ML21088A344
| ML21088A344 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 03/29/2021 |
| From: | Huey D Energy Harbor Nuclear Corp |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| GL-04-002, L-21-101, TAC MC4681 | |
| Download: ML21088A344 (20) | |
Text
Energy Harbor Nuclear Corp.
Davis-Besse Nuclear Power Station 5501 N. State Route 2 Oak Harbor, Ohio 43449 419-321-7676 March 29, 2021 10 CFR 50.54(f)
L-21-101 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
Davis-Besse Nuclear Power Station, Unit No. 1 Docket No. 50-346, License No. NPF-3 Final Supplemental Response to Nuclear Regulatory Commission Generic Letter 2004-02 (TAC No. MC4681)
This submittal provides the Energy Harbor Nuclear Corp. final supplemental response for the Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS), to Generic Letter (GL) 2004-02, dated September 13, 2004, (ADAMS Accession No. ML042360586) Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors.
By letter dated February 27, 2020 (ADAMS Accession No. ML20030A440), the Nuclear Regulatory Commission (NRC) staff issued amendment number 299 to the renewed facility operating license for DBNPS. The amendment documents that FirstEnergy Nuclear Operating Company has been renamed Energy Harbor Nuclear Corp. The following paragraphs refer to letters regarding DBNPS submitted by FirstEnergy Nuclear Operating Company (FENOC).
On May 15, 2013 (ADAMS Accession No. ML13135A456), FENOC submitted a letter of intent per SECY-12-0093, Closure Options for Generic Safety Issue - 191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance, indicating DBNPS would pursue Closure Option 2 - Deterministic of the SECY recommendations (refinements to evaluation methods and acceptance criteria). The final outstanding issue for DBNPS with respect to GL 2004-02 is the in-vessel downstream effects evaluation, which addresses that long-term core cooling (LTCC) can be adequately maintained for all postulated accident scenarios that require sump recirculation.
The in-vessel downstream effects evaluation has been completed for DBNPS and is enclosed. This satisfies the GSI-191 Regulatory Commitment No. 2 identified in the May 15, 2013 submittal.
Enclosure L-21-101 Davis-Besse Nuclear Power Station, Unit No. 1 Final Supplemental Response to Nuclear Regulatory Commission Generic Letter 2004-02 (17 pages follow)
Energy Harbor Nuclear Corp.
Davis-Besse Nuclear Power Station, Unit No. 1 Final Supplemental Response to Nuclear Regulatory Commission Generic Letter 2004-02 Prepared by _______________________________
Reviewed by _______________________________
Approved by _______________________________
Davis-Besse Nuclear Power Station, Unit No. 1 1
Table of Contents 1.0 Overall Compliance 1.1 Overview of DBNPS Resolution to GL 2004-02 1.2 Correspondence Background 1.3 General Plant System Description 1.4 General Description of Containment Sump Strainers 2.0 General Description and Schedule for Corrective Actions 3.0 Specific Information for Review Areas 3.n Downstream Effects - Fuel and Vessel 3.o Chemical Effects 3.p Licensing Basis 4.0 References Davis-Besse Nuclear Power Station, Unit No. 1 2
1.0 Overall Compliance NRC Issue:
Provide information requested in GL 2004-02, "Requested Information." Item 2(a) regarding compliance with regulations. That is, provide confirmation that the Emergency Core Cooling System (ECCS) and Containment Spray System (CSS) recirculation functions under debris loading conditions are or will be in compliance with the regulatory requirements listed in the Applicable Regulatory Requirements section of this generic letter. This submittal should address the configuration of the plant that will exist once all modifications required for regulatory compliance have been made and this licensing basis has been updated to reflect the results of the analysis described above.
