L-2018-205, St. Lucie Unit 2, Response to Request for Additional Information Regarding License Amendment Request to Reduce the Number of Control Element Assemblies: Difference between revisions

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{{Adams
#REDIRECT [[L-2018-205, Response to Request for Additional Information Regarding License Amendment Request to Reduce the Number of Control Element Assemblies]]
| number = ML18319A043
| issue date = 11/15/2018
| title = St. Lucie Unit 2, Response to Request for Additional Information Regarding License Amendment Request to Reduce the Number of Control Element Assemblies
| author name = Deboer D
| author affiliation = Florida Power & Light Co
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000389
| license number = NPF-016
| contact person =
| case reference number = L-2018-205
| document type = Letter type:L, Response to Request for Additional Information (RAI)
| page count = 8
| project =
| stage = Response to RAI
}}
 
=Text=
{{#Wiki_filter:* I= PL .. U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington DC 20555-0001 RE: St. Lucie Unit 2 Docket No. 50-389 NOV 1 5 2018 Renewed Facility Operating Licenses NPF-16 L-2018-205 10 CFR 50.90 Response to Request for Additional Information Regarding License Amendment Request to Reduce the Number of Control Element Assemblies
 
==References:==
: 1. Florida Power & Light Company letter L-2018-121, License Amendment Request to Reduce the Number of Control Element Assemblies, June 29,2018 (ADAMS Accession No. ML18180A094)
: 2. Florida Power & Light Company letter L-2018-153, Supplemental Information for License Amendment Request to Reduce the Number of Control Element Assemblies, August 17, 2018 (ADAMS Accession No. ML18229A050)
: 3. St. Lucie Plant, Unit No.2, Request for Additional Information Regarding License Amendment Request to Reduce the Number of Control Element Assemblies (EPID L-2018-0181)
October 22, 2018 (ADAMS Accession No. ML18296A205)
In Reference 1, as supplemented by Reference 2, Florida Power & Light Company (FPL) requested an amendment to Renewed Facility Operating License NPF-16 for St. Lucie Unit 2. The proposed license amendment modifies the St. Lucie Unit 2 Technical Specifications (TS) by reducing the specified number of control element assemblies (CEAs) from 91 to 87 in support of a planned modification to permanendy remove four 4-element (mini-dual)
CEAs from the reactor core. The proposed license amendment relatedly deletes a reference to the 4-element CEAs in a TS definition.
In Reference 3, the NRC requested additional information deemed necessary to complete its review. ------+h&-@ndesu+/-&-te
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&-NRGs-reEJU@St-fer-additien-al
-infe:tcm.a-tien(RAI). The response provides additional information that clarifies the application, does not expand the scope of the application as originally noticed, and does not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register. This letter contains no new or revised regulatory commitments.
Should you have any questions regarding this submittal, please contact Mr. Michael Snyder, St. Lucie Licensing Manager, at (772) 467-7036.
St. Lucie Nuclear Plant Docket No. 50-389 I declare under penalty of perjury that the foregoing is true and correct. Executed on NOV 1 5 2018 Sincerely, Daniel DeBoer Site Director-St. Lucie Nuclear Plant, Units 1 and 2 Florida Power & Light Company
 
==Enclosure:==
 
FPL Response to Request for Additional Information cc: USNRC Regional Administrator, Region II USNRC Project Manager, St. Lucie Nuclear Plant, Units 1 and 2 USNRC Senior Resident Inspector, St. Lucie Nuclear Plant, Units 1 and 2 Ms. Cindy Becker, Florida Department of Health L-2018-205 Page 2 of2 St. Lucie Nuclear Plant L-201 8-205 Docket No. 50-389  Enclosure Page 1 of 6  Enclosure  Florida Power & Light Company Response to NRC Request for Additional Information (RAI)
Regarding St. Lucie Unit 2 License Amendment Request to Reduce the Number of Control Element Assemblies
 
