ML16188A368
| ML16188A368 | |
| Person / Time | |
|---|---|
| Site: | South Texas  (NPF-076) |
| Issue date: | 07/06/2016 |
| From: | Guthrie B, Schmitt D Westinghouse |
| To: | Lisa Regner Plant Licensing Branch IV |
| Regner L, NRR/DORL/LPLIV-1, 415-1906 | |
| References | |
| WCAP-9272-P-A | |
| Download: ML16188A368 (8) | |
Text
Westinghouse Non-Proprietary Class 3
© 2016 Westinghouse Electric Company LLC. All Rights Reserved.
1 Brian Guthrie, Transient Analysis Danielle Schmitt, Nuclear Design Westinghouse Methodology Impacts from the Removal of the RCCA from Core Location D-6
2 Westinghouse Non-Proprietary Class 3
© 2016 Westinghouse Electric Company LLC. All Rights Reserved.
Overview of Reload Method (WCAP-9272-P-A)
- The STP engineering design change process is used to implement the removal of RCCA D-6
- Impacts of the change on UFSAR Chapter 6 and Chapter 15 analyses have been evaluated through a detailed review of the AORs and the reload design process (WCAP-9272-P-A).
- As detailed in Table 7 of NOC-AE-16003351, the existing key safety parameters continue to apply (and are confirmed on a cycle specific basis) and no other aspects of the AORs are impacted.
WCAP-9272 methodology is still applicable when control rod pattern is changed
3 Westinghouse Non-Proprietary Class 3
© 2016 Westinghouse Electric Company LLC. All Rights Reserved.
Overview of Reload Method (WCAP-9272-P-A)
4 Westinghouse Non-Proprietary Class 3
© 2016 Westinghouse Electric Company LLC. All Rights Reserved.
Overview of Reload Method (WCAP-9272-P-A)
- WCAP-9272-P-A Process
- The AORs are established using conservative inputs that are expected to bound future cycles.
- The key safety parameters that could be impacted by a reload become the Reload Safety Analysis Checklist (RSAC) which is issued every cycle by the Safety Analysis groups to Nuclear and Thermal Hydraulic Design for confirmation
- Any RSAC limit that is not met for a given cycle is provided to the impacted Safety Analysis group for evaluation
- Evaluation or re-analysis of the AOR is performed to address the RSAC exception and the results are reported in the cycle specific Reload Evaluation.
WCAP-9272 methodology is still applicable when control rod pattern is changed
5 Westinghouse Non-Proprietary Class 3
© 2016 Westinghouse Electric Company LLC. All Rights Reserved.
Safety Analysis Accidents
- Example - HZP Steamline Break
- RETRAN-02 uses a point kinetics neutronics model and does not directly model a control rod pattern
- Limiting case is generated using assumed physics characteristics.
- Nuclear Design implicitly confirms acceptability of the physics characteristics by comparing ANC calculated parameters to RETRAN-02 calculated parameters.
- An unacceptable mismatch requires evaluation or reanalysis of the event using adjusted physics parameters.
- This approach is consistent with the analysis methodology for the HZP steamline break core response described in WCAP-9226-P-A, Revision 1; the reload methodology for this event is described in WCAP-9272-P-A; and the qualification of RETRAN-02 for use in analyzing HZP steamline break core response is described in WCAP-14882-P-A.
No UFSAR safety analysis calculations explicitly model control rod patterns.
6 Westinghouse Non-Proprietary Class 3
© 2016 Westinghouse Electric Company LLC. All Rights Reserved.
Cycle-Specific Core Design Analysis
- Each cycle the reload core is modeled by Nuclear Design in detail (fuel pattern, control rod pattern, burnable absorbers, etc.)
- This detailed model is used to calculate cycle-specific values for comparison to the RSAC key safety parameters
- Key safety parameters assumed in the AOR include values that are impacted by RCCAs such as shutdown margin, total rod worth, and trip reactivity
- Key safety parameters assumed in the AOR do not include the number of RCCAs, RCCA configuration, or a symmetric RCCA pattern Nuclear Design calculations explicitly model control rod pattern.
7 Westinghouse Non-Proprietary Class 3
© 2016 Westinghouse Electric Company LLC. All Rights Reserved.
Cycle-Specific Core Design Analysis
- Nuclear Design performs the cycle-specific analysis using:
- ANC [WCAP-10965-P-A]
- APOLLO [WCAP-13524-P-A]
- PARAGON/NEXUS [WCAP-16045-P-A, Addendum 1-A]
- These codes were rigorously benchmarked and qualified for a variety of reactor types, fuel types, and burnable poisons
- These codes are capable of modeling an asymmetric core:
- Highest worth rod stuck out of the core (N-1) is a conservatism assumption for many RSAC calculations
- Asymmetric core temperature distribution with an N-1 condition is modeled for the Steamline Break accident
- N-2, N-3, etc are calculated for use in plant Operations Nuclear Design codes are capable of modeling the removal of RCCA D-6.
8 Westinghouse Non-Proprietary Class 3
© 2016 Westinghouse Electric Company LLC. All Rights Reserved.
Conclusions
- Plant changes are handled by STP engineering design change process
- As a part of the engineering design change process, Westinghouse and the STP Nuclear Fuel and Analysis group performed an impact review of the respective AORs including the continued applicability of WCAP-9272-P-A
- WCAP-9272-P-A methodology is still applicable when control rod pattern is changed
- No UFSAR safety analysis calculations explicitly model control rod patterns
- Nuclear Design codes are capable of modeling the removal of RCCA D-6