L-19-213, Application to Revise Technical Specifications to Adopt TSTF-564. Safety Limit MCPR
| ML19352D673 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 12/18/2019 |
| From: | Payne F FirstEnergy Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-19-213 | |
| Download: ML19352D673 (21) | |
Text
FirstEnergy Nuclear Operating Company Perry Nuclear Power Plant PO. Box 97 1 D Center Road Perry, Ohio 44081 Frank Payne Vice President 440-280-5382 December 18, 2019 L-19-213 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
Perry Nuclear Power Plant Docket No. 50-440, License No. NPF-58 Application to Revise Technical Specifications to Adopt TSTF-564. "Safety Limit MCPR" Pursuant to 10 CFR 50.90, the FirstEnergy Nuclear Operating Company (FENOC) is submitting a request for an amendment to the Technical Specifications (TS) for the Perry Nuclear Power Plant (PNPP).
FENOC requests adoption of TSTF-564, "Safety Limit MCPR," Revision 2, which is an approved change to the Improved Standard Technical Specifications (ISTS), into the PNPP Technical Specifications (TS). The proposed amendment revises the Technical Specification (TS) safety limit (SL) on minimum critical power ratio (MCPR) to reduce the need for cycle-specific changes to the value while still meeting the regulatory requirement for an SL. provides a description and assessment of the proposed changes. provides the existing TS pages marked to show the proposed changes. provides the revised TS pages. provides the existing TS Bases pages marked to show the proposed changes for information only.
Approval of the proposed amendment is requested by December 31, 2020. Once approved, the amendment shall be implemented within 90 days.
In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State of Ohio Official.
There are no regulatory commitments contained in this submittal. If there are any questions or if additional information is required, please contact Mr. Phil H. Lashley, Acting Manager - Nuclear Licensing and Regulatory Affairs, at (330) 315-6808.
Perry Nuclear Power Plant L-19-213 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on December I~, 2019.
rely, /\\(\\
,1~~
R. PaynJ/
Attachments:
- 1. Description and Assessment of License Amendment Request to Adopt TSTF-564
- 2. Proposed Technical Specification Changes (Mark-Up)
- 3. Revised Technical Specification Pages
- 4. Proposed Technical Specification Bases Changes (Mark-Up) for Information Only cc:
NRC Regional Ill Administrator NRC Resident Inspector N RR Project Manager Executive Director, Ohio Emergency Management Agency, State of Ohio (NRC Liaison)
Utility Radiological Safety Board L-19-213 Description and Assessment of License Amendment Request to Adopt TSTF-564 Page 1 of 4
Subject:
License Amendment Request for Adoption of Technical Specification Task Force Traveler TSTF-564, Revision 2, Safety Limit MCPR 1.0
SUMMARY
DESCRIPTION
2.0 ASSESSMENT
2.1 Applicability of Safety Evaluation 2.2 Variations
3.0 REGULATORY ANALYSIS
3.1 Significant Hazards Consideration Analysis 3.2 Conclusion 4.0 ENVIRONMENTAL EVALUATION L-19-213 Page 2 of 4 1.0
SUMMARY
DESCRIPTION FirstEnergy Nuclear Operating Company (FENOC) requests adoption of TSTF-564, "Safety Limit MCPR," Revision 2, which is an approved change to the Improved Standard Technical Specifications (ISTS), into the Perry Nuclear Power Plant (PNPP),
Technical Specifications (TS). The proposed amendment revises the Technical Specification (TS) safety limit (SL) on minimum critical power ratio (MCPR) to reduce the need for cycle-specific changes to the value while still meeting the regulatory requirement for an SL.
2.0 ASSESSMENT
2.1 Applicability of Safety Evaluation FENOC has reviewed the safety evaluation for TSTF-564 provided to the Technical Specifications Task Force in a letter dated November 19, 2018. This review included a review of the NRC staffs evaluation, as well as the information provided in TSTF-564.
As described herein, FENOC has concluded that the justifications presented in TSTF-564 and the safety evaluation prepared by the NRC staff are applicable to PNPP and justify this amendment for the incorporation of the changes to the PNPP TS.
