ML21238A035
| ML21238A035 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 09/07/2021 |
| From: | James Kim Plant Licensing Branch 1 |
| To: | Carr E Public Service Enterprise Group |
| Kim J | |
| References | |
| EPID L-2020-LLA-0210 | |
| Download: ML21238A035 (28) | |
Text
OFFICIAL USE ONLY - PROPRIETARY INFORMATION
September 7, 2021
Mr. Eric Carr President and Chief Nuclear Officer PSEG Nuclear LLC - N09 Hope Creek Generating Station P.O. Box 236 Hancocks Bridge, NJ 08038
SUBJECT:
HOPE CREEK GENERATING STATION CORRECTION TO SAFETY EVALUATION FOR AMENDMENT NO. 229 RE: REVISE LOW PRESSURE SAFETY LIMIT TO ADDRESS GENERAL ELECTRIC NUCLEAR ENERGY PART-21 SAFETY COMMUNICATION SC05-03 (EPID L-2020-LLA-0210)
Dear Mr. Carr:
By letter dated August 17, 2021 (Agencywide Documents and Access System (ADAMS)
Accession No. ML21181A056), the U.S. Nuclear Regulatory Commission (NRC, the Commission) issued the Amendment No. 229 to Renewed Facility Operating License No. NPF-57 for the Hope Creek Generating Station. The amendment revised Technical Specification 2.1, SAFETY LIMITS, specifical ly, Safety Limits 2.1.1, THERMAL POWER, Low Pressure or Low Flow, and 2.1.2, THER MAL POWER, High Pressure and High Flow, to reduce the reactor vessel steam dome pressure value to address General Electric Nuclear Energy 10 CFR Part 21 Safety Communication SC05-03, 10 CFR [Part] 21 Reportable Condition Notification: Potential to Exceed Low Pressure Technical Specification Safety Limit, dated March 29, 2005, regarding the potential to violate the low pressure safety limit following a pressure regulator failure-open transient.
Subsequent to the issuance of the amendment, the licensee identified concerns in certain sections of the safety evaluation (SE) issued with the amendment that may, as written, impact licensees ability to implement the amendmen t. The licensee also identified some minor errors/oversights. Upon further review, the NRC staff determined that some statements in the SE could potentially be misinterpreted and thus should be clarified. The clarifications do not change any of the conclusions associated with the issuance of Amendment No. 229 and do not affect the associated notice to the public. Enclosed are the corrected proprietary and non-proprietary SEs. The revised pages contain marginal lines indicating the areas of change.
Clarifications and corrections were made on pages 4, 5, 7, 8, 17, 18, 19, 20, 21, and Reference 6 on page 23 of the SE.
Enclosure 1 to this letter contains proprietary information. When separated from Enclosure 1, this document is DECONTROLLED.
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E. Carr Please replace the NRC Safety Evaluation to Amendment No. 229 for Hope Creek Generating Station with the enclosed corrected safety evaluation.
If you have any questions, please contact me at 301-415-4125 or James.Kim @nrc.gov.
Sincerely,
/RA/
James S. Kim, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Docket No. 50-354
Enclosures:
- 1. Corrected Safety Evaluation (Proprietary)
- 2. Corrected Safety Evaluation (Non-Proprietary)
cc without Enclosure 1: Listserv
OFFICIAL USE ONLY PROPRIETARY INFORMATION ENCLOSURE 2
(NON-PROPRIETARY)
CORRECTED SAFETY EVALUATION
RELATED TO AMENDMENT NO. 229
TO RENEWED FACILITY OPERATING LICENSE NO. NPF-57
HOPE CREEK GENERATING STATION
DOCKET NO. 50-354
Proprietary information has been redacted from this document pursuant to Section 2.390 of Title 10 of the Code of Federal Regulations.
Redacted information is identified by blank space enclosed within ((double brackets)).
