusion Region.

flow as a function of core THERMAL POWER in the Stability Exclusion Region of the power/flow map.

(continued)

Cooper 3.4-1 Amendment No.

Main Turbine Bypass System 3.7.7 3.7 PLANT SYSTEMS 3.7.7 The Main Turbine Bypass System LCO 3.7.7 The Main Turbine Bypass System shall be OPERABLE.

OR LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR),"

and LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," limits for one inoperable main turbine bypass valve, as specified in the COLR, are made applicable.

APPLICABILITY: THERMAL POWER ->25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO A.1 Satisfy the requirements of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> not met. the LCO.

B. Required Action and B.1 Reduce THERMAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion Time to < 25% RTP.

not met.

OR Two or more main turbine bypass valves inoperable.

Cooper 3.7-14 Amendment No.

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.3 Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODAM and the Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.

5.6.4 (Deleted) 5.6.5 Core Operatinq Limits Report (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1. The Average Planar Linear Heat Generation Rates for Specifications 3.2.1 and 3.7.7.

2. The Minimum Critical Power Ratio for Specifications 3.2.2 and 3.7.7.

3. The Linear Heat Generation Rates for Specifications 3.2.3 and 3.7.7.

4. The three Rod Block Monitor Upscale Allowable Values for Specification 3.3.2.1.

5. The power/flow map defining the Stability Exclusion Region for Specification 3.4.1.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. NEDE-2401 1-P-A, "General Electric Standard Application for Reactor Fuel" (Revision specified in the COLR).

(continued)

Cooper 5.0-20 Amendment No.

NLS2014028 Page 1 of 17 Attachment 4 Proposed Technical Specification Bases Revisions (Information Only)

Cooper Nuclear Station, Docket No. 50-298, DPR-46 Revised Technical Specification Bases Pages B 3.2-1, B 3.2-2, B 3.2-3, B 3.2-4, B 3.2-5, B 3.4-2, B 3.4-3, B 3.4-4, B 3.4-5, B 3.4-8, B 3.7-31 and B 3.7-32 New Technical Specification Bases Pages B 3.2-11, B 3.2-12, B 3.2-13, B 3.2-14

APLHGR B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)

BASES BACKGROUND The APLHGR is a measure of the average LHGR of all the fuel rods in a fuel assembly at any axial location. Limits on the APLHGR are specified to ensure that the fu*l design limit. identified in Referane. I .ar.nt

..... d@ad duing bnorm..al po.....on*, tR..ncicnt and tha the peak cladding temperature (PCT) during the postulated design basis loss of coolant accident (LOCA) does not exceed the limits specified in 10 CFR 50.46.

APPLICABLE The anal';tie ...- he ds and assumptianc weed in evaluatirnq SAFETY ANALYSES the fuol d 4 limiptoarc prusentedin Refaruatin 1 and 2. The analytical method~~ assumptions used in evaluating Dersign Basis Aeeidentsa (PBAs,, dnconormal operational transients, and normal operation that determine the APLHGR limits are presented in References 1, 2 3, 4, 5, e6"--'-and<- 10 ,---10 CFR 51 0.461 APLHCP Iimit~ ~rr ýi':nlent to~ the IH NR lirrit fen---r-I-rfnlridAIpd RAc I LOCA ~~by tlho Iccal peaking f66tcr ef the fuel acob.APGRimtar developed as a function of exposure and the various operg core flow a.power states to ensure adherence to ,e-, during the limitir* abnormal . p... ti.nal t.an.i.An. (Refs. 5 and 6). Fleepeine 14L--R lImFite are detorrnincd using the thrac dimensienol BWR Giulto cde (Ref. 7) to9 analyze 6l9W floW runcubt trancsients. The fie~

depondont multiplier, MAPF^AC,, is deopendent on the maximum,, .er. flew rUn.ut Gapability. ThFmim At flew ic dependent en the cxislin caetting of the care flaw limitar in the Racirculetian Flew Central Systemn.

B~aced. an nalyrses of limiting plant transiants (ether th98n eera fla incFeroacoc) o9or a rango Of power and flaW sonditione. pawar dependent multipliere, MAPF=AC,, are alsc Cooper B 3.2-1

APLHGR B 3.2.1 BASES APPLICABLE ge...ated. Due trthe scncitivity of the t.aint r-spen..

