PLA-7236, Response to Request for Additional Information on the Low-Pressure Safety Limit Amendment Request
| ML14268A510 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 09/25/2014 |
| From: | Franke J Susquehanna |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| PLA-7236, TAC MF0410, TAC MF0411 | |
| Download: ML14268A510 (7) | |
Text
SEP 2 5 2 014 Jon A. Franke Site Vice President U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 PPL Susquehanna, LLC 769 Salem Boulevard Berwick , PA 18603 Tel. 570.542.2904 Fax 570.542.1504 jfranke@ pplweb.com SUSQUEHANNA STEAM ELECTRIC STATION RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE LOW-PRESSURE SAFETY LIMIT LICENSE AMENDMENT REQUEST PLA-7236 Docket No 50-387 and No. 50-388
References:
I. PPLLetter (PLA-6I95), "Proposed Amendment No. 312 to Unit I License NPF-I4 and Proposed Amendment No. 284 to Unit 2 License NPF-22: Low Pres s ure Safety Limit and Reference Changes," dated December I9, 2012 (Accession MLI2355A351).
- 2. [SC05-03]
GE Nuclear Part 2I Report Involving Potential to Exceed Low Pressure Technical Specifications Safety Limit, dated March 29 , 2005 (Accession MW50950428).
- 3. NRC Letter, " Request for Additional Information re: License Amendment Request to Change the Low-Pressure Safety Limit (TAC Nos. MF04IO and MF04Il)," dated August 27, 2014 (Accession ML14233A277).
The purpose of this letter is for PPL Susquehanna, LLC (PPL) to provide the requested additional information (RAI). By Reference 1, PPL submitted a License Amendment Request (LAR) for review and approval of a revision to the Susquehanna Steam Electric Station (SSES), Units 1 and 2, Technical Specifications (TS). Specifically, PPL requested a change toTS Section 2.1.1, "Reactor Core SLs [Safety Limits]," to reflect a lower reactor steam dome pressure for Reactor Core SLs 2.1.1.1 and 2.1.1.2. PPL stated that the revision became necessary as a result of Reference 2, the General Electric (GE) 10 CFR 21 Communication, SC05-03, "Potential to Exceed Low Pressure Technical Specification Safety Limit." In Reference 3, the NRC requested additional information.
The response to the RAI questions is in attachment to this letter. There are no new regulatory commitments associated with this response. Document Control Desk PLA-7236 If you have any questions or require additional information, please contact Mr. Duane L. Filchner (570) 542-6501.
Sincerely, ,A.___
Response to Request for Additional Information Copy: NRC Region I Mr. J. Greives, NRC Sr. Resident Inspector Mr. J. Whited, NRC Project Manager Mr. L. Winker, PA DEP/BRP Attachment to PLA-7236 Response to Request for Additional Information Attachment to PLA-7236 Page 1 of 4 Response to Request for Additional Information By letter dated December 19, 2012 (I) PPL Susquehanna, LLC (PPL), submitted a License Amendment Request (LAR) for review and approval of a revision to the Susquehanna Steam Electric Station (SSES), Units 1 and 2, Technical Specifications (TS). Specifically, PPL requested a change toTS Section 2.1.1, "Reactor Core SLs [Safety Limits]," to reflect a lower reactor steam dome pressure for Reactor Core SLs 2.1.1.1 and 2.1.1.2.
PPL stated that the revision became necessary as a result of the General Electric (GE) 10 CFR 21 Communication, SC05-03, "Potential to Exceed Low Pressure Technical Specification Safety Limit," submitted by letter dated March 29, 2005. (Z) In a letter dated August 27, 2014, (3)the NRC requested additional information (RAI). The response to the RAI questions is in the balance of this attachment.
RAI-01: Siemens Power Corporation B (SPCB), AREVA NP Inc.'s (AREVA) critical power correlation for boiling-water reactors is applicable to AREVA (formerly SPCB) ATRIUM-9B and ATRIUM-10 fuel designs. TS Section 4.2.1, "Fuel Assemblies," states, in part, that: ... A limited number of lead use assemblies that have not completed representative testing may be used in SSES, Units 1 and 2, cores for the term of this amendment.