Energy Harbor Nuclear Corp. Response:
In accordance with SECY-12-0093 (Reference 1) and as identified in FENOC letter to the NRC dated May 15, 2013 (ADAMS Accession No. ML13135A456) (Reference 4),
DBNPS elected to pursue GSI-191 Closure Option 2 - Deterministic and identified in-vessel downstream effects as the last outstanding issue. Topical Report (TR) WCAP-17788-P, Rev. 1 provides evaluation methods and results to address in-vessel downstream effects. As discussed in NRC Technical Evaluation Report of In-Vessel Debris Effects, (ADAMS Accession No. ML19178A252), the NRC staff has performed a detailed review of WCAP-17788-P. Although the NRC staff did not issue a Safety Evaluation for WCAP-17788, as discussed further in U.S. Nuclear Regulatory Commission Staff Review Guidance for In-Vessel Downstream Effects Supporting Review of Generic Letter 2004-02 Responses (ADAMS Accession No. ML19228A011),
the staff expects that many of the methods developed in the TR can be used by PWR licensees to demonstrate adequate long-term core cooling (LTCC). Completion of the analyses demonstrate compliance with 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power plants, (b)(5), Long-term cooling, as it relates to in-vessel downstream debris effects for DBNPS.
1.1 Overview of DBNPS Resolution to GL 2004-02 On May 15, 2013 (Reference 4), FENOC submitted a Supplemental Response to GL 2004-02 for DBNPS which summarized changes implemented to support the resolution of GSI-191.
The changes that were implemented and identified in the DBNPS Supplemental response remain valid and no further changes are being made to address in-vessel downstream effects.
Davis-Besse Nuclear Power Station, Unit No. 1 3
1.2 Correspondence Background The following provides a listing of correspondence issued by the NRC or submitted by FENOC for DBNPS, on GL 2004-02:
Table 1.2-1: Generic Letter 2004-02 Correspondences ADAMS Accession Document Date Document Number Nuclear Regulatory Commission (NRC) Generic September 13, Letter (GL) 2004-02, "Potential Impact of Debris 2004 ML042360586 Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors" First FirstEnergy Nuclear Operating Company March 4, 2005 ML050670489 (FENOC) Response to GL 2004-02 June 2, 2005 ML051530197 First NRC Request for Additional information (RAI)
July 26, 2005 ML052090194 First FENOC RAI Response September 1, 2005 ML052490423 Second FENOC Response to GL 2004-02 February 9, 2006 ML060385725 Second NRC RAI NRC Letter to Nuclear Energy Institute (NEI),
March 3, 2006 ML060650335 GL 2004-02 RAI Responses NRC letter to Licensees, Alternative Approach for March 28, 2006 ML060870274 Responding to NRC RAI Letter Regarding GL 2004-02 SECY-06-0078, "Status of Resolution of GSI -191, "Assessment of Effect of Debris Accumulation on March 31, 2006 ML053620174 PWR Pressurized Water Reactor Sump Performance,""
NRC E-Mail to NEI, NRC Staff Review March 29, 2007 ML071280350 Considerations for Buffer Changes NRC Letter to NEI, Content Guide for Generic August 15, 2007 ML071060091 Letter 2004-02 Supplemental Responses Updated Communication Plan for GSI-191 October 5, 2007 ML060230398 Assessment of Debris Accumulation on PWR Sump Performance FENOC letter to NRC Davis Besse Nuclear Power Station NRC to NEI, Plant-Specific Requests for November 7, 2007 ML073060581 Extension of Time to Complete One or More Corrective Actions for GL 2004-02 Davis-Besse Nuclear Power Station, Unit No. 1 4
Table 1.2-1: Generic Letter 2004-02 Correspondences ADAMS Accession Document Date Document Number NRC letter to NEI, Plant-Specific Requests for November 8, 2007 ML073060581 Extension of Time to Complete One or More Corrective Actions for GL 2004-02.