OFFICIAL USE ONLY PROPRIETARY INFORMATION
CORRECTED SAFETY EVALUATION BY THE OFFICE OF
NUCLEAR REACTOR REGULATION
RELATED TO AMENDMENT NO. 229
TO RENEWED FACILITY OPERATING LICENSE NO. NPF-57
HOPE CREEK GENERATING STATION
DOCKET NO. 50-354
1.0 INTRODUCTION
By letter dated September 24, 2020 (Reference 1) with enclosure (Reference 2), as supplemented by letters dated April 29, 2021 (Reference 3) with enclosure (Reference 4) and May 27, 2021 (Reference 5) with enclosure (Reference 6), PSEG Nuclear LLC (PSEG, the licensee) submitted a license amendment reques t for the Hope Creek Generating Station (Hope Creek). The amendment would revise Technical Specification (TS) Safety Limit (SL) 2.1.1, THERMAL POWER, Low Pressure or Low Flow, and SL 2.1.2, THERMAL POWER, High Pressure and High Flow. The proposed changes would address the issue identified in a notification pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Part 21, in General Electric (GE) Safety Communication (SC)05-03 (Reference 7). Specifically, the proposed changes would reduce the low steam dome pressure safety limit (LPSL) specified in SLs 2.1.1 and 2.1.2 from 785 pounds per square inch gauge (psig) to 585 psig. As described in GE SC05-03, the issue concerns a potential to momentarily violate this limit during a Pressure Regulator Failure Maximum Demand - Open (PRFO) transient event. The PRFO event is an analyzed transient in Chapter 15 of the Hope Creek Updated Final Safety Analysis Report (UFSAR) (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20142A521). The GE-Hitachi (GEH) report NEDC-33928P, Revision 0 (Reference 2),
enclosed with Reference 1, refers to the LPSL as the Thermal Power Safety Limit (TPSL)
Pressure Boundary (TPSLPB). The licensee stated that LPSL and TPSLPB describe the same parameter and are interchangeable. Consistent with NEDC-33928P (Reference 2), this Safety Evaluation (SE) uses the abbreviation TPSLPB instead of LPSL.
The supplemental letters dated April 29, 2021, and May 27, 2021, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC, the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on December 1, 2020 (85 FR 77264).
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2.0 REGULATORY EVALUATION
2.1 Description of Proposed Changes
The licensee proposed to revise TS 2.1, Safety Limits, to lower the TPSLPB from 785 psig to 585 psig. The proposed changes are as follows (deletions in strike-out; additions in underline):
2.1 SAFETY LIMITS
THERMAL POWER, Low Pressure or Low Flow
2.1.1 THERMAL POWER shall not exceed 24% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 585 psig or core flow less than 10%
of rated flow.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
ACTION:
With THERMAL POWER exceeding 24% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 585 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.
THERMAL POWER, High Pressure and High Flow
2.1.2 With reactor steam dome pressure greater than 785 585 psig and core flow greater than 10% of rated flow:
The MINIMUM CRITICAL POWER RATIO (MCPR) shall be > 1.07.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
ACTION:
With reactor steam dome pressure greater than 785 585 psig or core flow greater than 10% of rated flow and the MCPR below the value for the fuel stated in LCO 2.1.2, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.
2.2 Applicable Regulatory Requirements and Guidance
As discussed in the Hope Creek UFSAR, Section 3.1, Conformance with the NRC General Design Criteria, for each of the 64 General Design Criteria (GDCs) in 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, a specific assessment of the plant design has been made. The Hope Creek UFSAR Section 3.1 also provides a list of UFSAR sections that include further information related to each GDC.
The NRC staff considered the following regulator y requirements and guidanc e during its review of the application:
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The Atomic Energy Act of 1954, as amended, Section 182a requires applicants for nuclear power plant operating licenses to include TSs as part of the license. The TSs ensure the operational capability of structures, systems, and components that are required to protect the health and safety of the public. The regulatory requirements related to the content of the TSs are contained in 10 CFR 50.36.
Safety limits are described in 10 CFR 50.36(c)(1)(i)(A) as follows:
Safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity. If any safety limit is exceeded, the reactor must be shut down. The licensee shall notify the Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence. Operation must not be resumed until authorized by the Commission.
