ML20045E147
| ML20045E147 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 06/23/1993 |
| From: | Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML19303F669 | List: |
| References | |
| NUDOCS 9307010179 | |
| Download: ML20045E147 (26) | |
Text
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10 CFR 50.90 PHILADELPHIA ELECTRIC COMPANY NUCLEAR GROUP HEADQUARTERS 955-65 CHESTERBROOK BLVD.
WAYNE, PA 19087-5691 June 23, 1993 Docket Nos. 50-277 50-278 STATION SUPPORT DEPARTMENT License Nos. DPR-44 DPR-56 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC J0555
Subject:
Peach Bottom Atomic Power Station, Units 2 and 3 Operating License Change Request 93-12-0 (entlemen:
Philadelphia Electric Company (PECo) hereby requests a change to Operating License Nos. DPR-44 and DPR-56 and Appendices A and G thereto, for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3 respectively. The proposed changes to the Operating Licenses and Appendices A and B reflect the planned implementation of the Power Rerate Program at PBAPS Units 2 and 3, and the corresponding increase in the authorized maximum reactor core power level by five percent to 3458 megawatts thermal (MWt) from the current limit of 3293 MWt. to this letter describes the proposed changes, and contains information supporting a finding that the proposed changes do not involve a Significant Hazards Consideration and information supporting an Environmental Assessment. Attachment 2 contains the Operating License and Appendices A and B pages showing the proposed changes. Attachment 3, NEDC-32183P, " Power Rerate Safety Analysis Report for Peach Bottom 2 & 3," dated May 1993, contains the safety analysis prepared by General Electric (GE) to support this Change Request and the implementation of Power Rerate at PBAPS Units 2 and 3.
The analyses and evaluations supporting these proposed changes were completed using the guidelines in GE Topical Report NEDC-31897P-A, " Generic Guidelines-for General Electric Boiling Water Reactor Power Uprate," dated May 1992. - The resolution of generic issues associated with power uprate was addressed in GE Topical Report NEDC-31984P, " Generic Evaluations of General Electric Boiling Water Reactor Power Uprate," dated July 1991. These reports are proprietary and have been submitted separately to the NRC by GE. The NRC reviewed and approved these Topical Reports by letters from the NRC to GE dated September-7 30, 1991 and July 31, 1992.
The safety analysis in Attachment 3 complies with the guidelines of NEDC-31897P-A and supplements the generic evaluations of-NEDC-31984P with PBAPS specific information.
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'U.S. Nuclear Regulatory Commission June 23, 1993 License Change Request 93-12-0 Page 2
( contains information proprietary to GE. GE requests that the document be withheld from public disclosure in accordance with 10 CFR 2.790(a)(4). An affidavit supporting this request in accordance with 10 CFR 2.790(b)(1) is provided with Attachment 3.
We are requesting that, if approved, the NRC make the amendments effective by September 11, 1994, for PBAPS Unit 2, and September 8, 1995, for PBAPS Unit 3.
If you have any questions or require additional information, please contact us.
Very truly yours,
. 0. L h.
G. A. Hunger, irector Licensing Section Attachments cc:
T. T. Martin, Administrator, Region I, USNRC w/ attachments USNRC Senior Resident Inspector, PBAPS W. P. Dornsife, Director, Bureau of Radiological Protection 4
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ATTACHMENT 1 l
l PEACH BOTTOM ATOMIC POWER STATION Units 2 and 3 Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 OPERATING LICENSE CHANGE REQUEST
" Power Rerate Program for Peach Bottom Atomic Power Station, Units 2 and 3" Supporting Information for Changes - 22 pages
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C0MMONWEALTH OF PENNSYLVANIA :
ss.
COUNTY OF CHESTER G. R. Rainey, being first duly sworn, deposes and says:
1 i
That he is Vice President of Philadelphia Electric Company; the Applicant herein; that he has read the foregoing Application for Amendment of Facility Operating License Nos'. DPR-44 and DPR-56 (Operating License Change Request No. 93-12-0) for the Power Rerate Program to be implemented,at PBAPS Units 2 and 3, and knows the contents thereof; and that the statements and matters set forth therein are true and correct to the best of his knowledge, 1
information and belief.
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Subscribedandswor7to
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Page 1 Philadelphia Electric Company (PECo), licensee under Facility Operating License Nos. OPR-44 and DPR-56 for Peach Bottom Atomic Power Station (PBAPS),
Units 2 and 3, requests that these licenses be amended as proposed herein to reflect the Power Rerate Program to be implemented at PBAPS, Units 2 and 3, specifically to increase the maximum reactor core power level by five percent (5%), to 3458 megawatts thermal (MWt) from the current limit of 3293 MWt.
The analyses and evaluations supporting these changes were completed using the guidelines in General Electric (GE) Topical Report NEDC-31897P-A, " Generic Guidelines for General Electric Boiling Water Reactnr Power Uprate," dated May 1992. This Topical Report was reviewed and approved by the NRC, by letter to GE, dated September 30, 1991.
Resolution of generic issues associated with power uprate was addressed in GE Topical Report NEDC-31984P, " Generic-Evaluations of General Electric Boiling Water Reactor Power Uprate," dated July 1991. This Topical Report was reviewed and approved by the NRC, by letter to GE, dated July 31, 1992.
An increase in electrical output is accomplished primarily by generation and supply of higher steam flow to the turbine generator. Continuing improvements in the analytical techniques (i.e., computer codes and data) based on several decades of Boiling Water Reactor (BWR) safety technology, plant performance feedback, and improved fuel and core design, have resulted in a significant increase in the margin between calculated safety analysis results and the licensing limits. This available safety analysis margin, combined with the excess capability of as-designed equipment, systems and components, provide the potential for an increase of 5% in the full power rating of a plant without the need to perform major Nuclear Steam Supply System (NSSS) or Balance-of-Plant (B0P) hardware modifications. The full power level can be increased safely, and the installed systems and equipment are capable of performing required functions at the rerated conditions. The method for achieving higher power is to extend the reactor core power-flow map by increasing reactor core flow along existing flow control lines. However, there will not be an increase in the maximum recirculation flow limit over the pre-rerate value. Most of the original safety analyses, such as the transient (i.e., abnormal operating events) analyses, were based on 105% steam flow, which coincides closely with the steam flow at the proposed rerated full power level.