NRC Letter to NEI, Revised Content Guide for November 21, 2007 ML073110389 GL 2004-02 Supplemental Responses FENOC letter to NRC Request for Extension of December 27, 2007 ML073650213 Completion Date for Davis-Besse Nuclear Power Station Unit No. 1 NRC Letter to FENOC, Davis-Besse Nuclear December 28, 2007 ML062830044 Power Station Unit No.1 GL 2004-02 Extension Request Evaluation FENOC to NRC Letter, Supplemental Response to GL 2004-02 Potential Impact of Debris of February 28, 2008 ML080650368 Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors NRC Letter to NEI, Revised Guidance for Review March 28, 2008 ML080230112 of Final Licensee Responses to GL 2004-02 Closure RC to FENOC Letter Request for Additional July 10, 2008 ML081780331 Information related to GL 2004-02, FENOC Letter to NRC, Supplemental Response to April 30, 2010 ML101250217 GL 2004-02 NRC Letter to NEI, NRC Review of NEI Clean May 2, 2012 ML120730181 Plant Acceptance Criteria for ECCS May 15, 2013 ML13135A456 FENOC to NRC Letter, GSI-191 Resolution Plan FENOC Letter to NRC, Notification of Closure of July 3, 2013 ML13186A051 a Commitment Related to the GSI-191 Resolution Plan Memo, NRC Review Guidance for IVDE September 4, 2019 ML19228A011 Supporting Review of GL 2004-02 Davis-Besse Nuclear Power Station, Unit No. 1 5
1.3 General Plant System Description Description of DBNPS Plant Systems DBNPS Unit 1 is a B&W two loop Pressurized Water Reactor (PWR) design. The reactor coolant (RC) system consists of the reactor vessel, two vertical once-through steam generators, four shaft-sealed reactor coolant pumps, an electrically heated pressurizer, and interconnecting piping. The system, located entirely within the containment vessel (CV), is arranged in two heat transport loops, each with two RC pumps and one steam generator. Reactor coolant is transported through piping connecting the reactor vessel to the steam generators and flows downward through the steam generator tubes transferring heat to the steam and water on the shell side of the steam generator. In each loop the reactor coolant is returned to the reactor through two lines, each containing an RC pump. In addition to serving as a heat transport medium, the coolant also serves as a neutron moderator and reflector and as a solvent for the soluble poison (boron in the form of boric acid) utilized in chemical shim reactivity control.
The emergency core cooling system (ECCS) is made up of the high-pressure injection (HPI) system, the low-pressure injection (LPI) system, and the core flooding system.
The HPI system provides pumped injection of water into the cold leg piping. The system operates independently of other ECCS subsystems and is used to provide cooling water to the core when the low-pressure system is unavailable, which could occur as a result of small leaks.
The core flooding system, made up of two pressurized storage tanks, provides water at medium to low pressure directly to the reactor vessel downcomer through a set of nozzles. The system is short term and ceases to operate when the tanks empty which may take forty to several hundred seconds depending on the nature of the accident.
The LPI system provides pumped injection of low-pressure water directly into the vessel downcomer. The system is used for long term cooling of the reactor core during a loss-of-coolant accident. The system injects into the vessel through the same nozzle as the core flooding system.
The containment spray (CS) system consists of two half-capacity pumps, two half-capacity spray headers, isolation valves, and the necessary piping, instrumentation, and controls. The pumps and valves can be remote manually operated from the control room.
High containment vessel pressure or low reactor coolant pressure will actuate Level 2 trip to open the spray isolation valves. High-high containment pressure will actuate Level 4 trip to start two containment spray pumps. The pumps take suction initially from the borated water storage tank (BWST). The CS system shares the BWST and the suction lines from the tank with the high and low pressure injection systems.
ECCS systems will start following a loss-of-coolant accident (LOCA). Initially, two HPI, two LPI pumps take suction from the BWST. When the BWST level reaches the low-Davis-Besse Nuclear Power Station, Unit No. 1 6
level set point, the suction for LPI and CS pumps will be manually transferred from the BWST to the containment emergency sump.
Schematics of the ECCS and CS System were provided in Reference 2.
1.4 General Description of Containment Sump Strainers The containment vessel emergency sump, constructed of concrete, is located inside the CV on elevation 565 feet 0 inches. The sump is protected from missiles by the secondary shield wall, the refueling canal wall, the floor at elevation 578 feet 0 inches, and the containment vessel wall. The strainers over the sump are constructed of tubular steel framing and grating to which 3/16-inch diameter stainless steel perforated plate is attached. The incore tunnel portion of the emergency sump strainer is also supported by tubular steel. It is protected from missile hazards by surrounding concrete structures. In the event of a LOCA in the vessel cavity area the sump strainer is designed to support recirculation without the incore tunnel portion of the strainer intact.
Both the containment vessel emergency sump, its strainers, and the anti-vortex grating are Seismic Class I and Q-listed.
The containment vessel emergency sump (upper strainer) consists of one sump, two horizontal exit openings, vertical strainer assemblies, and anti-vortexing grates.