Compliance with GDC 10, Reactor design, is achieved by preventing the violation of fuel design limits. GDC 10 states:
The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition, provides guidance on the acceptability of the reactivity control systems, the reactor core, and fuel system design. Specifically, Section 4.2, Fuel System Design (Reference 19), provides assurance that the fuel system is not damaged as a result of normal operation and anticipated operational occurrences (AOOs) and meets the specified acceptable fuel design limits (SAFDLs). Section 4.4, Thermal and Hydraulic Design (Reference 8), provides guidance on the review of thermal-hydraulic design in meeting the requirement of GDC 10 and the fuel design criteria of Section 4.2. Acceptable approaches to meeting these criteria are as follows:
A. For departure from nucleate boiling ratio (DNBR), CHFR [critical heat flux ratio] or CPR [critical power ratio] correlations, there should be a 95-percent probability at the 95-percent confidence level that the hot rod in the core does not experience a DNB [departure from nucleate boiling] or boiling transition condition during normal operation or AOOs.
B. The limiting (minimum) value of DNBR, CHFR, or CPR correlations is to be established such that at least 99.9 percent of the fuel rods in the core will not experience a DNB or boiling transition during normal operation or AOOs.
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3.0 TECHNICAL EVALUATION
3.1 Background
The PRFO transient event is an AOO described in Hope Creek UFSAR Section 15.1.3. During this event, failure of the reactor pressure regulator in the open position would lead to a loss of reactor pressure control. In this condition, the turbine control valve (TCV) could be in a fully open position and the turbine bypass valves (TBVs) could be in a partially open position until the maximum flow is established by the maximum combined flow limiter (MCFL).
In GE SC05-03 (Reference 7), by using NRC-approved methods, GE identified that during a PRFO event, there could potentially be a momentary decrease in the reactor steam dome pressure to below 785 psig while the thermal power is greater than 24 percent of the rated thermal power (RTP) specified in the current TS 2.1.1. Therefore, a PRFO event could potentially cause a violation of the current TS 2.1.1.
To resolve this issue, the licensee proposed to lower the TPSLPB from its current value of 785 psig to 585 psig. The licensee stated that revising the TPSLPB in SLs 2.1.1 and 2.1.2 to 585 psig would resolve the issue identified in GE SC05-03 regarding the potential to violate the SL during a PRFO transient. To confirm the resolution of this issue, the licensee evaluated the following:
The use of the GEXL14 and GEXL17 correlations for the GE14 and GNF2 fuels, respectively, up to the proposed TPSLPB of 585 psig.
Normal plant operation and AOOs to confirm that the PRFO event is limiting with respect to challenging the TPSLPB for low pressure or low core flow.
The PRFO event to confirm that exceeding the proposed TPSLPB of 585 psig for low pressure or low core flow will be avoided.
GEXL Correlations
In boiling-water reactors (BWRs) such as Hope Creek, the thermal-hydraulic conditions resulting in a DNB have been used as limiting conditions to avoid entering a region where fuel damage could occur. Although it is recognized that a DNB would not necessarily result in damage to the BWR fuel rods, the critical heat flux (CHF) at which the onset of transition boiling (OTB) is calculated to occur has been adopted as a limiting condition. GE has developed a critical steam quality versus boiling length correlation, termed the GEXL correlation, for accurately predicting the OTB in BWR fuel assemblies during both steady-state operation and reactor transient conditions. NEDO-32851P-A, Revision 5 (Reference 9) and (Reference 10), Section 5.4 defines the following 6 input parameters to the GEXL correlation for the calculation of bundle critical power: boiling length, thermal diameter, mass flux, system pressure, R-factor, and annular flow length. The GEXL correlation is an integral part of the transient analysis used to determine the MCPR operating limits resulting fr om transient analysis, the MCPR safety limit analysis, and core operating performance and design. As stated in NEDO-32851P-A, the development of the correlation was based on test data from full-scale simulations of 7x7, 8x8, 9x9, and 10x10 fuel assemblies. The tests were performed to demonstrate that the correlation
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can be used to predict the OTB during postulated transient conditions that are analyzed in the safety analysis.
The GEXL correlations for the 10x10 GE14, GNF2, and GNF3 fuel designs, denominated as GEXL14, GEXL17, and GEXL21, are provided in NEDO-32851P-A, NEDC-33292P (Reference
- 11) and (Reference 12), and NEDC-33880P (Reference 13) and (Reference 14), respectively.