St. Lucie Nuclear Plant L-201 8-205 Docket No. 50-389  Enclosure Page 2 of 6  In Reference 1, as supplemented by Reference 2, FPL requested a license amendment to modif y the St. Lucie Unit 2 TS by reducing the specified number of CEAs from 91 to 87 in support of a planned modification to permanently remove four 4
-element (mini
-dual) CEAs from the reactor core. The proposed license amendment relatedly delete s a reference to the 4
-element CEAs in a TS definition.
In Reference 3, the Reactor Systems Branch (SRXB) of the NRC Office of Nuclear Reactor Regulation requested additional information, as indicated below.
FPL's response follows:
 
SRXB-RAI-1:  Shutdown Margin Regulatory Basis
- Appendix A, "General Design Criteria [GDC] for Nuclear Power Plants," to Title 10 of the Code of Federal Regulations, Part 50; GDC-25, "Protection System Requirements for Reactivity Control Malfunctio n s",  GDC-26, "Reactivity Control System Redun dancy and Capability
" a nd GDC-27, "Combined Reactivity Control Systems Capability"
.
GDC-25 re quires, in par t , that the p rotection sys tem shall be designed to assure that t he speci fied acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems.
GDC-26 re quires, in par t , that two i ndependent reactivity control systems of different design principles sh a ll b e provided. One of the systems shall use control r ods. The s econd reacti vity control system shall b e capable of reliably con trolling the r ate of reactivity changes resulting f r o m planned, normal power changes t o assure acceptable fuel design limits are not exceeded. One of the sys tems shall be capable of holding t h e reactor c o re sub critical under cold c onditions.
GDC-27 re quires, in par t , that the r eactivity control systems shall be des igned to have a combined capability, in conjuncti o n with poison addition by the emergency core cooling system, of reliably controlling r eactivity changes to assure that t he capability to cool the core is mai ntained.
The current St. Lucie 2 design bases that satisfy these design criteria are to provide the amount of reactivity available from insertion of withdrawn CEAs under all power operating conditions, even when the highest worth CEA fails to insert, at least one percent shutdown margin after cooldown to hot zero power, and any additional shutdown reactivity requirements assumed in the safety analyses.
Request As stated in the LAR, a calculation specific to St. Lucie 2 was performed to evaluate the impact of the removal of the four 4
-element CEAs on the total worth of reactivity systems and the subsequent available shutdown margin. Provide the following information related to the calculation:
 
(1) Brief description including computer codes for the approved methodology to perform the calculation, (2) Summary of the calculation assumptions and conditions, e.g. the highest worth CEA fails to insert etc.,
St. Lucie Nuclear Plant L-201 8-205 Docket No. 50-389  Enclosure Page 3 of 6  (3) Calculation results on the shutdown margin showing the difference between "With" and "Without" four 4-element CEAs,  (4) Based on (3), a justification for the removal of four 4
-element CEAs by taking into account the calculation uncertainties (e.g. uncertainties applied to Updated Final Safety Analysis Report (UFSAR)
Table 4.3.6) and the compensatory measure, if required, for the reduction in the shutdown margin due to the removal of four 4
-element CEAs.
FPL Response: 
(1) The methodology used in performing the CEA reactivity worth and the subsequent shutdown margin calculation is the same as currently used in all the St. Lucie Unit 2 neutronics calculations. This methodology, WCAP
-11596-P-A (Reference 4), is NRC approved and listed in St. Lucie Unit 2 TS 6.9.1.11.b.1. The computer code used is the PHOENIX-P/ANC code package.
 
(2) The neutronics calculation for total rod worth and shutdown margin included conservative methodology assumptions such as skewing xenon to the limits of Axial Shape Index (ASI) in Core Operating Limits Report (COLR) Figure 3.2
-4, highest worth CEA stuck in the fully withdrawn position and 10% total rod worth uncertainty.
 
(3) To ensure the elimination of four 4
-element CEAs does not pose problems in meeting the COLR limits, a separate sample core loading pattern was developed for Cycle 21 without using the four 4
-element CEAs. The excess shutdown margin in this calculation, with the highest worth CEA stuck in the fully withdrawn position, was 1059 pcm at the beginning
-of-cycle (BOC) and 330 pcm at the end
-of-cycle (EOC).  (The excess shutdown margin is the margin above the COLR limit of 3600 pcm). The corresponding excess shutdown margins with the original Cycle 21 design, i.e. with the four 4
-element CEAs, was 1245 pcm at BOC and 356 pcm at EOC. Since the worth of the highest stuck rod is loading pattern dependent, the excess shutdown margin varies from cycle to cycle and can be controlled via the core design.
 