SAFETY ANALYSES to initial r. fow .. y.l. at powcr levels I I lcw thes.et (continued) which turine stop valve closure and t.rbin.

, ontrol valve fast ileauire LCrA mtripsaro bypassed, both high and loW core flow MAPFtAbov lidnite ame provided for oporation at power lovels between 265A RT-P and the provioucly montioned bypass power level. The cxpeazur dependent APLHGR limits are reduced by metAPFACand MAPFACmat iations operating conditions to onsuro that all fuel design editeuia ac maet for noemal tperatieon anid abnormal perhatienal transients. 0Cmpt0 A

diccuccixn of the analysis code is prrvided in Reforenee 8.

LOCA analyses are #ie~iperformed to ensure that the above determined APLHGR limits are adequate to meet the POT and maximum oxidation limits of 10 CFR 50.46. The analysis is performed using calculational models that are consistent with the requirements of 10 CFR 50, Appendix K. A complete discussion of the analysis code is provided in References 9 and 10. The PCT following a postulated LOCA is 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. The APLHGR limits specified are equivalent to the LHGR of the highest powered fuel rod assumed in The exposure dependent the LOCA analysis divided by its local peaking factor. A.senoor-=tivc APLHGR limits are multiplier is applied to the LHGR assumed in the LOCGA analyci t ri*rl llrlr l

hK IAPPA, =lnri account for the uncertainty associated with the mneaur~ement of thoe MAPFACp at various >

operating conditions to For single recirculation loop operation, the MAPFAC multiplier is ensure that all fuel design contained in the COLR. This maximum limit is due to the conservative criteria are met for normal analysis assumption of an earlier departure from nucleate boiling with operation and LOCA. one recirculation loop available, resulting in a more severe cladding heatup during a LOCA (Refs. 5 and 9).  :

The APLHGR satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii) (Ref. 12).

LCO The APLHGR limits (for each type of fuel as a function of average planar exposure) specified in the COLR are the result of the fuelde.sgn, D*-,

andtFRsie1continuedr (continued)

Cooper B 3.2-2 Coopr J B3.2- 0,4~999 I

APLHGR B 3.2.1 BASES LCO two recirculation loops operating, the limit is determined (continued) by multiplying the smaller of the MAPFAC, and MAPFACf factors times the exposure dependent APLHGR limits. With only one recirculation loop in operation, in conformance with the requirements of LCO 3.4.1, "Recirculation Loops Operating," the limit is determined by multiplying the exposure dependent APLHGR limit by the one recirculation loop lI operation multiplier contained in the COLR.

APPLICABILITY The APLHGR limits are primarily derived from fuel design evaluations and LOCA and tFrsiei.,t analyses that are assumed to occur at high power levels. Design calculations (Ref. 6) and operating experience have shown that as power is reduced, the margin to the required APLHGR limits increases. This trend continues down to the power range of 5% to 15% RTP when entry into MODE 2 occurs. When in MODE 2, the intermediate range monitor scram function provides prompt scram initiation during any significant transient, thereby effectively removing any APLHGR limit compliance concern in MODE 2. Therefore, at THERMAL POWER levels

• 25% RTP, the reactor is operating with substantial margin to the APLHGR limits; thus, this LCO is not required.

ACTIONS A.1I LOC1 If any APLHGR exce d limits, an assumption regarding an initial condition of the nalyses may not be met.

Therefore, prompt action should be taken to restore the APLHGR(s) to within the required limits such that the plant operates within analyzed LOCA conditions and within design limits of the fuel rods. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is sufficient to restore the APLHGR(s) to within its F'its and is acceptable based on the low probability of a troncicnt Or DB i occurring simultaneously with the APLHGR out of specification.

6-1 If the APLHGR cannot be restored to within its required limits within the associated Completion Time, the plant must (continued)

Cooper B 3.2-3 Cooperun B0 3.2- 1~n~Q

APLHGR B 3.2.1 BASES ACTIONS B.1 (continued) be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 25% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.1.1 REQUIREMENTS APLHGRs are required to be initially calculated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER is > 25% RTP and then every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.