In the core reload design analysis, these correlations are used in determining the thermal margin for the operating cycle. In the safety analysis, these correlations are used in determining the change in CPR during postulated transients and an acceptable MCPR safety limit (SLMCPR).
As stated in NEDO-32851P-A, Section 7, to facilitate the statistical evaluation of the predictive capability of the GEXL correlation, the concept of an experimental critical power ratio (ECPR) is used, which is determined from the following relationship:
ECPR = Predicted Critical Power / Measured Critical Power
As stated in NEDO-32851P-A, Section 5.4.5, the R-factor, an input parameter to the GEXL correlation, accounts for the effects of the fuel rod power distributions as well as the fuel assembly local spacer and lattice critical power c haracteristics. Its formulation for a given fuel rod location depends on the power of that fuel rod, as well as the power of the surrounding fuel rods. In addition, there is an additive constant applied to each fuel rod location ((
))
3.2 Licensees Evaluation of the GEXL14 and GEXL17 Correlations to Justify their Application up to TPSLPB of 585 psig ( 600 pounds per square inch absolute (psia)).
NEDO-32851P-A, NEDC-33292P, and NEDC-33880P provide the following application range for the GEXL correlations based on pressure as determined from their test data:
GEXL14 correlation for GE14 fuel - (( )) (NEDC-32851P-A, Table 2)
GEXL17 correlation for GNF2 fuel - (( )) (NEDC-33292P, Table 5-4)
GEXL21 correlation for GNF3 fuel - (( )) (NEDC-33880P, Table 5-4)
Since the application of the GEXL14 and GEXL17 correlations does not extend up to the proposed TPSLPB of 585 psig (600 psia), the licensee performed the following to justify the application of these correlations up to the proposed TPSLPB of 585 psig:
(a) Analysis to evaluate the performance of the GEXL14 and GEXL17 correlations against the GNF3 test data to confirm that these correlations can predict the GNF3 critical power down to a pressure of 585 psig (600 psia).
(b) Develop measured critical power trend with pressure for GE14, GNF2, and GNF3 to confirm GEXL14 and GEXL17 can predict critical power down to a pressure of 585 psig (600 psia)
For items (a) and (b) above, the licensee used the following method described in NEDC-33928P, Sections 3.0 and 3.1, and evaluated the performance of the GEXL14 and GEXL17
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correlations against the GNF3 test data to confirm their accuracy to predict critical power from these correlations down to 600 psia:
((
))
The licensee selected (( )) for analyzing the performance of GEXL14 and GEXL17 correlations. These included test points ((
))
In a letter dated April 29, 2021 (References 3 and 4), the licensee provided additional detail to justify its approach to adjust the R-factor additive constants for the GEXL14 and GEXL17 correlations and the data selected to analyze the performance of the resulting correlation at lower pressures.
NRC Staff Conclusion on the Method
The NRC staff determined that the licensees method of evaluating the performance of the adjusted GEXL14 and GEXL17 correlations against the GNF3 test data and confirming their accuracy to predict critical power from these correlations down to 600 psia is acceptable because:
The licensee justified the ((
))
The licensees selection of ((
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))
GEXL14 Results
The licensee provided results for the GEXL14 correlation in the following tables and figures of NEDC-33928P:
Table 3-2 provides the Mean ECPR (Mean (GEXL14 calculation for critical power /
GNF3 critical power data)) obtained using the GEXL14 correlation and GNF3 test data at pressures from 600 psia to 1400 psia.
Figure 3-3 provides the plot of data points of ECPR versus pressure.
Figure 3-4 provides the plots of calculated critical power by the GEXL14 correlation versus the GNF3 test critical power for pressures from 600 psia to 1400 psia.
Figure 3-6 shows the measured critical power as a function of pressure for GE14 fuel at different combinations of G (mass flux) and H (inlet subcooling).
Figure 3-7 shows the GE14, GNF2, and GNF3 fuels available test data trends with respect to decreasing pressure.