The safety analysis prepared by GE to support this Change Request and the implementation of the Power Rerate Program at PBAPS Units 2 and 3 is provided in Attachment 3, and demonstrates that the PBAPS, Units 2 and 3 can operate safely with the proposed 5% increase of the maximum reactor thermal power and an associated 30 psi increase in the operating reactor vessel pressure, with a corresponding increase in main turbine inlet steam flow and the corresponding increases of the flow, temperature, pressure, and capacity required in supporting systems and components at these rerated conditions. This safety analysis has been performed taking into account the current 24 month fuel cycles for PBAPS, Units 2-and 3, and the implementation of Average Power Range Monitor-Rod Block Monitor Technical Specifications / Maximum Extended Load Line Limit Analysis (ARTS /MELLLA) at PBAPS Units 2 and 3, prior to power rerate.
On April 1,1993, PEco submitted Technical Specifications Change Request (TSCR) No. 93-01, proposing the ARTS /MELLLA TS changes for PBAPS, Units 2 and 3, to the NRC.
Page.2 3
This Operating License Change Request for PBAPS, Units 2 and 3, provides a discussion and description of the proposed changes, a safety assessment, information supporting a finding of No Significant Hazards Consideration, and information supporting an Environmental Assessment.
We request that, if approved, the changes to the Operating Licenses for PBAPS be effective by September 11, 1994, for PBAPS Unit 2, and September 8, 1995, for PBAPS Unit 3.
I. Discussion and Description of the Proposed Chanaes A.
The proposed Operating License and Appendices A and B changes associated with the planned implementation of the Power Rerate Program at PBAPS Units 2 and 3 are as follows.
Operating License (0L) a) Revise Operating License " Maximum Power Level," Page 5.
OL Appendix A " Technical Specifications" (TS) a) Definitions i)
Revise Hot Standby Condition paragraph in TS Section 1.0, "DEFINIl:9NS," Page 2.
ii) Revise " Rated Power" paragraph in TS Section 1.0,
" DEFINITIONS," Page 6.
b) Fuel Cladding Integrity i)
Revise "APRM Flux Scram Trip Setting (Run Mode)," TS Section 2.1.A.1, Page 9.
ii)
Revise "APRM Rod Block Trip Setting," TS Section 2.1.B, Page 11.
iii) Revise "APRM Flow Bias Scram Relationship To Normal Operating Conditions," TS Figure 1.1-1, Page 16.
iv)
Revise and add reference to " Fuel Cladding Integrity," TS 1
Bases 2.1, Pages 17 and 18.
c) References i) Add reference to." References," TS Bases 2.1.L, Page 24.
d) Reactor Coolant System Integrity
- 1) Revise " Protective Action / Limiting Safety System Setting," TS Section 2.2.1, Pages 29 and 30.
i e)
Reactor Protection System (RPS) i) Revise "RPS Instrumentation Requirement," TS Table 3.1.1, Pages 37, 39 and 40.
u
Page 3 ii) Revise TS Bases 3.1, Pages 49 and 50.
f)
Control Rod Block Actuation
- 1) Revise " Instrumentation That Initiates Control Rod Blocks," TS Table 3.2.0, Pages 73 and 74.
g)
Standby Liquid Control System (SLCS) i) Revise " Standby Liquid Control System," TS Sections 3.4.3 and 4.4.3, Page 117.
h) High Pressure Coolant Injection (HPCI) System i) Revise "HPCI Subsystem," TS Section 4.5.C, Page 129, ii) Revise TS Bases 3.5.C, Page 137.
i)
Reactor Core Isolation Cooling (RCIC) System i) Revise " Reactor Core Isolation (RCIC) Subsystem," TS Section 4.5.D, Page 130.
i j) Minimum Critical Power Ratio (MCPR)
- 1) Add reference to " Operating Limit MCPR" paragraph, TS Bases 3.5.K, Page 140a.
k)
References i
i) Add reference to " References," TS Bases 3.5.M, Page 140c.
f 1)
Safety and Relief Valves
- 1) Revise TS Bases 3.6.D and 4.6.D, Page 157.
m)
TS figures i) Revise " Thermal Power and Core Flow limits," TS Figure 3.6.5, Page 164d.
n)
Primary Containment Isolation l
i) Revise " Primary Containment," TS Bases 3.7.4 and 4.7.4, Page 189.
ii) Revise " Leak Rate Testing," TS Bases 3.7.A and 4.7.A, Page 193.
iii) Revise " Post-LOCA Atmosphere Dilution," TS Bases 3.7.A and 4.7.A, Page 195.
i i
1
Page 4 1
8 OL Appendix B " Environmental Technical Specifications" (ETS) a) Cover Sheet b) Definitions 1)
Revise " DEFINITIONS," ETS Section 1.1, Page 2.
B.
The proposed OL and Appendices A and B changes specified above for PBAPS Units 2 and 3, have been grouped by categories and are discussed below. This discussion is a summary of information provided in the safety analysis provided in Attachment 3.
1.
Revise the Average Power Range Monitor (APRM) equation to reflect the rated thermal power increase of 5%.
OL, Appendix A TS Section 2.1.A.1, pg. 9 TS Section 2.1.B, pg. 11 TS Table 3.1.1, pg. 37 TS Table 3.2.C, pg. 73 The APRM signals will be recalibrated to the rerated maximum power level, and the percentage setpoints will be lowered. Since the Maximum Extended Load Line Limit (MELLL) of the reactor core power-flow map region will be changed as a result of the planned power rerate, the APRM flow-biased control rod block and reactor SCRAM setpoints will be changed accordingly. These changes will maintain the same margin from the upper limit of the MELLL region of the power-flow map to assure that the change is conservative.
The current Rod Block Monitor (RBM) setpoints were assumed in the Rod Withdrawal Error (RWE) transient analysis at the rerated maximum power level. This analysis demonstrates that RBM performance at the rerated power conditions is adequate to ensure that local power excursions within the reactor core due to any potential RWE are maintained within acceptable limits.
Power rerate will have little effect on the Intermediate Range Monitors (IRMs) overlap with the Source Range Monitors (SRMs) and the APRMs. Using normal plant procedures, the IRMs will be adjusted, as required, so that overlap with the SRMs and APRMs remains adequate. No change is needed in the APRM downscale setting.