They are compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution during normal operation. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance after THERMAL POWER

> 25% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels.

REFERENCES 1. NEDE-2401 1-P-A, "General Electric Standard Application for Reactor Fuel," (Revision specified in the COLR).

2. USAR, Sictn .Deleted.

3. USAR,Section VI.

4. USAR,Section XIV.

5. NEDO-24258, "Cooper Nuclear Station Single-Loop Operation,"

May 1980.

6. NEDC-32914P, "Maximum Extended Load Line Limit and Increased Core Flow for Cooper Nuclear Station," Revision 0, January 2000.

7. NEDO 30130 A, "Steady State OMuclear Methods," May 1985.

(oiDeleted.I (continued)

Cooper B 3.2-4 Co*fper B2, 2099-2G.§

APLHGR B 3.2.1 BASES Deleted.

REFERENCES 8. NED, 24154, "Qualifi,,, i^n ^f the ,n, ,imens.,nal Core (continued) T-r-n.i.nt Modeol

,iingfoer Wat^r Raet,*

,. ," GrOtbe 4197-*

9. NEDC-32687P, Revision 1, "Cooper Nuclear Station SAFERIGESTR-LOCA Loss-of-Coolant Accident Analysis," March 1997.

10. NEDE-23785-1-PA, "The GESTR-LOCA and SAFER Models for the Evaluation of. Loss-of-Coolant Accident," Volume Ill, Revision 1, October 1984.

11. Deleted.

12. 10 CFR 50.36(c)(2)(ii).

Cooper B 3.2-5 Cooperun B0 325ne*49Q 1

LHGR B 3.2.3 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)

BASES BACKGROUND The LHGR is a measure of the heat generation rate of a fuel rod in a fuel assembly at any axial location. Limits on LHGR are specified to ensure that fuel design limits are not exceeded anywhere in the core during normal operation, including anticipated operational occurrences (AOOs),

and to ensure that the peak clad temperature (PCT) during postulated design basis loss of coolant accident (LOCA) does not exceed the limits specified in 10 CFR 50.46. Exceeding the LHGR limit could potentially result in fuel damage and subsequent release of radioactive materials.

Fuel design limits are specified to ensure that fuel system damage, fuel rod failure, or inability to cool the fuel does not occur during the anticipated operating conditions identified in References 1 and 2.

APPLICABLE SAFETY ANALYSIS The analytical methods and assumptions used in evaluating the fuel system design limits are presented in Reference 1. The analytical methods and assumptions used in evaluating AOOs and normal operation that determine the LHGR limits are presented in Reference 2.

The fuel assembly is designed to ensure (in conjunction with the core nuclear and thermal hydraulic design, plant equipment, instrumentation, and protection system) that fuel damage will not result in the release of radioactive materials in excess of the guidelines of 10 CFR, Parts 20, 50, and 100. The mechanisms that could cause fuel damage during operational transients and that are considered in fuel evaluations are:

a. Rupture of the fuel rod and cladding caused by strain from the relative pellet and expansion of the U0 2 .

b. Severe overheating of the fuel rod cladding caused by inadequate cooling.

A value of 1% plastic strain of the fuel cladding has been defined as the limit below which fuel damage caused by overstraining of the fuel cladding is not expected to occur (Ref. 3).

Fuel design evaluations have been performed and demonstrate that the 1% fuel cladding plastic strain design limit is not exceeded during continuous operation with LHGRs up to the operating limit specified in the COLR. The analysis also includes allowances for short-term transient operation above the operating limit to account for AOOs, plus an allowance for densification power spiking.

Cooper B 3.2-11 xx/xx/xx I

LHGR B 3.2.3 BASES APPLICABLE SAFETY ANALYSIS (continued)

LHGR limits are developed as a function of exposure and the various operating core flow and power states to ensure adherence to fuel design limits during the limiting AQOs (Refs. 4 and 5). Flow dependent LHGR limits are determined (Ref. 5) using the three dimensional BWR simulator code (Ref. 6) to analyze slow flow runout transients. The flow dependent multiplier, LHGRFACf, is dependent on the maximum core flow runout capability. The maximum runout flow is dependent on the existing setting of the core flow limiter in the Recirculation Flow Control System.

