ML20056E799

Request for Additional Information for LAR on Primary Containment Hydrodynamic Loads

ML20056E799
Person / Time
Site:	FitzPatrick
Issue date:	02/25/2020
From:	Samson Lee Plant Licensing Branch 1
To:	Enrique Villar Exelon Generation Co
Lee S
References
EPID L-2019-LLA-0197
Download: ML20056E799 (4)

Text

From: Lee, Samson To: Villar, Enrique:(Exelon Nuclear)

Subject:

FitzPatrick request for additional information for LAR on primary containment hydrodynamic loads (EPID: L-2019-LLA-0197)

Date: Tuesday, February 25, 2020 10:01:00 AM By letter dated September 12, 2019, Agencywide Documents Access and Management System (ADAMS) Accession No. ML19255D988, Exelon Generation Company, LLC (Exelon, the licensee) submitted a license amendment request (LAR) for changes to the James A. FitzPatrick Nuclear Power Plant (FitzPatrick) Technical Specifications (TSs). The proposed changes would delete TS Limiting Condition for Operation (LCO) 3.6.2.4, Drywell-to-Suppression Chamber Differential Pressure, associated Actions and Surveillance Requirements; revise the upper level in LCO 3.6.2.2, Suppression Pool Water Level from 14 feet (ft) to 14.25 ft; and revise the Allowable Value for Table 3.3.5.1-1, Emergency Core Cooling System Instrumentation Function 3.e. Suppression Pool Water Level - High from 14.5 ft to 14.75 ft. The NRC staff has reviewed the LAR and determined that additional information is required to complete the review. The NRC staffs requests for additional information (RAIs) are listed below. These RAIs are in the nuclear systems performance area. A clarification call was held on February 19, 2020. The Exelon staff indicated that there was no proprietary or sensitive information. The Exelon staff requested, and NRC agreed, to a RAI response by April 10, 2020.

The NRC staff considers that timely responses to RAIs help ensure sufficient time is available for staff review and contribute toward the NRCs goal of efficient and effective use of staff resources. Please note that if you do not respond to this request by the agreed-upon date or provide an acceptable alternate date, we may deny your application for amendment under the provisions of Title 10 of the Code of Federal Regulations, Section 2.108. If circumstances result in the need to revise the agreed upon response date, please contact me at (301) 415-3168 or via e-mail Samson.Lee@nrc.gov.

Containment Pressure and Temperature Response Regulatory Basis:

GDC-16 as it relates to the containment and associated systems be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

GDC-50 as it relates to the reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any loss-of-coolant accident.

SNSB RAI-1:

Define the most limiting design basis loss-of-coolant accident (DBLOCA) on which the proposed short-term containment response is based. Provide all initial conditions and key inputs, and assumptions used in the DBLOCA short-term containment pressure and temperature response. Justify if the conservatism in any of the inputs and assumption is

reduced from those used in the NRC accepted current analysis of record (AOR) documented in NEDC-33087P, Revision 1.

SNSB RAI-2:

For the suppression pool swell response to a DBLOCA, the postulated break is the reactor recirculation suction line break (RSLB) that draws water from the annulus area of the reactor in which the break effluent is subcooled. The LOCA mass and energy (M&E) analysis would be non-conservative if the break fluid is assumed saturated because of its lower density as compared to subcooled. For the subcooled break fluid, the higher mass released should result in a higher drywell peak pressure, and therefore higher suppression pool swell loads acting upon structures and components located in the wetwell within the suppression pool swell zone. Provide the reactor water pressure and temperature input for the M&E release analysis and confirm that the break fluid is subcooled. Provide justification if the break fluid is not assumed to be subcooled.

SNSB RAI-3:

The AOR documented in NEDC-33087P, Revision 1, Section 9.0 provides evaluation and results of the following short-term analysis cases of DBLOCA containment pressure and temperature response for normal feedwater temperature (NFWT) and final feedwater temperature reduction (FFWTR). The results of these analyses were used for evaluating the containment hydrodynamic loads.

Case 1 which corresponds to 102% of current licensed thermal power (CLTP) and 100% of rated core flow (RCF).

Case 2 which corresponds to 102% of CLTP, with 105% of RCF (i.e., increased core flow (ICF)).

Case 3 which corresponds to 102% of CLTP, with 79.8% of RCF (i.e., on the maximum extended load line limit analysis (MELLLA) line).

Case 4 which corresponds to 62% of CLTP, with 36.8% of RCF (Minimum Pump Speed (MPS) on the MELLLA line).

Provide containment pressure and temperature response evaluation and results, both graphs and peak values, for the above cases based on the proposed change in the suppression pool TS maximum level. In case all of the above 4 cases are not re-analyzed, provide the results of the most limiting case with quantitative justification showing that it bounds the remaining 3 cases.

LOCA Containment Hydrodynamic Loads Regulatory Basis:

GDC-4 as it relates to structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and

discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit.

SNSB RAI-4:

The AOR documented in NEDC-33087P, Revision 1, Section 9.0 provides assessment of the following containment LOCA hydrodynamic loads based on the short-term containment pressure and temperature response analysis:

Pool Swell Vent Thrust Condensation Oscillation Chugging

a. Provide assessment of these loads based on the revised containment pressure and temperature response for the bounding case.

b. The pool swell loads depend on the drywell pressurization rate. NEDC-33087P, Revision 1, the third paragraph in Section 9.3.2 states:

The test condition pressurization rate and scale factor for the JAF quarter scale tests (Reference 19 [NEDE-21944-P, "Mark I Containment Program 1/4 Scale Pressure Suppression Pool Swell Test Program: Plant Unique Tests," Volume 1, March 1979.]) are 31 psi/sec and 0.2627, respectively.

From Reference 19, the composite scaling factor then becomes 0.2627 so that the pressurization rate in full scale is 31.0 / 0.2627 = 60.48 psi/sec. The pool swell loads evaluation reviews the drywell pressurization rate obtained from Case 5 and compares it to the Reference 19 pool swell test condition scaled up to full-scale.

An initial drywell pressurization rate of 58.9 psi/sec was calculated for Case 5.

As stated above, in the current analysis, the initial pressurization rate of 58.9 psi/sec is bounded by the derived full-scale test value of 60.48 psi/sec. Provide the most limiting drywell pressurization rate based on the proposed TS changes to confirm that it remains bounded by the full-scale drywell pressurization rate of 60.48 psi/sec.

c. Based on the proposed TS changes, provide bounding values of the vent thrust loads based on which the structural analysis given in Attachment 6 of the LAR is performed.

Main Steam Line Break Response Regulatory Basis:

GDC-16 as it relates to the containment and associated systems be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are

not exceeded for as long as postulated accident conditions require.

GDC-50 as it relates to the reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any loss-of-coolant accident.

SNSB RAI-5:

Confirm that based on the proposed TS changes, the containment pressure and temperature response and the drywell pressurization rate for the most limiting main steam line break is bounded by the RSLB DBLOCA.