NL-13-1751, Application to Revise Tech Specs to Adopt TSTF-535, Revise Shutdown Margin Definition to Address Advanced Fuel Designs
| ML14076A141 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 03/17/2014 |
| From: | Pierce C Southern Co, Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-13-1751 | |
| Download: ML14076A141 (13) | |
Text
Charles R. Pierce Southern Nuclear Regulatory Affairs Director Operating Company, Inc.
40 Inverness Center Parkway Post Office Box 1295 Birmingham, Alabama 35201 March 17, 2014 Docket Nos.: 50-321 50-366 Tel 205.992.7872 Fax 205.992.7601 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant SOUTHERN,\\
COMPANY NL-13-1751 Application to Revise Technical Specifications to Adopt TSTF-535,.. Revise Shutdown Margin Definition to Address Advanced Fuel Designs..
Ladies and Gentlemen:
Pursuant to 10 CFR 50.90, Southern Nuclear Operating Company (SNC) is submitting a request for an amendment to the Technical Specifications (TS) for Edwin I. Hatch Nuclear Plant (HNP), Units 1 and 2. The proposed amendment modifies the TS definition of.. Shutdown Margin" (SDM) to require calculation of the SDM at a reactor moderator temperature of 68°F or a higher temperature that represents the most reactive state throughout the operating cycle. This change is needed to address new Boiling Water Reactor (BWR) fuel designs which may be more reactive at shutdown temperatures above 68°F. provides a description and assessment of the proposed changes. provides the existing TS pages marked up to show the proposed changes. Attachment 3 provides revised (clean) TS pages.
Approval of the proposed amendment is requested by December 1, 2014. Once approved, the Unit 2 amendment shall be implemented prior to reactor startup following Unit 2 refueling outage 2R23 (spring 2015), and the Unit 1 amendment shall be implemented prior to reactor startup following Unit 1 refueling outage 1 R27 (spring 2016). The requested implementation schedule coincides with HNP's planned schedule for the introduction of an advanced fuel design which may require SDM calculations at higher temperatures.
In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Georgia Official.
This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at (205) 992-7369.
U. S. Nuclear Regulatory Commission NL-13-1751 Page 2 Mr. C. R. Pierce states he is Regulatory Affairs Director of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and, to the best of his knowledge and belief, the facts set forth in this letter are true.
Respectfully submitted, c.t~
C. R. Pierce Regulatory Affairs Director CRP/RMJ c1'-::::;e:.:::;::.re me this J 1 day of Nl~c..-L.
Notary Public My commission expires: /6 -~... Zot I Attachments:
1. Description and Assessment
- 2. Proposed Technical Specification Changes (Mark-Up)
- 3. Revised Technical Specification Pages cc:
Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. D. R. Vineyard, Vice President-Hatch Mr. B. L. lvey, Vice President-Regulatory Affairs Mr. D. R. Madison, Vice President-Fleet Operations RTYPE: CHA02.004 U. S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. R. E. Martin, NRR Senior Project Manager-Hatch Mr. E. D. Morris, Senior Resident Inspector-Hatch State of Georgia Mr. J. H. Turner, Environmental Director Protection Division
'2014.
Edwin I. Hatch Nuclear Plant Application to Revise Technical Specifications to Adopt TSTF-535,.. Revise Shutdown Margin Definition to Address Advanced Fuel Designs..
Description and Assessment to NL-13-1751 Description and Assessment
1.0 DESCRIPTION
The proposed amendment modifies the Technical Specifications (TS) definition of.. Shutdown Margin.. (SDM) to require calculation of the SDM at a reactor moderator temperature of 68°F or a higher temperature that represents the most reactive state throughout the operating cycle.
This change is needed to address new Boiling Water Reactor (BWR) fuel designs which may be more reactive at shutdown temperatures above 68°F.