2.
Revise the values of the Safety Relief Valves (SRVs) settings and Safety Valve settings by increasing the value of the setpoint by the increase in nominal operating pressure (i.e., 30 psi) at the rerated maximum power level.
OL, Appendix A TS Section 2.2.1.B, pg. 29 TS Section 2.2.1.C, pg. 30 1
Page 5 The installed safety valves and the SRVs have been determined to be acceptable under the rerated operating conditions. The safety valves and SRVs setpoints will, however, be reset to higher pressure values.
PECo has determined that the Nuclear Boiler Pressure Relief System has the capability to accommodate operation at the rerated maximum power level.
3.
Revise Reactor Core Isolation Cooling (RCIC) system operating requirements to reflect rerate conditions.
OL, Appendix A TS Section 4.5.D, pg. 130 The RCIC system provides core cooling when the Reactor Pressure Vessel (RPV) is isolated from the main condenser, and the RPV pressure is greater than maximum allowable for initiation of a low pressure core cooling system. The generic evaluation documented in-Section 4.2.2 of the GE Topical Report NEDC-31984P, " Generic Evaluations of General Electric Boiling Water Reactor Power Uprates," dated July 1991, examined RCIC system pump and drive turbine operational requirements for plant operation with an SRV setpoint increase of about 40 psi above nominal RPV pressure, and concludes that the RCIC system is capable of delivering its design '
flow for RPV pressure increases up to 40 psi above nominal operating pressure. As the PBAPS RPV operating and SRV setpoint pressures will increase by only 30 psi, the evaluation of the PBAPS RCIC system adequacy under rerated power conditions is consistent with the bases and conclusions in NEDC-31984P.
The RCIC turbine has the capacity to develop the horsepower and speed that will be required by the pump to meet the new RPV pressure conditions. The increase in pump and turbine speed requires a new rated speed of 4,550 rpm. This speed is below the maximum continuous operating speed specified by the pump and turbine manufacturers. The increased turbine rated speed will require that the overspeed trip margin be reduced in order to maintain the maximum trip speed below the limit specified by the manufacturer.
4.
Revise High Pressure Coolant Injection (HPCI) system operating requirements to reflect rerate conditions.
OL, Appendix A TS Section 4.5.C, pg. 129 TS Bases 3.5.C, pg. 137 The HPCI Syrtem was determined to have the capability to deliver its design steady state flow of 5,000 gpm at rerated power conditions.
These conditions are based on a proposed SRV setpoint increase of 30 psi, accounting for the current SRV allowable setpoint tolerance of 11%.
)
Page 6 The HPCI turbine has the capacity to develop the horsepower and speed required by the pump to meet the new RPV pressure requirements. The increase in pump and turbine speed requires a new rated speed of 4,100 rpm. This speed is below the maximum continuous operating speed specified by the pump manufacturer and is below the maximum operating speed used by other similar HPCI turbines. The increased turbine rated speed requires the overspeed trip margin be reduced in order to maintain the maximum trip speed below the limit specified by the manufacturer.
The remainder of the system components were determined not to be impacted significantly by power rerate because of the small increase in operating pressure and/or temperature.
5.
Revise TS to reflect the increase in nominal operating pressure (i.e., 30 psi) at the rerated maximum power level.
OL, Appendix A TS Section 1.0, pg. 2 TS Section 2.2.1.A, pg. 29 TS Table 3.1.1, pg. 37 TS Bases 3.7.A and 4.6.A, pg. 189 As a result of the RPV pressure increase associated with power rerate, the analytical limit for the reactor SCRAM at high RPV pressure will be increased accordingly.
During a pressure increase transient not terminated by a direct SCRAM signal or high reactor core flux SCRAM, the high RPV pressure SCRAM will terminate the transient. The high RPV pressure SCRAM signal settings will be maintained slightly above the RPV maximum normal operating pressure and below the specified analytical limit..This setting permits i
normal operation without causing inadvertent SCRAM, yet provides adequate margin to the marimum allowable RPV pressure.
6.
Revise Standby Liquid Control System (SLCS) operating requirements to reflect rerate conditions.
OL, Appendix A TS Sections 3.4.3 and 4.4.3, pg. 117 The ability of the SLCS boron solution to achieve and maintain safe shutdown is not a direct function of core thermal power. However, due to increased fuel loading for higher power and extended operating cycle, the required concentration of boron may change.
The shutdown capability (i.e., boron concentration) of the SLCS is re-evaluated for each core reload to ensure sufficient shutdown margin is available. The SLCS is designed for injection at a maximum RPV pressure equal to the minimum SRV setpoint pressure.
Since the SLCS pumps are positive displacement pumps, a small pressure increase in the SRV setpoint has no effect on the rated SLCS injection flow to the reactor. Therefore, the capability of the SLCS to provide its backup reactor shutdown function is not i
affected by power rerate conditions.
Page 7 7.
Revise the reactor SCRAM bypass value to be based on 30% of the increased rated thermal power.
OL, Appendix A TS Table 3.1.1, pg. 39 TS Bases 3.1, pg. 49 and 50 I
The transient analysis for operation at rerate conditions was performed with the analytical limit raised in accordance with the current basis for this setpoint (e.g., equivalent to 30% of full power). This approach maintains the safety basis for the setpoint and the small increase in the turbine first stage pressure used for the setpoint does not make any significant difference in the transient analysis results.
8.
Revise the containment leak rate testing to reflect rerated conditions (i.e., offsite radiological consequences of a Loss of Coolant Accident (LOCA)).
OL, Appendix A TS Bases 3.7.A and 4.7.A, pg. 193 The impact of power rerate affects the offsite radiological consequences of a LOCA that is discussed in TS Bases 3.7.A.
A re-evaluation of the original analysis was performed. The resultant offsite thyroid dose increased by 19% from 201 rem to 239 rem; however, only about 3% of this increase is due to rerated conditions and 16% is due to reconstitution of the analysis model.
This result continues to preserve adequate margin between expected offsite doses due to a LOCA and 10CFR100 guidelines.
9.
Editorial changes a.
Correct spelling of the word "0PERATING" on OL, Appendix B -
cover sheet.
1 b.
Revise the value of current rated thermal power (i.e., 3293 MWt) to rerated power level (i.e., 3458 MWt).