The PNPP reactor is currently fueled with GNF2 fuel bundles. The proposed Safety Limit in SL 2.1.1.2 is 1.07, consistent with Table 1 of TSTF-564.
The MCPR value calculated as the point at which 99.9% of the fuel rods would not be susceptible to boiling transition (i.e., reduced heat transfer) during normal operation and anticipated operational occurrences is referred to as MCPR99.9%. Technical Specification 5.6.5, "Core Operating Limits Report (COLR)," is revised to require the MCPR99.9% value to be included in the cycle specific COLR.
2.2 Variations The PNPP TS utilize different numbering than the Standard Technical Specifications on which TSTF-564 was based. Specifically, TSTF-564 revises Standard Technical Specification 5.6.3, "Core Operating Limits Report to add the MCPR99.9% value for anticipated operational occurrences. PNPP Technical Specification 5.6.5 is the applicable section to the PNPP TS. These differences are administrative and do not affect the applicability of TSTF-564 to the PNPP TS.
3.0 REGULATORY ANALYSIS
3.1 No Significant Hazards Consideration Analysis FENOC requests adoption of TSTF-564, "Safety Limit MCPR," which is an approved change to the Improved Standard Technical Specifications (ISTS), into the Perry Nuclear Power Plant (PNPP) Technical Specifications (TS). The proposed change revises the Technical Specifications (TS) safety limit on minimum critical power ratio L-19-213 Page 3 of 4 (SLMCPR). The revised limit calculation method is based on using the Critical Power Ratio (CPR) data statistics and is revised from ensuring that 99.9% of the rods would not be susceptible to boiling transition to ensuring that there is a 95% probability at a 95% confidence level that no rods will be susceptible to transition boiling. A single SLMCPR value will be used instead of two values applicable when one or two recirculation loops are in operation. TS 5.6.5, "Core Operating Limits Report (COLR),"
is revised to require the current SLMCPR value to be included in the COLR.
FENOC has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
- 1.
Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The proposed amendment revises the TS SLMCPR and the list of core operating limits to be included in the Core Operating Limits Report (COLR). The SLMCPR is not an initiator of any accident previously evaluated. The revised safety limit values continue to ensure for all accidents previously evaluated that the fuel cladding will be protected from failure due to transition boiling. The proposed change does not affect plant operation or any procedural or administrative controls on plant operation that affect the functions of preventing or mitigating any accidents previously evaluated.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2.
Does the proposed amendment create the possibility of a new or different kind of accident from any previously evaluated?
Response: No The proposed amendment revises the TS SLMCPR and the list of core operating limits to be included in the COLR. The proposed change will not affect the design function or operation of any structures, systems or components (SSCs).
No new equipment will be installed. As a result, the proposed change will not create any credible new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
L-19-213 Page 4 of 4
- 3.
Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No The proposed amendment revises the TS SLMCPR and the list of core operating limits to be included in the COLR. This will result in a change to a safety limit but will not result in a significant reduction in the margin of safety provided by the safety limit. As discussed in the application, changing the SLMCPR methodology to one based on a 95% probability with 95% confidence that no fuel rods experience transition boiling during an anticipated transient instead of the current limit based on ensuring that 99.9% of the fuel rods are not susceptible to boiling transition does not have a significant effect on plant response to any analyzed accident. The SLMCPR and the TS Limiting Condition for Operation (LCO) on MCPR continue to provide the same level of assurance as the current limits and do not reduce a margin of safety.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, FENOC concludes that the proposed change 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.
3.2 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 Commissions 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.
4.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change 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 change.
L-19-213 Proposed Technical Specification Changes (Mark-up)
(4 pages follow)
SLs 2.0 PERRY - UNIT 1 2.0-1 Amendment No. 176 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 686 psig or core flow < 10%
rated core flow:
THERMAL POWER shall be 23.8% RTP.
2.1.1.2 With the reactor steam dome pressure 686 psig and core flow 10%
rated core flow:
The Minimum Critical Power Ratio (MCPR) shall be 1.071.10 for two recirculation loop operation or 1.13 for single recirculation loop operation.