The containment sump screen is a vertical strainer design located within the existing sump pit and incore tunnel and is designed to assure full submergence at the minimum calculated recirculation pool water level.
The strainer is a passive design, and has been designed to withstand applicable structural loads, including seismic and head loss. A trash rack has been provided for the upper strainer to preclude the introduction of large debris that might cover a portion of the strainer surface.
The entire containment vessel emergency sump consisting of the elevation 565 feet and the incore tunnel area portions are required to mitigate a LOCA; however, for breaks located inside the reactor cavity the incore portion of the strainer is assumed to fail. For breaks located inside the reactor cavity with the incore (lower) portion failed, the elevation 565 feet (upper) strainer remains intact and is designed to handle the postulated debris loading.
The structural frame supporting the strainers assures that large debris carried in the water following a LOCA will not readily damage the strainers.
A significantly larger strainer was installed at DBNPS within the containment. The total strainer surface area is calculated to be 1226 ft2. This is made up of 394 ft2 in the upper strainer structure and 832 ft2 in the lower strainer structure. Both the upper and lower Davis-Besse Nuclear Power Station, Unit No. 1 7
strainer structure are fully submerged at the time of switchover to sump recirculation for all scenarios. (Reference 2)
The surface areas for the containment sump strainers are summarized below.
Table 1.4-1: Containment Sump Strainer Parameters Strainer Manufacturer Strainer Type Surface Area (ft2)
Upper Strainer Enercon Top Hat 394 Lower Strainer Enercon Top Hat 832 Total Strainer Enercon Top Hat 1226 2.0 General Description and Schedule for Corrective Actions NRC Issue:
Provide a general description of actions taken or planned, and dates for each. For actions planned beyond December 31, 2007, reference approved extension requests or explain how regulatory requirements will be met as per "Requested Information" Item 2(b). That is provide a general description of and implementation schedule for all corrective actions, including any plant modifications, that you identified while responding to this generic letter. Efforts to implement the identified actions should be initiated no later than the first refueling outage starting after April 1, 2006. All actions should be completed by December 31, 2007. Provide justification for not implementing the identified actions during the first refueling outage starting after April 1, 2006. If all corrective actions will not be completed by December 31, 2007, describe how the regulatory requirements discussed in the Applicable Regulatory Requirements section will be met until the corrective actions are completed.
Energy Harbor Nuclear Corp. Response:
Energy Harbor Nuclear Corp. has performed analyses to determine the susceptibility of the ECCS and CS system recirculation functions for DBNPS to the adverse effects of post-accident debris blockage and operation with debris-laden fluids. These analyses conform to the greatest extent practical to the NEI 04-07 methodology (Reference 11) as approved by the NRC Safety Evaluation dated December 6, 2004 (Reference 12). As of February 28, 2008, Energy Harbor Nuclear Corp. has completed the following GL 2004-02 actions, analyses and modifications as described in the previous supplemental response (Reference 2):
Summary of Plant Modifications and Programmatic Changes Completed:
During the DBNPS 13th refueling outage (13RFO) strainer replacements were installed. The new replacement strainer increased the available surface area from approximately 50 square feet to 1226 square feet.
Davis-Besse Nuclear Power Station, Unit No. 1 8
Modification of HPI pump was completed prior to the restart of 13 RFO in 2004.
Modification of cyclone separator for ECCS pumps seals and the installation of the cyclone separators on the CS pump seal supply lines were completed prior to the restart of 13RFO.
During 13RFO replaced all fibrous piping insulation in the containment with Reflective Metal Insulation with exception of three locations that total less than 1ft3 was replaced with NukonTM insulation. Containment coatings were restored including the containment dome. The containment air coolers were rebuilt using stainless steel coils.
Established the following debris source term programmatic controls:
o Updated the procedure for containment storage o Updated the procedure for tags, labels, and signs o Attached new equipment labels with stainless steel aircraft cable o Updated the procedure for containment cleanliness inspections o Coatings control program established - linked to sump strainer design by Engineering Change Process (ECP)
LOCA Deposition Model (LOCADM) analyses were conducted for DBNPS in accordance with WCAP-16793-NP, Revision 2 (Reference 9). DBNPS satisfies the maximum clad temperature and total deposition thickness limits of WCAP-16793-NP.