OL, pg. 5 OL, Appendix A TS Section 1.0, pg. 6 TS Section 2.1.A.1, pg. 9 TS Section 2.1.B, pg. 11 TS Bases 2.1, pg. 17 TS Bases 2.1, pg. 18 l
TS Table 3.1.1, Note 12, pg. 40 TS Table 3.2.C, Note 2, pg. 74 OL, Appendix B ETS Section 1.1, pg. 2
Page 8 c.
Add the proper references to NEDC-32183P, Power Rerate Safety Analysis Report for Peach Bottom 2 & 3," dated May 1993.
OL, Appendix A TS Bases 2.1, pg. 17 TS Bases 2.1, pg. 24 TS Bases 3.5.K, pg. 140a TS Bases 3.5.M, pg. 140c d.
Revise figures to reflect power rerate conditions.
OL, Appendix A TS Figure 1.1-1, pg. 16 TS Figure 3.6.5, pg. 164d e.
Revise TS Bases to reflect the containment venting operability under rerate conditions.
OL, Appendix A TS Bases 3.7.A and 4.7.A, pg. 195 f.
Revise TS Bases to reflect the rescaling of the SRVs and safety valves capacity to the rerate conditions.
OL, Appendix A TS Bases 3.6.D and 4.6.D, pg. 157 Additional discussions of various systems, structures, and components that have been evaluated for rerated conditions but do not involve TS changes, are provided in Attachment 3, NEDC-32183P.
II.
Safety Assessment of the Proposed Chances A.
Increasing the power level of PBAPS, Units 2 and 3, by 5% to 3,458 MWt can be done safely within certain limits.
Several light water reactors have already been rerated worldwide, including 17 Boiling Water Reactors (BWRs) in the United States, Switzerland and Spain.
Most of the original safety analyses, such as the transient (i.e.,
abnormal operating events) analyses, were based on 105% of full steam flow, which coincides with about the same power level as 'is requested for the power rerate of the maximum power for PBAPS, Units 2 and 3.
Those safety analyses exceeded the requirement to perform analyses at 102% of full power by about 2.2%. The power dependent safety analyses summarized below are based on 102% of rerated full power. Therefore, this safety assessment justifies only about an additional 3% power increase over the original analyzed power level.
-l
Page 9 An increase in electrical output of a BWR plant is accomplished primarily by generation and supply of higher steam flow for the turbine generator.
PBAPS, like most BWR plants, as originally licensed, has an as-designed equipment and system capability to accommodate a steam flow rate at least 5% above the original rating.
In addition, continuing improvements in the analytical techniques (i.e., computer codes and data) based on several decades of BWR safety technology, plant performance feedback, and improved fuel and core designs, have resulted in a significant increase in the difference between calculated safety analysis results and the licensing limits. This available safety analysis margin, combined with the excess capability of as-designed equipment, systems and components, provide BWR plants with the potential for an increase in their full power rating of between 5% and 10% without major hardware modifications and with no significant increase in the postulated hazards presented by the plant as approved by the NRC at the original license stage.
NEDC-32183P, " Power Rerate Safety Analysis Report for Peach Bottom 2 & 3," dated May 1993, is provided in Attachment 3 and contains the safety analysis prepared by General Electric (GE) to support the proposed OL changes and implementation of Power Rerate at PBAPS, Units 2 and 3.
This analysis is based on 24 month fuel cycles. The analyses and evaluations supporting power rerate changes were completed using the guidelines in General Electric Topical Report, NEDC-31897P-A, " Generic Guidelines for General Electric Boiling Water Reactor Power Uprate," dated May 1992.
Resolution of generic issues associated with power uprate was addressed in General Electric Topical Report, NEDC-31984P, " Generic Evaluations of General Electric Boiling Water Reactor Power Uprate," dated July 1991, and PBAPS specific evaluations are provided in NEDC-32183P. The NRC reviewed and approved these Topical Reports by letters from the NRC to GE dated September 30, 1991 and July 31, 1992.
Plant performance and responses to hypothetical accidents and transients at PBAPS, Units 2 and 3 have been evaltated for a power rerate of 5% of full power.
This safety assessment summarizes the information provided in NEDC-32183P (i.e., the safety significant plant reactions to events analyzed for licensing the plant and potential effects on various margins of safety).
PBAPS, Units 2 and 3 were originally licensed for operation at 100%
power (i.e., 3293 MWt). The original safety analysis basis was that the reactor would be operating continuously at a porer level at least 1.02 times the licensed power level; however, many of the original analyses had already been performed for the 1056 full steam flow conditions. The rerate power level included in this evaluation is a 5% thermal power rerate.
Therefore, a 53 power rerate is analyzed at a power level at least 1.02 x 1.05 x (original licensed power level), or 1.02 x (rerated power level).
Page 10 FRACTION OF RATED Rerate P_QE!1 POWER ANALYSIS *"
0 1.0 X"
21.02X 5%
1.05 1.05X=Y 21.071X 5%
1.05 Y"
21.02Y Original power level (3293 MWt)
Rerated power level (3458 MWt)
Some analyses are performed at 100% rated power, because the 2% power factor is already accounted for-in the analysis methods.
The above rerate analysis basis assures that the power dependent margins prescribed by the applicable regulations will be maintained by meeting the appropriate regulatory acceptance criteria. NRC accepted computer codes and calculational techniques were used to make the calculations that demonstrate that the stipulated criteria are met.
Similarly, margin specified by application of the pertinent sections of American Society of Mechanical Engineers (ASME) Code and American National Standards Institute (ANSI)
Standard B31.1 design requirements will be maintained, as will other margin assuring criteria used to judge the acceptability of the plant.
Environmental margins will be maintained by not increasing any of the present limits for releases such as residual chloride concentrations in the plant effluents or plant vent radiological limits, as a consequence of power rerate.
No change is required in the basic fuel design to achieve the rerated power levels or to meet the plant licensing limits. The current fuel operating limits will continue to be met at the rerated power level. Analyses for each fuel reload will continue to meet the criteria accepted by the NRC as specific in NED0-24011, "GESTAR II."
The plant's design concept includes a variety of ways to pump water into the reactor vessel to deal with all types of events. -These cooling water sources will maintain core integrity by providing adequate water for core cooling.