Based on analyses of limiting plant transients (other than core flow increases) over a range of power and flow conditions, power dependent multipliers, LHGRFACp, also are generated. Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which turbine stop valve closure and turbine control valve fast closure scram trips are bypassed, both high and low core flow LHGRFACp limits are provided for operation at power levels between 25% RTP and the previously mentioned bypass power level.

The exposure dependent LHGR limits are reduced by LHGRFACp and LHGRFACf at various operating conditions to ensure that all fuel design criteria are met for normal operation and AQOs. A complete discussion of the analysis code is provided in Reference 7.

For single recirculation loop operation, the LHGRFAC multiplier is limited to a maximum of 0.75 (Refs. 8 and 10). This maximum limit is due to the conservative analysis assumption of an earlier departure from nucleate boiling with one recirculation loop available, resulting in a more severe cladding heatup during a LOCA.

The LHGR satisfies Criterion 2 of the NRC Policy Statement (Ref. 9).

LCO The LHGR is a basic assumption in the fuel design analysis. The fuel has been designed to operate at rated core power with sufficient design margin to the LHGR limit calculated to cause a 1% fuel cladding plastic strain. For two recirculation loops operating, the limit is determined by multiplying the smaller of the LHGRFACf and LHGRFACp factors times the exposure dependent LHGR limits. With only one recirculation loop in operation, in conformance with the requirements of LCO 3.4.1, "Recirculation Loops Operating," the limit is determined by multiplying the exposure dependent LHGR limit by the smaller of either LHGRFACf, LHGRFACp, and 0.75, where 0.75 has been determined by a specific single recirculation loop analysis (Refs. 8 and 10).

Cooper B 3.2-12 xx/xxlxx I

LHGR B 3.2.3 BASES (continued)

APPLICABILITY The LHGR limits are derived from fuel design analysis that is limiting at high power level conditions. At core thermal power levels < 25% RTP, the reactor is operating with a substantial margin to the LHGR limits and, therefore, the Specification is only required when the reactor is operating at > 25% RTP.

ACTIONS A.1 If any LHGR exceeds its required limit, an assumption regarding an initial condition of the fuel design analysis is not met. Therefore, prompt action should be taken to restore the LHGR(s) to within its required limits such that the plant is operating within analyzed conditions and within the design limits of the fuel rods. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is normally sufficient to restore the LHGR(s) to within its limits and is acceptable based on the low probability of a transient or LOCA occurring simultaneously with the LHGR out of specification.

B. I If the LHGR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER is reduced to < 25% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems.

SURVEILLANCE REQUIREMENTS SR 3.2.3.1 The LHGR is required to be initially calculated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER is Z 25% RTP and then every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter. It is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on both engineering judgment and recognition of the slow changes in power distribution during normal operation. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance after THERMAL POWER > 25% RTP is achieved is acceptable given the large inherent margin to operating limits at lower power levels.

Cooper B 3.2-13 xxlxxlxx I

LHGR B 3.2.3 BASES (continued)

REFERENCES 1. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel" version specified in COLR.

2. Current Cycle COLR.

3. NUREG-0800,Section II.A.2(g), Revision 2, July 1981.

4. NEDC-32914P, "Maximum Extended Load Line Limit and Increased Core Flow for Cooper Nuclear Station," Revision 0, January 2000.

5. NEDC-31892P, "Extended Load Line Limit ARTS Improvement Program Analysis for Cooper Nuclear Station Cycle 14,"

Revision 1.

6. NRC approval of "Amendment 26 to GE Licensing Topical Report NEDE-2401 1-P-A, "GESTAR II"- Implementing Improved GE Steady-State Methods (TAC No. MA6481 )," November 10, 1999.

7. NEDO-24154-A, "Qualification of the One-Dimensional Core Transient Model (ODYN) for Boiling Water Reactors," August 1986, and NEDE-24154-P-A, Supplement 1, Volume 4, Revision 1, February 2000.

8. NEDO-24258, "Cooper Nuclear Station Single Loop Operation,"

May 1980.

9. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

10. NEDC-32687P, Revision 1, "Cooper Nuclear Station SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis,"

March 1997.