Table 3-2 in NEDC-33928P shows that ((
)) Therefore, GEXL14 can predict the critical power of GE14 fuel bundles down to 600 psia ((
)) Since the GEXL14 correlation slightly overpredicts GNF3 data at 600 psia compared to the overall GEXL14 statistics, the licensee provided that the remaining GE14 fuel bundles at the current cycle (cycle 23) of Hope Creek are 4 th and 5th cycle bundles located on the outer edge of the core with high exposure and large margin to the operating limit minimum critical power ratio (OLMCPR). These GE14 fuel bundles are most likely to be discharged in the next cycle (cycle
- 24) or they will reside on the periphery or other low power locations. Therefore, the GE14 fuel bundles at the current cycle or future cycles w ill have significant margin preventing them from being the limiting bundles.
The NRC staff determined that the GEXL14 correlation with adjusted R-factor additive constants can predict the CPR down to the proposed TPSLPB of 585 psig (600 psia) for the GNF3 fuel because the licensee adjusted the additive constants using an acceptable method and addressed the slight overprediction. Therefore, the staff concludes that the GEXL14 correlation is acceptable for predicting the CPR for the GE14 fuel down to the proposed TPSLPB of 585 psig (600 psia).
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GEXL17 Results
The licensee provided results for the GEXL17 correlation in the following tables and figures of NEDC-33928P:
Table 3-1 provides the Mean ECPR (Mean (GEXL17 calculation for critical power /
GNF3 critical power data)) obtained using the GEXL17 correlation and GNF3 test data at pressures from 600 psia to 1400 psia.
Figure 3-1 provides the plot of data points of ECPR versus pressure.
Figure 3-2 provides the plots of calculated critical power by the GEXL17 correlation versus the GNF3 test critical power for pressures from 600 psia to 1400 psia.
Figure 3-5 shows the measured critical power as a function of pressure for GNF2 fuel at different combinations of G (mass flux) and H (inlet subcooling).
Figure 3-7 shows the GE14, GNF2, and GNF3 fuels available test data trends with respect to decreasing pressure.
Table 3-1 in NEDC-33928P show that ((
)) Therefore, GEXL17 can predict the critical power of GNF2 fuel bundles down to 600 psia with (( ))
The NRC staff determined that the GEXL17 correlation with adjusted R-factor additive constants can predict the CPR down to the proposed TPSLPB of 585 psig (600 psia) for the GNF3 fuel because the licensee adjusted the additive constants using an acceptable method. Therefore, the staff concludes that the GEXL17 correlation is acceptable for predicting the CPR for the GNF2 fuel down to the proposed TPSLPB of 585 psig (600 psia).
3.3 Evaluation of Normal Plant Operation and AOOs
Normal Plant Operation
In NEDC-33928P, Section 4.1, for normal plant operation the licensee states:
During reactor startup, normal pressure control is established via the main turbine Electro-Hydraulic Control (EHC) system prior to power reaching the TPSL. Once established, three identical pressure regulators within EHC are provided to maintain primary system pressure control. They independently sense pressure just upstream of the main turbine stop valves and use the pressure to control the position of the TCVs.
The pressure is controlled well above the LPIS [low pressure isolation setpoint] and the TPSLPB. With the pressure regulator system operating properly there is no possibility that pressure would reduce below the TPSLPB.
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event. If a reactor high-water level (L8) setpoint is reached first due to item (b), then the turbine trip and reactor scram would occur. The licensee determined that the ((
))
3.4.1 Analysis
The licensee used the NRC-approved TRACG04 methodology for transient analysis documented in NEDE-32906P-A, Revision 3 (Reference 15) and (Reference 16) and in NEDE-32906P, Supplement 3-A, Revision 1 (Reference 17) and (Reference 18).
NEDC-33928P, Table 2-1 provides the key values of design inputs. NEDC-33928P, Table 2-2 provides the major assumptions, which are:
- 1. ((
))
- 2. To calculate the simulated thermal power (STP), the licensee conservatively used the thermal power scram time constant of 6.6 seconds as an input. The licensee stated that the STP time constant used is (( )).
Therefore, the use of the STP time constant is conservative for the purpose of the PRFO analysis.