(4) The calculation demonstrated that the shutdown margin requirements with the highest worth CEA stuck in the fully withdrawn position can be easily met without the use of the four 4-element CEAs. Additionally, if needed, excess shutdown margin can be controlled by altering the core design and fuel load during the design phase of a reload cycle.
 
SRXB-RAI-2:  Steam Line Break Analysis Regulatory Basis - GDC-10, "Reactor Design", GDC-15, "Reactor Coolant System Design", GDC-20 "Protection System Fu nction", and GDC-26. GDC-10 re quires, in par t , that the r eactor core a nd associa ted coolant, control, and protection systems shall be designed wi t h appropria t e margin to assure that specified acce ptable fu e l design St. Lucie Nuclear Plant L-201 8-205 Docket No. 50-389  Enclosure Page 4 of 6  limits are not exceeded during any condition of normal operation, includi n g the effects of anticipa ted operatio n al occurrences.
GDC-15 re quires, in par t , that the r eactor coola n t system and assoc i at e d auxiliary, control, and protection systems be designed wi t h sufficient margin to assure that the design co nditions of t h e reactor c oolant press u re boundary are not exceeded during a ny condition of normal opera tion, including anticipated operational occurrences.
 
GDC-20 re quires, in par t , that the p rotection sys tem shall be designed (1) to initiate automatically the operati o n of appropri ate systems including the reactivity control systems, and (2) to sense acci dent conditi o ns and initiate the operation of systems and components important to safety.
GDC-26 requires , in par t , that two i ndependent reactivity control systems of different design principles sh a ll be provided.
One of the systems shall use control r ods. The s econd reacti vity control system shall b e capable of reliably con trolling the r ate of reactivity changes resulting f r o m planned, normal power changes t o assure acceptable fuel design limits are not exceeded. One of the sys tems shall be capable of holding t h e reactor c o re sub critical under cold c onditions.
The current St. Lucie 2 design bas e s that satisfy these desi g n criteria are to provide the amount of reactivity available from insertion of withdrawn CEAs under all power operating c onditions, e ven when the highest wo rth CEA fails to insert, at le a st one perc e nt shutdown margin after cooldown t o hot zero power, and any additional shutdown reactivity requirements assumed in the safety analyse s. Request As stated in the LAR, a calculation s pecific to St. Lucie 2 was performed to evaluate the impact of the removal of the four 4-element CEAs on the s team line break analysis. P rovide the following in formation related to the calcula tion: 
(1) Brief descri ption inclu d i n g computer codes for t h e approved methodology to perform the calc ulation, (2) Summary of the calc ulation assumptions and co nditions, e.g. the highest worth CEA fails to i nsert etc., 
(3) Calculation results on t h e total reactivity balance for the rated thermal power and associated fuel peaking fac tors showing the differe nce between "With" and "Without" fo u r 4-element CEAs, (4) Based on (3), a justi fication for the removal of fo u r 4-element CEAs by ta king into account the calcula tion uncertainti e s. FPL Response
: 
(1) The methodology used in performing the steam line break reactivity balance was the same as that used for other St. Lucie Unit 2 neutronics calculations. This methodology, WCAP
-
St. Lucie Nuclear Plant L-201 8-205 Docket No. 50-389  Enclosure Page 5 of 6  11596-P-A (Reference 4), is NRC approved and is listed in St. Lucie Unit 2 T S 6.9.1.11.b.1. The computer code used is the PHOENIX
-P/ANC code package.
  (2) The Cycle 21 steam line break reactivity balance calculation performed using the Reference 4 methodology, assumed conservative methodology assumptions which include highest worth rod stuck in the fully withdrawn position in the limiting (coldest) region of the core, and also analyzing the impact of other high worth rods being stuck in the fully withdrawn position.
The statepoint used is conservative with respect to parameters such as the reactor coolant system pressure, temperature and flow.
 