Provide the fuel type(s) that will be used in SSES, Units 1 and 2, cores for the term of this amendment.
PPL's Response:
Susquehanna does not have any Lead Use Assemblies in Unit 1 or Unit 2. Reload quantities of AREVA ATRIUM-10 fuel will be utilized in Unit 2 in 2015 and Unit 1 in 2016. Susquehanna's reload design and analysis process requires that new fuel designs introduced either as Lead Use assemblies or in reload quantities are compared against the applicable fuel-related Technical Specifications.
If the comparison identifies any necessary changes, a License Amendment Request will be submitted following Susquehanna's licensing document change request process. (1) PPL Letter (PLA-6195), "Proposed Amendment No. 312 to Unit 1 License NPF-14 and Proposed Amendment No. 284 to Unit 2 License NPF-22: Low Pressure Safety Limit and Reference Changes," dated December 19, 2012 (Accession ML12355A351).
(2) [SC05-03]
GE-Nuclear Part 21 Report Involving Potential to Exceed Low Pressure Technical Specifications Safety Limit, dated March 29, 2005 (Accession ML050950428). (3) NRC Letter, "Request for Additional Information re: License Amendment Request to Change the Low-Pressure Safety Limit (TAC Nos. MF0410 and MF0411)," dated August 27, 2014 (Accession ML14233A277).
RAI-02: Attachment to PLA-7236 Page 2 of 4 Discuss how operating SSES, Units 1 and 2, with a core with mixed fuel design would impact analyses associated with the LAR. Address, in particular, the scenario where the critical power correlations for the different fuel types have different lower bound pressure ranges. PPL's Response:
SSES, Units 1 and 2, do not currently have mixed cores. If Susquehanna introduces a new fuel design, then the reload design and analysis process requires a comparison against the applicable fuel-related Technical Specifications.
Regardless of the fuel design that is used, the results of the Anticipated Operational Occurrences are not permitted to violate the low pressure safety limit. The critical power correlation must also be used within the approved limits. If either of these conditions cannot be met, then a revised low pressure safety limit is necessary and a License Amendment Request will be submitted following Susquehanna's licensing document change request process. RAI-03: TS 2.1.1.2 specifies the SL on the minimum critical power ratio (MCPR). The proposed change in TS 2.1.1.2 expands the range of applicability of the SL on the MCPR to a lower pressure.
Discuss the impact of the proposed change in TS 2.1.1.2 on the determination of the MCPR core operating limits (Specification 3.2.2) in the core operating limits report as specified in TS Section 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)." PPL's Response:
There is no impact of the proposed change in TS 2.1.1.2 on the determination of the MCPR core operating limits in the COLR. The expanded range of applicability of proposed TS 2.1.1.2 is consistent with the already approved range of applicability of the AREV A SPCB critical power correlation.
The proposed change provides consistency between the approved methodology and the Technical Specifications without requiring the modification of any of the NRC approved methods listed in TS Section 5.6.5.b used to calculate the MCPR operating limits. The proposed change allows for critical power calculations for Anticipated Operational Occurrences specified in the Susquehanna Final Safety Analysis Report (FSAR) that are applicable over a range that is consistent between the proposed TS 2.1.1.2 and the SPCB critical power correlation.
RAI-04: Attachment to PLA-7236 Page 3 of 4 The proposed change in TS 2.1.1.2 expands the range of applicability of the SL on the MCPR to a lower pressure.