Consequently, there are high and low pressure, high and low volume, safety-and nonsafety-related means of delivering water to the reactor vessel. These means include the feedwater and/or condensate pumps, the Low Pressure Coolant Injection / Core Spray (LPCI/CS) system pumps, the HPCI pump, the RCIC pump, the SLC pumps, and the Control Rod Drive (CRD) pumps. Many of these diverse water supply means are redundant in equipment and also redundant in system (e.g., there are several LPCI/CS pumps and complete redundant piping systems).
Power rerate does not result in an increase or decrease in the available water sources, nor does it change the selection of those assumed to function in the safety analyses. NRC approved methods were used for_ analyzing the performance of the Emergency Core Cooling Systems (ECCS) during LOCAs.
i Page 11 Power rerate results in a small (i.e., 5%) increase in reactor decay heat, and thus, the core cooling time to reach cold shutdown conditions is increased. However, the existing cooling capacity can bring the plant to cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Design basis accidents (DBAs) are very low probability hypothetical events whose characteristics and consequences are used in the design of the plant, so that the plant can mitigate their consequences to within acceptable regulatory limits.
For BWR licensing evaluations, capability is demonstrated for coping with the range of hypothetical pipe break sizes in the largest recirculation, steam, and feedwater lines, a postulated break in one of the ECCS lines, and even down to breaks the size of instrument lines. This break range bounds the full spectrum of large and small, high energy line breaks and the success of the plant systems in mitigating these events while accommodating a single active equipment failure in addition to the postulated LOCA.
Several of the most significant licensing assessments are made using these LOCA ground rules. These assessments are as follows.
Challenges to fuel or the ECCS performance analyses in accordance with the requirements and criteria of 10CFR50.46 and 10CFR50, Appendix K, wherein the principal concern is the fuel Peak Clad Temperature (PCT).
Challenges to the containment wherein the primary concerns are the maximum containment pressure calculated during the course of the LOCA and maximum suppression pool temperature for long term cooling in accordance with 10CFR50, Appendix A, General Design Criterion (GDC) 38.
The calculated radiological consequences of DBAs compared to the criteria of 10CFR100.
The ECCS are described in Chapter 6 of the PBAPS Updated Final Safety Analysis Report (UFSAR). As mentioned above, a complete spectrum of pipe breaks is investigated from the largest recirculation line down to instrument lines. The ECCS performance evaluation for rerated conditions was conducted through application of the 10CFR50 Appendix K evaluation models and then showing conformance to the acceptance criteria of 10CFR50.46. Therefore, the ECCS safety margin is not impacted by power rerate.
Analyses provide the results of the plant containment response to various hypothetical LOCAs.
For power rerate conditions, the containment response will increase slightly. However, containment pressures and temperatures following any DBA, have been evaluated to be acceptable. Thus, the containment and its cooling systems were determined to be satisfactory for rerated power operation.
The PBAPS UFSAR provides the radiological consequences for each of the DBAs.
These events have been re-evaluated using the methodology described in UFSAR Subsection 14.9, and the results j
remain below the 10CFR100 guidelines values.
1
Page 12 The effects of plant transients were evaluated by investigating a number of disturbances of process variables (e.g., temperature, pressure, flow), and malfunctions or failures of equipment according to a scheme of postulating initiating events. These events are primarily evaluated against the Safety Limit Minimum Critical Power Ratio (SLMCPR). The SLMCPR is determined using NRC approved methods. The Operating Limit MCPR will be increased very slightly to assure that the SLMCPR is maintained, when a transient is initiated from the rerated power level. Therefore, the margin of safety will not be affected by operating at power rerate conditions.
All of the other radiological releases previously evaluated are either unchanged because they are not power dependent, or increase by at most the percent increase in power. The dose consequences for all of the non-LOCA radiological release accident events'are bounded by the radiological release events discussed above and remain within the regulatory limits for each release.
Plant equipment and instrumentation have been evaluated against the i
criteria in the current equipment Environmental Qualification (EQ) program at rerated conditions. The majority of equipment was determined to be qualified for the revised environmental conditions.
Equipment that was determined to not remain qualified for the rerated conditions will be replaced prior to operation at the rerated power level.
Other systems and/or equipment used to perform safety-related and normal operation functions have been reviewed relative to rerated conditions in a manner comparable to that for safety-related NSSS systems and/or equipment. This includes, but is not necessarily limited to, all or portions of the Main Steam, Feedwater, Main i
Turbine, Condenser, Condensate, Essential and Non-essential Service Water, Emergency Diesel Generators, Standby Gas Treatment, Fuel Pool Cooling, piping, and supporting systems.
Significant groups and/or types of equipment and/or systems are justified for rerated conditions by bounding evaluations.
Unique evaluations justify power rerate operation for systems and/or equipment that are not generically justified.
Other special events and features such as Station Blackout, Fire Protection and Motor Operated Valves (MOVs) have been evaluated to ensure safe plant operation for rerated conditions.
The PBAPS Station Blackout analyses were reviewed to ensure that operation at rerated conditions does not com?romise the requirements of 10CFR50.63 and associated co..tments. The effects of operation at rerated conditions were evalt ied for areas required to mitigate the Station Blackout event. Pertinent equipment was shown to remain operable and have adequate capacity to mitigate for the event ensuring safe shutdown under Station Blackout conditions.
i Page 13 The plant fire protection features have been reviewed and there is no impact to the fire detection or suppression systems due to rerated conditions. The fire safe shutdown evaluation was reviewed to evaluate the impact of rerated conditions.
ECCS pump room.
heatup rates are impacted and temperature responses were updated for the rerate effects of power rerate.
Equipment was verified to
]
remain qualified and to have adequate mitigation capacity for rerated conditions.
All valves in the PBAPS MOV program are being evaluated, and the majority have been shown to be adequate for rerated conditions.
The remaining few H0Vs will be reconciled or modified to be adequate for rerated conditions.
B. Information Supporting a Finding of No Significant Hazards Consideration We have concluded that the proposed changes to the PBAPS, Units 2 and 3 Operating Licenses and Appendices A and B, to reflect the planned implementation of the Power Rerate Program and the corresponding increase in the authorized maximum reactor core power level by 5%, do not constitute a significant hazards consideration.