Summary of Tests and Evaluations Performed Walkdowns were performed to identify potential debris sources in containment.
Inventories of insulation, unqualified coatings, aluminum, and latent debris were created from the information gathered.
Debris generation and transport analyses were performed to determine the quantities of debris that may reach the sump strainers following a variety of LOCA scenarios.
Net Positive Suction Head margin analysis including debris sources and chemical effects were performed.
Downstream effects testing for debris was performed. The tested used a 100 percent bypass assumption for debris loading. The source was loaded into the tank and the devices similar to the system pumps were taking suction on the tank.
Enercon Report DBE004-RPT-004, Assessment of Debris Size Acceptance on ECCS Components determined that the cyclone separators that provide clean water to the LPI pump seals and the CS pump seals as well as the HPI pump internal passages could be adversely impacted by debris.
Energy Harbor Nuclear Corp. has no outstanding corrective actions associated with GL 2004-02 for DBNPS.
Davis-Besse Nuclear Power Station, Unit No. 1 9
3.0 Specific Information for Review Areas As stated in DBNPS Supplemental Response dated February 28, 2008 (Reference 2) as well as subsequent RAI responses submitted on April 30, 2010 (Reference 3),
DBNPS has addressed review areas 3.a through 3.m, and the only outstanding review areas 3.n through 3.p are addressed in this submittal.
3.n Downstream Effects - Fuel and Vessel NRC Issue:
The objective of the downstream effects, fuel and vessel section is to evaluate the effects that debris carried downstream of the containment sump screen and into the reactor vessel has on core cooling.
Show that the in-vessel effects evaluation is consistent with, or bounded by, the industry generic guidance (WCAP-16793), as modified by NRC staff comments on that document. Briefly summarize the application of the methods. Indicate where the WCAP methods were not used or exceptions were taken and summarize the evaluation of those areas.
Energy Harbor Nuclear Corp. Response:
Topical Report WCAP-17788-P, Rev. 1 (References 6,7,8) provides evaluation methods and results to address in-vessel downstream effects. As discussed in NRC Technical Evaluation Report of In-Vessel Debris Effects, (Reference 5), the NRC staff has performed a detailed review of WCAP-17788-P. Although the NRC staff did not issue a Safety Evaluation for WCAP-17788-P for use, as discussed further in U.S. Nuclear Regulatory Commission Staff Review Guidance for In-Vessel Downstream Effects Supporting Review of Generic Letter 2004-02 Responses (Reference 10), the staff expects that many of the methods developed in the TR may be used by PWR licensees in demonstrating adequate LTCC. Energy Harbor Nuclear Corp. used methods and analytical results developed in WCAP-17788-P, Rev. 1 to address in-vessel downstream debris effects for DBNPS and has evaluated the applicability of the methods and analytical results from WCAP-17788-P, Rev. 1 for DBNPS.
3.n.1 Sump Strainer Fiber Penetration Energy Harbor Nuclear Corp. has applied the NEI clean plant criteria to determine the amount of fibrous debris penetrating the sump strainers for use in the downstream in-vessel debris analysis for DBNPS. The clean plant criteria, as applied to in-vessel effects, utilize a fiber penetration (bypass) fraction of 45% and a debris transport fraction of 100%.
Davis-Besse Nuclear Power Station, Unit No. 1 10
Based on the clean plant criteria values for the debris transport fraction to the strainer (T) and the fiber penetration fraction (P), the following DBNPS specific in-vessel debris load is determined using:
g M. T. P
=
FA N Where:
g/FA grams of fiber per fuel assembly M mass of fibrous debris (latent + generated from worst-case break) [grams]
T transport fraction to the strainer 100%
P strainer penetration fraction 45% bypass fraction N number of fuel assemblies The mass of fibrous debris, M, is calculated as described below.
g [63lbm x.15x 453.5 gm/lbm fiber mass] . 1.00 . 0.45
= = [in-vessel fiber load 10.9 g/FA FA [177 FAs]
This is the DBNPS specific in-vessel fiber load (approximately 11 g/FA) that is below the applicable WCAP-17788-P, Rev. 1 in-vessel debris acceptance criterion (Volume 4, Part 4 Reference 7).