2.0 ASSESSMENT
2.1 Applicability of Published Safety Evaluation Southern Nuclear Operating Company (SNC) has reviewed the model safety evaluation, dated February 19, 2013, as part of the Federal Register Notice of Availability. This review included a review of the NRC staff's evaluation, as well as the information provided in TSTF-535. As described in the subsequent paragraphs, SNC has concluded that the justifications presented in the TSTF-535 proposal and the model safety evaluation prepared by the NRC staff are applicable to Edwin I. Hatch Nuclear Plant (HNP), Unit 1 and Unit 2, and justify this amendment for the incorporation of the changes to the HNP Unit 1 and Unit 2 TS.
The Traveler and model Safety Evaluation (SE) discuss the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). HNP Unit 1 was not licensed to the 10 CFR 50, Appendix A, GDC. The HNP Unit 1 construction permit was received under the 70 general design criteria issued for comment in July 1967, as discussed in section F.3 of the UFSAR. (Appendix F has since been designated as historical).
The HNP Unit 1 licensing basis criteria which are equivalent to the referenced GDCs in the model SE are:
1 0 CFR 50 Appendix A Criterion 26, "Reactivity Control System Redundancy and Capability" Two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions.
HNP Unit 1 10 CFR 50 Appendix A Criterion 26 Reconciliation:
The HNP Unit 1 Updated Final Safety Analysis Report (UFSAR) chapter that discusses the reactivity control system redundancy and capability is Chapter 3.0, "Reactor." This chapter of the Unit 1 UFSAR refers to Chapter 4.0, "Reactor," of the Unit 2 UFSAR in its entirety. As discussed in Chapter 4.0 of the Unit 2 UFSAR, the two independent reactivity control systems of different design principles provided for A-1 to NL-13-1751 Description and Assessment HNP Unit 1 and Unit 2 are by the control rods and standby liquid control system (SBLC). HNP Unit 2 was licensed under the 10 CFR 50 Appendix A GDC. The UFSAR licensing basis for HNP Unit 1 also meets 10 CFR 50 Appendix A Criterion
- 26.
1 0 CFR 50 Appendix A Criterion 27, "Combined Reactivity Control Systems Capability
The reactivity control systems shall be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained.
HNP Unit 1 10 CFR 50 Appendix A Criterion 27 Reconciliation:
As stated above, the HNP Unit 2 reactor (including the control rods and SBLC) was licensing under the current GDC. Given that the UFSAR licensing basis for the reactivity control systems (control rods and SBLC) applies to both Unit 1 and Unit 2, Unit 1 is therefore licensed to this Criterion.
Therefore, the original licensing basis for HNP Unit 1 does not alter the conclusion that the proposed change is applicable to Unit 1. As previously stated, Unit 2 was licensed under the current GDC.
2.2 Optional Changes and Variations SNC is not proposing any variations or deviations from the TS changes described in the TSTF-535, Revision 0, or the applicable parts of the NRC staff's model safety evaluation dated February 19, 2013.
3.0 REGULATORY ANALYSIS
3.1 No Significant Hazards Consideration Determination SNC requests adoption of TSTF-535, Revision 0,.. Revise Shutdown Margin Definition to Address Advanced Fuel Designs,.. which is an approved change to the standard technical specifications {STS), into the Edwin I. Hatch Nuclear Plant (HNP) Units 1 and 2 Technical Specifications (TS). The proposed amendment modifies the TS definition of.. Shutdown Margin..
(SDM) to require calculation of the SDM at a reactor moderator temperature of 68°F or a higher temperature that represents the most reactive state throughout the operating cycle.
A-2 to NL-13-1751 Description and Assessment SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92,
.. Issuance of amendment,.. as discussed below:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the definition of SDM. SDM is not an initiator to any accident previously evaluated. Accordingly, the proposed change to the definition of SDM has no effect on the probability of any accident previously evaluated. SDM is an assumption in the analysis of some previously evaluated accidents and inadequate SDM could lead to an increase in consequences for those accidents. However, the proposed change revises the SDM definition to ensure that the correct SDM is determined for all fuel types at all times during the fuel cycle. As a result, the proposed change does not adversely affect the consequences of any accident previously evaluated.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises the definition of SDM. The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operations. The change does not alter assumptions made in the safety analysis regarding SDM.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed change revises the definition of SDM. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The proposed change ensures that the SDM assumed in determining safety limits, limiting safety system settings or limiting conditions for operation is correct for all BWR fuel types at all times during the fuel cycle.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
A-3 to NL-13-1751 Description and Assessment Based on the above, SNC concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of 11nO significant hazards consideration.. is justified.