The licensee provided additional information regarding the development and use of the thermal power scram time constant via letter dated May 27, 2021. Specifically, the licensee provided that:
The definition of the plant STP time constant is consistent with its classical definition.
The parameter in consideration for the time constant is the fuel surface heat flux which represents the time delay between the neutron flux response and the fuel surface heat flux response.
In the PRFO analysis, the time constant provides a conservative approximation of the fuel surface heat flux based upon the neutron flux measured by the APRMs.
The licensee demonstrated that the time consta nt value utilized for the PRFO analysis is conservative in predicting the fuel surface heat flux.
Other Considerations
The licensee stated that the key parameters affecting the minimum reactor dome pressure are in the categories of operating parameters and the parameters related to plant configuration.
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The input parameters related to plant operation are initial core power, core flow, cycle exposure, feedwater temperature, and main steam flow. For conservative analysis the licensee used ((
))
The input parameters related to plant configuration are LPIS, main steam line (MSL) pressure drop, MSIV closure time, and delay time of the pressure sensor. For conservative analysis, the licensee used the ((
))
Additional considerations for conservative analysis included ((
)) because it will lead to a higher main steam flow resulting in a greater steam dome pressure drop during the event.
The licensee analyzed the PRFO event for the following:
((
))
The licensee used ((
)) The licensee stated that the ((
))
3.4.2 Results
NEDC-33928P, Table 5-2 tabulates the PRFO analysis results for all cases analyzed ((
)) NEDC-33928P, Figure 5-1 presents the reactor dome and turbine inlet pressure responses for the initial thermal power of 55 percent and 85 percent of RTP cases.
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The key results to be examined for the reactor dome pressure response are at the lower end of the initial power level, which are as follows:
((
))
NEDC-33928P, Figure 5-2 presents graphs of ((
] The licensee stated that the reactor scram occurred at about 7.5 and 9.5 seconds (( )) respectively, at which point the CPR has greatly increased.
NEDC-33928P, Table 5-3 shows that ((
)) The licensee therefore stated that the effect of uncertainty ((
)) is not significant because there is a significantly large SLMCPR margin during the PRFO event.
The NRC staff determined that the results presented in NEDC-33928P, Tables 5-2 and 5-3, and Figures 5-1 and 5-2 are acceptable because of the following:
The minimum dome pressure stays above the proposed value of TPSLPB (585 psig) ((
)) with significant margin for the conservatively analyzed ((
))
When the thermal power is equal to or less than the TPSL (24 percent of RTP), the proposed TPSLPB (585 psig) does not apply.
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The PRFO event does not affect the fuel integrity as the CPR responses indicate significantly increased margin in CPR as the reactor dome pressure decreases.
Table 2 of NEDC-33743P recommends a ((
)) The licensees analysis results show significant margin in the minimum reactor dome pressure during the analyzed PRFO event.
Based on the technical evaluations presented above, the NRC staff concludes the followings:
The GEXL14 and GEXL17 correlations can predict the critical power trend with pressure of the GE14 and GNF2 fuels, respectively, to the proposed TPSLPB of 585 psig.
The proposed TPSLPB is not challenged during normal plant operation.
The evaluation of AOOs shows that the PRFO event is not only the limiting transient but the only AOO that can credibly challenge the TPSLPB.
The PRFO event evaluation results show that the MCPR increases substantially as the power and pressure decrease and, therefore, is a non-limiting event for the fuel integrity.
The PRFO event evaluation confirmed that the lowest reactor dome pressure reached during this event is greater than (bounded by) the proposed TPSLPB of 585 psig with a significant margin while the STP is 24 percent of RTP or greater.
3.5 Technical Conclusions
With respect to the proposed change of the TPSLPB from 785 psig to 585 psig in Hope Creek SL 2.1.1 and SL 2.1.2, the NRC staff concludes as follows:
For the AOOs described in the Hope Creek UFSAR, including the PRFO transient, the GEXL14 and GEXL17 correlations can be used to predict the critical power versus pressure trend down to the TPSLPB of 585 psig for the GE14 and GNF2 fuels, respectively.