Davis-Besse Nuclear Power Station, Unit No. 1 11
3.n.2 Applicability to WCAP-17788 Methods and Analysis Results DBNPS is a B&W plant which is less susceptible to an inhibition of long-term core cooling due to in-vessel debris effects because of the extensive flow communication between the barrel-baffle region and the core. Per Section 3.0 of the NRC Staff Review Guidance (Reference 5), it is only necessary to confirm that the maximum combined amount of fiber that may arrive at the core inlet and heated core for the worst-case hot leg break scenario is less than the WCAP-17788-P, Rev. 1 in-vessel fibrous debris acceptance criterion.
As shown in the sump strainer fiber penetration section, the DBNPS maximum amount of fiber calculated to potentially reach the reactor vessel is 11 g/FA, which is less than the Proprietary in-vessel fibrous debris limit provided in WCAP-17788-P Volume 4 Part 4, Rev. 1 (Reference 7).
3.n.3 Fuel Design DBNPS uses the Framatome fuel design Mark-B-HTP fuel assemblies.
3.n.4 WCAP-17788 Debris Limit The proprietary total in-vessel (core inlet and heated core) fibrous debris limit contained in WCAP-17788-P Volume 4 Part 4, Rev. 1 (Reference 7) applies to DBNPS.
3.n.5 Methodology used to calculate the fibrous debris amounts The amount of fibrous debris calculated to arrive at the reactor vessel is determined for DBNPS following the method described in WCAP-17788-P, Volume 4 Part 4.
(Reference 7).
The response to 3.n.1 describes the methodology used to calculate the fibrous debris.
The mass (M) of fibrous debris is described in DBNPS calculation, Containment Latent Debris and Latent Fiber Determination, and is the limit for dirt loading in containment with 15% assumed to be fiber.
3.n.6 Confirm maximum combined amount of fiber that may arrive at the core inlet and heated core for hot leg break is below the WCAP-17788 fiber limit As shown in the sump strainer fiber penetration section, the DBNPS maximum amount of fiber calculated to potentially reach the reactor vessel is 11 g/FA which is less than the Proprietary in-vessel fibrous debris limit provided in WCAP-17788-P Volume 4 Part 4.
3.n.7 Confirmation that the core inlet fiber amount is less than the WCAP-17788-P, Rev. 1 threshold Davis-Besse Nuclear Power Station, Unit No. 1 12
DBNPS is a B&W two loop plant with Framatome fuel. The applicable WCAP-17788-P, Rev. 1 core inlet fiber threshold is provided in WCAP-17788-P Volume 4 Part 4.
(Reference 7). The core inlet fiber amount for DBNPS is calculated to be 11 g/FA, which is less than the applicable WCAP-17788-P, Rev. 1, core inlet fiber threshold.
3.n.8 Confirmation that the earliest sump switchover (SSO) time is 20 minutes or greater The earliest possible SSO time for DBNPS is 40 minutes. In DBNPS UFSAR Section 15.4.6.4, Maximum Hypothetical Accident, the injection phase is assumed to terminate at 40 minutes, the minimum time at which the water from the BWST could be exhausted.
3.n.9 Predict chemical precipitation timing from WCAP-17788-P, Rev. 1, Volume 5 testing and the specific test group considered to be representative of the plant.
Chemical precipitation timing is dependent on the plant buffer, sump pool pH, volume and temperature, and debris types and quantities. Table 3.n.9.1 summarizes the key chemical precipitation parameters and values for DBNPS and compares them to test group 42 from WCAP-17788-P, Rev. 1, Volume 5. Based on the comparison in Table 3.n.9.1, test group 42 is representative of DBNPS and the predicted chemical precipitation timing (tchem) is greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Table 3.n.9.1 Key Parameter Values for Chemical Precipitation Timing Parameter DBNPS Value Test Group 42 Value Buffer TSP TSP 8.042 pH 8.042 Minimum Sump Volume (ft3) 46,125 46125 Max Sump Pool 255 260 Temperature (°F)
CalSil (g) 0 0 E-glass (ft3) 6 6 Silica (g) 0 0 Mineral Wool (g) 0 0 Al Silicate (g) 0 0 Concrete (g) 1300 1300 Interam (g) 0 0 Al (ft2) 7286 7286 Galvanized Steel (ft2) N/A N/A Davis-Besse Nuclear Power Station, Unit No. 1 13
3.n.10 Confirm that chemical effects will not occur earlier than latest time to implement Boric Acid Precipitation (BAP) mitigation measures.