Cooper B 3.2-14 xx/xx/xx I

Recirculation Loops Operating B 3.4.1 BASES BACKGROUND is transferred to the coolant. As it rises, the coolant (continued) begins to boil, creating steam voids within the fuel channel that continue until the coolant exits the core. Because of reduced moderation, the steam voiding introduces negative

-reactivity that must be compensated for to maintain or to increase reactor power. The recirculation flow control allows operators to increase recirculation flow and sweep some of the voids from the fuel channel, overcoming the negative reactivity void effect. Thus, the reason for having variable recirculation flow is to compensate for reactivity effects of boiling over a wide range of power generation (i.e., 55 to 100% of RTP) without having to move control rods and disturb desirable flux patterns.

Each recirculation loop is manually started from the control room. The MG set provides regulation of individual recirculation loop drive flows. The flow in each loop is manually controlled.

APPLICABLE The operation of the Reactor Recirculation System is SAFETY ANALYSES an initial condition assumed in the design basis loss of coolant accident (LOCA) (Ref. 1). During a LOCA caused by a recirculation loop pipe break, the intact loop is assumed to provide coolant flow during the first few seconds of the accident. The initial core flow decrease is rapid because the recirculation pump in the broken loop ceases to pump reactor coolant to the vessel almost immediately. The pump in the intact loop coasts down relatively slowly. This pump coastdown governs the core flow response for the next several seconds until the jet pump suction is uncovered (Ref. 1). The analyses assume that both loops are operating at the same flow prior to the accident. However, the LOCA analysis was reviewed for the case with a flow mismatch between the two loops, with the pipe break assumed to be in the loop with the higher flow. While the flow coastdown and core response are potentially more severe in this assumed case (since the intact loop starts at a lower flow rate and the core response is the same as if both loops were operating at a lower flow rate), a small mismatch has been determined to be acceptable based on engineering judgement.

The recirculation system is also assumed to have sufficient flow coastdown characteristics to maintain fuel thermal margins during abnormal operational trnsie--n (Ref. 2),

which are analyzed in Section XIV of the USA loccurrences (AOOs)

(continued)

Cooper B 3.4-2 Revis4em 0

Recirculation Loops Operating B 3.4.1 BASES (Ref. 6)

APPLICABLE A plant specific LOCA analysis has been performed assuming SAFETY ANALYSES only one operating recirculation loop. This analysis has (continued) demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core cooling, provided the APLHGR requirements are modified accordingly (Ref. 3 The transient analyses -- ction XIV of the USAR have also been performed for single recirculation loop operation (Ref. 3) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR requirements are modified. During single recirculation loop operation, modification to the Reactor Protection System (RPS) average power range monitor (APRM) instrument Allowable Values is also required to account for the different relationships between recirculation drive flow and reactor core flow. The APLHGR and MCPR limits for single loop operation are specified in the COIR. The APRM Neutron Flux-High (Flow Biased) Allowable Value is in LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation."

The reactor is designed such that thermal hydraulic oscillations are prevented or can be readily detected and suppressed without exceeding specified fuel design limits.

To minimize the likelihood of a thermal hydraulic instability, a Stability Exclusion Region, to be avoided during normal power operation, is calculated using approved methodology. Since the Stability Exclusion Region may change each fuel cycle, the Stability Exclusion Region is contained in the COLR. Specific directions are provided to avoid operation in this region and to immediately exit upon entry. Entries Into the Stability Exclusion Region are not part of normal operation. An entry may occur as the result of an abnormal event, such as a single recirculation pump trip. In these events, operation in the Stability Exclusion Region may be needed to prevent equipment damage, but actual time spent inside the Region is minimized. Although operator action can prevent the occurrence of and protect the reactor from an instability, the APRM Neutron Flux-High (Flow Biased) scram function will suppress oscillations prior to exceeding the Safety Limit MCPR. While core-wide reactor instability is the predominate mode and regional mode oscillations are not expected to occur, the reactor is (continued)

Cooper B 3.4-3 CooprB.4-3ReV494efl-0

Recirculation Loops Operating B 34.1 BASES APPLICABLE SAFETY ANALYSES (continued) protected from regional mode oscillations through avoidance of the Stability Exclusion Region and administrative controls on reactor conditions which are primary factors affecting reactor stability.