In support of this determination, an evaluation of each of the three (3) standards set forth in 10 CFR 50.92 is provided below.
- 1) The proposed OL changes do not involve a significant increase _ in the probability or consequences of an accident previously evaluated.
The proposed power rerate imposes only minor increases in the plant operating conditions.
Plant systems, components, and structures have 1
been verified to be capable of performing their intended functions under rerated conditions. Where necessary, some components will be modified or replaced prior to implementation of the Power Rerate Program to accommodate the revised operating conditions.
No new component or system interactions that could lead to an accident are created. As discussed below, no transient events result in a new sequence of events which could lead to a new accident scenario.
Anticipated Transients Without Scram (ATWS) Analysis The changes to plant parameters are consistent with the results in NEDC-31984P, " Generic Evaluations of General Electric Boiling Water Reactor Power Uprate," dated July 1991. Therefore, the response to an ATWS event at rerated power will be consistent with the generic response and is acceptable.
ECCS-LOCA Analysis The current ECCS-LOCA performance analysis already bounds the rerated i
power conditions. The peak clad temperature for rerated conditions is 1,516*F which is below the 10CFR50.46 required limit of 2,200*F.
Therefore, the analysis demonstrates that PBAPS, Units 2 and 3 will continue to comply with 10CFR50.46 and 10CFR50, Appendix K at rerated conditions.
Page 14 Abnormal Operating Transient Analysis The results of the evaluation of transients indicate that the margin to the Safety Limit Minimum Critical Power Ratio (SLMCPR) is unchanged for the 8x8 array fuel types such as the GE9 product line currently in the Unit 2 and Unit 3 cores, and will increase by 0.01 for the Gell fuel design. The fuel thermal-mechanical limits at power rerate conditions are within the specific design criteria for the GE fuels currently loaded in the PBAPS Unit 2 Cycle 10 core.
Also, the power-dependent and flow-dependent MCPR and Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits developed as part of the core performance improvement program are applicable to rerated conditions. The peak RPV bottom head pressure is still within the ASME i
requirement for RPV overpressure protection.
The analysis performed focused on the most limiting transient events in each disturbance category selected specifically for the power rerated evaluations.
The results demonstrate that PBAPS Units 2 and 3 core thermal power output can be safely increased to the power rerate level without significant impact on the plant safety during a postulated transient event.
a)
Events Resulting in a Nuclear System Pressure Increase
- 1) Main Generator Load Rejection with No Steam Bypass At rerated conditions, the fuel transient therul and mechanical overpower results remain below the NRC accepted design criteria.
ii) Main Turbine Trip with No Steam Bypass The fuel transient thermal responses are less severe than for the Generator Load Rejection event. Therefore, at rerated conditions, this event remains bounded by the Generator Load Rejection eventg iii) Main Steam Isolation Valve Closure, Flux Scram The peak RPV bottom head pressure for rerated conditions is slightly higher than the RPV bottom head pressure at current rated conditions due to the higher initial system pressure.
However, the resultant pressure is still below the ASME overpressure limit of 1,375 psig by a margin of 68 psi.
b) Events Resulting in a Reactor Vessel Water Temperature Decrease 1)
Inadvertent HPCI Actuation For the condition analyzed, both the high water level setpoint and the high RPV steam dome pressure SCRAM setpoint are not reached. Based on the peak average fuel surface heat flux results, the HPCI actuation event will be bounded by the limiting pressurization event with respect to delta Critical Power Ratio (ACPR) considerations.
In addition, the fuel
Page 15 transient thermal and mechanical overpower limits remain within the allowable NRC accepted design values.
ii) Feedwater Controller Failure-Maximum Demand The ACPR calculated for this event at rerated conditions is about 0.01 higher than the corresponding value for the current rated power. However, the trend for the Feedwater Controller Failure-Maximum Demand event is consistent with the analysis for the current rated power. This event continues to be the limiting event at the low core flow condition and is bounded by the limiting Generator Load Rejection event. The fuel thermal margin results are within the acceptable limits for the fuel type analyzed.
iii) Loss of Feedwater Heating The ACPR for this event at the rerated conditions is bounded by the result estimated for this event at the current rated power level, and remains significantly less than the cycle operating MCPR limit.
Because of the round-off process, there is no change between the ACPR results for high and low core flow conditions. However, the results at low core flow conditions are actually slightly higher than for the high core flow condition because of increased inlet coolant _subcooling into the reactor core.
The calculated thermal and mechanical overpower limits for this event at power rerate conditions also meet the fuel design criteria.
c)
Event Resulting in a Positive Reactivity Insertion 1)
Rod Withdrawal Error (RWE)
The ACPR calculated for this event at rerated conditions is slightly less than the value for this event at the current rated power and is bounded by the generic RWE limits of 0.13 based on implementation of the APRM-Rod Block Monitor TS (ARTS) changes. Therefore, the generic ARTS-based RWE analysis ACPR result is verified for applicability to PBAPS power rerate conditions.
d)
Event Resulting in a Reactor Vessel Coolant Inventory Decrease 1)
Loss of Feedwater Flow This transient event does not pose any direct threat to the fuel in terms of a power increase from the initial conditions.
However, it is included in the power rerate evaluation to provide assurance that sufficient water make-up capability is available to keep the core covered when all normal feedwater is lost.
Page 16 The generic analysis results in NEDC-31984P, " Generic Evaluations of General Electric Boiling Water Reactor Power Uprate," dated July 1991, show that at )ower rerate conditions, the minimum water level is reduced by a)out 1.5 feet from that previously calculated for current rated power, but a large amount of water, more than 5 feet, remains above the top of the active fuel. The sensed water level outside of the core shroud has also been checked to show adequate operational flexibility exists for setting the Level 1 RPV water level setpoint so that it is not expected to be reached even in the conservative case of a HPCI failure. Therefore, PBAPS, Units 2 and 3 will maintain adequate reactor water level during a postulated Loss of Feedwater Flow event at power rerate conditions.
e) Event Resulting in a Core Coolant Flow Decrease i) Recirculation Pump Seizure The recirculation pump seizure assumes instantaneous stoppage of the pump motor shaft of one recirculation pump. As a result, the core flow decreases rapidly. The RPV water level swell due to the rapid core flow reduction reaches the high RPV water level setpoint, causing a feedwater pump trip, a turbine trip and subsequently a reactor SCRAM on turbine stop valves closure. The peak neutron flux and average fuel surface heat flux do not increase significantly above the initial conditions; therefore, no impact on the fuel thermal margin is postulated to occur.
f)
Event Resulting in a Core Coolant Flow Increase i) Recirculation Flow Controller Failure Increasing Flow The results of this transient for PBAPS, Units 2 and 3 power rerate remain non-limiting as compared with other more severe pressurization events.
g)
Performance Improvements
- 1) Main Turbine Bypass Out-of-Service The main turbine steam bypass out-of-service condition is included in the input assumptions used in the Abnormal Operating Transient Occurrences analyses for power rerate application. The transient analyses results at power rerate conditions reflect the plant response accounting for this condition, ii) Single Loop Operation (SLO)
The safety analysis for rerated conditions shows that the SLO mode is valid for power rerate conditions and remains unchanged-from the current rated power conditions.