The proposed TS change would require determination of the CPR with the reactor at steam dome pressure 2:: 557 psig and core 10 million lbrnlhr. a. Discuss if there is any reduction in the margin of safety because of the expanded range of applicability of the SL on the MCPR. b. Provide the MCPRs, as determined by the SPCB correlation, at steam dome pressures of 785 psig and 557 psig for SSES, Units 1 and 2, respectively, when the reactor is at 23% of rated thermal power (RTP) and a core flow of 10 million lbrnlhr. c. Provide the corresponding mass flux in lbrnlhr-ft 2 for the lowest assembly flow at 23% of RTP and a core flow of 10 million lbrnlhr. PPL's Response:
- a. There is no reduction in the margin of safety because of the expanded range of applicability of the SL on MCPR. The critical power correlation is valid over the range of pressures as specified in AREVA's NRC approved topical report listed in TS 5.6.5.b. Therefore, the proposed change ensures a valid critical power is calculated for normal operation and Anticipated Operational Occurrences defined in the Susquehanna Final Safety Analysis Report (FSAR). A valid critical power calculation permits the calculation of appropriate MCPR Operating Limits for the Core Operating Limits Report (COLR) and ensures the margin of safety is maintained. b. The Technical Specification bases states, in part, that: The use of the SPCB(4) correlation is valid for critical power calculations at pressures 2::571.4 psia and bundle mass fluxes> 0.087 x 10 6 lblhr-ft 2. Provided that the water level in the vessel downcomer is maintained above the top of the active fuel, natural circulation is sufficient to ensure a minimum bundle flow for all fuel assemblies that have a relatively high power and potentially can approach a critical heat flux condition.
For the AREV A NP ATRIUM -10 fuel design, the coolant minimum bundle flow and maximum area are such that the mass flux is always> 0.25 x 10 6 lblhr-ft 2. The natural circulation conditions at 23% core thermal power produce a core flow of approximately 30 Mlblhr which is sufficient to maintain a mass flux of 0.25 x 10 6 lblhr-ft 2. c. Refer to item b discussion.
<4 l EMF-2209(P)(A), "SPCB Critical Power Correlation," AREV A NP, [See Core Operating Limits Report for Revision Level]
RAI-05: Attachment to PLA-7236 Page 4 of 4 SCOS-03 presented a typical PRFO [pressure regulator failure open] response for a low pressure isolation setpoint-analytical limit of 720 psig. The LAR noted that for SSES, Units 1 and 2, the nominal and allowable trip setpoints for the Main Steam Line Pressure -Low are 861 psig and 841 psig, respectively.
- a. Discuss the applicability of the generic transient analysis in SC05-03 to SSES, Units 1 and2. b. Discuss expected differences, if any, in the SSES, Units 1 and 2, plant response as compared to the generic transient analysis in SC05-03. PPL's Response:
- a. The transient described in SCOS-03 is the Pressure Regulator Failure-Open (PRFO) event which is documented in the Susquehanna Final Safety Analysis Report (FSAR). As described in SC05-03 and the Susquehanna FSAR, the transient is initiated when the pressure regulator fails to a high demand allowing all the valves controlling steam flow (turbine control valves and bypass valves) to open as permitted by the Maximum Combined Flow Limiter (MCFL) setting. Opening the valves controlling steam flow to the MCFL setting results in a reactor vessel depressurization that is terminated by a reactor scram on high water level or Main Steam Isolation Valve (MSIV) closure on low steam line pressure.
Figure 1 of SCOS-03 shows a typical plant response to a PRFO. Susquehanna's response to a PRFO will be similar to Figure 1 of SCOS-03, however, there will be differences in the magnitude of the response based on Susquehanna's plant configuration.
Therefore, the generic transient described in SC05-03 is applicable to Susquehanna.
- b. The purpose of the generic analysis in SC05-03 was to demonstrate the potential for violating the Technical Specification low pressure safety limit during a PRFO event. The minimum pressure reached during a PRFO event is dependent on several inputs as described in SCOS-03:
- Low-Pressure Isolation Setpoint (LPIS) Analytical Limit,
- MCFL Setting,
- Feedwater temperature,
- Steam line pressure drop,
- MSIV closure time, and
- Turbine inlet pressure sensor delay. Susquehanna specific values for the above inputs are within the ranges specified in SC05-03. Therefore, the results of Susquehanna specific analyses are expected to be similar to those presented in SC05-03, and the potential exists for Susquehanna to exceed the low pressure safety limit during a PRFO event especially at off-rated power conditions.