2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.
2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be 1325 psig.
2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.
3.2 POWER DISTRIBUTION LIMITS 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)
MCPR 3.2.2 LCD 3. 2. 2 All MCPRs sha 11 be greater than or equal to the MCPR operating limits specified in the COLR.
APPLI CAB IL ITV:
THERMAL POWER t 23. 8% RTP.
ACTIONS CONDITION REQUIRED ACTION A.
Any MCPR not within A.l Restore MCPR(s) to limits.
within limits.
B.
Required Action and B.l Reduce THERMAL POWER associated Completion to< 23.8% RTP.
Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE COMPLETION TIME 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 4 hours FREQUENCY SR 3.2.2.1 Verify all MCPRs are greater than or equal Once within to the limits specified in the COLR.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after
23.8% RTP PERRY - UNIT 1 3.2-2 AND In accordance with the Surveillance Frequency Control Program Amendment No.171 THIS PAGE INCLUDED FOR CONTEXT ONLY
THIS PAGE INCLUDED FOR CONTEXT ONLY Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.2 5.6.3 5.6.4 5.6.5 Annual Radiological Environmental Operating Report (continued) results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position.
Revision 1. November 1979.
In the event that some individual results are not available for inclusion with the report. the report shall be submitted noting and explaining the reasons for the missing results.
The missing data shall be submitted in a supplementary report as soon as possible.
Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit during the previous year shall be submitted by May 1 of each year.
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 ODCM and process control program and in conformance with 10 CFR 50.36a and 10 CFR 50.
Appendix I,Section IV. B.1.
Monthly Operating Reports Deleted Core Operating 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.
LCO 3.2.1. Average Planar Linear Heat Generation Rate (APLHGR).
- 2.
LCD 3.2.2. Minimum Critical Power Ratio (MCPR).
- 3.
LCD 3.2.3. Linear Heat Generation Rate (LHGR).
(continued)
PERRY - UNIT 1 5.0-17 Amendment No. 136
Reporting Requirements 5.6 PERRY - UNIT 1 5.0-18 Amendment No. 140 5.6 Reporting Requirements 5.6.5 Core Operating Limits Report (COLR) (continued) 4.
LCO 3.3.1.1, RPS Instrumentation (SR 3.3.1.1.14), and 5.
LCO 3.3.1.3, Oscillation Power Range Monitor (OPRM)
Instrumentation., and 6.
The MCPR99.9% value used to calculate the LCO 3.2.2, MCPR, limit.
b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in 1). NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel or 2). NEDO-32465 Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications. (The approved revision at the time reload analyses are performed shall be identified in the COLR.)
c.
The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
d.
The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.6 Special Reports Deleted.
L-19-213 Revised Technical Specification Pages (2 pages follow)
SLs 2.0 PERRY - UNIT 1 2.0-1 Amendment No. 176 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 686 psig or core flow < 10%
rated core flow:
THERMAL POWER shall be 23.8% RTP.
2.1.1.2 With the reactor steam dome pressure 686 psig and core flow 10%
rated core flow:
The Minimum Critical Power Ratio (MCPR) shall be 1.07.
2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.
2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be 1325 psig.
2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.
Reporting Requirements 5.6 PERRY - UNIT 1 5.0-18 Amendment No. 140 5.6 Reporting Requirements 5.6.5 Core Operating Limits Report (COLR) (continued)
- 4.
LCO 3.3.1.1, RPS Instrumentation (SR 3.3.1.1.14),
- 5.
LCO 3.3.1.3, Oscillation Power Range Monitor (OPRM)
Instrumentation, and
- 6.
The MCPR99.9% value used to calculate the LCO 3.2.2, MCPR, limit.
- b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in 1). NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel or 2). NEDO-32465 Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications. (The approved revision at the time reload analyses are performed shall be identified in the COLR.)
- c.
The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
- d.
The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.6 Special Reports Deleted.