Per station procedure, DB-OP-02000, RPS, SFAS, SFRCS Trip or SG Tube Rupture, Attachment 12, Establishing Long Term Boron Dilution is time critical action that is performed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> when required. This is less than the predicted chemical precipitation time of > 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.n.11 WCAP-17788 tblock value for the RCS design category DBNPS is a B&W plant. Based on WCAP-17788, Rev. 1, Volume 1, section 11-2, (Reference 6) tblock for DBNPS is 20 minutes.
3.n.12 Confirmation that chemical effects do not occur prior to tblock The earliest time of chemical precipitation for DBNPS was determined to be >24 hours hours, which is greater than the applicable tblock value of 20 minutes.
3.n.13 Plant rated thermal power compared to the analyzed power level for the RCS design category DBNPS has a rated thermal power of 2817 MWt. DBNPS is a B&W plant and the applicable analyzed thermal power is 2827 MWt as provided in WCAP 17788-P, Rev. 1, Volume 4, Table 6-4 (Reference 7). The DBNPS rated thermal power is less than the analyzed power, this parameter is bounded by the WCAP-17788-P, Rev. 1 alternate flow path analysis.
3.n.14 Plant alternate flow path (AFP) resistance compared to the analyzed AFP resistance for the plant RCS design category DBNPS is a B&W plant. The proprietary analyzed AFP resistance is provided in Table 6-4 of WCAP-17788-P Volume 4, Rev. 1 (Reference 7). The barrel/baffle design for all B&W plants is the same and the barrel/baffle flow resistances shown in Table 6-4 represents all B&W plants. Therefore, the AFP value for DBNPS aligns with the analyzed value.
3.n.15 Consistency between the minimum ECCS flow per FA assumed in the AFP analyses and that at the plant DBNPS is a B&W barrel/baffle plant and B&W plants have the same barrel/baffle design. Table 11-1 of WCAP-17788-P shows a Sump Recirculation Rate of 8.5 gpm/FA which equals 1504.5 gpm (8.5 gpm x 177 FA). All B&W plants are bounded by 8.5 gpm/FA. The DBNPS Minimum ECCS Recirculation Flow is 19.8 gpm/FA.
Davis-Besse Nuclear Power Station, Unit No. 1 14
3.n.16 Summary The comparison of key parameters used in the WCAP-17788 AFP analysis to the DBNPS specific values is summarized in Table 3.n.16-1. Based on these comparisons DBNPS is bounded by the key parameters and the WCAP-17788 methods and results are applicable.
Table 3.n.16-1 Key Parameter Values for In-Vessel Debris Effects WCAP-17788 Parameter DBNPS Value Evaluation Value Maximum Total In-Vessel Volume 4 Maximum in-vessel fiber load is 11 Fiber Load (g/FA) Part 4 less than WCAP-17788 limit.
Maximum Core Inlet Volume 4 Maximum core inlet fiber load is 11 Fiber Load (g/FA) Part 4 less than WCAP-17788 threshold.
Later switchover time results in a lower decay heat at the time of Minimum Sump 20 40 debris arrival, reducing the Switchover Time (min) potential for debris induced core uncovery and heatup.
Potential for complete core inlet Minimum Chemical blockage due to chemical product 0.33 (tblock) >24 (tchem)
Precipitate Time (hr) generation would occur much later than assumed.
Latest hot leg switchover occurs Maximum Hot Leg
>24 (tchem) 4 well before earliest potential Switchover Time (hr) chemical product generation.
Rated Thermal Power DBNPS is bounded by the 2827 2817 (MWt) analyzed value.
DBNPS is bounded by the Maximum AFP Volume 4 Volume 4 analyzed value since all B&W Resistance Table 6-4 Table 6-4 plants are the same Minimum ECCS All B&W plants are bounded by Recirculation Flow 8.5 19.8 1504.5 gpm minimum ECCS (gpm/FA) Recirculation flow (i.e. 8.5 gpm/FA).