Page 17 iii) Final Feedwater Temperature Reduction Final Feedwater Temperature Reduction is a cycle extension mode of operation, used in conjunction with increased core flow (ICF) at the end of a normal operating cycle. The analyses show that for a temperature reduction up to 55'F, this mode of operation is applicable for operation of PBAPS, Units 2 and 3 at the power rerate conditions.
h) Other evaluations These evaluations included the effect of power rerate on the radiological consequences of accidents presented in UFSAR Subsections 5.2, 14.6 and 14.9.
The following bounding analyses were performed: 1) Loss-of-Coolant Accident (LOCA); 2) Main Steam Line Break (MSLB) Accident; 3) Fuel Handling Accident; 4) Control Rod Drop Accident; and 5) Instrument Line Break Accident.
The analyses shows the offsite radiological consequences for the bounding accidents increase, but remain well within the guidelines of 10CFR100 as discussed in the UFSAR Section 14.9 and the NRC Safety Evaluation Reports for PBAPS, Units 2 and 3.
In general, offsite doses are expected to increase proportionally with reactor power. However, a direct comparison between the original and rerate values has limited meaning because the original analyses could not be fully reconstituted.
For the fuel handling accident, control rod drop accident, and instrument line break accident, the offsite doses increase by less than 1 rem.
For the MSLB accident, the whole body dose remains less than 1 rem and the thyroid dose increases by only 3% from 85 rem to 88 rem.
For the LOCA, a re-evaluation of the original analysis was performed.
The resultant thyroid dose increased by 19% from 201 rem to 239 rem; however, only about 3% of the increase is due to rerated conditions and 16%
due to changes in the analysis model reconstitution.
Whole body dose increases slightly to 3.9 rem.
Accident radiological consequences in the Control Room and Technical Support Center (TSC) were also evaluated.
The results show doses are well below the 30-day limit of GDC 19 of Appendix A to 10CFR50 (i.e., 5 rem whole body and 30 rem thyroid). A re-evaluation of the original analysis was performed.
The highest dose consequence is from a main steam line break which results in a dose of 18 rem thyroid compared to 1.5 rem in the UFSAR.
- However, only about 3% of this increase is due to rerated conditions and 16%
is due to analysis model reconstitution. All whole body doses are less than 1 rem.
An evaluation was performed to address the impact of power rerate on accident mitigation features, structures, systems, and components within the balance of plant. The results are as follows.
Page 18 Auxiliary systems such as primary containment chilled water, building Heating, Ventilation, and Air Conditioning (HVAC) systems, reactor building closed loop cooling, service water and emergency service water, high pressure service water, spent fuel pool cooling, process auxiliaries such as instrument air and makeup water and the post-accident sampling system were confirmed to operate acceptably under normal and accident conditions at rerated conditions.
Combustible gas control systems were confirmed to be capable of maintaining oxygen concentrations inside the primary containment within limits under post accident conditions after implementation of the Power Rerate Program.
The secondary containment and standby gas treatment system were confirmed to be able to adequately contain, process, and control the release of normal and post-accidant levels of radioactivity at rerated conditions.
Instrumentation was reviewed and confirmed to be capable of performing its control and monitoring functions under rerated conditions.
Electric power systems including the turbine generator and switchgear components were verified as being capable of providing the electrical load as a result of the rerated power level s.
No safety-related electrical loads were affected which would impact the emergency diesel generators.
Piping systems were evaluated for the effect of operation at higher power levels, including transient loadings. The evaluation confirmed that with few exceptions piping and supports are adequate to accommodate the increased loadings resulting from operation at rerated power conditions.
In a few cases, piping supports will be modified to accept the higher forces due to rerated conditions.
The effect of rerated conditions on high energy line break (HELB) for all NSSS and B0P systems was evaluated. The evaluation confirmed structures, systems, and components important to safety are capable of accommodating the effects of jet impingement and blowdown forces and the environmental effects resulting from HELB events at rerated conditions.
Control room habitability was evaluated. Post-accident Control Room and TSC doses at rerated conditions were confirmed to be l
within the limits of GDC 19 of 10CFR50, Appendix A.
Doses for normal operation at rerated conditions were reviewed and confirmed to remain within the limits of 10CFR20 and i
10CFR50, Appendix 1.
The impact on post-accident sampling _
activities and post-accident access to vital areas was also confirmed to be acceptable.
l
Page 19 The environmental qualification.of equipment important to safety was evaluated for the impact of normal and accident operating conditions at rerated power levels.
The majority of equipment remains qualified for the new conditions.
For equipment not qualified, corrective actions will be taken to ensure the plant equipment will perform their intended functions under rerated conditions. No new equipment will be added for power rerate which would increase the potential for component failure. The Preventative Maintenance Program (PMP) is not power dependent and will continue to provide for equipment repair or replacement at rerated power conditions.
The impact of operation at rerated power levels was evaluated for Station Blackout and fire safe shutdown area heat-up concerns. The evaluation confirmed there is no adverse impact from rerated conditions on the ability of the plant to achieve safe shutdown under these conditions.
The consequences of all transients and special events (i.e., ATWS and Station Blackout) remain within NRC accepted criteria for rerated conditions. Concurrent malfunctions assumed to occur during accidents have been ' accounted for in the safety analyses for rerated conditions. The consequences of these equipment malfunctions do not change with implementation of the Power Rerate Program.