L-19-213 Proposed Technical Specification Bases Changes (Mark-up) for Information Only (6 pages follow)
Reactor Core SLs B 2.1.1 PERRY - UNIT 1 B 2.0-1 Revision No. 0 B 2.0 SAFETY LIMITS (SLs)
B 2.1.1 Reactor Core SLs BASES BACKGROUND GDC (Ref. 1) requires, and SLs ensure, that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).
The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a stepback approach is used to establish an SL, such that the MCPR is not less than the limit specified in Specification 2.1.1.2. MCPR greater than the specified limit 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 measureable. Fuel cladding perforations, however, can result from thermal stresses, which occur from reactor operation significantly above design conditions.
While 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 SL is defined with a margin to the conditions that would produce onset of transition boiling (i.e., MCPR = 1.00). These conditions represent a significant departure from the condition intended by design for planned operation. The MCPR fuel cladding integrity SL ensures that during normal operation and during AOOs, at least 99.9% of the fuel rods in the core do not experience transition boiling.This is accomplished by having a Safety Limit Minimum Critical Power Ratio (SLMCPR) design basis, referred to as SLMCPR95/95, which corresponds to a 95% probability at a 95% confidence level (the 95/95 MCPR criterion) that transition boiling will not occur.
(continued)
Information Only
Reactor Core SLs B 2.1.1 PERRY - UNIT 1 B 2.0-2 Revision No. 11 BASES BACKGROUND Operation above the boundary of the nucleate boiling regime could result (continued) in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient.
Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.
APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY operation and AOOs. The reactor core SLs are established to ANALYSES preclude violation of the fuel design criterion that an MCPR SL is to be established, such that at least 99.9% of the fuel rods in the core would not be expected to experience the onset of transition boiling.The Tech Spec SL is set generically on a fuel product MCPR correlation basis as the MCPR which corresponds to a 95% probability at a 95% confidence level that transition boiling will not occur, referred to as SLMCPR95/95.
The Reactor Protection System setpoints (LCO 3.3.1.1, Reactor Protection System (RPS) Instrumentation), in combination with other LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERMAL POWER level that would result in reaching the MCPR SL.
2.1.1.1 Fuel Cladding Integrity GE critical power correlations are applicable for all critical power calculations at pressures 686 psig and core flows 10% of rated flow.
For operation at low pressures or low flows, another basis is used, as follows:
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 > 4.5 psi. Analyses (Ref. 2) show that with a bundle flow of 28 x 103 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 > 28 x 103 lb/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia (continued)
Information Only
Reactor Core SLs B 2.1.1 PERRY - UNIT 1 B 2.0-3 Revision No. 11 BASES APPLICABLE 2.1.1.1 Fuel Cladding Integrity (continued)
SAFETY ANALYSES 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 > 47.6% RTP. Thus, a THERMAL POWER limit of 23.8% RTP for reactor pressure
< 686 psig is conservative.
2.1.1.2 MCPR The fuel cladding integrity SL is set such that no significant 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 that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur. Although it is recognized that the onset of transition boiling would not 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 SL is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties.The Technical Specification SL value is dependent on the fuel product line and the corresponding MCPR correlation, which is cycle independent. The value is based on the Critical Power Ratio (CPR) data statistics and a 95%
probability with 95% confidence that rods are not susceptible to boiling transition, referred to as MCPR95/95.
The MCPR SL is determined using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved General Electric critical power correlations. Details of the fuel cladding integrity SL calculation are given in Reference 2. Reference 2 also includes a tabulation of the uncertainties used in the determination of the MCPR SL and of the nominal values of the parameters used in the MCPR SL statistical analysis.The SL is based on GNF2 fuel. For cores with a single fuel product line, the SLMCPR95/95 is the MCPR95/95 for the fuel type. For cores loaded with a mix of applicable fuel types, the SLMCPR95/95 is based on the largest (i.e., most limiting) of the MCPR values for the fuel product lines that are fresh or once-burnt at the start of the cycle.
(continued)
Information Only
MCPR B 3.2.2 PERRY - UNIT 1 B 3.2-6 Revision No. 0 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)
BASES BACKGROUND 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 (refer to the Bases for SL 2.1.1.2). The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (AOOs), and that 99.9% of the fuel rods avoid boiling transition if the limit is not violated. Although fuel damage does not necessarily occur if a fuel rod actually experiences boiling transition (Ref. 1), 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.