The issue reported in GE SC05-03 regarding the potential to violate the SLs during a PRFO event is resolved for Hope Creek. This event is non-limiting for fuel cladding integrity and the proposed change will have no negative impact on the MCPR.
The SAFDLS are not exceeded during normal plant operation and AOOs and the fuel cladding integrity is maintained.
The NRC staff evaluated the proposed change against the applicable regulatory requirements and acceptance criteria and concludes that as long as the reactor pressure and core flow are within the revised range of the approved GEXL14 and GEXL17 correlations, the proposed TPSLPB changes in Hope Creek SL 2.1.1 and SL 2.1.2 will continue to ensure that the MCPR of the fuel rods in the core is within the required limit. Therefore, as required by GDC 10 and 10 CFR 50.36(c)(1)(i)(A), the criteria of the SLs are met and, thus, the proposed amendment is acceptable.
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4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the New Jersey State official was notified of the proposed issuance of the amendment on July 2, 2021. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, as published in the Federal Register (85 FR 77264; December 1, 2020), and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for catego rical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that 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.
7.0 REFERENCE 1 PSEG Nuclear LLC, letter LR-N20-0023 to U. S. Nuclear Regulatory Commission (NRC),
"Hope Creek Generating Station - License Amendment Request: Revise Hope Creek Generating Station Low Pressure Safety Limit to Address General Electric Nuclear Energy Part 21 Safety Communication SC05-03," September 24, 2020 (ADAMS Accession No. ML20272A063).
2 PSEG Nuclear LLC, Enclosure 6 of letter LR-N20-0023 to U. S. Nuclear Regulatory Commission (NRC), "GE-Hitachi Nuclear Energy (GEH) Report NEDC-33928P, SC05-03 Evaluation for Hope Creek Generating Station, Revision 0," September 24, 2020 (ADAMS Accession No. ML20272A064, Proprietary - Non-Public).
3 PSEG Nuclear LLC, letter LR-N21-0018 to U. S. Nuclear Regulatory Commission (NRC),
"Hope Creek Generating Station - Response to Requests for Additional Information SNSB-RAI 2 and SNSB RAI 3 Re: License Amendment Request to Revise Low Pressure Safety Limit to Address General Electric Part 21 Safety Communication," April 29, 2021 (ADAMS Accession No. ML21119A367).
4 PSEG Nuclear LLC, Enclosure 3 of letter LR-N21-0018 to U. S. Nuclear Regulatory Commission (NRC), "Response to Questions RAI 2 and RAI 3 in Request for Additional Information by Nuclear Systems Performance Branch on Changes in Technical Specification 2.1.1 Due to General Electric Safety Communication SC05-03 for Hope Creek
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION
Generating Station," April 29, 2021 (ADAMS Accession No. ML21119A368 - Proprietary, Non-Public).
5 PSEG Nuclear LLC, letter LRN-21-0040 to U. S. Nuclear Regulatory Commission (NRC),
"Hope Creek Generating Station - Response to Request for Additional Information SNSB-RAI 1 Re: License Amendment Request to Revise Low Pressure Safety Limit to Address General Electric Part 21 Safety Communication," May 27, 2021 (ADAMS Accession No. ML21147A110).
6 PSEG Nuclear LLC, Enclosure 3 of letter LR-N21-0040 to U. S. Nuclear Regulatory Commission (NRC), "SC05-03 Evaluation for H ope Creek Generating Station - Response to SNSB-RAI 1," May 27, 2021 (ADAMS Accession No. ML21147A111 - Proprietary, Non-Public).
7 GE Energy-Nuclear, General Electric Company, letter MFN-05-21 to U. S. Nuclear Regulatory Commission (NRC), "10CFR21 Reportable Condition Notification: Potential to Exceed Low Pressure Technical Specifications Safety Limit - U.S. Nuclear Regulatory Commission Operations Center Event Report," March 29, 2005 (ADAMS Accession No. ML050950428.
8 U. S. Nuclear Regulatory Commission (NRC), "NUREG-0800, Standard Review Plan (SRP), Section 4.4 - Thermal and Hydraulic Design," March 2007 (ADAMS Accession No. ML070550060).