All equipment "Important to Safety" is capable of or will be modified / replaced to be capable of performing its intended function. The availability of redundant systems to provide safety functions in the event of component malfunction is not impacted as a result of rerated conditions.
Furthermore, the impact of power rerate on the consequences of abnormal transients and accident conditions which are a result of component malfunctions has been shown to be acceptable.
The probability (i.e., frequency of occurrence) of DBAs occurring is not affected by the increased power level, as the applicable regulatory criteria established for plant equipment (e.g., ANSI Standard B31.1, ASME code, NRC Regulatory Guides) will still be followed as the plant is operated at the rerated power level.
Reactor SCRAM setpoints will be established such that there is no significant increase in scram frequency due to rerated conditions. No new challenges to safety-related equipment will result from power rerate.
The changes in consequences of hypothetical accidents which would occur from 102% of the rerated power, compared to those previously evaluated, are in all uses not significant, because the accident evaluations from a power reratt to 105% of original rated power will not result in exceeding the applicable NRC approved acceptance limits. The spectrum i
of hypothetical accidents and transients has been investigated, and have been determined to meet the current regulatory criteria for.PBAPS, Units 2 ano 3 at rerated conditions. The offsite doses resulting from DBAs are calcubted to increase only a few percent (i.e., approximately 3%) because of the verated power level and remain below 10CFR100 limits.
In the area of core design, the fuel operating limits will
Page 20 still be met at the rerated power level, and fuel reload analyses will show plant transients meet the criteria accepted by the NRC as specified in NEDO-240ll, "GESTAR II."
Challenges to fuel or ECCS performance were evaluated and shown to still meet the criteria of 10CFR50.46 and 10CFR50, Appendix K.
Challenges to the containment have been evaluated and still meet 10CFR50, Appendix A GDC 38, Long Term Cooling, and GDC 50, Containment.
Radiological Release events have been evaluated and shown to meet the guidelines of 10CFR100. Therefore, the proposed OL changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2) The proposed OL changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
All actions to ensure that safety-related structures, systems, and components will remain within their design allowable values and ensure they can perform their intended functions under rerated conditions will be taken prior to implementation of power rerate.
Power rerate does not increase challenges to or create any new challenge to safety-related equipment or other equipment whose failure could cause an accident. No new equipment is added as a result of implementing the Power Rerate Program which would create the possibility of a new type of accident.
In addition, power rerate does not create any new sequence of events or failure modes that lead to a new type of accident.
No new operating mode, safety-related equipment lineup, accident scenario, or equipment failure mode was identified as resulting from the implementation of the Power Rerate Program. The full spectrum of accident considerations defined in NRC Regulatory Guide 1.70 have been evaluated for rerated conditions and no new or different kind of accident has been identified.
Implementation of the Power Rerate Program uses already-developed technology and applies it within the capabilities of already existing plant equipment in accordance with presently existing regulatory criteria to include applicable NRC approved codes, standards, and methods. GE has designed BWRs of higher power levels than the rerated power of any of the currently operating BWR fleet and no new power dependent accidents have been identified.
Therefore, the proposed OL changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3) The proposed OL changes do not involve a significant reduction in a margin of safety.
Power rerate will not involve a significant reduction in a margin of safety, as plant equipment and reactions to transients and hypothetical accidents will not result in exceeding the presently approved NRC acceptance limits.
{
Page 21
+
For systems addressed in the TS Sections 2.1, 2.2, 3.1, 3.2, 3.4, 3.5, j
3.6, and 3.7 (i.e., RPS, Protective Instrumentation, SLCS, HPCI, RCIC, Primary System Boundary and Containment Systems) all components will be operable and capable of performing their intended functions under power-rerate conditions such that the existing margin of safety is not impacted.
For TS Bases 3.7.A and 4.7.A, the impact of rerated conditions affects LOCA offsite radiological consequences discussed in that section. A re-evaluation of the original analysis was performed. The resultant offsite thyroid dose increased by 19% from 201 rem to 239 rem; however, only about 3% of the increase is due to rerated conditions and 16% is due to the analysis model reconstitution. This preserves adequate margin between expected offsite doses and 10CFR100 guidelines.
The events (i.e., transients and accidents) that form the TS Bases (e.g. TS Bases 2.1, 3.1) were evaluated for rerated conditions.
Although some changes to the TS are required for power rerate, no NRC acceptance limit will-be exceeded. Therefore, the margins of safety to the safety limits and other TS limits will be maintained.
Therefore, the proposed OL changes do not involve a significant reduction in a margin of safety.
III.
Information Sucoortina an Environmental Assessment An environmental assessment is not required for the changes proposed by this OL Change Request because the requested changes conform to the criteria for " actions eligible for categorical exclusion," as specified in 10CFR51.22(c)(9). The requested changes will have no impact on the environment.
The non-radiological environmental impacts of the proposed power rerate were reviewed based on the information submitted in the Environmental Report, Operating License Stage (ER/0L), the NRC Final Environmental Statement (FES), OL Appendix B (i.e., the Environmental Technical Specifications) and the requirements of the National Pollutant Discharge Elimination System (NPDES) permit, which includes the outfall limits.
Based on this review, the proposed power rerate will have insignificant impacts on the non-radiological elements of concern and the plant will be operated in an environmentally acceptable manner as established by the FES.
Existing Federal, State and Local regulatory permits presently in effect will accommodate power rerate conditions without modification, except for the NPDES permit which will require modification. This NPDES permit includes a tabulation indicating how many functioning cooling towers are required at various power levels and other relevant variables. This permit will be amended to reflect the higher power level of the units. There are essentially no impacts to air, water, and land resources.
The proposed changes do not involve a significant hazards consideration as discussed in the preceding section. The proposed changes do not involve a significant change in the types or significant increase in the amounts of any effluents that may be released offsite.
In addition, the proposed changes do not involve a significant increase in individual or cumulative occupational radiation exposure.
Page 22 IV.
Conclusion The Plant Operations Review Comittee and the Nuclear Review Board have reviewed these proposed changes to the PBAPS Units 2 and 3 Operating-License Nos. DPR-44 and DPR-56 and Appendices A and B thereto and have concluded that they do involve an unreviewed safety. question, but that they do not involve significant hazards consideration, and will not endanger the health and safety of the public.
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