Recirculation loops operating satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii) (Ref. 4).

LCO Two recirculation loops are required to be in operation with their flows LHGR limits (LCO matched within the limits specified in SR 3.4.1.1 to ensure that during a 3.2.3, "LINEAR HEAT LOCA caused by a break of the piping of one recirculation loop the GENERATION RATE assumptions of the L CA analysis are satisfied. Alternatively, with only (LHGR)"), one recirculation loop N operation, modifications to the required APLHGR limits (LCO 3.2.1, 'AVENAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), MCPlimits (LCO 3.2.2, "MINIMUM CRITIs3and4 POWER RATIO (MCPR)"r,and APRM Neutron Flux-High Biased) setpoint (LCO 3.3.1.1) must be applied to allow co i*ed operation consistent with the assumptions of Referenc . During single recirculation loop operation, the recirculation system controls are placed in the manual flow control mode. In addition, during two loop or single loop operation, core flow as a function of core THERMAL POWER must not be in the Stability Exclusion Region of the power/flow map specified in the COLR.

APPLICABILITY In MODES 1 and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur. F0or AOO In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important.

Cooper B 3.4-4 C010&42

Recirculation Loops Operating B 3.4.1 BASES (continued)

ACTIONS A.

Because of thermal hydraulic stability concerns, operation of the plant is controlled by restricting core flow and power to the unrestricted region of the power/flow map specified in the COLR. If core flow as a function of core THERMAL POWER is in the Stability Exclusion Region of the power/flow map, action must be initiated immediately to restore the power/flow combination to outside the Stability Exclusion Region. The operator must ither insert control rods to reduce THERMAL POWER, or incr se the speed of the operating recirculation pump(s). Action ' ust continue until the Stability Exclusion Region has been eited.

_ Je LB.

With the requirements of the LCO not met for reasons other than Condition A, the recirculation loops must be restored to operation with matched flows within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. A recirculation loop is considered not i peration when the pump in that loop is idle or when t mismatch between total jet pump flows of the two loops i greater than required orthA I limits. The loop with the lo flow must be considered in operation. Should a LOCAaccur with one recirculatio loop not in operation, the core flow coastdown and re0 tant core response may not be bounded by the LOCA analyses.

Therefore, only a limited time is allowed to restore the inoperable loop to operating status. In addition, following one recirculation pump operation, the discharge valve of the low speed pump may not be opened unless the speed of the faster pump is s 50% of its rated speed. This provides assurance that excessive vibration of the reactor vessel internals will not occur.

Alternatively, if the single loop requirements of the LCO are applied to operating limits and RPS setpoints, operation with only one recirculation loop would satisfy the requirements of the LCO and the initial conditions of the acciden sequence.

Ior AQO0 The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time is based on the low probability of an accident occurring during this time period, on a reasonable time to complete the Required Action, and on frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected.

(continued)

Cooper B 3.4-5 Reviio

Recirculation Loops Operating B 3.4.1 BASES REFERENCES 5. GENE-A13-00395-01, "Application of the Regional (continued) Exclusion with Flow-Biased APRM Neutron Flux Scram Stability Solution (Option I-D) to the Cooper Nuclear Station," November 1996.

6. NEDC-32687P, Revision 1,"Cooper Nuclear Station SAFER/

\rGESTR-LOCA Loss-of-Coolant Accident Analysis," March 1997 Cooper B 3.4-8

Main Turbine Bypass System B 3.7.7 B 3.7 PLANT SYSTEMS B 3.7.7 Main Turbine Bypass System BASES BACKGROUND The Main Turbine Bypass System is designed to control steam pressure when reactor steam generation exceeds turbine requirements during unit startup, sudden load reduction, and cooldown. It allows excess steam flow from the reactor to the condenser without going through the turbine.