9 Global Nuclear Fuel, "NEDO-32851-A, Revision 5 - GEXL14 Correlation for GE14 Fuel,"
April 2011 (ADAMS Accession No. ML111290532).
10 Global Nuclear Fuel, "NEDC-32851P-A, Revision 5 - GEXL14 Correlation for GE14 Fuel,"
April 2011 (ADAMS Accession No. ML111290535 - Proprietary, Non-Public).
11 Global Nuclear Fuel, letter and Report to U. S. Nuclear Regulatory Commission (NRC),
"MFN-09-436 - GNF2 Advantage Generic Compliance with NEDE-24011-P-A (GESTAR II),
NEDC-33270P, Revision 2 an GEXL Correlation for GNF2 Fuel, NEDG-33292P, Revision 3," June 2009 (ADAMS Accession Nos. ML091830614 and ML091830624).
12 Global Nuclear Fuel, Enclosure 4 of letter to U. S. Nuclear Regulatory Commission (NRC),
"MFN 09-436 - GEXL17 Correlation for GNF2 Fuel NEDC-33292P. Revision 3," June 2009 (ADAMS Accession No. ML091830641 - Proprietary, Non-Public).
13 Global Nuclear Fuel, letter and Report M170253 to U. S. Nuclear Regulatory Commission (NRC), "Revision 1 of the GEXL21 Correlation for GNF3 Fuel NEDC-33880P," November 7, 2017 (ADAMS Accession Nos. ML17311A130 and ML17311A132).
14 Global Nuclear Fuel, Enclosure 1 of M170253 letter to U. S. Nuclear Regulatory Commission (NRC), "GEXL21 Correlation for GNF3 Fuel, NEDC-33880P Revision 1,"
November 2017 (ADAMS Accession No. ML17311A131 - Proprietary, Non-Public).
15 GE Energy, Letter MFN 06-327 and Report to U. S. Nuclear Regulatory Commission (NRC),
"Accptance Version of TRACG Application for Anticipated Operational Occurrences Analyses, NEDE-32906P, Revision," September 25, 2006 (ADAMS Accession Nos.
MML062720165 and ML062720174).
16 General Electric, Enclosure 1, Part 1 and Part 2 of letter MFN 06-327 to U. S. Nuclear Regulatory Commission (NRC), "GE Lic ensing Topical Report NEDE-32906P-A, Revision 3, TRACG Application for Anticipated Operational Occurrences (AOO) Transient Analyses,"
September 2006 (ADAMS Accession Nos. ML062720300 and ML062720309 - Proprietary Information, Non-Public).
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION
17 GE Hitachi Nuclear Energy, letter MFN 10-140 and Report to U. S. Nuclear Regulatory Commission (NRC), "Accepted Version of NEDE-32906P, Supplement 3 - MIgration to TRACG04 / PANAC11 from TRACG02 / PANAC10 for TRACG AOO and ATWS Overpressure Transients," April 16, 2010 (ADAMS Accession Nos. ML110970402 and ML110970406).
18 GE Hitachi Nuclear Energy, Enclosure 1 of letter MFN 10-140 to U. S. Nuclear Regulatory Commission (NRC), "Migration to TRACG04 / PANAC11 from TRACG02 / PANAC10 for TRACG AOO and ATWS Overpressure Transients NEDE-32906P Supplement 3-A, Revision 1," April 2010 (ADAMS Accession No. ML11970407 - Proprietary, Non-Public).
19 U. S. Nuclear Regulatory Commission (NRC), "NUREG-0800, Standard Review Plan (SRP), Section 4.2 - Fuel System Design," March 2007 (ADAMS Accession No. ML070740002).
Principal Contributor: A. Sallman Date: August 17, 2021
OFFICIAL USE ONLY PROPRIETARY INFORMATION
(Package-ML21238A020), (Proprietary - ML21238A009)
(Non-Proprietary - ML21238A035)
OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LAiT NRR/DSS/SNSB/BC NAME JKim KEntz SKrepel DATE 9/02/2021 9/02/2021 9/2/2021 OFFICE NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME JDanna JKim DATE 9/3/2021 9/7/2021