APPLICABLE The analytical methods and assumptions used in evaluating the AOOs to SAFETY establish the operating limit MCPR are presented in the USAR, Chapters ANALYSES 4, 6, and 15, and References 2, 3, 4, 5, and 6. To ensure that the MCPR Safety Limit (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 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 (CPR). When the largest CPR is added tocombined with the SLMCPR99.9% SL, the required operating limit MCPR is obtained. MCPR99.9% is determined to ensure more than 99.9% of the fuel rods in the core are not susceptible to boiling transition using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved General Electric Critical Power correlations. Details of the MCPR99.9% calculation are given in Reference 2. Reference 2 also includes a tabulation of the uncertainties and the nominal values of the parameters used in the MCPR99.9%
statistical analysis.
(continued)
MCPR B 3.2.2 PERRY - UNIT 1 B 3.2-7 Revision No. 4 BASES APPLICABLE The MCPR operating limits are derived from the MCPR99.9% value and the SAFETY transient analysis and, are dependent on the operating core flow and ANALYSES power state (MCPRf and MCPRp, respectively) to ensure adherence to (continued) fuel design limits during the worst transient that occurs with moderate frequency (Refs. 4, 5, and 6).
Flow dependent MCPR limits (MCPRf) are determined by steady state thermal hydraulic methods using the three dimensional BWR simulator code (Ref. 7). MCPRf curves are provided based on the maximum credible flow runout transient for Loop Manual and Non Loop Manual operation. The result of a single failure or single operator error during Loop Manual operation is the runout of only one loop because both recirculation loops are under independent control. Non Loop Manual operational modes allow simultaneous runout of both loops because a single controller regulates core flow.
Power dependent MCPR limits (MCPRp) are determined by the three dimensional BWR simulator code and the one dimensional transient codeapproved transient analysis models (Ref. 8). 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 scram trips are bypassed, high and low flow MCPRp operating limits are provided for operating between 23.8% RTP and the previously mentioned bypass power level.
Pressure Regulator Out of Service (PROOS) option is an analysis using the Pressure Regulator Downscale Failure (PRDF) at off-rated conditions.
At full power, the PRDF is bounded by other pressurization transients.
However, as the reactor power at the beginning of the transient decreases, the impact of the PRDF to MCPR increases.
During a PRDF transient, the pressure regulator closes the turbine control valves. This increases pressure, which increases power in the reactor.
When the reactor is at full power, the pressure and power increases quickly, causing a SCRAM. As the reactor power is decreased, the power is further from the SCRAM setpoint so it takes more time to SCRAM. This longer time to SCRAM increases the amount of specific heat in the fuel and impacts the CPR. There is a range of initial reactor power where the CPR is no longer bounded by the normal MCPRp limits.
(continued)
MCPR B 3.2.2 PERRY - UNIT 1 B 3.2-7a Revision No. 7 BASES APPLICABLE There are two independent channels in the pressure regulating system SAFETY and the PRDF transient is not applicable when both channels are ANALYSES operable.
(continued)
The COLR identifies the range of the modified MCPR limits and the new limits. These limits may be incorporated by either a revision to the monitoring system or appropriate administrative limits.
The MCPR satisfies Criterion 2 of the NRC Final Policy Statement on Technical Specification Improvements (58 FR 39132).
LCO The MCPR operating limits specified in the COLR (MCPR99.9% values, MCPRf values, and MCPRp values) are the result of the Design Basis Accident (DBA) and transient analysis. The MCPR operating limits are determined by the larger of the MCPRf and MCPRp limits, which are based on the MCPR99.9% limit specified in the COLR.
APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 23.8% RTP, the reactor is operating at a slow recirculation pump speed and the moderator void ratio is small. Surveillance of thermal limits below 23.8% RTP is unnecessary due to the large inherent margin that ensures that the MCPR95/95 SL is not exceeded even if a limiting transient occurs.
(continued)