3.o Chemical Effects NRC Issue:
The objective of the chemical effects section is to evaluate the effect that chemical precipitates have on head loss and core cooling.
- 1) Provide a summary of evaluation results that show that chemical precipitates formed in the post-LOCA containment environment, either by themselves or Davis-Besse Nuclear Power Station, Unit No. 1 15
combined with debris, do not deposit at the sump screen to the extent that an unacceptable head loss results, or deposit downstream of the sump screen to the extent that long-term core cooling is unacceptably impeded.
Energy Harbor Nuclear Corp. Response:
The DBNPS chemical effects analysis of the sump strainers was submitted in Supplemental Response dated February 28, 2008 (Reference 2) as well as subsequent RAI response submitted on April 30, 2010 (Reference 3). The DBNPS sump strainer chemical effects analysis is unchanged.
The DBNPS in-vessel chemical effects analysis is described in Sections 3.n.9 through 3.n.12.
3.p Licensing Basis NRC Issue:
The objective of the licensing basis section is to provide information regarding any changes to the plant licensing basis due to the sump evaluation or plant modifications.
- 1) Provide the information requested in GL 04-02 Requested Information Item 2(e) regarding changes to the plant licensing basis. The effective date for changes to the licensing basis should be specified. This date should correspond to that specified in the 10 CFR 50.59 evaluation for the change to the licensing basis.
Energy Harbor Nuclear Corp. Response:
The DBNPS licensing basis will be changed in accordance with the requirements of 10 CFR 50.71(e) to incorporate the GL 2004-02 response within six months of receiving NRC acceptance of the updated final supplemental response to the Generic Letter as committed to in Reference 4.
Davis-Besse Nuclear Power Station, Unit No. 1 16
List of References
- 1. SECY-12-0093, Closure Options for Generic Safety Issue - 191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance, dated July 9, 2012 (ADAMS Accession No. ML121310648)
- 2. FENOC Letter to NRC L-08-036, Supplemental Response to Generic Letter 2004-02 (TAC Nos. MC4681), dated February 28, 2008 (ADAMS Accession No. ML080650368)
- 3. FENOC Letter to NRC L-10-111, Response to Request for Additional Information Related to Generic Letter 2004-02 (TAC Nos. MC4681), dated April 30, 2010 (ADAMS Accession No. ML101250217)
MC4681), dated May 15, 2013 (ADAMS Accession No. ML13135A456)
- 5. NRC Document, Technical Evaluation Report of In-Vessel Debris Effects, dated June 13, 2019 (ADAMS Accession No. ML19178A252)
- 6. Topical Report WCAP-17788-P, Volume 1, Revision 1, Comprehensive Analysis and Test Program for GSI-191 Closure (PA-SEE-1090), dated December 12, 2019
- 7. Westinghouse Topical Report WCAP-17788-P, Volume 4, Revision 1, Comprehensive Analysis and Test Program for GSI-191 Closure (PA-SEE-1090) - Thermal-Hydraulic Analysis of Large Hot Leg Break with Simulation of Core Inlet Blockage, dated February 28, 2020
- 8. Topical Report WCAP-17788-P, Volume 5, Revision 1, Comprehensive Analysis and Test Program for GSI-191 Closure (PA-SEE-1090) - Autoclave Chemical Effects Testing for GSI-191 Long-Term Cooling, dated December 13, 2019
- 9. Westinghouse Report WCAP-16793-NP, Revision 2, Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid, dated October 2011 (ADAMS Accession No. ML11292A021)
- 10. NRC Document, U.S. Nuclear Regulatory Commission Staff Review Guidance for In-Vessel Downstream Effects Supporting Review of Generic Letter 2004-02 Responses, dated September 4, 2019 (ADAMS Accession No. ML19228A011)
- 11. Nuclear Energy Institute Guidance Report NEI 04-07, Revision 0, Pressurized Water Reactor Sump Performance Methodology, dated December 2004
- 12. Safety Evaluation by the Office of the Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02, Nuclear Energy Institute Guidance Report NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology, dated December 6, 2004 Davis-Besse Nuclear Power Station, Unit No. 1 17