3.2 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
4.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 1 0 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.
A-4
Edwin I. Hatch Nuclear Plant Application to _Revise Technical Specifications to Adopt TSTF-535,.. Revise Shutdown Margin Definition to Address Advanced Fuel Designs..
Proposed Technical Specification Changes (Mark-Up)
Definitions 1.1 1.1 Definitions (continued)
SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:
- a.
The reactor is xenon free;
- b.
The moderator temperature is ~ 68°F) corresponding to the m_o_st reactive state; and
- c.
All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.
With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.
THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
TURBINE BYPASS The TURBINE BYPASS SYSTEM RESPONSE TIME consists of two SYSTEM components:
RESPONSE
TIME HATCH UNIT 1
- a.
The time from initial movement of the main turbine stop valve or control valve until 80o/o of the turbine bypass capacity is established; and
- b.
The time from initial movement of the main turbine stop valve or control valve until initial movement of the turbine bypass valve.
The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.
1.1-5 Amendment No. 266
Definitions 1.1 1.1 Definitions (continued)
PHYSICS TESTS RATED THERMAL POWER {RTP)
REACTOR PROTECTION SYSTEM (RPS)
RESPONSE TIME SHUTDOWN MARGIN (SDM)
PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:
- a.
Described in Chapter 14, Initial Tests and Operation, of the FSAR;
- b.
Authorized under the provisions of 10 CFR 50.59; or
- c.
Otherwise approved by the Nuclear Regulatory Commission.
RTP shall be a total reactor core heat transfer rate to the reactor coolant of 2804 MWt.
The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.
SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:
- a.
The reactor is xenon free;
- b.
The moderator temperature is,~0,68°F,,,,=~2JI~,~PQJ1Q1D,Q,,J2.J.Q@.",.!119,§1 reactive state; and
- c.
All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.
With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.
THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
(continued)
HATCH UNIT2 1.1-5 Amendment No. 210
Edwin I. Hatch Nuclear Plant Application to Revise Technical Specifications to Adopt TSTF-535,.. Revise Shutdown Margin Definition to Address Advanced Fuel Designs..
Revised Technical Specification Pages
Definitions 1.1 1.1 Definitions (continued)
SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:
- a.
The reactor is xenon free;
- b.
The moderator temperature is ~ 68°F, corresponding to the most reactive state; and
- c.
All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.
With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.
THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
TURBINE BYPASS The TURBINE BYPASS SYSTEM RESPONSE TIME consists of two SYSTEM components:
RESPONSE
TIME HATCH UNIT 1
- a.
The time from initial movement of the main turbine stop valve or control valve until 80°/o of the turbine bypass capacity is established; and
- b.
The time from initial movement of the main turbine stop valve or control valve until initial movement of the turbine bypass valve.
The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.
1.1-5 Amendment No.
Definitions 1.1 1.1 Definitions (continued)
PHYSICS TESTS RATED THERMAL POWER (RTP)
REACTOR PROTECTION SYSTEM (RPS)
RESPONSE TIME SHUTDOWN MARGIN (SDM)
PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:
- a.
Described in Chapter 14, Initial Tests and Operation, of the FSAR;
- b.
Authorized under the provisions of 10 CFR 50.59; or
- c.
Otherwise approved by the Nuclear Regulatory Commission.
RTP shall be a total reactor core heat transfer rate to the reactor coolant of 2804 MWt.
The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.
SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:
- a.
The reactor is xenon free;
- b.
The moderator temperature is;::: 68°F, corresponding to the most reactive state; and
- c.
All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.
With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.
THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
(continued)
HATCH UNIT 2 1.1-5 Amendment No.