The bypass capacity of the system is 25% of the Nuclear Steam Supply System rated steam flow. Sudden load reductions within the capacity of the steam bypass can be accommodated without safety relief valves opening or a reactor scram. The Main Turbine Bypass System consists of three valves connected to the main steam lines between the main steam isolation valves and the turbine stop valves. Each of these three valves is operated by hydraulic cylinders. The bypass valves are controlled by the pressure regulation function of the Turbine Digital Electro Hydraulic (DEH) Control System, as discussed in the USAR, Section VII-1 1.3 (Ref. 1). The bypass valves are normally closed, and the DEH Control System controls the turbine control valves that direct all steam flow to the turbine. If the speed govemor or the load limiter restricts steam flow to the turbine, the DEH Control System controls the system pressure by opening the bypass valves. When the bypass valves open, the steam flows from the bypass chest, through connecting piping, to the pressure breakdown assemblies used to further reduce the steam pressure before the steam enters the condenser.

APPLICABLE SAFETY ANALYSES The Main Turbine Bypass System is assumed to function during the high energy line break analysis, as discussed in References 2 and 3, and the feedwater controller failure maximum demand transient, as discussed in Reference 4. However, the feedwater controller failure maximum An inoperable Main Turbine demand transient defines the MCPR operating limits if one Main Turbine Bypass Valve may result in Bypass----* Va-!'ve isim.

e.>

APLHGR, LHGR, or MCPR penalties. The Main Turbine Bypass System satisfies Criterion 3 of 10 CFR 50.36(cX2)(ii) (Ref. 5).

LCO The Main Turbine Bypass System is required to be OPERABLE to limit the APLHGR limits (LCO 3.2.1, peak pressure in the main steam lines and maintain reactor pressure "AVERAGE PLANAR LINEAR within acceptable limits during events that cause rapid pressurization, so HEAT GENERATION RATE that the Safety Limit MCPR is not exceeded. With one Main Turbine (APLHGR)"), and the LHGR -odificýaions to the MCPR operating limits limits (LCO 3.2.3, "LINEAR (LCO 3.2.2, "MINIMUM CRITICAL may be HEAT GENERATION RATE (LHGR)")

Cooper B 3.7-31 344I2ll qI

Main Turbine Bypass System B 3.7.7 BASES I'the APLHGR LCO (continued) IlimitadLG ilimit, n H Rý o n applied to allow this LCO to be met. The MCPR operatinglimn for one inoperable Main Turbine Bypass Valve are specified in the COLR. An OPERABLE Main Turbine Bypass System requires all three bypass valves to open in response to increasing main steam line pressure. This response is within the assumptions of the applicable analyses (Ref. 4).

APPLICABILITY The Main Turbine Bypass System is required to be OPERABLE at a 25%

RTP to ensure that the fuel cladding integrity Safety Limit and the cladding 1% plastic strain limit are not violated during the Applicable Safety Analyses transients. As discussed in the Bases for TLGCO 3.2. 1 "LINEA*R HEAT G.ENEP.ATION RATE (LHGR)," a.nd LGO 3.2.2, sufficient margin to these limits exists at < 25% RTP. Therefore, these rain requirements are only necessary when operating at or above thisfi wer level. ILCO 3.2.1, LCO 13.2.2, and LCO ACIOS .1APLHGR limit, I 13.2.3, ACTIONSA.._.*ILHGR limit and Ifone Main Turbine Bypass Valve is inoperable, and the MCPR operating limits for one inoperable Main Turbine Bypass Valve, as specified in the COLR, are not applied, the assumptions of the design basis transient analyses may not be met. Under such circumstances, prompt action should be taken to restore the inoperable Main Turbine Bypass Valve to OPERABLE status or adjust the MCPR operating limits accordingly. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is reas ble, based on the time to complete the Required Action and the low proba flity of an event occurring during this period requiring the Main Turbine By ss System.

"*~APLHGR limit, APLH R lriiýý,ILHGR limit and LHGR limit anda TrI If the inoperab e Tur ine Bypass Valve cannot be restored to OPERABLE status and t MCPR operating limits for one inoperable Main Turbine Bypass Valve are not applied within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or two or more Main Turbine Bypass Valves are inoperable, THERMAL POWER must be reduced to < 25% RTP. As discussed in the Applicability section, operation at < 25% RTP results in sufficient margin to the required limits, and the Main Turbine Bypass System is not required to protect fuel integrity during the Applicable Safety Analyses transients. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time is reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

Cooper B 3.7-32 1 B3.7-32 Cooper42 Ia