(3) The Cycle 21 steam line break reactivity balance calculation showed that the worst reactivity change was 3661 pcm for the case with the four 4-element CEAs, while the change was 3643 pcm for the case without the four 4
-element CEAs. The peak linear heat rate was 12.52 kW/ft for the case with the four 4-element CEAs, while the peak linear heat rate was 13.26 kW/ft for the case without the four 4
-element CEAs. The peak linear heat rate value without the four 4
-element CEAs remained well below the limit of 22 kW/ft.  (4) The calculation showed that the steam line break analysis reactivity balance impact is minimal with the four 4
-element CEAs removed, using the same analysis methodology and assumptions. The removal of the four 4-element CEAs is thus justified as being acceptable from core design considerations.
 
SRXB-RAI-3: Core Bypass Fl o w:  Regulatory Basis - GDC-10 GDC-10 re quires, in par t , that the r eactor core a nd associa ted coolant, control, and protection systems shall be designed wi t h appropria t e margin to assure that specified acce ptable fu e l design limits are not exceeded during any condition of normal operation, includi n g the effects of anticipa ted operatio n al occurrences.
Request Clarify the following inf ormation as described in LAR: 
(1) The design bypass flow fraction of t otal core flow at rated co r e flow condi tions that is used in the current St. Lucie 2 UFSAR C h apter 15 safe t y analysis, (2) The new bypass flow (f raction of ra ted core flo w) after removal of the four 4-element CEAs, (3) The bypass flow fraction (of rated c o re flow) assumed in the fuel vendor's reload an alysis, (4) The measured current total bypass flow fraction (of rated c o re flow) at rated core fl o w conditions.
 
St. Lucie Nuclear Plant L-201 8-205 Docket No. 50-389  Enclosure Page 6 of 6  FPL Response
:  (1) St. Lucie Unit 2 transitioned to Framatome fuel beginning with Cycle 23 in Spring 2017. The core bypass flow prior to the implementation of fuel design change was a conservative value of 3.7% of the design reactor coolant system (RCS) flow. The transition to Framatome fuel increased the core bypass flow by approximately 0.17% to a design bypass flow value of 3.87%. The St. Lucie Unit 2 current Chapter 15 safety analysis conservatively uses a bypass flow of 4.2% of design RCS flow for all fuel related analyses. 
(2) The removal of four 4-element CEAs will increase the core bypass flow conservatively by approximately 0.04% of the design RCS flow. The new bypass flow is estimated to become (3.87% + 0.04%) = 3.91%.
 
(3) The Framatome (current fuel vendor) reload analysis conservatively assumes a core bypass flow of 4.2% of design RCS flow, which bounds the bypass flow of 3.91% of design RCS flow resulting from the removal of the four 4
-element CEAs.
 
(4) There is no current measurement of core bypass flow. The value of 3.7%, used in analysis prior to the transition to Framatome fuel, was justified to be conservative for the previous fuel vendor's Chapter 15 analyses based on the values presented in UFSAR Table 4.4
-3, "Reactor Coolant Flows in Bypass Channels
". The currently used new bypass flow of 4.2% has more than sufficient margin to cover the impact of the Framatome fuel and bound the minor impact on bypass flow resulting from the removal of the four 4
-element CEAs, as discussed above.
 
==References:==
 
(1) Florida Power & Light Company letter L
-2018-121, License Amendment Request to Reduce the Number of Control Element Assemblies, June 29, 2018 (ADAMS Accession No. ML18180A094)
 
(2) Florida Power & Light Company letter L
-2018-153, Supplemental Information for License Amendment Request to Reduce the Number of Control Element Assemblies, August 17, 2018 (ADAMS Accession No. ML18229A050)
 
(3) St. Lucie Plant, Unit No. 2, Request for Additional Information Regarding License Amendment Request to Reduce the Number of Control Element Assemblies (EPID L
-2018-0181) October 22, 2018 (ADAMS Accession No. ML18296A205)
  (4) WCAP-11596-P-A, "Qualification of the PHOENIX
-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988 (Westinghouse Proprietary).}}

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